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At least 1,063 records · Page 59

Development of U isotope fractionation as an indictor or U(VI) reduction in uranium plumes

This is the final report for a university research project that advanced development of a new technology for identifying chemical reduction of uranium contamination in groundwater at the Rifle Field Challenge site. Reduction changes mobile hexavalent uranium into immobile U(IV). The stable isotope ratio (238U/235U) measurements of U using multicollector ICP-mass spectrometry were performed to understand the chemical reduction and sorption processes during various field experiments. In addition laboratory experiments were performed to better understand the isotopic fractionations. The main objectives of this project were completed during the project period and two peer-reviewed articles were published to disseminate the information gained.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Uranium Conversion & Enrichment

The isotopes of uranium that are found in nature, and hence in ‘fresh’ Yellowcake’, are not in relative proportions that are suitable for power or weapons applications. The goal of conversion then is to transform the U 3 O 8 yellowcake into UF 6 . Conversion and enrichment of uranium is usually required to obtain material with enough 235 U to be usable as fuel in a reactor or weapon. The cost, size, and complexity of practical conversion and enrichment facilities aid in nonproliferation by design.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Accident Analyses for Conversion of the University of Missouri Research Reactor (MURR) from Highly-Enriched to Low-Enriched Uranium

This report contains the results of reactor accident analyses for the University of Missouri Research Reactor (MURR). The calculations were performed as part of the conversion from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members in the Research and Test Reactor Department at the Argonne National Laboratory (ANL) and the MURR Facility. MURR LEU conversion is part of an overall effort to develop and qualify high-density fuel within the U.S. High Performance Research Reactor Conversion (USHPRR) program conducted by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization (M 3 ).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Optimization of Uranium Molecular Deposition for Alpha-Counting Sources

Method development for molecular deposition of uranium onto aluminum 1100 plates was conducted with custom plating cells at Lawrence Livermore National Laboratory. The method development focused primarily on variation of electrode type, which was expected to directly influence plated sample homogeneity. Solid disc platinum and mesh platinum anodes were compared and data revealed that solid disc platinum anodes produced more homogenous uranium oxide films. However, the activity distribution also depended on the orientation of the platinum electrode relative to the aluminum cathode, starting current, and material composition of the plating cell. Experiments demonstrated these variables were difficult to control under the conditions available. Variation of plating parameters among a series of ten deposited plates yielded variations up to 30% in deposition efficiency. Teflon particles were observed on samples plated in Teflon cells, which poses a problem for alpha activity measurements of the plates. Preliminary electropolishing and chemical polishing studies were also conducted on the aluminum 1100 cathode plates.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Investigation of uranium molecular species using laser ablation

The goal of this project is to investigate the dynamic evolution of uranium oxide (UOx) molecular species in a rapidly cooling low-temperature plasma using a coupled experimental and modeling approach. Our purpose is to develop quantitative constraints on the UOx phase chemistry under physical conditions similar to that of a nuclear fireball at the time of debris condensation. This work is motivated by a need to better understand the factors controlling uranium chemical fractionation in post-detonation nuclear debris.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Thermal Conductivity of Metallic Uranium

This project has developed a modeling and simulation approaches to predict the thermal conductivity of metallic fuels and their alloys. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary. The models incorporated both phonon and electron contributions. Good agreement with experimental data over a wide temperature range were found. The models also provided insight into the different physical factors that govern the thermal conductivity under different temperatures. The models were general enough to incorporate more complex effects like additional alloying species, defects, transmutation products and noble gas bubbles to predict the behavior of complex metallic alloys like U-alloy fuel systems under burnup. 3 Introduction Thermal conductivity is an important thermal physical property affecting the performance and efficiency of metallic fuels. Some experimental measurement of thermal conductivity and its correlation with composition and temperature from empirical fitting are available for U, Zr and their alloys with Pu and other minor actinides. However, as reviewed in by Kim, Cho and Sohn, due to the difficulty in doing experiments on actinide materials, thermal conductivities of metallic fuels have only been measured at limited alloy compositions and temperatures, some of them even being negative and unphysical. Furthermore, the correlations developed so far are empirical in nature and may not be accurate when used for prediction at conditions far from those used in the original fitting. Moreover, as fuels burn up in the reactor and fission products are built up, thermal conductivity is also significantly changed. Unfortunately, fundamental understanding of the effect of fission products is also currently lacking. In this project, we probe thermal conductivity of metallic fuels with ab initio calculations, a theoretical tool with the potential to yield better accuracy and predictive power than empirical fitting. This work will both complement experimental data by determining thermal conductivity in wider composition and temperature ranges than is available experimentally, and also develop mechanistic understanding to guide better design of metallic fuels in the future. So far, we focused on α-U perfect crystal, the ground-state phase of U metal. We focus on two methods. The first method has been developed by the team at the University of Wisconsin Madison. They developed a practical and general modeling approach for thermal conductivity of metals and metal alloys that integrates ab-initio and semi-empirical physics-based models to maximize the strengths of both techniques. The second method has been developed by the team at Virginia Tech. This approach consists of a determining the thermal conductivity using only ab-initio methods without any fitting parameters. Both methods were complementary and very helpful to understand the physics behind the thermal conductivity in metallic uranium and other materials with similar characteristics. In Section I, the combined model developed at UWM is explained. In Section II, the ab-initio method developed at VT is described along with the uranium pseudo-potential and its validation. Section III is devoted to the work done by Jianguo Yu at INL. Finally, we will present the performance of the project in terms of milestones, publications, and presentations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Uranium Mining and Milling

The starting point for material accountancy under INFCIRC/153-type comprehensive safeguards agreements (paragraph 34(c)) with an Additional Protocol envisions a shifting toward earlier in the nuclear fuel cycle than current practice. The design of effective and efficient safeguards requires a more detailed process-level understanding of front-end activities, to include conducting a survey of uranium mining, concentration, and purification processes in commercial use to catalog technological options for producing uranium ore concentrate, also known as yellow cake. Based on this survey, a process flow diagram (PFD) illustrating front-end process options will be created to illustrate pre-34(c) material flows and support the identification of material control points.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Alternate Design Concepts for the High Flux Isotope Reactor using Low-Enriched Uranium Fuel Systems: Scoping Analysis

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a multi-mission research reactor dedicated mainly to isotope production, neutron scattering experiments, material irradiation, testing and activation. In the framework of non-proliferation policies, the international community is currently aiming to reduce or even eliminate, when and where possible, the use of Highly Enriched Uranium (HEU: 235 U/U ≥ 20 wt. %) in civilian applications. In that context, most research reactors worldwide still operating with HEU fuel are actively engaged in an effort to convert to Low Enriched Uranium fuel system (LEU: 235 U/U < 20 wt. %). Within the US, the High Performance Research Reactor (USHPRR) fleet (which includes HFIR) is expected to convert to LEU fuel using the UMo “monolithic” fuel system currently under development. ORNL is responsible for the development of a LEU design for HFIR. In the past years, Argonne National Laboratory (ANL) offered technical assistance to the ORNL LEU design team whenever deemed appropriate. Due to their complex nature, acceptable HFIR LEU designs identified so far, all based on the UMo monolithic fuel system, would require a substantial fabrication development effort before being proven commercially viable. In order to mitigate technical risks and reduce cost/schedule uncertainties, it appears appropriate to study alternate design solutions and alternate fuel systems as possible backup options for the HFIR conversion to LEU fuel. In that context, ANL has been exploring alternate design concepts for HFIR. Complementing the ORNL design activities, the goal is to find the best design solution for HFIR in term of cost, performance and safety. The present report summarizes the findings regarding two types of alternate conceptual design: U 3 Si 2 designs and un-contoured UMo design. Based on available tools and methods, preliminary analysis show that both concepts could probably meet performance and steady-state safety requirements but element fabrication would remain complex and would require a non-trivial fabrication development effort.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Safety Analysis of the Mo-99 Production Upgrade to the University of Missouri Research Reactor (MURR) with Highly-Enriched and Low-Enriched Uranium Fuel

The University of Missouri has been working in conjunction with Argonne National Laboratory (ANL) in the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Reactor Conversion Program to support conversion of the University of Missouri- Columbia Research Reactor (MURR®) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. MURR is one of five U.S. High Performance Research Reactors (USHPRR) plus a critical facility that plans to convert to the use of LEU fuel. The purpose of this report is to document the safety margins for both HEU and LEU cores with the Mo-99 production upgrade proposed, designated as the 2017-RBM-99 device in this report. In addition to the 2017-RBM-99 device, the neutronics models also include descriptions of the experimental facilities and experiment loading that are currently typical for MURR and which are different from the experiments that were modeled in the analysis performed for the PSAR. Lastly, there have been changes to the LEU fuel element design with regard to the coolant channel thickness tolerances for the innermost and outermost coolant channels based on feedback from the fuel fabricator. This design change did not change the nominal dimensions of the coolant channels/fuel plates or fuel loading of the fuel plates in the element. Thus, for the purpose of the neutronics analysis for the 2017-RBM-99 upgrade, which is based on nominal dimensions and fuel loadings, the components of the LEU element are modeled the same as in the PSAR analysis. However, the design tolerance changes were incorporated in the calculations of the steady-state safety margins for both HEU and LEU. While the use of fewer significant figures has simplified fabrication, yet there were no negative impacts, including on safety margins, since the end channels were not limiting locations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

M3MS-19IN0101201 Modeling Intra-granular Fission Gas Bubble Evolution in Uranium Dioxide during Transients

The SciDAC milestone M3MS-19IN0101201 on Modeling Intra-granular Fission Gas Bubble Evolution in Uranium Dioxide during Transients was completed with the development, testing and implementation in the Bison code of an engineering model for intra-granular fission gas bubble evolution that is applicable to transient conditions. The model extends the previous work on modeling bubble evolution during normal operating conditions to include additional complexity that arises during transient situations, in particular, the so-called intra-granular bubble coarsening. This phenomenon refers to the appearance of a second population of coarsened bubbles with diameters of tens to hundreds of nm, which co-exist with nanometric bubbles present also during normal operating conditions and are responsible for a large portion of transient fuel swelling. The developed model considers growth of bubbles along dislocations, associated with the absorption of vacancies available in the dislocation core region, and favored by pipe diffusion of fission gas atoms. The model also utilizes the formulation for the gas atom resolution rate developed in the present SciDAC project through molecular dynamics simulations by other institutions. The present work therefore demonstrates an operational multiscale approach, based on collaboration within the project. The model is validated to an extensive experimental dataset of ramp tested uranium dioxide fuel samples. Additionally, we present new applications of the base model for bubble evolution during normal operating conditions to Bison engineering simulations of Cr2O3-doped fuel rods irradiated in the Halden reactor. Simulation results in terms of integral fission gas release in the fuel rods are compared to the available experimental data, pointing out significant improvements in the engineering predictions compared to application of the empirical model that was available in Bison before the SciDAC developments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Conversion of Uranium Metal to Uranyl Sulfate Solution

Experiments involving conversion of uranium (U) metal to U 3 O 8 , followed by dissolution to form uranyl sulfate solution, were performed with batches of up to 300 g of U metal. The focus was to identify major U species formed during the oxidation process, determine the particle size distribution of the uranium oxide product, and identify the size of the dissolution vessel needed to accommodate the exothermic reaction during the conversion of U 3 O 8 to uranyl sulfate using 30% hydrogen peroxide and H 2 SO 4 . The information obtained in this study will be used for designing full-scale processes to convert kilogram quantities of U metal to uranyl sulfate solution with a concentration of 300 g U/L.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Characterization of Uranium Foil Irradiations at the WSU TRIGA Reactor Using a New Reactor Model in SCALE [Slides]

A depleted uranium foil and a highly enriched uranium foil that were irradiated at the Washington State University TRIGA reactor were used to measure the relative production of 237 U and 239 U, two key isotopes of interest in nuclear forensics, as well as their decay daughters 237 Np and 239 Pu. Total fissions in the two foils were also estimated using 99 Mo measurement. A reactor model of the TRIGA was built in the SCALE neutron transport code using dimensions obtained from an existing MCNP geometry. The model was successfully used to simulate foil fissions and the reaction rate ratios in 238 U, using the laboratory-measured isotope ratios as a benchmark. The flux distribution in the foils was determined. Finally, using the continuous energy Tsunami-3d module of SCALE, a sensitivity/uncertainty analysis was performed to determine the effects on the reaction ratio caused by other reactions in the reactor and by uncertainties in the known cross-sections.

07 ISOTOPE AND RADIATION SOURCES↗

Small-Scale Recycling of Irradiated Uranium and Transuranic Elements (Np, Pu) Using 3D-Printed Centrifugal Contactors

Niowave’s technology employs super-conducting linear accelerators to induce fission on low-enriched uranium targets followed by subsequent radiochemical processing to purify 99 Mo as well as a number of other fission products. Argonne’s Radiochemistry group (CFCT), in partnership with Niowave and UNLV, will be managing the TCF project to develop a closed-cycle loop for Niowave’s uranium targets. The roles of the DOE national laboratory partner will be to 1) develop a basic chemical understanding of the separations and purifications required to meet industry standards, 2) initiate the additive manufacturing (AM) of the centrifugal contactors, 3) develop a process flowsheet using Argonne computer codes.

43 PARTICLE ACCELERATORS↗

Biogenic uranium isotope fractionation [Slides]

Presentation Outline: Background and relevance; Project objectives; Tasks and timeline; Current Activities (Task 1: Controlled culture growth, Task 2: Enzymatic reduction of uranium oxides, Task 3: Cellular location of uranium reduction and precipitation); Future work; Achievements

07 ISOTOPE AND RADIATION SOURCES↗

MITR Low-Enriched Uranium Conversion Fluid-Structure Interaction Preliminary Design Verification

The U.S. National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M 3 ) has developed and is pursuing an integrated approach to address the persistent threat posed by unintentional proliferation of nuclear materials. The NNSA M 3 approach reduces the risk of highly enriched uranium (HEU) and plutonium falling into the hands of non-state actors by minimizing the use of and, when possible, eliminating weapons-usable nuclear material around the world. In this geopolitical context, most research and test reactors, both domestic and international, have completed or started a program of conversion from the use of HEU to low-enriched uranium (LEU) as fuel. Conversion of civilian research reactors from HEU to LEU, and the return of the HEU to the country of origin, are important components of the NNSA non-proliferation program. Worldwide, 71 reactors have been converted to the use of LEU fuel, and an additional 32 have been confirmed to be permanently shut down. The U.S., with 20 reactor conversions, is among the 39 countries on six continents where conversions have occurred. With recent conversions in Ghana and Nigeria, an important milestone was reached in completing conversion of all reactors on the continent of Africa to LEU fuel. Africa thus becomes the third continent to have completed conversion of all HEU reactors to LEU, following Australia and South America

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

US-DOE and Republic of South Africa Uranium Benchmarking: Results From Los Alamos National Laboratory (Study Report)

This report details work performed at Los Alamos National Laboratory (LANL) in support of an interlaboratory uranium (U) benchmarking collaboration between LANL, Lawrence Livermore National Laboratory (LLNL), and South African Nuclear Energy Corporation (Necsa). The goal of this interlaboratory benchmarking exercise is to compare analytical methods and data obtained on uranium reference materials CUP-2 and New Brunswick Laboratory (NBL) certified reference material (CRM) 124-2. The LANL commitments for this project include performing microwave digestion of fresh CUP-2 and CRM 124-2 powders, measuring the U isotope composition of CUP-2 and CRM 124-2 by both single-collector and multi-collector inductively coupled plasma mass spectrometry (SC-ICP-MS and MC-ICP-MS, respectively), and determining trace-element abundances in CUP-2 and CRM 124-2 by inductively coupled plasma mass spectrometry (ICP-MS). This report details the methods employed by LANL in support of these objectives and presents the data obtained during this collaborative exercise.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Tensile Property and Lifetime Prediction for Low-Temperature Aged Uranium-Niobium Alloys

Thermal aging models and lifetime predictions for uranium-niobium (U-Nb) alloys were created using an approach similar to those previously employed. Lifetime estimates for generic U-6Nb components were thus updated; the reported value being 800 years. This update represents a small change in lifetime vs. that of the 2012 assessment (540 years). This lifetime estimate emerged from consideration of several model fits specific to the aging datasets and properties chosen. Aging was quantified using quasi-static tensile properties measured on specimens artificially aged for up to 10 years. The major change relative to the most recent 2012 LANL assessment was that a more comprehensive body of U-Nb literature data was mined, in addition to being augmented by the latest LANL and UK AWE data. The tensile data compilation was published separately (LANL report LA-14493, December 2016). Recognizing the chemical banding of industrially produced U-6Nb, models were developed for the mid-range (6 wt.%) and extrema (4 and 8 wt.%) compositions. Lifetime estimates were calculated for all three alloy classes (4, 6, 8 wt.% nominal) and two measures of total tensile elongation (TE) to failure, namely TE-ext. — extensometer method, and TE-NCD — normalized crosshead displacement method. The conservative assumption was made that whichever composition (4, 6, or 8 wt.% Nb) and property (TE-ext or TE-NCD) was the first to cross the ductility failure threshold would limit the lifetime of the entire component. Tensile strength properties did not figure into the lifetime predictions, but could be useful as age-sensitive diagnostics and were also modeled. Of these, only first yield strength is expected to show a change at 40°C aging vs. time = 0 over the ~100-year timespan of engineering interest. Second yield strength evolves more slowly, and ultimate tensile strength slower still. Among all the models, the apparent activation energies for aging were mostly in the narrow 29– 37 kcal/mol range, which is close to that for diffusion of Nb in gamma-uranium. This agreement may be coincidental. The data from recent long-term aging studies substantially improved the model fit quality and robustness of the lifetimes. Appendices document sensitivity studies of the model fits and lifetimes with respect to using more limited datasets. These results highlight the limitations of relying solely on data from scattered literature studies and smaller datasets more generally.

36 MATERIALS SCIENCE↗