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Journal Articles

Validation of the fast reactor plant dynamics analysis code Super-COPD using FFTF loss of flow without scram test #13

Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa

Annals of Nuclear Energy, 195, p.110157_1 - 110157_14, 2024/01

To validate the fast reactor plant dynamics analysis code Super-COPD for the loss of flow without scram (LOFWOS) event, we participated in the IAEA benchmark for the LOFWOS test No.13 performed at the FFTF as one of the passive safety demonstration test. In the blind phase, there were challenges to reproduce outlet temperatures of fuel assemblies and the total reactivity. To improve the evaluation accuracy of them, the whole core model considering the radial heat transfer and interwrapper flow and the simplified assembly bowing reactivity model were introduced. As a result of the final phase, the second peak of outlet temperatures was reproduced successfully, and the total reactivity could generally follow the measured data. Super-COPD was validated for the LOFWOS event.

Journal Articles

Validation of feedback reactivity evaluation models for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04

Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.

Journal Articles

Verification of fuel assembly bowing analysis model for core deformation reactivity evaluation

Doda, Norihiro; Uwaba, Tomoyuki; Ohgama, Kazuya; Yoshimura, Kazuo; Nemoto, Toshiyuki*; Tanaka, Masaaki; Yamano, Hidemasa

Nihon Kikai Gakkai Kanto Shibu Dai-29-Ki Sokai, Koenkai Koen Rombunshu (Internet), 5 Pages, 2023/03

An evaluation method for reactivity feedback due to core deformation during reactor power increase in sodium-cooled fast reactors is being developed for realistic core design evaluation. In this evaluation method, fuel assembly bowing was modeled with a beam element of the finite element method, and the assembly's pad contact between adjacent assemblies was modeled with a dedicated element which could consider the wrapper tube cross-sectional distortion and the pad stiffness depending on pad contact conditions. This fuel assembly bowing analysis model was verified for thermal bowing of a single assembly and assembly pad contact between adjacent assemblies in a core as past benchmark problems. The calculation results by this model showed good agreement with those of reference solutions of theoretical solutions or results by participating institutions in the benchmark. This study confirmed that the analysis model was able to calculate thermal assembly bowing appropriately.

Journal Articles

Benchmark analysis of FFTF Loss of Flow Without Scram Test No.13 using fast reactor plant dynamics analysis code Super-COPD

Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki

Proceedings of 12th Japan-Korea Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS12) (Internet), 9 Pages, 2022/10

To improve the prediction accuracy of the plant dynamics analysis code named Super-COPD, JAEA has joined the IAEA benchmark for the FFTF Loss of Flow Without Scram Test No.13. In the first blind phase, there was the challenge to perform outlet temperatures of fuel assemblies more accurately. Hence, the renewed analysis was performed with the whole core multi-channel model in which each assembly was modelled to simulate the radial heat transfer among assemblies and the flow redistribution induced by the buoyancy in the NC conditions. Then, to validate the coupled transient analysis between the whole core multi-channel model and the one-point kinetics model, the analysis considering major reactivity feedbacks such as GEM, assembly bowing was performed. As a result, the second peak of outlet temperatures was reproduced successfully, and it was observed that the plant dynamics analysis could follow the measured data.

Journal Articles

Application of 1D-CFD coupling method to unprotected loss of heat sink event in EBR-II focusing on thermal stratification in cold pool

Yoshimura, Kazuo; Doda, Norihiro; Fujisaki, Tatsuya*; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Proceedings of 29th International Conference on Nuclear Engineering (ICONE 29) (Internet), 10 Pages, 2022/08

To confirm the applicability of the reactivity model, the authors have been conducting the benchmark exercises of the unprotected loss of heat sink event tests in a pool-type experimental fast reactor EBR-II. In the blind phase in the benchmark analyses using the plant dynamics analysis (1D) code in which the cold pool was modeled by means of the perfect mixing volume, it was found the increase of the core inlet temperature was evaluated lower than that of the measured data and the feedback reactivity was underestimated, because the thermal stratification in the cold pool was ignored. Then, the detailed model of the cold pool for the computational fluid dynamics (CFD) code was introduced and the 1D-CFD codes coupling method was applied to the benchmark analyses. It was confirmed that both the thermal stratification in the cold pool and the increase of the core inlet temperature were successfully reproduced.

Journal Articles

Sodium-cooled Fast Reactors

Ohshima, Hiroyuki; Morishita, Masaki*; Aizawa, Kosuke; Ando, Masanori; Ashida, Takashi; Chikazawa, Yoshitaka; Doda, Norihiro; Enuma, Yasuhiro; Ezure, Toshiki; Fukano, Yoshitaka; et al.

Sodium-cooled Fast Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.3, 631 Pages, 2022/07

This book is a collection of the past experience of design, construction, and operation of two reactors, the latest knowledge and technology for SFR designs, and the future prospects of SFR development in Japan. It is intended to provide the perspective and the relevant knowledge to enable readers to become more familiar with SFR technology.

Journal Articles

Validation of evaluation method of feedback reactivity for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Tanaka, Masaaki; Yamano, Hidemasa; Igawa, Kenichi*

Proceedings of 28th International Conference on Nuclear Engineering (ICONE 28) (Internet), 8 Pages, 2021/08

The numerical results of the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests in the pool-type experimental SFR in the United States, EBR-II (BOP-302R and BOP-301) are discussed in order to validate the evaluation method of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD. By comparing the numerical results and the experimental data, the profiles of the increase of the core inlet temperature and the decrease of the reactor power calculated by Super-COPD were comparable with those of the experimental data and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS event.

Journal Articles

The Effect of system constraint on coolant injection mode of energetic fuel-coolant interactions

Park, H.; Yamano, Norihiro; Maruyama, Yu; Moriyama, Kiyofumi; Yang, Y.; Sugimoto, Jun

Dai-35-Kai Nihon Dennetsu Shimpojiumu Koen Rombunshu, 3, p.803 - 804, 1998/00

no abstracts in English

Journal Articles

Findings from CSARP; Cooperative severe accident research program

Sugimoto, Jun; Hashimoto, Kazuichiro*; Yamano, Norihiro; Hidaka, Akihide; Maruyama, Yu; Uetsuka, Hiroshi; Fuketa, Toyoshi; Nakamura, Takehiko; Soda, Kunihisa; Katanishi, Shoji*

Nihon Genshiryoku Gakkai-Shi, 39(2), p.123 - 134, 1997/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Research plan on in-vessel debris coolability in ALPHA program

Maruyama, Yu; Yamano, Norihiro; Kudo, Tamotsu; Moriyama, Kiyofumi; Sugimoto, Jun

JAERI-memo 08-127, p.269 - 275, 1996/06

no abstracts in English

Journal Articles

Phenomenological studies on melt-coolant interactions in the ALPHA program

Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Hidaka, Akihide; Sugimoto, Jun

Nuclear Engineering and Design, 155(1-2), p.369 - 389, 1995/04

 Times Cited Count:47 Percentile:96.3(Nuclear Science & Technology)

no abstracts in English

Journal Articles

Small-scale component experiments of the penetration leak characterization test in the ALPHA program

Yamano, Norihiro; Sugimoto, Jun; Maruyama, Yu; Hidaka, Akihide; Kudo, Tamotsu; Soda, Kunihisa

Nuclear Engineering and Design, 145(3), p.365 - 374, 1993/12

 Times Cited Count:2 Percentile:29.81(Nuclear Science & Technology)

no abstracts in English

Journal Articles

OCCD/NEA Specialist Mtg. on Fuel-Coolant Interactions

Akiyama, Mamoru*; Yamano, Norihiro

Nihon Genshiryoku Gakkai-Shi, 35(7), p.630 - 631, 1993/07

no abstracts in English

Journal Articles

Effect of gravity-fed water from a downcomer on coolability of a debris bed

Yamano, Norihiro; Maruyama, Yu; Abe, Yutaka; Soda, Kunihisa

AIChE Symp.Ser., 83(257), p.341 - 346, 1987/00

Experiments to investigate the effect of coolant injection from the bottom of a devris bed have been performed, and an analytical model has been developed.

Oral presentation

Neutronics benchmark analysis of FFTF loss of flow without scram test No.13

Ohgama, Kazuya; Takegoshi, Atsushi*; Hamase, Erina; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki

no journal, , 

no abstracts in English

Oral presentation

Benchmark analysis of FFTF unprotected loss of flow without scram test No.13 with fast reactor plant dynamics analysis code Super-COPD

Hamase, Erina; Ohgama, Kazuya; Kawamura, Takumi*; Doda, Norihiro; Yamano, Hidemasa; Tanaka, Masaaki

no journal, , 

Validation of an analysis model for a plant dynamic analysis code named Super-COPD including neutronics calculation of a one-point reactor kinetics model necessitates the further work on the beyond design basis accident. Therefore, JAEA participated in IAEA benchmark for Loss of Flow without Scram (LOFWOS) test No.13 performed at the Fast Flux Test Facility (FFTF), and the transient analysis at the first blind phase considering with major reactivity feedback mechanisms was carried out. It was observed that the whole plant dynamics analysis could follow the measured data. As a future work, the gap conductance model for transient, the upper plenum of reactor vessel with dividing several regions or multi-dimensional modeling, and the core model that can evaluate the radial heat transfer rate more accurately will be refined.

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