Verification of MCNP6 model of the Jordan Research and Training Reactor (JRTR) for calculations of neutronic parameters

The Jordan Research and Training Reactor (JRTR) is a multi-purpose reactor with power of 5 MW currently under commissioning. The core consists of standard MTR plate type fuel assemblies with low enriched fuel of 19.75% U-235 enrichment. A new computational model has been developed for JRTR using the MCNP6 Monte Carlo code. The purpose of this paper is to validate the MCNP6 model by comparison of the calculation results of important neutronics parameters like keff, flux distribution, kinetics parameters, power peaking factors, and control rod worth to the reference results which were obtained by the McCARD Monte Carlo code and presented in the safety analysis report. Same modeling assumptions were adopted in the MCNP6 model in order to check the computational differences. In addition various up-to-date nuclear data libraries were used in calculations to assess their effect on calculated quantities. The calculation results showed good agreement and the difference in the effective multiplication factor is up 50 pcm.