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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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At least 271 records · Page 15

Clad to coolant heat transfer during RIA in PWR conditions

The power pulse during an insertion of reactivity accident (RIA) leads to a transient heating of the fuel. Heat transfer toward the coolant through the clad and associated temperature transient of the clad is of safety significance since it may largely impact the risk of clad failure. For sufficiently large enthalpy deposited within the fuel, clad temperature increases rapidly and this can lead to coolant boiling. This paper concerns the critical review of the present understanding and of the models for the clad to coolant heat transfer in RIA for PWR conditions. The results of past experiments are reviewed and the reliability of current models dedicated to these conditions is analyzed. It appears that the use of classical models for wall to fluid heat transfer does not allow predicting the clad temperature transient as measured in RIA-related experiments. The deviation is all the more large when film boiling conditions are reached. To reproduce the experimental data, some correlations of heat transfer can be fitted but the reliability of such model out of the range of conditions of the experimental data is low. It is mainly due to a rather low level of understanding of the heat transfer in very transient conditions (i.e. when the time scale of variation of the main heat transfer parameters, like a wall temperature, is less than the typical internal time scale of the heat transfer process itself). It results in a very large scattering in the numerical simulation of clad temperature when several fuel codes are considered. To improve the reliability of these calculations, it is believed that the study of transient boiling heat transfer thanks to devoted experimental facilities is required.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of R&D GREET 2023 Rev1 to Estimate Greenhouse Gas Emissions of Sustainable Aviation Fuels for 40B Provision of the Inflation Reduction Act

The federal Interagency Working Group on sustainable aviation fuels (SAF) tasked Argonne National Laboratory with developing a modified version of the Greenhouse Gases, Regulated Emissions, and Energy Use in Technologies (GREET) model based on R&D GREET 2023. The goal of the new GREET version is to simulate the life-cycle greenhouse gas (GHG) emissions associated with seven sustainable aviation fuel (SAF) pathways for consideration under the 40B Provision of the Inflation Reduction Act – Sustainable aviation fuel credit. The Provision includes a new GHG-based tax credit to incentivize SAF production and reduce the costs of these fuels.

02 PETROLEUM↗

Sensitivity analysis of VERA-CS and FRAPCON coupling in a multiphysics environment

A demonstration and description of the LOCA Toolkit for US light water reactors (LOTUS) is presented. Through LOTUS, the core simulator VERA-CS developed by CASL is coupled with the fuel performance code FRAPCON. The coupling is performed with consistent uncertainty propagation with all model inconsistencies being well-documented. Monte Carlo sampling is performed on a single 17 × 17 fuel assembly with a three cycle depletion case. Both uncertainty quantification (UQ) and sensitivity analysis (SA) are used at multiple states within the simulation to elucidate the behavior of minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). The SA metrics used are the Pearson correlation coefficient, Sobol sensitivity indices, and the density-based, delta moment independent measures. Results for MDNBR show consistency among all SA measures, as well for all states throughout the fuel lifecycle. MFCT results contain consistent rankings between SA measures, but show differences throughout the lifecycle. GCPP exhibits predominantly linear relations at low and high burnup, but highly nonlinear relations at intermediate burnup due to abrupt shifts between models. As a result, such behavior is largely undetectable to traditional regression or variance-based methods and demonstrates the utility of density-based methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Impact of Grain Size on Performance Degradation of TREAT LEU

We argue that radiation damage induced degradation of thermal conductivity does not set a lower limit on fuel grain sizes for the low enriched uranium fuel design of the Transient Reactor Test Facility (TREAT). Earlier work reports that smaller grains cause a larger degradation of thermal conductivity than larger grains constraining the smallest feasible size of fuel grains. This work assesses TREAT’s transient performance in the presence of radiation damage. The difference between the two studies is in treating damaged and fresh graphite as serial (this work) or parallel (previous) heat resistors. We use a multiphysics model of TREAT fuel grains to compute the reduction in transient capability measured by the total deposited energy as a function of irradiation dose. We find that radiation damage has a negligible effect on energy deposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Exploration of the Phase Field Framework Marmot to Include Anisotropic Grain Boundaries with Molecular Dynamics [Thesis]

Using molecular dynamic (MD) calculations, meso-scale phase field simulations behave differently by including anisotropic grain boundary energy in the model for transient energy minimization in uranium-dioxide. As a preliminary step, MD will be used to find face centered cubic (FCC) copper grain boundary energies and to explore its anisotropic behavior in phase field simulations. MARMOT is a C++ code that uses object-oriented programming to implement the Allen-Cahn and Cahn-Hilliard equations for phase field modeling. MARMOT uses constants for mobility and grain boundary energy for the entire domain of the simulation. For MARMOT to reach its full capability, these constants need to vary within the domain. This paper explores the expansion of MARMOT to include an anisotropic view of grain boundary energy in transient phase field simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Results and Analysis of the Research and Development Work Scope Request for Information (DE-SOL-0008246)

The Department of Energy (DOE) Office of Nuclear Energy (NE) released a request for information (RFI) (DE-SOL-0008246) for “University, National Laboratory, Industry and International Input to the Office of Nuclear Energy’s Competitive Research and Development Work Scope Development” on April 13, 2015. DOE-NE solicited information for work scopes for the four main program areas as well as any others suggested by the community. The RFI proposal period closed on June 19, 2015. From the 124 responses, 238 individual work scopes were extracted. Thirty-three were associated with a DOE national laboratory, including Argonne National Laboratory (ANL), Brookhaven National Laboratory (BNL), Idaho National Laboratory (INL), Los Alamos National Laboratory (LANL), Pacific Northwest National Laboratory (PNNL) and Oak Ridge National Laboratory (ORNL). Thirty US universities submitted proposals as well as ten industrial/commercial institutions. Four major R&D areas emerged from the submissions, appearing in more than 15% of the proposed work scopes. These were: nuclear fuel studies, safety and risk analysis, nuclear systems analysis and design and advanced instrumentation and controls. Structural materials for nuclear power plants, used nuclear fuel disposition and various types of systems analysis were also popular, each appearing in more than 10% of the proposals. Nuclear Energy Enabling Technologies (NEET) was the most popular program area with 42% of the proposals referencing the NEET-CTD program. The order of the remaining programs was Fuel Cycle Technologies (FC) at 34%, Nuclear Energy Advanced Modeling and Simulation (NEAMS) at 29% and Reactor Concepts at 17%.

99 GENERAL AND MISCELLANEOUS↗

A Systems Processing Model for Molten Salt Reactors

The chemistry group within the U.S. Department of Energy Molten Salt Reactor Campaign is developing a model for the molten salt reactor system for assessing the suitability of various used fuel processing schemes. In the long run, the developed model is expected to be a tool for better understanding the interplay between the components of the molten salt reactor system including uranium resources, the molten salt reactor, the used fuel processor (e.g., reprocessing plant), and waste management. In this regard, the report proposes a theoretical framework for integrating the various components of the molten salt reactor system. Rather than being buried in implementation details for seemingly heterogeneous components of the reactor system asking for various modeling techniques, this report proposes to approximate and integrate all components with a unified mathematical framework, a discrete-time linear system. An abstracted molten salt reactor system is given as an example to illustrate the effectiveness of the proposed theoretical framework. A fully developed molten salt reactor system model (based on the proposed framework) is expected to be beneficial for various objectives including the transient and equilibrium analysis of fuel cycle scenarios and the identification of potential technology bottlenecks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

MST e-News Summer 2023

David Andersson (Materials Science in Radiation and Dynamic Extremes, MST-8) is the recipient of the 2023 American Nuclear Society (ANS) Mishima Award. The award recognizes outstanding contributions of an individual in research and development work on nuclear fuels and materials. Andersson is a staff scientist on MST-8’s Modeling of Radiation Effects team. An expert in atomistic modeling of nuclear fuels—both traditional UO 2 light water reactor fuel and advanced accident tolerant fuel concepts—he has been involved in numerous international and national multiscale simulation efforts to develop models for use in nuclear fuel performance simulations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Spent fuel storage and waste management fuel cycle optimization using CAFCA

Spent fuel storage modeling is at the intersection of nuclear fuel cycle system dynamics and waste management policy. A model that captures the economic parameters affecting used nuclear fuel storage location options, which complements fuel cycle economic assessment has been created using CAFCA (Code for Advanced Fuel Cycles Assessment) of MIT. Research has also expanded to the study on dependency of used nuclear fuel storage economics, environmental impact, and proliferation risk. Three options of local, regional, and national storage were studied. The preliminary product of this research is the creation of a system dynamics tool known as the Waste Management Module which provides an easy to use interface for education on fuel cycle waste management economic impacts. Storage options costs can be compared to literature values with simple variation available for sensitivity study. Additionally, a first of a kind optimization scheme for the nuclear fuel cycle analysis is proposed and the applications of such an optimization are discussed. The main tradeoff for fuel cycle optimization was found to be between economics and most of the other identified metrics. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Light-Duty Drive Cycle Simulations of Diesel Engine-Out Exhaust Properties for an RCCI-Enabled Vehicle

In-cylinder blending of gasoline and diesel fuels to achieve low-temperature reactivity controlled compression ignition (RCCI) can reduce NOx and PM emissions while maintaining or improving brake thermal efficiency compared to conventional diesel combustion (CDC). Moreover, the dual-fueling RCCI is able to achieve these benefits by tailoring combustion reactivity over a wider range of engine operation than is possible with a single fuel. However, the currently demonstrated range of stable RCCI combustion just covers a portion of the engine speed-load range required in several light-duty drive cycles. This means that engines must switch from RCCI to CDC when speed and load fall outside of the stable RCCI range. In this study we investigated the impact of RCCI as it has recently been demonstrated on practical engine-out exhaust temperature and emissions by simulating a multi-mode RCCI-enabled vehicle operating over two urban and two highway driving cycles. To implement our simulations, we employed experimental engine maps for a multi-mode RCCI/CDC engine combined with a standard mid-size, automatic transmission, passenger vehicle in the Autonomie vehicle simulation platform. Our results include both detailed transient and cycle-averaged engine exhaust temperature and emissions for each case, and we note the potential implications of the modified exhaust properties on catalytic emissions control and utilization of waste heat recovery on future RCCI-enabled vehicles.

Gao, Zhiming↗

Carbon Dioxide Absorption Modeling for Off-Gas Treatment in the Nuclear Fuel Cycle

The absorption of carbon dioxide is an important process in many practical applications such as reduction of greenhouse gases, separation and purification processes in the chemical and petroleum industries, and capture of radioactive isotopes in the nuclear fuel cycle The goal of this research is to develop a dynamic model to simulate CO 2 absorption by using different alkanolamines as absorption solvents. The model is based upon transient mass and energy balances for the chemical species commonly present in CO 2 gas-liquid absorption. A computer code has been written to implement the proposed model. Simulation results are discussed. The reported model simulates well the response to dynamic changes in input conditions. The proposed model can be used to optimize and control the separation of carbon-14 in the form of CO 2 in the nuclear industry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Compare Mechanistic Predictions for Doped UO 2 Mechanical Response and Other Properties with Empirical Models and Experimental Measurements

The U.S. Department of Energy’s Nuclear Energy Advanced Modeling and Simulation program develops predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON, a high-fidelity, high resolution fuel performance tool at the engineering scale. As part of its development, additional modeling capabilities and improvements have been developed for relevant fuel forms. In this work, a fuel creep deformation model for Cr-doped fuel has been implemented into BISON, along with improvements to the empirical UO 2 fuel creep model based on experimental data and improvements to the radial power factor calculation for doped fuels. This work allows for more accurate simulation analyses for both UO 2 and doped-UO 2 fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High-level fuel fabrication facility designs from discrete-event simulation

Like other industrial processes, the production of metallic nuclear fuels (MNF) requires that fabrication facilities be able to reliably meet production demands, operate efficiently, and adhere to federal and local safety regulations. In turn, the set of design variables employed by such a facility, such as operations policies, infrastructure and machinery purchased, and the type and number of staff hired, directly impact a facility’s ability to satisfy these goals. Therefore, facility designers must carefully determine which set of design variable values optimally satisfies these constraints. In this paper, we explore how values for these high-level design variables, namely hiring requirements, can be determined in the context of nuclear fuel manufacturing through the coupling of physics-based and discrete-event simulation technologies. Using the Versatile Test Reactor (VTR) program as a case study, we demonstrate how SCALE and MCNP nuclear physics model outputs can be integrated into ExtendSim discrete-event simulation (DES) models of the fuel fabrication process to determine the optimal number of staff hired to ensure fuel production goals are met, operations comply with effective dose limit regulations, and overall project costs are reduced.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Program Computes Performances Of Scramjets

SRGULL, a scramjet-cycle-analysis computer code, engineer's software tool capable of nose-to-tail simulation of hydrogen-fueled, scramjet engine integrated with air-frame. Simulated flow that of real gas with equilibrium thermodynamic properties. Program facilities initial estimates of performance of scramjet cycle by linking code for two-dimensional forebody, inlet, and nozzle with code for one-dimensional combustor. Five computer codes (SCRAM, SEAGUL, INLET, Program HUD, and GASH) integrated in program to provide capability for analysis of changing flow conditions. Written in FORTRAN 77.

Walton, J. T.↗

Fuel Cycle Performance of Thermal Spectrum Small Modular Reactors

Small modular reactors may offer potential benefits, such as enhanced operational flexibility. However, it is vital to understand the holistic impact of small modular reactors on the nuclear fuel cycle and fuel cycle performance. The focus of this paper is on the fuel cycle impacts of light water small modular reactors in a once-through fuel cycle with low-enriched uranium fuel. A key objective of this paper is to describe preliminary reactor core physics and fuel cycle analyses conducted in support of the U.S. Department of Energy Office of Nuclear Energy Fuel Cycle Options Campaign. Challenges with small modular reactors include: increased neutron leakage, fewer assemblies in the core (and therefore fewer degrees of freedom in the core design), complex enrichment and burnable absorber loadings, full power operation with inserted control rods, the potential for frequent load-following operation, and shortened core height. Each of these will impact the achievable discharge burn-up in the reactor and the fuel cycle performance. This paper summarizes the results of an expert elicitation focused on developing a list of the factors relevant to small modular reactor fuel, core, and operation that will impact fuel cycle performance. Preliminary scoping analyses were performed using a regulatory-grade reactor core simulator. The hypothetical light water small modular reactor considered in these preliminary scoping studies is a cartridge type one-batch core with 4.9% enrichment. Some core parameters, such as the size of the reactor and general assembly layout, are similar to an example small modular reactor concept from industry. The high-level issues identified and preliminary scoping calculations in this paper are intended to inform on potential fuel cycle impacts of one-batch thermal spectrum SMRs. In particular, this paper highlights the impact of increased neutron leakage and reduced number of batches on the achievable burn-up of the reactor. Fuel cycle performance metrics for a small modular reactor are compared to a conventional three-batch light water reactor in the following areas: nuclear waste management, environmental impact, and resource utilization. Metrics performance for a small modular reactor are degraded for mass of spent nuclear fuel and high level waste disposed, mass of depleted uranium disposed, land use per energy generated, and carbon emission per energy generated

Worrall, Andrew↗

A Turbine Based Combined Cycle Engine Inlet Model and Mode Transition Simulation Based on HiTECC Tool

An inlet system is being tested to evaluate methodologies for a turbine based combined cycle propulsion system to perform a controlled inlet mode transition. Prior to wind tunnel based hardware testing of controlled mode transitions, simulation models are used to test, debug, and validate potential control algorithms. One candidate simulation package for this purpose is the High Mach Transient Engine Cycle Code (HiTECC). The HiTECC simulation package models the inlet system, propulsion systems, thermal energy, geometry, nozzle, and fuel systems. This paper discusses the modification and redesign of the simulation package and control system to represent the NASA large-scale inlet model for Combined Cycle Engine mode transition studies, mounted in NASA Glenn s 10- by 10-Foot Supersonic Wind Tunnel. This model will be used for designing and testing candidate control algorithms before implementation.

Csank, Jeffrey T.↗

A Turbine Based Combined Cycle Engine Inlet Model and Mode Transition Simulation Based on HiTECC Tool

An inlet system is being tested to evaluate methodologies for a turbine based combined cycle propulsion system to perform a controlled inlet mode transition. Prior to wind tunnel based hardware testing of controlled mode transitions, simulation models are used to test, debug, and validate potential control algorithms. One candidate simulation package for this purpose is the High Mach Transient Engine Cycle Code (HiTECC). The HiTECC simulation package models the inlet system, propulsion systems, thermal energy, geometry, nozzle, and fuel systems. This paper discusses the modification and redesign of the simulation package and control system to represent the NASA large-scale inlet model for Combined Cycle Engine mode transition studies, mounted in NASA Glenn s 10-foot by 10-foot Supersonic Wind Tunnel. This model will be used for designing and testing candidate control algorithms before implementation.

Csank, Jeffrey↗

Report on Advanced Detector Development

Neutron, gamma and charged particle detection improvements are key to supporting many of the foreseen measurements and systems envisioned in the R&D programs and the future fuel cycle requirements, such as basic nuclear physics and data, modeling and simulation, reactor instrumentation, criticality safety, materials management and safeguards. This task will focus on the developmental needs of the FCR&D experimental programs, such as elastic/inelastic scattering, total cross sections and fission neutron spectra measurements, and will leverage a number of existing neutron detector development efforts and programs, such as those at LANL, PNNL, INL, and IAC as well as those at many universities, some of whom are funded under NE grants and contracts. Novel materials and fabrication processes combined with state-of-the-art electronics and computing provide new opportunities for revolutionary detector systems that will be able to meet the high precision needs of the program. This work will be closely coordinated with the Nuclear Data Crosscut. The Advanced Detector Development effort is a broadly-focused activity that supports the development of improved nuclear data measurements and improved detection of nuclear reactions and reactor conditions. This work supports the design and construction of large-scale, multiple component detectors to provide nuclear reaction data of unprecedented quality and precision. Examples include the Time Projection Chamber (TPC) and the DANCE detector at LANL. This work also supports the fabrication and end-user application of novel scintillator materials detection and monitoring.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗