Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of Spent Fuel Pools (SFPs) in Taiwan after Fukushima event. In order to estimate the safety of Kuosheng NPP SFP, by using MELCOR2.1 and SNAP, the safety analysis of Kuosheng NPP SFP was performed combined with the mitigation strategy of NEI 06-12 report. There were several steps in this research. First, the Kuosheng NPP SFP models were established by MELCOR2.1/SNAP. Second, the Station Blackout (SBO) analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition. The results showed that the calculations of MELCOR and TRACE were very similar in this case. Second, the mitigation strategy analysis was done with the MELCOR model by following the NEI 06-12 report. The results showed the effectiveness of NEI 06-12 strategy in Kuosheng NPP SFP. Finally, a sensitivity study of SFP quenching was done to check the differences of different water injection time and the phenomena during the quenching. The results showed that if the cladding temperature was over 1600 K, the water injection may have chance to cause the accident more severe with more hydrogen generation. It was because of the oxidation heat and the “Breakaway” effect of the zirconium-water reaction. An animation model built by SNAP was also shown in this study.
Abstract: Heat pipe is considered to be applied as a passive system to remove residual heat that generated from reactor core when incident occur or from spent fuel storage pool. The objectives are to characterized the heat transfer phenomena, performance of heat pipe, and as a model for large heat pipe will be applied as passive cooling system on nuclear spent fuel pool storage. In this experimental wickless heat pipe or two-phase closed thermosyphon (TPCT) is used. Variation of heat flux are 611.24 Watt/m2 - 3291.29 Watt/m2. Variation of filling ratio are 45 - 70%. Variation of initial pressure are -62 to -74 cm Hg. Demineralized water is used as working fluid in the TPCT. The results showed that increasing of heat load leads to an increase of evaporation of the working fluid. The optimum filling ratio obtained for 60% of TPCT evaporator volume, and initial pressure variation gave different TPCT wall temperature characteristic. TPCT showed best performance with 60% filling ratio and can be consider to be applied as passive residual heat removal system or passive cooling system on spent fuel storage pool.
Abstract: Kuosheng nuclear power plant (NPP) is a BWR/6 plant in Taiwan. There is more concern for the safety of NPPs in Taiwan after Japan Fukushima NPP disaster occurred. Hence, in order to estimate the safety of Kuosheng NPP spent fuel pool (SFP), by using TRACE, MELCOR, and SNAP codes, the safety analysis of Kuosheng NPP SFP was performed. There were two main steps in this research. First, the Kuosheng NPP SFP models were established. Second, the transient analysis of Kuosheng SFP was done by TRACE and MELCOR under the cooling system failure condition (Fukushima-like condition). The results showed that the calculations of MELCOR and TRACE were very similar in this case, and the fuel uncover happened roughly at 4th day after the failure of cooling system. The above results indicated that Kuosheng NPP SFP may be unsafe in the case of long-term SBO situation. In addition, future calculations were needed to be done by the other codes like FRAPTRAN for the cladding calculations.
Abstract: TRACE is developed by U.S. NRC for the nuclear
power plants (NPPs) safety analysis. We focus on the establishment
and application of TRACE/FRAPTRAN/SNAP models for Chinshan
NPP (BWR/4) spent fuel pool in this research. The geometry is 12.17
m × 7.87 m × 11.61 m for the spent fuel pool. In this study, there are
three TRACE/SNAP models: one-channel, two-channel, and
multi-channel TRACE/SNAP model. Additionally, the cooling system
failure of the spent fuel pool was simulated and analyzed by using the
above models. According to the analysis results, the peak cladding
temperature response was more accurate in the multi-channel
TRACE/SNAP model. The results depicted that the uncovered of the
fuels occurred at 2.7 day after the cooling system failed. In order to
estimate the detailed fuel rods performance, FRAPTRAN code was
used in this research. According to the results of FRAPTRAN, the
highest cladding temperature located on the node 21 of the fuel rod
(the highest node at node 23) and the cladding burst roughly after 3.7
day.
Abstract: This paper focuses on assessing sloshing-induced overflow of the seismically-isolated nuclear tanks based on Fluid-Structure Interaction (FSI) analysis. Typically, fluid motion in the seismically-isolated nuclear tank systems may be rather amplified and even overflowed under earthquake. Sloshing-induced overflow in those structures has to be reliably assessed and predicted since it can often cause critical damages to humans and environments. FSI analysis is herein performed to compute the total cumulative overflowed water volume more accurately, by coupling ANSYS with CFX for structural and fluid analyses, respectively. The approach is illustrated on a nuclear liquid storage tank, Spent Fuel Pool (SFP), forgiven conditions under consideration: different liquid levels, Peak Ground Accelerations (PGAs), and post earthquakes.
Abstract: Accident in spent fuel pool (SFP) of Fukushima
Daiichi Unit 4 showed the importance of continuous monitoring of the
key environmental parameters such as water temperature, water level,
and radiation level in the SFP at accident conditions. Because the SFP
water temperature is one of the key parameters indicating SFP
conditions, its behavior at accident conditions shall be understood to
prepare appropriate measures. This study estimated temporal change
in the SFP water temperature at Kori Unit 1 with 587 MWe for 1 hour
after initiation of a loss-of-pool-cooling accident. For the estimation,
ANSYS CFX 13.0 code was used. The estimation showed that the
increasing rate of the water temperature was 3.90C per hour and the
SFP water temperature could reach 1000C in 25.6 hours after the
initiation of loss-of-pool-cooling accident.