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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Characterization of Fuel and Cladding In and Near the Pellet-Pellet Gap of a High-Burnup Pressurized Water Reactor Fuel Rod

Oak Ridge National Laboratory (ORNL) is performing extensive destructive examinations of 15 high-burnup (HBU) spent nuclear fuel (SNF) rods from the North Anna Power Station (NAPS), which is operated by Dominion Energy Virginia [1]. The examinations are being conducted to obtain a baseline condition of the HBU rods before dry storage and are focused on understanding overall SNF rod strength and durability [2,3]. The HBU rods, referred to as sister rods or sibling rods, are similar to rods placed into dry storage at NAPS that are planned to be examined after one decade. The sister rods include several ZIRLO®-clad rods manufactured by Westinghouse Electric Company. One of the ZIRLO®-clad rods examined includes a pellet-pellet gap of 1 mm that was identified during the nondestructive examinations (NDEs) [4]. The rod was sectioned axially at that elevation to reveal the pellet-pellet interfaces and the pellet-pellet gap. The section was mounted and polished to reveal the distribution of hydride precipitates in the cladding above, below, and within the gap [5]. Total cladding hydrogen measurements will be performed to quantify the total cladding hydrogen through the gap and any additional in-solution or precipitated hydrogen in the pellet-pellet gap region

Montgomery, Rose↗
DOE OSTI2023

Investigating microstructural evolution in SolidStir™ extruded oxide dispersion strengthened 14YWT alloy fuel cladding tube

Structural components for extreme environments require advanced materials and manufacturing processes. One such advanced material is oxide dispersion strengthened (ODS) Fe–14 wt.%Cr–3W–0.4Ti–0.3Y 2 O 3 (14YWT), which is developed for components to be used in Gen IV and fusion nuclear reactors. However, the conventional manufacturing processes to produce components, such as fuel cladding tubes, either do not retain the desired microstructural attributes or are costly and less efficient. As a result, a more attractive option is a novel manufacturing process developed on the principle of friction stir welding or processing (FSW/P) and commercially referred to as SolidStir™ Extrusion (SSE). Here this study used the SSE technique to study the manufacturing and microstructural evolution of a fuel cladding tube made of ODS 14YWT alloy. The cladding extrusion by the SSE technique involved the use of ball-milled 14YWT powders with Y, Ti, and O in the solid solution and the use of a specially designed W-25Re-Hf tool for consolidation and extrusion of the powders. A scanning electron microscope (SEM), electron backscattered diffraction (EBSD), and transmission electron microscope (TEM) were utilized for microstructural characterization. A Keyence microscope captured macrostructural photographs of the extruded tube. EBSD examination of the extruded tube on both transverse and longitudinal cross-sections showed the presence of dynamically recrystallized grains and revealed that the average grain size of the transverse cross-section was smaller than that of the longitudinal cross-section. The texture was weak and consisted of some amount of shear texture. The presence of nano-oxide clusters or precipitates of Y, Ti, and O (pyrochlores) in the extruded tube was determined by TEM. Aging heat treatment caused the average precipitate size to decrease and the density to increase compared to the as-processed condition. The results indicated that SSE is a viable tool for manufacturing fuel cladding tubes with the desired microstructural attributes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Zirconium nanoparticle coating development for FCCI diffusion barrier in nuclear cladding

Fuel clad chemical interaction (FCCI) is a pervasive issue for metallic nuclear fuels. FCCI can damage cladding, reduce thermal efficiency, and compromise the structural integrity of the fuel rod. In this study, we evaluated the FCCI mitigation capabilities of electrophoretically-deposited Zr nanoparticle coatings on ferritic-martensitic stainless-steel cladding (HT9). Zr was deposited on the internal surface of 6 mm outer diameter cladding using electrophoretic deposition (EPD). Thickness varied among samples with similar EPD conditions. A uniformly coated sample was selected and sent to the Argonne Tandem Linac Accelerator System (ATLAS) to simulate the coating's behavior under irradiation conditions. Furthermore, the coating prevented FCCI in diffusion couple studies. Results of this study confirm that Zr could potentially make an effective FCCI barrier in metallic fuel rods but the deposition process needs improvement.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Step-loaded creep testing of Zircaloy-4 cladding at higher temperatures in the α-phase

A refined understanding of zirconium-based cladding thermomechanical performance during rapid transients is essential for enhancing the safety and operation of light-water reactors. Traditional models for zirconium alloys under accident conditions generally assume that creep dominates fuel cladding performance. Here, these historic models have largely remained unchanged and serve as the basis for safety criteria development. As the U.S. nuclear industry pursues higher burnup levels, the increased release of fission gases during transients raises the risk of cladding failure in the low-temperature hcp α-phase, making the fidelity of these models of greater importance. Creep testing was conducted from 550–700°C with 25–120 MPa applied hoop stresses to investigate Zircaloy-4 deformation at accident-relevant temperatures in the α-phase. Step-loading was employed to capture creep behavior across a wide stress range from a single sample. The stress-strain rate data at higher temperatures (650 and 700°C) were well-described by isotropic versions of the Erbacher and Kaddour models, while the lower temperature data (550 and 600°C) were underpredicted by both anisotropic and isotropic model variants. Greater strain rates during the initial loading step at 650 and 700°C were attributed to recrystallization and grain growth of sub-micron crystallites. Yet, texture analysis revealed the basal split texture remained after testing. These observations produced results suggesting Zircaloy-4 claddings experience higher creep rates across the α-phase than previously thought, possibly related to dynamic anisotropy due to temperature dependent activation of deformation mechanisms, effects of biaxial loading, and variation in material condition between the current testing used in previous model development.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Experimental characterization and potential energy savings of insulated cladding for U.S. residential buildings

This study evaluates a novel insulated cladding composed of foamed cement and vinyl siding that is designed as a retrofit product to enhance thermal performance in U.S. residential buildings. Experimental measurements showed that foamed cement exhibits a thermal conductivity of 0.0334–0.0365 W/m-K (R-value of 4.0–4.3 °F⋅ft 2 ⋅h/BTU), which is significantly higher than that of conventional fiber cement boards. Although mechanical testing confirmed the material’s brittleness and low strength, it also indicated its suitability for nonstructural insulation applications. Water vapor permeability testing demonstrated effective moisture resistance with the polymer coating. To evaluate the insulated cladding’s energy performance, we used the U.S. Department of Energy’s prototype single-family building model and simulated across 50 locations spanning 16 International Energy Conservation Code climate zones. The R-7 (R-value of 7 °F⋅ft 2 ⋅h/BTU per inch) retrofit cladding resulted in heating energy savings up to 275 therms (Fairbanks, AK) and cooling energy savings up to 1,547 kWh (Phoenix, AZ). The energy cost savings varied by region, but the highest annual cost savings ($\$$434/year) were observed in Santa Maria, CA, and the highest annual cost savings percentage (37 %) were noted in Monterey, CA. On a national scale, the insulated retrofit provided an average annual energy cost reduction of 16 %, highlighting the cladding system’s broad applicability and financial viability.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

A Comparative Tribo-Mechanical Behavior Analysis of Laser Cladded Nitronic 60 Coating Against Wrought Nitronic 60 Alloy

In this study the tribological and mechanical behavior of a laser cladded Nitronic 60 coating on maraging steel is compared to that of a wrought Nitronic 60 alloy. A multi-layer Nitronic 60 coating was deposited using laser directed energy deposition, and its microstructure, porosity, and microhardness were characterized. The microhardness of the coated region in the laser cladded Nitronic 60 varied between 270-300 HV whereas that for wrought form was 230 HV. Controlled tribological tests under motor oil lubricated conditions were conducted at both room and elevated temperatures. The laser cladded Nitronic coating exhibited poorer wear resistance at room temperature but higher wear resistance at elevated temperature as compared to the wrought Nitronic alloy. Severe plastic deformation and fracture was observed on the laser cladded coating after the wear tests at both temperatures. Wear mechanisms were investigated based on the morphological and compositional characterization of the wear tracks.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment

New fuel design and development currently requires 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance experiments to help reduce the timeline associated with fuel design and development. This study uses the Dynamical System Scaling (DSS) method to analyze cladding temperature data from the recent SETH-C experiment in the TREAT facility and accompanying BISON simulations to assess dynamic distortions occurring throughout the fast power excursion transient. The DSS analysis revealed that on the cool down from peak cladding temperature that the fuel radial power profile is the most sensitive modeling parameter with a heterogenous radial peaking factor corresponding to the lowest distortion compared to a uniform energy deposition. For the heat up to peak cladding temperature the heterogeneous radial power profile corresponded to the shortest process action. Finally, for the heat up to peak cladding temperature, the gap conductance model sensitivity was quantified using process action and shows that the default Light Water Reactor gap conductance model corresponded to the longest process action.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Interaction of Polymethyl Methacrylate with Boehmite-Filmed Aluminum Cladding Under Gamma Irradiation

This paper presents an overview of ongoing work to qualify the Advanced Test Reactor (ATR) driver fuel elements that have been affected by irradiation-degraded polymethyl methacrylate (PMMA) flux wands. Irradiation testing was performed on PMMA material in contact with aluminum clad material. The cladding was prefilmed with a boehmite oxide layer, an important feature of the ATR driver fuel. The effects on the boehmite layer due to gamma irradiation of the PMMA-aluminum clad system were investigated. PMMA embrittlement, followed by softening and degradation, occurred at high radiation levels. Adhesion between the cladding and irradiated PMMA was observed. Flow testing at prototypic ATR flow rates demonstrated the effective removal of the adhered material. Measurements of the boehmite layer thickness were performed, and Raman spectroscopy was utilized to detect the presence of boehmite in the irradiated PMMA material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Implementing Fuel Cladding Bonding and Assessing its Impact on Axial Gas Communication

Fuel rods irradiated in light-water reactors to burnup values above 45 MWd/kgU are subject to the formation of a chemical and mechanical bond between the fuel and cladding upon gap closure. The formation of the bond subsequently inhibits the ability of fission gases released from the fuel to flow freely to the plenum of the rod. The flowing of gases within fuel rods is referred to as axial gas communication. During transients, such as loss of coolant accidents, the bond may influence cladding deformation prior to breaking. Upon bond breakage, gases are able to more freely communicate to the lower pressure regions of the rod. The impact of bonding on gas communication and the ballooning behavior of the cladding during a loss of coolant accident is of interest to the nuclear industry in support of burnup extension for the existing light-water reactor fleet. In this report, a model to capture the effects of fuel cladding bonding in the BISON fuel performance code is presented. The theory of the model along with implementation testing is provided. A summary of a previously developed axial gas communication is given to set the stage for how the two models may be coupled together. A full-length pressurized-water reactor fuel rod demonstration is highlighted to evaluate the impact of including bonding on axial gas communication calculations using a preliminary coupling methodology. An overview of the next steps regarding the modeling of bonding, axial gas communication, and a more tightly coupled framework is also provided.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Unraveling the Mechanics: Computational Modeling of Residual Stresses in U10Mo Fuel with Aluminum Cladding

LEU (Low Enriched Uranium) plate-type fuel elements consist of a high-density, low-enrichment U–Mo alloy-based fuel foil encapsulated in an aluminum cladding and is fabricated through the Hot Isostatic Pressing (HIP) technique. Understanding the mechanism of possible failure modes of nuclear fuel is critical to mitigate potential consequences. One of the major contributing factors in various failure modes is stress and it greatly affects the integrity under temperature, pressure, and irradiation. These stresses could originate from various sources, including manufacturing, and operation, and can lead to deformation, cracking, or even complete failure of fuel elements. This study focuses on the development of a computational model that accurately predicts the residual stresses generated in the U10Mo fuel during fabrication. It has been observed that cladding creep has a substantial impact on the residual stresses in the U10Mo fuel post the HIP fabrication process. Furthermore, during the HIP bonding process the fuel plate system is heated to a temperature of 560 oC and as a result, the aluminum cladding transitions from (Al 6061-T6 to -O). This presents a challenge in capturing the change in material properties accurately. Therefore, a temperature dependent creep model such as hyperbolic sine creep model is considered to estimate the creep properties of the Aluminum cladding. The proposed calibrated creep model accurately predicts the residual stresses in the U10Mo fuel foil and agrees well with the experiments.

42 ENGINEERING↗
DOE OSTI2020

Thermal Dehydration of Aluminum (Oxy)hydroxides on Fuel Cladding Material - 20200

The aluminum cladding of research-reactor fuel undergoes general corrosion with resulting formation of adherent aluminum (oxy)hydroxide films during in-reactor and post-discharge exposure to water under various conditions and temperatures. These (oxy)hydroxides contain chemically-bound water that poses challenges for extended dry storage due to the risk of thermal or radiolytic decomposition releasing free water and/or hydrogen and oxygen gases. This study comprised laboratory experiments of thermal drying behavior of aluminum (oxy)hydroxides to identify approaches for reducing bound water on fuel cladding prior to sealed dry storage. Effective drying strategies for adherent (oxy)hydroxides on fuel cladding will improve the safety of dry storage by mitigating potential avenues for additional corrosion and/or generation of flammable gases inside the storage canister. This work is part of a broader investigation to address knowledge gaps and technical data needs for dry storage of aluminum-clad spent nuclear fuel (ASNF), which included investigation of (oxy)hydroxide formation on aluminum alloy substrates immersed in water, characterization of service-grown films from ASNF, and radiolytic yield of hydrogen from (oxy)hydroxide powders and films. The current work comprises experimental thermal dehydration of aluminum trihydroxides (bayerite and gibbsite) characteristic of low-temperature (<80 deg. C) corrosion in water and aluminum oxyhydroxide (boehmite) characteristic of high temperature (>80 deg. C) corrosion in water. First, commercially produced (oxy)hydroxide powders (boehmite and gibbsite) were tested via thermogravimetric analysis (TGA) and differential scanning calorimetry (DSC). The effort aimed to identify or confirm key temperature ranges/thresholds for the thermal decomposition reactions, as well as impacts of ramp rate and hold times, for drying of isolated, high-surface-area (oxy)hydroxides. The information gleaned from powder tests was used to guide subsequent drying tests on adherent (oxy)hydroxide films grown on aluminum alloy substrates. TGA was used to analyze small samples of the aluminum coupons with adherent (oxy)hydroxide films. Specimens were characterized both pre- and post-drying using X-ray diffraction (XRD) to determine the film composition and scanning electron microscopy (SEM) to determine its morphology. Drying of adherent (oxy)hydroxide films was anticipated to pose additional complications relative to drying of (oxy)hydroxide powders. Anticipated challenges include achieving sufficient drying of the (oxy)hydroxide at low-enough temperature to avoid melting or other undesirable phase changes in the aluminum substrate and overcoming transport limitations of water through the thickness of a dense film or through a tortuous pore structure to reach the outer surface of the film. Inn addition, dehydration of (oxy)hydroxides significantly changes the film density, which may lead to substantial alterations in the morphology, including potential cracking and spalling of the film. TGA/DSC tests of gibbsite powders resulted in successful conversion to boehmite or alumina, depending on the maximum temperature reached. At low ramp rates (≤5 deg. C/min), the conversion to boehmite occurred around 300 deg. C (210-340 deg. C). XRD confirmed that boehmite was the only phase detected after tests reaching 450 deg. C. The transition to alumina occurred around 510 deg. C (470-550 deg. C), with XRD detecting only alumina after tests reaching at least 600 deg. C. Boehmite powders dehydrated to alumina at about 400 deg. C (330-460 deg. C) for coarse (77-μm particle diameter) powder and about 490 deg. C (420-520 deg. C) for fine (0.7-μm particle diameter) powder. For all powders tested, the maximum percent mass loss after drying to high temperature slightly exceeded the theoretical mass loss for complete dehydration of the stoichiometric (oxy)hydroxide to alumina, which is likely attributable to physisorbed water in the powder. Drying of a thick (∼8.6 μm), predominantly bayerite adherent film displayed a dramatic change in film morphology, with the initially continuous trihydroxide film cracking into sections on the order of 50 μm square and partially delaminating. The outermost layer of the film completely faked of in some regions. The layer exposed under the spalled oxide also displayed prominent cracking, with spacing on the order of 10 μm, but this surface layer appeared to remain completely adhered to the aluminum substrate. TGA showed mass losses per unit surface area up to ∼1.0 mg/cm{sup 2}. XRD characterization of the remaining oxide was inconclusive, with no crystalline phases detected. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490

A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Anisotropic Effects of Loading on Zircaloy Cladding Performance

This presentation provides recent results related to the mechanical behavior of Zircaloy cladding for the Light Water Reactor Fuels Workshop/CRAFT Initiative Update presented by the Advanced Fuels Campaign. This presentation focuses on Zircaloy cladding response to pellet-cladding interactions and how mechanical anisotropy makes the cladding sensitive to the testing mode. Results are compared to existing material models used by BISON, and implications for reactor performance are briefly discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Enhanced Stress Relaxation Behavior via Basal ⟨a⟩ Dislocation Activity in Zircaloy-4 Cladding

Herein this work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25%, 1%, and 2% equivalent strains in the cladding samples at a temperature of 300°C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal ⟨a⟩ dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Postirradiation characterization of palladium as an additive for fuel cladding chemical interaction mitigation in metallic fuel

This work describes the microstructural and elemental characterization of irradiated metallic fuels containing palladium as an additive. The use of additives has been proposed to control Fuel-Cladding Chemical Interaction (FCCI) and thus to promote higher fuel utilization (i.e., higher burnup). In this work, Pd has been investigated as a potential additive to metallic fuel to bind lanthanides, impeding their migration and attack on the cladding. The influence of Pd on the microstructure, chemistry and performance of metallic fuel has been characterized via scanning electron microscopy for two metallic fuel designs—namely, annular and solid fuel. Pd was observed to play an important role in the chemistry of the fuel. Indeed, the addition of Pd leads to the formation of new phases. Pd was detected to combine not only with the lanthanides, as intended, but also with Zr, a main element of the fuel matrix. While Pd proved to be effective in preventing lanthanide migration and their attack on the cladding, the Pd-Zr compound may potentially lead to other unexpected fuel-performance issues, such as the formation of low-melting point phases and increased unalloyed U available for FCCI interaction with Fe in the cladding. Even the increase of Zr to 13wt%. did not completely mitigate this adverse phenomenon generated by the Pd-Zr interaction. Furthermore, the efficacy of using this additive needs further investigation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of grain size and porosity on cladding failure in high-burnup UO 2 : A sensitivity and uncertainty study

Isotopic taggants are being studied to aid in the provenance assessment of nuclear materials. However, these taggants must be selected such that they do not adversely affect fuel performance during normal operation or accident scenarios. Taggants are known to affect the fuel’s grain size and porosity. In the work described in this paper, the BISON fuel performance code was used to assess the potential effects of taggants (i.e., grain size and porosity) on fuel rod behavior and cladding failure during a high-burnup, large-break loss-of-coolant accident. Here, 281 individual fuel rods from the same reactor core were modeled for a sensitivity study, a parametric study, and uncertainty quantification. The cladding failure predictions often exhibited stochastic behavior. After additional study, it was found that the cladding failure model is highly sensitive to residual error inherent to numerical approximation solvers. Some strategies to mitigate this sensitivity are discussed. The study found no relationship between known taggant effects and cladding failure status. However, taggants were found to affect the time and location of failure in certain rods. In conclusion, future work to continue investigating and validating these findings is briefly discussed.

Doped UO 2↗
DOE OSTI2020

3D printed transparent ceramic YAG laser rods: Matching the core-clad refractive index

Yttrium Aluminum Garnet (YAG) solid state laser gain media rods with an active Neodymium-doped core and an optically-clear cladding region were additively manufactured via direct-ink-writing (DIW), followed by sintering and hot isostatic pressing to form fully dense optical ceramics. Lutetium and Gadolinium were chosen as optically-inert co-doping ions in the clad to match the increase in refractive index caused by the Neodymium in the core. Additionally, either 11.6% Lutetium or 3.8% Gadolinium can be used to match the index change from 2% neodymium; however, differences in diffusion distances across the core-clad interface lead to large fluctuations in index in that region. These index fluctuations can be minimized either by matching dopants with similar diffusion distances, or by implementing a gradual gradient in the doping profile, possible through DIW, rather than a sharp compositional interface attainable via more standard fabrication methods. This improvement in index homogeneity resulted in a 40% improvement in lasing performance compared with that of a core-clad rod fabricated with a sharp interface between the doped and undoped regions.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Investigating the Role of Accident Tolerant Cladding on Source Term Reduction for High-Burnup PWRs Using MELCOR

The use of accident tolerant fuel (ATF) cladding can increase coping times during and beyond design basis accidents. While such gains may be incremental, they provide a margin that can potentially be recovered to enable high-burnup (HBU) operation. Realizing such a margin requires demonstrating that the combination of HBU and ATF has not led to an overall increase in source term. This study investigates the influence of cladding technology (Zr-based, Cr-coated Zr, and FeCrAl) and fuel cycle length (18 and 24 months) on radiological dose at the boundary of the exclusion zone for a four-loop pressurized water reactor to investigate whether ATF claddings can provide such benefits. We analyze a recovered large break loss-of-coolant accident scenario to investigate the impact of transient timescale on the benefits of such coping time increases. The simulations have been performed using the MELCOR and MELCOR Accident Consequence Code System codes. For the cases analyzed, increased fuel cycle length did not necessarily increase radionuclide release and hydrogen generation, as these were found to be sensitive to the core power distribution. Similarly, off-site dose consequence is dominated by short-lived radionuclides that tend to saturate earlier in the burnup, so higher burnup operation did not necessarily increase the source term for the phenomena and transients analyzed here. Delays in recovery of the lowpressure safety injection system increase hydrogen production and radionuclide release, especially between 780 s and 1620 s, due to the nonlinear oxidation and core degradation behavior. Results show that Cr-coated Zr enhances safety by delaying heatup and gap release. Here, when uncertainty propagation on oxidation properties is considered, FeCrAl exhibits the lowest overall radionuclide release and off-site dose throughout the spectrum. However, while the considered “base model” performance is superior under delayed injection scenarios, upper-bound cases display hydrogen generation risk comparable to the Zr-based cladding.

Accident Tolerant Fuel↗