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Pre-detonation technical nuclear forensics techniques for research reactor spent fuel were developed in a collaborative project with Savannah River National Lab ratory. An inverse analysis method was employed to reconstruct reactor parameters from a spent fuel sample using results from a radiochemical analysis. In the inverse analysis, a reactor physics code is used as a forward model. Verification and validation of different reactor physics codes was performed for usage in the inverse analysis. The verification and validation process consisted of two parts. The first is a variance analysis of Monte Carlo reactor physics burnup simulation results. The codes used in this work are MONTEBURNS and MCNPX/CINDER. Both utilize Monte Carlo transport calculations for reaction rate and flux results. Neither code has a variance analysis that will propagate through depletion steps, so a method to quantify and understand the variance propagation through these depletion calculations was developed. The second verification and validation process consisted of comparing reactor physics code output isotopic compositions to radiochemical analysis results. A sample from an Oak Ridge Research Reactor spent fuel assembly was acquired through a drilling process. This sample was then dissolved in nitric acid and diluted in three different quantities, creating three separate samples. A radiochemical analysis was completed and the results were compared to simulation outputs at different levels ofdetail. After establishing a forward model, an inverse analysis was developed to re-construct the burnup, initial uranium isotopic compositions, and cooling time of a research reactor spent fuel sample. A convergence acceleration technique was used that consisted of an analytical calculation to predict burnup, initial 235U, and 236U enrichments. The analytic calculation results may also be used stand alone or in a database search algorithm. In this work, a reactor physics code is used as a for- ward model with the analytic results as initial conditions in a numerical optimization algorithm. In the numerical analysis, the burnup and initial uranium isotopic com- positions are reconstructed until the iterative spent fuel characteristics converge with the measured data. Upon convergence of the sample’s burnup and initial uranium isotopic composition, the cooling time can be reconstructed. To reconstruct cooling time, the standard decay equation is inverted and solved for time. Two methods were developed. One method uses the converged burnup and initial uranium isotopic compositions along in a reactor depletion simulation. The second method uses an isotopic signature that does not decay out of its mass bin and has a simple production chain. An example would be 137Cs which decays into the stable 137Ba. Similar results are achieved with both methods, but extended shutdown time or time away from power results in over prediction of the cooling time. The over prediction of cooling time and comparison of different burnup reconstruction isotope results are indicator signatures of extended shutdown or time away from power. Due to dynamic operation in time and function, detailed power history reconstruction for research reactors is very challenging. Frequent variations in power, repeated variable shutdown time length, and experimentation history affect the spectrum an individual assembly is burned with such that full reactor parameter reconstruction is difficult. The results from this technical nuclear forensic analysis may be used with law enforcement, intelligence data, macroscopic and microscopic sample characteristics in a process called attribution to suggest or exclude possible sources of origin for a sample.

A high-efficiency gamma-ray counting system has been built at Los Alamos National Laboratory for use in analyzing nuclear forensics samples. This system consists of two clover high-purity germanium detectors and is surrounded by a thallium-doped sodium iodide annulus. Special precautions have been taken to ensure the system has a low background. The system is connected to XIA Pixie-4 fast digitizers and collects data in list-mode. This work is split into two main parts. The first part describes the proper steps and techniques to initialize the settings of a detector system connected to fast digitizers in order to optimize the system for resolution and throughput. The various counting modes for this particular system are described in detail, including the benefits and drawbacks of each mode. Steps are then shown to characterize the system by obtaining efficiency curves for various counting modes and sample geometries. Because of the close counting geometry involved with this system, true-coincidence summing factors must be calculated, and are done so in part by measuring the peak-to-total ratios of the system in its various counting modes across a wide energy range. The dead-time for the system can be complicated due to the multiple inputs of the system. Techniques for calculating the dead-time of multiple-detector systems are discussed. The second part of this work shows the system's usefulness in analyzing nuclear forensics samples, specifically irradiated enriched uranium. Three fission product parent-daughter pairs of different lifetimes are analyzed over a course of six months. The activities of each nuclide are calculated at each time step. Age dating techniques using the parent-daughter pairs are discussed, as well as the detection limits of each nuclide for a range of sample ages. Finally, avenues for further research are presented, as well as potential sources of error or uncertainty for this work. / text

The work presented here develops a gamma-ray spectral construction and analysis software tool that was used to analyze multi-detector data collected using a digital spectrometer with list mode capabilities. The tool was used to parse the output from three detectors and generate new spectra that the user chooses from post-processing suppression routines, such as simulated anticoincidence and coincidence spectra. Part of this research was also to characterize the improvements in the detection limits and the various detector efficiencies from this method as opposed to creating these spectra using traditional electronic gating systems. A focus is placed on the detection capability improvements for nuclear forensics purposes, particularly the identification and quantification of fission product samples, and structuring the code framework for handling these types of time-dependent samples while increasing the versatility of the detector system. Improvements to the minimum detectable activity for a series of fission products was accomplished through post-processing suppression methods and multi-dimensional spectral data structures are now achievable. / text

It is important to have a method of experimentally calculating fission product yields. Statistical calculations and simulations produce very large uncertainties. Experimental calculations, depending on the methods used, tend to produce lower uncertainties. This work set up a method to calculate fission product yields using gamma ray spectroscopy. In order to produce a method that was theoretically sound, a simulation was set up using OrigenArp to calculate theoretical concentrations of fission products from the irradiation of natural uranium. From these concentrations, the fission product yields were calculated to verify that they would agree with expected values. Moving forward in the work, the total flux at the point of irradiation, in the pneumatic transfer system, was calculated and determined to be 3.9070E+11 ± 6.9570E+10 n/cm^2/s at 100 kW. Once the flux was calculated, the method for calculating fission product yields was implemented and yields were calculated for 10 fission products. The yields calculated were in very good agreement (within 10.04%) with expected values taken from the ENDF-349 library. This method has strong potential in nuclear forensics as it can provide a means for developing a library of experimentally-determined fission product yields, as well as rapid post-nuclear detonation analysis. / text

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