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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Computational Design of Advanced Nuclear Fuels

The objective of the project was to develop a method for theoretical understanding of nuclear fuel materials whose physical and thermophysical properties can be predicted from first principles using a novel dynamical mean field method for electronic structure calculations. We concentrated our study on uranium, plutonium, their oxides, nitrides, carbides, as well as some rare earth materials whose 4f eletrons provide a simplified framework for understanding complex behavior of the f electrons. We addressed the issues connected to the electronic structure, lattice instabilities, phonon and magnon dynamics as well as thermal conductivity. This allowed us to evaluate characteristics of advanced nuclear fuel systems using computer based simulations and avoid costly experiments.

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 OSTI2013

Time cycle analysis and simulation of material flow in MOX process layout

The (U,Pu)O{sub 2} MOX fuel is the driver fuel for the upcoming PFBR (Prototype Fast Breeder Reactor). The fuel has around 30% PuO{sub 2}. The presence of high percentages of reprocessed PuO{sub 2} necessitates the design of optimized fuel fabrication process line which will address both production need as well as meet regulatory norms regarding radiological safety criteria. The powder pellet route has highly unbalanced time cycle. This difficulty can be overcome by optimizing process layout in terms of equipment redundancy and scheduling of input powder batches. Different schemes are tested before implementing in the process line with the help of a software. This software simulates the material movement through the optimized process layout. The different material processing schemes have been devised and validity of the schemes are tested with the software. Schemes in which production batches are meeting at any glove box location are considered invalid. A valid scheme ensures adequate spacing between the production batches and at the same time it meets the production target. This software can be further improved by accurately calculating material movement time through glove box train. One important factor is considering material handling time with automation systems in place.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Determining the Release of Radionuclides from Tank Waste Residual Solids: FY2015 Report

Methodology development for pore water leaching studies has been continued to support Savannah River Site High Level Waste tank closure efforts. For FY2015, the primary goal of this testing was the achievement of target pH and Eh values for pore water solutions representative of local groundwater in the presence of grout or grout-representative (CaCO 3 or FeS) solids as well as waste surrogate solids representative of residual solids expected to be present in a closed tank. For oxidizing conditions representative of a closed tank after aging, a focus was placed on using solid phases believed to be controlling pH and E h at equilibrium conditions. For three pore water conditions (shown below), the target pH values were achieved to within 0.5 pH units. Tank 18 residual surrogate solids leaching studies were conducted over an E h range of approximately 630 mV. Significantly higher Eh values were achieved for the oxidizing conditions (ORII and ORIII) than were previously observed. For the ORII condition, the target Eh value was nearly achieved (within 50 mV). However, E h values observed for the ORIII condition were approximately 160 mV less positive than the target. E h values observed for the RRII condition were approximately 370 mV less negative than the target. Achievement of more positive and more negative E h values is believed to require the addition of non-representative oxidants and reductants, respectively. Plutonium and uranium concentrations measured during Tank 18 residual surrogate solids leaching studies under these conditions (shown below) followed the general trends predicted for plutonium and uranium oxide phases, assuming equilibrium with dissolved oxygen. The highest plutonium and uranium concentrations were observed for the ORIII condition and the lowest concentrations were observed for the RRII condition. Based on these results, it is recommended that these test methodologies be used to conduct leaching studies with actual Tank 18 residual solids material. Actual waste testing will include leaching evaluations of technetium and neptunium, as well as plutonium and uranium.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Flammability Analysis For Actinide Oxides Packaged In 9975 Shipping Containers

Packaging options are evaluated for compliance with safety requirements for shipment of mixed actinide oxides packaged in a 9975 Primary Containment Vessel (PCV). Radiolytic gas generation rates, PCV internal gas pressures, and shipping windows (times to reach unacceptable gas compositions or pressures after closure of the PCV) are calculated for shipment of a 9975 PCV containing a plastic bottle filled with plutonium and uranium oxides with a selected isotopic composition. G-values for radiolytic hydrogen generation from adsorbed moisture are estimated from the results of gas generation tests for plutonium oxide and uranium oxide doped with curium-244. The radiolytic generation of hydrogen from the plastic bottle is calculated using a geometric model for alpha particle deposition in the bottle wall. The temperature of the PCV during shipment is estimated from the results of finite element heat transfer analyses.

42 ENGINEERING↗
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 OSTI2011

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

The ENDF/B-VII.1 library is the latest revision to the United States Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 423 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [M. B. Chadwick et al., 'ENDF/B-VII.1 Nuclear Data for Science and Technology: Cross Sections, Covariances, Fission Product Yields and Decay Data,' Nuclear Data Sheets, 112, 2887 (2011)]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected {sup 235}U and {sup 239}Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as {sup 236}U, {sup 238,242}Pu and {sup 241,243}Am capture in fast systems. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for {sup 233}U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2011

ENDF/B-VII.1 Neutron Cross Section Data Testing with Critical Assembly Benchmarks and Reactor Experiments

The ENDF/B-VII.1 library is the latest revision to the United States' Evaluated Nuclear Data File (ENDF). The ENDF library is currently in its seventh generation, with ENDF/B-VII.0 being released in 2006. This revision expands upon that library, including the addition of new evaluated files (was 393 neutron files previously, now 418 including replacement of elemental vanadium and zinc evaluations with isotopic evaluations) and extension or updating of many existing neutron data files. Complete details are provided in the companion paper [1]. This paper focuses on how accurately application libraries may be expected to perform in criticality calculations with these data. Continuous energy cross section libraries, suitable for use with the MCNP Monte Carlo transport code, have been generated and applied to a suite of nearly one thousand critical benchmark assemblies defined in the International Criticality Safety Benchmark Evaluation Project's International Handbook of Evaluated Criticality Safety Benchmark Experiments. This suite covers uranium and plutonium fuel systems in a variety of forms such as metallic, oxide or solution, and under a variety of spectral conditions, including unmoderated (i.e., bare), metal reflected and water or other light element reflected. Assembly eigenvalues that were accurately predicted with ENDF/B-VII.0 cross sections such as unmoderated and uranium reflected 235U and 239Pu assemblies, HEU solution systems and LEU oxide lattice systems that mimic commercial PWR configurations continue to be accurately calculated with ENDF/B-VII.1 cross sections, and deficiencies in predicted eigenvalues for assemblies containing selected materials, including titanium, manganese, cadmium and tungsten are greatly reduced. Improvements are also confirmed for selected actinide reaction rates such as 236U capture. Other deficiencies, such as the overprediction of Pu solution system critical eigenvalues and a decreasing trend in calculated eigenvalue for 233U fueled systems as a function of Above-Thermal Fission Fraction remain. The comprehensive nature of this critical benchmark suite and the generally accurate calculated eigenvalues obtained with ENDF/B-VII.1 neutron cross sections support the conclusion that this is the most accurate general purpose ENDF/B cross section library yet released to the technical community.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2011

Ostwald Ripening and Its Effect on PuO2 Particle Size in Hanford Tank Waste

Between 1944 and 1989, the Hanford Site produced 60 percent (54.5 metric tons) of the United States weapons plutonium and produced an additional 12.9 metric tons of fuels-grade plutonium. High activity wastes, including plutonium lost from the separations processes used to isolate the plutonium, were discharged to underground storage tanks during these operations. Plutonium in the Hanford tank farms is estimated to be {approx}700 kg but may be up to {approx}1000 kg. Despite these apparent large quantities, the average plutonium concentration in the {approx}200 million liter tank waste volume is only about 0.003 grams per liter ({approx}0.0002 wt%). The plutonium is largely associated with low solubility metal hydroxide/oxide sludges where its low concentration and intimate mixture with neutron-absorbing elements (e.g., iron) are credited in nuclear criticality safety. However, concerns have been expressed that plutonium, in the form of plutonium hydrous oxide, PuO{sub 2} {center_dot} xH{sub 2}O, could undergo sufficient crystal growth through Ostwald ripening in the alkaline tank waste to potentially be separable from neutron absorbing constituents by settling or sedimentation. It was found that plutonium that entered the alkaline tank waste by precipitation through neutralization from acid solution is initially present as 2- to 3-nm (0.002- to 0.003-{mu}m) scale PuO{sub 2} {center_dot} xH{sub 2}O crystallite particles and grows from that point at exceedingly slow rates, posing no risk to physical segregation. These conclusions are reached by both general considerations of Ostwald ripening and specific observations of the behaviors of PuO{sub 2} and PuO{sub 2} {center_dot} xH{sub 2}O upon aging in alkaline solution.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

Relationships between experimental signatures and processing history for a variety of PuO 2 materials

The primary oxide of plutonium, PuO 2 , is a complex material due to ongoing change in lattice structure resulting from radioactive alpha decay. In addition, the element plutonium is complex, existing in various allotropes, which can influence formation of oxides due to different nucleation and growth mechanisms. The juxtaposition of these unique properties, along with formation conditions of the initial oxide, makes for PuO 2 materials with a wide variety of local and bulk experimental signatures. This work measures these signatures with X-ray diffraction and Raman spectroscopy from five different oxide materials, and relates these signatures to specific PuO 2 production and aging routes. Furthermore this information is of potential use in nuclear fuels processing and production, nuclear non-proliferation, forensics analyses, and interplanetary power sources.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Milestone report: The simulation of radiation driven gas diffusion in UO 2 at low temperature

Below 1000 K it is thought that fission gas diffusion in nuclear fuel during irradiation occurs through atomic mixing due to radiation damage. This is an important process for nuclear reactor performance as it affects fission gas release, particularly from the periphery of the pellet where such temperatures are normal. Here we present a molecular dynamics study of Xe and Kr diffusion due to irradiation. Thermal spikes and cascades have been used to study the electronic stopping and ballistic phases of damage, respectively. Our results predict that O and Kr exhibit the greatest diffusivity and U the least, while Xe lies in between. It is concluded that the ballistic phase does not sufficiently account for the experimentally observed diffusion. Preliminary thermal spike calculations indicate that the electronic stopping phase generates greater fission gas displacement than the ballistic phase, although further calculation must be carried out to confirm this. A good description of the system by the empirical potentials is important over the very wide temperatures induced during thermal spike and damage cascade simulations. This has motivated the development of a parameter set for gas-actinide and gas-oxygen interactions that is complementary for use with a recent many-body potential set. A comprehensive set of density functional theory (DFT) calculations were used to study Xe and Kr incorporation at a number of sites in CeO 2 , ThO 2 , UO 2 and PuO 2 . These structures were used to fit a potential, which was used to generate molecular dynamics (MD) configurations incorporating Xe and Kr at 300 K, 1500 K, 3000 K and 5000 K. Subsequent matching to the forces predicted by DFT for these MD configurations was used to refine the potential set. This fitting approach ensured weighted fitting to configurations that are thermodynamically significant over a broad temperature range, while avoiding computationally expensive DFT-MD calculations. The resultant gas potentials were validated against DFT binding energies and are suitable for simulating combinations of Xe and Kr in solid solutions of CeO 2 , ThO 2 , UO 2 and PuO 2 , providing a powerful tool for the atomistic simulation of conventional nuclear reactor fuel UO 2 as well as advanced MOX fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Americium Transmutation Feasibility When Used as Burnable Absorbers - 12392

The use of plutonium in Mixed Oxide (MOX) fuel in traditional Pressurized Water Reactor (PWR) assemblies leads to greater americium production which is not addressed in MOX recycling. The transuranic nuclides (TRU) contribute the most to the radiotoxicity of nuclear waste and a reduction of the TRU stockpile would greatly reduce the overall radiotoxicity of what must be managed. Am-241 is a TRU of particular concern because it is the dominant contributor of total radiotoxicity for the first 1000 years in a repository. This research explored the feasibility of transmuting Am-241 by using varying amounts in MOX rods being used in place of burnable absorbers and evaluated with respect to the impact on incineration and transmutation of transuranics in MOX fuel as well as the impact on safety. This research concludes that the addition of americium to a non-uniform fuel assembly is a viable method of transmuting Am-241, holding down excess reactivity in the core while serving as a burnable poison, as well as reducing the radiotoxicity of high level waste that must be managed. The use of Am/MOX hybrid fuel assemblies to transmute americium was researched using multiple computer codes. Am-241 was shown in this study to be able to hold down excess reactivity at the beginning of cycle and shape the power distribution in the core with assemblies of varying americium content loaded in a pattern similar to the traditional use of assemblies with varying amounts of burnable absorbers. The feasibility, safety, and utility of using americium to create an Am/MOX hybrid non-uniform core were also evaluated. The core remained critical to a burnup of 22,000 MWD/MTM. The power coefficient of reactivity as well as the temperature and power defects were sufficiently negative to provide a prompt feedback mechanism in case of a transient and prevent a power excursion, thus ensuring inherent safety and protection of the core. As shown here as well as many other studies, this non-uniform assembly type successfully addresses the concerns of reduced control rod worth within advanced MOX assemblies because the Am/MOX hybrid fuel rods are on the periphery of the assembly. The radiotoxicity of the high level waste that must be managed from a single use of an Am/MOX hybrid batch is reduced for the first 1000 years in comparison to a regular UO{sub 2} batch. However, there is no heat-load advantage without multi-recycling the Am/MOX hybrid assemblies. As shown through numerous studies, the heterogeneous assembly model used as the basis for this research is capable of multiple recycles while still maintaining negative temperature and power coefficients of reactivity. Multi-recycling would provide an even greater reduction to the radiotoxicity of the high level waste as well as provide a heat-load advantage compared to a once through UO{sub 2} fuel cycle. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Accident tolerant, non-oxidizable MOX fuel fabrication

Thorium dioxide (ThO{sub 2}) has a set of physical-chemical properties that make it well suited to serve as a nuclear fuel matrix: it has a very high melting point, it cannot be further oxidized, it has low leachability in water, reasonable thermal conductivity (better than UO{sub 2}), and of course it is fertile. Collectively, these properties allow for a fuel ceramic with increased in / out of core stability. As such, Thor Energy and a consortium of entities that include Westinghouse, KAERI, Fortum, NNL, ITU and IFE have recently fabricated (Th,Pu)O{sub 2} pellet specimens in Norway. The Th-MOX production methodology was developed over the course of 2 years of ceramic production trials involving Ce-Th oxide mixtures. It was understood that parameters developed for CeO{sub 2} material might need to have been modified if trial results with PuO{sub 2} led to sub-optimal pellets. Among observations made in this work, the most important production parameter was the preparation of constituent / mixed powders. Correct powder morphology was key to achieving high fuel densities. The process parameters developed for (Th,Ce)O{sub 2} pellets were adopted with the expectation that these would serve as a sound basis for obtaining dense, homogeneous (Th,Pu)O{sub 2} fuel ceramic. The Th-MOX fuel manufacture campaign was successful in producing enough (Th,Pu)O{sub 2} pellets to assemble two 30 cm long fuel pins for testing in the Halden test reactor. These fuel rods will be fitted with various instruments to enable temperature, fission gas release and dimensional changes to be measured while operating at power in conditions similar to those of a BWR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Study of the U{sub 3}O{sub 8} effect on MOX fuel sintering mechanisms

The manufacturing of uranium-plutonium mixed oxides fuel (also known as MOX) for light water reactor (LWR), starting generally from UO{sub 2} and PuO{sub 2} raw powders and resulting in (U,Pu)O{sub 2} pellets, involves several steps: i) preparation of a UO{sub 2}+PuO{sub 2} master blend with a Pu content of about 25-30 at% by milling and sieving, ii) dilution by mixing the master blend with non-milled UO{sub 2} to reach a Pu content of about 10 at%, iii) pressing and iv) sintering under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture. In this paper, triuranium octoxide (U{sub 3}O{sub 8}) is used as an additional raw powder in the LWR MOX fuel manufacturing. One of the objectives of this work is to understand the sintering behavior of U{sub 3}O{sub 8} under Ar / 4 % H{sub 2} + H{sub 2}O gas mixture and its effect when added to a UO{sub 2}-PuO{sub 2} mixture. A special attention is given on U{sub 3}O{sub 8} reduction, phase transitions, redox reactions and oxygen partial pressure (pO{sub 2}) evolution during shrinkage: the impact of these phenomena on the densification kinetics is discussed, especially the de-densification phenomenon which could occur at high temperature. For that purpose a specific dilatometer, coupled with a zirconia probe implemented at furnace inlet and outlet has been used which allows following continuously and simultaneously the fuel shrinkage and the oxygen partial pressure variation as a function of temperature. The experiments have shown that, contrary to UO{sub 2}, the sintering of milled PuO{sub 2} is accompanied by a de-densification at high temperature. This de-densification phenomenon already occurs with co-milled 71 at% UO{sub 2} + 29 at% PuO{sub 2}, but with a smaller magnitude. When pure milled U{sub 3}O{sub 8} is sintered, no de-densification phenomenon is observed but the low density of the sintered pellet could hide such a phenomenon. The incorporation of U{sub 3}O{sub 8} into a UO{sub 2}-PuO{sub 2} master blend containing 29 at% of plutonium leads to an exacerbation of the de-densification phenomenon with a higher loss on ignition suggesting that U{sub 3}O{sub 8} contributes also to the de-densification phenomenon. This could be due to high temperature reactions involving the initial carbon content of the raw U{sub 3}O{sub 8} powder. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Performance evaluation of two-stage fuel cycle from SFR to PWR

One potential fuel cycle option being considered is a two-stage fuel cycle system involving the continuous recycle of transuranics in a fast reactor and the use of bred plutonium in a thermal reactor. The first stage is a Sodium-cooled Fast Reactor (SFR) fuel cycle with metallic U-TRU-Zr fuel. The SFRs need to have a breeding ratio greater than 1.0 in order to produce fissile material for use in the second stage. The second stage is a PWR fuel cycle with uranium and plutonium mixed oxide fuel based on the design and performance of the current state-of-the-art commercial PWRs with an average discharge burnup of 50 MWd/kgHM. This paper evaluates the possibility of this fuel cycle option and discusses its fuel cycle performance characteristics. The study focuses on an equilibrium stage of the fuel cycle. Results indicate that, in order to avoid a positive coolant void reactivity feedback in the stage-2 PWR, the reactor requires high quality of plutonium from the first stage and minor actinides in the discharge fuel of the PWR needs to be separated and sent back to the stage-1 SFR. The electricity-sharing ratio between the 2 stages is 87.0% (SFR) to 13.0% (PWR) for a TRU inventory ratio (the mass of TRU in the discharge fuel divided by the mass of TRU in the fresh fuel) of 1.06. A sensitivity study indicated that by increasing the TRU inventory ratio to 1.13, The electricity generation fraction of stage-2 PWR is increased to 28.9%. The two-stage fuel cycle system considered in this study was found to provide a high uranium utilization (>80%). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Simultaneous Thermal Analysis of Remediated Nitrate Salt Surrogates

The actinide engineering and science group (MET-1) have completed simultaneous thermal analysis and offgas analysis by mass spectrometry (STA-MS) of remediated nitrate salt (RNS) surrogates formulated by the high explosives science and technology group (M-7). The 1.0 to 1.5g surrogate samples were first analyzed as received, then a new set was analyzed with 100-200mL 10M HNO 3 +0.3 MHF added, and a third set was analyzed after 200 mL of a concentrated Pu-AM spike (in 10M HNO 3 +0.3 MHF) was added. The acid and spike solutions were formulated by the actinide analytical chemistry group (C-AAC) using reagent-grade HNO 3 and HF, which was also used to dissolve a small quantity of mixed, high-fired PuO 2 / AmO 2 oxide.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗