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Journal Articles

Validation study of initiating phase evaluation method for the core disruptive accident in an 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.

Journal Articles

Development and validation of SAS4A code and its application to analyses on severe flow blockage accidents in a sodium-cooled fast reactor

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.

Journal Articles

Development of LORL evaluation method and its application to a loop-type sodium-cooled fast reactor

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.

Journal Articles

Updating of local blockage frequency in the reactor core of SFR and PRA on consequent severe accident in Monju

Nishimura, Masahiro; Fukano, Yoshitaka; Kurisaka, Kenichi; Naruto, Kenichi*

Journal of Nuclear Science and Technology, 54(11), p.1178 - 1189, 2017/11

 Times Cited Count:2 Percentile:42.02(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.

Journal Articles

Development and validation of evaluation method on hypothetical total instantaneous flow blockage in sodium-cooled fast reactors and its application to a middle size SFR

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.

Journal Articles

Development of the severe accident evaluation method on second coolant leakages from the PHTS in a loop-type sodium-cooled fast reactor

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.

Journal Articles

Analytical study on safety margins against significant core damage during loss-of-heat-removal-system events in a sodium-cooled fast reactor

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.

Journal Articles

Validation and applicability of reactor core modeling in a plant dynamics code during station blackout

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.

Journal Articles

Analytical studies on fuel element failure propagation due to adventitious fuel pin failure in small to large size sodium-cooled fast reactors

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.

Journal Articles

PRA on mixed foreign substances into core of Japanese prototype FBR

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.

Journal Articles

SAS4A analyses of SCARABEE in-pile experiments simulating hypothetical total instantaneous flow blockages in SFRs

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.

Journal Articles

SAS4A analyses of CABRI in-pile experiments simulating unprotected-loss-of-flow accidents in SFRs

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.

Journal Articles

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

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:53.75(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.

Journal Articles

Safety margins after failure of fuel cladding during protected loss-of-heat-sink accidents in a sodium-cooled fast reactor

Fukano, Yoshitaka; Nishimura, Masahiro; Yamada, Fumiaki

Proceedings of 16th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-16) (USB Flash Drive), p.5687 - 5698, 2015/08

The following safety criteria for anticipated operational occurrences are commonly and uniformly employed for all the DBAs in the Japanese prototype sodium-cooled fast reactor to prevent fuel melting and cladding failure:(a) Maximum fuel temperature shall be below the melting point,(b) Maximum cladding temperature shall be below 830$$^{circ}$$C, and (c) Maximum coolant temperature shall be below the boiling point. Cladding failure is allowed, on the contrary to that, in beyond DBAs (BDBAs) or severe accidents (SAs), whereas the core cooling capability is also needed to be secured as in DBAs. No fuel melting enables this by keeping the core in a coolable geometry, and is thus conservatively required even under such a condition. Protected loss-of-heat-sink (PLOHS) events are identified as one of the most dominant sequences. Safety margins for significant core damage in PLOHS events were therefore studied in this paper assuming fuel cladding failure. The following three possible mechanisms leading to degradation of the core were then identified to be scrutinized by a thorough and state-of-the-art review of open papers on the phenomena anticipated to occur under cladding failure conditions:(1) Fuel melting due to fuel-sodium reaction product (FSRP) formation, (2) Thermal transient due to FP gas impingement from adjacent failed fuel pins, and (3) Mechanical load due to the same FP gas impingement. It was clarified through simulation analyses on each phenomenon mentioned above using the FUCA code that there was no significant core damage at the coolant temperatures of up to 950$$^{circ}$$C. It was therefore concluded that large safety margins are provided during PLOHS events even in failure of fuel cladding.

Journal Articles

Validation of core cooling capability analysis in Monju during guillotine pipe break at primary heat transport system

Yamada, Fumiaki; Arikawa, Mitsuhiro*; Fukano, Yoshitaka

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 7 Pages, 2015/05

In sodium-cooled fast reactor, since the coolant does not need to be pressurized, a pipe break due to the internal pressure does not occur physically. For safety margin in Japanese prototype fast breeder reactor (Monju), the guillotine pipe break accident, i.e., loss of integrity (LOPI) has been analyzed as an extreme assumption for beyond design basis accidents (B-DBAs) in the licensing application for the permit. The cooling capability of the core was re-evaluated in this paper during a large-scale, more specifically guillotine pipe break at the primary heat transport system (PHTS) in Monju, newly considering the following latest findings: (a) Experimental data on sodium boiling in fuel assemblies, (b) Actual PHTS pump coast-down characteristics, and (c) Transient burst test data on irradiated fuel claddings. The analysis models were the validated and simulations were re-performed also using the actual Monju data such as the response time to the trip signals, etc. As a result, it was clarified that the ratio of failed fuel claddings does not exceed around 3% of all of fuel assemblies, as in the past licensing analysis. The safety has been reconfirmed to be secured without significant core damage even under an extreme assumption of a double-ended guillotine pipe break at the PHTS in Monju.

Journal Articles

SAS4A analysis on hypothetical total instantaneous flow blockage in SFRs based on in-pile experiments

Fukano, Yoshitaka

Annals of Nuclear Energy, 77, p.376 - 392, 2015/03

 Times Cited Count:7 Percentile:22.72(Nuclear Science & Technology)

Local fault (LF) has been historically considered as one of the possible causes of severe accidents in sodium-cooled fast reactors (SFRs). Safety assessments on the consequences of three major LF initiators, adventitious-fuel-pin-failures, local over-powers and large flow blockages, have already been performed for the Japanese prototype fast breeder reactor (Monju) and resulted in limited damage of within a fuel subassembly. Although the frequency of hypothetical total instantaneous flow blockage at the coolant inlet of a fuel subassembly (HTIB) is negligible, in-pile experiments on this event have been performed because it gives the most severe consequences among the above-mentioned LF initiators. Therefore a most rapid damage propagation scenario was newly proposed in this study based on past experiments, and a safety evaluation was performed on the consequences of HTIB in Monju using the SAS4A code, additionally validated by an in-pile experiment especially on the HTIB-related phenomena. It was clarified through the evaluation that the consequences of HTIB were much less severe than that of the unprotected loss of flow (ULOF) events which represent the most severe accidents in Monju. Therefore it was concluded that all the consequences of LFs in Monju can be comprised in that of ULOF. This evaluation method would be also applicable to enveloping safety assessments on LFs in arbitrary SFRs.

Journal Articles

Development of safety assessment methodology on fuel element failure propagation in SFR and its application to Monju

Fukano, Yoshitaka

Journal of Nuclear Science and Technology, 52(2), p.178 - 192, 2015/02

 Times Cited Count:3 Percentile:53.75(Nuclear Science & Technology)

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 were considered to be the most dominant initiators of LFs in these probabilistic assessments because of high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Four possible mechanisms of fuel element failure propagation from adventitious fuel pin failure (FEFPA) are extracted from the state-of-the-art review of open papers. All the methods for FEFPA including thermal, mechanical, chemical propagation are modeled into a safety assessment code which is applicable to all the SFRs with the development of some methods. Furthermore the assessment on FEFPA in Monju was performed using this methodology. It was clarified that FEFPA was highly unlikely and limited within at most one sub-assembly in Monju by the multiple and diverse detection and shutdown systems for FEFPA even assuming the propagation. This result also suggests future possibility of run beyond cladding breach which enhances the economic efficiency in Monju.

Journal Articles

Development of natural circulation analytical model in Super-COPD code and evaluation of core cooling capability in Monju during a station blackout

Yamada, Fumiaki; Fukano, Yoshitaka; Nishi, Hiroshi; Konomura, Mamoru

Nuclear Technology, 188(3), p.292 - 321, 2014/12

 Times Cited Count:8 Percentile:25.95(Nuclear Science & Technology)

The capability of natural circulation for core cooling has been evaluated in detail for a station blackout (SBO) event induced by an earthquake and a subsequent tsunami hit. The evaluation was prompted by the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. The plant dynamics computer code Super-COPD was used for the evaluation, which has been validated by analyses of preliminary test results on the natural circulation in Monju. As a result, it was concluded that natural circulation of the sodium coolant will enable the decay heat from the core to be removed under such an SBO condition.

Journal Articles

Local flow blockage analysis with checkerboard configuration in a wire wrapped fuel subassembly using the ASFRE code

Nishimura, Masahiro; Fukano, Yoshitaka

Proceedings of 10th International Topical Meeting on Nuclear Thermal Hydraulics, Operation and Safety (NUTHOS-10) (USB Flash Drive), 11 Pages, 2014/12

Deterministic evaluation of localflow blockage (LB) on the basis of state-of-the-art knowledge was performed using ASFRE code. In order to evaluate the effect of the realistic accidental condition, nominal power and flow rate were used for the analyses. Moreover the realistic blockage feature in the subassembly was newly adopted on the basis of existing experimental data which means LB in hound's-tooth pattern at a cross-section was assumed for the fuel subassemblies of wire spacer type. As the result, it was founds that the temperature increase in the downstream of LB was smaller than that in the past safety licensing because the flow pass is available around the blockage. And it was concluded that LB never lead to the large core damage from the evaluation results even if the blockage conditions beyond design criteria are assumed.

Journal Articles

Probability of adventitious fuel pin failures in fast breeder reactors and event tree analysis on damage propagation up to severe accident in Monju

Fukano, Yoshitaka; Naruto, Kenichi*; Kurisaka, Kenichi; Nishimura, Masahiro

Proceedings of 12th Probabilistic Safety Assessment and Management Conference (PSAM-12) (USB Flash Drive), 12 Pages, 2014/06

Experimental studies, deterministic and probabilistic and 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 were considered to be the most dominant initiators of LFs in these PRAs because of high frequency of occurrence during reactor operation and possibility of subsequent pin-to-pin failure propagation. Therefore event tree analysis (ETA) on fuel element failure propagation initiated from adventitious fuel pin failure (FEFPA) in Monju was performed in this study based on state-of-the-art knowledge on experimental and analytical studies on FEFPA and reflecting latest operation procedure at emergency in Monju. Probability of adventitious fuel pin failures in SFRs which is the initiating event of this ETA was also updated in this study. It was clarified that FEFPA in Monju was negligible and could be included in core damage fraction of the anticipated transient without scram and protected loss of heat sink in the viewpoint of both probability and consequence.

53 (Records 1-20 displayed on this page)