Absolute total and thermal neutron flux of the U.S. Naval Postgraduate School's AGN-201 reactor was determined by neutron activation of thin gold foils. Foil activities were measured with a gamma-ray scintillation spectrometer, using methods designed to minimize the effect of changes in spectrometer gain. Flux values were calculated for nominal power levels of 0.1 watt and 1, 10, 100, and 750 watts. Methods and results are compared with those of previous investigations. The flux level was found to be a linear function of power within this range; total and thermal average fluxes were determined to be respectively 6.64 x 10 to the 7th power and 5.41 x 10 to the 7th power neutrons per square centimeter per second per watt. (Author).

The purpose of this experiment was to determine the neutron spectrum of the ISU AGN201 reactor using both computer calculations and physical measurements and then to make comparisons between the results provided by the different methods. The regions of interest in the reactor for neutron spectral measurements were the core center region and the reflector region. Computer simulations performed using MCNP5 and DISNEL were both used to obtain neutron spectra. For the experimental portions, several foils were irradiated in the regions of interest. Each foil had a unique response function and could thus provide independent information that could then be used to obtain a neutron spectrum. Since there were fewer foil activation responses than the desired number of energy groups, the problem was under-determined; therefore, unfolding methods had to be utilized. The modified least-squares method and the maximum entropy method were both used. The MAXED software package was used for the maximum entropy method. MAXED indicated spectra with higher thermal neutron contributions and lower fast neutron contributions than the default MCNP spectra indicated for both the center and reflector regions. The modified least-squares method was very consistent with default MCNP spectra and provided significant uncertainty reduction. DISNEL and MCNP agreed very well for the center spectrum and fairly well for the reflector spectrum. This consistency illustrates that MCNP, with the most current set of cross-sections, and spectral measurement methods using the analysis tools described herein, are equally reliable methods of determining neutron spectra in a reactor of the relative simplicity of the AGN reactor.