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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Multi-Functional Zirconium-Silicide Coatings on Zirconium-alloy for Improved Accident Tolerance

Recently, development of protective coatings on zirconium-alloy (Zr-alloy) fuel cladding in light water reactors has been extensively investigated to enhance safety margins in accident scenarios. One accident scenario is degradation of the reactor core under a coolant-limited condition, where a large amount of Zr-alloy claddings would react with high temperature steam, leading to oxidation, hydrogen production, and loss of intrinsic mechanical properties of the fuel cladding. Although, total replacement of Zr-alloy with SiC-SiCf composites or FeCrAl is being considered for the long-term, application of oxidation resistant coatings on zirconium-alloy is being investigated as a near-term solution to mitigate oxidation of Zr-alloy and increase the accident coping time. Coatings, such as chromium, Ti{sub 2}AlC, SiC, and Cr{sub 3}C{sub 2}- NiCr, deposited by various deposition techniques have been investigated. In addition, recent studies on modified heating surface for heat transfer enhancement provides basis on the optimal design of coated Zr-alloy claddings so that the new cladding can be achieve accident tolerance and heat transfer efficiency for both accident and normal operation conditions. Therefore, boiling heat transfer from the coated surface to water environment must be evaluated in parallel. In heat transfer performance, boiling heat transfer is characterized by nucleate boiling, critical heat flux (CHF), transition boiling, minimum film boiling temperature (T{sub min}), and film boiling. Zirconium-silicide (ZrSi{sub 2}) is a potentially attractive protective coating material for cladding applications. Although this material has not been studied extensively, it appears to be a promising candidate because of the high melting points, and the low neutron penalty, and the generally good oxidation resistance to transition metal silicides. It has been reported that oxidation of ZrSi{sub 2} powder in a flowing air environment at elevated temperatures produced ZrO{sub 2}, SiO{sub 2}, and ZrSiO{sub 4}, pointing to the possibility of achieving highly protective surface layers of SiO{sub 2} and ZrSiO{sub 4} that would protect the material from further oxidation. In this study, magnetron sputter deposition of ZrSi{sub 2} on zirconium-alloy flats and cylindrical rods was performed. The coated samples were oxidized in high temperature air (700 deg. C, 1000 deg. C and 1200 deg. C) to evaluate their oxidation resistance. The coated cylindrical rods were quenched in saturated water to observe transition boiling, T{sub min}, film boiling regime, while the coated flats were tested in pool boiling facility to investigate nucleation boiling and CHF.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Characterization of PVD Cr, CrN, and TiN coatings on SiC

SiC ceramic matrix composites are a potential replacement for current light water nuclear reactor fuel cladding material. However, loss of fission gas via micro-cracks and corrosion remain an issue. Cathodic arc Cr, CrN, and TiN coatings were deposited on SiC tubes and plates to provide hermeticity and corrosion resistance. These coatings were characterized to determine as-deposited quality. Cross-sectional microscopy, X-ray diffraction, glow discharge optical emission spectroscopy, and scratch tests were performed to evaluate the purity, structure, and mechanical performance of the coatings. Nitride coatings had stable interfaces, but larger defects in the coatings as compared to the Cr coatings which showed cracking at the interface, but less deposition-induced defects. Despite the local state of the interface, the mechanical properties of the metallic coatings versus ceramic coatings enabled the Cr coatings to resist loads three times that of the nitride coatings during scratch tests. Finally, glow-discharge optical emission spectroscopy showed that improvement in elemental purity is needed for future coatings.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Radial Thermal Conductivity Measurement for SiC Composite Tubes

Thermal conductivity is a key property for design and performance assessment of silicon carbide composite (SiC/SiC) cladding that is being developed for accident-tolerant fuel to be used in light water reactors. The thermal conductivity of SiC plate has previously been studied, but it was unknown whether it would be different for tubular SiC. The purpose of this report is to provide the thermal conductivity of prototypic SiC composite tubes under temperatures ranging from ~25°C to ~300°C. Both nonirradiated and neutron-irradiated specimens were evaluated. Thermal diffusivity of curved coupons was measured by a laser flash method. First, the effect of the specimen geometry on thermal diffusivity was investigated using a glass ceramic with known thermal diffusivity. The results showed an insignificant impact of the specimen geometry on the thermal diffusivity in the case of a curvature radius of ~4 mm. Subsequently, nonirradiated and neutron-irradiated SiC composite coupons with the same geometry as the glass ceramic specimens were evaluated. Thermal conductivity was also obtained using additional data for density and specific heat capacity. Equations were written to express the temperature-dependent thermal conductivities of the nonirradiated and irradiated SiC composite tubes.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Pressure Loading with In-Situ Permeability Measurement of SiC Ceramic Matrix Composite Tube

Silicon carbide fiber in a silicon carbide matrix composite (SiC-SiC) has been actively researched as a potential structural material to be used in advanced high temperature gas reactors (HTGR) and as an accident tolerant cladding material for light water reactors (LWR). This ceramic composite material demonstrates desirable properties for use in nuclear reactor core that include retention of strength from ambient to LWR and HTGR operating and accident temperatures and stability under neutron radiation. However, the ceramic composite material must be designed to meet all the necessary, application driven requirements before it could be implemented in a reactor design. As one of the requirements for a fuel cladding material, the composite has to maintain high level of impermeability as to retain fission gases throughout the fuel cycle. General Atomics (GA) had previously reported results on permeability levels obtained by CVI produced SiC-SiC composites. It has been determined that for LWR size cladding tube that is 51 mm long the permeability allocation equivalent to currently used claddings is 7.3 E-10 atm*cc/sec. With high level of attention to possible sources of leaks in a test setup, it has been demonstrated that a CVI SiC-SiC composites produced by GA can achieve a leak rate of 1 E-12 atm*cc/sec, and therefore it meets the permeability allocation for a fuel cladding. Further study showed that a particular SiC-SiC composite tube can remain impermeable after being pressure loaded well above LWR normal operating conditions. In the study done by Deck et al., hoop stress was applied to a composite tube using pressure load on the inside surface in a load-unload stepwise sequence with increasing maximum applied stress at each step and measuring permeability between steps. The tested composite comprised of 70% thickness of SiC-SiC composite which provides fracture toughness and bears a majority of the applied stress and a 30% thickness of monolithic SiC layer on the outside surface to provide an impermeable barrier. The composite remained impermeable up to internal pressure of 56 MPa which is more than 2.5 times greater than a typical pressure at the end of life in a fuel rod.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of hydrothermal corrosion model and BWR metal coating for CVD SiC in light water reactors

SiC/SiC fiber composites with CVD SiC overcoats are potential candidates for light water reactor advanced accident tolerant cladding materials. Understanding its corrosion kinetics in Light Water Reactor (LWR) conditions is essential to evaluate the concept's viability. Existing models only account for the temperature and oxygen concentration effect on the hydrothermal corrosion behavior applicable to LWR operating conditions. However, the development of a general corrosion rate for CVD SiC that accounts for the impact of irradiated microstructure, flow rate, electrical resistivity, pH, and surface roughness is critical for the practical realization of SiC/SiC-based cladding concepts. After a rigorous experimental campaign, this work updates the existing hydrothermal corrosion model to predict hydrothermal corrosion in LWRs. Numerical radiation and coolant chemistry analysis for LWRs conducted based on the updated corrosion kinetic models suggests that CVD SiC is likely a viable environmental barrier coating for Pressurized Water Reactors while questionable for Boiling Water Reactors (BWR) when the effect of irradiation damage on SiC corrosion is considered. An effective mitigation strategy for the double-layer metal coating is proposed for BWR applications. The double-layer metal coating comprising a FeCrAl overcoat with an intermediate Cr bond coating was observed to provide a stable protective barrier against SiC dissolution in BWR conditions. Finally, the proposed metal coating was also fully adherent following quench and burst tests.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Pool boiling critical heat flux studies of accident tolerant fuel cladding materials

Pool boiling experiments have been performed at atmospheric pressure on potential accident tolerant fuel (ATF) cladding materials (Zirlo®, Cr coated Zirlo®, FeCrAl coated Zirlo®, and monolithic SiC) to evaluate the Critical Heat Flux (CHF) points. Corrosion resistant coatings of Cr and FeCrAl alloy were deposited on flat Zirlo® samples using cold spray technology. The as-prepared samples after uniform surface polishing were subjected to autoclave tests at 360 °C water and 18.6 MPa for 360 h to simulate prototypic corrosion of the ATF cladding materials in LWR normal operation. Surface characteristics potentially influencing CHF such as surface morphology, roughness, static contact angle, and surface chemistry were characterized using a suite of characterization methods including scanning electron microscopy, 3D optical profilometry, optical contact angle measurements, and x-ray photoelectron spectroscopy (XPS). Thermo-physical properties of the samples such as density, thermal conductivity, and heat capacity were measured by differential scanning calorimetry and laser flash thermal diffusivity measurement. FeCrAl coatings showed CHF values comparable to bare Zirlo® samples, while slightly lower CHF values were observed for Cr coated samples and bulk SiC flats (produced by chemical vapor deposition, CVD). As-deposited Cr coatings showed 67% higher CHF than the as-polished Cr coatings due to its higher surface roughness level. The autoclave testing for the prepared samples generally increased the CHF except for the surface polished Cr coatings. Similar CHF values of bare Zirlo®, FeCrAl coating, and CVD SiC samples were observed after autoclave test but there were negligible change CHF of Cr coatings. XPS studies indicated that the formation or deposition of a few monolayers of hydrophobic carbonaceous species on surface can affect CHF values. The trends in CHF data are discussed in terms of evolution of the surface characteristics of the ATF materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Dynamic Probabilistic Risk Assessment Based Response Surface Approach for FLEX and Accident Tolerant Fuels for Medium Break LOCA Spectrum

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Measuring the elastic tensor of a monolithic SiC hollow cylinder with resonant ultrasound spectroscopy

Resonant Ultrasound Spectroscopy (RUS) has become a standard tool in material science to measure nondestructively the elastic properties of materials. The widespread adoption of this tool for applications related to nondestructive testing and structural health monitoring in the industry and other places has been restricted thus far by some requirements imposed by RUS on the sample to be tested, in particular the fact that the sample geometry should be relatively simple (e.g., cuboid, cylinder, or sphere). Here, we use RUS to determine the baseline material properties of a hollow cylinder made of monolithic SiC, a surrogate material and geometry representing nuclear fuel cladding for this study. Typically, elastic properties of the nuclear fuel cladding samples are inspected destructively (e.g., indentation) or with acoustic techniquesrequiring multiple and cumbersome measurements. Here, a relatively new RUS approach specifically developed to measure the elastic tensor of samples with arbitrary geometries is used. The results obtained from this RUS approach are compared to values reported in the literature. As a second verification step, vibrational mode shapes computed numerically with the elastic constants are compared to those measured on the sample using a 3D laser Doppler vibrometer, showing an excellent agreement.

36 MATERIALS SCIENCE↗
NASA NTRS2012

Demonstration of Minimally Machined Honeycomb Silicon Carbide Mirrors

Honeycomb silicon carbide composite mirrors are made from a carbon fiber preform that is molded into a honeycomb shape using a rigid mold. The carbon fiber honeycomb is densified by using polymer infiltration pyrolysis, or through a reaction with liquid silicon. A chemical vapor deposit, or chemical vapor composite (CVC), process is used to deposit a polishable silicon or silicon carbide cladding on the honeycomb structure. Alternatively, the cladding may be replaced by a freestanding, replicated CVC SiC facesheet that is bonded to the honeycomb. The resulting carbon fiber-reinforced silicon carbide honeycomb structure is a ceramic matrix composite material with high stiffness and mechanical strength, high thermal conductivity, and low CTE (coefficient of thermal expansion). This innovation enables rapid, inexpensive manufacturing. The web thickness of the new material is less than 1 millimeter, and core geometries tailored. These parameters are based on precursor carbon-carbon honeycomb material made and patented by Ultracor. It is estimated at the time of this reporting that the HoneySiC(Trademark) will have a net production cost on the order of $38,000 per square meter. This includes an Ultracor raw material cost of about $97,000 per square meter, and a Trex silicon carbide deposition cost of $27,000 per square meter. Even at double this price, HoneySiC would beat NASA's goal of $100,000 per square meter. Cost savings are estimated to be 40 to 100 times that of current mirror technologies. The organic, rich prepreg material has a density of 56 kilograms per cubic meter. A charred carbon-carbon panel (volatile organics burnt off) has a density of 270 kilograms per cubic meter. Therefore, it is estimated that a HoneySiC panel would have a density of no more than 900 kilograms per cubic meter, which is about half that of beryllium and about onethird the density of bulk silicon carbide. It is also estimated that larger mirrors could be produced in a matter of weeks. Each cell is completely uniform, maintaining the shape of the inserted mandrel. Furthermore, the layup creates pressure that insures node bond strength. Each node is a composite laminate using only the inherent resin system to form the bond. This contrasts starkly with the other known method of producing composite honeycomb, in which individual corrugations are formed, cured, and then bonded together in a secondary process. By varying the size of the mandrels within the layup, varying degrees of density can be achieved. Typical sizes are 3/8 and 3/16 in. (approximately 10 and 5 millimeters). Cell sizes up to 1 in. (approximately 25 millimeters) have been manufactured. Similarly, the shape of the core can be altered for a flexible honeycomb structure.

Goodman, William↗
DOE OSTI2015

AREVA Team Enhanced Accident Tolerant Fuels Concepts (Phase 1A Final Report)

In response to the Department of Energy (DOE) funded initiative to develop and deploy lead fuel assemblies (LFAs) of Enhanced Accident Tolerant Fuel (EATF) into a US reactor within 10 years, AREVA put together a team to develop promising technologies for improved fuel performance during off normal operations. This team consisted of the University of Florida (UF) and the University of Wisconsin (UW), Savannah River National Laboratory (SRNL), Duke Energy and Tennessee Valley Authority (TVA). This team brought broad experience and expertise to bear on EATF development. AREVA has been designing; manufacturing and testing nuclear fuel for over 50 years and is one of the 3 large international companies supplying fuel to the nuclear industry. The university and National Laboratory team members brought expertise in nuclear fuel concepts and materials development. Duke and TVA brought practical utility operating experience. This report documents the results from the initial “discovery phase” where the team explored options for EATF concepts that provide enhanced accident tolerance for both Design Basis (DB) and Beyond Design Basis Events (BDB). The main driver for the concepts under development were that they could be implemented in a 10 year time frame and be economically viable and acceptable to the nuclear fuel marketplace. The economics of fuel design make this DOE funded project very important to the nuclear industry. Even incremental changes to an existing fuel design can cost in the range of $100M to implement through to LFAs. If this money is invested evenly over 10 years then it can take the fuel vendor several decades after the start of the project to recover their initial investment and reach a breakeven point on the initial investment. Step or radical changes to a fuel assembly design can cost upwards of $500M and will take even longer for the fuel vendor to recover their investment. With the projected lifetimes of the current generation of nuclear power plants large scale investment by the fuel vendors is difficult to justify. Specific EATF enhancements considered by the AREVA team were; Improved performance in DB and BDB conditions; Reduced release to the environment in a catastrophic accident; Improved performance during normal operating conditions; Improved performance if US reactors start to load follow; Equal or improved economics of the fuel; and Improvements to the fuel behavior to support future transportation and storage of the used nuclear fuel (UNF). In pursuit of the above enhancements, EATF technology concepts that our team considered were; Additives to the fuel pellets which included; Chromia doping to increase fission gas retention. Chromia doping has the potential to improve load following characteristics, improve performance of the fuel pellet during clad failure, and potentially lock up cesium into the fuel matrix; Silicon Carbide (SiC) Fibers to improve thermal heat transfer in normal operating conditions which also improves margin in accident conditions and the potential to lock up iodine into the fuel matrix; Nano-diamond particles to enhance thermal conductivity; Coatings on the fuel cladding; and Nine coatings on the existing Zircaloy cladding to increase coping time and reduce clad oxidation and hydrogen generation during accident conditions, as well as reduce hydrogen pickup and mitigate hydride reorientation in the cladding. To facilitate the development process AREVA adopted a formal “Gate Review Process” (GR) that was used to review results and focus resources onto promising technologies to reduce costs and identify the technologies that would potentially be carried forward to LFAs within a 10 year period. During the initial discovery phase of the project AREVA took the decision to be relatively hands off and allow our university and National Laboratory partners to be free thinking and consider options that would not be constrained by preconceived ideas from the fuel vendor. To counter this and to keep the partners focused, the GR process was utilized. During this GR process each of the team members presented their findings to a board made up of technical experts from utilities, fuel manufacturing experts, fuel technical experts, and fuel research and development (R&D) experts. During the initial 2 years of the project there were several major accomplishments. These accomplishments, along with the implications for successfully implementing EATF, are; The experimental spark plasma sintering process (SPS) process was successfully used to produce fuel pellets containing either 10% SiC whiskers or nano-diamond particles. The ability to use this process enables the thermal margin enhancements of the fuel additives to be realized. Without the SPS process, the conventional process cannot support adding pellet additives in the required quantities; Coatings of Ti2AlC were successfully applied to Zircaloy-4 cladding. Testing of Ti2AlC coatings at Loss of Cooling Accident (LOCA) conditions showed reduced cladding oxidation compared to present un-coated Zircaloy-4 cladding. This achievement allows the presently used cladding system to be retained so that the 10 year schedule can be met. Having to implement a new cladding material will extend the development schedule beyond 10 years; Several documents were produced to support future development, testing, and licensing of EATF, including a design requirements traceability matrix, a draft business plan, a draft test plan, a draft regulatory plan, and the acceptance criteria for lead fuel assembly insertion into a commercial reactor. This preparatory work lays the foundation for ensuring the future development plans address all the areas required to test, license, and manufacture the new EATF; and In addition, the high velocity oxy-fuel and electrophoretic deposition (EPD) coating application processes were dropped from further consideration due to their inability to meet manufacturing criteria. This allows the resources to be focused on the most promising EATF concepts identified. Future development opportunities that were identified during this work include; The use of SiC or diamond requires that a new pellet production technique (Spark Plasma Sintering), be developed. This entails investment in developing, proving and implementing a new commercial pellet production process. Development of the process to apply thinner coatings is required; Coatings cannot be too “thick” or they will displace a significant volume of water in the core resulting in reduced thermal hydraulic characteristics; Application of the coating at high temperature can affect the Zircaloy substrate. This will require the development and implementation of a new cladding coating manufacturing process; and Replace the Cold Spray (CS) cladding coating application with the Physical Vapor Deposition (PVD) process to eliminate duplication of work and provide greater control over coating thicknesses. This can result in a reduction in the final cycle economic penalty of coatings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Competing effects of electronic and nuclear energy loss on microstructural evolution in ionic-covalent materials

Ever increasing energy needs have raised the demands for advanced fuels and cladding materials that withstand the extreme radiation environments with improved accident tolerance over a long period of time. Ceria (CeO2) is a well known ionic conductor that is isostructural with urania and plutonia-based nuclear fuels. In the context of nuclear fuels, immobilization and transmutation of actinides, CeO2 is a model system for radiation effect studies. Covalent silicon carbide (SiC) is a candidate for use as structural material in fusion, cladding material for fission reactors, and an inert matrix for the transmutation of plutonium and other radioactive actinides. Understanding microstructural change of these ionic-covalent materials to irradiation is important for advanced nuclear energy systems. While displacements from nuclear energy loss may be the primary contribution to damage accumulation in a crystalline matrix and a driving force for the grain boundary evolution in nanostructured materials, local non-equilibrium disorder and excitation through electronic energy loss may, however, produce additional damage or anneal pre-existing defect. At intermediate transit energies where electronic and nuclear energy losses are both significant, synergistic, additive or competitive processes may evolve that affect the dynamic response of materials to irradiation. The response of crystalline and nanostructured CeO2 and SiC to ion irradiation are studied under different nuclear and electronic stopping powers to describe some general material response in this transit energy regime. Although fast radiation-induced grain growth in CeO2 is evident with no phase transformation, different fluence and dose dependence on the growth rate is observed under Si and Au irradiations. While grain shrinkage and amorphization are observed in the nano-engineered 3C SiC with a high-density of stacking faults embedded in nanosize columnar grains, significantly enhanced radiation resistance is attributed to stacking faults that promote efficient point defect annihilation. Moreover, competing effects of electronic and nuclear energy loss on the damage accumulation and annihilation are observed in crystalline 4H-SiC. Systematic experiments and simulation effort are needed to understand the competitive or synergistic effects.

Zhang, Yanwen↗
DOE OSTI2017

Evaluating the irradiation effects on the elastic properties of miniature monolithic SiC tubular specimens

The initial results of a post-irradiation examination study conducted on CVD SiC tubular specimens irradiated under a high radial heat flux are presented herein. The elastic moduli were found to decrease more than that estimated based on previous studies. The significant decreases in modulus are attributed to the cracks present in the specimens. The stresses in the specimens, calculated through finite element analyses, were found to be greater than the expected strength of irradiated specimens, indicating that the irradiation-induced stresses caused these cracks. In conclusion, the optical microscopy images and predicted stress distributions indicate that the cracks initiated at the inner surface and propagated outward.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Steam Corrosion Investigation of Single Silicon Carbide Nanostructures

Silicon carbide (SiC) has been advanced as an alternative to zircaloy for use as an accident-tolerant fuel cladding material in next-generation water-cooled reactors. This is due, in large part, to the predicted superior corrosion resistance of SiC, as compared to zircaloy, particularly at the high temperatures that are obtained in a nuclear reactor under accident conditions. The other favorable performance metrics of SiC include its enhanced mechanical stability, low neutron absorption, low chemical reactivity, and high radiation stability. Past reports, which include theoretical and isolated experimental activities, have studied the corrosion of bulk-scale SiC in reactor-like conditions. This report will present high-temperature steam corrosion results from single SiC nanofiber-based constructs. The changes in microstructure, and the evolution of oxide thickness and SiC recession with progressive corrosion will be characterized at the atomic-to-nanoscale using transmission electron microscopy (TEM). Furthermore, the corrosion-induced degradation in mechanical properties of the SiC nanofibers will be characterized using three-point bending tests involving an atomic-force microscope (AFM) in the contact mode.

36 MATERIALS SCIENCE↗
DOE OSTI2014

Neutronic analysis of candidate accident-tolerant cladding concepts in pressurized water reactors

A study analyzed the neutronics of alternate cladding materials in a pressurized water reactor (PWR) environment. Austenitic type 310 (310SS) and 304 stainless steels, ferritic Fe-20Cr-5Al (FeCrAl) and APMT™ alloys, and silicon carbide (SiC)-based materials were considered and compared with Zircaloy-4. SCALE 6.1 was used to analyze the associated neutronics penalty/advantage, changes in reactivity coefficients, and spectral variations once a transition in the cladding was made. In the cases examined, materials containing higher absorbing isotopes invoked a reduction in reactivity due to an increase in neutron absorption in the cladding. Higher absorbing materials produced a harder neutron spectrum in the fuel pellet, leading to a slight increase in plutonium production. A parametric study determined the geometric conditions required to match cycle length requirements for each alternate cladding material in a PWR. A method for estimating the end of cycle reactivity was implemented to compare each model to that of standard Zircaloy-4 cladding. By using a thinner cladding of 350 μm and keeping a constant outer diameter, austenitic stainless steels require an increase of no more than 0.5 wt% enriched 235 U to match fuel cycle requirements, while the required increase for FeCrAl was about 0.1%. When modeling SiC (with slightly lower thermal absorption properties than that of Zircaloy), a standard cladding thickness could be implemented with marginally less enriched uranium (~0.1%). Moderator temperature and void coefficients were calculated throughout the depletion cycle. Nearly identical reactivity responses were found when coolant temperature and void properties were perturbed for each cladding material. By splitting the pellet into 10 equal areal sections, relative fission power as a function of radius was found to be similar for each cladding material. FeCrAl and 310SS cladding have a slightly higher fission power near the pellet’s periphery due to the harder neutron spectrum in the system, causing more 239 Pu breeding. An economic assessment calculated the change in fuel pellet production costs for use of each cladding. Furthermore, implementing FeCrAl alloys would increase fuel pellet production costs about 15% because of increased 235 U enrichment and the additional UO 2 pellet volume enabled by using thinner cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Interlaboratory Round Robin Testing of SiC/SiC Composite Tubes

With the objective to develop accident tolerant fuel-clad systems for light water reactors, extensive research and development work is being conducted at different governmental, university and industrial labs. Continuous SiC fiber-reinforced SiC matrix (SiC/SiC) composites is one of the leading candidate materials for replacing the traditional Zr alloy cladding due to its high temperature stability, chemical inertness and stability under neutron irradiation. A rich database on the statistics of the mechanical properties of SiC/SiC composite is much needed to carry out the feasibility studies as well as rigorous analysis of the material’s behavior under normal and off-normal reactor conditions. In addition to the properties database, test standards, which ensure the unbiased characterization of the material, are needed. The mechanical properties database, particularly for tubular SiC/SiC specimens, is limited and standardized test methods are still under development. The current work focuses on filling these gaps through an interlaboratory round robin testing program of SiC/SiC composite tube specimen involving several organizations from academia, government laboratories and industry. The difficulties and issues in the testing were identified and addressed through the ORNL lead test. This report summarizes the current status and forward planning on the interlaboratory study.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Uncertainty quantification in elastic constants of SiC f /SiC m tubular composites using global sensitivity analysis

Silicon carbide fiber and silicon carbide matrix (SiC f /SiC m ) tubes produced through the chemical vapor infiltration process have become a candidate cladding material in nuclear applications. The performance of this composite is influenced by many variables such as braiding angle, porosity, material properties, etc., which vary over a range of values due to the inherent fluctuations in the manufacturing process. In this study, the variability in elastic constants of SiC f /SiC m composite has been quantified through multiscale finite element (FE) simulations, variable screening, and high-fidelity surrogate modeling. The key variables dominantly affecting the elastic constants of SiC f /SiC m tubes were identified using global sensitivity analysis. A surrogate to the high-fidelity FE-based model was used in Monte Carlo simulations to generate a hundred thousand samples from which the uncertainty in elastic constants was assessed. It turned out that the coefficient of variation was less than 10%.

Materials Science↗
DOE OSTI2024

Elongate SiC fuel elements

An elongate fuel element is described that has a silicon carbide cladding enclosing a fuel, such as UO 2 , wherein the fuel is dimensioned relative to the cladding to define gaps at each lateral end of the enclosure sufficiently large such that upon swelling in use, the fuel does not increase the strain on the cladding beyond the limits of the claddings strain tolerance. The lateral gaps at the ends of the fuel allow lateral expansion during swelling that reduces the strain on the cladding.

Long, Yun-Bo↗