共查询到17条相似文献,搜索用时 156 毫秒
1.
SCDAP/RELAP5是一种常见的机理性严重事故分析程序,能够分析多种类型的堆芯构件.通过对比分析SCDAP/RELAP5程序模拟棒形燃料元件与板型燃料元件堆芯在严重事故下行为的分析模型,结合UO2-Zr板型状元件堆芯的特性,提出了运用并改进SCDAP/RELAP5程序模拟UO2-Zr板型元件堆芯在严重事故下行为的研究方案.对程序结构的分析结果表明,SCDAP/RELAP5程序部分结构和模型适用于对UO2-Zr板型元件进行基本的严重事故分析,但需要通过创建新部件、研究新模型,并与已有模型的重新组合搭配才能较为精准地模拟UO2-Zr板型元件严重事故的实际行为. 相似文献
2.
压水堆核电厂的高压熔堆事故覆盖了大部分的严重事故序列,且具有很大的潜在威胁。根据我国900MW压水堆核电厂的概率安全分析(PSA)结果选取了丧失厂外电、未能紧急停堆的预期瞬态、二回路管道破口、一回路小破口和蒸汽发生器传热管破裂5种典型的高压熔堆严重事故序列,并使用SCDAP/RELAP5程序对这些事故序列的进程和后果进行了计算分析。计算结果表明:5种典型高压熔堆事故序列可能导致高压熔喷和安全壳直接加热风险,可能引起安全壳早期失效,很有必要采取相应的一回路卸压措施。 相似文献
3.
4.
5.
6.
7.
以典型的3环路压水堆为参考对象,建立了详细的严重事故计算模型。选择一回路热段当量直径为18 cm的失水事故(LOCA)作为初始事件,采用RELAP5/SCDAP/MOD3.2为分析工具,对无注水、无缓解措施下的基准事故进程进行计算分析,研究3种不同注水时机对严重事故进程的影响。3种注水时机分别为堆芯表面峰值温度达到1100 K、1300 K、1500 K时开始注水。计算结果显示,压水堆严重事故进程对于注水的时机非常敏感。较早阶段的注水对于阻止堆芯熔化十分有效,注水较晚会恶化事故进程,加速堆芯熔化。 相似文献
8.
采用严重事故最佳估算程序RELAP5/SCDAPSIM/MOD3.2,建立美国Surry-2核电站的详细计算模型,对完全丧失给水(TLFW)引发的堆芯熔化事故进行研究分析。为准确预测压力容器内堆芯熔化的进程,为二级概率安全评价提供可信的初始条件,计算中考虑了一回路压力边界的蠕变破裂失效,并评价了人为干预对堆芯熔化进程及事故后果的影响。计算结果表明,由完全丧失给水引发的压水堆核电站严重事故不会出现人们担心的高压熔堆;反应堆压力容器下封头的失效位置不是在其底部,而是在其侧面;通过打开稳压器释放阀对一回路实施主动卸压能够大大推迟事故的进程。 相似文献
9.
10.
11.
Rae-Joon Sang-Baik Kim Seong-Wan Hong Hee-Dong Kim 《Nuclear Engineering and Design》2009,239(11):2484-2490
A coolant injection into the reactor vessel with depressurization of the reactor coolant system (RCS) has been evaluated as part of the evaluation for a strategy of the severe accident management guidance (SAMG). Two high pressure sequences of a small break loss of coolant accident (LOCA) without safety injection (SI) and a total loss of feedwater (LOFW) accident in Optimized Power Reactor (OPR)1000 have been analyzed by using the SCDAP/RELAP5 computer code. The SCDAP/RELAP5 results have shown that only one train operation of a high pressure safety injection at 30,000 s with indirect RCS depressurization by using one condenser dump valve (CDV) at 6 min after implementation of the SAMG prevents reactor vessel failure for the small break LOCA without SI. In this case, only one train operation of the low pressure safety injection (LPSI) without the high pressure safety injection (HPSI) does not prevent reactor vessel failure. Only one train operation of the HPSI at 20,208 s with direct RCS depressurization by using two SDS valves at 40 min after an initial opening of the safety relief valve (SRV) prevents reactor vessel failure for the total LOFW. 相似文献
12.
As part of the evaluation for a severe accident management strategy, a reactor coolant system (RCS) depressurization in optimized power reactor (OPR)1000 has been evaluated by using the SCDAP/RELAP5 computer code. An indirect RCS depressurization by a secondary depressurization by using a feed and bleed operation has been estimated for a small break loss of coolant accident (LOCA) without a safety injection (SI). Also, a direct RCS depressurization by using the safety depressurization system (SDS) has been estimated for the total loss of feed water (LOFW). The SCDAP/RELAP5 results have shown that the secondary feed and bleed operation can depressurize the RCS, but it cannot depressurize the RCS sufficiently enough. For this reason, a greater direct RCS depressurization by using the SDS is necessary for the 1.35 in. break LOCA without SI. A proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time from 7.5 to 10.7 h. An opening of two SDS valves can depressurize the RCS sufficiently enough and the proper RCS depressurization time and capacity leads to a delay in the reactor vessel failure time of approximately 5 h for the total LOFW. An opening of one SDS valve cannot depressurize the RCS sufficiently enough. 相似文献
13.
《Journal of Nuclear Science and Technology》2013,50(11):1396-1402
A depressurization possibility of the reactor coolant system (RCS) before a reactor vessel rupture during a high-pressure severe accident sequence has been evaluated for the consideration of direct containment heating (DCH) and containment bypass. A total loss of feed water (TLOFW) and a station blackout (SBO) of the advanced power reactor 1400 (APR1400) has been evaluated from an initiating event to a creep rupture of the RCS boundary by using the SCDAP/RELAP5 computer code. In addition, intentional depressurization of the RCS using power-operated safety relief valves (POSRVs) has been evaluated. The SCDAPRELAP5 results have shown that the pressurizer surge line broke before the reactor vessel rupture failure, but a containment bypass did not occur because steam generator U tubes did not break. The intentional depressurization of the RCS using POSRV was effective for the DCH prevention at a reactor vessel rupture. 相似文献
14.
采用RELAP5/MOD3.2系统程序建立一体化小型反应堆的事故分析模型,包括反应堆冷却剂系统(RCS)、简化的二回路系统和专设安全设施。一体化多用途的非能动小型压水反应堆(SIMPLE)热功率为660 MWt(电功率大于200 MWe)。针对SIMPLE的直接安注管线(DVI)双端断裂事故和DVI2英寸(50.8mm)小破口失水事故(SBLOCA)进行分析。计算结果表明:对于直接安注管线双端断裂事故,破口和自动降压系统(ADS)能有效地使反应堆冷却系统降压,安注箱(ACC)和安全壳内置换料水箱(IRWST)能实现堆芯补水,确保堆芯冷却;对于DVI的SBLOCA,非能动专设安全设施能有效对RCS进行冷却和降压,防止堆芯过热。 相似文献
15.
16.
Using SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4), a model with postulated boundary conditions has been developed to simulate the evolution of the fuel channel in a CANada Deuterium Uranium reactor type (CANDU6) during a large loss of coolant accident (LLOCA) with a coincidence of a loss of emergency cooling (LOECC). The accident simulation is initiated from the steady-state flow regime and different steam mass flow rates are imposed in order to run sensitivity calculations of the heatup phase. Results are compared to referenced CHAN II code results for the same accident boundary conditions, concerning the fuel and pressure tube temperatures, power components (generated and exchanged to the moderator) and hydrogen production. The input model is applied both to the intact and to the disassembled bundle with 37 fuel elements. The paper includes a brief discussion of the capabilities of the present SCDAP component models, dedicated to PWR-BWR reactor components, to treat the degradation phenomena in the fuel channel during severe accidents in CANDU reactors, and also of the developments needed to enhance the quality of the code predictions. 相似文献
17.
采取堆腔注水策略冷却熔融池对缓解严重事故后果、降低安全壳的失效概率具有十分重要的作用。本文采用SCDAP/RELAP5程序,首先以韩国APR1400相关实验结果对堆腔外部注水自然对流冷却能力进行比对分析,然后建立了耦合堆腔注水措施的融熔池冷却的核电厂模型,以非能动压水堆为研究对象,针对冷段大破口失水事故(LBLOCA)始发严重事故序列,分析堆芯熔融进展过程中实施堆腔注水策略后融熔池的冷却特性及堆腔外部注水的自然循环能力。分析结果表明,LBLOCA下,当堆芯出口温度达到923K时,实施堆腔注水后能有效冷却下封头内的熔融池,从而保持压力容器的完整性。 相似文献