Thermal-hydraulic and safety analysis of alternative ceramic fuels in next generation of VVER-1000 reactor
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Ahmad Rahimi
,
Ebrahim Heidari
and
Kavoos Abbasi
Abstract
The main aim of this study is the thermohydraulic and safety analysis of alternative UN and UC ceramic fuels in the next generation of VVER-1000/V446 reactor core. One of the advantages of these alternative fuels is their higher thermal conductivity and density relative to the UO2 conventional fuels. Thermohydraulic analysis has been performed under full power conditions. Also, the safety evaluation of alternative ceramic fuels has been researched by the Rod Ejection Accident (REA) scenario. RELAP5 Mod3.2 code is used for thermo-hydraulic analysis in both steady state and transient state in VVER-1000/V446 reactor. In addition, calculations of the fuel center temperature in mean and hot channels for steady state, the temperature of the hottest rod in the fuel assembly, the radial distribution of fuel temperature for both channels, and the maximum clad temperature during the reactivity injection have been performed. The results showed that in the steady state with full power, the fuel centreline temperature in UO2 is about twice the corresponding values for UN and UC alternative fuels which is very attractive from a safety point of view. In general, due to the large difference between the hot spot point and the melting point in alternative fuels, the thermal power and safety factors can be significantly increased in the next generation of VVER-1000/V446 reactors.
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Research ethics: Not applicable.
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Informed consent: Not applicable.
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Author contributions: A. Rahimi: Methodology, Software, Investigation, Resources, Writing – Original Draft. M. Kheradmand Saadi: Conceptualization, Methodology, Software, Writing – Review & Editing. E. Heidari: Methodology, Writing – Review & Editing. K. Abbasi: Methodology, Writing – Review & Editing.
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Use of Large Language Models, AI and Machine Learning Tools: None declared.
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Conflict of interests: The authors state no conflict of interest.
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Research funding: None declared.
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Data availability: Data available upon request.
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Articles in the same Issue
- Frontmatter
- Study of reliability prediction method and uncertainty analysis of the pump and valve of a floating nuclear power plant
- Thermal-hydraulic and safety analysis of alternative ceramic fuels in next generation of VVER-1000 reactor
- Transient thermal analysis for optimal armor material in upgraded mockup for PST using MOOSE Framework
- Analysis of hypothetical channel flow blockage of a tubular fuel assembly in a 10 MWth nuclear research reactor
- Preliminary neutronic analyses for annular VHTR with different coated fuel particle designs
- Improved intelligent model predictive controller for the nuclear power reactor system
- Monte Carlo simulation of photo-peak efficiency response function and calculation of total efficiency for a NaI (Tl) scintillator detector
- Calculation and application research on real-time source term diffusion in NPP auxiliary building under accident circumstances
- Calendar of events – issue 6 of KERN
