Reaction Rate Ratios for Recent Fast Metal Experiments with Large Plutonium Masses
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73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
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Caustic Neutralization and Precipitation of Acidic Dissolved Simulated Stainless Steel–Clad Plutonium and Plutonium/Uranium Nuclear Fuel
Simulated dissolved stainless steel (SS) clad Pu and Pu/U nuclear fuel in HNO 3 was neutralized to a free hydroxide (OH - ) concentration of 0.6 M. A thermal neutron poison, Gd, was added to the simulants at concentrations of either ~3 - 6 g/L or ~37 - 38 g/L. The supernate Pu concentration the day of neutralization ranged from 0.48 to 8.75 mg/L. The supernate Pu concentration of a simplified simulant neutralized to 0.6 M OH - above precipitated solids containing Pu was demonstrated to decrease over 18 days. A significant portion of precipitated Pu was found to be insoluble in 8 M HNO 3 at ambient temperature, but essentially quantitative Pu dissolution was achieved in 11.5 M HNO 3 /0.1 M KF at 100 °C. The difficulty in dissolving the Pu precipitate is believed to be due to the formation of refractory PuO 2 •xH 2 O during the neutralization process. Initial Gd concentrations of ~37 – 38 g/L were found to result in a greater Al precipitation when neutralized to 0.6 M OH - than initial Gd concentrations of ~3 – 6 g/L. Physical properties of the resultant slurries were measured and used to calculate limiting flowrates and slurry velocities by gravity only in transfer piping between the Savannah River Site’s H-Canyon Facility and the Concentration, Storage, and Transfer Facility (CSTF). These results were compared to calculated deposition velocities to predict if solids would settle during the transfer. The Newtonian model was found to be reasonable for each diluted slurry evaluated. Deposition velocities of Pu containing slurries are lower than nuclear fuel slurries primarily composed of U due to the high density of Pu solids. In conclusion, dilution of slurries reduces the margin between the slurry and deposition velocities due to the reduction in viscosity because higher viscous forces on the particles promote maintained suspension.
Actinide Neutralization↗
Interpretation of Urinary Excretion Data From Plutonium Wound Cases Treated With DTPA: Application of Different Models and Approaches
After a chelation treatment, assessment of intake and doses is the primary concern of an internal dosimetrist. Using the urinary excretion data from two actual wound cases encountered at Los Alamos National Laboratory (LANL), this study discusses several methods that can be used to interpret intakes from the urinary data collected after one or multiple chelation treatments. One of the methods uses only the data assumed to be unaffected by chelation (data collected beyond 100 d after the last treatment). This method, used by many facilities for official dose records, was implemented by employing maximum likelihood analysis and Bayesian analysis methods. The impacts of an improper assumption about the physicochemical behavior of a radioactive material and the importance of the use of a facility-specific biokinetic model when available have also been demonstrated. Another method analyzed both the affected and unaffected urinary data using an empirical urinary excretion model. This method, although case-specific, was useful in determining the actual intakes and the doses averted or the reduction in body burdens due to chelation treatments. Finally, this approach was important in determining the enhancement factors, the behavior of the chelate, and other observations that may be pertinent to several DTPA compartmental modeling approaches being conducted by the health physics community.
62 RADIOLOGY AND NUCLEAR MEDICINE↗
Analysis of features of hydrodynamics and heat transfer in the fuel assembly of prospective sodium reactor with a high rate of reproduction in the uranium-plutonium fuel cycle
The fast sodium reactor fuel assembly (FA) with U–Pu–Zr metallic fuel is described. In comparison with a “classical” fast reactor, this FA contains thin fuel rods and a wider fuel rod grid. Studies of the fluid dynamics and the heat transfer were carried out for such a new FA design. The verification of the ANSYS CFX code was provided for determination of the velocity, pressure, and temperature fields in the different channels. The calculations in the cells and in the FA were carried out using the model of shear stress transport (SST) selected at the stage of verification. The results of the hydrodynamics and heat transfer calculations have been analyzed.
42 ENGINEERING↗
Determination of the Galaxy age by the method of uranium–thorium–plutonium isotopic ratios
The dependence of the Galaxy age (Т{sub G}), as determined by the method of uranium–thorium isotopic ratios, on the parameters of the nucleosynthesis model is studied within the theory of galactic nucleosynthesis. It is shown that ТG depends strongly both on the scenario of the production of nuclei in the r-process and those features of neutron-rich nuclei that are used in the respective analysis and on galactic-nucleosynthesis parameters. The effect of a sudden nucleosynthesis spike before the formation of a solar system on the Galaxy age is evaluated. The region of admissible values of the parameters of galacticnucleosynthesis theory is discussed. The method of uranium–thorium isotopic ratios is supplemented with the {sup 244}Pu/{sup 238}U ratio for yet another cosmochronometer pair, and the Galaxy age is estimated on the basis of the model modified in this way.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Redox potential evolution of nitric species and plutonium in HNO{sub 3}-HNO{sub 2} system
In France, the reprocessing process of spent fuel is carried out using the Purex process. The first chemical step of this process is the dissolution of the spent fuel in aqueous concentrated nitric acid. This dissolution solution is composed of many oxidizing species and it is necessary to understand the chemistry of these species to apprehend the behavior of structural materials. The redox potential is a useful discriminate variable to differentiate the effect of each species. In this framework, only the effect of NO{sub 3}{sup -}/HNO{sub 2} couple and PuO{sub 2}{sup 2+}/Pu{sup 4+} couple were investigated. These species were chosen because one is the main couple of medium and the other has a high standard potential. Their influence was investigated from the Nernst's equation. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Toward an Improved Understanding of Structure and Magnetism in Neptunium and Plutonium Phosphonates and Sulfonates
This grant supported the exploratory synthesis of new actinide materials with all of the actinides from thorium to californium with the exceptions of protactinium and berkelium. We developed detailed structure-property relationships that allowed for the identification of novel materials with selective ion-exchange, selective oxidation, and long-range magnetic ordering. We found novel bonding motifs and identified periodic trends across the actinide series. We identified structural building units that would lead to desired structural features and novel topologies. We also characterized many different spectroscopic trends across the actinide series. The grant support the preparation of approximately 1200 new compounds all of which were structurally characterized.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Literature Review on the Sorption of Plutonium, Uranium, Neptunium, Americium and Technetium to Corrosion Products on Waste Tank Liners
The Savannah River Site (SRS) has conducted performance assessment (PA) calculations to determine the risk associated with closing liquid waste tanks. The PA estimates the risk associated with a number of scenarios, making various assumptions. Throughout all of these scenarios, it is assumed that the carbon-steel tank liners holding the liquid waste do not sorb the radionuclides. Tank liners have been shown to form corrosion products, such as Fe-oxyhydroxides (Wiersma and Subramanian 2002). Many corrosion products, including Fe-oxyhydroxides, at the high pH values of tank effluent, take on a very strong negative charge. Given that many radionuclides may have net positive charges, either as free ions or complexed species, it is expected that many radionuclides will sorb to corrosion products associated with tank liners.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Uncertainty Measurement for Trace Element Analysis of Uranium and Plutonium Samples by Inductively Coupled Plasma-Atomic Emission Spectrometry (ICP-AES) and Inductively Coupled Plasma-Mass Spectrometry (ICP-MS)
The measurement uncertainty estimation associated with trace element analysis of impurities in U and Pu was evaluated using the Guide to the Expression of Uncertainty Measurement (GUM). I this evaluation the uncertainty sources were identified and standard uncertainties for the components were categorized as either Type A or B. The combined standard uncertainty was calculated and a coverage factor k = 2 was applied to obtain the expanded uncertainty, U. The ICP-AES and ICP-MS methods used were developed for the multi-element analysis of U and Pu samples. A typical analytical run consists of standards, process blanks, samples, matrix spiked samples, post digestion spiked samples and independent calibration verification standards. The uncertainty estimation was performed on U and Pu samples that have been analyzed previously as part of the U and Pu Sample Exchange Programs. Control chart results and data from the U and Pu metal exchange programs were combined with the GUM into a concentration dependent estimate of the expanded uncertainty. Comparison of trace element uncertainties obtained using this model was compared to those obtained for trace element results as part of the Exchange programs. This process was completed for all trace elements that were determined to be above the detection limit for the U and Pu samples.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Sorption Behavior and Morphology of Plutonium in the Presence of Goethite at 25 and 80°C
In this study, we examined the sorption behavior of Pu at elevated temperatures in the presence of one relevant mineral, goethite (α-FeOOH), over a range of concentrations that span solubility-controlled to adsorption-controlled concentrations. We focused on the sorptive behavior of two common forms of Pu: aqueous Pu(IV) and intrinsic Pu(IV) nano-colloids at 25 and 80°C in a dilute pH 8 NaCl/NaHCO 3 solution. The morphology of Pu sorbed to goethite was characterized using transmission electron microscopy (TEM). We examined the relative stability of PuO 2 precipitates, PuO 2 nano-colloids, Pu 4 O 7 surface precipitates, and monomeric sorbed Pu as a function of temperature and over a time scale of months.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Decontamination of Plutonium from Fluoride and Chloride during Oxalate Precipitation, Filtration and Calcination Processes
Due to analytical limitations for the determination of fluoride (F) and chloride (Cl) in a previous anion exchange study, an additional study of the decontamination of Pu from F and Cl by oxalate precipitation, filtration and calcination was performed. Anion product solution from the previous impurity study was precipitated as an oxalate, filtered, and calcined to produce an oxide for analysis by pyrohydrolysis for total Cl and F. Analysis of samples from this experiment achieved the purity specification for Cl and F for the proposed AFS-2 process. Decontamination factors (DF's) for the overall process (including anion exchange) achieved a DF of ~5000 for F and a DF of ~100 for Cl. Similar experiments where both HF and HCl were spiked into the anion product solution to a ~5000 μg /g Pu concentration showed a DF of 5 for F and a DF of 35 for Cl across the combined precipitation-filtration-calcination process steps.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Plutonium Metallurgy [Slides]
Due to its nuclear properties, Pu will remain a material of global interest well into the future. Processing, Structure, Properties and Performance remains a good framework for discussion of Pu materials science Self-irradiation and aging effects continue to be central in discussions of Pu metallurgy Pu in its elemental form is extremely unstable, but alloying helps to stabilize Pu; but, questions remain as to how and why this stabilization occurs. Which is true Pu-Ga binary phase diagram: US or Russian? Metallurgical issues such as solute coring, phase instability, crystallographic texture, etc. result in challenges to casting, processing, and properties modeling and experiments. For Ga alloyed FCC stabilized Pu, temperature and pressure remain as variables impacting phase stability.
36 MATERIALS SCIENCE↗
Impact of Fission Products Impurity on the Plutonium Content in PWR MOX Fuels
This report presents the results of a neutronics analysis done in response to the charter IFCA-SAT-2 entitled 'Fuel impurity physics calculations'. This charter specifies that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies (UOX SNF) is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate PWR MOX fuel assemblies. Only non-gaseous FP have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1). This mixture of Pu and FP is called Pu FP . Note that, in this preliminary analysis, the FP losses are considered element-independent, i.e., for example, 1% of FP losses mean that 1% of all non-gaseous FP leak into the Pu stream.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Technical Basis for DOE Standard 3013 Equivalency Supporting Reduced Temperature Stabilization of Oxalate-Derived Plutonium Dioxide Produced by the HB-Line Facility at Savannah River Site
This report documents the technical basis for determining that stabilizing highpurity PuO 2 derived from oxalate precipitation at the SRS HB-Line facility at a minimum of 625 °C for at least four hours in an oxidizing atmosphere is equivalent to stabilizing at a minimum of 950 °C for at least two hours as regards meeting the objectives of stabilization defined by DOE-STD-3013 if the material is handled in a way to prevent excessive absorption of water.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Corrective Action Investigation Plan for Corrective Action Unit 571: Area 9 Yucca Flat Plutonium Dispersion Sites, Nevada National Security Site, Nevada, Revision 0
CAU 571 is a grouping of sites where there has been a suspected release of contamination associated with nuclear testing. This document describes the planned investigation of CAU 571, which comprises the following corrective action sites (CASs): • 09-23-03, Atmospheric Test Site S-9F • 09-23-04, Atmospheric Test Site T9-C • 09-23-12, Atmospheric Test Site S-9E • 09-23-13, Atmospheric Test Site T-9D • 09-45-01, Windrows Crater These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the investigation report. The sites will be investigated based on the data quality objectives (DQOs) developed on March 6, 2013, by representatives of the Nevada Division of Environmental Protection and the U.S. Department of Energy (DOE), National Nuclear Security Administration Nevada Site Office (now the Nevada Field Office). The DQO process was used to identify and define the type, amount, and quality of data needed to develop and evaluate appropriate corrective actions for CAU 571. The site investigation process will also be conducted in accordance with the Soils Activity Quality Assurance Plan, which establishes requirements, technical planning, and general quality practices to be applied to this activity. The potential contamination sources associated with CAU 571 CASs are from nuclear testing activities. The DQO process resulted in an assumption that total effective dose (TED) within a default contamination boundary exceeds the final action level (FAL) and requires corrective action. The presence and nature of contamination outside the default contamination boundaries at CAU 571 will be evaluated based on information collected from a field investigation. Radiological contamination will be evaluated based on a comparison of the TED at sample locations to the dose-based FAL. The TED will be calculated as the total of separate estimates of internal and external dose. Results from the analysis of soil samples will be used to calculate internal radiological dose. Thermoluminescent dosimeters placed at the center of each sample location will be used to measure external radiological dose. Chemical contamination will be evaluated by comparing soil sample results to the FAL. Appendix A provides a detailed discussion of the DQO methodology and the DQOs specific to each CAS.
54 ENVIRONMENTAL SCIENCES↗
Summary of Plutonium-238 Production Alternatives Analysis (Final Report)
The Team implemented a two-phase evaluation process. During the first phase, a wide variety of past and new candidate facilities and processing methods were assessed against the criteria established by DOE for this assessment. Any system or system element selected for consideration as an alternative within the project to reestablish domestic production of Pu-238 must meet the following minimum criteria: Any required source material must be readily available in the United States, without requiring the development of reprocessing technologies or investments in systems to separate material from identified sources. It must be cost, schedule, and risk competitive with existing baseline technology. Any identified facilities required to support the concept must be available to the program for the entire project life cycle (notionally 35 years, unless the concept is so novel as to require a shorter duration). It must present a solution that can generate at least 1.5 Kg of Pu-238 oxide per year, for at least 35 years. It must present a low-risk, near-term solution to the National Aeronautics and Space Administration’s urgent mission need. DOE has implemented this requirement by eliminating from project consideration any alternative with key technologies at less than Technology Readiness Level 5. The Team evaluated the options meeting these criteria using a more detailed assessment of the reasonable facility variations and compared them to the preferred option, which consists of target irradiation at the Advanced Test Reactor (ATR) and the High Flux Isotope Reactor (HFIR), target fabrication and chemical separations processing at the ORNL Radiochemical Engineering Development Center, and neptunium 237 storage at the Materials and Fuels Complex at INL. This preferred option is consistent with the Records of Decision from the earlier National Environmental Policy Act (NEPA) documentation
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Corrective Action Investigation Plan for Corrective Action Unit 568: Area 3 Plutonium Dispersion Sites Nevada National Security Site, Nevada
CAU 568 is a grouping of sites where there has been a suspected release of contamination associated with nuclear testing. This document describes the planned investigation of CAU 568, which comprises the following corrective action sites (CASs): (1) 03-23-17, S-3I Contamination Area; (2) 03-23-19, T-3U Contamination Area; (3) 03-23-20, Otero Contamination Area; (4) 03-23-22, Platypus Contamination Area; (5) 03-23-23, San Juan Contamination Area; and (6) 03-23-26, Shrew/Wolverine Contamination Area. These sites are being investigated because existing information on the nature and extent of potential contamination is insufficient to evaluate and recommend corrective action alternatives (CAAs). Additional information will be obtained by conducting a corrective action investigation before evaluating CAAs and selecting the appropriate corrective action for each CAS. The results of the field investigation will support a defensible evaluation of viable CAAs that will be presented in the investigation report.
54 ENVIRONMENTAL SCIENCES↗
Corrective Action Decision Document/Closure Report for Corrective Action Unit 571: Area 9 Yucca Flat Plutonium Dispersion Sites, Nevada National Security Site, Nevada, Revision 0 with ROTC 1
The purpose of this CADD/CR is to provide documentation and justification that no further corrective action is needed for the closure of CAU 571 based on the implementation of corrective actions. This includes a description of investigation activities, an evaluation of the data, and a description of corrective actions that were performed. The CAIP provides information relating to the scope and planning of the investigation. Therefore, that information will not be repeated in this document.
54 ENVIRONMENTAL SCIENCES↗