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Yoshikawa, Ryuji; Imai, Yasutomo*; Kikuchi, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Nuclear Technology, 210(5), p.814 - 835, 2024/05
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)In the study of safety enhancement on advanced sodium-cooled fast reactor, it is essential to clarify the thermal-hydraulics under various operation conditions in a fuel assembly (FA) with the wire-wrapped fuel pins to assess the structural integrity of the fuel pin. A finite element thermal-hydraulics analysis code named SPIRAL has been developed to analyze the detailed thermal-hydraulics phenomena in a FA. In this study, the numerical simulations of the 37-pin bundle sodium experiments at different Re number conditions, including a transitional condition between laminar and turbulent flows and turbulent flow conditions, were performed to validate the hybrid turbulence model equipped in SPIRAL. The temperature distributions predicted by SPIRAL was consistent with those measured in the experiments. Through the validation study, the applicability of the hybrid turbulence model in SPIRAL to thermal-hydraulic evaluation of sodium-cooled FA in the wide range of Re number was confirmed.
Ishida, Shinya; Tagami, Hirotaka; Okano, Yasushi; Yamano, Hidemasa; Kubo, Shigenobu; Tobita, Yoshiharu
Proceedings of 11th European Review Meeting on Severe Accident Research Conference (ERMSAR 2024) (Internet), 10 Pages, 2024/05
Uwaba, Tomoyuki; Ito, Masahiro*; Ishitani, Ikuo*
JAEA-Technology 2023-006, 36 Pages, 2023/05
JAEA-Technology-2023-006.pdf:3.45MB
The BAMBOO code developed by the Japan Atomic Energy Agency is a computer code to analyze fuel pin bundle deformation in a fast reactor wire-spaced type fuel pin bundle subassembly. In this study we developed an analysis model to consider friction at the contact points between adjacent fuel pins, and at these between outermost fuel pins and a duct that are due to bundle-duct interaction. This model deals with friction forces at contact points in the contact and separation analysis of the code, and employs a convergent calculation where contact forces are gradually determined to avoid numerical instability when the friction occurs. Analyses of BAMBOO with the model showed very slight effects on the onset of contact between outer most pins and a duct, and on directions of pin displacements, within the range of practical friction coefficients.
Okita, Shoichiro; Mizuta, Naoki; Takamatsu, Kuniyoshi; Goto, Minoru; Yoshida, Katsumi*; Nishimura, Yosuke*; Okamoto, Koji*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 10 Pages, 2023/05
Uwaba, Tomoyuki; Nemoto, Junichi*; Ito, Masahiro*; Ishitani, Ikuo*; Doda, Norihiro; Tanaka, Masaaki; Otsuka, Satoshi
Nuclear Technology, 207(8), p.1280 - 1289, 2021/08
Times Cited Count：3 Percentile：28.52(Nuclear Science & Technology)Computer codes for irradiation behavior analysis of a fuel pin and a fuel pin bundle and for coolant thermal hydraulics analysis were coupled into an integrated code system. In the system, each code provides data required by other codes and the analyzed results are shared among them. The system allows for the synthesizing of analyses of thermal, chemical and mechanical behaviors in a fuel subassembly under irradiation. A test analysis was made for a fuel subassembly containing a mixed oxide fuel pin bundle irradiated in a fast reactor. The results of the analysis were presented with transverse cross-sectional images of the fuel subassembly and three-dimensional images of a fuel pin and fuel pin bundle models. For detailed evaluation, various irradiation behaviors of all fuel pins in the subassembly were analyzed and correlated with irradiation conditions.
Oka, Hiroshi*; Kaito, Takeji; Ikusawa, Yoshihisa; Otsuka, Satoshi
Nuclear Engineering and Design, 370, p.110894_1 - 110894_8, 2020/12
Times Cited Count：1 Percentile：9.42(Nuclear Science & Technology)The objective of this study is to evaluate the reliability of a cumulative damage fraction (CDF) analysis for the prediction of fuel pin breach in fast rector using experimentally obtained fuel pin breach data for the first time. Six breached fuel pins were obtained from steady state irradiation in the EBR-II. Post irradiation examinations revealed that FP gas pressure was the main cause of creep damage in cladding, and that the stress contribution from FCMI was negligible. CDFs evaluated for these pins using in-reactor creep rupture equation, taking into account the irradiation history of cladding temperature and hoop stress due to FP gas pressure, were in the range of 0.7 to 1.4 at the occurrence of breach. This shows clearly that fuel pin breach occurs when the CDF approaches 1.0. The results indicate that CDF analysis would be a reliable method for the prediction of fuel pin breach when appropriate material strength and environmental effects are adopted.
Sato, Ikken
Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05
Times Cited Count：12 Percentile：74.50(Nuclear Science & Technology)Water columns were adopted in the pressure measurement system of Fukushima-Daiichi Unit-3. Part of these water columns evaporated during the accident condition jeopardizing correct understanding on actual pressure. Through comparison of RPV (Reactor Pressure Vessel) and S/C pressures with D/W pressure, such water-column effect was evaluated. Correction for this effect was developed enabling clarification of slight pressure difference among RPV, S/C and D/W. This information was then integrated with other available data such as, water level, CAMS and environmental dose rate, into an interpretation of accident focusing on RPV and PCV pressurization/depressurization and radioactive material release to environment. It is suggested that dryout of in-vessel and ex-vessel debris was likely causing pressure decrease. S/C water poured into pedestal heated by relocated debris was the likely cause of pressurization. Cyclic reflooding of pedestal debris and dryout was likely.
Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*
Nuclear Engineering and Design, 317, p.133 - 145, 2017/06
Times Cited Count：9 Percentile：61.43(Nuclear Science & Technology)The coupled numerical analysis of mechanical and thermal behaviors was performed for a wire-wrap fuel pin bundle subassembly irradiated in a fast reactor. For the analysis, the fuel pin bundle deformation analysis code BAMBOO and the thermal hydraulics analysis code ASFRE exchanged the deformation and temperature analysis results through the iterative calculations to attain convergence corresponding to the static balance between deformation and temperature. The analysis by the coupled code system showed that radial distribution of coolant temperatures in a subassembly tended to be flattened as a result of the fuel pin bundle deformation governed by cladding void swelling and irradiation creep. Such temperature distribution was slightly analyzed as a result of the small bowing of the fuel pins due to the cladding-wire interaction even when no bundle-duct interaction occurred. The effect of the spacer wire-pitch on deformation and thermal hydraulics was also investigated in this study.
Kikuchi, Norihiro; Ohshima, Hiroyuki; Tanaka, Masaaki; Hashimoto, Akihiko*
Dai-21-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 4 Pages, 2016/06
For the thermal-hydraulic design and safety assessment regarding a fuel assembly of sodium-cooled fast reactors, a subchannel analysis code ASFRE has been and is continuously developed in JAEA. In the numerical simulation of ASFRE confirmed that the tendency to overestimate the maximum coolant temperature in a fuel assembly still remains. In this study, Distributed Resistance Model (DRM), which deals with wire-spacer wrap volumetric effect in subchannels on peripheral and axial directions, was modified and its calibration factor was optimized in order to improve the prediction accuracy of the maximum coolant temperature. A numerical simulation of a 37-pin bundle sodium experiment was also carried out and the result showed the validity of the modified DRM.
Nakajima, Kunihisa; Iwai, Takashi; Kikuchi, Hironobu; Serizawa, Hiroyuki; Arai, Yasuo
JAERI-Research 2005-027, 42 Pages, 2005/09
JAERI-Research-2005-027.pdf:4.15MB
Nitride fuel pins containing inert matrix such as ZrN and TiN were fabricated for the irradiation test at JMTR, aiming at understanding irradiation behavior of nitride fuel for transmutation of minor actinides. Minor actinides are surrogated by plutonium in the present fuel pin. This report describes the preparation and characterization of fuel pellets, and fabrication of fuel pins. The irradiation for 11 cycles from May 2002 to November 2004 at JMTR was completed without any failure of fuel pins.
Yamashita, Toshiyuki; Kuramoto, Kenichi; Shirasu, Noriko; Nakano, Yoshihiro; Akie, Hiroshi; Nagashima, Hisao; Kimura, Yasuhiko; Omichi, Toshihiko*
Journal of Nuclear Materials, 320(1-2), p.126 - 132, 2003/07
Times Cited Count：10 Percentile：55.68(Materials Science, Multidisciplinary)Two irradiation tests on the rock-like oxide (ROX) fuels, small disk-shape fuel targets and pellet-type fuels, were performed in order to clarify in-pile irradiation stabilities. Swelling, fractional fission gas release (FGR) and phase change were examined by puncture test, profilometry and ceramography. YSZ single-phase fuel showed an excellent irradiation behavior, ie. low fission gas release (less than 3%), negligible swelling and no appreciable restructuring. The particle dispersed fuels showed lower swelling and higher fission gas release than those of mechanically blended fuels. Spinel decomposition and subsequence restructuring in the spinel matrix fuels was observed for the first time in the present investigation. It would be possible to reduce the FGR of the spinel matrix fuels, if the maximum fuels temperature is limited below 1700 K where neither spinel decomposition nor restructuring was observed. Damaged area of spinel matrix due to fission fragment irradiation seemed to be confined to thin layers around the surface of YSZ particles.
Okuno, Hiroshi
Journal of Nuclear Science and Technology, 40(7), p.544 - 551, 2003/07
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)A method for classifying benchmark results of criticality calculations according to similarity was proposed in this paper. After formulation of the method utilizing correlation coefficients, it was applied to burnup credit criticality benchmarks Phase III-A and II-A, which were conducted by the Expert Group on Burnup Credit Criticality Safety under auspices of the OECD/NEA. Phase III-A benchmark was a series of criticality calculations for irradiated BWR fuel assemblies, whereas Phase II-A benchmark was a suite of criticality calculations for irradiated PWR fuel pins. These benchmark problems and their results were summarized. The correlation coefficients were calculated and sets of benchmark results were classified according to the criterion that the correlation coefficients were no less than 0.15 for Phase III-A and 0.10 for Phase II-A benchmarks. When a couple of results were in a same group, one result was found predictable from the other. An example was shown for each of the Benchmarks. The evaluated nuclear data seemed the main factor of errors.
Yamashita, Toshiyuki; Kuramoto, Kenichi; Akie, Hiroshi; Nakano, Yoshihiro; Shirasu, Noriko; Nakamura, Takehiko; Kusagaya, Kazuyuki*; Omichi, Toshihiko*
Journal of Nuclear Science and Technology, 39(8), p.865 - 871, 2002/08
Times Cited Count：26 Percentile：81.80(Nuclear Science & Technology)Research on the plutonium rock-like oxide (ROX) fuels and their once-through burning in light water reactors has been performed to establish an option for utilizing and disposing effectively the excess plutonium. The ROX fuel is a sort of the inert matrix fuels and consists of mineral-like compounds such as yttria stabilized zirconia, spinel and corundum. A particle-dispersed fuel was devised to reduce damage by heavy fission fragments. Some preliminary results on swelling, fractional gas release and microstructure change for five ROX fuels were obtained from the irradiation test and successive post-irradiation examinations. Inherent disadvantages of the Pu-ROX fuel cores could be improved by adding 238U or 232Th as resonant materials, and all improved cores showed a nearly the same characteristics as the conventional UO2 core during transient conditions. The threshold enthalpy of the ROX fuel rod failure was found to be comparable to the fresh UO2 rod by pulse-irradiation tests simulating reactivity initiated accident conditions.
Kugo, Teruhiko; Nakakawa, Masayuki
JAERI-Data/Code 2000-004, p.97 - 0, 2000/02
JAERI-Data-Code-2000-004.pdf:10.51MB
no abstracts in English
Sugihara, Masayoshi; Federici, G.*; C.Grisolia*; P.Ghendrih*; T.Loarer*; Nakamura, Hiroo; Y.Igitkhanov*; G.Janeschitz*; G.W.Pacher*; H.D.Pacher*; et al.
24th EPS Conf. on Controlled Fusion and Plasma Physics, 21A, p.997 - 1000, 1997/00
no abstracts in English
Omichi, Toshihiko
JAERI-Review 96-008, 13 Pages, 1996/07
JAERI-Review-96-008.pdf:0.92MB
no abstracts in English
; ;
PATRAM 95: 11th Int. Conf. on the Packaging and Transportation of Radioactive Materials, 3, p.1174 - 1181, 1996/00
no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki
PHYSOR 96: Int. Conf. on the Physics of Reactors, 1, p.B73 - B81, 1996/00
no abstracts in English
; ; ; ; Kawamura, Hiroshi; ; ; Nakagawa, Tetsuya; ;
JAERI-M 93-206, 48 Pages, 1993/10
no abstracts in English
Kugo, Teruhiko; Nakagawa, Masayuki;
JAERI-M 92-117, 70 Pages, 1992/08
no abstracts in English
