Publications

Radiation transport calculations were performed to compute the angular tallies for scattered gamma-rays as a function of distance, height, and environment. Greens Functions were then used to encapsulate the results a reusable transformation function. The calculations represent the transport of photons throughout scattering surfaces that surround sources and detectors, such as the ground and walls. Utilization of these calculations in GADRAS (Gamma Detector Response and Analysis Software) enables accurate computation of environmental scattering for a variety of environments and source configurations. This capability, which agrees well with numerous experimental benchmark measurements, is now deployed with GADRAS Version 18.2 as the basis for the computation of scattered radiation.

The Gamma Detector Response and Analysis Software-Detector Response Function (GADRAS-DRF) application computes the response of gamma-ray detectors to incoming radiation. This manual provides step-by-step procedures to acquaint new users with the use of the application. The capabilities include characterization of detector response parameters, plotting and viewing measured and computed spectra, and analyzing spectra to identify isotopes or to estimate flux profiles. GADRAS-DRF can compute and provide detector responses quickly and accurately, giving researchers and other users the ability to obtain usable results in a timely manner (a matter of seconds or minutes).

T to compute the radiography properties of various materials, the flux profiles of X-ray sources must be characterized. This report describes the characterization of X-ray beam profiles from a Kimtron industrial 450 kVp radiography system with a Comet MXC-45 HP/11 bipolar oil-cooled X-ray tube. The empirical method described here uses a detector response function to derive photon flux profiles based on data collected with a small cadmium telluride detector. The flux profiles are then reduced to a simple parametric form that enables computation of beam profiles for arbitrary accelerator energies.

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The Gamma Detector Response and Analysis Software (GADRAS) software package is capable of simulating the radiation transport physics for one-dimensional models. Spherical shells are naturally one-dimensional, and have been the focus of development and benchmarking. However, some objects are not spherical in shape, such as cylinders and boxes. These are not one-dimensional. Simulating the radiation transport in two or three dimensions is unattractive because of the extra computation time required. To maintain computational efficiency, higher-dimensional geometries require approximations to simulate them in one-dimension. This report summarizes the theory behind these approximations, tests the theory against other simulations, and compares the results to experimental data. Based on the results, it is recommended that GADRAS users always attempt to approximate reality using spherical shells. However, if fissile material is present, it is imperative that the shape of the one-dimensional model matches the fissile material, including the use of slab and cylinder geometry.

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A common source of neutrons for calibration and testing is alpha-neutron material, named for the alpha-neutron nuclear reaction that occurs within. This material contains a long-lived alpha-emitter and a lighter target element. When the alpha particle from the emitter is absorbed by the target, neutrons and gamma rays are released. Gamma Detector Response and Analysis Software (GADRAS) includes built-in alpha-neutron source definitions for AcC, AmB, AmBe, AmF, AmLi, CmC, and PuC. In addition, GADRAS users may create their own alpha-neutron sources by placing valid alpha-emitters and target elements in materials within their one-dimensional models (1DModel). GADRAS has the ability to use pre-built alpha-neutron sources for plotting or as trace-sources in 1D models. In addition, if any material (existing or user-defined) specified in a 1D model contains both an alpha emitter in conjunction with a target nuclide, or there is an interface between such materials, then the appropriate neutron-emission rate from the alpha-neutron reaction will be computed. The gamma-emissions from these sources are also computed, but are limited to a subset of nine target nuclides. If a user has experimental data to contribute to the alpha-neutron gamma emission database, it may be added directly or submitted to the GADRAS developers for inclusion. The gadras.exe.config file will be replaced when GADRAS updates are installed, so sending the information to the GADRAS developers is the preferred method for updating the database. This is also preferable because it enables other users to benefit from your efforts.

This document is a final report for the polyvinyl toluene (PVT) neutron-gamma (PVT-NG) project, which was sponsored by the Domestic Nuclear Detection Office (DNDO). The PVT-NG sensor uses PVT detectors for both gamma and neutron detection. The sensor exhibits excellent spectral resolution and gain stabilization, which are features that are beneficial for detection of both gamma-ray and neutron sources. In fact, the ability to perform isotope identification based on spectra that were measured by the PVT-NG sensor was demonstrated. As described in a previous report, the neutron sensitivity of the first version of the prototype was about 25% less than the DNDO requirement of 2.5 cps/ng for bare Cf-252. This document describes design modifications that were expected to improve the neutron sensitivity by about 50% relative to the PVT-NG prototype. However, the project was terminated before execution of the design modifications after portal vendors demonstrated other technologies that enable neutron detection without the use of He-3. Nevertheless, the PVT-NG sensor development demonstrated several performance goals that may be useful in future portal designs.

Radiation portals normally incorporate a dedicated neutron counter and a gamma-ray detector with at least some spectroscopic capability. This paper describes the design and presents characterization data for a detection system called PVT-NG, which uses large polyvinyl toluene (PVT) detectors to monitor both types of radiation. The detector material is surrounded by polyvinyl chloride (PVC), which emits high-energy gamma rays following neutron capture reactions. Assessments based on high-energy gamma rays are well suited for the detection of neutron sources, particularly in border security applications, because few isotopes in the normal stream of commerce have significant gamma ray yields above 3 MeV. Therefore, an increased count rate for high-energy gamma rays is a strong indicator for the presence of a neutron source. The sensitivity of the PVT-NG sensor to bare {sup 252}Cf is 1.9 counts per second per nanogram (cps/ng) and the sensitivity for {sup 252}Cf surrounded by 2.5 cm of polyethylene is 2.3 cps/ng. The PVT-NG sensor is a proof-of-principal sensor that was not fully optimized. The neutron detector sensitivity could be improved, for instance, by using additional moderator. The PVT-NG detectors and associated electronics are designed to provide improved resolution, gain stability, and performance at high-count rates relative to PVT detectors in typical radiation portals. As well as addressing the needs for neutron detection, these characteristics are also desirable for analysis of the gamma-ray spectra. Accurate isotope identification results were obtained despite the common impression that the absence of photopeaks makes data collected by PVT detectors unsuitable for spectroscopic analysis. The PVT detectors in the PVT-NG unit are used for both gamma-ray and neutron detection, so the sensitive volume exceeds the volume of the detection elements in portals that use dedicated components to detect each type of radiation.

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Variance estimates that are used in the analysis of radiation measurements must represent all of the measurement and computational uncertainties in order to obtain accurate parameter and uncertainty estimates. This report describes an approach for estimating components of the variance associated with both statistical and computational uncertainties. A multi-sensor fusion method is presented that renders parameter estimates for one-dimensional source models based on input from different types of sensors. Data obtained with multiple types of sensors improve the accuracy of the parameter estimates, and inconsistencies in measurements are also reflected in the uncertainties for the estimated parameter. Specific analysis examples are presented that incorporate a single gross neutron measurement with gamma-ray spectra that contain thousands of channels. The parameter estimation approach is tolerant of computational errors associated with detector response functions and source model approximations.

Sandia National Laboratories has developed an inverse radiation transport solver that applies nonlinear regression to coupled neutron-photon deterministic transport models. The inverse solver uses nonlinear regression to fit a radiation transport model to gamma spectrometry and neutron multiplicity counting measurements. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5 kg sphere of {alpha}-phase, weapons-grade plutonium. The source was measured bare and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses between 1.27 and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to evaluate the solver's ability to correctly infer the configuration of the source from its measured radiation signatures.

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Forward radiation transport is the problem of calculating the radiation field given a description of the radiation source and transport medium. In contrast, inverse transport is the problem of inferring the configuration of the radiation source and transport medium from measurements of the radiation field. As such, the identification and characterization of special nuclear materials (SNM) is a problem of inverse radiation transport, and numerous techniques to solve this problem have been previously developed. The authors have developed a solver based on nonlinear regression applied to deterministic coupled neutron-photon transport calculations. The subject of this paper is the experimental validation of that solver. This paper describes a series of experiments conducted with a 4.5-kg sphere of alpha-phase, weapons-grade plutonium. The source was measured in six different configurations: bare, and reflected by high-density polyethylene (HDPE) spherical shells with total thicknesses of 1.27, 2.54, 3.81, 7.62, and 15.24 cm. Neutron and photon emissions from the source were measured using three instruments: a gross neutron counter, a portable neutron multiplicity counter, and a high-resolution gamma spectrometer. These measurements were used as input to the inverse radiation transport solver to characterize the solver's ability to correctly infer the configuration of the source from its measured signatures.

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The performance of the Gamma Detector Response and Analysis Software (GADRAS) was validated by comparing GADRAS model results to experimental measurements for a series of benchmark sources. Sources for the benchmark include a plutonium metal sphere, bare and shielded in polyethylene, plutonium oxide in cans, a highly enriched uranium sphere, bare and shielded in polyethylene, a depleted uranium shell and spheres, and a natural uranium sphere. The benchmark experimental data were previously acquired and consist of careful collection of background and calibration source spectra along with the source spectra. The calibration data were fit with GADRAS to determine response functions for the detector in each experiment. A one-dimensional model (pie chart) was constructed for each source based on the dimensions of the benchmark source. The GADRAS code made a forward calculation from each model to predict the radiation spectrum for the detector used in the benchmark experiment. The comparisons between the GADRAS calculation and the experimental measurements are excellent, validating that GADRAS can correctly predict the radiation spectra for these well-defined benchmark sources.

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