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Sakai, Takaaki; Ito, Kei; Yamaguchi, Akira; Iwasaki, Takashi*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), P. 5626, 2005/07
Numerical analysis method for a sodium-heated steam generator (SG) was developed for thermal-hydraulics evaluation of the whole region in SG including a cover gas volume. Two-dimensional analysis method for the SG (MSG code) was modified to extend to the cover gas region. In order to verify the calculated temperature profile in the cover gas region, three-dimensional analysis by the Fluent-3D code was performed for the cover gas region. In addition, shroud temperatures measured at the SG in the prototype Fast Breeder Reactor (FBR)
Yamaji, Akifumi*; Oka, Yoshiaki*; Ishiwatari, Yuki*; Liu, J.*; Koshizuka, Seiichi*; Suzuki, Motoe
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 7 Pages, 2005/05
Ensuring the fuel integrities is one of the most fundamental parts in the High Temperature Supercritical-Pressure Light Water Reactor. Most abnormal transient events of SCLWR-H last for a short period of time and the fuel rods are replaced after being irradiated in the core. In this study, the fuel integrity criteria are rationalized based on the fact that the fuel rod mechanical failures can be represented by the strain of the fuel rod cladding. A new fuel rod is designed with a Stainless Steel cladding. It is internally pressurized to reduce the stress on the cladding and also to increase the gap conductance between the pellet and the cladding. The fuel integrities both at normal operation and abnormal transient conditions are evaluated using the fuel analysis code FEMAXI-6 of JAERI.
Kugo, Teruhiko; Kojima, Kensuke; Ando, Masaki; Okajima, Shigeaki; Mori, Takamasa; Takeda, Toshikazu*; Kitada, Takanori*; Matsuoka, Shogo*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05
We have preliminarily evaluated the reduction of prediction errors of the core characteristics of the breeding light water reactor core based on the bias factor method by utilizing the FCA critical experiments carried out for MOX fueled tight lattice light water reactor cores. The prediction uncertainty of k
is reduced from 0.62% to 0.39% by utilizing the FCA-XV-2 (65V) result. As for the reaction rate ratio of 
U capture and 
Pu fission, it is found that the FCA XXII-1 (95V) and XV (95V) results are suitable for the upper core and the upper blanket of the real core and the FCA XXII-1 (65V) and XV-2 (65V) results are suitable for the lower core and the internal blanket.
Moriyama, Kiyofumi; Takagi, Seiji; Muramatsu, Ken; Nakamura, Hideo; Maruyama, Yu
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 9 Pages, 2005/05
The containment failure probability due to ex-vessel steam explosions were evaluated for a BWR Mk-II model plant. The evaluation was made for two scenarios: a steam explosion in the pedestal area, or in the suppression pool. A probabilistic approach, Latin Hypercube Sampling (LHS), was applied for the evaluation of steam explosion loads, in which a steam explosion simulation code JASMINE was used as a physics model. The fragility curves connecting the steam explosion loads and containment failure were developed based on simplified assumptions on the containment failure scenarios. The mean conditional probabilities of containment failure per occurrence of a steam explosion were 
for suppression pool and 
for pedestal area. Note that the results depend on the assumed range of input parameters and fragility curves that involve conservatism and simplification.
Tamai, Hidesada; Onuki, Akira; Kureta, Masatoshi*; Liu, W.; Sato, Takashi; Akimoto, Hajime
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 8 Pages, 2005/05
no abstracts in English
C operation of the HTTRSakaba, Nariaki; Matsuzawa, Takaharu*; Hirayama, Yoshiaki*; Nakagawa, Shigeaki; Nishihara, Tetsuo; Takeda, Tetsuaki
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 8 Pages, 2005/05
The permeation of hydrogen isotopes through the Hastelloy XR high-temperature alloy adopted for the heat exchanger pipes of the intermediate heat exchanger in the HTTR (High Temperature Engineering Test Reactor) is one of the concerns in the hydrogen production system, which will be connected to the HTTR in the near future. An evaluation of the hydrogen permeation between the primary and secondary coolant through the Hastelloy XR was performed using the hydrogen concentration data observed during the initial 950C operation of the HTTR. The hydrogen permeability of the Hastelloy XR was estimated conservatively high as follows. The activation energy E
and pre-exponential factor F of the permeability of hydrogen were E = 62 kJ/mol and F = 3.6
10
cm
(NTP)/(cm s Pa
), respectively, in the temperature range from 735K to 940K. The results implied that some oxidized film had been formed on the surface of the heat exchanger pipes of the intermediate heat exchanger.
Liu, W.; Tamai, Hidesada; Onuki, Akira; Kureta, Masatoshi*; Sato, Takashi; Akimoto, Hajime
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 10 Pages, 2005/05
A major concern in the design of RMWR is that sufficient cooling capability be provided to keep fuel cladding temperature below specified values, even for a postulated abnormal transient process. In this research, centered the postulated transient cases that may be possibly met in the RMWR running, transient BT tests are performed in 7-rod and 37-rod double-humped tight lattice bundles, under the RMWR nominal operating condition (P = 7.2 MPa, Tin =556 K) for mass velocity G = 300 - 800 kg / (m
s). Experiments are analyzed with TRAC code, in which JAERI critical power correlation is implemented for BT judgment. The traditional quasi-steady-state prediction of BT in transient process is confirmed being applicable for the postulated nominal transients in the RMWR cores.
Hishida, Masahiko; Konomura, Mamoru; Uchita, Masato; Iitsuka, Toru*; Kamishima, Yoshio*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), P. 5112, 2005/05
An innovative concept of medium-scale sodium-cooled modular reactor, named M-JSFR, has been created as based on the large-scale advanced loop type fast reactor concept. M-JSFR employs other concepts such as standardization and learning effects by designing as a modular plant and reduction of secondary loop number for the purpose of dissolving the scale-demerit. On this M-JSFR, some improvements are performed to overcome the weak point (strong chemical activity of sodium) of a sodium-cooled reactor and to achieve in-service inspection (ISI) and repair as easily as in light water reactors. Based on the ISI guidelines for light water reactors, the ISI procedures are reviewed reflecting the characteristics of M-JSFR. A guideline for ISI with the same grade of that of the light water reactors is established and major components subjected to ISI are selected. Moreover, suitable ISI procedures for each selected major component are proposed, and a plant concept amenable to ISI is studied. As the result of these studies, the construction cost of ISI&R reinforcement M-JSFR increases about 3% mainly because of the diameter expansion of reactor vessel.
Chikazawa, Yoshitaka; Hori, Toru*; Konomura, Mamoru; Hori, Toru*; Uchida, Shoji*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 0 Pages, 2005/05
The thermochemical and electrolytic hybrid cycle is one of hydrogen production methods using sulfuric acid cycle. The maximum temperature in all of the processes is kept lower than 500C because electrolysis is partially utilized in the thermochemical process. In this study, a hydrogen production plant with the thermochemical and electrolytic hybrid cycle has been designed and the hydrogen production efficiency has been evaluated. In this design, components in hydrogen production system are designed to be made of steels such as high Si cast iron which has good toughness against sulfuric acid. High hydrogen production efficiency of 42% (high heating value) is achieved assuming development of high efficiency electrolysis.
Usui, Shinichi; Chikazawa, Yoshitaka; Konomura, Mamoru; Tanaka, Toshihiko; Hori, Toru*; Ohkubo, Toshiyuki*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), 0 Pages, 2005/05
The Japan Nuclear Cycle Development Institute and the Electric Utilities have been studying for concepts of commercialized FBR plant system. In this study, a small sized sodium cooled reactor has been studied from standpoints of having a potential to use a power source applicable to diversified social needs and reduced capital risks. The concept is pursued to satisfy various requirements: economical competitiveness, reactor safety, very long lived core, etc. The electrical power is set to 165 MWe based on previous studies of small sized sodium cooled reactor which was aim of cost reduction of the plant pursued. A three regional Zr concentration with one Pu enrichment core has been designed. The burn-up reactivity is kept 1.17% dk/k and the core refueling occurs after 20 years operation with 77,000 MWd/t average burn-up. The reactor system is enhanced a passive safety to avoid a core melt even an ATWS and the design has possible capability to withstand UTOP and ULOHS events, although it requires reactor shutdown action by operator within a certain time. Tank-type primary system has been designed. A circular type IHX and a primary EMP are located in series inside of the reactor vessel. The steam generator and the secondary EMP are located outside of the reactor vessel and connected to the IHX by piping. A diameter of the core is minimized in order to reduce a reactor vessel size. Two PRACS driven by natural circulation is chosen as a decay heat removal system in order to satisfy a heat removal condition of Category IV in Japanese standard.The construction cost of the whole plant evaluates to be achieved an economical goal considering a mass production.
Ishikawa, Hiroyasu; Miyahara, Shinya; Yoshizawa, Yoshio*
Proceedings of 2005 International Congress on Advances in Nuclear Power Plants (ICAPP '05) (CD-ROM), P. 5688, 2005/05
Focusing on the cover layer materials (as the Radon Barrier Materials), which could have the effect to restrain the radon from scattering into the air and the effect of the radiation shielding, we produced the radon barrier materials with crude bentonite on an experimental basis, using the rotary type comprehensive unit for grinding and mixing, through which we carried out the evaluation of the characteristics thereof.