NEUTRONIC ANALYSIS OF A LIQUID SODIUM FAST REACTOR EXPERIMENT FOR THE LARGE-I POSITION OF THE ADVANCED TEST REACTOR

摘要

Material testing experiments are needed to enable thenext generation of nuclear power reactors. This paperdiscusses the neutronic analysis of using a modifiedBoosted Fast Flux Loop (BFFL) design. This Fast Loopdesign features both booster fuel plates and a thermalneutron filter.Attila is a 3D multigroup deterministic neutrontransport code. It is used to mesh CAD model geometriesfor use in both the Attila deterministic calculation and theMonte Carlo calculation. MC**2-generates a multi-groupfast reactor cross section library which is converted to aform that is used by Attila for deterministic numericalsolutions. Monte Carlo neutron transport results generatedby MCNP are used as reference solutions for comparison.The Large-I position Fast Loop fast neutron (>0.1MeV) flux magnitude is 2 x 10~(13) n/cm~2-s, which issignificantly less than the desired BFFL experimentcriteria of 3 x 10~(15) n/cm~2-s. The fast to thermal ratio of 8 isalso below the minimum criteria of 15. Further designiteration is necessary to achieve adequate materials testingperformance in the Large-I position. Attila default thermallibrary calculations yielded group-by-group errors of<±50% compared to the MCNP reference solution whileAttila-MC**2 calculations had significantly reducederrors (<±15%).