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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. Here, a type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15—which tested ZIRLO cladding behavior—so that B136Y3 and ZIRLO cladding could be compared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 released significantly less H 2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455°C vs. 1880°C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO 2 spacers detrimentally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expansion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that FeCrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Autoclave grid-to-rod fretting wear evaluation of a candidate cladding coating for accident-tolerant fuel

In pressurized water reactors (PWRs), water flow induced vibrations cause contact and rubbing between the fuel rods and the supporting grid, a phenomenon known as Grid-to-Rod-Fretting (GTRF). GTRF may produce progressive wear damage on the fuel claddings leading to subsequent leakage of radioactive fission products. Various accident-tolerant fuel (ATF) concepts are being developed for higher resistance to the high temperature steam and one approach is to apply a cladding coating. Here, fretting wear behavior of a candidate Cr-coating was investigated using a unique bench-scale autoclave testing rig mimicking the environment in an industrial full-assembly PWR simulator. The contact was under a realistically low load (~0.5 N) lubricated by deionized water at a temperature of 204 °C under a pressure of 20-23 bars. Results demonstrated that the Cr-coating significantly improved the cladding's wear resistance when tested against a commercial ZIRLO grid with or without pre-oxidization. In addition, the Cr-coating also reduced wear on the non-oxidized ZIRLO grid but slightly increased the wear on the pre-oxidized grid.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Grid-to-rod fretting wear study of SiC/SiC composite accident-tolerant fuel claddings using an autoclave fretting bench test

Grid-to-rod-fretting (GTRF) in pressurized water reactors (PWRs) is known to cause wear and surface damage on the fuel claddings, potentially leading to radioactive leakage. One of the accident-tolerant fuel (ATF) concepts is to use advanced cladding materials that could withstand higher temperatures. Here, we investigated the wear behavior of candidate silicon carbide (SiC)-based composite claddings with different levels of surface finish in fretting against a commercial ZIRLO alloy grid using a unique bench-scale autoclave GTRF rig. The experiments mimicked the environment in an industrial full-assembly PWR simulator. Fretting tests were conducted with a realistic load (~0.5 N) in deionized water under a pressure of 20–23 bar at 204 °C for 100 h. While the SiC/SiC composite claddings showed significantly higher wear resistance than the commercial ZIRLO alloy cladding as expected, the smoother versions experienced surprisingly higher wear than the much softer counterface, ZIRLO grid. The wear mechanism of the SiC/SiC cladding was attributed to the SiC wear debris that was trapped at the fretting interface causing both 3-body and 2-body (embedded into the grid surface) abrasion of the cladding. Rougher SiC/SiC claddings had less material loss but caused more wear on the ZIRLO grid. Finally, pre-oxidized ZIRLO grid showed better compatibility with the SiC/SiC cladding to protect both the cladding and grid as a result of reduced wear debris trapping.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Post-irradiation examination of legacy high burnup fuel to support safety testing

Safety/transient testing to evaluate performance under off-normal conditions is an essential pillar for both the development of Accident Tolerant Fuels (ATF) and the optimization of fuel operation economics beyond current discharge burnups. Among other factors, the successful interpretation of the transient testing results relies upon the knowledge of the initial conditions of the test, including the characteristics of the fuel system under scrutiny. When testing pre-irradiated material, the assumptions that the fuel and the cladding still have the same properties as in the pre-irradiation stage is obviously wrong and could affect the results of the test. This is particularly true the more burnup accumulates in the fuel rod and irradiation progresses. The knowledge of the initial microstructure of both fuel and cladding allows a clearer interpretation of the subsequent transient testing results, provides validation of the physical phenomena underlying the model predictions and eliminates the uncertainties related to the limited knowledge of the sample status before the test. One example is the phenomenon of fine fragmentation that occurs in Light Water reactor (LWR) fuel. During a Loss of Coolant Accident (LOCA) or Reactivity Initiated Transient (RIA) the fuel can severely fragment. During LOCA, high burnup fuel tend to finely fragment, which has raised safety concerns due to the increased likelihood of dispersal of such small particles once the cladding has burst and due to the increased fission gas release. Therefore, efforts have been devoted to the assessment of a pulverization threshold that could determine the conditions under which fine fragmentation is predominant. However, the lack of information regarding the initial conditions of the fuel, and the connections between those conditions and the pre-transient irradiation history, have hindered the development of a fully mechanistic fragmentation and pulverization criterion. The empirical relationships rely on conservative estimations, due to the lack of information on critical material properties and characteristics. More generally, experimental evidence of the irradiation-induced modifications at microstructural scale are necessary to determine the behavior of the material at the macroscopic scale, with the latter being the one of technological interest. Significant progress has been made in the last two decades in the developments of analytical materials science techniques that can be applied to highly radioactive materials, such as high burnup fuels. The availability of new techniques and the improvement of existing ones has enabled investigations previously not possible that can deepen the understanding of the fuel characteristics and properties at high burnup. The better knowledge of material behavior and irradiation-induced phenomena could help the prediction of its performance. In this context, the scope of the present work is to apply a wide portfolio of advanced characterization techniques to determine properties that are relevant for safety and performance. The results are interpreted in the context of engineering scale post-irradiation examinations and available information on the irradiation conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Status of Eddy Current Developments at INL

Both standard and advanced Light Water Reactor (LWR) fuel systems are composed of fuel pellets enclosed in a metallic cladding. The fuel cladding represents the first containment of the fission products and the first safety barrier. Upon progression of irradiation, the cladding properties deteriorate, mainly due to the pick-up of hydrogen, which causes embrittlement, and the consequence corrosion that reduced the metallic substrate. Those changes in the properties combined with the harsh environment, e.g., fatigue and fretting occurring during operation, might eventually lead to loss of hermeticity. Therefore, it is important to be able to quantify the amount of corrosion taking place and to detect and size structural discontinuity that can promote cladding failure. Eddy current testing is one of the non-destructive methods employed for this purpose. Inspections are carried out using special probes, which are designed and optimized to inspect a tube with well-defined geometrical and physical characteristics, i.e. external diameter, wall thickness, material electric conductivity and magnetic permeability. The inspection goals are mainly twofold: first, the determination of the amount of corrosion by measurement of the external oxide thickness; second, the detection of flaws in the tube wall, ether on the external or the internal surface of the tube. This report summarizes the available capabilities for eddy current testing of LWR fuel claddings, identifies needs and are of developments for testing of ATF claddings.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Hydrothermal Corrosion of Latest Generation of FeCrAl Alloys for Nuclear Fuel Cladding

After the Fukushima nuclear disaster, the nuclear materials community has been vastly investing in accident tolerant fuel (ATF) concepts to modify/replace Zircaloy cladding material. Iron–chromium–aluminum (FeCrAl) alloys are one of the leading contenders in this race. In this study, we investigated FA-SMT (or APMT-2), PM-C26M, and Fe17Cr5.5Al over a time period of 6 months in simulated BWR environments and compared their performance with standard Zirc-2 and SS316 materials. Our results implied that water chemistry along with alloy chemistry has a profound effect on the corrosion rate of FeCrAl alloys. Apart from SS316 and Zirc-2 tube specimens, all FeCrAl alloys showed a mass loss in hydrogen water chemistry (HWC). FA-SMT displayed minimal mass loss compared to PM-C26M and Fe17Cr5.5Al because of its higher Cr content. The mass gain of FeCrAl alloys in normal water chemistry (NWC) is significantly less when compared to Zirc-2.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Development of Mechanical Testing Capability for HFEF

This report depicts the development of mechanical testing capability for ATF fuels at Idaho National Laboratory (INL). An aim of the study is to demonstrate the pathway of defueling an irradiated cladding segment, machining of segments into specimens with unique tensile geometries, and mechanical testing using INL’s facilities. The report consists of three parts: 1) defueling of a fuel segment, 2) mechanical test specimen machining, and 3) mechanical test demonstration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Wettability and CHF limits of Accident-Tolerant nuclear fuel cladding materials in light water reactor conditions

Here, we present the results of experimental investigations aimed at evaluating the thermal–hydraulic performance of chromium-coated zircaloy, i.e., one of the most promising accident tolerant fuel (ATF) cladding material for light water nuclear reactors. Precisely, we investigate the wettability and critical heat flux (CHF) limits of chromium-coated and conventional zircaloy surfaces in prototypical reactor conditions. For both surface types, we measure the contact angle in a vapor-saturated atmosphere from ambient pressure to the operating pressure of pressurized water reactors (PWRs), i.e., ~15 MPa. We measure the ambient-pressure steady-state flow boiling CHF with a spatially uniform heat flux. We measure the high-pressure steady-state flow boiling CHF with a cosine shape heat flux (up to 20 MPa) and with a uniform heat flux (up to 15 MPa), also exploring the effect of CRUD deposits on the chromium-coated surface. Our results reveal that the chromium surface and the bare zircaloy surface have similar wettability and both become super-hydrophilic in PWR conditions, and that there is practically no difference in the steady-state CHF limits, both at low-pressure and high-pressure conditions, also when the chromium-coated surface is covered by a CRUD deposit. However, while the chromium-coating does not improve the CHF compared to the bare zircaloy surface, it improves the post-CHF behavior. The chromium coating prevents the reaction between zircaloy and steam, which results in the formation of a brittle zirconium oxide through the surface of the cladding. We also measure the transient CHF under exponentially escalating heat flux inputs of a nano-smooth and a rough surface mimicking a chromium-coated zircaloy cladding. Interestingly, the results of the transient heat flux tests suggest that the CHF limit for very short periods (i.e., fast transients) is independent of the surface finish, being the same for a rough chromium surface or a nano-smooth surface.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Steam oxidation of chromium corrosion barrier coatings for sic-based accident tolerant fuel cladding

SiC-based materials are currently being considered for accident tolerant fuel (ATF) cladding, however, there are concerns regarding hydrothermal corrosion rates and hermeticity. These issues may be addressed with the utilization of a Cr corrosion barrier coating (CBC) but the interaction of Cr with SiC under loss of coolant accident (LOCA) conditions needs to be investigated. In the present work, Cr coatings were deposited onto chemically vapor deposited (CVD) SiC with either high power impulse magnetron sputtering (HiPIMS), cathodic arc (CA), or a combination of both. Annealing and steam thermogravimetric analysis (TGA) was used to study reaction product formation with and without oxidation. Additionally, after 1200 °C annealing, formation of a Cr x C y carbide layer at the ambient interface and a Cr x Si y C z silicide layer at the SiC interface was observed. After 1200 °C TGA exposure, similar carbide and silicide layers were observed but with a Cr 2 O 3 outer reaction layer and a Si and O rich interfacial layer forming between the carbide and silicide layers. Initial mass gain of the coatings during the 1200 °C TGA exposure was parabolic and were similar to reported rates of chromia formation. Mass gain behavior and observed transitions in chromia microstructure indicate that metallic Cr may be consumed within the first ~1 h at 1200 °C and that subsequent chromia formation occurs at the expense of the underlying Cr carbide layer. Overall, the results show no detrimental impact of Cr coatings on the steam oxidation resistance of SiC up to 4 h at 1200 °C.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Improvement of the $\mathrm{BISON U_3Si_2}$ modeling capabilities based on multiscale developments to modeling fission gas behavior

Uranium silicide (U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel, but operated at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. These lower length scale informed models are then utilized in the validation of BISON U 3 Si 2 modeling capabilities to simulate the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the validation process to identify where further experiments and lower length scale modeling would be beneficial. Here, the multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride, uranium carbide).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO 2 fuel were irradiated at a temperature of ~300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. Finally, the study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO 2 interface than C35M due to the higher bulk Cr:Al ratio.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2020

Ion irradiation effects on Cr-coated zircaloy-4 surface wettability and pool boiling critical heat flux.

The concept of coating the currently used nuclear fuel cladding (zirconium-based alloy, typically Zircaloy-4 or Zirc-4) with an oxidation preventive layer is a progressing Accident tolerant Fuel (ATF) candidate alloys. The coated Zirc-4-based alloys could be a solution to suppress undesirable fast reaction kinetics with high-temperature steam. Zirc-4 has been the most preferred cladding material in pressurized water reactors (PWRs). Chromium (Cr) based alloys as a coating material provides excellent corrosion protection and good strength and wear resistance. In this paper, we present the surface wettability measurements and pool boiling Critical Heat Flux (CHF) for Cr-coated Zirc-4 claddings pre- and post-exposure to an ion irradiation environment. The wettability measurements, including static contact angle (contact angle, θ) and average surface roughness (surface roughness, Ra), are introduced for samples of different coating thicknesses (5–30 μm thick). The coatings fabricated by the cold spray of Cr-Al particles to 10 mm × 10 mm × 1.95 mm Zirc-4 substrates. Post fabrication, a Pilgering (cold rolling) process, was applied to finalize the coating thickness and resulted in a significant reduction in surface roughness of initially fabricated rough surfaces. The process produced three distinguished samples 5-μm unpolished (as machined), 5-μm, and 30-μm polished (cold rolled). The measurements are presented for the three surfaces and bare Zirc-4 as a baseline surface. The contact angle analyses were implemented in theoretical models from the literature to predict pool boiling CHF. Pool boiling experiments were conducted to measure the pool boiling CHF values and compare them to the predicted values. Scanning Electron Microscope (SEM) images and Energy Dispersive X-ray Spectroscopy (EDS) analysis was performed to characterize the surfaces for better understanding and interpreting the results. The SEM images showed localized surface damage due to ion irradiation. No recognized change in the measured surface roughness due to ion irradiation. The contact angles of irradiated Cr-coated surfaces are consistently higher (10°) than pre-irradiated surfaces. Decreasing the Cr-coating layer thickness resulted in lower contact angle pre- and post- ion irradiation. The predicted pool boiling CHF using the Kandlikar model is in good agreement with the experimentally measured CHF values within ±12% for all samples.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Reactivity-initiated accidents in two pressurized water reactor high burnup core designs

Here, this paper presents a safety analysis of two proposed core loadings for 24-month Pressurized Water Reactor (PWR) fuel cycles. This analysis focuses on reactivity-initiated accidents (RIAs) and evaluates core safety performance impacts of rod-averaged burnup limits up to 75 GWD/MTU and less than 7 % enriched UO 2 . The capabilities of Polaris, PARCS, and RELAP5-3D are leveraged to evaluate the core neutronic and thermal–hydraulic behavior for normal-operation, uncontrolled control rod withdrawal (CRW) transients, and control rod ejection (CRE) accidents. The two core designs are compared to identify features of realistic high burnup/extended enrichment core design approaches which have significant safety impact, identify experimental data needs for high-fidelity predictive modeling, and provide recommendations for future high burnup core designs. The first core design evaluated in this study was developed by Southern Nuclear Company and used an ZrB 2 Integral Fuel Burnable Absorber (IFBA) and B 4 C Wet-Annular Burnable Absorber (WABA)-based burnable poison strategy. The second core design assessed in this work used a Gd 2 O 3 -doped UO 2 burnable poison, similar to that used in boiling water reactors or French PWRs. Results indicate that fuel thermal limits are maintained for limiting CRW and hot full power (HFP) CRE transients. Cladding failure is predicted for the highest energy deposition rods in each core during limiting hot zero power (HZP) CRE accidents (where maximum radially averaged enthalpy exceeds 120 cal/g), though licensing may be permissible with a limited number of failed rods. While concerns exist regarding high critical boron concentration during steady state for the IFBA core and large plenum pressures for the gadolinia core design, the analysis demonstrates adequate safety performance during limiting RIA accident scenarios for two representative high burnup core designs. Design changes limiting plenum pressures and implementation of accident tolerant fuel (ATF) cladding features which minimize hydriding and susceptibility to pellet-cladding mechanical interaction (PCMI) are recommended for future high burnup fuel concepts. To support the technical basis for burnup limit increases, high-fidelity fuel performance models are needed to address physical effects not considered in this analysis, and high burnup irradiated fuel tests are required to extend applicability of the fuel failure limits and validate existing and future models.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

A Coupled Experimental and Simulation Approach to Investigate the Impact of Grain Growth, Amorphization, and Grain Subdivision in Accident Tolerant U 3 Si 2 Light Water Reactor Fuel

Triuranium disilicide (U 3 Si 2 ) is being considered as an alternative accident tolerant fuel (ATF) due to its higher thermal conductivity. However, there is uncertainty in its use due to a lack of irradiation data at light water reactor (LWR) conditions. In this project, we used an integrated experimental and simulation approach to answer two significant questions pertaining to U 3 Si 2 that were not investigated by the Fuel Cycle Research and Development (FCRD) and Nuclear Energy Advanced Modeling and Simulation (NEAMS) programs: (1) Will grain growth in the hotter portions of the fuel significantly impact U 3 Si 2 LWR fuel behavior? (2) Under what conditions do grain subdivision and amorphization occur in U 3 Si 2 and will either occur at LWR conditions? Our approach to answer these questions used in situ ion irradiation and annealing to investigate grain growth, grain subdivision, and amorphization, along with mesoscale simulations using the MARMOT tool. We found that while the grain boundary mobility of U 3 Si 2 is larger than that of UO 2 , grain growth in U 3 Si 2 fuel pellets will be less than in UO 2 and may be neglected due to the lower thermal conductivity and thus lower centerline temperature. We also found that amorphization will not occur in U 3 Si 2 above 600 K, such that it is not likely to occur at typical LWR operating conditions. Grain subdivision will occur at LWR conditions and will likely occur at lower burnups than in UO 2 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BISON BWR Fuel Modeling Capability: Material Models (FY2020)

Until this year the BISON fuel performance code has been developed primarily for pressurized water reactor (PWR), accident tolerant fuel (ATF), and advanced reactor (e.g., tri-structural isotropic (TRISO), metallic) analyses. The main focus of this work is to develop and document required material models that would aid predicting cladding failure with a specific hydride distribution to be investigated for boiling water reactor (BWR) applications in the next fiscal year. In terms of materials, the primary differences between a BWR and PWR is the use of gadolinia doping of the UO 2 fuel, the use of Zircaloy-2 for the cladding, and the addition of a liner on the inner surface of the cladding for improved pellet-clad mechanical interaction (PCMI)performance. The liner is typically made of pure zirconium or a lower tin content zirconium alloy (e.g. Zr-0.3%Sn). This work details the development and documentation of material and behavioral models for BWR analysis such as the addition of material and behavior models for gadolinia doped UO 2 , pure zirconium, Zr-0.3%Sn, and Zircaloy-2 as well as the ease-of-use additions to allow internal meshing capabilities to include liners. Another important aspect affecting cladding performance is hydrogen diffusion and hydride precipitation in the cladding. A model was added previously for non-lined, homogeneous claddings. In this work, instructions are provided on how to apply the model to BWR liner cladding. Representative experimental values are provided for the hydrogen diffusion parameters and precipitation-dissolution kinetics for typical liner materials of interest.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Fracture Toughness Characterization of Generation II FeCrAl Alloys after ~18 dpa Irradiation

FeCrAl alloys are promising candidate materials for the accident tolerant fuel (ATF) cladding applications due to their excellent corrosion resistance to the elevated temperature steam environment. Currently, the handbook on FeCrAl material properties contains only limited data regarding the fracture toughness properties of any FeCrAl alloy. This includes alloys currently under investigation within the Advanced Fuels Campaign (AFC) at Oak Ridge National Laboratory (ORNL). In this project, a series of irradiation capsules have been irradiated in the High Flux Isotope Reactor (HFIR) at ORNL with two Generation II FeCrAl candidate alloys, i.e., C06M and C36M, to assess the fracture response of these alloys after neutron irradiation. These alloys represent the “book-end” compositions for C26M, the alloy currently being developed as the leading candidate for LWR cladding. A total of six irradiation capsules were irradiated in HFIR at target temperatures of 200°C, 330°C, and 500°C up to target damage doses of 8 displacements per atom (dpa) and 16 dpa. These damage doses represent the expected middle and end of life damage levels for typical LWR cladding while the irradiation temperature regimes will provide insight into the role of varying microstructural features on the fracture toughness properties of neutron irradiated FeCrAl alloys. To date, irradiation of all capsules has been completed in HFIR. This report summarizes the latest results of microhardness and fracture toughness PIE for the 16 dpa capsules (FCAB2, FCAB4, and FCAB6), for which the measured irradiation conditions were: 204°C/17.6dpa, 343°C/18.3dpa, and 507°C/18.6dpa. The main conclusions of this study can be summarized as follows: 1) After the 204°C/17.6dpa irradiation, both C06M and C36M exhibited significant irradiation hardening and embrittlement 2) After the 343°C/18.3dpa irradiation, both C06M and C36M exhibited small irradiation hardening without irradiation embrittlement 3) After the 507°C/18.6dpa irradiation, both C06M and C36M exhibited irradiation softening without irradiation embrittlement 4) Comparing the microhardness and Master Curve reference temperature T 0q before and after neutron irradiation, we did not observe a linear correlation between the two parameters for both C06M and C36M. This should be mainly due to a flat response of the Master Curve reference temperature T 0q to the irradiations at 166-204°C and 315-343°C ranges 5) C06M showed a lower T 0q , meaning better toughness, than C36M at the unirradiated condition and such trend was kept even after neutron irradiation except for the 166-204°C irradiation where both materials had similar T 0q . 6) In terms of hardening and embrittlement, the irradiation effect on both C06M and C36M appeared to saturate after an irradiation dose of 7 dpa.

36 MATERIALS SCIENCE↗
DOE OSTI2021

FY21 burst activities with coated Zircaloy-4 under accident conditions

This report summarizes the results of Advanced Fuels Campaign (AFC) accident-tolerant fuel (ATF) burst activities. Nuclear service grade Zry-4 was procured and coated with a 7-micron thick Cr coating. The coating quality was investigated, and there were several defects at the Cr/Zry-4 interface due to the surface roughness of the as received Zry-4 tubing. To provide insight into the effect of coating defects on the cladding performance under accident scenarios, the unirradiated uncoated and coated material was tested under loss of coolant accident (LOCA) and pellet cladding mechanical interaction (PCMI) reactivity insertion accident (RIA) conditions. The defected coating appeared to have no impact on the cladding performance under these scenarios when compared to the as-received cladding material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗