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Nuclear Safeguards Education Portal

Advanced NGSI NDA Techniques

In 2009, the Next Generation Safeguards Initiative (NGSI)* commenced a 5 year project to develop better instrumentation for nondestructive assay of spent nuclear fuel (Ref. 18).  The primary goals of the spent fuel NDA program was to improve the ability to determine plutonium quantity and to detect partial defects (missing fuel rods) in spent fuel assemblies.  Fourteen NDA methods were evaluated and in 2012, eight methods were selected for further study as part of 4 integrated technique instruments.  This section reviews the 8 methods selected for further study and presents the combinations for the integrated NDA instruments.  The eight independent NDA methods are

Aside from PG and TN, which represent traditional spent fuel safeguards techniques already covered in this course, each of these methods are described in the following interactive table.

Table 15. Independent Advanced NGSI NDA Techniques


Passive Neutron Albedo Reactivity (PNAR)

Figure 26. PNAR-FC cross-sections. A horizontal slice is shown on the left and a vertical slice is shown on the right. The two diagrams are not on the same scale. (from Ref. 19)

PNAR uses reflected neutrons from a spent nuclear fuel assembly to induce fission and characterize total fissile content (Ref. 19).  The PNAR system surrounds a fuel assembly on all sides with polyethylene to reflect emitted spent nuclear fuel neutrons.  Embedded in the polyethylene are four fission chambers (1 on each side).  PNAR evaluates the ratio of two neutron counts.  This first count is with the system as just described.  The second measurement is recorded with a cadmium layer inserted between the fuel assembly and the polyethylene.  Total fissile material content is characterized as a weighted average of the mass of U-235, Pu-239, and Pu-241.  Figure 26 illustrates horizontal and vertical cross-sectional views of the PNAR system.  PNAR has also been studied with He-3 neutron detectors instead of fission chambers.


Differential Die-Away (DDA)

Figure 27. Schematic cross-sectional view of the DDA instrument (from (Ref. 20)

The DDA method uses the difference in die-away (or decay) time between epithermal (medium energy) and thermal (low energy) neutrons to characterize the amount of fissile material in spent nuclear fuel assemblies (Ref. 20).  DDA uses a neutron generator to induce fissions in the spent fuel.  He-3 tubes in polyethylene moderator detect neutrons.  A cadmium liner around the polyethylene is used for some of He-3 tubes to absorb thermal neutrons and make the Cd-lined He-3 positions sensitive to only fast (high energy) and epi-thermal neutrons.   Without the presence of fissile material in the spent fuel, epithermal and fast neutrons die-away very quickly while thermal neutrons die-away more slowly.  As the amount of fissile material and neutron production from fission increases, the die-away time of epi-thermal neutrons gets closer to that of the thermal neutron population.  The shift in the epi-thermal neutron population die-away time is proportional to that amount of fissile material in the fuel. Figure 27 shows a diagram of the DDA system.


Delayed Neutrons (DN)

The DN system was integrated with the DDA to make use of greater sensitivity of delayed neutron count rate to U-235 concentration (Ref. 21).  Delayed neutrons are emitted by very short lived fission products.  Delayed neutrons emission and detection lacks the coincident neutrons which are simultaneously emitted and detected for prompt fission neutrons.  Thus, delayed neutrons are counted by the detection system as single neutron events minus any background.  U-235 fission emits approximately 2.6 times as many delayed neutrons as Pu-239, so the signal is more dependent on the U-235 concentration.  The DN system works by repeatedly pulsing with a neutron source and counting a spent fuel assembly to produce a steady emission rate of delayed neutrons.  The amount of delayed neutrons emitted is proportional to the amount of fissile material in the spent fuel.  Due to the shared need for a strong neutron source, DN was integrated with DDA from the beginning of design.  The two He-3 tubes without a surrounding Cd liner in DDA system of Figure 25 are used for delayed neutron counting.  Given the greater dependence of the DN signal on U-235, the combined DN and DDA system may be able to resolve individual quantities for U-235 and Pu-239.


Differential Die-Away Self-Interrogation (DDSI)

Ddsi -1
Figure 28. Neutron capture time distributions from spontaneous and induced fission events. (From Ref. 22)

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 (from Ref. 22). 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 28 illustrates this difference.  The fission neutron count rate can then be correlated to the amount of fissile material in the fuel.  Figure 29 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.

Ddsi -2
Figure 29. Cross-sectional view of the DDSI design (From Ref. 22)


Delayed Gamma (DG)

Similar to the delayed neutron measurements, delayed gamma techniques utilize measurements of beta-delayed isotopes that are produced in fission events (from Ref. 23, 24). Delayed gamma rays are emitted from particular fission products following beta-decays, and therefore have distinguishing time scales associated with emission.  The differences in the distribution of fission product yields can be used to quantify and distinguish each fissile element.  The DG system works by interrogating a spent fuel assembly with a neutron source, commonly a Deuterium-Tritium source compatible with most active interrogation techniques, and using a high-purity germanium (HPGe) detector to measure the delayed gamma peaks.  An example of a DG spectra is shown in Figure 30 for a depleted uranium sample (i.e. almost entirely 238U). 

Figure 30. Calculated (bottom) and measured (top) spectra of delayed-gamma from depleted uranium (From Ref. 24)

In order to isolate the relevant gamma peaks from the spectra, previous knowledge is needed of the energy of the delayed gammas specific to the material being investigated.  The isolation of the DG peaks can be a challenge due to interferences with passive gamma radiation from the spent fuel.  Current work is concentration on minimizing interferences, such as by better optimizing the detector setup.  Other efforts are focused on better characterizing the differences in the DG spectra for unique spent fuel compositions.


Cf-252 Interrogation with Prompt Neutrons (CIPN)

Figure 31. (a) Horizontal cross section of CIPN; (b) Vertical cross section of CIPN. (from Ref. 25)

CIPN uses fission chambers in polyethylene in a Fork detector-like configuration with active interrogation from a Cf-252 neutron source to determine the total fissile content of a spent nuclear fuel assembly (from Ref. 25). Figure 31 shows a diagram with the horizontal cross section and vertical cross section of CIPN.  CIPN takes one measurement without the source and another measurement with the Cf-252 source.  The difference between the background and active count rates corresponds to the total fissile material content.  If a baseline measurement is taken prior to a partial defect diversion, CIPN has the capability to detect a partial defect of as little as 3 % of the total spent fuel mass (8 fuel rods from a 17 x 17 assembly).


Self-Interrogation Neutron Resonance Densitometry (SINRD)

Inside Sinrd Pod 2
Figure 32. View of the SINRD unit internals used in experimental measurements at LANL

SINRD relies on solely on the neutron emissions of the spent fuel without any active interrogation (from Ref. 26). SINRD uses four fission chambers.  Three of the fission chambers are covered with different filtering materials that absorb neutrons in different energy ranges.  The neutron energy spectrum from the spent fuel was quantified by subtracting the count rate in different fission chambers.  A ratio of the difference between the count rates of two fission chambers to the count rate another of another fission chamber was used to minimize systematic uncertainties.  The filtering materials of the fission chambers were specifically selected to make the fission chambers sensitive to neutron energy ranges corresponding to resonances in the neutron absorption cross sections of U-235, Pu-239, and Pu-241.  Thus, depressions in the neutron energy ranges quantified by the SINRD detector correspond to the amount of U-235, Pu-239, and Pu-241 in the spent fuel.   This allows SINRD to quantify the amount of those fissile isotopes in the fuel.   


*The NGSI is a program of the U.S. Department of Energy's National Nuclear Security Administration (DOE/NNSA)

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