Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO
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V. Hovi
Abstract
The 7th dynamic AER benchmark is the first in which three-dimensional thermal hydraulics codes are supposed to be applied. The aim is to get a more precise core inlet temperature profile than the sector temperatures available typically with system codes. The benchmark consists of a start-up of the sixth, isolated loop in a VVER-440 plant. The isolated loop initially contains cold water without boric acid and the start-up leads to a somewhat asymmetrical core power increase due to feedbacks in the core. In this study, the 7th AER benchmark is calculated with the three-dimensional nodal reactor dynamics code HEXTRAN-SMABRE coupled with the porous computational fluid dynamics code PORFLO. These three codes are developed at VTT. A novel two-way coupled simulation of the 7th AER benchmark was performed successfully demonstrating the feasibility and advantages of the new reactor analysis framework. The modelling issues for this benchmark are reported and some evaluation against the previously reported comparisons between the system codes is provided.
Kurzfassung
Das 7. AER-Benchmark-Problem ist das erste für die Anwendung von 3D-Thermohydraulikcodes. Ziel ist es, ein präziseres Temperaturprofil am Kerneintritt zu erhalten als die typischen auf Sektoren bezogenen Temperaturen, die mit Systemcodes berechnet werden. Der Benchmark besteht aus der Inbetriebnahme der sechsten, isolierten Schleife in einer WWER-440-Anlage. Die isolierte Schleife enthält zunächst kaltes Wasser ohne Borsäure und die Inbetriebnahme führt zu einer leicht asymmetrischen Leistungserhöhung im Kern durch Rückkopplungseffekte. Für die hier vorgestellte Berechnung des 7. AER-Benchmarks wurden das 3D-Nodal-Kerndynamikprogramm HEXTRAN-SMABRE mit dem porösen 3D-Fluiddynamikcode PORFLO gekoppelt. Diese drei Codes werden bei VTT entwickelt. Diese neuartige Zwei-Wege-Kopplung konnte bei der Berechnung des Benchmarks erfolgreich getestet werden und demonstriert nicht nur Anwendbarkeit sondern auch die damit verbundenen Vorteile. Im Beitrag wird die Modellierung vorgestellt und es werden auch Vergleiche zu bereits veröffentlichen Rechnungen mit verschiedenen Systemcodes präsentiert.
References
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© 2017, 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 2016
- Technical Contributions/Fachbeiträge
- Physical startup tests for VVER-1200 of Novovoronezh NPP: advanced technique and some results
- Experimental study of asymmetric boron dilution at VVER-1000 of Kudankulam NPP and its simulation
- Study on the impact of transition from 3-batch to 4-batch loading at Loviisa NPP on the long-term decay heat and activity inventory
- New engineering safety factors for Loviisa NPP core calculations
- Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440: Preliminary assessment of operating efficiency
- Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors
- Recalculating the steady state conditions of the V-1000 zero-power facility at Kurchatov Institute using Monte Carlo and nodal diffusion codes
- Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO
- Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants
- Advances in HELIOS2 nuclear data library
- ANDREA 2.2 and 2.3 – Advances in modelling of VVER cores
- CFD analyses of the rod bowing effect on the subchannel outlet temperature distribution
- A methodology for the estimation of the radiological consequences of a Loss of Coolant Accident
- Neutron balance as indicator of long-term resource availability in growing nuclear energy system
- Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system
- Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core
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 2016
- Technical Contributions/Fachbeiträge
- Physical startup tests for VVER-1200 of Novovoronezh NPP: advanced technique and some results
- Experimental study of asymmetric boron dilution at VVER-1000 of Kudankulam NPP and its simulation
- Study on the impact of transition from 3-batch to 4-batch loading at Loviisa NPP on the long-term decay heat and activity inventory
- New engineering safety factors for Loviisa NPP core calculations
- Development of fuel cycles with new fuel with 8.9 mm external diameter for VVER-440: Preliminary assessment of operating efficiency
- Investigation of circulating temperature fluctuations of the primary coolant in order to develop an enhanced MTC estimator for VVER-440 reactors
- Recalculating the steady state conditions of the V-1000 zero-power facility at Kurchatov Institute using Monte Carlo and nodal diffusion codes
- Start-up of a cold loop in a VVER-440, the 7th AER benchmark calculation with HEXTRAN-SMABRE-PORFLO
- Verification results of methodology for determining the weighted mean coolant temperature in the primary circuit hot legs of WWER-1000 reactor plants
- Advances in HELIOS2 nuclear data library
- ANDREA 2.2 and 2.3 – Advances in modelling of VVER cores
- CFD analyses of the rod bowing effect on the subchannel outlet temperature distribution
- A methodology for the estimation of the radiological consequences of a Loss of Coolant Accident
- Neutron balance as indicator of long-term resource availability in growing nuclear energy system
- Analysis of changes in the fuel component of the cost of electricity in the transition to a closed fuel cycle in nuclear power system
- Experimental and numerical thermal-hydraulics investigation of a molten salt reactor concept core
