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Journal Articles

Friction factor and Reynolds number correlation for finned tube bundle of the air cooler of Monju reactor

Sotsu, Masutake

Nuclear Engineering and Design, 396( ), p.111893_1 - 111893_27, 2022/09

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

Applicability of the analysis model of the cooling ability under a high temperature and low flowrate condition should be improved to evaluate the plant safety in case of the severe accident due to long-term station blackout. The present study reviews experiments on pressure drop behavior for complicated tube bundle geometry to the flow path and then develops a new correlation equation based on a computational fluid dynamics analysis of the Monju air cooler reflecting the actual geometry and plant data. After a basic simulation model being developed for a typical pressure drop experiment, the simulation is applied to the Monju air cooler that has a finned tube bundle. The obtained relationship between friction factors and Reynolds numbers ranging from 10 to 10,000 are fitted to a power function to derive a correlation equation of the fin tube bundle friction factor. The derived correlation equation can estimate pressure loss in the finned tube bundle more precisely than that in the Monju design. It is applicable to the future reactor design of the air cooler, especially when the cooling ability in low Reynolds number is requested.

Journal Articles

Investigation of thermal expansion model for evaluation of core support plate reactivity in ATWS event

Sotsu, Masutake

Journal of Energy and Power Engineering, 14(8), p.251 - 258, 2020/08

Thermal expansion behavior was investigated for evaluation of the core support plate expansion reactivity in the Unprotected Loss of Heat Sink reactor trip failure event. A possibility of mechanical restraint was investigated in thermal expansion of the core structure for the prototype fast breeder reactor Monju. The reactor core expansion was simulated in a three-dimensional finite element analysis model of the reactor vessel considering detailed temperature distribution of the sodium coolant based on the thermal-hydraulic analysis result of the whole core model. It was found that the thermal expansion of the core was not restrained in the ULOHS evert, although part of the core structure is mechanically restrained.

Journal Articles

Uncertainty evaluation of anticipated transient without scram plant response in the Monju reactor considering reactivity coefficients within the design range

Sotsu, Masutake; Hazama, Taira

Journal of Energy and Power Engineering, 13(11), p.393 - 403, 2019/11

This paper describes methods and results of the uncertainty evaluation of the significant plant response analysis of the reactor trip failure event, i.e. anticipated transient without scram of the Japanese prototype fast breeder reactor Monju. Unprotected loss of heat sink has a relatively large contribution to the core damage frequency due to reactor trip failure. The uncertainty of the allowable time to core damage in this event by plant transient response analysis, so far, has been estimated with considering the range of reactivity coefficients. There are some cases where core damage is considered to be avoided. Specifically, it is assumed that the core damage due to the ULOHS event would be avoided if the sodium temperature at the pump inlet stays below 650$$^{circ}$$C for 1 h; otherwise the possibility of cavitation occurring in the hydrostatic bearing increases. In this study, a method is developed to search for a solution as an inverse problem of multiple input variables that satisfy the temperature condition. This paper, as a first step, describes input conditions and probability to satisfy the temperature are evaluated through analyses treating input parameters, reactivity coefficients and kinetic parameters, as variables within the design range.

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

Unplanned shutdown frequency prediction of FBR MONJU using fault tree analysis method

Sotsu, Masutake

Journal of Energy and Power Engineering, 8(7), p.1286 - 1292, 2014/07

In order to evaluate the operational reliability of Japanese fast breeder reactor MONJU, frequencies of important intermediate events and equipment failures resulting during reactor automatic trip are predicted using fault tree analysis technique for the plant system model. The targeted devices are the following: primary heat transport system (PHTS), secondary heat transport system (SHTS), water and steam system (WS), plant protection system (PPS) and plant control system (PCS). In this paper was estimated the frequency of automatic reactor trips by extracting and analyzing the important intermediate events and equipment failures covering all the derived fault trees of these systems. The analyses predicted 1.2/reactor year (RY) the value of unplanned shut down frequency by the internal factor of the system.

Journal Articles

Multidimensional thermal-hydraulic analysis on natural circulation behavior in ex-vessel fuel storage tank of MONJU

Ono, Jun; Mori, Takero; Sotsu, Masutake; Ohira, Hiroaki

Proceedings of ASME 2013 International Mechanical Engineering Congress and Exposition (IMECE 2013) (DVD-ROM), 9 Pages, 2013/11

The severe accident evaluation on the EVST of MONJU has ever been performed by one-dimensional flow network code "Super-COPD". However, it is difficult to predict thermal-hydraulics in the EVST accurately because the fluid in the EVST is driven by natural circulation. Thus we have performed multidimensional thermal-hydraulic analysis in order to clarify the thermal-hydraulic behavior and evaluate the appropriateness of the flow network model. As a result, it was noted that the multidimensionality on temperature and velocity in the EVST was small enough and the flow network model would be almost appropriate. It should be noted that flow resistance of the supporting plates or the heat transfer center of the cooling coils should be set conservatively for the safety analysis.

Journal Articles

Plant dynamics evaluation of a MONJU ex-vessel fuel storage system during a station blackout

Mori, Takero; Sotsu, Masutake; Honda, Kei; Suzuki, Satoshi*; Ohira, Hiroaki

Journal of Energy and Power Engineering, 7(9), p.1644 - 1655, 2013/09

The prototype fast breeder reactor "MONJU" has an ex-vessel fuel storage system which consists mainly of an ex-vessel fuel storage tank (EVST) and an EVST sodium cooling system. EVST sodium cooling system consists of three independent loops. In this study, an analysis and evaluation of the plant dynamics for the spent fuel and the EVSS structural integrity during an station blackout (SBO) were performed. When the number of cooling loops was not changed and natural circulation occurred in only two loops, the sodium temperature in the EVST increased to approximately 450$$^{circ}$$C. However, the structural integrity of the EVSS was maintained. The analytical results, therefore, help clarify the number of necessary cooling loops for efficient decay heat removal and sodium temperature behavior in an SBO.

Journal Articles

Numerical simulations of upper plenum thermal-hydraulics of Monju reactor vessel using high resolution mesh models

Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake

Journal of Energy and Power Engineering, 7(4), p.679 - 688, 2013/04

Journal Articles

Evaluation of MONJU core damage risk due to control rod function failure

Sotsu, Masutake; Kurisaka, Kenichi

Journal of Power and Energy Systems (Internet), 6(3), p.462 - 471, 2012/12

The limiting conditions of operation defined in the safety regulations for MONJU given the allowed outage time were evaluated by a probabilistic safety assessment technique in our previous study. This paper describes a method to assess the validity of the 24 h allowable time in view of core damage risk, it is necessary to analyze the conditions to be changed when a stuck rod is discovered. The results showed that the timeframe defined in the present safety regulations was concluded to be appropriate.

Journal Articles

Numerical simulations of upper plenum thermal-hydraulics of Monju reactor vessel using high resolution mesh models

Ohira, Hiroaki; Honda, Kei; Sotsu, Masutake

Proceedings of 14th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-14) (CD-ROM), 12 Pages, 2011/09

In order to evaluate the upper plenum thermal-hydraulics of the Monju reactor vessel, we have performed detail calculations under the 40% rated power operational condition using high resolution mesh models by a commercial FVM code, FrontFlow/Red. In this study, we applied a high resolution meshes around the flow holes (FHs) on the inner barrel. We mainly made clear that the thermal-hydraulics did not change largely since the flow rates through the FHs were small enough to the total coolant flow rate but were affected largely incase without FHs on the honeycomb structure.

Journal Articles

Assessment of FBR MONJU accident management reliability in causing reactor trips

Sotsu, Masutake; Kurisaka, Kenichi

Journal of Nuclear Science and Technology, 47(10), p.867 - 883, 2010/10

 Times Cited Count:1 Percentile:10.01(Nuclear Science & Technology)

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor which can supply 280 MW of electricity. The Accident Management (AM) in MONJU is based on three functions: the reactor trip function, the reactor liquid level retaining function, and the decay heat removal function. These are basic safety features, and it is necessary to evaluate the AM capability of these features quantitatively using a PSA technique. This paper describes the AM reactor trip evaluation method comprising plant transient response analysis using the Super-COPD code developed for a best estimate of the plant dynamics of MONJU, the results of this evaluation, and the results of simulator training of plant operators.

Journal Articles

Thermal-hydraulic analysis of MONJU upper plenum under 40% rated power operational condition

Honda, Kei; Ohira, Hiroaki; Sotsu, Masutake; Yoshikawa, Shinji

Proceedings of 8th International Topical Meeting on Nuclear Thermal-Hydraulics, Operation and Safety (NUTHOS-8) (CD-ROM), 12 Pages, 2010/10

In this study, we calculated the thermal hydraulics of the upper plenum of MONJU by the detailed analysis model using commercial FVM code, FrontFlow/Red. The present analysis model simulates all structures with high resolution meshes. The 1st order upwind and 2nd order central difference scheme were applied to the advection and diffusion terms, respectively. And RNG $$k$$-$$epsilon$$ model was applied to turbulence modeling. These calculation results indicated that the structures installed in the plenum except for UIS did not affect largely to the temperature and velocity, the flow characteristics in the present results had similar tendencies with porous media approached applied to the UCS region and that the difference between the temperature measured in the UCS region and that of SA outlets is relatively small.

Journal Articles

Evaluation of Monju core damage risk with change of AOT using probabilistic method

Sotsu, Masutake; Kurisaka, Kenichi

Journal of Power and Energy Systems (Internet), 4(1), p.84 - 93, 2010/02

Monju is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.

Journal Articles

Evaluation of MONJU core damage risk with change of AOT using probabilistic method

Sotsu, Masutake; Kurisaka, Kenichi

Proceedings of 17th International Conference on Nuclear Engineering (ICONE-17) (CD-ROM), 9 Pages, 2009/06

MONJU is a sodium-cooled, loop-type prototype fast breeder reactor with three primary cooling loops which can supply 280 MW of electricity. Limiting conditions of operation (LCO) defined in the safety regulations in MONJU given the allowed outage time (AOT) are evaluated using a PSA technique. The result indicates the possibility of limit extension and some prospects that we should examine.

Journal Articles

Unplanned shutdown frequency prediction of FBR MONJU using fault tree analysis method

Sotsu, Masutake; Yamada, Fumiaki*

Proceedings of 13th International Conference on Nuclear Engineering (ICONE-13), 0 Pages, 2005/05

MONJU is a sodium cooled, loop-type prototype fast breeder reactor which can supply 280MW of electricity to the grid. The generated heat at the reactor core is removed by three loops of primary heat transport system (PHTS), each of those is thermally connected through individual intermediate heat exchanger (IHX) to another clloant eirculation loop of secondary heat transport system (SHTS). The turbine generator is driven by steam generated at three evaporators and super heaters installed at the SHTS.

Journal Articles

Development of living PSA system for FUGEN NPS

Sotsu, Masutake; Iguchi, Yukihiro

PSAM-5, 0 Pages, 2000/00

None

Journal Articles

DEVEIOPMENT OF AN EMERGENCY RESPONSE SUPPORT SYSTEM FOR THE FUGEN NPS

Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Koichi

PSA'99, 0 Pages, 1999/00

None

Journal Articles

Development of living PSA system for FUGEN NPS

Sotsu, Masutake; Iguchi, Yukihiro

PSAM-5, 0 Pages, 1999/00

None

Journal Articles

On-line support of the FUGEN plant using MARS

Iguchi, Yukihiro; Sotsu, Masutake; Mizuno, Koichi

Proceedings of 7th International Conference on Nuclear Engineering (ICONE-7), 0 Pages, 1999/00

None

Journal Articles

PSA related activities and application to the maintenance of FUGEN

Sotsu, Masutake; Iguchi, Yukihiro;

Proceedings of 3rd International Conference on Nuclear Engineering (ICONE-3), 0 Pages, 1995/00

None

30 (Records 1-20 displayed on this page)