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Onoda, Yuichi; Ishida, Shinya; Fukano, Yoshitaka; Kamiyama, Kenji; Yamano, Hidemasa; Kubo, Shigenobu; Shibata, Akihiro*; Bertrand, F.*; Seiler, N.*
Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10
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:34.39(Nuclear Science & Technology)Ishida, Shinya; Fukano, Yoshitaka; Tobita, Yoshiharu; Okano, Yasushi
Proceedings of 2023 International Congress on Advanced in Nuclear Power Plants (ICAPP 2023) (Internet), 8 Pages, 2023/04
Ishida, Shinya; Fukano, Yoshitaka
Nihon Kikai Gakkai Rombunshu (Internet), 88(911), p.21-00304_1 - 21-00304_11, 2022/07
In previous studies, the reliability and validity of the SAS4A code was enhanced by applying Phenomena Identification and Ranking Table (PIRT) approach to the Unprotected Loss of Flow (ULOF). SAS4A code has been developed to analyze the early stage of Core Disruptive Accident (CDA), which is named Initiating Phase (IP). In this study, PIRT approach was applied to Unprotected Transient over Power (UTOP), which was one of the most important and typical events in CDA as well as ULOF. The phenomena were identified by the investigation of UTOP event progression and physical phenomena relating to UTOP were ranked. 8 key phenomena were identified and the differences in ranking between UTOP and ULOF were clarified. The code validation matrix was completed and an SAS4A model, which was not validated in ULOF, was identified and validated. SAS4A code became applicable to various scenarios by using PIRT approach to UTOP and the reliability and validity of SAS4A code were significantly enhanced.
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.
Ishida, Shinya; Fukano, Yoshitaka
Dai-25-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (Internet), 4 Pages, 2021/07
no abstracts in English
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Mechanical Engineering Journal (Internet), 7(3), p.19-00523_1 - 19-00523_17, 2020/06
The Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to indicate the reliability of SAS4A code sufficiently and objectively. Based on this approach, issue and objective were clarified, plant design and scenario were defined, FOM and key phenomena were selected, and the code validation test matrix was completed with the results of investigation about analysis models and test cases. The results of the test analysis corresponding to this matrix show that the SAS4A models required for the IP evaluation were sufficiently validated. Furthermore, the validation with this matrix is highly reliable, since this matrix represents the comprehensive validation that also considers the relation between physical phenomena. In this study, the reliability and validity of SAS4A code were significantly enhanced by using PIRT approach to the sufficient level for CDA analyses in SFR.
Ishida, Shinya; Kawada, Kenichi; Fukano, Yoshitaka
Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 10 Pages, 2019/05
Core Disruptive Accident (CDA) has been considered as one of the important safety issues in the severe accident evaluation of Sodium-cooled Fast Reactor (SFR), and SAS4A code is developed for Initiating Phase (IP) of CDA. Phenomena Identification and Ranking Table (PIRT) approach was applied to the validation of SAS4A code in order to enhance its reliability in this study. SAS4A was validated in the following steps: (1) selection of the figure of merit (FOM) corresponding to Unprotected Loss Of Flow (ULOF) which is one of the most important and typical events in CDA, (2) identification of the phenomena involved in ULOF, (3) ranking the important phenomena, (4) development of the code validation test matrix, and (5) test analyses for validation corresponding to the test matrix. The reliability and validity of SAS4A code were significantly enhanced by this validation with PIRT approach.
Fukano, Yoshitaka
Journal of Nuclear Engineering and Radiation Science, 5(1), p.011001_1 - 011001_13, 2019/01
Local subassembly faults (LFs) have been considered to be of greater importance in safety evaluation in sodium-cooled fast reactors (SFRs) because fuel elements were generally densely arranged in the subassemblies (SAs) in this type of reactors, and because power densities were higher compared with those in light water reactors. A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of LFs. Although an evaluation on the consequences of HTIB using SAS4A code was performed in the past study, SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in an SFR by this developed SAS4A code clarified that the conclusion in the past study was almost same as that in this study. Furthermore SAS4A code was newly validated using four in-pile experiments which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study. Thus the methodology of HTIB evaluation was established in this study together with the past study and is applicable to HTIB evaluations in other SFRs.
Imaizumi, Yuya; Yamada, Fumiaki; Arikawa, Mitsuhiro*; Yada, Hiroki; Fukano, Yoshitaka
Mechanical Engineering Journal (Internet), 5(4), p.18-00083_1 - 18-00083_11, 2018/08
A calculation program was developed to evaluate and discuss the effectiveness of the countermeasures such as sodium pump-up and siphon-breaking against the loss-of-reactor-level (LORL) where the coolant circulation path is lost in loop-type sodium-cooled fast reactors. Due to the non-negligible possibility obtained by probabilistic risk assessment (PRA), sodium leakages in two points both occurred in primary heat transport system (PHTS) was assumed in this study. In addition, the crack size was discussed and evaluated realistically, instead of the value that was assumed in the conventional studies. Representative sequences and leakage positions were chosen, and the sodium level transient in reactor vessel (RV) was calculated. The calculations were also conducted where the larger crack size was set for the second leakage, in order to investigate additional requirements to maintain the RV sodium level. The evaluation results clarified that the coolant circulation loop can be maintained even after the second leakage in PHTS, taking into account the effects by the countermeasures.
Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*
Journal of Nuclear Science and Technology, 54(11), p.1178 - 1189, 2017/11
Times Cited Count:3 Percentile:26.37(Nuclear Science & Technology)Fuel subassemblies (FSAs) of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be negligible compared with the core damage due to ATWS or PLOHS in the viewpoint of both frequency and consequence.
Yamada, Fumiaki; Imaizumi, Yuya; Nishimura, Masahiro; Fukano, Yoshitaka; Arikawa, Mitsuhiro*
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
The loss-of-reactor-level (LORL) is one of the loss-of-heat-removal-system (LOHRS) of beyond-DBA (BDBA) severe accident. An evaluation method for the LORL which is caused by the coolant leakage in two positions of the primary heat transport system (PHTS) was developed for prototype JSFR which is loop-type sodium-cooled fast reactor. The secondary leakage in cold standby which occurred in different loop from that of the first leakage in rated power operation can lead LORL by excessive declining of the sodium level. Therefore, the sodium level behavior in RV was studied in a representative accident sequence by considering the sodium pumping up into RV, siphon-breaking to stop pumping out from RV and maintain the sodium level, and calculation programs for the transient sodium level in RV. The representative sequence with lowest sodium level was selected by considering combinations of possible leakage positions. As a result of the evaluation considering the countermeasures above, it was revealed that the LOHRS can be prevented by maintaining the sodium level for the operation of decay heat removal system, even in the leakages in two positions of PHTS which corresponds to BDBA.
Fukano, Yoshitaka
Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07
An evaluation on the consequences of a hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) using SAS4A code was also performed in the past study. SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in Monju by this developed SAS4A code was performed. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the plant protection system triggered by either of 116 percent over power or delayed neutron detector trip signals. Therefore the conclusion in the past study that the consequences of HTIB event would be much less severe than that of unprotected loss-of-flow event was strongly supported by this study. Furthermore SAS4A code was newly validated using an in-pile experiment which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study.
Mori, Takero; Ohira, Hiroaki; Sotsu, Masutake; Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Since safety measures against severe accidents (SAs) such as a long-term station blackout (SBO) are required for Japanese prototype fast breeder reactor Monju, a validation is necessary for the plant dynamics code during SBO. In order to take into account the phenomena in natural circulation: a heat transfer among subassemblies and a flow redistribution, a whole core model has been developed for the plant dynamics code, Super-COPD. This model has been validated by test results of natural circulation in actual facility. In this study, this whole core model was applied to Monju core to evaluate safety measures against SBO, and the pressure loss model of Monju was validated by comparing with results of the plant trip test from the power of 40%. In addition, an analysis was conducted for SBO to investigate the applicability of this model to Monju. The applicability of this model was confirmed by comparing with analytical results using the model without heat transfer between assemblies.
Fukano, Yoshitaka
Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 9 Pages, 2017/04
Loss-of-heat-removal-system (LOHRS) events are identified as some of most dominant severe accident sequences in a sodium-cooled fast reactor. Safety margins against significant core damage in LOHRS events were therefore studied in this paper assuming large fuel-cladding gap and fuel cladding failure. It was clarified through analyses by the developed code that neither fuel melting nor further mechanical pin failure occurs owing to large fuel-cladding gap and fuel cladding failure. It was therefore concluded that large safety margins against significant core damage are provided during LOHRS events. These results will be effectively used in formulating the safety criteria for severe accidents or beyond-design-basis-accidents as one of the supporting evidences to be seriously considered.
Fukano, Yoshitaka
Proceedings of 11th International Topical Meeting on Nuclear Reactor Thermal Hydraulics, Operation and Safety (NUTHOS-11) (USB Flash Drive), 12 Pages, 2016/10
Probabilistic and deterministic safety assessments and experimental studies on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious fuel pin failures have been considered to be the most dominant initiators of LFs in these probabilistic assessments because of its high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. The four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) were identified in the past study. All the mechanisms for FEFPA analysis including thermal, mechanical and chemical propagation were modeled into a safety assessment code which was applicable to arbitrary SFRs. Safety analyses on FEFPA of Japanese experimental fast reactor (JOYO), Japanese prototype fast breeder reactor (Monju), Japanese prototype fast breeder reactor with upgraded reactor core (Upgraded Monju) and Japan sodium-cooled fast reactor (JSFR) were performed using this methodology. Although analytical results were different owing to the different core designs in four SFRs, it was clarified in this study that FEFPA was highly unlikely in these SFRs. These results also suggest future possibility of long-term run-beyond-cladding-breach operation which would enhance the economic efficiency in SFRs.
Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*
Proceedings of 13th Probabilistic Safety Assessment and Management Conference (PSAM-13) (USB Flash Drive), 12 Pages, 2016/10
Fuel subassemblies of fast breeder reactors (FBRs) are densely arranged and have high power densities. Therefore, the local fault (LF) has been considered as one of the possible initiating events of severe accidents. According to the LF evaluation under the condition of total flow blockage of one sub-channel in the analyses of design basis accident (DBA) for Monju, it was confirmed that the pin failures were limited locally without severe core damage. In addition, local flow blockage (LB) of 66% central planar in the subassembly was investigated as one of the beyond-DBA. However, it became clear that these deterministic analyses were not based on a realistic assumption by experimental studies. Therefore, PRA on LF which was initiated from LB was performed reflecting the state-of-the-art knowledge in this study. As the result, damage propagation from LF caused by LB in Monju can be included in CDF of ATWS or PLOHS in the viewpoint of both probability and consequence.
Fukano, Yoshitaka
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.347 - 356, 2016/04
A hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) gives most severe consequences among a variety of initiating event for local faults (LFs) such as adventitious fuel pin failure, local overpower and flow blockages. An evaluation on the consequences of HTIB was performed in the past study as an enveloping event analysis among a wide spectrum of initiating events for LFs. Although the SAS4A code has been validated by a number of in-pile experiments in the French CABRI and U.S. TREAT reactors, most of them were performed under loss-of-flow (LOF) combined with transient-overpower (TOP) conditions. The changing rate of flow in HTIB is much more rapid than that in these LOF test. Therefore additional and more expedient validation was undergone in this study using the TIB experiments which were performed in the French SCARABEE reactor especially for its modeling of coolant boiling, cladding melting, molten cladding motion, fuel melting and wrapper tube failure anticipated to occur during an HTIB condition. Four TIB experiments were performed with 19 or 37 pin bundles in the SCARABEE reactor. SAS4A analyses on these experiments showed good agreement with those experimental results in the following phenomena which were anticipated to occur during an HTIB condition: (1) Timing and progress of coolant boiling and cladding dryout; (2) Timing of cladding melting and behavior of molten cladding relocation; (3) Timing and progress of fuel melting, disruption and relocation; (4) Timing of wrapper tube melt-through. Therefore it can be concluded that the validity of SAS4A application to safety evaluation on the consequence of HTIB in the past study is enhanced in this study.
Imaizumi, Yuya; Fukano, Yoshitaka
Proceedings of 2016 International Congress on Advances in Nuclear Power Plants (ICAPP 2016) (CD-ROM), p.357 - 363, 2016/04
SAS4A is the code which has been developed to analyze the initiation phase of the core-disruptive accident in SFRs. The code of which can be adopted in a safety licensing needs to be validated through the experimental results. In this study, the code was validated by the experimental results of CABRI project which was conducted in the framework of international collaboration. The selected three CABRI tests of this validation target were all conducted using annular fuel pellets with middle burn-up (6.4 at%). Severe conditions consisted of loss of flow (LOF) and transient overpower (TOP) was imposed in the tests to reproduce similar conditions when unprotected-loss-of-flow (ULOF) occurred in SFRs. The TOP were imposed when coolant temperature reached around the boiling point or several seconds after the cladding melting. The results of the SAS4A analyses agreed well with the CABRI results such as the timing of coolant boiling, voiding extension during the coolant boiling, and the relocation and refrozen behaviors of the molten fuel. Consequently, the coolant boiling and fuel relocation models of SAS4A were validated by these analyses.
Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro
Journal of Nuclear Science and Technology, 52(9), p.1122 - 1132, 2015/09
Times Cited Count:3 Percentile:24.31(Nuclear Science & Technology)Experimental studies, deterministic approaches, and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated-transient-without-scram or protected-loss-of-heat-sink events from both the viewpoint of occurrence probability and consequences.