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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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DOE OSTI2016

Corrosion Testing of Carbon Steel Exposed to Sludge Heel Chemical Cleaning Solutions - 16109

As efforts continue to treat and dispose of millions of gallons of legacy radioactive materials from the production of nuclear weapons, non-compliant waste storage tanks will gradually be emptied of the bulk waste volume leaving heel materials requiring removal prior to tank closure. The waste heel slurries are distributed on the floor of large, million gallon tanks, which frequently contain numerous obstructions that limit the effectiveness of mechanical removal methods. As a result, chemical cleaning methods are needed for the effective removal of the heels as well as chemical scales that are present on the tank walls and other interior surfaces. Two tank cleaning technologies have already been implemented at the Savannah River Site (SRS): Low Temperature Aluminum Dissolution (LTAD) and Bulk Oxalic Acid Cleaning (BOAC). Recent chemical cleaning efforts on SRS Tank 12 were very successful with regard to bulk sludge heel (especially for Al, Fe, and U phases) and beta/gamma radionuclide removal. Although chemical cleaning using these technologies has been shown to be effective, no disposition path has been identified for oxalate added during BOAC, and insoluble oxalate salts are accumulating within the SRS tank farm and waste processing facilities (evaporators, etc.). Extensive sludge washing is also required to remove moderately soluble sodium oxalate salts prior to sludge vitrification in the SRS Defense Waste Processing Facility (DWPF). As a result, oxalate additions to the tank farm need to be minimized by the use of supplementary acids to assist sludge removal in OA or the use of other cleaning reagents or processing strategies. Previous SRNL testing revealed the importance of pH control for BOAC and recommended the use of a supplementary acid (dilute HNO{sub 3}) with OA to minimize oxalate additions. Primary drivers in SRS Tank Closure Performance Assessments, which evaluate the fate and impact of tank sludge residuals on a geological timescale, are the removal of alpha emitting radionuclides present at low concentrations such as Pu, Am, and Np, which are not highly soluble in currently utilized chemical cleaning reagents. Scoping studies conducted at SRNL revealed promising methods to dissolve the actinides within the HLW tank heels. Oxidation of the actinides with permanganate in either strong caustic (10 M NaOH) or dilute acidic (0.2 M HNO{sub 3}) solutions was shown to result in dissolution of oxy/hydroxide phases of these metals in the absence of major sludge phases. The waste tanks at SRS and Hanford are constructed of carbon steel. Carbon steel corrodes rapidly in most acidic and/or oxidizing environments. Previous electrochemical corrosion testing indicated that the corrosion rates in dilute HNO{sub 3} with OA may result in manageable corrosion rates, over short exposure periods. Additional corrosion testing was performed to provide data for the nitric acid/OA blend as well as the acid/permanganate and caustic permanganate based chemical cleaning solutions. Corrosion rates from the passive coupon tests for the nitric acid/oxalic acid blend were significantly greater than those observed previously with the electrochemical testing. This result may be due to simultaneous electrochemical and chemical dissolution of the steel. The corrosion rate data from the coupon tests in the nitric acid/oxalic acid blend were evaluated to determine the degree of potential structural damage. The corrosion rates, although relatively high, would not be expected to cause damage that would reduce the capacity of the tank primary wall if the process is completed within a month. The corrosivity of sodium permanganate in four proposed cleaning solutions, 1 nitric acid solution and 3 sodium hydroxide solutions, was studied by electrochemical methods at room temperature. The corrosion rates were significantly less aggressive than in the nitric acid/oxalic acid blend corrosion tests. Passive coupon tests needs to be performed in these solutions to ensure that chemical dissolution of the metal is not occurring simultaneously. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Integrity Assessment of Hanford 242-A Evaporator Facility E-A-1 Re-boiler Vessel via Fluorescein Liquid Tracer Test Methods - 17357

The 242-A Evaporator Facility is critical to the United States Department of Energy's (DOE) Hanford Site clean-up mission due to limited space within the existing Double-Shell Tanks (DSTs) used to store nuclear and chemical waste until future processing. The 242-A Evaporator Facility is currently the sole waste volume reduction process used at Hanford, and the E-A-1 Re-boiler is essential to the Evaporator process. Due to the potential for cross-contamination between the Re-boiler tubes through which waste passes and the tube sheets of uncontaminated steam, an assessment of the Re-boiler vessel's integrity is required for continuing operations of the facility. In order to meet these requirements, an independent integrity assessment was designed to use fluorescein liquid tracer (LT) to verify the leak-tightness of the Re-boiler vessel in accordance with Washington River Protection Solutions, LLC (WRPS) tightness test standards, as well as, the ASME Boiler and Pressure Vessel (B and PV) Code, Section III, Division 1, NB-6300. LT test methodologies were evaluated and determined to be appropriate for a vessel with restricted access, such as the E-A-1 Re-boiler, as it provided the needed sensitivity, reliability, and cost; as well as, exceeded the leak-tightness specifications required by both WRPS and ASME B and PV Code standards. The integrity test requires water laced with fluorescein be circulated through the waste processing loop of the 242-A Evaporator and samples of steam condensate be collected and analyzed for traces of fluorescein to determine leak-tightness. Liquid condensate sampling was conducted prior to and during testing and the samples were analyzed for fluorescein tracer to determine the amount of fluorescein that flowed out of the tubes of the Re-boiler into the steam side of the vessel. Given the established criteria for 'leak-tight,' test results determined that the Re-boiler passed the integrity assessment. After further scrutiny of the 2014 test and data results, it was determined that although the Re-boiler was 'leak-tight,' as established by the test criteria, there was concern of future leakage. The DOE and the DOE Office of River Protection (ORP) requested that WRPS repeat the integrity assessment in two and a half years to determine if the potential leak was growing, which could threaten continuing operations of the 242-A Evaporator. The 2017 test has been redesigned using lessons learned from the 2014 test and was modified to account for ergonomic and safety considerations, ease of repeatability, and accuracy of results. The liquid tracer methodology and test plan developed for the Re-boiler will continue to be implemented in 5 year increments to monitor current and future integrity. The method may also prove useful for other aging components across the Hanford Site, where appropriate. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Integrated soft snake robot for gamma spectroscopy of radiologically contaminated ducting

Anticipated growth in the number and complexity of nuclear facilities entering the decommissioning phase of their lifespan requires innovative methods of mitigating radiological and safety hazards for workers. Advancements in robotics provide opportunities for remote characterization of radiologically contaminated ducts and pipes, reducing worker exposure. However, the nuclear decommissioning industry has shown a hesitancy in adopting new technology due to concerns regarding capital costs and a lack of demonstration data in representative environments. This work aims to leverage novel soft robotics to develop and demonstrate a sensor-integrated characterization tool for contaminated ducting systems. Two gamma spectrometers – one scintillation-based and one solid-state – were selected and integrated into a pneumatically-actuated snake-like soft robot using stretchable electronic cabling. The robot was then demonstrated at the Idaho National Laboratory’s Critical Infrastructure Test Range Complex. The robot maneuvered through a pseudo-constrained corridor containing several sealed sources and the integrated spectrometers collected spectra at various points along the path of travel. Both spectrometers successfully collected spectra with identifiable, isotope-specific peaks. To evaluate the effect of contamination buildup on the collected spectra, the robot with the integrated scintillation detector was tested using a liquid radioactive source. The resulting spectra showed little detection interference from the liquid contamination; however, a short-lived isotope was used which may not be representative of longer-lived contamination.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Verification of Monte Carlo transport codes against measured small angle p-, d-, and t-emission in carbon fragmentation at 600 MeV/nucleon

Momentum spectra of hydrogen isotopes have been measured at 3.5° from 12 C fragmentation on a Be target. Momentum spectra cover both the region of fragmentation maximum and the cumulative region. Differential cross sections span five orders of magnitude. The data are compared to predictions of four Monte Carlo codes: QMD, LAQGSM, BC, and INCL++. There are large differences between the data and predictions of some models in the high momentum region. The INCL++ code gives the best and almost perfect description of the data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Online Waste Library (OWL) and Waste Forms Characteristics Annual Report

This report represents completion of milestone deliverable M2SF-19SNO10309013 "Online Waste Library (OWL) and Waste Forms Characteristics Annual Report" that reports annual status on fiscal year (FY) 2019 activities for the work package SF-19SN01030901 and is due on August 2, 2019. The online waste library (OWL) has been designed to contain information regarding United States (U.S.) Department of Energy (DOE)-managed (as) high-level waste (DHLW), spent nuclear fuel (SNF), and other wastes that are likely candidates for deep geologic disposal, with links to the current supporting documents for the data (when possible; note that no classified or official-use-only (OUO) data are planned to be included in OWL). There may be up to several hundred different DOE-managed wastes that are likely to require deep geologic disposal. This annual report on FY2019 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the management processes for the OWL. Updates to the OWL include an updated user's guide, additions to the OWL database content for wastes and waste forms, results of the beta testing and changes implemented from it. Also added are descriptions of the management/control processes for the OWL development, version control, and archiving. These processes have been implemented as part of the full production release of OWL (i.e., OWL Version 1.0), which has been developed on, and will be hosted and managed on, Sandia National Laboratories (SNL) systems. The version control/update processes will be implemented for updates to the OWL in the future. Additionally, another process covering methods for interfacing with the DOE SNF Database (DOE 2007) at Idaho National Laboratory on the numerous entries for DOE-managed SNF (DSNF) has been pushed forward by defining data exchanges and is planned to be implemented sometime in FY2020. The INL database is also sometimes referred to as the Spent Fuel Database or the SFDB, which is the acronym that will be used in this report. Once fully implemented, this integration effort will serve as a template for interfacing with additional databases throughout the DOE complex.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

NSRD-06. Computational Capability to Substantiate DOE-HDBK-3010 Data

Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Hand book, DOE-HDBK-3010, Airborne Release Fractions/Rates and Resp irable Fractions for Nonreactor Nuclear Facilities , to determine source terms. In calcula ting source terms, analysts tend to use the DOE Handbook's bounding values on airbor ne release fractions (ARFs) and respirable fractions (RFs) for various cat egories of insults (representing potential accident release categories). This is typica lly due to both time constraints and the avoidance of regulatory critique. Unfort unately, these bounding ARFs/RFs represent extremely conservative values. Moreover, th ey were derived from very limited small- scale table-top and bench/labo ratory experiments and/or fr om engineered judgment. Thus the basis for the data may not be re presentative to the actual unique accident conditions and configura tions being evaluated. The goal of this res earch is to develop a more ac curate method to identify bounding values for the DOE Handbook using the st ate-of-art multi-physics-based high performance computer codes. This enable s us to better understand the fundamental physics and phenomena associated with the ty pes of accidents for the data described in it. This research has examined two of the DOE Handbook's liquid fire experiments to substantiate the airborne release frac tion data. We found th at additional physical phenomena (i.e., resuspension) need to be included to derive bounding values. For the specific cases of solid powder under pre ssurized condition and mechanical insult conditions the codes demonstrated that we can simulate the phenomena. This work thus provides a low-cost method to establis h physics-justified sa fety bounds by taking into account specific geometri es and conditions that may not have been previously measured and/or are too costly to do so.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

A proposed method for addressing large unphysical uncertainties in mubar

Mubar uncertainties in Section MF 34 MT 2 of ENDF/B-VIII.0 data are sometimes too large. Physical limitations of the bounds of mubar limit the maximum uncertainty to be < 1.0 and usually $\ll$ 1.0. Two ad hoc methods are proposed to artificially constrain the randomly sampled values of mubar to allowable values. The application of NJOY’s mubar covariance matrix to the “sandwich rule” is also discussed and a consistent way to apply sensitivity coefficients is described. Mubar uncertainties for the Jezebel k eff were found to be 156 pcm – a value comparable to other evaluated cross section uncertainties.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

RADAI: A Large-Scale Realistic Dataset for Radiation Detection Algorithm Development

Open, realistic datasets are essential for developing and benchmarking radiation detection algorithms, yet they remain scarce. The Radiological Anomaly Detection and Identification (RADAI) project was develop to create datasets that meet the training and testing needs for sophisticated radiation detection algorithms. The RADAI dataset is a large-scale synthetic resource that integrates high-fidelity Monte Carlo simulations with realistic urban scenarios to capture both background variability and source signatures. RADAI models construction-material NORM, people and vehicles, urban clutter, and dynamic environmental effects such as cosmic-ray and rain-induced transients, and they provide list-mode detector data with motion and response modeling suitable for algorithm training and evaluation. The RADAI project resulted in three publicly-released complementary datasets together with an online scoring portal for standardized performance assessment and an open software toolkit that supports data access, augmentation, model development, and evaluation. These resources enable reproducible comparisons across methods and promote rigorous studies at the scale required by contemporary machine learning. By grounding algorithm development in realistic, well-documented conditions, RADAI supports progress toward more robust detection, identification, and localization in complex urban environments.

Ghawaly, James M. [Division of Computer Science an↗
DOE OSTI2020

Update on the Benchmark Analysis of the Jupiter Experiments: Plutonium Moderated by Lead

The Jupiter experiment was a series of measurements in the summer of 2017 to analyze the effects of voids in lead moderated plutonium assemblies. This experiment was part of a larger campaign that took place at the National Criticality Experiments Research Center (NCERC) in collaboration with the Japan Atomic Energy Agency (JAEA) that also investigated assemblies with differing uranium enrichments. The purpose for this campaign was to provide data useful for a potential lead-bismuth cooled accelerator driven system. This data is needed due to the low number of benchmarks involving lead. Given the wide use of lead for shielding in addition to new reactor designs, there is a need for further benchmarking that this experiment could help fulfill. In addition, adding more benchmarks to the International Criticality Safety Benchmark Evaluation Project (ICSBEP) is helpful for further validation of simulation methods used for criticality safety. This summary is an update on the current progress of this analysis, and all results shown are preliminary.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Quantification of LEU Holdup using gamma ray imaging and inverse transport solver

Holdup is the residual amount of special nuclear material (SNM) remaining in a processing facility after the bulk materials have been cleaned out. In commercial uranium processing facilities, quantification of holdup is a major challenge because of the highly variable shapes and sizes of the deposits. Any method that attempts to generalize and calibrate deposit shapes in order to quantify holdup will be prone to high uncertainties. Uncertainties on the order of ±50% are typical in holdup results. In international safeguards applications, a ±50% uncertainty can result in a large amount of material unaccounted for (MUF) thereby increasing the difficulty of detecting material diversion and facility misuse. An imaging-based methodology has been developed with the objective of significantly reducing this uncertainty by using the true deposit shape, instead of relying on oversimplified geometric assumptions. The project is a collaboration between ORNL, Y-12, and the University of Tennessee, Knoxville, TN. Uranium sources of known masses were measured using the Germanium Gamma-ray Imager (GeGI), a high-resolution imaging spectrometer, creating a pixelated map for each spectral bin. Two different gamma imaging methods are employed in this work: coded aperture imaging and Compton imaging. A validated MonteCarlo model of the detector has been developed using the GEANT4 code for determining the intrinsic response of the detector, its enclosure, and the coded aperture mask. An inverse transport solver based on the Markov Chain Monte-Carlo approach known as Differential Evolution Adaptive Metropolis (DREAM) is employed to use the measurement data from the image pixels (coded aperture or Compton) to solve for the mass of 235 U in the deposit. A reliable method based on the DREAM solver has been developed to flag the infinite thickness condition of a uranium deposit. The project team is working towards improving the image reconstruction for Compton imaging so that a better localization of the source can be achieved. Besides treating the coded aperture and Compton imaging methods independently, the project is also evaluating a combined method that uses the Compton scatter data from a coded aperture measurement. GEANT4 simulations are being performed to evaluate the combined approach. The impact on the DREAM optimization as the source thickness progressively approaches infinite thickness is being evaluated. A number of uranium sources available at ORNL have been measured, and the DREAM results have been tested and validated for the coded aperture imaging. A similar effort will be carried out to validate the Compton based method once the development of algorithms for better localization are complete. The imaging based quantification is very amenable to unattended monitoring of holdup accumulation at key measurement points. A proof of concept measurement has been completed to demonstrate this capability The current work used the high energy resolution imager GeGI. However, the approach and methodologies are applicable to other imagers such as the cadmium zin telluride (CZT) based imager manufactured by H3D, Inc.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Water dynamics on ice and hydrate lattices studied by second-order central-line stimulated-echo oxygen-17 nuclear magnetic resonance

Oxygen-17 stimulated-echo spectroscopy is a novel nuclear magnetic resonance (NMR) technique that allows one to investigate the time scale and geometry of ultraslow molecular motions in materials containing oxygen. The method is based on detecting orientationally encoded frequency changes within oxygen’s central-transition NMR line that are caused by second-order quadrupolar interactions. In addition to the latter, the present theoretical analysis of various two-pulse echo and stimulated-echo pulse sequences takes also heteronuclear dipolar interactions into account. As an experimental example, the ultraslow water motion in polycrystals of tetrahydrofuran clathrate hydrate is studied via two-time oxygen-17 stimulated-echo correlation functions. The resulting correlation times and those of hexagonal ice are similar to those from previous deuteron NMR measurements. Calculations of the echo functions’ final-state correlations for various motional models are compared with the experimental data of the clathrate hydrate. It is found that a six-site model including the oxygen-proton dipolar interaction describes the present results.

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

Evaluation of Delayed Critical ORNL Unreflected HEU Metal Sphere (ORSphere)

In 1971 and 1972 experimenters at the Oak Ridge Critical Experiment Facility performed critical experiments using an unreflected metal sphere of highly enriched uranium (HEU). The sphere used for the criticality experiments, originally used for neutron leakage spectrum measurements by General Atomic Company, consisted of three main parts and were assembled with a vertical assembly machine. Two configurations were tested. The first was nearly spherical with a nominal radius of 3.467 inches and had a reactivity of 68.1 ± 2.0 cents. The sphere parts were then re-machined as a sphere with a nominal radius of 3.4425 inches. This assembly had a reactivity of -23 cents. The method, dimensions, and uncertainty of the critical experiment were extensively recorded and documented. The original purpose of the experiments was for comparison to GODIVA I experiments. The ORNL unreflected HEU Metal Spheres have been evaluated for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (scheduled for inclusion in the September 2013 edition).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Machine learning molecular dynamics simulations toward exploration of high-temperature properties of nuclear fuel materials: case study of thorium dioxide

Predicting materials properties of nuclear fuel compounds is a challenging task in materials science. Their thermodynamical behaviors around and above the operational temperature are essential for the design of nuclear reactors. However, they are not easy to measure, because the target temperature range is too high to perform various standard experiments safely and accurately. Moreover, theoretical methods such as first-principles calculations also suffer from the computational limitations in calculating thermodynamical properties due to their high calculation-costs and complicated electronic structures stemming from f-orbital occupations of valence electrons in actinide elements. Here, we demonstrate, for the first time, machine-learning molecular-dynamics to theoretically explore high-temperature thermodynamical properties of a nuclear fuel material, thorium dioxide. The target compound satisfies first-principles calculation accuracy because f-electron occupation coincidentally diminishes and the scheme meets sampling sufficiency because it works at the computational cost of classical molecular-dynamics levels. We prepare a set of training data using first-principles molecular dynamics with small number of atoms, which cannot directly evaluate thermodynamical properties but captures essential atomistic dynamics at the high temperature range. Then, we construct a machine-learning molecular-dynamics potential and carry out large-scale molecular-dynamics calculations. Consequently, we successfully access two kinds of thermodynamic phase transitions, namely the melting and the anomalous λ transition induced by large diffusions of oxygen atoms. Furthermore, we quantitatively reproduce various experimental data in the best agreement manner by selecting a density functional scheme known as SCAN. Our results suggest that the present scale-up simulation-scheme using machine-learning techniques opens up a new pathway on theoretical studies of not only nuclear fuel compounds, but also a variety of similar materials that contain both heavy and light elements, like thorium dioxide.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2015

Regulatory perspective of boiling water reactor peak reactivity credit in spent fuel storage and transportation - 14592

The U.S. Nuclear Regulatory Commission (NRC) performs reviews of applications for spent nuclear fuel transportation packages and storage casks to ensure pertinent safety regulations are met. Among other acceptance criteria, these regulations require that the package be subcritical under a variety of normal, off-normal and accident conditions. Currently, all spent nuclear fuel transportation packages and storage casks assume fresh fuel (i.e., the fuel is unirradiated) in their criticality safety analyses for boiling water reactor (BWR) fuels. Reviews using this conservative assumption are less complex. The fresh fuel assumption for BWR fuel does not allow the presence of gadolinium (often called the 'fresh fuel no-Gad' assumption), an integral burnable absorber present in nearly all BWR fuel assemblies. More recently, applicants began requesting credit for the reactivity reduction due to depletion. This credit is commonly referred to as 'burnup credit' (BUC). As of today, NRC recommends that only analyses for Pressurized Water Reactor (PWR) spent fuel implement BUC. This is due to the fact there is limited directly applicable data available to benchmark codes for depletion and reactivity calculations for BWR BUC analyses, and determining the most reactive irradiation conditions is much less straightforward than for PWR BUC analyses. Similar to what was done for PWR BUC, the NRC staff and its contractors have begun identifying and prioritizing significant technical issues so that a technical basis for the allowance of BWR BUC can be developed. BWR fuel exhibits a peak in reactivity where initial gadolinium is depleted. Crediting burnup at this point is generally referred to as using a 'peak reactivity method'. This is a conservative approach and a method routinely used in BWR spent fuel pool criticality analyses. NRC is implementing a two-phased approach in investigating BWR BUC. Phase 1 investigates peak reactivity credit and Phase 2 evaluates BUC at a typical discharge exposure. This paper discusses peak reactivity credit and its history in regulatory analyses, challenges the NRC expects in reviewing an application requesting peak BWR reactivity credit for a spent fuel transportation package or storage cask, the work that is being performed by NRC contractors to address these challenges, and future work on peak reactivity credit and full BWR BUC. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Comparison of Burnup Credit Uncertainty Quantification Methods

The recent revision of Interim Staff Guidance (ISG-8 revision 3), issued by the US Nuclear Regulatory Commission, on burnup credit for criticality safety analysis of pressurized water reactor spent fuel in transportation and storage casks expands the nuclides that can be credited in the fuel composition to include minor actinides and fission products. The technical basis for development of the uncertainties associated with calculated nuclide concentrations in spent nuclear fuel that are recommended in ISG-8 was established through the use of measured isotopic data acquired from destructive radiochemical assay data for spent fuel samples. These data provide a direct experimental basis for determining uncertainties in the calculated nuclide concentrations. To provide additional insight into code uncertainties resulting from nuclear data and also the level of conservatism associated with using experimental assay benchmark data for uncertainty evaluation, a new uncertainty analysis tool in the SCALE code system that performs stochastic sampling of nuclear data has been applied to representative burnup credit calculations for spent fuel pool and dry cask storage applications methodology. The nuclear data uncertainty analysis capabilities in Sampler can account for uncertainty contributions from cross sections, fission yields, and decay data by using correlated sampling of covariance files developed for all nuclear data used by SCALE. These uncertainties associated with nuclear data evaluations are compared to corresponding values developed in ISG-8 that are based on destructive assay data, as well as to values obtained using an alternate methodology that was developed by the Electric Power Research Institute based on reactor core simulation data.

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

Assessment of approximate computational methods for conical intersections and branching plane vectors in organic molecules

Quantum-chemical computational methods are benchmarked for their ability to describe conical intersections in a series of organic molecules and models of biological chromophores. Reference results for the geometries, relative energies, and branching planes of conical intersections are obtained using ab initio multireference configuration interaction with single and double excitations (MRCISD). They are compared with the results from more approximate methods, namely, the state-interaction state-averaged restricted ensemble-referenced Kohn-Sham method, spin-flip time-dependent density functional theory, and a semiempirical MRCISD approach using an orthogonalization-corrected model. It is demonstrated that these approximate methods reproduce the ab initio reference data very well, with root-mean-square deviations in the optimized geometries of the order of 0.1 Å or less and with reasonable agreement in the computed relative energies. A detailed analysis of the branching plane vectors shows that all currently applied methods yield similar nuclear displacements for escaping the strong non-adiabatic coupling region near the conical intersections. Our comparisons support the use of the tested quantum-chemical methods for modeling the photochemistry of large organic and biological systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor

This work studies the hydrodynamics of the seven pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. Here, the scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number Re. A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of Re values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors f Exp agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within ±7% but are significantly underpredicted by Rehme's correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall y + values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method (y + ~ 1) with the Shear Stress Transport k - ω turbulence model and the scalable wall functions (y + > 12) with the realizable k - ε turbulence model can predict f Exp within ±5% The standard wall functions (y + > 30) and nonequilibrium wall functions (y + > 30) with the realizable k - ε model can predict f Exp within ±10%.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Use of Percutaneous Aspiration Thrombectomy vs. Anticoagulation Therapy to Treat Acute Iliofemoral Venous Thrombosis: 1-year Follow-up Results of a Randomised, Clinical Trial

PurposeThe purpose of this study was to compare the efficacy of percutaneous aspiration thrombectomy (PAT) followed by standard anticoagulant therapy, with anticoagulation therapy alone, for the treatment of acute proximal lower extremity deep vein thrombosis.MethodsIn this randomised, prospective study, 42 patients with acute proximal iliofemoral deep vein thrombosis documented via Doppler ultrasound examination, were separated into an interventional treatment group (16 males, 5 females, average age 51 years) and a medical treatment group (13 males, 8 females, average age 59 years). In the interventional group, PAT with large-lumen 9-F diameter catheterisation was applied, after initiation of standard anticoagulant therapy. Balloon angioplasty (n 19) and stent implementation (n: 14) were used to treat patients with residual stenosis (>50 %) after PAT. Prophylactic IVC filters were placed in two patients. The thrombus clearance status of the venous system was evaluated by venography. In both the medical and interventional groups, venous patency rates and clinical symptom scores were evaluated at months 1, 3, and 12 after treatment.ResultsDeep venous systems became totally cleared of thrombi in 12 patients treated with PAT. The venous patency rates in month 12 were 57.1 and 4.76 % in the interventional and medical treatment groups, respectively. A statistically significant improvement was observed in clinical symptom scores of the interventional group (PAT) with or without stenting (4.23 ± 0.51 before treatment; 0.81 ± 0.92 at month 12) compared with the medical treatment group (4.00 ± 0.63 before treatment; 2.43 ± 0.67 at month 12). During follow-up, four patients in the medical treatment and one in the interventional group developed pulmonary embolisms.ConclusionsFor treatment of acute deep vein thrombosis, PAT with or without stenting is superior to anticoagulant therapy alone in terms of both ensuring venous patency and improving clinical symptoms. PAT is a safe, inexpensive, and easily performed method of endovascular treatment with a low rate of major complications. Our present findings and literature data suggest that PAT can be used as first-line treatment in proximal deep vein thrombosis patients, especially when thrombolytic treatment is contraindicated.

62 RADIOLOGY AND NUCLEAR MEDICINE↗