**1. Introduction**

Two-phase flow is a prevailing condition, which applies to a wide range of industrial applications [1]. Especially 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 addition, in a pressurized water reactor, the relevance exists to heat exchangers, such as steam generators [2]. However, it is well known that the characteristics of two-phase flow depend on the flow regime, which differs depending on flow conditions and geometry [3–5]. Conventionally, the void fraction is employed to distinguish the transition between the specified regimes. As depicted in Figure 1, 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. Therefore, 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.

**Figure 1.** Schematic of flow regime map in MARS-KS [6]: (**a**) horizontal flow regime; (**b**) vertical flow regime.

In the previous study [7], the void fraction predictability of three different best-estimate system analysis codes was assessed against the experimental data from the NUPEC (Nuclear Power Engineering Corporation) test facility used for the OECD/NRC PSBT benchmark [8]. The assessment was performed by comparing the result from one-dimensional components in each system code. The results showed that all codes generally predicted the void fraction greater than the experiment. Especially in the case of TRACE V5.0 Patch 5 [9], a significant overprediction tendency in bundles was revealed, especially compared to MARS-KS 1.4 [10] and RELAP5/MOD3.3 Patch 5 [11]. Both MARS-KS and RELAP5 showed almost identical predictions, but they also showed an overprediction tendency at low void conditions. The general overprediction tendency of the system codes was also illustrated from the benchmark results [12]. Because all codes were utilized for the best-estimate safety analysis for nuclear reactor systems, it is necessary to figure out the root cause of the overprediction and to improve the predictability of the void distribution.

Meanwhile, as the system codes were improved with the multi-dimensional capacity for more realistic analyses of complex components, such as a reactor pressure vessel, it was decided that the follow-up study should be expanded to include an assessment with multi-dimensional components in the codes. Thus, it is of interest in this study to figure out the characteristics of one- and multi-dimensional components in the prediction of void distribution and to find out the root cause of the systematic overprediction in void fraction by the system codes.

In this study, a series of assessments on void distribution predictability of the system codes was performed using one- and multi-dimensional components in order to figure out the root cause of the overprediction in void fractions by the system codes. Because RELAP5 does not have the multi-dimensional component, MARS-KS and TRACE have been employed for the assessment. The results from both codes are analyzed both physically and statistically.
