Steam–gas pressurizers are self-pressurizing, and since steam and noncondensable gas are used to sustain their pressure, they experience very complicated thermal–hydraulic phenomena owing to the presence of the latt...Steam–gas pressurizers are self-pressurizing, and since steam and noncondensable gas are used to sustain their pressure, they experience very complicated thermal–hydraulic phenomena owing to the presence of the latter. A steam–gas pressurizer model was developed using Relap5 code to investigate such a pressurizer's thermal–hydraulic characteristics.The important thermal–hydraulic processes occurring in the pressurizer model include bulk flashing, rainout, wall condensation with noncondensable gas, and interfacial heat and mass transfer. The pressurizer model was verified using results from insurge experiments performed at the Massachusetts Institute of Technology. It was found that noncondensable gas was one of the important factors governing the pressure response, and the accuracy of the developed model would change with different mass fractions and types of noncondensable gas.展开更多
This paper reports on the modeling and simulation of flashing-induced instabilities in naturalcirculation systems, with special emphasis on simplified boiling water reactors (SBWRs). In this work, flashing-induced osc...This paper reports on the modeling and simulation of flashing-induced instabilities in naturalcirculation systems, with special emphasis on simplified boiling water reactors (SBWRs). In this work, flashing-induced oscillations have been studied by using an experimental test facility (SIRIUS-N) and RELAP5/MOD3.2 thermal hydraulic code. The behavior of the test facility is investigated for different values of core inlet temperature value. The results of the simulations have been compared qualitatively and quantitatively with experiments. In general, deviations are found between the numerical and experimental results, in spite of the close similarity between the SIRIUS-N facility and the definition of the system in the RELAP code. This result indicates that predictions regarding experimental facility, based on modeled system, should be carefully considered.展开更多
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
利用修改后的适用于固态熔盐堆的RELAP5/MOD4.0系统分析程序,对固态熔盐堆全厂断电ATWS(Anticipated Transient Without Scram)事故进行了分析。主回路系统进行了合理简化建模,模拟系统在全厂断电ATWS事故时非能动余热排出系统有效与否...利用修改后的适用于固态熔盐堆的RELAP5/MOD4.0系统分析程序,对固态熔盐堆全厂断电ATWS(Anticipated Transient Without Scram)事故进行了分析。主回路系统进行了合理简化建模,模拟系统在全厂断电ATWS事故时非能动余热排出系统有效与否两种情况下的瞬态响应过程。分析结果表明:非能动余热排出系统在全厂断电ATWS事故初期作用不明显,但长期作用较明显,投入使用后最终将使堆芯温度和主冷却剂温度达到稳定;对于固态熔盐堆来说,即使非能动余热排出系统失效,燃料元件温度上升也很缓慢,给人员干预采取必要措施提供了超过20天的宽限时间。分析结果表明了固态熔盐堆在应对极端事件时具有高的安全性。展开更多
文摘Steam–gas pressurizers are self-pressurizing, and since steam and noncondensable gas are used to sustain their pressure, they experience very complicated thermal–hydraulic phenomena owing to the presence of the latter. A steam–gas pressurizer model was developed using Relap5 code to investigate such a pressurizer's thermal–hydraulic characteristics.The important thermal–hydraulic processes occurring in the pressurizer model include bulk flashing, rainout, wall condensation with noncondensable gas, and interfacial heat and mass transfer. The pressurizer model was verified using results from insurge experiments performed at the Massachusetts Institute of Technology. It was found that noncondensable gas was one of the important factors governing the pressure response, and the accuracy of the developed model would change with different mass fractions and types of noncondensable gas.
文摘This paper reports on the modeling and simulation of flashing-induced instabilities in naturalcirculation systems, with special emphasis on simplified boiling water reactors (SBWRs). In this work, flashing-induced oscillations have been studied by using an experimental test facility (SIRIUS-N) and RELAP5/MOD3.2 thermal hydraulic code. The behavior of the test facility is investigated for different values of core inlet temperature value. The results of the simulations have been compared qualitatively and quantitatively with experiments. In general, deviations are found between the numerical and experimental results, in spite of the close similarity between the SIRIUS-N facility and the definition of the system in the RELAP code. This result indicates that predictions regarding experimental facility, based on modeled system, should be carefully considered.
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.
文摘利用修改后的适用于固态熔盐堆的RELAP5/MOD4.0系统分析程序,对固态熔盐堆全厂断电ATWS(Anticipated Transient Without Scram)事故进行了分析。主回路系统进行了合理简化建模,模拟系统在全厂断电ATWS事故时非能动余热排出系统有效与否两种情况下的瞬态响应过程。分析结果表明:非能动余热排出系统在全厂断电ATWS事故初期作用不明显,但长期作用较明显,投入使用后最终将使堆芯温度和主冷却剂温度达到稳定;对于固态熔盐堆来说,即使非能动余热排出系统失效,燃料元件温度上升也很缓慢,给人员干预采取必要措施提供了超过20天的宽限时间。分析结果表明了固态熔盐堆在应对极端事件时具有高的安全性。
