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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 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2012

Reanalysis of the gas-cooled fast reactor experiments at the zero power facility proteus - Spectral indices

The gas-cooled fast reactor (GCFR) concept was investigated experimentally in the PROTEUS zero power facility at the Paul Scherrer Inst. during the 1970's. The experimental program was aimed at neutronics studies specific to the GCFR and at the validation of nuclear data in fast spectra. A significant part of the program used thorium oxide and thorium metal fuel either distributed quasi-homogeneously in the reference PuO{sub 2}/UO{sub 2} lattice or introduced in the form of radial and axial blanket zones. Experimental results obtained at the time are still of high relevance in view of the current consideration of the Gas-cooled Fast Reactor (GFR) as a Generation-IV nuclear system, as also of the renewed interest in the thorium cycle. In this context, some of the experiments have been modeled with modern Monte Carlo codes to better account for the complex PROTEUS whole-reactor geometry and to allow validating recent continuous neutron cross-section libraries. As a first step, the MCNPX model was used to test the JEFF-3.1, JEFF-3.1.1, ENDF/B-VII.0 and JENDL-3.3 libraries against spectral indices, notably involving fission and capture of {sup 232}Th and {sup 237}Np, measured in GFR-like lattices. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fragmentation Prediction for a Brittle Ceramic: a Two-Scale Model Approach

Glass and oxide-ceramic radioactive material forms (e.g., PuO{sub 2}) are susceptible to brittle fracture during handling and transportation. Fracture of these materials may result in the generation of fragments small enough to become airborne, and thus easily dispersed and inhaled. Entrainment and dispersal of particles with dimensions less than 10 μm is generally anticipated under normal aerodynamic conditions. The assessment of the risk to the public and the environment from these materials due to events resulting in their fracture is based on the determination of the source term for environmental release. This source term is typically quantified by an airborne release fraction (ARF) and a respirable fraction (RF) of material generated by a given event. While the recommended empirical bounding values prove satisfactory to comply with the regulatory process for handling and transportation of these materials, they are based on test data for only a few materials subjected to a limited range of loading conditions. Because the final state of fragmentation (in terms of the size and distribution of fragments) is strongly dependent not only on the properties of the material and the energetic loading conditions that lead to breakup, but also on the many intrinsic length scales associated with fragmentation, it is difficult to precisely predict fragmentation characteristics for other brittle solid wastes forms subjected to a wider range of loading conditions based on the test data alone. The purpose of this manuscript is to present a two-scale modeling approach that may be used to simulate the dynamic fragmentation of brittle solid radioactive material forms subjected to general loading and boundary conditions. The method provides the resulting fragment characteristics (average size and size distribution) across the entire fragment range of interest. At the macro-scale, the approach uses a gradient damage elastic continuum mechanics material model within a finite element framework to simulate dynamic fracture under general loading and boundary conditions. Results from the macroscopic calculations are passed to a one-dimensional (1-D) lower length scale model that incorporates the relevant fine scale effects to determine the micro-scale fragmentation. This 1-D micro-scale fragmentation model simulates the internal crack nucleation and opening processes to predict the average fragment size and the fragment size distribution as a function of the material properties, loading conditions, and microscopic characteristic length scale of the material. Results from the lower length scale model are then combined with the macro-scale model fragmentation information to determine the resulting fragment size distribution spanning both length scales. This two-scale model has been exercised on a representative problem drawn from experimental studies, consisting of an impact of a falling weight onto a brittle ceramic. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2019

Sensitivity of an (α,n) neutron source to (α,n) cross sections and stopping powers

The first-order sensitivities of the (α,n) neutron source rate density in a homogeneous material to isotopic (α,n) cross sections, elemental electronic stopping power data, and general coefficients for the nuclear stopping power are derived for the total neutron source and the energy-dependent neutron source. The physics bases for some of the equations in the SOURCES4C calculation of the stopping power are briefly discussed. The derivatives have been implemented in SOURCES4C. In simple PuO 2 and PuB 2 test problems, the computed derivatives compare extremely well with central differences. As a result, the first-order sensitivities are applied to show that improving the value of one of SOURCES4C’s stopping-power coefficients changes the (α,n) neutron source rate by 1.738 × 10 –6 % in the PuO 2 problem.

(α,n) neutron source↗
DOE OSTI2022

Solubility controls on plutonium and americium release in subsurface environments exposed to acidic processing wastes

To identify the role of waste composition and sediment interactions in controlling Pu and Am mobility in contaminated sediments at Hanford, a legacy nuclear site, Pu and Am concentrations in solutions equilibrated with contaminated sediments from beneath the 216-Z-9 (Z-9) Trench were compared to the solubilities of PuO 2 materials, synthesized by methods representative of the disposed wastes, in the absence of sediments. Furthermore, this work shows that the solubilities of PuO 2 materials synthesized by different methods, and with varying particle sizes, agree with PuO 2 (am,hyd), although dissolution kinetics differed between materials. According to saturation index (SI) calculations, PuO 2 (am,hyd) is likely also controlling Pu release from sediments under conditions where phosphate concentrations are low. However, both Pu-phosphate and Am-phosphate phases, identified in SI calculations and by high resolution transmission electron microscopy, play roles in controlling release in low pH, high phosphate, shallow sediments just below the Z-9 Trench. The elevated phosphate is likely due to decomposition of tributyl phosphate from waste solutions over time. Sediments from deeper in the subsurface beneath the Z-9 Trench are less acidic and contain less phosphate, with Pu solubility likely controlled by PuO 2 (am,hyd) that precipitated following neutralization of the acidic waste stream. Controls on Am concentrations in deeper sediments are more complex and potentially involve sediment adsorption and/or release from Pu (1-x) Am x O 2 following Am in-growth. The concentrations of both Pu and Am were elevated in the colloidal fraction associated with shallow sediments, but not in PuO 2 experiments, suggesting the presence of Pu/Am pseudocolloids (e.g., Pu/Am associated with mineral colloids). However, Pu and Am association with the colloidal size fraction was not observed in deeper sediments, suggesting transport of Pu and Am to these depths beneath the Z-9 Trench was not due to colloidal transport.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Crystal growth methods dedicated to low solubility actinide oxalates

Two novel crystal growth syntheses dedicated to low solubility actinide-oxalate systems and adapted to glove box handling are described. These methods based on the use of precursors of either actinide metal or oxalic acid have been optimized on lanthanide systems (analogue of actinides(III)) and then assessed on real actinide systems. They allow the synthesis of several actinide oxalate single crystals, Am{sub 2}(C{sub 2}O{sub 4}){sub 3}(H{sub 2}O){sub 3}·xH{sub 2}O, Th(C{sub 2}O{sub 4}){sub 2}·6H{sub 2}O, M{sub 2+x}[Pu{sup IV}{sub 2−x}Pu{sup III}{sub x}(C{sub 2}O{sub 4}){sub 5}]·nH{sub 2}O and M{sub 1−x}[Pu{sup III}{sub 1−x}Pu{sup IV}{sub x}(C{sub 2}O{sub 4}){sub 2}·H{sub 2}O]·nH{sub 2}O. It is the first time that these well-known compounds are formed by crystal growth methods, thus enabling direct structural studies on transuranic element systems and acquisition of basic data beyond deductions from isomorphic (or not) lanthanide compounds. Characterizations by X-ray diffraction, UV–visible solid spectroscopy, demonstrate the potentialities of these two crystal growth methods to obtain oxalate compounds. - Graphical abstract: Two new single crystal growth methods dedicated to actinide oxalate compounds. - Highlights: • Use of diester as oxalate precursor for crystal growth of actinide oxalates. • Use of actinide oxide as precursor for crystal growth of actinide oxalates. • Crystal growth of Pu(III) and Am(III) oxalates. • Crystal growth of mixed Pu(III)/Pu(IV) oxalates.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Host-guest prospects of neodymium and gadolinium ultraphosphate frameworks for nuclear waste storage: Multi-temperature topological analysis of nanoporous cages in RP{sub 5}O{sub 14}

Highlights: • The temperature-dependent cage topologies of RP{sub 5}O{sub 14} (R = Nd or Gd) were examined. • A topological analysis revealed similar cage structure for RP{sub 5}O{sub 14} (R = Nd or Gd). • One exception: the 8/4 cage topology of NdP{sub 5}O{sub 14} lacks structural cross-linking. • Volume (UO{sub 2} or PuO{sub 2}) ≫ 8/4 topology in RP{sub 5}O{sub 14} (R = Nd or Gd). • RP{sub 5}O{sub 14} (R = Nd or Gd) are good prospective host media for nuclear waste storage. - Abstract: Rare-earth ultraphosphate (RP{sub 5}O{sub 14}) framework materials are potential host media for nuclear waste storage, since cages within their nanoporous structures have volumes that match well those of prospective guests such as uranium or plutonium ions. Good volume matches of host cages and guest compounds are nonetheless not the only structural requirement for ensuring viable nuclear waste storage. Host structures also need to be stable enough to withstand the typical environmental conditions of long-term storage of spent nuclear fuel. To this end, the nanoporous cage topologies of neodymium and gadolinium ultraphosphate, NdP{sub 5}O{sub 14} and GdP{sub 5}O{sub 14}, are investigated as a function of temperature. Topological analysis shows that, while both compounds are essentially isostructural, thus displaying the same type of cage structures, the cage volumes of NdP{sub 5}O{sub 14} are significantly larger than those of GdP{sub 5}O{sub 14}, with one stark exception. This exception concerns the cage of NdP{sub 5}O{sub 14} whose 8/4 topology lacks structural cross-linking that would otherwise give it much more strength. This 'squashed' cage appears to owe its origins to the specific nature of crystallographic twinning in NdP{sub 5}O{sub 14}, which causes strain that needs alleviating; squashing the 8/4 cage in NdP{sub 5}O{sub 14} would be the most susceptible option towards this end, since its cage manifests the weakest structural construct. GdP{sub 5}O{sub 14} succumbs to a gradual contraction of its 8/4 cage only with increasing temperature above 300 K; this is well below its second-order monoclinic-to-orthorhombic phase transition at 420–430 K; which is 420 K for NdP{sub 5}O{sub 14}. Fortunately, the volumes of heavy metal ions that arise from spent nuclear fuels do not match the size requirements of cages with 8/4 topology in RP{sub 5}O{sub 14} hosts needed for encapsulation; otherwise, radiation leakage of such containment would present a risk. Both NdP{sub 5}O{sub 14} and GdP{sub 5}O{sub 14} would therefore seem to offer good prospects as host media for nuclear waste storage.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Renewability and sustainability aspects of nuclear energy

Renewability and sustainability aspects of nuclear energy have been presented on the basis of two different technologies: (1) Conventional nuclear technology; CANDU reactors. (2) Emerging nuclear technology; fusion/fission (hybrid) reactors. Reactor grade (RG) plutonium, {sup 233}U fuels and heavy water moderator have given a good combination with respect to neutron economy so that mixed fuel made of (ThO{sub 2}/RG‐PuO{sub 2}) or (ThC/RG-PuC) has lead to very high burn up grades. Five different mixed fuel have been selected for CANDU reactors composed of 4 % RG‐PuO{sub 2} + 96 % ThO{sub 2}; 6 % RG‐PuO{sub 2} + 94 % ThO{sub 2}; 10 % RG‐PuO{sub 2} + 90 % ThO{sub 2}; 20 % RG‐PuO{sub 2} + 80 % ThO{sub 2}; 30 % RG‐PuO{sub 2} + 70 % ThO{sub 2}, uniformly taken in each fuel rod in a fuel channel. Corresponding operation lifetimes have been found as ∼ 0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼ 30 000, 60 000, 100 000, 200 000 and 290 000 MW.d/ton, respectively. Increase of RG‐PuO{sub 2} fraction in radial direction for the purpose of power flattening in the CANDU fuel bundle has driven the burn up grade to 580 000 MW.d/ton level. A laser fusion driver power of 500 MW{sub th} has been investigated to burn the minor actinides (MA) out of the nuclear waste of LWRs. MA have been homogenously dispersed as carbide fuel in form of TRISO particles with volume fractions of 0, 2, 3, 4 and 5 % in the Flibe coolant zone in the blanket surrounding the fusion chamber. Tritium breeding for a continuous operation of the fusion reactor is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. Fission reactions in the MA fuel under high energetic fusion neutrons have lead to the multiplication of the fusion energy by a factor of M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5 % TRISO volume fraction at start up, respectively. Alternatively with thorium, the same fusion driver would produce ∼160 kg {sup 233}U per year in addition to fission energy production in situ, multiplying the fusion energy by a factor of ∼1.3.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Chemical speciation of U, Fe, and Pu in melt glass from nuclear weapons testing

Nuclear weapons testing generates large volumes of glassy materials that influence the transport of dispersed actinides in the environment and may carry information on the composition of the detonated device. We determine the oxidation state of U and Fe (which is known to buffer the oxidation state of actinide elements and to affect the redox state of groundwater) in samples of melt glass collected from three U.S. nuclear weapons tests. For selected samples, we also determine the coordination geometry of U and Fe, and we report the oxidation state of Pu from one melt glass sample. We find significant variations among the melt glass samples and, in particular, find a clear deviation in one sample from the expected buffering effect of Fe(II)/Fe(III) on the oxidation state of uranium. In the first direct measurement of Pu oxidation state in a nuclear test melt glass, we obtain a result consistent with existing literature that proposes Pu is primarily present as Pu(IV) in post-detonation material. In addition, our measurements imply that highly mobile U(VI) may be produced in significant quantities when melt glass is quenched rapidly following a nuclear detonation, though these products may remain immobile in the vitrified matrices. The observed differences in chemical state among the three samples show that redox conditions can vary dramatically across different nuclear test conditions. The local soil composition, associated device materials, and the rate of quenching are all likely to affect the final redox state of the glass. The resulting variations in glass chemistry are significant for understanding and interpreting debris chemistry and the later environmental mobility of dispersed material.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2013

Impact of Fission Products Impurity on the Plutonium Content of Metal- and Oxide- Fuels in Sodium Cooled Fast Reactors

This short report presents the neutronic analysis to evaluate the impact of fission product impurity on the Pu content of Sodium-cooled Fast Reactor (SFR) metal- and oxide- fuel fabrication. The similar work has been previously done for PWR MOX fuel. The analysis will be performed based on the assumption that the separation of the fission products (FP) during the reprocessing of UOX spent nuclear fuel assemblies is not perfect and that, consequently, a certain amount of FP goes into the Pu stream used to fabricate SFR fuels. Only non-gaseous FPs have been considered (see the list of 176 isotopes considered in the calculations in Appendix 1 of Reference 1). Throughout of this report, we define the mixture of Pu and FPs as Pu FP . The main objective of this analysis is to quantify the increase of the Pu content of SFR fuels necessary to maintain the same average burnup at discharge independently of the amount of FP in the Pu stream, i.e. independently of the Pu FP composition. 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↗
DOE OSTI2016

Destructive examination of shipping package 9975-02101

Destructive and non-destructive examinations have been performed on the components of shipping package 9975-02101 as part of the comprehensive Model 9975 package surveillance program. This package is one of ten high-wattage packages that were selected for field surveillance in FY15, and was identified to contain several non-conforming conditions. Most of these conditions (mold, stains, drum corrosion, calculated fiberboard dimensions and fiberboard damage) relate to the accumulation of water in the outer and lower portions of the cane fiberboard assembly. In the short term, this causes local but reversible changes in the fiberboard properties. Long-term effects can include the permanent loss of fiberboard properties (thus far observed only in the bottom fiberboard layers) and reduced drum integrity due to corrosion. The observed conditions must be fully evaluated by KAC to ensure the safety function of the package is being maintained. Three of the other nine FY15 high-wattage packages examined in the K-Area Complex showed similar behavior. Corrosion of the overpack drum has been seen primarily in those packages with relatively severe fiberboard degradation. Visual examination of the drums in storage for external corrosion should be considered as a screening tool to identify additional packages with potential fiberboard degradation. Where overpack drum corrosion has been observed, it is typically heaviest adjacent to the stitch welds along the bottom edge. It is possible that changes to the stitch weld design would reduce the degree of corrosion in this area, but would not eliminate it. Several factors can contribute to the concentration of moisture in the fiberboard, including higher than average initial moisture content, higher internal temperature (due to internal heat load and placement with the array of packages), and the creation of additional moisture as the fiberboard begins to degrade.

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

Measuring Monodisperse Aerosol Transmission in the Los Alamos Respirable Release Fraction Measurement Chamber

The transmission ratio of monodisperse aerosol (between 1 and 20 μm aerodynamic equivalent diameter) was measured in a modified configuration of the Los Alamos RRFMC (Respirable Release Fraction Measurement Chamber). The RRFMC is a newly built, 20- foot tall, airtight chamber for the drop testing of nuclear material storage containers. The RRFMC was integrated into an existing HEPA filtered wind tunnel system, and is capable of direct measurement of Airborne Release Fraction (ARF) and Respirable Release Fraction (RRF) for plutonium and uranium under various accident conditions. Cerium oxide powder is used as a surrogate for plutonium oxide. Real-time aerosol measurements for aerodynamic diameters from 0.5 to 20 μm are performed with a TSI Inc (Shoreview, MN) Model 3321 APS (Aerodynamic Particle Sizer) spectrometer.

47 OTHER INSTRUMENTATION↗
DOE OSTI2022

Improbability of Nuclear Criticality in Compacted Criticality Control Overpacks after Room Closure by Salt Creep at Waste Isolation Pilot Plant

Based on the rationale presented, nuclear criticality is improbable after salt creep causes compaction of criticality control overpacks (CCOs) disposed at the Waste Isolation Pilot Plant, an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality potential has been updated here because of the introduction of CCOs, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCOs during two representative conditions: (1) large salt block fall, and (2) gradual salt compaction (without brine seepage and subsequent gas generation to permit maximum room closure). Geomechanical models of rock fall demonstrate three tiers of CCOs are not greatly disrupted. Geomechanical models of gradual room closure from salt creep predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Criticality models of spheres and cylinders of 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular spacing demonstrate that an array of CCOs is not critical when surrounded by salt and magnesium oxide, provided the amount of hydrogenous material shipped in the CCO (usually water and plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Understanding Biases in Sample Preparation Techniques for Coupled Scanning Electron Microscopy and MAMA PuO 2 Morphological Analysis

In this project, the scanning electron microscopy (SEM) sampling method used during the statistical design study (SDS) was investigated to determine if any sampling biases were present in the analyzed data. Using standard particle size distribution powders from the National Institute of Standards and Technology (NIST 1984 standard reference material) with the origin wet dispersion method, it was determined that a bias to smaller particles was present. This was supported by theoretical calculations using Stokes’ law to determine the settling rate of spherical particles of roughly the same size and mass as those found in the SDS. Based on the theoretical calculations, it was determined that the settling rate for each of the 76 powder sets in the SDS could be unique based on specific particle shape and mass distributions, making a universal correction factor/formula not applicable. Therefore, priority shifted to developing an improved wet dispersion method that significantly reduced the particle settling rate for all particle size and shapes. This was achieved by replacing the original solvent (isopropyl alcohol) with a heavy liquid (lithium heteropolytungstates), which dramatically slowed the settling rate and allowed for the capture of a suitable homogeneous aliquot. SEM imaging and Morphological Analysis for Material Attribution (MAMA) software analysis were conducted on the NIST standard, and the SEM/MAMA data were compared to data captured by a dynamic image analysis particle size analyzer. The resulting data confirmed that the new wet dispersion method does indeed deliver an improved representative aliquot to the SEM stub. For instance, in the NIST certificate, the average particle size is ~17.1 µm ± 2.2 µm with a normal distribution. The initial wet dispersion method resulted in a drastically reduced average particle size of 6.1 µm in addition to a non-representative heavy bi-modal distribution whereas the improved LST wet dispersion method resulting in an average particle size that was much closer to the NIST certificate (12.7 µm) with a similar normal distribution. Although the improved method was still short of the NIST certificate average, atomic force microscopy analysis determined that the resulting ~20-25% reduction in size was due to particles sinking into the carbon sticky tape used for SEM imaging. It is believed that that this bias can be calibrated in a much more predicable manner than the original settling rate bias. In addition, the matching normal distribution curves between the NIST certificate and the heavy liquid method indicate a much-improved representative aliquot has been sampled and imaged. A surrogate CeO 2 powder was used to reflect PuO 2 more accurately and to aid in implementing radiological controls and shielding. The resulting data sets from the SEM/MAMA method and the particle size analyzer give almost identical average particle sizes and particle distribution statistics. Future work will re-analyze several select runs from the SDS to determine if morphological signatures can be found with the improved sampling method.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

Improbability of Post-Closure Criticality in Compacted Criticality Control Overpacks after Room Closure at Waste Isolation Pilot Plant

As part of its periodic re-certification of the Waste Isolation Pilot Plant (WIPP), an operating repository in bedded salt for the disposal of transuranic (TRU) waste from atomic energy defense activities, the United States Environmental Protection Agency expects a re-evaluation of features, events, and processes, such as post-closure nuclear criticality. Although salt creep beneficially encapsulates the TRU waste in the closed WIPP repository, the spacing between an array of waste packages is disrupted as the salt creep closes disposal rooms and containers lose structural integrity. For most TRU waste, the possibility of post-closure criticality is exceedingly small either because the salt neutronically isolates TRU waste canisters or because closure of a disposal room from salt creep does not sufficiently compact the low mass of fissile material. The criticality evaluation was updated, however, because of the introduction of criticality control overpack (CCO) containers, which may dispose up to 380 fissile gram equivalent plutonium-239 in each container. The criticality potential is evaluated through high-fidelity geomechanical modeling of a disposal room filled with CCO containers during two representative conditions: (1) large salt block fall, and (2) gradual disposal room closure from salt creep. Geomechanical models of roof fall demonstrate three tiers of CCO containers are not greatly disrupted. Geomechanical models of gradual room closure from salt creep (without brine seepage and subsequent gas generation to permit maximum room closure) were used to predict irregular arrays of closely packed CCOs after 1000 years, when room closure has asymptotically approached maximum compaction. Models of spheres or cylinders with 380 fissile gram equivalent of plutonium (as oxide) at the predicted irregular compacted spacing demonstrate that an array of CCO containers is not critical when surrounded by salt and magnesium oxide, provided the mass of hydrogenous material shipped in CCO containers (usually plastics) is controlled or boron carbide (a neutron poison) is mixed with the fissile contents.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Estimates of Quantum Tunneling Effects for Hydrogen Diffusion in PuO 2

We detail the estimation of activation energies and quantum nuclear vibrational tunneling effects for hydrogen diffusion in PuO 2 based on Density Functional Theory calculations and a quantum double well approximation. We find that results are relatively insensitive to choice of exchange correlation functional. In addition, the representation of spin in the system and use of an extended Hubbard U correction has only a small effect on hydrogen point defect formation energies when the PuO 2 lattice is held fixed at the experimental density. We then compute approximate activation energies for transitions between hydrogen interstitial sites seeded by a semi-empirical quantum model and determine the quantum tunneling enhancement relative to classical kinetic rates. Our model indicates that diffusion rates in H/PuO 2 systems could be enhanced by more than one order of magnitude at ambient conditions and that these effects persist at high temperature. The method we propose here can be used as a fast screening tool for assessing possible quantum nuclear vibrational effects in any number of condensed phase materials and surfaces, where hydrogen hopping tends to follow well defined minimum energy pathways.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2011

Shipment of Non-traditional Contents in the 9977 Type B Package

The 9977 is a certified Type B Packaging authorized to ship uranium and plutonium in metal and oxide forms. These materials are typically confined within metallic containers designed for ease of handling and to prevent the spread of contamination. The Pacific Northwest National Laboratory (PNNL) uses Pu and U sources for the training of domestic and international customs agents in the identification and detection of radioactive materials (RAM). These materials are packed in polycarbonate containers which permit the trainees to view the RAM. The safety basis was made to authorize the use of these unusual containers. The inclusion of the PNNL Training Source Contents into the 9977 Packaging imposed unique conditions previously unanalyzed. The use of polycarbonate as a content container material, while different from any configuration previously considered, does not raise any safety issues with the package which continues to operate with a large safety margin for temperatures, pressures, containment, dose rates, and subcriticality.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Enhancing the Material Control & Accounting Measurement System at the State Scientific Center of the Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky

Nuclear material control and accounting (NMCA) system is improving under cooperation with USA national laboratories. Standard reference materials (RMs) and measurement techniques certified at IPPE level are required for: instrument calibration, verification measurements of parameters of items and materials, measurement error estimation, and quality control measurements. We present the main results for development of nuclear RMs for two uranium strata and the results for certification of three measurement techniques (MT) for U-235 mass fraction in uranium and U-235 mass in items. We present the results for developing measurement techniques for Pu-239 in PuO{sub 2}.

ACCOUNTING↗