Analysis of Fuel Temperature Reactivity Coefficient of the PWR Using Wims Code

Abstract

The Fuel Temperature Reactivity Coefficient (FTRC) is an important parameter in design, control, and safety, particularly in PWR reactor. It is then very important to validate any new library for an accurate prediction of this parameter. The objective of this work is to determine the value of the FTRC parameter using the new WIMDS library based on ENDF/BVIII.0 nuclear data files. For this purpose, it is used a set of light water moderated lattice experiments as the PWR-1175 MWe experiment critical reactors, the reactor using UO2 fuel pellet. The analysis is used with WIMSD-5B lattice code with original cross-section libraries and WIMSD-5B with ENDF/B-VIII.0 new cross-section libraries. The results showed that the fuel temperatures reactivity coefficients for the PWR reactor using original libraries is – 3.10 pcm/K with enrichment of 3.1% but for ENDF/B-VlII.0 libraries is – 3.00 pcm/K. Compared to the experimental data of the reactor core, the difference is in the range of 6.9 % for ENDF/B-VIII.0 libraries. It can be concluded that for the reactor, it is better to use ENDF/B-VIII.0 libraries because the original library is not accurate anymore.

Effect of Neutronic Parameters on the Cladding of the AP1000 MOX Fuel by Calculations Using MCNP Code

“…Table 1 presents the design parameters of the AP1000 reactor which is a description of the reactor core. A UO2 fuel assembly is arranged in a 17 × 17 matrix and in the middle is a guide tube [12][13][14]. To control the reactivity, a combustible poison (PYREX) is given and the fuel pellet is coated with a neutron absorber (Integral fuel burnable absorber).…”