The accuracy and quality of neutron-physical calculations of the active core characteristics depend heavily on the few-group constant preparation procedure. The method, based on using average in the fuel assembly fuel and coolant parameters is currently used for preparing macroscopic cross-sections. The question is what impact would considering the uneven distribution of those parameters, made on the few-group constant preparation stage exert on further analysis of the reactor facility behavior during steady-state and transients operation. The study carries out comparative analysis of the neutron-physical characteristics of the VVER-1000 core using the standard approach and using distributed in the fuel assembly fuel and coolant parameters while preparing few-group constants. It’s revealed that the fuel pellet and coolant radial temperature distributions affect the multiplication factor and temperature reactivity effect values.
Skip Nav Destination
2013 21st International Conference on Nuclear Engineering
July 29–August 2, 2013
Chengdu, China
Conference Sponsors:
- Nuclear Engineering Division
ISBN:
978-0-7918-5583-6
PROCEEDINGS PAPER
The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics
A. A. Mishin,
A. A. Mishin
National Technical University of Ukraine “Kyiv Polytechnic Institute”, Kyiv, Ukraine
Search for other works by this author on:
V. V. Galchenko
V. V. Galchenko
OJSC “KIEP”, Kyiv, Ukraine
Search for other works by this author on:
A. A. Mishin
National Technical University of Ukraine “Kyiv Polytechnic Institute”, Kyiv, Ukraine
V. V. Galchenko
OJSC “KIEP”, Kyiv, Ukraine
Paper No:
ICONE21-16387, V006T16A036; 8 pages
Published Online:
February 7, 2014
Citation
Mishin, AA, & Galchenko, VV. "The Analysis of Influence of Fuel Pellet and Coolant Temperature Distributions on the VVER-1000 Active Core Characteristics." Proceedings of the 2013 21st International Conference on Nuclear Engineering. Volume 6: Beyond Design Basis Events; Student Paper Competition. Chengdu, China. July 29–August 2, 2013. V006T16A036. ASME. https://doi.org/10.1115/ICONE21-16387
Download citation file:
5
Views
Related Proceedings Papers
Related Articles
Development and Processing of Thermal-Hydraulic Model for GFR 2400 Fast Reactor Design and NESTLE Coupled Transient Code
ASME J of Nuclear Rad Sci (October,2022)
Study on the Coupled Neutronic and Thermal-Hydraulic Characteristics of the New Concept Molten Salt Reactor
J. Eng. Gas Turbines Power (October,2010)
Related Chapters
New Generation Reactors
Energy and Power Generation Handbook: Established and Emerging Technologies
Studies Performed
Closed-Cycle Gas Turbines: Operating Experience and Future Potential
Long.Term Reactivity Change and Control: On.Power Refuelling
Fundamentals of CANDU Reactor Physics