Many research reactors with plate-type fuel are operated worldwide and the thermal hydraulic core analysis is done with dedicated codes or with 1D system thermal hydraulic codes. At the Karlsruhe Institute of Technology (KIT), investigations are underway to develop advanced coupled neutronic/thermal hydraulic simulation tools for the Safety analysis of research reactors with plate-type fuel. For this purpose, the subchannel thermal hydraulic code SubChanFlow was extended and validated using relevant tests. The SubChanFlow extension consists of the addition of a heat conduction module for a thin plate and the implementation of heat transfer correlations for rectangular channels. For the validation of the modified SubChanFlow code, temperature data from RA-6 experiment obtained in a Reynolds range of $7\times {10}^3<Re<1.4\times {10}^4$ were used.

The RA-6 experiment was carried out at the Bariloche Atomic Center to support the development of MTR research reactors. Different heat transfer correlations e.g., the Colburn, Dittus-Boelter, Gnielinski, Sieder-Tate, and Y-Sudo were used to simulate the RA-6 tests. The comparison of the plate temperatures measured along the active height with the predicted values using the extended SubChanFlow code shows a deviation ranging from ± 4 °C to ± 18 °C for the Colburn and Y-Sudo correlations, respectively, which is promising compared to results of other codes.

在全球范围内，许多使用板式燃料的研究堆都在运行，并且热液压堆芯分析是通过专用代码或一维系统热液压代码完成的。在卡尔斯鲁厄技术学院（KIT），正在进行研究以开发先进的中子/热耦合水力模拟工具，以对板式燃料研究堆的安全性进行分析。为此，使用相关测试扩展并验证了子通道热液压代码SubChanFlow。SubChanFlow扩展包括为薄板增加一个导热模块，并为矩形通道实现传热相关性。为了验证修改后的SubChanFlow代码，在雷诺范围内从RA-6实验获得的温度数据为$7\times {10}^3<\mathrm{[R}Ë<1.4\times {10}^4$ 被使用。

RA-6实验在巴里洛切原子中心进行，以支持MTR研究反应堆的开发。使用不同的传热相关性，例如Colburn，Dittus-Boelter，Gnielinski，Sieder-Tate和Y-Sudo，来模拟RA-6测试。使用扩展的SubChanFlow代码将沿有效高度测量的板温度与预测值进行比较，结果表明，对于Colburn和Y-Sudo相关性，偏差分别在±4°C至±18°C范围内，与其他代码的结果。