Comparative thermohydraulic analyses of VVER 1000 active core for two different construction types of assemblies
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Y. Perin
Abstract
The OECD/NEA Benchmark on NPP Kalinin Unit 3 “Switching-off of one of the four operating main circulation pumps at nominal power” based on measurements is at its end phase. A new benchmark is starting based on measurements from the NPP Rostov Unit 2 “Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core”. The Rostov-2 core is loaded with assemblies of a new modern type “TVS-2M” which differs in construction in comparison with the “TVSA”-type implemented in the Kalinin-3 core. The goal of the performed study is to determine the thermohydraulic differences, during steady-state operation, between the two cores based on loads with different assembly types. Results of steady-state simulations are compared for Kalinin-3 and Rostov-2 cores, taken from the respective Benchmarks specifications. To ensure that the observed differences are coming only from the fuel assembly construction, the same axial power density is used for all cases. Calculations are done with an under-development version of the GRS system code ATHLET, which has sub-channel modeling capabilities. Thus, in addition to the typical assembly-wise models, calculations with seven assemblies modeled at the sub-channel (pin-by-pin) level were also performed and analyzed. This work will help the Benchmarks' Rostov-2 participants to set up more accurate core thermohydraulic models.
Kurzfassung
Der OECD/NEA-Benchmark „Switching-off of one of the four operating main circulation pumps at nominal power“, der auf experimentellen Daten des KKWs Kalinin Block 3 basiert, befindet sich in der Endphase der Auswertung. Ein neuer Benchmark „Reactivity compensation with diluted boron by stepwise insertion of control rod cluster into the VVER-1000 core“, der auf experimentellen Daten aus dem KKW Rostov Block 2 basiert, startet derzeit. Der Rostov-2-Kern wird mit Brennelementen des neuen modernen Typs „TVS-2M“ beladen. Dieser Typ unterscheidet sich in der Konstruktion vom „TVSA“-Typ im Kalinin-3-Kern. Das Ziel der durchgeführten Studie ist, die thermohydraulischen Unterschiede zwischen den Kernen im stationären Zustand aufgrund der Implementierung des neuen Brennelementtyps zu ermitteln. Die Ergebnisse von stationären Rechnungen werden für die beiden Kerne (Kalinin-3 und Rostov-2) verglichen. Um sicherzustellen, dass die beobachteten Unterschiede nur von der Brennelementkonstruktion stammen, wird für beide Fälle die gleiche axiale Leistungsverteilung verwendet. Berechnungen werden mit einer Version des GRS-Systemcodes ATHLET, die Entwicklungen zur Unterkanalmodellierung enthält, durchgeführt. Deshalb wurden zusätzlich zu den üblichen brennelementweisen Modellen auch Berechnungen mit sieben Brennelemente, die auf Unterkanalebene (brennstabweise) modelliert werden, durchgeführt. Diese Arbeit soll den Teilnehmern des Rostov-2 Benchmarks helfen, genauere thermohydraulische Kernmodelle zu erstellen.
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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
