Yng-Ruey Yuann

Implementation and Assessment of Moody Homogeneous Equilibrium Critical Flow Model of RELAP5-3D

Abstract According to “Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants” (NUREG-0800) of the U.S. Nuclear Regulatory Commission, the homogeneous and thermal equilibrium critical flow model (HEM model) is acceptable for pressure and temperature analysis of the subcompartment of the containment. However, it was not built into the RELAP5-3D code. In order to provide the blowdown boundary conditions that meet the acceptance criteria for the subcompartment pressure and temperature response analysis, Institute of Nuclear Energy Research implemented and assessed the Moody HEM model of RELAP5-3D. The assessment phase was subsequent to the implementation of the Moody HEM model of RELAP5-3D. Three experiments of Marviken critical flow tests (CFTs) were selected as the assessment cases. They were CFT 15, CFT 22, and CFT 24. The assessment input decks of RELAP5-3D had been modified from the appendixes of the references. Additional comparisons with the results of the RELAP5-3D built-in Ransom-Trapp and Henry-Fauske critical flow models were also included. The comparisons of the calculated blowdown mass flow rate with the test data assessed the newly implemented model, which gave good prediction. Moreover, the comparisons between the results of the critical flow models of RELAP5-3D and the test data provided a measure of the relative conservatism of the critical flow calculations.

Loss of Cooling Thermal Analysis for the Spent Fuel Pool of the Chinshan Nuclear Power Plant

The Chinshan Nuclear Power Plant owned and operated by the Taiwan Power Company is a twin-unit BWR-4 plant. Unit 1 and unit 2 began their commercial operation in 1978 and 1979, respectively. Since commercial operation, all the fuels discharged from reactor core at each cycle are stored in the spent fuel pool (SFP). An engineering analysis is performed to predict the SFP water temperature and pool water level during a postulated loss of forced cooling accident. A full-core discharged loading is considered, and the fuel assemblies are moved to the SFP just after 7 days of cooling. The pool temperature and level are calculated using lumped energy and mass balances. Calculation results show that the water temperature reaches the saturation temperature at 9.4 hours after the onset of the accident, and the pool level drops to the top of the active fuels at 76.8 hours. After the pool level drops to the top of the active fuels, the cladding temperature increases dramatically because the convective heat transfer of steam is much weaker than that of liquid water. The peak cladding temperature after fuel uncovery is calculated by detailed CFD simulations, and the results show that the peak cladding temperature reaches 600°C in 3 hours and 1200°C in 9.5 hours after the fuels are uncovered. Additionally, the check-board arrangement for fuels is also investigated. Through enhanced the radiation heat transfer, the check-board fuel arrangement can have slower heating rate for the fuels. For the Chinshan SFP, extra 2.5 hours can be gained by employing such an arrangement for necessary actions.Copyright

Effects of the RHR Return Line Elevation to the Suppression Pool Temperature of the Lungmen ABWR Containment

Lungmen Nuclear Power Plant in Taiwan is a twin-unit Advanced Boiling Water Reactor (ABWR) plant. In this study, a long-term GOTHIC model for the Lungmen ABWR primary containment response analysis is established. The wetwell space is vertically divided into several volumes to catch the pool temperature stratification effect. The long-term containment responses for a double-ended feedwater line break (FWLB) accident are calculated. The fuel decay heat is absorbed by the reactor coolant, and the coolant flows to the containment via the broken line. The suppression pool is gradually heated up by the high-temperature gas-water mixture following through horizontal vents. To reduce the pool temperature, the Residual Heat Removal (RHR) system will be required to operate in the suppression pool cooling mode. The RHR pumps have suction flow from suppression pool and discharge it to the RHR heat exchangers for cooling. The cooled water then returns to the pool. An elevated RHR return line is desired to avoid the cooled water being directly sucked again. The wetwell temperature stratification associated with the RHR return line elevation is investigated in this study. Effects of the RHR return line elevation on the pool temperature can be determined since the whole wetwell space is not lumped as a node only. The calculated peak pool temperature is 92.6°C based on the plant piping configuration. The peak temperature can be reduced to 88.9°C by returning the water via the wetwell spray spargers located in the top of the wetwell. However, it should be noted that using the wetwell spray also pressurizes the wetwell because the pool water temperature is higher than that of airspace during the late period of the event. Returning the pool water via the wetwell spray spargers is not suggested because it causes long-term wetwell pressurization.Copyright