RELAP5 （reactor excursion and leak analysis program, version 5） code analyses were performed on two ROSA/LSTF （rig of safety assessment/large scale test facility） experiments on PWR （pressurized water reactor） s...RELAP5 （reactor excursion and leak analysis program, version 5） code analyses were performed on two ROSA/LSTF （rig of safety assessment/large scale test facility） experiments on PWR （pressurized water reactor） safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG （steam generator） secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by core boil-off. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. In the 4-in. break test, on the other hand, there was no core uncovery and heatup due to smaller break flow rate than in the 8-in. break test. Adjustment of Cd （break discharge coefficient） for two-phase discharge flow predicted the break flow rate reasonably well. The code well predicted the overall trend of the major thermal-hydraulic response observed in the two LSTF tests by the Cd adjustment. The code, however, overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.展开更多
This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating s...This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.展开更多
基于轻水堆最佳估算系统分析程序RELAP/SCDAPSIM/MOD4.0,添加新的FLi Na K熔盐热物性参数和适用于熔盐的对流换热系数,开发了适用于FHR系统的热工水力分析程序RELAP5-FHR。通过FLi Na K高温熔盐实验回路对RELAP5-FHR程序进行实验验证。...基于轻水堆最佳估算系统分析程序RELAP/SCDAPSIM/MOD4.0,添加新的FLi Na K熔盐热物性参数和适用于熔盐的对流换热系数,开发了适用于FHR系统的热工水力分析程序RELAP5-FHR。通过FLi Na K高温熔盐实验回路对RELAP5-FHR程序进行实验验证。结果表明:RELAP5-FHR程序计算值与实验值吻合较好,验证了程序的适用性。展开更多
文摘RELAP5 （reactor excursion and leak analysis program, version 5） code analyses were performed on two ROSA/LSTF （rig of safety assessment/large scale test facility） experiments on PWR （pressurized water reactor） safety system that simulated cold leg small-break loss-of-coolant accidents with 8-in. or 4-in. diameter break using SG （steam generator） secondary-side depressurization. The SG depressurization was initiated by fully opening the depressurization valves in both SGs immediately after a safety injection signal. In the 8-in. break test, loop seal clearing occurred and then core uncovery and heatup took place by core boil-off. Core collapsed liquid level recovered after the initiation of accumulator coolant injection, and long-term core cooling was ensured by the actuation of low-pressure injection system. In the 4-in. break test, on the other hand, there was no core uncovery and heatup due to smaller break flow rate than in the 8-in. break test. Adjustment of Cd （break discharge coefficient） for two-phase discharge flow predicted the break flow rate reasonably well. The code well predicted the overall trend of the major thermal-hydraulic response observed in the two LSTF tests by the Cd adjustment. The code, however, overpredicted the peak cladding temperature because of underprediction of the core collapsed liquid level due to inadequate prediction of the accumulator flow rate in the 8-in. break case.
文摘This paper discusses the results obtained during an investigation of WWER-1000 Nuclear Power Plant (NPP) behavior at shutdown reactor during maintenance. For the purpose of the analysis is selected a plant operating state with unsealed primary circuit by removing the MCP head. The reference nuclear power plant is Unit 6 at Kozloduy NPP (KNPP) site. RELAP5/ MOD3.2 computer code has been used to simulate the transient for WWER-1000/V320 NPP model. A model of WWER-1000 based on Unit 6 of KNPP has been developed for the RELAP5/MOD3.2 code at the Institute for Nuclear Research and Nuclear Energy-Bulgarian Academy of Sciences (INRNE-BAS), Sofia. The plant modifications performed in frame of modernization program have been taken into account for the investigated conditions for the unsealed primary circuit. The most specific in this analysis compared to the analyses of NPP accidents at full power is the unavailability of some important safety systems. For the purpose of the present investigation two scenarios have been studied, involving a different number of safety systems with and without operator actions. The selected initiating event and scenarios are used in support of analytical validation of Emergency Operating Procedures (EOP) at low power and they are based on the suggestions of leading KNPP experts and are important in support of analytical validation of EOP at low power.
文摘基于轻水堆最佳估算系统分析程序RELAP/SCDAPSIM/MOD4.0,添加新的FLi Na K熔盐热物性参数和适用于熔盐的对流换热系数,开发了适用于FHR系统的热工水力分析程序RELAP5-FHR。通过FLi Na K高温熔盐实验回路对RELAP5-FHR程序进行实验验证。结果表明:RELAP5-FHR程序计算值与实验值吻合较好,验证了程序的适用性。
