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Miyazawa, Takeshi; Uwaba, Tomoyuki; Yano, Yasuhide; Tanno, Takashi; Otsuka, Satoshi; Onizawa, Takashi; Ando, Masanori; Kaito, Takeji
JAEA-Technology 2024-009, 140 Pages, 2024/10
JAEA-Technology-2024-009.pdf:8.03MB
For the purpose of enhancing the reliability of fast reactor fuel designing using modified type 316 steel, the out-of-pile and in-pile mechanical data of modified type 316 steel cladding tubes and wrapper tubes were statistically analyzed with the knowledge on material science and engineering; the high-temperature strength equations of modified type 316 steel, which can be applied to high-dose neutron irradiation environment, were derived. The out-of-pile high-temperature tensile and creep data of modified type 316 steel cladding tubes and wrapper tubes were derived up to 900
C, which is higher than the upper limit temperature of anticipated transient event of fast reactor. Using the extended database, the best-fit equation and the lower limit equation were derived for out-of-pile 0.2% proof strength, ultimate tensile strength and creep rupture strength while the best-fit equation and the upper and lower limit equations for creep strain. Furthermore, the degradation factors for tensile and creep strength, which will be produced by in-reactor environment (i.e., neutron irradiation in liquid sodium), were evaluated using the existing neutron irradiation data of modified type 316 steel, which were derived using the experimental fast reactor Joyo, the French proto-type fast reactor Phenix, the American experimental fast reactor FFTF. The derived equations were validated by the comparison with the experimental data.
Pu
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powder using master sintering curve theoryNakamichi, Shinya; Sunaoshi, Takeo*; Hirooka, Shun; Vauchy, R.; Murakami, Tatsutoshi
Journal of Nuclear Materials, 595, p.155072_1 - 155072_11, 2024/07
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Miyazawa, Takeshi; Tanno, Takashi; Imagawa, Yuya; Hashidate, Ryuta; Yano, Yasuhide; Kaito, Takeji; Otsuka, Satoshi; Mitsuhara, Masatoshi*; Toyama, Takeshi*; Onuma, Masato*; et al.
Journal of Nuclear Materials, 593, p.155008_1 - 155008_16, 2024/05
Times Cited Count：0 Percentile：0.00(Materials Science, Multidisciplinary)Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Journal of Nuclear Science and Technology, 61(5), p.582 - 594, 2024/05
Times Cited Count：1 Percentile：41.04(Nuclear Science & Technology)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.
Kawaguchi, Munemichi; Hirakawa, Yasushi; Sugita, Yusuke; Yamaguchi, Yutaka
Nuclear Technology, 210(1), p.55 - 71, 2024/01
Times Cited Count：0 Percentile：0.00(Nuclear Science & Technology)This study has developed an estimation method for residual sodium film and sodium lumps on dummy fuel pins in Monju and demonstrated sodium draining behavior through gaps among the pins, experimentally. The amounts of the residual sodium on the surface of the pins were measured using the three-type test specimens: (a) single pin, (b) 7-pin assembly, and (c) 169-pin assembly. The experiments revealed that the withdrawal speed of the pins and improvement of the sodium wetting increased drastically the amounts of the residual sodium. Furthermore, the experiments using the 169-pin assembly measured the practical amounts of the residual sodium in the dummy fuel assembly of short length and demonstrated sodium draining behavior through the dummy fuel assembly. The estimation method includes four models: a viscosity flow model, Landau-Levich-Derjaguin (LLD) model, an empirical equation related to the Bretherton model, and a capillary force model in a tube. The calculation predicted comparable amounts of the residual sodium with the experiments. An uncertain of the sodium wetting effects were close to 1.8 times the estimation values of the LLD model. With this estimation method, the amounts of the residual sodium on the unloaded Monju dummy fuel assembly can be evaluated.
Tsai, T.-H.; Sasaki, Shinji; Maeda, Koji
Journal of Nuclear Science and Technology, 60(6), p.715 - 723, 2023/06
Times Cited Count：1 Percentile：19.69(Nuclear Science & Technology)Ohgama, Kazuya; Takegoshi, Atsushi*; Katagiri, Hiroki; Hazama, Taira
Nuclear Technology, 208(10), p.1619 - 1633, 2022/10
Times Cited Count：4 Percentile：63.92(Nuclear Science & Technology)Yoshida, Hiroyuki; Horiguchi, Naoki; Ono, Ayako; Furuichi, Hajime*; Katono, Kenichi*
Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 7 Pages, 2022/08
Johnson, M.*; Delacroix, J.*; Journeau, C.*; Brayer, C.*; Clavier, R.*; Montazel, A.*; Pluyette, E.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji
Proceedings of International Conference on Fast Reactors and Related Fuel Cycles; Sustainable Clean Energy for the Future (FR22) (Internet), 8 Pages, 2022/04
Fuel-coolant interactions in the event of molten fuel discharge to the lower plenum of a sodium cooled fast reactor is under investigation as part of a French-Japanese experimental collaboration on severe accidents. The MELT facility enables the X-ray visualisation of the quenching of molten core material jets in sodium at kilogram-scale. The SERUA facility, currently under preparation, is presented for the investigation of boiling heat transfer at elevated melt-coolant interface temperatures. In this article, the status of the collaboration using these facilities is presented.
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：30.60(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：10.11(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.
Amaya, Masaki; Kakiuchi, Kazuo; Mihara, Takeshi
Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.1048 - 1056, 2019/09
Uwaba, Tomoyuki; Nemoto, Junichi*; Ishitani, Ikuo*; Ito, Masahiro*
Nuclear Engineering and Design, 331, p.186 - 193, 2018/05
Times Cited Count：4 Percentile：35.54(Nuclear Science & Technology)A computer code for the analysis of the overall irradiation performance of a fast reactor mixed-oxide (MOX) fuel element was coupled with a specialized code for the analysis of fission product cesium behaviors in a MOX fuel element. The coupled code system allowed for the analysis of the radial and axial Cs migrations, the generation of Cs chemical compounds and fuel swelling due to Cs-fuel-reactions in association with the thermal and mechanical behaviors of the fuel element. The coupled code analysis was applied to the irradiation performance of a fast reactor MOX fuel element attaining high burnup for discussion on the axial distribution of Cs, fuel-to-cladding mechanical interaction owing to the Cs-fuel-reactions by comparing the calculated results with post irradiation examinations.
Kikuchi, Norihiro; Imai, Yasutomo*; Yoshikawa, Ryuji; Doda, Norihiro; Tanaka, Masaaki; Ohshima, Hiroyuki
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 12 Pages, 2017/07
In the design study of an advanced loop-type SFR in JAEA, a specific fuel assembly (FA) named FAIDUS (Fuel Assembly with Inner DUct Structure) has been adopted as one of the measures to enhance safety of the reactor. Thermal-hydraulics evaluations of FAIDUS under various operation conditions are required to confirm its design feasibility. In this study, after the applicability of ASFRE to FAs was confirmed through the numerical analysis using simulated FA tests, thermal-hydraulic analyses of a FA without an inner duct and a FAIDUS were conducted. Through the numerical analyses, it was indicated that asymmetric temperature distribution in a FAIDUS would not be occurred and characteristics of the temperature distribution was almost the same as that in a FA without an inner duct. Under the low flow rate condition, it was expected that the local flow acceleration caused by the buoyancy force in a FAIDUS could bring the flow redistribution and make the temperature distribution flat.
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Mechanical Engineering Journal (Internet), 4(3), p.16-00592_1 - 16-00592_9, 2017/06
Uwaba, Tomoyuki; Ohshima, Hiroyuki; Ito, Masahiro*
Nuclear Engineering and Design, 317, p.133 - 145, 2017/06
Times Cited Count：9 Percentile：62.79(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.
Ohgama, Kazuya; Aliberti, G.*; Stauff, N. E.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 8 Pages, 2017/04
Ohgama, Kazuya; Ota, Hirokazu*; Ikusawa, Yoshihisa; Oki, Shigeo; Ogata, Takanari*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 6 Pages, 2017/04
Stauff, N. E.*; Ohgama, Kazuya; Aliberti, G.*; Oki, Shigeo; Kim, T. K.*
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 10 Pages, 2017/04
