Thermal-hydraulic simulation of loss of flow accident for WWR-S research reactor
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Salah El-Din El-Morshedy
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Mohamed Moawed
Abstract
A thermal-hydraulic model has been developed by RELAP5 code to simulate the thermal-hydraulic behavior of a WWR-S research reactor under loss of flow accident (LOFA). The reactor power is 2 MW with downward flow direction and different types of fuel bundles of different power densities and different coolant flow-rates. The simulation is performed for two scenarios; protected and unprotected LOFA. In the protected LOFA scenario; Scram is triggered as soon as the coolant flow rate reaches 85 % of its nominal value while in the unprotected LOFA scenario; the reactor continues operation during pump cost-down and natural circulation flow. Once the flow is low enough, the buoyancy force increases in the core leading to flow inversion phenomenon and establishment of a natural circulation mechanism within the core coolant channels in both scenarios. In the protected LOFA scenario, the coolant remains subcooled by a vast margin during transient while bulk boiling is predicted at the upper part of the core for the unprotected LOFA scenario. The heat fluxes leading to the onset of nucleate boiling and the critical heat flux are predicted and the safety margins are determined. The results for both scenarios are analyzed and discussed.
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Research ethics: Not applicable.
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Informed consent: Not applicable.
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Author contributions: El-Morshedy: 50 %, Moawed: 15 %, Abdelrahman: 5 %, Elnour: 10 % and Taha: 20 %.
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Use of Large Language Models, AI and Machine Learning Tools: None declared.
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Conflict of interest: The authors state no conflict of interest.
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Research funding: None declared.
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Data availability: Not applicable.
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Articles in the same Issue
- Frontmatter
- Comprehensive review of surface contamination in nuclear waste waters: identification, quantification, and mitigation strategies
- Methodology of probabilistic safety assessment for transportation of radioactive material
- A new approach to determine abnormality of radioactive discharges from pressurized water reactors and to derive abnormality indicators correlated with a specific causal event
- A critical analysis of the role of artificial intelligence and machine learning in enhancing nuclear waste management
- Design study of gas-cooled fast reactor with natural uranium as fuel employing modified CANDLE shuffling strategy in the axial direction
- Synthesis, structural transformation and magnetic properties of the Nd(III)-doped Fe3−xNd x O4 (0 ≤ x ≤ 0.9): an analogue for actinicles immobilization
- Examination of the use of thorium-based fuel for burning minor actinides in European sodium cooled fast reactor
- Solitary wave form of reaction rate in graphite diffusive medium using different neutron absorbers
- Evaluation of the unavailability of the primary circuit of Triga SSR reactor, importance factors and risk criteria for its components
- Thermal-hydraulic simulation of loss of flow accident for WWR-S research reactor
- A quick parameter configuration tool for SCHISM’s ocean transport simulation of radioactive materials
- Main heat transport system configuration influence on steam drum level control and safety for a pressure tube type boiling water reactor with multiple interconnected loops
- Testing the thermal performance of water cooling towers
- Design a robust intelligent power controller for pressurized water reactor using particle swarm optimization algorithm
- Calendar of events
