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At least 55 records · Page 3
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Partial completion of post-irradiation examination of the ORNL samples

A total of six Grade 92 and two 800H samples were identified and secured as the selected ORNL samples, which were irradiated in the High Flux Isotope Reactor (HFIR) of ORNL to 0.46–14.66 dpa at 400 to ~720°C. The Grade 92 is an optimized heat and named as G92-2b. The two 800H samples have one sample as the standard solution-annealed condition and the other sample as a thermos-mechanically processed (TMP) condition with a maximized fraction of low-Σ coincidence site lattice (CSL) boundaries. Tensile tests were completed for the selected ORNL samples using the Irradiated Materials Examination and Testing (IMET) hot cell facility at ORNL. The G92-2b samples were tested at room temperature. The lower temperature irradiation at 400–496.7°C to 0.52–14.66 dpa resulted in minimal changes in tensile properties. However, the higher temperature irradiation at ~683.3–720°C to 0.46–14.63 dpa tended to reduce yield and tensile strength by about –136 to –311 MPa and plastic elongation with the increasing dose. The 800H and 800H-TMP samples were tensile-tested at their irradiation temperature 580°C, which showed >40% increase in yield and tensile strength with nearly 9% reduction in elongation for 800H-TMP compared with 800H.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Impact of corrosion test container material in molten fluorides

The effects of crucible material choice on alloy corrosion rates in immersion tests in molten LiF–NaF–KF (46.5–11.5-42 mol. %) salt held at 850 °C for 500 hrs are described. Four crucible materials were studied. Molten salt exposures of Incoloy-800H in graphite, Ni, Incoloy-800H, and pyrolytic boron nitride (PyBN) crucibles all led to weight-loss in the Incoloy-800H coupons. Alloy weight loss was ~30 times higher in the graphite and Ni crucibles in comparison to the Incoloy-800H and PyBN crucibles. It is hypothesized galvanic coupling between the alloy coupons and crucible materials contributed to the higher corrosion rates. Alloy salt immersion in graphite and Ni crucibles had similar weight-loss hypothesized to occur due to the rate limiting out diffusion of Cr in the alloys to the surface where it reacts with and dissolves into the molten salt, followed by the reduction of Cr from solution at the molten salt and graphite/Ni interfaces. As a result, both the graphite and the Ni crucibles provided sinks for the Cr, in the formation of a Ni–Cr alloy in the case of the Ni crucible, and Cr carbide in the case of the graphite crucible.

14 SOLAR ENERGY↗
DOE OSTI2019

Identification and Shipment Progress of the INL and LANL Samples to Be Received at ORNL

Ferritic-martensitic steel Grade 92, with T91 as a reference, together with austenitic stainless steel (or Incoloy Alloy) 800H and its grain-boundary-engineered version 800H-TMP (ThermoMechanical Processing), is investigated in this project. A total of fifteen Grade 92 samples from two or three heats, four T91 samples from two heats, and six 800H and six 800H-TMP samples from one heat were selected, which were primarily irradiated in the Advanced Test Reactor of Idaho National Laboratory (INL), with the rest in the High Flux Isotope Reactor of Oak Ridge National Laboratory (ORNL), BOR-60 of Russia, and Phénix reactor of France. The samples were irradiated in the temperature range of 241 to 720°C and a dose range of 1.28 to ~70 displacements per atom. All the INL samples, most of the ORNL samples, and part of the LANL (Los Alamos National Laboratory) samples from the ATR, BOR-60 and Phénix reactors have been received at the Irradiated Materials Examination and Testing (IMET) hot cell facility and Low Activation Materials Design and Analysis (LAMDA) laboratory of ORNL. Post-irradiation examination of the received samples is in progress at ORNL. The effects of irradiation temperature, dose, reactor (e.g., dose rate, neutron-spectrum-induced difference in transmutation), TMP, and heats variation on the microstructures and mechanical properties of the samples will be evaluated.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Post-Irradiation-Examination of the ORNL and LANL Samples

Ferritic-martensitic steel G92-2b (an optimized Grade 92 heat) and austenitic stainless steel 800H and its Grain Boundary Engineering (GBE)-treated version 800H-TMP (ThermoMechanical Processing) were irradiated in the HFIR. Ferritic-martensitic steel T91 was irradiated in the ATR. Selected G92-2b samples were irradiated up to 14.66 dpa at two temperature ranges: 400–496.7°C and 683.3–720°C. Selected 800H and 800H-TMP samples were irradiated to 1.28 dpa at 580°C. A selected T91 sample was irradiated to 6.5 dpa at 295°C. Vickers hardness measurements, fractography, and microstructural characterization were performed on the selected samples in LAMDA. Further systematic data analyses, together with some complementary experiments, will be pursued for these samples to foster peer-reviewed journal article publications.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Update on the Current R&D Activity in the U.S.

This presentation is to fulfil DOE commitment to GIF VHTR Materials PMB. Notched creep rupture testing: Weldment V-notch specimen test finished. 800C 60MPa, 15,716 hour rupture life. Two double V-notch specimen tests have finished. Crack growth tests: System modification was required for creep and creep-fatigue testing, Epsilon 3548COD Extensometer has been ordered, to allow load-line displacement measurements, to be paired with direct current potential drop already used to measure crack length, Sample redesigned in accordance to ASTM E2760 and E1457 to allow for both creep and creep-fatigue crack growth rate testing. Alloy 800H Weldments (1/2) and Alloy 800H Weldments (2/2), Variable Fatigue Testing, and Multi=axial Stress Relaxation. Continue work to finish 800H model, examine ?universal? constitutive model form.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Fabrication and Design Aspects of High-Temperature Compact Diffusion Bonded Heat Exchangers

The Very High Temperature Reactor (VHTR) using gas-cooled reactor technology is anticipated to be the reactor type for the Next Generation Nuclear Plant (NGNP). In this reactor concept with an indirect power cycle system, a high-temperature and high integrity Intermediate Heat Exchanger (IHX) with high effectiveness is required to efficiently transfer the core thermal output to a secondary fluid for electricity generation, hydrogen production, and/or industrial process heat applications. At present, there is no proven IHX concept for VHTRs. The current Technology Readiness Level (TRL) status issued by NGNP to all components associated with the IHX for reduced nominal reactor outlet temperatures of 750–800 degrees C is 3 on a 1–10 scale, with 10 indicating omplete technological maturity. Among the various potential IHX concepts available, diffusion bonded heat exchangers (henceforth called printed circuit heat exchangers, or PCHEs) appear promising for NGNP applications. The design and fabrication of this key component of NGNP with Alloy 617, a candidate high-temperature structural material for NGNP applications, are the primary focus of this paper. In the current study, diffusion bonding of Alloy 617 has been demonstrated, although the optimum diffusion bonding process parameters to engineer a quasi interface-free joint are yet to be determined. The PCHE fabrication related processes, i.e., photochemical etching and diffusion bonding are discussed for Alloy 617 plates. In addition, the authors’ experiences with these non-conventional machining and joining techniques are discussed. Two PCHEs are fabricated using Alloy 617 plates and are being experimentally investigated for their thermal-hydraulic performance in a High-Temperature Helium Facility (HTHF). The HTHF is primarily of Alloy 800H construction and is designed to facilitate experiments at temperatures and pressures up to 800 degrees C and 3 MPa, respectively. Furthermore, some preliminary microstructural and mechanical property characterization studies of representative diffusion bonded Alloy 617 specimens are presented. The characterization studies are restricted and less severe from an NGNP perspective but provide sufficient confidence to ensure safe operation of the heat exchangers in the HTHF. The test results are used to determine the design operating conditions for the PCHEs fabricated.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2011

Progress Report for Diffusion Welding of the NGNP Process Application Heat Exchangers (Revision 1)

The U.S. Department of Energy selected the high temperature gas-cooled reactor as the basis for the Next Generation Nuclear Plant (NGNP). The NGNP will demonstrate the use of nuclear power for electricity, hydrogen production, and process heat applications. The NGNP Project is currently investigating the use of metallic, diffusion welded, compact heat exchangers to transfer heat from the primary (reactor side) heat transport system to the secondary heat transport system. An intermediate heat exchanger will transfer this heat to downstream applications such as hydrogen production, process heat, and electricity generation. The channeled plates that make up the heat transfer surfaces of the intermediate heat exchanger will have to be assembled into an array by diffusion welding. This report describes the preliminary results of a scoping study that evaluated the diffusion welding process parameters and the resultant mechanical properties of diffusion welded joints using Alloy 800H. The long-term goal of the program is to progress towards demonstration of small heat exchanger unit cells fabricated with diffusion welds. Demonstration through mechanical testing of the unit cells will support American Society of Mechanical Engineers rules and standards development, reduce technical risk, and provide proof of concept for heat exchanger fabrication methods needed to deploy heat exchangers in several potential NGNP configurations.1 Researchers also evaluated the usefulness of modern thermodynamic and diffusion computational tools (Thermo-Calc and Dictra) in optimizing the parameters for diffusion welding of Alloy 800H. The modeling efforts suggested a temperature of 1150°C for 1 hour with an applied pressure of 5 MPa using 15 μm nickel foil as joint filler to reduce chromium oxidation on the welded surfaces. Good agreement between modeled and experimentally determined concentration gradients was achieved

08 HYDROGEN↗
DOE OSTI2012

Ways of improvement for the materials of sodium cooled fast reactors

The French sodium cooled prototype reactor ASTRID will take into account 'Generation IV' requirements, especially a long operational life-time (60 years) and a high efficiency. The good behavior of austenitic steel AISI316L(N), should be confirmed for a use, in moderately irradiated and unirradiated parts of ASTRID. Parts recovered from dismantled French sodium-cooled reactors will be characterized. Further experiments must be carried out concerning ageing of these components. Other materials will be chosen for fuel wrapping and cladding, in order to reduce creep and swelling under irradiation, (either conventional, or oxide-dispersed strengthened steels (ODSS). Corrosion of ODSS in the presence of sodium needs a serious assessment The lifetime of primary pumps components made of Duplex steels should also be assessed. The disruptions in steam generator tubes should be minimized and controlled; therefore, optimised designs and geometries must be established before defining the corresponding materials. Either Modified 9Cr1Mo or Incoloy 800H, might be candidates;it will be necessary to check whether austenitic steels are compatible with Modified 9Cr1Mo or Incoloy 800H in the same circuit. For all materials, the best manufacturing processes must be combined with thermal, mechanical treatments; calculations of phase diagrams (CALPHAD) might be used to optimise both treatments and chemical compositions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Re-establishment of Sandia National Labs super-critical carbon dioxide testing autoclave capability for exposure of metal alloys and polymers (Level 4 Milestone Report)

The sCO 2 system located in 916/160A, Sandia National Laboratories, CA, was constructed in 2014, for testing of materials in the presence of supercritical carbon dioxide (sCO 2 ) at high pressures (up to 3500 psi) and temperatures (up to 650°C). The basic design of the system consists of a thermally insulated IN625 autoclave, a high-pressure supercritical CO 2 compressor, autoclave heaters, temperature controllers, gas manifold, and temperature and pressure diagnostics. This system was modified in 2016 (sCO 2 compressor was removed) to enable corrosion studies with metal alloys in gaseous CO 2 at lower pressure (up to 300 psi) at 500°C. The capability was not used much afterwards until 2020, when preliminary tests using this capability (again without the supercritical CO 2 compressor) involved the exposure of fatigue and tensile specimens of HN 230 and 800H alloys to CO 2 gas for 168 hours in gaseous CO 2 . Using this capability, we finished experiments with low pressure (450 psi/ 3 MPa), high temperature (650°C) exposure of fatigue and tensile specimens of HN 230 and 800H alloys to CO 2 gas for 168 hours. The data from these experiments will be compared to that gathered from experiments performed in 2020 using the tube furnace and presented in a future report. It is to be noted that the tube furnace experiments ran 500-1500 hours, unlike the 168 hours of exposure in the recent experiment. This can help validate the use of the sCO 2 autoclave for both CO 2 and sCO 2 experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Summary of FY2024 Experimental Results to Support Development of New Inelastic Material Models and Validation of Section III, Division 5, Class A Design Rules

To address the data gap crucial for updating existing viscoplastic constitutive material models for the Class A materials in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors (ASME, 2023) for inelastic design analysis, Oak Ridge National Laboratory (ORNL) conducted experimental studies on Alloy 800H and Alloy 617 across a range of temperatures up to their maximum temperature limits in Section III, Division 5. The studies aim to characterize the materials' deformation behavior under both strain-controlled mechanical cyclic loading and thermal cycling conditions. This report summarizes ORNL's FY 2024 experimental findings on Alloy 800H and Alloy 617, focusing on pure fatigue tests, cyclic stress-strain curves, thermomechanical fatigue experiments, and verification of the high temperature cyclic damage summation design rules.

36 MATERIALS SCIENCE↗
DOE OSTI2012

CHARACTERIZATION OF ELEVATED TEMPERATURE PROPERTIES OF HEAT EXCHANGER AND STEAM GENERATOR ALLOYS

The Next Generation Nuclear Plant project is considering Alloy 800H and Alloy 617 for steam generator and intermediate heat exchangers. It is envisioned that a steam generator would operate with reactor outlet temperatures from 750 to 800 C, while an intermediate heat exchanger for primary to secondary helium would operate up to an outlet temperature of 950 C. Although both alloys are of interest due in part to their technical maturity, a number of specific properties require further characterization for design of nuclear components. Strain rate sensitivity of both alloys has been characterized and is found to be significant above 600 C. Both alloys also exhibit dynamic strain aging, characterized by serrated flow, over a wide range of temperatures and strain rates. High temperature tensile testing of Alloy 617 has been conducted over a range of temperatures. Dynamic strain aging is a concern for these materials since it is observed to result in reduced ductility for many solid solution alloys. Creep, fatigue, and creep-fatigue properties of Alloy 617 have been measured as well, with the goal of determining the influence of the temperature, strain rate and atmosphere on the creep fatigue life of Alloy 617. Elevated temperature properties and implications for codification of the alloys will be described.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Understanding Fundamental Material Degradation Processes in High Temperature Aggressive Chemomechanical Environments

The objective of this project is to develop a fundamental understanding of the mechanisms that limit materials durability for very high-temperature applications. Current design limitations are based on material strength and corrosion resistance. This project will characterize the interactions of high-temperature creep, fatigue, and environmental attack in structural metallic alloys of interest for the very high-temperature gas-cooled reactor (VHTR) or Next–Generation Nuclear Plant (NGNP) and for the associated thermo-chemical processing systems for hydrogen generation. Each of these degradation processes presents a major materials design challenge on its own, but in combination, they can act synergistically to rapidly degrade materials and limit component lives. This research and development effort will provide experimental results to characterize creep-fatigue-environment interactions and develop predictive models to define operation limits for high-temperature structural material applications. Researchers will study individually and in combination creep-fatigue-environmental attack processes in Alloys 617, 230, and 800H, as well as in an advanced Ni-Cr oxide dispersion strengthened steel (ODS) system. For comparison, the study will also examine basic degradation processes in nichrome (Ni-20Cr), which is a basis for most high-temperature structural materials, as well as many of the superalloys. These materials are selected to represent primary candidate alloys, one advanced developmental alloy that may have superior high-temperature durability, and one model system on which basic performance and modeling efforts can be based. The research program is presented in four parts, which all complement each other. The first three are primarily experimental in nature, and the last will tie the work together in a coordinated modeling effort. The sections are (1) dynamic creep-fatigue-environment process, (2) subcritical crack processes, (3) dynamic corrosion – crack initiation processes, and (4) modeling.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems

This research project, which includes collaborators from INL and ORNL, focuses on the study of alloy 617 and alloy 800H that are candidates for applications as intermediate heat exchangers in GEN IV nuclear reactors, with an emphasis on the effects of grain size, grain boundaries and second phases on the creep properties; the mechanisms of dislocation creep, diffusional creep and cavitation; the onset of tertiary creep; and theoretical modeling for long-term predictions of materials behavior and for high temperature alloy design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Challenges and successes in establishing a multipurpose fuels and materials irradiation capsule [Slides]

This presentation discusses the challenges and success related to designing and fabricating multipurpose capsules for irradiation experiments for irradiation within the Advanced Test Reactor (ATR) to be presented at tje DEVICE-MTR irradiation experiment design working group meeting hosted by the Japanese Atomic Energy Agency (JAEA). The information herein is limited to the design of the capsules only and no programmatic information from any project supporting the design of the multipurpose capsule is discussed. The intent of the design team was establish a capsule design that could accommodate multiple different irradiation project needs for the irradiation of nuclear fuel and materials of interest in the nuclear field. The main points of the presentation are as follows: use of tin as a molten thermal bond material was selected for the fueled capsule design and was found to be inadequate as it would corrode the stainless steel at high temperatures, this capsule design has high sensitivity to manufacturing of the gas gap and control of the gas mixture, and the fueled capsule has experienced welding challenges with the inner capsule material 800H Incoloy.

42 ENGINEERING↗
DOE OSTI2022

MRP: Qualification of Powder Metallurgy Hot Isostatic Pressed (PM HIP) Materials for Elevated-Temperature Nuclear Construction

PM HIP is a mature technology that offers many advantages that are attractive to the microreactor industry. The elevated-temperature creep-fatigue properties of the PM HIP 316H SS bar that has been characterized are reduced compared to the wrought material. Work will be carried out in fiscal year 2022 to identify the mechanisms responsible for the reduced creep-fatigue properties. Confirmatory testing of optimized material will be conducted to demonstrate long-term PM HIP properties comparable to the wrought-product form. The lessons learned for PM HIP 316H SS will be applied to the procurement and confirmatory testing of PM HIP Alloy 800H.

316H stainless steel↗
DOE OSTI2018

Multilayered Coatings for High-Temperature Steam Oxidation: TGA Studies up to 1000 °C

The aim of the study was to obtain a preliminary data focused on the steam oxidation performance of multilayered coatings deposited on 310 stainless steel and 800H Incoloy. The coatings were obtained through the aluminizing thermal diffusion technology with the formation of aluminides; some coating options had an additional top layer of selected composition. In this study, five different coating systems, as well as uncoated metals, were exposed at 1000 °C using thermogravimetric analysis instrumentation simulating an initial stage of steam oxidation corrosion. In this work, along with determination of the mass change of the samples, microanalyses using light optical microscopy, scanning electron microscopy (SEM) equipped with energy x-ray dispersive spectrometry (EDS), as well as surface micro-hardness determination, have been carried out. The designed coating systems demonstrated promising high-temperature steam oxidation performance with no breakaway degradation issues.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Heterogeneous creep deformation behavior of functionally graded transition joints (GTJs)

Dissimilar metal welds (DMWs) made between ferritic steels and austenitic alloys are used extensively in high-temperature power plant components. These DMWs experience premature creep failure in the ferritic steel’s heat affected zone (HAZ), close to the weld interface. Steep variations in microstructure and chemical composition across the dissimilar weld interface have been the contributing factors for the failure. Graded transition joints (GTJs), developed by functionally grading the chemical composition in layers, have been proposed as potential candidates to replace DMWs. In this research, GTJ coupons were fabricated between 2.25Cr-1Mo steel and Alloy 800H base material using two filler materials: (i) Inconel 82 and (ii) P87. These samples were aged at 600°C for 2000h to accelerate high-temperature microstructural evolution seen in service conditions, before subjecting them to short-term (~1 month) creep tests. Surface strains were measured using digital image correlation (DIC) technique to extract creep strain rates at different locations within GTJs. Both the GTJs exhibited heterogeneous creep strain distribution. Creep strain accumulated in the FGHAZ of 2.25Cr-1Mo steel, similar to type IV failure associated with Cr-Mo steel weldments. Microstructure based creep model framework was developed to describe the discrete creep strain rates observed in various 2.25Cr-1Mo steel regions of GTJs.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Building blocks for a digital twin of additive manufacturing

Properties and serviceability of additively manufactured components are affected by their geometry, microstructure and defects. These important attributes are now optimized by trial and error because the essential process variables cannot currently be selected from scientific principles. A recourse is to build and rigorously validate a digital twin of the additive manufacturing process that can provide accurate predictions of the spatial and temporal variations of metallurgical parameters that affect the structure and properties of components. Key building blocks of a computationally efficient first-generation digital twin of laser-based directed energy deposition additive manufacturing utilize a transient, three-dimensional model that calculates temperature and velocity fields, cooling rates, solidification parameters and deposit geometry. The measured profiles of stainless steel 316L and Alloy 800H deposits as well as the secondary dendrite arm spacing (SDAS) and Vickers hardness measurements are used to validate the proposed digital twin. The predicted cooling rates, temperature gradients, solidification rates, SDAS and micro-hardness values are shown to be more accurate than those obtained from a commonly used heat conduction calculation. Finally, these metallurgical building blocks serve as a phenomenological framework for the development of a digital twin that will make the expanding knowledge base of additive manufacturing usable in a practical way for all scientists and engineers.

36 MATERIALS SCIENCE↗