Criticality and safety parameter studies of a 3-MW TRIGA MARK-II research reactor and validation of the generated cross-section library and computational method

This study deals with the analysis of some neutronics and safety parameters of the current core of a 3-MW TRIGA MARK-II research reactor and validatio n of the generated macroscopic cross-section library and calculational tech niques by benchmarking with experimental, operational, and available Safety Analysis Report (SAR) values. The overall strategy is: (a) generation of t he problem-dependent cross-section library from basic Evaluated Nuclear Dat a Files such as ENDF/B-VI and JENDL-3.2 with NJOY94.10+, (b) use of the WIM SD-5 package to generate a few-group neutron macroscopic cross section for all of the materials in the core and its immediate neighborhood, (c) use th e three-dimensional CITATION code to perform the global analysis of the cor e, and (d) checking of the validity of the CITATION diffusion code with the MCNP4B2 Monte Carlo code. The ultimate objective is to establish methods f or reshuffling the current core configuration to upgrade the thermal flux a t irradiation locations for increased isotope production. The computational methods, tools and techniques, customization of cross-section libraries, v arious models for cells and supercells, and many associated utilities are s tandardized and established/validated for the overall neutronic analysis. T he excess reactivity, neutron flux, power distribution, power peaking facto rs, determination of the hot spot, and fuel temperature reactivity coeffici ents alpha(f) in the temperature range of 45 to 1000 degrees C are studied. All the analyses are performed using the 4- and 7-group libraries of the m acroscopic cross sections generated from the 69-group WIMSD-5 library. The 7-group calculations yield comparatively better agreement with the experime ntal value of k(eff) and the other core parameters. The CITATION test runs using different cross-section sets based on the different models applied in the WIMSD-5 calculations show a strong influence of those models on the fi nal integral parameter. Some of the cells are specially treated with the Pr ize options available in WIMSD-5 to take into account the fine structure of the flux gradient in the fuel-reflector interface region. The hot spot is found physically at the fuel position C4 with a maximum power density of 1. 044559 X 10(2) W/cm(3). The calculated total peaking factor is 5.8867 compa red to the original SAR value of 5.6325. The curve of alpha(f) with the tem perature at zero burnup shows that the curve deviates somewhat with that re ported in the original SAR for low-enriched uranium fuel. The MCNP calculat ions establish that the CITATION calculations and the generated cross-secti on library are reasonably good for neutronic analysis of TRIGA reactors. Th e results obtained from the neutronic analysis will be used to analyze the thermal-hydraulic behavior and the safety margins of the core both for stea dy-state and pulse-mode operations.