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[en] The future expansion of nuclear power will require not just electricity production but fuel cycle facilities such as fuel fabrication and reprocessing plants. As large reprocessing facilities are built in various states, they must be built and operated in a manner to minimize the risk of nuclear proliferation. Process monitoring has returned to the spotlight as an added measure that can increase confidence in the safeguards of special nuclear material (SNM). Process monitoring can be demonstrated to lengthen the allowable inventory period by reducing accountancy requirements, and to reduce the false positive indications. The next logical step is the creation of a Safeguards Envelope, a set of operational parameters and models to maximize anomaly detection and inventory period by process monitoring while minimizing operator impact and false positive rates. A brief example of a rudimentary Safeguards Envelope is presented, and shown to detect synthetic diversions overlaying a measured processing plant data set. This demonstration Safeguards Envelope is shown to increase the confidence that no SNM has been diverted with minimal operator impact, even though it is based on an information sparse environment. While the foundation on which a full Safeguards Envelope can be built has been presented in historical demonstrations of process monitoring, several requirements remain yet unfulfilled. Future work will require reprocessing plant transient models, inclusion of 'non-traditional' operating data, and exploration of new methods of identifying subtle events in transient processes
[en] A thermal neutron imaging facility (TNIF) was developed at the University of Texas Nuclear Engineering Teaching Laboratory from 1994 to 1998 using a 1-MW TRIGA reactor. Currently, neutron radiography is being investigated as a method to detect flaws in large carbon composite flywheels using the TNIF. Thermal neutrons have successfully been used to detect flaws in thin carbon composites (<1 cm thick), but image quality can degrade as composite thickness increases and neutrons are attenuated. To improve the TNIF capabilities for this application, the facility shielding is being redesigned to accommodate operation at the maximum reactor power to achieve the highest possible neutron flux. Secondary goals of the shielding redesign are to minimize the backscatter of neutrons from external shielding into the imaging system, another source of image degradation, and to improve overall facility usability. Monte Carlo calculations using the MCNP code have been performed to verify the redesign of each component before construction. The shutter system is composed of a lead gamma shutter embedded in the reactor biological shielding and an external neutron shutter composed of 8-in.-thick 5% borated polyethylene housed in 8-in.-thick polyethylene. Only the neutron shutter can be modified. At a reactor power level of 500 kW, the average radiation level on the surface of the neutron shutter was measured at 120 mR/h. Scaling linearly to the full reactor power of 1 MW, the transmitted radiation level is expected to be 240 mR/h. The goal of the redesigned shutter is to reduce the exposure through the neutron shutter by a factor of at least 48 to 5 mR/h without increasing the shutter thickness. MCNP calculations have shown that a laminated shutter composed of boron carbide powder, lead, and aluminum can successfully attenuate the beam. Increasing the shutter thickness by 50% provides a safety factor near 3. The current shielding cave is constructed from blocks of concrete stacked to form ∼30-in.-thick walls. In the current configuration, an external exposure of 5 mR/h is achieved at a reactor power of 250 kW. Assuming that the dose scales linearly with power, the external dose must be reduced by a factor of 4 to operate the TNIF at the full power of 1 MW. An important note on the physical structure of the current cave that is not simulated is that the bricks used are somewhat misshapen and poorly aligned in some cases. This may allow streaming paths in several locations that could be the primary cause of the external radiation fields. Reassembling the cave with new bricks and great care may be sufficient to reduce the external dose to acceptable levels at full power. A second motivation to redesign the existing external radiography cave is to reduce the amount of neutrons that reflect from the inside wall of the shield back into the experimental area. Neutrons that reflect back into the neutron imaging equipment can degrade image quality and reduce resolution, which is of primary concern for the carbon composite analysis. Simulations of the current wall indicate that >60% of the neutrons that enter the shield walls are reflected back into the experimental area. MCNP calculations indicate that the addition of a 1.25-cm Boral liner on the inner wall is sufficient to lower the external dose to acceptable levels and reduce the percentage of neutrons reflected back into the experimental area to <2%. MCNP simulations have been a valuable tool to test shielding configurations before construction. The redesigned shutter is composed of aluminum, lead, and boron carbide. MCNP simulations for the external shielding have shown that the addition of a Boral liner on the inner shield wall is sufficient to reduce external radiation exposure to acceptable levels. The Boral liner also greatly reduces the amount of neutrons reflected back into the experimental region. The implementation of the redesigned neutron shutter and external shielding should greatly enhance the TNIF capabilities and overall usability. The new neutron shutter will allow work to be performed inside the shielding cave while the reactor is at power. The improved external shielding will enable radiographs to be taken at higher flux levels, which will be beneficial when imaging thick carbon composites. The reduction of neutron scattering within the experimental area will also enhance image quality and improve the TNIF resolution. (authors)
[en] The preliminary results from a series of nuclear fluorescence imaging experiments using a variety of radioactive sources and shielding are given. These experiments were done as part of a proof of concept to determine if nuclear fluorescence imaging could be used as a safeguards measurements tool or for nuclear warhead verification for nuclear arms control treaties such as the New Strategic Arms Reduction Treaty and the Fissile Material Cut-Off Treaty. An off-the-shelf Princeton Instruments charged coupled device camera system was used to image the emission of fluorescence photons from the de-excitation of nitrogen molecules in air that have been excited by ionizing radiation. The fluorescence emissions are primarily in the near ultraviolet range; between the wavelengths of 300 and 400 nm. Fluorescent imaging techniques are currently being investigated in a number of applications. A French research team has successfully demonstrated this concept for remote imaging of alpha contamination. It has also been shown that the phenomenon can be seen through translucent materials and that alpha radiation can be seen in the presence of large gamma backgrounds. Additionally, fluorescence telescopes and satellites utilize the de-excitation of nitrogen molecules to observe cosmic ray showers in the atmosphere. In cosmic ray shower detection, electrons are the main contributor to the excitation of the of nitrogen molecules in air. The experiments presented in this paper were designed to determine if the imaging system could observe beta emitting sources, differentiate between beta emitters and alpha emitting materials such as uranium oxide and uranium metal, and to further investigate the phenomenon through translucent and non-translucent materials. The initial results show that differentiation can be made between beta and alpha emitting sources and that the device can observe the phenomenon through very thin non-transparent material. Additionally, information is given on the detection of the fluorescent photons through translucent materials. Camera images, analysis, and results of the initial laboratory experiments are presented. (authors)
[en] A methodology for determining alternate nuclear material (237Np, 241Am, and 243Am) concentrations in spent nuclear fuel based on the use of various monitors was developed and validated for use with several pressurized water reactor fuels. The monitors studied included the fuel burnup, the total plutonium concentration, the 240Pu/239Pu isotopic ratio, the 148Nd/238U isotopic ratio, and the 137Cs activity. Calculations were performed using the HELIOS-1.4 lattice physics code for spent fuel from the Mihama Unit 3, Genkai Unit 2, and Calvert Cliffs Unit 1 reactors. These calculations were compared to measured values for the fuel. It was determined that the 240Pu/239Pu isotopic ratio and the 137Cs activity were the most useful and accurate for use in predicting alternate nuclear material concentrations at reprocessing facilities for safeguards purposes. Based on these comparisons, it was determined that measurements of these monitors would allow for determination of 237Np, 241Am, and 243Am concentrations to within ±4, ±6, and ±15%, respectively. It is suggested that these uncertainties may be decreased through improvements in measurement techniques and additional benchmarking. These monitors may be used to provide an accurate prediction of the concentrations of the alternate nuclear materials while decreasing the need for direct measurement of these isotopes. This will translate into a monetary savings for reprocessing facility safeguards
[en] While gamma measurements are a well understood technique for estimating burn up and cooling time in spent nuclear fuel, only a few isotopes are currently used and calculations are dependent on operator declarations. Utilizing a larger set of nuclide measurements as well as including low energy measurements (around 100 keV) could provide a significant increase in the information gained from such measurements. Nuclides investigated in this work include 235U, 238U, 239Pu, 240Pu, 137Cs, 154Eu, 241Am, 155Eu, 149Sm, and 155Gd. Low energy, broad range, and high energy gamma measurements have been performed on a variety of spent uranium dioxide and MOX fuel from PWRs. Selected measured fuel locations have been simulated with Origen, TransLAT, SCALE, and Monteburns to benchmark each of these programs with destructive analysis results. Monteburns was identified as the most accurate program for prediction of isotopes of interest and will be used for future analysis. Origen was chosen as a quick but relatively accurate program to perform a preliminary sensitivity analysis of the effects of fuel parameters (burnup, cooling time, and initial enrichment) on each nuclide measurement. (author)
[en] Radiation Portal Monitors (RPMs) are our primary border defense against nuclear smuggling, but are they still the best way to spend limited funds? The purpose of this research is to strategically compare RPM defense at the border with state-side mobile detectors. Limiting the problem to a comparison of two technologies, a decision-maker can prioritize how to best allocate resources, by reinforcing the border with stationary overt RPMs, or by investing in Mobile Radiation Detection Systems (MRDs) which are harder for an adversary to detect but may have other weaknesses. An abstract, symmetric network was studied to understand the impact of initial conditions on a network. An asymmetric network, loosely modeled on a state transportation system, is then examined for the technology that will maximally suppress the adversary's success rate. We conclude that MRDs, which have the advantage of discrete operation, outperform RPMs deployed to a border. We also conclude that MRDs maintain this strategic advantage if they operate with one-tenth the relative efficiency of their stationary counter-parts or better.
[en] The International Atomic Energy Agency has recently drawn attention to the fact that neptunium (Np), a byproduct of the nuclear power industry, can be used to make nuclear weapons. Current monitoring approaches for Np do not rely on material balance accounting as is used for uranium and plutonium. In the future this may change. Although full material balance accounting is not anticipated for Np, it is informative to evaluate the impact and benefit of full material balance accounting when considering other options. Therefore, this paper will apply systems analysis to evaluate ways to convert the current system to full materials balance accounting that will minimize the intrusiveness of the verification system and minimize costs to both the facility operator and the inspection agency. We then compare full material balance accounting to partial material balance accounting and to a ratio-monitoring technique referred to as flow sheet verification. We conclude that sampling approximately 25% of the batches is likely to be adequate and that Pu (or perhaps 137Cs) will be the most effective surrogate for estimating the Np in the input accountability tank. (author)
[en] Radiation portal monitors are being deployed at border crossings throughout the world to prevent the smuggling of nuclear and radiological materials; however, a tension exists between security and the free-flow of commerce. Delays at ports-of-entry have major economic implications, so it is imperative to minimize portal monitor screening time. We have developed an algorithm to locate a radioactive source using a distributed array of detectors, specifically for use at border crossings. To locate the source, we formulated an optimization problem where the objective function describes the least-squares difference between the actual and predicted detector measurements. The predicted measurements are calculated by solving the 3-D deterministic neutron transport equation given an estimated source position. The source position is updated using the steepest descent method, where the gradient of the objective function with respect to the source position is calculated using adjoint transport calculations. If the objective function is smaller than the convergence criterion, then the source position has been identified. This paper presents the derivation of the underlying equations in the algorithm as well as several computational test cases used to characterize its accuracy.
[en] The ability to quickly quantify the Pu content within spent nuclear fuel (SNF) is essential to nuclear forensics. Analysis of the Pu to U ratio can provide information on fuel origin which could contribute to the attribution of a fuel sample. Plutonium concentration data can be acquired through non-destructive analysis (NDA) by detecting self-induced x-ray fluorescence (XRF) from Pu in the fuel. However, during conventional spectroscopy, the characteristic Pu x-ray peak of interest lies beneath background and requires an extended count time. Crystal spectrometers allow x-rays of selected energies, obeying Bragg's law for coherent scattering of incident photons, to be focused directly onto a detector. This provides a high signal with limited background by decreasing the possible Compton interaction in the detector. The crystal design and the experimental geometry that would allow for the study of high energy x-rays were investigated. In addition, a preliminary MCNP simulation and external routine to determine the energy-direction coupled photon source from the quartz crystal was used to calculate the improved signal-to-noise ratio of the Pu x-ray peak above background. (author)
[en] As new reprocessing methods for spent nuclear fuel are developed, such as the uranium extraction (UREX) process, methods using nondestructive assay (NDA) techniques must also be developed to allow for quantitative measurements of product materials. Currently developed NDA techniques cannot directly quantify materials containing U, Np, Pu, and Am. This research investigates the ability to quantify these actinides in an oxide form using neutron multiplicity measurements. This technique assumes that the isotopic composition of the sample is known, either through gamma spectroscopy or other means. This measurement technique is based on performing three different neutron measurements and analyzing their neutron multiplicity response. The first is a passive measurement of the product material to determine the effective plutonium-240 (240Pueff) content, self multiplication (M), and alpha-neutron reaction rate (α). The second is an active, AmLi (α, n) source, measurement of the product material to determine the effective 235U content. The third is an active, AmB (α, n) source, measurement of the product material to determine the effective 237Np content. The quantity of Am in the sample can be determined from α. Simulated results using Monte Carlo N-Particle eXtended (MCNPX) version 2.6 will illustrate the viability of this technique and its practical limitations. (author)