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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

Simulations of Failure via Three-Dimensional Cracking in Fuel Cladding for Advanced Nuclear Fuels

Enhancing performance of fuel cladding and duct alloys is a key means of increasing fuel burnup. This project will address the failure of fuel cladding via three-dimensional cracking models. Researchers will develop a simulation code for the failure of the fuel cladding and validate the code through experiments. The objective is to develop an algorithm to determine the failure of fuel cladding in the form of three-dimensional cracking due to prolonged exposure under varying conditions of pressure, temperature, chemical environment, and irradiation. This project encompasses the following tasks: 1. Simulate 3D crack initiation and growth under instantaneous and/or fatigue loads using a new variant of the material point method (MPM); 2. Simulate debonding of the materials in the crack path using cohesive elements, considering normal and shear traction separation laws; 3. Determine the crack propagation path, considering damage of the materials incorporated in the cohesive elements to allow the energy release rate to be minimized; 4. Simulate the three-dimensional fatigue crack growth as a function of loading histories; 5. Verify the simulation code by comparing results to theoretical and numerical studies available in the literature; 6. Conduct experiments to observe the crack path and surface profile in unused fuel cladding and validate against simulation results; and 7. Expand the adaptive mesh refinement infrastructure parallel processing environment to allow adaptive mesh refinement at the 3D crack fronts and adaptive mesh merging in the wake of cracks. Fuel cladding is made of materials such as stainless steels and ferritic steels with added alloying elements, which increase stability and durability under irradiation. As fuel cladding is subjected to water, chemicals, fission gas, pressure, high temperatures, and irradiation while in service, understanding performance is essential. In the fast fuel used in advanced burner reactors, simulations of the nuclear fuels are critical to understand the burnup, and thus the fuel efficiency.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Strength and rupture geometry of un-irradiated C26M FeCrAl under LOCA burst testing conditions

Ferritic iron-chromium-aluminum (FeCrAl) alloys are an accident tolerant fuel candidate to replace the incumbent Zr-based claddings. Nuclear grade FeCrAl alloys are marked by superior high temperature mechanical behavior and exceptional steam oxidation resistance, both of which increase safety margins during accident scenarios. In the present study, the loss of coolant accident (LOCA) burst behaviors of three un-irradiated commercially fabricated cladding materials in a simulated LOCA environment were compared: (1) T35Y2, a 1 st generation nuclear grade FeCrAl, (2) C26M, a 2 nd generation nuclear grade FeCrAl, and (3) Zircaloy-2. Both FeCrAl alloys showed improved mechanical strength and steam oxidation resistance compared to Zircaloy-2. C26M claddings burst at significantly higher temperatures for all tested engineering hoop stresses, had limited ballooning, and demonstrated preferential fuel retention behavior in terms of burst opening area and length. High temperature tensile data for C26M is also presented. For both un-irradiated FeCrAl alloys, it was found that a distinct “threshold” burst stress signified the transition between small and large openings. Finally, higher threshold hoop stresses were associated with higher uniaxial strength for the FeCrAl alloys, indicating that tensile data, rather than creep data, could be useful for predicting rupture size and assessing fuel dispersal concerns.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Drying Wet Stored and Corroded Magnox Fuel for Interim Dry Storage - 20187

To investigate the feasibility of drying corroded Magnesium alloy clad nuclear fuel, the corrosion products of inactive and unirradiated Magnox simulant dry corrosion product and dried corroded Magnox sludge were characterised by X-ray diffraction and thermogravimetric analysis. XRD identified brucite in the dried corrosion product and both brucite and hydromagnesite in the sludge. TGA identified mass loss on dehydration of ∼9.1% up to 250 deg. C and 27.9% from dehydroxylation up to 440 deg. C, with total mass loss of 41.9% up to 800 deg. C for the dried corrosion product. The sludge TGA showed 8.3% mass loss up to 250 deg. C from dehydration, 7.5% up to 330 deg. C from dehydroxylation and 31.4% mass loss up to 47.2% for decarbonation, with total mass loss of 49.2% up to 800 deg. C. Water removal up to 1.8 g was performed by cold vacuum drying (40 deg. C-120 deg. C) on a ≅14.8 g (wet) corroded Magnox sample. The process was monitored by observing pressure, dew point, temperature and gas flow changes supported by measuring the sample mass loss as water is removed. From these tests, it was observed that the dried corroded Magnox displayed some hygroscopic capacity which has implications for water retention following vacuum drying and the length of exposure required to achieve the desired dryness. Whilst at higher temperatures the water was removed faster, the increased temperature also risks the fuel cladding undergoing unwanted chemical reactions with the residual water during the process. However, at lower temperatures the achievable levels of dryness were reduced. Therefore, from these observations this work has identified that there is a temperature balance that may be necessary to optimise the drying process with respect to allowing the greatest level of dryness whilst restricting unwanted chemical reactions. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Atomistic modeling of lanthanide diffusion in refractory body-centered cubic molybdenum

Lanthanide fission products can strongly interact with candidate cladding alloys, but their transport properties in refractory metals remain poorly understood. Here, in this work, we investigate the atomic-scale diffusion behavior of La, Ce, Pr, and Nd in body-centered cubic (bcc) molybdenum, a potential candidate for advanced nuclear cladding. Self-consistent mean-field transport modeling is performed to evaluate the fission product transport and vacancy mobility, informed by first-principles and nudged elastic band calculations of vacancy formation energies, migration barriers, and solute–vacancy binding characteristics. Compared with bcc Fe, lanthanide solutes in bcc Mo exhibit slower tracer diffusion due to higher vacancy formation and migration energies. Furthermore, the calculations reveal that the influence of fission products on migration barriers in bcc Mo are not as extensive in range compared to bcc Fe. Among the studied lanthanides, La exhibits the strongest vacancy binding while also being the fastest diffuser in Mo. These findings highlight how refractory bcc alloys can reduce fission product infiltration, offering valuable insight into the development of durable cladding systems for advanced reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2021

Microstructure Investigation and Mechanical Properties of Coated Zircaloy Cladding

Accident Tolerant Fuel concepts have been developed and tested in diverse research programs presently around the world. Industry teams have developed proprietary coating compositions and geometries, but significant knowledge gaps exist in terms of the resulting properties of the coating and the cladding and how those components impact each other in reactor conditions. Furthermore, the national laboratory and university teams currently studying the industry concepts are limited in their ability to freely disseminate and discuss results obtained on commercial coatings. The Advanced Fuels Campaign is therefore working to develop an understanding of ‘generic’ coatings applied to commercial nuclear cladding alloys free of these constraints. This milestone report provides insights in the microstructure of a 7-micrometer Cr coating on Zircaloy-2 and a 5- micrometer coating on Zry-4 as reference materials. SEM and TEM methods were utilized to describe and identify the interface, while nanoindentation was used to determine hardness and E-Moduli of coating and cladding material. This milestone summarizes the acquired knowledge to build a necessary verified foundation for future simulation and qualification efforts.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fabrication of New NFA-SiC Composites for Nuclear Applications

Cladding materials play the critical and fundamental function of fuel confinement during the safe operation of a nuclear power system. With the development of a new generation of nuclear reactors, investigation of the corresponding advanced cladding materials is attracting intensive scrutiny. Nanostructured ferritic alloy (NFA), which has excellent irradiation resistance and mechanical performance, is one primary candidate for future fission and fusion reactors. The NFA alloys are composed of nano-grained Fe-Cr alloy domains and Y-Ti-O-Cr-Fe enriched nano-clusters. The nano-clusters can not only prevent dislocation glide, grain growth, and grain boundary slip, but also act as sinks to trap helium atoms and radiation-generated point defects. In addition, NFAs can provide high temperature strength and prominent thermal stability, which are beneficial to nuclear cladding applications. Silicon carbide (SiC) is another nuclear cladding candidate. Due to its physical and chemical stability, it shows very good refractory properties at high temperatures up to 2000 deg. C, which has enabled its wide applications in combustion chamber and aerospace engines. In addition, SiC has the remarkable characteristic of radiation stability. SiCf/SiC composite has been developed as an alternative cladding material to metallic alloys. Theoretical simulation is also being developed to understand irradiation induced defects in SiC. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods

A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Thermal conductivity of self-ion irradiated nanocrystalline zirconium thin films

Thermomechanical stability and high thermal conductivity are important for nuclear cladding material performance and reliability, which degrade over time under irradiation. The literature suggests nanocrystalline materials as radiation tolerant, but little or no evidence is present from thermal transport perspective. In this study, we irradiated 10 nm grain size zirconium thin films with 800 keV Zr + beam from a 6 MV HVE Tandem accelerator to achieve various doses of 3 × 10 10 to 3.26 × 10 14 ions/cm 2 , corresponding to displacement per atom (dpa) of 2.1 × 10 –4 to 2.28. Transmission electron microscopy showed significant grain growth, texture evolution and oxidation in addition to the creation of displacement defects due to the irradiation. The specimens were co-fabricated with micro-heaters to establish thermal gradients that were mapped using infrared thermometry. An energy balance approach was used to estimate the thermal conductivity of the specimens, as function of irradiation dosage. As a result, up to 32% reduction of thermal conductivity was measured for the sample exposed to a dose of 2.1 dpa (3 × 10 14 ions/cm 2 ).

36 MATERIALS SCIENCE↗
DOE OSTI2013

Final LDRD report :

We present the results of a three-year LDRD project focused on understanding microstructural evolution and related property changes in Zr-based nuclear cladding materials towards the development of high fidelity predictive simulations for long term dry storage. Experiments and modeling efforts have focused on the effects of hydride formation and accumulation of irradiation defects. Key results include: determination of the influence of composition and defect structures on hydride formation; measurement of the electrochemical property differences between hydride and parent material for understanding and predicting corrosion resistance; in situ environmental transmission electron microscope observation of hydride formation; development of a predictive simulation for mechanical property changes as a function of irradiation dose; novel test method development for microtensile testing of ionirradiated material to simulate the effect of neutron irradiation on mechanical properties; and successful demonstration of an Idaho National Labs-based sample preparation and shipping method for subsequent Sandia-based analysis of post-reactor cladding.

Clark, Blythe G.↗
DOE OSTI2018

Survey on cladded component design procedures for nuclear and non-nuclear applications

Corrosion resistant weld overlay and integral claddings are used to provide corrosion resistance to components made of structural materials that are susceptible to corrosion. Generally, cladding is used in industry to offset the cost of expensive corrosion resistance alloys. In the application to nuclear, cladding provides a mechanism to allow use of limited code approved base materials in corrosive environments. Applicable design rules will need to be developed to support the design and construction of these cladded components for MSR applications. This document provides a survey of the state-of-the-art in design practice of cladded components for both the nuclear and non-nuclear industry. Design practice and service experience in the nuclear and non-nuclear is investigated. A gap analysis on applying these design practice to nuclear applications is included.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Development of Innovative Non-Concentrated Alloy (NCA) Fuel Cladding for Advanced Nuclear Reactors

As the nuclear energy sector advances toward next-generation reactors, the need for high-performance fuel cladding materials has become increasingly urgent. Traditional alloys like zirconium and stainless steel are reaching their performance limits under higher temperatures, more corrosive coolants, and extended irradiation. This report presents the development of a new class of fuel cladding materials based on Non-Concentrated Alloys (NCAs) ? multi-element systems designed to deliver enhanced mechanical strength, corrosion resistance, and radiation tolerance. Through a combination of computational modeling (Computer Coupling of Phase Diagrams and Thermochemistry: CALPHAD), simulation-guided alloy selection, and experimental fabrication via arc melting and spark plasma sintering (SPS), three strategic alloy design paths were explored: (1) FeCrAl-based NCAs, (2) refractory-lean neutron-efficient alloys, and (3) equimolar high-entropy compositions. Microstructural analysis confirmed the formation of stable body-centered cubic BCC_A2 phases, while mechanical testing demonstrated hardness values significantly exceeding those of conventional cladding materials. The results highlight the tunability of NCA systems and their potential for balancing strength and ductility ? a critical consideration for in-reactor performance. Looking forward, future work will focus on thermomechanical optimization, CALPHAD refinement, and benchmarking against industry standards to enable scalable deployment. This work not only advances the science of nuclear materials but also supports broader goals in nuclear safety, performance, and nuclear energy innovation.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

Caustic Neutralization and Precipitation of Acidic Dissolved Simulated Stainless Steel–Clad Plutonium and Plutonium/Uranium Nuclear Fuel

Simulated dissolved stainless steel (SS) clad Pu and Pu/U nuclear fuel in HNO 3 was neutralized to a free hydroxide (OH - ) concentration of 0.6 M. A thermal neutron poison, Gd, was added to the simulants at concentrations of either ~3 - 6 g/L or ~37 - 38 g/L. The supernate Pu concentration the day of neutralization ranged from 0.48 to 8.75 mg/L. The supernate Pu concentration of a simplified simulant neutralized to 0.6 M OH - above precipitated solids containing Pu was demonstrated to decrease over 18 days. A significant portion of precipitated Pu was found to be insoluble in 8 M HNO 3 at ambient temperature, but essentially quantitative Pu dissolution was achieved in 11.5 M HNO 3 /0.1 M KF at 100 °C. The difficulty in dissolving the Pu precipitate is believed to be due to the formation of refractory PuO 2 •xH 2 O during the neutralization process. Initial Gd concentrations of ~37 – 38 g/L were found to result in a greater Al precipitation when neutralized to 0.6 M OH - than initial Gd concentrations of ~3 – 6 g/L. Physical properties of the resultant slurries were measured and used to calculate limiting flowrates and slurry velocities by gravity only in transfer piping between the Savannah River Site’s H-Canyon Facility and the Concentration, Storage, and Transfer Facility (CSTF). These results were compared to calculated deposition velocities to predict if solids would settle during the transfer. The Newtonian model was found to be reasonable for each diluted slurry evaluated. Deposition velocities of Pu containing slurries are lower than nuclear fuel slurries primarily composed of U due to the high density of Pu solids. In conclusion, dilution of slurries reduces the margin between the slurry and deposition velocities due to the reduction in viscosity because higher viscous forces on the particles promote maintained suspension.

Actinide Neutralization↗
DOE OSTI2016

A Radial Heat Flow Apparatus for Thermal Conductivity Characterisation of Cylindrical Samples

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. In this frame, the thermal conductivity (TC) of cladding prototype sections are measured both before and after neutron irradiation with the purpose of better understanding the effects of neutron irradiation on the pyrolytic carbon interphase linking the SiC fibres to the SiC matrix. To this end, an apparatus has been designed and built to measure the TC of short cylindrical samples, whose lengths range from 5 to 20 cm, by the radial heat flow method. To avoid oxidation issues, the experiments are carried out in vacuum. Today the most common measurement method used to characterise the TC of materials is the so-called laser-flash technique, whereby a short high intensity laser pulse is shot on one surface of a thin flat sample and the energy transfer through the sample is analysed on the other side. Though much faster and easier to implement than the apparatus presented in this work, laser-flash equipment does have issues when dealing with highly anisotropic materials or capturing shape effects such as those present in large aspect ratio samples such as cladding tube sections. The two aforementioned effects are exactly what the developed equipment has been built to capture. The radial heat flow method is based on the principle of measuring the temperature difference building up through a piece of material as this material acts as a thermal resistance against a known heat flow crossing it. This paper presents a new implementation of one of the classical thermal conductivity measurement techniques. Since the development of the laser-flash method in the sixties, steady-state measurements such as the radial heat flow have fallen in disuse. In this discussion, it is shown that a direct measurement method can be used and even be preferable in the case of highly anisotropic materials such as SiC/SiC composites. The advantage of this experimental setup is that the actual transverse conductivity is obtained, whereas laser-flash measurements yield a response in which both the transverse and axial conductivities are mixed. Further steps of this study will include the measurements of samples irradiated with neutrons in the MITR test-reactor. The aim will be to evidence the role played by the pyrolytic carbon interphase by linking the TC measurements with energy filtered electron microscopy. We would like to acknowledge General Atomics and Westinghouse for their support in the frame of the CARAT research program. Providing us with actual prototype clad sections is a unique opportunity. Furthermore the Swiss CCEM.ch project MeAWaT is acknowledged for partially financing the development of the measurement device. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Effect of PyC interlayer amorphization on the thermal conductivity of SiC/SiC composites - 18682

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. This paper focuses on the consequences of the microstructural changes occurring at the pyrolytic (PyC) carbon interphases linking silicon carbide matrix and fibres in the composite. The experimental approach is two-fold: measurements of the thermal conductivity of short sections of prototype SiC/SiC cladding tubes deal with the macro-scale of the question, whereas the effects of radiation damage on the PyC interlayer are studied by energy filtered transmission electron microscopy (EFTEM) as well as spectrum imaging in STEM (ESI). The results from both parts of the experimental work are brought together in an effective medium model based on the work of Markworth and Youngblood et al. A detailed outlining of the radial heat flow experimental apparatus developed to this end has been presented at the 2016 ANS meeting in New Orleans. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

MECHANISTIC MULTIPHYSICS MODELING OF CLADDING RUPTURE IN NUCLEAR FUEL RODS DURING LOSS-OF-COOLANT ACCIDENT CONDITIONS

The Loss of Coolant Accident (LOCA) is a design basis accident that is included as part of the safety analysis of nuclear power plants. As the nuclear industry desires to increase the cycle length and discharge burnup of existing nuclear power plants they must demonstrate safe operation during a LOCA on high burnup fuel. During a LOCA transient on high burnup fuel rods, the rods may undergo a process known as fuel fragmentation, relocation, and dispersal (FFRD). To permit dispersal, the cladding encapsulating the fuel must rupture with an opening size large enough to allow the fragmented fuel particles to release. Current licensing tools used by industry and the United States Nuclear Regulatory Commission are limited in geometric fidelity and materials that can be analyzed. These simulation tools generally employ a quasi-two-dimensional (1.5D or Layered1D) or 2D-RZ axisymmetric geometric representations exclusively. While a valid approach under some instances, there are times when important physics have an asymmetric behavior in the fuel rod. Examples include fuel fragmentation, thermal-hydraulic boundary conditions, and cladding rupture, all of which are important for LOCA analysis. As industry pursues burnup extensions it must be demonstrated that fuel dispersal can be mitigated or eliminated. To do this, an understanding of the rupture opening after cladding failure is required. This work presents the development of a model for predicting the size and location of the rupture opening in failed fuel rods during LOCA conditions using advanced modeling and simulation tools. In order to supply the rupture model with appropriate boundary conditions, improvements to fuel fragmentation, axial relocation and oxidation modeling were required. First, the eXtended Finite Element Method (XFEM) is used to mechanistically predict the number of fuel fragments that form due to material strength randomization, criteria for strength randomization, mesh density, power ramping rates and irradiation effects. These predictions with associated uncertainty were compared to empirical correlations developed for UO2 verifying that they can be used with increased confidence in subsequent axial relocation analyses. Secondly, a new first-of-its-kind Layered2D computational framework was developed that provides the ability to apply azimuthally varying boundary conditions while providing discrete layers to track fuel movement during the LOCA. An existing fuel axial relocation model developed for Layered1D was extended to work within the Layered2D framework. A large sensitivity study was performed on the initial version of the model to identify modeling parameters of particular importance, with the emissivity used for radiation after blowdown being the primary source of uncertainty. Then, a simplistic approach to incorporate mechanical degradation of the cladding due to oxidation was also developed to investigate the impact of reduced cladding thickness on predictions of the time to failure of cladding tubes. It was found that the cladding will typically rupture prior to a reduction in thickness that is sufficient to impact the rupture behavior. A model was then developed for predicting cladding rupture that transfers the cladding surface temperatures, rod internal and external pressures, fast neutron fluence, and fast neutron flux from a more detailed Layered1D, Layered2D, or 2D-RZ analysis to a 3D cladding only analysis. Comparisons of the rupture model to a few experiments indicate reasonable predictions. The rupture model was then applied to two accident tolerant fuel concepts (FeCrAl and Cr-coated Zircaloy) where it predicted that both ATF concepts would have smaller rupture openings and delayed rupture times than the standard Zircaloy-4 cladding material under identical loading conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Charge Optimized Many Body (COMB) potentials for simulation of nuclear fuel and clad

The Charge Optimized Many Body (COMB) potential formalism enables the molecular dynamics simulation of complex materials structures in which multiple types of bonding (metallic, covalent, ionic and secondary bonding) coexist. Here we briefly review the COMB formalism and illustrate its capabilities to address critical issues in the area of nuclear fuel. In particular, we In Section 3, we look at U, UO2 and the process of oxidation of U. In Section 4, we look at the mechaincal behavior of Zr, representing LWR clad, and explore the effects of oxidation and hydridation on the mechanical response. In Section 5, we describe the capabilities of COMB simulations of corrosion. Since there has been relatveily little work done in this area, we also describe some work performed on non-nuclear materials, which illustrate the capabilities. Section 6 briefly describes some of the known limitations of the COMB approach.

36 MATERIALS SCIENCE↗