Mark W. Shaver

Coupling Deterministic and Monte Carlo Transport Methods for the Simulation of Gamma-Ray Spectroscopy Scenarios

Simulation is often used to predict the response of gamma-ray spectrometers in technology viability and comparative studies for homeland and national security scenarios. Candidate radiation transport methods generally fall into one of two broad categories: stochastic (Monte Carlo) and deterministic. Monte Carlo methods are the most heavily used in the detection community and are particularly effective for calculating pulse-height spectra in instruments. However, computational times for scattering- and attenuation-dominated problems can be extremely long - many hours or more on a typical desktop computer. Deterministic codes that discretize the transport in space, angle, and energy offer potential advantages in computational efficiency for these same kinds of problems, but pulse-height calculations are not readily accessible. This paper investigates a method for coupling angular flux data produced by a three-dimensional deterministic code to a Monte Carlo model of a gamma-ray spectrometer. Techniques used to mitigate ray effects, a potential source of inaccuracy in deterministic field calculations, are described. Strengths and limitations of the coupled methods, as compared to purely Monte Carlo simulations, are highlighted using example gamma-ray detection problems and two metrics: (1) accuracy when compared to empirical data and (2) computational time on a typical desktop computer.

Signatures and Methods for the Automated Nondestructive Assay of

International Atomic Energy Agency (IAEA) inspectors currently perform periodic inspections at uranium enrichment plants to verify UF6 cylinder enrichment declarations. Measurements are typically performed with handheld high-resolution sensors on a sampling of cylinders taken to be representative of the facilitys entire cylinder inventory. These measurements are time-consuming, expensive, and assay only a small fraction of the total cylinder volume. An automated nondestructive assay system capable of providing enrichment measurements over the full volume of the cylinder could improve upon current verification practices in terms of manpower and assay accuracy. The 185-keV emission from U-235 is utilized in todays cylinder measurements, but augmenting this “traditional” signature with more-penetrating “non-traditional” signatures could help to achieve full-volume assay in an automated system. This paper describes the study of non-traditional signatures that include neutrons produced by F-19 (α, n) reactions (spawned primarily from U-234 alpha emission) and the high-energy gamma rays (extending up to 8 MeV) induced by those neutrons when they interact in the cylinder wall and nearby materials. The potential of these non-traditional signatures and assay methods for automated cylinder verification is explored using field measurements on a small population of cylinders ranging from 2.0% to 5% in U-235 enrichment. The standard deviation of the non-traditional high-energy gamma-ray assay approach was 4.7% relative to the declared cylinder enrichments; the standard deviation of the traditional enrichment meter approach using a well-collimated high-resolution spectrometer was 4.3%. The prospect of using the non-traditional high-energy gamma-ray signature in concert with the traditional 185-keV signature to reduce the uncertainty of automated cylinder assay is discussed.

{\rm UF}_{6}

Nondestructive techniques for measuring the mass of fissile isotopes in spent nuclear fuel is a considerable challenge in the safeguarding of nuclear fuel cycles. A nondestructive assay technology that could provide direct measurement of fissile mass, particularly for the plutonium (Pu) isotopes, and improve upon the uncertainty of todays confirmatory methods is needed. Lead slowing-down spectroscopy (LSDS) has been studied for the spent fuel application previously, but the nonlinear effects of assembly self shielding (of the interrogating neutron population) have led to discouraging assay accuracy for realistic pressurized water reactor fuels. In this paper, we describe the development of time-spectral analysis algorithms for LSDS intended to overcome these self-shielding effects. The algorithm incorporates the tabulated energy-dependent cross sections from key fissile and absorbing isotopes, but leaves their mass as free variables. Multi-parameter regression analysis is then used to directly calculate not only the mass of fissile isotopes in the fuel assembly (e.g., Pu-239, U-235, and Pu-241), but also the mass of key absorbing isotopes such as Pu-240 and U-238. Modeling-based assay results using this self-shielding relationship indicate that LSDS has the potential to directly measure fissile isotopes with less than 5% average relative error for pressurized water reactor assemblies with burnup as high as 60 GWd/MTU. Shortcomings in the initial self-shielding model and potential improvements to the formulation are described.

Cylinders at Uranium Enrichment Plants

Abstract Monte Carlo methods are typically used for simulating radiation fields around gamma-ray spectrometers and pulse-height tallies within those spectrometers. Deterministic codes that discretize the linear Boltzmann transport equation can offer significant advantages in computational efficiency for calculating radiation fields, but stochastic codes remain the most dependable tools for calculating the response within spectrometers. For a deterministic field solution to become useful to radiation detection analysts, it must be coupled to a method for calculating spectrometer response functions. This coupling is done in the RADSAT toolbox. Previous work has been successful using a Monte Carlo boundary sphere around a handheld detector. It is desirable to extend this coupling to larger detector systems such as the portal monitors now being used to screen vehicles crossing borders. Challenges to providing an accurate Monte Carlo boundary condition from the deterministic field solution include the greater possibility of large radiation gradients along the detector and the detector itself perturbing the field solution, unlike smaller detector systems. The method of coupling the deterministic results to a stochastic code for large detector systems can be described as spatially defined rectangular patches that minimize gradients. The coupled method was compared to purely stochastic simulation data of identical problems, showing the methods produce consistent detector responses while the purely stochastic run times are substantially longer in some cases, such as highly shielded geometries. For certain cases, this method has the ability to faithfully emulate large sensors in a more reasonable amount of time than other methods.

Time-Spectral Analysis Methods for Spent Fuel Assay Using Lead Slowing-Down Spectroscopy

This report presents 9 test problems to guide testing and development of hybrid calculations for the ADVANTG code at ORNL. These test cases can be used for comparing different types of radiation transport calculations, as well as for guiding the development of variance reduction methods. Cases are drawn primarily from existing or previous calculations with a preference for cases which include experimental data, or otherwise have results with a high level of confidence, are non-sensitive, and represent problem sets of interest to NA-22.