Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “cladding”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

FAST-1.2.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation (including creep), fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1973

Irradiation of TZM: Uranium dioxide fuel pin at 1700 K

A fuel pin clad with TZM and containing solid pellets of uranium dioxide was fission heated in a static helium-cooled capsule at a maximum surface temperature of 1700 K for approximately 1000 hr and to a total burnup of 2.0 percent of the uranium-235. The results of the postirradiation examination indicated: (1) A transverse, intergranular failure of the fuel pin occurred when the fuel pin reached 2.0-percent burnup. This corresponds to 1330 kW-hr/cu cm, where the volume is the sum of the fuel, clad, and void volumes in the fuel region. (2) The maximum swelling of the fuel pin was less than 1.5 percent on the fuel-pin diameter. (3) There was no visible interaction between the TZM clad and the UO2. (4) Irradiation at 1700 K produced a course-grained structure, with an average grain diameter of 0.02 centimeter and with some of the grains extending one-half of the thickness of the clad. (5) Below approximately 1500 K, the irradiation of the clad produced a moderately fine-grained structure, with an average grain diameter of 0.004 centimeter.

Mcdonald, G. E.↗
NASA NTRS1974

Electrical heating tests of uranium dioxide external fuel configuration at emitter temperature of 1900 K

Testing of two fuel clad specimens for thermionic reactor application is described. The annular UO2 fuel was clad on both sides with tungsten; heat rejection was radially inward. The tests were intended to study inner clad stability, fuel redistribution, and fuel melting problems. The specimens were tested in a vacuum chamber using electron bombardment heating. Fuel structural changes were studied using periodic gammagraphs and posttest metallography. The first specimen test was terminated at 50 hours because of a braze failure. The second specimen was tested for 240 hours when an outer clad leak developed due to a tungsten-water reaction. The fuel developed numerous cracks on cooldown but the inner clad remained dimensionally stable. The fuel cover gas did not impede the rate of fuel redistribution. Posttest examination showed the fuel had not melted during operation.

Diianni, D. C.↗
NASA NTRS1981

Investigation of direct integrated optics modulators

Direct optical modulation techniques applicable to integrated optical data preprocessors were studied. Emphasis was placed on the analysis and fabrication of a field effect type modulator. A series of computer modeling studies were performed to determine the effects of semiconductor cladding on the fields of propagating waves in planar dielectric waveguides. These studies predicted that changes in the propagation characteristics of waveguides clad with silicon and gallium arsenide could be made large enough to be useful in modulators. These effects are dependent on the complex permittivity and thickness of the cladding. Since the conductivity of the cladding can be changed by the photon generation of hole-electron pairs, incoherent light may be used as the input to modulate a coherent light beam. Waveguides were fabricated and silicon claddings were applied to verify the theoretical predictions.

Batchman, T. E.↗
NASA NTRS1984

Single mode acoustic fiber waveguide

The single mode operation of a clad rod acoustic waveguide is described. Unlike conventional clad optical and acoustic waveguiding structures which use modes confined to a central core surrounded by a cladding, this guide supports neither core nor cladding modes but a single interface wave field on the core-cladding boundary. The propagation of this bound field and the potential improved freedom from spurious responses is discussed.

Jackson, B. S.↗
NASA NTRS1988

Cracking of a layered medium on an elastic foundation under thermal shock

The cladded pressure vessel under thermal shock conditions which is simulated by using two simpler models was studied. The first model (Model 1) assumes that, if the crack size is very small compared to the vessel thickness, the problem can be treated as a semi-infinite elastic medium bonded to a very thin layer of different material. However, if the crack size is of the same order as the vessel thickness, the curvature effects may not be negligible. In this case it is assumed that the relatively thin walled hollow cylinder with cladding can be treated as a composite beam on an elastic foundation (Model 2). In both models, the effect of surface cooling rate is studied by assuming the temperature boundary condition to be a ramp function. The calculated results include the transient temperature, thermal stresses in the uncracked medium and stress intensity factors which are presented as a function of time, and the duration of cooling ramp. The stress intensity factors are also presented as a function of the size and the location of the crack. The problem is solved for two bonded materials of different thermal and mechanical properties. The mathematical formulation results in two singular integral equations which are solved numerically. The results are given for two material pairs, namely an austenitic steel layer welded on a ferritic steel substrate, and a ceramic coating on ferritic steel. In the case of the yielded clad, the stress intensity factors for a crack under the clad are determined by using a plastic strip model and are compared with elastic clad results.

Rizk, Abd El-Fattah A.↗
NASA NTRS1991

Thermal analysis of conceptual designs for GPHS/FPSE power systems of 250 We and 500 We

Thermal analyses were performed for two distinct configurations of a proposed space nuclear power system which combines General Purpose Heat Source (GPHS) modules with the state of the art Free-Piston Stirling Engines (FPSEs). The two configurations correspond to systems with power levels of 250 and 500 W(sub e). The 250 W(sub e) GPHS/FPSE power system utilizes four GPHS modules and one FPSE, and the 500 W(sub e) contains eight GPHS modules and two FPSEs. The configurations of the systems and the bases for selecting the configurations are described. Brief introductory sections are included to describe the GPHS modules and free piston Stirling engines. The primary focus of the thermal analyses is on the temperature of the iridium fuel clad within the GPHS modules. A design goal temperature of 1573 K was selected as the upper limit for the fuel clad during normal operating conditions. The basis for selecting this temperature limit is discussed in detail. Results obtained from thermal analysis of the 250 W(sub e) GPHS/FPSE power system indicate fuel clad temperatures which slightly exceed the design goal temperature of 1573 K. The results are considered favorable due to the numerous conservative assumptions used in developing the thermal model and performing the thermal analysis. To demonstrate the effects of the conservatism, a brief sensitivity analysis is performed in which a few of the key system parameters are varied to determine their effect on the fuel clad temperatures. It is concluded that thermal analysis of a more detailed thermal model would be expected to yield fuel clad temperatures below the design foal temperature limiy 1573 K.

Mccomas, Thomas J.↗
NASA NTRS1993

Optical fibers and Fluorosensors having improved power efficiency and methods of producing same

Optical fibers may have applications including fluorosensors which sense the concentration of an analyte. Like communication fibers, these fluorosensors are modeled using a weakly guiding approximation which is only effective when the difference between the respective refractive indices of the fiber core and surrounding cladding are minimal. An optical fiber fluorosensor is provided having a portion of a fiber core which is surrounded by an active cladding which is permeable by the analyte to be sensed and containing substances which emit light waves upon excitation. A remaining portion of the fiber core is surrounded by a guide cladding which guides these light waves to a sensor which detects the intensity of waves, which is a function of the analyte concentration. Contrary to conventional weakly guiding principles, the difference between the respective indices of refraction of the fiber core is surrounded by an active cladding which is thin enough such that its index of refraction is effectively that of the surrounding atmosphere, thereby the atmosphere guides the injective indices of the fiber core and the cladding results in an unexpected increase in the power efficiency of the fiber core.

Egalon, Claudio O.↗
NASA NTRS1966

Development of Protective Coatings for Chromium-Base Alloys

Chromium alloy sheet was clad with 5 to 10-mil-thick oxidation-resistant nickel-base alloy foils. Specimens also contained 1/2 to 1-mil-thick intermediate layers of platinum, tungsten, and/or W-25Re. Cladding was done by the isostatic hot gas-pressure bonding,.process. The clad chromium-alloy specimens were cyclic oxidation tested at 2100 F and 2300 F for up to 200 hours to determine the effectiveness of these metal claddings in protecting the chromium alloy Cr-5W from oxidation and contamination. Cladding systems consisting of 5-mil-thick Ni-20Cr-20W modified with 3 to 5 weight percent aluminum and containing a 1 /2-mil tungsten diffusion barrier demonstrated potential for long-time service at temperatures as high as 2300 F.

English, J. J.↗
NASA NTRS2017

Compact Hybrid Laser Rod and Laser System

A hybrid fiber rod includes a fiber core and inner and outer cladding layers. The core is doped with an active element. The inner cladding layer surrounds the core, and has a refractive index substantially equal to that of the core. The outer cladding layer surrounds the inner cladding layer, and has a refractive index less than that of the core and inner cladding layer. The core length is about 30 to 2000 times the core diameter. A hybrid fiber rod laser system includes an oscillator laser, modulating device, the rod, and pump laser diode(s) energizing the rod from opposite ends. The rod acts as a waveguide for pump radiation but allows for free-space propagation of laser radiation. The rod may be used in a laser resonator. The core length is less than about twice the Rayleigh range. Degradation from single-mode to multi-mode beam propagation is thus avoided.

Busch, George E.↗
DOE OSTI2022

ANS Winter 2022: ATF-2C Physics Safety and Scoping Analysis

ATF-2C comprises four tiers with six rodlets each. The experiment test train (TT) is situated in Loop 2A in the center flux trap (CFT) of the ATR, in which typical pressurized water reactor (PWR) conditions shall be approximated. Tier 1 (bottom) of the TT tests Framatome silicon carbide (SiC) cladding, containing molybdenum heater rods. Low-enriched uranium (LEU) Framatome rodlets with chrome-coated M5 cladding have been modeled in their place for the physics safety analysis to allow the substitution of this tier with fuel if needed. Tier 2 tests LEU Framatome rodlets with M5 cladding, which had irradiated in ATR in prior ATF experiments. Tier 3 tests General Atomics SiC composite cladding, also containing molybdenum heater rods. Tier 4/5/6 (top) tests the Japan Atomic Energy Agency (JAEA)/Mitsubishi Heavy Industries (MHI) fueled rodlets. This upper tier tests LEU rodlets with chrome-coated MDA cladding—several with temperature or pressure instrumentation. To flatten the axial power profile, lower portion of Tier 4/5/6 is surrounded by a hafnium (Hf) shroud, and the bottom two UO2 pellets are of natural enrichment. The ATF-2C TT is to be irradiated in the CFT of ATR (Fig. 1) for three or four 60-day cycles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Removal of Bond Sodium from Full-Length Unirradiated Fermi-1 Blanket Elements and Assembly via Melt-Drain-Evaporate Process

Equipment was designed, fabricated, tested, and operated in an inert atmosphere radiological glovebox at Idaho National Laboratory to demonstrate the removal of bond sodium from full-length unirradiated Fermi-1 radial blanket elements and an entire radial blanket assembly using a Melt-Drain-Evaporate process. A series of three runs was performed with individual and multiple radial Fermi-1 blanket elements, and a fourth run was conducted with an entire Fermi-1 radial blanket assembly. After each run, the depleted uranium alloy slugs in every element slid out of its cladding, mechanically exhibiting the effectual absence of bond sodium. Quantitative and further qualitative analyses of the treated Fermi-1 material revealed the substantive, if not complete, absence of sodium metal in blanket element components. Indeed, entire columns of depleted uranium alloy slugs and associated cladding from select treated elements were separately contacted with water to react with residual sodium metal on the element surfaces, forming hydrogen gas. Samples of the resultant gas were analyzed to quantify the hydrogen concentration, which correlated to a residual sodium metal content for the treated element. Accordingly, no detectable sodium metal (<7 µg) was found on the surfaces of multiple depleted uranium alloy slugs after removal from its cladding. Detectable sodium metal, ranging from 14 to 30 µg, was found on the surfaces of one separated column of slugs and two separated full-length cladding segments. Each element originally contained ~25 g of bond sodium, resulting in quantitative sodium metal removal efficiencies of =99.9998%. The remaining separated uranium alloy slugs and cladding segments from the four runs were piece-wise immersed in alcohol while videorecording possible gas bubble formation as an additional qualitative analysis for residual sodium metal on treated surfaces, which identified the predominant absence of sodium reactivity on the balance of treated Fermi-1 blanket material. The results of this demonstration substantiate a path forward for treatment and disposal of 34 metric tons heavy metal of irradiated sodium-bonded Fermi-1 blanket material currently stored at Idaho National Laboratory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Comparison of FAST Experiments to BISON Simulations

The Fission Accelerated Steady-State Test (FAST) experiments utilize the effects of scaling to accelerate the burnup experienced by the fuel. This experimental technique offers a unique opportunity to evaluate the effects of scaling on fuel and cladding material modelling. Currently, models used in the BISON fuel performance code for metallic fuel performance are calibrated for EBR-II conditions and geometries. Comparing BISON simulations to FAST experimental data will allow for the identification of shortcomings in the models for conditions outside the EBR-II paradigm. FAST experiments have been performed on U-Zr and U-10Mo fuel types with HT9 cladding. Initial comparisons of simulated cladding strains to experimental results show that the BISON simulations consistently overpredict cladding strain in the fuel region but underpredict the cladding strain in the plenum region. This indicates potential issues with the simulation of swelling and gas pressure at smaller geometries. Further comparison with fission gas release measurements is planned as data becomes available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Elongate SiC fuel elements

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

Long, Yun-Bo↗
DOE OSTI2025

Hydride Reprecipitation in Prototypical 17x17 PWR Fuel Rods after Simulated Vacuum Drying Conditions

Light water reactor (LWR) fuel rod cladding picks up hydrogen as it oxidizes in reactor. Once the hydrogen concentration exceeds the solubility limit, zirconium hydride platelets precipitate, generally in the circumferential direction. In preparation for dry storage, elevated temperatures can dissolve some of the circumferential hydrides, which during cooling can then reprecipitate in a direction perpendicular to the hoop stress (radially). From 2019 to 2024, ORNL published a series of reports that included metallographic observations of cladding after a thermal transient to 400°C followed by slow cooling. The simulated dry storage process was performed on three full-length 17x17 PWR fuel rods at their as-discharged rod internal pressure. There were two important observations from the metallographic examinations of these heat-treated rods: (1) the propensity for hydride reorientation is strongly influenced by nearby as-discharged precipitated hydrides, (2) the radial hydrides observed near the cladding inner diameter were often associated with cracks in the pellet. Subsequent modeling using the BISON fuel performance code that considered the effects of pellet cracks and pellet-cladding bonding observed in high-burnup fuel support the conclusion that pellet cracks enhance the local hoop stress in the cladding creating preferential sites for hydride reorientation, and that the influence of the pellet cannot be ignored in accurate fuel rod performance predictions.

Montgomery, Rose [ORNL] (ORCID:0000000286038936)↗
DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FeCrAl fuel/clad chemical interaction in light water reactor environments

This article investigates the fuel-cladding chemical interaction (FCCI) behavior of two commercial FeCrAl alloys, APMT composition (Fe-21Cr-5Al-3Mo wt.%) and C35M (Fe-13Cr-5Al-2Mo-0.2Si-0.03Y wt.%), after neutron irradiation. “H-cup” diffusion multiples of FeCrAl alloys and ceramic UO 2 fuel were irradiated at a temperature of ~300 °C to a total estimated burnup of 26 GWd/tHM. Post-irradiation Examination results demonstrate the excellent degradation resistance of FeCrAl alloys as accident tolerant fuel (ATF) cladding materials in light water reactor conditions. The study concludes that there was no irradiation-induced defects observed in either of the two commercial FeCrAl claddings. The formation of amorphous Al/U mixed oxide was observed at the fuel-clad interface, which can serve as a tritium permeation barrier and protect against potential chemical attack from the fuel. Finally, the study attributed the formation of amorphous Al/U mixed oxide to the low temperature and limited time of neutron irradiation. APMT forms more distinct Cr and Cr-Fe intermetallic at the FeCrAl-UO 2 interface than C35M due to the higher bulk Cr:Al ratio.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Out-of-pile and postirradiated examination of lanthanide and lanthanide-palladium interactions for metallic fuel

Palladium is being investigated as a fuel additive to bind with and potentially immobilize lanthanide fission products. A primary cause of fuel-cladding chemical interaction (FCCI) is the lanthanide fission products migrating to the fuel periphery and interacting with the cladding. This interaction will lead to wastage of the cladding and eventually to a cladding breach. Palladium has previously been identified as a promising additive used to prevent or decrease FCCI by reacting with the lanthanide fission products. In the current study, an alloy cast from the four highest abundant lanthanides found in irradiated metallic fuel, Nd, Ce, Pr, and La, with and without Pd, has been characterized using neutron diffraction, scanning electron microscopy, and electron probe microanalysis. In the lanthanide-Pd intermetallic compounds, all of the constituent compounds, i.e. Nd-Pd, Ce-Pd, La-Pd and Pr-Pd are known. There is very good agreement, both structurally and compositionally, between the out-of-pile lanthanide alloy and lanthanide fission products characterized in irradiated fuels. In both cases, the lanthanide elements form a solid solution in a hexagonal crystal structure. The out-of-pile lanthanide alloy follows Vegard's Law, with the measured and calculated (weighted average of constituents) lattice parameters being within 1% for both the a and c parameters. Pd bonds with the lanthanides (Ln) forming the phases LnPd and Ln7Pd3. The results indicate the properties of lanthanide compounds in irradiated metallic fuel can be reliably simulated in out-of-pile experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗