Biaxial Thermal Creep of Alloy 617 and Alloy 230 for VHTR Applications

In this study, we employed pressurized creep tubes to investigate the biaxial thermal creep behavior of Inconel 617 (Alloy 617) and Haynes 230 (Alloy 230). Both alloys have been considered to be the primary candidate structural materials for very high temperature reactors (VHTRs) due to their exceptional high-temperature mechanical properties. The current creep experiments were conducted at 900°C for the effective stress range of 15–35 MPa. For both alloys, complete creep strain development with primary, secondary, and tertiary regimes were observed in all studied conditions. The tertiary creep was found to be dominant in the entire creep lives of both alloys. With increasing applied creep stress, the fraction of the secondary creep regime decreases. The nucleation, diffusion, and coarsening of creep voids and carbides on grain boundaries was found to be the main reason for the limited secondary regime, and was also found to be the major cause of creep fracture. The creep curves computed using the adjusted creep equation of the form ε = Aσ cosh−1(1 + rt) + Pσntm agree well with the experimental results for both alloys at the temperatures of 850–950°C. Paper published with permission.

Development of Welding P92 Pipes Using the Reduced Pressure Electron Beam Welding Process for a Study of Creep Performance

Boiler and steam piping components in power plants are fabricated using creep strength enhanced ferritic (CSEF) steels, which often operate at temperatures above 550°C. Modification of alloy content within these steels has produced better creep performance and higher operating temperatures, which increases the process efficiency of power plants. The improved materials, however, are susceptible to type IV cracking at the welded regions. A better understanding of type IV cracking in these materials is required and is the basis of the Technology Strategy Board (TSB) UK funded VALID (Verified Approaches to Life Management & Improved Design of High Temperature Steels for Advanced Steam Plants) project. In order to study the relationship between creep performance and heat input during welding, several welds with varying amounts of heat input and resultant HAZ widths were produced using the electron beam welding process. The welding parameters were developed with the aid of weld process modeling using the finite element (FE) method, in which the welding parameters were optimized to produce low, medium and high heat input welds. In this paper, the modeling approach and the development of electron beam welds in ASTM A387 grade P92 pipe material are presented. Creep specimens were extracted from the welded pipes and testing is ongoing. The authors acknowledge the VALID project partners, contributors and funding body: Air Liquide, Metrode, Polysoude, E.ON New Build & Technology Ltd, UKE.ON, Doosan, Centrica Energy, SSE, Tenaris, TU Chemnitz, The University of Nottingham, The Open University and UK TSB. Paper published with permission.

Microstructure Stability of Alloy 740H and its Effect on Material Properties

Alloy 740H was developed specifically for boiler tube and steam transfer pipe for the Advanced-Ultrasupercritical steam boiler application. The composition was formulated to provide creep strength, weldability and oxidation and coal ash corrosion resistance. It is the first nickel-base age-hardened alloy to be approved under Section 1 of the ASME Boiler and Pressure Vessel Code. Although a significant amount of microstructure and creep data was submitted to support the code case, it was largely collected on tube and light plate, much of this from pilot-scale material. The data package did not cover other product forms, sizes, hot working methods and weld joint configurations that may be required to construct a power plant. These include large diameter pipe, bends, fittings, heavy section welds, dissimilar metal welds, etc. that have complex thermal cycles and composition gradients. An ongoing effort is underway to characterize the microstructure and properties of alloy 740H under a range of manufacturing processes and to project these results through the potential life cycle of a plant. This paper describes the current status of this work. Paper published with permission.

A Combined Experimental-Computational Approach to Design Optimization of High Temperature Alloys

Experimental data were used to develop metamodels to predict high temperature alloy chemistry trends influencing stress-to-rupture and time-to-rupture of Nickel based superalloys. Chemistry optimization utilized evolutionary neural networks, bi-objective genetic programming and pruning algorithm. Optimization results were compared with the experimental data and IOSO optimization algorithm. Response surfaces were developed through various modules available in a commercial optimization package. Pareto optimized chemistries were tested using thermodynamic database, FactSage™, by studying the phase distribution as a function of temperature of manufacture and exposure. Uniformity in the amount of critical phases over 0–1200 °C range confirmed high temperature stability for optimized alloys. Paper published with permission.

Hastelloy® N for Molten Salt Reactors Used for Power Generation

Hastelloy® N alloy was developed in the 1950’s as ‘INOR 8’ by the Oak Ridge National Laboratory to resist molten salts used as a fuel and coolant in the early development of molten salt nuclear reactors for propulsion and power generation. China has recently expressed interest in Hastelloy N for use in prototype and demonstration components for a high-temperature, uranium-fueled, molten-salt cooled reactor for the production of electricity. An ASME Section III NH Code Case will be necessary to move Alloy N forward commercially. This paper discusses the guidelines for design data requirements necessary to satisfy the Boiler Code for elevated temperature nuclear applications where creep effects are significant. The historic tensile and creep properties data for Alloy N (N10003) were collected and re-analyzed in accordance with current ASME procedures. The collected data will be uploaded into the ASME Materials Properties Database to support the NH Code Case development. Paper published with permission.

Structural Materials and Code Development for Japanese Sodium-Cooled Fast Reactors

This paper describes the latest status on the development of elevated temperature materials and structural codes for Japanese sodium-cooled fast reactors (SFRs). Based on the extensive research and development activities in the last decades in Japan, two materials, 316FR and Modified 9Cr-1Mo steels were recently incorporated into the 2012 Edition of Fast Reactor Design and Construction Code of the Japan Society of Mechanical Engineers (JSME). Structural design methodologies are continuously being improved towards the next major revision planed in 2016 Edition where methodologies for a 60-year design of Japanese demonstration fast reactor will be provided. Codes and guidelines for fitness-for-service, leak-before-break evaluation and reliability assessment are concurrently being developed utilizing the System Based Code concept aiming at establishing an integrated code system that encompasses a life cycle of SFRs. Paper published with permission.

Long-Term Creep-Rupture Behavior of Inconel® 740 and Haynes® 282

Creep testing and microstructural analysis were used to assess the properties and time-dependent deformation behavior of precipitation-strengthened nickel-based alloys, specifically, Inconel® alloy 740 and Haynes® 282® alloy, for use as pressure components in boilers operating under advanced steam conditions (750°C, >35 MPa). In support of the need for extended service of steam boiler tubing, piping and vessels, the ability of simple Larson-Miller estimates and a modified power-law model (Wilshire et al.) to predict creep lifetimes on the order of 105 hours based on experimental data out to almost 50,000 h was evaluated for these alloys. Even under conservative conditions, both Inconel alloy 740 and Haynes 282 project to have creep lifetimes exceeding 100,000 hours at 750°C and 100 MPa. Paper published with permission.

High Temperature Testing and Analysis of a Class 300 Bolted Flange Joint

When class 300 flange bolted joints are held at temperatures in the material creep range, it is documented that the bolt loads can relax. Tests and analysis are being performed with the goal of developing a validated FEA simulation approach to predicting the impact of creep on the bolt loads. The bolt load and gasket geometry are evaluated upon bolt up and after being heated to 1100 deg F. Tests are performed with and without a gasket to separate the impact of the gasket relaxation and flange material creep. The results of the tests and analysis approaches will be presented. Paper published with permission.

Microstructural Stability of Alloy 617 Mod. During Thermal Aging

The microstructural stability of Alloy 617 mod. during aging for up to 3000 h at temperatures of 700°C and 750°C was investigated. The precipitates of the aged alloy included M23C6 carbides located both inside grains (intragranular) and at grain boundaries (intergranular) and γ′ phase dispersed within grains. During aging, the intergranular precipitates showed a good stability. Intragranular particles increased substantially after the aging for 3000h at 750°C. Inter and intragranular carbide particles resulted in the precipitation hardening of the aged alloys. The precipitation of γ′ phase particles during aging at 700°C and 750°C is also an important factor for an enhanced hardness and an obvious decrease of the impact absorbed energy. Additionally, the intergranular cracks apparently lead to a decrease in the impact absorbed energy for the aged alloys due to carbide particles at grain boundaries. The impact absorbed energies of the aged alloys were fairly stable within the dwelling time from 300 h to 3000 h and were in the range of 63∼65J and 75∼83J for the 700°C and 750°C aging, respectively. Paper published with permission.

Simplified Methods for Elevated Temperature Structural Design: An Overview of Some Current Activities

Elevated temperature design criteria for Class 1 nuclear components employ two fundamental approaches for evaluation of structural integrity in the temperature regime where creep effects are significant: full inelastic analysis to predict the actual stress and strain resulting from time dependent loading conditions and simplified methods which bound the actual response with, conceptually, simpler material models and analytical procedures. However, the current simplified methods have been found to be more complex for real component design applications than originally envisioned. There is an added complication that the current simplified methods are considered inappropriate in the very high temperature regime where there is no distinction between plasticity and creep. Recently, some improved, less complex methods have been proposed which would overcome these objections. One set of criteria are based on elastic-perfectly plastic (E-PP) analysis methods. Draft code cases have been prepared which address the use of the E-PP methodology to primary loading, strain limits and creep-fatigue damage evaluation. Another proposed criterion is based on the use of test specimens which include the effects of stress and strain redistribution due to plasticity and creep to develop creep-fatigue damage evaluation design curves. An overview of the key features, associated analytical and experimental verification, status and path forward are presented. Although targeted to nuclear components, these criteria also have potential application to non-nuclear components and operating temperatures below the creep regime. Paper published with permission.