Upgrade and validation of PHX2MCNP for criticality analysis calculations for spent fuel storage pools

Abstract: A few years ago Westinghouse started the development of a new method for criticality calculations for spent nuclear fuel storage pools called “PHOENIX-to–MCNP” (PHX2MCNP). PHX2MCNP transfers burn-up data from the code PHOENIX to use in MCNP in order to calculate the criticality. This thesis describes a work with the purpose to further validate the new method first by validating the software MCNP5 at higher water temperatures than room temperature and, in a second step, continue the development of the method by adding a new feature to the old script. Finally two studies were made to examine the effect from decay time on criticality and to study the possibility to limit the number of transferred isotopes used in the calculations. MCNP was validated against 31 experiments and a statistical evaluation of the results was done. The evaluation showed no correlation between the water temperature of the pool and the criticality. This proved that MCNP5 can be used in criticality calculations in storage pools at higher water temperature. The new version of the PHX2MCNP script is called PHX2MCNP version 2 and has the capability to distribute the burnable absorber gadolinium into several radial zones in one pin. The decay time study showed that the maximum criticality occurs immediately after the takeout from the reactor as expected. The last study, done to evaluate the possibility to limit the isotopes transferred from PHOENIX to MCNP showed that Case A, a case with the smallest number of isotopes, is conservative for all sections of the fuel element. Case A, which contains only some of the actinides and the strongest absorber of the burnable absorbers gadolinium 155, could therefore be used in future calculations. Finally, the need for further validation of the method is discussed.