Figure 26. Neutron capture time distributions from spontaneous and induced fission events. 

The DDSI system exploits a very slight difference in the arrival time of neutrons from spontaneous fission in spent fuel Curium and neutrons from induced fission in spent fuel fissile material. The primary source of neutrons from spent fuel is from spontaneous fission (SF) in Cm-244.  The SF neutrons arrive at system He-3 neutron detectors, experience one thermalization (slow to thermal neutron speeds), and are absorbed in the detector.  Induced fission neutrons from spent fuel fissile material absorbed in the system detectors are the result of two neutron thermalizations.  First an SF neutron must be thermalized and absorbed in spent fuel fissile material to produce a fissile material fission neutron.  Then the fissile material fission neutron must be thermalized and absorbed in the detector.   By setting an early gate and a late gate for neutron detector counts DDSI can distinguish between SF neutrons, which dominate the early gate, and self-interrogation fission neutron, which dominate the late gate.  Figure 26 illustrates this difference.  The fission neutron count rate can then be correlated to the amount of fissile material in the fuel.  Figure 27 shows a cross-sectional view of the DDSI with He-3 tubes in a fork-like block of high density polyethylene (HDPE).  The HDPE is lined with cadmium, and the whole apparatus is surrounded by lead to shield the detectors from the intense gamma radiation of the spent fuel.

Figure 27. Cross-sectional view of the DDSI design

Source: A. Belian, H.O. Menlove, M.T. Swinhoe, and S.J. Tobin, "New Design of the Differential Die-away Self-interrogation Instrument for Spent Fuel Assay," Journal of Nuclear Materials Management   40:3 (2012).