Feasibility of Classic Multiplicity Analysis applied to Spent Nuclear Fuel Assemblies

摘要

We address whether it is feasible to perform multiplicity measurements on commercial light water reactor spent nuclear fuel assemblies. A principle concern is the viability of counting statistics and how these propagate through the point-model, used in classic multiplicity analysis (CMA) assuming known efficiency, to estimate spontaneous fissile mass, multiplication and a, the random source neutron to spontaneous fission prompt neutron ratio. We take a bespoke ~3He proportional counter based neutron detector design simulated in the Monte Carlo N-Particle eXtended (MCNPX) radiation transport code as the source of our anticipated experimental count rates. Using semi-empirical relationships we have estimated the experimental uncertainties that would result from an experimental count of 1-hr duration given that the pulse train is not random but rather populated with correlated events and additionally gives rise to intense accidentals. We consider eleven Westinghouse PWR 17×17 pin fuel assemblies ranging in initial ~(235)U enrichment from 2 wt% to 5 wt% expressed relative to the total uranium content with realistic burnups from 15 GWd/tU to 60 GWd/tU, all 5 years cooled. In accord with CMA the predicted rates, along with the uncertainty estimates, are analyzed via the point-model to extract values for the point model parameters, along with associated uncertainties, which can be compared with the known physical values obtained from the fuel library or MCNPX simulation, as appropriate. We find spontaneous fission mass and multiplication to be the main parameters of interest with a, the ratio of random to spontaneous fission neutron production, the least interesting parameter having a far lesser role due to its small value. Efficiency varies to a modest extent across the eleven cases but here is assumed fixed during inversion. For the higher burnups, precision on the mass is unacceptably poor due to poor triples precision. But applying a simple singles and doubles analysis with known efficiency and known a to extract ~(244)Cm_(eff), the effective mass of ~(244)Cm, and multiplication (that is the traditional known-α neutron coincidence counting (NCC) approach) remains robust. This work is part of a larger effort sponsored by the Next Generation Safeguards Initiative to develop an integrated instrument, comprised of individual NDA techniques with complementary features, that is fully capable of determining Pu mass in spent fuel assemblies.