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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Thermal Conductivity and Diffusivity for SiC Fibers for Use in ATF Cladding Composites

As part of the accident tolerant fuels (ATF) initiative, there is a need to understand the thermophysical properties of cladding for these fuels. One potential cladding material is silicon carbide (SiC) ceramic matrix composite reinforced with SiC fibers because of the corrosion resistance, resistance to hydrogen formation, high bulk thermal conductivity, and favorable mechanical properties of SiC. However, data about the thermal properties of individual SiC fibers in these composites is absent in the literature. This paper presents the measured axial thermal conductivity and thermal diffusivity of a leading SiC fiber (HiNicalon Type S) via a thermal measurement technique (transient electrothermal technique, TET) optimized for thin fibers. The results show that the measured properties are significantly lower than bulk values for SiC. Because of the low neutron absorption, as well as excellent thermal and mechanical properties, zircaloy has primarily been used for the fuel cladding in many nuclear reactors. A major drawback of the zircaloy is its hydrogen generation vulnerability. This issue has been known for some time, but the explosion of hydrogen accident in the Fukushima Daiichi nuclear disaster has motivated efforts for accident tolerant fuels (ATF) and their cladding as suitable substitutes for zircaloy. Fuel cladding made of silicon carbide (SiC) ceramic matrix composite (CMC) is an attractive alternative to the zircaloy. Investigation on the tubular monolithic a-SiC polytropes indicates that it has superior oxidation (hydrogen production) resistance (three orders of magnitude slower) than zircaloy. Additionally, it has well-documented, favorable mechanical properties, along with high temperature corrosion resistance. However, its bulk thermal properties can vary significantly depending on polytrope, the grain size, grain boundary, or impurities. At room temperature, SiC has a thermal conductivity as high as 490 Wm{sup -1}K{sup -1}, where the {sup 6}H polytrope has the highest conductivity. Thermal measurements of the SiC CMC are usually accomplished by laser flash method for thermal diffusivity (a) and thermal conductivity is indirectly calculated through the relationship k=aρcp. Depending on the manufacturing method and the reinforced fiber, the composite thermal conductivity varies from <10 Wm{sup -1}K{sup -1} to ∼70 Wm{sup -1}K{sup -1}. While the flash method is convenient for the measurement of the composite, it is difficult to apply to the measurement of individual fibers. Because of this, the thermal properties of individual SiC fibers is not present in the literature. Because the SiC fiber manufacturing process is different than that of the SiC matrix in the CMC, the properties of the bulk material might be erroneous when applied to the fiber. Being able to measure the fiber properties independently and directly allows for an understanding of the effect of the manufacturing process on the fiber and the opportunity to potentially improve or control the thermal properties of the composite. This improved accuracy can be achieved by measuring the fiber with a technique optimized for individual fiber measurement, known as the Transient Electrothermal Technique or TET. This paper presents the thermal conductivity and diffusivity of a leading SiC fiber (HiNicalon Type S). Measurements are performed with the TET, with the resulting properties being significantly lower than the bulk values of SiC, which is one possible explanation for relatively low CMC values. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Duplex accident tolerant coating for nuclear fuel rods

A method is described for forming duplex layers including an interlayer and a corrosion resistant boundary layer on a nuclear fuel rod cladding tube for use in a water cooled nuclear reactor. The method includes forming, by thermal deposition or physical vapor deposition, on the exterior of a substrate, an inner interlayer with Mo, Ta, W or Nb or other particles, and forming, by thermal deposition or physical vapor deposition, on the interlayer, an outer corrosion resistant layer with particles selected from the group consisting of Cr, a Cr alloy, and combinations thereof. The interlayer prevents eutectic formation between the corrosion resistant layer and the substrate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Photoluminescence in Tagging: A Glowing Review

This literature review explores the options of using photoluminescence for tagging uranium fuel and zirconium cladding in order to easily and quickly associate interdicted material to its facility of origin or pathway after loss of institutional control. Photoluminescence is the spontaneous emission of radiation by a substance when excited by interaction with light. The source of photoluminescence can be from organic, inorganic, and organometallic compounds. Each luminophore, or compound responsible for luminescent properties, has individual strengths and weaknesses, depending on the intended use. Many factors can affect the brightness, lifetime, color, and detectability of luminophores, such as humidity, temperature, oxygen availability, noise from radiation–induced luminescence, and the presence of quenching substances like iron. Application of luminescent compound to the surface of an item can be done using paint, dye, thin film, thick film, or electrochemical coating. Several of the application techniques may have secondary functions such as corrosion resistance and detection, heat resistance, protective properties, and more. When considering cladding, there are various successful studies on the application of luminescent coatings on zirconium. However, significant research will be necessary to explore which luminophore and which application technique would best fit the intended use. In order to proceed with this research, a baseline of the natural luminescence of the intended base material (nuclear fuel or cladding) is necessary. While there may be issues with background noise, quenching from the environment, quenching from the substrate, signal to noise weakness, etc. the multitude of options may make photoluminescent tagging an option. In summary, this review has demonstrated that photoluminescence is a potential option to tag nuclear fuel and cladding, but additional paper–studies, testing, and research is required to understand the compatibility of luminescent tags in nuclear reactor environments, manufacturability, and detectability.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Advanced 3D characterisation of iodine induced stress corrosion cracks in zirconium alloys

Highlights: • Macroscopic crack direction is affected by metallurgical state. • High resolution XCT reveals complex crack morphology. • 3D-EBSD used to fully characterise attacked grain boundaries. • Evidence of iodine segregation ahead of a crack tip. - Abstract: In this study, a range of advanced techniques have been used to characterise iodine-induced stress corrosion cracks, and investigate their interaction with microstructure in zirconium alloys used as cladding material in nuclear power reactors. C-ring samples, machined from cladding tubes in a recrystallised and cold worked condition, were subjected to compressive stress while submerged in an iodine-ethanol solution to simulate stress corrosion cracks arising from pellet cladding interaction. The complex morphology of these stress corrosion cracks was imaged in 3D using X-ray computed tomography (XCT), and subsequent analysis using both 2D and 3D electron backscatter diffraction (EBSD) allowed for the interaction of these cracks with microstructure to be investigated. Intergranular cracking modes were observed to dominate in the recrystallised microstructure, resulting in cracking in the radial direction with variations on the order of the grain size. Crack propagation in the cold worked microstructure was observed to be correlated with the direction of the elongated grains, resulting in a more tangential crack path with less local variation and a microstructure potentially more susceptible to iodine-induced stress corrosion cracking. 3D-EBSD using serial sectioning allows for the determination of the full five-parameter grain boundary description, and opens up the possibility of a statistical analysis of susceptible boundaries to SCC. Nanoscale secondary ion mass spectrometry (NanoSIMS) was also used to map the chemical distribution in the crack tip region and provides possible evidence of iodine segregation ahead of the crack tip. The results provide new insight into crack propagation mechanisms in recrystallised and cold worked microstructures, and pave the way for a more mechanistic understanding of crack propagation through advanced characterisation of iodine-induced stress corrosion cracks in zirconium alloys.

36 MATERIALS SCIENCE↗
DOE OSTI2016

First-principles study of mechanical properties of zirconium alloys and hydrides

Zirconium alloys have been extensively used as fuel cladding materials in nuclear light water reactors (LWRs) due to their favorable mechanical properties, corrosion resistance, low thermal neutron capture cross section and criticality. Because the main functions of these structural materials are to maintain the integrity of UO{sub 2} fuel rods and prevent the release of fission products into the biosphere, a detailed knowledge of their structural stability is of critical importance to ensure the safe operation of nuclear reactors, and the safety of storage, transportation or disposition of used nuclear fuel assemblies. While the Zr-H phase diagram has been known since the 1950's and has remained largely unchanged, the structures, formation mechanisms, and stability of Zr-hydride phases in pure α-Zr or Zr alloys matrices have been subjects of much debate over the last few decades. The formation of hydride phases in the Zr matrix has been linked in previous studies to the degradation of mechanical properties of Zr/Zr-alloys materials. Therefore, an accurate knowledge of the interplay between the structures, stability and mechanical properties of Zr/Zr-alloys and Zr hydrides is of paramount significance to ensure materials reliability during their in- service lifetime. Despite few first-principles studies have investigated the mechanical stability of Zr hydrides polymorphs proposed so far, no ab initio studies have been carried out, to the best of our knowledge, on the mechanical properties of modern nuclear-grade zirconium alloys such as Zircaloy-4 (Zry-4; 1.2-1.7% Sn), Zirconium low oxidation alloy (ZIRLO; 0.7-1.0% Sn and ∼1.0% Nb), or M5 alloy (0.8-1.2% Nb). Therefore, it is important to carry out a series of first-principles calculations systematically to investigate the structure-property relationship of zirconium hydrides and zirconium alloys, that can help explain differences in elastic properties between high-purity zirconium and nuclear-grade zirconium alloys. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Modeling Approach for the Aluminum-clad Dry Storage Pilot using HFIR Fuel

To confirm that the dry storage of aluminum-clad research reactor spent nuclear fuel (ASNF) will remain within the safety envelope after applied drying schemes and that the resulting evolution of the gas space composition, temperature, and pressure conditions are understood, a dry storage pilot project is being established. The pilot will incorporate an instrumented lid for discrete interval or for on-demand gas composition and temperature monitoring of two DOE Standard Canisters (DSCs) loaded with three High Flux Isotope Reactor (HFIR) inner cores per DSC. Each DSC would be subjected to a separate alternative candidate drying scheme. Canisters will undergo 1 to 5 years of monitoring, including internal temperature and gas sampling to track pressure and composition changes. This report outlines the approach for modeling the ASNF-in-canister behavior in terms of evolving gas space conditions for the ASNF dry storage pilot using HFIR fuel. The ASNF has an adherent surface oxyhydroxide layer comprised of boehmite/bayerite that generates hydrogen when subjected to irradiation. Three-dimensional multi-physics computational fluid dynamics simulations will be executed to compute the thermal field within the DSC and provide inputs to a chemical model employed to compute pressure buildup as hydrogen is generated in the system. Implemented in Cantera, the chemical model solves gas phase and aluminum oxyhydroxide surface-mediated radiolysis reactions. Gas phase reactions are sourced from Wittman and Hanson (2015), whereas surface-mediated reactions are incorporated by fitting experimental data using an optimization algorithm (Abboud, 2023). Water radiolysis reactions from Wren and Ball (2001) are adopted with modifications as described in Abboud (2023c). Understanding the effect of the hydrogen buildup over time is important for long-term storage safety considerations. Modeling results will include the canister pressure, temperature, and composition evolution from the initial helium backfill with the addition of radiolytically-evolved chemical species (e.g., hydrogen and oxygen). The specific HFIR cores for the pilot program have not yet been selected, and the overall design is still in development. The CFD-chemical model used for this work will be based on prior models with necessary updates to allow for improved accuracy and efficiency. The experimental data obtained from the HFIR demonstration will be used to improve and validate the computational models to predict the ASNF-in-canister behavior.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Zr Nuclear Data Campaign: Measurement of (n,γ) cross section of 90,91 Zr [Abstract]

The isotopes of Zr with mass number A = [90; 91; 92; 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications due to their desirable mechanical and nuclear properties. For example, the zirconium alloys are used as cladding material in nuclear reactors. In addition to the mechanical properties, Zr-90 has a closed neutron shell, hence the capture cross sections of the zirconium isotopes are relatively small, such that they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σ γ (E), of Zr isotopes. For incident neutron energies < 0.1 MeV, the ENDF library reports a relative uncertainty of approximately 20% and uncertainty greater than 20% above this energy. This motivates the Nuclear Criticality Safety Program (NCSP) to perform new neutron induced cross section measurements to accurately evaluate these four isotopes of Zr.

ENDF↗
DOE OSTI2017

Establishing New Capability of High Temperature Electrochemical Impedance Spectroscopy Techniques for Equilibrium and Kinetic Experiments

In nuclear fuel and cladding materials, changes in structure and chemistry over the lifetime of the fuel can have a significant impact on fuel and reactor performance. Currently, chemical evolution over the lifetime of the fuel can only be inferred using post-irradiation examination. As an alternative to post-irradiation measurements, electrochemical impedance spectroscopy (EIS) can provide an in-pile measurement of changes in chemistry in oxides and hydrides coupling with modeling such as finite element (FE) and other material characterization techniques such as atom probe tomography (APT), Raman spectroscopy, in-situ Transmission Electron Microscopy (TEM), and synchrotron X-ray microscopy. The objective of this work package is to develop EIS techniques for the in-pile measurement of changes in solid state chemistry of fuels and materials in coupling with model simulation and material characterization. This work package will involve the initial development of electrochemical sensing technologies for measuring spatial- and time-resolved changes in fuel and cladding chemistry. Specific attention will be paid to monitoring changes in fuel stoichiometry, monitor cladding hydride formation and deformation, and cladding corrosion. This report summarized the FY17 achievements of this work package. A laboratory was begun to set up from scratch. The laboratory place was located; necessary equipment and materials were purchased and some were delivered; laboratory instruction (LI) for conducting experiments was drafted for review; radiation level of depleted UO2 that will be investigated as the potential testing sample was valued for safety issue; a testing plan for measuring changes in stoichiometry of oxide fuels has been made and roadblocks for making measurements at high temperature have been pinpointed; in addition, collaboration with Boise State University (BSU) and Massachusetts Institute of Technology (MIT) has been established. In short, the capability of applying EIS for in-pile measurement of changes of fuels and cladding materials was initiated to be established.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Nanocrystalline SiC and Ti 3 SiC 2 Alloys for Reactor Materials: Annual Report

A new dual-phase nanocomposite of Ti 3 SiC 2 /SiC is being synthesized using preceramic polymers, ceramic powders, and carbon nanotubes (CNTs) designed to be suitable for advanced nuclear reactors and perhaps as fuel cladding. The material is being designed to have superior fracture toughness compared to SiC, adequate thermal conductivity, and higher density than SiC/SiC composites. This annual report summarizes the progress towards this goal and reports progress in understanding certain aspects of the material behavior but some shortcomings in achieving full density or in achieving adequate incorporation of CNTs. The measured thermal conductivity is adequate and falls into an expected range based on SiC and Ti 3 SiC 2 . Part of this study makes an initial assessment for Ti 3 SiC 2 as a barrier to fission product transport. Ion implantation was used to introduce fission product surrogates (Ag and Cs) and a noble metal (Au) in Ti 3 SiC 2 , SiC, and a synthesized at PNNL. The experimental results indicate that the implanted Ag in SiC is immobile up to the highest temperature (1273 K) applied in this study; in contrast, significant out-diffusion of both Ag and Au in MAX phase Ti₃SiC₂ occurs during ion implantation at 873 K. Cs in Ti₃SiC₂ is found to diffuse during post-irradiation annealing at 973 K, and noticeable Cs release from the sample is observed. This study may suggest caution in using Ti 3 SiC 2 as a fuel cladding material for advanced nuclear reactors operating at very high temperatures. Progress is reported in thermal conductivity modeling of SiC-based materials that is relevant to this research, as is progress in modeling the effects of CNTs on fracture strength of SiC-based materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2016

Study on Modeling of Nuclear Fuel End Drop during a Hypothetical Accident

It is well known that shell elements have been used to model the cladding of single nuclear fuel pins. These have been used, for example, to analyze the response of a fuel assembly by assuming that all fuel pins in a fuel assembly deform in the same pattern. It is increasingly significant to model the entire fuel assembly in order to capture more realistically the behavior of each individual nuclear fuel during a hypothetical accident in a manner that can account for interactions among single pins. However, modeling each individual fuel pin in an entire fuel assembly using shell elements will result in models with large amount of elements and degrees of freedom that take correspondingly larger amount of computational times to execute. Therefore, modeling the nuclear fuel assembly with beam elements may become an option. This paper presents the results of a set of analysis with the LS-DYNA finite element software that investigated the differences and similarities of modeling nuclear fuel assemblies using either shell elements or beam elements. This study provides additional insights on the appropriateness of using beam element models of the fuel pins in the simulation of a hypothetical accident. In this study, two single pin models, a single pin model using shell elements and a single beam model using beam elements, are constructed to analyze end drop scenarios with the LS-DYNA code. Each model is then analyzed for a 19 ft end drop scenario (linear behavior) and 30 ft end drop scenario (non-linear behavior). Comparisons are made between the results obtained with the single pin shell model and with the single pin beam model to study the differences and similarities resulting from the two modeling approaches. In this study, a B and W 15x15 PWR fuel is analyzed for both 30 ft and 19 ft end hypothetical accident drops using both single pin finite element model with shell elements and beam elements for the fuel cladding. It is found that for the 19 ft end drop, the difference in the Max/Min total axial strains is about 12% between the two models. For the 30 ft end drop, the maximum total axial strains calculated with both models is similar, and a difference of about 12% is observed for the minimum total axial strains calculated by the two models. Advantages of the single pin shell model are that it permits consideration of internal pressures and may capture local buckling. Therefore, use of beam elements to model the fuel cladding needs to take into account the possible significance of local buckling and internal pressurization for the problem being analyzed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Estimation of Core Recovery Time with TRACE

Following the severe accident in Fukushima Daiichi Nuclear Power Plant in Japan, safety related concerns about the zirconium based claddings used in current light water reactors have increased. High corrosion resistance and low neutron absorption cross section make zirconium a suitable cladding material for normal nuclear plant operational conditions. However, for nuclear accident scenarios when safety functions of the plant are lost, Zr claddings undergo exothermic reaction with steam and exacerbate the accident conditions. The exothermic reaction increases the temperature of the cladding and accelerates core melting in severe accident progression. And the generation of explosive hydrogen gas accumulates and challenges the structural integrity. The accident tolerant fuel (ATF) concept is the fuel and cladding system that gives a longer time margin, compared to standard Zr claddings, to cope with the accident when core cooling systems are lost in a nuclear power plant. The ATF cladding should have slower oxidation kinetics and less hydrogen generation than Zr to alleviate consequences of cladding interaction with steam at high temperatures. The maximum time available to mitigate progression of severe accident is an important measure to evaluate benefits of ATF claddings. For this work, the maximum time available to mitigate progression of severe accident is selected as the time at which cladding melting reaction starts. Melting reaction is selected as an upper limit because of the degradation of coolable geometry. Maintaining the coolable geometry is a regulation criteria for both loss of coolant (LOCA) analyses and outside the LOCA analyses. Once the melting of the cladding starts, it exacerbates the deformation of the fuel assembly and hottest rods within the core cannot be cooled by water reflood and results in fuel melting progression. The aim of this work is to evaluate time margin gain of ATF cladding in a severe accident scenario of LOCA without safety injections and station blackout considering oxidation kinetics only. Also it is assumed that safety injections are recovered before melting of the claddings start and the sensitivity of safety injection actuation time on core recovery is analyzed. The core recovery time is the time when makeup water from safety injections quench the fuel rods and halt the progression of the severe accident. This time is estimated by initiating safety injections at different times and observing the hot rod behavior which represents 1/5. of the core in this work. It is considered that core is not recoverable when 20% of the core is damaged.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary analysis of core heat-up for ATFs under accident conditions

The accident-tolerant fuel (ATF) system is aim at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the conventional UO{sub 2}-Zr fuel system when the reactor is in normal operation. In this paper, the improved performance of UO{sub 2}-SiC, UO{sub 2}-FeCrAl fuel systems, and of fully ceramic micro-encapsulated (FCM) fuels: FCM-SiC and FCM-FeCrAl, on the core heat-up progress in a pressurized water reactor (PWR) is analyzed by using the best-estimate system code RELAP5 and severe accident analysis program. For this purpose two accident scenarios are selected: a large break LOCA as a design base accident and an extended station blackout (SBO) severe accident. For the LBLOCA with just low-pressure safety injection, larger volumetric heat capacity and thermal conductivity of ATF materials contribute to a larger safety margin in terms of the peak cladding temperature compared with UO{sub 2}-Zr. Thanks to higher melting temperatures, FCM-SiC is failed by melting almost 1200 s later than UO{sub 2}-Zr. The accumulated production of hydrogen for ATF materials is around 10 times smaller than that for UO{sub 2}-Zr. However, high temperature oxidation of SiC cladding generates carbon monoxide which is combustible, while the melting point of FeCrAl is not high enough. In the SBO scenario with mitigation measure, the advantage of ATF is more evident. If the mitigation measure could not be taken, core heatup is inevitable for all types of fuel system, but the temperature rising of ATF core is much slower. In general, improved high-temperature properties and resistance to oxidation play important roles for ATF cores.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Anisotropic Azimuthal Power and Temperature distribution on FuelRod: Impact on Hydride Distribution

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗