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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.

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At least 757 records · Page 42
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

ATF Cladding Mechanical Properties Report: Capability Demonstration

This report documents post-irradiation examination (PIE) activities performed in FY 2025 at Oak Ridge National Laboratory on chromium-coated (Cr-coated) and uncoated advanced zirconium alloy claddings irradiated in the High Flux Isotope Reactor (HFIR) to approximately 4 displacements per atom (dpa), corresponding to ~40 GWd/t burnup. Specimens were prepared in axial tension (ATT) and ring tension (RTT) geometries, and passive silicon carbide thermometry was employed to determine irradiation temperatures, which averaged 38–43 °C below the 330 °C design target.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Prioritizing Nuclear Materials for SAM-3 Neutron Irradiation Campaign: Structural and Cladding Materials Candidates

This report outlines a framework for selecting structural and cladding materials for the Nuclear Science User Facilities (NSUF) SAM-3 neutron irradiation campaign to support the advancement of nuclear energy technologies. The document begins with an introduction that provides background context, highlights the motivations for launching a new irradiation campaign, and defines the overall objectives. The core of the report describes the design considerations for the irradiation campaign, including capsule configurations, irradiation temperature ranges, and target dose levels (defined by displacements per atom, or dpa). The material recommendation was guided by the Specimen Identification and Prioritization (SIP) Working Group, a multidisciplinary team of experts representing national laboratories, academia, industry, federal government and agency. This group played a central role in identifying candidate materials, evaluating technical justifications, and ensuring alignment with boarder programmatic goals. A detailed set of criteria for material prioritization is then presented, taking into account reactor relevance, performance gaps, advanced manufacturing methods, and emerging material classes. Based on the input of SIP working group, specific materials were selected and justified for inclusion in the irradiation campaign by the NSUF leadership and its U.S. Department of Energy (DOE)-Office of Nuclear Energy (NE) management. The final section provides recommended capsule designs, summarizing critical parameters such as material type, fabrication method, sample geometry, irradiation conditions, and specimen quantities. This report serves as a foundation for executing a focused and high-impact neutron irradiation campaign aimed at addressing key materials challenges for both existing and advanced nuclear reactors.

36 - MATERIALS SCIENCE↗
DOE OSTI2018

Design Requirements for Ultrasonic Deformation Sensor for TREAT Experiments

Ultrasonic measurements of deformation can provide nondestructive measurements of dimensional changes rapidly (within tens to hundreds of microseconds). Further these methods are sensitive to both microstructural changes due to damage (from thermal, mechanical, and irradiation environments) and gross structural changes (such as swelling). As a result, ultrasonic methods have been applied to address needs in the nondestructive evaluation of structural components in nuclear power plants (including fuel cladding) during periodic pre-service and in-service inspection inspections. Ultrasonic measurements have been successfully used for nondestructive materials characterization, including nondestructive evaluation (NDE) of degradation and damage, microstructure characterization, and process control. Ultrasonic NDE is a critical element of the nuclear power industry’s in-service inspection program for maintaining the integrity of the pressure boundary, and is being actively investigated for post-irradiation examination of fuels. Ultrasonic measurements, typically performed at 10 MHz or higher, performed post irradiation show that fuel microstructural parameters, such as porosity and grain size, can be correlated to ultrasonic velocity. Ultrasonic methods historically have seen limited applicability to environments with high temperatures and irradiation. Though some environmental factors (such as temperature) affect the measurement (sound speed, for instance), the limitation is primarily due to the probes themselves. Most commonly, lead-zirconate-titanate (PZT) is used as the piezoelectric sensor material for ultrasonic nondestructive measurements. PZT is limited in its applicability at elevated temperatures (approximately above 300C). However, recent tests (through DOE-NE’s NEET-ASI program) have identified a number of alternatives that can operate at elevated temperatures (in excess of 400C) and can survive irradiation. Certain grades of PZT have also been demonstrated for use in imaging under-sodium in sodium fast reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Mechanical Behavior and Structure of Advanced Fe-Cr-Al Alloy Weldments

FeCrAl alloys are promising for developing accident tolerant nuclear fuel claddings. These alloys showed good environmental compatibility and oxidation resistance in elevated-temperature water and steam, as well as low radiation-induced swelling. However, FeCrAl alloys may suffer from several degradation mechanisms, one of which may be a susceptibility to cracking during welding. Here, a set of advanced modified FeCrAl alloys were designed and produced by varying Al-content and employing additions of Nb and TiC. Strength, ductility, and deformation hardening behavior of the advanced FeCrAl alloys and their weldments are discussed.

Gussev, Maxim N.↗
DOE OSTI2021

Evaluation of ceria as a surrogate material for UO 2 in experiments on fuel cracking driven by resistive heating

A variety of normal operation and accident scenarios can generate thermal stresses large enough to cause cracking in light-water reactor (LWR) fuel pellets. Cracking of fuel pellets can lead to reduced heat removal, larger centerline temperatures, and localized stress in cladding all of which impact fuel performance. Furthermore, pellet cracking also contributes to a temperature reduction in the pellet since the pellet fragments tend to move towards the heat sink (cladding), and the heat flow remains predominantly radial despite the presence of cracks. It is important to understand the temperature profile on the pellet before and after cracking to improve cracking models in fuel performance codes However, in-reactor observation and measurement of cracking is very challenging owing to the harsh environment and design of fuel rods. Recently, an experimental pellet cracking test stand was developed for separate effects testing of normal operations and accident temperature conditions, using thermal imaging to capture the pellet surface temperature for evaluation of thermal stresses and optical imaging to capture the evolution of cracking in real time. Cracking experiments were initially performed using ceria (CeO 2 ) as a surrogate fuel material, which is useful for developing and demonstrating the experimental approaches but is also valuable in its own right for cracking model development and validation. A combination of induction and resistance heating was used for volumetric heat generation in the pellet creating a thermal gradient. The material properties of CeO 2 and UO 2 are reviewed and compared for use in model development. Simulations of the experiment were performed to evaluate the behavior of the surrogate (CeO 2 ) fuel in BISON. The measured temperature profiles from BISON models match reasonably well with the observed experiments for the ceria pellets before cracking. The findings from this work will help improve confidence in fracture models used for fuel pellets under similar in-reactor conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Encapsulated silicene: A robust large-gap topological insulator

The quantum spin Hall (QSH) effect predicted in silicene has raised exciting prospects of new device applications compatible with current microelectronic technology. Efforts to explore this novel phenomenon, however, have been impeded by fundamental challenges imposed by silicene’s small topologically nontrivial band gap and fragile electronic properties susceptible to environmental degradation effects. Here we propose a strategy to circumvent these challenges by encapsulating silicene between transition-metal dichalcogenides (TMDCs) layers. First-principles calculations show that such encapsulated silicene exhibit a two-orders-of-magnitude enhancement in its nontrivial band gap, which is driven by the strong spin–orbit coupling effect in TMDCs via the proximity effect. Moreover, the cladding TMDCs layers also shield silicene from environmental gases that are detrimental to the QSH state in free-standing silicene. In conclusion, the encapsulated silicene represents a novel two-dimensional topological insulator with a robust nontrivial band gap suitable for room-temperature applications, which has significant implications for innovative QSH device design and fabrication.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Intrinsic thermal interfacial resistance measurement in bonded metal–polymer foils

Heat conduction through bonded metal–polymer interfaces often limits the overall heat transfer in electronic packaging, batteries, and heat recovery systems. To design the thermal circuit in such systems, it is essential to measure the thermal interfacial resistance (TIR) across ~1 µm to 100 µm junctions. Previously reported TIR of metal–polymer junctions utilize ASTM E1530-based two-block systems that measure the TIR by applying pressure across the interface through external heating and cooling blocks. Here, we report a novel modification of the ASTM-E1530 technique that employs integrated heaters and sensors to provide an intrinsic TIR measurement of an adhesively bonded metal–polymer junction. We design the measurement technique using finite element simulations to either passively suppress or actively compensate the lateral heat diffusion through the polymer, which can minimize the systematic error to ≲5%. Through proof-of-concept experiments, we report the TIR of metal–polymer interfaces made from DuPont’s Pyralux double-side copper-clad laminates, commonly used in flexible printed circuit boards. Our TIR measurement errors are <10%. We highlight additional sources of errors due to non-idealities in the experiment and discuss possible ways to overcome them. Our measurement technique is also applicable to interfaces that are electrically insulating such as adhesively joined metal–metal junctions and sputter-coated or welded metal–polymer junctions. Altogether, the technique is capable of measuring TIR ≳10 –5 m 2 KW –1 in bonded metal–polymer foils and can be tailored for in situ measurements in flexible electronics, circuit packaging, and other hybrid metal–polymer systems.

47 OTHER INSTRUMENTATION↗
DOE OSTI2016

Complete Non-Radioactive Operability Tests for Cladding Hull Chlorination

Non-radioactive operability tests were made to test the metal chlorination reactor and condenser and their accessories using batch chlorinations of non-radioactive cladding samples and to identify optimum operating practices and components that need further modifications prior to installation of the equipment into the hot cell for tests on actual used nuclear fuel (UNF) cladding. The operability tests included (1) modifications to provide the desired heating and reactor temperature profile; and (2) three batch chlorination tests using, respectively, 100, 250, and 500 g of cladding. During the batch chlorinations, metal corrosion of the equipment was assessed, pressurization of the gas inlet was examined and the best method for maintaining solid salt product transfer through the condenser was determined. Also, additional accessing equipment for collection of residual ash and positioning of the unit within the hot cell were identified, designed, and are being fabricated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Benchmark Specifications for TREAT Tests M5, M6, and M7

Detailed information describing three TREAT tests performed on metallic fuels has been collected and organized for use as a benchmark in evaluating the performance of metallic fuel models and codes. The tests, designated M5, M6, and M7, subjected EBR-II-irradiated fuel pins to a single type of overpower transient (at prototypical conditions with full coolant flow and an exponential power rise on an 8 second period) in flowing sodium loops. Six fuel pins were tested; five were ternary (U-19Pu-10Zr) alloy fuel clad in D9 with burnup ranging from 0.8 to 9.8 at. %, and one was binary alloy fuel (U-10Zr) clad in HT9 with 2.9 at. % burnup. The information gathered from the test records is expected to be useful for pre-transient characterization of the irradiated fuel pins as well as the transient analysis of the metallic fuel when subjected to severe accident conditions. This report presents benchmark specifications for the M5, M6, and M7 TREAT tests, identifies where primary sources of benchmark-related information can be found, and includes background information to help a user of the data understand their applicability and limitations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Characterization of Large Batch 14YWT Heat (NFA-2) for JOYO Irradiations (Rev.1)

The development and deployment of advanced cladding materials in advanced reactors requires the collection of relevant performance data for each cladding concept. Novel cladding concepts typically undergo significant innovation and iteration over the course of their development, and intention should be given to providing a snapshot of present performance. Thus, a modern heat of the nanostructured ferritic alloy (NFA) 14YWT is produced using state-of-the-art procedures for impurity minimization combined with thermomechanical processing methods designed to optimize fracture toughness, tensile strength, and ductility. This new heat of 14YWT, NFA-2, will provide the Advanced Fuels Campaign with adequate material for upcoming irradiation and mechanical testing campaigns. This report outlines procurement of new gas atomized 14YWT powder, production of four new bars of 14YWT, and subsequent mechanical and microstructural property characterization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

CTF Preprocessor User's Manual

This document describes how a user should go about using the CTF pre- processor tool to create an input deck for modeling rod-bundle geometry in CTF. The tool was designed to generate input decks in a quick and less error-prone manner for CTF. The pre-processor is a completely independent utility, written in Fortran, that takes a reduced amount of input from the user. The information that the user must supply is basic information on bundle geometry, such as rod pitch, clad thickness, and axial location of spacer grids--the pre-processor takes this basic information and determines channel placement and connection information to be written to the input deck, which is the most time-consuming and error-prone segment of creating a deck. Creation of the model is also more intuitive, as the user can specify assembly and water-tube placement using visual maps instead of having to place them by determining channel/channel and rod/channel connections. As an example of the benefit of the pre-processor, a quarter-core model that contains 500,000 scalar-mesh cells was read into CTF from an input deck containing 200,000 lines of data. This 200,000 line input deck was produced automatically from a set of pre-processor decks that contained only 300 lines of data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Handbook of the Materials Properties of FeCrAl Alloys For Nuclear Power Production Applications

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including as accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the state-of-the-art on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Handbook on the Material Properties of FeCrAl Alloys for Nuclear Power Production Applications (FY18 Version: Revision 1)

FeCrAl alloys are a class of alloys that have seen increased interest for nuclear power applications including accident tolerant fuel cladding, structural components for fast fission reactors, and as first wall and blanket structures for fusion reactors. FeCrAl alloys are under consideration for these applications due to their inherent corrosion resistance, stress corrosion cracking resistance, radiation-induced swelling resistance, and high temperature oxidation resistance. A substantial amount of research effort has been completed to design, develop, and begin commercial scaling of FeCrAl alloys for nuclear power applications over the past half a century. These efforts have led to the development of an extensive database on material properties and process knowledge for FeCrAl alloys but not within a consolidated format. The following report is the first edition of a materials handbook to consolidate the current state of knowledge on FeCrAl alloys for nuclear power applications. This centralized database focuses solely on wrought FeCrAl alloys, oxide dispersion strengthened alloys, although discussed in brief, are not covered. Where appropriate, recommendations for applications of the data is provided and current knowledge gaps are identified.

36 MATERIALS SCIENCE↗
DOE OSTI2020

CTF User's Manual (V.4.2)

This document describes how a user should go about using the CTF pre-processor tool to create an input deck for modeling rod-bundle geometry in CTF. The tool was designed to generate input decks in a quick and less error-prone manner for CTF. The pre-processor is a completely independent utility, written in Fortran, that takes a reduced amount of input from the user. The information that the user must supply is basic information on bundle geometry, such as rod pitch, clad thickness, and axial location of spacer grids - the pre-processor takes this basic information and determines channel placement and connection information to be written to the input deck, which is the most time-consuming and error-prone segment of creating a deck. Creation of the model is also more intuitive, as the user can specify assembly and water-tube placement using visual maps instead of having to place them by determining channel/channel and rod/channel connections. As an example of the benefit of the pre-processor, a quarter-core model that contains 500,000 scalar-mesh cells was read into CTF from an input deck containing 200,000 lines of data. This 200,000 line input deck was produced automatically from a set of pre-processor decks that contained only 300 lines of data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Status Report on Development of a Cask to Enable Pulsed Neutron Characterization of Irradiated Fuel

We present the design of a shielding cask that allows pulsed neutron characterization of irradiated fuel rodlets prior to their destructive examination in hot cells. The goal is to provide complementary and informative measurements that will inform subsequent hot cell PIE examinations by identifying typical and atypical regions with respect to microstructure, tomographic data, or isotope densities. Time-of-flight neutron diffraction has the potential to offer efficient, non-destructive and non-contact microstructural characterization of irradiated fuel pins with spatial resolution of 1 mm3 to 1 cm3 while for energy-resolved neutron imaging (and by extension tomography) a resolution of 100 μm was demonstrated. The potential results include crystallinity vs. amorphous volumes and microstructural information such as phase compositions and textures from the diffraction data as well as distances (e.g. pellet to cladding), cracks, and isotopic distributions of minor actinides, fission products as well as fission gas partial pressures e.g. in the plenum from energy resolved neutron imaging. The capability will identify regions of atypical behavior and provide cost effective bulk characterization of entire samples prior to destructive evaluation in hot cells. This effort is a collaboration between LANL, INL, ORNL, and UC Berkeley. A key facet of the initiative is the ability to cost effectively ship irradiated fuels from INL to the pulsed spallation neutron source at LANSCE. The irradiated fuel would be loaded in the custom designed cask (designated RaMHaM) at INL hot cells which then could be shipped in a BRR Type B shipment container between INL and LANL. No hot cell is required after the loading of the sample at INL in the pathway laid out in this report, greatly simplifying handling at LANL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Resumption of water capsule reactivity-initiated accident testing at TREAT

A series of integral reactivity-initiated accident (RIA) commissioning experiments were completed in a new static water capsule in the Transient Reactor Test Facility (TREAT), marking the first such tests in the U.S. in more than 40 years. The test campaign included a verification test followed by five tests in the Static Environment Rodlet Transient Test Apparatus (SERTTA) capsule. The capsule initial conditions varied from room temperature and pressure up to 200°C and 2.5 MPa, with energy depositions varying between ~500–1100 J/gUO 2 . The series of tests allowed for a number of instrumentation qualifications and demonstrations, including cladding thermometry, rodlet plenum pressure, cladding elongation, and an electro-impedance boiling detector. Post transient examinations such as gamma emission spectroscopy, profilometry, and microscopy were performed to document the end state of the fuel rods. Furthermore, the results from the experiments show that the fuel rodlets behaved very similar to historical tests under similar energy depositions. This paper documents the design of the capsule and highlights some results from the commissioning tests and post-transient examination.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Neutron and Thermal Embrittlement of RPV Steels: An Overview

Because the reactor pressure vessel (RPV) represents the first structural line of defense against the release of radiation to the public, the design and fabrication of the RPV for any nuclear reactor facility is performed at very high standards in accordance with consensus codes that are based on mechanical and physical properties of the steels used to construct the vessel. Nuclear RPVs may weigh up to 800 tons with wall thicknesses up to approximately 330 mm and are clad on the inside with stainless-steel weld metal and given a final post-weld heat treatment. The RPV is a unique structural component in that it operates under high pressures and temperatures and is exposed to relatively high neutron radiation. Although typical RPV steels and welds have excellent fracture toughness at room temperature and above when put into service, the degrading effects of high-energy neutron irradiation can cause levels of irradiation-induced embrittlement in radiation-sensitive materials of concern for the structural integrity of the RPV. In recent decades, remarkable progress has been made in developing a mechanistic understanding of irradiation embrittlement. This progress includes developing physically based and statistically calibrated models of Charpy V-notch-indexed transition temperature shifts based on results from RPV surveillance programs complemented by significant results from comprehensive research experiments performed in test reactors. In addition, advances in elastic-plastic fracture mechanics allow for a relatively small number of relatively small specimens to characterize the fracture toughness of RPV steels with statistical confidence. This paper presents a review of the primary mechanical properties, test procedures, examples of applicable codes and standards, and specimen types used to characterize RPV steels and welds, the effects of neutron irradiation on those most relevant mechanical properties, and a brief review of the effects of thermal aging on RPV materials. The paper closes with a summary.

Nanstad, Randy K.↗
DOE OSTI2018

Verification of Pilger Mill Performance

The Nuclear Technology Research & Development program of the Office of Nuclear Energy has implemented a program to develop an advanced reactor cladding fabrication capability by using the extrusion and pilgering processes. Oxide-dispersion strengthened (ODS) steels are a promising class of advanced materials, but processes need to be developed to produce tubing for these advanced alloys. The Pacific Northwest National Laboratory (PNNL) has converted its rolling mill into a lab-scale pilger mill, which required designing pilger tooling and rolls that contained multiple pilgering grooves and designing a servo-driven feed mechanism. Upon assembling and checking out the system, initial pilgering runs were completed. This report will provide verification of the design and performance of the lab-scale pilger mill.

36 MATERIALS SCIENCE↗