使用 SOCRAT 代码支持严重事故建模的 FA 水反灌计算和实验研究结果摘要

IF 0.9 Q4 ENERGY & FUELS
I. S. Akhmedov, K. S. Dolganov, N. I. Ryzhov, D. Yu. Tomashchik, A. E. Kiselev
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引用次数: 0

摘要

在现代理论知识和代表性实验研究结果的基础上,考虑了燃料组件再充水的现象。实验中保持的参数(试验段压力、水过冷度、充水开始时的包壳峰值温度、束功率等)与压水反应堆实施管理假定严重事故的措施时预计的参数接近。已经制定了燃料棒组件再充水过程的清单,并确定了可能导致燃料棒模拟器包壳与蒸汽-水混合物之间热交换局部条件发生变化并影响其淬火的具体影响。对实验研究结果的比较表明,冷却水流速对燃料组件上部淬火时间测量值的分布有影响。从再充水物理学的角度,我们分析了 SOCRAT 代码在不同现象复杂性实验(在完整堆芯、强烈的蒸汽-锆反应、熔体形成)中的验证结果。分析表明,SOCRAT 代码正确预测了燃料棒模拟器包壳的温度历史、淬火时间和实验过程中释放的氢气总质量,但有轻微低估的趋势;建模结果与实验数据并不矛盾。在验证过程中,确定了在对不同现象复杂度的实验进行建模时,热水力学模型在计算淬火时间和氢气产生总质量方面对评估模型误差的贡献最大。SOCRAT 代码的良好预测能力证实了一维方法在燃料组件再充水建模中的适用性。
本文章由计算机程序翻译,如有差异,请以英文原文为准。

Summary of the Results of Computational and Experimental Studies of Water Reflood of FA in Support of Modeling of Severe Accidents using the SOCRAT Code

Summary of the Results of Computational and Experimental Studies of Water Reflood of FA in Support of Modeling of Severe Accidents using the SOCRAT Code

Based on modern theoretical knowledge and the results of representative experimental studies, the phenomenology of reflooding of fuel assemblies is considered. The parameters (test section pressure, water subcooling, peak cladding temperature at the start of the flooding, bundle power, etc.) maintained in the experiments under consideration are close to those expected when implementing measures to manage hypothetical severe accidents at pressurized water reactors. A list of processes accompanying the reflood of fuel-rod assemblies has been formulated, and specific effects have been established that can lead to a change in the local conditions of heat exchange between the cladding of fuel-rod simulators and the steam-water mixture and affect their quenching. A comparison of the results of experimental studies showed the influence of cooling water flow rate on the spread of measured values of quench time in the upper part of the fuel assembly. The view of reflood physics allowed us to analyze the results of validation of the SOCRAT code in experiments of varying phenomenological complexity (in an intact core, with an intense steam-zirconium reaction, formation of a melt). The analysis showed that the SOCRAT code correctly predicts the temperature histories of the fuel-rod simulator claddings, the quench time, and the total mass of hydrogen released during the experiment with a tendency toward slight underestimation; the modeling results do not contradict the experimental data. During validation, it was established that the thermal hydraulics model makes the greatest contribution to the assessment of the model error in calculating the quench time and the total mass of hydrogen production when modeling experiments of varying phenomenological complexity. Good predictive capabilities of the SOCRAT code confirmed the applicability of a one-dimensional approach to modeling the reflooding of fuel assemblies.

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来源期刊
CiteScore
1.30
自引率
20.00%
发文量
94
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