Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor
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 surroundi...
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| Format: | Article |
| Language: | English |
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Wiley
2021-01-01
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| Series: | Science and Technology of Nuclear Installations |
| Online Access: | http://dx.doi.org/10.1155/2021/6673162 |
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| author | Kien-Cuong Nguyen Vinh-Vinh Le Ton-Nghiem Huynh Ba-Vien Luong Nhi-Dien Nguyen |
| author_facet | Kien-Cuong Nguyen Vinh-Vinh Le Ton-Nghiem Huynh Ba-Vien Luong Nhi-Dien Nguyen |
| author_sort | Kien-Cuong Nguyen |
| collection | DOAJ |
| description | 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. |
| format | Article |
| id | doaj-art-f97d2f2599e34823bf4b4ae88c8ccdc0 |
| institution | DOAJ |
| issn | 1687-6075 1687-6083 |
| language | English |
| publishDate | 2021-01-01 |
| publisher | Wiley |
| record_format | Article |
| series | Science and Technology of Nuclear Installations |
| spelling | doaj-art-f97d2f2599e34823bf4b4ae88c8ccdc02025-08-20T03:20:09ZengWileyScience and Technology of Nuclear Installations1687-60751687-60832021-01-01202110.1155/2021/66731626673162Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research ReactorKien-Cuong Nguyen0Vinh-Vinh Le1Ton-Nghiem Huynh2Ba-Vien Luong3Nhi-Dien Nguyen4Dalat Nuclear Research Institute, Vinatom, 01 Nguyen Tu Luc Street, Dalat 670000, Lam Dong, VietnamDalat Nuclear Research Institute, Vinatom, 01 Nguyen Tu Luc Street, Dalat 670000, Lam Dong, VietnamDalat Nuclear Research Institute, Vinatom, 01 Nguyen Tu Luc Street, Dalat 670000, Lam Dong, VietnamDalat Nuclear Research Institute, Vinatom, 01 Nguyen Tu Luc Street, Dalat 670000, Lam Dong, VietnamDalat Nuclear Research Institute, Vinatom, 01 Nguyen Tu Luc Street, Dalat 670000, Lam Dong, VietnamThis 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.http://dx.doi.org/10.1155/2021/6673162 |
| spellingShingle | Kien-Cuong Nguyen Vinh-Vinh Le Ton-Nghiem Huynh Ba-Vien Luong Nhi-Dien Nguyen Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor Science and Technology of Nuclear Installations |
| title | Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor |
| title_full | Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor |
| title_fullStr | Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor |
| title_full_unstemmed | Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor |
| title_short | Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor |
| title_sort | steady state thermal hydraulic analysis of the leu fueled dalat nuclear research reactor |
| url | http://dx.doi.org/10.1155/2021/6673162 |
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