NO.3 保留

A Review on Current Status of Alloys 617and230for Gen IV Nuclear Reactor Internals and Heat Exchangers Weiju RenMetals Science and Technology Division,Oak Ridge National Laboratory,MS-6155,Building4500-S,Oak Ridge,TN37831e-mail:renw@Robert SwindemanCromtech,125Amanda Drive,Oak Ridge,TN37831e-mail:rswindeman@Alloys617and230are currently identified as two leading candi-date metallic materials in the down selection for applications at temperatures above760°C in the Gen IV nuclear reactor systems. Qualifying the materials requires significant information related to codification,mechanical behavior modeling,metallurgical sta-bility,environmental resistance,and many other aspects.In the present paper,material requirements for the Gen IV nuclear reac-tor systems are discussed;available information regarding the two alloys for the intended applications are reviewed and ana-lyzed;and further R&D activities are suggested.In the United States the major requirement for qualifying the materials is to satisfy the ASME Subsection NH,with adequate considerations for NRC,ASME NQA-1,and Section XI.In comparison,Alloy617 is more studied with larger existing databases in air and helium, while Alloy230may have highly desired potentials but needs more exploration.To provide a sound technical basis for the ma-terial selection decision,more data should be generated to char-acterize behaviors of both alloys in creep,loading rate sensitivity, fatigue,creep-fatigue,crack resistance,toughness,product form dependency,and metallurgical stability.͓DOI:10.1115/1.3121522͔1IntroductionAlloys617and230are currently considered as primary candi-date materials for Gen IV nuclear reactor components serving in the temperature range of760–950°C.1The leading Gen IV reac-tor concept,namely,the very high temperature reactor͑VHTR͒,is a gas-cooled reactor with the goal to produce helium at tempera-tures up to950°C and pressures up to7MPa for a design life of 60years.Approximately90%of the VHTR heat will be used to generate electricity and10%to produce hydrogen.Several other Gen IV reactor concepts under development around the world also require high service temperatures and long design lives.Applica-tions for the two alloys may include heat exchanger and other reactor components.For the VHTR,the structural alloys must face mainly three challenges including the effects of high temperature exposure,helium contaminants,and long-term irradiation on mi-crostructure stability,mechanical properties,and service perfor-mances.The other Gen IV reactor systems require similar consid-erations for materials selection.To qualify the materials,a significant amount of data must be acquired;various mechanical behavior models must be developed;strengthening mechanisms must be well understood;metallurgical evolution and stability must be accurately predicted;and most importantly,the materials must be approved by American Society of Mechanical Engineers Boiler and Pressure Vessel Code͑ASME B&PV Code͒for nuclear applications at elevated temperatures.Development of the Gen IV nuclear reactor systems is a joint effort of more than ten international partners under the Gen IV International Forum͑GIF͒.The materials qualification involves significant R&D,decision making,as well as collaboration of par-ticipants from different organizations and countries.To facilitate various R&D activities for the materials selection and qualifica-tion,it is highly important to periodically review the status of the candidate materials,and to provide guidelines and adjustments for further actions.The present paper is intended to give a preliminary review of the material requirements and available information on Alloys617and230that are related to R&D activities and appli-cations to the Gen IV nuclear reactor systems,with a focus on the needs for the VHTR.2The MaterialsThe development of Alloys617and230can be traced back to cobalt base Alloy L605,a.k.a.Haynes25͑UNS R30605͓͒1͔.To achieve improved temperature performance,Haynes International introduced Alloy188͑UNS R30188͒as a development from Al-loy L605with increases significantly in Ni and slightly in Cr,and addition of La to enhance corrosion resistance͓2͔.The develop-ment was soon followed by special metals with a Ni-base alloy of much lower Co,Alloy617͓3͔.Haynes later produced Alloy230 with even lower Co content͓4͔as a competitor of Alloy617.Both Alloy617and Alloy230have very attractive properties for high temperature applications.Standard specifications have since been developed and accepted by American Society for Testing and Ma-terials͑ASTM͒and American Society of Mechanical Engineers ͑ASME͒for the alloys to be produced by any steel manufacturers with their own specific production techniques.Alloy617,also designated as Inconel617,UNS N06617,or popularly known as W.Nr.2.4663a in Europe,was initially de-veloped for high temperature applications above800°C.It is of-ten considered for use in aircraft and land-based gas turbines, chemical manufacturing components,metallurgical processing fa-cilities,and power generation structures.In the late1970s and early1980s,the alloy was also investigated for the high tempera-ture gas-cooled reactor͑HTGR͒programs in the United States and Germany.The chemical composition of Alloy617is given in Table1.The high Ni and Cr contents provide the alloy with high resistance to a variety of reducing and oxidizing environments.The Al,in con-junction with Cr,offers oxidation resistance at high temperatures. In addition,the Al also forms intermetallic compound gamma prime͑␥Ј͒over a range of temperatures,which results in precipi-tation strengthening on top of the solid solution strengthening im-parted by the Co and Mo.It should be noted that since Co is immediately next to Ni in the periodic table,its solid solution strengthening effect is limited in the Ni base due to similar atom sizes.The high Co is not beneficial for applications in radiation environments since isotope Co-60is a high-energy gamma ray emitter.Strengthening is also derived from M23C6,M6C,Ti͑C,N͒and other precipitates.The M23C6and M6C are mostly found to be rich in Cr.Observations of the precipitates and their existing temperature ranges in Alloy617have been controversial.Detailed reviews can be found in Refs.͓5,6͔.Alloy230,also designated as Haynes230,UNS N06230,or popularly known as W.Nr.2.4733in Europe,is considered a relatively new alloy compared with Alloy617and serves similar applications as a competitor material of Alloy617.1Reduced from the initial1000°C consideration based on analysis results that 950°C can satisfy the application requirements.Contributed by the Pressure Vessel and Piping Division of ASME for publication in the J OURNAL OF P RESSURE V ESSEL T ECHNOLOGY.Manuscript received August13, 2007;final manuscript received February24,2009;published online May26,2009. Review conducted by T.L.͑Sam͒Sham.The chemical composition of Alloy230is compared with that of Alloy617in Table1.Its Ni base and high Cr content impart great resistance to high temperature corrosion in various environ-ments.Oxidation resistance is further enhanced by the micro-addition of rare earth element pared with Alloy617,Al-loy230has a high W concentration.The W and Mo in conjunction with C are largely responsible for the strength of the material.In Ni-base superalloys,B is generally considered as a beneficial trace element for improving hot working and creep properties͓7–9͔.The relatively high B content in Alloy230can be controlled to achieve optimized ductility and creep resistance. Usually B acts as an electron donor to affect the grain boundary energy,thus improving the ductility;it also segregates into grain boundaries and contributes to slowing down grain boundary dif-fusion to reduce the creep process.On the other hand,because B has a high cross section for thermal neutrons,the level of B must be carefully controlled.Under radiation,B may transmute to other elements and cause degradation of material properties,e.g.,grain boundary embrittlement͓10͔and irradiation assisted stress corro-sion cracking͑IASCC͒due to He͓11͔,and embrittlement due to Li;both elements are products resulting from the B10͑n,␣͒Li7 transmutation reaction with thermal neutrons:5B10+0n1→3Li7 +2He4.3Requirements for Gen IV ApplicationsOf all the technical requirements,a decisive step in qualifying candidate materials for applications in the Gen IV reactors is the acceptance into ASME B&PV Section III Division1Subsection NH for Class1Components in elevated temperature service͓12͔. Subsection NH covers construction of nuclear facility compo-nents in elevated temperature region and contains rules for mate-rials,design,fabrication,examination,testing,and overpressure relief of Class1components,parts,and appurtenances.At present, the temperature coverage in NH is far less than what is required for the VHTR reactor system.However,the rules it contains for elevated temperature design makes it the most important subsec-tion in delineating the property requirements for the design and construction of the VHTR and other Gen IV reactor systems.It is also worth mentioning that a United States draft code case͓13͔and a German code͓14͔that have higher temperature coverage already exist.Beside NH,properties of interest to the VHTR operating con-ditions also requires codification of candidate materials for the following code sections:Section III Division1Subsection NB for Class1Components,Subsection NC for Class2Components, Subsection ND for Class3Components,Subsection NG for core support structures,and relevant Code Cases,most importantly CC N-499and CC N-201;Section VIII Division1and Division2;and the piping Codes B31.1and B31.3where applicable.Due to the unprecedented working conditions for the VHTR components,the current ASME Code does not cover all the properties needed for the VHTR design and construction.As requirements emerge in the course of the reactor system design,new properties must be added.In addition to codification for design and construction men-tioned above,the candidate materials must also be prepared for licensing from the Nuclear Regulatory Commission͑NRC͒.Fur-thermore,they must meet the ASME NQA-1requirements for quality assurance͓15͔,and ASME Code Section XI requirements for in-service inspection of nuclear power plant components. 3.1Requirements for Subsection NH.Subsection NH was developed from Code Case N-47,primarily for a liquid-metal re-actor͑LMR͒program.The development started in the late1960s with the recognition that the low-temperature structural design methodology for light-water reactors would not be adequate for the LMR due to its higher operating temperature.The intention of NH is to provide design-by-analysis rules to guard against four failure modes at high temperature including͑1͒ductile rupture from short-term loadings,͑2͒creep rupture from long-term load-ings,͑3͒creep-fatigue failure,and͑4͒gross distortion due to in-cremental collapse and ratcheting.In general,NH requires the use of inelastic design analyses to reflect plasticity and time-dependent creep effects.Because of high operating temperatures up to950°C expected for the VHTR,the codification require-ments for Alloys617and230are considerably more extensive than those in current NH.There is little application experience to draw on,and the required60year design life for the Gen IV systems is nearly double that currently permitted by the NH rules. Of course,possibility also exists that components may be de-signed as replaceable so that their service times are much shorter than the60years of the reactor system design life.3.1.1Cold Forming Limits.When certain limits for cold form-ing are exceeded in component manufacturing,the cold work can impair material properties such as fatigue,creep rupture,impact toughness,and more.To ensure adequate service of the compo-nent,NH requires a post fabrication heat treatment depending on the amount of the cold work.With cold work less than5%,no such heat treatment is necessary,but when it is greater than20%, the heat treatment is a must.Between5%and20%,NH Figure NH-4212–1:“Permissible Time/Temperature Conditions for Ma-terial Which Has Been Cold WorkedϾ5%andϽ20%and Sub-jected to Short-Time High Temperature Transient”specifies the permissible total time of high temperature excursions beyond which the heat treatment is required.For Alloys617and230,the current NH limits of5%and20% should be reconsidered for the Gen IV systems due to the severe operating conditions,particularly the extremely high service tem-peratures.The accident temperature that may occur in a given component should be used to determine the maximum short-time excursion temperature.In the codification process,sufficient data from both alloys will be needed to generate respective curves,asTable1Chemical composition…wt%…of Alloy617Ni Cr Co Mo Fe Mn Al C617min44.520.010.08.0--0.80.05 230min-20.0- 1.0-0.300.200.05 617max-24.015.010.0 3.0 1.0 1.50.15 230max Bal24.0 5.0 3.0 3.0 1.000.500.15Cu Si S Ti B La P W617min-------230min-0.25---0.005-13.0 617max0.5 1.00.0150.60.006---230max-0.750.015-0.0150.050.03015.0shown in Fig.1.The curve provides an envelope for permissible total time of high temperature excursions,within which no heat treatment is required for a determined cold strain range.To generate such an envelope,the relationship between cold work and recrystallization or other metallurgical degradations,such as strengthening precipitates dissolution and/or coarsening,for various temperature excursions and times must be clearly de-termined.Test data for creep rupture,toughness,and ductility properties of the alloys with various amounts of cold work are also required.The values of x%and y%should be defined based on assessment of cold work effects on these properties as well as microstructural evolutions during long-term services.Some stud-ies have been in progress on the cold work limits of Alloys 617and 230in the United States and Japan,respectively,͓16,17͔for fossil energy applications.3.1.2Tensile Properties .The NH requires data of ultimate ten-sile strength S u and the yield strength S y as a function of tempera-ture.For applications to the Gen IV systems,the S u values of Alloys 617and 230should be provided for Code Table NH-3225-1at temperatures ranging from 550°C to at least 950°C with intervals of 25°C,and the values for lower temperatures are covered in Section II Part D Table U.The S y values of both alloys should be provided for Table I-14.5of Mandatory Appendix 1-14at temperatures ranging from 550°C to at least 950°C with inter-vals of 25°C,and the values for lower temperatures are covered in Section II Part D Table Y-1.ASME requires that for the purpose of codification,the maxi-mum temperature at which the value of a given property is pro-vided should be 50°C higher than the intended design tempera-ture.Therefore,to qualify Alloys 617and 230for Gen IV nuclear reactor applications,tensile properties ͑S u and S y ͒of both alloys must be generated from room temperature up to the possible high-est accidental temperature excursion at intervals of 25°C.The possible highest accidental temperature excursion should be well defined in the conceptual-design stage through theoretical deriva-tion and computational simulation.3.1.3Tensile Reduction Factors for Aging .To ensure safe op-eration of the Gen IV systems,metallurgical aging effects on ul-timate tensile and yield strengths of the structural materials must be taken into account in design.For Alloys 617and 230to be codified in NH,the TS reduction factor for ultimate tensile strength and the YS reduction factor for yield strength must be provided for Code Table NH-3225–2:“Tensile and Yield Strength Reduction Factor Due to Long Time Prior Elevated Temperature Service.”To satisfy the proposed design life of 60years for the Gen IV nuclear reactor systems,the reduction factor values for 60years at various possible component operating temperatures must be developed.Furthermore,for applications where 60years of operation is obviously too risky for the alloys,plans for replace-able components must be considered and additional reduction fac-tor values at time intervals of 5–10years should be generated.3.1.4Allowable Stress Intensity Values .The NH requires al-lowable stress intensity values for various loaded durations and temperatures.Sufficient test data of both Alloys 617and 230must be provided to produce values for two required parameters that are presented as curves in Figure I-14.3of Mandatory Appendix 1-14.These two parameters include the time-dependent stress intensity S t ,and the time-independent stress intensity S m ;both should be given as a function of temperature up to 1000°C.To produce such curves for Alloys 617and 230,information must be derived from tensile and creep test results over the temperature range of 425–1000°C.The allowable stress intensity value S mt can then be determined by the lower of S t and S m at a given temperature,and tabulated in Tables 14.3of Mandatory Appendix 1-14.These data should produce plots,as shown in the template in Fig.2.3.1.5Minimum Stress-to-Rupture .To guard against creep rup-ture,expected minimum stress-to-rupture values of Alloys 617and 230must be presented as a function of time and temperature numerically in Table 1-14.6and graphically in Figures I-14.6of NH Mandatory Appendix 1-14to cover durations up to 600,000h for the 60years of Gen IV nuclear reactor design life and a tem-perature range from 425°C to 1000°C at intervals of 25°C.Be-cause it is impossible to test the alloys for 600,000h,data beyond practical testing time must be determined through a combination of testing and modeling efforts.The developed data should be sufficient to produce plots similar to Fig.3.Due to possible dra-matic property degradations from extremely long-term aging,the modeling efforts should not merely be based on curve fitting or mathematical extrapolation.Studies must be conducted on aging and creep property degradation mechanisms and rates to deter-mine metallurgically based and experimentally viable testing time,and to estimate the risks in various extrapolations and modeling schemes.Furthermore,if extrapolation and modeling risksareFig.1Template plot for permissible time/temperature data re-quired for Subsection NH Codification of Alloys 617and 230that have been cold worked >x%and <y%and subjected to short-time high temperaturetransientFig.2Template plot for curves required for allowable stress intensity S mt as a function of temperature for Alloys 617and 230in Subsection NH Codificationconsidered high for long-term services,components made of these alloys should be designed as replaceable parts to limit their ser-vice times.3.1.6Stress-Rupture Factors for Weldments .A weldment usu-ally has different stress-rupture properties than that of the base metal due to its different microstructures resulting from the fusion and/or heating during the welding process.To addresses this dif-ference for Alloys 617and 230in NH,values of the stress-rupture factor for welds from specified filler metals must be provided in Table 1-14.10of Mandatory Appendix 1-14.These values should be derived from the ratio of the stress-rupture strength of the weld metal or weldment to that of the base metal away from the weld.Due to the impractical long testing times needed,the derivation must be combined with modeling based on data provided by stress-rupture tests on the weld metal,base metal,and weldment of each alloy.3.1.7Strain Range Fatigue Curves .To design against fatigue damage,NH requires providing the designers with curves of strain range versus the number of allowable cycles with tabulated values appended in Figures T-1420-1of Nonmandatory Appendix T:“Rules for Strain,Deformation,and Fatigue Limits at Elevated Temperatures.”For Alloys 617and 230,strain range versus allow-able cycle number curves similar to Fig.4must be generated to cover the temperature range from 400°C to 1000°C,preferably at intervals of 50–100°C depending on the severity of property variation within given temperature intervals.Small intervals will be required if the fatigue property varies sharply over a given temperature range.As a baseline,a curve for room temperature should also be provided.Sufficient information must be provided to generate NH Figures T-1420-1for Alloys 617and 230,respec-tively.For tabulated curve values appended to the figures,the number of cycles at each temperature should be provided at a preferable interval pattern of 10,2ϫ10,4ϫ10;102,2ϫ102,4ϫ102...up to 106.The curves for the code figures can be derived using data generated from strain-controlled fatigue tests or com-bined strain-and stress-controlled fatigue tests.The fatigue tests should be conducted at a strain rate of 10−3/s at a completely reversed stress ratio R =−1,and strain ranges for various numbers of cycles at failure up to 106should be determined.3.1.8Creep-Fatigue Damage Envelope .Both creep and fa-tigue cause accumulative damage in materials at high tempera-tures.NH requires that accumulated creep and fatigue damage be evaluated by the linear damage rule,and the evaluation includes hold time and strain rate effects.For a design to be acceptable,the creep and fatigue damage shall satisfy Eq.͑1͓͒12͔originally rec-ommended in Code Case 1592͓18͔.The approach combines the damage summations of Robinson ͓19͔for creep and of Miner ͓20͔for fatigue modified by Taira ͓21͔.͚j =1P ͩn N dͪj +͚k =1qͩ⌬t T dͪkՅD͑1͒where n is the actual fatigue cycles during event j ;N d is the fatigue life of the material;⌬t is the creep time during event k ;T d is the creep-rupture life of the material;and D is the total allow-able creep-fatigue damage.The first term represents the calculated fraction of fatigue life consumed in a component,and the second term represents the calculation of the consumed creep life.The equation indicates that the total accumulated creep and fatigue damage should be controlled within an envelope.For NH Codification of Alloys 617and 230,creep,fatigue,and creep-fatigue test data should be generated to produce their re-spective parameter values for Eq.͑1͒,which can then be summa-rized graphically in corresponding creep-fatigue damage enve-lopes for NH Figure T-1420-2,as exemplified by Fig.5.It should be noted that this creep-fatigue damage model is only a simple extension of the linear cumulative damage theory,a.k.a.Palmgren–Miner rule,for which path independence is assumed ͓20,22͔.For the intended Gen IV reactor applications at very high temperatures where creep damage plays an important role,the creep-fatigue damages are path dependent and interactive.The linear summation in Eq.͑1͒is no longer adequate for accurately describing the true material behavior.Depending on how the pa-rameters are developed,the results may become overconservative,leaving designers little room for design;or too speculative,putting reactor operations at risk.This is one of the issues that the ASME elevated temperature design task intends to resolve.There is in-terest in developing a new approach that may allow for variations in design to lead to different allowable limits based on enhanced modeling tools available to designers.To address many technical issues in the Gen IV Nuclear Reactor Program,United States De-partment of Energy ͑DOE ͒has been sponsoring a Cooperative Research and Development Agreement ͑CRADA ͒between ASME and United States DOE.Several tasks established in this CRADA are related to the creep-fatigue issue ͓23͔.Although the CRADA does not specifically address the creep-fatigue issue of Alloys617Fig.3Template plot for expected minimum stress-to-rupture values required for Subsection NH Codification of Alloys 617and230Fig.4Template plot for curves of fatigue strain range at vari-ous temperatures required for Subsection NH Codification of Alloys 617and 230and 230but of 9Cr-1Mo-V ,the new approach developed will likely provide a basis for further improvement to satisfy the needs of these two Ni-based superalloys.3.1.9External Pressure Time-Temperature Limits .At high temperatures,cylindrical and spherical components subjected to sustained external pressure may eventually buckle when creep de-formation exceeds certain limits.To design against time-dependent creep buckling,time-temperature limits for the appli-cation of Section II external pressure charts are provided in NH Figures T-1522.For NH Codification of Alloys 617and 230,the time-temperature limits for temperature range starting from 400°C to 1000°C,and time durations up to 1,000,000h should be developed,and time-dependent buckling charts required for NH Figures T-1522,as exemplified by Fig.6,must be generated.3.1.10Isochronous Stress-Strain Curves .To provide designers with information regarding total strain caused by stress under el-evated temperature conditions,isochronous stress-strain curves for Alloys 617and 230at temperatures ranging from 425°C up to1000°C at intervals of 25°C and times up to 600,000h for the required Gen IV reactor design life must be provided in NH Fig-ures T-1800.Hot tensile stress-strain curves should also be in-cluded in these figures as a baseline.Creep tests will be needed to produce data for generating the isochronous stress-strain curves.Because of the long design life of 60years,modeling must be conducted to assist creep testing efforts and to produce sufficient information to generate the curves required for NH Figures T-1800,as exemplified by Fig.7.Again,because extremely long-term aging may cause dramatic property degradations,the model-ing efforts should not merely be based on mathematical deriva-tion,but combined with considerations of material aging and property degradation rates.These curves will be particularly use-ful in simplified design methods permitted by NH.Full inelastic analyses,e.g.,time and rate dependent analyses,with unified con-stitutive equations are being considered for conditions when tem-peratures reach levels where plastic and creep deformations are difficult to differentiate and are highly strain rate dependent ͓24͔.3.2Desirable High Temperature Data Not in Subsection NH.It is expected that new requirements of material properties data that are not covered in current NH will emerge in the course of the design.The following material properties data are some of interest.3.2.1Biaxial Fatigue .To perform consistent and valid design analyses required by NH,constitutive equations ͑for multiaxial inelastic response of the alloys to complex time-varying,thermal,and mechanical loadings ͒and inelastic analysis guidelines need to be developed.Biaxial fatigue data are desirable for providing the basis for the constitutive equations,and are used to confirm the adequacy of application of the von Mises effective strain or other approaches for fatigue analysis.3.2.2Biaxial Creep-Rupture .Like the biaxial fatigue data,bi-axial creep-rupture data are desirable for providing the basis for the development of constitutive equations.They are also used for developing material specific multiaxial creep-rupture strength theories used in the code.3.2.3Loading Rate Dependency .Unlike the response at lower temperatures,Ni-base alloys such as Alloys 617and 230may exhibit high sensitivity to loading rate at very high temperatures ͓25,26͔.An example of this loading rate dependency is shown in Fig.8.Variations in loading rate can result in significant changes in the stress-strain curve,and the information is highly desirable for developing mechanical property models of the alloys forGenFig.5Template plot for creep-fatigue damage envelope re-quired for Alloys 617and 230in Subsection NH Codification under the currentcriteriaFig.6Template plot for time-temperature limits for application of Section II external pressure charts required for Alloys 617and 230in Subsection NHCodificationFig.7Template plot for isochronous stress-strain curves re-quired for Alloys 617and 230in Subsection NH Codification。