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Journal Articles

Development of evaluation method for in-place cooling of residual core materials in core disruptive accidents of SFRs

Imaizumi, Yuya; Aoyagi, Mitsuhiro; Kamiyama, Kenji; Matsuba, Kenichi; Ganovichev, D. A.*; Baklanov, V. V.*

Proceedings of 27th International Conference on Nuclear Engineering (ICONE-27) (Internet), 11 Pages, 2019/05

The cooling of the residual core materials after the fuel discharge from the core in the accident of SFRs can significantly affect the distribution fraction of the core materials, which is an important factor for the in-vessel retention (IVR). For the evaluation of the cooling of the residual core materials which is called ``in-place cooling'', behavior in a SFR core was analyzed preliminary by SIMMER-III. Based on the analysis result, method of phenomena identification and ranking table (PIRT) was applied. Fundamental experiment focusing on three thermal-hydraulic phenomena those were extracted by PIRT was considered in order to investigate them and utilize it for validation of the SIMMER-III. To achieve continuous oscillation of sodium level which can occur in the phase of in-place cooling of SFRs, analytical survey was conducted by SIMMER-III. As a result of that, the effects of experimental conditions on the oscillation amplitude and the duration time were clarified quantitatively, which are necessary to determine the specific experimental conditions.

Journal Articles

Effect of porosity distribution on two-phase pressure drop in a packed bed

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 3 Pages, 2018/11

In the evaluation of the in-place cooling which is for the residual core materials in the severe accident of sodium-cooled fast reactors, pressure loss of two-phase flow in debris bed is one of the important factors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, experiment to measure the pressure loss in a packed bed of non-homogeneous porosity distribution was conducted, and the Lipinski model was modified dividing the cross section to evaluate the pressure loss in it. As a result, it was confirmed that agreement of the experimental values with the values by modified Lipinski model was better than that with the original Lipinski model.

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

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

Fundamental experiments of jet impingement and fragmentation simulating the fuel relocation in the core disruptive accident of sodium-cooled fast reactors

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki

Proceedings of 2017 International Congress on Advances in Nuclear Power Plants (ICAPP 2017) (CD-ROM), 5 Pages, 2017/04

In order to simulate the typical accident conditions of the fuel relocation phase in SFRs, the molten alloy of low melting point was discharged into a shallow water pool. The distance between the nozzle exit and the bottom plate was set to a value which was indicated to be insufficient to fragment. As the experimental result, the melt jet reached the bottom plate, and dispersed in all directions along the plate, together with the progress of fragmentation. In addition, the melt temperature on the bottom plate decreased rapidly along the radius direction. These results suggest that the fragmentation which would accompany this rapid cooling would be enhanced by the plate. This enhancement would be caused by the extension of the melt-water interface when the melt was dispersed forcibly by the plate. The solidified debris remained after the discharge showed remarkable fragmentation which was assumed to be caused by the formations of small vapor bubbles in the interface.

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

Flight demonstration of Cu(In,Ga)Se$$_{2}$$ thin-film solar cells using micro-satellite

Kawakita, Shiro*; Imaizumi, Mitsuru*; Kibe, Koichi*; Oshima, Takeshi; Ito, Hisayoshi; Yoda, Shinichi*; Nakamura, Yuya*; Nakasuka, Shinichi*

Proceedings of 7th International Workshop on Radiation Effects on Semiconductor Devices for Space Application (RASEDA-7), p.61 - 64, 2006/10

no abstracts in English

Oral presentation

Oral presentation

Analyses of in-pile experiments by an analysis code on initiating phase of core disruptive accident in sodium cooled fast reactors, 4 Analysis of EFM1 test

Imaizumi, Yuya; Fukano, Yoshitaka

no journal, , 

The test result of EFM1 in the CABRI in-pile experiment which was conducted as an internationally collaborated project was analyzed by SAS4A code. The code was developed for the analysis of initiation phase of core disruptive accident. In the EFM1 test, transient overpower was imposed after the cladding melting and coolant boiling which was due to the previously imposed loss of flow. The behavior of large relocation and refreezing of the molten fuel which was caused by the heating peak after the fuel melting was one of the important point in the analysis. As a result, good agreements were observed between the results of experiment and analysis such as timings of coolant boiling, extension of boiling area and behavior of molten fuel motion.

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 2; Experimental study on molten-metal discharge into a shallow water pool

Imaizumi, Yuya; Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Suzuki, Toru; Emura, Yuki

no journal, , 

For the study of the fuel-coolant interaction and deposition behavior of the molten core material on the in-vessel structure in the relocation phase of the core disruptive accident in SFR, several experiments discharging low-melting-point alloy into a shallow water pool were conducted. The experimental results and the mechanism discussion are reported.

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 1; Overall plan

Kamiyama, Kenji; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Suzuki, Toru; Emura, Yuki

no journal, , 

no abstracts in English

Oral presentation

Oral presentation

Safety evaluations for MONJU during decommissioning phase, 2; Conservative thermal evaluations on fuel integrity

Mori, Takero; Sotsu, Masutake; Imaizumi, Yuya; Yoshimura, Kazuo; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Studies on relocation behavior of molten core materials in the core disruptive accident of sodium-cooled fast reactors, 4; Effects of test conditions on dispersion and sedimentation of molten-material in a shallow water pool

Emura, Yuki; Matsuba, Kenichi; Isozaki, Mikio; Imaizumi, Yuya; Kamiyama, Kenji

no journal, , 

A series of simulation tests discharging a molten alloy with low-melting point into a shallow water pool is being conducted in order to investigate fuel-coolant interaction and sedimentation behavior of molten core materials on the structure of core inlet coolant plenum in the core disruptive accident of sodium-cooled fast reactors. The present study shows effects of the initial temperature of the molten alloy, the confined coolant volume, etc. on the test results.

Oral presentation

Study on effectiveness evaluations of countermeasures against severe accidents in fast reactor, 6; Effectiveness evaluations of countermeasures against anticipated transient without scram

Imaizumi, Yuya; Yamada, Fumiaki; Nishimura, Masahiro; Mori, Takero; Fukano, Yoshitaka

no journal, , 

no abstracts in English

Oral presentation

Study on coolant behavior in damaged core of sodium-cooled fast reactor, 4; Evaluation of pressure drop through simulated debris bed in consideration of porosity distribution

Kurisaki, Tatsuya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

no journal, , 

The pressure loss of two-phase flow in debris bed is one of the important factor for coolability evaluation of the inplace-cooling for the residual core-materials in the severe accident of sodium-cooled fast reactors. Although Lipinski model is already proposed for the pressure loss evaluation, the accuracy would decrease when the porosity is not homogeneous. Thus, the evaluation method by dividing the cross section was developed for the two-phase flow in packed bed of non-homogenious porosity, and compared the calculated values with experimental data for the validation.

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