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首页> 外文会议>International congress on advances in nuclear power plants;ICAPP 2010 >Applicability of Sub-scale Integral Test Data and TRACG Computer Code to Loss-of-Coolant Accidents of Uprated BWRs
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Applicability of Sub-scale Integral Test Data and TRACG Computer Code to Loss-of-Coolant Accidents of Uprated BWRs

机译：子规模的整体测试数据和TRACG计算机代码对升级的BWR的冷却液损失事故的适用性

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摘要

In the 1980s, the General Electric Company (GE) and the Japan Atomic Energy Research Institute (JAERI) conducted transient thermal hydraulic experiments in the FIST and ROSA-III sub-scale integral test facilities, respectively, to simulate loss-of-coolant accidents (LOCAs) in jet pump Boiling Water Reactors (BWRs). In subsequent years, the BWR fuel assemblies changed from 8×8 to 10×10 rod arrays and many BWRs were or are being uprated to as much as 120% of the Original Licensed Thermal Power (OLTP). This paper utilizes a current scaling methodology, in combination with TRACG analyses, to examine the applicability of the FIST and ROSA-III data to an uprated (120% of OLTP) BWR/6 loaded with 10×10 fuel assemblies for the case of a guillotine rupture of a recirculation suction line with a single failure. The results show that (1) the FIST and ROSA-III test data remain applicable to uprated BWRs; and (2) TRACG is capable of simulating the important phenomena that govern BWR vessel depressurization and inventory change, and the peak cladding temperature during a LOCA in an uprated BWR.
机译：在1980年代，通用电气公司（GE）和日本原子能研究院（JAERI）分别在FIST和ROSA-III次级规模整体测试设施中进行了瞬态热工水力实验，以模拟冷却液流失事故喷射泵沸水反应堆（BWR）中的（LOCA）。在随后的几年中，BWR燃料组件从8×8变为10×10棒状阵列，并且许多BWR已经或正在被升级到原始许可热功率（OLTP）的120％。本文利用当前的定标方法，结合TRACG分析，来检验FIST和ROSA-III数据对装载有10×10燃料组件的升级版（OLTP的120％）BWR / 6的适用性。循环吸气管路断头台破裂，一次失败。结果表明：（1）FIST和ROSA-III测试数据仍然适用于升级后的BWR; （2）TRACG能够模拟控制BWR容器降压和库存变化的重要现象，以及提高BWR的LOCA期间的包层峰值温度。

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