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 865 records · Page 48

Advanced Fuels Campaign Time-at-Temperature Irradiation Campaign and Post-Irradiation Examination Plan

With increasing per capita energy consumption and rapidly growing energy consumption predicted for data centers, the US faces challenges in meeting projected energy demands. The current administration is attempting to accelerate the deployment of new nuclear construction by providing significant investments in advanced nuclear and by modernizing regulatory approaches for the licensing new reactors. However, enabling extended power uprates (EPUs) for the current nuclear reactor fleet would provide a shorter-term solution to better calibrate near-term energy generation capacity to the ever-growing energy demands of the modern era. One pathway to achieve power uprates is to increase the operational window by reassessing fuel safety limits around anticipated transients. Existing light-water reactors (LWRs) utilize a targeted operational window with well-defined operational efficiencies, yet this window is also often bound by conservative safety limits that result in suboptimal operational performance. For example, the operation of the existing fleet is highly constrained by safeguards to operation related to anticipated operational occurrences (AOOs), which are moderate-frequency transients expected with a frequency greater than 0.01 per reactor-year. These AOOs include temperature excursions where the critical heat flux for the system is exceeded, resulting in a departure from nucleate boiling (DNB) at the cladding surface. This DNB event is a thermohydraulic condition that, with current conservative fuel safety limits, results in the fuel rods being classified as “failed,” meaning that the cladding may not be returned to service.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Summary of the Initial Post-Irradiation Characterization of HFIR-Irradiated Low-N and High-N HT-9 Steel

Reference cladding systems for sodium fast reactors are based on the historical steel, HT-9. HT-9 is a Fe12Cr ferritic/martensitic steel with additions of Mo, W, V, and other minor elements and demonstrates low irradiation swelling and adequate mechanical properties. Extensive irradiation literature exists on the use of HT-9 as cladding for metal fuel, primarily irradiation on the U-Zr/HT-9 system from the Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) sodium fast reactor, and as a structural material from experiments in the FFTF. The large amount of historical data makes the U-Zr/HT-9 system the reference fuel specification for many nuclear reactor vendors that seek to license modern sodium-cooled fast reactors in the United States. However, it is yet unclear how variations in impurity content within HT-9 fundamentally affect irradiation performance at various irradiation temperatures. Recent work suggests that impurity content may noticeably alter the production of helium through nuclear transmutation. For these reasons, High-Flux Isotope Reactor (HFIR) irradiation of HT-9 steels with known variations in the impurity content is particularly timely to generate data to enable more accurate refinement of the chemical specification for nuclear-grade HT-9 material. This report summarizes the initial transmission electron microscopy characterization of HFIR-irradiated HT-9 steels following mechanical property measurements by the Advanced Fuels Campaign (AFC). This report includes qualitative results of the cavity, dislocation loop and cluster/precipitate microstructures as well as radiation-induced segregation. Quantitative results are being shared with partner institutions and will be included in more detail in a future report in FY2026.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Material Characterization Report of EBR II Mark IV Fuel Element

The focus of this report is the characterization of an unirradiated as-built fuel element from the Experimental Breeder Reactor II (EBR-II) Mark IV driver fuel design. The fuel in this element is a metallic uranium-zirconium alloy, and the cladding material is HT9 stainless steel. Both the metallic fuel and cladding materials are characterized, including chemical composition, grain size, morphology, phase composition, precipitate composition, crystallography, density, and hardness.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

The advanced characterization, post-irradiation examination, and materials informatics for the development of ultra high-burnup annular U-10Zr metallic fuel

U-Zr metallic fuel is a promising fuel candidate for Gen Ⅳ fast spectrum reactors. Previous experimental irradiation campaigns showed that the sodium thermal bonded U-10Zr fuel design can achieve a burnup of 10% fissions per initial heavy metal atom (FIMA). Advanced metallic fuel designs are pushing the burnup limit to 20% or even 30% FIMA. To achieve the higher burnup and eliminate the pyrophoric sodium, a prototypical annular fuel has been designed, fabricated, clad with HT-9 in the Materials and Fuels Complex, and irradiated in the Advanced Test Reactors of Idaho National Laboratory (INL) to a peak burnup of 3.3% FIMA. During irradiation, the mechanical contact between fuel and cladding acts as a thermal bond. The irradiation lasted for 132 days in the reactor. Recently, the archived fresh and irradiated fuel samples were characterized using advanced characterization capabilities in the Irradiated Materials Characterization Laboratory (IMCL) of INL. This article summarizes the results of advanced characterization and computer vision-based materials informatics to reveal the irradiation effects on U-Zr metallic fuel. Future work will focus on further implementation of advanced characterization and statistical data mining to improve the fidelity of fuel performance modeling and support U-Zr metallic fuel qualification for fast spectrum reactors.

Yao, Tiankai↗

Full-core high-burnup BWR LOCA fuel performance analysis and FFRD susceptibility

The susceptibility of the boiling water reactor (BWR) Limerick Unit 1 to fuel fragmentation, relocation, and dispersal during a postulated large-break loss-of-coolant accident (LBLOCA) was calculated using a multiphysics framework. The simulations include full-core, rod-resolved neutronic, thermal hydraulic, and fuel performance models using the VERA, TRACE, and BISON codes. This work focused on the transient BISON simulations, which include both the normal operation and LBLOCA periods in the same simulations. Cladding integrity was assessed using two correlations that are included with BISON. make page break Several new BWR-specific features were recently added to BISON. This work represents the first time these features have been included in a core-scale set of simulations. This study hence evaluates the performance of these new models for an operating reactor with realistic operating conditions. Simulation results showed that cladding integrity was maintained (i.e., no rods burst). Finally, future work to improve BWR and PWR predictions using this framework is suggested.

BISON↗

Dynamic Probabilistic Risk Assessment Based Response Surface Approach for FLEX and Accident Tolerant Fuels for Medium Break LOCA Spectrum

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Carburization Kinetics of Zircalloy-4 and Its Implication for Small Modular Reactor Performance

Carburization of cladding materials has long been a concern for the nuclear industry and has led to the restricted use of high-thermal conductivity fuels such as uranium carbides. With the rise of small modular reactors (SMRs) that frequently implement a graphite core-block, carburization of reactor components is once more in the foreground as a potential failure mechanism. To ensure commercial viability for SMRs, neutron-friendly cladding materials such as Zr-based alloys are required. In this work, the carburization kinetics of Zircaloy-4 (Zry-4), for the temperature range 1073–1673 K (covering typical operating temperatures and off-normal scenarios) are established. The following Arrhenius relationship for the parabolic constant describing ZrC growth is derived: Kp (in μm2/s) = 609.35 exp(−1.505 × 105/RT)). Overall, the ZrC growth is sluggish below 1473 K which is within the operational temperature range of SMRs. In all cases the ZrC that forms from solid state reaction is hypo-stoichiometric, as confirmed through XRD. The hardness and elastic modulus of carburized Zry-4 are also examined and it is shown that despite the formation of a ZrC layer, C ingress in the Zry-4 bulk does not impact the mechanical response after carburization at 1073 K and 1473 K for 96 h.

36 MATERIALS SCIENCE↗

Innovative Accident Tolerant Fuel Materials May Help Extending the Life of Light Water Reactors: A Perspective

One of the main sources of clean electricity in the civilian power grid comes from light water reactors (LWRs). Some of these LWRs are being decommissioned prematurely because they may not be economically sustainable. It is proposed that the use of accident-tolerant fuel concepts may enable for LWR plant operation extension by making them safer and more economically competitive to operate. For example, the use of iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel would not only reduce the risk of catastrophic reaction between the cladding and its environment in the case of a loss of coolant accident situation but also, for example, allow for higher fuel burnup during normal plant operation.

Materials Science↗

Development of a highly reliable composite board for printed circuitry for use in space environment

Materials, processes and fabrication techniques have been investigated for the development of a high-temperature circuit-board laminate. High quality, void-free copper-clad laminates have been made using 7628/HS-1 style fiberglas reinforcements with filled polyimide matrices. The fabricating characteristics of P13N resin appear suitable for use as a filled matrix in this circuit board development. High-fired, ball-milled alumina appears to be necessary to obtain the desired effects in the circuit board system. Nickel-clad copper foil bonding surfaces appear to be another requirement for retention of good bond strengths after art work and plating sequences. The fabrication cycle for this circuit board system is very dependent on the heating profile. Very rapid heating with quick loading is recommended. A stack approach to lamination was successfully used.

Bradbury, E. J.↗

Results of a preliminary experimental investigation of a vapor transport fuel pin

In-Pile experimental tests were conducted on two fuel pins mounted side-by-side in a holder assembly. The fuel pins consisted of stacked fully enriched UO2 pellets enclosed in a type 316 stainless steel clad 1/2 inch outside diameter and 3 inches in length. Each fuel pin contained four pellets for a total of 22 grams of fuel weight. The experiment was operated to maintain the maximum clad temperature at 395 K corresponding to a fuel temperature at the central void surface of 2350 K. Total primary coolant water flow past the pin holder was calculated to be 110 gallons per minute. Forty-four gallons per minute at a velocity of 36.5 feet per second flowed around each fuel pin. Calculations indicated the maximum heat flux was at the point of the highest temperature and was equal to 0.24 kW/sq cm or 7.62 Btu/hr sq ft. A series of high resolution neutron radiographs of the nuclear fueled vapor transport capsule was taken. The thermal neutrons emitted from the core of the 60 megawatt Plum Brook reactor facility were used.

Lanzo, C. D.↗

Fabrication of (U, Zr) C-fueled/tungsten-clad specimens for irradiation in the Plum Brook Reactor Facility

Fuel samples, 90UC - 10 ZrC, and chemically vapor deposited tungsten fuel cups were fabricated for the study of the long term dimensional stability and compatibility of the carbide-tungsten fuel-cladding systems under irradiation. These fuel samples and fuel cups were assembled into the fuel pins of two capsules, designated as V-2E and V-2F, for irradiation in NASA Plum Brook Reactor Facility at a fission power density of 172 watts/c.c. and a miximum cladding temperature of 1823 K. Fabrication methods and characteristics of the fuel samples and fuel cups prepared are described.

Source record↗

Studies of thermionic materials for space power applications

The effect of microstructures of tungsten cladding on the transport rates of carbide fuel components was studied at 2073 K. hyperstoichiometric 90UC-10ZrC containing 4 wt% tungsten was clad with six types of tungsten material of 40 mil thickness. Screening tests of 1000 hours were carried out, and then selected samples were subjected to long-term tests up to 10,000 hours. The results indicate that the microstructures strongly affect the transport rates of carbide fuel components. The conditions for preparing (110) oriented cylindrical chloride tungsten emitters of high vacuum work functions were also investigated. Specimen sets were deposited on fluoride tungsten substrates for evaluating the effects of various deposition parameters on the degree and uniformity of the (110) preferred orientation and the vacuum work function. Long-term tests showed that the high vacuum work function of a cylindrical emitter was stable and the chloride tungsten to fluoride tungsten bond remained in excellent shape after 4850 hours at 2073 K.

Source record↗

Molybdenum-UO2 cermet irradiation at 1145 K.

Two molybdenum-uranium dioxide cermet fuel pins with molybdenum clad were fission-heated in a forced-convection helium coolant for sufficient time to achieve 5.3% burnup. The cermet core contained 20 wt % of 93.2% enriched uranium dioxide. The results were as follows: there was no visible change in the appearance of the molybdenum clad during irradiation; the maximum increase in diameter of the fuel pins was 0.8%; there was no migration of uranium dioxide along grain boundaries and no evident interaction between molybdenum and uranium dioxide; and, finally, approximately 12% of the fission gas formed was released from the cermet core into the gas plenum.

Mcdonald, G.↗

Design synthesis of a boron/epoxy reinforced metal shear web.

An advanced composite shear web design concept has been developed for the Space Shuttle Orbiter main engine thrust beam structure. Various web concepts were synthesized by a computer-aided adaptive random search procedure. A practical concept is identified having a titanium-clad, boron/epoxy plate with vertical boron/epoxy reinforced stiffeners. Baseline composite and titanium shear resistant designs are compared; the composite concept is 28% lighter than the titanium web. Element test results show the metal cladding effectively reinforces critical composite load transfer and fastener hole areas making the composite web concept practical for other shear structure applications.-

Laakso, J. H.↗

Fabrication of fuel pin assemblies, phase 3

Five full size and eight reduced length fuel pins were fabricated for irradiation testing to evaluate design concepts for a fast spectrum lithium cooled compact space power reactor. These assemblies consisted of uranium mononitride fuel pellets encased in a T-111 (Ta-8W-2Hf) clad with a tungsten barrier separating fuel and clad. Fabrication procedures were fully qualified by process development and assembly qualification tests. Detailed specifications and procedures were written for the fabrication and assembly of prototype fuel pins.

Keeton, A. R.↗

Experimental evaluation of thermal ratcheting behavior in UO2 fuel elements

The effects of thermal cycling of UO2 at high temperatures has been experimentally evaluated to determine the rates of distortion of UO2/clad fuel elements. Two capsules were rested in the 1500 C range, one with a 50 C thermal cycle, the other with a 100 C thermal cycle. It was observed that eight hours at the lower cycle temperature produced sufficient UO2 redistribution to cause clad distortion. The amount of distortion produced by the 100 C cycle was less than double that produced by the 50 C, indicating smaller thermal cycles would result in clad distortion. An incubation period was observed to occur before the onset of distortion with cycling similar to fuel swelling observed in-pile at these temperatures.

Phillips, W. M.↗

Preliminary analysis of hot spot factors in an advanced reactor for space electric power systems

The maximum fuel pin temperature for nominal operation in an advanced power reactor is 1370 K. Because of possible nitrogen embrittlement of the clad, the fuel temperature was limited to 1622 K. Assuming simultaneous occurrence of the most adverse conditions a deterministic analysis gave a maximum fuel temperature of 1610 K. A statistical analysis, using a synthesized estimate of the standard deviation for the highest fuel pin temperature, showed probabilities of 0.015 of that pin exceeding the temperature limit by the distribution free Chebyshev inequality and virtually nil assuming a normal distribution. The latter assumption gives a 1463 K maximum temperature at 3 standard deviations, the usually assumed cutoff. Further, the distribution and standard deviation of the fuel-clad gap are the most significant contributions to the uncertainty in the fuel temperature.

Lustig, P. H.↗

Carbide fuel pin and capsule design for irradiations at thermionic temperatures

The design of a capsule assembly to evaluate tungsten-emitter - carbide-fuel combinations for thermionic fuel elements is presented. An inpile fuel pin evaluation program concerned with clad temperture, neutron spectrum, carbide fuel composition, fuel geometry,fuel density, and clad thickness is discussed. The capsule design was a compromise involving considerations between heat transfer, instrumentation, materials compatibility, and test location. Heat-transfer calculations were instrumental in determining the method of support of the fuel pin to minimize axial temperature variations. The capsule design was easily fabricable and utilized existing state-of-the-art experience from previous programs.

Siegel, B. L.↗