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All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Selective Extraction of Uranium from Liquid or Supercritical Carbon Dioxide

Current liquid-liquid extraction processes used in recycling irradiated nuclear fuel rely on (1) strong nitric acid to dissolve uranium oxide fuel, and (2) the use of aliphatic hydrocarbons as a diluent in formulating the solvent used to extract uranium. The nitric acid dissolution process is not selective. It dissolves virtually the entire fuel meat which complicates the uranium extraction process. In addition, a solvent washing process is used to remove TBP degradation products, which adds complexity to the recycling plant and increases the overall plant footprint and cost. A liquid or supercritical carbon dioxide (l/sc-CO 2 ) system was designed to mitigate these problems. Indeed, TBP nitric acid complexes are highly soluble in l/sc-CO 2 and are capable of extracting uranium directly from UO 2 , UO 3 and U 3 O 8 powders. This eliminates the need for total acid dissolution of the irradiated fuel. Furthermore, since CO 2 is easily recycled by evaporation at room temperature and pressure, it eliminates the complex solvent washing process. In this report, we demonstrate: (1) A reprocessing scheme starting with the selective extraction of uranium from solid uranium oxides into a TBP-HNO 3 loaded Sc-CO 2 phase, (2) Back extraction of uranium into an aqueous phase, and (3) Conversion of recovered purified uranium into uranium oxide. The purified uranium product from step 3 can be disposed of as low level waste, or mixed with enriched uranium for use in a reactor for another fuel cycle. After an introduction on the concept and properties of supercritical fluids, we first report the characterization of the different oxides used for this project. Our extraction system and our online monitoring capability using UV-Vis absorbance spectroscopy directly in sc-CO 2 is then presented. Next, the uranium extraction efficiencies and kinetics is demonstrated for different oxides and under different physical and chemical conditions: l/sc-CO 2 pressure and temperature, TBP/HNO 3 complex used, reductant or complexant used for selectivity, and ionic liquids used as supportive media. To complete the extraction and recovery cycle, we then demonstrate uranium back extraction from the TBP loaded sc-CO 2 phase into an aqueous phase and the characterization of the uranium complex formed at the end of this process. Another aspect of this project was to limit proliferation risks by either co-extracting uranium and plutonium, or by leaving plutonium behind by selectively extracting uranium. We report that the former is easily achieved, since plutonium is in the tetravalent or hexavalent oxidation state in the oxidizing environment created by the TBP-nitric acid complex, and is therefore co-extracted. The latter is more challenging, as a reductant or complexant to plutonium has to be used to selectively extract uranium. After undertaking experiments on different reducing or complexing systems (e.g., AcetoHydroxamic Acid (AHA), Fe(II), ascorbic acid), oxalic acid was chosen as it can complex tetravalent actinides (Pu, Np, Th) in the aqueous phase while allowing the extraction of hexavalent uranium in the sc-CO 2 phase. Finally, we show results using an alternative media to commonly used aqueous phases: ionic liquids. We show the dissolution of uranium in ionic liquids and its extraction using sc-CO 2 with and without the presence of AHA. The possible separation of trivalent actinides from uranium is also demonstrated in ionic liquids using neodymium as a surrogate and diglycolamides as the extractant.

carbone dioxide↗
DOE OSTI2011

Decay Heat Calculations for PWR and BWR Assemblies Fueled with Uranium and Plutonium Mixed Oxide Fuel using SCALE

In currently operating commercial nuclear power plants (NPP), there are two main types of nuclear fuel, low enriched uranium (LEU) fuel, and mixed-oxide uranium-plutonium (MOX) fuel. The LEU fuel is made of pure uranium dioxide (UO{sub 2} or UOX) and has been the fuel of choice in commercial light water reactors (LWRs) for a number of years. Naturally occurring uranium contains a mixture of different uranium isotopes, primarily, {sup 235}U and {sup 238}U. {sup 235}U is a fissile isotope, and will readily undergo a fission reaction upon interaction with a thermal neutron. {sup 235}U has an isotopic concentration of 0.71% in naturally occurring uranium. For most reactors to maintain a fission chain reaction, the natural isotopic concentration of {sup 235}U must be increased (enriched) to a level greater than 0.71%. Modern nuclear reactor fuel assemblies contain a number of fuel pins potentially having different {sup 235}U enrichments varying from {approx}2.0% to {approx}5% enriched in {sup 235}U. Currently in the United States (US), all commercial nuclear power plants use UO{sub 2} fuel. In the rest of the world, UO{sub 2} fuel is still commonly used, but MOX fuel is also used in a number of reactors. MOX fuel contains a mixture of both UO{sub 2} and PuO{sub 2}. Because the plutonium provides the fissile content of the fuel, the uranium used in MOX is either natural or depleted uranium. PuO{sub 2} is added to effectively replace the fissile content of {sup 235}U so that the level of fissile content is sufficiently high to maintain the chain reaction in an LWR. Both reactor-grade and weapons-grade plutonium contains a number of fissile and non-fissile plutonium isotopes, with the fraction of fissile and non-fissile plutonium isotopes being dependent on the source of the plutonium. While only RG plutonium is currently used in MOX, there is the possibility that WG plutonium from dismantled weapons will be used to make MOX for use in US reactors. Reactor-grade plutonium in MOX fuel is generally obtained from reprocessed irradiated nuclear fuel, whereas weapons-grade plutonium is obtained from decommissioned nuclear weapons material and thus has a different plutonium (and other actinides) concentration. Using MOX fuel instead of UOX fuel has potential impacts on the neutronic performance of the nuclear fuel and the design of the nuclear fuel must take these differences into account. Each of the plutonium sources (RG and WG) has different implications on the neutronic behavior of the fuel because each contains a different blend of plutonium nuclides. The amount of heat and the number of neutrons produced from fission of plutonium nuclides is different from fission of {sup 235}U. These differences in UOX and MOX do not end at discharge of the fuel from the reactor core - the short- and long-term storage of MOX fuel may have different requirements than UOX fuel because of the different discharged fuel decay heat characteristics. The research documented in this report compares MOX and UOX fuel during storage and disposal of the fuel by comparing decay heat rates for typical pressurized water reactor (PWR) and boiling water reactor (BWR) fuel assemblies with and without weapons-grade (WG) and reactor-grade (RG) MOX fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Defining the Shape Density of Cerium Oxide and Titanium Dioxide Aerosol

The “shape density” of CeO 2 (cerium oxide) powder (American Elements, Los Angeles CA) is estimated to be 0.82 ± 0.02 g cm -3 and the shape density for TiO 2 (titanium dioxide) powder (American Elements, Los Angeles CA) is estimated as 0.54 ± 0.01 g cm -3 . The “shape density” is defined as the measured loose density (ASTM 2018 and ASTM 2021) of a powder divided by the ICRP (International Commission on Radiological Protection) default aerosol shape factor (Valetin 2002). This information is used for experimental measurements of the RRF (respirable release factor) and other respirable source term parameter (US DOE 1994) experiments at the Los Alamos National Laboratory (LANL). The shape density is the input parameter to the TSI Inc (Shoreview MN) AIM (Aerosol Instrument Manager) software for the TSI model 3321 APS (Aerodynamic Particle Sizer) (Figure 1).

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2012

Analysis of burnup and isotopic compositions of BWR 9 x 9 UO{sub 2} fuel assemblies

In order to extend isotopic composition data focusing on fission product nuclides, measurements are progressing using facilities of JAEA for five samples taken from high burnup BWR 9 x 9 UO{sub 2} fuel assemblies. Neutronics analysis with an infinite assembly model was applied to the preliminary measurement data using a continuous-energy Monte Carlo burnup calculation code MVP-BURN with nuclear libraries based on JENDL-3.3 and JENDL-4.0. The burnups of the samples were determined to be 28.0, 39.3, 56.6, 68.1, and 64.0 GWd/t by the Nd-148 method. They were compared with those calculated using node-average irradiation histories of power and in-channel void fractions which were taken from the plant data. The comparison results showed that the deviations of the calculated burnups from the measurements were -4 to 3%. It was confirmed that adopting the nuclear data library based on JENDL-4.0 reduced the deviations of the calculated isotopic compositions from the measurements for {sup 238}Pu, {sup 144}Nd, {sup 145}Nd, {sup 146}Nd, {sup 148}Nd, {sup 134}Cs, {sup 154}Eu, {sup 152}Sm, {sup 154}Gd, and {sup 157}Gd. On the other hand, the effect of the revision in the nuclear. data library on the neutronics analysis was not significant for major U and Pu isotopes. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Dose reconstruction for plutonium-239 intakes at the Rocky Flats Plant

Purpose: The Rocky Flats (RF) Plant was a weapons manufacturing facility that operated from the early 1950s to 1989. Its primary missions were the production of plutonium (Pu) pits for thermonuclear weapons and the processing of retired weapons for Pu recovery. The purpose of this study was to estimate radiation doses to a cohort of 4499 RF workers from an intake of 239 Pu, the primary plutonium isotope handled at the site. Materials and methods: The latest biokinetic models of the International Commission on Radiological Protection, or site-specific variations of those models, were used to estimate 239 Pu intakes for each worker based on model fits to bioassay data often coupled with lung measurements. Results: Urinary excretion and lung retention data for most 239 Pu intakes could be fit reasonably well by a mixture of Pu dioxide and moderately soluble material. For some workers, better fits were obtained by application of other absorption types including Type S, 239 Pu nitrate, or pure 239 Pu dioxide, or by assuming intake via a wound. The lungs typically received the highest tissue doses, with fifty-year committed equivalent doses in the range of 0.5–1 Sv for 275 workers, 1–5 Sv for 115 workers, and greater than 5 Sv for 12 workers. Conclusions: RF was a unique site regarding a large number of lung measurements available for determining the appropriate absorption types for inhaled material. Further, this provided higher confidence in reconstructed 239 Pu doses than is generally gained from urinary data alone.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Cocrystallization of Uranyl and Plutonyl Nitrate Hexahydrates

In the system UO{sub 2}(NO{sub 3}){sub 2}–PuO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O with the ratio Pu/(U + Pu) = 0.3–0.5 and stabilization of the pair U(VI)–Pu(VI), the crystallization of (U, Pu)O{sub 2}(NO{sub 3}){sub 2}·6H{sub 2}O occurs in accordance with the metric of the phase diagram of the system UO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O with coefficient of cocrystallization of plutonium 0.804 ± 0.028. This condition holds in the sector of the working lines of the crystallization process that correspond to the minimum content of uranium in the mother solution. The degree of enrichment of the mother solution is determined by the crystallization temperature, i.e., the mass fraction of the mother solution in the total mass of the system. The constancy of the coefficient of cocrystallization with the technologically important parameters of the process makes it possible to use the metric of the solubility diagram of the system UO{sub 2}(NO{sub 3}){sub 2}–HNO{sub 3}–H{sub 2}O to calculate the crystallization of (U, Pu)O{sub 2}(NO{sub 3}){sub 2}·6H{sub 2}O with the crystalline phase with the required composition being obtained.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Evolution of Chemical Conditions and Estimated Plutonium Solubility in the Residual Waste Layer During Post-Closure Aging of Tank 18

This document updates the Eh-pH transitions from grout aging simulations and the plutonium waste release model of Denham (2007, Rev. 1) based on new data. New thermodynamic data for cementitious minerals are used for the grout simulations. Newer thermodynamic data, recommended by plutonium experts (Plutonium Solubility Peer Review Report, LA-UR-12-00079), are used to estimate solubilities of plutonium at various pore water compositions expected during grout aging. In addition, a new grout formula is used in the grout aging simulations and apparent solubilities of coprecipitated plutonium are estimated using data from analysis of Tank 18 residual waste. The conceptual model of waste release and the grout aging simulations are done in a manner similar to that of Denham (2007, Rev. 1). It is assumed that the pore fluid composition passing from the tank grout into the residual waste layer controls the solubility, and hence the waste release concentration of plutonium. Pore volumes of infiltrating fluid of an assumed composition are reacted with a hypothetical grout block using The Geochemist's Workbench{reg_sign} and changes in pore fluid chemistry correspond to the number of pore fluid volumes reacted. As in the earlier document, this results in three states of grout pore fluid composition throughout the simulation period that are termed Reduced Region II, Oxidized Region II, and Oxidized Region III. The one major difference from the earlier document is that pyrite is used to account for reducing capacity of the tank grout rather than pyrrhotite. This poises Eh at -0.47 volts during Reduced Region II. The major transitions in pore fluid composition are shown. Plutonium solubilities are estimated for discrete PuO2(am,hyd) particles and for plutonium coprecipitated with iron phases in the residual waste. Thermodynamic data for plutonium from the Nuclear Energy Agency are used to estimate the solubilities of the discrete particles for the three stages of pore fluid evolution. In Denham (2007, Rev. 1), the solubilities in the oxidized regions were estimated at Eh values in equilibrium with dissolved oxygen. Here, these are considered to be maximum possible solubilities because Eh values are unlikely to be in equilibrium with dissolved oxygen. More realistic Eh values are estimated here and plutonium solubilities calculated at these are considered more realistic. Apparent solubilities of plutonium that coprecipitated with iron phases are estimated from Pu:Fe ratios in Tank 18 residual waste and the solubilities of the host iron phases. The estimated plutonium solubilities are shown. Uncertainties in the grout simulations and plutonium solubility estimates are discussed. The primary uncertainty in the grout simulations is that little is known about the physical state of the grout as it ages. The simulations done here are pertinent to a porous medium, which may or may not be applicable to fractured grout, depending on the degree and nature of the fractures. Other uncertainties that are considered are the assumptions about the reducing capacity imparted by blast furnace slag, the effects of varying dissolved carbon dioxide and oxygen concentrations, and the treatment of silica in the simulations. The primary uncertainty in the estimates of plutonium solubility is that little is known about the exact form of plutonium in the residual waste. Other uncertainties include those inherent in the thermodynamic data, pH variations from those estimated in the grout simulations, the effects of the treatment of silica in the grout simulations, and the effect of varying total dissolved carbonate concentrations. The objective of this document is to update the model for solubility controls on release of plutonium from residual waste in closed F-Area waste tanks. The update is based on new information including a new proposed grout formulation, chemical analysis of Tank 18 samples and more current thermodynamic data for plutonium and grout minerals. In addition, minor changes to the modeling of the grout chemical evolution have been made. It should be noted that the intent is to provide bounding solubilities for plutonium to be used in Performance Assessment modeling rather than trying to identify an exact concentration of plutonium in pore fluids released from a tank at any given time. This document also considers suggestions and opportunities for improvement regarding the plutonium modeling assumptions from the Plutonium Solubility Peer Review Report, LA-UR-12-00079.

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

Quantification of calculation accuracy for code systems in burn-up credit applications by recalculations of experimental data - 14480

For application in criticality safety, numerical code systems require profound validation against experimental data. In terms of burn-up credit, this refers to the inventory determination as well as the subsequent criticality calculation. Within this framework, a dedicated, conservative estimation of uncertainties, bias and bias uncertainties for the resulting calculated multiplication factor k{sub eff} is indispensable. However, measures to reduce excess conservatism may become desirable, depending on the application case. For this reason GRS implemented a methodology to validate a calculation code system for criticality safety analyses for systems with irradiated nuclear fuel. The work at hand demonstrates this methodology by means of the GRS home-built inventory determination system KENOREST and ORNL's well-known SCALE 6 CSASS criticality safety calculation sequence. A number of 40 radiochemical analysis samples taken from SFCOMPO and elsewhere have been calculated to derive isotopic correction factors in different ways, depending on the required level of conservatism and the calculational efforts to be carried out. This includes bounding methods as well as different Monte Carlo sampling approaches. Furthermore, from the ICSBEP a number of 515 critical benchmark experiment configurations comprising water moderated low enriched UO{sub 2} fuel rods and water moderated mixed oxide (MOX) fuel rods have been selected and evaluated. Important absorber and structure materials are also covered. Various trending and sensitivity and uncertainty analysis methods have been applied, again to derive conservative estimates on the various contributions to k{sub eff} uncertainties. The overall methodology is applicable to low enriched, water moderated PWR uranium dioxide fuels in a burn-up range of about 10 to 60 GWd/tHM, and validates the reactivity contributions of the major uranium and plutonium isotopes. The methodology itself has no limitation which requires the exclusion of further isotopes as e.g. fission products. However, its application is constrained by the experimental database being publicly available. After the potential provision of an extended database in the future, the range of applicability can easily be extended to more nuclides. In addition, it is not limited to the code systems being used here but can also be applied using different numerical tools. For a generic PWR spent fuel pool rack and different enrichment and burn-up combinations, the impact of the current level of validation on k{sub eff} is demonstrated and quantified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Nondestructive and destructive assay for forensics characterization of weapons-grade plutonium produced in LEU irradiated in a thermal neutron spectrum

A post-irradiation examination (PIE) of an enriched uranium sample was performed to advance a nuclear forensics methodology, based on intra-elemental nuclide ratios. The PIE was carried out on a few milligrams of low enriched uranium dioxide (LEUO 2 ) sample, irradiated to a burnup of 1 GWd/MTU. The PIE of the LEUO 2 sample provided the concentration of selected plutonium and fission product nuclides ( 239 Pu, 240 Pu, 241 Pu, 91 Y, 95 Zr, 95 Nb, 103 Ru, 133 Cs, 134 Cs, 135 Cs, 137 Cs, 136 Ba, 138 Ba, 140 Ba, 140 La, 141 Ce, 144 Ce, 149 Sm, 150 Sm, 152 Sm, 153 Eu, and 154 Eu). The PIE supported validation of the MCNP (Monte Carlo N-Particle neutronics simulation code) predicted concentration of these nuclides in the irradiated LEUO 2 sample. This study showed that most of the MCNP predicted nuclide concentrations were within 15% and several within 10% compared to the PIE results. Furthermore, gamma and mass spectrometry of nuclides in irradiated LEUO 2 enabled the determination of uranium burnup and time since irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Static electric dipole polarizabilities of An{sup 5+/6+} and AnO{sub 2}{sup +/2+} (An = U, Np, and Pu) ions

The parallel components of static electric dipole polarizabilities have been calculated for the lowest lying spin-orbit states of the penta- and hexavalent oxidation states of the actinides (An) U, Np, and Pu, in both their atomic and molecular diyl ion forms (An{sup 5+/6+} and AnO{sub 2}{sup +/2+}) using the numerical finite-field technique within a four-component relativistic framework. The four-component Dirac-Hartree-Fock method formed the reference for MP2 and CCSD(T) calculations, while multireference Fock space coupled-cluster (FSCC), intermediate Hamiltonian Fock space coupled-cluster (IH-FSCC) and Kramers restricted configuration interaction (KRCI) methods were used to incorporate additional electron correlation. It is observed that electron correlation has significant (∼5 a.u.{sup 3}) impact upon the parallel component of the polarizabilities of the diyls. To the best of our knowledge, these quantities have not been previously reported and they can serve as reference values in the determination of various electronic and response properties (for example intermolecular forces, optical properties, etc.) relevant to the nuclear fuel cycle and material science applications. The highest quality numbers for the parallel components (α{sub zz}) of the polarizability for the lowest Ω levels corresponding to the ground electronic states are (in a.u.{sup 3}) 44.15 and 41.17 for UO{sub 2}{sup +} and UO{sub 2}{sup 2+}, respectively, 45.64 and 41.42 for NpO{sub 2}{sup +} and NpO{sub 2}{sup 2+}, respectively, and 47.15 for the PuO{sub 2}{sup +} ion.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Building flexibility into the design of a pilot plant for the immobilisation of Pu containing residues and wastes

NNL and ANSTO on behalf of Sellafield Ltd have developed a process for the immobilisation of a range of Pu containing wastes and residues. Following the inactive demonstration of the technology the project is now focusing on the design of an active pilot plant capable of validating the technology and ultimately immobilising a waste inventory containing around 100 kg plutonium. The diverse wastes from which it is uneconomic to recover Pu, require a flexible process with a wide product envelope capable of producing a wasteform suitable for disposal in a UK repository. Ceramics, glass ceramics and metal encapsulated waste-forms can be delivered by the process line which incorporates size reduction and heat treatment techniques with the aim of feeding a hot isostatic pressing process designed to deliver the highly durable waste-forms. Following a demonstration of feasibility, flowsheet development is progressing to support the design which has the aim of a fully flexible facility based in NNL's Central Laboratory on the Sellafield site. Optimisation of the size reduction, mixing and blending operations is being carried out using UO{sub 2} as a surrogate for PuO{sub 2}. This work is supporting the potential of using an enhanced glass ceramic formulation in place of the full ceramic with the aim of simplifying glove box operations. Heat treatment and subsequent HIPing strategies are being explored in order to eliminate any carbon from the feeds without increasing the valence state of the uranium present in some of the inventory which can result in an unwanted increase in wasteform volumes. The HIP and ancillary systems are being specifically designed to meet the requirements of the Sellafield site and within the constraints of the NNL Central Laboratory. The HIP is being configured to produce consolidated product cans consistent with the requirements of ongoing storage and disposal. With the aim of one cycle per day, the facility will deliver its mission of immobilising the identified waste and residues inventory within 3 years. During that period it will also be used to demonstrate the potential of this technology to deliver the immobilisation of a proportion of the UK plutonium stockpile that may not be suitable for use as MOx fuel should that decision be taken. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2012

Advanced Lithium Ion Venus Explorer (ALIVE)

The COncurrent Multidisciplinary Preliminary Assessment of Space Systems (COMPASS) Team partnered with the Applied Research Laboratory to perform a NASA Innovative Advanced Concepts (NIAC) Program study to evaluate chemical based power systems for keeping a Venus lander alive (power and cooling) and functional for a period of days. The mission class targeted was either a Discovery ($500M) or New Frontiers ($750M to $780M) class mission. Historic Soviet Venus landers have only lasted on the order of 2 hours in the extreme Venus environment:temperatures of 460 degrees Centigrade and pressures of 93 bar. Longer duration missions have been studied using plutonium powered systems to operate and cool landers for up to a year. However, the plutonium load is very large. This NIAC study sought to still provide power and cooling but without the plutonium. Batteries are far too heavy but a system which uses the atmosphere (primarily carbon dioxide) and on on-board fuel to power a power generation and cooling system was sought. The resuling design was the Advanced Long-Life Lander Investigating the Venus Environment (ALIVE) Spacecraft (S/C) which burns lithium (Li) with the CO2 atmosphere to heat a Duplex Stirling to power and cool the lander for a 5-day duration (until the Li is exhausted). While it does not last years a chemical powered system surviving days eliminates the cost associated with utilizing a flyby relay S/C and allows a continuous low data rate direct to earth (DTE) link in this instance from the Ovda Regio of Venus. The five-day collection time provided by the chemical power systems also enables science personnel on earth to interact and retarget science - something not possible with an approximately 2-hour spacecraft lifetime. It also allows for contingency operations directed by the ground (reduced risk). The science package was based on that envisioned by the Venus Intrepid Tessera Lander (VITaL) Decadal Survey Study. The Li Burner within the long duration power system creates approximately 14000 W of heat. This 1300 degree Centigrade heat using Li in the bottom "ballast" tank is melted to liquid by the Venus temperature, drawn into a furnace by a wick and burned with atmospheric CO2. The Li carbonate exhaust is liquid at 1300 degrees Centigrade and being denser than Li drains into the the Li tank and solidifies. Since the exhaust product is a dense liquid no "chimney" is required which conserves the heat for the stirling power convertor. The Duplex Stirling provides about 300 W of power and removes about 300 W of heat from the avionics and heat that leaks into the 1-bar-insulated payload pressure vessel kept at 25 degrees Centigrade. The Na K radiator is run to the top of the drag flap.The ALIVE vehicle is carried to Venus via an Atlas 411 launch vehicle (LV) with a C3 of 7 km2/s2. An Aeroshell, derived from the Genesis mission, enables a direct entry into the atmosphere of Venus (-10 degrees Centigrade, 40 g max) and 6 m/s for landing (44 g) using a drag ring. For surface science and communication, a 100 WRF (WebEx Recording Format), X-Band 0.6-meter pointable DTE (Direct-to-Earth) antenna provides 2 kbps (kilobits per second) to DSN (Deep-Space Network) 34-meter antenna clusters.Table 1.1 summarizes the top-level details of each subsystem that was incorporated into the design. Cost estimates of the ALIVE mission show it at approximately $760M which puts it into the New Frontiers class.The ALIVE landed duration is only limited by the amount of Li which can be carried by the lander. Further studies are needed to investigate how additional mass can be carried, perhaps by a larger launcher and larger aeroshell.

Exploration↗
DOE OSTI2015

Characterization of the kinetics of NF 3 -fluorination of NpO 2

The exploitation of selected actinide and fission product fluoride volatilities has long been considered as a potentially attractive compact method for recycling used nuclear fuels to avoid generating the large volumes of radioactive waste arising from aqueous reprocessing . The most developed process uses the aggressive and hazardous fluorinating agents hydrogen fluoride (HF) and/or molecular fluorine (F 2 ) at high temperatures to volatilize the greatest fraction of the used nuclear fuel into a single gas stream. The volatilized fluorides are subsequently separated using a series of fractionation and condensation columns to recover the valuable fuel constituents and fission products. In pursuit of a safer and less complicated approach, we investigated an alternative fluoride volatility-based process using the less hazardous fluorinating agent nitrogen trifluoride (NF 3 ) and leveraging its less aggressive nature to selectively evolve fission product and actinide fluorides from the solid phase based on their reaction temperatures into a single recycle stream. In this approach, successive isothermal treatments using NF 3 will first evolve the more thermally susceptible used nuclear fuel constituents leaving the other constituents in the residual solids until subsequent isothermal temperature treatments cause these others to volatilize. During investigation of this process, individual neat used fuel components were treated with isothermal NF 3 in an attempt to characterize the kinetics of each fluorination reaction to provide input into the design of a new volatile fluoride separations approach. In these directed investigations, complex behavior was observed between NF 3 and certain solid reactants such as the actinide oxides of uranium, plutonium, and neptunium. Given the similar thermal reaction susceptibilities of neptunium oxide (NpO 2 ) and uranium dioxide (UO 2 ) and the importance of Np and U, we initially focused our efforts on determining the reaction kinetic parameters for NpO 2 . Characterizing the NF3 fluorination of NpO 2 using established models for gas-solid reactions [16] proved unsuccessful so we developed a series of successive fundamental reaction mechanisms to characterize the observed successive fluorination reactions leading to production of the volatile neptunium hexafluoride (NpF 6 ).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

MOX recycling in GEN 3 + EPR Reactor homogeneous and stable full MOX core

In the case of the EPR (European Pressurized Reactor) reactor, 100% MOX core management is possible with simple design adaptations which are not significantly costly. 100% MOX core management offers several highly attractive advantages. First, it is possible to have the same plutonium content in all the rods of a fuel assembly instead of having rods with 3 different plutonium contents, as in MOX assemblies in current PWRs. Secondly, the full MOX core is more homogeneous. Thirdly, the stability of the core is significantly increased due to a large reduction in the Xe effect. Fourthly, there is a potential for the performance of the MOX fuel to match that of new high performance UO{sub 2} fuel (enrichment up to 4.95 %) in terms of increased burn up and cycle length. Fifthly, since there is only one plutonium content, the manufacturing costs are reduced. Sixthly, there is an increase in the operating margins of the reactor, and in the safety margins in accident conditions. The use of 100% MOX core will improve both utilisation of natural uranium resources and reductions in high level radioactive waste inventory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

UO{sub 2} and PuO{sub 2} utilization in high temperature engineering test reactor with helium coolant

High temperature engineering test reactor (HTTR) is one of high temperature gas cooled reactor (HTGR) types which has been developed by Japanese Atomic Energy Research Institute (JAERI). The HTTR is a graphite moderator, helium gas coolant, 30 MW thermal output and 950 °C outlet coolant temperature for high temperature test operation. Original HTTR uses UO{sub 2} fuel. In this study, we have evaluated the use of UO{sub 2} and PuO{sub 2} in form of mixed oxide (MOX) fuel in HTTR. The reactor cell calculation was performed by using SRAC 2002 code, with nuclear data library was derived from JENDL3.2. The result shows that HTTR can obtain its criticality condition if the enrichment of {sup 235}U in loaded fuel is 18.0% or above.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Water Interactions with Actinide Oxides from First Principles: A Computational Study - 16470

The interactions between water and the actinide oxides UO{sub 2} and PuO{sub 2} are important when considering the long-term storage of spent nuclear fuel. However, experimental studies in this area are severely limited by plutonium's intense radioactivity, and hence we have recently begun to investigate these interactions computationally. In this contribution we report the results of first principles calculations of the electronic and geometric structures of AnO{sub 2} bulk, AnO{sub 2} clusters, and AnO{sub 2} surfaces (An = U, Np, Pu), as well as the interaction of water with the latter. Strongly-correlated effects are taken into account using a Hubbard corrected potential, which enables us to perform efficacious plane-wave density functional calculations of extended systems. In particular, we compare results of water adsorption on UO{sub 2} (111) with the corresponding results on CeO{sub 2}, focusing on the energetics and structural properties of molecular versus dissociative adsorption, for both partially and fully covered surfaces. (authors)

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

Comparison of Advanced Processing of Irradiated UO{sub 2} Fuels and NpO{sub 2} Targets

Advanced processing methods for UO{sub 2}-based used nuclear fuels have been developed and demonstrated in several experimental programs at the Oak Ridge National Laboratory's Radiochemical Engineering Development Center (REDC). The Consolidated Fuel Recycle Program (1979-1986) and the Advanced Fuel Cycle Initiative (2002- 2008) were sponsored by the U.S. Department of Energy Office of Nuclear Energy (DOE/NE). Currently, DOE/NE is administering a new program for the National Aeronautics and Space Administration to produce {sup 238}Pu by irradiating NpO{sub 2}, and the irradiated targets are to be chemically processed using the same equipment and shielded facilities in the REDC that were used previously to process irradiated UO{sub 2}-based fuels. Flowsheets for processing irradiated NpO{sub 2} targets are now being developed and demonstrated. There are both similarities and differences in the processing of the two types of irradiated fuels/targets, and a comparison of these is useful for application to new types of feed material. A comparison of similarities and differences for processing UO{sub 2}- and NpO{sub 2}-based feed materials has been useful in planning for advanced processing methods and new requirements. Use of existing multistage solvent extraction equipment and shielded processing facilities at the ORNL REDC has enabled development, demonstration, and application of new and improved separations methods. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Sintering effects on NpO 2 grain size and morphology: The role of precursor Np phase

Neptunium dioxide (NpO 2 ) is a key phase in nuclear material processing as a target material for the production of plutonium-238 ( 238 Pu) and has been historically synthesized via the calcination of a Np oxalate precursor. Alternative synthesis methods for NpO 2 are now more prevalent, necessitating their study and comparison with the more common oxalate route. The purpose of this work was to investigate the microstructural properties of NpO 2 synthesized from a nitrate-based Np precursor phase via the assessment of NpO 2 particle size and morphology as a function of calcination temperature and time. Scanning electron microscopy (SEM) was used to probe the primary grain size and morphology of NpO 2 after calcination at temperatures ranging from 700 to 1100 °C and hold times ranging from 1 to 10 h. Post-image analysis using ImageJ software enabled the quantification of mean particle diameter. This analysis indicated that particle diameter increases with both increasing calcination temperature and hold time. Primary particles were shown to be clumped in irregular patterns into the overall rough, blocky aggregates, but this macroscopic morphology was not affected by calcination time or temperature. Although trends in primary grain size of NpO 2 were consistent with available literature from other Np precursor phases, the macroscopic morphology of the NpO 2 aggregates was quite different than reported for other precursors. Through comparison with historical literature on Np oxalate, this work emphasizes the importance of Np precursor on the physical properties of NpO 2 .

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗