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At least 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss of coolant accident (LOCA), there is the possibility of nuclear fuel rods to undergo a three step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light water reactors (LWRs) it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling modeling approach is used to garner insight into underlying mechanisms leading to the ne fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering scale Bison fuel performance code. Details are also provided on cladding modeling improvements related to hydrogen/hydride embrittlement and damage, and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Advancements in modeling fuel pulverization and cladding behavior during a LOCA

During a loss-of-coolant accident (LOCA), it is possible for nuclear fuel rods to undergo a three-step process known as fuel fragmentation, relocation, and dispersal (FFRD). The chance of FFRD occurring increases as the fuel burnup increases. To support the nuclear industry's desire to increase the discharge burnup of nuclear fuels in light-water reactors (LWRs), it is imperative to understand the mechanisms driving the evolution of FFRD. In this work, a multiscale modeling approach is used to garner insight into underlying mechanisms leading to the fine fragmentation (also known as pulverization) of nuclear fuel during a LOCA. This report includes a summary of the atomistic and phase-field studies to develop a new pulverization criterion for use in the engineering-scale Bison fuel performance code. Details are also provided on cladding model improvements related to hydrogen/hydride embrittlement and damage and anisotropic thermal creep. The new models are used on the existing integral and separate effects LOCA validation cases available in Bison.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Description of a Dry Cask Simulator for Measuring Internal and External Thermal-Hydraulic Performance

The thermal performance of a commercial nuclear spent fuel dry storage cask is evaluated through detailed analytical modeling. These modeling efforts are performed by the vendor to demonstrate the performance as well as regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). The majority of commercial dry storage casks in use today are aboveground. Cooling of the assemblies located inside the sealed canister is achieved by air flowing in the bottom of the cask and exiting out the top. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating design and performance models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. Thus the enhanced performance of modern dry storage casks cannot be fully validated using previous studies. The purpose of the investigation described in this report is to produce data sets for the validation of assumptions associated with the thermal-hydraulic simulation of modern dry cask storage systems. Of particular interest are the steady-state cladding temperatures and the effect of elevated helium pressures in sealed canisters in both above and belowground designs. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly will be deployed inside of a representative storage basket and cylindrical pressure vessel that will represent the canister. The arrangement of a single assembly with well-controlled boundary conditions should simplify interpretation of results. Various arrangements of outer concentric ducting will be used to mimic conditions for above and belowground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. As of this writing, the apparatus is currently under construction at Sandia National Laboratories in Albuquerque, NM and is expected to be operational starting in April 2016. Pre-test CFD calculations have been performed using best available information. These calculations are being used to assist in test design but will need to be validated and potentially reevaluated. The planned testing will be adjusted to reflect as-built conditions and measured system limitations. Dimensional analyses indicate that the anticipated ranges of relevant dimensionless groups (Reynolds, Modified Rayleigh, and Nusselt numbers) bracket or closely approach prototypic values for both the aboveground and belowground configurations. While designed to match prototypic values, the expected test matrix will include values that exceed currently acceptable values for decay heat, internal helium pressure, and peak cladding temperatures in order to gain more insight into the underlying behavior of the system. (authors)

42 ENGINEERING↗
DOE OSTI2017

Design and modeling for hydrogen fuel cell conversion of parcel delivery trucks

With an emphasis on reducing fossil fuel consumption and emissions, major delivery service corporations have begun deploying and operating Class 6 zero emissions delivery vans. Battery electric delivery vans are currently available, but their range is a limitation and doesn’t offer flexibility in assigning vehicles to delivery routes. Fuel cell hybrid delivery vans offer a zero emissions solution with extended range and route planning flexibility. In a competitive market, a delivery van design that trades the fuel cell and battery size, performance, and cost is critical to the success and adoption of these zero emissions trucks. This paper outlines a systematic approach using real world operational data and duty cycles to properly size the fuel cell and battery for a zero emissions hybrid delivery van. The approach uses real world GPS data to define routes and duty cycles, along with vehicle powertrain models to assess performance capabilities. Furthermore, the study suggests that a 32 kW fuel cell power system paired with a 49 kWh battery energy storage system and 15 kg of hydrogen storage is a desired option to complete the most demanding parcel and delivery duty cycles.

08 HYDROGEN↗
DOE OSTI2014

Performance of solid oxide fuel cells approaching the two-dimensional limit

We model electrochemical kinetics and physical conduction mechanisms for carrier transport in electrolyte membranes to determine the limits of dimensionality reduction (down to 1 nm) on maximum power output of solid oxide fuel cells with symmetric Pt electrodes. Using Y-doped ZrO{sub 2}, we find a minimum thickness of ∼6 nm to realize near ideal chemical potential in such fuel cells, which is limited by electronic breakdown when approaching the dielectric breakdown strength. For larger electrolyte thicknesses, the greatest source of electronic leakage influencing power loss is from Ohmic transport of minority carriers and emission of trapped carriers. For porous metallic electrodes, an ideal microstructure with the particle size comparable to particle spacing dimensions is found to accurately model experimental results. The role of electronic trap states in the electrolyte band gap on power density characteristics is highlighted.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A layered 2D computational framework: Theory and applications to nuclear fuel behavior

Nuclear fuel performance computer codes have been developed over the last 50 years to analyze fuel behavior under various operating conditions. Traditionally, these codes used quasi-two-dimensional (also commonly known as 1.5D) representations of the fuel rod, which model the rod using a set of one-dimensional axisymmetric models that represent the behavior at specific axial positions on the rod. Modern fuel performance codes have the ability to investigate full three-dimensional (3D) effects and couple to other physics-based codes for true multiphysics simulations. However, with increasing complexity comes increasing computational costs. Many phenomena of interest involve azimuthally-varying behavior that cannot be represented using the aforementioned quasi-two-dimensional approach, but do not require the use of a full 3D model. To efficiently address these problems, there is a need for a computational framework that provides a compromise between the quasi-two-dimensional and full 3D models. In this paper, we present a new quasi-three-dimensional approach that represents the fuel as a set of 2D planar models that represent the behavior of the fuel cross-section at various axial positions. Presented here are the theory behind the methodology, test cases to illustrate proper implementation, and practical applications of its use in the BISON fuel performance code for a variety of cases in nuclear fuel analysis, such as fuel fracture, axial fuel relocation, and cladding distension and oxidation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Modeling and Simulations for the High Flux Isotope Reactor Cycle 400

A concerted effort over the past few years has been focused on enhancing the core model for the High Flux Isotope Reactor (HFIR), as part of a comprehensive study for HFIR conversion from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel. At this time, the core model used to perform analyses in support of HFIR operation is an MCNP model for the beginning of Cycle 400, which was documented in detail in a 2005 technical report. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed to serve as reference for the design of an LEU fuel for HFIR. The recent enhancements in modeling and simulations for HFIR that are discussed in the present report include: (1) revision of the 2005 MCNP model for the beginning of Cycle 400 to improve the modeling data and assumptions as necessary based on appropriate primary reference sources HFIR drawings and reports; (2) improvement of the fuel region model, including an explicit representation for the involute fuel plate geometry that is characteristic to HFIR fuel; and (3) revision of the Monte Carlo-based depletion model for HFIR in use since 2009 but never documented in detail, with the development of a new depletion model for the HFIR explicit fuel plate representation. The new HFIR models for Cycle 400 are used to determine various metrics of relevance to reactor performance and safety assessments. The calculated metrics are compared, where possible, with measurement data from preconstruction critical experiments at HFIR, data included in the current HFIR safety analysis report, and/or data from previous calculations performed with different methods or codes. The results of the analyses show that the models presented in this report provide a robust and reliable basis for HFIR analyses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Modeling and validation of single-chamber microbial fuel cell cathode biofilm growth and response to oxidant gas composition

This work describes experiments and computational simulations to analyze single-chamber, air-cathode microbial fuel cell (MFC) performance and cathodic limitations in terms of current generation, power output, mass transport, biomass competition, and biofilm growth. Steady-state and transient cathode models were developed and experimentally validated. Two cathode gas mixtures were used to explore oxygen transport in the cathode: the MFCs exposed to a helium-oxygen mixture (heliox) produced higher current and power output than the group of MFCs exposed to air or a nitrogen-oxygen mixture (nitrox), indicating a dependence on gas-phase transport in the cathode. Multi-substance transport, biological reactions, and electrochemical reactions in a multi-layer and multi-biomass cathode biofilm were also simulated in a transient model. The transient model described biofilm growth over 15 days while providing insight into mass transport and cathodic dissolved species concentration profiles during biofilm growth. Lastly, simulation results predict that the dissolved oxygen content and diffusion in the cathode are key parameters affecting the power output of the air-cathode MFC system, with greater oxygen content in the cathode resulting in increased power output and fully-matured biomass.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2017

A kinetic modeling study on octane rating and fuel sensitivity under HCCI conditions

With the development of advanced compression ignition engines such as the homogeneous charge compression ignition (HCCI) and the reactivity controlled compression ignition (RCCI), the long existing research and motor octane rating for fuels, i.e., RON and MON, might be insufficient to describe the fuel reactivity and performance with these new combustion strategies due to their distinctive engine operation conditions. It is also noted that with the stringent regulation for CO2 emission, a detailed knowledge of fuel reactivity is urgently needed for the selection of novel alternative bio-fuels for transportation, and optimization of their engine performance in the future. It is therefore very important to understand the reactivity and sensitivity for general classes of fuels as a function of the operating conditions in these advanced combustion engines. In this modeling study, using validated kinetic mechanism for ignition, we have simulated the performance of different classes of fuels including hydrogen, alkanes (straight and branched), alcohols and potential bio-fuels under typical HCCI engine conditions using detailed chemical kinetics. The crank angle corresponding to 50% total heat release (CA50) is utilized as an indicator for the fuel reactivity. Iso-contours of CA50 are then plotted in the engine operating parameter coordinate, such as the intake temperature, and pressure. It is shown that CA50 value can vary significantly with different engine operating conditions, and that fuels with the same RON/MON even exhibit qualitatively different CA50 iso-contours in specific operating regimes, implying differences in the intrinsic reactivity of different classes of fuels, as well as the insufficiency of RON/MON as comprehensive metrics of fuel reactivity in advanced compression ignition (ACI) strategies. In addition, the operating regime with high and low sensitivity, as well as potential non-monotonic change in fuel reactivity, has been identified by the CA50 iso-contour map for all studied fuels.

Wu, Tong↗
DOE OSTI2015

Evaluation of DUSTRAN Software System for Modeling Chloride Deposition on Steel Canisters

The degradation of steel by stress corrosion cracking (SCC) when exposed to atmospheric conditions for decades is a significant challenge in the fossil fuel and nuclear industries. SCC can occur when corrosive contaminants such as chlorides are deposited on a susceptible material in a tensile stress state. The Nuclear Regulatory Commission has identified chloride-induced SCC as a potential cause for concern in stainless steel used nuclear fuel (UNF) canisters in dry storage. The modeling of contaminant deposition is the first step in predictive multiscale modeling of SCC that is essential to develop mitigation strategies, prioritize inspection, and ensure the integrity and performance of canisters, pipelines, and structural materials. A multiscale simulation approach can be developed to determine the likelihood that a canister would undergo SCC in a certain period of time. This study investigates the potential of DUSTRAN, a dust dispersion modeling system developed by Pacific Northwest National Laboratory, to model the deposition of chloride contaminants from sea salt aerosols on a steel canister. Results from DUSTRAN simulations run with historical meteorological data were compared against measured chloride data at a coastal site in Maine. DUSTRAN’s CALPUFF model tended to simulate concentrations higher than those measured; however, the closest estimations were within the same order of magnitude as the measured values. The decrease in discrepancies between measured and simulated values as the level of abstraction in wind speed decreased suggest that the model is very sensitive to wind speed. However, the influence of other parameters such as the distinction between open-ocean and surf-zone sources needs to be explored further. Deposition values predicted by the DUSTRAN system were not in agreement with concentration values and suggest that the deposition calculations may not fully represent physical processes. Overall, results indicate that with parameter refinement, DUSTRAN has the potential to simulate atmospheric chloride dispersion on steel canisters.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS1981

Simplified off-design performance model of a day turbofan engine cycle

The specific thrust and fuel-air ratio for a dry turbofan engine cycle were calculated for several power levels over a range of altitudes and Mach numbers. The engine has a design fan pressure ratio of 2.9, compressor pressure ratio of 8.0, and bypass ratio of 0.6. Nominal engine component curves were picked to approximate the calculated data to construct a simplified model of the off-design performance of the engine. The model was then used to construct a simplified design-point engine model for the full-power condition.

Lallman, F. J.↗
DOE OSTI2016

Comparison between Flownex and Corona thermal-hydraulic models for a single HTR prismatic fuel assembly - 18544

Computational tools and methodologies that can perform thermal-hydraulic analyses to reasonably predict the temperature distribution for the whole core of HTRs are useful for operational and safety analyses. In this paper, Flownex (a system code based on a one-dimensional thermal-fluid network methodology) and CORONA (Core Reliable Optimization and thermo-fluid Network Analysis) thermal-hydraulic models and results for a vertically stacked single fuel assembly will be compared. In Flownex, a basic unit cell containing part of the graphite and the associated fuel elements and coolant channels was developed. CORONA uses a basic unit cell concept for generating computational grids using several types of basic unit cells to model the heat transfer through the hexagonal blocks and a 1-D formulation to model the flow through the coolant channels. In CORONA, computational grids are generated using a combination of pre-generated mesh cells. The coolant channels and fuel rods are modelled explicitly. Flownex in contrast, models fuel and coolant channels implicitly by defining representative coolant channels, fuel rods and coated particle layers associated with each unit cell. Flownex can therefore model temperatures at coated particle level and also through each graphite block. The computational mesh employed by Flownex is much coarser than the mesh employed by CORONA. In both models, the helium gap between the fuel compact and the graphite block is modelled as stagnant. Flownex results were found to be of the same order of magnitude as those obtained using CORONA. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

An approach to grid-to-rod fretting wear modeling using dynamic mortar contact

This article presents the development of a grid-to-rod fretting (GTRF) modeling methodology that utilizes dynamic mortar contact. Here, we leverage a recently developed computational framework for modeling thermomechanical contact in dynamic simulations from the nuclear fuel performance code Bison. Usage of mortar contact ensures a smooth distribution of normal and frictional contact forces, displacements, and velocities on the contact interface, thus facilitating an accurate computation of wear. The integration within an advanced nuclear fuel performance code enables analysis of complex interactions between the fuel, cladding, and spacer grid. Such interactions include swelling, creep, fuel fragmentation, burnup, plenum pressure, and fission gas release. Our methodology is demonstrated via the modeling of two spacer grid geometries and the influence of fuel-cladding mechanical contact on the generation of fretting wear.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermo-Mechanical Analysis of SBLOCA with Loss of One Safety Injection System Using SCDAP/RELAP5

Design Extension Condition (DEC) for nuclear power plants has been being deliberated to bring it within a safety regulation boundary, following the 2011 Fukushima Daiichi accident. Moreover, a new regulation on the LOCA accident analysis designated as 10CFR50.45(c) has been proposed based on a concern of the non-conservatism of current criterion neglecting fuel burn-up effect on the cladding embrittlement. Therefore, a more comprehensive and realistic simulation of fuel rod behavior is required to cope with the stringent environment under LOCA accidents. A SCDAP/RELAP5 system analysis code has been used mainly on the prediction of severe accident phenomena. It has a set of mechanistic models for the fuel deformation, which plays an important role particularly in the early phase progression of the fuel melting. However, limited work has been done focusing on the models only to the fuel failure estimation through the DBA or candidates of DEC. Therefore, some analyses were performed here with a burnup dependent core model to understand the capability of SCDAP/RELAP5 to predict the thermo-mechanical behavior of the fuel cladding such as ballooning, burst and external/internal oxidation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

The Feasibility and the Benefits of the Advanced Nuclear Fuel Pellet Designs with Radially Varying Fuel Zoning and Burnable Poison Concentration (Final Summary Report)

As part of the work supported by the US Department of Energy (DOE) Office of Nuclear Energy (NE) Gateway for Accelerated Innovation in Nuclear (GAIN) FY 2021 Voucher, Exelon Generation, now Constellation, and Oak Ridge National Laboratory (ORNL) entered into a cooperative research and development agreement (CRADA) to evaluate and assess the feasibility and impact of various conceptual advanced nuclear fuel pellet designs (ANFPDs). The objective of this project was to perform modeling and simulation and analyses using the advanced modeling and simulation capabilities of VERA/BISON, developed by DOE, to determine the viability and benefits of numerous advanced nuclear fuel pellet design concepts in terms of fuel cycle costs, operational safety, and margin improvement. Whereas the detailed coupled neutronic and thermal hydraulic analyses performed using VERA provided in-depth knowledge in terms of fuel cycle performance, the detailed VERA results were used in subsequent fuel performance analyses using BISON. These subsequent analyses focused on several key fuel performance criteria, such as peak fuel centerline temperature (FCT), fission gas release (FGR), gap closure, plenum pressure, and cladding hoop stress. These key fuel performance criteria were analyzed for some of the conceptual fuel designs and were compared with the results obtained for UO 2 fuel. The results provided detailed information to enable better understanding of the performance of the fuel types analyzed. Understanding the advantage of loading these conceptual fuel designs into the core is important not only to Constellation but also to the entire light-water reactor (LWR) fleet in the United States.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

Predictive modeling and simulation of the performance of a nuclear reactor and its fuel is a challenging task because of the large number of coupled physical phenomena that must be addressed. In addition to this intrinsic complexity, model uncertainty must be accounted for in any analysis if the model will be used to facilitate design or operational decisions which may impact safety and performance. Rigorous, structured uncertainty analyses are performed by first characterizing the model's input uncertainties and, by propagating the uncertainty through the model in order to estimate the output's uncertainty. This paper presents the state of advancement of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the CFD code Nek5000 using the Dakota package. Nek5000 is a high order spectral element computational fluid dynamics (CFD) code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (RANS) simulations. The Dakota package developed at Sandia National Laboratory can be integrated with many scientific and engineering codes to facilitate efficient, effective uncertainty quantification (UQ) and sensitivity analyses. Dakota is designed to interface with external packages/codes such as Nek5000 through a script-driven interface called fork. Such interface provides maximum flexibility to the users for pre- and post-processing tasks preceding and following each function evaluation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development of Macroscale Models of UO 2 Fuel Sintering and Densification using Multiscale Modeling and Simulation

Light water reactor fuel pellets are fabricated using sintering to final densities of 95% or greater. During reactor operation, the porosity remaining in the fuel after fabrication decreases further due to irradiation-assisted densification. While empirical models have been developed to describe this densification process, a mechanistic model is needed as part of the ongoing work by the NEAMS program to develop a more predictive fuel performance code. In this work we will develop a phase field model of sintering of UO 2 in the MARMOT code, and validate it by comparing to published sintering data. We will then add the capability to capture irradiation effects into the model, and use it to develop a mechanistic model of densification that will go into the BISON code and add another essential piece to the microstructure-based materials models. The final step will be to add the effects of applied fields, to model field-assisted sintering of UO 2 . The results of the phase field model will be validated by comparing to data from field-assisted sintering. Tasks over three years: 1. Develop a sintering model for UO 2 in MARMOT 2. Expand model to account for irradiation effects 3. Develop a mechanistic macroscale model of densification for BISON

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗