Abstract
The previous study, where the void fraction predictability of three different thermal-hydraulic system codes was assessed against PSBT (PWR Subchannel and Bundle Test) benchmark data, indicated a general overprediction tendency of all system codes, especially in bundles. Because all codes have been utilized for best-estimate analyses, it is necessary to conduct further assessments in order to find the root cause of the overprediction. A further assessment has been performed using two thermal-hydraulic system codes, TRACE V5.0 patch 5 and MARS-KS 1.4, and the assessment has been carried out for both one- and multi-dimensional components. The results indicate that there is no significant difference in the predictability of the void fraction between one- and multi-dimensional components. In addition, it is found that the turbulent mixing model implemented for the multi-dimensional component of MARS-KS does not play an important role in the prediction of void distribution. Meanwhile, TRACE reveals a significant overprediction due to much less crossflow calculation compared to MARS-KS. By conducting an additional analysis with the modified one-dimensional models, it is clearly confirmed that crossflow significantly affects the void distribution. Therefore, it is concluded that the model for the thermal hydraulic mixing by crossflow in each system code should be improved in order to predict the void distribution in bundles appropriately.
Highlights
Two-phase flow is a prevailing condition, which applies to a wide range of industrial applications [1]
It is concluded that the model for the thermal hydraulic mixing by crossflow in each system code should be improved in order to predict the void distribution in bundles appropriately
It is well known that the characteristics of two-phase flow depend on the flow regime, which differs depending on flow conditions and geometry [3,4,5]
Summary
Two-phase flow is a prevailing condition, which applies to a wide range of industrial applications [1]. In the nuclear field, the most relevant application is the boiling water reactor, in which a phase change of flowing liquid directly occurs within the reactor core. In a pressurized water reactor, the relevance exists to heat exchangers, such as steam generators [2]. It is applied as one of the criteria for flow regime prediction, and this form of flow map is generally utilized in the best-estimate system analysis codes in the nuclear field. The precise prediction of the void fraction has great importance in the best-estimate safety analysis methodology since additional conservatism could be induced by the inaccurate prediction of the void fraction, which plays a negative role from a coolability point of view
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