Exact analytical solutions are presented for the evolution of the aerosol particle mass density function in a control volume for particle deposition due to gravitational settling, thermophoresis, and diffusion. The solutions are for arbitrary initial mass density functions and are applied for an initial lognormal density function. Integration of these solutions provides the suspended mass in the control volume as a function of time. These solutions serve as an exact benchmark to assess the accuracy of numerical methods. For the numerical algorithm used in MELCOR, excellent agreement is obtained for gravitational settling, diffusive deposition, and thermophoretic deposition for the suspended aerosol mass. In all cases, the default number of discrete particle size bins of 10 is shown to converge, with hardly any advantage to using 20 size bins.
CFD (Computational Fluid Dynamic) simulation of aerosol-laden natural convective flow and particle deposition in a spent fuel storage canister with 37 assemblies is currently computationally prohibitive. PWR (Pressurized Water Reactor) assemblies have up to 289 pins or tubes with several spacer grids to align the pins. Spacer grids with mixing vanes induce swirling during operation to increase heat transfer. Each spacer grid contains hundreds of small structures such as retaining clips, channel walls, and openings. The largest canisters store 37 PWR assemblies thus, there are numerous pins, tubes, and spacer grids for which the flow region between and around these structures needs to be determined along with the movement and deposition of aerosol particles. Because of the complicated geometry, modeling the intricate flow even for just one assembly is currently impractical. Nonetheless, we are developing techniques for a practical model to assess the natural aerosol particle deposition process in a canister in the event that a release occurs from one or more fuel pins. In the previous work it was demonstrated that CFD can model the flow through a PWR spacer grid with mixing vanes, including particle deposition, in a reasonable amount of time on a personal computer. In this work, the analysis is extended to include the bypass region between an assembly and the canister basket walls. It is shown that the flow velocity in the bypass region is about three times that of the interstitial region between the pins. The lengths before and after the spacer grid are also extended to determine when the flow becomes fully developed. In addition, the approach of computationally “stitching together” segments of an assembly is demonstrated with the plan to ultimately model a full assembly. The fraction of particles that are deposited in a segment with a spacer grid is determined as a function of particle size and flow velocity.
The flow and particle deposition patterns on surfaces in an idealized spacer grid for a 17x17 pressurized water reactor (PWR) assembly in a spent fuel canister are modeled using computational fluid dynamics (CFD) with laminar flow. The effects of gravitational settling, non-Stokesian flow, and particle slip are first rigorously analyzed. From the analysis, non-Stokesian effects and slip may be neglected for the particle sizes and conditions expected in a canister. For particles that do not settle out, a swirling flow pattern at the corners of a spacer grid channel directs particles to the leeward side of the flow vanes where much of the deposition occurs. Particle deposition increases with increasing particle diameter. Deposition also increases with decreasing flow velocity as this provides more time for particles to settle and deposit on the leeward side of the flow vanes. The fraction of particles that are transmitted through a spacer grid is determined as a function of inlet gas velocity and particle diameter by running the CFD calculation for each set of conditions and for each particle diameter. Curve fits of the transmission curve as a function of particle diameter for a specified spacer grid and flow velocity are applied to a lognormal particle mass density function for the inlet particles. The resulting mass density function and aerosol mass fraction that passes through the spacer grid can be determined analytically without resorting to numerical iteration. A sample calculation of the analytical solution is demonstrated for a lognormal particle mass density function.
Molten Salt Reactor (MSR) systems can be divided into two basic categories: liquid-fueled MSRs in which the fuel is dissolved in the salt, and solid-fueled systems such as the Fluoride-salt-cooled High-temperature Reactor (FHR). The molten salt provides an impediment to fission product release as actinides and many fission products are soluble in molten salt. Nonetheless, under accident conditions, some radionuclides may escape the salt by vaporization and aerosol formation, which may lead to release into the environment. We present recent enhancements to MELCOR to represent the transport of radionuclides in the salt and releases from the salt. Some soluble but volatile radionuclides may vaporize and subsequently condense to aerosol. Insoluble fission products can deposit on structures. Thermochimica, an open-source Gibbs Energy Minimization (GEM) code, has been integrated into MELCOR. With the appropriate thermochemical database, Thermochimica provides the solubility and vapor pressure of species as a function of temperature, pressure, and composition, which are needed to characterize the vaporization rate and the state of the salt with fission products. Since thermochemical databases are still under active development for molten salt systems, thermodynamic data for fission product solubility and vapor pressure may be user specified. This enables preliminary assessments of fission product transport in molten salt systems. In this paper, we discuss modeling of soluble and insoluble fission product releases in a MSR with Thermochimica incorporated into MELCOR. Separate-effects experiments performed as part of the Molten Salt Reactor Experiment in which radioactive aerosol was released are discussed as needed for determining the source term.
This document summarily provides brief descriptions of the MELCOR code enhancement made between code revision number 18019and 21440. Revision 18019 represents the previous official code release; therefore, the modeling features described within this document are provided to assist users that update to the newest official MELCOR code release, 21440. Along with the newly updated MELCOR Users’ Guide [2] and Reference Manual [3], users are aware and able to assess the new capabilities for their modeling and analysis applications.
This report represents an assessment of the gaps in Mechanistic Source Term (MST) and consequence assessment modeling for Molten Salt Reactors (MSRs). The current capabilities for MELCOR and the MELCOR Accident Code System (MACCS) are discussed, along with updates needed in order to address specific needs for MSRs. A test plan developed by Argonne National Laboratories is discussed as addressing some of these gaps, while some will require additional attention. Further recommendations are made on addressing these gaps. This report satisfies the DOE NE Milestone M2RD-21SN0601061 to leverage MELCOR and MACCS to identify parameters of importance for source term assessments for salt spill experiments.
This report aids in the development of models to perform characterization studies of aerosol dispersal and deposition within a spent fuel cask system. Due to the complex geometry in a spent-fuel canister, direct simulation of buoyancy-driven flow through the fuel assemblies to model aerosol deposition within the fuel canister is computationally expensive. Identification of an effective permeability as given in this work for a nuclear fuel assembly greatly simplifies the requirements for thermal hydraulic computations. The results of computations performed using OpenFOAM® to solve the Navier-Stokes Equations for laminar flow are used to determine an effective permeability by applying Darcy's Law. The computations are validated against an analytical solution for the special case of an infinite array of pins for which the numerical and analytical solutions have excellent agreement. The effective permeability of a 1717 PWR nuclear fuel assembly in a basket without spacer grids is numerically determined to be 1.85010 -6 m 2 for the range of fluid viscosities and pressure drops expected in a spent fuel storage canister. However, the flow is not uniform on the scale of multiple pins. Instead, significantly higher velocities are attained in the space between the assembly and the basket walls compared to the flow between the fuel pins within the assembly. Comparison with an analytical solution for fully developed flow through an infinite array of pins shows that the larger spacing near the basket walls results in about a 20% larger permeability compared to the analytical solution which does not include the enhanced flow in the space between the assembly and basket wall, or entrance and exit effects. A preliminary assessment of turbulence effects shows that with a k-epsilon model, significantly higher flow velocities are attained between the fuel pins within the assembly compared to the flow velocity in the space between the assembly and the basket walls. This is the opposite of what is determined for laminar flow.
This report documents the progress in preparing the ANSYS/Fluent® and MELCOR models to perform characterization studies of aerosols dispersal and deposition within a select spent fuel cask system. Steady state thermal-hydraulic cask response is modeled with both codes, at present, while the MELCOR source code is being modified to allow imposed thermal-hydraulic conditions with aerosol physics calculations. This will allow the MELCOR model to assume the thermal-hydraulic calculation from the ANSYS/Fluent®, while only computing the aerosol physics. Detailed results are presented on the thermal-hydraulic analysis of the MAGNASTOR® cask for the current ANSYS/Fluent® model, with convergent conditions observed over two fidelities. While the MELCOR computation computes the steady state conditions, they differ sufficiently from that computed with ANSYS/Fluent®. The MELCOR analysis include a set of UO2 sources to investigate system response for interim reporting. The airborne concentrations and evolving distributions are presented. Model development is anticipated to continue as additional components impacting flow conditions or aerosol deposition within the MAGNASTOR® cask and Westinghouse fuel assemblies are identified.
This report represents completion of milestone deliverable M2SF-21SN010309012 “Annual Status Update for OWL and Waste Form Characteristics” that provides an annual update on status of fiscal year (FY 2020) activities for the work package SF-20SN01030901 and is due on January 29, 2021. 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 draft report contains versions of the OWL model architecture for vessel information (Appendix A) and an excerpt from the OWL User’s Guide (Appendix B and SNL 2020), which are for the current OWL Version 2.0 on the Sandia External Collaboration Network (ECN).
One of the objectives of the United States (U.S.) Department of Energy's (DOE) Office of Nuclear Energy's Spent Fuel and Waste Science and Technology Campaign is to better understand the technical basis, risks, and uncertainty associated with the safe and secure disposition of spent nuclear fuel (SNF) and high-level radioactive waste. Commercial nuclear power generation in the U.S. has resulted in thousands of metric tons of SNF, the disposal of which is the responsibility of the DOE (Nuclear Waste Policy Act 1982). Any repository licensed to dispose the SNF must meet requirements regarding the longterm performance of that repository. For an evaluation of the long-term performance of the repository, one of the events that may need to be considered is the SNF achieving a critical configuration. Of particular interest is the potential behavior of SNF in dual-purpose canisters (DPCs), which are currently being used to store and transport SNF but were not designed for permanent geologic disposal. A two-phase study has been initiated to begin examining the potential consequences, with respect to longterm repository performance, of criticality events that might occur during the postclosure period in a hypothetical repository containing DPCs. Phase I, a scoping phase, consisted of developing an approach intended to be a starting point for the development of the modeling tools and techniques that may eventually be required either to exclude criticality from or to include criticality in a performance assessment (PA) as appropriate; Phase I is documented in Price et al. (2019). The Phase I approach guided the analyses and simulations done in Phase II to further the development of these modeling tools and techniques as well as the overall knowledge base. The purpose of this report is to document the results of the analyses conducted during Phase II. The remainder of Section 1 presents the background, objective, and scope of this report, as well as the relevant key assumptions used in the Phase II analyses and simulations. Subsequent sections discuss the analyses that were conducted (Section 2), the results of those analyses (Section 3), and the summary and conclusions (Section 4). This report fulfills the Spent Fuel and Waste Science and Technology Campaign deliverable M2SF-20SN010305061.
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.
The U.S. Department of Energy is conducting research and development on generic concepts for disposal of spent nuclear fuel and high-level radioactive waste in multiple lithologies, including salt, crystalline (granite/metamorphic), and argillaceous (clay/shale) host rock. These investigations benefit greatly from international experience gained in disposal programs in many countries around the world. The focus of this study is the post-closure degradation and radionuclide-release rates for tristructural-isotropic (TRISO) coated particle spent fuels for various generic geologic repository environments.1,2,3 The TRISO particle coatings provide safety features during and after reactor operations, with the SiC layer representing the primary barrier. Three mechanisms that may lead to release of radionuclides from the TRISO particles are: (1) helium pressure buildup4 that may eventually rupture the SiC layer, (2) diffusive transport through the layers (solid-state diffusion in reactor, aqueous diffusion in porous media at repository conditions), and (3) corrosion5 of the layers in groundwater/brine. For TRISO particles in a graphite fuel element, the degradation in an oxidizing geologic repository was concluded to be directly dependent on the oxidative corrosion rate of the graphite matrix4, which was analyzed as much slower than SiC layer corrosion processes. However, accumulated physical damage to the graphite fuel element may decrease its post-closure barrier capability more rapidly. Our initial performance model focuses on the TRISO particles and includes SiC failure from pressure increase via alpha-decay helium, as exacerbated by SiC layer corrosion5. This corrosion mechanism is found to be much faster than solid-state diffusion at repository temperatures but includes no benefit of protection by the other outer layers, which may prolong lifetime. Our current model enhancements include constraining the material properties of the layers for porous media diffusion analyses. In addition to evaluating the SiC layer porosity structure, this work focuses on the pyrolytic carbon layers (inner/outer-IPyC/OPyC) layers, and the graphite compact, which are to be analyzed with the SiC layer in two modes: (a) intact SiC barrier until corrosion failure and (b) SiC with porous media transport. Our detailed performance analyses will consider these processes together with uncertainties in the properties of the layers to assess radionuclide release from TRISO particles and their graphite compacts.
TRISO nuclear fuel particles that are less than 1 mm in diameter are designed with multiple barrier layers to retain fission products both during reactor operations and for long-term geological disposal. The primary barrier is a 35 μm thick silicon carbide (SiC-a highly impermeable semi-metal) layer for which data are available on the diffusion of short-lived fission products at high temperatures (> 1000 °C). However, for a geological repository, this layer may contact brine and hence corrode even at ambient temperatures. As an initial approach to assess the effectiveness of the SiC barrier for geological repositories, ranges of fission product diffusivities and corrosion rates for SiC are modeled concurrently with the simultaneous effect of radioactive decay. Using measured corrosion rates of SiC, if the diffusivity is more than about 10-20 m2/s, fission product releases may occur before the SiC barrier has corroded to the point of breach. For diffusivities less than about 10-21m2 /s there may not be significant diffusional releases prior to SiC barrier removal/breach by corrosion. This work shows the importance of estimating diffusivities in SiC at geological repository temperatures, and highlights the relevance of evaluating the porosity/permeability evolution of the SiC layer in a geologic environment.
This report represents completion of milestone deliverable M2SF-18SNO10309013 "Inventory and Waste Characterization Status Report and OWL Update that reports on FY2018 activities for the work package (WP) SF-18SNO1030901. This report provides the detailed final information for completed FY2018 work activities for WP SF-18SN01030901, and a summary of priorities for FY2019. This status report on FY2018 activities includes evaluations of waste form characteristics and waste form performance models, updates to the OWL development, and descriptions of the two planned 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. There are two processes being planned in FY2018, which will be implemented in FY2019. One process covers methods for interfacing with the DOE SNF DB (DOE 2007) at INL on the numerous entries for DOE managed SNF, and the other process covers the management of updates to, and version control/archiving of, the OWL database. In FY2018, we have pursued three studies to evaluate/redefine waste form characteristics and/or performance models. First characteristic isotopic ratios for various waste forms included in postclosure performance studies are being evaluated to delineate isotope ratio tags that quantitatively identify each particular waste form. This evaluation arose due to questions regarding the relative contributions of radionuclides from disparate waste forms in GDSA results, particularly, radionuclide contributions of DOE-managed SNF vs HLW glass. In our second study we are evaluating the bases of glass waste degradation rate models to the HIP calcine waste form. The HIP calcine may likely be a ceramic matrix material, with multiple ceramic phases with/without a glass phase. The ceramic phases are likely to have different degradation performance from the glass portion. The distribution of radionuclides among those various phases may also be a factor in the radionuclide release rates. Additionally, we have an ongoing investigation of the performance behavior of TRISO particle fuels and are developing a stochastic model for the degradation of those fuels that accounts for simultaneous corrosion of the silicon carbide (SiC) layer and radionuclide diffusion through it. The detailed model of the TRISO particles themselves, will be merged with models of the degradation behavior(s) of the graphite matrix (either prismatic compacts or spherical "pebbles") containing the particles and the hexagonal graphite elements holding the compacts.
Safety basis analysts throughout the U.S. Department of Energy (DOE) complex rely heavily on the information provided in the DOE Handbook, DOE - HDBK - 3010, Airborne Release Fractions/Rates and Respirable Fractions for Nonreactor Nuclear Facilities, to determine radionuclide source terms. In calculating source terms, analysts tend to use the DOE Handbook's bounding values on airborne release fractions (ARFs) and respirable fractions (RFs) for various categories of insults (representing potential accident release categories). This is typically due to both time constraints and the avoidance of regulatory critique. Unfortunately, these bounding ARFs/RFs represent extremely conservative values. Moreover, they were derived from very limited small-scale bench/laboratory experiments and/or from engineered judgment. Thus, the basis for the data may not be representative of the actual unique accident conditions and configurations being evaluated. The goal of this research is to develop a more accurate and defensible method to determine bounding values for the DOE Handbook using state-of-art multi-physics-based computer codes. This enables us to better understand the fundamental physics and phenomena associated with the types of accidents in the handbook. In this year, this research included improvements of the high-fidelity codes to model particle resuspension and multi-component evaporation for fire scenarios. We also began to model ceramic fragmentation experiments, and to reanalyze the liquid fire and powder release experiments that were done last year. The results show that the added physics better describes the fragmentation phenomena. Thus, this work provides a low-cost method to establish physics-justified safety bounds by taking into account specific geometries and conditions that may not have been previously measured and/or are too costly to perform.
In this work we have presented a particle resuspension model implemented in the SNL code SIERRA/Fuego, which can be used to model particle dispersal and resuspension from surfaces. The method demonstrated is applicable to a class of particles, but would require additional parametric fits or physics models for extension to other applications, such as wetted particles or walls. We have demonstrated the importance of turbulent variations in the wall shear stress when considering resuspension, and implemented both shear stress variation models and stochastic resuspension models (not shown in this work). These models can be used in simulations with of physically realistic scenarios to augment lab-scale DOE Handbook data for airborne release fractions and respirable fractions in order to provide confidences for safety analysts and facility designers to apply in their analyses at DOE sites. Future work on this topic will involve validation of the presented model against experimental data and extension of the empirical models to be applicable to different classes of particles and surfaces.
In this work we have presented a particle resuspension model implemented in the SNL code SIERRA/Fuego, which can be used to model particle dispersal and resuspension from surfaces. The method demonstrated is applicable to a class of particles, but would require additional parametric fits or physics models for extension to other applications, such as wetted particles or walls. We have demonstrated the importance of turbulent variations in the wall shear stress when considering resuspension, and implemented both shear stress variation models and stochastic resuspension models (not shown in this work). These models can be used in simulations with of physically realistic scenarios to augment lab-scale DOE Handbook data for airborne release fractions and respirable fractions in order to provide confidences for safety analysts and facility designers to apply in their analyses at DOE sites. Future work on this topic will involve validation of the presented model against experimental data and extension of the empirical models to be applicable to different classes of particles and surfaces.
In a submerged environment, power cables may experience accelerated insulation degradation due to water-related aging mechanisms. Direct contact with water or moisture intrusion in the cable insulation system has been identified in the literature as a significant aging stressor that can affect performance and lifetime of electric cables. Progressive reduction of the dielectric strength is commonly a result of water treeing which involves the development of permanent hydrophilic structures in the insulation coinciding with the absorption of water into the cable. Water treeing is a phenomenon in which dendritic microvoids are formed in electric cable insulation due to electrochemical reactions, electromechanical forces, and diffusion of contaminants over time. These reactions are caused by the combined effects of water presence and high electrical stresses in the material. Water tree growth follows a tree-like branching pattern, increasing in volume and length over time. Although these cables can be “dried out,” water tree degradation, specifically the growth of hydrophilic regions, is believed to be permanent and typically worsens over time. Based on established research, water treeing or water induced damage can occur in a variety of electric cables including XLPE, TR-XLPE and other insulating materials, such as EPR and butyl rubber. Once water trees or water induced damage form, the dielectric strength of an insulation material will decrease gradually with time as the water trees grow in length, which could eventually result in failure of the insulating material. Under wet conditions or in submerged environments, several environmental and operational parameters can influence water tree initiation and affect water tree growth. These parameters include voltage cycling, field frequency, temperature, ion concentration and chemistry, type of insulation material, and the characteristics of its defects. In this effort, a review of academic and industrial literature was performed to identify: 1) findings regarding the degradation mechanisms of submerged cabling and 2) condition monitoring methods that may prove useful in predicting the remaining lifetime of submerged medium voltage power cables. The research was conducted by a multi-disciplinary team, and sources included official NRC reports, national laboratory reports, IEEE standards, conference and journal proceedings, magazine articles, PhD dissertations, and discussions with experts. The purpose of this work was to establish the current state-of-the-art in material degradation modeling and cable condition monitoring techniques and to identify research gaps. Subsequently, future areas of focus are recommended to address these research gaps and thus strengthen the efficacy of the NRC’s developing cable condition monitoring program. Results of this literature review and details of the testing recommendations are presented in this report.
In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. NASA has prepared an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS includes information on the risks of mission accidents to the general public and on-site workers at the launch complex. The Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG option to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of the MMRTG option for the EIS. This paper provides a summary of the methods and results used in the NRA.
Spent nuclear fuel reprocessing may involve some hazardous liquids that may explode under accident conditions. Explosive accidents may result in energetic dispersion of the liquid. The atomized liquid represents a major hazard of this class of event. The magnitude of the aerosol source term is difficult to predict, and historically has been estimated from correlations based on marginally relevant data. A technique employing a coupled finite element structural dynamics and control volume computational fluid dynamics has been demonstrated previously for a similar class of problems. The technique was subsequently evaluated for detonation events. Key to the calculations is the use of a Taylor Analogy Break-up (TAB) based model for predicting the aerodynamic break-up of the liquid drops in the air environment, and a dimensionless parameter for defining the chronology of the mass and momentum coupling. This paper presents results of liquid aerosolization from an explosive event.
Spent nuclear fuel reprocessing may involve some hazardous liquids that may explode under accident conditions. Explosive accidents may result in energetic dispersion of the liquid. The atomized liquid represents a major hazard of this class of event. The magnitude of the aerosol source term is difficult to predict, and historically has been estimated from correlations based on marginally relevant data. A technique employing a coupled finite element structural dynamics and control volume computational fluid dynamics has been demonstrated previously for a similar class of problems. The technique was subsequently evaluated for detonation events. Key to the calculations is the use of a Taylor Analogy Break-up (TAB) based model for predicting the aerodynamic break-up of the liquid drops in the air environment, and a dimensionless parameter for defining the chronology of the mass and momentum coupling. This paper presents results of liquid aerosolization from an explosive event.
In the summer of 2020, the National Aeronautics and Space Administration (NASA) plans to launch a spacecraft as part of the Mars 2020 mission. One option for the rover on the proposed spacecraft uses a Multi-Mission Radioisotope Thermoelectric Generator (MMRTG) to provide continuous electrical and thermal power for the mission. An alternative option being considered is a set of solar panels for electrical power with up to 80 Light-Weight Radioisotope Heater Units (LWRHUs) for local component heating. Both the MMRTG and the LWRHUs use radioactive plutonium dioxide. NASA is preparing an Environmental Impact Statement (EIS) in accordance with the National Environmental Policy Act. The EIS will include information on the risks of mission accidents to the general public and on-site workers at the launch complex. This Nuclear Risk Assessment (NRA) addresses the responses of the MMRTG or LWRHU options to potential accident and abort conditions during the launch opportunity for the Mars 2020 mission and the associated consequences. This information provides the technical basis for the radiological risks of both options for the EIS.
A comprehensive test program to evaluate nonmetallic materials use in the Hanford Tank Farms is described in detail. This test program determines the effects of simultaneous multiple stressors at reasonable conditions on in-service configuration components by engineering performance testing.
The consumption of petroleum by the transportation sector in the United States is roughly equivalent to petroleum imports into the country, which have totaled over 12 million barrels a day every year since 2004. This reliance on foreign oil is a strategic vulnerability for the economy and national security. Further, the effect of unmitigated CO{sub 2} releases on the global climate is a growing concern both here and abroad. Independence from problematic oil producers can be achieved to a great degree through the utilization of non-conventional hydrocarbon resources such as coal, oil-shale and tarsands. However, tapping into and converting these resources into liquid fuels exacerbates green house gas (GHG) emissions as they are carbon rich, but hydrogen deficient. Revolutionary thinking about energy and fuels must be adopted. We must recognize that hydrocarbon fuels are ideal energy carriers, but not primary energy sources. The energy stored in a chemical fuel is released for utilization by oxidation. In the case of hydrogen fuel the chemical product is water; in the case of a hydrocarbon fuel, water and carbon dioxide are produced. The hydrogen economy envisions a cycle in which H{sub 2}O is re-energized by splitting water into H{sub 2} and O{sub 2}, by electrolysis for example. We envision a hydrocarbon analogy in which both carbon dioxide and water are re-energized through the application of a persistent energy source (e.g. solar or nuclear). This is of course essentially what the process of photosynthesis accomplishes, albeit with a relatively low sunlight-to-hydrocarbon efficiency. The goal of this project then was the creation of a direct and efficient process for the solar or nuclear driven thermochemical conversion of CO{sub 2} to CO (and O{sub 2}), one of the basic building blocks of synthetic fuels. This process would potentially provide the basis for an alternate hydrocarbon economy that is carbon neutral, provides a pathway to energy independence, and is compatible with much of the existing fuel infrastructure.
The Arrhenius parameters for graphite oxidation in air are reviewed and compared. One-dimensional models of graphite oxidation coupled with mass transfer of oxidant are presented in dimensionless form for rectangular and spherical geometries. A single dimensionless group is shown to encapsulate the coupled phenomena, and is used to determine the effective reaction rate when mass transfer can impede the oxidation process. For integer reaction order kinetics, analytical expressions are presented for the effective reaction rate. For noninteger reaction orders, a numerical solution is developed and compared to data for oxidation of a graphite sphere in air. Very good agreement is obtained with the data without any adjustable parameters. An analytical model for surface burn-off is also presented, and results from the model are within an order of magnitude of the measurements of burn-off in air and in steam.
A sulfuric acid catalytic decomposer section was assembled and tested for the Integrated Laboratory Scale experiments of the Sulfur-Iodine Thermochemical Cycle. This cycle is being studied as part of the U. S. Department of Energy Nuclear Hydrogen Initiative. Tests confirmed that the 54-inch long silicon carbide bayonet could produce in excess of the design objective of 100 liters/hr of SO{sub 2} at 2 bar. Furthermore, at 3 bar the system produced 135 liters/hr of SO{sub 2} with only 31 mol% acid. The gas production rate was close to the theoretical maximum determined by equilibrium, which indicates that the design provides adequate catalyst contact and heat transfer. Several design improvements were also implemented to greatly minimize leakage of SO{sub 2} out of the apparatus. The primary modifications were a separate additional enclosure within the skid enclosure, and replacement of Teflon tubing with glass-lined steel pipes.
