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Best-Estimate Evaluation of Large-Break Loss-of-Coolant Accident for Advanced Natural Circulation Nuclear Reactor

机译:先进自然循环核反应堆大面积冷却水损失事故的最佳估计

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This paper presents the methodology, which will be adopted, for quantifying the effect of uncertainties on the peak clad temperature of an advanced natural circulation nuclear reactor. The method relies on probabilistic analysis, treating uncertain parameters as random variables. The paper will cover a case of a loss-of-coolant accident due to a 200% (that is, double ended) break of the largest pipe with partial unavailability of the low-pressure emergency core cooling system. The break has been postulated at the inlet header, which is the largest pipe in the main heat transport system. For this assessment a two-step procedure has been adopted. In the first step the probability of the peak clad temperature exceeding 800℃ has been evaluated using the response surface, generated from the results of thermal-hydraulic analyses. One of the fuel failure criteria for this reactor is the peak clad temperature exceeding 800℃. Such a high temperature is expected during typical large-break loss-of-coolant accident conditions. The thermal-hydraulic analyses, using the computer code RELAP5/MOD3.2, were done for several cases involving different combinations of six selected uncertain parameters. The probabilistic analysis was carried out using Monte Carlo and first-order reliability methods. The first step results in conditional probability of the peak clad temperature exceeding the criteria subject to the condition of a 200% break in the inlet header. The probability of a 200% break is calculated in the second step. The probability of an inlet header pipe weld rupture has been evaluated based on probabilistic fracture assessment. The pipe break analysis considers the uncertainties in strength, fracture, and stress corrosion properties and initial crack/flaw sizes produced during fabrication or welding. It also accounts pre-service and in-service inspection, inspection quality, and different damage mechanisms such as fatigue and intergranular stress corrosion cracking. The combined results of both these steps give the overall probability of the peak clad temperature exceeding 800℃.
机译:本文介绍了将用于量化不确定性对先进自然循环核反应堆峰值包层温度影响的方法。该方法依赖于概率分析,将不确定参数视为随机变量。本文将介绍由于最大的管道200%(即双头)断裂而部分低压应急堆芯冷却系统不可用而导致的冷却液流失事故。折断处已假定在入口集管处,后者是主传热系统中最大的管道。为了进行该评估,采用了两步程序。第一步,使用热工水力分析结果生成的响应面,评估了包层峰值温度超过800℃的可能性。该反应堆的燃料失效标准之一是包层的峰值温度超过800℃。在典型的大流量冷却液损失事故情况下,预计会出现如此高的温度。使用计算机代码RELAP5 / MOD3.2,对涉及六个选定不确定参数的不同组合的几种情况进行了热工水力分析。概率分析使用蒙特卡洛和一阶可靠性方法进行。第一步导致包层峰值温度超过条件的概率,条件是入口集管破裂200%。在第二步中计算200%断裂的概率。基于概率断裂评估,已经评估了进气总管焊接破裂的可能性。管道断裂分析考虑了强度,断裂,应力腐蚀性能以及在制造或焊接过程中产生的初始裂纹/缺陷尺寸的不确定性。它还说明了服务前和服务中的检查,检查质量以及不同的损坏机制,例如疲劳和晶间应力腐蚀开裂。这两个步骤的综合结果给出了最高包层温度超过800℃的总概率。

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