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Development of a Coupled Neutronics/Thermal-Hydraulics Calculation Algorithm for Analyzing Fast Transients of Power in Pressurized Water Reactors: A Case Study of Bushehr Nuclear Power Plant

Naghavi Dizaji, Davod | 2024

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  1. Type of Document: Ph.D. Dissertation
  2. Language: Farsi
  3. Document No: 57775 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Vosoughi, Naser; Ghafari, Mohsen; Ghofrani, Mohamad Bagher
  7. Abstract:
  8. A comprehensive analysis of the interactions between neutronic and thermal-hydraulics phenomena is crucial to avoid unnecessary costs in nuclear power plants resulting from the implementation of overly conservative approaches in design and operation. A recommended strategy involves transitioning from one-dimensional neutronics calculations, which are based on point kinetic equations, to three-dimensional calculations through the integration of neutronic-thermal-hydraulics simulations. While thermal-hydraulic and safety analyses are commonly performed using system codes like RELAP5, their reliance on one-dimensional calculations limits their ability to model complex phenomena such as mixing and cross-flow. Enhancing system codes through multi-physics and multi-domain simulations offers a promising avenue for improving the accuracy of their outcomes. The current study investigates the impact of asymmetries within the reactor core region, primary and secondary circuits, through the utilization of coupled RELAP5/PARCS calculations to analyze the interplay between neutronics and thermal-hydraulics phenomena. The research involves simulations aimed at exploring various aspects such as cross-flow between fuel assemblies, operation under asymmetric coolant flow conditions, the influence of asymmetric fuel loading in both steady-state and transient scenarios, and a detailed examination of a postulated SBLOCA with an asymmetric SCRAM. The analysis revealed that the cross-flow between two fuel assemblies with the largest power difference during steady-state and symmetrical transients is minimal (approximately 0.16% of the total mass flow rate of a fuel assembly). Additionally, in the asymmetric coolant flow mode, the worst-case scenario occurs when two pumps are placed contiguously. In this situation, the maximum fuel temperature can reach 988°C. Furthermore, simulations of the rod ejection accident during the asymmetric loading mode of the third cycle at the Bushehr nuclear power plant indicated that the amount of positive reactivity injection upon the exit of rods in symmetrical positions varies. In the worst-case scenario, the maximum power can reach 174.5% of the nominal power, and the maximum relative power of the assemblies can reach 2.25. The findings indicate that asymmetries in coolant distribution can significantly influence the accuracy and precision of 1-D thermal-hydraulics calculations, necessitating the use of more comprehensive tools when required. Moreover, it is noted that in certain accident scenarios, the utilization of solely coupled models may not enhance calculation accuracy, suggesting the incorporation of more precise models, such as the best-estimate critical mass flow model, within single physics codes
  9. Keywords:
  10. RELAP5 Code ; Nonsymetry ; SB-LOCA Accident ; Cross Flow ; Purdue Advanced Reactor Core Simulator (PARCS) ; Coupled Neutronics/Thermal-Hydraulics ; Partial Loss of Flow Accident (PLOFA) ; Bushehr Nuclear Power Plant

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