This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

中文翻译：

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低浓铀燃料大叻核反应堆的稳态热工水力分析

本文介绍了使用PLTEMP / ANL代码对大叻核研究堆（DNRR）设计的工作堆进行稳态热工水力分析的结果。堆芯设计为装有92个低浓缩铀（LEU）VVR-M2燃料束（FB）和12个铍棒，围绕着堆芯中心的中子阱，用104个高浓缩铀替代了先前的堆芯（ HEU）VVR-M2 FB。在将此代码用于设计的LEU工作核心的热工液压分析之前，通过将计算结果与从DNRR的HEU工作核心收集的实验数据进行比较来对其进行验证。计算结果与实测数据之间的差异分别是燃料包壳和出口冷却液温度分别最高约0.8°C和1.5°C。在设计计算中，通过评估燃料包壳和冷却液温度以及其他安全参数（如偏离核沸腾比（DNBR）和发生核沸腾比（ONBR））来确定热工液压安全性。计算结果表明，在正常运行条件下，没有任何不确定性参数的情况下，在满载额定功率为500 kW的情况下，最热的FB的最高燃料包壳温度约为90.4°C，低于其极限值103°C。最小DNBR为32.0，远高于建议值1.5，而最小ONBR为1.43，高于VVR-M2 LEU燃料类型的建议值1.4。如果考虑到全局和局部热通道因素，则最热的FB处燃料包壳的最高温度约为98.4°C，仅对于全球而言，最高温度为114.3°C，再加上局部热通道因素。计算结果证实了装有92个新鲜VVR-M2 FB的设计LEU内核的安全运行。