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Journal Articles

Comparative methodology between actual RCCS and downscaled heat-removal test facility

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 133, p.830 - 836, 2019/11

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.

Journal Articles

Improvement of heat-removal capability using heat conduction on a novel reactor cavity cooling system (RCCS) design with passive safety features through radiation and natural convection

Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*

Annals of Nuclear Energy, 122, p.201 - 206, 2018/12

 Percentile:100(Nuclear Science & Technology)

A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. This study addresses an improvement of heat-removal capability using heat conduction on the RCCS. As a result, a heat flux removed by the RCCS could be doubled; therefore, it is possible to halve the height of the RCCS or increase the thermal reactor power.

Journal Articles

Experimental study on heat removal performance of a new Reactor Cavity Cooling System (RCCS)

Hosomi, Seisuke*; Akashi, Tomoyasu*; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*; Takamatsu, Kuniyoshi

Proceedings of 11th Korea-Japan Symposium on Nuclear Thermal Hydraulics and Safety (NTHAS-11) (Internet), 7 Pages, 2018/11

A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We started experiment research with using a scaled-down test section. Three experimental cases under different emissivity conditions were performed. We used Monte Carlo method to evaluate the contribution of radiation to the total heat released from the heater. As a result, after the heater wall was painted black, the contribution of radiation to the total heat could be increased to about 60%. A high emissivity of RPV surface is very effective to remove more heat from the reactor. A high emissivity of the cooling part wall is also effective because it not only increases the radiation emitted to the ambient air, but also may increase the temperature difference among the walls and enhance the convection heat transfer in the RCCS.

Journal Articles

New reactor cavity cooling system using novel shape for HTGRs and VHTRs

Takamatsu, Kuniyoshi; Hu, R.*

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

continuous closed regions; one is an ex-reactor pressure vessel (RPV) region and another is a cooling region having heat transfer area to ambient air assumed at 40 ($$^{circ}$$C). The RCCS uses novel shape so that the heat released from the RPV can be removed efficiently with radiation and natural convection. Employing the air as the working fluid and the ambient air as the ultimate heat sink, the new RCCS design greatly reduces the possibility of losing the heat sink for decay heat removal. Therefore, HTGRs and VHTRs adopting the new RCCS design can avoid core melting owing to overheating the fuels.

JAEA Reports

An Evaluation study of measures for prevention of Re-criticality in sodium-cooled large FBR with MOX fuel

JNC-TN9400 2000-038, 98 Pages, 2000/04

JNC-TN9400-2000-038.pdf:7.49MB

As an effort in the feasibility study on commercialized Fast Breeder Reactor cycle systems, an evaluation of the measures to prevent the energetic re-criticality in sodium-cooled large MOX core, which is one of the candidates for the commercialized reactor, has been performed. The core disruptive accident analysis of Demonstration FBR showed that the fuel compaction of the molten fuel by radial motion in a large molten core pool had a potential to drive the severe super-prompt re-criticality phenomena in ULOF sequence. ln order to prevent occurrence of the energetic re-criticality, a subassembly with an inner duct and the removal of a part of LAB are suggested based on CMR (Controlled Material Relocation) concept. The objective of this study is the comparison of the effectiveness of CMR among these measures by the analysis using SIMMER-III. The molten fuel in the subassembly with inner duct flows out faster than that from other measures. The subassembly with inner duct will work effectively in preventing energetic re-criticality. Though the molten fuel in the subassembly without a part of LAB flows out a little slower, it is still one of the promising measures. However, the UAB should be also removed from the same pin to prevent the fuel re-entries into the core region due to the pressurization by FCl below the core, unless it disturbs the core performance. The effect of the axial fuel length of the center pin to CMR behavior is small, compared to the effect of the existence of UAB.

JAEA Reports

Phase Change Predictions for Liquid Fuel in Contact with Steel Structure using the Heat Conduction Equation

Brear, D. J.

PNC-TN9410 98-005, 53 Pages, 1998/01

PNC-TN9410-98-005.pdf:2.09MB

When liquid fuel makes contact with steel structure the liquid can freeze as a crust and the structure can melt at the surface. The melting and freezing processes that occur can influence the mode of fuel freezing and hence fuel relocation. Furthermore the temperature gradients established in the fuel and steel phases determine the rate at which heat is transferred from fuel to steel. In this memo the 1-D transient heat conduction equations are applied to the case of initially liquid UO$$_{2}$$ brought into contact with solid steel using up-to-date materials properties. The solutions predict criteria for fuel crust formation and steel melting and provide a simple algorithm to determine the interface temperature when one or both of the materials is undergoing phase change. The predicted steel melting criterion is compared with available experimental results.

JAEA Reports

None

*; Hibiki, Takashi*; *; Tobita, Yoshiharu

PNC-TY9604 96-003, 10 Pages, 1996/05

PNC-TY9604-96-003.pdf:0.34MB

no abstracts in English

Journal Articles

Research on core melt accident analysis of LWRs

Abe, Kiyoharu

Tokyo Daigaku Gakui Rombun, 0, 245 Pages, 1994/09

no abstracts in English

Journal Articles

Comparative study of source terms of a BWR severe accident by THALES-2, STCP and MELCOR

Hidaka, Akihide; *; Soda, Kunihisa; Muramatsu, Ken; Sakamoto, Toru*

ANS Proc. of the 1992 National Heat Transfer Conf., p.408 - 416, 1993/00

no abstracts in English

JAEA Reports

Study on the safety of an large scale fast breeder reactor

; ; ;

PNC-TN9410 92-068, 73 Pages, 1992/03

PNC-TN9410-92-068.pdf:2.12MB

ln order to be useful for selecting specifications about the safety of the large scale fast breeder rector on and after Monju, following items were studied. (1)Design conditions of the reactor containment, (2)scenarios as to evaluation of core disruptive accident, and (3)applicability of the method of PSA. Technical documents provided for these studies are su㎜arized in this report.

JAEA Reports

None

PNC-TN1410 92-026, 113 Pages, 1992/01

PNC-TN1410-92-026.pdf:11.01MB

no abstracts in English

Journal Articles

PSA research and severe accident research at JAERI

*; Tobioka, Toshiaki; Soda, Kunihisa; Abe, Kiyoharu

Proc. of the 8th Pacific Basin Nuclear Conf., p.3-A-1 - 3-A-9, 1992/00

no abstracts in English

JAEA Reports

The plant thermohydraulic analysis for the monju PRA study; Recovery from PLOHS or LORL using the maintenance cooling system

*; *

PNC-TN9410 88-055, 111 Pages, 1988/06

PNC-TN9410-88-055.pdf:5.87MB

In this study, decay heat removal capability of the Maintenance Cooling System (MCS) of Monju has been investigated with respect to protected accidents. The protected accidents of the Liquid Metal Fast Breeder Reactors (LMFBRs), such as Protected Loss-of-Heat-Sink (PLOHS) or Loss-of-Reactor-Level (LORL), are of great importance from the viewpoint of the annual frequency of core damage. The progression of the protected accidents is mild in general because reactor decay heat can be dispersed from the core by natural circulation. The decay heat for Monju is to be removed by the Intermediate Reactor Auxiliary Cooling system (IRACS). It is essential to keep the intactness of coolant flow path from the reactor core to the heat sink and the availability of heat sink itself. If the either of them is degraded, it is taken for granted d that protected slow meltdown follows. However, the reactor core can be prevented from any damage or meltdown if the decay heat can be removed through MCS. The plant thermohydraulics of the procected accidents is analyzed using SSC-L to develop success criteria in the decay heat removal by the MCS. Parametric calculations are performed with respect to: available heat capacity in the heat transport system, cooling time before the loss-of-heat-sink and MCS starting time. It has been found, for example, that (1)MCS can remove the decay heat immediately after the reactor shutdown if heat capacity of more than two main coolant loops is available; (2)after two hours cooling time by natual circulation, MCS can remove the decay heat even if no coolant flow is assumed in all the main heat transport system; (3)LORL caused by the failure in sodium make-up can be recovered by the MCS operation. In the PLOHS condition, the coolant temperature may exceed conservative design limit of the MCS piping. However, the conservativeness of the design limit and the method of qualification make compensation for the deterioration in structural strength. Finally, ...

JAEA Reports

Users manual of ART code for analyzing fissin product transport behaviour during core meltdown accident

Ishigami, Tsutomu; Sakamoto, Toru*; Kobayashi, Kensuke; *

JAERI-M 88-093, 89 Pages, 1988/05

JAERI-M-88-093.pdf:2.2MB

no abstracts in English

Journal Articles

Sensitivity study on PWR source terms with THALES/ART code package and effects of in-vessel thermal-hydraulic models

Abe, Kiyoharu; Watanabe, Norio; Ida, Mitsuo; Sakamoto, Toru; Ishigami, Tsutomu; Harami, Taikan

Proc.on Probabilistic Safety Assessment and Risk Management PSA87, Vol.3, p.945 - 950, 1987/00

no abstracts in English

Journal Articles

Overview of development and application of THALES code system for analyzing progression of core meltdown accident of LWR´s

; *; ; *; *

Proc.2nd Int.Topical Meeting on Nuclear Power Plant Thermal Hydraulics and Operatiions, p.6 - 49, 1986/00

no abstracts in English

Journal Articles

Development of computer code system THALES for thermal-hydrauice analysis of core meltdown accident, I; Outlines of code system and analytical models in each code

; *; Watanabe, Norio; *

Nippon Genshiryoku Gakkai-Shi, 27(11), p.1035 - 1046, 1985/00

 Times Cited Count:1 Percentile:74.86(Nuclear Science & Technology)

no abstracts in English

Journal Articles

PWR S$$_{2}$$D sequence analysis by THALES code system

; *; Watanabe, Norio; ; *

Proc.of ANS/ENS Int.Topical Meeting on Probabilistic Safety Methods and Applications, p.30 - 1, 1985/00

no abstracts in English

Journal Articles

An Experimental study of reactivity change and flux distortion in simulated LMFBR meltdown cores

; *;

Nuclear Science and Engineering, 87, p.283 - 294, 1984/00

 Times Cited Count:5 Percentile:46.79(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

33 (Records 1-20 displayed on this page)