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At least 217 records · Page 12

Steam Oxidation of FeCrAl and SiC in the Severe Accident Test Station (SATS)

Numerous research projects are directed towards developing accident tolerant fuel (ATF) concepts that will enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the U.S. program, the high temperature steam oxidation performance of ATF solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012 and this facility continues to support those efforts in the ATF community. Compared to the current UO 2 /Zr-based alloy fuel system, alternative cladding materials can offer slower oxidation kinetics and a smaller enthalpy of oxidation that can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. Thus, steam oxidation behavior is a key aspect of the evaluation of ATF concepts. This report summarizes recent work to measure steam oxidation kinetics of FeCrAl and SiC specimens in the SATS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Interlaboratory round robin study on axial tensile properties of SiC-SiC CMC tubular test specimens [Interlaboratory round robin study on axial tensile properties of SiC/SiC tubes]

An interlaboratory round robin study was conducted on the tensile strength of SiC–SiC ceramic matrix composite (CMC) tubular test specimens at room temperature with the objective of expanding the database of mechanical properties of nuclear grade SiC–SiC and establishing the precision and bias statement for standard test method ASTM C1773. The mechanical properties statistics from the round robin study and the precision statistics and precision statement are presented herein. The data show reasonable consistency across the laboratories, indicating that the current C1773–13 ASTM standard is adequate for testing ceramic fiber reinforced ceramic matrix composite tubular test specimen. Furthermore, it was found that the distribution of ultimate tensile strength data was best described with a two–parameter Weibull distribution, while a lognormal distribution provided a good description of the distribution of proportional limit stress data.

36 MATERIALS SCIENCE↗

Severe Accident Test Station (SATS) Activity Report

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO 2 /Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000°C compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development of mechanistic, microstructure-informed BISON models for fission product-induced failure mechanisms in advanced nuclear fuels

The US continues to prioritize commercial deployment of advanced reactors—particularly those utilizing U-Zr metallic and TRISO particle fuels. Fission product-induced failure mechanisms for these systems include fuel–cladding chemical interaction in metallic fuel rods, which threatens cladding integrity, and Pd penetration in TRISO particles, which degrades SiC layer properties. Empirical models can be applied within the bounds of existing irradiation databases, but their utility is limited when considering new designs or investigating fundamental material behaviors. The Nuclear Energy Advanced Modeling and Simulation Program is therefore developing mechanistic models for these behaviors, which are expected to aid in fuel design, assist in development of failure mitigation strategies, and provide support for qualification and licensing. These efforts leverage multiscale capabilities to capture the microstructural features and processes that govern these behaviors. This talk summarizes past and ongoing efforts to develop and validate these models for the BISON fuel performance code.

FCCI↗

Hydrothermal Corrosion of First-Generation Dual-Purpose Coatings on Silicon Carbide for Accident-Tolerant Fuel Cladding

In this study, four different commercial coatings, TiN, Cr, CrN, and Cr/CrN (multilayer) were deposited by physical vapor deposition (PVD) on high resistivity chemical vapor deposited (CVD) SiC for the dual purpose of corrosion mitigation and fission gas hermeticity. The coatings were exposed for up to 2600 hours in flowing high temperature (288-350°C), high purity, liquid water with dissolved hydrogen (0.15-3ppm) or dissolved oxygen (1-2ppm). The Cr/CrN coatings did not adhere well and spalled quickly during exposure. In oxygenated conditions, none of the coatings were protective . CrN coatings spalled preferentially at defect sites, but reacted slowly enough in the absence of defects to potentially be chemically protective. TiN coatings fully oxidized within a few hundred hours and spalled. In hydrogen, all the coatings were chemically stable. Coating failures were identified and attributed to localized spallation, potentially related to coating defects. Coating improvements are recommended to reduce spallation and improve coating protection for future generations of dual-purpose coatings.

36 MATERIALS SCIENCE↗

Potential Impact of Cladding Wettability on LWR Transient Progression

The US Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance. This type of fuel is also known as accident tolerant fuel (ATF). Two promising cladding material candidates with potentially enhanced accident tolerance are silicon carbide (SiC) and an iron-chromium-aluminum (FeCrAl) alloy. The FeCrAl alloy features significantly enhanced oxidation resistance versus the Zr-based alloys currently used in state-of-the-art light water reactors (LWRs). Thermal-fluid characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One characteristic related to boiling heat transfer is the critical heat flux (CHF). At CHF, the efficacy of heat transfer rapidly degrades and components can heat up rapidly. CHF is the heat flux at which the departure from nucleate boiling (DNB) occurs in PWRs. DNB is directly relevant to the departure from nuclear boiling ratio (DNBR), which is the ratio between the critical heat flux and the actual heat flux. Surface wettability is an important factor that influences CHF. CHF plays a significant role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. Significant changes in CHF properties and boiling heat transfer characteristics will impact the progression of transient and accident events and reactor safety characteristics. Additionally, the behavior of the candidate cladding materials post-CHF impacts reactor safety. A brief description of the relationship between cladding material properties and CHF is provided, followed by an analysis of two hypothetical transient scenarios: a super-prompt reactivity-initiated accident (RIA) and an anticipated transient without scram with instability (ATWS-I). The analysis illustrates the potential impact of cladding material properties on safety margins via direct alteration of calculated CHF values. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

An Investigation of the Use of Fully Ceramic Microencapsulated Fuel for Transuranic Waste Recycling in Pressurized Water Reactors

An investigation of the utilization of TRistructural- ISOtropic (TRISO)-coated fuel particles for the burning of plutonium/neptunium (Pu/Np) isotopes in typical Westinghouse four-loop pressurized water reactors is presented. Though numerous studies have evaluated the burning of transuranic isotopes in light water reactors (LWRs), this work differentiates itself by employing Pu/Np-loaded TRISO particles embedded within a silicon carbide (SiC) matrix and formed into pellets, constituting the fully ceramic microencapsulated (FCM) fuel concept that can be loaded into standard LWR fuel element cladding. This approach provides the capability of Pu/Np burning and, by virtue of the multibarrier TRISO particle design and SiC matrix properties, will allow for greater burnup of Pu/Np material, plus improved fuel reliability and thermal performance. In this study, a variety of heterogeneous assembly layouts, which utilize a mix of FCM rods and typical UO2 rods, and core loading patterns were analyzed to demonstrate the neutronic feasibility of Pu/Np-loaded TRISO fuel. The assembly and core designs herein reported are not fully optimized and require fine-tuning to flatten power peaks; however, the progress achieved thus far strongly supports the conclusion that with further rod/assembly/core loading and placement optimization, Pu/Np-loaded TRISO fuel and core designs that are capable of balancing Pu/Np production and destruction can be designed within the standard constraints for thermal and reactivity performance in pressurized water reactors.

Gentry, Cole A↗

Fabrication and Testing of Complex Silicon Carbide Composite Structures

The ability to fabricate larger complex silicon carbide composites is an enabling technology for advanced reactors. Processing techniques to make complex SiC/SiC structures must be versatile enough to facilitate fabrication of both very large components (>1 m), as well as very small feature sizes (<1 mm), minimize the amount of machining required to complete the final part, and must also provide composite material with similar performance to bulk properties previously established for planar SiC/SiC composite material. This paper presents results on fabrication of complex silicon carbide structures and targeted testing aimed specifically at understanding how complex features impact composite performance. Silicon carbide (SiC) and silicon carbide matrix, silicon carbide fiber-reinforced composites (SiC/SiC) are a strong candidate for cladding and core structure used in advanced reactor concepts such as General Atomics' (GA) Energy Multiplier Module (EM{sup 2}) as well as accident tolerant fuel (ATF) in LWRs. SiC retains its strength up to very high temperatures, is stable under irradiation, and shows favorable neutronics and oxidation resistance compared to alloy alternatives. However, unlike metals, SiC cannot be easily machined, formed or welded to make the more complex core components needed to fully realize the benefits of advanced high temperature reactors. TEP processing provides an appealing route to achieving SiC/SiC composite structures with enhanced feature complexity. This is a critical development in the maturation of SiC/SiC for in-core structural materials in advanced reactors. This processing route provides an avenue to make large, dense, complex structures. Experiments have shown retention of strength and pseudo-ductility with hole spacing as small as 4 mm with potential for still smaller feature spacing. The results indicate that feature spacing is not limited by processing effects, but instead by the fiber architecture. The lower limit of feature spacing will likely have implications on potential core loading limits in advanced nuclear reactors utilizing SiC/SiC. (authors)

36 MATERIALS SCIENCE↗

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part B: required research in present tokamaks

Abstract The companion part A paper (Stangeby et al 2022) reports a number of independent estimates indicating that high-duty-cycle DT tokamaks starting with pilot plants will likely experience rates of net erosion and deposition of solid PFC, plasma facing component, material in the range of 10 3 to 10 4 kg yr −1 , regardless of the material used. The subsequent redeposition of such large quantities of material has the potential for major interference with tokamak operation. Similar levels and issues will be involved if ∼continuous low-Z powder dropping is used for surface conditioning of DT tokamaks, independent of the material used for the PFC armor. In Stangeby et al (2022) (part A) it is proposed that for high-duty-cycle DT tokamaks, non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for protecting the main walls, while reducing the risk of degrading the confined plasma. Assessment of whether such an approach is viable will require information, much of which is not available today. Section 6 of part A identifies a partial list of major physics questions that will need to be answered in order to make an informed assessment. This part B report describes R&D needed to be done in present tokamaks in order to answer many of these questions. Most of the required R&D is to establish better understanding of low-Z slag generation and to identify means to safely manage it. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4 ) and silica (SiO 2 ) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Ceramic reinforced zirconium alloy nuclear fuel cladding with intermediate oxidation resistant layer

The invention relates to compositions and methods for coating a zirconium alloy (e.g., ceramic-containing) cladding tube for use with fuel rods in a nuclear water reactor. The coating includes an intermediate oxidation resistant layer and a SiC containing layer at least partially deposited on the intermediate oxidation resistant layer. The SiC containing layer can include a plurality of fibers. The SiC containing layer may also be on the outer surface of end plugs. The invention provides improved capability for the zirconium alloy cladding to withstand normal and accident conditions in the nuclear water reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

EBSD Characterization of Tubular Cladding Plasticity after Bend Testing: A Feasibility Study

The Light Water Reactor Sustainability (LWRS) Program developed a bend test method for tubular samples during FY12. The purpose of this report is to provide details of the Electron Backscattered Diffraction (EBSD) measurements performed on the various sample sets to provide a comparative baseline of the impact that the silicon carbide ceramic matrix composites (SiC-CMC) sleeves had on the deformation and plasticity properties of the Zircaloy-4 (Zr-4) cladding tube.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Mechanical properties of SiC composites neutron irradiated under light water reactor relevant temperature and dose conditions

Silicon carbide (SiC) fiber–reinforced SiC matrix (SiC/SiC) composites are being actively investigated for use in accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this paper examined SiC/SiC composites following neutron irradiation at 230–340 °C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials were chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC) -coated Hi-Nicalon™ Type-S (HNS), Tyranno™ SA3 (SA3), and SCS-Ultra™ (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young's modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young's moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. Finally, this study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Hydrothermal corrosion behavior of CVD SiC in high temperature water

Here, the hydrothermal corrosion of polished and as-cut high purity chemical vapor deposited (CVD) SiC was studied in a constantly refreshing water loop. Light water reactor (LWR) conditions were simulated at 288, 320, and 350 °C with dissolved gas concentrations between 0.15 and 3 ppm H 2 or between 1 and 4 ppm O 2 . In hydrogenated water, the rate of material loss was low, calculated to be ~1.3 μm of recession after 5 years of service in 320 °C water. Moreover, there was no observed localized attack at any temperature. In oxygenated conditions, the corrosion rate was higher, with a calculated material loss >10 μm after 5 years of service in 1 ppm O 2 , 320 °C water. Mass loss significantly increased when grain fallout became significant (as early as 200h with 4 ppm O 2 at 350 °C or after 1000–2000h with 2 ppm O 2 at 288 °C). Grain fallout more than doubled the corrosion rate and a steady state corrosion rate in the grain fallout regime was not observed but expected to eventually occur once large grains begin to be removed. Polished specimens had lower mass loss than unpolished coupons. A kinetic analysis of the data in this work suggests that the corrosion rates are controlled by a single activation step in both oxygenated and deoxygenated conditions, with the reaction order with respect to oxygen being 1. A resulting reaction rate equation to predict corrosion of SiC (in mg/cm 2 s) in high purity water from 288 to 350 °C and up to 4 ppm O 2 was constructed: Rate = $\frac{0.1458}{1+SA}$ T (1.09(1–10 -3 T)[O 2 ]e– $\frac{1.275x10^4}{T}$ + 7.91x10 -6 e– $\frac{7.39x10^3}{T}$).

36 MATERIALS SCIENCE↗

PIE of nuclear grade SiC/SiC flexural coupons irradiated to 10 dpa at LWR temperature

Silicon carbide fiber-reinforced SiC matrix (SiC/SiC) composites are being actively investigated for accident-tolerant core structures of light water reactors (LWRs). Owing to the limited number of irradiation studies previously conducted at LWR-coolant temperature, this study examined SiC/SiC composites following neutron irradiation at 230–340°C to 2.0 and 11.8 dpa in the High Flux Isotope Reactor. The investigated materials are chemical vapor infiltrated (CVI) SiC/SiC composites with three different reinforcement fibers. The fiber materials were monolayer pyrolytic carbon (PyC)-coated Hi-NicalonTM Type-S (HNS), TyrannoTM SA3 (SA3), and SCS-Ultra TM (SCS) SiC fibers. The irradiation resistance of these composites was investigated based on flexural behavior, dynamic Young’s modulus, swelling, and microstructures. There was no notable mechanical properties degradation of the irradiated HNS and SA3 SiC/SiC composites except for reduction of the Young’s moduli by up to 18%. The microstructural stability of these composites supported the absence of degradation. In addition, no progressive swelling from 2.0 to 11.8 dpa was confirmed for these composites. On the other hand, the SCS composite showed significant mechanical degradation associated with cracking within the fiber. This study determined that SiC/SiC composites with HNS or SA3 SiC/SiC fibers, a PyC interphase, and a CVI SiC matrix retain their properties beyond the lifetime dose for LWR fuel cladding at the relevant temperature.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Hydrothermal Corrosion of Coatings on Silicon Carbide in Boiling Water Reactor Conditions

SiC/SiC ceramic matrix composites are an attractive material for use as accident tolerant fuel cladding. However, despite its resistance to corrosion in high-temperature steam, SiC is known to dissolve in high-temperature water, necessitating the use of corrosion-resistant coatings. Here, samples were exposed in 288°C water with 2 ppmw dissolved oxygen to simulate boiling water reactor (BWR) normal water chemistry (NWC) and 288°C water with 150 ppbw dissolved hydrogen to simulate BWR hydrogen water chemistry. Several candidate coatings were tested: Cr, CrN, TiN, and ZrN. All coatings were applied by physical vapor deposition onto chemical vapor deposition SiC (CVD-SiC) substrates. Two uncoated SiC/SiC variants were also exposed and degraded more rapidly compared to CVD-SiC, especially in NWC. Cr and CrN coated specimens were found to have low mass change, suggesting they are promising candidate coatings. Lastly, TiN coatings were found to oxidize rapidly, suggesting that coatings would fully oxidize during service in BWR environments. ZrN coatings dissolved completely during exposure.

36 MATERIALS SCIENCE↗

X-ray characterization of atomistic defects causing irradiation creep of SIC

Silicon carbide (SiC) fiber–reinforced SiC matrix composites have been widely studied for use in developing an accident-tolerant fuel (ATF) cladding for light water reactors (LWRs) in recent years because of their superior oxidation and irradiation resistance and low neutron absorption cross section. An important task is the establishment of the thermo-mechanical analysis capability to predict the in-pile performance and failure probability of the cladding. Irradiation creep is known to be a key property in modeling the stress state under irradiation. To better understand the radiation creep mechanism from an atomic point of view, this study conducted synchrotron-based X-ray diffraction (XRD) experiments on SiC neutron-irradiated with and without applied stress at the National Synchrotron Light Source-II (NSLS-II). This report presents progress of the XRD study and future plans. The important finding obtained was that the macroscopic creep strain and strain due to the lattice expansion were similar in case of neutron irradiation at 300°C to 0.1 displacement per atom (dpa), which indicates that the lattice strain is a major source of the transient creep strain.

36 MATERIALS SCIENCE↗