为更好地支持国产先进压水堆核安全审评工作，针对国产先进压水堆卡轴事故开展了计算分析，并研究了丧失厂外电（loss of offsite power, LOOP）假设和慢化剂温度反馈对事故后果的影响。最终得到：①事故过程中得到的反应堆冷却剂系统峰值压力、热点包壳峰值温度均满足事故验收准则的要求，核电厂最终可达到一个稳定状态；②LOOP假设是保守的一种假设条件，但对于本事故，LOOP假设引起的后果差异都体现在事故中后期，对在事故早期出现的系统压力峰值、最小偏离泡核沸腾比（departure from nucleate boiling ratio, DNBR）和热点包壳峰值温度等主要结果参数不会产生影响，是否引入LOOP假设都可接受；③慢化剂温度反馈的引入是一种不保守的假设条件。然而，在不引入慢化剂温度反馈情况下得到的反应堆冷却剂系统峰值压力、热点包壳峰值温度也均满足事故验收准则的要求，且保有很大裕量。同时，引入慢化剂温度反馈使得计算更接近现实。综上，在本事故中引入慢化剂温度反馈也是可以接受的。

To enhance the support for the nuclear safety review of domestic advanced pressurized water reactor (PWR), the reactor coolant pump (RCP) shaft seizure accident for this reactor type was conducted. The impact of the loss of offsite power (LOOP) assumption and moderator temperature feedback on the accident consequences was examined. The findings are as follows. The peak pressure in the reactor coolant system and the peak temperature of the hot spot cladding, recorded during the accident process, met the accident acceptance criteria, enabling the nuclear power plant to eventually stabilize. The LOOP assumption was found to be conservative. For this accident, however, the differences arising from this assumption manifested in the later stages and did not influence key outcome parameters such as system pressure peak, minimum departure from nucleate boiling ratio (DNBR), and peak temperature of hot spot cladding, which occurred early in the accident. Therefore, the inclusion or exclusion of the LOOP assumption was deemed acceptable. The introduction of moderator temperature feedback was considered an unconservative assumption. Nevertheless, even without this feedback, the peak pressure in the reactor coolant system and the peak temperature of the hot spot cladding still met the accident acceptance criteria, maintaining a substantial margin. Additionally, incorporating moderator temperature feedback brought the calculations closer to reality. Overall, the introduction of moderator temperature feedback in this accident was also judged to be acceptable.