engleski

Influence of Thermal-Hydraulic Model to Fuel Management Core Calculation

The integration of neutronic, fuel rod and thermal-hydraulic calculations for both, steady-state core design type of the calculation and for transient and safety analyses, is used to improve the response of Nuclear Power Plants (NPP) both from the point of view of safe and economic plant operation. That process assumes improvement of current calculational tools and application of results acquired from operational experience. The objective of this paper is to explore influence of improved thermal-hydraulics core model to overall in-core fuel management parameters (reactivity, power distribution, burn-up distribution) and to take it into account in systematic way. New core thermal-hydraulics model based on codes COBRA III C and COBRA-EN was included within the PARCS depletion loop to calculate the behavior of representative fuel rods for each assembly. Modified code is called COBRA-VIP and exists as both standalone version and part of PARCS code. Some programming changes were necessary to make possible dual use of COBRA-VIP. Core fuel management calculation was performed for NPP Krsko cycle 23 to show influence of performed change to selected core parameters. The benefit of the described approach is that in addition to normal depletion calculation, the behavior of any fuel rod in each fuel assembly can be studied from thermal-hydraulics point of view. The average fuel rod per assembly can be used to improve TH feedback calculations and the limiting fuel rod per assembly can be used to perform DNBR or fuel centre line temperature calculation.

burn-up; fuel properties; in-core fuel management

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