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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Codes and standards for ceramic composite core materials for High Temperature Reactor applications

Fiber-reinforced ceramic matrix composites are attractive for high-temperature nuclear applications due to excellent thermal and mechanical properties as well as reasonable-to-outstanding radiation resistance. Over the past 20 years, the use of ceramic matrix composite applications expanded to many commercial non-nuclear industries as fabrication and application of the technologies mature. The ASME Boiler Pressure Vessel Code, under Section III Division 5, provides the design and construction rules for High Temperature Reactor components. It published the first rules for ceramic matrix composites to be used for reactor core components. The rules lay out the quality requirements together with the design and materials criteria for the use and application of silicon carbide- and carbon-based matrix material technologies. As with the established graphite rules, the ceramic composite material rules are structured in Subsection HH (from Section III), that addresses the criteria for class SN nonmetallic core components. The code rules rely heavily on the development and publication of standards for composite specification, classification, and testing of mechanical, thermal, and other properties. These test methods are developed in ASTM Committee C28 on Advanced Ceramics, with a current focus on ceramic composite tubes. This article describes the detail of the composites code, the design methodology and similarities to the graphite code, the guidance for the development of specifications for ceramic composites (for nuclear applications) including recent standard developments, and it mentions the next steps to support licensing aspects by validating the code with benchmarking data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Design of an Additively Manufactured Recuperator With 800 °C Inlet Temperature for sCO2 Power Cycle Application

Abstract Recuperators with design temperatures at and above 800 °C can further increase the thermal efficiency of supercritical CO2 power cycles by enabling higher turbine exhaust temperatures. Mar-M247 is a well-suited nickel-based superalloy for high temperature service due to its high creep strength that prevents excessive material thickness being required for pressure containment. Additive manufacturing using a high-speed laser-directed energy deposition (L-DED) process presents a promising solution, with build trials demonstrating the ability to produce nonconventional flow channels for enhanced heat transfer. A design process is presented that includes aerothermal and mechanical evaluation to maximize performance within the constraints of the manufacturing process. A 2-D heat transfer network and pressure drop code allows prediction of flow distribution and its effect on overall thermal performance. Established literature correlations, along with CFD simulation, inform the prediction of heat transfer coefficients and friction factors for the flowpaths and enhancement features in the heat exchanger core. Mechanical evaluation using finite element analysis (FEA) modeling with the intent of the ASME Boiler & Pressure Vessel Code (BPVC) Section VIII, Div. 2 assesses the operational safety of the design. The detailed design features annular finned passages that take advantage of helical flow paths to distribute the flow from separated headers to shared heat transfer surfaces. Performance predictions for the recuperator at a 50 kW scale provide insights into the feasibility of the additively manufactured (AM) process to produce recuperators on a commercial scale that extend existing operating envelopes.

Engineering↗
DOE OSTI2020

Example Evaluation of a Representative Heat Pipe Test Article Design for Structural Acceptability using ASME Design Rules

This report describes a design analysis of a test article mimicking a core block of a heat-pipe mi-croreactor using the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B rules. These rules cover the design and construction of elevated temperature nu-clear reactor structural components, like the core block of a heat-pipe microreactor emulated by the test article. The purposes of the report are to: (1) provide a step-by-step example of applying the ASME rules to a reasonable core block geometry and corresponding loading conditions, (2) to provide design feedback on the test article, and (3) to evaluate the ASME rules for use in the design of complex heat-pipe microreactor components. The test article design has a relatively short ASME design life, limited by high thermal stresses. The current ASME rules are adequate for evaluating the core block component designs, though they could be optimized to increase the design efficiency and reduce the amount of designer/analyst time required to execute the design process. The report covers both the current, 2019 edition ASME rules and modified, currently unapproved, rules designed to improve the design process for components with complex geometries and high secondary stresses, like the core block test article.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Experimental and Analytical Verification of ASME Section IiII Division 5 Creep-Fatigue Design Rules

The continuous advancement of structural materials and the growing demands for more reliable and economical structural components in high-temperature reactor applications have necessitated the development of comprehensive design methodologies and design rules. Mechanical degradation of structural components at elevated temperatures subjected to cyclic deformation is controlled by the creep-fatigue damage. Over the past few decades, diligent research efforts have been dedicated to refining the development of elevated temperature design rules in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5 and to develop conservative design rules that can effectively guard against the risk of creep-fatigue failure. In ASME Section III, Division 5, for a design to pass the creep-fatigue acceptance criteria, creep damage and fatigue damage are evaluated separately, and these damages must not violate the bi-linear creep-fatigue interaction diagram, i.e., the so-called D-diagram. The creep-fatigue damage evaluation procedure assumes that the effects of the actual cyclic loading sequence can be bounded by assuming that the individual loading cycles are uniformly distributed throughout the component design life. In this study, creep-fatigue experiments with variable amplitudes and loading sequencies were designed and performed on Alloy 617 at high temperatures. The results were analyzed to evaluate the loading history effect on creep-fatigue damage accumulation and to verify the assumptions for the creep-fatigue evaluation design rules.

36 - MATERIALS SCIENCE↗
DOE OSTI2024

Fatigue and Creep-Fatigue Evaluation of Alloy 709 at 760 and 816°C

A significant research and development effort is underway to support the qualification of Alloy 709 as a Class A construction material in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5, High Temperature Reactors. This initiative includes a comprehensive Alloy 709 code qualification plan aimed at generating extensive material testing data crucial for compiling the code case data package. The data package is essential in establishing material-specific design parameters for Alloy 709 to be used as Section III, Division 5 Class A construction material for fast reactors, molten salt reactors and gas-cooled reactors. An ASME Section III, Division 5 material code case requires the evaluation of mechanical properties from a minimum of three commercial heats, covering anticipated compositional ranges. A key part of the data package involves fatigue and creep-fatigue testing at elevated temperatures, needed for developing the fatigue design curves and the damage envelope of the creep-fatigue interaction diagram (D-diagram). This paper summarizes the strain-controlled fatigue testing on three commercial heats of Alloy 709 at 760 and 816°C with strain ranges between 0.25% and 3%. The fatigue failure data are used to generate a preliminary fatigue design curve. Additionally, the creep-fatigue testing results at 816°C with tensile hold times of 10, 30, and 60 minutes are presented in support of developing the D-diagram for Alloy 709.

Wang, Yanli↗
DOE OSTI2025

Automated qualification data tool for high temperature metallic materials

This report describes a framework for storing, processing, and displaying qualification data for high temperature mechanical properties. The framework automates the process of generating design data from mechanical test results, for example for a data qualification report for the ASME Boiler \& Pressure Vessel Code. The framework has three parts: a data storage model with common formats for several types of typical mechanical property tests, a backend based on the \pycreep Python library for correlating and extrapolating the data to generate design material properties and allowable stresses, and a demonstration user interface for displaying, sorting, and filtering the data and exploring different options for modeling the design mechanical properties. The report discusses the options available for data processing, with illustrations from real test data on Alloy 617, Alloy 709, Alloy 740H, and Laser-Powder Bed Fusion 316H. The framework is complete for ASME type data analysis and will be used to store test data generated by the Department of Energy, Office of Nuclear Energy, Advanced Materials and Manufacturing Technologies sponsored qualification programs. Future work could extend the tool to other types of material properties and/or expand the demo user interface to make it accessible across the AMMT program.

36 MATERIALS SCIENCE↗
DOE OSTI2023

High temperature inelastic constitutive models for the ASME Section III, Division 5 Class A materials

This report describes the exploration of a universal high temperature inelastic constitutive model for use with the ASME Boiler & Pressure Vessel Code Section III, Division 5, Class A design rules. The idea, developed based on feedback from reactor vendors, is to simplify the current bespoke material models for Grade 91, 316H, and Alloy 617 and the new model for Alloy 800H into a simple, single model form. The report describes a new parallel time integration technique implemented in the pyoptmat package which allowed us to explore a wide variety of model forms, searching for a suitable common model. The report then describes preliminary work on models for monontonic deformation and then the development of a set of models suitable for capturing high temperature cyclic deformation. These models are suitable for use with the ASME design rules, except potentially for a narrow, material-specific band of temperatures at the edge of the creep regime. An incremental improvement the current models could overcome this discrepancy and provide a new, simpler set of constitutive model for all four materials.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Updated ASME design correlations and qualification plan for powder bed fusion 316H stainless steel

This report provides an update on the Advanced Materials and Manufacturing Technologies (AMMT) program effort to qualify Laser-Powder Bed Fusion (L-PBF) 316H stainless steel for use with the ASME Boiler & Pressure Vessel Code Section III, Division 5 rules. The report summarizes progress in testing and characterizing L-PBF material at elevated temperatures by providing preliminary design data for L-PBF 316H and by comparing the elevated temperature performance of the L-PBF material to wrought and conventional fusion welded 316H. The report then updates the initial AMMT qualification plan for L-PBF 316H, originally developed in 2023, to update the accelerated qualification strategy adopted in that plan to account for the new high temperature test data. The report also explores a few methods for further accelerating the qualification process using machine learning techniques to supplement the more conventional, empirical analysis methods typically used by ASME to correlate and extrapolate time-dependent material test data.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Initial Development of an Improved Creep-Fatigue Design Method that Avoids the Separate Evaluation of Creep and Fatigue Damage and Eliminates the Requirement for Stress Classification

This report describes a method for capturing the effect of strain range, temperature, hold time, and elastic follow up on the cyclic life of a material undergoing uniaxial load. Developing design correlations of this type requires methods for extrapolating experimental data as a function of hold time and follow up factor. Various extrapolation methods are developed and discussed here and recommendations are developed for the final approaches to be adopted for design use. The report also describes comparisons between the new creep-fatigue design method, described here, and current creep-fatigue design methods in the ASME Boiler and Pressure Vessel Code and associated nuclear Code Cases. These comparisons demonstrate several advantages of the new method: it is easier to execute and is less over conservative than many current approaches. This report, along with a companion on extending these results to multiaxial load and on combining load cycles, represents the accumulation of several years of work on the development of this new approach to creep-fatigue design, combining an elastic perfectly plastic analysis method with simplified model tests measuring the effect of elastic follow up on creep-fatigue cyclic life. The conclusions to this report describe a path forward to developing the required design data for high temperature advanced reactor structural materials and guiding the adoption of the new method in the ASME Boiler and Pressure Vessel Code and related standards.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Low Cost Hydrogen Storage at 875 Bar Using Steel Liners and Steel Wire Wraps

The Department of Energy’s (DOE) Fuel Cell Technologies Office (FCTO) within the Office of Energy Efficiency and Renewable Energy (EERE) supports research and development activities leading to the development of low-cost, high-pressure hydrogen storage systems for use in hydrogen refueling stations (forecourt). The specific goals of this project were to develop and produce a prototype of a pressure vessel to safely store gaseous hydrogen at 875 bar that meets the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section VIII, Division 3 design code for service in hydrogen (Article KD-10) and meets the DOE storage tank cost target of <$1000/kg stored hydrogen (H 2 ). The pressure vessel designed and prototyped as part of this study utilizes a standard ASME approved, 406 mm diameter, 31.7 mm thick wall, and 9.14m long, liner with a capacity of 765 liters and rated for a pressure of up to 460 bar. The liner was wrapped with high strength steel wires to boost its pressure rating to 875 bars. The process of wrapping metal liners with a high strength steel wire, a WireTough patented technology, was first demonstrated on short cylinders that were 1.9 m long but of the same diameter and wall thickness as the 9.1 m long, 765 L cylinders. Subsequently, full length cylinders that were 9.14 m long were designed and certified under the applicable ASME Design Codes. Two prototype cylinders were wire wrapped to demonstrate the manufacturability. Following is a bulleted summary of the accomplishments of the project that ended on August 31, 2018.

25 ENERGY STORAGE↗
DOE OSTI2023

Factors Influencing Propensity for Stress Relaxation Cracking in Inconel® Alloy 740H® and Practical Guidance for Applications

Inconel® alloy 740H® (UNS N07740) was the first age-hardenable nickel-based alloy approved by the ASME Boiler & Pressure Vessel Code for use in pressure-boundary applications. In recent years, advanced energy systems such as supercritical CO2 power cycles have utilized alloy 740H in large demonstration projects driven by the requirement for higher fluid temperatures and pressures. Stress relaxation cracking (SRxC) following post-weld heat-treatment (PWHT), also known as strain age cracking (SAC), has been identified in a limited number of weldments during these industrial builds resulting in focused research to further clarify factors influencing this cracking tendency. This paper will summarize some of the findings from shop and field fabrication leading to successful welds and characteristics of observed SRxC. Laboratory experiments supported by microstructural characterization will be presented to highlight the importance of variables such as strain, material starting condition, and PWHT temperatures. Finally, the results will be summarized within the context of practical guidance for industry to successfully weld the material in boiler, heat exchanger, and piping applications.

14 SOLAR ENERGY↗
DOE OSTI2021

A Minimum Creep Rate for 2-1/4Cr-1Mo Steel Consistent With the ASME Section III, Division 5 Rules

This technical note describes a minimum creep rate model for 2-1/4Cr-1Mo steel that is consistent with the current creep strain equation embedded in the ASME Boiler & Pressure Vessel Code Section III, Division 5, Subsection HB, Subpart B isochronous stress–strain curves. Minimum creep rate models for all the Section III, Division 5 Class A materials are required for the development of improved high temperature design methods. Of all the Class A materials, only 2-1/4Cr-1Mo does not have a readily identifying minimum creep rate term in the current isochronous stress–strain curve model.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Design Guidance for High Temperature Concentrating Solar Power Components

This report provides guidance for the design of components for concentrating power facilities operating at high temperatures and undergoing high, secondary thermal stresses relative to the applied primary pressure stress. The design rules were developed for the design of Generation 3 CSP tubular receivers manufactured from Alloy 740H, but are generally applicable to a wide range of component types undergoing similar loads. Part 1 of the report provides procedural design rules for components to be used in conjunction with the 2019 edition of the ASME Boiler & Pressure Vessel Code. Part 2 provides corresponding design data for Alloy 740H. Part 3 of the report is a commentary describing the rationale behind the design rules and the data underlying the design material properties. Finally, Part 4 provides an extensive set of worked sample problems detailing the application of the rules to CSP components.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2022

Innovative Method for Welding in Generation 3 CSP to Enable Reliable Manufacturing of Solar Receivers to withstand Daily Cycling at Temperatures Above 700°C (Final Technical Report)

Inconel® Alloy 740H® (alloy 740H) was the first age-hardenable nickel-based alloy approved by the ASME Boiler & Pressure Vessel Code for use in pressure-boundary applications. Over the past ~20 years the alloy has been optimized for weldability and high-temperature stability, approved for use in different applications. Development of a supply chain combined with the advantageous properties of the alloy (high-temperature creep strength, oxidation and corrosion resistance, etc.) have resulted in the alloy being applied to new high-temperature power cycle demonstration projects, and of particular interest are applications to concentrating solar power (CSP) to enable higher-efficiency Generation 3 CSP systems and the corresponding supercritical CO 2 (sCO 2 ) power cycle components (heat exchangers, piping, etc.). The high allowable stresses of alloy 740H also make it a desirable material for current Generation 2 CSP solar power receivers to improve cyclic capability and/or reduce receiver height. Recent experiences in demonstration projects utilizing alloy 740H identified cracking issues during welding and fabrication. In this project, a detailed study was done to confirm and clarify the Stress Relaxation Cracking (SRxC) mechanism, also known as stress relief cracking or strain-age cracking (SAC), during post-weld heat-treatment (PWHT). This involved detailed microscopy and advanced characterization to understand the root cause(s) of three failures obtained from industry. Based in-part on these findings, a targeted laboratory based SRxC test method was utilized to evaluate variables such as heat-to-heat variations, strain level, PWHT temperature, and starting material condition on three heats of alloy 740H. Industrial shop welding of cold-worked plates was also conducted. The research showed the following: SRxC was confirmed as the cracking mechanism for all field failures; Stress state (from residual stresses, constraint, deformation, and local stress concentrations) was playing a significant role in field failures and laboratory testing confirmed increasing susceptibility for all heats with increasing strain levels. High levels of microstructural strain were identified at crack initiation locations, in some cases leading to local recrystallization; Precipitate free zones (PFZs) at grain boundaries were found at relaxation cracks and crack initiation locations uniquely associated with SRxC in alloy 740H. Laboratory testing reproduced this microstructural feature which had only previously been reported in long-term creep testing of weldments. Advanced nano-scale characterization confirmed the presence of a moving boundary leading to coarsening of precipitates and PFZs where damage accumulated; The research suggested heat-to-heat variations due to local chemistry and processing may influence SRxC susceptibility, but more work is needed to fully clarify these effects. To disseminate the key learnings from this research to the scientific and engineering communities and alloy 740H end users, multiple technical publications and presentations were made, an industrial alloy 740H users meeting was held, and a new industry guideline specification document which can be directly implemented by end-users of alloy 740H was produced.

14 SOLAR ENERGY↗
DOE OSTI2022

GIF-VHTR Materials (Graphite) Project Plan: Progress since September 2021

¦1: Data, Design Methodology and Construction 1.1: Graphite Selection and Acquisition Strategy (Original tasks completed) 1.2: Graphite Properties (Virgin Materials) 1.3: Graphite Fracture Behaviour(Original tasks completed) 1.4: Graphite Oxidation Behaviour(Original tasks completed) 1.5: Graphite Component Testing 1.6: Graphite Irradiation Effects 1.7: Graphite Irradiation Induced Creep 1.8: Graphite Codes and Standards Development (ASME & ASTM) 1.9: Graphite BehaviourModels Development 1.10: Links to Existing Graphite Irradiation BehaviourDatabases 1.11: Materials Database ¦2: Operation and Inspection 2.1: Methodologies for determining requirements for inspection of graphite cores 2.2: Methodologies for ISI of graphite cores ¦3: Decommissioning and Disposal (New task added 2021) 3.1: Technical data for decommissioning and disposal issues

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Application of the Continuum Damage Mechanics Wilshire-Cano-Stewart (WCS) Model

In this study, the applications of the continuum damage mechanics-based Wilshire-Cano-Stewart (WCS) model are explored to predict rupture time, minimum-creep-strain-rate (MCSR), damage, damage evolution, and creep deformation. Increase knowledge in manufacturing methods has pushed the limit of material science and the development of new materials. Conventional testing is required to qualify materials against creep which according to the ASME B&PV III code, 10,000+ hours of experiments are necessary for each heat before materials are put into service. This process is costly and not feasible for new materials. As an alternative, models have been employed to predict creep behaviors and reduce the amount of time necessary for material qualification. Many models have been developed to predict distinct creep behaviors and the question of which model is best remains. Amongst current models, the WCS model has emerge with the ability to predict multiple behaviors using an explicit analytical approach with the ability to predict long-term creep. In this study the novel continuum damage mechanics WCS is employed in multiple applications. The goals of the study are (a) to discuss and determine the framework of the WCS model and validated it mathematically and using parametric simulations, (b) applied the model to accelerated creep data to show the capabilities of the model with non-conventional data, and (c) applied a novel numerical method, the datum temperature method (DTM) to show the model extrapolations and interpolations capabilities with limited and reduce data sets. To accomplish these goals, data is gathered for alloy P91 and Inconel 718 to develop and post-audit validate the model. The benefits of using the WCS model is that it provides an explicit stress and temperature dependency ideal for extrapolations, the coupled equations are suitable for finite element analysis (FEA) implementation, and it follows an explicit calibration approach. The model also proves that it can be applied to accelerated testing data and using the DTM.

Cano, Jaime A↗
DOE OSTI2022

Results of Initial Alloy 617 High Temperature Crack Growth Testing

Crack propagation data can provide valuable insights when performing a safety evaluation for a component. Alloy 617 is qualified in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear service up to 950°C. Little is known, however, about subcritical crack-growth phenomena in Alloy 617. Previous crack-growth studies of Alloy 617 did not investigate temperatures across the range which is currently qualified in Section III, Division 5. High-temperature crack-growth testing in air and in reactor-grade helium can provide data for establishing the crack-growth correlations in support of an ASME BPVC Section XI high-temperature flaw evaluation Code Case. Idaho National Laboratory (INL) has previously performed crack-growth testing, but the equipment requires revitalization. This report provides the status of crack-growth testing in Alloy 617 at INL.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Initial Assessment of Erosion/Abrasion Issues Related to Gas-Cooled Reactors

The US Department of Energy’s Oak Ridge National Laboratory (ORNL) is investigating the graphite erosion concern detailed in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), Section III, Division 5, “High Temperature Reactors.” The special consideration in Subsection HH, Subpart A (HHA)-3143, “Abrasion and Erosion,” was described in ASME BPVC Section III, Division 5, which was accepted with exception/limitation in a US Nuclear Regulatory Commission (NRC) Technical Review. This report describes ORNL’s FY 2023 comprehensive research of erosion-related reports and publications to further study the proposed gas flow velocity limitation to the erosion effect in graphite structure in the ASME BPVC. This report also proposes that a nondimensional fluid parameter, the Reynolds number, might be a better alternative than gas flow velocity for comparing erosion effects caused by carbonaceous dust in fast-flowing helium coolant in gas-cooled reactors (GCRs).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗