COAL reflooding experiments during a loss of coolant Accident: Effect of the water flow rate, the pressure and the rod power with ballooned rods

IF 1.9 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Nuclear Engineering and Design Pub Date : 2024-10-23 DOI:10.1016/j.nucengdes.2024.113641

G. Repetto , Q. Grando , S. Eymery , R. Van Lochem

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引用次数: 0

Abstract

During a loss of coolant accident (LOCA) in a pressurized water reactor, the drying of the fuel assemblies leads to an increase in the fuel temperature and a deformation of the fuel rod cladding.

The COAL experiments focused on the coolability issue of a partially deformed fuel assembly during water injection with the safety systems using a 7x7 bundle of electrically heated rods. The relocation of the fragmented fuel in the balloons is taken into account by a local increase in power by a factor of 1.5, and the effect of the flow area restriction is provided with various flow blockage (intact geometry up to moderate and long ballooning (100 and 300 mm) with different blockage ratios (80 and 90 %)).

These experiments, in the frame of the PERFROI project, were launched by the “Institut de Radioprotection et de Sureté Nucléaire” (IRSN).

This paper presents the thermal hydraulics parameters and the main results of some experiments carried out in a facility of the STERN Laboratories. We studied the effect of the inlet water flow rate which is the consequence of the amount of water entering the reactor core after the break of the primary circuit, the effect of the pressure and the effect of the rod power as a function of the moment of availability of the safety pumps after the reactor scram. We provide experiments data on the coolability limits for different rod powers, which is given by the minimum of water flow to consider that the reflooding may be not impaired (PCT below the LOCA criterium of 1204 °C). The needed flow is ranging from 7 7 kg/s/m² (with intact rods geometry) at low power up 35 kg/s/m² (with at the high power that remaining in the core 1 min after the reactor scram) with a strong effect of the presence a partially local area due to rod ballooning during the large break LOCA accident. We outlined also the effect of the system pressure with a strong effect on the reflooding process above 10 bar up to 30 (for medium break LOCA).

These results are used to improve and validate the heat exchange models of thermal hydraulics codes dealing with the complex reflooding processes in such a configuration.

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冷却剂损失事故期间的 COAL 再充水实验：水流速度、压力和气球棒功率的影响

在压水反应堆发生冷却剂损失事故（LOCA）期间，燃料组件的干燥会导致燃料温度升高和燃料棒包壳变形。COAL 实验的重点是在使用 7x7 电加热棒束的安全系统注水期间，部分变形燃料组件的可冷却性问题。通过局部功率增加 1.5 倍来考虑气球中碎裂燃料的重新定位，并通过各种流动阻塞（从完整几何形状到中等和较长的气球（100 和 300 毫米），以及不同的阻塞率（80% 和 90%））来提供流动区域限制的影响。这些实验是在 PERFROI 项目框架内进行的，由 "辐射防护与核安全研究所"（IRSN）发起。本文介绍了热水力学参数以及在 STERN 实验室设施中进行的一些实验的主要结果。我们研究了进水流量的影响（进水流量是一回路断开后进入堆芯的水量的结果）、压力的影响和棒功率的影响（棒功率是反应堆熄火后安全泵可用时刻的函数）。我们提供了不同棒功率下冷却极限的实验数据，即考虑到不影响再充水的最小水流量（PCT 低于 LOCA 标准 1204 °C）。所需的水流量从低功率时的 7.7 千克/平方米（棒材几何形状完好）到高功率时的 35 千克/平方米（反应堆熄火 1 分钟后留在堆芯中的水流量）不等，在大断裂 LOCA 事故中，棒材气球对局部区域的影响很大。我们还概述了系统压力的影响，该压力对高于 10 巴至 30 巴（中等断裂 LOCA）的再充水过程有很大影响。这些结果可用于改进和验证热工水力学代码的热交换模型，以处理这种配置下的复杂再充水过程。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Nuclear Engineering and Design 工程技术-核科学技术

CiteScore

3.40

自引率

11.80%

发文量

377

审稿时长

5 months

期刊介绍： Nuclear Engineering and Design covers the wide range of disciplines involved in the engineering, design, safety and construction of nuclear fission reactors. The Editors welcome papers both on applied and innovative aspects and developments in nuclear science and technology. Fundamentals of Reactor Design include: • Thermal-Hydraulics and Core Physics • Safety Analysis, Risk Assessment (PSA) • Structural and Mechanical Engineering • Materials Science • Fuel Behavior and Design • Structural Plant Design • Engineering of Reactor Components • Experiments Aspects beyond fundamentals of Reactor Design covered: • Accident Mitigation Measures • Reactor Control Systems • Licensing Issues • Safeguard Engineering • Economy of Plants • Reprocessing / Waste Disposal • Applications of Nuclear Energy • Maintenance • Decommissioning Papers on new reactor ideas and developments (Generation IV reactors) such as inherently safe modular HTRs, High Performance LWRs/HWRs and LMFBs/GFR will be considered; Actinide Burners, Accelerator Driven Systems, Energy Amplifiers and other special designs of power and research reactors and their applications are also encouraged.