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.
Highlights
Introduction eDalat Nuclear Research Reactor (DNRR) is a 500-kW, pool-type research reactor that uses light water as a moderator and coolant. e reactor was modified and upgraded from an original 250-kW TRIGA Mark II reactor built in 1963
In November 1983, the initial core of the upgraded reactor was loaded with Russian VVRM2 high-enriched uranium (HEU) fuel bundles (FBs) of 36% U-235 enrichment. e natural convection mechanism was reinforced by installing a 2 m high “chimney” above the reactor core, which permits the reactor to operate at a nominal thermal power of 500 kW [1]. e upgraded reactor retains some structures of the original TRIGA Mark II, including the graphite reflector, the horizontal beam tubes, the thermal column, and the biological shielding
When the systematic uncertainties were taken into account, the maximum fuel cladding temperature was predicted to be 98.4°C, as shown in Figure 9, which again was well below the limit value of 103°C. e ONB temperature was about 115.9°C, the minimum Departure from Nucleate Boiling Ratio (DNBR) value was 17.79, and the minimum ONBR was about 1.26, which was below recommended value of 1.4
Summary
Introduction eDNRR is a 500-kW, pool-type research reactor that uses light water as a moderator and coolant. e reactor was modified and upgraded from an original 250-kW TRIGA Mark II reactor built in 1963. E natural convection mechanism was reinforced by installing a 2 m high “chimney” above the reactor core, which permits the reactor to operate at a nominal thermal power of 500 kW [1]. From 2005 to 2012, the project on conversion of the DNRR core from HEU to LEU fuel was implemented. E maximum power peaking factor was used for thermal hydraulic analysis in a conservative method. E PLTEMP/ANL code has been used for steady-state thermal-hydraulic analysis of the DNRR’s LEU core with 92 FAs, because of its good correspondence to both the calculation model and numerical method [8]. After modification of the code, the calculation model of the DNRR with VVR-M2 fuel type, using the “chimney” and natural convection mode for heat removal, was satisfied
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