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DOE OSTI2015

Status Report on Activities of the Systems Assessment Task Force, OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs

The Organization for Economic Cooperation and Development /Nuclear Energy Agency (OECD/NEA) Nuclear Science Committee approved the formation of an Expert Group on Accident Tolerant Fuel (ATF) for LWRs (EGATFL) in 2014. Chaired by Kemal Pasamehmetoglu, INL Associate Laboratory Director for Nuclear Science and Technology, the mandate for the EGATFL defines work under three task forces: (1) Systems Assessment, (2) Cladding and Core Materials, and (3) Fuel Concepts. Scope for the Systems Assessment task force includes definition of evaluation metrics for ATF, technology readiness level definition, definition of illustrative scenarios for ATF evaluation, parametric studies, and selection of system codes. The Cladding and Core Materials and Fuel Concepts task forces will identify gaps and needs for modeling and experimental demonstration; define key properties of interest; identify the data necessary to perform concept evaluation under normal conditions and illustrative scenarios; identify available infrastructure (internationally) to support experimental needs; and make recommendations on priorities. Where possible, considering proprietary and other export restrictions (e.g., International Traffic in Arms Regulations), the Expert Group will facilitate the sharing of data and lessons learned across the international group membership. The Systems Assessment Task Force is chaired by Shannon Bragg-Sitton (INL), while the Cladding Task Force will be chaired by a representative from France (Marie Moatti, Electricite de France [EdF]) and the Fuels Task Force will be chaired by a representative from Japan (Masaki Kurata, Japan Atomic Energy Agency [JAEA]). This report provides an overview of the Systems Assessment Task Force charter and status of work accomplishment.

99 GENERAL AND MISCELLANEOUS↗

NASA NTRS1996

Near-Optimal Operation of Dual-Fuel Launch Vehicles

A near-optimal guidance law for the ascent trajectory from earth surface to earth orbit of a fully reusable single-stage-to-orbit pure rocket launch vehicle is derived. Of interest are both the optimal operation of the propulsion system and the optimal flight path. A methodology is developed to investigate the optimal throttle switching of dual-fuel engines. The method is based on selecting propulsion system modes and parameters that maximize a certain performance function. This function is derived from consideration of the energy-state model of the aircraft equations of motion. Because the density of liquid hydrogen is relatively low, the sensitivity of perturbations in volume need to be taken into consideration as well as weight sensitivity. The cost functional is a weighted sum of fuel mass and volume; the weighting factor is chosen to minimize vehicle empty weight for a given payload mass and volume in orbit.

Ardema, M. D.↗

DOE OSTI2019

Simulation of Spent Nuclear Fuel loading into a Final Waste Repository

Previous work towards studying repository loading have used: - spent nuclear fuel (SNF) with an average burnup composition; - a lumped capacitance thermal model for calculating temperature in a Cyclus repository model. The goal of this work is to improve on the repository models and use U.S. historical SNF inventory data in simulations to more accurately study the loading of a waste repository. These goals will be achieved by: - using UNF-ST-and-DARDS Unified Database (UDB) that has historic assembly-specific data (e.g, isotopic composition, heat) in Cyclus simulations; - implementing a more accurate thermal model within a Cyclus repository model. Objectives: - Create a Cyclus spent fuel conditioning model that packages spent fuel bundles into packages which have user-defined properties. - Create a Cyclus medium-fidelity repository model that gives accurate time and spatial dependent temperature values and loads the repository based on a user-selected loading strategy. Cyclus is an agent-based extensible framework for modeling flow of material through user-defined nuclear fuel cycles. In Cyclus, each facility in the fuel cycle is modeled individually and the facilities interact with one another as independent agents. The spent fuel conditioning model accepts spent fuel bundles and puts them into a cylindrical waste package. In the spent fuel conditioning model, the user can define variables: For each layer, radius, thermal conductivity, thermal diffusivity; For each package, Number of spent fuel bundles, Radius and height. The waste repository model accepts waste packages and emplaces them into available positions within the waste repository based on a thermal criteria. The thermal criteria is a temperature limit at the interface between the waste package surface and the host geology, that is set based on the repository's host geology. In the waste repository model, the user can define the variables: Capacity, Distance between waste packages, Distance between drifts, Repository host geology, Loading Strategy. After the addition of new waste packages at each time step, the waste repository model recalculates the temperature at each location in the repository. If the addition of this new package causes its temperature to exceed the thermal limit, it will be placed back into the buffer. A thermal model that relies on a transient 'outside' model and quasi-steady-state 'inside' model is used to accurately determine the temperature in the repository. The 'outside' model assumes a homogenous medium with the Engineered Barrier System (EBS) replaced by the geologic medium. Temperature solutions for the central waste package, adjacent point and line sources are superimposed to calculate the temperature at specific points in the repository. The central drift consists of one finite line source which represents the central waste package. The central drift also consists of point sources that represent neighboring waste packages in the central drift. The neighboring drifts are represented by infinite line sources. The 'inside' model is considered to be at a quasi-steady-state condition because EBS has a relatively low thermal mass compared to the infinite geologic medium. The steady state calculation is performed at each time step with the heat source and interface temperature as boundary conditions. Future Work: Run Cyclus simulations with U.S. historical SNF inventory data, the spent fuel conditioning and repository models to study how waste package acceptance strategies impact repository loading.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

JASMINE: a fuel rod thermal-mechanical performance's code

JASMINE, which is developed by China General Nuclear Power Corporation (CGNPC), is a code for calculating thermal-mechanical response of PWR fuel rods. The code models many physical phenomena, including heat conduction through the fuel and cladding to the coolant, cladding elastic and plastic deformation, fuel elastic deformation, fuel swelling and densification,fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, cladding oxidation and hydriding. Just as most fuel rod codes, JASMINE, which is a quasi-two dimensional code, includes a thermal module, a mechanical module, a fission gas release (FGR) module, and an internal pressure module. For thermal module, the time dependent thermal conduction equation of axisymmetric cylindrical geometry, Halden's fuel thermal conductivity, thermal conductivity integration method and multi-dimension Newton-Raphson procedure are used; for the mechanical module, generalized plane strain hypothesis, cladding-fuel axial lockup model, cladding thermal stress and transfer matrix method are applied; for fission gas release module, the time dependent diffusion equation for a spherical grain, Speight-Turnbull grain surface boundary condition, and Forsberg-Massih's method are used; for the internal pressure module, the ideal gas state equation is used. The code, which is on validation stage, will be used for fuel rod design of nuclear power plants in China in a near future. So far, the preliminary validation process for pellet centerline temperature and fission gas release, which is based on fuel rod test data in FUMEX-III project, has finished. According to the validation results, JASMINE has the ability to reasonably predicts fuel behavior and performance in-pile. JASMINE has good prediction in fuel temperature, and good prediction on FGR as a whole. However, for FGR in the range 1% to 15%, JASMINE over-predicts FGR; for other range, JASMINE underestimates FGR. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

Performance and Cost Assessment of a Natural Gas-Fueled Direct sCO 2 Power Plant

This final report details the cost and performance baseline developed at NETL for a natural gas-fueled direct sCO 2 plant, which holds promise as an efficient, cost-effective method of power production with carbon capture and storage (CCS). The steady state plant model includes turbine cooling flows based on a semi-empirical cooling model, and thermal integration with the air separation unit and oxygen compression train. Discussion of the performance of the baseline plant design is followed by details on the costing methodology, overall plant capital cost, and resulting cost of electricity (COE). Performance results yield a baseline sCO 2 plant thermal efficiency of 48.2% (HHV) with 98.2% carbon capture, an improvement over the reference NGCC plant with CCS (45.7% and 90.7% carbon capture). Cost analyses yield a 5% reduction in COE versus the NGCC plant, albeit with significant uncertainty. Finally, the effects of incomplete combustion on the overall plant performance are assessed.

03 NATURAL GAS↗

DOE OSTI2023

Unsaturated alluvium disposal modelling with improved geological realism

This report describes specific activities in the Fiscal Year (FY) 2023 associated with the Geologic Disposal Safety Assessment (GDSA) Repository Systems Analysis (RSA) work package funded by the Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Spent Fuel and Waste Disposition (SFWD). The overall objective of the GDSA RSA work package is to develop generic deep geologic repository concepts and system performance assessment (PA) models in several potential host-rock environments, and to simulate and analyze these generic repository concepts and models using the GDSA Framework toolkit and other tools as needed. The specific GDSA goal addressed in this report is reference case development, simulation, and analysis for the unsaturated alluvium (UZ), one of the four potential host-rocks considered by the GDSA. Further, we aim to exercise the simulation tools and methodologies under development by GDSA for PA modelling.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2017

Final Report: Utilizing Alternative Fuel Ignition Properties to Improve SI and CI Engine Efficiency

Experimental and modeling studies were completed to explore leveraging physical and chemical fuel properties for improved thermal efficiency of internal combustion engines. Fundamental studies of the ignition chemistry of ethanol and iso-octane blends and constant volume spray chamber studies of gasoline and diesel sprays supported the core research effort which used several reciprocating engine platforms. Single cylinder spark ignition (SI) engine studies were carried out to characterize the impact of ethanol/gasoline, syngas (H 2 and CO)/gasoline and other oxygenate/gasoline blends on engine performance. The results of the single-cylinder engine experiments and other data from the literature were used to train a GT Power model and to develop a knock criteria based on reaction chemistry. The models were used to interpret the experimental results and project future performance. Studies were also carried out using a state of the art, direct injection (DI) turbocharged multi- cylinder engine with piezo-actuated fuel injectors to demonstrate the promising spray and spark timing strategies from single-cylinder engine studies on the multi-cylinder engine. Key outcomes and conclusions of the studies were: 1. Efficiency benefits of ethanol and gasoline fuel blends were consistent and substantial (e.g. 5-8% absolute improvement in gross indicated thermal efficiency (GITE)). 2. The best ethanol/gasoline blend (based on maximum thermal efficiency) was determined by the engine hardware and limits based on component protection (e.g. peak in-cylinder pressure or maximum turbocharger inlet temperature) – and not by knock limits. Blends with <50% ethanol delivered significant thermal efficiency gains with conventional SI hardware while maintain good safety integrity to the engine hardware. 3. Other compositions of fuel blends including syngas (H 2 and CO) and other dilution strategies provided significant efficiency gains as well (e.g. 5% absolute improvement in ITE). 4. When the combination of engine and fuel system is not knock limited, multiple fuel injection events maintain thermal efficiency while improving engine-out emissions (e.g. CO, UHC, and particulate number).

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2019

A Preliminary Probabilistic Performance Assessment Model for Radiological and Chemical Contamination at the West Valley Site, New York - 19295

The New York State Energy Research and Development Authority (NYSERDA) is the owner of the Western New York Nuclear Service Center (WNYNSC), a 1,351-ha site located approximately 48 km south of Buffalo, New York. In 1962, Nuclear Fuel Services, Inc. (NFS) entered into agreements with the Atomic Energy Commission and New York State to construct the first commercial reprocessing plant of nuclear fuel in the United States at the WNYNSC. NFS built and operated the spent fuel reprocessing plant and waste disposal facilities, processing 640 Mg (640 metric tons) of spent nuclear fuel from 1966 to 1972 under an Atomic Energy Commission license. Nuclear fuel reprocessing operations halted in 1972 and never restarted, leaving behind radioactive and chemical wastes. The reprocessing wastes include high-level waste (HLW), so U.S. Department of Energy (DOE) was compelled to complete certain waste management activities under the West Valley Demonstration Project (WVDP) Act of 1980, including decommissioning of WVDP facilities. As collaborating agencies, NYSERDA and the DOE are tasked with making decisions about decommissioning and risk reduction for the West Valley Site. Neptune and Company, Inc. (Neptune) was contracted to develop a probabilistic performance assessment (PPA) model to assist the agencies in their decision making process for decommissioning the WVDP and WNYNSC. One important tool that is needed in order to inform the decision-making process is a science-based model of the contamination at the West Valley Site and its potential implications for human health and the environment. The West Valley PPA Model (the Model), developed using the GoldSim modeling software, is a tool intended to provide support for decision making with uncertainty, in a manner that is transparent, defensible, and robust. The PPA Model includes contaminant transport and health effects components, and is organized around geographically-organized contaminated facilities. These include the waste disposal areas licensed by the U.S. Nuclear Regulatory Commission and the State of New York, a waste tank farm for storage of HLW resulting from reprocessing operations, and several areas contaminated with radioactive and chemical constituents. The Model evaluates contaminant transport from these sources to points of exposure across the site and into receiving surface waters and sediments. Hypothetical people and wildlife are exposed to contamination at these locations, and the effects of these exposures are evaluated. Contaminant transport processes to be evaluated in the Model include groundwater and surface water transport, contaminant translocation by plants and animals, diffusion in subsurface air and water, and erosion of the Site. The evaluation of exposures to people in this preliminary model is limited to a resident farmer scenario, and ecological assessment is performed at the level of a screening analysis. The results of these preliminary evaluations inform future model developments. Model results are subjected to sensitivity analysis in order to determine those pathways and parameters that are most significant in influencing the results. This information allows analysts and decision makers to focus on those aspects of Site behavior and processes that contribute most to overall uncertainty. With this information, the Supplemental Environmental Impact Statement (SEIS) can inform defensible decisions regarding remediation of the Site. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Radiation-induced bowing of SiC/SiC composites under neutron flux gradients—integral experimental data for model validation

Here, the radiation-induced swelling of SiC and its composites, including strong dependencies on temperature and dose, can drive significant lateral bowing in the presence of temperature and/or dose gradients. In recent years, simulations have been performed to assess the extent of bowing in SiC composite light-water reactor (LWR) fuel cladding and boiling water reactor (BWR) channel boxes. However, to date, no integral experimental data exist to validate these models. This work provides the first experimental bowing evaluation of three ∼380 mm long SiC composite specimens irradiated under varying neutron dose gradients (∼50°C–60°C, 0.03–0.06 dpa): two tubes (∼9.8 mm diameter) and a miniature BWR channel box (∼30 mm square). The measured radiation-induced length swelling (∼0.3%–0.7% linear) was consistently 10%–21% higher than values obtained from 3D finite element structural analyses with inputs from 3D radiation transport calculations. This discrepancy could be at least partially explained by differences in dose rate (∼10 -8 dpa/s) compared to the literature data (∼10-6 dpa/s) used to establish the dose-to-swelling correlations in the model. Nevertheless, the modeled bowing magnitudes (<2 mm) obtained from finite element analyses and simple analytical equations were within the bounds of the experimental measurements for all specimens. With improved confidence in the ability to predict the structural response and measure the macroscopic deformations, future experiments will target transient bowing under neutron flux gradients at representative LWR temperatures and assess whether grid spacers can mitigate the tens of millimeters of bowing that would otherwise be expected in ∼4 m long LWR components.

bowing↗

DOE OSTI2024

Analysis of buffer-IPyC separation in TRISO fuel particles

During High Temperature Gas-cooled Reactor (HTGR) operation, and due to the neutron irradiation in the reactor core, damage of the nuclear fuel coating layers occurs. The mechanism of damage formation in the TRISO fuel is explored by the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program, in which the debonding process between coating layers was also investigated. The purpose of this paper is to report simulation results for two models. Firstly, the debonding restricted model, where no gap formation between buffer and IPyC layers is permitted. Secondly, a debonding enabled model, where the gap between those layers is created. The simulations were performed with the Bison code. The inputs of the simulated models are based on data from the AGR-1 experiment. Further, the research included simulations on spherical and aspherical fuel types. The simulations match results obtained by the AGR-1 experiment, which as such shows that the Bison code is a good computational method for simulating the behavior of the gap between buffer and IPyC layers in TRISO fuel. Based on the irradiation experiments, and the Bison simulations, it was concluded that the most common scenario is a gap formation along the buffer-IPyC interface, while the least possible scenario is the situation where there is no gap formation at the buffer-IPyC junction. The computational results confirmed that the sphericity of the fuel influences the thickness of the gap that occurs at the buffer-IPyC junction, in a way that with increasing aspect ratio the gap thickness increases. In addition, the fuel sphericity does not influence the Weibull failure probability. The results obtained for spherical and aspherical fuel are nearly identical.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Design, Analysis, and Experimental Testing of Hydrogen Lean Direct Injection Nozzles at Elevated Pressure

Abstract There are many challenges of commissioning a hydrogen combustor into future gas turbine engines; especially regarding achieving emissions goals. Previously, Escudero et al. and Tran et al. conducted a study to adapt the liquid fuel Lean Direct Injection (LDI) concept from Jet-A to gaseous natural gas-hydrogen blends and pure hydrogen [1], [2]. Experimental data was collected at atmospheric conditions using a Box Behnken design of experiments. The design of experiments suggested that biasing the air split in favor of the inner air circuit and increasing the swirl strength of this inner air passage resulted in improved NOx emissions, while the inverse was true for stability, which was quantified by studying the lean blowoff point (LBO) [1], [2]. The trends revealed by the original experiment [1], [2] provided a design direction for further iterations of the experimental hardware. The study presented herein describes the further investigation of such LDI injectors through experimental methods and computational fluid dynamic (CFD) simulations at atmospheric conditions, which were used to identify potential flow behaviors driving enhanced emissions performance. Further evaluation of select injectors from both studies was then conducted at elevated pressures up to 6 atmospheres. The results from both experiments are presented in this study, which include flame observations, emissions measurements, and operational challenges. NOx emissions results are reported on a volume basis in ppmvd corrected to 15% O2 and corrected for fuel. A predictive model for relating NOx emissions to test conditions at atmospheric conditions show high significance to adiabatic flame temperature while little to no significance to fuel composition for the best performing configurations. The results illustrate the connection between atmospheric testing and testing elevated pressures. The design direction indicated by the initial tests and CFD results in promising configurations for implementation into a Multi-point LDI array.

08 HYDROGEN↗

DOE OSTI2024

Towards a NEAMS-based high-fidelity model of the MARVEL reactor

This report outlines the progress of Idaho National Laboratory in developing a high-fidelity and high-resolution model of the Microreactor Applications Research Validation and Evaluation reactor. The model was developed under the Nuclear Energy Advanced Modeling and Simulation microreactor application driver at Idaho National Laboratory. The overarching objective of this activity is the development of a high-fidelity multiphysics MARVEL model using NEAMS tools, and to verify and validate NEAMS tools against MARVEL reference simulation and experimental data, respectively. This is a unique opportunity to conduct multiphysics analysis on a soon-to-be-deployed microreactor. This multiphysics model developed under the NEAMS-funded INL microreactor application driver leverages three single-physics models coupled via the MOOSE’s MultiApp and Transfer systems. The latter systems enable in-memory data transfer between MOOSE-based and MOOSE-wrapped applications. The first single-physics model, that functions as main application, leverages Griffin to model the neutron transport in the core through the discontinuous finite element (DFEM) discrete ordinates solver (SN). Several optimization flags that were developed by the Griffin developer team were beta-tested to enhance the solver’s performance. These include the combined use of using_average_xs and update_averaged_xs_on that enable to avoid expensive on-the-fly cross sections evaluations at each linear iterations in favor of evaluations of the macroscopic cross sections at each Picard iteration. The second single-physics model uses BISON to handle solid heat transfer and asymptotic hydrogen redistribution analysis in the fuel. While the model returns consistent results for the temperature and hydrogen distribution in the fuel, a mismatch was noticed in the calculated temperature in the reflector due to the value of the gap conductance used in our model. Ongoing investigations are being performed to assess the origin of this discrepancy. Finally, the System Analysis Module (SAM) was used to model the flow of the sodium-potassium eutectic in the primary loop. A first verification was also performed showing good agreement in terms of mass flow rate and inlet temperature. All mesh files were generated using the MOOSE Reactor module, removing the need for external meshing tools. Notably, this workscope represents one of the initial applications of the MOOSE Reactor module for modeling highly irregular geometries. The use of the reactor module significantly streamlined the mesh generation process. The full multiphysics mode, that combines all the single physics models, was leveraged to conduct initial steady-state multiphysics simulations to compute power, and temperature distribution in the reactor. Initial testing was performed for transient simulations as well. In this case, the new checkpoint restart capability for eigenvalue calculations was tested showing the capability for streamlined restart of transient calculations. Future work will focus on improving the fidelity of the model by performing comprehensive code-to-code comparisons. For instance, the full-core Griffin neutronics model will be benchmarked against MCNP reference results, that were provided by the MARVEL design team. Additionally, the SAM T/H model will be verified against reference RELAP-5 results for selected accident scenarios. Besides code-to-code verification exercises, the model fidelity will be improved by replacing the single-channel SAM model with a more complex SAM-Pronghorn coupled model, in which the sub-channel capability is deployed to obtain radial temperature resolution in the coolant. This model will be developed in synergy with the NEAMS thermal hydraulics team.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

STEM/EELS Study of Fission Product Transport in Neutron Irradiated TRISO Fuel Particles (NSUF report)

The objectives of this project are to study the distribution, composition and structure of fission products in neutron irradiated SiC layer of the TRISO particles, using electron microscopy including scanning transmission electron microscopy (STEM), energy dispersive x-ray spectroscopy (EDS), electron energy loss spectroscopy (EELS), high-resolution transmission electron microscopy (HRTEM) and precession electron diffraction (PED), in an effort to enhance the understanding of distribution, composition, structure and transport mechanisms of fission products. In the past 40 years, numerous studies, including reactor experiments, out-of-pile experiments, and simulations, have been performed to investigate the fission product transport behavior and mechanisms. Grain boundary (GB) diffusion, neutron-induced diffusion, catalytic Pd-assisted transport, and vapor-phase migration all have been suggested as possible governing transport mechanisms. The previous studies, however, have not been able to reproduce the transport behavior of fission products and the mechanisms remain poorly understood. This project is anticipated to contribute to enhancing the understanding of the mechanisms of fission product transport in irradiated SiC layer of the TRISO particles. Thorough understanding of fission product transport mechanisms in the SiC layer will improve fuel modeling and design and better SiC layers and TRISO coated particles can be designed to enhance retention of fission products and reduce their release. Eventually, TRISO coated particles with improved performance can be fabricated, and such advancement will expedite the qualifying and licensing processes for TRISO coated particles. Hence, this research is contributing to achieving the objectives of DOE-NE to develop the next generation nuclear reactors, particularly high-temperature gas reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

A wide range experimental and kinetic modeling study of the oxidation of 2,3-dimethyl-2-butene: Part 1

2,3-Dimethyl-2-butene (TME) is a potential fuel additive with high research octane number (RON) and octane sensitivity (S), which can improve internal combustion engine performance and efficiency. However, the combustion characteristics of TME have not been comprehensively investigated. Thus, it is essential to study the combustion characteristics of TME and construct a detailed chemical kinetic model to describe its combustion. In this paper, two high-pressure shock tubes and a constant-volume reactor are used to measure ignition delay times and laminar flame speeds of TME oxidation. Further, the ignition delay times were measured at equivalence ratios of 0.5, 1.0, and 2.0 in “air”, at pressures of 5 and 10 bar, in the temperature range of 950 – 1500 K. Flame speeds of the TME/ “air” mixtures were measured at atmospheric pressure, at a temperature of 325 K, for equivalence ratios ranging from 0.78 to 1.31. Two detailed kinetic mechanisms were constructed independently using different methodologies; the KAUST TME mechanism was constructed based on NUIGMech1.1, and the MIT TME mechanism was built using the Reaction Mechanism Generator (RMG). Both mechanisms were used to simulate the experimental results using Chemkin Pro. In the present work, reaction flux and sensitivity analyses were performed using the KAUST mechanism to determine the critical reactions controlling TME oxidation at the conditions studied.

33 ADVANCED PROPULSION SYSTEMS↗

DOE OSTI2021

(Invited) Understanding Multi-Component Transport and Reaction Phenomena in Bipolar Membranes Used for Electrosynthesis

Bipolar membranes (BPMs), which have long seen usage in electrodialysis reactors for the generation of acid and base, have recently demonstrated potential to become critical components in electrochemical synthesis devices. Because they can operate under large pH gradients, BPMs enable favorable environments for electrocatalysis at the individual electrodes. Critical to the implementation of BPMs in these devices is understanding the kinetics of water dissociation that occurs within the BPM junction as well as the co- and counter-ion crossover through the BPM, which both present significant obstacles to developing efficient and stable BPM-devices for electrosynthesis applications. Prior work has modeled ion transport in bipolar membranes in neutral salt solutions for electrodialysis. However, no model exists for the BPM under the harsh applied pH gradients that would be present in electrosynthesis, and there is significant need to explore the effects of the internal hydration on the lifetime and performance of BPMs in such environments. Additionally, a mechanistic understanding of water dissociation catalysis will be required to develop interfacial catalysts that enable the high current density operation required for scalable electrosynthesis of fuels. In this talk, we discuss modeling methodologies and physics inherent in BPMs and present our recent model of ion transport and water dissociation catalysis in BPMs across the pH scale. Specifically, we simulate multi-ion transport for a BPM with various electrolyte combinations on each side of the membrane, demonstrating the significance of co- and counter-ion crossover in BPMs operating under harsh pH gradients. We then investigate effects caused by hydration gradients that occur due to internal ion-exchange and examine potential methods for improving performance and mitigating crossover. Finally, we examine the impact of the interfacial water dissociation catalyst and perform sensitivity analysis on the key properties (catalyst point of zero charge and pKa) that dictate catalyst performance. These results provide information that is critical to developing a comprehensive understanding of multi-component phenomena in BPMs and to informing the design and implementation of BPMs in next-generation devices for the numerous electrosynthesis chemistries that benefit from operation under an applied pH gradient.

Bui, Justin C↗

DOE OSTI2017

Steady-State Subchannel Thermalhydraulic Assessment of Thorium-Based Fuel Concepts for Use in Pressure Tube Heavy Water Reactors

Thorium is an abundant alternative fertile nuclear fuel that could potentially be implemented in pressure tube heavy water reactors (PT-HWR) in the near future. One of the key stages in developing the thorium-based fuels for PT-HWRs is the reactor core concept, which requires the collaboration of different disciplines, such as neutron physics, thermal-hydraulics, materials, chemistry, safety, and others. Currently there are numerous core concepts under investigation at the Canadian Nuclear Laboratories that involve several types of thorium-based fuels, and that could be implemented in a 700-MWe-class PT-HWR. Four fuel bundle concepts are considered, including i) the reference (or nominal) 37-element bundle; ii) a 37-element modified bundle, with the center element using a different fuel material; iii) a 35-element bundle design, and iv) an 18-element internally cooled annular fuel (ICAF-18) fuel bundle. This study presents preliminary results of the steady-state thermalhydraulic subchannel analyses of the hot channels of different PT-HWR core concepts using thorium-augmented or thorium-based fuels. A follow up study will be initiated to improve the assessment of the proposed cores and associated fuel bundle concepts for thorium-augmented and thorium-based fuels. The computational model will be revisited to include; - Expanding the analysis to cover different power profiles at different burnup stages of the bundle. - Expand the analysis to cover 28.5 kg/s. - Reduce the eccentricity of the 35-element fuel bundle. - Assess the effect of the phenomena not yet included in the subchannel model (such as the effect of cold wall, and gamma heat). - Expand the analysis to include the effect of spacers and junctions. - Compare bundle concepts using other performance indicators (such as burnup, breeding ratio), as optimizing the fuel bundle geometry is a multidisciplinary problem.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Design and Testing for a New Thermosyphon Irradiation Vehicle

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory (ORNL) requires most materials and all fuel experiments to be placed in a pressure containment vessel to ensure that internal contaminants such as fission products cannot be released into the primary coolant. It also requires that all experiments be capable of withstanding various accident conditions (e.g., loss of coolant) without generating vapor bubbles on the surface of the experiment in the primary coolant. These requirements are intended to artificially increase experiment temperatures by introducing a barrier between the experimental materials and the HFIR coolant, and by reducing heat loads to the HFIR primary coolant, thus ensuring that no boiling can occur. A proposed design for materials irradiation would remove these limitations by providing the required primary containment with an internal cooling flow. This would allow for experiments to be irradiated without concern for coolant contamination (e.g., from cladding failure of advanced fuel pins) or for specimen heat load. This report describes a new materials irradiation experiment design that uses a thermosyphon cooling system to allow experimental materials direct access to a liquid coolant. The new design also increases the range of conditions that can be tested in HFIR. This design will provide a unique capability to validate the performance of current and advanced fuels and materials. Because of limited supporting data for this kind of irradiation vehicle, a test program was initiated to obtain operating data that can be used to (1) qualify the vehicle for operation in HFIR and (2) validate computer models used to perform design- and safety-basis calculations. This report also describes the test facility and experimental data, and it provides a comparison of the experimental data to computer simulations. A total of 51 tests have been completed: four tests with pure steam, 12 tests with argon, and 35 tests with helium. A total of 10 tests were performed at subatmospheric pressure, and four of these were performed with pure steam. One test was conducted at a high power of 92.7 kW, six tests were HFIR startups, and two tests were HFIR loss of offsite power (LOOP). Pressures up to 10 MPa, vapor temperatures up to 583 K (310°C), and heater temperatures above 600 K (327°C) have been reached in these tests. Two computer programs, RELAP5-3D and TRACE, have been used to simulate the tests. The TRACE code has shown good agreement with the test data and has been used to model a variety of tests. This experimental facility has been very useful in demonstrating the viability of this new type of irradiation facility.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2017

Manufacturing of Low Cost, Durable Membrane Electrode Assemblies Engineered for Rapid Conditioning

Over the past 20 years significant progress in membrane-electrode assembly (MEA) technology development for polymer electrolyte fuel cells (PEMFCs) has resulted in the PEMFC technology approaching a commercial reality for transportation applications. However, there remain two primary technical challenges to be addressed in the MEA. First and foremost is meeting the automotive cost targets: Producing a fuel cell stack cost competitive with today’s internal combustion engine. In addition to the material cost, MEA (and other components) and stack assembly production methods must be amenable for use in low cost, high speed, automotive assembly line. One impediment to this latter goal is that stack components must currently go through a long and tedious conditioning procedure before they produce optimal power. This so-called “break-in” can take many hours, and can involve quite complex voltage, temperature and/or pressure steps. These break-in procedures must be simplified and the time required reduced if fuel cells are to become a viable automotive engine. The second challenge is to achieve the durability targets in real-world automotive duty cycle operations. Significant improvements in cost, break-in time, and durability for the key component of fuel cell stacks, MEAs were achieved in this project. Advanced modeling was used to guide design of the new MEA to maximize performance and durability. A new, innovative process and manufacturing approach utilizing direct in-line coating using scalable, cost-competitive, continuous high volume 3-layer rolled-good manufacturing processes was developed and validated by single cell and short stack testing. In addition, the direct coating methods employed were shown to reduce the cost for sacrificial films. Furthermore, Gore has demonstrated a 10 µm reinforced membrane that is used in the new low-cost process and can meet automotive power density and durability targets. Across a wide range of operating conditions, the direct-coated MEA outperformed the commercial baseline MEA, and did so through a process that delivers MEAs at $92.35/m2 at a volume of 500,000 systems per year, according to Strategic Analysis (SA) estimates.

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