Kurt A. Terrani

Effect of steam on high temperature oxidation behaviour of alumina-forming alloys

Abstract Alternative light water reactor fuel cladding materials are being investigated to replace Zircaloy for enhanced accident tolerance, which involves oxidation resistance to steam environments at ≧1200°C for short times. As chromia-forming alloys and Ni-containing alloys are both undesirable for this application, the focus has been on FeCrAl, although NiAl was used to evaluate the effect of steam oxidation at 1600°C for this study. For commercial and model FeCrAlY alloys, a critical Cr–Al composition was identified for 1 bar isothermal steam (100% H2O) oxidation resistance at 1200°C, which differed for exposures in Ar–50%H2O at the same temperature. Alloys with lower Cr and Al contents were not able to form a protective alumina scale under these conditions. To simulate the accident scenario, exposures were also conducted in steam with the temperature rising 5°C min−1 to 1500°C for the most oxidation resistant alloys. Using thermogravimetry, the maximum use temperature for candidate alloys was determined for different Cr and Al contents. Minor additions such as Y and Ti appeared to be beneficial for oxidation resistance. Similar to prior studies, alumina scales formed in air and in steam appeared to have only subtle differences in microstructure.

Systematic Technology Evaluation Program for SiC/SiC Composite-based Accident-Tolerant LWR Fuel Cladding and Core Structures: Revision 2015

Fuels and core structures in current light water reactors (LWR’s) are vulnerable to catastrophic failure in severe accidents as unfortunately evidenced by the March 2011 Fukushima Dai-ichi Nuclear Power Plant Accident. This vulnerability is attributed primarily to the rapid oxidation kinetics of zirconium alloys in a water vapor environment at very high temperatures. Zr alloys are the primary material in LWR cores except for the fuel itself. Therefore, alternative materials with reduced oxidation kinetics as compared to zirconium alloys are sought to enable enhanced accident-tolerant fuels and cores.

Neutronics Studies of Uranium-Bearing Fully Ceramic Microencapsulated Fuel for Pressurized Water Reactors

Abstract This study evaluated the neutronics and some of the fuel cycle characteristics of using uranium-based fully ceramic microencapsulated (FCM) fuel in a pressurized water reactor (PWR). Specific PWR lattice designs with FCM fuel have been developed that are expected to achieve higher specific burnup levels in the fuel while also increasing the tolerance to reactor accidents. The SCALE software system was the primary analysis tool used to model the lattice designs. A parametric study was performed by varying tristructural isotropic particle design features (e.g., kernel diameter, coating layer thicknesses, and packing fraction) to understand the impact on reactivity and resulting operating cycle length. To match the lifetime of an 18-month PWR cycle, the FCM particle fuel design required roughly 10% additional fissile material at beginning of life compared with that of a standard uranium dioxide (UO2) rod. Uranium mononitride proved to be a favorable fuel for the fuel kernel due to its higher heavy metal loading density compared with UO2. The FCM fuel designs evaluated maintain acceptable neutronics design features for fuel lifetime, lattice peaking factors, and nonproliferation figure of merit.

Synthesis of phase-pure U2N3 microspheres and its decomposition into UN

Uranium mononitride (UN) is important as a nuclear fuel. Fabrication of UN in its microspherical form also has its own merits since the advent of the concept of accident-tolerant fuel, where UN is being considered as a potential fuel in the form of TRISO particles. However, not many processes have been well established to synthesize kernels of UN. Therefore, a process for synthesis of microspherical UN with a minimum amount of carbon is discussed herein. First, a series of single-phased microspheres of uranium sesquinitride (U2N3) were synthesized by nitridation of UO2+C microspheres at a few different temperatures. Resulting microspheres were of low-density U2N3 and decomposed into low-density UN. The variation of density of the synthesized sesquinitrides as a function of its chemical composition indicated the presence of extra (interstitial) nitrogen atoms corresponding to its hyperstoichiometry, which is normally indicated as α-U2N3. Average grain sizes of both U2N3 and UN varied in a range of 1-2.5 μm. These also had a considerably large amount of pore spacing, indicating the potential sinterability of UN toward its use as a nuclear fuel.

Accident Tolerant FeCrAl Fuel Cladding: Current Status Towards Commercialization

FeCrAl alloys are rapidly becoming mature candidate alloys for accident tolerant fuel applications. The FeCrAl material class has shown excellent oxidation resistance in high-temperature steam environments, a key aspect of any accident tolerant cladding concept, while also being corrosion resistant, stress corrosion cracking (SCC) resistant, irradiation-induced swelling resistant, weldable, and formable. Current research efforts are focused on design, development and commercial scaling of advanced FeCrAl alloys including large-scale, thin-walled seamless tube production followed by a broad spectrum of degradation evaluations in both normal and off-normal conditions. Included in this discussion is the theoretical analysis of the alloying principles and rules, alloy composition design, and overview of the most recent empirical database on possible degradation phenomena for FeCrAl alloys. The results are derived from extensive in-pile and out-of-pile experiments and form the basis for near-term deployment of a lead-test rod and/or assembly within a commercially operating nuclear power plant.

Application of X-ray microcomputed tomography in the characterization of irradiated nuclear fuel and material specimens

X‐ray microcomputed tomography (μCT) was applied in characterizing the internal structures of a number of irradiated materials, including carbon‐carbon fibre composites, nuclear‐grade graphite and tristructural isotropic‐coated fuel particles. Local cracks in carbon‐carbon fibre composites associated with their synthesis process were observed with μCT without any destructive sample preparation. Pore analysis of graphite samples was performed quantitatively, and qualitative analysis of pore distribution was accomplished. It was also shown that high‐resolution μCT can be used to probe internal layer defects of tristructural isotropic‐coated fuel particles to elucidate the resulting high release of radioisotopes. Layer defects of sizes ranging from 1 to 5 μm and up could be isolated by tomography. As an added advantage, μCT could also be used to identify regions with high densities of radioisotopes to determine the proper plane and orientation of particle mounting for further analytical characterization, such as materialographic sectioning followed by optical and electron microscopy. In fully ceramic matrix fuel forms, despite the highly absorbing matrix, characterization of tristructural isotropic‐coated particles embedded in a silicon carbide matrix was accomplished using μCT and related advanced image analysis techniques.

A Challenge to Multivariate Statistical Analysis: Spent Nuclear Fuel

Nuclear fission accounts for most of the non-polluting, non-fossil-fuel electrical power in the world. Higher burnup of fuel – that is, using a given fuel bundle for a longer time to produce more power – reduces the uranium resources needed, greatly enhances the economics of nuclear electricity, and reduces the amount of spent fuel for disposal. However, as the fuel burnup progresses, the fission process builds up large atomic fractions of fission products, consisting of most elements in the central region of the periodic table, in the fluorite UO2 matrix; and, a fuel/clad interaction (FCI) layer forms at the interface between the oxide fuel and the Zircaloy cladding. Providing a scientific basis for understating fuel behavior in the high burnup regime requires detailed characterization of high-burnup urania fuel. We have used X-ray spectrum imaging (SI) in SEM and STEM to analyze high-burnup (irradiated for 7 eighteen month long cycles to average burnup of 72 GWd/MTU) fuel from the H.B. Robinson pressurized water reactor. Multivariate statistical analysis (MVSA) is irreplaceable for understanding the extremely complex chemistry found.

Steam Oxidation Behavior of FeCrAl Cladding

In order to better understand the high temperature steam oxidation behavior of FeCrAl alloys, this study addressed two topics. The first is continuing to evaluate the effect of alloy composition on performance of commercial and laboratory-made candidate FeCrAl alloys. For a few optimized compositions, this includes the performance of commercially-made tubing where it is clear that dropping the Cr content from 20% to 10–13% reduces the maximum operating temperature in steam by ~50 °C. The second addresses more realistic accident conditions. Model FeCrAl compositions that were exposed in ~300 °C water for 1 year were subsequently “ramp” tested in steam at 5 °C/min to 1500 °C to assess the effect of the Fe-rich oxide formed in water on the subsequent steam oxidation resistance. For Fe-18Cr-3Al+Y, the 1 year exposures in three different LWR water chemistries did not affect the ability to form alumina to 1500 °C. However, for marginal alloys Fe-13Cr-4Al and Fe-10Cr-5Al, some specimens began forming voluminous Fe-rich oxide at lower temperatures.

Characterization of the Hydrothermal Corrosion Behavior of Ceramics for Accident Tolerant Fuel Cladding

Accident-tolerant fuel (ATF) is an increasingly important research topic for the nuclear industry, and ceramics such as SiC are strong contenders for deployment as ATF cladding. The hydrothermal corrosion characteristics of SiC and Al2O3 were investigated via constantly-refreshing autoclave corrosion and post exposure characterization. Four different types of chemical vapor deposited (CVD) SiC specimens were examined (two with high electrical resistance, one with low electrical resistance, and a single crystal 4H structural variant). Al2O3 specimens were prepared in single crystal and polycrystalline states. PWR primary water, BWR–HWC, and BWR–NWC environments were maintained throughout the experiments. Characterization conducted using SEM and EDS was used to determine factors affecting corrosion rates and susceptibility to grain boundary attack in each water chemistry condition. Raman spectroscopy was also used to determine chemical variation of the surface with corrosion. Grain boundary attack was found to be significant for both alumina and SiC polycrystalline variants.

Carbothermic Synthesis of ~820- m UN Kernels. Investigation of Process Variables

This report details the continued investigation of process variables involved in converting sol-gel-derived, urainia-carbon microspheres to ~820-μm-dia. UN fuel kernels in flow-through, vertical refractory-metal crucibles at temperatures up to 2123 K. Experiments included calcining of air-dried UO3-H2O-C microspheres in Ar and H2-containing gases, conversion of the resulting UO2-C kernels to dense UO2:2UC in the same gases and vacuum, and its conversion in N2 to in UC1-xNx. The thermodynamics of the relevant reactions were applied extensively to interpret and control the process variables. Producing the precursor UO2:2UC kernel of ~96% theoretical density was required, but its subsequent conversion to UC1-xNx at 2123 K was not accompanied by sintering and resulted in ~83-86% of theoretical density. Decreasing the UC1-xNx kernel carbide component via HCN evolution was shown to be quantitatively consistent with present and past experiments and the only useful application of H2 in the entire process.