On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR)

Neutron energy spectrum in Miniature Neutron Source Reactor (MNSR), called Pakistan Research Reactor (PARR-2), is measured employing threshold neutron activation detectors. The calculated neutron spectrum was obtained through modeling the core in detail in three-dimensions employing the transport theory based code WIMS-D/4 and the diffusion theory based code CITATION which was also used as pre-information in the adjustment procedure. A Number of threshold detectors in the form of thin foils are used for spectrum measurements. Gamma activity of irradiated foils was measured with the help of a gamma spectroscopic system consisting of a high efficiency HPGe detector and 8000 channels PC based multi-channel analyzer. STAYNL computer code supplied by International Atomic Energy Agency (IAEA) was used for neutron spectrum adjustment. The group cross-section values and their covariance matrices were derived from the data given in preprocessed cross section libraries in ENDF–6 format of IRDF-90/NMF-G. The comparison between theoretical and experimental work shows good agreement.