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Mochizuki, Hiroyasu
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05
This paper describes about the development of the plant simulation code NETFLOW. The code developed to simulate thermal-hydraulics in the auxiliary cooling system of a reactor has expanded to simulate plant dynamics not only for light water reactors but also for liquid metal cooled fast reactors (LMFRs). In order to improve the applicability for LMFRs, models affecting on characteristics of natural circulation are improved. They are a model of inter-subassembly heat transfer and a model of heat transfer coefficient of finned heater tubes in an air cooler. A natural circulation in the Joyo experimental fast reactor after the reactor trip is calculated, and good agreement has been obtained between the test and simulation results.
Nakano, Yoshihiro; Akie, Hiroshi; Okubo, Tsutomu; Uchikawa, Sadao
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05
no abstracts in English
Takata, Takashi*; Yamaguchi, Akira*; Uchibori, Akihiro; Ohshima, Hiroyuki
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 8 Pages, 2007/05
A numerical investigation of sodium-water reaction (SWR) phenomenon under a tube bundle configuration was carried out in order to evaluate characteristics of the reacting zone in an area nearby the leakage, such as distributions of the gas volume fraction, the temperature and the concentration of the sodium compounds, as well as to evaluate an applicability of the numerical methodology that has been developed in the previous authors' work to a complicated flow path. SWAT-1R experiment, in which 43 tubes were placed in a zigzag alignment and water vapor leaked vertically upward from the bottom center of the bundle, was chosen as a benchmark problem. As a result of the numerical investigation, it was demonstrated that one could predict the SWR phenomenon with a complicated pin bundle configuration numerically. In addition, it was also demonstrated that numerical quantification of a secondary failure possibility due to over-heating rupture was to be achieved using the present result.
Rugama, Y.*; Suyama, Kenya; Brady Raap, M.*
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 4 Pages, 2007/05
The importance of measured assay data from Post-Irradiation Examination (PIE) has long been understood by members of the OECD/NEA/NSC/WPNCS (Working Party on Nuclear Criticality Safety). These data are particularly important in criticality analyses related to any application of burnup credit as well as to evaluation of far-field criticality in geologic repositories and applications such as reprocessing where the fuel is not maintained intact. This report provide us the outline of the new expert group on assay data (PIE data) of spent nuclear fuel under OECD/NEA/NSC/WPNCS.
Kunitomi, Kazuhiko; Shiozawa, Shusaku; Yan, X.
Proceedings of 2007 International Congress on Advances in Nuclear Power Plants (ICAPP 2007) (CD-ROM), 9 Pages, 2007/05
Japan Atomic Energy Agency had been engaging in the basic design and development of Gas Turbine High Temperature Reactor 300 (GTHTR300). Costs of capital, fuel, and operation and maintenance have been estimated to confirm economical feasibility of the GTHTR300. The cost of electricity for the GTHTR300 is estimated to be below US 3.3 cents/kWh (4 yen/kWh), which is about two-third of that of the current LWRs in Japan. The results confirm that the net power generation cost of the GTHTR300 is much lower than that of the LWR, indicating that the GTHTR300 plant consisting of small-scale reactor units can be economically competitive to the latest large-scale LWR. The contents and results of the economical evaluation will be presented together with the outline of the GTHTR300 design.
Takamatsu, Misao; Kuroha, Takaya*; Aoyama, Takafumi
no journal, ,
Studies for passive safety tests to be conducted by using Joyo have been carried out to demonstrate the inherent safety of LMFBR. In these studies, emphasis has been placed on the improvement of the calculation accuracy of the feedback reactivity. Therefore, in the passive safety tests, the feedback reactivity measurement is planned under the condition simulating ATWS event. The Mimir-N2 which has been developed to analyze Joyo plant dynamic is used for predicting plant behavior during transients. In order to determine the passive safety tests plan, the improvement of the accuracy of Mimir-N2 was required. In this study, transient tests including manual reactor shutdown test, loss of power supply test and preliminary transient over power test were conducted to verify the updated model of the Joyo MK-III core and the heat transport system of Mimir-N2. As a result, a good agreement was obtained between calculated and measured sodium temperatures.