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At least 1,351 records · Page 75
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

An innovative accident tolerant LWR fuel rod design based on uranium-molybdenum metal alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The paper is organized as follows. First, an overview of the design features of the U-Mo fuel concept is summarized, including the key attributes and technical challenges to be addressed through the research and development program. Second, recent results are provided on the triple extrusion of a depleted U-Mo fuel rodlet with a niobium liner and SS-316 cladding. A status is presented next on the high temperature water corrosion testing of short samples with pre-drilled cladding defects. Lastly, an assessment of the enhanced accident tolerance features is discussed that highlights the key performance attributes of the U-Mo fuel rod design. Scoping calculations have identified that the reduced stored energy of the U-Mo fuel at the initiation of a LOCA event may reduce core temperatures below 500-600 Celsius degrees during the early phase of the LOCA event. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, anticipated operational occurrences (AOOs) and postulated accidents.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nano-scale mechanical behavior of uranium silicide compounds

Uranium silicide compounds exhibit a range of attractive material properties for nuclear fuel applications, but little is known regarding their mechanical behavior. The structure of nuclear fuels at high burnup can prove to be especially difficult given the significant changes in both chemistry and microstructure caused by fission products. Nano-indentation appears to be a valuable addition to post irradiation examination. The present focus is assessment of the room temperature modulus and hardness of U{sub 3}Si{sub 2}, U{sub 3}Si{sub 5} and USi materials. Each of the compositions under investigation was synthesized via arc melting of the pure metals in an Ar atmosphere using a titanium getter to remove O{sub 2} contamination. The resulting pellets following milling and sintering of the arc melted material were between 96 and 98% of density. Coupling different techniques (SEM and Chemical Analysis), microstructural investigations of the samples has been performed. Preliminary nano-mechanical response of U-Si fuel materials has been demonstrated, on a Hysitron Tribo-indenter which will impact the ability to design advanced fuel materials for enhanced high temperature performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Ammonia gas Treatment for Uranium Immobilization at DOE Hanford Site - 17067

Ammonia gas is a potential uranium (U) remediation technique for the vadose zone at the Department of Energy's (DOE) Hanford Site in Washington State via pH manipulation. The objective of this work was to investigate U removal from the aqueous phase and mineral dissolution with either NaOH or NH{sub 4}OH in order to understand ammonia as a remediation technique for the vadose zone. Batch experiments were designed to investigate the fate of U and mineral dissolution upon treatment with either NaOH or NH{sub 4}OH. These experiments investigated mineral dissolution and U partitioning in the presence of pure minerals (quartz, kaolinite, illite, montmorillonite, muscovite) and sediments relevant to the Hanford site in either NaCl or synthetic groundwater formulated based on the Hanford site groundwater components. Analysis of mineral dissolution via measurement of major cations has demonstrated that there is a significant increase in dissolution of minerals with basic treatment and likely greater secondary precipitation for the NH{sub 4}OH as compared to NaOH treatment for clay minerals and Hanford sediments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Radon Fluxes from an Earthen Barrier Over Uranium Mill Tailings After Two Decades of Service - 17234

Radon (Rn) fluxes were measured at a disposal facility for uranium mill tailings after two decades of service. The facility was closed with an earthen cover vegetated with grasses. Measurements were made on the surface of the Rn barrier and directly on the surface of the tailings. Fluxes were measured using extra small (area = 0.018 m{sup 2}), small (0.071 m{sup 2}), medium (0.59 m{sup 2}), and large (2.32 m{sup 2}) flux chambers to evaluate the impact of measurement scale on Rn flux. Activated carbon (AC) passive collectors and electric radon detectors (RAD7) were used to measure Rn concentrations. Tests were conducted at various locations on each cover representing conditions that can lead to different levels of soil structural development and different water content. Rn fluxes at the surface of the Rn barrier were much lower than Rn fluxes measured at the surface of the tailings, indicating the barrier remained effective for Rn containment. Geometric mean Rn fluxes measured at the surface of the Rn barrier in each test pit were below the regulatory requirement (0.74 Bq/m{sup 2}- s). Rn concentrations measured using AC samplers in the flux chambers were 60% of concentrations measured using the RAD7, on average, and Rn fluxes computed with the AC data were 9% of those computed with the RAD7 data, on average. Size of the flux chamber had no systematic effect on Rn flux, indicating that 20 y of soil forming processes had not created a pore network causing scale-dependent Rn flux. Geometric mean Rn fluxes in the test pits were similar regardless of differences in surface conditions (vegetation, thickness of soil cover) known to influence soil forming processes and soil structure, but were lower in areas where surface conditions promoted higher water content in the Rn barrier. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Understanding Uranium Binding Mechanisms and Speciation on Stainless Steel - 17462

Stainless steels have been exposed to nitric acid solutions containing natural uranium under similar conditions to those found in reprocessing plants. Solution and solid analysis of the stainless steel samples has been undertaken to describe the contamination phenomenon. General corrosion of the surface was seen to take place over the month-long experiments. Contamination has been shown on all exposed surfaces. Solution analysis indicates that an equilibrium state is reached after 2 weeks exposure to the 12 M HNO{sub 3}. Surface analysis through autoradiography and X-ray absorption near edge structure (XANES) indicate that multiple mechanisms of uptake may occur and consequently concentrated regions of contamination may be produced. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2015

Bio-sorption of uranium, Am-241 and Cs-137 by Rice Husk - 15437

New techniques involving the treatment of radioactive waste which combine simplicity and low cost have directed attention to bio-sorption. The bio-sorption term is used to describe the removal of metals, non-metals and/or radionuclides by a material from a biological source, regardless of its metabolic activity. Among the potential biomass, agricultural residues have very attractive features, as they allow for the removal of radionuclides present in the waste using a low cost bio-sorbent. The aim of this study was to evaluate the capacity of rice husk to remove uranium, Am-241 and Cs-137 in real radioactive liquid organic waste. This work was divided into two stages: i) Preparation and physical characterization of the rice husk and ii) Batch bio-sorption experiments. The parameters evaluated for physical characterization of biomass were morphological characteristics of rice husk, real and apparent density and surface area. The adopted contact times were 30 min, 1, 2 and 4 hours and the concentrations tested ranged between 10% and 100%. The results were evaluated by maximum experimental sorption capacity and isotherm and kinetics ternary models. The maximum capacity of the bio-sorption for the rice husk with approximate values was of 0.5 mg/g of U (total), 14 x10{sup -6} mg/g of Am-241 and 20 x10{sup -9} mg/g of Cs-137. These results suggest that bio-sorption with rice husk can be applied in the treatment of radioactive liquid organic waste. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Texas authorizes disposal of large quantities of depleted uranium at wcs - 15626

The U.S. Nuclear Regulatory Commission (NRC) has been actively engaged in a rulemaking process that would revise Title 10 of the Code of Federal Regulations (CFR), Part 61, Licensing Requirements for Land Disposal of Radioactive Waste, to govern the disposal of large quantities of Depleted Uranium (DU) and Blended Waste (referred to as Unique Waste Streams). While the NRC plans to propose a Site Specific Analysis (SSA) rulemaking in early 2015, they have already issued guidance to Agreement States which identified general disposal criteria that could be used should a licensee propose to dispose of Unique Waste Streams before their rulemaking was completed. On August 5, 2013, Waste Control Specialists LLC (WCS) submitted a major license amendment application for review and approval to the Texas Commission on Environmental Quality (TCEQ). This major license amendment request proposed, among other things, authorization to dispose of large quantities of DU at its Federal Facility Waste Disposal Facility (FWF). Pursuant to Title 30 of the Texas Administrative Code (TAC), Rule 336.709, Technical and Environmental Analyses, disposal of waste must be sufficient to meet the radiological dose criteria for a minimum period of performance of 1,000 years after closure or the period where peak dose occurs, whichever is longer. These regulations are more stringent than those of any other state that hosts a facility authorized to dispose of Low-Level Radioactive Waste (LLW) in the U.S. On August 28, 2014, the TCEQ approved the major amendment request authorizing disposal of large quantities of DU. During the licensing review process, significant technical issues were successfully resolved ensuring that large quantities of DU could be safely isolated from the biosphere for long time periods after site closure. Many important lessons were learned during the licensing proceedings that have potential significant impacts to the licensed community. Namely that a modern, well designed disposal facility that is properly sited geologically and in an arid environment is ideal for ensuring that long-lived alpha emitting radionuclides are sufficiently isolated to protect public health for long into the future. The DOE is currently finalizing plans to complete a Supplemental Environmental Analysis (EA) to support shipments of DU for disposal at WCS' FWF as required under the National Environmental Policy Act of 1970. The DOE may begin shipments of DU for disposal to WCS FWF that was licensed and constructed to support disposal of waste that is the responsibility of the federal government as stipulated in the Low-Level Radioactive Waste Policy Amendments Act of 1985. This paper will present the regulatory history of events that initiated the NRC's rulemaking for Unique Waste Streams and the technical analysis that were conducted in support of WCS' major amendment that has been approved by the TCEQ. A description of the geological characteristics and engineering designs used to support the Performance Assessment will be included. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Determination of critical experiment correlations for experiments involving highly enriched uranium solutions

Quantifying the correlations among critical experiments has become a topic of interest over the last 10-15 years. The correlations arise from the shared use of fissile material, non-fissile materials, fixtures, and procedures in multiple experiments or multiple experiment series. A Monte Carlo sampling technique proposed as a viable technical method to evaluate these correlations has been implemented within the SCALE 6.2 Sampler sequence. This paper presents the investigation of the use of this new tool and technique for highly enriched uranium solution experiments. One of the first attempts to quantify the correlations among critical experiments examined a series of highly enriched solution experiments, including the HST-001 series. These correlations are included in the International Criticality Safety Benchmark Evaluation Project Handbook, and they form a basis for methodology comparisons. Results generated using the Sampler implementation are compared with these previously published results. Additional studies involving the HST-001 experiment series are also presented. A range of different assumptions of the independence of the geometry components, enrichments, and other material compositions are implemented to investigate the impact of these assumptions on the correlation coefficients. These effects are investigated with these experiments because the calculation time is shorter for the solution experiments than for the low-enriched lattice experiments discussed in a companion paper. The magnitude of the effects of different independence assumptions on the correlation coefficients is important to establish which parameters must be clearly identified as shared or unique for accurate correlation determinations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Base Treatment for Uranium Immobilization at DOE's Hanford Site - 18279

Ammonia gas injection is currently being field tested for uranium remediation in the vadose zone at the Department of Energy's Hanford site in Washington State. U is a significant risk driver for the site because of its high mobility in the subsurface and the significant mass released due to Manhattan Project activities. In this research, we conducted laboratory batch experiments to evaluate the changes in mobility of U with different base treatments. Based on our current body of work, base treatment is a promising technique for U immobilization with ammonia gas being the optimal choice because (1) it increases partitioning coefficients for U and results in significant removal from the aqueous phase and (2) it does not add liquid to the vadose zone which could initially increase the flux of U into groundwater. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Results of Robotic Evaluation of Uranium-235 in Gaseous Diffusion Piping Holdup Deposits - 18303

Measurement of U-235 in pipes within old uranium enrichment facilities is time-consuming and very costly. Radiometric assay from outside the pipes suffers from manual deployment challenges, attenuation of detection through pipe walls, long counting times, approximate modeling, and shortfalls associated with transcription and human interpretation. The D and D process of cutting into these pipes creates a unique possibility of robotic in-pipe assay to address these problems. This paper presents the results of testing a first-of-kind radiometric non-destructive assay (NDA) robot for high-cadence, accurate, certain, paperless assay of U-235 in holdup deposits within process piping. The robot's enabling capability is an innovative radiometric collimation method that views only a one-foot moving annulus of pipe wall. This exploits the internal axisymmetry of pipes to provide a direct measurement of industry-standard grams per foot U-235. The initial proof-of-application robot was successfully hot tested by operators at the Gaseous Diffusion Enrichment Facility in Piketon, Ohio, in late 2017. This testing achieved all goals of traversing pipe, measuring U-235, deployment by facility personnel, and displaying analyzed radiation, visual, and geometric information. The success was a Kitty Hawk moment. The system demonstrated robotic in-pipe measurement of holdup deposit to a resolution, accuracy, and speed unachievable by alternate methods. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Advanced Waste Management Concepts for Uranium and Plutonium Multi-Recycling - 18533

The multiple recycling of plutonium and uranium will play a major role in gradually achieving a fully closed nuclear fuel cycle in the future, once the first Generation IV fast reactors (FR) are deployed in the NPP fleet. To reach this objective, optimizations of current processes and technologies and, in some cases, the development of innovative concepts are required to meet the scientific, technical and economic challenges imposed by the treatment of spent fuels with an increasingly higher Pu content. Based on the significant past experience acquired by France in LWR and FR MOX fuel treatment, the CEA is running R and D studies to develop new efficient processes for Pu recycling and for FR MOX fuel fabrication. More compact and economically competitive processes, ensuring the safety/security of the facilities (particularly as regards criticality and proliferation resistance) have to be developed taking into account that the amounts of plutonium involved will be one order of magnitude higher than in UOX fuels. New concepts of head-end treatment, dissolution, and separation have been proposed. Consequently, effluent and waste management must be adapted to these processes and to the characteristics of fast reactor MOX fuels. New waste treatment processes are suggested: 1) Cladding waste and undissolved metallic particles could be treated simultaneously in a melting process. The molten metal could be decontaminated by a specific slag and the oxides recovered, in particular PuO{sub 2}, would be recycled at the frond-end of the process, 2) Fission Products and Actinides (FPA) containment glass will also be optimized to accommodate significant amounts of iron, nickel and chromium coming from hull corrosion, and / or decontamination slag coming from the metal melting process. 3) New peraluminous-type glass formulations are under development to accept increased amounts of FPA, as well as iron, nickel, and chromium. The first trials to evaluate the melting process have indicated that metallic compounds from the dissolution residues could be easily incorporated into a molten metal at levels of 1, 2 and 3 w%. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Study of Synergetic Interactions between Uranium, Humic Acid, Silica Colloids and SRS Sediments - 18581

The Savannah River Site's (SRS) F/H Area Seepage Basins received approximately 1.8 billion gallons of acidic waste containing radionuclides and dissolved heavy metals. This led to the creation of a highly contaminated groundwater plume with uranium (VI) as a key contaminant of concern within the plume. Humic substances (HS) are major components of soil organic matter having the ability to influence migration behavior and fate of heavy metals and are being investigated for potential use in environmental remediation at SRS. HS are polyfunctional, organic macromolecules able to carry a large number of functional groups which is an important function in ion exchange and is a metal complexing ligand with a high complexation capacity. Humic acid (HA) represents the fraction of HS soluble at pH greater than 3.5. Silica, naturally found in soil, is the term applied to solid forms with the stoichiometric composition of SiO{sub 2}. The colloidal silica used in this work is amorphous and non-porous in suspension. This investigation studied the synergistic interactions between four key components: U (VI), humic acid (HA), colloidal silica and SRS sediment; the interactions were studied under pH 3 - 8. Multi-component batch systems were constructed to effectively analyze the U (VI) removal. The results show that filtered and unfiltered samples depicted similar trends; while, filtered samples had more overall removal than the unfiltered samples at the varying pH levels. Colloidal silica does not have a significant effect on the removal of U (VI) due to U (VI) bound to colloidal silica remaining in suspension. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Transition Benefits of using Low Enriched Uranium Fuel in Fast Reactors

An Evaluation and Screening (E and S) study of nuclear fuel cycle options was recently conducted for the U.S. Department of Energy to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the U.S. This study considered the entire fuel cycle, i.e., the complete nuclear energy system from mining to disposal including both once-through and recycle fuel cycles, and identified four options that were most promising, offering the highest benefits compared to the current cycle. The E and S study considered performance at steady-state when evaluating the potential benefits and challenges to achieving that eventual end state. Based on the steady-state findings from the E and S, a methodology was developed for extending the evaluation of the fuel cycle options over the period of transition from the current state of the U.S. fuel cycle to a completely new system. Such methods are being applied to a range of potential transition scenarios involving the aforementioned most promising options, with the goal of identifying issues and challenges associated with such transitions. For the first detailed analysis of transition, a single example scenario of transition was developed in order to implement the methodology. The timescales, assumptions, and specifications of this example scenario were chosen to be representative of a reasonable transition from the current state of the U.S. fuel cycle to a fuel cycle that will continuously recycle U/Pu in fast critical reactors (referred to as 'EG23'). The results represent just one transition pathway, but not necessarily the best or optimized solution for that specific scenario. These transition scenarios were modeled using nuclear fuel cycle systems codes in order to provide the time-dependent mass flows and inventories, reactor/facility capacities and throughputs, and system behavior that were used for performance analyses. These analyses of the example scenario identified a number of potential system issues that can be applied generally, but they did not consider the much broader transition space (scenarios involving various FR startup fuels, different cooling times, growth rates, recycling technologies, etc.). Therefore, additional studies were performed to inform over a broad range of these potential transition pathways. The main focus of this paper is on the results of these additional analyses for transition pathways that involve the use of low enriched uranium fuel as the initial startup fuel for fast reactors during transition. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Continuum Scale Modeling of Thermal Conductivity in Hyperstoichiometric Uranium Dioxide

Conventional UO{sub 2} nuclear fuel is susceptible to oxidation via the formation of hyperstoichiometric uranium dioxide (UO{sub 2+x}) when exposed to air or high temperature water vapor in the event of a cladding breach during operation or storage. Below 1130 deg. C intermediate oxides U{sub 4}O{sub 9} and U{sub 3}O{sub 7} form; which degrades the thermal conductivity of the fuel. Understanding how the thermal conductivity degrades as a function of oxygen content and phase fractions is an important engineering problem for future fuel performance codes for safe and efficient reactor operation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Ab Initio Molecular Dynamics Analysis of Xenon Incorporation in Uranium Dioxide

The fission process in nuclear fuel produces volatile fission product gases, such as xenon, that migrate out of the fuel and into surrounding structures. Consequently, these fission product gases may affect thermal performance and integrity of the fuel. Fission gas release models, used in modern fuel performance codes such as BISON, are currently under investigation to improve their accuracy and ability to simulate the fission gas release phenomenon and its consequences. For uranium dioxide (UO{sub 2}), the most commonly used fuel in commercial power reactors, fission gas release models are known to have significant uncertainties. In the interest of improving such models, several complex diffusional mechanisms have been examined using a variety of techniques such as Nudge Elastic Band (NEB), utilizing both Density Functional Theory (DFT) and semi-empirical inter-atomic potential functions. One proposed diffusion mechanism in UO{sub 2} is through tetra-vacancy (i.e., 2U{sub V} + 2O{sub V}) structures, which is expected to provide a suitable means for the incorporation and diffusion of xenon in the fuel crystal. In this work, a model was developed to study the incorporation of xenon in UO{sub 2} via a tetra-vacancy using ab initio molecular dynamics (AIMD). This is a first principles technique that abandons the use empirical force fields as usually is the case of classical MD analysis. Furthermore, in contrast with static transition analysis tools such as NEB, AIMD is able to capture the dynamics of a given process. The translation of xenon within a tetra-vacancy has been investigated using ab-initio molecular dynamics. The transition frequency, barrier energy, and incorporation energy were extracted from the model, demonstrating its ability to capture a variety of parameters and its potential for use in studying fission gas diffusive behavior. Future refinements to the model will seek to develop additional insight into Xe and Kr mobility and will likely involve executing more elaborate AIMD simulations to investigate this phenomenon. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Uranium Partitioning upon Ammonia Gas Treatment on Phyllosilicate Minerals

In the past, > 200,000 kg of U was released to the Hanford Site vadose zone from disposal and leakage of waste byproducts from plutonium production during World War II and the Cold War (Zachara et al., 2007). U is highly mobile due to the presence of carbonates and oxidizing conditions, with K{sub d} of 0.1 - 8 mL/g at the site (Zachara et al., 2007; Szecsody et al., 2013; Serne et al., 2008). NH{sub 3} (gas) injection is being investigated as a remediation technique as it is an in situ method that will not introduce additional liquid that could drive contaminants deeper into vadose zone towards groundwater. Objectives: To understand the impact of alkaline injection on geochemical processes in the Hanford Site vadose zone, including: 1)Long-term fate of uranium (U); 2)Dissolution and re-precipitation of secondary minerals. Apparent K{sub d}'s increase significantly with NH{sub 3} gas treatment, especially for six-month aging. - K{sub d} decreases after aeration, primarily due to U desorption at lower pH. U removal is still significantly greater than natural conditions (K{sub d} < 8 vs. > 100 mL/g). - Three-day aeration has higher loading than six-month after ammonia gas treatment and aeration. However, significantly greater desorption of U occurs with aeration in the six-month aged sample potentially due to transformation into different, more soluble solid phases with longer term aging. - U removal is primarily due to (1) co-precipitation of U with secondary minerals and (2) adsorption to minerals' surface, although further research is required to identify dominant mechanisms. Future experiments will focus on garnering a better understanding of the solid phase U speciation. Solids will be characterized using SEM-EDS, EMPA, XPS, XAFS, and TEM to better understand the effects of basic treatments on initial minerals and formation of secondary precipitates.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Investigating Uranium Corrosion in Magnesium-containing Sludge Using X-ray Tomography - 20042

One of the current primary concerns within the British nuclear community is the decommissioning of legacy storage ponds in Sellafield Cumbria, where nuclear fuel cladding material (a Mg-Al alloy) resides along with fuel metallic swarf. Corrosion of the immersed metallic elements, which has been occurring over the storage period, has resulted in the formation of what is commonly known as Corroded Magnox Sludge. The behaviour of uranium metal within this particular environment and, more significantly, the identity of the corrosion products that may have formed is currently unknown. The present study attempts to shed some light on these aspects by investigating simplified surrogate systems, mimicking storage ponds. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Ab initio generation of thermal neutron scattering law for uranium dioxide

Neutronic analysis of moderated nuclear fission systems, e.g. nuclear reactors, relies upon an accurate prediction of the thermal neutron spectrum. Thermal neutrons have wavelengths on the order of the interatomic spacing between atoms in matter. Consequently, the scattering cross-sections of these neutrons are affected by the scattering law of the material (i.e. S(α,β), where α and β are dimensionless momentum and energy transfer variables), which describes the atomic structure and thermodynamic motion of the atomic nuclei. Uranium-dioxide (UO{sub 2}) is the primary fuel source for nuclear power reactors moderated by light water. In such systems (e.g. nuclear power reactors, spent fuel pools, etc.) the nuclear fuel comprises a large fraction of the system volume, approximately 1/3 in light-water reactors. Therefore, the scattering law of the fuel material may impact the thermal neutron spectrum and criticality of these systems. In this work previously established ab initio lattice dynamics (AILD) methods were utilized to calculate the phonon density of states (DOS), which acts as a fundamental input to the scattering law. Subsequently, the S(α,β) of UO{sub 2} was generated for publication as an ENDF thermal scattering library (TSL), which will be available for use in standard neutronics computer applications. (authors)

07 ISOTOPE AND RADIATION SOURCES↗