Browsing by Subject "spent fuel"
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Item Development of a Real-Time Detection Strategy for Material Accountancy and Process Monitoring During Nuclear Fuel Reprocessing Using the Urex+3A Method(2010-07-14) Goddard, BradenReprocessing nuclear fuel is becoming more viable in the United States due to the anticipated increase in construction of nuclear power plants, the growing stockpile of existing used nuclear fuel, and a public desire to reduce the amount of this fuel. However, a new reprocessing facility in non-weapon states must be safeguarded and new reprocessing facilities in weapon states will likely have safeguards due to political and material accountancy reasons. These facilities will have state of the art controls and monitoring methods to safeguard special nuclear materials, as well as to provide real-time monitoring. The focus of this project is to enable the development of a safeguards strategy that uses well established photon measurement methods to characterize samples from the UREX+3a reprocessing method using a variety of detector types and measurement times. It was determined that the errors from quantitative measurements were too large for traditional safeguards methods; however, a safeguards strategy based on qualitative gamma ray and neutron measurements is proposed. The gamma ray detection equipment used in the safeguard strategy could also be used to improve the real-time process monitoring in a yet-to-be built facility. A facility that had real-time gamma detection equipment could improve product quality control and provide additional benefits, such as waste volume reduction. In addition to the spectral analyses, it was determined by Monte Carlo N Particle (MCNP) simulations that there is no noticeable self shielding for internal pipe diameters less than 2 inches, indicating that no self shielding correction factors are needed. Further, it was determined that HPGe N-type detectors would be suitable for a neutron radiation environment. Finally, the gamma ray spectra for the measured samples were simulated using MCNP and then the model was extended to predict the responses from an actual reprocessing scenario from UREX+3a applied to fuel that had a decay time of three years. The 3-year decayed fuel was more representative of commercially reprocessed fuel than the acquired UREX+3a samples. This research found that the safeguards approach proposed in this paper would be best suited as an addition to existing safeguard strategies. Real-time gamma ray detection for process monitoring would be beneficial to a reprocessing facility and could be done with commercially available detectors.Item Development of Self-Interrogation Neutron Resonance Densitometry (SINRD) to Measure the Fissile Content in Nuclear Fuel(2011-10-21) Lafleur, AdrienneThe development of non-destructive assay (NDA) capabilities to directly measure the fissile content in spent fuel is needed to improve the timely detection of the diversion of significant quantities of fissile material. Currently, the International Atomic Energy Agency (IAEA) does not have effective NDA methods to verify spent fuel and recover continuity of knowledge in the event of a containment and surveillance systems failure. This issue has become increasingly critical with the worldwide expansion of nuclear power, adoption of enhanced safeguards criteria for spent fuel verification, and recent efforts by the IAEA to incorporate an integrated safeguards regime. In order to address these issues, the use of Self-Interrogation Neutron Resonance Densitometry (SINRD) has been developed to improve existing nuclear safeguards and material accountability measurements. The following characteristics of SINRD were analyzed: (1) ability to measure the fissile content in Light Water Reactors (LWR) fuel assemblies and (2) sensitivity and penetrability of SINRD to the removal of fuel pins from an assembly. The Monte Carlo Neutral Particle eXtended (MCNPX) transport code was used to simulate SINRD for different geometries. Experimental measurements were also performed with SINRD and were compared to MCNPX simulations of the experiment to verify the accuracy of the MCNPX model of SINRD. Based on the results from these simulations and measurements, we have concluded that SINRD provides a number of improvements over current IAEA verification methods. These improvements include: 1) SINRD provides absolute measurements of burnup independent of the operator?s declaration. 2) SINRD is sensitive to pin removal over the entire burnup range and can verify the diversion of 6% of fuel pins within 3? from LWR spent LEU and MOX fuel. 3) SINRD is insensitive to the boron concentration and initial fuel enrichment and can therefore be used at multiple spent fuel storage facilities. 4) The calibration of SINRD at one reactor facility carries over to reactor sites in different countries because it uses the ratio of fission chambers (FCs) that are not facility dependent. 5) SINRD can distinguish fresh and 1-cycle spent MOX fuel from 3- and 4-cycles spent LEU fuel without using reactor burnup codes.Item Nationwide Used Fuel Inventory Analysis(2013-11-27) Yancey, KristinaThe goal of this research was to develop a methodology to collect inventory estimates for the analysis and characterization of used fuel in the United States. To accomplish this, the Spent Fuel Database (SFD) was created. Data was collected for the database from publicly available information on the 103 operating reactors in January 2012. Using this data, plant models were developed using ORIGEN-ARP, a point-depletion tool. The output for each reactor model included current inventory estimates for used fuel taken out of the reactor 0, 1, 3, 5, 10, and 20 years ago. To determine the applicability of the database, a methodology was developed to analyze and compare the SFD with mass values produced using knowledge of past fuel assembly designs for general reactor classes. The methodology was centered around the idea of the ?applicability range? (AR) of the database, which was defined as the degree to which a correct estimate can be made quantitatively. Pressurized Water Reactors (PWRs) were shown to have a much higher AR than Boiling Water Reactors (BWRs), and older assembly classes were shown to have a lower AR than newer classes. The fission products in the database were shown to consistently have a high AR. Berkelium and californium had low AR for all of the assembly classes, curium had low AR for BWR classes and mixed AR for PWR classes, and americium and some plutonium isotopes had low AR for BWR classes. An assessment of the inventory estimates considered the potential radiotoxicity and heat load from these masses. The radiotoxicity by ingestion decreased by about a factor of 10 from the newest used fuel to the oldest, and the radiotoxicity by inhalation decreased by a factor of 2. While one person could never eat or inhale a spent fuel assembly, radiotoxicity was used as a metric for the upper limit of possible harm. The heat load decreased by more than a factor of 100 over the same range of fuel assemblies. On a per assembly basis, the radiotoxicity and heat load showed similar trends, with newer PWR assemblies being the highest and BWR assemblies being the lowest in both categories. Considering these results, at a potential interim storage facility, priority should be given to the oldest BWR assemblies to reduce the radiotoxic risk and heating requirements. Also, reprocessing and transmuting is highly encouraged to reduce the radiotoxicity and heat of the waste entering storage. Finally, to continue improving the SFD, future work should seek to quantify the magnitude of the impact of variations in AR for curium and for BWR classes. Moreover, future work should incorporate the used fuel from all the shutdown reactors into the database. Even in its current form, though, the SFD is a useful reference tool.