【摘要】 为验证非能动应急堆芯冷却系统的设计和给国产核电自主软件验证提供支撑，基于安全系统整体效应试验装置，开展了冷管段小破口失水事故直接注入特性试验，主要结论如下：第1级快速卸压系统开启后，中压安注投入，高压安注并未受到中压安注抑制，反而因为回路卸压速率加快而增大；第4级快速卸压系统投入后，一回路卸压明显能力增强，中压安注流量显著增大，高压安注完全被抑制；受回路整体卸压速率影响，低压安注实际投入时间晚于信号给定时间102 s；试验装置有效模拟了非能动安全系统各系统的响应特性，主回路系统无偏环现象，堆芯未出现裸露，非能动应急堆芯冷却系统能有效带走堆芯余热，缓解小破口失水事故。更多还原

【Abstract】 To assess the capability of the passive emergency core cooling system and support the confirmation of domestic nuclear power software independently, the small break loss of coolant accident of the cold leg was conducted on the integral test facility of passive safety system facility. The first-stage rapid pressure relief system and the accumulator(ACC) were activated almost simultaneously. The injection flowrate of high pressure makeup tank(HPMT) was not suppressed by the accumulator safety injection. Conversely, the flowrate of HPMT increased due to the accelerated pressure relief rate of the primary coolant system. After the four-stage rapid pressure relief system was activated, the pressure relief capacity of the primary coolant system was significantly enhanced, the flowrate of accumulator increased, and the HPMT injection was completely suppressed. Affected by the pressure relief rate of th e primary coolant system, the actual activation time of out-containment refueling water storage tank(ORWST) was 102 seconds later than the given signal time. The integral test facility of the passive safety system facility simulated the characteristics of the prototype passive emergency core cooling system effectively. There is no deviation loop phenomenon in the primary coolant system, and the reactor core is not exposed. The passive emergency core cooling system could effectively remove the residual heat of the core and alleviate the coolant loss accident of small break.更多还原