Influence of the uncertainty of the microscopic cross sections of 238U and plutonium isotopes on the neutron multiplication coefficient in VVÉR

Abstract

The influence of the uncertainties in the microscopic cross sections of 238 U and 239-241 Pu on the neutron multiplication coefficient in VVÉR-1000 is studied. Data on the uncertainty of the cross sections were obtained by analyzing the ENDF/B files of the JENDL 3.3 system using the ERRORJ program. The characteristics of the reactor are calculated using the TRIFON 2.1 and SHERHAN programs. An approach associated with the calculation of the coefficients of sensitivity of the fuel assemblies and the reactor to standard samples in a small-group approximation is used.The purpose of this work is to calculate the uncertainty of the neutron multiplication coefficient in VVÉR-1000 due to the uncertainty in the microscopic cross sections of four isotopes -239 Pu, 240 Pu, 241 Pu, and 238 U. The three-dimensional model of the reactor contained 109 fuel assemblies with uranium and 54 fuel assemblies with mixed fuel, corresponding to the midpoint of the yearly fuel cycle, and uniform properties along the height. The 353 cm high core is surrounded by the elements of the compartment.The fuel assemblies were calculated using the TRIFON 2.1 program [1] in a multigroup approximation taking account of the fine structure of the neutron flux in the region of strong resonances (230 superthermal and 25 thermal groups), using a library of microscopic cross sections which is formed by the ASMS system from ENDF/B files. It has the BNAB-26 structure and includes data on the parameters of resonance levels and the dependence of the cross sections on the energy of neutrons in the thermal range. The results of the calculations were converted into data of the small-group representation with partitioning according to neutron energy 1.05·10 7 -4.65·10 3 -100-0.465-0 eV (for groups with boundaries corresponding to the lower limits of the 12th, 17th, 24th, and 26th groups of the BNAB-26 system). The calculations of the reactor were performed using the SHERHAN program [3] with four-group characteristics of the effective boundary conditions (Λ matrices), obtained from a calculation of the fuel assemblies using the TRIFON 2.1 program.Computational Method. In the base calculation of the fuel assemblies, the integrals T i m of the product of the neutron flux by radiative capture cross section of a "standard sample" σ s = 1 b and its concentration c s = 0.001·10 24 cm −3 were calculated for each type of physical zone m in each small-group energy interval i over the corresponding volumes V m of the physical zones and their relative values t i m : (1) T c F u r dUdV T T t T T i m s s U V i i m m i m i m i i m = = = ∫ ∫ ∑ σ ( , ) ; ; / .