A method for accurately calculating the nucleus density of nuclides in burnup calculation

Abstract

The invention provides a method for accurately calculating the nucleus density of nuclides in burnup calculation. The method comprises the steps of: 1, solving a multigroup neutron transport equation according to the nucleus density Nn of various nuclides at the moment of tn to obtain the neutron flux density (Phin, g) of various nuclides and calculating the micro-reaction rate Rn of each nuclide; 2, calculating a corrected predictor step micro-reaction rate [Rn(with an overhead transverse line)(p)] at the moment of tn according to Rn and a corrected corrector step micro-reaction rate [Rn-1(with an overhead transverse line)(c)] of each nuclide at the moment of tn-1; 3, calculating the nucleus density Nn+1 (p) of a predictor step of each nuclide at the moment of tn+1 by solving a burnup equation; (4) again solving the multigroup neutron transport equation to obtain the neutron flux density (Phin+1, g) at the moment of tn+1 and the micro-reaction rate Rn+1 of each nuclide; 5, obtaining a corrected corrector step micro-reaction rate [Rn+1(with an overhead transverse line)(c)] of each nuclide at the moment of tn+1 by using a linear interpolation method; 6, again solving the burnup equation to calculate the nucleus density Nn+1 of each nuclide at the moment of tn+1. The nucleus density of nuclides calculated by using the method is approximate to the nucleus density in an actual state, so that results of burnup calculation can be more accurate.

Description

technical field [0001] The invention relates to the fields of nuclear reactor design and reactor physical calculation, in particular to a method for accurately calculating nuclide atomic nucleus density in burnup calculation. Background technique [0002] During the operation of the reactor, the fissile nuclides in the nuclear fuel are continuously consumed through reactions such as fission or radiation capture, and the convertible nuclides such as 238 U, captured neutrons and converted to fissile nuclides 239 Pu, while some fission products (such as 135 I and 135 Xe) come into being and disappear to reach a balance state, and some are constantly accumulating. In conclusion, the nuclide composition in the fuel will change with the burnup depth. In the calculation of core fuel management, the calculation of the change of nuclear fuel composition with the burnup depth (that is, the calculation of burnup) is a very important content. In reactor design and fuel management c...

AI Technical Summary

Problems solved by technology

However, the traditional calculation method of estimated and corrected burnup considers that the nuclei density at the initial moment of the current burnup step is the average value of the nuclei density obtained from the previous burnup calculation in the estimated and corrected step, while the current burnup step obtained by the estimated and corrected method is There is a certain deviation between the atomic nucleus density at the initial moment and its real atomic nucleus density, so the traditional estimation and correction method still needs to divide the fuel consumption of the combustible poison component with a strong neutron absorption effect into a very fine division. It takes a step size to ensure the accuracy of the calculation, that is, it still takes a long time to calculate

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