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At least 487 records · Page 27
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Production of High-purity Radium-223 from Legacy Actinium-Beryllium Neutron Sources

Radium-223 is a short-lived alpha-particle-emitting radionuclides with potential applications in cancer treatment. Research to develop new radiopharmaceuticals employing 223Ra has been hindered by poor availability due to the small quantities of parent actinium-227 available world-wide. The purpose of this study was to develop innovative and cost-effective methods to obtain high-purity 223Ra from 227Ac. We obtained 227Ac from two surplus actinium-beryllium neutron generators. We retrieved the actinium/beryllium buttons from the sources and dissolved them in a sulfuric-nitric acid solution. A crude actinium solid was recovered from the solution by coprecipitation with thorium fluoride, leaving beryllium in solution. The crude actinium was purified to provide about 40 milligrams of actinium nitrate using anion exchange in methanol-water-nitric acid solution. The purified actinium was then used to generate high-purity 223Ra. We extracted 223Ra using anion exchange in a methanol-water-nitric acid solution. After the radium was separated, actinium and thorium were then eluted from the column and dried for interim storage. This single-pass separation produces high purity, carrier-free 223Ra product, and does not disturb the 227Ac/227Th equilibrium. A high purity, carrier-free 227Th was also obtained from the actinium using a similar anion exchange in nitric acid. These methods enable efficient production of 223Ra for research and new alpha-emitter radiopharmaceutical development.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Assessment for advanced fuel cycle options in CANDU

The possible options for advanced fuel cycles in CANDU reactors including actinide burning options and thorium cycles were explored and are feasible options to increase the efficiency of uranium utilization and help close the fuel cycle. The actinide burning TRUMOX approach uses a mixed oxide fuel of reprocessed transuranic actinides from PWR spent fuel blended with natural uranium in the CANDU-900 reactor. This system reduced actinide content by 35% and decreased natural uranium consumption by 24% over a PWR once through cycle. The thorium cycles evaluated used two CANDU-900 units, a generator and a burner unit along with a driver fuel feedstock. The driver fuels included plutonium reprocessed from PWR, from CANDU and low enriched uranium (LEU). All three cycles were effective options and reduced natural uranium consumption over a PWR once through cycle. The LEU driven system saw the largest reduction with a 94% savings while the plutonium driven cycles achieved 75% savings for PWR and 87% for CANDU. The high neutron economy, online fuelling and flexible compact fuel make the CANDU system an ideal reactor platform for many advanced fuel cycles.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Radioactive Water Treatment at a United States Environmental Protection Agency Superfund Site - 12322

A water treatment system at a United States Environmental Protection Agency (USEPA) Superfund site impacted by radiological contaminants is used to treat water entering the site. The United States Army Corps of Engineers (USACE) is actively managing the remedial action for the USEPA using contracts to support the multiple activities on site. The site is where former gas mantle production facilities operated around the turn of the century. The manufacturing facilities used thorium ores to develop the mantles and disposed of off-specification mantles and ore residuals in the surrounding areas. During Site remedial actions, both groundwater and surface water comes into contact with contaminated soils and must be collected and treated at an on-site treatment facility. The radionuclides thorium and radium with associated progeny are the main concern for treatment. Suspended solids, volatile organic compounds, and select metals are also monitored during water treatment. The water treatment process begins were water is pumped to a collection tank where debris and grit settle out. Stored water is pumped to a coagulant tank containing poly-aluminum chloride to collect dissolved solids. The water passes into a reaction tube where aspirated air is added or reagent added to remove Volatile Organic Compounds (VOC'S) by mass transfer and convert dissolved iron to a solid. The water enters the flocculent polymer tank to drop solids out. The flocculated water overflows to a fluidized bed contact chamber to increase precipitation. Flocculation is where colloids of material drop out of suspension and settle. The settled solids are periodically removed and disposed of as radioactive waste. The water is passed through filters and an ion exchange process to extract the radionuclides. Several million liters of water are processed each year from two water treatment plants servicing different areas of the remediation site. Ion exchange resin and filter material are periodically replaced and disposed of as radioactive waste. A total of 0.85 m{sup 3} of waste sludge per year requires disposal on average, in addition to another 6.6 m{sup 3} of waste cartridge filters. All water discharges are regulated by a state of New Jersey Pollutant Discharge Elimination System Permit implemented by the Federal Water Pollution Control Act (Clean Water Act). Laboratory analyses are required to satisfy requirements of the state NPDES permit. Specific monitoring parameters and discharge rates will be provided. Use of the water treatment systems drastically reduces the amount of contaminated water requiring solidification and water disposal to near zero. Millions of liters of potentially contaminated water from excavation activities is treated and released within permit limits. A small volume of solid radioactive waste (21 cubic meters) is generated annually from water treatment process operations. Management of ground and surface water is effectively controlled in remediation areas by the use of sumps, erosion control measures and pumping of water to storage vessels. Continued excavations can be made as water impacting the site is effectively controlled. (authors)

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

Radon Vapor Intrusion Screening Level Calculator - 17041

The U.S. Environmental Protection Agency has produced a web-based calculator for the assessment of cancer risk from radon in soil, soil gas, and pore water seeping into above-ground buildings. Like the chemical vapor intrusion screening level (VISL) calculator, the radon VISL (RVISL) provides screening levels in subsurface soil and water that are protective of preliminary remediation goals (PRGs) for radon in air that are based on either cancer risk or Applicable or Relevant and Appropriate Requirements (ARARs). In addition to the screening levels calculated, the RVISL can also calculate the risk from radon concentrations in soil, water, and air. The exposure routes evaluated in the RVISL are inhalation and external exposure from submersion in radon. Two source term options are provided for the user: one without radioactive decay and one with decay. Sources continually generating radon (Rn-222 and Rn-220), such as below ground uranium, would best be modeled with the 'without decay' models. Radon is the second leading cause of lung cancer. Radon is released from the normal decay of uranium, thorium, and radium in rocks and soil. It is an invisible, odorless, tasteless gas that seeps up through the ground and diffuses into the air. In areas without adequate ventilation, such as underground mines, radon can accumulate to levels that substantially increase the risk of lung cancer. Radon can enter homes through cracks in floors, walls, or foundations, and collect indoors. It can also be released from building materials, or from water obtained from wells that contain radon. Radon levels can be higher in homes that are well insulated, tightly sealed, and/or built on soil rich in the elements uranium, thorium, and radium. Basement and first floors typically have the highest radon levels because of their closeness to the ground. The RVISL will also allow for the calculation of screening levels for thoron (Rn-220) and Rn-219. The RVISL will be useful in assessing the final cleanup goals for subsurface soils where direct contact is not assumed. These sites can be landfills, buried waste cells or naturally occurring. By calculating protective levels of radon in subsurface soils, the likelihood of exceeding CERCLA protective standards (either risk range or ARARs) can be reduced. (authors)

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

Multi-Faceted Approach to Evaluate Discharge Locations at a FUSRAP Site - 18421

Following the completion of a Soil and Groundwater Remedial Investigation by USACE at a FUSRAP site located in Nassau County, New York, groundwater data indicated the presence of contaminant source(s) that had not been identified. As a result, a targeted investigation was undertaken to understand how materials were handled and disposed of at a site located in Nassau County, Long Island, New York. The FUSRAP site is a 4.3-hectare (10.5-acre) site encompassing three parcels which had historically been utilized as a nuclear materials manufacturing facility (e.g., cores, slug, and fuel elements) for reactors used in both research and electric power generation in early 1950's until late 1960's; however, the targeted investigation was on the southern-most parcel known as the '70 Property' (the Site). The 70 Property contained a building that was used for atomic fuel-element manufacturing from 1958 until 1966 under license to the Atomic Energy Commission (AEC), and involved the use of natural, enriched, and depleted uranium as well as lesser amounts of thorium. The major contaminants of concern are tetrachloroethene, nickel, uranium and thorium. During the RI, significant sources of contamination were not found in Site soils despite records indicating that chemicals and radionuclides had been disposed of on the property. In addition, downgradient groundwater data indicated a source(s) on the 70 Property that could not be identified by known sources on other the other FUSRAP parcels or from adjacent sites. This conclusion was further supported following development of groundwater contaminant transport and flow models. The main objective of the targeted investigation was to evaluate the potential discharge points on the 70 Property that may have been utilized for liquids waste disposal. However, several challenges existed including 1) lack of building construction details; 2) historical changes to Site building and property; 3) the use of the Site by various tenants; and 4) the commingling of Site contaminants with impacts from multiple sources. To overcome these challenges, historic documents/reports with information about the handling and disposal of liquid wastes were reviewed to evaluate quantities, contents and locations of waste disposal. Thereafter, an investigation was undertaken that included geophysical surveys; pipe tracing and video inspection; in-situ smoke and water testing; installation of test trenches and soil borings; and, sampling sludge residuals, soils and water within the various drainage structures that were located. This paper describes the multi-faceted investigation approach that was utilized to overcome the challenges faced during the investigation as well as insights into the waste disposal practices of the time period and the findings of the investigation. (authors)

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

Optimising National WM Strategies through early consideration of Nuclear Safeguards - 18607

International Atomic Energy Agency (IAEA) safeguards stand alongside safety and security as one of the three pillars for prudent management of all nuclear material, including nuclear waste. Although safety considerations, including public and environmental radiation protection, are the traditional drivers of waste management strategies, there is a requirement for nuclear safeguards in shaping and quantifying long-term solutions for special classes of radioactive waste. This is especially important when considering the disposal of nuclear material in a repository that could be released from regulatory oversight for radiation protection after a period of institutional control, as some safeguards oversight would still be required. Failure to consider nuclear safeguards at the early stages of planning may alternatively lead to excessive volumes for disposal or the need to recondition already treated wastes. At the very least, safeguards measures should be considered before countries seek to dispose of legacy nuclear materials, decommissioning wastes and other materials containing uranium, thorium or plutonium in an operating repository. Nuclear safeguards are the set of tools and techniques used by the IAEA to verify that countries comply with their international obligations not to divert nuclear materials to nuclear explosive devices. Under the Nuclear Non-Proliferation Treaty (NPT), nuclear safeguards apply to all uranium, thorium and plutonium held in non-weapons NPT states, with limited exceptions. In nuclear-weapon states, much of their material is likewise subject to IAEA safeguards through conditions in agreements with respective suppliers of the uranium. Under IAEA safeguards, such nuclear material must remain accounted for and accessible for periodic verification-analogous to the safety concepts of reversibility and retrievability. Nevertheless, safeguards requirements can be terminated - provided that the State and the [IAEA] agree that such nuclear material is practicably irrecoverable - through dilution or immobilisation in solid form (such as cement, glass or ceramic). This means candidate material can be disposed of as non- nuclear material, with no further safeguards requirements, provided this criterion is met. Unlike radiological wastes where repository design strongly influences pathways for radionuclide migration and overall safety performance, repository design is not a factor in considering the termination of safeguards on nuclear material. This paper looks at similarities and differences in the safety, security and safeguards regulatory requirements for the management of radioactive waste and nuclear material in Australia. In particular, it outlines recent experience gained from participation in Australia's waste acceptance criteria working group for a national low-level waste repository and intermediate-level waste store, as well as from Australia's overall waste management strategy. These items of work have coincided with the IAEA update of its internal policy on termination of safeguards on nuclear material. Australia's experience shows how inclusion of safeguards criteria in the planning process is leading to optimisations in choice of facility concepts, project sequencing and equitable distribution of economic risk. By considering the three pillars-safety, security and safeguards-early on in the planning process, Australia's waste management strategy will provide a benchmark for other states in establishing and updating their own waste acceptance criteria for the disposal of nuclear material. (authors)

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

A Rapid Sequential Separation Method for Determination of Actinides and TENORM in Fracking Wastes - 20223

With recent advances in unconventional drilling technology in the US and around the world, the risks of environmental contamination and exposure due to technologically enhanced naturally occurring radioactive materials (TENORM) from fracking wastes have also surfaced. Although the Permian Basin (southeastern New Mexico, western Texas) has long been a modest producer of oil, the advent of fracking technology a decade ago turned it into hot property for producers. A 2018 assessment of undiscovered, technically recoverable continuous oil and gas resources by the U.S. Geological Survey estimates an average 46.3 billion barrels of oil and 281 trillion cubic feet of gas. This is likely to increase the use of unconventional drilling and TENORM generation. TENORM concentrations in oil and gas exploration and production waste can be several hundred to several thousand times higher than background TENORM concentrations. Both the thorium and uranium decay chains contribute to airborne radionuclides arising from the radon gas escaping the ground and subsequently decaying as airborne particulates. Additionally, radon decay products {sup 210}Pb and {sup 210}Po can build up in scale on the internal surfaces of oil and gas handling pipes and in sludge in refineries, becoming potential inhalation and ingestion hazards for workers. Furthermore, southeastern NM is also home to world's only licensed and operating transuranic nuclear waste repository, the DoE's Waste Isolation Pilot Plant (WIPP). Plutonium isotopes ({sup 239+240}Pu) and {sup 241}Am, are expected to account for more than 99% of the total radioactivity scheduled for disposal in the WIPP repository. Thus, an improved understanding of the environmental fate and transport of TENORM, liberated by unconventional drilling, is essential to assess how best to protect individuals and the environment. In this context, accurate measurement of TENORM and actinides in environmental and biological samples is essential. In this presentation, a new sequential method for the separation and pre-concentration of actinides (Pu, Am, Np) and TENORM (Po, U and Th) derived from oil and gas exploration is proposed. The TEVA method involves a rapid co-precipitation step to remove matrix interferences from the samples, followed by plutonium oxidation state adjustment to Pu (IV) and an incubation period of ∼ 1 hour at 50-60 deg. C to allow the resultant Po (II) to oxidize into Po(IV). The polonium, neptunium, thorium and plutonium are then separated on a TEVA column, while americium and uranium are separated on a TRU column. After separation, the alpha counting source was prepared by micro-precipitation with copper sulfide (CuS) for polonium and neodymium fluoride (NdF{sub 3}) for actinides. The efficiency and reliability of the procedures were tested by analyzing filter, drinking water and frack sand samples. The method is simple, robust and can be performed quickly with excellent removal of interferences, high chemical recovery and very good alpha peak resolution. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Neutronic analysis of a PWR-type SMR core using duplex ThO{sub 2}-UO{sub 2} in TRISO fuel particles

Currently, studies on small modular reactors (SMRs) present an important development due to the potential they represent in terms of safety, operational flexibility, economy, and non-energy applications. Furthermore, there is increasing interest in studying the use of thorium as fuel, as an effective way to solve problems such as the shortage of uranium reserves, reduction of nuclear waste and nuclear proliferation. Also, the use of thorium combined with highly enriched uranium TRISO particles has been studied, showing proper performance. In this work, the concept of ThO{sub 2}-UO{sub 2} duplex fuel is used, for the core configuration of a PWR type SMR that uses TRISO fuel, designed to achieve extended fuel cycles. Three distribution cases of ThO{sub 2} and UO{sub 2} in TRISO particles inside the fuel rods are compared. First, particles composed by a homogeneous mixture of ThO{sub 2} and UO{sub 2} are distributed inside of fuel elements. Second, the fuel zone of the fuel elements is divided into two radially, an internal one where the TRISO particles composed of ThO{sub 2} are distributed and the external one where the TRISO particles composed of UO{sub 2} are distributed. The third case is like the previous one, except that the particles that contain UO{sub 2} are distributed in the inner zone and those that contain ThO{sub 2} in the outer zone of the fuel elements. The comparison of the cases is carried out in terms of cycle main isotopes' mass transmutation, power distribution and temperature reactivity coefficients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

LA-ICP/MS Analysis of Plastics as a Method to Support Polymer Assay in the Assessment of Materials for Low Background Detectors

Ultra low-background radiation measurements are essential to several large-scale physics investigations, such as those involving neutrinoless double-beta decay, dark matter detection (such as SuperCDMS), and solar neutrino detection. There is a need for electrically and thermally insulating dielectric materials with extremely low-background radioactivity for detector construction. This need is best met with plastics. Most currently available structural plastics have milliBecquerel-per-kilogram total intrinsic radioactivity. Modern low-level detection systems require a large variety of plastics with low microBecquerel-per-kilogram levels. However, the assay of polymer materials for extremely low levels of radioactive elements, uranium and thorium in particular, presents new challenges. It is only recently that any certified reference materials (CRMs) for toxic metals such as lead or cadmium in plastics have become available. However, there are no CRMs for uranium or thorium in thermoplastics. This paper discusses our assessment of the use of laser ablation (LA) for sampling and inductively coupled plasma mass spectrometry (ICP-MS) for analysis of polyethylene (PE) samples, with an emphasis on uranium determination. Using a CRM for lead in PE, we examine LA and ICP-MS parameters that determine whether the total atom efficiencies for uranium and lead are similar, and explore methods to use the lead content in a plastic as part of the process of estimating or determining the uranium content by LA-ICP-MS.

laser ablation↗
DOE OSTI2021

Neutronic Performance of a Compact 10 MWe Nuclear Reactor with Low Enrichment (ThxU1−x)N Fuel

Abstract Recent interest in compact nuclear reactors for applications in space or in remote locations drives innovation in nuclear fuel design, especially non-oxide ceramic nuclear fuels. This work details neutronic modeling designed to support the development of a new nuclear fuel concept based on a mixture of thorium and uranium nitride. A Monte Carlo N-Particle Version 6.2 (MCNP-6) model of a compact 10 MWe reactor design which incorporates (Th x U 1− x )N fuel is presented. In this context, a “compact” reactor is a completely assembled reactor which may be emptied of coolant and transported by specialized commercial vehicle, deployed by a C130J aircraft, or launched into space. Core geometry, reflector barrels, and the heat exchange zones are designed to support reduction of overall reactor volume of core components while maintaining criticality with a fixed total fuel mass of 4500 kg. Dense mixed nitrides of thorium nitride (ThN) additions in uranium nitride (UN) in 5 wt.% increments between $$0.05 \le x \le 0.5$$ 0.05 ≤ x ≤ 0.5 have been considered for calculation of $$k_{\infty }$$ k ∞ and $$k_{{{\text{effective}}}}$$ k effective . ThN additions in UN results in a slight increase in the magnitude of the temperature coefficient of reactivity, which is negative by design. The isotopic distribution of the principal actinide inventory as a function of burnup, time, and initial fuel composition is presented and discussed within the context of the proliferation risk of this core design.

Parker, S. S. (ORCID:000000018597440X)↗
DOE OSTI2017

Molten salt reactor neutronics and fuel cycle modeling and simulation with SCALE

Current interest in advanced nuclear energy and molten salt reactor (MSR) concepts has enhanced interest in building the tools necessary to analyze these systems. A Python script known as ChemTriton has been developed to simulate equilibrium MSR fuel cycle performance by modeling the changing isotopic composition of an irradiated fuel salt using SCALE for neutron transport and depletion calculations. Some capabilities in ChemTriton that have improved, include a generic geometry capable of modeling multi-zone and multi-fluid systems, enhanced time-dependent feed and separations, and a critical concentration search. Although more generally applicable, the capabilities developed to date are illustrated in this paper in three applied problems: (1) simulating the startup of a thorium-based MSR fuel cycle (a likely scenario requires the first of these MSRs to be started without available 233U); (2) determining the effect of the removal of different fission products on MSR operations; and (3) obtaining the equilibrium concentration of a mixed-oxide light-water reactor fuel in a two-stage fuel cycle with a sodium fast reactor. Moreover, the third problem is chosen to demonstrate versatility in an application to analyze the fuel cycle of a non-MSR system. During the first application, the initial fuel salt compositions fueled with different sources of fissile material are made feasible after (1) removing the associated nonfissile actinides after much of the initial fissile isotopes have burned and (2) optimizing the thorium concentration to maintain a critical configuration without significantly reducing breeding capability. In the second application, noble metal, volatile gas, and rare earth element fission products are shown to have a strong negative effect on criticality in a uranium-fueled thermal-spectrum MSR; their removal significantly increases core lifetime (by 30%) and fuel utilization. In the third application, the fuel of a mixed-oxide light-water reactor approaches an equilibrium composition after 20 depletion steps, demonstrating the potential for the longer time scales required to achieve equilibrium for solid-fueled systems over liquid fuel systems. This time to equilibrium can be reduced by starting with an initial fuel composition closer to that of the equilibrium fuel, reducing the need to handle time-dependent fuel compositions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Accelerating actinium-225 purification by high-pressure ion chromatography

Actinium-225 (t1/2 = 9.92 days) is an important radioisotope for targeted alpha therapy applications. The limited supply obtained through the decay of thorium-229 has motivated accelerator-based production routes, including irradiation of thorium targets. Irradiated targets can produce useful quantities of actinium-225, but the product requires final purification from chemically similar lanthanide contaminants. This work describes an automated high-pressure ion chromatography method for this final polishing step. The method uses a reusable strong-acid cation-exchange column bearing sulfonic acid functional groups. α-Hydroxyisobutyric acid (α-HIBA), adjusted to pH 4.3 with lithium hydroxide, complexes and elutes lanthanides, a dilute hydrochloric acid matrix-exchange step removes residual α-HIBA, and concentrated hydrochloric acid then elutes retained actinium(III). The protocol purified actinium-225 to >99% radiopurity across tracer-level samples and samples containing >150 µCi (5.6 MBq) of activity. A 10 min, 0.1 M hydrochloric acid matrix exchange substantially reduced organic eluent carryover, and in-line sodium iodide detection enabled real-time monitoring of actinium and lanthanide elution. The developed method can be completed in <1 h and provides a basis for automated purification workflows for accelerator-produced actinium-225.

Gaddis, Kevin [ORNL] (ORCID:0000000183398314)↗
DOE OSTI2016

Structure and thermodynamics of uranium-containing iron garnets

Use of crystalline garnet as a waste form phase appears to be advantageous for accommodating actinides from nuclear waste. Previous studies show that large amounts of uranium (U) and its analogues such as cerium (Ce) and thorium (Th) can be incorporated into the garnet structure. In this study, we synthesized U loaded garnet phases, Ca 3 U x Zr 2–x Fe 3 O 12 (x = 0.5–0.7), along with the endmember phase, Ca 3 (Zr 2 )SiFe 3+ 2 O 12 , for comparison. The oxidation states of U were determined by X-ray photoelectron and absorption spectroscopies, revealing the presence of mixed pentavalent and hexavalent uranium in the phases with x = 0.6 and 0.7. The oxidation states and coordination environments of Fe were measured using transmission 57 Fe-Mössbauer spectroscopy, which shows that all iron is tetrahedrally coordinated Fe 3+ . U substitution had a significant effect on local environments, the extent of U substitution within this range had a minimal effect on the structure, and unlike in the x = 0 sample, Fe exists in two different environments in the substituted garnets. The enthalpies of formation of garnet phases from constituent oxides and elements were first time determined by high temperature oxide melt solution calorimetry. The results indicate that these substituted garnets are thermodynamically stable under reducing conditions. Furthermore, our structural and thermodynamic analysis further provides explanation for the formation of natural uranium garnet, elbrusite-(Zr), and supports the potential use of Ca 3 U x Zr 2–x Fe 3 O 12 as viable waste form phases for U and other actinides.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Negative thermal expansion in Th{sub 2}O(PO{sub 4}){sub 2}

Highlights: {yields} Dithorium oxide phosphate shows a continuous negative thermal expansion over a 600 {sup o}C range. {yields} Negative expansion arises from oxygen rocking and cations repulsions. {yields} Big and high-charge thorium IV appears ideal for generating negative expansion. -- Abstract: High temperature X ray diffraction performed on recently discovered orthorhombic Th{sub 2}O(PO{sub 4}){sub 2} shows a continuous linear thermal contraction (-1.6 x 10{sup -6} {sup o}C{sup -1}) in 20-600 {sup o}C range and a near-zero expansion at higher temperatures resulting from a dual structural deformation involving oxygen oscillations and inter-cations repulsions. Although similar mechanisms were observed in isotypic Zr{sub 2}O(PO{sub 4}){sub 2} (+1.5 x 10{sup -6} {sup o}C{sup -1}) and U{sub 2}O(PO{sub 4}){sub 2} (-1.4 x 10{sup -6} {sup o}C{sup -1}), those observed in Th{sub 2}O(PO{sub 4}){sub 2} are particularly intense because of the high ionic radius of tetravalent thorium.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Reactor performance and safety characteristics of ThN-UN fuel concepts in a PWR

The reactor performance and safety characteristics of mixed thorium mononitride (ThN) and uranium mononitride (UN) fuels in a pressurized water reactor (PWR) are investigated to discern the potential nonproliferation, waste, and accident tolerance benefits provided by this fuel form. In this paper, we present results from an initial screening of mixed ThN-UN fuels in normal PWR operating conditions and compares their reactor performance to UO 2 in terms of fuel cycle length, reactivity coefficients, and thermal safety margin. ThN has been shown to have a significantly greater thermal conductivity than UO 2 and UN. Admixture with a UN phase is required because thorium initially contains no fissile isotopes. Results from this study show that ThN-UN mixtures exist that can match the cycle length of a UO 2 -fueled reactor by using 235 U enrichments greater than 5% but less than 20% in the UN phase. Reactivity coefficients were calculated for UO 2 , UN, and ThN-UN mixtures, and it was found that the fuel temperature and moderator temperature coefficients of the nitride-based fuels fall within the acceptable limits specified by the AP1000 Design Control Document. Reduced soluble boron and control rod worth for these fuel forms indicates that the shutdown margin may not be sufficient, and design changes to the control systems may need to be considered. The neutronic impact of 15 N enrichment on reactivity coefficients is also included. Due to the greatly enhanced thermal conductivity of the nitride-based fuels, the UN and ThN-UN fuels provide additional margin to fuel melting temperature relative to UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Radiometric evaluation of diglycolamide resins for the chromatographic separation of actinium from fission product lanthanides

Actinium-225 is a potential Targeted Alpha Therapy (TAT) isotope. It can be generated with high energy (≥ 100 MeV) proton irradiation of thorium targets. The main challenge in the chemical recovery of 225 Ac lies in the separation from thorium and many fission by-products most importantly radiolanthanides. We recently developed a separation strategy based on a combination of cation exchange and extraction chromatography to isolate and purify 225 Ac. In this study, actinium and lanthanide equilibrium distribution coefficients and column elution behavior for both TODGA (N,N,N',N'-tetra- n -octyldiglycolamide) and TEHDGA (N,N,N',N'-tetrakis-2-ethylhexyldiglycolamide) were determined. Density functional theory (DFT) calculations were performed and were in agreement with experimental observations providing the foundation for understanding of the selectivity for Ac and lanthanides on different DGA (diglycolamide) based resins. The results of Gibbs energy (ΔG aq ) calculations confirm significantly higher selectivity of DGA based resins for Ln III over Ac III in the presence of nitrate. As a result, DFT calculations and experimental results reveal that Ac chemistry cannot be predicted from lanthanide behavior under comparable circumstances.

225Ac↗
DOE OSTI2016

Mössbauer Spectral Properties of Yttrium Iron Garnet, Y3Fe5O12, and its Isovalent and Nonisovalent Yttrium-Substituted Solid Solutions

Several high-resolution Mössbauer spectra of yttrium iron garnet, Y3Fe5O12, have been fit as a function of temperature with a new model based on a detailed analysis of the spectral changes that result from a reduction from the cubic Ia–3d space group to the trigonal R–3 space group. These spectral fits, which are all statistically identical, indicate that the magnetic sextet arising from the 16a site in cubic symmetry is subdivided into three sextets arising from the 6f, and the 3d, 3d, and the 1a, 1b, and 2c sites in rhombohedral-axis trigonal symmetry. The 24d site in cubic symmetry is subdivided into four sextets arising from four different 6f sites in R–3 rhombohedral-axis trigonal symmetry, sites that differ only by the angles between the principal axis of the electric field gradient tensor and the magnetic hyperfine field assumed to be parallel with the magnetic easy axis. This analysis, when applied to the potential nuclear waste storage compounds, Y3-xCa0.5xTh0.5xFe5O12 and Y3- xCa0.5xCe0.5xFe5O12, indicates virtually no perturbation of the structural, electronic, and magnetic properties upon substitution of small amounts of calcium(II) and thorium(IV) or cerium(IV) onto the yttrium(III) 24c site as compared with Y3Fe5O12. The observed broadening of the four different 6f sites derived from the 24d site results from the substitution of yttrium(III) by calcium(II) and thorium(IV) or cerium(IV) cations on the next-nearest neighbor 24c site. In contrast, the same analysis, when applied to Y2.8Ce0.2Fe5O12, indicates a local perturbation of the magnetic exchange pathways as a result of the presence of cerium(IV) in the 24c next-nearest neighbor site of the iron(III) 24d site.

Environmental Molecular Sciences Laboratory↗
DOE OSTI2020

Facile Oxide to Chalcogenide Conversion for Actinides Using the Boron–Chalcogen Mixture Method

Actinide chalcogenides are of interest for fundamental studies of the behavior of 5f electrons in actinides located in a soft ligand coordination environment. As actinides exhibit an extremely high affinity for oxygen, the synthesis of phase pure actinide chalcogenide materials free of oxide impurities is a great challenge and, moreover, requires the availability and use of oxygen free starting materials. Herein, we report a new method, the Boron-Chalcogen Mixture (BCM) method, for the synthesis of phase pure uranium chalcogenides based on the use of a boron/chalcogen mixture, where boron functions as an “oxygen sponge” to remove oxygen from an oxide precursor and where the elemental chalcogen effects transformation of the oxide precursor into an oxygen free chalcogenide reagent. Here, the boron oxide can be separated from the reaction mixture that is left to react to form the desired chalcogenide product. Several syntheses are presented that demonstrate the broad functionali-ty of the technique and thermodynamic calculations that show the underlying driving force are discussed. Specifically, three classes of chalcogenides that include both new (rare earth uranium sulfides and alkali-thorium thiophosphates) and previously reported compounds were prepared to validate the approach: binary uranium and thorium sulfides, oxide to sulfide transfor-mation in solid state reactions, and in situ generation of actinide chalcogenides in flux crystal growth reactions.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗