Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

Abstract

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.

Determination of the Activity Inventory in the Structural Components of the Dalat Nuclear Research Reactor for Its Decommissioning Planning

“…From 1982 to 1984, under the former Soviet Union's assistance, this reactor was reconstructed, with its thermal power upgraded to 500 kW; the upgraded reactor was restarted in November 1983 using highly enriched uranium (HEU) fuel fabricated by the former Soviet Union. e reactor serves mainly for radioisotope production, neutron activation analysis, and manpower training [1,2].…”

“…Distilled light water is used as coolant, neutron moderator, re ector, and radiation shielding. After its reconstruction and upgrading, since March 1984, the reactor has been o cially put into operation for the purposes of radioisotope production, neutron activation analysis, fundamental and applied research, and manpower training but without having a preliminary decommissioning plan [1][2][3].…”

Analysis of control rod effect on the safety operation parameter of the RSG-GAS Core

“…It means that change of control rod positions is followed by the change of neutronic core parameters [7]. It changes the axial power distibution as well that makes non uniform distribution of power [8]. The high power peaking will result a high fission products at that point.…”