Analysis of initial core and time dependent fuel burnup for high temperature testing reactors (HTTRs)
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Moustafa Aziz
Abstract
HTTRs are high temperature testing reactors which are considered generation IV nuclear power reactors. The reactor core startup and fuel burnup time dependent are simulated and analyzed. For the present computer model core startup, fuel loading with approach to criticality, control rod critical positions, scram reactivity and control rod worth, different core excess reactivity are evaluated. The results are compared with both typical measurements and other previously published methods. Moreover, the time dependent core model and fuel burn up are simulated to calculate the core multiplication factor, cycle length, discharged burnup, core isotopic evolution with time and burnable poisons behavior.
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Research ethics: Not applicable.
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Author contributions: The author has accepted responsibility for the entire content of this manuscript and approved its submission.
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Competing interests: The author states no conflict of interest.
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Research funding: None declared.
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Data availability: The raw data can be obtained on request from the corresponding author.
References
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Articles in the same Issue
- Frontmatter
- Multilateral evaluation of the effects of utilizing thorium oxide in the Bushehr VVER-1000 reactor
- Comparison of modeling methods for the effective diffusivities of IO3− estimated in compacted bentonite using through-diffusion tests under aerobic conditions
- Analysis of initial core and time dependent fuel burnup for high temperature testing reactors (HTTRs)
- A detection and defense security system design for nuclear waste storage against stealth terrorists attack
- Optimization of ECR assisted pre-ionization in GLAST-III via Multiphysics simulation
- Fuzzy reliability algorithm for the shutdown system of research reactor
- System theory safety analysis of network malfunction in nuclear power plant distributed control systems
- Two phase flow analysis of micro channel evaporator to investigate effect of geometry on pressure and heat transfer coefficient with respect to volume of fraction
- Methodology for analyzing dose consequence using atmospheric dispersion code A2CDOSE
- Reliability analysis of digital reactor protection systems in floating nuclear power plants
- Study on comprehensive evaluation method of mental workload level
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