High Temperature Alloy 617 Properties for Engineering Design:

Transcription

1 High Temperature Alloy 617 Properties for Engineering Design: Program Overview & Approach to ASME Code Qualification Thomas Lillo Idaho National Laboratory Program Lead: Dr. R.N. Wright Technical Meeting on High-Temperature Qualification of High Temperature Gas Cooled Reactor Materials June 2014

2 Objectives Provide Technology Development to Support Future Design and Deployment of Very High Temperature Gas Cooled Reactors: Pressure Vessel Steam Generator and Intermediate Heat Exchanger (IHX) Support Codes and Standards Activities for SiC/SiC composites and Materials Handbook Program Goal Alloy 617 Code Case Submittal for ASME approval by FY15 Develop experimentally validated elevated temperature design methods applicable to any high temperature nuclear system Resolve Materials Issues Beyond Code Qualification that will allow design of components for life of plant Alloy 617 Plate Composition Element C S Cr Ni Mn Si Mo Ti Cu Fe Al Co B Concentration, wt% 0.05 < <0.001

3 Specific VHTR High-Temperature Materials Approach: Identify and resolve gaps in understanding by leveraging historical research and development and carrying out rigorously assessed research and analysis as necessary Current lower technical risk Outlet Temperature, ºC Dose RPV (dpa) Dose Internals (dpa) << Future << Primary Circuit Materials RPV Piping Internals Intermediate Heat Exchanger Current A508/533 A508/533 Graphite 304 and 316 SS 800H Current Lifetime Desired Lifetime 800H Steam Generator 2.25Cr 1 Mo 800H years years 60 years Future Grade 91 Grade 91 SiC/SiC Lifetime 12 years 60 years

4 Leveraging High Temperature Materials Research and Development Development and Demonstration in Germany and Japan: Extensive Alloy 800H steam generator materials research Alloy 617 and Hastelloy X (Alloy XR in Japan) steam generator and Intermediate Heat Exchanger (IHX) materials characterization Draft ASME Code Case submitted in US in 1990 Alloy 617 Code Case Submittal for ASME approval Code Case was withdrawn and did not receive final action Fossil Energy Ultrasupercritical Materials research in US and Europe 4

5 NGNP/VHTR Planning R&D Plans and Acquisition Plans Were Developed After Two Workshops, ASME Code Committee Input and Formal Program Peer Review Alloy 800H was identified as primary candidate alloy for steam generator based on current acceptance in ASME Code and likelihood that Code allowed temperature and time could be extended Alloy 617 was down selected as primary IHX alloy based on high temperature properties and technical maturity 5

6 Design Envelope for VHTR RPV >6m in diameter and 150 to 250 mm in thickness; Operating Temperature of 350C Due to weldability, Code qualification to 371ºC and technical maturity, downselection to A508/533 steel was made Steam Generator up to 750ºC and Sixty Year Life Alloy 800H currently Code Qualified up to 760C and 300,000 hours Low allowable stress at 750ºC for Alloy 800H suggests alloy with superior properties may be desirable to allow design for life of plant IHX up to 950ºC (with life up to 20 years desirable) Alloy 617 was downselected from possible alloys Code qualification for elastic design up to 427ºC Code qualification allowing design in inelastic regime up to 950ºC Detailed Analysis of Needs Specified in: PLN-2803 Next Generation Nuclear Plant Reactor Pressure Vessel Materials Research and Development Plan PLN-2804 Next Generation Nuclear Plant Steam Generator and Intermediate Heat Exchanger Research and Development Plan

7 ASME High Temperature Materials Code Qualification Approach Two aspects Elevated Temperature Design Methods Provide adequate design curves derived from experiments NB for temperatures up to 427ºC NH for temperatures up to 950ºC A Task Group on Alloy 617 Code Qualification has been established to provide guidance, review, and comment on the process Staff associated with the High Temperature Materials R&D have become members of relevant Code committees to facilitate the Code Case 7

8 Alloy 617 ASME Code Qualification Schedule Code Case for nuclear design in the elastic regime (Section III NB, T < 427ºC) will be submitted in June 2014 Alloy 617 Code Case for Subsection NB, includes tensile properties, modulus and fatigue design curves In September 2015 it is planned to submit Code Case for Alloy 617 for elevated temperature design Use temperature up to 950ºC for time up to 150,000 hours I-9.5M from ASME Section III Appendices (UNS N06003, N06007, N06455, and N10276 for T 425 C) 8

9 Interaction with NRC ASME Code does not address potential issues with in-service degradation that may be concerns to NRC NRC performed formal PIRT (Phenomena Identification and Ranking Table) evaluation Vendor teams provided a materials selection and qualification plan to the NRC at a public meeting VHTR project continues bi-monthly conference calls to update NRC on project status NRC staff participate in many of the relevant ASME Committees 9

10 Alloy 617 Code Qualification Combining Rigorously Assessed historical Data and Contemporary Data Developed Using NQA 1 Quality Level Tensile and fatigue properties from ambient temperature to 427ºC Creep, fatigue, creep-fatigue and tensile properties for temperatures above 427ºC Developing Code allowable values from experiment using Code methods Characterizing properties of weldments where appropriate Re-evaluated physical properties up to 1000ºC Investigating Mechanical Properties After Long Term Aging Determining Mechanisms Responsible for Observed Behavior to Address Potential for Extrapolating Behavior Beyond Time/Temperature/Stress Examined Experimentally Started ASME Code Committees interaction process Participate in Alloy 617 Task Group Presented information at the May ASME Code Week to the Subgroup on Materials, Fabrication and Evaluation Presented information at the May ASME Code Week to the Subgroup on Fatigue In Some Cases Suggesting Modification to Code Rules that Limit Performance Potentially relaxing 1% tertiary creep limit for example

11 Simplified Design Methods Developed methodologies based on elastic-perfectly plastic (E-PP) finite element analyses, covering Primary loads Strain limits Creep-fatigue Implemented methodologies into three draft Design Code Cases Started ASME Code Committees interaction process Presented technical basis to Working Group Analysis Methods and Subgroup Elevated Temperature Design Verification and qualification of draft code rules to be conducted

12 Supporting HTM Activities HTM 9: A508/533 Pressure Vessel Steel Characterization Julian Benz, INL HTM 10: Generation IV Materials Handbook Weiju Ren, ORNL HTM 11: FHR Materials (Supported by ORNL LDRD) Govingarajan Muralidharan, ORNL HTM 12 SiC/SiC composites Code Activities Yutai Katoh, ORNL

13 Related Nuclear Energy University Programs (NEUP) Corrosion and Creep of Candidate Alloys in High Temperature Helium and Steam Environments for the NGNP Creep-Fatigue and Creep-Ratcheting Failures of Alloy 617: Experiments and Unified Constitutive Modeling towards Addressing the ASME Code Issues Mechanisms Governing the Creep Behavior of High Temperature Alloys for Generation IV Nuclear Energy Systems Assessment of Embrittlement of VHTR Structural Alloys in Impure Helium Environments Identifying and Understanding Environment-Induced Crack Propagation Behavior in Ni-Based Superalloy INCONEL 617 Monitoring microstructural evolution of Alloy 617 with nonlinear acoustics for remaining useful life prediction; multiaxial creep-fatigue and creep-ratcheting Development of Barrier Layers for the Protection of Candidate Alloys in the VHTR New Mechanistic Models of Creep-Fatigue Crack Growth Interactions for Advanced High Temperature Reactor Components Multi-Resolution In Situ Testing and Multiscale Simulation for Creep Fatigue Damage Analysis of Alloy 617 University of Michigan North Carolina State University University of Cincinnati University of Wisconsin-Madison University of Nevada, Las Vegas Pennsylvania State University University of California, Santa Barbara Oregon State University Arizona State University 13

14 Tensile Properties and Mechanical Testing of Alloy 617 Lead -Nancy Lybeck Team Members: J. Simpson, Q.R. Lloyd, J.K.Wright Idaho National Laboratory Technical Meeting on High-Temperature Qualification of High Temperature Gas Cooled Reactor Materials June 2014

15 Introduction Qualifying data to put in the code cases for use of Alloy 617 in nuclear applications at low (NB) and elevated (NH) temperatures Investigating relevant phenomena Strain rate sensitivity at elevated temperatures Creep rupture Effects of weldments Combining data sources Legacy data Published data Newly generated data Establishing data provenance Adherence to relevant test standards Availability of supporting information Merging old and new data systematically Look for statistical differences between older and modern heats 15

16 STRAIN RATE SENSITIVITY 16

17 Importance of Strain Rate Sensitivity Why is this important? Strain rate defines stress allowables at elevated temperatures When combining data sets, need to make sure similar conditions were used Strain rate must be carried with the data 17

18 True Stress (MPa) Method Strain rate sensitivity was determined at ºC using strain rate jump tests Strain rate sensitivity, m, is determined by the change in stress resulting from an instantaneous change in strain rate 250 Fastest strain rate Slowest strain rate m = log ( σ 2 σ 1 ) log( ε 2 ) ε C True Strain (mm/mm) 18

19 Strain Rate Jump Tests, Alloy 617 Above 750ºC the increments of stress associated with strain rate jumps are of consistent magnitude Serrated flow associated with dynamic strain aging occurs at lower temperatures, but little strain rate sensitivity 19

20 True Stress (MPa) Strain Rate Jump Results, Alloy 617 m is relatively independent of strain rate in the range of ºC m increases slightly with increasing temperature C 850 C 900 C 950 C y = 1145x y = 1060x y = 842x y = 710x E E E E E E-02 Strain Rate (/s) 20

21 TENSILE DATA ANALYSIS 21

22 Tensile Data Yield strength and tensile strength at temperature are used to set time independent allowable stress for structural materials in the ASME Boiler and Pressure Vessel Code, Section III, Subsection NH Legacy database exists from suspended code case Additional data sets were identified Oak Ridge National Laboratory (Gen IV) CEA (literature) Idaho National Laboratory (generated) University of Nevada Las Vegas (literature, excluded from analysis because strain rate was too low) Source of the metal ORNL, INL (new heats from ThyssenKrupp VDM) ORNL, legacy data (older heats from Special Metals, Huntington, VA) Strain rate 8.3e-5/sec 22

23 Data Sources Heat Product Form Vendor Data Source XX00A1USL BAR Huntington Huntington XX00A4USL BAR Huntington Huntington XX00A5USL BAR Huntington Huntington XX05A4UK BAR Huntington Huntington XX07A7UK BAR Huntington Huntington XX00A1USL CR SHEET Huntington Huntington XX00A5USL CR SHEET Huntington Huntington XX20A5UK CR SHEET Huntington Huntington XX26A8UK CR SHEET Huntington Huntington XX00A3USL FORGING Huntington Huntington XX00A3USL PLATE Huntington Huntington BAR VDM INL Plate VDM INL XX01A3US Plate Huntington ORNL XX09A4UK Plate Huntington ORNL VDM Plate VDM CEA 23

24 Engineering Stress Strain Curves Tensile and Yield strength vary substantially with temperature Work hardening is observed at temperatures below 850ºC Dynamic strain aging is observed between ºC Tensile stress-strain curves for Alloy

25 Yield Strength Data 198 data points 14 heats 5 product forms Additional data appear to be consistent with the suspended code case data (Huntington) Do the modern heats have similar behavior to the older heats? 25

26 Statistical Evaluation Fit piecewise exponential decay model to data from older heats (Huntington) and modern heats (VDM) 95% confidence bounds (for the mean) are overlapping through the region of interest There is no evidence to suggest a difference in the data sets 26

27 Tensile Strength Data At lower temperatures, the additional data are consistent with the legacy data At higher temperatures, the additional data points are consistently lower than the legacy data 27

28 Statistical Evaluation Fit piecewise exponential decay model to data from older heats (Huntington) and modern heats (VDM) There is no evidence to suggest a difference in the data sets at lower temperatures Modern heats have lower tensile strength data than older heats at higher temperatures 28

29 ASME Method Normalize tensile and yield strength data using average room temperature tensile or yield strength for the associated heat Generate best-fit trend curve to the normalized data Traditionally use a polynomial Cubic is adequate for yield strength (Sham, Eno, and Jensen, PVP 2008) Quartic is adequate for tensile strength (Sham, Eno, and Jensen, PVP 2008) Scale the best-fit curves Multiply normalized yield strength curve by 240 MPa Multiply normalized tensile strength curve by Mpa 29

30 Strength Values (MPa) ASME Yield and Tensile Strength at Temperature Temperature (ºC) Resulting curve at low temperatures Lower bound for yield strength Conservative average for tensile strength Yield strength at temperature does not change much with the additional data Tensile strength at temperature is more conservative with the additional data 30

31 CREEP RUPTURE DATA ANALYSIS 31

32 Creep Data The low-temperature (NB) code case assumes no creep. The high-temperature (NH) code case requires creep data. Experimental creep tests have rupture times that are substantially shorter than the expected nuclear plant service life of 60 years. Material models are necessary to predict longer life e.g., Larson-Miller model Material model often used with creep data Based on the Arrhenius rate equation Used to predict rupture life for a specified temperature and stress The Larson-Miller parameter is given by L = T C + log t T is the temperature (ºKelvin) t is the time to rupture (hours) C is a material specific constant ( 20) 32

33 Creep Rupture Stress (Mpa) Larson-Miller Plot Larson-Miller Parameter (C=20) Linear regression was performed for older heats and modern heats 95% confidence bounds are not overlapping The modern heats have lower creep rupture stress 33

34 Creep Rupture Stress (Mpa) Larson-Miller Plot - Welds Larson-Miller Parameter Linear regression was performed for base and welded specimens 95% confidence bounds are overlapping over the regions with the most data There is no strength reduction caused by weldments 34

35 Conclusions Creating a code case requires a methodical, systematic approach to data collection Data provenance and completeness are critical Analysis should be performed to determine the data are all comparable INL generated data go through an NQA-1 qualification process Was the test procedure followed? Were all test specifications met? Data are stored in the Nuclear Data Management and Analysis System (NDMAS) with all supporting information Data are submitted to GEN IV from NDMAS Alloy 617 Tensile properties: Alloy 617 does not exhibit strain rate sensitivity below 800 o C Dynamic strain aging observed in the range of o C Modern heats of Alloy 617 exhibit lower elevated temperature tensile strength Modern heat of Alloy 617 exhibit lower creep rupture stress No apparent weld strength reduction factor for this alloy 35