SEARCH · Search NASA
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.
This is an update on the Alloy 617 notched specimen creep testing and crack growth rate testing performed in FY23. Long-term notch testing continues, with some new results on double V-notch specimens. Creep-fatigue crack growth rate testing has begun and some preliminary results are shown.
This report covers the results from crack growth rate testing performed at Idaho National Laboratory in the fiscal year 2024. It includes two creep crack growth rate tests and two creep-fatigue crack growth rate tests. Results are shown for each of these tests, included crack growth measured during the test using direct current potential drop, and measurements based off of the fracture surface. Load line displacement, as well as load, were also collected during the tests. Detailed analysis of the results will continue into the next fiscal year, however, a number of potential improvements to the equipment and testing procedures were identified that will increase testing reliability and throughput, as well as decrease complications with the analysis.
This report summarizes updates on the ongoing development of the new American Society of Mechanical Engineers Section III, Division 5, Class B rules to address the potential failure modes of Class B components. An initial design-by-analysis approach and supporting design rules are presented. The elastic-perfectly plastic (EPP) analysis approach has been adopted for primary stress limit and ratcheting check to prevent the component failure against the structural failure mode against primary load and strain accumulation due to cyclic loads. The creep-fatigue damage assessment uses explicitly defined elastic follow-up calculated from the stress concentration region of given component. Damage fraction calculation uses a novel coupled approach to capture the influence of the elastic follow-up and provide adequate conservatism for Class B components. The proposed Class B rules does not use the stress classification approach and uses the combined loads to assess component.
Overview of high temperature metals, including component construction rules for advanced reactor designs, assessment of filler metal for Alloy 800H weldments, high-temperature crack growth rate testing, work packages and contributors for FY24, creep-fatigue damage summation evaluation, GCR metal needs, and high temperature metals from other programs.
Conventional power-generation plants, including nuclear plants, have been traditionally designed to provide a steady baseload energy capacity, optimizing output efficiency while minimizing variable costs. However, the growing adoption of large-scale renewable energy-generation systems, which rely on intermittent sources such as solar and wind, has introduced more variability into the energy supply in interconnected electricity grids. As a result, the next generation of power plants needs to operate in what is known as the load-following mode, requiring flexible adjustments in electricity production to align with the energy demand on the grid. This transition from the steady baseload operation to load-following operating conditions can significantly increase the number of times various plant components are exposed to transient stresses. This increased thermo-mechanical cycling can lead to accelerated material degradation, thereby elevating the risk of premature failure of a component. It becomes imperative to conduct a comprehensive analysis of fatigue, creep-fatigue, and stress corrosion cracking life, to assess the resilience of the various engineering components under these flexible load-following operating conditions. This study aims to develop a comprehensive numerical model of a light-water reactor pressure vessel (RPV) to investigate its degradation under various operating scenarios. This coupled thermo-mechanical finite element analysis evaluated the stress response of the RPV caused by considering fluctuations in thermal and mechanical loads caused by the varying pressure and temperature occurring during the load-following operation. Critical locations on the RPV are subsequently identified based on the stress response. The stress intensity factors for the postulated flaws at those locations are then calculated, followed by an evaluation of the reactor's life in accordance with the ASME Boiler and Pressure Vessel Code Section XI. This comprehensive life assessment covers a number of transients expected during the flexible load-following operation, providing invaluable insights into the RPV's structural integrity. Moreover, the development methodology can be adapted to other reactor components, as well as components of conventional power stations that are affected by varying operating conditions.
Materials in molten salt reactors (MSR) undergo accelerated degradation from corrosion, irradiation, and cyclic loads at elevated temperatures. Establishing a materials surveillance program to enable the assessment of material deterioration is critical to assure structural integrity of MSRs components. This presentation summarizes recent efforts towards the development of surveillance test articles for collecting various damages for monitoring materials degradation. Surveillance test articles with reduced dimensions were designed to capture creep-fatigue damage from cyclic loading at elevated temperatures. The strain evolution of test articles during thermal cycling was analyzed both numerically and experimentally. Out-of-reactor thermal cycling demonstrated successful capture of strain range for materials assessment. Moreover, test articles were subject to both mechanical loads and molten salt exposure, and the damage due to stress and corrosion was investigated. Additionally, damage inference models were developed to predict the remaining life of materials based on the accumulated damage in surveillance test articles.
High operating temperatures along with diurnal cycling and high operating stresses bring many material and engineering challenges for concentrated solar power (CSP) receivers. Current design rules in ASME Boiler and Pressure Vessel (B&PV) Code for high temperature structural components were not intended for CSP components. The paper reviews the design rules in several sections of ASME B&PV Code in terms of their applicability in designing high temperature CSP receivers. With some restrictions and modifications, the conditions assumed in the design rules in Section III, Division 5 of ASME B&PV Code are found to closely resemble the loading conditions of CSP receiver, including accounting for the interaction between creep and fatigue damage. Further, the paper applies the proposed design rules to an external tubular receiver design and discusses several lessons learned from the design study. The creep damage accumulation due to high thermal stresses resulting from circumferentially non-uniform flux distribution is found to control the design of CSP receivers. Several mitigation actions to improve the design life of receivers are also discussed.
Here, this paper introduces srlife, a tool for estimating the structural service life of concentrating solar power (CSP) receivers operating at high temperatures. Supporting both metallic and ceramic receiver designs, srlife is available as open-source software at https://github.com/applied-material-modeling/srlife and can be installed via the PyPi package manager (https://pypi.org). Given basic receiver geometry and incident heat flux, the tool performs thermohydraulic and structural analysis and estimates the life of a receiver. Designed for easy integration into a software stack, including solar field and levelized cost analysis, the tool can be utilized for optimizing receiver designs to meet service life and economic targets. This paper is Part I in a two-part series. Part I discusses the analysis process used to estimate the life of metallic receivers, along with a description of the required input data. Additionally, several heuristics applied within srlife can reduce analysis time significantly while maintaining accurate life estimations for metallic receivers when compared to full analyses. Several examples demonstrating the utility of srlife in receiver design are also discussed. Part II focuses on the life estimation of ceramic receivers, using time-dependent reliability analysis and various ceramic failure models implemented in srlife.
Abstract This study presents incremental finite element computations of creep‐fatigue crack growth in Alloy 718 at in air. Alloy 718 is representative of creep‐brittle materials, in which viscoplastic deformation is restricted near the tip of a growing crack. The computations predict crack growth using a unique combination of an irreversible cohesive zone formulation and a strain gradient viscoplastic material model based on the Kocks–Mecking formulation. Cohesive zone damage parameters are estimated using sustained loading and constant‐amplitude cyclic loading experiments. Computations of crack extension under three different waveforms containing overloads all predict post‐overload retardation. The amount of retardation depends strongly on the overload ratio, consistent with experiments in the literature using similar waveforms. Analysis of the crack‐tip fields demonstrates retardation is associated with unloading in the highly deformed material near the advancing crack tip. Dynamic recovery and geometrically necessary dislocations are shown to significantly influence post‐overload crack extension.
Operational environments in generation IV reactors involve corrosive and irradiative conditions at elevated temperatures. Typical reactor operations consist of transients which impose cyclic loads on reactor components. These cyclic loads, combined with corrosive and irradiative environments, result in synergistic degradation of component materials. However, limited data exists on the coupled damage effects on materials for reactor environments. While the surrogate material surveillance concept has been used in light water reactors to assess irradiation damage, existing material surveillance technologies are not suitable for in-situ monitoring of coupled material degradation. The materials surveillance program focuses on material degradation management and the estimation of remaining life of reactor components through surveillance test articles. This paper presents the design and analysis methodology of a bi-metal surveillance test article, which uses difference in thermal expansion coefficient between two metals to induce in-situ cyclic loads. This report presents the work conducted in FY 25, to test the surveillance test article in air and salt environments. These test specimens were evaluated after thermal cycle exposure and remaining life is measured through creep test.
Explore the source record for details and available documents.
INL has performed simplified model testing using a variable in the software control program to compare with the testing performed at ORNL, which uses hardware to modify the control signal. The INL tests compared well with the ORNL tests, showing similar stress and strain values, as well as similarly shaped hysteresis loops. This demonstrates that possibility of performing simplified model testing without any hardware modifications.
Powder metallurgy hot isostatic pressing (PM-HIP) Microreactor Project presentation for the 2023 Advanced Reactor Technology (ART) program review.