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DOE OSTI2025

Fitness For Service Assessment of a Corroded Heat Exchanger

Within the Fermi National Accelerator complex, there exist various water systems that support accelerator operations. One of these systems is extremely vital to the operation of the machine; that is the cooling system. The cooling system consists of nine relatively large heat exchangers that take untreated pond water and use it to cool the process fluid that further cools machine components. Over the 30 years these heat exchangers have been in operation, they have undergone significant material loss on the channels. This material loss, due to various forms of corrosion such as galvanic and microbiologically influenced corrosion (MIC) and possibly others, has deteriorated more than 80% of the nominal wall thickness of some of the exchangers and placed them in a questionable state. ASME FFS-1 (API 579) has been applied to address the condition of the heat exchangers due to their noncompliance with the governing code, BPVC Sec. VIII Div. 1. The assessments encompassed ASME FFS-1 parts 4: General Metal Loss and 9: Crack Like Flaw using level 1, 2, and 3 analysis techniques based on inspection data obtained by API 510 inspections. Level 1 and 2 assessments were deemed unfit for the corroded regions due to their location relative to a major structural discontinuity (channel to tube-sheet joint), so a level 3 analysis was conducted according to ASME Sec. VIII Div. 2 (design by analysis) rules for pressure vessels. Supplemental information included pond water tests to determine an accurate future corrosion allowance due to lacking inspection history. A leak before break (LBB) route was chosen to evaluate the possibility of leaking prior to the onset of failure. The analysis of one heat exchanger shows that the possibility the channel will develop a pinhole leak over 2.5 more years of operation should not be overlooked, but burst was unlikely from operation. The use of fracture mechanics show, that if a through-wall crack were to develop, it would not propagate further than the channel geometry and cause a leak not greater than 35 GPM. Using ASME Section XI Code Case N-705-1, allowing us to operate with a leak until the next outage given certain operating conditions and developing a leak mitigation procedure, this heat exchanger is deemed fit-for-service.

Humenik, Alex [Fermilab]↗

DOE OSTI2019

Development of the Technical Basis of a Unified Viscoplastic Model of 316H Stainless Steel for Incorporation into ASME Division 5

This report describes the development of a unified viscoplastic model describing the deformation of 316 stainless steel. The model was developed for incorporation into Section III, Division 5 of the ASME Boiler and Pressure Vessel Code covering the design and fabrication of Class A high temperature nuclear reactor components. The current version of the Code does not provide reference inelastic models, leaving the responsibility of defining an adequate model up to either the plant owner or designer. This approach represents a substantial barrier to using the ASME design by inelastic analysis methods. As historical practice demonstrates these methods produce much more efficient designs, compared to the design by elastic analysis methods, eliminating this barrier could result in safer, more effective, and more efficient structural designs thereby lowering the cost of future advanced reactors. This report summarizes the process of gathering experimental data, calibrating the candidate inelastic model against that data, and validating the model. The report also summarizes work on a new method of calculating the sensitivities of history dependent nonlinear constitutive models that was used to calibrate the 316H model using gradient-based optimization.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Assessment of UTP A 2133 Mn as a Matching Filler Metal for Alloy 800H in Section III, Division 5 Applications

Six alloys are qualified in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear components. Due to reactor design alloy restrictions and motivation to improve operational efficiency and plant lifetimes, it is necessary to maximize the usefulness of qualified alloys. One of these alloys is Alloy 800H, which is qualified for a maximum temperature of 760°C and a maximum 300,000-hour service life. However, the welding filler metals qualified to join Alloy 800H have stress rupture factors that reduce the allowable weldment strength by factors as low as 0.59 times the Alloy 800H base metal. In an effort to improve the weldment creep-rupture performance, or to increase the stress reduction factor, non-code qualified filler metals are under investigation by the Department of Energy?s Advanced Reactor Technologies program. For this investigation, UTP A 2133 Mn filler metal (Fe-Cr-Ni-Mn-Nb) was used to join 0.5 in. thick Alloy 800H plate to demonstrate passing an ASME BPVC Section IX weld qualification. The multi-pass, pulsed GTAW process showed successful weld qualification results, and additional property measurements were conducted to compare previous analyses of Alloy 82 and Alloy 617 filler metals. Future work will evaluate the cross-weld creep-rupture performance of the UTP A 2133 Mn in comparison to Alloy 82 and Alloy 617 filler metals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Preliminary description of a new creep-fatigue design method that reduces over conservatism and simplifies the high temperature design process

The report provides the initial description of a new creep-fatigue design method for structural components in high temperature nuclear service. The new method is based on an integrated elastic-perfectly plastic (EPP) analysis and Simplified Model Test (SMT) approach that reduces over conservatism, improves the treatment of elastic follow up, and simplifies the design procedure, when compared with the current creep-fatigue design methods in ASME Boiler and Pressure Vessel Code. Developing the design charts for the EPP-SMT design method requires extrapolating SMT test data as a function of hold time and follow up factor. The report develops the preliminary design charts for Alloy 617 at temperatures between 800°C and 950°C by combining two extrapolation approaches developed in a previous work. The report also presents a comparative analysis between the EPP-SMT design method and the current ASME creep-fatigue design methods by evaluating design life of two sample geometries under different loading conditions. Results from the comparative analysis verify the EPP-SMT design charts but suggest the requirement of additional test data in the low strain range regime for improving the extrapolation procedure that will further reduce the over conservatism in the creep-fatigue damage evaluation. The report also concludes that the EPP-SMT design procedure can account for effect of primary load on creep-fatigue life by using a fixed, bounding value of follow up in constructing the design charts. The conclusions to this report describe the future work required to complete this new design method so it can be codified through a nuclear Code Case.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

Preliminary design analysis workflow for Division 5 HHA-3200 requirements for graphite core components

This report presents a design analysis workflow for graphite core components and assemblies, based on the design rules of ASME Boiler Pressure and Vessel Code, Section III, Division 5, Article HHA-3000. The workflow contains three stages: developing the design of the graphite core component, modeling the component with the finite element software MOOSE, and assessing if the component passes/fails the criteria of the HHA-3000 design rules. Since the design rules use probabilistic metrics specifically established to evaluate brittle materials, we developed a python library that performs all the statistical calculations necessary for the evaluations of the HHA-3000 criteria.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

Mechanical Properties of Aged A709

A709 plate is in the process of being developed and qualified through a collaboration by Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL). The goal is to qualify A709 plate in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). This would permit the construction of A709 plate for fast-reactor structural applications. A tradeoff between creep and creep-fatigue properties was established in prior work. A processing route was identified as providing the optimal balance between creep-rupture and elevated-temperature cyclic properties for the solution annealed condition. This condition, however, resulted in a shorter than desired creep-rupture life. Aging is probed to evaluate its potential at further improving mechanical properties. Aging statically precipitates the solutes in solution prior to service. The processing routes identified as providing the best creep-rupture properties and best balance in mechanical properties are investigated. This report discloses the results from tensile, creep, and elevated-temperature cyclic testing performed at INL on the aged material. The aged properties are compared to solution annealed properties. A709 offers improvement in performance compared to 316H particularly at high temperatures; 316H is a material qualified in Section III, Division 5 of the ASME BPVC. For certain high-temperature fast-reactor design conditions, 316H cannot be used while A709 can.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Failure Analysis for Molten Salt Thermal Energy Storage Tanks for In-Service CSP Plants

Thermal Energy Storage (TES) is a fundamental component in concentrating solar power (CSP) plants to increase the plant's dispatchability, capacity factor, while reducing the levelized cost of electricity. In central receivers CSP plants, nitrate molten salts have been used for several years for operation temperatures of up to 565 degrees C. Despite many efforts to advance nitrate salt to higher operation temperatures (even considering a replacement with molten chloride salts) to achieve higher energy conversion efficiencies, the 565 degrees C temperature is currently considered the state-of-the art. Although molten salt tanks have been broadly deployed in commercial CSP plants worldwide, several failures have been reported in these tanks after a few months or years of operation, causing significant economic loss and mistrust in CSP technologies. Most of these failures are associated with the infancy of the technology and multiple issues related to tank design, fabrication, commissioning, and aggressive operation. A technical standard dedicated to the design and fabrication of molten nitrate TES tanks does not exist today. Current in-service molten salt tanks have been generally designed based on the American Petroleum Institute's (API) 650 and ASME Section II standards. The API 650 code provides guidelines for dimensions and fabrication for oil storage tanks up to 260 degrees C. The ASME standard provides allowable stress values for various materials at a range of temperatures and conditions. Both standards seem to be limited for molten salt TES tanks where high temperatures, thermal cycling, and transient conditions are expected. In 2020, NREL released the Concentrating Solar Power Best Practices Study (NREL/TP-5500-75763) that summarized multiple issues in CSP plants, along with potential alternatives and recommendations to address those issues based on information collected from participants representing about 80% of operating CSP plants in the world. One of the recommendations from this study was the development of accurate and validated models to evaluate the plant's transient operation, capable of capturing the effect of short-term clouds and operator response, while being flexible in being adapted to various spatial and temporal resource data. The "Failure Analysis for Molten Salt Thermal Energy Tanks for In-Service CSP Plants" project was inspired on this recommendation and was focused on (1) the development and validation of a physics-based model for a representative, commercial-scale molten salt tank, (2) performing simulations to evaluate the behavior of the tank as a function of typical plant operation conditions, (3) understanding tank failures mechanisms, (4) determining the residual stress and distortion in the tank floor after welding fabrication and evaluating their impact in the stresses developed in the tank during operation, (5) assessing the impact of key operation parameters on the temperature and stress distribution, (6) conduct a preliminary evaluation of design features to reduce stress and improve tank's reliability, and (7) estimate the tank's service life based on the stress developed under diverse operation scenarios. From the analysis conducted in the project and presented in this report, it was found that maximum stresses surpassing the yield strength point of the stainless steel (SS) 347H are developed on the tank floor near the perimeter. These large stresses are strongly influenced by the initial residual stresses and distortion of the tank floor after welding fabrication. During operation, large stresses are developed in the tank floor at high operation temperatures with large salt inventory levels during transient operation. High stresses are also related to elevated temperature gradients in the tank floor that could be attributed to insufficient mixing within the salt inflow and the salt inventory. Based on the analysis, creep is the predominant failure mechanism. However, the large stress levels could favor the plastic deformation into buckles, and crack formation due to stress relaxation cracking during cycle operation. A lifetime below 3 years was estimated for the typical plant operation conditions and a specific initial residual stress and deformation distribution of the tank floor. The estimated life agrees with the service time to failure reported in several commercial molten salt tanks. Desing and operation guidelines can be extracted from the analysis presented in this report, which could be adopted by tank manufacturers and CSP operators to advance toward an ultimate solution for tank failures by reducing residual and operational stresses to achieve a tank service life of more than 30 years. Addressing failures in molten salt TES tanks is fundamental for the CSP industry's survivability, but it is also important for other industrial and power generation applications using this technology, including nuclear and concentrating solar thermal.

14 SOLAR ENERGY↗

DOE OSTI2019

Evaluating and modeling rate sensitivity in advanced reactor structural materials: 316H, Gr. 91, and A617

Advanced reactor systems operate at high temperatures to achieve better thermal efficiency. During transients from lower temperatures to high operating temperatures structural metals transition from rate insensitive behavior to rate dependent behavior. In the rate dependent regime creep and plastic deformation are not separable hence requiring a unified treatment with a viscoplastic model. Several bounding design methods in the ASME Boiler and Pressure Vessel Code, Section III, Division 5, Subsection HB, Subpart B assume a decoupled treatment of creep and plasticity so establishing the boundary of the rate insensitive regimes for advanced reactor structural materials is necessary to set limits for the use of these bounding design methods. In the rate sensitive regime new bounding methods or fully inelastic analysis are required for the design of Division 5 Class A components. During temperature transients these components may operate in both the rate insensitive and rate sensitive regions, requiring computational models that can represent both types of behavior. This work establishes criteria for material rate sensitivity and develops a new composite method for modeling constitutive response in both the rate sensitive and rate insensitive regimes for three Division 5, Class A materials: 316H austenitic stainless steel, Grade 91 ferritic-martensitic steel, and Alloy 617, a nickel-based alloy. Both the criteria and new modeling strategy apply the concept of a normalized activation energy for thermally activated mechanisms developed by Kocks, Mecking, and co-workers. Specifically, this work quantifies the average rate sensitivity of the three materials from an experimental database collected from the literature. Additionally, this work develops rate sensitivity criteria and a complete model for each material that can transition between both deformation regimes and provides several example simulations demonstrating the accuracy of the models and the numerical performance of the composite modeling method.

36 MATERIALS SCIENCE↗

DOE OSTI2021

A method for predicting failure statistics for steady state elevated temperature structural components

This paper presents the initial development of a high temperature life prediction method that accounts for the variability in the material properties of Grade 91 steel. The method accounts for material variability by fitting a variable 3-parameter Weibull distribution to experimental rupture data and accounts for the variability of creep deformation on the steady-state stresses via a Monte Carlo approach. To ensure reasonable computational times, the model represents the material as an extremely viscous Stokes fluid with a non-Newtonian viscosity, therefore solving the stress relaxation problem with a steady, static, instead of transient, analysis. Furthermore, the complete statistical analysis combines this model for creep deformation with a probabilistic model for creep rupture to evaluate the probability of premature failure for a set of sample problems, comparing the predicted failure statistics to the design life predicted by the ASME Boiler and Pressure Vessel Code rules.

42 ENGINEERING↗

DOE OSTI2022

Machine learning for ultrasonic nondestructive examination of welding defects: A systematic review

Recent years have seen a substantial increase in the application of machine learning (ML) for automated analysis of nondestructive examination (NDE) data. One of the applications of interest is the use of ML for the analysis of data from in-service inspection of welds in nuclear power and other industries. These types of inspections are performed in accordance with criteria described in the ASME Boiler and Pressure Vessel Code and require the use of reliable NDE techniques. The rapid growth in ML methods and the diversity of possible approaches indicate a need to assess the current capabilities of ML and automated data analysis for NDE and identify any gaps or shortcomings in current ML technologies as applied to the automated analysis of NDE data. In particular, there is a need to determine the impact of ML on the NDE reliability. This paper discusses the findings from a literature survey on the current state of ML for the automated analysis of data from ultrasonic NDE of weld flaws. It discusses an overview of ultrasonic NDE as used for weld inspections in nuclear power and other industries. Herein, data sets and ML models used in the literature are summarized, along with a generally applicable workflow for ML. Findings on the capabilities, limitations and potential gaps in feature selection, data selection, and ML model optimization are discussed. The paper identified several needs for quantifying and validating the performance of ML methods for ultrasonic NDE, including the need for common data sets.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Detection of ratcheting in finite element calculations

The distinction between a ratcheting and non-ratcheting response is critical for many high temperature design methods. Non-ratcheting is generally considered safe deformation remain bounded over the lifetime of the component while ratcheting is undesirable. As a particular example, the elastic perfectly plastic (EPP) design methods described in recent ASME Section III, Division 5 code cases require a designer to distinguish ratcheting from non-ratcheting for finite element analyses using a relatively simple, elastic perfectly-plastic constitutive response. However, it can be quite difficult to distinguish these two deformation regimes using finite element (FE) analysis particularly in the case where the actual ratcheting strain is small. In practice FE analysis of structures that are analytically in either the plastic shakedown or ratcheting regimes will result in small, cycle to -cycle accumulated strains characteristic of ratcheting. Distinguishing false ratcheting the structure is actually in the plastic shakedown regime from true ratcheting can be challenging. We describe the characteristics of nonlinear FE analysis that cause these false ratcheting strains and describe practical methods for distinguishing a ratcheting from a non-ratcheting response.

Messner, M. C.↗

DOE OSTI2019

Establishing temperature upper limits for the ASME Section III, Division 5 design by elastic analysis methods

ection III, Division 5 of the ASME Boiler and Pressure Vessel Code provides two broad paths for the design of high temperature, safety-critical nuclear components: design by elastic analysis and design by inelastic analysis. The design by elastic analysis approach, as the name suggests, uses a linear elastic stress analysis of the component and applies design rules designed to bound response of the actual structure, which will undergo both creep and plasticity. Currently, the Code allows the use of the elastic approach for all operating temperatures up to the maximum use temperatures in the Code. The bounds used in the elastic approach assume an uncoupled material response combining rate dependent creep with rate independent plasticity. However, at elevated temperatures creep and plasticity are coupled, rate dependent mechanisms and so the elastic analysis rules may become non-conservative. We present several examples of potential non-conservatism in the elastic analysis rules at high operating temperatures. Then we describe a systematic method for determining a temperature cutoff describing the transition from non-unified, rate independent plasticity material response to a rate dependent, unified plastic response. Logically, this transition temperature sets the upper bound for the allowable, conservative use of the design by elastic analysis approach and so we propose these temperatures, determined for all the Section III, Division 5 Class A materials, as Code limits for the applicability of the elastic approach.

Messner, M. C.↗

DOE OSTI2019

Welding Process Development for Spent Nuclear Fuel Canister Repair

The potential for stress corrosion cracking (SCC) of welded stainless-steel interim storage containers for spent nuclear fuel (SNF) has been identified as a high priority data gap. This paper presents a fusion welding process that was developed for SNF canister repair. Submerged arc welding (SAW) was developed to weld 12.7 mm (0.5 in.) thick 304L stainless steel plates to simulate the initial welds on SNF canisters. The SAW procedure was qualified following ASME Boiler and Pressure Vessel Code requirements. During SAW, the welding temperature was recorded at various locations by using thermocouples. After SAW, weld microstructures were characterized, joint mechanical properties were tested, and the maximum tensile residual stress direction was identified. After SAW procedure qualification, artificial cracks were excavated perpendicular to the maximum tensile residual stress direction in the SAW heat affected zone. Machine cold-wire gas tungsten arc welding (CW-GTAW) was developed and used for repair welding at cracked locations.

Tang, Wei↗

DOE OSTI2020

Structural Analysis Approach for the Defense Programs Package 3 (DPP-3)

The Pacific Northwest National Laboratory (PNNL) is the design authority for a new Type B hazardous materials transportation package designated as the Defense Programs Package 3 (DPP-3) for the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA). The DPP-3 has been developed using similar materials and fabrication methods employed in previous U.S. Nuclear Regulatory Commission (NRC), DOE, and NNSA certified packages. The DPP-3 design criteria are derived from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC), NNSA guidance and NRC regulatory guides in order to safely and securely transport a variety of payloads. Final regulatory approval by the NNSA will require physical testing to demonstrate that the containment vessel (CV) remains leaktight after enduring the entire regulatory testing sequence prescribed in Title 10 of the Code of Federal Regulations Part 71 (10 CFR 71). In order to gain confidence that the DPP-3 will remain leaktight after testing, the DPP-3 has been structurally analyzed using the Finite Element Analysis (FEA) software LS-DYNA. The FEA analyses serve two general purposes: first, they aid in design and development of the package, and second, they advise as to which drop orientations are expected to cause the most damage during regulatory testing. This paper will discuss how the design criteria are incorporated into analytical techniques needed to evaluate the FEA structural simulation results for 10 CFR 71 conditions to give confidence the DPP-3 testing campaign will be successful.

Sakalaukus Jr., Peter J.↗

DOE OSTI2024

STRUCTURAL MODELING TO SUPPORT POST-YIELD ACCEPTANCE CRITERIA FOR SPENT NUCLEAR FUEL CLADDING

Spent nuclear fuel (SNF) is evaluated for structural failure during storage and transportation scenarios. The U.S. Department of Energy’s Spent Fuel and Waste Science and Technology (SFWST) program has sponsored significant research in quantifying mechanical loads on SNF during storage and transportation scenarios using experimental and modeling methods. The SFWST program has also performed significant research on measuring the mechanical behavior of irradiated SNF as defueled cladding segments and cladding with fuel pellets to measure composite behavior. This paper considers some of the key material data from the Sibling Pin testing and uses structural modeling and analysis methods that have been informed by testing to consider post-yield acceptance criteria for SNF cladding structural analysis. Test data published by Oak Ridge National Laboratory (ORNL) and Pacific Northwest National Laboratory (PNNL) are the foundation for informing the material behavior of the models developed in this study. In particular, four-point bend (4PB) tests of fueled and defueled cladding segments provide significant information about the bending failure mode of SNF. ORNL’s 4PB test data is on fueled cladding segments, so the composite behavior of SNF is demonstrated. This paper describes PNNL’s coincident beam model that was developed to approximate the composite behavior of SNF. This paper also presents PNNL’s structural dynamic finite element models of a cask tip-over scenario, which is predicted to cause the strongest mechanical loads on SNF of all postulated storage and transportation scenarios. SNF bending loads predicted in the cask tip-over scenario and cladding acceptance criteria beyond yield are considered, with justification based on the Sibling Pin test data. ASME Boiler and Pressure Vessel code stress intensity limits are also considered. The ultimate goal of this work is to aid in the justification of structural acceptance criteria for SNF cladding beyond the cladding’s irradiated yield strength for use in structural analysis of all storage and transportation scenarios.

Klymyshyn, Nicholas A.↗

DOE OSTI2018

Notch Effects on the Short Term Rupture Properties of Alloy 617 Base Metal and Weldments

Alloy 617 is currently undergoing approval for inclusion in the ASME Boiler and Pressure Vessel Code in Section III Division 5 for use in high temperature nuclear reactors. Prior to use in construction of these reactors, however, it must be approved by the Nuclear Regulatory Commission. Several concerns relating to the effects of geometric discontinuities, such as notches and multi-axial stress states (as opposed to uniaxial stress states typically present in laboratory run creep tests), must first be addressed. Alloy 617 has been found to be notch-strengthening in short term (1000-2500 hour) creep testing for both base metal and weld metal and will fail preferentially in a straight gauge section rather than at a sharp notch, provided that the applied stresses are similar in both the straight gauge and notch sections. Additionally, specimens designed with semi-circular notches (or U notches, as they are referred to in this report) to impose multi-axial stress states on the specimen during creep-testing were found to either maintain similar creep-rupture-lives as uniaxial stress specimens tested at similar conditions, or have creep-rupture-lives that were significantly longer. This work shows that notches and multi-axial stress states do not appear to raise concerns with relation to the use of design rules created for Alloy 617 using smooth, straight-gauge test specimens and uniaxial creep tests.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Fundamental Understanding of Creep-Fatigue Interactions in 9Cr-1MoV Steel Welds. Final Report

Ferritic-martensitic steels constitute a vital class of materials for current and future generation of nuclear reactors as well as ultra supercritical coal fired power plants due to their superior creep resistant properties. One particular ferritic-martensitic steel, 9Cr-1MoV (Grade 91) is one of five materials approved for high temperature use in nuclear service by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC). Grade 91 is also a candidate material for proposed fission and fusion reactors such as the sodium cooled fast reactor (SFR). In service, Grade 91 components can be subjected to elevated temperatures and cyclic stresses, leading to the accumulation of creep-fatigue (CF) damage. While the monotonic creep properties of Grade 91 steels have been studied extensively, there is less knowledge concerning the mechanisms governing CF damage of this steel in the creep-dominated failure region. Moreover, the creep failure is especially pronounced in welded Grade 91 components, which are commonly joined using conventional arc welding techniques such as flux-cored arc welding (FCAW). In a Grade 91 weld, the creep failure typically occurs in the outer region of the heat affected zone (HAZ), specifically in the intercritical or fine-grained heat-affected zone (ICHAZ or FGHAZ), a phenomenon known as type IV failure. Currently, there exists a limited knowledge of CF damage mechanisms in Grade 91 welds. Improving the mechanistic understanding of the CF damage and failure of Grade 91 steel and its welds with inhomogeneous microstructure is crucial for the safety and design of key structural components in advanced, high-temperature reactors. The overall objectives of the project were to advance the state of knowledge and understanding of CF damage in both Grade 91 base metal and its welds, especially under loading conditions where creep is the dominant damage mechanism. Uncovering the mechanisms governing the creep-fatigue deformation and damage in Grade 91 base metal in creep-dominated region as well as type IV failure in Grade 91 welds contributes to a new knowledge that may lead to more accurate life prediction models than those currently in use. Moreover, the superior CF properties obtained by low heat input welding has a high potential to develop practical solutions to the premature failure of Grade 91 weld.

36 MATERIALS SCIENCE↗

DOE OSTI2018

Development of Grade 91 inelastic model for incorporation in ASME Division 5

This report describes a unified viscoplastic model for Grade 91 suitable for use with the Section III, Division 5 design by inelastic analysis procedures in the ASME Boiler and Pressure Vessel Code. The report identifies several key features a successful design inelastic model must capture along with critical aspects of the response of Grade 91 for elevated temperature loading conditions typical of expected reactor operating conditions. An inelastic model for Code design must capture the average response of a material including heat-to-heat and product form variations. This work develops a method for calibrating such a model to a large experimental database using genetic algorithm optimization. Finally, the report compares the final model to several specialized experimental tests to validate that it will be suitable for use in high temperature structural design. The techniques developed here can be applied to the remaining Section III, Division 5, Class A materials to develop a nonmandatory appendix to the Code describing inelastic material models suitable for use with the design by inelastic analysis procedures.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗