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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 1,135 records · Page 63
All sourcesNASA NTRSDOE OSTI
NASA NTRS2007

High Fidelity, Fuel-Like Thermal Simulators for Non-Nuclear Testing: Analysis and Initial Test Results

Non-nuclear testing can be a valuable tool in the development of a space nuclear power system, providing system characterization data and allowing one to work through various fabrication, assembly and integration issues without the cost and time associated with a full ground nuclear test. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Testing with non-optimized heater elements allows one to assess thermal, heat transfer. and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. High fidelity thermal simulators that match both the static and the dynamic fuel pin performance that would be observed in an operating, fueled nuclear reactor can vastly increase the value of non-nuclear test results. With optimized simulators, the integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and fueled nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics and assess potential design improvements at relatively small fiscal investment. Initial conceptual thermal simulator designs are determined by simple one-dimensional analysis at a single axial location and at steady state conditions; feasible concepts are then input into a detailed three-dimensional model for comparison to expected fuel pin performance. Static and dynamic fuel pin performance for a proposed reactor design is determined using SINDA/FLUINT thermal analysis software, and comparison is made between the expected nuclear performance and the performance of conceptual thermal simulator designs. Through a series of iterative analyses, a conceptual high fidelity design is developed: this is followed by engineering design, fabrication, and testing to validate the overall design process. Test results presented in this paper correspond to a "first cut" simulator design for a potential liquid metal (NaK) cooled reactor design that could be applied for Lunar surface power. Proposed refinements to this simulator design are also presented.

Bragg-Sitton, Shannon M.↗
DOE OSTI2016

Computer Based Procedures for Field Workers - FY16 Research Activities

The Computer-Based Procedure (CBP) research effort is a part of the Light-Water Reactor Sustainability (LWRS) Program, which provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants. One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. One area that could yield tremendous savings in increased efficiency and safety is in improving procedure use. A CBP provides the opportunity to incorporate context-driven job aids, such as drawings, photos, and just-in-time training. The presentation of information in CBPs can be much more flexible and tailored to the task, actual plant condition, and operation mode. The dynamic presentation of the procedure will guide the user down the path of relevant steps, thus minimizing time spent by the field worker to evaluate plant conditions and decisions related to the applicability of each step. This dynamic presentation of the procedure also minimizes the risk of conducting steps out of order and/or incorrectly assessed applicability of steps. This report provides a summary of the main research activities conducted in the Computer-Based Procedures for Field Workers effort since 2012. The main focus of the report is on the research activities conducted in fiscal year 2016. The activities discussed are the Nuclear Electronic Work Packages – Enterprise Requirements initiative, the development of a design guidance for CBPs (which compiles all insights gained through the years of CBP research), the facilitation of vendor studies at the Idaho National Laboratory (INL) Advanced Test Reactor (ATR), a pilot study for how to enhance the plant design modification work process, the collection of feedback from a field evaluation study at Plant Vogtle, and path forward to commercialize INL’s CBP system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

$\mathrm{EBR-II MOX}$ Fuel Characterization Enabling ARES Phase I Testing

Pretransient characterization was performed for the Experimental Breeder Reactor II (EBR-II) mixed-oxide (MOX) fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and the United States. Continued collaboration under the Advanced Reactor Experiments for Sodium Fast Reactor Fuels project will investigate the transient performance of these rods in the Transient Reactor Test facility at Idaho National Laboratory in the MOXTOP-THOR experiment. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of 14.3 at. % (~134.4 GWd/t) in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success of this effort. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination. The methodology development and analysis activities utilized we report enable current experiment design work and provide the pathway through which measured data of this type can be further evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Damage tolerance in the graphite cores of UK power reactors and implications for new build

The United Kingdom has been operating graphite-moderated gas-cooled commercial nuclear power reactors for more than 60 years. Both Generation I Magnox reactors and Generation II Advanced Gas-cooled Reactors have operated beyond their nominal design lifetimes and the safety cases for continued operation have had to address the issue of damage tolerance of the graphite cores. This review describes the designs of the reactors and the methodologies employed for their assessment. Emerging issues arising from extended operation and ageing of the cores are discussed together with the evolution of assessment methodologies to address predicted and observed damage. The UK nuclear regulator perspective is summarised and some conclusions are drawn on the implications for new build.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗
DOE OSTI2021

Development of a Pulsed Slowing-Down-Time Benchmark of Neutron Thermalization in Graphite

Graphite is a classic neutronic material that has been used as both a reactor reflector and moderator in various nuclear reactor systems. The ability to accurately predict the slowing down and thermalization of neutrons in graphite can have significant implications on the safety and operation of such reactor systems. In reactors, the neutron thermalization process is quantified using the thermal scattering law (TSL) and related cross sections for a given moderator. An ideal approach to assess the validity of TSL data is using benchmark measurement based on the pulsed Slowing-Down-Time technique and its comparison with the appropriate graphite nuclear library. In this work, experimental measurement and computational Monte Carlo simulations were performed to benchmark the slowing down characteristics and thermalization of neutrons in nuclear (reactor-grade) graphite. Given the density of graphite, various graphite libraries were selected for the benchmark analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

One-Step Ahead Prediction of Thermal Mixing Tee Sensors with Long Short Term Memory (LSTM) Neural Networks

High-temperature advanced reactors under development, such as sodium fast reactors (SFR) and molten salt cooled reactors (MSCR), are expected to offer lower levelized cost of energy (LCOE) compared to existing light water reactor (LWR’s). In the existing light water reactors (LWR’s), operation and maintenance (O&M) expenses constitute the largest fraction of the total operating cost. Some of the O&M costs are related maintenance of sensors which can fail due to exposure to harsh environment in a reactor. The O&M costs of Advanced Reactor (AR)’s are expected to constitute a significant fraction of the total cost as well, because of high temperature and radiation level in AR are likely to cause material fatigue and premature failure of sensors and components. The O&M costs in AR’s could be reduced through integration of advanced informatics of performance-related sensors into a digital twin designed for reactor monitoring. For example, machine learning (ML) could be employed for real-time validation and correction of performance-related sensors, and reducing the number of performance-related physical sensor units through virtual sensing. As part of the effort, we investigate real-time validation of thermal hydraulic sensors through one-step ahead forecasting of sensor values using long short-term memory (LSTM) recurrent neural networks (RNN). The sensors are installed in a flow loop containing a thermal mixing tee, which is a common experimental model to study thermal fatigue in a thermal hydraulic loop. In addition, nonlinear transients generated in a thermal mixing tee constitute a good challenge data set for training and validation of ML algorithms. Sensors in this study include thermocouples, flow meters, and optical fibers for distributed temperature sensing. In one experiment, measurement data sets were obtained for a loop was filled with water, and in another experiment, measurements were performed on a loop filled with liquid metal Galinstan. We have also conducted preliminary investigation of one-step ahead prediction of fiber optics-based distributed temperature sensing with LSTM networks. In predicting fiber-based temperature measurements, we treated each gauge pitch of the fiber as an independent sensor. Accuracy of one-step ahead forecasting was estimated by calculating root mean square error (RMSE) for the test segment of time series of each sensor. RMSE’s for temperature sensors in water loop were, for the most part, lower than for the same sensors in Galinstan loop. The RMSE’s for flow meters were similar for both loops. The RMSE’s for distributed temperature measured with the fiber optic sensor were similar to those of the point sensors. Results of this study demonstrated the capability of LSTM one-step ahead forecasting with RMSE comparable to uncertainty in sensor measurements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Kinetics of Hydrogen Desorption from Yttrium Hydride: The Development of Material Characterization Methods

Yttrium hydride possesses attractive neutronic and thermal properties for moderator applications as a high-temperature moderator material in advanced thermal neutron spectrum reactors that require small core volumes. For safe operation of the nuclear reactors, it is critical to understand the kinetics of H desorption from yttrium hydride at elevated temperatures. In previous studies, the H desorption flux as a function of temperature was measured by thermal desorption spectroscopy. To fully understand the H desorption kinetics, the amount of H within the specimen must be locally evaluated, especially if this uniformity strongly affects the thermomechanical properties and irradiation resistance of the bulk material.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Advances on Bacterial and Fungal Biofilms for the Production of Added-Value Compounds

In recent years, abundant research has been performed on biofilms for the production of compounds with biotechnological and industrial relevance. The use of biofilm platforms has been seen as a compelling approach to producing fine and bulk chemicals such as organic acids, alcohols, and solvents. However, the production of recombinant proteins using this system is still scarce. Biofilm reactors are known to have higher biomass density, operational stability, and potential for long-term operation than suspended cell reactors. In addition, there is an increasing demand to harness industrial and agricultural wastes and biorefinery residues to improve process sustainability and reduce production costs. The synthesis of recombinant proteins and other high-value compounds is mainly achieved using suspended cultures of bacteria, yeasts, and fungi. This review discusses the use of biofilm reactors for the production of recombinant proteins and other added-value compounds using bacteria and fungi.

Carvalho, Fábio M.↗
DOE OSTI2021

Nuclear reactors including heat exchangers and heat pipes extending from a core of the nuclear reactor into the heat exchanger and related methods

A nuclear reactor including a reactor core comprising a plurality of fuel materials and a plurality of heat pipes. The nuclear reactor further includes a heat exchanger coupled to the reactor core defining a flow path in an open volume including at least two heat pipes of the plurality of heat pipes. Methods of operating a nuclear reactor include passing fluid through an open volume in a heat exchanger including at least two heat pipes extending from a reactor core.

McKellar, Michael G.↗
DOE OSTI2020

Start-up of Silo 130 Waste Retrieval at La Hague: first Step Towards Reducing Legacy Inventory - 20020

On June 25, 2019, the first grapple of waste and debris was lifted from Silo 130, one of the most challenging legacy waste storage silos at La Hague. This was the achievement of more than ten years of studies and efforts focused on solving some very unique technical and safety challenges, and on designing an innovative, efficient solution. The waste is from de-cladding materials from past Gas-Cooled Reactor fuel reprocessing operations. It is an unsorted mixture of magnesium cladding pieces, graphite debris, and various other metallic waste types. These materials can be chemically reactive and therefore need specific dispositions for handling and storing. The building is more than 40 years old and does not meet modern standards. French safety authorities have required the silo to be emptied. After about three years of operation, the bulk of this waste will be removed and placed in safe storage, pending the selection of a final conditioning solution. Removal of bottom debris will then be possible. When finished, this operation will bring a conclusion to a significant legacy-related hazard on the site. The silo will then be available for final cleaning and de-activation. This paper will describe the technical, safety, and economical challenges, and the various decisions and optimizations that led to the definition of the process and technology to be implemented. Several innovative features were developed and implemented, in all sections of the project: project organization, progressive approach for the waste disposition strategy, light building design, customized and robust retrieval technology, processing technologies involving the most modern robotic and sorting techniques, storage method... Several of these first-of-a-kind innovative features have been developed and qualified specifically for this application, using the detailed qualification procedure specified for all La Hague projects. The facility was built in less than five years, overcoming various difficulties along the line. Commissioning is now finished with the combined efforts of the project, engineering, start-up, and operating teams assisted by dedicated suppliers. The facility is now ramping-up. The paper will also report on the commissioning and start-up of the facility and describe the first operational results. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Examining Two-Phase Flow with Optical Sensors

In lieu of the events that occurred at Fukushima in 2011, passive safety shutdowns for nuclear reactors has become a focus area for recent research. At Argonne National Laboratory this research has been simulated using air as a means of cooling down a experimental reactor core. This operates on the principals of natural convection and can be further studied in the report devised by Lisowski et al. in 2014. The air reactor cavity cooling system (RCCS) at Argonne National Laboratory is currently being transitioned into a water cooled system for passive safety reactor shutdown systems. In this system eight water pipes (1.5'' schedule 160 stainless steel) will transfer heat by means of natural convection. However, the primary network pipe will be 4 inch schedule 40 stainless steel pipe. The adiabatic chimney portion of this piping segment will be under analysis for void fraction. Based on the internal temperature of the water pipes, the characteristic regimes of two-phase flow that could exist inside of the pipe are an area of interest for experimental analysis. Determining which phase the water exists is instrumental to understanding the state of the processes inside of the water cooling pipe. Knowledge of these states allows for analysis of the efficiency of the RCCS as it shuts down. The state of two-phase flow can be measured by the analysis of the fluids void fraction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Simulation of the RPI Reactor Critical Facility Core Using High-Fidelity Neutronics Code PROTEUS

The PROTEUS code is a high-fidelity three-dimensional (3D) deterministic neutron transport code developed by ANL under the DOE Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. The code contains the SN2ND (second order discrete ordinates method) and two MOC transport solvers, MOCFE and MOCEX (first order methods of characteristics), based on unstructured finite element meshes which allow users to model complex or unconventional geometry reactor problems. Among the two MOC solvers which normally perform better than SN for heterogeneous geometry problems, MOCEX is based on a rigorous formulation using the two-dimensional (2D) MOC radially and the discontinuous Galerkin finite element method axially to overcome the drawbacks of MOCFE with the full 3D MOC representation which requires enormous memory and computational efforts. The Reactor Critical Facility (RCF), an open tank research reactor at the Rensselaer Polytechnic Institute (RPI), became operational in 1956 under the American Locomotive company and has been operating using low-enriched uranium fuel pins, which makes it the only university research facility to use fuel rods similar to operating commercial light water reactors (LWRs). This paper presents modeling and simulation results of the RCF using PROTEUS for code verification and validation which also supports an ongoing NEUP project in which a set of benchmark validation experiments for multiphysics coupling is being developed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

HFE in the design of the socio-technical system of the control room of a new generation reactor: issues faced during multistage validation - 126

This paper deals with the contribution of Human Factors and Ergonomics (HF and E) experts to the design of the control room (CR) of a new generation reactor (NGR), our view of the multistage validation (MSV) process and the issues faced during it. Our approach to MSV was to articulate two points of view: the designers' point of view (test and explore various design options) and the regulatory authorities' point of view (validate the performance of the socio-technical system of the CR). To perform this articulation, both tests and validation shared the same method and goals. The main issues faced during this process were to (1) get a coherent and integrated version of the full-scope simulator when the design of the socio-technical system of the CR was still in progress, (2) determine when to stop MSVs, and (3) define a criterion for the sampling of operational conditions. The first issue led us to the conclusion that specific project milestones synchronising the design of every part of the CR have to be planned early in the project in order to mitigate the risk of delaying WSTs (Whole System Tests). Regarding the second issue, we advocate that no further WSTs are required when the last modifications brought to the CR design have no significant impact on crews' activities. Finally, concerning the last issue, we think that performing WSTs in all classes of situations a crew may face during operating the plant is necessary. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Sensitive Resources Assessment and Forest Analysis for the Proposed Versatile Test Reactor, Oak Ridge, Tennessee

The US Department of Energy’s (DOE’s) Oak Ridge National Laboratory (ORNL) is a leading institution in advanced materials, supercomputing, neutrons, and nuclear science. As a research laboratory managed by UT-Battelle, LLC for DOE, ORNL has national priorities in energy, security, and scientific discovery that necessitate facility improvements and expansions. DOE is also committed to environmental stewardship. The laboratory is located on the ~32,000-acre (~13,000-ha) Oak Ridge Reservation (ORR), much of which is categorized as a National Environmental Research Park (NERP) and a state Wildlife Management Area. DOE works with the Tennessee Wildlife Resources Agency (TWRA), Tennessee Department of Environment and Conservation (TDEC), US Fish and Wildlife Service (USFWS), US Department of Agriculture, and other agencies to serve as an effective steward of the ORR. Accordingly, project managers must conform to environmental regulations, agreements, and policies at the federal, state, and institutional levels. Per 40 CFR (Code of Federal Regulations) 1508.14, potential effects on research and science education also represent potential effects of federal actions on the NERP, and impacts on, e.g., deer harvest, must be considered on the Oak Ridge Wildlife Management Area when other aspects of the human environment are affected. The United States currently has no fast neutron testing capability to support advanced nuclear research and development. The proposed Versatile Test Reactor (VTR) will take advantage of current investments by the US government and private industry in nuclear reactors to expedite the design and construction process, using proven technology to create a world-class scientific infrastructure. The VTR will take advantage of fast neutrons provided by this proven technology, along with a capability to rapidly insert, conduct, and remove state-of-the-art experiments. An advantage of the VTR is that it can support future innovations in experimental capabilities without modifying the facility. The VTR will support progress in a variety of science and technology areas, including testing and qualification of advanced reactor fuels; testing and qualification of innovative structural materials; testing of innovative components and instruments; validation of advanced modeling and simulation tools; and versatility for future technical missions. Through proven technology, the VTR can take advantage of existing reactor designs and operating experience to reduce the risk, cost, and time for design and construction. The top available resources of DOE laboratories, industry, and universities will be used to expedite reactor design and construction toward developing the scientific infrastructure that affords a strong testing capability that can be sustained over many years. This report summarizes current knowledge of natural and cultural resources primarily within the VTR construction area. At the time of this report, the proposed VTR site design includes a construction area of ~150.4 acres (~69.9 ha), which contains an ~51.3-acre (~20.8-ha) operations area, located within forested natural areas of the ORR. The primary goal of the work presented here was to evaluate potential effects on sensitive resources that might result from development and construction activities associated with VTR. In addition to on-the-ground surveys during spring and summer 2020 by the ORNL Natural Resources Management Program and Aquatic Ecology Group staff, this report makes use of historical (pre-1995) and contemporary (1995 to present) data from additional confirmed sources (e.g., TDEC). Likewise, forest conditions were compiled from a 2011 forest inventory and supplemented with limited ground observations in 2020. The individuals who obtained and compiled the data presented here are familiar with and routinely assess sensitive resources on the ORR. Anyone who references this report must consider that the timing of surveys did not permit a complete delineation of the resources that will be affected. If the VTR project proceeds, additional surveys will be required to account for the seasonal patterns of various threatened and endangered species. Data deficiencies and potential resources that likely went undetected are indicated where possible. Accordingly, this report should facilitate more environmentally sound decisions during planning and 2 development of the VTR site, provide a foundation for further assessment of sensitive and cultural resources, and help project managers better address regulatory guidance and DOE policies on sustainable development in compliance with, for example, the US Endangered Species Act (ESA), Migratory Bird Treaty Act (MBTA), Tennessee Rare Plant Protection and Conservation Act of 1985, Tennessee Nongame and Endangered or Threatened Wildlife Species Conservation Act of 1974, several federal and state regulations regarding aquatic resource protection, and site-specific policies as outlined in various ORR management plans developed by ORNL and TWRA for DOE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Instrumentation Framework for Molten Salt Reactors

This report provides an overview of the specialized instrumentation necessary to operate and maintain liquid-fueled molten salt reactors (MSRs). The report does not include non-specialized industrial process control instrumentation, fissile material tracking instrumentation, or digital control/communication technologies. The report focuses on identifying MSR measurement technology gaps, and recommending research, development, and demonstration (RD&D) activities to close the identified gaps. To a large extent modern MSRs can rely on the instrumentation demonstrated during prior MSR development efforts. No fundamental technology gaps have been identified that would prevent operating or maintaining MSRs. However, improved instrumentation technology combined with the integration of modern process modeling and simulation could significantly improve the reliability and decrease the operating costs of future plants. The largest potential impact of new instrumentation technologies is in the integration of maintenance automation into reactor operations. Flow measurement has the largest technology hurdles remaining for to enable deployment of robust systems. Also, the technology for reliable, on-line fuel salt sample acquisition remains at relatively low technology readiness.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Ideal MHD stability and characteristics of edge localized modes on CFETR

Investigation on the equilibrium operation regime, its ideal magnetohydrodynamics (MHD) stability and edge localized modes (ELM) characteristics is performed for the China Fusion Engineering Test Reactor (CFETR). The CFETR operation regime study starts with a baseline scenario (R = 5.7 m, B T = 5 T) derived from multi-code integrated modeling, with key parameters ${{\beta }_{{\rm N}}},{{\beta }_{{\rm T}}},{{\beta }_{{\rm p}}}$ varied to build a systematic database. These parameters, under profile and pedestal constraints, provide the foundation for the engineering design. The long wavelength low-n global ideal MHD stability of the CFETR baseline scenario, including the wall stabilization effect, is evaluated by GATO. It is found that the low-n core modes are stable with a wall at r/a = 1.2. An investigation of intermediate wavelength ideal MHD modes (peeling ballooning modes) is also carried out by multi-code benchmarking, including GATO, ELITE, BOUT++ and NIMROD. A good agreement is achieved in predicting edge-localized instabilities. Nonlinear behavior of ELMs for the baseline scenario is simulated using BOUT++. A mix of grassy and type I ELMs is identified. When the size and magnetic field of CFETR are increased (R = 6.6 m, B T = 6 T), collisionality correspondingly increases and the instability is expected to shift to grassy ELMs.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

Report Summarizing Boiling Transition (Dryout) Testing of FeCrAl Cladding

Nuclear fuel and cladding in a light water reactor must maintain integrity under all normal operational and anticipated operational occurrences. In a Boiling Water Reactor (BWR) some Anticipated Operational Occurrences (AOOs) and Anticipated Transients Without SCRAM (ATWS) events have been known to result in cladding damage due to dryout with conventional Zircaloy-2 claddings. This work presents an assessment of commercially-fabricated, accident tolerant iron-chromium-aluminum (FeCrAl) alloy C26M cladding in such events with regard to the ballooning failure mode for unirradiated cladding In the severe accident test station (SATS), an experiment has been developed to gauge accident tolerance of fuel cladding materials under simulated cyclic dryout conditions. Thermal cycling from 300°C to 650°C was implemented on C26M and Zircaloy-2 tubes. The C26M tube did not fail after 54 cycles. The Zircaloy- 2 tube run under the same parameters as the C26M tube failed during the fourth cycle, consistent with expectations from the literature. Due to thermal gradients present in the furnace, both tubes experienced a maximum temperature at ~780°C. Due to differences in tube geometry, the hoop stress of the C26M tube was significantly higher than that of the Zircaloy-2 tube, 76 MPa at the maximum peak temperature compared to 55 MPa at burst, respectively. The cyclic dryout experiments show an increased performance of C26M over Zircaloy-2 under simulated conditions, an indication of C26M possessing superior high temperature oxidation resistance, mechanical properties and potentially resistance to fatigue.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗