Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark
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I. Pasichnyk
Abstract
An uncertainty and sensitivity analysis is performed for the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). A switch off of one main coolant pump (MCP) at nominal reactor power is calculated using a coupled thermohydraulic and neutron-kinetic ATHLET-PARCS code. The objectives are to study uncertainty of total reactor power and to identify the main sources of reactor power uncertainty. The GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. A set of most important thermal-hydraulic parameters of the primary circuit is identified and a total of 23 thermohydraulic parameters are statistically varied using GRS code SUSA. The ATHLET model contains also a balance-of-plant (BOP) model which is simulated using ATHLET GCSM module. In particular the operation of the main steam generator regulators is modelled in detail. A set of 200 varied coupled ATHLET-PARCS calculations is analyzed. The results obtained show a clustering effect in the behavior of global reactor parameters. It is found that the GCSM system together with varied input parameters strongly influence the overall nuclear power plant behavior and can even lead to a new scenario. Possible reasons of the clustering effect are discussed in the paper. This work is a step forward in establishing a “best-estimate calculations in combination with performing uncertainty analysis” methodology for coupled full core calculations.
Kurzfassung
Die Unsicherheits- und Sensitivitätsanalyse wird an dem OECD/NEA-coolant-transient-Benchmark (K-3) basierend auf Messdaten im Kalinin-3-Kernkraftwerk durchgeführt. Die Abschaltung einer Hauptkühlmittelpumpe bei Volllast wird mit dem gekoppelten ATHLET-PARCS-Code simuliert. Das Ziel der vorliegenden Arbeit ist es, die Hauptbeiträge der Reaktorleitungsunsicherheit zu identifizieren. Zu diesem Zweck wird die GRS-Statistiksoftware XSUSA verwendet. Mit dieser Methode ist es möglich, die Berechnungsunsicherheiten in allen Phasen der nuklearen Rechenkette – von der Stabzellrechnung bis zur Kerntransiente – zu bestimmen. Darüber hinaus wird auf der Grundlage der früheren Untersuchungen ein Satz von 23 wichtigsten thermohydraulischen unsicheren Parametern ausgewählt und wird zu dem unsicheren Eingangsvektor hinzugefügt. Die Nachbildung der Regelung- und Leittechniksysteme erfolgt mit den General Simulation Control Module (GCSM) des ATHLET-Code. Insbesondere werden die Regelung, Kontroll- und Hilfssysteme des Dampferzeugers modelliert. Für die Unsicherheitsanalyse der Ausgangsparameter wurden insgesamt 200 ATHLET-PARCS-Rechenläufe ausgeführt. Die Ergebnisse zeigen einen Clustering-Effekt der globalen Reaktorparameter. Es wurde festgestellt, dass GCSM-System zusammen mit variierten Eingangsparametern das Gesamtverhalten des Kernkraftwerk stark beeinflussen. Die möglichen Gründe des Clustering-Effekts werden diskutiert. Diese Arbeit ist ein wesentlicher Schritt zur Herstellung einer Methodik für Best-Estimate-Berechnungen in Kombination mit der Durchführung von Unsicherheits- und Sensitivitätsanalysen („Best Estimate Plus Uncertainty“ – BEPU) für Ganzkerntransienten.
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© 2016, Carl Hanser Verlag, München
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2015
- Technical Contributions/Fachbeiträge
- Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes
- Xenon instability study of large core Monte Carlo calculations
- Error detection in core loading in the condition of asymmetrical distribution of power
- New models in VERONA 7.0 system
- Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants
- Verification of three-dimensional neutron kinetics model of TRAP-KS code regarding reactivity variations
- Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents
- Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark – re-connection of an isolated loop
- Uncertainties of the KIKO3D-ATHLET calculations using the Kalinin-3 benchmark (Phase II) data
- Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark
- Efficient introduction of natural uranium and thorium into nuclear energy system
- Economical aspects of multiple plutonium and uranium recycling in VVER reactors
- Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO
- Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2015
- Technical Contributions/Fachbeiträge
- Monte-Carlo code calculation of 3D reactor core model with usage of burnt fuel isotopic compositions, obtained by engineering codes
- Xenon instability study of large core Monte Carlo calculations
- Error detection in core loading in the condition of asymmetrical distribution of power
- New models in VERONA 7.0 system
- Methodology for determining of the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants
- Verification of three-dimensional neutron kinetics model of TRAP-KS code regarding reactivity variations
- Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents
- Qualification of coupled 3D neutron kinetic/thermal hydraulic code systems by the calculation of a VVER-440 benchmark – re-connection of an isolated loop
- Uncertainties of the KIKO3D-ATHLET calculations using the Kalinin-3 benchmark (Phase II) data
- Coupled code analysis of uncertainty and sensitivity of Kalinin-3 benchmark
- Efficient introduction of natural uranium and thorium into nuclear energy system
- Economical aspects of multiple plutonium and uranium recycling in VVER reactors
- Neutronic analysis of absorbing materials for the control rod system in reactor ALLEGRO
- Uncertainty analysis and flow measurements in an experimental mock-up of a molten salt reactor concept
