Search NASA⌕ Search

SEARCH · Search NASA

Results for “nuclear reactor materials”

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 379 records · Page 21
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Milestone Report: Effect of High Temperature CO 2 on Haynes 230 Alloy

The supercritical carbon dioxide (sCO 2 ) Brayton cycle is a promising candidate for future nuclear reactors due to its ability to improve power cycle energy conversion efficiency. The sCO 2 Brayton cycle can operate with an efficiency of 45-50% at operating temperatures of 550-700 C. One of the greatest hurdles currently faced by sCO 2 Brayton cycles is the extreme corrosivity of sCO 2 . This affects the longevity of the power cycle and thus the levelized cost of electricity. Past studies have shown that sCO 2 corrosion occurs through the formation of metal carbonates, oxide layers, and carburization, and alloys with Cr, Mo and Ni generally exhibit less corrosion. While stainless steels may offer sufficient corrosion resistance at the lower range of temperatures seen by the sCO 2 Brayton cycles, more expensive alloys such as Inconel and Haynes are typically needed for the higher temperature regions. This study investigates the effects of corrosion on the Haynes 230 alloy, focusing on changes in the mechanical properties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Progress Report on Performance of A709 and G91 Steels in Sodium

Specimens in six different processing and heat treatment conditions of A709 H58776 were exposed to sodium at 550, 600, and 650°C, respectively for various exposure times. G91 base metal was tested in sodium at 550, 600, and 650°C, and G91 weldment at 550 and 600°C, respectively. Parallel thermal aging experiments were conducted on G91 and A709 to obtain thermal aging data for comparison with sodium exposure results to separate the thermal and sodium effects. The corrosion data obtained on A709 steel continue to show low corrosion rates, which indicate good compatibility of A709 with sodium when oxygen content is controlled. G91 has also shown acceptable corrosion rates over the temperature range of 550-650°C in sodium. Thermal aging or sodium exposures reduced the tensile strength, uniform elongation, and total elongation for A709 ESR, AOD specimens. These specimens consistently showed higher yield stress and tensile strength and lower uniform and total elongations after sodium exposure than after thermal aging under comparable testing conditions. It implies that there was an additional effect of sodium exposure that gave rise to an increase in tensile strength of A709. Dynamic strain aginginduced flow serrations were completely removed by either thermal aging or sodium exposure at 650°C, which imply that carbon or nitrogen in solution was removed from the solution and formed precipitates during aging or sodium exposures. The tensile data ruled out the possibility of decarburization under the sodium exposure conditions. The HOMO specimens behaved somewhat differently from the ESR and AOD specimens in sodium. While thermal aging and sodium exposures at 550 and 600°C decreased the yield stress and the ultimate tensile strength of G91, there was virtually no additional effect resulting from sodium exposures at these two temperatures. In contrast, sodium exposures at 650°C had a drastic effect on the yield stress and the ultimate tensile strength of G91. It is suggested that G91 experienced decarburization in the 650°C sodium environments. Carbon concentrations in sodium in the SMT-1 and SMT-2 loops were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8-1.2 ppm in the SMT-1 loop and 0.3-0.7 ppm in the SMT-2 loop. The carbon activity in sodium in the SMT-2 loop was estimated to be 0.03-0.08 at 600°C, and 0.08-0.2 at 550°C. The carbon activity in sodium at 650°C in the SMT-1 loop was 0.04-0.07. Equilibrium simulation of the carburization – decarburization processes was conducted for A709 and G91 steels exposed in sodium environments at temperatures of 550-700°C. The carbon activity-concentration relationship for G91 was re-evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC carbides dominated the process at low carbon activities. Formation of NbC increases the decarburization resistance of G91 steel. It remains to be understood whether the decarburization resistance provided by MC carbides in G91 can be maintained during long-term operations of SFRs. The carbon concentration-activity relationship for A709 was calculated based on the equilibrium of fcc-austenite and M 23 C 6 carbide phase. The calculations showed that A709 decarburizes at 650°C in the SMT-1 loop environments, which is different from the experimental findings. The carbon concentration-activity relationship in A709 was further evaluated by considering four phases, including fcc-austenite, M 23 C 6 , NbC, and TiC carbide phases. TiC is the most stable carbides among the three carbide phases. The carburization-decarburization process in A709 is dictated by M 23 C 6 carbides at high carbon activities and by TiC at low carbon activities. Formation of TiC increases the decarburization resistance of A709 steel. Because of the complexity of the precipitation process in A709 during sodium exposures, detailed characterization of precipitates in sodium-exposed specimens and improved thermodynamic models are needed to understand the carburization-decarburization behavior of A709. Kinetic analysis of carbon transfer will be investigated for G91 and A709 in future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

FLiBe Capsule Testing of 316H

This milestone was originally envisioned for completion in FY20 and was delayed due to the COVID-19 response and lack of FLiBe salt. In March 2021, FLiBe salt was received from Kairos Power. The nine capsule experiments with type 316H stainless steel specimens and 316 capsules were completed at 550° and 650°C for 1000 h to bracket the conditions expected in the subsequent thermal convection loop (TCL) experiment. In three capsules, Be metal was added for redox control. The specimen mass losses were low, similar to those observed previously for FLiNaK salt and the addition of Be reduced the average mass loss. Initial characterization suggested that some reaction with Be may have occurred. Therefore, Be was not included in the TCL experiment.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Report on FY 2021 testing in support of integrating Alloy 800H and 2.25Cr-1Mo for EPP analysis

The use of simplified Elastic-Perfectly Plastic (EPP) design methods avoids the complexities and limitations of the elastic analysis approach in evaluating strain limits and creep-fatigue damage for Class A metallic coolant boundary components in ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B. No stress classification and linearization are required in the EPP methods. Further, the strain limits and creep-fatigue damage evaluation procedures based on elastic analysis results were developed for the reactor operating conditions where uncoupled, rate-independent plasticity and stationary creep can be used to approximate the deformation in the component adequately. But at very high temperatures, the deformation is rate-dependent, and it can only be approximated adequately by unified viscoplastic model. Thus, for example, the elastic analysis approach cannot be used to evaluate strain limits and creep-fatigue damage for Alloy 617 above 650 °C. The EPP methods remove the temperature restriction. They can be used to evaluate strain limits and creep-fatigue damage for the full range of temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Initial framework for engineering-scale statistical creep-fatigue modeling

This report describes the integration of new solid and interface-cohesive mechanics systems into MOOSE. The purpose of these new systems is to support the ability of MOOSE to run full-field crystal plasticity finite element method simulations of key material processes in high temperature metallic materials. These simulations could be used to help accurately predict the performance of key high temperature structural materials in future advanced nuclear reactor components. Previous work implemented preliminary versions of many of these systems in MOOSE Apps. The current work reports on their integration into the main MOOSE tensor mechanics module along with associated improvements to the basic formulations and numerical implementations. Finally, the report provides an example of the full-field crystal plasticity simulations now possible in MOOSE, including examples of realistic geometries requiring millions of degrees of freedom to resolve the microstructural features and macroscale geometry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Corrosion of 316H Stainless Steel Specimens in Two FLiBe (LiF-BeF 2 ) Salt Batches

This milestone was originally envisioned for completion in FY21 and was delayed due to the COVID-19 response and difficulties in fabricating fluoride salts. To complete the milestone, commercial FLiBe was compared to a batch of FLiBe produced by the conventional hydrofluorination process. However, the batch was stopped early due to an HF leak and the salt did not undergo the final H2 sparging. Subsequent static compatibility testing of 316H specimens in 316H capsules resulted in small mass losses for the commercial FLiBe and much larger mass losses for the 2nd batch of FLiBe.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Eddy Current Flow Meter Measurements and Modeling Validation

The goal of the eddy current flow meter (ECFM) is to monitor velocities and temperatures of conductive liquid metals. The ECFM can also detect voids in two-phase liquid metal flows. Measurement of these flows can greatly improve the accuracy of the empirical formulas used for fluid dynamics near the reactor core and aid in mapping any non-uniformities in fluid flow such as estimates of temperature and phase. While the eventual goal is to develop the ECFM for the liquid sodium surrounding the Versatile Test Reactor (VTR) core, this report focuses on modeling and experimental validation of ECFM for a moving solid test rod and for liquid mercury in the Target Test Facility (TTF) flow loop to better understand and further develop ECFM measurement capabilities by validating model predictions with benchmark experiments. Eventually the models could be used to design and optimize future ECFMs for specific applications such as monitoring flow within the VTR pool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Development of an accelerated creep testing technique via multi-specimen loading

This report summarizes a research effort undertaken by Argonne National Laboratory to accelerate the development of creep data package for code qualification. The goal is to explore new creep testing techniques to shorten the overall time needed to generate high-quality data to support an ASME code case on laser powder bed fusion 316H stainless steel. Progresses on the design and implementation of a 3-specimen load train and an innovative strain measurement method with direct potential current drop (DCPD) technique are reported.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Complete the first phase of the comprehensive characterization of repair welding performed on irradiated Ni alloy 182

This report describes the research activities on post-weld evaluation and characterization of the quality and properties of welds made on the first phase of repair welding campaign on irradiated Ni alloy 182. The equipment and capabilities of the campaign were developed jointly by the U.S. Department of Energy, Office of Nuclear Energy, Light Water Reactor Sustainability Program, the Electric Power Research Institute, Long Term Operations Program (and the Welding and Repair Technology Center), and Oak Ridge National Laboratory. Irradiated nickel alloy 182 with doped boron ranging from 5 wppm to 15 wppm, were laser welded in the hot cell. The weld samples were cross-sectioned and characterized. The significant, on-going effort to weld irradiated alloys with high helium concentrations and comprehensively analyze the results will eventually yield validated repair techniques and guidelines for use by the nuclear industry in extending the operational lifetimes of nuclear power plants.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Summary of FY 2024 A709 Code Case testing at ANL, INL and ORNL

A collaborative research and development effort in support of the Alloy 709 Code Case qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors is being carried out at Oak Ridge National Laboratory (ORNL), Idaho National Laboratory (INL), and Argonne National Laboratory (ANL). Key testing data for the Alloy 709 100,000-hr near-term Code Case submittal to ASME is expected to be completed by the end of 2024, with design parameters anticipated to be finalized in FY 2025. This report summarizes the testing results for three commercial heats of Alloy 709 conducted across three laboratories, reviews the current testing status, and outlines the remaining data needed to support the first Alloy 709 Code Case submittal to ASME. The Alloy 709 Code Case plan remains on schedule.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Performance Testing of Additively Manufactured 316L Stainless Steel in Light Water Reactor Environment

This report summarizes research activities conducted at Argonne National Laboratory in support of the development, qualification and certification of additively manufactured (AM) metallic components for the long-term sustainability of light water reactors. In this program, AM 316L stainless steel (SS) has been evaluated in light water reactor (LWR) environments for their fatigue, environmentally assisted fatigue, and stress corrosion cracking performances on specimens in as-printed condition, aiming for facilitating the regulatory acceptance and ultimately adoption of AM components in aging LWRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

First-Principles Studies of Tritium Species Diffusivity Across the Interface of Nickel-Plated Zircaloy-4

α-Zr and its alloys are known as best 3 H getters due to their excellent corrosion resistance under chemically corrosive environment, low thermal neutron absorption cross-section and robust mechanical strength at high temperature. In this work, we conducted a systematic study on constructing and optimizing Ni and Zr surfaces, interfacing the stable Ni and Zr surfaces to create an optimal Ni-Zr interface, and an understanding the diffusion mechanisms for 3 H across Ni(111)-Zr(0001) interface with and without vacancies and impurities such as Sn and O.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Metallurgical Analysis and Forward Modeling of Weld Distortion in SMR Containment Vessels

This work aimed to apply Sandia’s expertise in metallurgy and modeling to enable the use of hybrid laser arc welding for building nuclear reactor containment structures, via a collaboration with Holtec International. Experimental observations were coupled with finite element analysis to resolve microstructure development, mechanical properties, distortion, and residual stress in welds relevant to the production of the Holtec SMR-160. High residual stresses were observed in welds that were not subjected to preheat. Meanwhile, the microstructure of the welds generally exhibited a narrow heat affected zone relative to conventional arc welds. FEA appeared to be effective in simulating the thermal/mechanical conditions that occur during hybrid laser arc welding of simplified and instrumented test welds. Subsequently, FEA was used to perform sensitivity analyses for various weld geometries that would be prohibitively costly to assess with physical experiments. Insights from the study were used to inform Holtec’s welding process, and successful production welds were performed in 2025.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

The Effect of Thermal Aging on the Performance of High-Cr Ni-based Alloys in Primary Water Environment

Nickel-based Alloy 690, and the associated weld Alloys 52 and 152, are used for nozzle penetrations in replacement heads for pressurized water reactor (PWR) vessels. They are chosen for their excellent overall resistance to general corrosion and environmental degradation, primarily stress corrosion cracking (SCC). However, many of the existing PWRs are expected to operate beyond their first or second license renewal (e.g., 40 to 60 years or beyond 60 years). Hence, the thermal stability of Ni-Cr alloys is a potential concern for the long-term performance of both existing and advanced nuclear power plants. Moreover, Ni-Cr alloys are being considered for high-level radioactive waste disposal containers that will be required to contain the waste for at least 10,000 years.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Multilayer Architecture Barriers for High Temperature Nuclear Applications

Innovative thin film multilayer metal/ceramic barrier coatings have been developed by Argonne National Laboratory to protect fuel cladding against different phenomena that can be encountered under the harsh operating conditions of advanced nuclear reactors. One phenomenon is Fuel-Cladding Chemical Interaction (FCCI), which can represent a challenge to the performance of steel-clad metallic fuels in sodium-cooled fast reactors (SFRs), especially when operating to high burnups at elevated cladding temperatures, and long fuel residence time. Another phenomenon is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated cladding during off normal conditions, including loss of coolant accidents (LOCA). This work provides details of the development and testing of a thin film (few microns) multilayer metal/ceramic (CrY/Y 2 O 3 ) coatings on steel cladding and its performance against FCCI. The coating architecture evolved through systematic mechanical testing, radiation tolerance assessments, and high-temperature diffusion studies from a preliminary multilayer concept to an optimized high-performance design. The optimized structure leverages stoichiometric Y 2 O 3 for its exceptional radiation stability, chemical inertness, and ultra-low fission product diffusivity (e.g., Ce diffusivity ~10 -20 m 2 /s in irradiated Y 2 O 3 at 650 °C). This is paired with ductile chromium-rich interlayers (Cr 95 Y 5 ), which significantly enhance fracture toughness and mechanical resilience. Design refinements, including increased ceramic layer thickness and optimized Cr–Y interlayer composition, eliminated fracture propagation and preserved coating integrity under extreme mechanical stress (>10× typical reactor conditions) and prolonged radiation exposure (~300 dpa). Those conditions are well beyond expected conditions during normal operations in SFR, and cover some of the transient conditions as well. In addition, the coating deposition was demonstrated to be applied uniformly to the inner surface of steel tubes representing prototypic SFR cladding. Another phenomenon under consideration in this work is the corrosion of zirconium-based cladding under prototypic boiling water reactor (BWR) conditions and associated off normal conditions including loss of coolant accident (LOCA). Another multilayer thin film coating is considered for mitigation of this phenomenon, where a coating of yttrium aluminum and yttrium aluminum ceramic/metal thin film coating. Preliminary results on the coating performance during high temperature thermal cycling (up to 1000 C) have shown promising performance of the coating. On-going optimization of the coating will continue in FY26.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Embedded Aluminum Nitride Sensors for Advanced Reactors

This project aims to manipulate the of growth of aluminum nitride (AlN) inclusions in an iron-chromium-aluminum (FeCrAl) alloy, using the principle of powder metallurgy and heat treatment methods. to promote the formation of AlN phase within FeCrAl for embedded sensing. This effort seeks to fill in a gap with respect to robust sensor hardware for ubiquitous structural health monitoring of advanced nuclear reactors. A systematic evaluation of various solid-state methods will be performed to understand the influence of process conditions on AlN growth and to promote the growth of desirable AlN phase. Fabricated specimens will be analyzed to investigate the AlN structures that are formed using a suite of tools to characterize the concentration, morphology, and distribution of AlN within the FeCrAl substrate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗