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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Preliminary Analysis of SiC BWR Channel Box Performance under Normal Operation

SiC-SiC composites are being considered for applications in the core components, including BWR channel box and fuel rod cladding, of light water reactors to improve accident tolerance. In the extreme nuclear reactor environment, core components like the BWR channel box will be exposed to neutron damage and a corrosive environment. To ensure reliable and safe operation of a SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one cycle of irradiation. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions have been calculated using the neutronics and thermal-hydraulics codes Serpent2 and COBRA-TF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5mm. The channel box bowing behavior is time dependent, and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. However, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Deformation analysis of SiC-SiC channel box for BWR applications

Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composites are being considered as components in light water reactor cores to improve accident tolerance, including channel boxes and fuel cladding. In the nuclear reactor environment, core components like a channel box will be exposed to neutron and other radiation damage and temperature gradients. To ensure reliable and safe operation of a SiC-SiC channel box, it is important to assess its deformation behavior under in-reactor conditions including the expected neutron flux and temperature distributions. In particular, this work has evaluated the effect of non-uniform dimensional changes caused by spatially varying neutron flux and temperatures on the deformation behavior of the channel box over the course of one year. These analyses have been performed using the fuel performance modeling code BISON and the commercial finite element analysis code Abaqus, based on fast flux and temperature boundary conditions that have been calculated using the neutronics and thermal-hydraulics codes Serpent and CTF, respectively. The dependence of dimensions and thermophysical properties on fast flux and temperature has been incorporated into the material models. These initial results indicate significant bowing of the channel box with a lateral displacement greater than 6.5 mm. The channel box bowing behavior is time dependent and driven by the temperature dependence of the SiC irradiation-induced swelling and the neutron flux/fluence gradients. The bowing behavior gradually recovers during the course of the operating cycle as the swelling of the SiC-SiC material saturates. Furthermore, the bending relaxation due to temperature gradients does not fully recover and residual bending remains after the swelling saturates in the entire channel box.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

High Temperature Steam Oxidation Testing of Candidate Accident Tolerant Fuel Cladding Materials

The Fuel Cycle Research and Development (FCRD) program’s Advanced Fuels Campaign has initiated a multifold effort aimed at facilitating development of accident tolerant fuels in order to overcome the inherent shortcomings of light water reactor (LWR) fuels when exposed to beyond design basis accident conditions. The campaign has invested in development of experimental infrastructure within the Department of Energy complex capable of chronicling the performance of a wide range of concepts under prototypic accident conditions. This report summarizes progress made at Oak Ridge National Laboratory (ORNL) and Los Alamos National Laboratory (LANL) in FY13 toward these goals. Alternative fuel cladding materials to Zircaloy for accident tolerance and a significantly extended safety margin requires oxidation resistance to steam or steam-H 2 environments at ≥1200°C for short times. At ORNL, prior work focused attention on SiC, FeCr and FeCrAl as the most promising candidates for further development. Also, it was observed that elevated pressure and H2 additions had minor effects on alloy steam oxidation resistance, thus, 1 bar steam was adequate for screening potential candidates. Commercial Fe-20Cr-5Al alloys remain protective up to 1475°C in steam and CVD SiC up to 1700°C in steam. Alloy development has focused on Fe-Cr-Mn-Si-Y and Fe-Cr-Al-Y alloys with the aluminaforming alloys showing more promise. At 1200°C, ferritic binary Fe-Cr alloys required ≥25% Cr to be protective for this application. With minor alloy additions to Fe-Cr, more than 20%Cr was still required, which makes the alloy susceptible to α’ embrittlement. Based on current results, a Fe-15Cr-5Al-Y composition was selected for initial tube fabrication and welding for irradiation experiments in FY14. Evaluations of chemical vapor deposited (CVD) SiC were conducted up to 1700°C in steam. The reaction of H 2 O with the alumina reaction tube at 1700°C resulted in Al(OH) 3 formation, which reacted with the specimen to form a liquid reaction product. This behavior was not observed at lower temperatures where gas velocity and H 2 O content showed typical effects on the reaction rate. For LANL, the capabilities for oxidation testing as well as exploration of a methodology for measurement of hydrogen production of samples during oxidation under water vapor atmospheres is discussed. Results obtained for available commercial alloys are summarized, and data highlighting the performance of molybdenum, a recently proposed cladding material, are presented. Finally, leveraging of these techniques in conjunction with current and companion FCRD programs is discussed with respect to work in FY14.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Advanced synthesis for enhanced accident tolerant of LWR cladding materials

Advanced materials synthesis is utilized at Argonne National Laboratory (ANL) with the intended goal of improving performance of light water reactor (LWR) claddings under normal operation and accident conditions. The advanced synthesis utilizes atomic layer deposition (ALD) techniques to infiltrate SiC wires wrapped around ZIRLO tubes with SiC deposits, to form a SiC-SiC composite on the ZIRLO tubes. This paper reports on the effort to deposit SiC using ALD. The deposition was successful in infiltrating deep into the wires structure, reaching the surface of the ZIRLO tubes, as indicated through different characterization techniques. Because of the slow deposition rate of SiC, only a limited amount of SiC was deposited (a few hundreds of nm of SiC thick layers). This was followed by a few micrometers of ALD deposition of zircon (ZrSiO{sub 4}.) on top of the SiC deposition. The additional deposition was adequate to hold the wires together around the tube. Work is ongoing to increase the SiC ALD deposition rate, and infiltrate the larger pores within the wires with other SiC deposition methods such as low temperature chemical vapor deposition (CVD). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Overview of recent studies related to lead-alloy-cooled fast reactors

The recent progress of the studies related to lead alloy-cooled reactors (LFR) and the accelerator driven system (ADS) is summarized. The compatibility of materials with lead alloys has been clarified under steady and transient temperature conditions. Higher Cr content, Si and Al addition and Al-Fe alloy-coating improved the corrosion resistance of steels. The Al-Fe alloy-coated steel was not corroded even under high temperature transient conditions. The ceramics of SiC and Si{sub 3}N{sub 4} are expected to be used as cladding material for high temperature LFR. For the analytical consideration of mass transport in lead alloy circuit, the diffusion coefficient of Ni was measured using the capillary methods. A new bubble visualization method in LBE using gamma-ray radiography was developed. The thermal interaction of lead-bismuth eutectic (LBE) and lead droplets with sub-cooled water, and the behaviors of droplet fragmentation were investigated, and the visualization of volatile liquids in high temperature liquids was achieved.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Investigating Grain Structure and Microcracking in SiCf-SiCm Composites Using 4D-STEM

Silicon carbide (SiC) fiber-reinforced SiC matrix composites (SiCf-SiCm) are promising materials for accident-tolerant fuel (ATF) cladding in light water reactors. This paper, using four-dimensional scanning transmission electron microscopy (4D-STEM), studied the microstructure of SiCf-SiCm at nanoscale and provided grain statistics at the CVI/fiber region. Here, the results reveal that CVI region mainly contain irregular sub-micron grains (115 nm to 1800 nm) with a preferred orientation, while the fiber region has nano equiaxed grains (21 nm to 78 nm) with random orientations. Centered dark field imaging over the fiber regions verified the grain size measurement by 4D-STEM. Through 4D-STEM virtual dark field imaging and orientation mapping, intragranular propagation was identified as the fracture mechanism for a microcrack observed within the CVI region. The amorphous pyrolytic carbon layer was also shown to effectively arrest the microcrack.

4D-STEM↗
DOE OSTI2016

Mechanical and Thermal Properties of Irradiated SiC/SiC Composites: Bridge Continuum Models With Atomic-level Simulations

Due to their excellent high-temperature fracture, creep, corrosion, and thermal shock resistance in addition to their high thermo-chemical stability, reduced activation, and low radiation- induced after heat, silicon carbide-fiber/silicon carbide-matrix (SiC/SiC) composites are being developed as structural materials for first wall and breeder blanket applications in advanced fusion power plants. In such applications these materials will be subjected to a high heat flux that imposes extremely stringent requirements on the minimum thermal conductivity needed. Also, SiC offers a number of potential applications for both electron devices and structural components as an inert matrix for the transmutation of plutonium and as cladding material for gas-cooled fission reactor. There has been a rapid growth of activity in the experimental study and the modeling of energetic-beam interactions, such as those encountered in ion-beam processing or in irradiation environments, on the structure and properties of SiC. During the last decade, understanding the basics of ion-solid interaction and irradiation damage has led to significant developments in state-of-the-art atomic-level, meso-scale simulations and continuum models of mechanical and thermal property prediction, and these different scales studies have dramatically advanced the understanding of defect and defect processes in a number of materials, ranging from metals to semiconductors to insulators. Recent progress in multi-scale simulations of fundamental ion-solid interactions, defect generation and defect evolution in SiC, as well as its property degradation under irradiation, is reviewed, which will provide significant insights into the radiation resistance behavior of SiC and SiC/SiC composites. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2012

Application of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in PWRs

Presented is an investigation of the utilization of Tristructural-Isotropic (TRISO) particle-based fuel designs for the recycling of transuranic (TRU) wastes in typical Westinghouse four-loop pressurized water reactors (PWRs). Though numerous studies have evaluated the recycling of TRU in light water reactors (LWRs), this work differentiates itself by employing TRU-loaded TRISO particles embedded within a SiC matrix and formed into pellets that can be loaded into standard 17 x 17 fuel element cladding. This approach provides the capability of TRU recycling and, by virtue of the TRISO particle design, will allow for greater burnup (i.e., removal of the need for UO2 mixing) and improved fuel reliability. In this study, a variety of assembly layouts and core loading patterns were analyzed to demonstrate the feasibility of TRU-loaded TRISO fuel. The assembly and core design herein reported are a work in progress, so they still require some fine-tuning to further flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, TRU-loaded TRISO fuel and core designs that are capable of balancing TRU production and destruction can be designed within the standard constraints for thermal and reactivity performance in PWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

A Standardized Analysis Process Using Digital Image Correlation to Calculate In Situ Cladding Strain from Modified Burst Tests for Fuel Performance Code Validation

Historical data collection on nuclear fuel cladding materials has focused on generating a statistically significant amount of data to assess the material and its failure behavior. Furthermore, data generated to support material model and failure criteria development were previously posttest evaluations, so a large number of tests was required to gain new understanding. A way to expedite this process is to develop techniques capable of generating large, high-fidelity data sets from a single test with lower uncertainty or quantified uncertainty. One such example of this approach is Oak Ridge National Laboratory’s use of modified burst tests (MBTs) to analyze the mechanical behavior and failure conditions of cladding during a simulated reactivity-initiated accident (RIA). Each test incorporates digital image correlation (DIC) analysis techniques that are used to assess the accumulated strain in situ, as well as eventual cladding failure. This work has been fruitful in defining strain-to-failure conditions for materials like silicon carbide (SiC) fiber–reinforced/SiC matrix composite tubes (SiC/SiC), iron-chromium-aluminum (FeCrAl) alloy tubes, and chromium-coated Zircaloy-4 tubes. However, there are numerous DIC software available, including open-source and proprietary software. The different DIC software use various algorithms to process images and calculate displacement values. Using these different software and algorithms can lead to varying results, and perhaps larger-than-expected uncertainties. In the present study, previously published MBT data encompassing a variety of test conditions were reanalyzed with two different DIC software to assess the variance in the calculated strain results. The data consisted of SiC/SiC, FeCrAl, and chromium-coated Zircaloy-4 tubes. Plots of the calculated strains during the transient revealed good agreement between the two DIC software. The average root-mean-square errors between the two software was 0.20% strain, which is slightly larger than a previously reported error value for these tests. In conclusion, this variance in results is low enough that this analysis method can be used for code validation.

Reactivity-initiated accident↗
DOE OSTI2019

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis and Final Technical Report Deliverable for the Westinghouse Accident Tolerant Fuel Program (Rev. 2)

The beyond design basis accident at Fukushima Daiichi in March 2011 and the Three Mile Island accident in 1979 imply that severe nuclear accidents may occur at somewhat higher frequencies than previously predicted, and that the financial liabilities of such accidents can cripple a utility. While current fuel is more than adequate for design basis accidents, the potential for harm from beyond design basis accidents (BDBA) has been a driver for Westinghouse development efforts since 2003. After Fukushima, other fuel vendors also began pursuing new fuel materials that can provide significant increases in the time, enabling reactor operators to respond to unforeseen events before significant releases of fuel materials and fission products occur. In order to be commercially successful, accident tolerant fuel (ATF) products must also provide significant fuel cost savings, operating cost improvements as well as safety improvements. This report documents the engineering, design and testing that occurred from October 1, 2016, through September 30, 2018, by the Westinghouse team. In addition, it includes: 1. The results of modeling and simulation work completed to date that supports technical feasibility of the ATF concept in meeting the goals of the enhanced ATF program including a core level analysis predicting how an existing LWR will perform when loaded with ATF. 2. Regulatory Analysis showing path to LTA/LTR acceptance. 3. Business Analysis showing commercial viability of the ATF concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Micro-mechanical evaluation of SiC-SiC composite interphase properties and debond mechanisms

SiC-SiC composites exhibit exceptional high temperature strength and oxidation properties making them an advantageous choice for accident tolerant nuclear fuel cladding. In the present work, small scale mechanical testing along with AFM and TEM analysis were employed to evaluate PyC interphase properties that play a key role in the overall mechanical behavior of the composite. The Mohr-Coulomb formulation allowed for the extraction of the internal friction coefficient and debonding shear strength as a function of the PyC layer thickness, an additional parameter. Here, these results have led to re-evaluation of the Mohr-Coulomb failure criterion and adjustment via a new phenomenological equation.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Progress on ATF development of SiC for LWR

Since 2011, Toshiba has been participating in a study whose objective is the commercial utilization of SiC instead of Zircaloy as LWR fuel material. During the period (Phase-1) from 2012 to 2015, feasibility studies on material design and fabrication were performed. This paper reports the principal results of a comprehensive feasibility study on SiC involving experiments and trial fabrication in terms of material properties, core and fuel design, and analytical simulation for safety analysis. The principal results are summarized as follows. First, the fabrication trials for the reduced scales of channel box and cladding have been successfully performed. Secondly, with regard to fuel rod performance analysis, hoop stress during normal operation may be a potential issue in fuel mechanical design as it is relatively high. Thirdly, with regard to safety analysis, the hydrogen generation of SiC fuel tends to decrease significantly compared with that of Zircaloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

High Level Gap Analysis for Accident Tolerant and Advanced Fuels for Storage and Transportation

This initial gap analysis considers proposed accident tolerant fuel (ATF) options currently being irradiated in commercial reactors, since these are most likely for future batch implementation. Also, advanced fuel (AF) options that may be likely for use in advanced reactors are considered. The cladding technologies considered were chromium-coated zirconium-based alloys, FeCrAl, and both monolithic and matrix composite Silicide carbide (SiC). The fuel technologies considered were chromium-doped uranium dioxide fuel, uranium alloys, uranium nitride, and uranium silicide. Numerous national labs, industry, and countries are performing significant testing and modeling on these proposed technologies to establish performance, but at this time none of the prototypes being irradiated have achieved end-of-life (EOL) burnup. There are some testing results after one burnup cycle to verify in-reactor performance, but little data beyond that. As the ATF prototypes acquire more burnup, data will be produced that is relevant to storage and transportation. The DOE:NE Spent Fuel and Waste Science and Technology (SWFST) Storage and Transportation (ST) Control Account will evaluate the performance data as it becomes available for application to the identified gaps for ST.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

The response of accident tolerant fuel cladding to LOCA burst testing: A comparative study of leading concepts

Accident tolerant fuel claddings seek to improve safety margins during loss-of-coolant accident (LOCA) scenarios by reducing cladding steam oxidation rates, hydrogen production, delaying or mitigating cladding burst, and reducing cladding deformation. After rapid and extensive progress in recent years, Cr coated Zr-based alloys, ferritic FeCrAl, and SiC/SiC ceramic matrix composites (CMCs) have emerged as the most promising cladding concepts, but there are few studies comparing the accident tolerance of all three concepts simultaneously. Here, in this study, LOCA burst testing has been conducted on Cr high impulse power magnetron sputtered Zry-4 (Cr/Zry-4), C26M (FeCrAl), and SiC/SiC SiGA (CMC). Experimental observations indicate Cr/Zry-4 burst at higher temperatures than Zry-4 however the effectiveness of the coating on mitigating ZrO 2 formation via Cr 2 O 3 formation depended on physical proximity to the burst opening. No dramatic improvement in diametrical strain nor opening geometry was observed. C26M burst at higher temperatures than Zry-4 at lower pressures, exhibiting small amounts of strain and smaller openings, and formed slow-growing Al 2 O 3 regardless of location. The CMCs did not burst or show any macroscopic signs of deformation, all the way up to an internal cladding pressure of 15.2 MPa. High temperature expanding Nb plug testing indicated that CMC room temperature strength may persist up to 1650°C, with an approximately 40% loss of strength at 1900°C. CMC steam oxidation was also compared to reference chemically vapor deposited SiC up to 1700°C, and CMCs were found to oxidize at slightly accelerated rates that were still orders of magnitude lower than bare Zry-4. Projected H 2 generation during LOCA burst testing showed that when accounting for cladding burst exposing inner diameters to steam, Cr coated Zry-4 reduced H 2 generation by a factor of two compared with bare Zry-4, while both C26M and SiC/SiC reduced H 2 generation by orders of magnitude relative to bare Zry-4.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Stress Analysis and Structural Optimization of a Three-Layer Composite Cladding Tube Under Thermo-Mechanical Loads

A general solution for the stress and strain fields in a three-layer composite tube subjected to internal and external pressures and temperature changes is first derived using thermo-elasticity. The material in each layer is treated as orthotropic, and the composite tube is regarded to be in a generalized plane strain state. A three-layer ZRY4-SiCf/SiCSiC composite cladding tube under a combined pressure and thermal loading is then analyzed and optimized by applying the general solution. The effects of temperature changes, applied pressures, and layer thickness on the mechanical behavior of the tube are quantitatively studied. The von Mises failure criterion for isotropic materials and the Tsai-Wu's failure theory for composites are used, respectively, to predict the failure behavior of the monolithic ZRY4 (i.e., Zircaloy-4) inner layer and SiC outer layer and the composite SiCf/SiC core layer of the three-layer tube. The numerical results reveal that the maximum radial and circumferential stresses in each layer always occur on the bonding surfaces. By adjusting the thickness of each layer, the effective stress in the three-layer cladding tube under the prescribed thermal-mechanical loading can be changed, thereby making it possible to optimally design the cladding tube.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Design and Thermal Analysis for Irradiation of Silicon Carbide Joint Specimens in the High Flux Isotope Reactor

This report provides a summary of the irradiation vehicle design and thermal analysis of SiC joint specimens planned for irradiation in the flux trap of the High Flux Isotope Reactor (HFIR). Two different capsule designs will be used to accommodate the two different specimen geometries: a small torsion joint specimen geometry to measure mechanical and thermal properties, and a tube and end plug joint representative cladding geometry to demonstrate strength and hermeticity retention. The capsule designs, with target temperatures of 350° and 750°C, will accommodate either sixteen torsion joint specimens or one tube end plug joint specimen. Three joint variations will be studied in each capsule design: a hybrid SiC (preceramic polymer with chemical vapor deposition (CVD) SiC), a transient eutectic phase (TEP) process, and an oxide process.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗