On comparison of experimental and calculated neutron energy flux spectra at miniature neutron source reactor (MNSR) Masood Iqbal * , Atta Muhammad, Tayyab Mahmood, Naseer Ahmed Nuclear Engineering Division, PINSTECH, P.O. Nilore, Islamabad, Pakistan Received 13 April 2007; received in revised form 22 June 2007; accepted 25 June 2007 Available online 13 August 2007 Abstract 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. Ó 2007 Elsevier Ltd. All rights reserved. 1. Introduction The neutron spatial distribution and energy spectrum are some of the most important characteristics of a reactor for the calculation of radiation damage, neutron dosime- try, fast neutron physics, isotope production and safety analysis. The distribution of neutron energies in a reactor differs from the fission neutron spectrum due to the slowing down of neutrons in elastic and inelastic collisions with fuel, coolant and construction material. For continuous source of fission neutrons present in a reactor, the total spectrum can be considered to contain three different energy distributions of neutrons at given time. The first is fission neutron spectrum, which is likely to be Watt or Maxwell distribution. The second is slowing down spec- trum, which is characterized by a 1/E (1+a) distribution where alpha is the shape parameter (Yucel and Karadag, 2004). The third is a thermal neutron spectrum, which is almost a Maxwellian distribution characterized by the tem- perature of the medium. The true spectrum characteristics of a typical reactor will have more features than these in it (Profio, 1976). For measurement of neutron energy spectra, the method of multiple foil activation combined with neutron spectrum adjustment code is frequently used. Various unfolding codes based in neutron spectrometry are available. Seghour and Seghour (2005), compare their neutron spectrum adjusted by both with MINUIT and STAYNL at JOYO MK-II fast breeder reactor neutron spectrum and men- tioned that the unfolded spectrum obtained by STAYNL is very close to the guess solution. There is a number of studies for the neutron spectrum adjustment in MTR type research reactors like (Iqbal, 1995; Iqbal et al., 1996; Malkawi and Ahmad, 2000). In these studies, WIMSD/4 and CITATION has been used for theoretical calculations and spectrum was adjusted using SANDBP (Szondy and Zsolnay, 1992) and MSI- TER, a modified version of the code STAY’SL (Perry, 1977). The Pakistan Research Reactor-2 is a miniature neutron source rector 30 kW tank in pool research reactor, cooled, 0306-4549/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.06.017 * Corresponding author. Tel.: +92 51 2207264; fax: +92 51 9290275. E-mail address: masiqbal@hotmail.com (M. Iqbal). www.elsevier.com/locate/anucene Available online at www.sciencedirect.com Annals of Nuclear Energy 35 (2008) 209–215 annals of NUCLEAR ENERGY