MONITORING SPENT OR REPROCESSED NUCLEAR FUEL USING FAST NEUTRONS

摘要

Attempting to assay the Pu fraction in spent or reprocessed fuel by counting spontaneous fission neutrons immediately runs into the problem that the neutron signal from the spent fuel is dominated by the spontaneous fission neutrons from the ~(242)Cm, ~(244)Cm and ~(240)Pu isotopes. We have found that this problem can be overcome by using fast neutron correlations to measure the number of induced fissions. When the spent fuel is placed inside a polyethylene moderator blanket and lead shield, the number of neutron induced fissions in ~(239)Pu, ~(241)Pu and ~(235)U increases dramatically and this increase can be measured with an array of fast neutron counters. In the case of spent fuel, potential difficulties arise because of the need to shield the detectors from the very high gamma ray flux. However the gamma ray flux can be exponentially attenuated by using a lead shield about 1 mean free path thick for fast neutrons, which would still allow a fast neutron signal sufficient to allow one to determine the total amount of ~(239)Pu, ~(241)Pu and ~(235)U.