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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Multiphysics Demonstration of Temperature-Driven Assembly Bowing in SFRs using MOOSE-Based Codes

Core bowing is an important passive safety mechanism in liquid metal cooled fast reactors. When the core restraint system is properly designed, temperature and flux gradients influence assemblies in the core to bow into less reactive configurations during accident scenarios, resulting in negative reactivity feedback. Prediction of core bowing involves complex interplay of radiation transport, impacts of fluid flow and heat transfer on duct temperature, and mechanical responses to the induced temperature and flux gradients. Under the U.S. Department of Energy Office of Nuclear Energy’s Advanced Modeling and Simulation (NEAMS) Program [1], an integrated multiphysics approach is being developed to model the core bowing phenomena in liquid metal-cooled fast reactors with the Multiphysics Object Oriented Simulation Environment (MOOSE) [2]. In this methodology, the MOOSE-based reactor physics code Griffin [3] will solve the neutron transport equation and determine the power distribution. With the detailed power distribution from Griffin, the subchannel analysis codes MOOSE-Subchannel [4] and Pronghorn [5] are utilized to calculate the assembly temperature distribution. MOOSE’s Solid Mechanics [6] and Contact [7] Modules are leveraged to calculate the thermal expansion and duct bowing displacement with the duct wall temperature from thermal hydraulics calculation. In this work, an initial one-way coupling demonstration of the integrated multiphysics approach has been performed on a seven-assembly problem based on the sodium-cooled fast reactor ABR-1000 design [8]. The neutronics calculation with Griffin is not yet involved in the current simulation. MOOSE-Subchannel and Pronghorn evaluate fluid and solid temperature based on a fixed power distribution. In addition, one-way coupling is utilized in this coupled calculation, via Pronghorn passing the duct temperature data to the MOOSE Solid Mechanics calculation. An assessment of the Solid Mechanics module was performed in parallel to verify duct bowing behavior with duct-to-duct contact phenomenon [9]. The displacement from MOOSE Solid Mechanics is not yet transferred back and utilized in the Pronghorn and MOOSE-Subchannel calculation. This model will be available on the National Reactor Innovation Center (NRIC) Virtual Test Bed (VTB) repository [10]. Future stages of this work will involve solving problems of increasing complexity as well as adding more physics (e.g. reactor physics) to the integrated workflow to reach the end goal of modeling the core bowing phenomenon with an integrated multiphysics workflow.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS2004

Physics in NASA Exploration

The primary focus of the workshop was NASA's new concentration on sending crewed missions to the Moon by 2020, and then on to Mars and beyond. Several speakers, including JPL s Fred O Callaghan and NASA's Mark Lee, broached the problem that there is now a serious reduction of capability to perform experiments in the ISS, or to fly significant mass in microgravity by other means. By 2010, the shuttle fleet will be discontinued and Russian craft will provide the only access to the ISS. O Callaghan stated that the Fundamental Physics budget is being reduced by 70%. LTMPF and LCAP are slated for termination. However, ground-based experiments are continuing to be funded at present, and it will be possible to compete for $80-90 million in new money from the Human Research Initiative (HRI). The new program thrust is for exploration, not fundamental physics. Fundamental, we were told by Lee, does not ring well in Washington these days. Investigators were advised to consider how their work can benefit missions to the Moon and Mars. Work such as that regarding atomic clocks is looked upon with favor, for example, because it is considered important to navigation and planetary GPS. Mark Lee stressed that physicists must convey to NASA senior management that they are able and willing to contribute to the new exploration research programs. The new mentality must be we deliver products, not do research. This program needs to be able to say that it is doing at least 50% exploration-related research. JPL s Ulf Israelsson discussed the implications to OBPR, which will deliver methods and technology to assure human health and performance in extraterrestrial settings. The enterprise will provide advanced life-support systems and technology that are reliable, capable, simpler, less massive, smaller, and energy-efficient, and it may offer other necessary expertise in areas such as low-gravity behavior. Like Dr. Lee, he stated that the focus must be on products, not research. While there is not yet a formal direction, he said, LTMPF and PARCS ISS flight projects are slated to terminate in October 2004. All flight investigations are being returned to ground programs and phased out by the end of FY07. Physics ground programs are intact for now, but to survive we must shift about 50% of research to supporting exploration. Basic research programs in other disciplines are being cancelled. Product lines will support human health, safety and life-support, including countermeasures against radiation and other hazards, as well as advances in time-keeping, navigation and communications technologies. Israelsson said that the new Fundamental Physics for Exploration Roadmap points to how fundamental physics research can and does support exploration. JPL will use the roadmap to argue for support for fundamental physics research under several codes. Nicholas Bigelow of the University of Rochester encouraged attendees not to become discouraged, but rather to embrace the opportunities presented by NASA's new direction.

O'Callaghan, Fred↗
DOE OSTI2023

Autonomous Inversion of In Situ Deformation Measurement Data for Injection-Induced Stress Change

Geologic carbon storage (GCS) is likely to play a key part of the global effort to dramatically reduce CO2 emissions and perhaps even reduce atmospheric CO2 concentrations through carbon negative operations. A critical part of effort to commercialize and widely deploy this technology is developing the capability to rapidly assimilate real-time monitoring data into a form that will enable site operators to make decisions to manage the safe and efficient operations. Two of the risks associate with GCS are the risk of inducing fractures in the sealing formations that can create leakage pathways and the risk of inducing earthquakes of sufficient magnitude to cause public concern, property damage, or safety risks. To properly manage these risks the site operator needs to know the initial state of stress, the change in stress induced by injection, and the relationship between operational parameters such as injection rate and pressure and the change in stress. Current methods of estimating the change in stress require choosing the type of constitutive model and the model parameters based on core, log, and geophysical data during the characterization phase, with little feedback from operational observations to validate or refine these choices. These characterization methods interrogate the geologic formations using length scales, loading rates or magnitudes that are quite different from those encountered by the actual storage system. It is shown that errors in the assumed constitutive response, even when informed by laboratory tests on core samples, are likely to be common, large, and underestimate the magnitude of stress change caused by injection. Recent advances in borehole-based strain instruments and borehole and surface-based tilt and displacement instruments have now enabled monitoring of the deformation of the storage system throughout its operational lifespan. This data can enable validation and refinement of the knowledge of the geomechanical properties and state of the system, but brings with it a challenge to transform the raw data into actionable knowledge. We demonstrate a method that uses automatic differentiation and a finite-element based geomechanical model perform a gradient-based deterministic inversion of geomechanical monitoring data. This approach allows autonomous integration of the instrument data without the need for time consuming manual interpretation and selection of updated model parameters. Furthermore, only isotropic linear elasticity is considered in this paper, the approach presented is very flexible as to what type of geomechanical constitutive response can be used. The approach is easily adaptable to nonlinear physics-based constitutive models to account for common rock behaviors such as creep and plasticity. The approach also enables training of machine learning-based constitutive models by allowing back propagation of errors through the finite element calculations. This enables strongly enforcing known physics, such as conservation of momentum and continuity, while allowing data-driven models to learn the truly unknown physics such as the constitutive or petrophysical responses.

Burghardt, Jeffrey A.↗
DOE OSTI2015

Suitability of Synthetic Driving Profiles from Traffic Micro-Simulation for Real-World Energy Analysis: Preprint

A shift towards increased levels of driving automation is generally expected to result in improved safety and traffic congestion outcomes. However, little empirical data exists to estimate the impact that automated driving could have on energy consumption and greenhouse gas emissions. In the absence of empirical data on differences between drive cycles from present day vehicles (primarily operated by humans) and future vehicles (partially or fully operated by computers) one approach is to model both situations over identical traffic conditions. Such an exercise requires traffic micro-simulation to not only accurately model vehicle operation under high levels of automation, but also (and potentially more challenging) vehicle operation under present day human drivers. This work seeks to quantify the ability of a commercial traffic micro-simulation program to accurately model real-world drive cycles in vehicles operated primarily by humans in terms of driving speed, acceleration, and simulated fuel economy. Synthetic profiles from models of freeway and arterial facilities near Atlanta, Georgia, are compared to empirical data collected from real-world drivers on the same facilities. Empirical and synthetic drive cycles are then simulated in a powertrain efficiency model to enable comparison on the basis of fuel economy. Synthetic profiles from traffic micro-simulation were found to exhibit low levels of transient behavior relative to the empirical data. Even with these differences, the synthetic and empirical data in this study agree well in terms of driving speed and simulated fuel economy. The differences in transient behavior between simulated and empirical data suggest that larger stochastic contributions in traffic micro-simulation (relative to those present in the traffic micro-simulation tool used in this study) are required to fully capture the arbitrary elements of human driving. Interestingly, the lack of stochastic contributions from models of human drivers in this study did not result in a significant discrepancy between fuel economy simulations based on synthetic and empirical data; a finding with implications on the potential energy efficiency gains of automated vehicle technology.

33 ADVANCED PROPULSION SYSTEMS↗
NASA NTRS1999

NASGRO 3.0: A Software for Analyzing Aging Aircraft

Structural integrity analysis of aging aircraft is a critical necessity in view of the increasing numbers of such aircraft in general aviation, the airlines and the military. Efforts are in progress by NASA, the FAA and the DoD to focus attention on aging aircraft safety. The present paper describes the NASGRO software which is well-suited for effectively analyzing the behavior of defects that may be found in aging aircraft. The newly revised Version 3.0 has many features specifically implemented to suit the needs of the aircraft community. The fatigue crack growth computer program NASA/FLAGRO 2.0 was originally developed to analyze space hardware such as the Space Shuttle, the International Space Station and the associated payloads. Due to popular demand, the software was enhanced to suit the needs of the aircraft industry. Major improvements in Version 3.0 are the incorporation of the ability to read aircraft spectra of unlimited size, generation of common aircraft fatigue load blocks, and the incorporation of crack-growth models which include load-interaction effects such as retardation due to overloads and acceleration due to underloads. Five new crack-growth models, viz., generalized Willenborg, modified generalized Willenborg, constant closure model, Walker-Chang model and the deKoning-Newman strip-yield model, have been implemented. To facilitate easier input of geometry, material properties and load spectra, a Windows-style graphical user interface has been developed. Features to quickly change the input and rerun the problem as well as examine the output are incorporated. NASGRO has been organized into three modules, the crack-growth module being the primary one. The other two modules are the boundary element module and the material properties module. The boundary-element module provides the ability to model and analyze complex two-dimensional problems to obtain stresses and stress-intensity factors. The material properties module allows users to store and curve-fit fatigue-crack growth data. On-line help and documentation are provided for each of the modules. In addition to the popular PC windows version, a unix-based X-windows version of NASGRO is also available. A portable C++ class library called WxWindows was used to facilitate cross-platform availability of the software.

Mettu, S. R.↗
DOE OSTI2019

Comparison of experimental and simulated critical heat flux tests with various cladding alloys: Sensitivity of iron-chromium-aluminum (FeCrAl) to heat transfer coefficients and material properties

In this paper we analyze the differences between previous transient critical heat flux (CHF) experiments using iron-chromium-aluminum (FeCrAl), Inconel 600, and stainless steel 316 (SS316) alloy test sections and best-estimate modeling results from widely-used nuclear engineering systems and subchannel analysis tools. FeCrAl is an Accident Tolerant Fuel (ATF) candidate cladding material. The thermal hydraulic performance and safety characteristics of FeCrAl are being evaluated to determine viability as a cladding material in Light Water Reactors (LWRs). In this study, the results of the CHF experiments conducted at atmospheric pressure and fixed inlet coolant temperature and mass flux are compared to models built in the fifth version of the Reactor Excursion and Leak Analysis Program (RELAP5-3D) and CTF, the modernized version of COBRA-TF developed by the Consortium for Advanced Simulation of LWRs (CASL). Results from RELAP5-3D and CTF showed differences from the experiments and from each other in predicting CHF. In the Inconel 600 case, both computational tools overpredicted CHF, which led to an underprediction in the tube outer surface temperature. In the SS316 and FeCrAl cases, CHF was underpredicted by the codes, leading to an overprediction of the tube outer surface temperature. To understand the discrepancies in CHF and post-CHF predictions, studies were performed using RELAP5-3D and RAVEN to determine the sensitivity of CHF and peak test section temperature, an analog to peak cladding temperature (PCT), to heat transfer coefficients, a CHF multiplier, and uncertainties in the thermal conductivity and volumetric heat capacity. We found that CHF depends most strongly on the CHF multiplier and thermophysical properties. A combination of these factors that produced the best match to the experiment based on CHF, PCT, and the total energy deposited into the tube was determined. The best match parameters were able to provide best-estimate predictions of the CHF and integral heat flux, but were still conservative when predicting the PCT. The best match set of parameters developed in this paper are intended only as a demonstration of an approach that could be applied in the future with a larger set of experiments to produce more accurate models of CHF and post-CHF behavior.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Observation of Downward Facing Pool Boiling on a Hemispherical Vessel under External Reactor Vessel Cooling

It is a great challenge to ensure the safety of nuclear power plant during a severe accident. One of the major concerns is to overheat the reactor core which potentially may create a nuclear crisis if the core cooling fails. The concept of In-vessel retention (IVR) by passive external reactor vessel cooling (ERVC) in a flooded cavity of pressurized water reactor (PWR) is a viable and effective approach to retain radioactive core melt within the reactor pressure vessel (RPV). As long as the heat flux at the lower head of the RPV is lower than the critical heat flux (CHF) limit at the corresponding point, nucleate boiling will be the prevailing regime. In that case, the temperature of outer surface can be maintained near the saturated temperature of water such that the reactor vessel is able to be sufficiently cooled to sustain its thermal and structural integrity. Owing to the practical significance of this IVR-ERVC concept, there were many experimental and theoretical studies to investigate the two-phase flow natural circulation phenomena including the effect of inclination angle on the CHF, based on which many different empirical or semiempirical correlations were developed for CHF by fitting the experimental results. For the engineering applications, reactor-scale facilities such as ULPU series, KAIST facility, CYBL and FIRM were built to experimentally verify and validate the feasibility of ERVC for different design of reactors. Besides, for the research on physical mechanism of downward facing pool boiling, laboratory-scale experiments such as DELTA, CASA and SBLB were performed to visually and theoretically simulate the process of IVR-ERVC. All of the aforementioned test facilities and experiments provided a simulation of the downward facing boiling process and experimental study on the CHF limits during the ERVC conditions of different types of reactors. There were also some visualized investigations on the two-phase flow behavior in both quenching and steady state experiments. In this study, a scale-down three-dimensional downward facing pool boiling facility (iSBLB) was built to investigate the two phase natural circulation flow characteristics during the ERVC process. The general morphologies of vapor mass at different locations were visually recorded using the high-speed camera. The observation results showed that vapor behavior along the hemispherical curved surface was periodical. The boiling phenomenon was quite different from those on all other surface orientations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

SAS4A/SASSYS-1 VERSION 5.2 SAFETY ANALYSIS CODE SYSTEM

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolant flow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.«

Fanning, Thomas↗
DOE OSTI2026

Data for Filling the Cellulosic Bio-economy Gap by Utilizing a Wedge Approach Combined with Stakeholder Collaboration

The price gap between the market and breakeven prices of cellulosic biomass for farmers represents a significant barrier to the development of a low-carbon cellulosic bioeconomy. Using a bottom-up, agent-based modeling tool that replicates the behaviors and interactions of key stakeholders, this study analyzes the emergence of a cellulosic bioeconomy at the local scale through a wedge approach that examines an integrated portfolio of multiple policy options, including subsidies for small-scale bioproducts and environmental credits. The role of collaboration among multiple stakeholders, such as biomass producers (farmers), bio-refinery industry, government, and society, is assessed for filling the price gap. Using the Sangamon River Basin as a case study site, we evaluate the effectiveness of the wedge approach by comparing simulation results from multiple scenarios, each incorporating different combinations of bioeconomy wedges, with and without stakeholder collaboration. Results underscore that active collaboration among stakeholders acts as a catalyst enlarging the effectiveness of bioeconomy wedges. Including the carbon credits and environmental value in the policy portfolio is found to bridge the price gap through collective contributions from diverse stakeholders, where the cellulosic biofuel and bioproduct industry plays a pivotal role. Although this study is conducted at the local watershed scale, the methodology and findings offer valuable insights for market development in other watersheds and the potential scaling of local markets to regional and national levels.

Economics↗
DOE OSTI2023

Energy Optimization of Light and Heavy-Duty Vehicle Cohorts of Mixed Connectivity, Automation and Propulsion System Capabilities via Meshed V2V-V2I and Expanded Data Sharing (Final Scientific and Technical Report)

Vehicle connectivity and automated driving technologies individually have the potential to decrease energy consumption and/or increase safety on light, medium or heavy duty vehicles to varying degrees depending on the traffic infrastructure and specific driving scenarios. Due to advances in sensing, perception and computing power, research and development emphasis in the mobility sector has shifted away from connectivity. Prior research has shown that driving automation with the absence of connectivity can in certain circumstances increase energy consumption. The effectiveness of synergizing connectivity and driving automation technologies is the focus of this work, specifically applied to vehicle cohorts of mixed composition, light and heavy duty, and powertrains ranging from all electric to conventional internal combustion engine. The project team is led by Michigan Technological University (MTU) and partnered with AVL Mobility Technologies Inc. (AVL), Borg Warner (BW), Traffic Technology Services (TTS), American Center for Mobility (ACM) and Navistar (NAV). The main thrusts for the team are to develop a micro-traffic simulation environment with specific VD&PT system attributes and CAV capabilities, 2) field a vehicle test fleet of mixed classification, propulsion and CAV capacity, 3) develop artificial intelligence (AI) and machine learning (ML) based multi-agent optimization methods for various traffic infrastructures, 4) integrate the virtual environment and the optimization methods then deploy the system as a CAV hardware in the loop (HiL) for the vehicle test fleet and 5) conduct closed track and public road testing to validate simulation and demonstrated energy and mobility improvements at multiple scales. For a cohort of mixed vehicles, the team will demonstrate a reduction of energy consumption of 10-50% at intersection, arterial roadway and limited access highway scenarios through connectivity and automation in simulation and at a closed test track. The energy reduction objectives of the project are summarized in Table 1, indicating the infrastructure and over what distances are relevant considered. Single scenario energy reductions are not relevant and thus, the research team took the approach to vary parameters associated with the infrastructure, vehicle cohort composition and dynamic behavior to generate energy consumption distributions for both unconnected and connected scenarios.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2024

Challenging Common Assumptions of Thick-Wall Chamber Dynamics in Inertial Fusion Systems Using MOOSE-based Multiphysics Simulations

As an increasing number of companies look toward commercial Inertial Fusion Energy (IFE) designs, there is a pressing need to understand the physics of thick-wall chamber gas dynamics. The thick liquid wall approach implements a renewable wall to mitigate the fusion target emissions, thereby reducing the radiation damage rate and significantly extending the lifetime of chamber structures, leading to increased plant availability and reduced waste streams in comparison to dry wall chamber designs. It is necessary, however, to assess the critical performance and safety aspects of these systems. For example, it is crucial to predict (1) where the mass ablated from the liquid walls will vent, which determines the placement of condensing surfaces; (2) debris propagation up the beam lines, which provides essential information for design and protection requirements; (3) peak pressures and impulse on chamber walls, which affect chamber structural design; and (4) momentum transfer to the liquid jets, which constrains the shape and positioning of the jets. In turn, the chamber design and its liquid walls affect shielding requirements, material activation, and tritium fuel cycle. Currently available simulation tools, however, are unable to accurately capture key thick-wall chamber dynamics. Significant assumptions are often made to simplify the system and reduce computational cost and modeling capability needs, but the impact of these assumptions on simulation predictions has not been evaluated. For example, no three-dimensional simulations can be found in the open literature to evaluate gas venting and momentum transfer to the jets with simulations using two-dimensional domains to represent complex three-dimensional geometries. Moreover, limited studies have been dedicated to jet breakup due to both turbulence and neutron heating, and no studies have been found that evaluate how jet breakup can impact shock-jet interaction. Furthermore, effects of radiative heat transfer have rarely been included for the hydrodynamic phase of shock propagation, and integration of proper equations of state in shock dynamics codes has been mostly exploratory. In this study, we use the flexible, high-fidelity Multiphysics Object-Oriented Simulation Environment (MOOSE) to model these complex phenomena and inform design and safety studies. Capabilities to model thick-wall chamber gas dynamics are being developed, and the impact of the assumptions listed above (i.e., two-dimensional vs three-dimensional, absence of jet breakout, no radiative heat transfer, and ideal gas behavior) are being quantified.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

RELAP-7 Application and Enhancement for FLEX Strategies and ATF Behavior under Extended Loss of AC Power Conditions

This report summarizes the results of a three-year research project sponsored by the U.S. Department of Energy (DOE) Nuclear Energy University Program (NEUP) to enhance and apply the RELAP-7 code by adding and improving several important components (e.g., a mechanistic Reactor Core Isolation Cooling (RCIC) system model) for thermal hydraulic studies of LWRs under ELAP conditions and evaluating the time available for transition to portable FLEX equipment. The project team included University of Massachusetts–Lowell, The Ohio State University, Texas A&M University, Idaho National Laboratory and Oak Ridge National Laboratory. In the Fukushima accident, it was found that the RCIC system played a crucial role in delaying core meltdown by almost three days in Fukushima Daiichi Unit 2, because of self-regulated operation of the steam driven RCIC turbine-pump injection system. Steam flow in the convergent-divergent nozzles of the RCIC Terry turbine is two-phase non-equilibrium transonic flow with homogenous nucleation condensation. To more accurately predict the dynamic process and behavior of the transonic compressible steam flow, a one-dimensional transient two-phase analytical model is presented. A simplified four-fluid model was employed in the present work with the consideration of four separate fluid fields: vapor, liquid film, entrained droplets and condensed droplets. The mass, momentum and energy interactions between the fluids were considered and modeled. An extended seven-equation non-equilibrium critical flow model was developed to obtain the critical pressure and velocities of each phase at the nozzle throat. To predict the wetness in the divergent section, a mechanistic nucleation condensation model was integrated in the nozzle analysis model, considering the generation and consequent growth of droplets. The governing differential equations on a staggered grid were discretized using the second-order Lax-Wendroff scheme with a flux limiter, and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to solve the discrete linear system. To demonstrate the predictability and reliability of the physical models and the numerical method proposed in the present work, three representative nozzles were modeled and simulated. The results show good agreement with the available experimental data, even for condensation shock. Then, the 1D nozzle model was employed to obtain nozzle flow tables of the Terry turbine nozzle for different working pressures which can cover the operation pressure range of the RCIC system. A mechanistic RCIC turbine-pump system model was developed and implemented in the system code TRACE to simulate dynamic responses of the RCIC system under Beyond Design Basis Accident (BDBA) conditions. The turbine-pump governing equations are based on the control volume approach of the angular momentum balance. The physics based mechanistic RCIC model was developed using the TRACE control system components (i.e., signal variables, control blocks, and tables), and incorporated into a TRACE boiling water reactor (BWR) model. The TRACE model in this report has a detailed nodalization of the reactor pressure vessel (RPV), and all of the major flow paths and system components, including the safety relief valves (SRVs) and the containment suppression pool and drywell. Based on the nozzle flow tables generated from the 1D nozzle model developed, the turbine drive torque can be calculated from table lookup. Since the detailed specifications of the RCIC pump are unavailable, the homologous curves for a Bingham pump were used in the current pump component. A station black-out (SBO) accident test problem was selected to demonstrate the TRACE RCIC model. The short-term SBO simulations were performed for two cladding materials: Zircaloy and FeCrAl, to demonstrate the effect of the accident tolerant fuel cladding on fuel heat-up under BDBA accident conditions. The wetwell plays a vital safety role in SBO and other BWR accident scenarios in that it can reduce containment pressure and supply additional core make-up water. The suppression pool temperature distribution has a very large impact on both RPV and containment pressure. Thus, another novel contribution of the project comes mainly from an improved, systems-level wetwell model which can capture buoyancy-induced thermal stratification effects due to steam injection and condensation. A two-zone stratified wetwell model has been implemented in RELAP-7 and some results from that model are presented. This wetwell model is capable of simulating thermal stratification due to a low steam mass injection rate. With a low mass flow rate, the model assumes that all the steam condenses within the pipe and the resulting plume can be approximated with a purely buoyant, heat-source driven model. The wetwell model developed with these assumptions is adequate to simulate slow transients such as extended SBO transients.

42 ENGINEERING↗
DOE OSTI2025

Integrated Molten Salt Reactor Modeling Capabilities in NEAMS Thermal Hydraulics Tools

The DOE neams program supports a full range of computational thermal fluids analysis capabilities and code developments for a broad range of advanced reactor concepts. The research and development approach under the thermal fluids technical area synergistically combines three length and time scales in a hierarchical multi-scale approach. To enable multi-scale thermal fluids capability using these codes, a key joint effort has been underway to develop an integrated system- and engineering-scale thermal fluids analysis capability, through integration of SAM and Pronghorn codes, both based on the MOOSE framework. This report summarizes recent advances in developing an integrated system- and engineering-scale modeling capability for the msr concept, which has gained significant interest in recent years. A consistent framework was established by coupling Pronghorn and SAM through the Saline interface, with thermophysical properties provided by the Molten Salt Thermal Property Database (MSTDB-TP). Further improvements were made to the coupling schemes and domain-overlapping strategies, enhancing the stability and robustness of multi-code simulations. Verification and validation efforts demonstrate the accuracy of this integration across a range of benchmark problems, including one-dimensional heated pipe flows, three-dimensional natural convection loops with evolving isotopic compositions, and \gls{msre} demonstration cases. Within Pronghorn, new capabilities were introduced to model corrosion and noble-metal plating phenomena, supported by an extended thermal-hydraulics framework and refined turbulence treatments. To capture two-phase flow behavior, a multiphase Euler–Euler model was implemented in Pronghorn, including advanced closure relations, high-resolution advection techniques, and capillary force reconstruction. Preliminary verification cases confirm the fidelity of the approach, while planned validation efforts target canonical multiphase benchmarks and application to msr components such as the msre pump bowl. Finally, updates to SAM’s msr mass transfer modeling were extended to consider noble gas migration into porous structures like graphite. The point kinetics model was updated to include reactivity feedback contributions from any defined species, such as xenon. The gas transport model was expanded for applicability to gas mixtures, bubble efflux phenomena, and species transport between liquid and gas phases. A selection of multi-scale Sherwood number correlations from MOSCATO/NekRS and multi-phase correlations from literature have been added for improved accuracy in calculating mass transfer coefficients. A companion effort on developing system-level redox corrosion has also been incorporated into SAM. Collectively, these enhancements strengthen the predictive capability of SAM and Pronghorn for simulating MSR thermal-hydraulics, corrosion, multiphase behavior, and fission-product transport, providing a more complete toolset for design, safety analysis, and licensing support of next-generation \gls{msr}s.

42 - ENGINEERING↗
DOE OSTI2020

Requirements and Conceptual Design of Off-gas Systems for the Reprocessing of Metallic Fuels

An assessment has been conducted to determine how key regulations regarding volatile radionuclide emissions to the atmosphere may apply to the off-gas streams associated with electrochemical reprocessing. The scope of this assessment was based upon a generic electrochemical reprocessing scheme with a throughput rate of 200 MTIHM/y applied to metallic fuel discharged from a sodium fast reactor (SFR), but the findings are able to be translated to other advanced nuclear scenarios as merited. Air dispersion modeling was performed using the EPA CAP-88 model and evaluated the uncontrolled decontamination factors (DFs) that would be required to achieve regulatory compliance with the dose-based limits set forth by EPA regulation 40 CFR 190.10(a). These DFs were compared to those required by fuel cycle–based limits set forth by EPA regulation 40 CFR 190.10(b). Two theoretical sites with disparate climatological conditions were selected for air dispersion modeling (Idaho and Tennessee). The radionuclides modeled included 3 H, 85 Kr, 129 I, and selected alpha-emitting transuranic isotopes (referred to here as 239 Pu-TRU <1y ). It was found that the fuel cycle-based limits in 40 CFR 190.10(b) are most restrictive for 85 Kr and 239 Pu-TRU <1y , with DFs of 3 and 6.1E+09, respectively. The dose-based limit as derived from 40 CFR 190.10(a) could require mitigation of tritium in some scenarios, with an estimated DF of about 3 for the reference scenarios. The fuel cycle-based limit for 129 I resulted in a DF of about 240 for the reference scenario. The need for iodine mitigation based on dose to the public depended upon the physical form of iodine as either particulate or vapor-phase species. Emission of iodine from the facility as a vapor necessitated DFs of about 2 but emission as a particulate would require DFs >6,000 to meet thyroid dose-based limits. Effects of physical form on needed iodine mitigation are significant, but the understanding of speciation of iodine both during electrochemical reprocessing and after release to the atmosphere is limited. The electrochemical processing unit operations were evaluated to identify potential release points for the volatile radionuclides and to assess the potential for retention of the radionuclides within the process (thus decreasing the need for mitigation). Mitigation strategies for 3 H, 85 Kr, 129 I, and 239 Pu-TRU <1y were identified. In all cases, there are reasonably achievable pathways to regulatory compliance, although in some cases additional R&D is merited to verify the chemical speciation of these isotopes and to develop and demonstrate potential treatment technologies for this application. Whether or not additional off-gas controls (beyond common operations such as HEPA filtration and oxygen and moisture control) are needed for any of these regulated or volatile radionuclides depends on the (a) type of facility (NRC-regulated or DOE), (b) used fuel process rate, (c) used fuel burnup and composition, (d) speciation and retention of volatile radionuclides in the process and in the cell gas cleanup system, (e) site-specific parameters such as location, meteorology, stack height, and site boundaries, and (f) levels of conservatism and safety factors used in assessing compliance to air emissions regulations. Performance of this assessment revealed several areas where information is lacking or additional research is required in order to better determine if or what kinds of off-gas control might be needed. First, and most significantly, the understanding of the chemical speciation and physical form and partitioning of iodine during electrochemical processing operations is lacking and prevents the ability to accurately assess the potential iodine mitigation requirements. Future research in this area should be multifaceted and include thermodynamic modeling of iodine speciation in different process steps, experiments to quantify the kinetics of vapor-phase and melt-phase transitions, bench-scale experiments to determine the potential chemical and physical form of iodine emissions from the electrorefining process, and verification of iodine behavior with experiments utilizing operational facilities. Similarly, an improved understanding of iodine behavior in the environment after release from the facility stack will be required to refine dose estimations, as particulate and vapor-phase emissions result in significantly different doses to the MEI.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

HySafe research priorities workshop report Summary of the workshop organized in cooperation with US DOE and supported by EC JRC in Washington DC November 10-11 2014.

The HySafe research priorities workshop is held on the even years between the International Conference on Hydrogen Safety (ICHS) which is held on the odd years. The research priorities workshop is intended to identify the state-of-the-art in understanding of the physical behavior of hydrogen and hydrogen systems with a focus on safety. Typical issues addressed include behavior of unintended hydrogen releases, transient combustion phenomena, effectiveness of mitigation measures, and hydrogen effects in materials. In the workshop critical knowledge gaps are identified. Areas of research and coordinated actions for the near and medium term are derived and prioritized from these knowledge gaps. The stimulated research helps pave the way for the rapid and safe deployment of hydrogen technologies on a global scale. To support the idea of delivering globally accepted research priorities for hydrogen safety the workshop is organized as an internationally open meeting. In attendance are stakeholders from the academic community (universities, national laboratories), funding agencies, and industry. The industry participation is critically important to ensure that the research priorities align with the current needs of the industry responsible for the deployment of hydrogen technologies. This report presents the results of the HySafe Research Priorities Workshop held in Washing- ton, D.C. on November 10-11, 2014. At the workshop the participants presented updates (since the previous workshop organized two years before in Berlin, Germany) of their research and development work on hydrogen safety. Following the workshop, participants were asked to provide feedback on high-priority topics for each of the research areas discussed and to rank research area categories and individual research topics within these categories. The research areas were ranked as follows (with the percentage of the vote in parenthesis): 1. Quantitative Risk Assessment (QRA) Tools (23%) 2. Reduced Model Tools (15%) 3. Indoor (13%) 4. Unintended Release-Liquid (11%) 5. Unintended Release-Gas (8%) 6. Storage (8%) 7. Integration Platforms (7%) 8. Hydrogen Safety Training (7%) 9. Materials Compatibility/Sensors (7%) 10. Applications (2%) The workshop participants ranked the need for Quantitative Risk Analysis (QRA) tools as the top priority by a large margin. QRA tools enable an informed expert to quantify the risk asso- ciated with a particular hydrogen system in a particular scenario. With appropriate verification and validation such tools will enable: * system designers to achieve a desired level of risk with suitable risk mitigation strategies, * permitting officials to determine if a particular system installation meets the desired risk level (performance based Regulations, Codes, and Standards (RCS) rather than prescrip- tive RCS), and * allow code developers to develop code language based on rigorous and validated physical models, statistics and standardized QRA methodologies. Another important research topic identified is the development of validated reduced physical models for use in the QRA tools. Improvement of the understanding and modeling of specific release phenomena, in particular liquid releases, are also highly ranked research topics. Acknowledgement The International Association HySafe, represented here by the authors, would like to thank all participants of the workshop for their valuable contributions. Particularly appreciated is the active participation of the industry representatives and the steady support by the European Com- mission's Joint Research Centre (JRC). Deep gratitude is owed for the great support by the United States Department of Energy (DOE) Office of Energy Efficiency and Renewable Energy's Fuel Cell Technologies Office (EERE/FCTO) for the organization of the 2014 version of the hydrogen safety research priorities workshop. This page intentionally left blank.

08 HYDROGEN↗
DOE OSTI2023

Assessment of Sodium Thermal Stratification Models Utilizing the TSTF Benchmark

As a result of certain transient scenarios, a thermally stratified layer of liquid sodium can develop in the bulk coolant volumes of a sodium-cooled fast reactor (SFR). In addition to the effects a stratification layer has on the temperature of the heat transport system, a stratification layer can also influence the transition to and establishment of natural circulation flow, which plays an important role in passive cooling and the inherent safety of a pool-type SFR. Therefore, the ability to accurately capture thermal stratification phenomena is important when demonstrating the safety basis of a pool-type SFR during transient sequences. The present work assesses various computational models with different fidelities in their ability to predict thermal stratification in the upper plenum of an SFR. Each computational model will be assessed using the data generated at the Thermal Stratification Test Facility (TSTF) located at the University of Wisconsin-Madison. Using measured flow rate and inlet temperature data, the measured temperature distributions of the tests are compared to the predictions of the lumped volume-based models in SAS4A/SASSYS-1, a 1D-based model in SAM, and a 3-D computational fluid dynamics (CFD) model using STAR-CCM+. The relative performance of the various computational methods is assessed with respect to key metrics such as bulk coolant temperature distribution and plenum exit temperature. A total of eight tests are analyzed, covering different combinations of flow rates (3 and 10 GPM) and upper internal structure (UIS) configurations (none, solid, porous, and open) The perfect mixing model of SAS4A/SASSYS-1 provides the highest accuracy when the flow rate is high and there is no UIS in the test vessel, as high flow rate injection promotes thermal mixing of the sodium in the test vessel. For most of the analyzed tests, the stratified volume model of SAS4A/SASSYS-1 is able to predict the delay in the outlet temperature drop and temperature distribution in the test vessel by a small number of layers to represent thermal stratification. However, the stratified volume model can only simulate a maximum of three temperature layers within a volume and when a layer approaches the elevation of the outlet, the predicted outlet temperature can demonstrate rapid, non-physical changes. The 1-D axial mixing model of SAM provides results that agree reasonably well with the measured data in the prediction of the temporal evolution of the outlet temperature with the exception of the case with a high flow rate and no UIS. The SAM 1-D model has a similar level of accuracy to CFD results when it comes to predicting the outlet temperature. CFD shows overall good agreement in predicting the temperature distribution in the test vessel and outlet temperature. As CFD can model the test vessel geometry in detail, it performs well in the cases of complex geometries such as tests that included a UIS and internal flow through the UIS resulting in active mixing of the coolant in the test vessel. Each of the models discussed in the present work has the potential to be useful during the various stages of reactor design, analysis, and licensing. The lumped-volume approach can be applied for fast turnaround safety calculations to obtain overall reactor behavior during transients. The 1-D models provide improved accuracy when stratification is expected for a relatively low increase in the computational cost. The CFD model can be utilized for confirmatory analysis of the 1-D model, when experimental measurements are not available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Estimation of Core Recovery Time with TRACE

Following the severe accident in Fukushima Daiichi Nuclear Power Plant in Japan, safety related concerns about the zirconium based claddings used in current light water reactors have increased. High corrosion resistance and low neutron absorption cross section make zirconium a suitable cladding material for normal nuclear plant operational conditions. However, for nuclear accident scenarios when safety functions of the plant are lost, Zr claddings undergo exothermic reaction with steam and exacerbate the accident conditions. The exothermic reaction increases the temperature of the cladding and accelerates core melting in severe accident progression. And the generation of explosive hydrogen gas accumulates and challenges the structural integrity. The accident tolerant fuel (ATF) concept is the fuel and cladding system that gives a longer time margin, compared to standard Zr claddings, to cope with the accident when core cooling systems are lost in a nuclear power plant. The ATF cladding should have slower oxidation kinetics and less hydrogen generation than Zr to alleviate consequences of cladding interaction with steam at high temperatures. The maximum time available to mitigate progression of severe accident is an important measure to evaluate benefits of ATF claddings. For this work, the maximum time available to mitigate progression of severe accident is selected as the time at which cladding melting reaction starts. Melting reaction is selected as an upper limit because of the degradation of coolable geometry. Maintaining the coolable geometry is a regulation criteria for both loss of coolant (LOCA) analyses and outside the LOCA analyses. Once the melting of the cladding starts, it exacerbates the deformation of the fuel assembly and hottest rods within the core cannot be cooled by water reflood and results in fuel melting progression. The aim of this work is to evaluate time margin gain of ATF cladding in a severe accident scenario of LOCA without safety injections and station blackout considering oxidation kinetics only. Also it is assumed that safety injections are recovered before melting of the claddings start and the sensitivity of safety injection actuation time on core recovery is analyzed. The core recovery time is the time when makeup water from safety injections quench the fuel rods and halt the progression of the severe accident. This time is estimated by initiating safety injections at different times and observing the hot rod behavior which represents 1/5. of the core in this work. It is considered that core is not recoverable when 20% of the core is damaged.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Understanding the high-temperature behavior and corrosion resistance of Cr-Nb coated cladding for BWRs

While Cr-coated Accident Tolerant Fuel has proven successful in Pressurized Water Reactors, its dissolution in high dissolved oxygen Boiling Water Reactor environments represents a major technological barrier. Here, this study presents a comprehensive evaluation of novel Cr-Nb alloys as a coating breakthrough solution for Accident Tolerant Fuel cladding in Boiling Water Reactors. We systematically investigated Physical Vapor Deposited Cr-Nb coatings with 13% and 24 at% Nb content on commercial Zr-based cladding through rigorous testing under both normal and accident conditions. Our results demonstrate Cr-Nb coatings remained completely intact during extended Boiling Water Reactor autoclave testing, while pure Cr coatings failed catastrophically due to oxide dissolution. Under Loss-of-Coolant Accident conditions at 1100°C, Cr-Nb coatings maintained protective capability for 45 min. At 800°C, the alloy coatings matched pure Cr's excellent corrosion resistance. Despite microstructural changes during prolonged high-temperature exposure, no coating delamination occurred, ensuring continued protection of the underlying Zr substrate. These findings establish Cr-Nb coatings as a viable Accident Tolerant Fuel solution for Boiling Water Reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗