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NASA NTRS2003

Systems Analysis Initiated for All-Electric Aircraft Propulsion

A multidisciplinary effort is underway at the NASA Glenn Research Center to develop concepts for revolutionary, nontraditional fuel cell power and propulsion systems for aircraft applications. There is a growing interest in the use of fuel cells as a power source for electric propulsion as well as an auxiliary power unit to substantially reduce or eliminate environmentally harmful emissions. A systems analysis effort was initiated to assess potential concepts in an effort to identify those configurations with the highest payoff potential. Among the technologies under consideration are advanced proton exchange membrane (PEM) and solid oxide fuel cells, alternative fuels and fuel processing, and fuel storage. Prior to this effort, the majority of fuel cell analysis done at Glenn was done for space applications. Because of this, a new suite of models was developed. These models include the hydrogen-air PEM fuel cell; internal reforming solid oxide fuel cell; balance-of-plant components (compressor, humidifier, separator, and heat exchangers); compressed gas, cryogenic, and liquid fuel storage tanks; and gas turbine/generator models for hybrid system applications. Initial mass, volume, and performance estimates of a variety of PEM systems operating on hydrogen and reformate have been completed for a baseline general aviation aircraft. Solid oxide/turbine hybrid systems are being analyzed. In conjunction with the analysis efforts, a joint effort has been initiated with Glenn s Computer Services Division to integrate fuel cell stack and component models with the visualization environment that supports the GRUVE lab, Glenn s virtual reality facility. The objective of this work is to provide an environment to assist engineers in the integration of fuel cell propulsion systems into aircraft and provide a better understanding of the interaction between system components and the resulting effect on the overall design and performance of the aircraft. Initially, three-dimensional computer-aided design (CAD) models of representative PEM fuel cell stack and components were developed and integrated into the virtual reality environment along with an Excel-based model used to calculate fuel cell electrical performance on the basis of cell dimensions (see the figure). CAD models of a representative general aviation aircraft were also developed and added to the environment. With the use of special headgear, users will be able to virtually manipulate the fuel cell s physical characteristics and its placement within the aircraft while receiving information on the resultant fuel cell output power and performance. As the systems analysis effort progresses, we will add more component models to the GRUVE environment to help us more fully understand the effect of various system configurations on the aircraft.

Kohout, Lisa L.↗

DOE OSTI2014

Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Enhanced Performance Assessment Models for Generic Deep Geologic Repositories for High-Level Waste and Spent Nuclear Fuel - 16223

Development of an enhanced performance assessment (PA) capability for geologic disposal of spent nuclear fuel and high-level waste has been ongoing for several years in the U.S. repository program. The new Generic Disposal System Analysis (GDSA) modeling and software framework is intended to be flexible enough to evolve through the various phases of repository activities, beginning with generic PA activities in the current Concept Evaluation phase to site-specific PA modeling in the Repository Development phase. The GDSA Framework utilizes modern software and hardware capabilities by being based on open-source software architecture and being configured to run in a massively parallel, high-performance computing (HPC) environment. It consists of two main components, the open-source Dakota uncertainty sampling and analysis software and the PFLOTRAN reactive multi-phase flow and transport simulator. Reference cases or 'generic repositories' have been, and are being developed, based on typical properties for potential salt, clay, and granite host-rock formations and corresponding engineered design concepts for each medium. Past simulations have focused on a generic repository in bedded-salt host rock, while the most recent research has focused on a reference case for a typical clay/shale host rock. A variety of single-realization (i.e., deterministic) and multi-realization (probabilistic) results for the new clay reference case are presented, including an analysis of the effects of heat generation on repository performance, assuming a 100-year out-of-reactor commercial SNF waste form. Order-of-magnitude differences between predicted radionuclide concentrations in thermal versus isothermal simulations imply that mechanistic, coupled-process modeling in three-dimensional (3-D) domains can be important for building confidence in post-closure performance assessments. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Status of the Combined Third and Fourth NGNP Fuel Irradiations In the Advanced Test Reactor

The United States Department of Energy’s Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and is currently scheduled to be completed in September 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. Since the purpose of this combined experiment is to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment is significantly different from the first two experiments, though the control and monitoring systems are extremely similar. The design of the experiment will be discussed followed by its progress and status to date.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

STATUS OF TRISO FUEL IRRADIATIONS IN THE ADVANCED TEST REACTOR SUPPORTING HIGH-TEMPERATURE GAS-COOLED REACTOR DESIGNS

The United States Department of Energy’s Advanced Reactor Technologies (ART) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is irradiating up to seven low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The experiments will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of several independent capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also have on-line fission product monitoring on its effluent to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2) started irradiation in June 2010 and completed in October 2013. The third and fourth experiments have been combined into a single experiment designated (AGR-3/4), which started its irradiation in December 2011 and completed in April 2014. Since the purpose of this experiment was to provide data on fission product migration and retention in the NGNP reactor, the design of this experiment was significantly different from the first two experiments, though the control and monitoring systems are very similar. The final experiment, AGR-5/6/7, is scheduled to begin irradiation in early summer 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Advanced Gas Reactor (AGR)-5/6/7 Fuel Irradiation Experiments in the Advanced Test Reactor

The United States Department of Energy’s Very High Temperature Reactor (VHTR) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which each consist of at least five separate capsules, are being irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gases also have on-line fission product monitoring the effluent from each capsule to track performance of the fuel during irradiation. The first two experiments (designated AGR-1 and AGR-2), have been completed. The third and fourth experiments have been combined into a single experiment designated AGR-3/4, which started its irradiation in December 2011 and is currently scheduled to be completed in April 2014. The design of the fuel qualification experiment, designated AGR-5/6/7, is well underway and incorporates lessons learned from the three previous experiments. Various design issues will be discussed with particular details related to selection of thermometry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

AGR-2 Daily As-run Thermal Analyses

This report documents the daily as-run thermal analyses performed on Advanced Gas-cooled Reactor (AGR) experiment AGR-2 in the Advanced Test Reactor (ATR) for capsules 2,3,5, and 6. Capsule 1 (French) results are in Appendix F, while Capsule 4 (South African) results are in Appendix G. Appendix F, and G are separate documents Several fuel and material irradiation experiments, which support the development of the Next Generation Nuclear Plant (NGNP), are planned for the Advanced Gas Reactor Fuel Development and Qualification Program. The goals of these experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post-irradiation examination (PIE) and safety testing. AGR-2 was the second in this series of planned experiments to test tristructural-isotropic (TRISO)-coated, low enriched uranium oxycarbide (LEUCO) fuel. The AGR-2 experiment was intended to serve as a follow-on to the AGR-1 experiment. The AGR-2 experiment is a high temperature fuel particle test. The test train planned for AGR-2 is based on the experience gained from previous irradiations in the ATR, using instrumented lead experiments. Instrumented lead experiments are used for irradiations requiring a controlled environment and monitored parameters. The experiment test train positions the fuel within the test location and contains sweep gas lines and thermocouple wiring that is routed through access ports to external support systems. The AGR-2 experiment is comprised of six individual capsules, approximately 1.375-in. in diameter by 6-in. long, stacked on top of each other to form the test train. Each capsule contains 12 fueled compacts that are approximately 0.5-in. in diameter by 1-in. long. The compacts are composed of fuel particles bound together by a carbon matrix. Each compact contains approximately 4,150 fissile particles (35 vol% particle packing fraction). Each capsule will be supplied with a flowing helium/neon gas mixture to control the test temperature and to sweep any fission gases that are released to the fission product monitoring system. Temperature control is accomplished by adjusting the gas mixture ratio of the two gases (helium and neon) with differing thermal conductivities. Many similarities exist between this thermal analysis for AGR-2 with the AGR-1 daily as-run report in [1]. Gas gaps were taken from as built drawings in Reference [2]. The finite element heat transfer code ABAQUS [3] was used to perform the thermal analysis. The AGR-2 experiment was placed in the B-12 position in the ATR core as shown in Figure 1. The AGR-2 experiment capsule assembly consists of six capsules axially stacked. Each capsule contains a graphite holder with three equally spaced fuel compact holder openings as shown in Figure 2. Each holder opening accommodates four axially stacked fuel compacts. Thus, each capsule has three stacks by four fuel compacts per stack for a total of 12 fuel compacts per capsule. So, within the entire AGR-2 experiment capsule assembly, there are six capsules by 12 fuel compacts per capsule for a total of 72 fuel compacts. Six finite element heat transfer models were created (one for each capsule), each with a corresponding gas gap.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

The application of MAMMOTH for a detailed tightly coupled fuel pin simulation with a station blackout

The accurate calculation of desired quantities to predict fuel behavior in a nuclear reactor requires the solution of inter linked equations representing different physics phenomena. Traditional fuels performance codes often rely on internal empirical models for the pin power density and a simplified boundary condition on the cladding outer surface. These simplifications are performed because of the difficulty of coupling applications or codes on differing domains and mapping the required data. In this work the MAMMOTH application is employed in the demonstration of an approach that more consistent with first principles, in a simulation that couples the neutronics application Rattlesnake, the thermal fluids application RELAP-7 and the fuels performance application BISON. A single fuel pin was modeled based on the dimensions of a fuel rod from a Westinghouse 17*17 optimized PWR fuel assembly. The simulation consisted of a depletion period of 1343 days, representing three full operating cycles, followed by a station blackout (SBO) event. The fuel rod was depleted with a total pin power of 65.81 kW. After 1343 days the fission power was reduced to zero (simulating a reactor shutdown). Decay heat calculations provided the time-varying energy source after this time. For this problem, Rattlesnake, BISON, and RELAP-7 are coupled under MAMMOTH in a tightly coupled approach. Each system solves its physics on a separate mesh and, for RELAP-7 and BISON, on only a subset of the full problem domain. Rattlesnake solves the neutronics over the whole domain, which includes the fuel, cladding, gaps, water, and top and bottom rod holders. Here BISON is applied to the fuel and cladding with a 2-D axisymmetric domain, and RELAP-7 is applied to the flow of the circular outer water channel with a set of 1-D flow equations. For this work, the mesh on the Rattlesnake side is either be 3-D (for low order PN transport or diffusion) or 2-D axisymmetric (for diffusion). BISON has a matching ring structure mesh for the fuel so both the power density and local burnup are transferred accurately from Rattlesnake. At each depletion time step, Rattlesnake calculates a power density, fission density rate, burnup distribution and fast flux based on the current water density and fuel temperature. These are then mapped to the BISON mesh to perform a fuels performance solution. BISON calculates the fuel temperature and cladding surface temperature based upon the current power density and bulk fluid temperature. RELAP-7 then calculates the fluid temperature, water density fraction and water phase velocity based upon the cladding surface temperature. The fuel temperature and the fluid density are then passed back to Rattlesnake for another neutronics calculation. For this paper, sets of results from the detailed calculation are provided for both depletion and the SBO event. We demonstrate that a detailed simulation akin to first principles can be achieved using MAMMOTH and the various MOOSE application on differing domains. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

"Source Term Modeling for Advanced Gas Micro-Reactors"

Maintaining the safety of the public, environment, and operating personnel is the most important factor in designing, operating, maintaining, and decommissioning nuclear reactors. In recent years, there has been a growing interest in the development of micro-reactors employing TRi-structural ISOtropic (TRISO)-coated particle fuel. In gas reactors, TRISO fuel plays an important role in the safety case for high temperature reactors because of the fission product retention properties of the fuel. This ability enables the use of a functional containment strategy for the reactor where multiple barriers are used to prevent fission product release to the environment. Part of the safety analysis of these advanced reactors is the assessment of radionuclide releases under normal and accident conditions through the multiple credited safety barriers. Using conservative assumptions, a mechanistic analysis can be performed to quantify these releases that combines the probabilistic assessment of failure with analytic solutions to radionuclide transport equations. Source term modeling for TRISO fuel has been performed for previous reactor designs; however, these models are outdated, in many cases proprietary, and need updates to be applied to the current state of TRISO fuel technology and alternative gas reactor core configurations [1]. Currently, the only publicly available source term assessment for gas reactors is an expert-based Monte Carlo simulation based on the effectiveness of the fuel kernel, coating layers, and graphite block in a modular high temperature gas reactor [2]. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that could be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. The release is calculated by the diffusion of the key safety important fission products through the kernel, silicon carbide (SiC), graphite for both intact and defective TRISO particles based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system to remove fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. The model then can calculate the fission product release for any transient temperature profile and fission product releases can then be used to assess radiological dose to the workers and the public using conventional dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy (DOE) Advanced Gas Reactor (AGR) TRISO fuel development program. The model is coded in python with inputs and outputs in excel spreadsheets, as well as python plotting utilities to aid in the interpretation of the results. References: [1] INL, NGNP Mechanistic Source Term White Paper, INL-10-17997, July 2010. [2] David A. Petti, Richard R. Hobbins, Peter Lowry, Hans Gougar, “Representative Source Terms and The Influence of Reactor Attributes on Functional Containment in Modular High Temperature Gas-cooled Reactors,” Nuclear Technology, Vol. 184, p. 181-197, Nov. 2013.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2022

Assessment of BISON capabilities for component-level prediction of tritium transport in fusion and fission applications

Here we assessed the hydrogen migration and redistribution model implemented in the BISON fuel performance code for component-level tritium transport applications in fusion and fission example scenarios. We developed BISON models of an ITER heat exchanger, light water reactor fuel cladding, and a fluoride salt cooled high temperature reactor heat exchanger. Reasonable agreement between reported values in reference studies and BISON predictions demonstrated the ability of the BISON models to predict tritium transport behavior through steel materials. Next, sensitivity and uncertainty analysis was used to understand key parameter sensitivities within the BISON model for evaluation of tritium migration and redistribution. . The sensitivity analysis showed that the diffusivity activation energy was the most important parameter and that the heat of transport was the least important parameter for tritium migration in steels. Overall, the capabilities of the BISON code for component level modeling of tritium transport are promising.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

DOE OSTI2019

A ROBUST MECHANISTIC APPROACH TO PREDICTION OF DEPARTURE FROM NUCLEATE BOILING

departure from nucleate boiling (DNB) sets a regulatory limit for licensing pressurized water reactors (PWRs). Under such conditions, the heated surface is permanently blanketed by a vapor film, leading to a sharp deterioration of the heat transfer coefficient at the heater/coolant interface and an abrupt temperature rise. Unfortunately, the path for an accurate, robust prediction of DNB has been elusive due to lack of consensus on its triggering mechanism. This work reviews existing physics-driven modeling tools in the first place. An evolutionary channel-scale mechanistic model that leverages key assumptions in the relatively well-accepted mechanisms of liquid sublayer dryout and near-wall bubble crowding is then proposed. Detailed validation of the proposed model has demonstrated its improved predictive capabilities over previous data/physics-driven models with regard to an extensive DNB-specific CHF test matrix covering a wide range of flow conditions. The unique feature of the proposed model lies in its ability to predict DNB for different heater geometries (including round tube, annulus, and rod bundle), which is essential in deciphering fuel performance metrics from different facilities and reactor types. The proposed model will be further implemented in fuel performance codes to help improve modeling of transient DNB scenarios such as during a reactivity-initiated accident (RIA).

Zhao, Xingang↗

DOE OSTI2013

IMPLEMENTING PRIMARY CREEP CALCULATIONS DURING STRESS CHANGES AND REVERSALS IN THE FUEL PERFORMANCE CODE FRAPCON-3

Recently, the creep models in FRAPCON-3.4 were used to model the creep tests performed in IFA-699 up through cycle 4. IFA-699 is an instrumented test rig in the Halden Boiling Water Reactor that is being used to perform in-reactor creep experiments under changing stress conditions. The cladding creep tests in IFA-699 were on four fresh fuel segments with Zircaloy-4, E110, M-MDA, and M5™ cladding. These fuel segments were loaded in a test rig and placed in a PWR loop and operated under PWR conditions. Each of the test fuel rods were connected to gas lines in order to apply tensile and compressive stresses to the cladding. Diameters of the cladding were measured by diameter gages at several intervals to measure the creep behavior of the cladding as a function of time. The creep tests were performed at a temperature of around 360°C and fast neutron flux of 1.8 to 2.4×1017 n/m²-s (depends on specimen location). During the four cycles of operation, hoop stresses of -75, -50, +30, and +110 MPa were applied to the specimens. In addition to providing assessment of the FRAPCON-3.4 creep modeling, the creep model comparisons were also used to provide insight into the nature of primary creep following stress changes, whether creep behavior is symmetric under tensile and compressive loading, and the effect of fast fluence on creep rate. It was determined that creep strain should be calculated as a function of effective stress rather than hoop stress. This accounts for the observed difference between creep under compressive and tensile stress for the same hoop stress. It was also determined that the primary creep changes in proportion to the change in stress following stress change or stress reversal. Finally, it was observed that the irradiation creep rate decreases somewhat with increasing fluence. The creep model in FRAPCON-3.4 has already been developed to be a function of effective stress and includes a term that decreases the irradiation creep rate with increasing fluence. A new mathematical formulation has been developed that is suitable for a fuel performance code to calculate the primary creep strain under conditions of changing stress and stress reversal. The creep model in FRAPCON-3.4 and the formulation for calculating primary creep under changing stress conditions will be described in this paper. Comparisons to data will also be provided.

Geelhood, Kenneth J.↗

DOE OSTI2019

Demonstration of the Coating Analysis Capabilities in BISON

In order to improve the accident tolerance of light water reactor (LWR) fuel, coated cladding materials have been proposed which exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high-temperature steam, making accident mitigation more likely. In order to determine how these coating materials will adjust the behavior of the Zr-alloy cladding during normal reactor operation, simulations have been performed using the Bison fuel performance code. This analysis focused on modeling and comparing the integral thermo-mechanical performance of coated and uncoated zircaloy cladding with uranium dioxide fuel under normal operating conditions. This analysis demonstrates the ability to model a discrete meshed coating bonded to the cladding exterior and provides a comparison with an uncoated fuel rod. A generic range of coating material properties are implemented given the differences and unknowns regarding the proprietary coating properties used by the vendors. A main-effects analysis was performed to identify trends and sensitivities in how coating material properties affect key fuel performance metrics. Lastly, because residual stresses are expected to form in the coating, arising from the high temperature coating application process and application techniques, an additional analysis was performed to identify the impact of these stresses on the cladding behavior. Future analysis efforts will target more comprehensive documentation of this main effects analysis by sampling the results at the beginning and middle-of-life in order to capture important behavior before and during the onset of gap closure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Bayesian discovery of optimal reduced order models from mechanistic and experimental data: A case study of Pd penetration in TRISO fuels using BISON

TRistructural ISOtropic (TRISO) particles rely on a silicon carbide (SiC) layer as the primary structural material and barrier to metallic fission products (FPs) release. Accurate prediction of palladium (Pd) transport and penetration is therefore critical for qualifying TRISO fuels for advanced reactors. The empirical correlation for Pd penetration in BISON is derived from historical particle-fuel data, but cannot explain the large scatter in the experimental data that arises from varying experimental conditions. To aid fuel qualification, we previously developed a mechanistic reduced order model (ROM) using BISON that resolves these dependencies. Here, in this work we build on that mechanistic ROM and perform validation and quantify its uncertainty using Bayesian uncertainty quantification (UQ). calibration against a suite of in-pile and out-of-pile experiments spanning particle compositions, geometries, and operating conditions, and we benchmark it against the empirical correlation. Bayesian UQ identifies influential parameters, calibrates them to data, and yields predictive intervals. Results show that while the empirical correlation can be tuned to fit a single experiment type, it transfers poorly; the mechanistic ROM sustains accuracy with credible uncertainty across disparate conditions. This demonstrates a practical path—via Bayesian UQ applied to mechanistic ROMs—to leverage single-effect experiments for inferring in-reactor behavior and supporting TRISO fuel qualification.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

A Phenomenological Thermal Spray Wall Interaction Modeling Framework Applied to a High Temperature Ignition Assistant Device

Airborne compression ignition engines must operate with reliable ignition systems to achieve proper ignition at every cycle, particularly at high altitudes. Glow-plug-based ignition-assistant (IA) devices can provide the necessary energy to preheat the fuel and ensure ignitability of the fuel-air mixture. Ignitability of liquid sprays can be facilitated via direct impingement onto the hot IA surface, however this comes with adverse effects on the IA durability. Therefore, optimizing an IA’s design requires detailed understanding of the physics of fuel spray impingement of superheated surfaces. While spray impingement on relatively low wall temperatures has been extensively studied and appropriate numerical models have been proposed through the years, fundamental understanding of high-speed liquid spray impingement on superheated walls is still elusive. This work aims to formulate a phenomenological thermal spray-wall interaction framework for modeling the film-boiling-induced heat transfer, atomization, and dispersion of fuel spray droplets impinging onto a superheated IA device. A qualitative comparison of the new phenomenological model is performed against optical experiments from the literature of an F-24 fuel spray injected onto an IA device located 12 mm away from the injector tip. The temperature of the IA was set at 1400 K. The fuel injection pressure was 400 bar, while the ambient gas pressure and temperature were 30 bar and 800 K, respectively. The performance of the phenomenological model is evaluated in comparison with two other state-of-art models from the literature. A qualitative analysis of the different spray and fuel-air mixture characteristics is performed to outline the differences in the predictions offered by the new phenomenological model and the two state-of-art spray-wall interaction models.

droplet dispersion↗

NASA NTRS1991

Cold flow testing of the Space Shuttle Main Engine high pressure fuel turbine model

In order to experimentally determine the performance of the Space Shuttle Main Engine (SSME) High Pressure Fuel Turbopump (HPFTP) turbine, a 'cold' air flow turbine test program was established at NASA's Marshall Space Flight Center. As part of this test program, a baseline test of Rocketdyne's HPFTP turbine has been completed. The turbine performance and turbine diagnostics such as airfoil surface static pressure distributions, static pressure drops through the turbine, and exit swirl angles were investigated at the turbine design point, over its operating range, and at extreme off-design points. The data was compared to pretest predictions with good results. The test data has been used to improve meanline prediction codes and is now being used to validate various three-dimensional codes. The data will also be scaled to engine conditions and used to improve the SSME steady-state performance model.

Hudson, Susan T.↗

DOE OSTI2019

Fuel Matrix Degradation Model Development Update: Alloy Corrosion Rates and Hydrogen Generation

This work is being performed as part of the DOE NE Spent Fuel and Waste Science and Technology Campaign, Argillite and Crystalline Rock R&D work packages: SF-19AN01030101 and SF-19AN01030201. This document meets the milestone M4SF-19AN010301013 for Argillite R&D and the milestone M4SF-19AN010302013 for Crystalline R&D. The fuel matrix degradation (FMD) model calculates the degradation rate of spent UO2 fuel based on fundamental electrochemical theory and thermodynamics. The model provides the quantitative basis for radionuclide source term predictions in the generic disposal system analysis (GDSA) performance assessment (PA) model. The FMD model has been implemented in a manner that facilitates its integration with the GDSA-PA code and a preliminary, integrated FMD-GDSA model has been successfully tested (Jerden et al., 2017). The specific focus of on-going work is to accurately and quantitatively represent the generation of H2 in a breached waste package and its effect on the spent fuel degradation rate so that this key process can be accurately represented in PA models. It has been shown experimentally that millimolar concentrations of dissolved H2 in contact with spent fuel will inhibit oxidative dissolution and decrease the fuel degradation rates by as much as three orders of magnitude (e.g., Röllin et al., 2001, Ollila, 2008). This will lead to a significant decrease in the radionuclide source term values used in repository PA models. Sensitivity studies using the FMD model have shown that the dissolved H2 concentration is the dominant environmental variable affecting the UO2 spent fuel dissolution rate as long as a source of H2 persists in the breached waste package (Jerden et al., 2015). Therefore, on-going experimental and modeling efforts focus on H2 production, accumulation and interfacial reactions within a breached waste package. The anoxic corrosion of metallic engineering materials within a breached waste package will be the main source of H2 in crystalline and argillite rock repository systems. The most important H2-producing alloys present within a waste package include stainless steel, carbon steel, and aluminum alloys and the Zircaloy fuel cladding. A steel corrosion module was added to the FMD model to account for these H2 sources and couple the H2 generation rate with fuel degradation processes (Jerden et al., 2017). As discussed below, there is a need for experimental data from electrochemical corrosion tests to quantify the H2 generation rates due to the corrosion of these waste package alloys as a function of key environmental variables (Eh, pH, groundwater chemistry). This report presents results from series of electrochemical corrosion tests with materials present in waste packages that provide information needed to parameterize and validate the FMD model. The alloys tested include 316L stainless steel (316SS), AISI 4320 carbon steel (C-steel) and Zircaloy-4 fuel cladding (unirradiated). Potentiodynamic polarization and potentiostatic polarization tests were performed in buffer solutions at pH 4, 7 and 10 with NaCl concentrations of 0, 4 and 100 millimolar (mM). The potentiostatic tests were performed at voltages varying from 0.5 volts down to -0.66 volts, vs the standard hydrogen electrode (SHE), and all tests were performed at laboratory ambient temperatures (~22 °C). The corrosion rates measured during the electrochemical tests were used as direct input to a new FMD/in-package chemistry model that is being developed to provide accurate long-term spent fuel degradation rates as chemical conditions (Eh, pH, speciation) evolve within a breached waste package. This new model was developed by coupling the FMD model with the reactive transport code X1t, which is a module within the Geochemist’s Workbench (GWB) reaction path modeling code (Jerden et al., 2019). The reactive transport model consists of a 1D domain discretized with 21 reaction diffusion cells and includes a single waste packag

Jerden, James↗

DOE OSTI2020

BISON TRISO Modeling Advancements and Validation to AGR-1 Data

BISON is a finite element-based nuclear fuel performance code. Among its unique characteristics are its ability to model 1D, 2D, and 3D geometries and its applicability to a wide variety of nuclear fuels. For the last eight years, BISON has included a beginning capability to model tri-structural isotropic (TRISO) fuel. Recently, interest in TRISO fuel has grown, and a significant effort has been made to improve BISON’s capabilities in this area. Capability development has occurred for each material present in TRISO fuel particles: the buffer, inner pyrolytic carbon, silicon carbide, and outer pyrolytic carbon layers, as well as the fuel kernel. New elastic, creep, swelling, thermal expansion, thermal conductivity, and fission gas release (FGR) models are available. New models for the graphite matrix are also now available. Another important addition is the ability to perform statistical failure analysis of large samples of fuel particles. This new capability, which continues to grow, enables evaluation of failure due to pressure or crack formation by analyzing many thousands of particles. This enables realistic calculations of fission product release from the many particles in a TRISO-fueled reactor. These capabilities were checked via regression and verification tests. A large number of code benchmarking problems were also run, showing that BISON’s results closely match those of other software tools. Finally, a significant validation effort was completed in which fission product release, measured as part of the AGR-1 capsule experiments, was compared to BISON outputs. BISON outputs compared very well to the experimental data and to PARFUME results. Interest in BISON’s TRISO capabilities is growing, with the U.S. Nuclear Regulatory Commission (NRC) and Westinghouse Electric Company receiving training during the past year. Multiple other entities have expressed interest in or are actively using BISON. Kairos Power, LLC, has a strong partnership with Idaho National Laboratory (INL) regarding the use of BISON for TRISO analysis. While its capabilities still continue to grow, BISON has already become a powerful tool for TRISO analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗