Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries
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V. Yu. Blandinskiy
Abstract
This article presents a comparison of results of simulation of VVER-440 reactor fuel burnup by means of ISTAR set of tools using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries. Code verification is required on the basis of both experimental data and computational uncertainty analysis. A VVER-440 fuel rod burnup simulation using ENDF/B-VII.0 and JEFF-3.1 evaluated nuclear data libraries was performed as preliminary step before comparison of nuclear data uncertainty analysis and calculation/experiment. Fuel irradiation conditions and SNF nuclide composition experimental data were obtained from radiochemical assays of irradiated VVER-440 fuel report. Fuel burnup simulation was carried out by means of ISTAR set of tools using Monte-Carlo code for spectrum averaged reaction rates evaluation required for solution of Bateman's equations. As a result, Kinf and nuclide composition across burnup were estimated for the sample considered. On the basis of two nuclear data libraries’ implementation for depletion simulation it was shown that Kinf uncertainties rise across burnup from 0.4% at the beginning of life up to 0.7% at the end of life. Heavy nuclides composition uncertainties run up to 9.5% for Am-241 at the end of the 4th life time; however for main uranium and plutonium isotopes the uncertainties do not exceed 1.2%. Most fission products composition uncertainties stay below 7.3% except Sm-150 nuclide (15.1%). The next step will be a comparison with experimental spent nuclear fuel isotopic composition data considering the above mentioned uncertainties.
Kurzfassung
Dieser Artikel stellt einen Vergleich der Ergebnisse der Simulation des VVER-440-Reaktorbrennstoffabbrands mit Hilfe von ISTAR-Werkzeugen unter Verwendung von ENDF/B-VII.0 und JEFF-3.1 ausgewerteten Nukleardatenbibliotheken dar. Die Überprüfung der berechneten Ergebnisse ist sowohl mit Hilfe experimenteller Daten als auch mit Hilfe der Bestimmung der rechnerischen Unsicherheit durchzuführen. Zur Vorbereitung dieser Überprüfung wurde das Verhalten eines VVER-440 Brennelements unter Verwendung der Bibliotheken ENDF/B-VII.0 und JEFF-3.1 berechnet. Die Bestrahlungsbedingungen und die SNF-Nuklidzusammensetzung wurden aus radiochemischen Untersuchungen des bestrahlten VVER-440-Brennstoffs bestimmt. Die Berechnung des Brennstoffabbrands wurde mit Hilfe von ISTAR-Werkzeugen, die zur Lösung der Bateman-Gleichungen Monte-Carlo-Code für die Auswertung der spektral gemittelten Reaktionsgeschwindigkeiten, berechnet. Infolgedessen wurden Kinf und die Nuklidzusammensetzung für die betrachtete Probe bestimmt. Mit zwei implementierten Nukleardatenbibliotheken für die Simulation der Verarmung wurde gezeigt, dass die Unsicherheiten von 0,4% zu Beginn des Zyklus auf 0,7% am Ende des Zyklus steigen. Größere Unsicherheiten in der Zusammensetzung der Nuklide liegen bei Am-241 am Ende der 4. Zyklus bei bis zu 9,5%, bei den Haupt-Uran- und Plutoniumisotopen jedoch nicht über 1,2%. Die meisten Spaltprodukte bleiben mit Ausnahme des Sm-150-Nuklids (15,1%) unter 7,3%. Der nächste Schritt wird ein Vergleich mit experimentellen Daten zur Isotopenzusammensetzung abgebrannter Brennelemente unter Berücksichtigung der oben genannten Unsicherheiten sein.
References
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© 2018, 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 2017
- Technical Contributions/Fachbeiträge
- SIMULATE5-HEX extension for VVER analyses
- Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code
- “Full-Core” VVER-440 extended calculation benchmark
- Calculation of “full core” VVER-1000 benchmark
- Study of neutron-physical characteristics of VVER-1200 considering feedbacks using MCU Monte Carlo code
- Advantages of VVER-440 fuel cycles with new fuel assemblies
- A neutronics feasibility study on utilization of a thinned cladding fuel design at Loviisa NPP
- Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors
- Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries
- Simulation of standard temperature control indications at the outlet of a fuel assembly of VVER1000 reactor of Rostov NPP unit No. 2
- Power transient calculations with VERONA
- Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode
- Renewing the refueling neutron monitoring and reactivity measurement systems at Paks NPP
- Hot channel calculation methodologies in case of VVER-1000/1200 reactors
- Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes
- Simulation of a hypothetical MSLB core transient in VVER-1000 with several stuck rods
Artikel in diesem Heft
- Contents/Inhalt
- Contents
- Editorial
- Research on the reactor physics and reactor safety of VVER reactors – AER Symposium 2017
- Technical Contributions/Fachbeiträge
- SIMULATE5-HEX extension for VVER analyses
- Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code
- “Full-Core” VVER-440 extended calculation benchmark
- Calculation of “full core” VVER-1000 benchmark
- Study of neutron-physical characteristics of VVER-1200 considering feedbacks using MCU Monte Carlo code
- Advantages of VVER-440 fuel cycles with new fuel assemblies
- A neutronics feasibility study on utilization of a thinned cladding fuel design at Loviisa NPP
- Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors
- Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries
- Simulation of standard temperature control indications at the outlet of a fuel assembly of VVER1000 reactor of Rostov NPP unit No. 2
- Power transient calculations with VERONA
- Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode
- Renewing the refueling neutron monitoring and reactivity measurement systems at Paks NPP
- Hot channel calculation methodologies in case of VVER-1000/1200 reactors
- Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes
- Simulation of a hypothetical MSLB core transient in VVER-1000 with several stuck rods
