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A Review on Current Status of Alloys 617and230for Gen IV Nuclear Reactor Internals and Heat Exchangers Weiju Ren

Metals Science and Technology Division,

Oak Ridge National Laboratory,

MS-6155,Building4500-S,

Oak Ridge,TN37831

e-mail:renw@https://www.doczj.com/doc/226599820.html,

Robert Swindeman

Cromtech,

125Amanda Drive,

Oak Ridge,TN37831

e-mail:rswindeman@https://www.doczj.com/doc/226599820.html,

Alloys617and230are currently identi?ed 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 signi?cant information related to codi?cation,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:10.1115/1.3121522?

1Introduction

Alloys617and230are 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 reactor?VHTR?,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 signi?cant 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 Code?ASME B&PV Code?for 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 Forum?GIF?.The materials quali?cation involves signi?cant 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 quali?ca-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 Materials

The development of Alloys617and230can be traced back to cobalt base Alloy L605,a.k.a.Haynes25?UNS R30605??1?.To achieve improved temperature performance,Haynes International introduced Alloy188?UNS R30188?as a development from Al-loy L605with increases signi?cantly in Ni and slightly in Cr,and addition of La to enhance corrosion resistance?2?.The develop-ment was soon followed by special metals with a Ni-base alloy of much lower Co,Alloy617?3?.Haynes later produced Alloy230 with even lower Co content?4?as a competitor of Alloy617.Both Alloy617and Alloy230have very attractive properties for high temperature applications.Standard speci?cations have since been developed and accepted by American Society for Testing and Ma-terials?ASTM?and American Society of Mechanical Engineers ?ASME?for the alloys to be produced by any steel manufacturers with their own speci?c 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 reactor?HTGR?programs 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 prime????over 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 bene?cial for applications in radiation environments since isotope Co-60is a high-energy gamma ray emitter.Strengthening is also derived from M23C6,M6C,Ti?C,N?and 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.?5,6?.

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;?nal manuscript received February24,2009;published online May26,2009. Review conducted by T.L.?Sam?Sham.

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 https://www.doczj.com/doc/226599820.html,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 bene?cial trace element for improving hot working and creep properties?7–9?.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 embrittlement?10?and irradiation assisted stress corro-sion cracking?IASCC?due to He?11?,and embrittlement due to Li;both elements are products resulting from the B10?n,??Li7 transmutation reaction with thermal neutrons:5B10+0n1→3Li7 +2He4.

3Requirements for Gen IV Applications

Of 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 service?12?. 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 case?13?and a German code?14?that have higher temperature coverage already exist.

Beside NH,properties of interest to the VHTR operating con-ditions also requires codi?cation 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 codi?cation for design and construction men-tioned above,the candidate materials must also be prepared for licensing from the Nuclear Regulatory Commission?NRC?.Fur-thermore,they must meet the ASME NQA-1requirements for quality assurance?15?,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-actor?LMR?program.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 including?1?ductile rupture from short-term loadings,?2?creep rupture from long-term load-ings,?3?creep-fatigue failure,and?4?gross distortion due to in-cremental collapse and ratcheting.In general,NH requires the use of inelastic design analyses to re?ect plasticity and time-dependent creep effects.Because of high operating temperatures up to950°C expected for the VHTR,the codi?cation 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 Worked?5%and?20%and Sub-jected to Short-Time High Temperature Transient”speci?es 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 codi?cation process,suf?cient data from both alloys will be needed to generate respective curves,as

Table1Chemical composition…wt%…of Alloy617

Ni Cr Co Mo Fe Mn Al C

617min44.520.010.08.0--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 1.000.500.15

Cu Si S Ti B La P W

617min-------230min-0.25---0.005-13.0 617max0.5 1.00.0150.60.006---230max-0.750.015-0.0150.050.03015.0

shown 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 de?ned 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,?16,17?for 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-1

4.5of 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 codi?cation,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 de?ned 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 codi?ed 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.Suf?cient test data of both Alloys 617and 230must be provided to produce values for two required parameters that are presented as curves in Figure I-1

4.3of 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 suf?cient 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 ?tting 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 risks

are

Fig.1Template plot for permissible time/temperature data re-quired for Subsection NH Codi?cation of Alloys 617and 230that have been cold worked >x%and

transient

Fig.2Template plot for curves required for allowable stress intensity S mt as a function of temperature for Alloys 617and 230in Subsection NH Codi?cation

considered 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 speci?ed ?ller metals must be provided in Table 1-1

4.10of 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.Suf?cient information must be provided to generate NH Figures T-1420-1for Alloys 617and 230,respec-tively.For tabulated curve values appended to the ?gures,the number of cycles at each temperature should be provided at a preferable interval pattern of 10,2?10,4?10;102,2?102,4?102...up to 106.The curves for the code ?gures 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.?1??12?originally rec-ommended in Code Case 1592?18?.The approach combines the damage summations of Robinson ?19?for creep and of Miner ?20?for fatigue modi?ed by Taira ?21?.

?j =1

P ?

n N d

?j +

?k =1

q

??t T d

?

k

?D

?1?

where 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 ?rst 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 Codi?cation of Alloys 617and 230,creep,fatigue,and creep-fatigue test data should be generated to produce their re-spective parameter values for Eq.?1?,which can then be summa-rized graphically in corresponding creep-fatigue damage enve-lopes for NH Figure T-1420-2,as exempli?ed 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 ?20,22?.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.?1?is 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 ?23?.Although the CRADA does not speci?cally address the creep-fatigue issue of Alloys

617

Fig.3Template plot for expected minimum stress-to-rupture values required for Subsection NH Codi?cation of Alloys 617and

230

Fig.4Template plot for curves of fatigue strain range at vari-ous temperatures required for Subsection NH Codi?cation of Alloys 617and 230

and 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 Codi?cation 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 exempli?ed 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 to

1000°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 ?gures 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 suf?cient information to generate the curves required for NH Figures T-1800,as exempli?ed 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 simpli?ed design methods permitted by NH.Full inelastic analyses,e.g.,time and rate dependent analyses,with uni?ed con-stitutive equations are being considered for conditions when tem-peratures reach levels where plastic and creep deformations are dif?cult to differentiate and are highly strain rate dependent ?24?.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 con?rm 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 speci?c 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 ?25,26?.An example of this loading rate dependency is shown in Fig.8.Variations in loading rate can result in signi?cant changes in the stress-strain curve,and the information is highly desirable for developing mechanical property models of the alloys for

Gen

Fig.5Template plot for creep-fatigue damage envelope re-quired for Alloys 617and 230in Subsection NH Codi?cation under the current

criteria

Fig.6Template plot for time-temperature limits for application of Section II external pressure charts required for Alloys 617and 230in Subsection NH

Codi?cation

Fig.7Template plot for isochronous stress-strain curves re-quired for Alloys 617and 230in Subsection NH Codi?cation

IV nuclear reactor systems.To ef?ciently generate the data,tests at loading rates of 10?3/s,10?4/s,and 10?5/s representing typical fatigue,tensile,and creep testing loading rates,respectively,at temperatures from 750°C to 1050°C are suggested.Depending on initial test results,these rates may be adjusted to adequately characterize the material response;the characteristics,particularly the shape,of the tensile curves will provide information on which yield strength and tensile strength will be used for design.3.3Considerations for NRC,ASME NQA-1,and Section XI

3.3.1Crack Growth Rate .In large components of the Gen IV systems,pre-existing ?aws in structural materials can be expected as inevitable.Furthermore,with aging and property deterioration during long-term services at elevated temperatures,microcracks can be initiated as a result of accumulation of creep and fatigue damage in highly stressed areas.Stress concentration at the tips of these ?aws can then induce localized creep deformation at el-evated temperatures even under a very low applied load level.As time elapses,these ?aws may grow by a process of coalescence of creep cavities near the sharp tips,and fatigue cycling can accel-erate this process.It has been recognized,however,that although microcracks can pre-exist and initiate in a component during its service at elevated temperatures,a large portion of the compo-nent’s useful life can be spent in crack propagation.With the requirements of 60years of service life and a 950°C outlet tem-perature for the VHTR system,crack growth properties of the candidate materials could become very important factors that should be considered in the materials selection and reactor licens-ing process.

3.3.1.1Creep Crack Growth Rate .To estimate creep crack growth ?CCG ?resistance of Alloys 617and 230,creep crack growth rate data of both alloys should be provided for materials selection consideration.Several time-dependent fracture mechan-ics parameters exist and are appropriate for characterization under various creep conditions of growing creep cracks.The most popu-larly used two parameters are C ?for crack growth under extensive secondary creep conditions,and C ??t ?for the growth under exten-sive primary and/or secondary creep conditions.Creep crack growth rate data should be provided preferably at 50°C intervals from 750°C to 1000°C to cover the intended service temperature range.Possibility exists that such characterization is not affected by temperature.Then the maximum temperature of 1000°C should be used in testing to save testing time.Otherwise,suf?-cient data should be generated to produce a plot,as shown in Fig.9,at each temperature.

3.3.1.2Fatigue Crack Growth Rate .To compare fatigue crack growth resistance of Alloys 617and 230,fatigue crack growth rate data should be generated and characterized using the stress intensity factor range ?K at various temperatures preferably at 50°C intervals or less in the temperature range of 750–1000°C to cover the intended service temperatures.Suf?cient data should be generated to produce a plot,as shown in Fig.10,at each tempera-ture.Due to the increased sensitivity to the loading rate of both alloys at very high temperatures,tests at various loading rates should be considered.

3.3.1.3Creep-Fatigue Crack Growth Rate .At the tip of a pre-existing ?aw or a microcrack initiated due to long-term exposure at elevated temperature under cyclic loading,a fatigue damage zone can be developed within a creep zone.The overall damage may not be satisfactorily represented by a linear summation of

the

Fig.8Strain-controlled tensile test curves at 24°C and 950°C and different loading

rates

Fig.9Template plot for crack growth rate data desirable for estimating creep crack resistance of Alloys 617and

230

Fig.10Template plot for crack growth rate data desirable for estimating fatigue crack resistance of Alloys 617and 230

creep and fatigue damage effects.Synergism of the two damage mechanisms usually causes accelerated material deterioration.Therefore,creep-fatigue crack growth testing should be conducted to generate creep-fatigue crack growth rate data and damaged ma-terial specimens that provide information on the material deterio-ration process through microstructural characterization.Several time-dependent fracture mechanics parameters exist for character-izing the creep-fatigue crack growth rate.For each alloy,creep crack growth testing should be conducted preferably at 50°C in-tervals from 750°C to 1,000°C to cover the intended service temperature range,if temperature effects are exhibited in the char-acterization.Suf?cient data should be generated to produce a plot,as shown in Fig.11,at each temperature.

3.3.1.4Crack Growth Rate of Welds .Fusion welding pro-cesses are usually primary techniques employed in joining metal-lic materials in the construction of elevated temperature compo-nents.Due to their dendritic microstructures ?like a miniature casting ?and the heat affected zone ?HAZ ?,welds are usually more vulnerable to creep,fatigue,and creep-fatigue crack growth dam-age.Suf?cient test data on crack growth rates in the welds of the Alloys 617and 230should be provided to determine their resis-tance to these crack growth mechanisms for materials selection and licensing considerations.

3.3.2Fracture Toughness .The severe working environment of the VHTR is expected to promote several mechanisms that can cause embrittlement.Therefore,to minimize the risk of structural failure during shutdown and restart,the capability of the alloys to resist embrittlement must be considered,and fracture toughness of the components for which toughness will be an issue should be evaluated.Appropriate fracture toughness characterization of Al-loys 617and 230should be required for the material selection consideration.Based on available preliminary experimental data,Alloy 617demonstrated rapid decreases in room temperature im-pact energy after aging at temperatures above approximately 600°C ?27?.Although transition temperature behavior is normally not expected in face centered cubic ?FCC ?materials,such as Al-loys 617and 230,temperature dependency of impact energy is still possible and should be investigated.Therefore,impact energy absorption data for various temperatures from Charpy tests on both alloys are desirable for comparing the toughness of the two alloys.To evaluate the effects of the service environment,tough-ness measurements should be made before and after exposure to simulated VHTR working conditions.Suf?cient data should be provided for both alloys to produce curves,as shown in the ex-ample in Fig.12,in which two hypothetical alloys are evaluated and their relative toughness can be compared.Such comparison can also be conducted on the same alloy with different exposures to simulated VHTR working conditions.

3.3.3Metallurgical Data on Aging Effects .Knowledge of ag-ing effects on long-term properties is important in understanding damage mechanisms associated with long-term exposure to creep and fatigue.As was previously discussed,the 60years of design life of the Gen IV systems make it impractical to conduct very long-term testing for aging effects.Therefore,metallurgical data and modeling must be employed to supplement testing in predic-tion of full-term materials behavior under the intended operating conditions.The time-temperature-transformation ?TTT ?diagrams are considered an ef?cient tool for this purpose.Generally,calcu-lations that predict stable phases as a function of alloy chemistry say very little about kinetics and the development of nonequilib-rium transition phases.Such information may be derived from TTT diagrams.The TTT diagrams are useful in understanding the metallurgical factors that control deformation mechanisms,failure mechanisms,strength,and ductility.For Gen IV nuclear reactor applications of 60years,the TTT diagrams of Alloys 617and 230should be produced up to 600,000h,as shown in Fig.13,unless the alloys are used only for replaceable components of shorter operation lives.Since experimentally generate data for 600,000h before the reactors are constructed is impossible,studies are needed to predict the potential long-term metallurgical changes,for example,through a combination of computational simulation and accelerated testing at adequately higher temperatures.Risks and uncertainties of such results should also be estimated to de-velop safety margins and measures.Some existing TTT diagrams,such as those provided in Refs.?6,28,29?,may be used as a base for an extension work.Findings in metallurgical changes during aging by various researchers also provide references ?29–34?.Fur-thermore,it is known that the compositional variables,fabrication variables,thermal-mechanical loadings,and radiation may affect the kinetics of precipitation and resolution of precipitates in these two alloys.To be useful for modeling deformation and fracture,the diagrams should be developed to indicate the initiation time,50%completion time and 95%completion time for important phases,as well as their precipitation sites ?grain boundaries,twins,cell boundaries,dislocation networks,etc.?.Histograms revealing the distribution of precipitate sizes provide useful information as well.Other histograms revealing the size distribution of laths and subgrains,as well as dislocation distributions and densities,

would

Fig.11Template plot for crack growth rate data desirable for estimating creep-fatigue crack resistance of Alloys 617and

230

Fig.12Template plot for energy-temperature curves from Charpy tests desired for evaluating toughness of Alloys 617and 230

be helpful.Meanwhile,a monitoring program that starts a few years prior to the reactor operation with continued aging and test-ing of the structural materials under the service conditions may also be considered to provide continuous forecast information on metallurgical changes in the reactor structural materials through-out the reactor’s operational life.

Besides aging effects on long-term properties,knowledge of the aging effects on short-time properties,such as yield,?ow,and hardening properties,is also very important.Furthermore,estima-tion of toughness is highly desirable for risk assessment purposes.As mentioned previously on properties required for NH Codi?ca-tion,it is necessary to develop tables of strength reduction factors for short-time properties.

3.3.4Environmental Effects .In the VHTR operating environ-ment,the large quantity of coolant helium circulating within the reactor system will not be ideally pure,but inevitably carry vari-ous gaseous impurities such as N 2,CO,CO 2,H 2,H 2O,and CH

4.These impurities can cause undesirable chemical reactions with the component materials,resulting in environmental degradation.Depending on the helium impurity chemistry,the structural alloys may be oxidized,carburized,or decarburized.As a matter of fact,to keep the inevitable environmental degradation under control within the reactor system,some elements that form these impure gases may be intentionally added to maintain a desired chemical composition.A slightly oxidizing environment may be desired for maintaining the protective surface oxide scale.To address the en-vironmental effects on mechanical properties in design,mechani-cal test data of both Alloys 617and 230from the impure helium environment of the VHTR and other Gen IV reactor coolants must be provided.

3.3.5Radiation Effects .Inevitably,some of the reactor com-ponents will be exposed to radiation in service ?e.g.,control rod sleeves may experience a ?uence ?0.15?1021n /cm 2/yr fast,?2.5?1021n /cm 2/yr thermal.?and experience irradiation dam-age.The long-term irradiation can cause operation of micro-mechanisms such as abnormal absorption of interstitials at dislo-cations,accumulation of vacancies at cavities,asymmetrical partitioning of self-interstitials and vacancies to dislocations dif-ferently oriented to stresses,creation of extremely small obstacles,and weakening of grain boundaries.These damaging mechanisms can result in swelling ?35,36?,irradiation creep ?37?,and irradia-tion embrittlement ?38–43?.If Alloys 617and 230are planned for potential applications in radiation regions of the Gen IV nuclear reactor systems,to ensure knowledgeable design and construction,test data generated after exposure to simulated radiation condi-tions are desired for materials selection and quali?cation.Realis-tically,the two alloys are likely to be excluded from such appli-cations.In addition to potential radiation damages discussed above,Co in both 617and 230can cause another problem:expo-sure to radiation can generate radioactive cobalt-60,which will pose a health and environmental hazard during maintenance or accidental leakage.

4Status of Alloys 617and 230

Since the introduction of Alloys 617and 230into the market-place,signi?cant research activities have been conducted around the world to characterize their properties and explore their poten-tials.The results of these investigations exist in various public and nonpublic records,documents,and databases,many of which con-tain information applicable to Gen IV nuclear reactor design and construction.Generally speaking,Alloy 617is better investigated with more existing data compared with Alloy 230due to its earlier introduction into the marketplace.On the other hand,as a newer alloy,Alloy 230may have potentials that are highly desired for the intended application and need exploration.At this stage of Gen IV nuclear reactor development,it is impossible for the present paper to cover and scrutinize all the existing information and ongoing research activities.The following status review can only be considered preliminary.

4.1Data on Mechanical Properties.The ?rst data set on Alloy 617was produced by its manufacturer Huntington Alloys,Inc.for general applications of the alloy,which included manu-facturing such components as ducting,combustion cans,transition liners in both aircraft and land-based gas turbines,catalyst-grid supports for the production of nitric acid,heat-treating baskets in metallurgical processing,reduction boats in the re?ning of molyb-denum,and many others ?44?.The Huntington data set include tensile properties data generated from 179tensile tests at tempera-tures from 25°C to 1093°C,and creep property data generated from 249creep tests at temperatures from 593°C to 1093°C.The tensile properties include yield strength,ultimate tensile strength,total strain,and reduction in area.The creep properties include creep-rupture time,creep-rupture stress,creep-rupture strain,minimum creep rate,time to 1%total strain,and time to 0.2%offset tertiary creep.The longest creep test lasted 28,735h.

Thirteen heats plus one heat for welding ?ller metal were used to generate the Huntington data set.All the heats were solution annealed.The welds were produced using Alloy 617welding elec-trodes of four different diameters.The welds were produced using gas tungsten arc welding ?GTAW ?,a.k.a.tungsten inert gas ?TIG ?welding,and pulsed gas metal arc welding processes.

A summary of the Huntington data set is given in Table 2.It should be pointed out that Table 2may not include all the tests conducted by the Huntington alloys.Because these data were gen-erated more than two decades ago and ownership of the company has changed hands more than once,some valuable information became irretrievable.For example,the original curves of creep and tensile tests listed in Table 2,which are highly desired for constitutive equation development,have not been located.

During the years when Alloy 617was considered for the HTGR applications,the alloy was investigated by the HTGR programs at the Oak Ridge National Laboratory ?ORNL ?and General Electric Co.?GE ?in the United States.Signi?cant data applicable to the Gen IV systems were generated from these investigations ?27,45?.The ORNL-HTGR data set includes tensile data from 73tests at temperatures from 24°C to 871°C,creep data from 51tests at temperatures from 593°C to 871°C,and toughness data from 20Charpy V-notch impact tests at room temperature.It also includes the result of 1tensile test on a specimen after long-term creep testing at 871°C.A summary of the ORNL data set is given in Table 3.The tensile data include 0.2%yield strength,ultimate tensile strength,uniform strain,total strain,and reduction in

area.

Fig.13A schematic of time-temperature-transformation de-sired for Alloys 617and 230

The creep data include times to 1%,2%,and 5%total strains,time to tertiary creep,time to creep rupture,creep-rupture stress,mini-mum creep rate,loading strain,creep-rupture strain,and reduction in area.The longest creep test lasted 34,231h.The toughness data include Charpy V-notch impact energy.

Four commercial heats were used to generate the ORNL-HTGR data set.Three heats in plate form,9.5mm and 13mm thick,were used as the base metal;and one heat in wire form,1.1mm in diameter,was used as the weld ?ller metal.The welds were pro-duced using the GTAW process.

The GE-HTGR data set includes creep data from 36tests at temperatures from 750°C to 1100°C;creep-fatigue data from 7tests at 950°C;fatigue data from 40tests at temperatures of 850°C and 950°C,which consist of 27low cycle fatigue ?LCF ?tests conducted under strain control and 13high cycle fatigue ?HCF ?tests conducted under load control.A summary of the GE-HTGR data set is given in Table 4.The creep properties include the elastic and plastic components of the loading strains,times to 0.1%,0.2%,0.5%,1%,2%,and 5%total strain,time to onset of tertiary creep,time to 0.2%offset tertiary creep,minimum creep rate,total creep strain,reduction in area,creep-rupture time,and creep-rupture stress.The longest test lasted 28,920h.The fatigue and creep-fatigue properties include fatigue life data and strain-life curves.

Two commercial heats were used to generate the GE data set.The specimens were machined from product forms of 16mm

Table 2Summary of the Huntington alloy data set of Alloy 617Temperature

?°C ?Base metal Weld metal Tensile Creep

Tensile Creep

242693839362149320462260331662371342762

48235388459357264992247044827609244816328338718654392736298262942

10001610383510936314Total

129

241

50

8

Table 3Summary of ORNL data set of Alloy 617

Test type Specimen treatment Test env.Test temperature

?°C ?

25d 538593649

704760

871Tensile base metal He aging a Air 7222Inert aging b Air 11433Solution anneal Air 2c 1

11212Tensile weld e As-welded Air 1111

1He aging a Air 6

222Creep base metal

He aging a He 32

1

Inert aging b Air 2He 111Solution anneal

Air 21124He 14333Creep weld e As-welded He 33

21

1

He aging a He 1

2

Charpy base metal

He aging a Air 2Inert aging b Air 11Solution anneal

Air

5

a

Specimens were aged in simulated HTGR helium mostly for 10,000h or 20,000h at the same temperature as the test temperature,except room temperature.b

Block material was aged in steam,then the specimens were machined.c

One of the two was reported with unknown sample treatment,solution annealing was assumed.d

Test temperatures of 22°C,24°C,and 25°C were reported;all are summarized as room temperature=25°C.e

Weld represents weld metal or weldment.

Table 4Summary of GE data set of Alloy 617

Test type Specimen treatment Test env.Test temperature

?°C ?

75085095010501100Creep

Solution annealed Air 1101031

He 1

334

Low cycle fatigue He aging a

He 33Solution annealing Air 37He 110High cycle fatigue He aging a

He 33Solution annealing Air 1He 3

3Creep-fatigue

Solution annealing

Air

7

a

Aged in simulated HTGR helium for 6000h at the same temperature as the test temperature.

plate and44.5mm round bar,respectively.

Both the ORNL-HTGR and GE-HTGR data sets contain data generated from Alloy617aged and/or tested in simulated HTGR helium environments.The impurities in the simulated HTGR he-lium included N2,CO,CO2,H2,H2O,and CH4,same as those considered for the VHTR systems.However,the concentration of each impurity was different from that under consideration for the VHTR systems.Therefore,how well these data represent the im-pure helium effects in the intended Gen IV VHTR working envi-ronment needs further investigation,especially for the extended long service time and most likely nonequilibrium gaseous chemi-cal conditions of the VHTR helium coolant.

It is unfortunate that the original raw test curves for both ORNL-HTGR and GE-HTGR data sets became irretrievable in the past two decades.Only the processed data are available. When the United States was generating data on Alloy617for the HTGR project,Germany also investigated the alloy and pro-duced a large amount of data for its nuclear reactor and other programs.The data generated from these programs were later col-lected in the Online Data and Information Network?ODIN?de-veloped under the leadership of European Commission Joint Re-search Centre?JRC?.The ODIN data on Alloy617consist of three data sets,which include tensile data from302tests at tempera-tures from room temperature up to1000°C,creep data from1947 tests at the temperature range from500°C to over1000°C,CCG data from29tests at the temperature range of700–1000°C,and low cycle fatigue data from261tests at the temperature range of lower than500–1000°C.A summary of the German data is pre-sented in Table5.Access to the ODIN data on Alloy617is re-stricted.Original test curves,if not all,are stored in the database. However,the strain measurements of these creep test curves were not all conducted with?ne resolution,and may not all be ideal for constitutive equation development.

In the current Gen IV Program,Alloy617has been included for investigation.Major participants of the testing activities include several GIF member countries?46?.Preliminary tests conducted by the United States Gen IV Program include creep-fatigue data from80tests at temperatures of800°C and1000°C.Test envi-ronments included air,pure helium,and vacuum.A summary of the tests that generated the preliminary United States Gen IV data is presented in Table6.All original test records are preserved. More testing activities are planned to further investigate the po-tentials of the alloy?47,48?.

Alloy230is a relatively new material compared with Alloy 617.The material has been actively investigated for various ap-plications since its commercialization.The major known existing data of Alloy230were generated by its manufacturer Haynes International.The Haynes data set includes tensile data from36 tests at temperatures ranging from room temperature to1149°C and creep data from261tests at temperatures ranging from593°C to1149°C.The creep-rupture times ranged from15.3h to28,391 h.All tests were conducted in air environment.A summary of the Haynes data set is given in Table7.Because these are relatively new tests,the test records are well preserved.

Like Alloy617,Alloy230has also been included for investi-gation in the Gen IV Program.Major participants of the testing activities include several GIF member countries?46?.Preliminary tests conducted by the United States Gen IV Program include tensile data from24tests at temperatures from25°C to950°C; and toughness data from4Charpy tests?26?.All tests were con-ducted in air environment.A summary of the tests that generated the preliminary United States Gen IV data set is presented in Table8.All original test records are preserved.More testing ac-tivities are planned to further investigate the potentials of the alloy ?47,48?.

All the data generated in U.S.from the two alloys have been collected in a web-accessible materials property database dubbed “Gen IV Materials Handbook,”which provides powerful analyti-cal and data processing tools for materials selection,component design,and information management?49–54?.These and other data contributed from GIF members including Canada,European

Table5Summary of German data sets of Alloy617

Test type?500500–600600–700700–800800–900900–1000?1000

FZJ-COST/MATFO data set

CCG0000030 Creep0132629996436 Creep a022******* Tensile a0050620

FZJ-HTR data set

CCG000410120 LCF1512275972760 Creep058711735063260 Creep b023********* Creep a0492573410 Tensile b7024129120 Tensile c2000000 Tensile58711517430 Tensile a,b2444660

PE-TUD data set

Creep073320000 Tensile0254000

PE-THERMIE data set

CCG0002000 CCG d0002000 CCG=creep crack growth.

LCF=low cycle fatigue.

a Similar joint.

b Irradiated.

c Service exposed.

d Cast.

Union,France,Japan,Korea,South Africa,Switzerland,and the United States will be shared through this database?55,56?.

4.2ASME B&PV Codi?cation Status.Both Alloy617and Alloy230have been accepted in the ASME B&PV Code for construction under the rules of Section I for power boilers to 900°C and Section VIII Division1for un?red pressure vessels to 982°C.These construction codes list the acceptable product forms,which include forgings,plate,sheet,tube,and piping.The criteria for establishing the time-dependent allowable stresses for these two construction codes are provided in Appendix5of Code Section II-D and include67%or less of the rupture strength at 100,000h,80%of the minimum rupture strength at100,000h, and100%of the average stress to produce a creep rate of1%per 100,000h.Procedures for the assigning of the time-dependent allowable stress values in Code Table1for Alloy617and Alloy 230has been described in a background document?57?.

The criteria for setting the allowable stress intensities for Code Section III Subsection NH were described earlier in the this paper and differ signi?cantly from criteria of Section I and VIII.First and foremost,the time-dependent allowable stress intensities in NH depend on time and are not uniquely related to the100,000 strengths.Second,the three time-dependent criteria are different: one is based on67%of the minimum rupture strength,one is based on80%of the minimum strength to tertiary creep,and one is based on the average stress to produce1%total deformation.As mentioned previously,neither Alloy617nor Alloy230is covered in NH.For Alloy617,Corum and Blass?25?outlined the essential ingredients needed for a code case consistent with the Section III-NH philosophy.They summarized a draft code case and iden-ti?ed de?ciencies in the Alloy617database including weldment fatigue data,effects of VHTR environments on properties,a better understanding of the synergistic effects of aging,environment, loading,and temperature,and effects of aging on toughness.

4.3Speci?cation Modi?cation Efforts.Motivated by the un-precedented high temperature and impure helium environment re-quirements of the VHTR system,efforts have been made in the Gen IV Program to improve properties of Alloy617by modifying its speci?cation.Similar modi?cation efforts are also underway by DOE fossil energy programs.The major goals of these efforts include improved high temperature strength,improved corrosion resistance,and reduced mechanical data scatter.The DOE fossil energy programs have attempted at improved creep strength.A single heat of chemistry-controlled version of Alloy617,dubbed as CCA617,was designed and produced for investigation pur-pose.Preliminary testing has indicated moderate improvement in creep strength up to760°C.Under the Gen IV Programs,the United States established a task to re?ne the alloy within its ASTM speci?cation to reduce mechanical data scatter and to im-prove creep strength?6?;France has intended to restrict the con-centrations of Al and Ti in the alloy chemistry to improve its

Table6Summary of preliminary Gen IV data set of Alloy617

Testing parameter Base metal Weld metal Testing environment and temperature

?°C?

e ?%?t hold

?s?800air1000air1000vacuum1000pure He800air1000air

0.30322012 0.318030001 0.360221312 0.3180020001 0.3600123012 0.31800020001 0.50110000 0.80110000 10223012 118020001 160221012 1180020001 1600223012 11800020000 Table7Summary of the Haynes data set of Alloy230

Temp.

?°C?Tensile Creep Temp.

?°C?Tensile Creep

242070429 9320760261 14920816217 2042085701 26020871242 31620927261 3712096802 42720982241 48220103805 53820109305 59322114903 649212

corrosion resistance.Because of the inherent slow turnover of

metallurgical development cycles,such efforts are facing great

time pressure.

Similar efforts in improving the alloy performance have been

made for Alloy230by Haynes International.The efforts are

aimed mainly at improving corrosion resistance.Chemistry of the

ASTM speci?cation of Alloy230has been modi?ed by eliminat-

ing the minimum concentration requirement of Al.The modi?ed

speci?cation has been approved by ASTM?58?.Effects of this

modi?cation will have to be investigated.

5R&D Needs for Gen IV Reactors

The information above provides guidelines for R&D needs of

the two alloys.Limited by space,the following discussion is pre-

liminary.It is expected that new material R&D demands will con-

tinue to emerge during the reactor conceptual-design and design

process.Detailed assessment of the R&D needs should be con-

ducted as the Gen IV Program advances.

Generally speaking,more mechanical properties data must be

produced for both alloys.In this regard,Alloy617obviously has

certain advantages in its abundance of existing data.However,

data gaps exist for application in the Gen IV systems.For its creep

properties,some selected creep curves from annealed heat treat-

ment condition must be generated in air environment with high

creep strain measurement resolution to provide crucial informa-

tion for constructing constitutive equations and conducting analy-

ses to support the high temperature design methodology?HTDM?

development.Such needs also exist for Alloy230,but the interest

is mainly in acquiring more data for statistical evaluations.To

facilitate HTDM development,the investigation on creep proper-

ties should not be limited to axial loading,data on biaxial creep

properties are also highly desired.Experiments for such investi-

gations can be conducted using tension-torsion testing machines

and tube burst testing facilities.

Because data from various sources will be used for HTDM

development,data quality assurance poses a signi?cant challenge.

In generating new data,a uni?ed quality assurance plan or closely

coordinated quality assurance plans from different data contribu-

tors may be imposed.However,such measures are obviously not

viable for historical data,which were mostly generated decades

ago.This is particularly true for the historical data of Alloy617.A

survey reveals that considerable data scatter exist in the creep data

generated in the German HTGR program?6?.For example,under 13.5MPa at950°C in helium,creep strains of three different

heats with similar grain sizes scattered from less than1%to more than8%at20,000h?59?;at850°C under about50MPa the time

to1%creep strain scattered from approximately100–10,000h,

and generally the scatter band widths for the1%creep strain limit resulting from tests at800–1000°C were greater than?30%?60?.Such scatter could have originated from various sources,and will inevitably impose large knock down factors on allowable

stresses,leaving very limited room for design,especially at very

high temperatures where the material strength is already low.At

the present time,it is obviously impractical to identify all the

sources for this“historical”scatter.However,efforts may still be

possible to exclude some contributory factors.For example,qual-

ity assurance plans under which these data were generated could be reviewed.Also,if newly generated data no longer exhibit such considerable scatter,improvements in steel manufacturing tech-nologies may be investigated to verify the cause,assuming both historical and present data have met strict nuclear program quality assurance requirements for data generation;and then a technically sound judgment can be made in using the historical data. Another signi?cant challenge is the very low available strength of the alloys at the desired service temperature range.A review of Code Case N-47-28indicates that at927°C the expected mini-mum stress-to-rupture value of Alloy617for100,000h is12MPa ?1.75ksi?,and after a safety margin to cover the data scatter the allowable stress intensity is only7MPa?0.98ksi??13?.Note that

100,000h is less than1/5of the expected design life of60years.

Some efforts were made with preliminary success to re?ne the

Alloy167standard speci?cation,aimed at reducing data scatter

and improving high temperature strength?6,61,62?.Although the

timeframe for demonstrating the?rst prototype Gen IV reactor ?year2017–2021?is not in favor of such time-consuming en-deavor,for the long-term bene?ts of Gen IV reactors after the?rst prototype the efforts should still be continued to reach a de?nitive conclusion.

In the intended operating temperature range,tensile response of

Alloy617becomes very sensitive to the loading rate?25?.Testing

is needed to verify such behavior in Alloy230.Loading rate sen-

sitivity must be characterized and factored into the HTDM devel-

opment.Due to massive plastic?ow and large elongation at very

high temperature,the testing faces signi?cant challenges in accu-

rately capturing the true stress and strain evolution process.Novel

testing techniques,such as video recording of specimen shape

variation,plastic surface strain mapping,or multiple extensom-

etries each covering a given elongation range,must be developed

to acquire reliable data for model development.

Limited LCF data exist for Alloy617,and none has been col-

lected for Alloy230in the Gen IV Program.Obviously more LCF

data must be generated for design against cyclic damage during

reactor operational?uctuations such as heating up and cooling

down.As for creep properties,data on biaxial fatigue properties

should also be generated to facilitate HTDM development.In the

GE data set,as shown in Table4,a few HCF data were included.

Compared with the LCF damage,the HCF damage may be less of

a problem in reactor systems.The HCF damage in reactor sys-

tems,if any,may occur from vibrations of some components

when coolant?ows through at high speed.The HCF damage usu-

ally spends a long time in crack initiation and grows to a detect-

able size,but because of the high frequency that normally comes

with the HCF,a very short time is needed for a crack of critical

size to propagate to rupture.In the FCC structure of Alloys617

and230,the HCF damage develops mainly from LCF and plastic

deformation damages,which signi?cantly shorten the time needed

for the HCF crack initiation?63?.Therefore,design against LCF

and plastic deformation damages can largely guard against the

occurrence of HCF rupture.However,it is in the best interest of

the Gen IV reactor safety assurance that the HCF resistance of the

two alloys are investigated and compared during the materials

selection.

Compared with fatigue properties of the two alloys,the resis-

tance to creep-fatigue damage should be more of a concern for

Gen IV nuclear reactor applications.Interactions between the

creep and fatigue mechanisms usually accelerates the damaging

process in materials.Very limited creep-fatigue investigations

were conducted in the past on creep-fatigue effects on the two

alloys?26,64–66?.Moreover,as previously mentioned,because

the creep-fatigue damage model currently used in ASME B&PV

Code has not been satisfactorily representing the true material

behavior,future investigation should not be limited to generating

more creep-fatigue data for both alloys,but should develop new

approaches that may allow for variations in design to lead to dif-

ferent allowable limits based upon enhanced modeling tools.

During the60years of design life,initiation and growth of

Table8Summary of the preliminary Gen IV data set of Alloy 230

Temperature

?°C?Charpy Tensile

2546

75006

85006

95006

creep cracks and creep-fatigue cracks in certain high stress con-centration locations or pre-existing material defects in reactor components are almost inevitable.Understanding of crack initia-tion and growth of the structural materials will become extremely important under such circumstances for life prediction,accident prevention,and determination of inspection intervals.So far,very limited crack growth data are known to exist for Alloy 617and none for Alloy 230.Both creep crack and creep-fatigue crack data of the two alloys should be generated for materials selection and future component life prediction.Because some components may be designed to operate in the creep regime,for these two alloys that exhibit early tertiary creep behavior and for the intended nuclear application that prohibits extensive creep conditions,cor-rectly characterizing the creep crack and creep-fatigue crack be-haviors poses a signi?cant challenge.Creep conditions for time-dependent fracture mechanics are summarized in Table 9?67?.The aforementioned popular parameters C ?and C ??t ?,which are by and large technically mature,are for creep crack growth under extensive secondary creep conditions and extensive primary and/or secondary creep conditions,and cover the regions of EC-SC and EC-PC/SC in Table 9,respectively.For Gen IV reac-tor systems,if design in creep regime is allowed,it is most likely to be restricted to the SSC or TC regions;which of the PC,SC,and TC regions are allowed may still need to be determined,de-spite the early exhibition of tertiary creep behavior of the two alloys.Although C ?and C ??t ?can provide comparative informa-tion on creep crack resistance for material selection between the two alloys,adequate parameters and methods will need to be de-veloped to characterize the crack behavior in the regions of Table 9that represent the Gen IV reactor operation conditions.Based on the experiences in the past few decades,development in this area is not an easy task,and dedicated new tasks must be identi?ed and established.

It is envisioned that the impure helium coolant at 950°C in VHTR will be ?owing at velocities on the order of 40m/s and higher;and most of the core components will be made of graphite materials ?68?.A possibility exists that some small particles may break off from the graphite and become circulating at high veloci-ties in the coolant ?ow.When these particles hit metal compo-nents at high speed,foreign object damage ?FOD ?could be in-?icted on the metal.To guard against serious property deterioration due to FOD,notch sensitivity and FOD resistance of the two alloys may need to be investigated.Furthermore,small FOD,if repetitive,can also become erosion damage unless such mechanisms can be eliminated through considerations in graphite selection.

In the known existing data of both alloys,very little informa-tion on toughness is available.Although both alloys are intended for very high temperature applications,toughness properties are still of interest for reactor shutdowns.This is especially of a con-cern after the alloys have been exposed to the severe VHTR op-eration environments.Furthermore,because the environment will not only affect toughness,its effects on most of the mechanical properties discussed above should also be investigated.The major environmental effects that should be studied include that of im-pure helium,thermal aging,and radiation unless use of the alloys in radiation region is completely excluded.

All the R&D needs discussed so far are for the base metal of the two alloys.Similar studies should also be conducted on the welds.Although both alloys have fairly good weldability,R&D activities are still needed for special welding methods,especially the welding methods that are required for joining particular prod-uct forms such as very thin sheets for compact intermediate heat exchanger and thick sections for tube-and-shell or helical heat exchanger designs.

In the late phase of the materials testing,some of the mechani-cal properties should be studied on product forms that are needed for reactor construction.Effects of grain size,heat treatment,cold work,etc.should also be investigated.While signi?cantly more data regarding larger grain sizes and product forms for tube-and-shell heat exchangers are desired,efforts should also be made to characterize ?ner grain sizes and thinner product forms to assist materials and heat exchanger designs selection and to facilitate codi?cation.For example,limited test data generated by ORNL indicate that thin sheets have much lower creep resistance than the plate form,as shown in Fig.14.If the compact heat exchanger concept with very thin walls between the primary and secondary loops is considered,creep properties of thin sheets must be exten-sively investigated.The signi?cant difference in creep behavior shown in Fig.14results from the microstructures in the thin prod-uct forms,mainly the average grain sizes.For the compact heat exchanger concept,not only creep and relaxation data from thin sheets must be generated for the HTDM development,R&D must also be conducted to investigate optimization of processing pa-rameters for fabricating thin sheets and foils to achieve the desired creep properties.Some reference in this regard can be found in Refs.?69,70?.

Although metallurgical information on both alloys exist in fair amount,further investigations are still necessary.The metallurgi-cal studies should be focused on material stability,especially on the evolution process of strengthening mechanisms after exposure to simulated reactor working environments including irradiation,thermal aging,and impure helium contact.For safe long-term ser-vices,the further investigations should provide data to reconcile and/or to clarify the controversial ?ndings in Alloy 617hitherto reported by different researchers ?5,6,29–34?.The studies should also provide information from accelerated testing for metallurgi-cally based materials long-term behavior modeling.

Table 9Summary of the conditions in time-dependent fracture mechanics

Scale of creep

Level of load

Stage of creep deformation Small-scale creep ?SSC ?Elastic loading ?EL ?Primary creep ?PC ?Transition creep ?TC ?Elastic-plastic loading ?EPL ?

Secondary creep ?SC ?Extensive creep ?EC ?

Plastic loading ?PL ?

Tertiary creep ?TC ?

Fig.14Signi?cant difference is observed between creep re-sistance of tube and foil product forms of Alloy 230

Acknowledgment

This work is sponsored by the U.S.Department of Energy,Of-?ce of Nuclear Energy Science and Technology under Contract No.DE-AC05-00OR22725with Oak Ridge National Laboratory, managed by UT-Battelle,LLC.

References

?1?Haynes International,Inc.,2004,“Haynes25Alloy,”High-Temperature Al-loys Brochure,Paper No.H-3057D.

?2?Haynes International,Inc.,1991,“Haynes25Alloy,”High-Temperature Al-loys Brochure,Paper No.H-3001B.

?3?Special Metals Corporation,2005,“Inconel Alloy617,”Paper No.SMC-029.?4?Haynes International,Inc.,2007,“Haynes230Alloy,”High-Temperature Al-loys Brochure,Paper No.H-3000H.

?5?Ren,W.,and Swindeman,R.,2006,“A Review of Aging Effects in Alloy617 for Gen IV Nuclear Reactor Applications,”Proceedings of the2006ASME Pressure Vessels and Piping Division Conference,PVP2006-ICPVT11–93128, July23–27,Vancouver,BC Canada.

?6?Ren,W.,and Swindeman,R.,2005,“Development of A Controlled Material Speci?cation for Alloy617for Nuclear Applications,”U.S.Department of Energy,Report No.ORNL/TM-2005/504.

?7?Brown,E.E.,and Muzyka,D.R.,1987,Superalloys II,Wiley,New York,p.

165.

?8?Holt,R.T.,and Wallace,W.,1976,“Impurities and Trace Elements in Nickel-Base Superalloys,”Int.Met.Rev.,203,pp.1–24.

?9?Ross,E.W.,and Sims,C.T.,1987,Superalloys II,Wiley,New York,p.97.?10?Shogan,R.P.,1971,“Neutron Irradiation Effects on the Tensile Properties of Inconel718,Waspaloy and A286,”Westinghouse Electric Corporation,Paper No.WANL-TME-2791.

?11?Bajaj,R.,Mills,W.J.,Lebo,M.R.,Hyatt,B.Z.,and Burke,M.G.,1995,“Irradiation-Assisted Stress Corrosion Cracking of HTH Alloy X-750and Al-loy625,”International Symposium on Environmental Degradation of Materi-als in Nuclear Power Plants:Water Reactors,Breckenridge,CO,Aug.6–10.?12?American Society of Mechanical Engineers,2004,“ASME Boiler and Pres-sure Vessel Code,an International Code,”ASME,New York,NY.

?13?Task Force Very High-Temperature Design,1988,“Draft Alloy617Code Case,”Case N-47–28,Class1Components in Elevated Temperature Service, Section III,Division1,Jul.27.

?14?Schubert,F.,Breitbach,G.,and Nickel,H.,1993,“German Structural Design Rule KTA3221for Metallic HTR-Components,”High-Temperature Service and Time-Dependent Failure,PVP-V ol.262,ASME,New York,pp.9–18.?15?American Society of Mechanical Engineers,“Quality Assurance Requirements for Nuclear Facility Applications,”American National Standard,Paper No.

ASME NQA-1-2000.

?16?Viswanathan,R.,Coleman,K.,Shingledecker,J.,Sarver,J.,Stanko,G.,Bor-den,M.,Mohn,W.,Goodstine,S.,and Perrin,I.,2006,“Boiler Materials for Ultrasupercritical Coal Power Plants,”U.S.Department of Energy,Report No.

DE-FG26-01NT41175.

?17?Viswanathan,R.,Henry,J.F.,Tanzosh,J.,Stanko,G.,Shingledecker,J.,and Vitalis,B.,2005,“U.S.Program on Materials Technology for Ultrasupercriti-cal Coal Power Plants,”ECCC Creep Conference,London,Sept.12–14.

?18?1974,“Nuclear Vessels in High Temperature Service,”Code Case1592,Sec-tion III,ASME Boiler Pressure Vessels Code,ASME,New York.

?19?Robinson,E.,1952,“Effect of Temperature Variation on the Long-Time Rup-ture Strength of Steels,”Trans.ASME,74?5?,pp.777–780.

?20?Miner,M.A.,1945,“Cumulative Damage in Fatigue,”ASME J.Appl.Mech., 12?3?,pp.A159–A167.

?21?Taira,S.,1962,“Life Time of Structures Subjected to Varying Load and Tem-perature,”Creep in Structures,Academic,New York.

?22?Palmgren,A.,1924,“Die Lebensdauer von Kugellagern,”Z.Ver.Dtsch.Ing., 68,Heft14,pp.339–341.

?23?McGreevy,T.E.,and Jetter,R.I.,2006,“DOE-ASME Generation IV Mate-rials Tasks,”Proceedings of ASME2006Pressure Vessels and Piping Division Conference,PVP2006-ICPVT-11,July23–27,Vancouver,BC Canada.

?24?McGreevy,T.E.,and Abou-Hanna,J.,2007,“Applicability of Simpli?ed Methods to Alloy617in Excess of649°C,”Proceedings of ASME2006Pres-sure Vessels and Piping Division Conference,PVP2007-26672,July22–26, San Antonio,TX.

?25?Corum,J.M.,and Blass,J.J.,1991,“Rules For Design of Alloy617Nuclear Components to Very High Temperatures,”PVP?Am.Soc.Mech.Eng.?,215, pp.147–153.

?26?Ren,W.,Totemeier,T.,Santella,M.,Battiste,R.,and Clark,D.,2006,“Status of Testing and Characterization of CMS Alloy617and Alloy230,”U.S.De-partment of Energy,Report No.ORNL/TM-2006-547.

?27?McCoy,H.E.,and King,J.F.,1985,“Mechanical Properties of Inconel617 and618,”U.S.Department of Energy,Report No.ORNL/TM-9337.

?28?Swindeman,R.W.,and Schubert,F.,private communication.

?29?Kirchhofer,H.,Schubert,F.,and Nickel,H.,1984,“Precipitation Behavior of Ni-Cr-22Fe-18Mo?HASTELLOY X?and Ni-Cr-2Co-12Mo?INCONEL617?After Isothermal Aging,”Nucl.Technol.,66,pp.139–148.

?30?Santella,M.L.,and Ren,W.,2005,Oak Ridge National Laboratory,private communication.

?31?Mankins,W.L.,Hosier,J.C.,and Bassford,T.H.,1974,“Microstructure and Phase Stability of INCONEL Alloy617,”Metall.Trans.,5,pp.2579–2590.?32?Kimball,G.F.,Lai,G.Y.,and Reynolds,G.H.,1976,“Effects of Thermal Aging on Microstructural and Mechanical Properties of a Commercial Ni-Cr-Co-Mo Alloy?Inconel617?,”Metall.Trans.A,7A,pp.1951–1952.

?33?Kihara,S.,Newkirk,J.B.,Ohtomo,A.,and Saiga,Y.,1980,“Morphological Changes of Carbides During Creep and Their Effects on the Creep Properties of Inconel617at1000C,”Metall.Trans.A,11A,pp.1019–1031.

?34?Wu,Q.,and Vasudevan,V.K.,2004,“Characterization of Boiler Materials for Ultracritical Coal Power Plants,”Annual Progress Report for Period Aug.1, 2002to July30,2003,under UT-Battelle Sub Contract Number4000017043.?35?Mansur,L.K.,1987,“Mechanisms of Kinetics of Radiation Effects in Metals and Alloys,”Kinetics of Nonhomogeneous Processes:A Practical Introduction for Chemists,Biologists,Physicists,and Materials Scientists,G.R.Freeman, ed.,Wiley,New York,p.377–463.

?36?Shewmon,P.G.,1971,“Radiation-Induced Swelling of Stainless Steel,”Sci-ence,173?4001?,pp.987–991.

?37?Lewthwaite,G.W.,Mosedale,D.,and Ward,I.R.,1967,“Irradiation Creep in Several Metals and Alloys at100°C,”Nature?London?,216,pp.472–473.?38?Harries,D.R.,1966,“Neutron Irradiation Embrittlement of Austenitic Stain-less Steels and Nickel Base Alloys,”J.Br.Nucl.Energy Soc.,5?1–4?,pp.

74–87.

?39?Harries,D.R.,1979,“Neutron Irradiation-Induced Embrittlement in Type316 and Other Austenitic Steels and Alloys,”J.Nucl.Mater.,82,pp.2–21.

?40?Mansur,L.K.,and Grossbeck,M.L.,1988,“Mechanical Property Changes Induced in Structural Alloys by Neutron Irradiations With Different Helium to Displacement Ratios,”J.Nucl.Mater.,155–157,pp.130–147.

?41?Klueh,R.L.,and Vitek,J.M.,1983,“The Resistance of9Cr-1MoVNb and 12Cr-1MoVW Steel to Helium Embrittlement,”J.Nucl.Mater.,117,pp.295–302.

?42?Bloom,E.E.,1976,“Irridiation Strengthening and Embrittlement,”Radiation Damage in Metals,S.D.Hackneys and N.L.Petersen,eds.,American Society for Metals,Metals Park,OH,pp.295–392.

?43?Boothby,R.M.,1990,“Modeling Grain Boundary Cavity Growth in Irridiated Nimonic PE16,”J.Nucl.Mater.,171,pp.215–222.

?44?Huntington Alloys,Inc.,1983,“Inconel Alloy617Stress Rupture and Tensile Data Package.”

?45?Baldwin,D.H.,Kimball,O.F.,and Williams,R.A.,1986,“Design Data for Reference Alloys:Inconel617and Alloy800H,”General Electric Company for the Department of Energy,Contract No.DE-AC03-80ET34034.

?46?2007,“Generation IV International Forum Very High Temperature Reactor System Materials Project–Project Plan,”Revision1.02,Aug.13.

?47?Ren,W.,and Swindeman,R.W.,2004,“High Temperature Metallic Materials Test Plan for Generation IV Nuclear Reactors,”U.S.Department of Energy, Report No.ORNL/TM-2005/507.

?48?2008,“Next Generation Nuclear Plant Intermediate Heat Exchanger Materials Research and Development Plan,”Revision0,Apr.,Report No.PLN-2804.?49?Rittenhouse,P.,and Ren,W.,2005,“Gen IV Materials Handbook Implemen-tation Plan,”U.S.Department of Energy,Report No.ORNL/TM-2005/77.?50?Ren,W.,and Swindeman,R.W.,2005,“Assessment of Existing Alloy617 Data for Gen IV Materials Handbook,”U.S.Department of Energy,Report No.ORNL/TM-2005/510.

?51?Ren,W.,and Rittenhouse,P.,2005,“Construction of Web-Accessible Materi-als Handbook for Generation IV Nuclear Reactors,”Proceedings of ASME 2005Pressure Vessels and Piping Division Conference,PVP2005-71780,July 17–21,Denver,CO.

?52?Ren,W.,and Rittenhouse,P.,2005,“Initial Development of the Gen IV Ma-terials Handbook,”U.S.Department of Energy,Report No.ORNL-GEN4/ LTR-05/012.

?53?Ren,W.,2006,“Gen IV Materials Handbook Architecture and System De-sign,”U.S.Department of Energy,Report No.ORNL-GEN4/LTR-06–004.?54?Ren,W.,and Luttrell,C.,2006,“Gen IV Materials Handbook Beta Release for Structural and Functional Evaluation,”U.S.Department of Energy,Report No.

ORNL/GEN4/LTR-06–027.

?55?Ren,W.,2008,“Development of Digital Materials Database for Design and Construction of New Power Plants,”Proceedings of the2008International Congress on Advances in Nuclear Power Plants(ICAPP‘08),American Nuclear Society Annual Meeting,Anaheim,CA,Jun.8–12,Paper No.8019.?56?Ren,W.,2008,“Summary Report on FY2008Gen IV Materials Handbook Development,”U.S.Department of Energy,Report No.ORNL/TM-2008/106.?57?Prager,M.,1996,“Proposed Implementation of Criteria for the Assignment of Allowable Stresses High in the Creep Range,”PVP?Am.Soc.Mech.Eng.?, 336,pp.273–293.

?58?American Society for Testing and Materials,2006,“Standard Speci?cation for UNS N06002,UNS N06230,UNS N12160,and UNS R30556Plate,Sheet, and Strip,”Annual Book of ASTM Standards,V ol.02.04.

?59?Ennis,F.J.,Mohr,K.P.,and Schuster,H.,1984,“Effect of Carburizing Ser-vice Environments on the Mechanical Properties of High-Temperature Al-loys,”Nucl.Technol.,66,pp.363–368.

?60?Schubert,F.,Bruch,U.,Cook,R.,Diehl,H.,Ennis,P.L.,Jakobeit,W.,Pen-kalla,H.J.,Heesen,E.,and Ullrich,G.,1984,“Creep Rupture Behavior of Candidate Materials for Nuclear Process Heat Applications,”Nucl.Technol., 66,pp.227–240.

?61?Totemeier,T.,and Ren,W.,2006,“Procurement and Initial Characterization of Alloy230and CMS Alloy617,”U.S.Department of Energy,Report No.

INL/EXT-06-11290.

?62?Shingledecker,J.P.,Swindeman,R.W.,Wu,Q.,and Vasudevan,V.K.,2004,“Creep Strength of High-Temperature Alloys for Ultrasupercritical Steam Boilers,”Proceedings of the Fourth International Conference on Advances in

Materials Technology for Fossil Power Plants,Hilton Head,SC,Oct.25–28.?63?Ren,W.,and Nicholas,T.,2002,“Effects and Mechanisms of Low Cycle Fatigue and Plastic Deformation on High Cycle Fatigue Limit Reduction in Nickel-Base Superalloy Udimet720,”Mater.Sci.Eng.,A,332,pp.236–248.?64?Penkalla,H.J.,Nickel,H.,and Schubert,F.,1989,“Multiaxial Creep of Tubes From Incoloy800H And Inconel617Under Static and Cyclic Loading Con-ditions,”Nucl.Eng.Des.,112,pp.279–289.

?65?Penkalla,H.J.,Schubert,F.,and Nickel,H.,1993,“Description Of Creep and Fatigue Exposed Tubes of Alloy617,”PVP?Am.Soc.Mech.Eng.?,262,pp.

221–224.

?66?Totemeier,T.C.,2007,“High-Temperature Creep-Fatigue of Alloy617Base Metal and Weldments,”Proceedings of ASME2007Pressure Vessels and Pip-ing Division Conference,PVP2007-26372,July22–26,San Antonio,TX.

?67?Ren,W.,1995,“Time-Dependent Fracture Mechanics Characterization of Haynes HR160Superalloy,”Ph.D.thesis,University of Tennessee,Knoxville.?68?Hayner,G.O.,Shaber,E.L.,Mizia,R.E.,Bratton,R.L.,Sowder,W.K., Wright,R.N.,Windes,W.E.,Totemeier,T.C.,Moore,K.A.,Corwin,W.R., Burchell,T.D.,Corum,J.M.,Klett,J.W.,Nanstad,R.K.,Snead,L.L., Rittenhouse,P.L.,Swindeman,R.W.,Wilson,D.F.,McGreevy,T.E.,Jones, R.,and Gardner,F.,2004,“Next Generation Nuclear Plant Materials Research and Development Program Plan,”U.S.Department of Energy,Report No.

INEEL/EXT-04-02347.

?69?Maziasz,P.J.,Pint,B.A.,Shingledecker,J.P.,Evans,N.D.,Yamamoto,Y., More,K.L.,and Lara-Curzio,E.,2006,“Advanced Alloys for Compact,High-Ef?ciency,High-Temperature Heat-Exchangers,”Int.J.Hydrogen Energy, 32?16?,pp.3622–3630.

?70?Maziasz,P.J.,Shingledecker,J.P.,Evans,N.D.,Yamamoto,Y.,More,K.L., Trejo,R.,and Lara-Curzio,E.,2007,“Creep Strength and Microstructure of AL20–25+Nb Alloy Sheets and Foils for Advanced Microturbine Recupera-tors,”ASME J.Eng.Gas Turbines Power,129?3?,pp.798–805.

法律保留原则

法律保留原则 法律保留的思想产生于19世纪初，最早提出该概念的是德国行政法学之父奥托·迈耶。根据迈耶的经典定义，法律保留是指在特定范围内对行政自行作用的排除。因此，法律保留本质上决定着立法权与行政权的界限，从而也决定着行政自主性的大小。 目录 一、什么是法律保留？ （一）法律保留与宪法保留 （二）法律保留与行政保留 （三）法律保留中的“法律”为何？ 1二、为什么要法律保留？总论 1（一）功能结构理论 1（二）法律保留在我国宪法上的依据 三、保留什么？ 1（一）重要性保留总论 11、基本权重要性的标准 12、公共事务重要性的标准 13、消极标准 1总论 11、积极标准 12、消极标准 11、在法律保留的范围之外 12、在法律保留的范围之内 13、行政立法不作为 1总论 11、立法不作为 12、授权明确性 1（一）有关一般保留标准的规定 1（二）有关绝对保留标准的规定 1（三）有关授权明确性的规定

展开 一、什么是法律保留？ （一）法律保留与宪法保留 （二）法律保留与行政保留 （三）法律保留中的“法律”为何？ 1二、为什么要法律保留？总论 1（一）功能结构理论 1（二）法律保留在我国宪法上的依据 三、保留什么？ 1（一）重要性保留总论 11、基本权重要性的标准 12、公共事务重要性的标准 13、消极标准 1总论 11、积极标准 12、消极标准 11、在法律保留的范围之外 12、在法律保留的范围之内 13、行政立法不作为 1总论 11、立法不作为 12、授权明确性 1（一）有关一般保留标准的规定 1（二）有关绝对保留标准的规定 1（三）有关授权明确性的规定 一、什么是法律保留？ 是指宪法关于人民基本权利限制等专属立法事项，必须由立法机关通过法律规定，行政机关不得代为规定，行政机关实施任何行政行为皆必须有法律授权，否则，其合法性将受到质疑。行政行为的做出必须有法律依据，法律没规定的行政主体不得擅自做出行政行为。 （一）法律保留与宪法保留

公共秩序保留制度在我国的运用

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9．动物先天性行为的操纵因素是〔〕A．环境因素 B．后天〝学习〞所得 C．遗传物质 D．亲代训练 10．以下哪种行为属于动物的学习行为〔〕A．刚出生的小袋鼠爬到母袋鼠的育儿袋中吃奶 B．小鸟在池边喂鱼 C．蜜蜂筑巢 D．小鸡绕道取食 11．关于动物先天性行为的表达，错误的选项是〔〕A见于所有动物 B生来就有的行为 C是可遗传的 D在同种动物的不同个性间有不同表现 12．专门多年前，在英格兰有一只大山雀，一次碰巧打开了放在门外的奶瓶偷喝了牛奶，不久，那儿的其他大山雀也学会了偷喝牛奶；幼小的黑猩猩能仿照成年黑猩猩，会利用一根蘸水的树枝从洞穴中钓取白蚁作为食物；成年黑猩猩会利用体会来解决咨询题，当香蕉被挂在高处，徒手拿不到时，黑猩猩会把几个木箱子叠起来，然后爬到木箱顶上去摘香蕉，这些例子能够讲明〔〕 A．不同动物的学习能力是有差不的 B．动物越高等，学习能力越强 C．动物越高等，适应复杂环境的能力也越强 D．上述三个选项都正确 13．实验人员想了解新孵化的鸟类是如何样获得对它们母亲的情感的。他们选择了6只刚孵化的小鸭。小鸭①、②、③刚孵化出来就被从它们的出生地移走了，使它们没有见到自己的母亲。小鸭④、⑤、⑥还和它们的母亲在一起。实验人员给小鸭①、②、③展现了一个气球，同时放母鸭叫的录音，通过假设干天后，实验人员发觉小鸭①、②、③仿照母鸭的行为并随时跟随在气球左右。请选择： ⑴小鸭①、②、③的行为进展过程能够用以下哪个术语描述: ( ) A先天行为 B条件反射 C本能 D学习行为 ⑵小鸭④、⑤、⑥最好被称为 ( ) A实验模型 B对比组 C 实验组 D一个变量 ⑶假如不是把小鸭①、②、③放在气球周围并放母鸭叫的录音，而是将它们放在一只猫的周围并放母鸭叫的录音，那么小鸭①、②、③： ( ) A它们可能恐吓其他鸭子 B它们将失去繁育能力 C它们会发出类似猫的声音 D它们会认为猫是它们的母亲 参考答案 1.C 2.D 3.C 4.D 5.C 6.A 7.A 8.D 9.C 10.D 11.D 12.D 13.(1)D (2)B (3)D

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论国际私法中的公共秩序保留原则引言 在全球化的大背景下，每个国家要获得发展，都必须把本国置于国际社会当中。随着国际社会交流的加强，从而使以涉外民商事关系为调整对象的国际私法的应用范围越来越广。作为国际私法当中一个主要制度——公共秩序保留制度，在解决国际法律冲突中发挥着重要作用。公共秩序保留制度无论是在理论上还是在实践上，都得到了国际社会的普遍认可。但就公共秩序的内涵、本质、以及在什么情况下可以应用公共秩序保留的相关条款，各国理论界看法不一，在实际操作上也不尽相同。因此，探索出符合国情的公共秩序保留理论尤为重要。本文在介绍公共秩序保留的相关理论时，将从公共秩序保留的含义、本质、意义、作用出发，详尽公共秩序保留理论的发展趋势及各国对公共秩序保留原则合理的限制适用情形，并从我国立法和司法两方面介绍公共秩序保留的实践，反映此制度在我国的发展状态。在此基础上，对我国公共秩序保留制度提出一些意见和建议，这些对公共秩序的探索和研究，可以视为国内学者研究成果的一点体现。 一、公共秩序保留的概述 （一）公共秩序保留原则的含义 公共秩序保留原则，作为我国现行国际私法的一项基本原则，乃是对所有可导致外国法适用的双边冲突规则的例外规定。 公共秩序保留，在英美法系中被称为“公共政策”；法语中称为“公共秩序”、在德语中称“保留条款”；而在我国大陆地区的法律

规定中，则用“国家和社会的公共利益”来表述。关于这一原则的经典定义出自卡多佐之口。1918年在审理Loucks V.standiard Oilco 案中,他首次较为完整的提出：“法院不应对外国法的适用闭上大门，除非适用该外国法，将会与正义的重大原则、道德的基本观念或使馆大众福祉的传统相抵触。”我国学者将这一原则定义为：“一国法院依冲突规范应该适用外国法时；或者依法应该承认与执行外国法院判决或仲裁裁决时；或者依法应该提供司法协助时；因这种适用，承认与执行或者提供司法协助会与法院地国的重大利益、基本政策、法律的基本原则或道德的基本观念相抵触，而有权排除和拒绝的保留制度。[1] 公共秩序保留原则对于维护法院地国的道德传统、社会秩序和根本利益起着重要作用。该原则的最大特点是灵活性与不确定性。一方面它有利于法官根据本国利益的需要，随机应变地适用之；另一方面由于没有一个确定的、统一的概念与适用标准，导致这一原则在实践中易被滥用。所以有学者无不担忧地说：“公共秩序保留好比一匹性格暴虐的烈马，一旦骑上它，就会失去控制。不知会被带向何方。”[2]同时，国际私法中的识别、反致、法律规避、直接适用的发等制度也能在一定程度上排除外国法，以至有人提出公共秩序保留原则作用不大。因此，我们很有必要进一步剖析公共秩序保留原则。 （二）公共秩序的含义 国际私法上的公共秩序与国内民法上的公共秩序是不同的，较

色谱分析教案

色谱分析教案

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第一节色谱分析法概述 一、色谱法的定义及用途 1定义： 色谱分析法简称为色谱法或称层析法(chromatography)，是一种物理或物理化学的分离分析方法。其分离原理是利用物质在固定相和流动相中的分配系数不同，使混合物中的各组分离。 回顾一下以前学过的分离方法 沉淀法：蒸馏法：萃取法： 2色谱法的用途： 色谱法已广泛用于各个领域，成为多组分混合物的最重要的分离分析方法。因此，对于复杂的药物分析的，首选是色谱方法，在各国药典中都大量收载色谱法。现在色谱法形成一门专门的科学，广泛的用于各个领域，成为多组分混合物的最重要的分离分析方法。 二、色谱法的起源 1．创立： 1906 年，俄国植物学家Tsweet 发现色谱分离现象（植物色素分离见图示）。在碳酸钙上出现了具有不同颜色的色带，并将由此色带组成的谱图命名为“色谱图”。

10 2．现状：一种重要的分离、分析技术，分离混合物各组分并加以分析固定相——除了固体，还可以是液体 流动相——液体或气体 色谱柱——各种材质和尺寸 被分离组分——不再仅局限于有色物质教学要求、方法及时间分配 三、色谱法的特点 1 优点： “三高”、“一快”、“一广” 2 缺点： 四、色谱法的分类 1．按两相分子的聚集状态分类 2．按固定相的固定方式分类 3 按分离机制分类 色谱法简单分类： 五、色谱法的发展 1、历史 2、在我国的发展 3、展望 第二节色谱过程和基本术语

一、色谱过程、分离原理及特点 ㈠色谱过程 ㈡色谱分离原理 ㈢色谱分离特点 1．不同组分通过色谱柱时的迁移速度不等→提供了分离的可能性2．各组分沿柱子扩散分布→峰宽↑→不利于不同组分分离。 二、色谱流出曲线和有关概念 ㈠色谱流出曲线和色谱峰 1.流出曲线(色谱图)：电信号强度随时间变化曲线 2.基线：无样品时的电信号 （二）保留值：色谱定性参数 1. 保留时间tR： 2. 死时间tm(或t0)： 3.调整保留时间tR'： 4.保留体积VR： 5.死体积Vm (或V0)： 6.调整保留体积VR'： 7.相对保留值ris(选择性系数α)：调整保留值之比 （三）色谱峰高和峰面积 1.峰高h： 2.峰面积A： （四）色谱区域宽度： 1.标准差σ： ： 2.半峰宽W 1/2 3.峰宽W ： 4. 三、分配系数与色谱分离 ㈠分配系数和容量因子：相平衡参数 1.分配系数K(平衡常数)： 2..容量因子k (容量比，分配比)： ㈡分配系数和容量因子与保留时间的关系 第三节色谱基本原理