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At least 775 records · Page 43

Heterogeneous reactions of chlorine nitrate and hydrogen chloride on type I polar stratospheric clouds

A fast-flow reactor coupled with a quadrupole mass spectrometer was used to study the heterogeneous reactions ClONO2 + HCl yields Cl2 + HNO3 (1) and ClONO2 + H2O yields HOCl + HNO3 (2) on vapor-deposited HNO3-H2O ice substrates. It was found that the sticking coefficient of HCl on these substrates was a strong function of the substrate composition, ranging from about 2 x 10 exp -5 at nitric acid trihydrate composition to 6 x 10 exp -3 at 45 wt pct HNO3. The HNO3-H2O ice substrates were found to have large internal surface areas, and corrections for gas-phase diffusion within the porous ices were applied to observed loss rates.

Leu, Ming-Taun↗

Heterogeneous Reactions of ClONO2, HCl, and HOCl on Liquid Sulfuric Acid Surfaces

The heterogeneous reactions of ClONO2 + H2O yields HNO3 + HOCl (1), ClONO2 + HCl yields C12 + HNO3 (2), and HOCl + HCl yields Cl2 + H2O (3) on liquid sulfuric acid surfaces have been studied using a fast flow reactor coupled to a quadrupole mass spectrometer. The main objectives of the study are to investigate: (a) the temperature dependence of these reactions at a fixed H2O partial pressure typical of the lower stratosphere (that is, by changing temperature at a constant water partial pressure, the H2SO4 content of the surfaces is also changed), (b) the relative importance or competition between reactions 1 and 2, and (c) the effect of HNO3 on the reaction probabilities due to the formation of a H2SO4/HNO3/H2O ternary system. The measurements show that all the reactions depend markedly on temperature at a fixed H2O partial pressure: they proceed efficiently at temperatures near 200 K and much slower at temperatures near 220 K. The reaction probability (gamma(sub 1)) for ClONO2 hydrolysis approaches 0.01 at temperatures below 200 K, whereas the values for gamma(sub 2) and gamma(sub 3) are on the order of a few tenths at 200 K. Although detailed mechanisms for these reactions are still unknown, the present data indicate that the competition between ClONO2 hydrolysis and ClONO2 reaction with HCl may depend on temperature (or H2SO4 Wt %): in the presence of gaseous HCl at stratospheric concentrations, reaction 2 is dominant at lower temperatures (less than 200 K), but reaction 1 becomes important at temperatures above 210 K. Furthermore, reaction probability measurements performed on the H2SO4/HNO3/ H2O ternary solutions do not exhibit noticeable deviation from those performed on the H2SO4/H2O binary system, suggesting little effect of HNO3 in sulfate aerosols on the ClONO2 and HOCl reactions with HCl. The results reveal that significant reductions in the chlorine-containing reservoir species (such as ClONO2 and HCl) can take place on stratospheric sulfate aerosols at high latitudes in winter and early spring, even at temperatures too warm for Polar Stratospheric Clouds (PSCs) to form or in regions where nucleation of PSCs is sparse. This is particularly true under elevated sulfuric acid loading, such as that after the eruption of Mt. Pinatubo. Comparisons between our results and those presently available have also been made.

Zhang, Renyi↗

Heterogeneous Interactions of ClONO2 and HCl with Sulfuric Acid Tetrahydrate: Implications for the Stratosphere

The reaction probabilities for ClONO2+H2O- HOCl + HNO3 and ClONO2+ HCl Cl2 +HNO3 have been investigated on sulfuric acid tetrahydrate (SAT, H2SO4-4H2O)surfaces at temperatures between 190 and 230 K and at reactant concentrations that are typical in the lower stratosphere, using a fast-flow reactor coupled to a quadrupole mass spectrometer. The results indicate that the reaction probabilities as well as HCl uptake depend strongly on the thermodynamic state of SAT surface: they decrease significantly with decreasing H2O partial pressure at a given temperature, and decrease with increasing temperature at a given H2O partial pressure, as the SAT changes from the H2O-rich form to the H2SO4-rich form. For H2O-rich SAT at 195 K gamma(sub 1) approx. = -0.01 and gamma(sub 2) greater or equal to 0.1, whereas the values for H2SO4-rich SAT decrease by more than 2 orders of magnitude. At low concentrations of HCl, close to those found in the stratosphere, the amount of HCl taken up by H2O-rich SAT films corresponds to a coverage of the order of a tenth of a monolayer (approx. = 10(exp 14) molecules/sq cm); H2SO4-rich SAT films take up 2 orders of magnitude less HCl (les than 10(exp 12) molecules/sq cm). Substantial HCl uptake at high HCl concentrations is also observed, as a result of surface melting. The data reveal that frozen stratospheric sulfate aerosols may play an important role in chlorine activation in the winter polar stratosphere via processes similar to those occurring on the surfaces of polar stratospheric cloud particles.

Zhang, Renyi↗

Sulfuric Acid Monohydrate: Formation and Heterogeneous Chemistry in the Stratosphere

We have investigated some thermodynamic properties (i.e., freezing/melting points) and heterogeneous chemistry of sulfuric acid monohydrate (SAM, H2SO4.H2O), using a fast flow reactor coupled to a quadrupole mass spectrometer. The freezing point observations of thin liquid sulfuric acid films show that for acid contents between 75 and 85 wt % the monohydrate crystallizes readily at temperatures between 220 and 240 K on a glass substrate. Once formed, SAM can be thermodynamically stable in the H2O partial pressure range of (1-4) x 10(exp -4) torr and in the temperature range of 220-240 K. For a constant H2O partial pressure, lowering the temperature causes SAM to melt when the temperature and water partial pressure conditions are out of its stability regime. The reaction probability measurements indicate that the hydrolysis of N2O5 is significantly suppressed owing to the formation of crystalline SAM: The reaction probability on water-rich SAM (with higher relative humidity, or RH) is of the order of 10(exp -3) at 210 K and decreases by more than an order of magnitude for the acid-rich form (with lower RH). The hydrolysis rate of ClONO2 on water-rich SAM is even smaller, of the order of 10(exp -4) at 195 K. These reported values on crystalline SAM are much smaller than those on liquid solutions. No enhancement of these reactions is observed in the presence of HCl vapor at the stratospheric concentrations. In addition, Brunauer, Emmett, and Teller analysis of gas adsorption isotherms and photomicrography have been performed to characterize the surface roughness and porosities of the SAM substrate. The results suggest the possible formation of SAM in some regions of the middle- or low-latitude stratosphere and, consequently, much slower heterogeneous reactions on the frozen aerosols.

Zhang, Renyi↗

Deflection Measurements of a Thermally Simulated Nuclear Core Using a High-Resolution CCD-Camera

Space fission systems under consideration for near-term missions all use compact. fast-spectrum reactor cores. Reactor dimensional change with increasing temperature, which affects neutron leakage. is the dominant source of reactivity feedback in these systems. Accurately measuring core dimensional changes during realistic non-nuclear testing is therefore necessary in predicting the system nuclear equivalent behavior. This paper discusses one key technique being evaluated for measuring such changes. The proposed technique is to use a Charged Couple Device (CCD) sensor to obtain deformation readings of electrically heated prototypic reactor core geometry. This paper introduces a technique by which a single high spatial resolution CCD camera is used to measure core deformation in Real-Time (RT). Initial system checkout results are presented along with a discussion on how additional cameras could be used to achieve a three- dimensional deformation profile of the core during test.

Stanojev, B. J.↗

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7 percent resulting in a net system power of 8.1 kW and a system level efficiency of 17.2 percent. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to the NASA Glenn Research Center (GRC). The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3 percent. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 percent.

Facility Cooling System↗

Fission Surface Power Technology Demonstration Unit Test Results

The Fission Surface Power (FSP) Technology Demonstration Unit (TDU) is a system-level demonstration of fission power technology intended for use on manned missions to Mars. The Baseline FSP systems consists of a 190 kWt UO2 fast-spectrum reactor cooled by a primary pumped liquid metal loop. This liquid metal loop transfers heat to two intermediate liquid metal loops designed to isolate fission products in the primary loop from the balance of plant. The intermediate liquid metal loops transfer heat to four Stirling Power Conversion Units (PCU), each of which produce 12 kWe (48 kW total) and reject waste heat to two pumped water loops, which transfer the waste heat to titanium-water heat pipe radiators. The FSP TDU simulates a single leg of the baseline FSP system using an electrically heater core simulator, a single liquid metal loop, a single PCU, and a pumped water loop which rejects the waste heat to a Facility Cooling System (FCS). When operated at the nominal operating conditions (modified for low liquid metal flow) during TDU testing the PCU produced 8.9 kW of power at an efficiency of 21.7% resulting in a net system power of 8.1 kW and a system level efficiency of 17.2%. The reduction in PCU power from levels seen during electrically heated testing is the result of insufficient heat transfer from the NaK heater head to the Stirling acceptor, which could not be tested at Sunpower prior to delivery to GRC. The maximum PCU power of 10.4 kW was achieved at the maximum liquid metal temperature of 875 K, minimum water temperature of 350 K, 1.1 kg/s liquid metal flow, 0.39 kg/s water flow, and 15.0 mm amplitude at an efficiency of 23.3%. This resulted in a system net power of 9.7 kW and a system efficiency of 18.7 %.

fission power technology↗

Sodium Extraction from Full-Length Fermi-1 Blanket Assembly

Enrico Fermi Nuclear Generating Station is a nuclear power plant on the shore of Lake Erie. On October 5, 1966, Fermi 1, a prototype fast breeder reactor, suffered a partial fuel meltdown (no radioactive material was released). As a result, 34 metric tons of sodium-bonded blanket material from the decommissioned Fermi-1 reactor were shipped to and are currently stored at the Idaho Nation Lab (INL). This material is currently awaiting removal from the state of Idaho by 2035. In its current state, the Fermi-1 material is not suitable for disposal due to the reactive characteristic of its bond sodium. This work demonstrates removal of bond sodium from an entire full-length unirradiated Fermi-1 outer radial blanket assembly using a gravity assisted melt-drain-evaporate (MEDE) treatment process. Quantitative and qualitative analyses were performed to characterize the extent of sodium removal from the blanket material following MEDE treatment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Godiva IV Central Cavity Neutron Environment Characterization with Threshold Neutron Detectors

Godiva IV is a cylindrical fast burst reactor comprised of approximately 65 kg of highly enriched uranium that is operated by Los Alamos National Laboratory and sited at the National Criticality Experiments Research Center at the Nevada National Security Site in Nevada in the United States. Godiva IV is typically operated at delayed critical and in the regime spanning from sub-prompt to super-prompt bursts. Godiva IV is used for sample irradiations, criticality safety demonstrations, dosimetry studies, and for studying super-prompt behavior. In preparation for both an upcoming experiment to reduce uncertainties in the prompt fission spectrum for 235 U using threshold neutron detectors, and for future research using Godiva IV, it was desired to exercise the process of the selection of threshold neutron detectors/activation foils, radiation metrology, and the subsequent adjustment of the neutron spectrum. For this exercise, nine high purity threshold neutron detectors/activation foils were irradiated in a Godiva IV burst. The foils were then analyzed using a high purity germanium detector in the NCERC counting laboratory to determine end of irradiation specific activities for available IRDFF-II reactions. This work summarizes the Godiva IV foil irradiation, radiation metrology results, and adjusted neutron spectrum. The results of this exercise ultimately characterized the neutron environment inside the sample irradiation cavity inside Godiva IV to a higher degree than previously performed, informed decisions for the upcoming larger scale experiment, and will inform future neutron spectrum characterizations at NCERC.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Status Report on Fast Flux Test Facility Mechanistic Fuel Failure Experiment Analysis with BISON for Post Irradiation Examination Support

The renewed interest in metallic U-Zr nuclear fuel alloy has led to a drive for deeper understanding of the mechanisms driving the phenomena observed under irradiation conditions. The Department of Energy Advanced Fuel Campaign has developed infrastructure to support metallic fuel development, including Post Irradiation Examination (PIE) of legacy Fast Flux Test Facility (FFTF) Mechanistic Fuel Failure (MFF) experiments. The PIE performed on legacy FFTF MFF experiments gives insight on metallic fuel performance and can address the lack of knowledge and scarcity of reliable data identified in several studies over recent years. Unfortunately, PIE efforts can cost significant time and resources which can impede the progress of metallic U-Zr fuel development. Metallic U-Zr fuel performance modeling can be used to inform PIE efforts on regions of interest for relevant investigations and can help understand phenomena observed in PIE. This report demonstrates the current progress of FFTF MFF fuel performance simulations using the BISON fuel performance code and discusses the support provided by simulation to PIE efforts. Progress in temperature, profilometry, fission gas release, plenum pressure, and zirconium redistribution simulation results have been demonstrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Quantitative Insight to Fission Gas Bubble Distribution in Irradiated Annular U-10Zr Metallic Fuel Using Deep Learning

U-10Zr Metal fuel is a promising nuclear fuel candidate for next-generation sodium-cooled fast spectrum reactors. Porosity is one of the key facts to impact the performance of metallic fuel. Additionally, a mechanical understanding of fission gas bubbles evolution behavior is a prerequisite for fuel development and qualification. Previous study of fission gas bubbles relied on a simple threshold method working on low resolution optical microscopy images, which has challenges in recognizing bubble boundaries, and caused inaccurate statistics of bubble properties. In this paper, a pre-trained deep learning model on Scanning Electron Microscopy (SEM) images from an annular U-10Zr fuel (AF1), was applied to another U-10Zr annular fuel (AF2). More accurate fission gas bubble segmentation results were generated, which leads to more precise qualitative analysis on the morphology, size, density, and orientation of bubbles. Furthermore, we investigated the lanthanide movement along the radial temperature gradient and obtained conclusive findings.

36 MATERIALS SCIENCE↗

Insight into the Radiolytic Degradation Mechanism of TODGA

Partitioning and transmutation schemes, where americium is separated from other components of used nuclear fuel and burned in a fast neutron reactor, offer a path to dramatically decrease the space requirements for storage of nuclear fuel cycle wastes, allowing more efficient usage of potentially scarce storage resources.[1] Improvements in the efficiency of processes for separating americium from used nuclear fuels can thus have a significant impact on the costs of future nuclear fuel cycles. Understanding separation ligand radiation chemistry is important for development of new ligands and processes, as the ligand degradation products can have a deleterious effect on a separation process. Tetraoctyl diglycolamide (TODGA) has been studied as a ligand for lanthanide/minor actinide extraction for partitioning and transmutation schemes. However, the mechanism of initial energy transfer from products of solvent radiolysis to TODGA is still under debate. One proposed mechanism is attack by a n-dodecane radical cation resulting in electron abstraction and formation of a TODGA radical cation,[2] while later work has proposed hydrogen abstraction from sites adjacent to the central ether oxygen by other radical species.[3] The latter mechanism is proposed to result in degradation products from rupture of the ether bond. However, radiolytic degradation of TODGA in n-dodecane shows products that would result from rupture of all the backbone bonds. There is little information on what degradation products would be expected to form from the TODGA radical cation. In this work, we have begun to investigate this by examining the electron impact (EI) ionization mass spectrum of TODGA, which initially produces a TODGA radical cation in the gas phase. The EI spectrum shows fragments that would result from most backbone bond cleavages, similar to what is observed in the radiolysis of TODGA in n-dodecane. As the lifetime of the solvent radical cation decreases from n-dodecane to n-hexane,[4] irradiation of TODGA in n-hexane should have less solvent radical cations available for reaction with TODGA, and instead should favor reactions with longer-lived radicals. In contrast to n-dodecane, radiolytic degradation of TODGA in n-hexane produces only N,N-dioctylacetamide, which would result from rupture of the ether bond. The n-dodecane radical cation lifetime is long enough to produce TODGA radical cations but the n-hexane radical cation is mostly consumed before reaction with TODGA, so other radical processes that abstract hydrogen from TODGA dominate. This suggests that electron abstraction by the n-dodecane radical cation is an important component of TODGA radiolysis, although it does not rule out the presence of mechanisms involving hydrogen abstraction adjacent to the ether oxygen. References [1] J. Serp et al., Energies 2017, 10 (9), DOI: 10.3390/en10091445. [2] C. Zarzana et al., Solvent Extr. Ion Exch. 2015, 33 (5), 431–447, DOI: 10.1080/07366299.2015.1012885. [3] T. Koubský et al., Prog. Nucl. Energy 2017, 94, 208–215, DOI: 10.1016/j.pnucene.2016.07.010. [4] F. Sviridenko et al., Chem. Phys. Lett. 1998, 297 (3), 343–349, DOI: 10.1016/S0009-2614(98)01099-9.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗

Perspectives on Tailoring Neutron Energy Spectra in Material Test Reactors

Material test reactors (MTRs) are used to irradiate nuclear fuels and materials to develop data for how they endure neutron bombardment in or near reactor cores. Most historic MTRs, and all that remain operational in the US today, are water-cooled types and produce a thermalized neutron flux. New irradiation facilities are needed which can produce neutron energy spectra relevant to fast and fusion reactor environments. Construction of these facilities will take several years of steadfast funding to complete, which poses a formidable schedule challenge for current fast and fusion reactor developers. Irradiation designs which modify the neutron energy spectra delivered to test specimens in thermal spectrum MTRs, an approach referred to as “spectral tailoring”, can be used to approximate several relevant phenomena in the materials needed to enable fast and fusion reactor technologies. This approach is imperfect, but still valuable in the present situation. Here, the two highest flux MTRs operational in the United States, the Advanced Test Reactor (ATR) and High Flux Isotope Reactor (HFIR), have rich histories, ongoing developments, and new potentials for spectral tailoring that will be reviewed and discussed in this paper.

36 MATERIALS SCIENCE↗

SAM Theory Manual

The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR reactor safety analysis. It aims to provide fast-running, modest-fidelity, whole-plant transient analyses capabilities, which are essential for fast turnaround design scoping and engineering analyses of advanced reactor concepts. While SAM is being developed as a system-level modeling and simulation tool, advanced modeling techniques being implemented include a reduced-order three-dimensional module, pseudo 3-D conjugate heat transfer modeling in reactor core, flexible and multi-scale modeling of heat transfer between fluid and structures, in addition to the advances in software environments and design, and numerical methods. SAM aims to be a generic system-level safety analysis tool for advanced non-LWRs, including Liquid-Metal-cooled fast Reactors (LMR), Molten Salt Reactors (MSR), Fluoride-salt-cooled High- temperature Reactors (FHR), and High-Temperature Gas-cooled Reactors (HTGR). SAM takes ad- vantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented computational framework (MOOSE), and its underlying meshing and finite-element library and linear and non-linear solvers, to leverage the modern advanced software environments and numerical methods. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As new code capabilities and features are added, the SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SAM Theory Manual

The System Analysis Module (SAM) is an advanced and modern system analysis tool under development at Argonne National Laboratory for advanced non-LWR reactor safety analysis. It aims to provide fast-running, modest-fidelity, whole-plant transient analyses capabilities, which are essential for fast turnaround design scoping and engineering analyses of advanced reactor concepts. While SAM is being developed as a system-level modeling and simulation tool, advanced modeling techniques being implemented include a reduced-order three-dimensional module, pseudo 3-D conjugate heat transfer modeling in reactor core, flexible and multi-scale modeling of heat transfer between fluid and structures, in addition to the advances in software environments and design, and numerical methods. SAM aims to be a generic system-level safety analysis tool for advanced non-LWRs, including Liquid-Metal-cooled fast Reactors (LMR), Molten Salt Reactors (MSR), Fluoride-salt-cooled High-temperature Reactors (FHR), and High-Temperature Gas-cooled Reactors (HTGR). SAM takes advantage of advances in physical modeling, numerical methods, and software engineering to enhance its user experience and usability. It utilizes an object-oriented computational framework (MOOSE), and its underlying meshing and finite-element library and linear and non-linear solvers, to leverage the modern advanced software environments and numerical methods. This document provides the theoretical and technical basis of the code to help users understand the underlying physical models (such as governing equations, closure models, and component models), system modeling approaches, numerical discretization and solution methods, and the overall capabilities in SAM. As new code capabilities and features are added, the SAM Theory Manual will be updated periodically to keep it consistent with the state of the development.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗