SINGLE DETECTOR DEFINITION LEFT BANK ***** c ***** 4 atm =-4.99e-4, atm =-7.488e-4, 7.5 atm =-9.359e-4 g/cc 10 atm-1.248e-3 ***** 940-7.488e-4-901 904-905 u=500 imp:n=1 $ He3 tube at 6.0 atm 901 870-2.699 901-902-906 u=500 imp:n=1 $ Al tube 902 920-9.58e-4 902 u=500 imp:n=1 $ air gap around He-3 tube 903 870-2.699-902 906 u=500 imp:n=1 $ Void above each tube 904 940-7.488e-4-901-904 u=500 imp:n=1 $ Lower dead 905 940-7.488e-4-901 905-906 u=500 imp:n=1 $ Upper dead c c SINGLE DETECTOR DEFINITION LEFT BANK NON-DDSI tube c ***** 4 atm =-4.99e-4, 6 atm =-7.488e-4, 7.5 atm =-9.359e-4 g/cc 10 atm-1.248e-3 ***** 950 940-7.488e-4-911 904-905 u=501 imp:n=1 $ He3 tube at 6.0 atm 951 870-2.699 911-912-906 u=501 imp:n=1 $ Al tube 952 920-9.58e-4 912 u=501 imp:n=1 $ air gap around He-3 tube 953 870-2.699-912 906 u=501 imp:n=1 $ Void above tube 954 940-7.488e-4-911-904 u=501 imp:n=1 $ Lower dead 955 940-7.488e-4-911 905-906 u=501 imp:n=1 $ Upper dead c c DETECTOR ASSEMBLY, LEFT BANK 800 0 900 u=2 fill=500 imp:n=1 $ Initial detector # 60 (at left) 811 0-910 u=2 fill=501 imp:n=1 $ NON-DDSI TUBE c Left bank of He-3 tubes, first row 801 like 800 but trcl=(0-3.5 0) u=2 $ 58 802 like 800 but trcl=(0-7.0 0) u=2 $ 56 803 like 800 but trcl=(0-10.5 0) u=2 $ 54 804 like 800 but trcl=(0-14.0 0) u=2 $ 52 805 like 800 but trcl=(0-17.5 0) u=2 $ 50 806 like 800 but trcl=(0-21.0 0) u=2 $ 48 807 like 811 but trcl=(3 12.25 0) u=2 $ 62 DOWN IN OCTOBER c Left bank of He-3 tubes, second row 808 like 800 but trcl=(-3.0 1.75 0) u=2 $ 61 809 like 811 but trcl=(0 7.00 0) u=2 $ 59 DOWN IN OCTOBER 810 like 800 but trcl=(-3.0-5.25 0) u=2 $ 57 812 like 800 but trcl=(-3.0-12.25 0) u=2 $ 53 813 like 800 but trcl=(-3.0-15.75 0) u=2 $ 51 814 like 800 but trcl=(-3.0-19.25 0) u=2 $ 49

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There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source (252 Cf) for spent fuel interrogations was evaluated.