Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
-
I. Pasichnyk
Abstract
The effects of nuclear data covariance on important reactor parameters are investigated. The analyses are performed on the base of the OECD/NEA coolant transient Benchmark (K-3) on measured data at Kalinin-3 Nuclear Power Plant (NPP). For this purpose the GRS uncertainty and sensitivity software package XSUSA is applied to propagate uncertainties in nuclear data libraries to the full core coupled transient calculations. Moreover, based on the previous thermo-hydraulic studies a set of most important thermo-hydraulic parameters is chosen and added to the uncertain input vector. A statistically representative set of coupled ATHLET-PARCS code steady state calculations is analyzed and both integral and local output quantities are compared with the measurements available in the benchmark. The 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 vorliegende Veröffentlichung untersucht die Auswirkung der nuklearen Kovarianzdaten in wichtigen berechneten Werten für nukleare Systeme. Die Unsicherheitsanalyse wird an dem OECD/NEA coolant transient Benchmark (K-3), der auf Messdaten im Kalinin-3-Kernkraftwerk basiert, durchgeführt. 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 wichtigsten thermohydraulischen unsicheren Parametern ausgewählt und wird zu dem unsicheren Eingangsvektor hinzugefügt. Statistisch repräsentativer Satz von gekoppelten ATHLET-PARCS 3D-Ganzkernberechnungen für den stationären Zustand wurde erstellt und sowohl globale als auch lokale berechnete Größen wurden mit den gemessenen Werten verglichen. 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.
References
1 Tereshonok, V. A.; Nikonov, S. P.; Lizorkin, M. P.; Velkov, K.; Pautz, A.; Ivanov, K.: Specification: Kalinin-3 Coolant Transient Benchmark – Switching-off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power. OECD/NEA-DEC, 2008Search in Google Scholar
2 Pasichnyk, I.; Nikonov, S. P.; Velkov, K.: Uncertainty and Sensitivity Analysis of Fuel Assembly Head Parameters in the Framework of Kalinin-3 Benchmark Transient. Proceedings of the 8th International Conference “Safety Assurance of NPP with WWER”, (May 2013, Podolsk, Russia)Search in Google Scholar
3 Pasichnyk, I.; Nikonov, S. P.; Velkov, K.: Sensitivity of Hydrodynamic Parameters' Distributions In VVER-1000 Reactor Presure Vessel (RPV) with Respect to Uncertainty of The Local Hydraulic Resistance Coefficients. In Proceedings of the 23rd AER Symposium (Strbske Pleso, October 2013)10.3139/124.110461Search in Google Scholar
4 Pasichnyk, I.; Klein, M.; Velkov, K.; Zwermann, W.; Pautz, A.: Nuclear Data Uncertainties by the PWR MOX/UO2 Core Rod Ejection Benchmark. Nuclear Technology183 (2013) 46410.13182/NT183-464Search in Google Scholar
5 Zwermann, W.; Gallner, L.; Klein, M.; Krzykacz-Hausmann, B.; Pasichnyk, I.; Pautz, A.; Velkov, K.: Status of XSUSA for Sampling Based Nuclear Data Uncertainty and Sensitivity Analysis. EPJ Web of Conferences42 (2013) 0300310.1051/epjconf/20134203003Search in Google Scholar
6 Scale: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design. ORNL/TM-2005/39Version 6.1, 2011Search in Google Scholar
7 Austregesilo, H. etal.: ATHLET Mod 3.0 Cycle A – Code Documentation, Vol. 4: Models and Methods. GRS-P-1, November 2012Search in Google Scholar
8 Downar, T. etal.: PARCS v3.0 U.S. NRC Core Neutronics Simulator. Theory manual, University of Michigan/U.S. NRS, 2009Search in Google Scholar
9 Nikonov, S.; Kotsarev, A.; Lizorkin, M.: 3D Distribution of Coolant Characteristics in the Reactor Pressure Vessel by Coupled Code ATHLET/BIPR8KN. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark. First Workshop (V1000-CT1), Saclay (Paris), France, 12–13 May, 2003Search in Google Scholar
10 Danilin, S.; Nikonov, S.; Lizorkin, M.; Krukov, S.: Comparative analysis of consistent coast-down of one of four and one of three working main circulation pumps with ATHLET/BIPR8KN and TIGER-1. OECD/DOE/CEA VVER-1000 Coolant Transient Benchmark, First Workshop (V1000-CT1), Saclay (Paris), France, 12–13 May, 2003Search in Google Scholar
11 Mittag, S.; Kliem, S.; Weiss, F. P.; Kyrki-Rajamäki, R.; Hämäläinen, A.; Langenbuch, S.; Danilin, S.; Hadek, J.; Hegyi, G.: Validation of coupled neutron kinetic/thermal-hydraulic codes, Part 1: Analysis of a VVER-1000 transient (Balakovo-4). Annals of Nuclear Energy28 (2001) 857–87310.1016/S0306-4549(00)00095-5Search in Google Scholar
12 NikonovS.; LizorkinM.; LangenbuchS.; VelkovK.: Kinetics and Thermal-Hydraulic Analysis of Asymmetric Transients in a VVER-1000 by the Coupled Code ATHLET-BIPR8KN. 15th Symposium of AER on VVER Reactor Physics and Reactor Safety, Znojmo, Czech Republic, Oct. 3–7, 2005Search in Google Scholar
13 Nikonov, S.; Langenbuch, S.; Velkov, K.: Flow Mixing Modeling by the System Code ATHLET for a VVER-1000 Reactor Vessel Applied for a Main Steam Line Break Transient. Jahrestagung Kerntechnik (Annual Meeting on Nuclear Technology), Aachen, 16–18 May, 2006Search in Google Scholar
14 Nikonov, S. P.; Langenbuch, S.; Lizorkin, M. S.; Velkov, K.: Analyses of the MSLB Benchmark V1000-CT2 by the Coupled System Code ATHLET-BIPR8KN, PHYSOR-2006. Advances in Nuclear Analysis and Simulation, Vancouver, BC, Canada, Sept. 10–14, 2006Search in Google Scholar
15 Trostel, I.; Hegyi, Gy.; Keresztúri, A.; Nikonov, S.: Solution of the OECD NEA KALININ-3 Coolant Transient Benchmark Phase 1 Problem by using the ATHLET code. 19th Symposium of AER on VVER Reactor Physics and Reactor Safety, Varna, Bulgaria, September, 21–25, 2009Search in Google Scholar
16 Ivanov, K.; Avramova, M.; Kodeli, I.; Sartori, E.: Benchmark for Uncertainty Analysis in Modelling (UAM) for Design, Operation and Safety Analysis of LWRs. Volume 1 – Specification and Supporting Data for the Neutronics Cases (Phase I), Version 1.0, NEA/NSC/DOC(2007)23Search in Google Scholar
17 Lizorkin, M. P.: Two-group sparse-grid nodal neutron balance equation of the BIPR-8 computer program. Atomic Energy105 (2008) 8–1710.1007/s10512-008-9059-0Search in Google Scholar
18 Lizorkin, M. P.; Semenov, V. N.; Ionov, V. S.; Lebedev, V. I.: Time dependent spatial neutron kinetic algorithm for BIPR8 and ist verification. Proceedings of the 2nd Symposium of AER (Atomic Energy Research) for Investigating Neutron Physics and Thermohydraulics Problems of Reactor Safety, Paks, Hungary, Sept. 21–26, 1992, Budapest, KFKI, 1992, pp. 389–408Search in Google Scholar
19 Kloos, M.: SUSA Version 3.6: User's Guide and Tutorial, GRSP5, October 2008Search in Google Scholar
© 2015, Carl Hanser Verlag, München
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
Articles in the same Issue
- Contents/Inhalt
- Contents
- Summaries/Kurzfassungen
- Summaries
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2014
- Technical Contributions/Fachbeiträge
- Assessment of the uncertainties of MULTICELL calculations by the OECD NEA UAM PWR pin cell burnup benchmark
- Development of codes and KASKAD complex
- Applying full multigroup cell characteristics from MCU code to finite difference calculations of neutron field in VVER core
- Calculations of 3D full-scale VVER fuel assembly and core models using MCU and BIPR-7A codes
- An analysis of reactivity prediction during the reactor start-up process
- Experimental and computational investigations of heat and mass transfer of intensifier grids
- Implementation of CFD module in the KORSAR thermal-hydraulic system code
- Numerical and experimental investigation of 3D coolant temperature distribution in the hot legs of primary circuit of reactor plant with WWER-1000
- Analyses of Beyond Design Basis Accident Homogeneous Boron Dilution Scenarios
- Analysis of heterogeneous boron dilution transients during outages with APROS 3D nodal core model
- Prospects of subcritical molten salt reactor for minor actinides incineration in closed fuel cycle
- Usage of burnt fuel isotopic compositions from engineering codes in Monte-Carlo code calculations
- Neutron-kinetic and thermo-hydraulic uncertainties in the study of Kalinin-3 benchmark
- Inter-assembly gap deviations in VVER-1000: Accounting for effects on engineering margin factors
