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DOE OSTI2012

Creation of a Full-Core HTR Benchmark with the Fort St. Vrain Initial Core and Assessment of Uncertainties in the FSV Fuel Composition and Geometry

Information and measured data from the intial Fort St. Vrain (FSV) high temperature gas reactor core is used to develop a benchmark configuration to validate computational methods for analysis of a full-core, commercial HTR configuration. Large uncertainties in the geometry and composition data for the FSV fuel and core are identified, including: (1) the relative numbers of fuel particles for the four particle types, (2) the distribution of fuel kernel diameters for the four particle types, (3) the Th:U ratio in the initial FSV core, (4) and the buffer thickness for the fissile and fertile particles. Sensitivity studies were performed to assess each of these uncertainties. A number of methods were developed to assist in these studies, including: (1) the automation of MCNP5 input files for FSV using Python scripts, (2) a simple method to verify isotopic loadings in MCNP5 input files, (3) an automated procedure to conduct a coupled MCNP5-RELAP5 analysis for a full-core FSV configuration with thermal-hydraulic feedback, and (4) a methodology for sampling kernel diameters from arbitrary power law and Gaussian PDFs that preserved fuel loading and packing factor constraints. A reference FSV fuel configuration was developed based on having a single diameter kernel for each of the four particle types, preserving known uranium and thorium loadings and packing factor (58%). Three fuel models were developed, based on representing the fuel as a mixture of kernels with two diameters, four diameters, or a continuous range of diameters. The fuel particles were put into a fuel compact using either a lattice-bsed approach or a stochastic packing methodology from RPI, and simulated with MCNP5. The results of the sensitivity studies indicated that the uncertainties in the relative numbers and sizes of fissile and fertile kernels were not important nor were the distributions of kernel diameters within their diameter ranges. The uncertainty in the Th:U ratio in the intial FSV core was found to be important with a crude study. The uncertainty in the TRISO buffer thickness was estimated to be unimportant but the study was not conclusive. FSV fuel compacts and a regular FSV fuel element were analyzed with MCNP5 and compared with predictions using a modified version of HELIOS that is capable of analyzing TRISO fuel configurations. The HELIOS analyses were performed by SSP. The eigenvalue discrepancies between HELIOS and MCNP5 are currently on the order of 1% but these are still being evaluated. Full-core FSV configurations were developed for two initial critical configurations - a cold, clean critical loading and a critical configuration at 70% power. MCNP5 predictions are compared to experimental data and the results are mixed. Analyses were also done for the pulsed neutron experiments that were conducted by GA for the initial FSV core. MCNP5 was used to model these experiments and reasonable agreement with measured results has been observed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2012

Chemical Differences Between Sludge Solids at the F and H Area Tank Farms

The primary source of waste solids received into the F Area Tank Farm (FTF) was from PUREX processing performed to recover uranium and plutonium from irradiated depleted uranium targets. In contrast, two primary sources of waste solids were received into the H Area Tank Farm (HTF): a) waste from PUREX processing; and b) waste from H-modified (HM) processing performed to recover uranium and neptunium from burned enriched uranium fuel. Due to the differences between the irradiated depleted uranium targets and the burned enriched uranium fuel, the average compositions of the F and H Area wastes are markedly different from one another. Both F and H Area wastes contain significant amounts of iron and aluminum compounds. However, because the iron content of PUREX waste is higher than that of HM waste, and the aluminum content of PUREX waste is lower than that of HM waste, the iron to aluminum ratios of typical FTF waste solids are appreciably higher than those of typical HTF waste solids. Other constituents present at significantly higher concentrations in the typical FTF waste solids include uranium, nickel, ruthenium, zinc, silver, cobalt and copper. In contrast, constituents present at significantly higher concentrations in the typical HTF waste solids include mercury, thorium, oxalate, and radionuclides U-233, U-234, U-235, U-236, Pu-238, Pu-242, Cm-244, and Cm-245. Because of the higher concentrations of Pu-238 in HTF, the long-term concentrations of Th-230 and Ra-226 (from Pu-238 decay) will also be higher in HTF. The uranium and plutonium distributions of the average FTF waste were found to be consistent with depleted uranium and weapons grade plutonium, respectively (U-235 comprised 0.3 wt% of the FTF uranium, and Pu-240 comprised 6 wt% of the FTF plutonium). In contrast, at HTF, U-235 comprised 5 wt% of the uranium, and Pu-240 comprised 17 wt% of the plutonium, consistent with enriched uranium and high burn-up plutonium. X-ray diffraction analyses of various FTF and HTF samples indicated that the primary crystalline compounds of iron in sludge solids are Fe 2 O 3 , Fe 3 O 4 , and FeO(OH), and the primary crystalline compounds of aluminum are Al(OH) 3 and AlO(OH). Also identified were carbonate compounds of calcium, magnesium, and sodium; a nitrated sodium aluminosilicate; and various uranium compounds. Consistent with expectations, oxalate compounds were identified in solids associated with oxalic acid cleaning operations. The most likely oxidation states and chemical forms of technetium are assessed in the context of solubility, since technetium-99 is a key risk driver from an environmental fate and transport perspective. The primary oxidation state of technetium in SRS sludge solids is expected to be Tc(IV). In salt waste, the primary oxidation state is expected to be Tc(VII). The primary form of technetium in sludge is expected to be a hydrated technetium dioxide, TcO 2 ⋅ xH 2 O, which is relatively insoluble and likely co-precipitated with iron. In salt waste solutions, the primary form of technetium is expected to be the very soluble pertechnetate anion, TcO 4 - . The relative differences between the F and H Tank Farm waste provide a basis for anticipating differences that will occur as constituents of FTF and HTF waste residue enter the environment over the long-term future. If a constituent is significantly more dominant in one of the Tank Farms, its long-term environmental contribution will likely be commensurately higher, assuming the environmental transport conditions of the two Tank Farms share some commonality. It is in this vein that the information cited in this document is provided - for use during the generation, assessment, and validation of Performance Assessment modeling results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

MPACT LSDS Progress Report: LANL and LANL-Funded Activities in Support of Spent Fuel Measurements Using a Lead Slowing Down Spectrometer in FY 2012

During FY 2012, Lead Slowing Down Spectrometer (LSDS) measurement campaigns took place at the Los Alamos Neutron Science Center, and the Rensselaer Polytechnic Institute. At Los Alamos, we measured the response of a thorium fission chamber to a 25 gram plutonium sample, and evaluated the performance of a 4 He-filled proportional counter. At RPI, the effort lead by Professor Danon, performed an assay of fuel pins, and used a fission chamber with plutonium to evaluate the impact of room return.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2012

Actinide Foil Production for MPACT Research

Sensitive fast-neutron detectors are required for use in lead slowing down spectrometry (LSDS), an active interrogation technique for used nuclear fuel assay for Materials Protection, Accounting, and Controls Technologies (MPACT). During the past several years UNLV sponsored a research project at RPI to investigate LSDS; began development of fission chamber detectors for use in LSDS experiments in collaboration with INL, LANL, and Oregon State U.; and participated in a LSDS experiment at LANL. In the LSDS technique, research has demonstrated that these fission chamber detectors must be sensitive to fission energy neutrons but insensitive to thermal-energy neutrons. Because most systems are highly sensitive to large thermal neutron populations due to the well-known large thermal cross section of 235U, even a miniscule amount of this isotope in a fission chamber will overwhelm the small population of higher-energy neutrons. Thus, fast-fission chamber detectors must be fabricated with highly depleted uranium (DU) or ultra-pure thorium (Th), which is about half as efficient as DU. Previous research conducted at RPI demonstrated that the required purity of DU for assay of used nuclear fuel using LSDS is less than 4 ppm 235U, material that until recently was not available in the U.S. In 2009 the PI purchased 3 grams of ultra-depleted uranium (uDU, 99.99998% 238U with just 0.2 ± 0.1 ppm 235U) from VNIIEF in Sarov, Russia. We received the material in the form of U3O8 powder in August of 2009, and verified its purity and depletion in a FY10 MPACT collaboration project. In addition, chemical processing for use in FC R&D was initiated, fission chamber detectors and a scanning alpha-particle spectrometer were developed, and foils were used in a preliminary LSDS experiment at a LANL/LANSCE in Sept. of 2010. The as-received U3O8 powder must be chemically processed to convert it to another chemical form while maintaining its purity, which then must be used to electro-deposit U or UO2 in extremely thin layers (1 to 2 mg/cm2) on various media such as films, foils, or discs. After many months of investigation and trials in FY10 and 11, UNLV researchers developed a new method to produce pure UO2 deposits on foils using a unique approach, which has never been demonstrated, that involves dissolution of U3O8 directly into room temperature ionic liquid (RTIL) followed by electrodeposition from the RTIL-uDU solution (Th deposition from RTIL had been previously demonstrated). The high-purity dissolution of the U3O8 permits the use of RTIL solutions for deposition of U on metal foils in layers without introducing contamination. In FY10 and early FY11 a natural U surrogate for the uDU was used to investigate this and other techniques. In this research project UNLV will deposit directly from RTIL to produce uDU and Th foils devoid of possible contaminants. After these layers have been deposited, they will be examined for purity and uniformity. UNLV will complete the development and demonstration of the RTIL technology/ methodology to prepare uDU and Th samples for use in constructing fast-neutron detectors. Although this material was purchased for use in research using fast-fission chamber detectors for active inspection techniques for MPACT, it could also contribute to R&D for other applications, such as cross section measurements or neutron spectroscopy for national security

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2012

Flowsheet Evaluation for the Dissolving and Neutralization of Sodium Reactor Experiment Used Nuclear Fuel

This report includes the literature review, hydrogen off-gas calculations, and hydrogen generation tests to determine that H-Canyon can safely dissolve the Sodium Reactor Experiment (SRE; thorium fuel), Ford Nuclear Reactor (FNR; aluminum alloy fuel), and Denmark Reactor (DR-3; silicide fuel, aluminum alloy fuel, and aluminum oxide fuel) assemblies in the L-Bundles with respect to the hydrogen levels in the projected peak off-gas rates. This is provided that the number of L-Bundles charged to the dissolver is controlled. Examination of SRE dissolution for potential issues has aided in predicting the optimal batching scenario. The calculations detailed in this report demonstrate that the FNR, SRE, and DR-3 used nuclear fuel (UNF) are bounded by MURR UNF and may be charged using the controls outlined for MURR dissolution in a prior report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

Experimentally characterizing the electronic structures of f-electron systems using advanced high resolution Fourier transform microwave spectroscopies

We aim to (i) provide data that directly addresses the fundamental roles of actinide valence electrons in chemical bonding, and (ii) serve to provide prototypical data for the heavy element computational chemistry community. These goals will be achieved through the first pure rotational spectroscopic measurements on prototypical systems at ultra-high resolution. These systems encompass low oxidation state uranium and thorium compounds including, but not limited to, UX and ThX, X = F, Cl, Br, I, and UY and ThY, Y = O, S, and other simple U and Th-containing compounds. Our primary experimental tools involve time-domain rotational spectroscopy achieving line widths and resolutions of a few kHz.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2013

Special Analysis for the Disposal of the Consolidated Edison Uranium Solidification Project Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

The purpose of this Special Analysis (SA) is to determine if the Oak Ridge (OR) Consolidated Edison Uranium Solidification Project (CEUSP) uranium-233( 233 U) waste stream (DRTK000000050, Revision 0) is acceptable for shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS) on the Nevada National Security Site (NNSS). The CEUSP 233 U waste stream requires a special analysis because the concentrations of thorium- 229 ( 229 Th), 230 Th, 232 U, 233 U, and 234 U exceeded their NNSS Waste Acceptance Criteria action levels. The acceptability of the waste stream is evaluated by determining if performance assessment (PA) modeling provides a reasonable expectation that SLB disposal is protective of human health and the environment. The CEUSP 233 U waste stream is a long-lived waste with unique radiological hazards. The SA evaluates the long-term acceptability of the CEUSP 233 U waste stream for near-surface disposal as a two tier process. The first tier, which is the usual SA process, uses the approved probabilistic PA model to determine if there is a reasonable expectation that disposal of the CEUSP 233 U waste stream can meet the performance objectives of U.S. Department of Energy Manual DOE M 435.1-1, “Radioactive Waste Management,” for a period of 1,000 years (y) after closure. The second tier addresses the acceptability of the OR CEUSP 233 U waste stream for near-surface disposal by evaluating long-term site stability and security, by performing extended (i.e., 10,000 and 60,000 y) modeling analyses, and by evaluating the effect of containers and the depth of burial on performance. Tier I results indicate that there is a reasonable expectation of compliance with all performance objectives if the OR CEUSP 233 U waste stream is disposed in the Area 5 RWMS SLB disposal units. The maximum mean and 95 th percentile PA results are all less than the performance objective for 1,000 y. Monte Carlo uncertainty analysis indicates that there is a high likelihood of compliance with all performance objectives. Tier II results indicate that the long-term performance of the OR CEUSP 233 U waste stream is protective of human health and the environment. The Area 5 RWMS is located in one of the least populated and most arid regions of the U.S. Site characterization data indicate that infiltration of precipitation below the plant root zone at 2.5 meters (8.2 feet) ceased 10,000 to 15,000 y ago. The site is not expected to have a groundwater pathway as long as the current arid climate persists. The national security mission of the NNSS and the location of the Area 5 RWMS within the Frenchman Flat Corrective Action Unit require that access controls and land use restrictions be maintained indefinitely. PA modeling results for 10,000 to 60,000 y also indicate that the OR CEUSP 233 U waste stream is acceptable for near-surface disposal. The mean resident air pathway annual total effective dose (TED), the resident all-pathways annual TED, and the acute drilling TED are less than their performance objectives for 10,000 y after closure. The mean radon-222 (222Rn) flux density exceeds the performance objective at 4,200 y, but this is due to waste already disposed at the Area 5 RWMS and is only slightly affected by disposal of the CEUSP 233 U. The peak resident all-pathways annual TED from CEUSP key radionuclides occurs at 48,000 y and is less than the 0.25 millisievert performance objective. Disposal of the OR CEUSP 233 U waste stream in a typical SLB trench slightly increases PA results. Increasing the depth was found to eliminate any impacts of the OR CEUSP 233 U waste stream. Containers could not be shown to have any significant impact on performance due to the long half-life of the waste stream and a lack of data for pitting corrosion rates of stainless steel in soil. The results of the SA indicate that all performance objectives can be met with disposal of the OR CEUSP 233 U waste stream in the SLB units at the Area 5 RWMS. The long-term performance of the OR CEUSP 233 U waste stream disposed in the near surface is protective of human health and the environment. The waste stream is recommended for disposal without conditions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

COMPARISON OF RESULTS FOR QUARTER 4 SURFACE WATER SPLIT SAMPLES COLLECTED AT THE NUCLEAR FUELS SERVICES SITE, ERWIN, TN

Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, collected split surface water samples with Nuclear Fuel Services (NFS) representatives on June 12, 2013. Representatives from the U.S. Nuclear Regulatory Commission (NRC) and the Tennessee Department of Environment and Conservation were also in attendance. Samples were collected at four surface water stations, as required in the approved Request for Technical Assistance number 11-018. These stations included Nolichucky River upstream (NRU), Nolichucky River downstream (NRD), Martin Creek upstream (MCU), and Martin Creek downstream (MCD). Both ORAU and NFS performed gross alpha and gross beta analyses, and Table 1 presents the comparison of results using the duplicate error ratio (DER), also known as the normalized absolute difference. A DER ≤ 3 indicates at a 99% confidence interval that split sample results do not differ significantly when compared to their respective one standard deviation (sigma) uncertainty (ANSI N42.22). The NFS split sample report specifies 95% confidence level of reported uncertainties (NFS 2013). Therefore, standard two sigma reporting values were divided by 1.96. In conclusion, most DER values were less than 3 and results are consistent with low (e.g., background) concentrations. The gross beta result for sample 5198W0014 was the exception. The ORAU gross beta result of 6.30 &plusmn 0.65 pCi/L from location NRD is well above NFS's non-detected result of 1.56 &plusmn 0.59 pCi/L. NFS's data package includes no detected result for any radionuclide at location NRD. At NRC's request, ORAU performed gamma spectroscopic analysis of sample 5198W0014 to identify analytes contributing to the relatively elevated gross beta results. This analysis identified detected amounts of naturally-occurring constituents, most notably Ac-228 from the thorium decay series, and does not suggest the presence of site-related contamination.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2013

Letter Report for Analytical Results for Two Soil Samples Associated with the Westinghouse Hematite Decommisioning Project in Hematite Missouri

Oak Ridge Associated Universities (ORAU), under the Oak Ridge Institute for Science and Education (ORISE) contract, received two soil samples on September 26, 2013 from the Westinghouse Hetnatite Decomminsioning project in Hematite, Missouri. The samples were analyzed for thorium-232, radium-226, uranium-235, and uranium-238 by gamma spectrometry and technetium-99 by liquid scintillation analysis. The samples were received in good condition. The sample collection data and identification numbers are tabulated. Also presented are the gamma spectrometry and technetium-99 data, respectively. The pertinent procedure references are included with the data tables.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2013

Project X: Accelerator Reference Design

Project X is a high-intensity proton facility being developed to support a world-leading program of Intensity Frontier physics over the next two decades at Fermilab. Project X is an integral part of the U.S. Intensity Frontier Roadmap as described in the P5 report of May 2008 [1] and within the Fermilab Strategic Plan of November 2011 [2]. This document represents Part I of the “Project X Book” describing the Project X accelerator facility and the broad range of physics research opportunities enabled by Project X. Parts II and III provide in-depth descriptions of the physics research program, both within and beyond particle physics [3]. The primary elements of the U.S. program to be supported by Project X include: Neutrino Experiments: Experimental studies of neutrino oscillations and neutrino interaction physics with ultra-intense neutrino beams provided by a high-power proton source with energies up to 120 GeV, utilizing near detectors at the Fermilab site and massive detectors at distant underground laboratories. Goal: At least 2 MW of proton beam power at any energy between 60 to 120 GeV; several hundred kW of proton beam power on target at 8 GeV. Kaon, Muon, Nucleon, and Neutron Precision Experiments: World-leading experiments studying ultra-rare kaon decays, searching for muon-to-electron conversion and nuclear electron dipole moments (EDMs), and exploring neutron properties at very high precision. Goal: MW-class proton beams supporting multiple experiments at 1 and 3 GeV, with flexible capability for providing distinct beam formats to concurrent users while allowing simultaneous operations with the neutrino program. Material Science and Nuclear Energy Applications: High-intensity accelerator, spallation, target and transmutation technology demonstrations will provide critical input into the design of future energy systems, including next generation fission reactors, nuclear waste transmutation systems and future thorium fuel-cycle power systems. Possible applications of muon spin rotation techniques provide sensitive probes of the magnetic structure of materials. Goal: Provide MW-class proton beams at 1 GeV, coupled with novel targets required to support a broad range of materials science and energy applications. Platform for Evolution to Future Frontier Facilities: A high-intensity proton source will strengthen and modernize the Fermilab injector complex, providing a robust platform upon which to build future frontier facilities. The Neutrino Factory and Muon Collider are examples that would provide world-leading capabilities at the Intensity and Energy Frontiers for many decades to come. Goal: Provide a straightforward upgrade path for a 4 MW, low-duty-factor source of protons at energies between 5 and 15 GeV. These four elements are expected to form the basis of the Mission Need statement required for the Department of Energy (DOE) Critical Decision 0 (CD-0), and represent the fundamental design criteria for Project X. The following chapters present the Reference Design for the Project X accelerator facility. The Reference Design is based on a continuous wave (CW) superconducting (SC) linac providing up to 1 and 3 MW of beam power at 1 and 3 GeV respectively. A pulsed linac provides acceleration of roughly 4% of the beam delivered from the CW linac to the 8 GeV injection energy of the existing Recycler/Main Injector complex. Upgrades to the Recycler and Main Injector support a factor of three increase, beyond current capabilities, in proton beam power at 60 to 120 GeV. The Reference Design represents a facility that will be unique in the world with unmatched capabilities for the delivery of very high beam power, with flexible beam formats, to multiple users at multiple energies. The utilization of linacs within the Project X facility enables capabilities beyond what is achievable with circular accelerators. It is anticipated that the final configuration and operating parameters of the complex will be further refined through the R&D program in advance of CD-2.

Holmes, Stephen D.↗

DOE OSTI2014

Matrix Infrared Spectroscopic and Computational Investigations of Novel Small Uranium Containing Molecules - Final Technical Report

Direct reactions of f-element uranium, thorium and lanthanide metal atoms were investigated with small molecules. These metal atoms were generated by laser ablation and mixed with the reagent molecules then condensed with noble gases at 4K. The products were analyzed by absorption of infrared light to measure vibrational frequencies which were confirmed by quantum chemical calculations. We have learned more about the reactivity of uranium atoms with common molecules, which will aid in the develolpment of further applications of uranium.

Andrews, Lester↗

DOE OSTI2014

Special Analysis for the Disposal of the Idaho National Laboratory Unirradiated Light Water Breeder Reactor Rods and Pellets Waste Stream at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada

The purpose of this special analysis (SA) is to determine if the Idaho National Laboratory (INL) Unirradiated Light Water Breeder Reactor (LWBR) Rods and Pellets waste stream (INEL103597TR2, Revision 2) is suitable for disposal by shallow land burial (SLB) at the Area 5 Radioactive Waste Management Site (RWMS). The INL Unirradiated LWBR Rods and Pellets waste stream consists of 24 containers with unirradiated fabricated rods and pellets composed of uranium oxide (UO2) and thorium oxide (ThO2) fuel in zirconium cladding. The INL Unirradiated LWBR Rods and Pellets waste stream requires an SA because the 229 Th, 230 Th, 232 U, 233 U, and 234 U activity concentrations exceed the Nevada National Security Site (NNSS) Waste Acceptance Criteria (WAC) Action Levels.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2015

Flowsheet Evaluation for the Dissolving and Neutralization of Sodium Reactor Experiment Used Nuclear Fuel

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Nuclear forensics: Soil content

Nuclear Forensics is a growing field that is concerned with all stages of the process of creating and detonating a nuclear weapon. The main goal is to prevent nuclear attack by locating and securing nuclear material before it can be used in an aggressive manner. This stage of the process is mostly paperwork; laws, regulations, treaties, and declarations made by individual countries or by the UN Security Council. There is some preliminary leg work done in the form of field testing detection equipment and tracking down orphan materials; however, none of these have yielded any spectacular or useful results. In the event of a nuclear attack, the first step is to analyze the post detonation debris to aid in the identification of the responsible party. This aspect of the nuclear forensics process, while reactive in nature, is more scientific. A rock sample taken from the detonation site can be dissolved into liquid form and analyzed to determine its chemical composition. The chemical analysis of spent nuclear material can provide valuable information if properly processed and analyzed. In order to accurately evaluate the results, scientists require information on the natural occurring elements in the detonation zone. From this information, scientists can determine what percentage of the element originated in the bomb itself rather than the environment. To this end, element concentrations in soils from sixty-nine different cities are given, along with activity concentrations for uranium, thorium, potassium, and radium in various building materials. These data are used in the analysis program Python.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2015

The thermal conductivity of mixed fuel U x Pu 1-x O 2 : molecular dynamics simulations

Mixed oxides (MOX), in the context of nuclear fuels, are a mixture of the oxides of heavy actinide elements such as uranium, plutonium and thorium. The interest in the UO 2 -PuO 2 system arises from the fact that these oxides are used both in fast breeder reactors (FBRs) as well as in pressurized water reactors (PWRs). The thermal conductivity of UO 2 fuel is an important material property that affects fuel performance since it is the key parameter determining the temperature distribution in the fuel, thus governing, e.g., dimensional changes due to thermal expansion, fission gas release rates, etc. For this reason it is important to understand the thermal conductivity of MOX fuel and how it differs from UO 2 . Here, molecular dynamics (MD) simulations are carried out to determine quantitatively, the effect of mixing on the thermal conductivity of U x Pu 1-x O 2 , as a function of PuO 2 concentrations, for a range of temperatures, 300 – 1500 K. The results will be used to develop enhanced continuum thermal conductivity models for MARMOT and BISON by INL. These models express the thermal conductivity as a function of microstructure state-variables, thus enabling thermal conductivity models with closer connection to the physical state of the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Final Technical Report of Project DE-FG02-96ER14647

This is the final technical report of work completed under DOE support over the period Sept. 1, 1996 until May 31, 2015. The title of the project was "Ion/Excited Atom Collision Studies with a Rydberg Target and a CO2 Laser" from 9/1/96 to 10/31/06, and "Properties of Actinide Ions from Measurements of Rydberg Ion Fine Structure" from 11/1/06 until 5/31/15. The primary technical results were a detailed experimental study of resonant charge transfer between Rydberg atoms and highly-charged ions, and unique measurements of many properties of multiply-charged Thorium ions.

74 ATOMIC AND MOLECULAR PHYSICS↗

DOE OSTI2015

Modeling Background Radiation Using Geochemical Data

Aerial gamma ray surveys have many applications in geology and science in general, such as locating mining prospects, defining radioactive plumes, and detecting nuclear weapons. Unfortunately there is currently no simple way to separate the natural gamma radiation of soil and rocks from that of contaminants such as radioactive plumes. This project used geochemical data (uranium, potassium and thorium concentrations) collected from national databases, private companies, and the NURE (National Uranium Resource Evaluation) Survey, to create forward models of exposure rates measured by aerial gamma ray surveys. We developed these techniques using an area in north central Arizona known as the Navajo Mines area, chosen for its optimal conditions for aerial gamma ray surveys and readily available survey data. Models based on geochemical analyses from databases were not found to be successful, in part due to lack of data for some units. Models based on NURE data sorted by geologic unit were successful at replicating the aerial gamma ray survey, though units with Uranium mineralization proved difficult. ASTER visualizations were effectively used to create subunits of similar exposure rate within the Chinle Formation which contains multiple lithologies. For alluvial units, models based on drainage basin were attempted with success. With these models we are developing techniques to estimate the natural radiation generated by the rocks and soils of an area, making aerial ray surveys more effective.

58 GEOSCIENCES↗

DOE OSTI2016

MST e-News June 2016

This is Los Alamos National Laboratory's (LANL) June 2016 newsletter of the Materials Science and Technology Division. The following are major topics in this newsletter: MST-8 scientists guide national efforts to overcome nuclear energy technical challenges, first-ever scanning probe microscopy capabilities for plutonium, laboratory metallurgists make thorium targets for production of cancer-fighting isotopes, and a spotlight on Veronica Livescu.

99 GENERAL AND MISCELLANEOUS↗