JOPSS - Search Results

Journal Articles

Thermal hydraulic characteristics during ingress of coolant and loss of vacuum events in fusion reactors

Takase, Kazuyuki; Kunugi, Tomoaki*; Seki, Yasushi; Akimoto, Hajime

Nuclear Fusion, 40(3Y), p.527 - 535, 2000/03

https://doi.org/10.1088/0029-5515/40/3Y/313

Times Cited Count：11 Percentile：34.99(Physics, Fluids & Plasmas)

no abstracts in English

JAEA Reports

Thermal-Hydraulic investigation on severaI fast reactor design concepts

Ohshima, Hiroyuki; Sakai, Takaaki; ; Yamaguchi, Akira; Nishi, Yoshihisa*; Ueda, Nobuyuki*; *

JNC TN9400 2000-077, 223 Pages, 1999/05

JNC-TN9400-2000-077.pdf:6.24MB

The feasibility study (Phase l) is being carried out at JNC to build up new design concepts of practical fast reactors (FRs) from the viewpoint of economy, safety, effective use of resources, reduction of environmental burden and non-proliferation. This report describes the results of the investigation, related to decay heat removal, core/fuel-assembly thermal-hydraulics and thermal-hydraulic correlations, that was performed in fiscal l999 as a part of the feasibility study. ln the study of the decay heat removal, the effects of several design parameters on the performance of the reactor vessel auxiliary cooling system (RVACS) in a middle-scale sodium-cooled FR were clarified by using a plant dynamic analysis code. The upper limit of RVACS performance was preliminarily estimated at approximately 0.50.6 MWe. Numerical methods for the plant dynamic analysis of gas-and heavy-metal-cooled FRs were also developed. They were applied to the preliminary calculations of the transition from scram to natural circulation and the transient characteristics in tentative plant design concepts were clarified. ln addition, a dimensionless number indicating natural circulation performance was deduced for the comparison of several plant design concepts. With respect to the core/fuel-assembly thermal-hydraulics, numerical analysis methods were improved for the pin-type fuel assembly of gas-and heavy-metal-cooled FRs, the coated-particle- type fuel assembly of helium-gas-cooled FR, and the ductless core of sodium-and heavy-metal-cooled FRs. As preliminary evaluations, thermal-hydraulics in the heavy-metal-cooled FR fuel assembly was compared with sodium-cooled one and thermal-hydraulic analyses of carbon-dioxide- and helium-gas-cooled FR fuel assemblies were performed. The analysis for the fuel assembly with inside duct of sodium-cooled FR was also carried out. The correlations of pressure loss and heat transfer coefficient were investigated for the thermal-hydraulic ...

Journal Articles

Preliminary numerical analysis on dust transport in fusion reactors during the loss-of-vacuum accident

Takase, Kazuyuki; Kunugi, Tomoaki*

Proc. of 5th ASME/JSME Joint Thermal Engineering Conf. (CD-ROM), 8 Pages, 1999/00

no abstracts in English

JAEA Reports

None

PNC TN1000 98-001, 73 Pages, 1998/05

PNC-TN1000-98-001.pdf:5.65MB

no abstracts in English

Journal Articles

Analysis of pressure transient during ingress-of-coolant event in fusion reactor with TRAC-code

Takase, Kazuyuki; Kunugi, Tomoaki*; Akimoto, Hajime

Proc. of 6th Int. Conf. on Nucl. Eng. (CD-ROM), 12 Pages, 1998/00

no abstracts in English

JAEA Reports

None

PNC TN1410 97-031, 638 Pages, 1997/08

PNC-TN1410-97-031.pdf:12.12MB

no abstracts in English

Journal Articles

Three-dimensional numerical simulations of heat transfer in an annular fuel channel with periodic spacer ribs under a fully developed turbulent flow

Takase, Kazuyuki

Nuclear Technology, 118(2), p.175 - 185, 1997/05

https://doi.org/10.13182/NT97-A35377

Times Cited Count：5 Percentile：42.84(Nuclear Science & Technology)

no abstracts in English

JAEA Reports

None

Yamaguchi, Akira

PNC TN9420 96-049, 15 Pages, 1996/07

PNC-TN9420-96-049.pdf:0.34MB

None

JAEA Reports

None

PNC TN9420 96-048, 10 Pages, 1996/07

PNC-TN9420-96-048.pdf:0.29MB

None

Journal Articles

Numerical simulation of turbulent heat transfer in an annular fuel channel augmented by spacer ribs

Takase, Kazuyuki; Akino, Norio

Proc. of the 30th Intersociety Energy Conversion Engineering Conf., 0, P. 95_169, 1995/00

no abstracts in English

JAEA Reports

Preliminary design for reconstruction of SWAT-3 facility

*; *; *; *; *; *; *

PNC TJ9164 94-006, 133 Pages, 1994/03

PNC-TJ9164-94-006.pdf:3.4MB

This report gives an applicability of SWAT-3 facility and contents of the reconstruction in order to confirm a DBL (Design Basis Leak) for the demonstration reactor SG. (1).Test Cndition and test case. Evaluation of the wall temperature for adjacent heat transfer tubes under the sodium-water reaction event was performed. (a)As the effect of tube rupture due to overheating, failure of upper part of the helical coil was severer than one of the lower part. (b)The wall temperature depends on the water side condition. (c)Reference test condition, whici is water leak rale about 1 kg/s, failure of upper part of the helical coil and 30% partial load, was selected. A total of ten test cases were decided. (2).System and Components Design. (a)Large leak sodium-water reaction analyses including water injection rate analysis and quasi-steady pressure analysis were performed. The maximum water leak rate of 1 DEG was 7.2 kg/s and the water leak rate at the quasi-steady state was 3.1 kg/s. The maximum pressure was 18.1kg/cm $^{2}$ a at the piping between the reaction vessel and IHX, the pressure was within the design condition of SWAT-3 facility. (b)Based on the results of the large leak sodium-water reaction analyses, a reaction vessel, water heaters and a dump tank were designed and their design specification were clarified. The reaction vessel was a scale of one third of the demonstration reactor SG and it was designed to be able to conduct the water injection test twice with one test unit. (c)A system and piping diagram, and many kinds of list (Piping list, Valve list, instrumentlist) were made up. (3).Reconstruction scope and arrangement plan. The reconstruction scope and a layout for the components and piping were clarfied and the arrange ment plans were made up. (4)Reconstruction period. The recoastruction period and man power for the design, fabrication, inspection and installation were studied and the reconstruction schedule was made up.

JAEA Reports

Preliminary study on modification of LEAP

*; *; *; *

PNC TJ9124 94-009, 164 Pages, 1994/03

PNC-TJ9124-94-009.pdf:4.63MB

In selecting the reasonable DBL on steam generator, it is indicated that the possibility of failure propagation due to overheating should be evaluated. In this study, the general plan for the next models to evaluate the reasonable DBL have been designed; a)overheating tube bursting models (structural/fractural dynamics), b)unsteady heat conduction analysis models, c)blow down analysis models and d)reaction zone temperature distribution analysis models. Then blow down analysis models were developed to evaluate the overheating tube bursting and analysis code was preliminarily designed in which the module construction of this code and link of each modules were described. Furthermore, easy coupling of this code and LEAP in future was fully considered.

Journal Articles