Core loading optimisation in Slovak VVER-440 reactors
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R. Zajac
Abstract
VVER-440 reactors have been utilized in Slovakia since 1978. So far, the vast majority of their core loadings were designed in VUJE institute. This paper presents a description of the methods and procedures, which have been used for this purpose in the last decade. Main attention is focused on the calculating tools for core refuelling scheme optimization.
Kurzfassung
VVER-440-Reaktoren werden in der Slowakei seit 1978 eingesetzt. Bisher wurde die überwiegende Mehrheit ihrer Kernbeladungen im VUJE-Institut entworfen. Dieses Papier stellt eine Beschreibung der Methoden und Verfahren dar, die in den letzten zehn Jahren zu diesem Zweck eingesetzt wurden. Das Hauptaugenmerk liegt dabei auf den Berechnungswerkzeugen zur Optimierung der Brennelementnachladungen.
References
1 BIPR-7 code, Algorithm description and user manual. (in Russian) RNC “Kurchatovskiy institut”, Moscow, 1995Search in Google Scholar
2 PERMAK code, Algorithm description and user manual. (in Russian) Kurchatov Nuclear Energy Institute, Moscow, 1985Search in Google Scholar
3 Ananyev, J. A.; Bogachev, G. A.: Calculation of power distribution within hexagonal assemblies – SHESTIGRANNIK code. (in Russian) Kurchatov Nuclear Energy Institute, Moscow, 1974Search in Google Scholar
4 Melice, M.: Pressurized Water Reactor Optimal Core Management and Reactivity Profiles. Nucl. Sci. Eng.37 (1969) 451–45710.13182/NSE69-A19119Search in Google Scholar
© 2019, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018
- Technical Contributions/Fachbeiträge
- Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
- C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
- Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
- A procedure for verification of Studsvik's spent nuclear fuel code SNF
- Extension of nodal diffusion solver of Ants to hexagonal geometry
- VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
- Fuel cycles with PK-3+ FAs for VVER-440 reactors
- Prospects for implementation of VVER nuclear fuel enriched above 5%
- Core loading optimisation in Slovak VVER-440 reactors
- Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
- Optimized 18-months low-leakage core loadings for uprated VVER-1000
- Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
- Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
- Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
- Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
- Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
- Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
- Neutron balance in two-component nuclear energy system
Articles in the same Issue
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2018
- Technical Contributions/Fachbeiträge
- Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
- C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
- Adaptation of the gas gap simplified model in DYN3D code to new types of fuel
- A procedure for verification of Studsvik's spent nuclear fuel code SNF
- Extension of nodal diffusion solver of Ants to hexagonal geometry
- VVER-1000 fuel assembly model in CAD-based unstructured mesh for MCNP6
- Fuel cycles with PK-3+ FAs for VVER-440 reactors
- Prospects for implementation of VVER nuclear fuel enriched above 5%
- Core loading optimisation in Slovak VVER-440 reactors
- Statistical evaluation of C-15 cycles in Paks NPP, based on measured in-core data
- Optimized 18-months low-leakage core loadings for uprated VVER-1000
- Leningrad NPP-2 start-up loss of power test and its simulation with use of KORSAR/GP code
- Assessment of the fuel assembly pin-by-pin model in the KORSAR/GP code
- Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
- Analysis of uncontrolled dilution of boric acid concentration in the reactor VVER-1000/320
- Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies
- Criticality safety analysis for GNS IQ® – The Integrated Quiver System for damaged fuel
- Neutron balance in two-component nuclear energy system
