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Low Level Radon Concentration and Radon Daughter Dose Monitoring in an Arid-zone Waste Store - 20081

Radioactive waste containing predominantly soil and rubble from the clean-up of a former uranium and thorium minerals processing research facility is currently stored at the Woomera Test Range in South Australia awaiting final disposal. Prior to commencing any manipulation of the nearly 10,000 waste drums (205 L or 55 gallon each), radon concentration monitoring and radon daughter dosimetry has been undertaken to (i) gain understanding of the temporal and spatial variation of radon and its daughters within the storage facility, which could provide insights into the radiological source term and (ii) confirm that radiological risks to workers are below the guideline values. A time series of nearly 12 months of detailed observations shows low level radon concentrations with annual average values that are slightly above the average radon concentrations in Australian homes. Peak values in summer are about three times higher than the annual average based on the CRM measurements. Both radon concentrations and annual effective dose are well below the regulatory limits. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

GPS-Based Gamma Survey for Characterizing and Decommissioning NORM Sites - 20389

Gamma survey techniques are an especially powerful decommissioning tool at naturally occurring radioactive material (NORM) sites due to both the low cost to obtain data over a large spatial scale and the abundance of gamma emitters in the uranium and thorium decay series. Gamma surveys are executed by coupling a detector - most often a sodium iodide crystal - to a global positioning system (GPS), then reporting a location and gross gamma reading coincidentally to a data logger. Systems may be carried by workers or mounted to a car, all-terrain vehicle, or unmanned aerial system (UAS). The resulting data set provides a high-resolution but low precision map of the gamma radiation field over the area surveyed. Frequently this map is also correlated to soil concentrations of NORM radionuclides (most often, Ra-226) and/or exposure rate. Gamma survey parameters such as movement speed, transect spacing, and data logging frequency define the spatial resolution of the resulting surface, and can be optimized depending on the desired survey sensitivity. This paper examines gamma survey as a tool for decommissioning NORM sites and provides an overview of current gamma survey technology designed to improve the efficiency and effectiveness of the decommissioning process. Topics to be discussed in the paper include: - An overview of gamma survey systems, and the utility of different delivery vehicles depending on desired cost, desired spatial resolution, and site topography. - The influence of physical detector characteristics on detection sensitivity and survey planning. - The tradeoff between high-resolution and large spatial extent, but inherently uncertain data, and low-resolution, low spatial extent, but highly certain data, as well as the specific utility of each of these types of data during NORM facility decommissioning. - Confounding variables that may limit the utility of gamma survey at some sites (e.g., radon gas and spatial heterogeneity / hot spots), and methods to plan for and control these conditions. Results show that the confounding variables, such as radon and data output can greatly influence the overall data quality associated with the decommissioning process. In addition, the use of real-time and aerial survey platforms provides a method for ensuring proper spatial extent of the data. When applied thoughtfully, gamma survey is a powerful tool for detecting NORM radionuclides in the environment and a cost-effective technique for identifying areas requiring remediation. However, entities performing or using gamma survey as a decommissioning tool must be aware of both its advantages and its limitations before basing remediation or regulatory action on gamma survey results. (authors)

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Development of an In-Situ Leaching Technology for Extracting Residual Uranium from Remediated Soil - 20099

Site-specific chemical solutions for extracting uranium were developed and evaluated for treating previously remediated soil at a former thorium mill tailings site, known as the Former Rare Earths Facility (REF), in West Chicago, Illinois. Prior remediation efforts had resulted in residual sand and gravel aggregate that met the soil cleanup standards. The target soil contains residual levels of uranium, which continues to leach into the site groundwater at levels exceeding groundwater protection standards (GWPS). A series of bench-scale tests were conducted to evaluate the efficacy and viability of in-situ leaching (ISL) technology to treat the target soil for the purposes of attaining uranium GWPS at the REF. Specific objectives of the bench-scale tests entailed identification and selection of optimal leaching, quantification of the extent and rate of uranium released as a function of chemical composition, chemical concentrations, chemical residence times, and obtaining critical data to estimate treatment costs. A combination of sodium bicarbonate and sodium carbonate reagent was the most effective leaching solution. In general, leaching solutions containing a higher concentration of sodium carbonate demonstrated superior uranium leaching capabilities. Removal of 50% of the uranium mass from the soil matrix was adequate for achieving compliance with the site-specific GWPS. After six to seven pore volume flushes (PVFs), the uranium leaching effectiveness of all leaching solutions was similar. The amount of initial uranium removed with each subsequent PVF was minimal, suggesting limited solubility of the remaining mass of uranium. The effects of leaching duration were significant and persisted through at least the first five consecutive PVFs. Leaching of uranium also resulted in leaching of several non-uranium constituents, many of which exceeded applicable National Pollutant Discharge Elimination System (NPDES) standards. Given historical operations and waste disposal practices at the site, leaching of non-uranium constituents was expected. Hydrochloric acid was used to stabilize treated soil. Savings of approximately $148 million in direct and indirect remediation costs are expected with an ISL alternative compared to the excavation and landfill disposal remediation option. (authors)

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Uranium Mill Tailings Radiation Control Act Title II DOE Due Diligence and Lessons Learned from a Previous Site Transfer - 20352

Title II of the Uranium Mill Tailings Radiation Control Act (UMTRCA) established that a government agency will provide perpetual care for closed uranium and thorium ore-processing sites that were operating under an NRC source material license in 1978 or were licensed thereafter. Commercial owners (licensees) operating under an NRC or agreement state specific license when UMTRCA was passed are responsible for conducting reclamation of any byproduct material remaining from uranium-ore processing operations in accordance with an NRC or agreement state approved reclamation plan. Reclamation includes both surface and groundwater remedies. Upon completion of reclamation and approval by NRC, the site is required to be transferred to either the host state or the DOE for long-term surveillance and maintenance. Since UMTRCA's enactment, six Title II sites have been transferred to DOE; an additional 24 Title II sites are anticipated to be transferred before 2050. DoE's role mandated under UMTRCA Title II as the long-term care custodian is to perform 'monitoring, maintenance, and emergency measures necessary to protect the public health and safety.' UMTRCA requires that the licensee pay a long-term surveillance charge 'sufficient to cover the annual costs of site surveillance.' However, at some sites such as the Bluewater, New Mexico, Disposal Site, this mandate has required additional effort and expense by DOE, beyond the originally anticipated and intended scope within UMTRCA, but within the authority of DOE under UMTRCA. In 1997, the Bluewater site became the second UMTRCA Title II site to be transferred to DOE. The site was the location of a uranium mill operated from 1953 until 1982. The specific licensee began site reclamation in 1991, and by 1995 all tailings and contaminated materials were encapsulated in two tailings disposal cells and other disposal areas. In addition to surface contamination, milling activities impacted groundwater in the two upper aquifers. In 1989, the specific licensee attempted active groundwater remediation; however, no significant reduction in contaminant concentrations was observed. As a result, the specific licensee applied to NRC for alternate concentration limits (ACLs) in 1990, which were approved in 1996 as being protective, after additional corrective actions were performed. Since transfer of the Bluewater site to DOE, unforeseen challenges have occurred, requiring additional actions. The first challenge is the occurrence of surface depressions located on the northern section of the main tailings disposal cell. The depressions were first observed during DoE's initial inspection in 1998; however, evidence of these can be observed on satellite images taken prior to transfer. Since being first observed, the depressions have continued to grow both in depth and areal extent. Due to the design of the main tailings disposal cell, the depressions impede storm water from being effectively shed off the 101- hectare (250-acre) top slope of the main tailings disposal cell. Instead, storm water accumulates in the depressions, forming a large ephemeral pond that has stored up to 16.3 x 10{sup 6} liters (4.3 million gallons) of stormwater. The ponding poses a potential risk to the integrity of the main tailings disposal cell in the case of a large storm event, with the potential to cause the pond to overtop and erode the cover material and underlying waste. DOE has taken a number of short-term actions to monitor, measure, and reduce the ponding and is currently working with the US Army Corps of Engineers to design and construct a repair. Additional challenges are associated with groundwater at the Bluewater site. Nine wells were present on the 1335-hectare (3300-acre) site upon transfer. Groundwater compliance was called into question after the State of New Mexico reduced its uranium groundwater standard from 5.0 to 0.03 milligrams per liter in 2004, and when an ACL for uranium was exceeded in a site monitoring well in 2010. Acquiring historical groundwater data and subsequent evaluations as well as additional DOE groundwater monitoring led to installing 10 new monitoring wells and performing additional site hydrogeology recharacterization. DOE continues to evaluate groundwater conditions at the site and works with the NRC to determine regulatory requirements and a path forward. As a result of lessons learned at the Bluewater site and other Title II sites, improved processes have been implemented at a programmatic level to increase due diligence before site transfer and prevent similar issues from occurring at other UMTRCA Title II sites under long-term management. DoE's due diligence process is documented in the Process for Transition of UMTRCA Title II Disposal Sites to DOE for Long-Term Surveillance and Maintenance and is designed to ensure that DOE has no technical or compliance concerns with regulatory decisions that might compromise protectiveness following site transfer to DOE. Implementation of the due diligence process has increased DoE's role prior to site transfer and has been effective in identifying potential issues. Actions by NRC and specific licensees, in response to enhanced due diligence efforts by DOE, are expected to minimize, if not totally prevent, the need for unanticipated actions by DOE pertaining to the surface and groundwater remedies after site transfer. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543

This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)

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Benchmark Study of Infinite Lattice of Pebble Fuel within FLiBe Coolant for TMSR-SF1 Design

Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1) is Chinese Fluoride-salt-cooled High Temperature Reactor (FHR) designed by the TMSR group. The construction of this test reactor is expected in the near years. Using TRISO particle based pebble fuel leads to various challenges in neutronics modeling, such as double heterogeneity effect, preserving exact number of particles, their random distributions in pebble, and etc. For the purpose of verifying the calculations, benchmark study of infinite lattice of pebble fuel within FLiBe coolant for the TMSR-SF1 design has been performed. Back-to-back comparisons have been made between TMSR and MIT researchers using different Monte-Carlo based codes. Very good agreements have been found and the results quantitatively evaluate the neutronic effect of various modeling approaches. In addition, sensitivity study has been carried out for the reference pebble-bed fuel lattice. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Separation of Actinides from Lanthanides in Molten Fluorides by the Modulated-Current Electrolysis

Recently, the Molten Salt Reactor (MSR) is taking a renewed attraction as one of the reactor systems highlighted by the Generation IV International Forum. It is a non-classical (liquid fuel) system which can be operated as a breeder within the {sup 232}Th-{sup 233}U fuel cycle with the very interesting breeding factor. The necessary precondition for its proper functionality is the online reprocessing within the MSR's fuel cycle. A lot of effort is given to the development of the on-line pyrochemical reprocessing technology. Among the processes considered to be suitable for this process, the electrochemical separations are highlighted. Within this context, it is necessary to study electrochemical behavior of both fissile material and fission products in the molten salt systems and to develop the processes of their extraction from the system at various stages of the whole cycle. In this contribution, the study of the electrochemical separation of U from Gd in the LiF-CaF{sub 2} molten system by the modulated current electrolysis is presented. The study of MSR-related electrochemistry will continue as a part of the new Czech MSR R and D program. The knowledge of the electrochemical behavior of actinides and lanthanides in molten fluorides is necessary to the development of any electrochemical separation process. In our previous work, electrochemical characteristics of uranium, thorium, as well as that of several lanthanides, were measured in LiF-BeF{sub 2} (the basic melt considered for the MSR concept) and Be-free melts (LiF-NaFKF, LiF-NaF, LiF-CaF{sub 2}), and added to available literature data. If the fluoride carrier melt is electrochemically stable enough, it is possible to deposit the element of interest in its basic, metallic state (e.g. uranium in all types of fluoride melts tested) by the electrolysis. In the case of insufficient electrochemical stability of the fluoride melt in which the actinides and lanthanides are dissolved for the experiments, it is possible to use a reactive working electrode made of a material which creates alloys with the element to be deposited. An alloying reaction can shift the deposition potential in a favorable direction and therefore achieve the separation of an element from the system. The use of reactive electrode can also improve the adhesion of the deposit. When two or more electroactive species are present in the system at the same time, the efficiency of their separation by an electrolysis depends on how different the deposition potentials of these two species are. If we input certain modulation to the electrolytic current, we can improve overall result of an electrolytic experiment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

New Czech R and D Program on Fluoride Salt-cooled Nuclear Reactor Systems

A technology of nuclear reactor systems with liquid molten salt fuel has been investigated in the Czech Republic since 1999. The original effort came from the national Partitioning and Transmutation concept based on the subcritical Accelerator Driven System for incineration of transuranium elements with liquid fluoride fuel and pyrochemical partitioning fuel cycle technology. After 2005, the original R and D intentions were gradually converted to classical Molten Salt Reactor (MSR) technology and to thorium - uranium fuel cycle. The aim of this choice was to contribute to the development of an advanced nuclear technology, which can minimize environmental impact of nuclear power, save the natural resources and which has some potential to be deployed also in a non-superpower country. The basic technological development of selected areas of MSR and Th - U fuel cycle technology was realized in the frame of the SPHINX project solved in 2005 - 2012. In addition to the domestic activities of the MSR technology development, in 2012 the Czech Ministry of Industry and Trade and the US - Department of Energy concluded an agreement about the collaborative R and D on Molten Salt Reactor and Fluoride-salt-cooled High-temperature Reactor (FHR) technologies. These facts and results created the principal background for the present national program on the MSR/FHR technology development which was approved by the Ministry of Industry and Trade in 2016 and is granted by the Technological Agency of the Czech Republic.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Neutronic Tests of Fluoride Salt Based MSR/FHR Coolants

Research on molten salt reactor (MSR) technology based on fluoride salts as a convenient way of higher actinides incineration and/or utilization of thorium fuel has been performed at UJV Rez and later Research Centre Rez in Czech Republic since 1999. Necessary steps towards to the deployment and possible commercialization were examined in past investigations. Neutronic testing is an inseparable part of the development and was at first carried out in a series of experiments with LiF-NaF salt (60 mol% LiF, 40 mol% NaF) subjected to a neutron field shaped by different moderators within the framework of the EROS experiments. Later, neutron spectrum was measured in this salt and significant variations in calculation to experiment C/E comparison were discovered. Further tests were performed with Flibe (66 mol% LiF, 33 mol% BeF{sub 2}) reactor coolant salt containing enriched lithium (99.994 mol% Li-7), which was provided by Oak Ridge National Laboratory (ORNL). This Flibe salt originally comes from the Molten Salt Reactor Experiment (MSRE) and was supplied to the Czech Republic within the framework of a collaborative agreement on MSR and fluoride-salt-cooled high-temperature reactor (FHR) research and development (R and D) concluded between the Czech Ministry of Industry and Trade and US Department of Energy in 2012. All experiments in this work were carried out at the LR-0 reactor at Research Centre Rez. The LR-0 is a pool-type zero power light water reactor with large and flexible experimental capabilities due to its design and fuel construction.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Feasibility of a Breed-and-Burn Molten Salt Reactor

This work addresses the feasibility and characteristics of the breed and burn (B and B) cycle for Molten Salt Reactors. In the general B and B cycle, natural Uranium or Thorium is used as fuel for a reactor which breeds a sufficient amount to maintain criticality at equilibrium without reprocessing. Interest in applying this scheme to an MSR has developed recently For the MSR B and B scheme considered here, fresh salt containing fertile material is continually fed into the core while salt with the composition of the homogeneous fuel is expelled without reprocessing. Through an analysis of the equilibrium conditions and startup requirements, the feasibility of such a cycle operating with various salts is assessed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

RADIOACTIVE WASTE STREAMS FROM VARIOUS POTENTIAL NUCLEAR FUEL CYCLE OPTIONS

Five fuel cycle options, about which little is known compared to more commonly known options, have been studied in the past year for the United States Department of Energy. These fuel cycle options, and their features relative to uranium-fueled light water reactor (LWR)-based fuel cycles, include: • Advanced once-through reactor concepts (Advanced Once-Through, or AOT) – intended for high uranium utilization and long reactor operating life, use depleted uranium in some cases, and avoid or minimize used fuel reprocessing • Fission-fusion hybrid (FFH) reactor concepts – potential variations are intended for high uranium or thorium utilization, produce fissile material for use in power generating reactors, or transmute transuranic (TRU) and some radioactive fission product (FP) isotopes • High temperature gas reactor (HTGR) concepts - intended for high uranium utilization, high reactor thermal efficiencies; they have unique fuel designs • Molten salt reactor (MSR) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, use on-line reprocessing of the used fuel, produce lesser amounts of long-lived, highly radiotoxic TRU elements, and avoid fuel assembly fabrication • Thorium/U-233 fueled LWR (Th/U-233) concepts – can breed fissile U-233 from Th fuel and avoid or minimize U fuel enrichment, and produce lesser amounts of long-lived, highly radiotoxic TRU elements. These fuel cycle options could result in widely different types and amounts of used or spent fuels, spent reactor core materials, and waste streams from used fuel reprocessing, such as: • Highly radioactive, high-burnup used metal, oxide, or inert matrix U and/or Th fuels, clad in Zr, steel, or composite non-metal cladding or coatings • Spent radioactive-contaminated graphite, SiC, carbon-carbon-composite, metal, and Be reactor core materials • Li-Be-F salts containing U, TRU, Th, and fission products • Ranges of separated or un-separated activation products, fission products, and actinides. Waste forms now used or studied for used LWR fuels can be used for some of these waste streams – but some waste forms may need to be developed for unique waste streams.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Current Status of VHTR Technology Development

Abstract – High Temperature Gas-cooled Reactors (HTGRs) featuring particle fuel reached the stage of commercial deployment in the mid-1980s with the Fort St.Vrain and Thorium HochTemperatur Reaktor feeding electricity to the grids in the United States and West Germany, respectively. The technology was then adopted by Japan and China with the operation of the High Temperature Test Reactor in Oarai, Japan and the High Temperature Reactor (HTR-10) in China. Increasing the outlet temperature of the HTGR to even higher temperatures above 900°C will improve the thermodynamic efficiency of the system and enable application of a new class of gas reactor, the very high temperature reactor, to provide process heat, electricity, and hydrogen to chemical industries with the attendant benefits of improved energy security and reduced CO2 emissions. However, the increase in coolant outlet temperature presents a number of technical challenges associated with fuel, materials, power conversion, and analysis methods for the reactor and hydrogen production. The U.S. Department of Energy is sponsoring a broad program of research and development with a goal of addressing the technical challenges over a broad range of outlet temperatures as part of the Next Generation Nuclear Plant Project. This paper describes the research and development activities that are currently underway to realize the technologies needed for an HTGR that features outlet temperatures of 750 to 950°C.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Selectivity, Kinetics, and Efficiency of Reversible Anion Exchange with TcO 4 - in a Supertetrahedral Cationic Framework

[ThB5O6(OH)6][BO(OH)2]·2.5H2O (Notre Dame Thorium Borate-1, NDTB-1) is an inorganic supertetrahedral cationic framework material that is derived from boric acid flux reactions. NDTB-1 exhibits facile single crystal to single crystal anion exchange with a variety of common anions such as Cl-, Br-, NO3-, IO3-, ClO4-, MnO4-, and CrO42-. More importantly, NDTB-1 is selective for the removal of TcO4- from nuclear waste streams even though there are large excesses of competing anions such as Cl-, NO3-, and NO2-. Competing anion exchange experiments and magic-angle spinning (MAS)-NMR spectroscopy of anion-exchanged NDTB-1 demonstrate that this unprecedented selectivity originates from the ability of NDTB-1 to trap TcO4- within cavities, whereas others remain mobile within channels in the material. The exchange kinetics of TcO4- in NDTB-1 are second-order with the rate constant k2 of 0.059 s-1 M-1. The anion exchange capacity of NDTB-1 for TcO4- is 162.2 mg g-1 (0.5421 mol mol-1) with a maximum distribution coefficient Kd of 1.0534 × 104 mL g-1. Finally, it is demonstrated that the exchange for TcO4- in NDTB-1 is reversible. TcO4- trapped in NDTB-1 can be exchanged out using higher-charged anions with a similar size such as PO43- and SeO42-, and therefore the material can be easily recycled and reused.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Synthesis of Actinide Fluoride Complexes Using Trimethyltin Fluoride as a Mild and Selective Fluorinating Reagent

Trimethyltin fluoride (Me₃SnF) is a mild and selective reagent for the installation of actinide fluoride bonds as demonstrated by the room temperature synthesis of a variety of organometallic and inorganic thorium(IV), uranium(IV), and uranium(V) fluoride complexes ((1,2,4-tBu₃C₅H₂)₂ThF₂, (C₅Me₅)₂U(F)(O-2,6-iPr₂C₆H₃), U(F)(O-2,6-tBu₂C₆H₃)₃, U(F)[N(SiMe₃)₂]₃ (C₅Me₅)₂UF₂(L) (L = O=PMe₃, O=PPh₃, O=PCy₃), and (C₅Me₅)₂U(F)(=N-2,6-iPr₂C₆H₃)) from their corresponding chloride, bromide, and iodide analogues. From these reactions, the new (C₅Me₅)₂UF₂(L) (L = O=PPh₃, O=PCy₃) uranium fluoride complexes were isolated and characterized by NMR spectroscopy and X-ray crystallography.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Strategies for the Photoreduction of Tc-99 Pertechnetate to Low-Valent Tc by Keggin Polyoxometalates

Technetium-99 (half-life of 2.1 × 10 5 yrs, β max = 0.29 MeV) is a hazardous radiological contaminant, which, in its predominant form of pertechnetate (TcO 4 – ), is highly mobile in the environment. Most strategies for the removal of pertechnetate from the environment involve uptake and/or absorption of pertechnetate using resins, clays, cationic metal-organic frameworks and even thorium borate ceramic like materials. Alternative approaches have involved the reduction and subsequent sequestration of lower valent technetium species using iron, sulfides, or iron sulfides. Here, our lab has explored this strategy using the lacunary alpha-2 Wells–Dawson polyoxometalate (α 2 -[P 2 W 17 O 61 ] 10– ) to both reduce and sequester lower valent technetium, and we have reported on the ligand features that stabilize the reduced species. In this work we investigate the potential of “plenary” Keggin POMs (XW 12 O 40 n – ) (X = P, Si, Al, n = 3, 4, 5, respectively) to both reduce TcO 4 – and stabilize the reduced Tc species. Specifically, we report on the mechanism by which the reduction of technetium occurs and find that PW 12 , SiW 12 , and AlW 12 promote the reduction of TcO 4 – to lower valent states. X-ray absorption spectroscopy was used to confirm a combination of Tc IV {in the form of TcO 2 · 2H 2 O and Tc 2 (µ-O) 2 4+ } and Tc V , which is subsequently complexed into a POM defect as a Tc V =O species.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Helium Release During Fracture and Granular Fragmentation of Rocks

Geogenic Helium-4 ( 4 He) in-situ increases locally in regions of large deformation generated naturally or anthropogenically. This gas release by deformation is a potential geochemical precursor signal for subsurface deformation. To evaluate the applicability of 4 He degassing for correlating deformation in different lithologies, we conducted high force crush tests, up to 97,800 N axial load, to assess the total 4 He released during fragmentation of the rocks. We observed that the highest 4 He released occurred in the sedimentary rocks and that release correlated strongly with lithologic age and U/Th content. Microstructural changes of the pre- and post-test rocks indicate that the degree of grain size reduction relates directly to the total 4 He released during crushing. The range of in-place 4 He was calculated based XRF measurements of uranium and thorium in each lithology, with the results indicating that the majority of the trapped 4 He was not released. However, the 4 He released by deformation depended upon how the each rock deformed during deformation and the degree of grain size reduction. We postulate that 4 He precursor signals can be used to understand subsurface deformation only if geomechanical and geochemical conditions for 4 He enrichment in a lithology are met.

Deformation signals↗

Rapid determination of 226Ra in environmental samples

A new rapid method for the determination of {sup 228}Ra in natural water samples has been developed at the SRNL/EBL (Savannah River National Lab/ Environmental Bioassay Laboratory) that can be used for emergency response or routine samples. While gamma spectrometry can be employed with sufficient detection limits to determine {sup 228}Ra in solid samples (via {sup 228}Ac) , radiochemical methods that employ gas flow proportional counting techniques typically provide lower MDA (Minimal Detectable Activity) levels for the determination of {sup 228}Ra in water samples. Most radiochemical methods for {sup 228}Ra collect and purify {sup 228}Ra and allow for {sup 228}Ac daughter ingrowth for ~36 hours. In this new SRNL/EBL approach, {sup 228}Ac is collected and purified from the water sample without waiting to eliminate this delay. The sample preparation requires only about 4 hours so that {sup 228}Ra assay results on water samples can be achieved in < 6 hours. The method uses a rapid calcium carbonate precipitation enhanced with a small amount of phosphate added to enhance chemical yields (typically >90%), followed by rapid cation exchange removal of calcium. Lead, bismuth, uranium, thorium and protactinium isotopes are also removed by the cation exchange separation. {sup 228}Ac is eluted from the cation resin directly onto a DGA Resin cartridge attached to the bottom of the cation column to purify {sup 228}Ac. DGA Resin also removes lead and bismuth isotopes, along with Sr isotopes and {sup 90}Y. La is used to determine {sup 228}Ac chemical yield via ICP-MS, but {sup 133}Ba can also be used instead if ICP-MS assay is not available. Unlike some older methods, no lead or strontium holdback carriers or continual readjustment of sample pH is required.

54 ENVIRONMENTAL SCIENCES↗