Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments
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R. Ferrer
Abstract
Studsvik has recently extended the CASMO5 advanced lattice physics code for the analysis of VVER 1000 and 1200 reactors. These extensions form the basis of CASMO5-VVER, which is primarily intended to compute homogenized nodal data for SIMULATE5-VVER. CASMO5-VVER leverages the latest nuclear data and numerical methods to VVER analyses. The current CASMO5 data library, based on the ENDF/B-VII.1 nuclear data evaluation, features a 586 energy group structure and nuclear data for hundreds of unique nuclides. Resonance self-shielding, based on the Equivalence Theory and an Optimal Two-Term Rational (OTTR) method, has been extended to support hexagonal geometry. The solution to the two-dimensional transport equation is based on the new Linear Source (LS) approximation for the Method of Characteristics (MOC). The acceleration of the MOC solution is attained through the implementation of a Coarse-Mesh Nonlinear Diffusion (CMND) acceleration. Results from preliminary validation against various critical experiments are presented in this work.
Kurzfassung
Studsvik hat kürzlich den fortgeschrittenen Gittercode CASMO5 für die Analyse von VVER 1000 und 1200 Reaktoren erweitert. Diese Erweiterungen bilden die Grundlage von CASMO5-VVER, das in erster Linie zur Berechnung homogenisierter nodaler Daten für SIMULATE5-VVER gedacht ist. CASMO5-VVER nutzt die neuesten nuklearen Daten und numerischen Methoden für die VVER-Analyse. Die aktuelle CASMO5-Datenbibliothek, die auf den ENDF/B-VII.1 Kerndaten basiert, verfügt über eine 586 Energiegruppenstruktur und Kerndaten für Hunderte von einzelnen Nukliden. Die Resonanz-Selbstabschirmung, basierend auf der Äquivalenztheorie und einer optimalen Zwei-Term-Rational-Methode (OTTR), wurde erweitert, um die hexagonale Geometrie zu unterstützen. Die Lösung der zweidimensionalen Transportgleichung basiert auf einer neuen linearen Quellenapproximation für die Charakteristikenmethode (MOC). Die Beschleunigung der MOC-Lösung wird durch die Implementierung einer nichtlinearen Grobgitter Diffusionsbeschleunigung erreicht. Ergebnisse der vorläufigen Validierung anhand verschiedener kritischer Experimente werden in dieser Arbeit vorgestellt.
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© 2019, Carl Hanser Verlag, München
Artikel in diesem Heft
- 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
Artikel in diesem Heft
- 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