Jeffrey King

A New Modeling Technique to Analyze Safeguards Measurements in Large Systems

Abstract SafeGuards Analysis (SGA) is a computational toolbox able to simulate different safeguards scenarios across a number of different fuel cycles and at many different scales within the MATLAB Simulink framework. SGA functions by simulating Material Balance Areas (MBAs) under safeguards materials control and accountability and allows the user to define the uncertainty parameters of the associated flow and inventory measurements. The simulated safeguard system uses the uncertain measurement estimates to calculate a mass-balance across the MBA. This mass balance is then evaluated by one of a number of different statistical tests to determine if a significant amount of material has been removed from the MBA. This paper describes the design of SGA, the results of testing each element of the toolbox, and a number of single MBA example scenarios. In all of the test cases, SGA performed as expected and produced acceptable results from the single MBA scenario.

Design, Construction, and Demonstration of the Colorado School of Mines Neutron Imaging Facility

Abstract The MInes NEutron Radiography facility (MINER facility) installed at the United States Geological Survey TRIGA Reactor provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a microchannel plate imaging detector, and the associated electronics. Radiographs of a sensitivity indicator–a resolution indicator developed by the American Society for Testing and Materials–taken using both the digital detector and the transfer method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a comparison between digital and film radiography at the new facility. The calibration pins contain simulated defects of known dimensions, including pellet-clad gaps, gaps between pellets, and central voids within the pellets. Comparison of the radiographs taken by the two methods reveals that the digital detector does not produce high-quality images when compared to film radiography. Additionally, there are a number of artifacts in the digital images produced by the image acquisition system. The quality of the film images demonstrates that the problems with the digital images are a product of the digital imaging system and not the neutron beam.

Installation of a New Neutron Beam Facility at the U.S. Geological Survey TRIGA Reactor

The fleet of research and training reactors is aging, and no new research reactors are planned in the United States; thus, there is a need to expand the capabilities of existing reactors to meet users’ needs. To address these needs, the Colorado School of Mines added a neutron beamline facility to the U.S. Geological Survey TRIGA Reactor (GSTR), a 1-MW(thermal) Mark-I TRIGA reactor located at the Denver Federal Center in Lakewood, Colorado. The original GSTR design did not include any beam ports, and future research efforts will benefit from a neutron beam at the GSTR. Adding new beamline facilities to existing research reactors is both rare and challenging, and this paper describes the design and installation of a new neutron beamline facility at a Mark-I TRIGA reactor with no existing beamline facilities. The design and construction of a radiation beamstop for the new beamline is described in detail. A neutronics model of the neutron beam provides researchers with a useful tool for experiment design. The new neutron beam has a measured length-to-diameter ratio of 200 ± 10, a neutron flux of 2.2×106 ± 6.4×105 n/cm2-s, and an average cadmium ratio of 7.4 using copper, gold, manganese, and indium foils.

Radiation Shielding Options for a Nuclear Reactor Power System Landed on the Lunar Surface

Abstract A survey of neutron-attenuating materials is conducted, followed by a systematic optimization of the radiation shield configuration for the Affordable Fission Surface Power System. Water, borated water, boron carbide, boron-doped beryllium, zirconium hydride, and lithium hydride are evaluated for neutron shielding, and tungsten is considered for gamma shielding. Lithium hydride, borated water, and boron carbide are selected for further consideration, and radial, upper axial, and lower axial shield sections are developed separately from these materials and then combined to form complete shields. Two competing effects determine the optimal position of the tungsten layer: increasing secondary gamma production due to fast neutron scattering when the tungsten layer is placed closer to the core, and radially increasing mass when placed farther from the core. The optimal position of the tungsten layer is found for each shield configuration and material. The as-landed configuration of each radiation shield allows a maximum dose of 5 rem/yr to an outpost 1 km from the reactor core. The shield also protects the SmCo magnets in the alternators of the Stirling power converters, allowing a maximum dose of 2 Mrad gamma and 1014 n/cm2 fast neutron fluence to the magnets over the 8-yr design lifetime. A minimum mass is found for each shield section while meeting these dose limits. The radial shield section is cylindrical, and the upper and lower axial shield sections are conical in shape. Axial shields with a range of pitch and thickness are analyzed, and the optimal shapes of the upper and lower axial shields for each material are found. The three sections of the shield are combined to form a complete shield. The lithium hydride shield is the lightest of the final shields at 6215 kg. The borated water shield is the second lightest at 6663 kg, which is 448 kg more than the lithium hydride shield. The boron carbide shield is the most massive at 8315 kg, which is 2100 kg more than the lithium hydride shield.

Optimizing Nuclear Material Accounting and Measurement Systems

Abstract The ability to create nuclear weapons from 235U and 239Pu makes it imperative to closely account for these materials as they progress through a nuclear fuel cycle. Improved measurement systems provide more accurate estimates of material quantities and material unaccounted for (MUF). This paper provides examples of how two safeguards computational toolboxes can optimize and analyze hypothetical nuclear fuel cycle scenarios. The NUclear Measurement System Optimization (NUMSO) toolbox uses operations research techniques to find optimal solutions to safeguards measurement problems based on minimizing the variance of the estimated MUF. The SafeGuards Analysis (SGA) toolbox employs Monte Carlo techniques to analyze a given configuration of measurement methods and material flows to determine the probabilities of Type I (false detection) and Type II (missed detection) errors. Applying these toolboxes to a realistic fuel cycle scenario demonstrates the capability of NUMSO and SGA to address nuclear safeguards problems. Working in tandem, both toolboxes are able to determine how to quickly improve upon an existing safeguards measurement system and to calculate the resulting improvement in the error probabilities of the system. This information shows engineers not only how to develop new measurement systems but also how to improve existing systems in the most efficient manner.

Examining Fuel-Cycle Scenarios with the Safeguards Analysis Toolbox

Abstract SafeGuards Analysis (SGA) is a toolbox developed to allow engineers and scientists to create detailed simulations of safeguards material control and accountability simulations. SGA accepts material flow data from an external material flow model and can be used with any existing fuel cycle or material control code. This paper examines some new developments to the SGA code that allow the code to consider material losses over long time frames. The first scenario described in this paper examined an enrichment facility consisting of two material balance areas (MBAs). Cumulative sum and basic control chart tests were evaluated for a case involving a loss of material from both MBAs simultaneously and a case in which material is removed from the facility over a timescale of double the one that the tests were calibrated to detect. A second scenario represents an entire fuel cycle consisting of four MBAs and two materials of interest (low-enriched uranium and plutonium). This scenario evaluated the calibrated safeguards system with three blind unidentified stream cases, with the goal of determining the calibrated system’s ability to detect where the material loss occurred in each case. SGA was able to produce the expected results for all of the examples examined in this paper, demonstrating that modules produced using the toolbox are capable of examining larger systems in realistic multi-MBA scenarios.

Beam Characterization at the Neutron Radiography Facility

The quality of a neutron imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam’s effective length-to-diameter ratio, neutron flux profile, energy spectrum, image quality, and beam divergence, is vital for producing quality radiographic images. This project characterized the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam’s effective length-to-diameter ratio and image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. Improvement of the existing NRAD MCNP beamline model includes validation of the model’s energy spectrum and the development of enhanced image simulation methods. The image simulation methods predict the radiographic image of an object based on the foil reaction rate data obtained by placing a model of the object in front of the image plane in an MCNP beamline model.