In-core fuel management, safety, and thermal hydraulics studies for upgrading TRIGA MARK II research reactor

Bangladesh Atomic Energy Commission has approved a project to upgrade the research reactor to higher flux to meet the growing demand of medical radio-isotopes production and other irradiation facilities. Preliminary studies with the various core parameters showed that it might be possible to create new irradiation flux traps, increase the neutron flux at desired location, and at the same time the fuel burn-up can be made optimal. This will need major reshuffling and reconfiguration of the core with fuel rods initially loaded. The principal objective of this study is focused to make the above improvements in the core without disturbing the safety parameters. This presentation deals with the neutronic and thermal hydraulic analysis of the 3 MW TRIGA MARK II research reactor to upgrade it to a higher flux. To realize this objective, the overall strategy followed is: (I) generation of problem dependent cross section library from basic Evaluated Nuclear Data Files such as ENDF/B-VI, JENDL 3.2 with NJOY94.10+, (ii) use WIMSD-5 package to generate cell constants for all of the materials in the core and its immediate neighborhood, (iii) use CITATION to perform 3-D global analysis of the core to study multiplication factor, neutron flux and power distribution, power peaking factors, temperature reactivity coefficients, etc., (iv) check the validity of the deterministic codes with the Monte Carlo code MCNP-4B2, (v) couple output of CITATION with PARET to study thermal hydraulic behavior to predict safety margins, and (vi) reshuffle the current core configuration to achieve the desired objectives. The computational methods, tools and techniques, customization of cross section libraries, various models for cells and super cells, and a lot of associated utilities have been standardized and established/validated for the overall core analysis. Analyses using the 4-group, and 7-group libraries of macroscopic cross sections generated from the 69-group WIMSD-5 library were performed to study the effect of group structure on neutronics parameters. Various studies showed that a 7-group structure is more suitable for TRIGA calculations considering its LEU fuel composition. The 7-group calculation predicts the experimental value of keff (1.077459) with an accuracy of 0.2% whereas the 4-group calculation yields keff with 0.4% accuracy, in the case of wet CT. So the neutronic analyses were performed using 7-group structure. Table I shows a comparison of the values of the flux distributions obtained from the calculations and experiment at irradiation location in the CT. It may be observed that our calculations overpredict in case of the dry CT and underpredict in the case of wet CT. The maximum error with the experiment was found for wet CT at core midplane with 32% underprediction. (author)