C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant
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I. Pós
Abstract
The C-PORCA/HELIOS models have been used at NPP Paks as basic core neutron physics calculation tools for many years. C-PORCA is a node-wise diffusion model for the purpose of 3D core analysis. HELIOS is a well-known neutron transport code. Its utilisation at Paks NPP has a dual use. This code is a basic tool for preparation of homogenised few-group neutron cross sections inside fuel nodes and areas without fuel and the flexibility of HELIOS allows using it for testing. During the last decade some new kind of fuel assemblies were utilised in Paks. In order to ensure the accuracy and performance requirements of the off-line core analysis and in-core monitoring, continuous development and testing of the codes have been performed. In this paper the main characteristics of the diffusion solver applied in the C-PORCA model are described. The accuracy of this solver is also demonstrated on the basis of comparisons with different international references available in hexagonal geometry. The C-PORCA results have been compared against benchmark data produced in the framework of the AER (Atomic Energy Research) community in recent decades. All presented comparisons illustrate that the accuracy of the C-PORCA diffusion solver is excellent.
Kurzfassung
Die C-PORCA/HELIOS-Modelle werden im KKW Paks seit vielen Jahren als Basiswerkzeuge zur Berechnung der Neutronenphysik eingesetzt. C-PORCA ist ein nodales Diffusionsprogramm für die 3D-Kernanalyse. HELIOS ist ein bekannter Neutronentransportcode. HELIOS wird dabei sowohl als grundlegendes Werkzeug zur Berechnung homogenisierter Weniggruppen-Wirkungsquerschnitte für Brennstoff- und brennstofffreie Bereiche als auch für Testzwecke eingesetzt. In den letzten zehn Jahren wurden in Paks neuartige Brennelemente eingesetzt. Um die Genauigkeits- und Leistungsanforderungen der Offline-Kernanalyse und des In-Core-Monitorings zu gewährleisten, wurden kontinuierliche Entwicklungen und Tests der Codes durchgeführt. In diesem Beitrag werden die wichtigsten Eigenschaften des im C-PORCA-Modell verwendeten Diffusionslösers beschrieben. Die Genauigkeit dieses Lösers wird auch anhand von Vergleichen mit verschiedenen internationalen Referenzen in hexagonaler Geometrie nachgewiesen. Die C-PORCA-Ergebnisse wurden mit Benchmark-Daten verglichen, die im Rahmen der AER (Atomic Energy Research) Community in den letzten Jahrzehnten erstellt wurden. Alle vorgestellten Vergleiche zeigen, dass die Genauigkeit des C-PORCA Diffusionslösers ausgezeichnet ist.
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© 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
