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Asahi, Yoshiro; Okumura, Keisuke; Ose, Yasuo*
Nuclear Science and Engineering, 139(1), p.78 - 95, 2001/09
Times Cited Count：1 Percentile：12.01(Nuclear Science & Technology)no abstracts in English
Kinoshita, Hidetaka; Kaminaga, Masanori; Hino, Ryutaro
JAERI-Tech 2001-052, 41 Pages, 2001/08
JAERI-Tech-2001-052.pdf:1.99MB
no abstracts in English
Odano, Naoteru; Kusunoki, Tsuyoshi; Yoritsune, Tsutomu; Fukuhara, Yoshifumi*; Saito, Kazuo*; Takahashi, Teruo*; Ishida, Toshihisa
Proceedings of International Workshop on Utilization of Nuclear Power in Oceans (N'ocean 2000), p.164 - 169, 2000/02
no abstracts in English
*; Watanabe, Norio; Hirano, Masashi
JAERI-Data/Code 98-037, 193 Pages, 1998/11
JAERI-Data-Code-98-037.pdf:6.14MB
no abstracts in English
Kawasaki, Nobuchika; kasahara, Naoto
PNC TN9410 96-294, 47 Pages, 1996/07
As main components of fast reactors, there are reactor vessels, pipes, heat exchangers, nozzles. In order to keep their structural integrity, the elevated temperature structural design guide evaluates primary stress, strain, and creep-fatigue damage of them. Especially in fast reactors which operate under low pressure and high temperature conditions, creep-fatigue damage is dominant, and limits a design range. For the purpose of extending a design range, author evaluated strength of cylindrical structures by an advanced proposed method based on the generalized elastic follow-up model. Furthermore we studied evaluation accuracy of a method of 'monjyu" and a proposed method, through comparison of thermal transient strength test data with damage calculated by two strength evaluation methods. Results are summarized as follows. (1)An advanced proposed method predicted lower fatigue damage (Df) than a method of 'monjyu'. (However, in structural discontinuities where an elastic follow up is large, a proposed method shows higher damage than a method of 'monjyu'.) (2)An advanced proposed method calculated lower creep damage (Dc) than a method of 'monjyu'. (About 0.7 times lower in a cylindrical structure with a structural discontinuity (STF-3). About 0.1 times lower in a cylindrical structure without a structural discontinuity (STF-10). (3)On no-crack portions in STF-3, prediction of a method of 'monjyu' is out of a limit of crack initiation, nevertheless one of an advanced proposed method is within the range. Through above results, an advanced proposed method was evaluated to be more accurate in prediction of creep-fatigue damage. We concluded from the above investigation that we calculated more rational and lower damage by proposed method, and it has the possibilities to get extension of a design range.
gas turbine power generation system (User's manual); Sekiguchi, Nobutada
PNC TN9520 95-002, 66 Pages, 1995/02
This analysis program code STEDFAST; Space, TErrestrial and Deep sea FAST reactor gas tubine system; is used to get the adequate values of system parameters on fast reactor gas turbine power generation systems used as power sources for deep sea, space and terrestrial cogeneration. Characteristics of the code are as follows. Objective systems of the code are a deep sea, a space and a terrestrial reactors. Primary coolants of the systems are NaK, Na, Pb and Li. Secondary coolant is the mixture gas of He and Xe. The ratio of He and Xe is arbitrary. Modeling of components in the systems was performed so that detailed modeling might be capable in future and that a transient analytical code could be easily made by using the code. A progra㎜ing language is MAC-FORTRAN. The code can be easily used in a personal computer. The code made possible instant calculation of various state values in a Brayton cycle, understanding the effects of many parameters on thermal efficiency and finding the most adequate values of the parameters. From now on, detailed modeling of the components will be performed. After that, the transient program code will be made.
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PNC TN9410 93-092, 93 Pages, 1993/04
In-vessel thermohydraulic analysis using a multi-dimensional code AQUA was conducted to investigate thermohydralic conditions in the lower plenum of a large scale liquid metal fast breeder reactor (LMFBR). In the investigation, it was focused in a effective mixing volume in the plenum and in a loose parts trapping by hydraulic forces. Through the analysis by the AQUA code and discussions based on their results, the following results have been obtained: [In-Flow Allotment Characteristics at the Truss] (1)Mass flow rate between lower flow holes and side flow holes to the truss was balanced in the case of 22.5
for the coolant inlet angle to the plenum under a full flow condition. (2)The above result did not change with the transient conditions simulating a manual reactor trip event. [Effective Mixing volume] (1)From the transient calculation, it was clear that the effective mixing volume was corresponded to 77-79% of the total volume in the plenum. [Loose Parts Trapping] (1)A condition (W
1.0m/s) to trap loose parts by hydraulic forces was not satisfied under parametric calculations only considered here.
*; Fumizawa, Motoo
Kashika Joho Gakkai-Shi, 12(SUPPL.1), p.211 - 214, 1992/07
no abstracts in English
*; Haraguchi, Tetsuharu*; *; Tanigawa, Shingo*; Yamaguchi, Katsuhisa
PNC TN9410 88-107, 121 Pages, 1988/09
In the studies using PLANDTL, it would be planned to valid the thermal-hydraulic analysis codes which were developed each for whole system, plenum and subassembly, and also to evaluate the reactor plant in the future using these codes. SSC-L is to be as the main code in these studies and is used for design analysis through test analysis. In the first step of this study, model development and modification of SSC-L has been achieved for PLANDTL and predictive analyses have been applied as to validate the models and examine the design of PLANDTL. The estimated transient curves have been obtained about flow rate and temperatures at subassembly and loop of PLANDTL. As a result, the design conditions have been given to be able to perform the programmed tests. It have been validated that the conditions of tests would be within the design value, and the characteristics of PLANDTL and operational conditions have been obtained from the predictive analyses using design data of the plant. The modification and validation of SSC-L will be applied using the results of various kinds of functional tests, and test analyses will be performed in future.
*; *
PNC TN9410 88-055, 111 Pages, 1988/06
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, ...
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Anticipated and Abnormal Plant Transients in Light Water Reactor,Vol.2, p.727 - 738, 1984/00
no abstracts in English
