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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS Weiju Ren and Robert W. Swindeman Oak Ridge National Laboratory November 30, 2004

ORNL/TM-2004/507

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS

Weiju Ren and Robert W. Swindeman

Date Published – November 30, 2004

Prepared for Office of Nuclear Energy Science and Technology AF3410000

Prepared by OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831-6285 managed by UT-BATTELLE, LLC for the U.S. DEPARTMENT OF ENERGY Under contract DE-AC05-00OR22725

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ ABTRACT Activities for acquiring metallic materials property data are planned to support the design and construction of the Generation IV (Gen IV) nuclear reactors with the focus on the Next Generation Nuclear Plant (NGNP) reactor system. Testing priorities are given to the candidate materials that are the most desirable and have near term feasibility for use in the reactors. Material properties required for the design and construction are discussed through a review of the ASME B&PV Code. Possible data needs for the candidate materials are identified by comparing the requirements with existing data reviewed and/or present knowledge of the materials. Guidelines are discussed for further assessment of existing data. Analysis and processing for data derivation are briefed for better understanding of data needs. Requirements for NRC licensing and ASME quality assurance are considered, and testing method requirements are described. Guidelines and helpful information are provided for developing test matrices of priority 1 materials. Coordination within the data acquisition activities and with other tasks of the Gen IV materials project is discussed.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________

CONTENTS 1. INTRODUCTION ------------------------------------------------------------------------------ 1 1.1 1.2 1.3 1.4 1.5

Background---------------------------------------------------------------------------------- 1 Scope ---------------------------------------------------------------------------------------- 3 Assessment of Existing Data ------------------------------------------------------------- 4 Coordination and Plan Change Control ------------------------------------------------- 5 Unit System --------------------------------------------------------------------------------- 6

2. BASES FOR THE SELECTION OF MATERIALS --------------------------------------- 7 2.1 General Requirements --------------------------------------------------------------------- 7 2.2 Grouping and Priority --------------------------------------------------------------------- 9 3. CLASSIFICATION OF PROPERTIES REQUIREMENTS---------------------------- 11 3.1

ASME Section III, Division 1, Subsection NH - Class 1 Components in Elevated Temperature Service -------------------------------------------------------------------- 11

3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.1.9 3.1.10 3.1.11

Cold forming limits---------------------------------------------------------------- 11 Tensile properties as a function of temperature ------------------------------- 13 Ultimate tensile and yield strength reduction factors for aging ------------- 14 Stress intensity values as a function of time and temperature --------------- 15 Minimum stress-to-rupture as a function of time and temperature --------- 16 Stress rupture factors for weldments ------------------------------------------- 17 Strain range fatigue curves ------------------------------------------------------- 18 Creep-fatigue damage envelope ------------------------------------------------- 19 Time-temperature limits for external pressure charts ------------------------ 20 Isochronous stress versus strain curves ---------------------------------------- 21 Other properties that may be needed for Subsection NH acceptance ------ 22

3.2 ASME Section III, Class 2, Class 3, and Code Cases ------------------------------- 23 3.3 ASME Section VIII, Div. 1 ------------------------------------------------------------- 24 4. IDENTIFICATION OF NEEDED DATA ------------------------------------------------ 25 4.1

Data Needs for Materials Properties for Alloys in Priority 1, 2, and 3 Rankings 25

4.1.1

Priority 1 alloys ------------------------------------------------------------------- 25

4.1.1.1 Data needs for Alloy 617 ------------------------------------------------------- 26 4.1.1.2 Data needs for Hastelloy X----------------------------------------------------- 27 4.1.1.3 Data needs for Hastelloy XR--------------------------------------------------- 29

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4.1.1.4 Data needs for Alloy 800H ----------------------------------------------------- 30 4.1.1.5 Data needs for stainless steel 316FR ------------------------------------------ 31 4.1.1.6 Data needs for stainless steel 316H ------------------------------------------- 32 4.1.1.7 Data needs for modified 9Cr-1Mo -------------------------------------------- 33 4.1.2 4.1.3 4.1.4 4.2

Priority 2 alloys ------------------------------------------------------------------- 33 Priority 3 alloys ------------------------------------------------------------------- 34 Group 5 alloys --------------------------------------------------------------------- 35

Guidelines for Assessment of Existing Data ----------------------------------------- 35

5. ANALYSIS AND PROCESSING FOR DATA DERIVATION ----------------------- 37 5.1 Review of Data Requirements for ASME Section II Acceptance------------------ 37 5.2 Review of Data Requirements for ASME Section III, Subsection NH------------ 38 6. CONSIDERATION OF NRC, ASME NQA-1, AND SECTION XI REQUIREMENTS---------------------------------------------------------------------------- 40 6.1

Crack Growth Rate ---------------------------------------------------------------------- 40

6.1.1 6.1.2 6.1.3 6.1.4 6.2

Creep crack growth rate ---------------------------------------------------------- 40 Fatigue crack growth rate -------------------------------------------------------- 41 Creep-fatigue crack growth rate ------------------------------------------------- 42 Crack growth rate of welds ------------------------------------------------------ 43

Fracture Toughness ---------------------------------------------------------------------- 43

6.2.1 6.2.2

Transition temperature for impact energy absorption ------------------------ 44 KJC parameter for reference temperature To ------------------------------------ 45

6.3 Aging Effects ----------------------------------------------------------------------------- 45 6.3.1 6.3.2

Construction of time-temperature-precipitation diagrams ------------------- 46 Effect of time and temperature on short and long time properties ---------- 47

6.4 Environmental Effects ------------------------------------------------------------------ 47 6.5 Radiation Effects ------------------------------------------------------------------------- 48 7. TESTING METHODS ----------------------------------------------------------------------- 49 7.1 Applicable ASTM Standard Documents ---------------------------------------------- 49 7.2 Equivalent Test Methods Applicable to Imported Test Data ----------------------- 50 8. TEST MATRIX DEVELOPMENT -------------------------------------------------------- 51

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 8.1 Test Matrices of Alloy 617 ------------------------------------------------------------- 51 8.2 Test Matrices of Hastelloy X------------------------------------------------------------ 52 8.3 Test Matrices of Hastelloy XR---------------------------------------------------------- 53 8.4 Test Matrices of Alloy 800H------------------------------------------------------------ 53 8.5 Test Matrices of Stainless Steel 316FR------------------------------------------------ 54 8.6 Test Matrices of Stainless Steel 316H ------------------------------------------------- 55 8.7 Test Matrices of Grade 91 --------------------------------------------------------------- 55 APPENDIX A: TEST PARTICIPANTS ------------------------------------------------------- 57 APPENDIX B: REFERENCES ----------------------------------------------------------------- 58 APPENDIX C: INFORMATION ABOUT ALLOY 617 ------------------------------------ 59 APPENDIX D: INFORATION ABOUT HASTELLOY X --------------------------------- 67 APPENDIX E: INFORMATION ABOUT HASTELLOY XR ----------------------------- 75 APPENDIX F: INFORMATION ABOUT ALLOY 800H ---------------------------------- 78 APPENDIX G: INFORMATION ABOUT 316FR ------------------------------------------- 89 APPENDIX H: INFORMATION ABOUT 316H--------------------------------------------- 92 APPENDIX I: INFORMATION ABOUT GRADE 91 -------------------------------------- 93

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ ACKNOLEDMENTS In the process of developing this test plan, the authors had discussions, required information, or requested review of the draft from the following people. Their cooperation and help are greatly appreciated. James Corum Robert Jetter Louis Mansur Timothy McGreevy Karen Moore Randy Nanstad Philip Rittenhouse Dane Wilson

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 1. INTRODUCTION

1.1

Background

In the development of the Gen IV nuclear reactors, the U.S. Department of Energy (DOE) has selected the Very High Temperature Reactor (VHTR) as the leading Next Generation Nuclear Plant (NGNP) reactor among six candidate reactor concepts. At the early stage of the Gen IV program, the research and development (R&D) efforts are focused on the NGNP reactor system, which will provide invaluable experience and lead the development of the other Gen IV reactors. The NGNP reference concept is a helium-cooled, graphite-moderated, thermal neutron spectrum reactor of 600 MWth with a once-through uranium fuel cycle and a target outlet temperature of 1000ºC (1832ºF) or higher. The high temperature helium output from the NGNP reactor will be directed to a helium gas turbine with or without an intermediate heat exchanger (IHX) for electricity generation; it can also be coupled through an IHX with heat application processes for hydrogen production. The design service life of the NGNP reactor system is 60 years. The basic technology for the NGNP has been established in earlier high-temperature gas-cooled reactor plants; techniques have been advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, the Pebble Bed Reactor (PBR), and Prismatic Modular Reactor (PMR) International Near-Term Deployment projects. The feasibility of some of the planned NGNP components and materials are being demonstrated in the Japanese High-Temperature Engineering Test Reactor (HTTR) and the Chinese HighTemperature Reactor (HTR)-10 projects. The NGNP project is focused on building a full-scale, Nuclear Regulatory Commission (NRC)-licensed demonstration reactor capable of producing electricity and low-cost hydrogen, rather than simply confirming the basic feasibility of the concept, in the 2015 to 2017 timeframe. The preliminary design, which will be completed by FY-09, will define the final materials requirements for the reactor design. These materials must be fully qualified for the final design process expected to start at the end of FY-09. Construction of long-lead items, such as the reactor pressure vessel, is planned to begin by FY-10. To support the design and construction of the NGNP, the Idaho National Engineering and Environmental Laboratory (INEEL), the Oak Ridge National Laboratory (ORNL) and other DOE contractors collaboratively developed a NGNP materials plan issued in November 2003 and entitled “Next Generation Nuclear Plant Materials Selection and Qualification Program Plan” [1], hereafter cited within the text as “the materials plan”. In the materials plan, the NGNP materials needs, rationale, and content of the NGNP Materials Program activities were described; and candidate materials for specific high temperature components were proposed for down-selection based on American Society of Mechanical Engineers (ASME) Codes and Standards, previous application experience, and recent development in materials technology. However, due to the unprecedented operating conditions specified for the NGNP, neither the previous application experience

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ nor the existing Codes data of the identified candidate materials can cover all the severe service conditions anticipated. For example; some of the NGNP component materials will be exposed to temperatures much higher than those currently allowed for any materials in the ASME Boiler and Pressure Vessel (B&PV) Code for nuclear construction. This is particularly true under accident conditions, which can expose the core structure components to temperatures up to 1200ºC (2192ºF). In this regard, Code Cases will have to be developed and many areas of the Code will have to be modified or expanded in order to address these design conditions. Furthermore, new materials and approaches may also be required. Because the thermal, environmental, and service life conditions in the NGNP design are unprecedented for many of the high temperature components, the modification and expansion of the Code and down-selection and eventual qualification of the candidate materials face significant challenges. The three major challenges identified in the materials plan were the effects of helium contaminants, high temperature exposure, and long-term irradiation on microstructure stability and mechanical properties of the materials. The helium contaminants include various gaseous impurities such as N, CO, CO2, H2, H2O, and CH4. These could cause undesirable chemical reactions, resulting in environmental degradation. The high temperature exposure can result in metallurgical structure changes such as coarsening or dissolving strengthening precipitates, phase transformation, and deprivation of solid solution elements at grain boundaries. These can result in loss of strength or embrittlement. The long-term irradiation can cause operation of micro mechanisms such as abnormal absorption of interstitials at dislocations, accumulation of vacancies at cavities, asymmetrical partitioning of self-interstitials and vacancies to dislocations differently oriented to stresses, creation of extremely small obstacles, and weakening of grain boundaries. These can result in swelling, irradiation creep or embrittlement. With time, all of these mechanisms can deteriorate or affect the mechanical properties such as tensile strength, creep strength, creep-fatigue resistance, stress-rupture resistance, high- and low-cycle fatigue resistance, creep crack growth rate, fatigue crack growth rate, and fracture toughness. Obviously, Code expansion and reactor design without full knowledge of these effects on microstructures and mechanical properties of the materials can compromise the safety and operation of the NGNP reactor system. Although modern computation and modeling technologies can often provide excellent predictions and simulations of physics phenomena including materials behavior, the predictions and simulations still have to be built on sound experimental data. Furthermore, the effects on materials behavior from various factors, interactions and synergisms in the NGNP operating conditions are far too complicated, if not impossible, to be understood completely through software experiments such as computational modeling. It is apparent that before the candidate materials can be down-selected and eventually qualified for the construction of the NGNP reactor system, much has to be learned about their behavior under the design service conditions through carefully designed hardware experiments. This is especially true for materials that are not even covered by the current ASME Codes and Standards, or of limited previous application experience and sparse existing data. On the other hand, even for materials with previous

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ application experience and existing data, the data from different sources need to be analyzed, verified, and accepted or rejected based on new data generated in the NGNP Materials Program. All this has to be accomplished by testing of materials in simulated NGNP service environments. Due to the number of the candidate materials and the numerous properties of importance, the testing of NGNP materials entails a large amount of work. This will be accomplished by collaboration among DOE laboratories, universities, and private sector research institutes. To ensure the efficiency of this collaboration and to avoid redundancy, all of the testing activities must be well coordinated under a single, mutually agreed materials test plan. The present “High Temperature Metallic Materials Test Plan for Generation IV Nuclear Reactors”, hereafter cited within the text as “the test plan”, is developed under the High Temperature Materials (HTM) Task for this purpose.

1.2

Scope

Although the materials testing activities will address materials issues for all the Gen IV reactor concepts, the current test plan is focused on the testing of metallic materials for the NGNP reactor system including the reactor vessel and internals, cross vessel, power conversion system, intermediate heat exchanger (IHX), and balance of plant system. Because NGNP is the leading Gen IV reactor candidate and shares many common materials issues with the other reactor concepts as well as the hydrogen production facilities, progress in its materials testing will cover the majority of the materials issues for the other reactor and facility concepts. Issues specific to the other reactor and facility concepts that are not covered here will also gain invaluable experiences from the NGNP materials testing. Therefore, detailed materials testing issues specific to the other reactor and facility concepts will not be discussed at present, but left to be addressed in future in order to draw upon experience that will be accumulated in the NGNP materials testing. Non-metallic materials, including ceramics and graphite, will be excluded from this plan, but covered in other specific documents. At present, the test plan includes eight major sections as shown below: •

Introduction, in which the background, scope, and organizing mechanism of materials data acquisition activities are discussed.



Bases for the Selection of Materials, in which materials to be tested are grouped and prioritized according to their existing data status and desirability in NGNP.



Classification of Data Requirements, in which ASME B&PV Codes and Standards are reviewed for properties of interest, and generic information needed for the NGNP reactor system are discussed.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Identification of Needed Data, in which properties data needed for each candidate material are identified by comparing the generic data needs and the existing data reviewed. •

Testing Methods, in which requirements for tests to generate the needed data are discussed. Test methods covered by the American Society for Testing and Materials (ASTM) are identified for the testing activities. The issue about equivalent test methods applicable to test data from international sources is also discussed.



Analysis and Processing for Derived Data, in which analysis on raw data, requirements for data derivation and ASME Code acceptance are described.



Consideration of NRC and ASME NQA-1 Requirements, in which properties and working condition effects not covered by current ASME Codes and Standards are identified and discussed.



Test Matrix Development, in which guidelines and helpful information for specific materials are provided for participants of the testing activities to develop their test matrices.

The schedule, cost, and participants of the testing activities will not be covered in this plan, but will follow those specified in the materials plan and the management of the Gen IV materials program. In general, most testing activities should be completed to meet the full qualification of the candidate materials for the final design process, which is expected to start at the end of FY-09. Long-term creep tests and tests for examining material properties of specific product forms and/or manufacturing processes that are specified during the reactor design will continue past FY-09 to meet the reactor components fabrication and reactor licensing needs.

1.3

Assessment of Existing Data

It should be pointed out that large amounts of data exist for many of the candidate materials proposed in the materials plan. These data may be found in ASME and ASTM documents, recognized international Codes and Standards, various materials handbooks, manufacturer records and open literature. As discussed above, identification of property testing needed for a candidate material is done by comparing the generic data needs with the existing data of the material. A complete understanding and full knowledge of existing data will help to address key issues and to avoid redundancy in testing. The benefit of taking advantage of existing data in reducing cost, saving time, and minimizing manpower requirements cannot be overstated. Needless to say, to gain such understanding entails substantial efforts in assembling, reviewing, analyzing, accepting or rejecting, compiling, and documenting the existing data. For this purpose guidelines and helpful information for assessment of existing data are fully discussed in this plan. This makes the plan a data acquisition plan rather than just a test plan.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 1.4 Coordination and Plan Change Control This test plan is developed based on requirements from the pre-conceptual and conceptual design, experience from previous nuclear reactor systems, current ASME Codes and Standards, and present understanding of the materials behavior and data needs. As the NGNP reactor design progresses to maturity and understanding of the materials behavior improves, new requirements for materials properties data will inevitably emerge. The new requirements will most likely come from the designers through the High Temperature Design Methodology (HTDM) Task. The test plan should respond to these new requirements and make changes accordingly in a timely manner. On the other hand, a materials database, entitled “The Gen IV Materials Handbook”, has been planned to facilitate the design activities of all Gen IV reactors including the NGNP reactor VHTR. The testing activities in this test plan will be a major data source for the Gen IV Materials Handbook. For this reason, the test plan should be closely coordinated with the development of the Gen IV Materials Handbook. Furthermore, due to the currently limited understanding of the existing data, the section for test matrices of this plan only contains preliminary and tentative guidelines. These will be subject to significant expansion and changes as the existing data assessment and testing activities progress. All test matrices are expected to be developed by participants of the testing activities based on their facility capabilities as well as the guidelines and requirements established in this plan. It is believed that this approach will give the participants enough freedom to explore to their full capacity while fulfilling the properties data needs identified in this test plan. Therefore, this test plan is virtually a “living” document, which will be revised in the form of addenda as needed and updated annually to reflect adjustments in the design requirements basis and their effect on the materials testing, selection, and qualification activities. To facilitate future changes, figures and tables are not sequentially numbered throughout the plan. Instead they are identified relatively independently to each other by their subsection number or an alphabetical suffix to their subsection number when more than one figures or tables are presented in one subsection. To coordinate various testing activities among all the participants, especially keep all the testing plan progress and changes on track, it is highly necessary that a coordinator be identified so that all the testing efforts can be directed fully towards the common final goals set in the test plan and can be adjusted in a timely manner in response to any new design needs. All the testing activities supported by the Gen IV funds should include the requirement that the test matrices be filed with the coordinator. The coordinator’s responsibilities should include the following: 1)

Keeping record of test matrices from all the testing activities.

2)

Updating the test matrix section of the test plan as needed.

3)

Identifying overlooked data needs; preventing redundancy and informing the participants of the others’ testing activities when necessary.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4) Serving as a point of contact with the High Temperature Design Methodology and Gen IV Materials Handbook activities.

1.5

5)

Documenting non ASTM standard test methods and analytical procedures if developed for new data requirements.

6)

Maintaining and revising the test plan annually.

Unit System

Based on the fact that data generated under this plan will be incorporated into the Gen IV Materials Handbook, which will be shared with international participants of the Generation IV International Forum, and the fact that ASME now requires data in the Standard International (SI) unit system, SI units will be employed as the primary unit system in this plan and all the testing activities.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 2. BASES FOR THE SELECTION OF MATERIALS Candidate materials for the NGNP reactor system have been proposed for downselection and qualification in the materials plan. As part of the materials plan activities, this test plan will proceed into further details of the down-selection and qualification process. The present section lays out the bases for the selection through grouping and prioritizing the candidate materials.

2.1

General Requirements

In the materials plan, components of the NGNP reactor system were described and various candidate materials were proposed for their construction. Those components have a wide variety of configurations and will be manufactured using many different materials processing technologies. Weldments will also be heavily employed in the construction processes. On the other hand, structural alloys are generally supplied in various forms such as wrought plates, bars, tubing, forgings, and castings etc., and not all forms are available for a particular alloy mostly due to various manufacturing issues. It is well recognized that for a given alloy, the product form may affect the compositional range that can be practically specified. As a result, variations in microstructures and mechanical properties are normally expected in different product forms of the same alloy. Therefore, in the testing program, data should be acquired not only for particular alloys, but also for particular product forms needed for constructing a specific component. The ASME normally adopts the ASTM specification of a material with little or no modifications by adding a prefix of S. However, because ASME deals heavily with structural materials used for pressure containment, it consequently requires higher standards for some alloys. For example, compared to ASTM specifications for structural plates, the ASME specifications for pressure vessel plates normally requires more stringent limits for allowable surface and edge imperfections, and demands meeting notch toughness requirements. In this testing program, data from various specifications for a given material should be acquired and compared; the specifications of the material for a given component application should be identified and documented. For the same specification, properties data should be acquired from at least three commercial-size heats in various product forms expected in typical service applications. Furthermore, for a candidate material to obtain approval under the ASME B&PV Code [2], it must meet the general requirements provided in Appendix 5 of Section II, Part D of the Code: “Guideline on the Approval of New Materials Under the ASME Boiler and Pressure vessel Code”. Included in this appendix is a “New Material Checklist” with fifteen issues that must be addressed. All the candidate materials proposed for the NGNP reactor system that have not been accepted into the ASME B&PV Code must comply with these general requirements. For the convenience of the readers, the New Material Checklist” is presented as follows: New Materials Checklist: To assist inquirers desiring Code coverage for new materials, or extending coverage of existing materials, the Committee has

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ developed the following checklist of items that ought to be addressed by each inquiry. The Committee reserves the right to request additional data and application information when considering new materials. (a)

Has a qualified inquirer request been provided?

(b)

Has a request either for revision to existing Code requirements or for a Code Case been defined?

(c)

Has a letter to ASTM or AWS been submitted requesting coverage of the new material in a specification, and has a copy been submitted to the Committee? Alternatively, is this material already covered by a specification issued by a recognized national or international organization and has an English language version been provided?

(d)

Has the construction Code and Division coverage been identified?

(e)

Has the material been defined as ferrous or nonferrous and has the application (product forms, size range, and specification) been defined?

(f)

Has the range (maximum/minimum) of temperature application been defined?

(g)

Has mechanical property data been submitted (ultimate tensile strength, yield strength, reduction of area, and elongation at 100°F or 50°C intervals, from room temperature to 100°F or 50°C above the maximum intended use temperature for three heats of appropriate product forms and sizes)?

(h)

If requested temperatures of coverage are above those at which timedependent properties begin to govern design values, has appropriate time-dependent property data for base metal, weld metal, and weldments been submitted?

(i)

If coverage below room temperature is requested, has appropriate mechanical property data below room temperature been submitted?

(j)

Have toughness considerations required by the construction Code been defined and has appropriate data been submitted?

(k)

Have external pressure considerations been defined and have stressstrain curves been submitted for the establishment of external pressure charts?

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ (l) Have cyclic service considerations and service limits been defined and has appropriate fatigue data been submitted?

2.2

(m)

Has physical properties data (coefficient of thermal expansion, thermal conductivity and diffusivity, Young's modulus, shear modulus, Poisson's ratio) been submitted?

(n)

Have welding requirements been defined and has procedure qualification test data been submitted?

(o)

Has influence of fabrication practices on material properties been defined?

Grouping and Priority

To facilitate coordinating among all the participants, to ensure efficient work flow, and to save time and cost in the testing activities, the candidate materials are organized in groups with different priorities for testing. The grouping is mainly based on the intended service temperature of the materials. In each group, priorities for testing are given to materials based mainly on their desirability to the NGNP components, but with consideration given to existing data and previous industrial experience. The groups and priorities are listed as follows: Group 1

Service above 760ºC (1400ºF) Priority 1: Alloy 617, Hastelloy X and XR Priority 2: Alloy 230 Priority 3: CCA 617, Inconel 263, Inconel 740

Group 2

Service from 650 to 760ºC (1202 to 1400ºF) Priority 1: Alloy 800H Priority 2: Alloy 120

Group 3

Service from 600 to 650ºC (1112 to 1020ºF) - stainless steels Priority 1: 316FR SS, 316H SS Priority 2: 316LN SS

Group 4

Service to 600ºC (1112ºF) - ferritic and ferritic/martensitic steels Priority 1: Grade 91 Priority 2: Grade 92 Priority 3: Grade 122, SAVE 12

Group 5

Special and advanced alloys Haynes 214, ODS alloys, Abe alloys (nano-nitride strengthened 9Cr martensitic steels)

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ In the first four groups, data for materials of higher priority should be acquired first. No priorities have been given to the materials in Group 5 because these are special and advanced alloys still under development. Although these are materials that may provide better performance in the unprecedented working conditions of the NGNP reactor system than the other materials, the existing data are sparse and industrial experience with such materials is lacking. They are not given top priority, but preliminary acquisition of their data should be initiated as soon as funds and testing facilities allow. It should also be noted that there is interest in extending the application temperature of Alloy 800H or its variants beyond 760ºC (1400ºF) for applications such as the core barrel. If this becomes a request from the designers in future, Alloy 800H or its variants will be relocated to Group 1.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 3. CLASSIFICATION OF PROPERTIES REQUIREMENTS Properties that may be needed for the design and construction of the NGNP reactor system are identified and discussed in this section. The identification is based mainly on the 2001 ASME Boiler and Pressure Vessel Code [2] with the 2002 and 2003 addenda, which is the most updated version at the time this plan was being developed. Code sections relevant to the NGNP operating conditions employed to identify the properties of interest include Code Section III, Division 1, Subsection NH – Class 1 Components in Elevated Temperature Service, Subsection NB – Class 1 Components, Subsection NC – Class 2 Components, Subsection ND – Class 3 Components, and relevant Code Cases; Section VIII, Division 1 and Division 2; and the piping Codes B31.1 and B31.3 where applicable. Due to the unprecedented working conditions for the NGNP components, the current ASME Code does not cover all the properties needed for the NGNP design and construction. New properties will be added to this section in future revisions as requirements emerge in the course of design. The objective of this section is only to identify what structural material properties should be acquired for the design and construction of the NGNP reactor system. Little effort is made to discuss how the properties should be determined; this is left to Section 5: Testing Methods. Furthermore, properties identified in this section are generic, without specific material identifications or property value descriptions. Property values required for a specific material will be identified in Section 4. Therefore, all the tables and figures for properties identified in this section can be considered as templates.

3.1 ASME Section III, Division 1, Subsection NH - Class 1 Components in Elevated Temperature Service Subsection NH covers the rules for construction of nuclear facility components in elevated temperature region. It contains rules for materials, design, fabrication, examination, testing, and overpressure relief of Class 1 components, parts, and appurtenances. Although its temperature coverage is far less than what is required for the NGNP reactor system, the rules it contains for elevated temperature design make it the most important subsection in delineating the property requirements for the design and construction of the NGNP reactor system.

3.1.1

Cold forming limits

In manufacturing a component, cold forming is frequently employed. When exceeding certain limits, the cold work can impair material properties such as fatigue, creep rupture, impact toughness, etc. To ensure that components are adequate for the intended service, Subsection NH requires a post fabrication heat treatment depending on the amount of the cold work. For components with cold work less than 5%, no post fabrication heat treatment is required. For those with cold work greater than 20%, the heat treatment is a must. Between 5% and 20%, limits and conditions for omitting heat

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ treatment are given in Figure NH-4212-1: “Permissible Time/Temperature Conditions for Material Which Has Been Cold Worked > 5% and < 20% and Subjected to ShortTime High Temperature Transient”, which specifies permissible total time of high temperature excursions within which no heat treatment is required. For some of the Gen IV candidate materials, the cold work limits of 5% and 20% defined in the current Subsection NH may need to be re-determined due to the severe operating condition, especially the extremely high service temperatures. The accident temperature that may occur to a given component should be used to determine the maximum short-time excursion temperature. For some core component materials, the short-time excursion temperature can run as high as 1200ºC (2192ºF). In the testing program, sufficient data should be acquired for each candidate material that may be subjected to cold forming during manufacturing and construction to generate a curve as shown in Figure 3.1.1. Such a 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.

1200

TEMPERATURE, C

1100

1000

900

800

700 t-TColdWork040825 NGNPTestPlan ORNL / W.Ren

600 10

1

10

2

10

3

10

4

10

5

TIME, h

Figure 3.1.1 Permissible time/temperature conditions for a given material which has been cold worked > x% and < y% and subjected to short-time high temperature transient To generate such an envelope for cold forming limits, the relationship between cold work and recrystallization is needed. Test data for creep-rupture, toughness, and ductility properties from materials with various amounts of cold work are also required.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 3.1.2 Tensile properties as a function of temperature Subsection NH provides designers with data of two major tensile properties, the ultimate tensile strength Su and the yield strength Sy, as a function of temperature. In Table NH-3225-1, Su values are listed at temperatures ranging from 371ºC (700ºF) to 816ºC (1500ºF) with intervals of (50ºF), plus that at room temperature. In Table I-14.5, Sy values are listed at temperatures ranging from 38ºC (100ºF) to 871ºC (1600ºF), with intervals of (100ºF) from 38ºC (100ºF) to 371ºC (700ºF), and intervals of (50ºF) from 371ºC (700ºF) to 871ºC (1600ºF), plus that at room temperature. These ranges can vary depending on materials. To support the NGNP design under Subsection NH, Su values at temperatures up to 1000ºC (1832ºF) are required for the materials of some components. ASME requires that for the purpose of codification, the maximum temperature at which the value of a given property is provided should be 50ºC higher than the intended application temperature. Therefore, the Su values for a given candidate material can be acquired at temperatures ranging from 350ºC (662ºF) to 50ºC higher than the design service temperature specified in Section 2 of this plan at intervals of 25ºC. If a candidate material has very rapidly changing properties in a certain temperature range, consideration should be given to performing tests at 15ºC intervals over that range. The Su value at room temperature should also be acquired. As an example, the Su values need to be acquired for a candidate material with the maximum design service temperature of 1000ºC (1832ºF) are given in Table 3.1.2 (a). Table 3.1.2 (a): Tensile strength values (Su) Temperature (ºC) Su (MPa) Temperature (ºC) Su (MPa) Temperature (ºC) Su (MPa)

25 350 375 400 425 450 x x x x x x 575 600 625 650 675 700 x x x x x x 825 850 875 900 925 950 x x x x x x

475 x 725 x 975 x

500 525 550 x x x 750 775 800 x x x 1000 1025 1050 x x x

Based on the argument above for Su, values of Sy should be acquired at temperatures ranging from 25ºC (77ºF) to the design service temperature specified in Section 2 of this plan plus 50ºC required by the ASME for codification. The temperatures should be at intervals of 50ºC from 50ºC (122ºF) to 350ºC (662ºF) and 25ºC from 350ºC (662ºF) to the maximum design service temperature plus 50ºC. For candidate materials with rapidly changing properties in a certain temperature range, consideration should be given to intervals of 15ºC over that range. Again, the Su value at room temperature should be acquired. An example of the data that need to be acquired for a candidate material with the maximum design service temperature of 1000ºC (1832ºF) is given in Table 3.1.2 (b). Tensile tests at various temperatures are required to generate the information. Because the service temperatures for some Gen IV components are very high, the

13

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ material may become very sensitive to loading rate. Therefore, depending on the results from preliminary testing, tensile tests with various loading rate may also be required. Table 3.1.2 (b): Tensile strength values (Sy) Temperature (ºC) Sy (MPa) Temperature (ºC) Sy (MPa) Temperature (ºC) Sy (MPa) Temperature (ºC) Sy (MPa)

25 x 400 x 625 x 850 x

50 x 425 x 650 x 875 x

100 x 450 x 675 x 900 x

150 x 475 x 700 x 925 x

200 x 500 x 725 x 950 x

250 x 525 x 750 x 975 x

300 x 550 x 775 x 1000 x

350 x 575 x 800 x 1025 x

375 x 600 x 825 x 1050 x

Although not presented in Subsection NH, tensile properties other than the ultimate tensile and yield strengths can also be very important for the design. In this testing program, tensile properties such as uniform and total elongation, Young’s modulus, reduction of area should also be reported for all the tensile tests. Related properties such as Poisson’s ratio and shear modulus should also be acquired.

3.1.3

Ultimate tensile and yield strength reduction factors for aging

The effects of aging on the deterioration of ultimate tensile strength and yield strength are treated in Subsection NH with strength reduction factors. For austenitic alloys, Table NH-3225-2 gives strength reduction factors to be applied to several alloys when Su and Sy are used. For ferritic/martensitic steel 2 1/4Cr–1Mo, Tables NH3225-3A and NH32253B list strength reduction factors for aging time ranging from 1 to 300,000 hours at aging temperatures ranging from 371ºC (700ºF) to 649ºC (1200ºF) at specified intervals. The NGNP reactor system requires a design life of 60 years, which is 525,600 hours, and it is not possible to age a specimen for that long in this testing program. Therefore, strength reduction factors for aging have to be acquired through a combination of testing and modeling efforts. The following actions should be taken: 1. Begin aging the specimens or specimen materials as soon as the materials are ready. 2. If the specified materials are already aged or being aged in other programs, efforts should be made to acquire the materials for NGNP. 3. Start modeling as soon as possible to identify information needed from testing and provide test requirements (temperature, inspection interval, minimum number of tests for a given testing condition, microstructural characterization requirements etc.).

14

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4. Test the aged materials to support the modeling process.

3.1.4

Stress intensity values as a function of time and temperature

To provide designers with information to design for adequate load capacity, Subsection NH offers allowable stress intensity values for various loaded durations and temperatures in Figures I-14.3A ~ D and Tables 14.3A ~ D. The time-dependent stress intensity St and the time-independent stress intensity Sm are both given as a function of temperature. Depending on the material, the temperature covers ranges from 316ºC (600ºF) to 816ºC (1500ºF). The allowable stress intensity value Smt for structural design is determined by the lower of St and Sm at a given temperature. For the design of the NGNP reactor system, the temperature range needs to be expanded to 1000ºC (1832ºF) plus the 50ºC required for codification for some components. For each candidate material, sufficient information should be acquired from tensile and creep tests to produce a curve of Sm and a series of curves of St over the temperature range of application. An example of the curves that need to be generated for a candidate material with the maximum design service temperature of 1000ºC (1832ºF) is given in Figure 3.1.4. 150

STRESS, MPa

125 S

100

1h m

10 h 5

3 x 10 h 75

30 h

5

100 h

10 h

300 h

4

3 x 10 h

50

4

10 h 25

3

3 x 10 h

Sts&TSmt040825 NGNPTestPlan ORNL / W.Ren

3

10 h

0 400

500

600

700

800

900

1000

TEMPERATURE, C

Fig. 3.1.4

Allowable stress intensity Smt as a function of temperature for a given material

15

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 3.1.5 Minimum stress-to-rupture as a function of time and temperature To design against creep rupture, Subsection NH provides designers with minimum rupture stresses under creep condition. Expected minimum stress-to-rupture values as a function of time and temperature are listed in Tables I-14.6A ~ E and plotted in Figures I14.6A ~ E for several alloys covering durations up to 300,000 hours and temperatures of 427ºC (800ºF) and above, increasing at intervals of 50ºF. To support the NGNP design and construction, the values of stress-to-rupture should be acquired at 350ºC (662ºF) and above for ferritic/martensitic steels, and 400ºC (752ºF) and above for high alloys, increasing at intervals of 25ºC. Because it is impossible to test a specimen for 525,600 hours in this testing program for the required NGNP design life of 60 years, data beyond practical testing time must be determined through a combination of testing and modeling efforts. Some modeling efforts, such as Larson-Miller method, may require testing at temperatures higher than the intended maximum service temperature. The following actions should be taken: 1. Start testing each candidate material at low stresses as soon as the materials are ready. Tests at higher stresses require shorter testing time, and therefore can be tested later in the program when more facilities become available. 2. If the specified materials are already being tested in other programs, efforts should be made to acquire the information for NGNP. 3. Start modeling as soon as possible to identify information needed from testing and provide test requirements (temperature, stress, minimum number of tests for a given testing condition, microstructural characterization requirements etc.). Table 3.1.5: Expected minimum stress-to-rupture values Temp. ºC 400 425 450 475 500 … 1050

1 x x x x x x x

10 x x x x x x x

30 x x x x x x x

102 x x x x x x x

3x102 x x x x x x x

103 x x x x x x x

Time (hr) 3x103 x x x x x x x

104 x x x x x x x

3x104 x x x x x x x

105 x x x x x x x

3x105 6x105 x x x x x x x x x x x x x x

The testing and modeling efforts should be able to provide enough information to generate a table and a plot shown as Table 3.1.5 and Figure 3.1.5 for each candidate material. The maximum temperature of 1050ºC (1922ºF) is used in the table and figure as an example for the maximum design service temperature of 1000ºC (1832ºF) plus

16

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 50ºC required for codification. It can be changed according to the maximum design service temperature of a given candidate material. 1000 o

400 C o

425 C

o

450 C o

STRESS, MPa

475 C

100

o

500 C o

525 C

o

550 C o

... C o

1000 C

10

o

1025 C o

1050 C MinSts&t050318 NGNPTestPlan ORNL / W.Ren

1 10

1

10

2

10

3

10

4

10

5

MINIMUM TIME TO RUPTURE, h

Figure 3.1.5

3.1.6

Expected minimum stress-to-rupture values for a given material

Stress rupture factors for weldments

A weld usually 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. Subsection NH addresses this difference with a stress rupture factor. Tables I-14.10 A-1 ~ D-1 provide designers with values of the stress rupture factor for welds of various base and filler metals, which is derived from the ratio of the rupture strength of the weld metal or weldment to that of the base metal away from the weld. For the design of the NGNP reactor system, the values of the stress rupture factor should be generated from derivation and modeling based on data provided by stress rupture tests conducted on the weld metal, the base metals, and weldments of the candidate materials. The welds should be tested and models should be developed in the same manner as described for base metals in 3.1.5 “Minimum stress-to-rupture as a function of time and temperature”. Because of the long testing times required for such tests, testing on welds should be started as early as possible. As soon as the specification of a candidate material is identified, weldability of the material should be investigated and welds should be produced for stress rupture testing.

17

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 3.1.7 Strain range fatigue curves At high temperature fatigue damage is usually considered to be under strain control. To design against fatigue damage, Subsection NH provides the designers with curves of strain range versus number of allowable cycles in Figures T-1420-1A ~ 1D, in which each curve stands for one temperature, and curves for several temperatures ranging from 427ºC (800ºF) up to 760ºC (1400ºF) plus that for 38ºC (100ºF) are presented. The same information is also listed in corresponding tables. The curves for all materials presented have a cyclic strain rate of 10-3 m/m/sec except one material with a cyclic strain rate of 4 x 10-3 m/m/sec. For the design and construction of the NGNP reactor system, the curves for strain range versus number of allowable cycles should be generated for each candidate material at 350ºC (662ºF) and above for ferritic/martensitic steels, 400ºC (752ºF) and above for high alloys, increasing at intervals of 50ºC, to 50ºC higher than the maximum design service temperature. As a baseline, a curve for room temperature should also be produced. The curves can be derived using data generated from strain controlled fatigue tests or strain and stress combined fatigue tests. Based on the information provided in Subsection NH, the fatigue tests should be conducted at a strain rate of 10-3 m/m/sec at a completely reversed stress ratio R = -1. Strain ranges for various numbers of cycles at failure up to 106 should be determined. The tests should provide sufficient information for generating a table and a plot as shown in Table 3.1.7 and Figure 3.1.7 for each candidate material. The maximum temperature of 1050ºC (1922ºF) is used in the table and figure as an example for the maximum design service temperature of 1000ºC (1832ºF) plus 50ºC required for codification. Table 3.1.7 Cycle No. 10 2x10 4x10 102 2x102 4x102 103 2x103 4x103 104 2x104 4x104 105 2x105 4x105 106

400 x x x x x x x x x x x x x x x x

Fatigue strain range at various temperatures 450 x x x x x x x x x x x x x x x x

Strain Range (m/m) at Temperature (ºC) 500 550 600 … 850 900 950 x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x x

18

1000 x x x x x x x x x x x x x x x x

1050 x x x x x x x x x x x x x x x x

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________

STRAIN RANGE, (m/m)

10

-1

25°C 400°C 450°C 10

-2

10

-3

500°C ...°C 1050°C

eRange&CycNo040816 NGNPTestPlan ORNL / W.Ren

10

1

10

2

10

3

10

4

10

5

10

6

NUMBER OF CYCLES, N

Figure 3.1.7

Fatigue strain range at various temperatures for a given material

Since materials usually become very sensitive to loading rate at very high temperature, strain rate other than 10-3 m/m/sec may also be considered in testing if preliminary results indicate that sensitivity to loading rate becomes a concern.

3.1.8

Creep-fatigue damage envelope

Both creep and fatigue cause accumulated damage in materials at high temperatures. Subsection 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 Equation 3.1.8 as follows: q ⎛ ⎛ n ⎞ ⎞ ⎜ ⎟ ⎜ Δt ⎟ ≤ D + ∑ ∑ ⎟ ⎜ ⎟ k =1 ⎜ j=1 ⎝ Nd ⎠ ⎝ Td ⎠ k j P

(3.1.8)

where, n = actual fatigue cycles during event j; Nd = fatigue life of the material; Δt = actual creep time during event k; Td = creep-rupture life of the material; and D = the total allowable creep-fatigue damage. The first term represents the calculated fraction of fatigue life consumed in a component, and the second term represents the calculated creep life consumed. The equation indicates that the total accumulated creep and fatigue damage should be controlled within an envelope.

19

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________

FRACTION OF CREEP LIFE CONSUMPTION

For the design of the NGNP reactor system, the creep-fatigue damage envelope should be developed for each candidate material. Creep, fatigue, and creep-fatigue tests should be conducted to provide information for the development. To facilitate the estimation of allowable creep and fatigue damage accumulation, microstructural analysis and modeling may also be necessary. For each candidate material, sufficient test data, analysis and modeling results should be acquired to produce a creep-fatigue damage envelope as shown in Figure 3.1.8.

Cp&FtgDamEvlop040816 NGNPTestPlan ORNL / W.Ren

1.0

0.8

0.6

0.4

0.2

0.0 0.0

0.2

0.4

0.6

0.8

1.0

FRACTION OF FATIGUE LIFE CONSUMPTION

Figure 3.1.8

Creep-fatigue damage envelope for a given material

It should be noted that this creep-fatigue damage model has not been satisfactorily representing true material behavior. It is one of the issues that ASME Elevated Temperature Design Task intends to resolve. There is interest in developing a new approach that may allow for variations in design to lead to different allowable limits based upon enhanced modeling tools available to designers. The testing activities will be adjusted accordingly to meet the new requirements.

3.1.9

Time-temperature limits for external pressure charts

At high temperatures, cylindrical and spherical components subjected to sustained external pressure may eventually buckle when creep deformation exceeds certain limits. To design against time-dependent creep buckling, Subsection NH provides timetemperature limits for application of Section II external pressure charts in Figures T-

20

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 1522-1 ~ 3 for several materials. These limit curves cover temperature ranges from 371 to 816ºC (700 to 1500ºF) and a time range from 10 to 1,000,000 hours. For the NGNP candidate materials, the time-temperature limits for temperature range starting from 350ºC (662ºF) to 50ºC higher than the maximum design service temperature and time durations up to 1,000,000 hours should be developed. Creep tests and modeling are needed to provide information for the development. For each candidate material, creep tests should be conducted at various temperatures to provide sufficient data for producing a plot as shown in Figure 3.1.9.

1000

o

TEMPERATURE, C

900

800

700

600

500

t&TExPressure040826 NGNPTestPlan ORNL / W.Ren

400 10

1

10

2

10

3

10

4

10

5

10

6

TIME, h

Figure 3.1.9

3.1.10

Time-temperature limits for application of Section II external pressure charts for a given material

Isochronous stress versus strain curves

To provide designers with information regarding total strain caused by stress under elevated temperature conditions, Subsection NH gives average isochronous stress-strain curves for several materials at temperatures ranging from 427ºC (800ºF) up to 816ºC (1500ºF) at intervals of 50ºF and times up to 300,000 hours in Figures T-1800-A1 ~ D11. Hot tensile stress-strain curves are also included in these figures as the baseline for each given temperature. For the design of NGNP reactor system, average isochronous stress-strain curves should be developed for temperatures ranging from 425ºC (797ºF) to 50ºC higher than the maximum design service temperature at intervals of 25ºC. For the 60 years of design

21

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ life, the curves should cover up to 600,000 hours. Creep tests are needed to produce data for generating the isochronous stress-strain curves. Because of the long design service life of 60 years, modeling must be conducted to assist creep testing efforts and produce sufficient information to generate isochronous stress-strain curves as shown in Fig 3.1.11. These curves will be particularly useful in simplified design methods permitted by Subsection NH. Full inelastic analysis, e.g. time and rate dependent analysis, with unified constitutive equations is currently required for some materials when temperatures reach levels where plastic and creep deformation are difficult to differentiate and are highly strain rate dependent. 180 IsoSts&Stn040817 NGNPPlan ORNL / W.Ren

Hot tensile

150

1 hr

STRESS, MPa

10 hr 30 hr

120

2

10 hr 2

3 x 10 hr

3

10 hr

90 3

3 x 10 hr 60

4

10 hr

4

3 x 10 hr

5

10 hr

5

3 x 10 hr 30

5

6 x 10 hr

0 0.0

0.2

0.4

0.6

0.8

1.0

1.2

1.4

1.6

1.8

2.0

2.2

STRAIN, %

Figure 3.1.10 Isochronous stress-strain curves for a given material at a given temperature

3.1.11

Other properties that may be needed for Subsection NH acceptance

Subsection NH was developed from its elevated-temperature Code Case predecessors, primarily for DOE’s liquid-metal reactor (LMR) program [3]. The development started in the late 1960s when it was recognized 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 Subsection NH is to provide design-byanalysis rules to guard against four failure modes that can occur at high temperature. These are 1) ductile rupture from short-term loadings, 2) creep rupture from long-term loadings, 3) creep-fatigue failure, and 4) gross distortion due to incremental collapse and ratcheting. In general, Subsection NH requires the use of inelastic design analyses to reflect plasticity and time-dependent creep effects. Because of the high operating

22

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ temperature of 1000ºC (1832ºF) expected for the NGNP reactor system, the requirements for a material to become qualified are considerably more extensive than those in current Subsection NH. There is little or no application experience to draw upon, and the required life of 60 years for the NGNP reactor system is nearly double that currently permitted by Subsection NH rules. New requirements of material properties data that are not covered in current Subsection NH will emerge in the course of the design. The following material properties data are some of those of interest: 1) Time to tertiary creep: Because of the required 60 years of life and 1000ºC (1832ºF) outlet temperature of the NGNP reactor, some components may necessarily be required to work under creep conditions. At low creep rate in the secondary creep regime, many materials can sustain load for a long period of time without rupture. However, if the deformation develops into the tertiary creep regime, these materials will deform at high rate and rupture may soon follow. Therefore, the time these materials can spend before reaching its tertiary creep region can become an important factor. However, it should be noted that other materials may exhibit limited primary and secondary creep but spend a significant amount of their lives sustaining load in the tertiary creep regime. For such materials, the time to reach a specified creep strain or creep-rupture strain may become an important factor for consideration in design. 2) Biaxial fatigue data: To perform consistent and valid design analyses required by Subsection NH, constitutive equations (for multiaxial inelastic response of the candidate materials to complex time-varying, thermal and mechanical loadings) and inelastic analysis guidelines need to be developed. Biaxial fatigue data are needed to provide basis for the constitutive equations, and are used to confirm the adequacy of application of Mises effective strain or other approaches for fatigue analysis. 3) Biaxial creep-rupture data: Like the biaxial fatigue data, biaxial creep-rupture data are needed to provide basis for the constitutive equations. They are also used for developing material specific multiaxial creep-rupture strength theory used in the Code.

3.2

ASME Section III, Class 2, Class 3, and Code Cases

The criteria for setting the allowable stresses for ASME Section III Class 2 and Class 3 components are identical to those for ASME Section I and Section VIII, Division 1. Values are provided in ASME Section II, Part D, Tables 1A and 1B. However, the usage of these tables is not permitted for temperatures above 371ºC (700ºF) for ferritic steels and 427ºC (800ºF) for austenitic stainless steels and nickel base alloys. For Class 2 and Class 3 component construction for higher temperatures, it is necessary to revert to Code

23

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Cases, such as N-253-11, in which stress intensity, rather than maximum stress is used and many of the rules set forth in Section III, Subsection NH must be followed. Data requirements for use of N-253-11, N-254, and the like are similar to those identified for Subsection NH. Other nuclear Code Cases intended to accommodate short time loading conditions that exceed the temperature limits of the governing construction may be needed. The data required to support these Code Cases will be specified when the product forms, materials, and loading conditions become known.

3.3

ASME Section VIII, Div. 1

For non-nuclear applications, the construction Code for unfired pressure vessels is Section VIII, Division 1 or Division 2. Division 1 provides rules for construction to temperatures in the time-dependent range. Division 2 is limited to lower temperatures. Again, ASME Section II, Part D, Appendix 1 provides the basis for establishing the stress values for Division 1. Although the criteria for setting the stresses may differ from ASME Section III and Subsection NH, the data needs for Section VIII, Division 1 are less and any material accepted for Subsection NH would have a sufficient database to gain Section VIII, Division 1 acceptance. Specifically, the needs for the following data have been identified. 1) Minimum room temperature tensile properties 2) Average tensile properties versus temperature 3) Average and minimum stress rupture strength at 100,000 hours 4) Slope of the log stress versus log rupture life at 100,000 hours 5) Typical (average) stress to produce 1 % creep in 100,000 hours 6) Toughness data for thicker products of some steels 7) Stress versus strain curves for buckling charts 8) Strength data for weldments 9) Information on alloys stability It should be noted that Subsection NH requires the time to produce 1% total strain as a function of stress and temperature, whereas Section VIII generally accepts the minimum creep rate of 0.01% creep per 1000 hours as a function of stress and temperature.

24

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4. IDENTIFICATION OF NEEDED DATA The focus of this section is to identify the data that need to be acquired, through testing or existing data mining, for each specific candidate material listed in Section 2 of this plan. The identification is based on a comparison between the generic properties requirements discussed in Section 3 of this plan and present knowledge of the existing data in ASME B&PV Code or experience with the material in previous projects. Needless to say, the identification presented in this section is only preliminary because the assessment of existing data has not been conducted to its full extent. Because the knowledge and understanding of existing data is considered an important part of the data-needs identification process, and furthermore, a prerequisite step for developing efficient test matrices, guidelines for assessment of existing data are also included in this section.

4.1

Data Needs for Materials Properties for Alloys in Priority 1, 2, and 3 Rankings

The candidate materials proposed in the materials plan are prioritized for testing based mainly on their desirability to the NGNP design and construction needs along with consideration of existing data and previous industrial experience including the information already contained in the ASME B&PV Codes and Standards. Their data needs are discussed in the order of priority in this section.

4.1.1

Priority 1 alloys

Seven alloys are ranked as Priority 1 materials: Alloy 617, Hastelloy X, and Hastelloy XR for service temperatures above 760ºC (1400ºF); Alloy 800H for service temperatures ranging from 650 to 760ºC (1202 to 1400ºF); stainless steels 316RF and 316H for service temperatures ranging from 600 to 650ºC (1112 to 1202ºF); and ferritic/martensitic steels Grade 91 (modified 9Cr-1Mo steel) for service temperatures up to 600ºC (1112ºF). Because of extremely high operating temperature, the major materials data needs for the design and construction of the NGNP reactor system lie in the temperature range where time-dependent material behavior plays a crucial role in structural failure. At present, five alloys qualified by ASME B&PV Code for nuclear service within this temperature range are included in Subsection NH. These are 304 and 316 stainless steels, 2.25Cr-1Mo ferritic steel (annealed), Alloy 800H, and modified 9Cr-1Mo steel. Evaluation results of some alloys for nuclear service in this temperature range can also be obtained through the work of previous High-Temperature Gas-Cooled Reactor (HTGR) project. In the project summary report [4], five materials (2.25Cr-1Mo ferritic steel, modified 9Cr-1Mo ferritic/martensitic steel, Alloy 800H, Alloy 617, and Hastelloy X) were examined extensively and four others (Nimonic-86, Hastelloy S, Manaurite-36 and Inconel 519) were investigated to a lesser extent.

25

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4.1.1.1 Data needs for Alloy 617 Current ASME B&PV Code Subsection NH does not include Alloy 617. Therefore, Alloy 617 is not a material currently qualified for nuclear service. All the generic data needs discussed in Section 3 of this plan should be acquired through testing and/or existing data mining. Since the alloy is proposed for service temperatures above 760ºC (1400ºF), all the data should be acquired for temperatures up to 1050ºC (1922ºF), 50ºC higher than the maximum intended service temperature of 1000ºC (1832ºF). Although Alloy 617 is not codified for nuclear service, it has been evaluated for nuclear service in the previous HTGR project [4]. According to the evaluation, data that need to be acquired are identified as follows: 1) Sufficient tensile properties data of Alloy 617 in solution-annealed and aged conditions at temperatures from 875 to 1050ºC (1607 to 1922ºF) may not exist, and should be investigated and acquired. Due to possible increased sensitivity to loading rate at very high temperatures, tensile properties at various loading rates should also be acquired in this temperature range. 2) Some existing tensile properties data of Alloy 617 in solution-annealed and aged conditions at temperatures from RT to 850ºC (1562ºF) are discussed in Reference [4]. The aging durations include 2,500, 10,000, and 20,000 hours. However, the discussed data are sparse. A few more data in this temperature range should be acquired for verification purpose, and sufficient amount of data should be obtained to fill the temperature intervals specified in Tables 3.1.2 (a) and (b). Data from samples with aging time longer than 20,000 hours are also desirable. 3) Significant amounts of creep deformation and rupture data of Alloy 617 exist at temperatures of 800, 850, 900, 950, and 1000ºC (1472, 1562, 1652 and 1832ºF) are discussed in Reference [4]. Stress-rupture data are needed to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan except those for these five temperatures. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 4) Sufficient creep deformation and rupture data of Alloy 617 in contaminated helium may not exist. Creep data in helium environments with various possible impurities should be acquired to cover all the creep properties requirements identified in Section 3 of this plan. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 5) Sufficient creep deformation and rupture data of Alloy 617 for welds may not exist. Creep properties data of welds should be acquired for the requirements identified in subsection 3.1.6 of this plan. Data should also be acquired to

26

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ investigate possible synergism of welding and helium impurity deterioration effects on creep properties. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 6) Sufficient fatigue data on Alloy 617 may not exist. Fatigue data are needed to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan. Data acquired should also provide sufficient information for developing the creep-fatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF).

4.1.1.2 Data needs for Hastelloy X The current ASME B&PV Code Subsection NH does not include Hastelloy X. Therefore, Hastelloy X is not a material qualified for nuclear service. All the generic data needs discussed in Section 3 of this plan should be acquired, through testing and/or existing data mining. Since the alloy is proposed for service temperatures above 760ºC (1400ºF), all the data should be acquired for temperatures up to 1050ºC (1922ºF), 50ºC higher than the maximum intended service temperature of 1000ºC (1832ºF). Although Hastelloy X is not codified for nuclear service, it has been evaluated for nuclear service in the previous HTGR project [4]. According to the evaluation, data that need to be acquired are identified as follows: 1) Tensile properties data of Hastelloy X in solution-annealed and aged conditions at temperatures from 875 to 1050ºC (1607 to 1922ºF) may not be sufficient and should be investigated and acquired. Due to possible increased sensitivity to loading rate at very high temperatures, tensile properties at various loading rates should also be acquired in this temperature range. 2) Some existing tensile properties data of Hastelloy X in solution-annealed and aged conditions at temperatures from RT to 850ºC (1562ºF) and are discussed in Reference [4]. The aging durations include 2,500, 10,000, and 20,000 hours. However, the discussed data are sparse. More data in this temperature range should be acquired for verification purpose, and a sufficient amount of data should be obtained to fill the temperature intervals specified in Tables 3.1.2 (a) and (b). Data from samples with aging time longer than 20,000 hours are also desirable. 3) Existing data of Hastelloy X evaluated in Reference [4] indicate that rapid change occurs in tensile properties around 700ºC (1292ºF). As suggested in Section 3 of this plan, small temperature intervals of 15ºC should be considered in this temperature region when data are acquired if such a rapid change is confirmed.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4) Some existing creep deformation and rupture data of Hastelloy X at temperatures of 760 and 871ºC (1400 and 1600ºF) are discussed in Reference [4]. Creep data are needed to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan except these 2 temperatures. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 5) Some existing creep deformation and rupture data of Hastelloy X at temperatures of 760 and 871ºC (1400 and 1600ºF) in helium are discussed in Reference [4]. Stress-rupture data are needed to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan except for these two temperatures. If future assessment indicates that the helium environment in which the existing data have been generated did not contain the possible impurities for the NGNP reactor system, all the temperatures identified in Table 3.1.5 should be covered in data acquisition. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 6) Sufficient creep deformation and rupture data of Hastelloy X for welds may not exist. Creep properties data of welds should be acquired for the requirements identified in subsection 3.1.6 of this plan. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties. 7) Some existing fatigue data of Hastelloy X at temperatures of RT, 538, 649, 760 and 871ºC (RT, 1000, 1200, 1400 and 1600ºF) are discussed in Reference [4]. More data are needed to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan. Data acquired should also provide sufficient information for developing the creep-fatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 8) Sufficient fatigue data for Hastelloy X in helium with various possible impurities may not exist. Fatigue data in contaminated helium environments are needed to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan. Data acquired should also provide sufficient information for developing creep-fatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF).

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4.1.1.3 Data needs for Hastelloy XR The current ASME B&PV Code Subsection NH does not include Hastelloy XR. Therefore, Hastelloy XR is not a material qualified for nuclear service. All the generic data needs discussed in Section 3 of this plan should be acquired, through testing and/or existing data mining, for codification. Since the alloy is proposed for service temperatures above 760ºC (1400ºF), all the data should be acquired for temperatures up to 1050ºC (1922ºF), 50ºC higher than the maximum intended service temperature of 1000ºC (1832ºF). According to the generic data needs discussed in Section 3 of this plan, data that need to be acquired are identified as follows: 1) Sufficient tensile properties data of Hastelloy XR in solution-annealed and aged conditions may not exist, and need to be acquired to fill Tables 3.1.2 (a) and (b) in Section 3 of this plan. Due to possible increased sensitivity to loading rate at very high temperature, acquisition of the data from tests at various loading rates should also be considered if such sensitivity is confirmed. For the aged conditions, the data should cover aging durations of 2,500, 10,000, 20,000, 40,000, and even more hours if possible. 2) Sufficient creep deformation and rupture data of Hastelloy XR may not exist. Stress-rupture data should be acquired to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 3) Sufficient creep deformation and rupture data of Hastelloy XR in helium environment with various possible impurities may not exist. Stress-rupture data from possible contaminated helium environments should be acquired to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). 4) Sufficient creep deformation and rupture data of Hastelloy XR welds may not exist. Creep properties data of welds should be acquired for the requirements identified in subsection 3.1.6 of this plan. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF). Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties. 5) Sufficient fatigue data of Hastelloy XR may not exist. Fatigue data should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan. Data acquired should also provide sufficient information for developing the creep-fatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF).

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 6) Sufficient fatigue data of Hastelloy XR in helium with various possible impurities may not exist. Fatigue data in possible contaminated helium environments should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan. Data acquired should also provide sufficient information for developing the creep-fatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. Majority of the data should be acquired for temperatures above the intended service temperature of 760ºC (1400ºF).

4.1.1.4

Data needs for Alloy 800H

Alloy 800H is one of the five materials qualified for nuclear service covered in current ASME B&PV Code Subsection NH. The alloy has also been evaluated for nuclear service in the previous HTGR project [4]. The evaluated data coverage did not exceed that in the Code. Based on the evaluation of its codified status and existing data, data that need to be further acquired are identified as follows: 1) Creep properties of Alloy 800H are covered for temperatures up to 760ºC (1400ºF) and durations up to 300,000 hours in Subsection NH for Smt and stress-to-rupture limits. For its application in the NGNP reactor system, the time coverage needs to be expanded to 600,000 hours at temperatures up to 810ºC (1490ºF). Note that the 600,000 hours coverage should also include the lower temperatures that are already covered up to 300,000 hours in the current Code. Although the 600,000 hours coverage will be developed through modeling, more long-term creep test data may be required for the modeling efforts. 2) Creep deformation and rupture data of Alloy 800H in helium environment with various possible impurities are not covered in Subsection NH and sufficient data may not exist elsewhere. Stress-rupture data from possible contaminated helium environments are needed to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan up to 810ºC (1490ºF). 3) Effects of welding on Alloy 800H are covered for temperatures up to 760ºC (1400ºF) and durations up to 300,000 hours in Subsection NH for stress-torupture limits. For its application in the NGNP reactor system, the duration coverage needs to be expanded to 600,000 hours at temperatures up to 810ºC (1490ºF). Note that data at lower temperatures that are already covered in the current Code should also be extended to 600,000 hours. Although the 600,000 hours coverage will be developed through modeling, more long-term creep test data may be required for the modeling efforts. Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 4) Fatigue properties of Alloy 800H are already covered for temperatures up to 760ºC (1400ºF) and cycle numbers up to 106 at a loading rate of 10-3 m/m/s for various fatigue strain ranges in Subsection NH. However, due to the concern for increased sensitivity to loading rate at high temperature, loading rates other than 10-3 m/m/s may be required for the design and construction of the NGNP reactor system. 5) Sufficient fatigue data for Alloy 800H in helium with various possible impurities may not exist. Fatigue data in possible contaminated helium environments are needed to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan for temperatures up to 800ºC (1472ºF). Data acquired should also provide sufficient information for developing creep-rupture envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. As mentioned in Section 2 of this test plan, interest exists in extending the application temperature of Alloy 800H beyond 760ºC (1400ºF). The data needs described above will be expanded when such interest becomes a request from the designers.

4.1.1.5

Data needs for stainless steel 316FR

Current ASME B&PV Code Subsection NH does not include stainless steel 316FR. Therefore, 316FR is not a material qualified for nuclear service. All the generic data needs discussed in Section 3 of this plan should be acquired through testing and/or existing data mining for codification. Since the alloy is proposed for service temperatures ranging from 600 to 650ºC (1112 to 1202ºF), all the data should be acquired for temperatures up to 700ºC (1292ºF), 50ºC higher than the maximum intended service temperature. According to the generic data needs discussed in Section 3 of this plan, data that need to be acquired for 316FR are identified as follows: 1) Sufficient tensile properties data for stainless steel 316FR in annealed and aged conditions may not exist, and should be acquired to fill Tables 3.1.2 (a) and (b) in Section 3 of this plan for temperatures up to 700ºC (1292ºF). Due to possible increased sensitivity to loading rate at very high temperatures, data from tests at various loading rates should also be considered. For the aged conditions, data acquisition should cover aging durations of 2,500, 10,000, 20,000, 40,000, and even more hours if possible. 2) Sufficient creep deformation and rupture data of stainless steel 316FR may not exist. Stress-rupture data should be acquired to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan up to 700ºC (1292ºF). 3) It is unlikely that creep deformation and rupture data of stainless steel 316FR in helium environment with various possible impurities exist. Stress-rupture data from possible contaminated helium environments should be acquired to

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ cover the temperatures identified in Table 3.1.5 in Section 3 of this plan up to 700ºC (1292ºF). 4) Sufficient creep deformation and rupture data of stainless steel 316 FR for welds may not exist. Creep properties data of welds should be acquired for the requirements identified in subsection 3.1.6 of this plan for temperatures up to 700ºC (1292ºF). Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties. 5) Sufficient fatigue data of stainless steel 316FR may not exist. Fatigue data should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan for temperatures up to 700ºC (1292ºF). Data acquired should also provide sufficient information for developing the creepfatigue damage envelope as shown in Figure 3.1.8 or any new damage models to replace the envelope. 6) It is unlikely that fatigue data for stainless steel 316FR in helium with various possible impurities exist. Fatigue data from possible contaminated helium environments should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan for temperatures up to 700ºC (1292ºF).

4.1.1.6

Data needs for stainless steel 316H

Stainless steel 316H is one of the five materials qualified for nuclear service covered in current ASME B&PV Code Subsection NH. The alloy is presented therein as 316 SS. For the NGNP reactor system, 316H is proposed for service in the temperature range of 600 to 650ºC (1112 to 1202ºF). The temperature coverage for 316H in current Subsection NH is up to 816ºC (1500ºF), exceeding what is required for its application in the NGNP reactor system. Therefore, unless new data requirements emerge in the design process, no further acquisition is needed for data produced in air environment. A possibility exists that more long-term creep data may be required for the modeling effort to predict the material behavior for the 60 years of service life. Data from various possible contaminated helium environments are definitely needed as follows: 1) Creep deformation and rupture data of stainless steel 316H in helium environment with various possible impurities may not be sufficient. Stressrupture data in the contaminated helium environments should be acquired to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan up to 700ºC (1292ºF). Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 2) Fatigue data of stainless steel 316H in helium with various possible impurities may not exist as needed. Fatigue data from possible contaminated helium environments should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan for temperatures up to 700ºC (1292ºF).

4.1.1.7

Data needs for modified 9Cr-1Mo

Ferritic/martensitic steel Grade 91 (modified 9Cr-1Mo steel) is one of the five materials qualified for nuclear service covered in current ASME B&PV Code Subsection NH. It is also evaluated for nuclear service in the HTGR project [4]. For the NGNP reactor system, Grade 91 is proposed to serve at temperatures up to 600ºC (1112ºF). The temperature coverage for Grade 91 in current Subsection NH is up to 649ºC (1200ºF), exceeding what is required for its intended applications in the NGNP reactor system. Therefore, no further acquisition is needed for its properties data in air environment unless new data requirements emerge from the design process. However, testing in various possible contaminated helium environments is still needed as follows: 1) Sufficient creep deformation and rupture data of Grade 91 in helium environment with various possible impurities may not exist. Stress-rupture data from possible contaminated helium environments should be acquired to cover the temperatures identified in Table 3.1.5 in Section 3 of this plan up to 650ºC (1202ºF). Data should also be acquired to investigate possible synergism of welding and helium impurity deterioration effects on creep properties. 2) Fatigue data of Grade 91 in helium with various possible impurities may not exist as needed. Fatigue data from contaminated helium environments should be acquired to fill Table 3.1.7 and produce specific curves for Figure 3.1.7 in Section 3 of this plan for temperatures up to 650ºC (1202ºF).

4.1.2

Priority 2 alloys

Four alloys are ranked as Priority 2 materials. These are Alloy 230 for service temperatures above 760ºC (1400ºF), Alloy 120 for service temperatures ranging from 650 to 760ºC (1202 to 1400ºF), 316LN SS for service temperatures ranging from 600 to 650ºC (1112 to 1202ºF), and Grade 92 for service temperatures up to 600ºC (1112ºF). Since these alloys have lower priority in the testing program, they will only be briefly discussed in this version of the test plan. More detailed planning will be conducted in future if the testing on Priority 1 materials indicates an increase in desirability of the Priority 2 materials. In alloy Group 1, Alloy 230 is the only Priority 2 material listed. The existing database for Alloy 230 is quite extensive, and the alloy experiences a broad range of

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ high-temperature applications. Alloy 230 is also Code approved for Section I Division 1 and Section VIII Division 1 for service to 899ºC (1650ºF) and the uniaxial database extends to even higher temperatures. Its fatigue properties appear to be superior to Alloy 617. At some temperatures the weld metal strength falls below that of base metal and the weldability of thick sections is an issue that should be addressed. Nothing is known about the performance of Alloy 230 in NGNP helium. At a minimum, Alloy 230 should be included in aging programs that expose materials to simulated NGNP helium environments. Since the aging process requires a long time, it should be started at the early stage of the testing program. In alloy Group 2, Alloy 120 has been included. This material shows improved strength over Alloy 800H, the Priority 1 alloy in its group. The existing creep-rupture database of Alloy 120 is quite expensive, and the alloy is Code approved for Section I Division 1 and Section VIII Division 1 for service to 899ºC (1650ºF). Although weldable in thick sections, there is no filler metal of matching compositions, so issues on weldment performance could arise. As with Alloy 230, it would be well to include Alloy 120 in the helium exposure tasks. In alloy Group 3, stainless steel 316LN is the only material listed. It is one of several austenitic stainless steels that could be included for service to 650ºC (1202ºF). These steels are included in ASME Section II along with the maximum temperatures permitted by the various Code books. These steels have had long-time exposure in fossil power plants as superheater tubing and much is known about their long-time high temperature stability. The development of these steels for Section III Class 1 construction in the creep range, however, would require an extensive experimental program. Grade 92 is listed as a Priority 2 steel for Group 4. This material is one of the most thoroughly studied high-temperature steels available today. It has superior strength to Grade 91 and service experience has been excellent. It is weldable in thick sections but suffers from the same problem as Grade 91 with respect to weakness in the fine-grained heat affected zone of weldments. Grade 92 steel should be included in the exposure tests that include Grade 91.

4.1.3

Priority 3 alloys

Five alloys are ranked as Priority 3 materials. These are Alloy CCA 617, Inconel 263, and Inconel 740 for service temperatures above 760ºC (1400ºF) and Grade 122 and SAVE 12 for service temperatures up to 600ºC (1112ºF). Since these alloys have the lowest priority in the testing program, their testing needs will not be discussed in this version of the test plan. Some of these alloys are being investigated under other programs in the United States and other countries. The progress of these investigations should be monitored. More detailed planning may be conducted in the future if the monitoring and the testing on Priorities 1 and 2 materials indicate an increase in desirability of the Priority 3 materials.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ It should be noted that significant interest exists in replacing the standard Alloy 617 with its variant Alloy CCA 617. Investigation on Alloy CCA 617 in other programs indicates that the CCA 617 exhibits improved high temperature strength compared to the standard Alloy 617. The improvement is attributed to its addition of boron, control of grain size, and lowering the maximum limits of certain elements. However, the material is still given Priority 3 at present mainly for three reasons: 1) The strength improvement is only up to 750ºC (1382ºF); 2) There is a concern about the interaction between Boron and neutron; 3) The material is currently under extensive investigation in the UltraSupercritical Steam Boiler Materials Program at ORNL. Preliminary tests should be conducted at the early stage of the testing program to compare the standard and CCA 617; and information on the effects of Boron on neutron should also be collected and reviewed before the CCA 617 can be given a higher priority.

4.1.4

Group 5 alloys

Group 5 include several special and advanced alloys including Haynes 214, ODS alloys, and Abe alloys. These alloys normally feature highly desirable properties for elevated temperature service, but have unsolved problems and may suffer from lack of existing data and experience as well. Like the priority 3 alloys, many of the Group 5 alloys are under development and investigation in other programs in the United States and other countries. The progress made for these alloys should be monitored, but no testing needs will be discussed in this version of the test plan. However, if ideas that could result in breakthrough for the development of these alloys and also address important material concerns of the NGNP reactor systems are proposed, they should be seriously considered.

4.2 Guidelines for Assessment of Existing Data The entire assessment process includes assembling, reviewing, analyzing, accepting or rejecting, compiling, and documenting the existing data. Because existing data are provided by various sources, decisions must be constantly made about their quality, reliability, relevancy, and acceptability. Without certain guidelines, these decisions will largely be left to the personal judgment of the individual data assessors and conflicts and inconsistency may be inevitable. Therefore, it is necessary that some commonly agreed guidelines are provided for the assessment activities. In regard to the quality of existing data, the assessment will follow the criteria proposed by the Gen IV Materials Handbook Task for existing data quality identification and classification as listed below [5]: Class 1 These materials data meet all DOE Gen IV Reactor Programs and NRC QA requirements (i.e., these are data generated in documented R&D programs that meet all of the requirements of 10CFR50 Appendix B and

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ DOE/NRC agreed versions of NQA-1). It is expected that the data collected in Gen IV materials programs will be of this category. Class 2 Materials data and data correlations provided in various sections of wellrecognized US Codes and Standards (e.g., ASME and ASTM) will be designated as Class 2. In many cases the raw data (i.e., individual data points) will not be available from these sources. Thus, the results of peer-approved analyses and resulting data correlations contained in these Codes and Standards may be the major Class 2 input to the Handbook. Class 3 Materials data provided in well-recognized international codes and standards will be categorized at present as Class 3. This may be revisited as the result of any international agreements reached relative to cooperation on the Handbook. For, example Classes 2 and 3 might be combined into a single class. Class 4 Materials data obtained from materials handbooks such as the Nuclear Systems Materials Handbook and the AFCI Materials Handbook will be identified as Class 4. The data contained in these two examples have had careful and extensive analysis and peer review (equivalent to Class 2) but the QA for the data generation range from “unknown” to Class 1. In the latter case the data from such handbooks would be listed as Class 1. Class 5 Materials data obtained from manufacturers brochures and from the open literature will be categorized as Class 5. Such data will be reviewed/approved as will be described in the Gen IV Materials Handbook Implementation Plan prior to its incorporation into the Gen IV Materials Handbook. Although these classifications do not provide an absolute measurement of the quality of existing data, they provide certain guidance on data acceptability and assurance. If the quality of any existing data appears questionable, testing should be conducted to verify the reliability before the data are rejected or accepted.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 5. ANALYSIS AND PROCESSING FOR DATA DERIVATION

5.1

Review of Data Requirements for ASME Section II Acceptance

The ASME Subcommittee II has the responsibility for setting the allowable stresses for all of the construction Codes that concern pressure boundary components. The responsibility for setting the criteria on which the stresses are determined, however, falls on the particular construction “book” or Code. These criteria differ from book to book but are provided in Appendices 1 and 2 of ASME Section II Part D. Generally, the data specified in Appendix 5 of Section II Part D are provided to the Subcommittee II, and these data are processed by a representative of the Subcommittee to derive the information that goes into the appropriate stress tables. Currently, there are no formal guidelines for such analyses. The individuals who process and analyze the data for Subcommittee II generally use methods that have evolved quite esoterically over a number of years. The data required for the Section II analysis consist of: 1)

Tables for tensile yield strength, ultimate strength, elongation, and reduction of area at the required temperature intervals for at least three heats;

2)

Tables of stress-rupture life for at least three heats at the required stress, temperature, and life intervals;

3)

Minimum creep rate data for at least one heat and over the range of stresses covered by the rupture data.

In regard to yield and tensile strength data, the analyses performed by Subcommittee II representatives generally consists of a normalization of the elevated temperature data by the room temperature value and the regressing on the ratio values as a function of temperature. This procedure produces a ratio curve whose values correspond to RY for the yield strength (ratio of the average temperature dependent trend curve value of yield strength to the room temperature yield strength) and RT for the ultimate strength (ratio of the average temperature dependent trend curve value of tensile strength to the room temperature tensile strength). These terms are defined in the Section I-D Appendix 1. The rupture data are generally analyzed on the basis of a time-temperature parameter that is a function of log stress. Typically, the Larson-Miller parameter is selected and a “lotcentering” method is used with log time as the dependent variable. The analysis leads to the determination of SRavg (average stress to cause rupture at the end of 100,000 hours) in the Section II-D Appendices 2 and 3. Generally, the analysis of weld metal and weldment data is not performed as above by Section II. Rather, data are compared to base metal trends and a judgment is necessary with respect to the relative strength of weld metal or weldment. Weld stress reduction factors may be assigned.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ A set of tensile curves to at least 2% total strain at the required temperature intervals are required for construction of buckling charts. Other data required but not used to set allowable stresses include Charpy V toughness energy for ferritic steels and some ferritecontaining austenitic steels for thickness about the exclusion limit.

5.2

Review of Data Requirements for ASME Section III, Subsection NH

As described in subsection 3.1 of this plan, the data requirements for ASME Section III Subsection NH are much more extensive than those for other Code books, and, except for the items specifically covered in Section II, Part D, the processing and analysis of the data fall within the responsibilities of the ASME Subgroup on Elevated Temperature Design (SG-ETD). Many of the tables and graphs represent derived data developed from various correlations. As an example, the isochronous stress strain curves are anchored from a tensile yield curve whose 0.2% off set yield strength is 1.25 times the value provided in the Section II-D Table Y-1. The creep components of the isochronous curves represent material of average creep strength and include only the primary and secondary stages of creep. The components are derived from a creep law applicable to the range of conditions covered by the data but are extended by extrapolation to stresses approaching zero. The curves do not included the tertiary creep components for the materials currently included in Section III Subsection NH. Different methods were used for processing and analyzing data to develop the creep law. In contrast to the methods used to develop the isochronous curves in the current Subsection NH, the draft Code Case for Alloy 617, requires unified equations be used that do not distinguish between “timeindependent” plasticity and “time” or “rate” plasticity (creep). Several unified equations currently exist for alloy 617. These differ significantly in regard to complexity and applicability. Some apply only to monotonic deformation under uniaxial stress. Others apply to monotonic and cyclic loading under uniaxial or multiaxial stresses. The derivation of materials-specific parameters from test data may require special efforts. Most of the material data needed to support ASME Section III Subsection NH and its Code Cases will be generated or mined as discussed in Sections 3 and 4 of this test plan. However, verification of design approaches will require either component or feature-like material testing. Such testing will be supported and conducted within the HTDM task. No efforts are made in present test plan that specifically address these specific HTDM needs, but much overlap exists between the HTM and HTDM tasks and the testing program outlined in this test plan must include the flexibility to adjust to the changing needs of the SG-ETD. The current version of Subsection NH includes guidelines for simplified design methods and analyses at high temperatures. Such approaches rely upon either elastic analysis or simplified inelastic analysis. As such, the material properties discussed in Sections 3 and 4 of this test plan, along with isochronous curves and creep-fatigue interaction diagrams, weld reduction factors, etc. are utilized. The modification or addition of simplified design methods for ASME Section III Subsection NH falls under the responsibility of the task on HTDM. Additional testing needs will be the

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ responsibility of HTDM. Close cooperation, communication, and planning between HTM and HTDM tasks will be required, especially to anticipate unforeseen test needs

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 6. CONSIDERATION OF NRC, ASME NQA-1, AND SECTION XI REQUIREMENTS

Although not required by the ASME B&PV Code for design and construction, some additional material properties data may be needed for obtaining operation license from the Nuclear Regulatory Commission (NRC). Some properties data may also be needed for meeting the ASME NQA-1 [6] and Section XI [7] requirements. In this section, material property information of interest to NRC licensing and ASME quality assurance and some issues related to testing are discussed.

6.1

Crack Growth Rate

During its long-term service at elevated temperature, the material of a structural component will inevitably become aged. Its microstructures and chemical properties can become degraded, and its mechanical properties can be gradually deteriorated by creep and fatigue damage. Eventually microcracks can be initiated as a result of accumulation of creep and fatigue damage in stress concentration areas. On the other hand, preexisting flaws such as cavities and microcracks are commonly found in engineering materials. These pre-existing flaws may result from crystal or material processing defects, which are almost inevitable. The stress concentration at the tips of these flaws can induce localized creep deformation when exposed to elevated temperatures even under very low applied load levels. As time elapses, these flaws may grow by a process of coalescence of creep cavities near the sharp tips, and fatigue cycling can accelerate this process. Because this process can occur for the combinations of temperature and net section stress where creep would not normally be of concern, conventional designs based upon creep rupture data may not be suitable to guard against such damage. It has been recognized that although microcracks can pre-exist and initiate in a component during its service at elevated temperatures, a large portion of the component’s useful life can be spent in crack propagation. In the energy generation industry, many components are operating safely with cracks nucleating and propagating at elevated service temperatures because their growth rates are predictable and closely monitored. With the requirements of 60 years of service life and a 1000ºC (1832ºF) outlet temperature for the NGNP reactor system, crack growth properties of the candidate materials could become very important factors that should be considered in the materials down-selection and licensing processes.

6.1.1

Creep crack growth rate

Creep crack growth testing should be conducted to generate creep crack growth rate data, which are characterized by a time-dependent fracture mechanics parameter. Several time-dependent fracture mechanics parameters exist and are appropriate for various creep conditions of the growing crack. The most popularly used two parameters are C* for

40

CREEP CRACK GROWTH RATE, da/dt (mm/h)

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ cracks growing under extensive secondary creep conditions, and C*(t) for those growing under extensive primary and/or secondary creep conditions. Creep crack growth testing should be conducted for each candidate material at various temperatures in the intended service temperature range at 50ºC intervals or less if temperature effects are exhibited in the characterization. Sufficient data should be generated to produce a plot as shown in Figure 6.1.1 at each temperature. For materials whose characterization is not affected by temperature, the maximum temperature should be used in testing to save testing time. Creep crack growth resistance of different materials can be compared based on their creep crack growth rates. 1.0

A 1.8441x10

q -2

0.9225

0.1

q

da/dt = A C*(t) 0.0 R = 0.994

CpCrackRate040831 NGNPTestPlan ORNL / W.Ren

0.0 0.1

1.0

10.0

100.0

2.

C*(t), Kilo Joule/m h

Figure 3.1.5

6.1.2

Creep crack growth rate

Fatigue crack growth rate

Fatigue crack growth testing should be conducted to generate fatigue crack growth rate data, which are characterized using stress the intensity factor range ΔK. For each candidate material, fatigue crack growth testing should be conducted at various temperatures at 50ºC intervals or less in the intended service temperature range, and sufficient data should be generated to produce a plot as shown in Figure 6.1.2 at each temperature. Due to the increased sensitivity to loading rate, tests at various loading rates should be considered. Fatigue crack growth resistance of different materials can be compared based on their fatigue crack growth rates.

41

FATIGUE CRACK GROWTH RATE, da/dN (mm/cycle)

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________

10

-3

n

C 4.1309 x 10

10

f

-12

6.8908

-4

R = 0.98668 n

da/dN = C ( Δ K)

f

FgCrackRate040831 NGNPTestPlan ORNL / W.Ren

10

-5

4 10

0

6 10

0

0

8 10 10

1

3 10

1

STRESS INTENSITY FACTOR RANGE, Δ K (MPa.m ) 1/2

Figure 6.1.2

6.1.3

Fatigue crack growth rate

Creep-fatigue crack growth rate

Components working in elevated temperature and cyclic loading conditions can suffer from both creep and fatigue damage. At the tip of a pre-exist flaw or a micro crack initiated due to long-term aging at elevated temperature under cyclic loading, a fatigue damage zone can be developed within a creep zone. Usually, the overall damage can not satisfactorily represented by a linear summation of the creep and fatigue damage effects. The synergism of the two damage mechanisms may cause accelerated material deterioration. Creep-fatigue crack growth testing should be conducted to generate creepfatigue crack growth rate data and damaged material specimens to provide knowledge about the material deterioration process through micro structural characterization. Several time-dependent fracture mechanics parameters exist for characterizing the creepfatigue crack growth rate. For each candidate material, creep crack growth testing should be conducted at 50ºC intervals or less in the intended service temperature range if temperature effects are exhibited in the characterization. Sufficient data should be generated to produce a plot as shown in Figure 6.1.3 at each temperature. For materials whose characterization is not affected by temperature, the maximum temperature should be used in testing to save testing time. Creep crack growth resistance of different materials can be compared based on their creep-fatigue crack growth rates.

42

avg

AVERAGE CREEP-FATIGUE CRACK GROWTH RATE (da/dt) (mm/h)

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________

10

-1

t = 10 seconds h

t = 100 seconds h

t = 3600 seconds h

10

-2

da/dt = A C*(t) R = 0.9941

A

10

q

1.7941 x 10

CpFgCrackRate040831 NGNPTestPlan ORNL / W. Ren

-2

0.8534

-3

10

-1

10

[C*(t)]

Figure 6.1.3

6.1.4

q

0

10

1

2

, (Kilo Joules/m .h)

avg

Creep-fatigue crack growth rate

Crack growth rate of welds

Fusion welding processes are primary techniques employed in joining metallic materials in the construction of elevated temperature components. Due to their castinglike microstructures and the heat affected zone (HAZ), welds are usually more vulnerable to creep, fatigue, and creep-fatigue crack growth damage. Limited testing on crack growth rates in the welds of the candidate materials should be conducted to determine their resistance to these crack growth mechanisms. If their vulnerability to these types of damage is verified, sufficient data should be generated to produce plots as shown in Figs. 4.1.1 – 4.1.3 for design and licensing considerations.

6.2

Fracture Toughness

As mentioned in the Introduction, structural materials are subjected to high temperature aging, irradiation damage, and helium impurities deterioration under the operating conditions of the NGNP reactor system. In these processes, there are several mechanisms that can cause embrittlement. To ensure safe operation without catastrophic structural failure in the system, the materials’ capability to resist embrittlement should be considered, and fracture toughness of the materials for which toughness will be an issue

43

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ should be evaluated. Fracture toughness of a material can be evaluated in several ways. At present, the transition temperature for impact energy absorption and KJC parameter are considered for evaluating the candidate materials of the NGNP reactor system. Other methods, such as the Drop Weight test, RTNDT, J-R etc., may also be considered when necessary. A detailed plan has been made by the Fracture Mechanics Task [8]

6.2.1

Transition temperature for impact energy absorption

Impact energy absorption data for various temperatures from Charpy tests should be acquired for candidate materials that may have toughness issues. Transition temperature for impact energy absorption can be derived from these data to evaluate and compare the toughness of the materials. For the NGNP candidate materials, toughness measurements before and after exposure to simulated NGNP working conditions will be desirable. Sufficient data should be acquired for candidate materials to produce curves as shown in Figure 6.2.1, in which three hypothetical steels are evaluated and their relative toughness can be compared. Such comparison can also be conducted on the same material with different exposures to simulated NGNP working conditions.

CHARPY ENERGY ABSORBED, J

140 120 Steel A

100 80

Steel B

60

Steel C

40 20 0 -100

CharpyE&T040902 NGNPTestPlan ORNL / W.Ren

-50

0

50

100

150

200

250

300

o

TEMPERATURE, C

Figure 6.2.1

Transition-temperature curve from Charpy tests

44

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 6.2.2 KJC parameter for reference temperature To Data of KJC parameter at various temperatures should be acquired for ferritic steels such as Grade 91. The KJC parameter is an elastic-plastic equivalent stress intensity factor derived from the J-integral at the point of onset of cleavage fracture, JC. With values of KJC at various temperatures, the reference temperature To can be determined. The reference temperature To characterizes the fracture toughness of ferritic steels that experience onset of cleavage cracking at elastic and/or elastic-plastic KJC instabilities. The relative toughness of different materials, or of a material before and after exposure to simulated NGNP working conditions, can be evaluated and compared based on the values of To. An example of KJC versus temperature from a material before and after exposure to radiation is given in Figure 6.2.2. The curves can also be produced from different materials for comparison. Sufficient data should be acquired for each material to produce curves as shown in Figure 6.2.2.

300 KJC&T040902 NGNPTestPlan ORNL / W.Ren

250

JC

K , MPa.m

1/2

As Received 200 Irridated Irridated/Annealed

150

100

50 -100

-50

0

50

100

o

TEMPERATURE, C

Figure 6.2.2

6.3

KJC versus temperature

Aging Effects

It was discussed in Section 3 of this test plan that data will be generated from aged candidate materials to provide information about the effects of aging on materials properties. However, it is impractical to test specimens to the 60 year aging time

45

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ characteristic of the NGNP. Therefore, modeling will be employed to supplement testing in the prediction of long-term materials behavior under the NGNP operating conditions.

6.3.1

Construction of time-temperature-precipitation diagrams

Generally, calculations that predict stable phases as a function of alloy chemistry say very little about kinetics and the development of non-equilibrium transition phases. Kinetics may be observed in time-temperature-precipitation (TTP) diagrams that have been developed for all of the Priority 1 alloys and several of the Priority 2 alloys. The diagrams are useful in understanding the metallurgical factors that control deformation mechanisms, failure mechanisms, strength, and ductility. Most of the existing TTP diagrams are limited to 10,000 hours, so much work remains to be done to extend the diagrams to times of interest to the NGNP reactor system, as shown in Figure 6.3.1. Further, it is known that the compositional variables, fabrication variables, thermalmechanical loadings, and radiation affect the kinetics of precipitation and re-solution of precipitates in the Priority 1 alloys. To be useful for modeling of deformation and fracture, the diagrams should 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 be helpful. 1100

o

TEMPERATURE, C

1000 900 M C + Laves M C x

800

x

M C +σ

y

x

y

y



M C + Laves x

y

700

+σ+β

600

M C + Laves x

500

y

PPT040913 NGNPPlan ORNL / W.Ren

400 10

-2

10

-1

10

0

10

1

10

2

10

3

10

4

10

5

10

TIME, h

Figure 6.3.1

Time-temperature-precipitation diagram

46

6

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 6.3.2 Effect of time and temperature on short and long time properties Knowledge of the effect of aging on short-time properties such as yield, flow, and hardening properties is very important. Estimation of toughness is also important for risk assessment purposes. As mentioned in subsection 3.1.3 of this test plan on properties required for ASME B&PV Code Section III, Subsection NH, it may be necessary to develop tables of strength reduction factors for short-time properties. If the requirement for the use of unified equations is to be satisfied, the influence of aging on the material parameters will be needed. Knowledge of the effects of aging on long-term properties is important in understanding damage mechanisms associated with long-term exposure to creep and fatigue. The specification of aging conditions and subsequent testing should be predicated on the requirements for the development of applicable constitutive equations, mechanistic models, and continuum damage mechanics models.

6.4

Environmental Effects

Some details about testing for environmental effects are discussed in the proposed FY-05 activities from the Materials Compatibility Task of the NGNP Materials Program [9]. Aging in NGNP helium and testing in NGNP helium should be emphasized with respect to the overall testing program on high-temperature metallic materials. Because of the high expense of environmental testing, the planning of test conditions takes on paramount importance. What is needed early on is a plan in regard to how helium effects will be factored into the design process. As an example, it was assumed early in the Japanese HTGR project that the major effect of the environment on the performance of Alloy 617 was decarburization and subsequent loss in strength. Correlations between carbon level and strength were developed along with a model for decarburization of Alloy 617 as a function of temperature, time, and product thickness. For design purposes, the stress intensities could then be adjusted according to the expected decarburization. Test schedules to validate environment-strength interaction models should be the emphasis of the environmental testing program rather than a large matrix of stresses, temperatures, and time for a number of product forms. Helium effects on cyclic behavior and crack growth are of major concern since surface effects are important to crack initiation and growth. Again, it would be best to propose a model for the damage process and build a testing program around the validation of the model. Helium effects on alloys other than those in Group 1 may be of lesser concern since their maximum operating temperatures will be lower.

47

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 6.5 Radiation Effects Many of the candidate materials will be exposed to radiation in service and experience irradiation damage as described in the Introduction of this plan. Testing in simulated radiation conditions of the NGNP reactor system is crucial for ensuring knowledgeable design and construction of the system. Special facilities will be required for such testing, and a detailed plan has been developed by the Materials for Radiation Service Task [8]. Activities under the present test plan will be coordinated closely as necessary with the Materials for Radiation Service Task, and provide the task with information and assistance as needed.

48

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 7. TESTING METHODS All tests to be conducted under this test plan should refer to the test methods and analytical procedures written by the American Society for Testing and Materials, hereafter cited within the text as “ASTM Standards”. Because the ASTM constantly reviews its Standards and publishes revisions annually, it is important that the updated test methods and analytical procedures are followed in this testing program. In addition, data generated in all the testing activities should be reported with the version of the ASTM Standards employed so that the data sources are accurately documented for future reference. Although English units are in many cases used in the ASTM Standards, SI units should be used as the primary unit system in all the testing activities under this test plan. If data needs emerge from the design activities such that the use of non ASTM Standard test methods or analytical procedures becomes necessary and inevitable, details of the methods and procedures should be thoroughly documented in the format of the ASTM Standards and filed with the test plan coordinator. The filed documents will be sent to relevant experts for review, and efforts may be made to incorporate the documents into the ASTM Standards at appropriate time.

7.1

Applicable ASTM Standard Documents

At present, ASTM Standard documents [10] that should be complied within this test plan are listed below. If additional ASTM Standard documents that have not been included in the current test plan are involved in the testing activities, the documents should be filed with the test plan coordinator so that revisions can be made accordingly. E 1856-97ε1(2002) Evaluating Computerized Data Acquisition Systems Used to Acquire Data from Universal Testing E 4-03 Force Verification of Testing Machines E 83-02 Verification and Classification of Extensometer System E 6 - 03 Methods of Mechanical Testing Conducting Creep, Creep-Rupture, and Stress-Rupture Tests E 139-00ε1 of Metallic Materials E 633 - 00 Use of Thermocouples in Creep and Stress Rupture Testing to 1000°C (1800°F) in Air E 21-03a Elevated Temperature Tension Tests of Metallic Materials E 111-97 Young's Modulus, Tangent Modulus, and Chord Modulus E 8M - 04 Tension Testing of Metallic Materials [Metric] E 8 - 04 Tension Testing of Metallic Materials E 467- 98a

Verification of Constant Amplitude Dynamic Forces in an Axial Fatigue Testing System

49

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Evaluating Data Acquisition Systems Used in Cyclic Fatigue E 1942- 98ε1 and Fracture Mechanics Testing E 1457 - 00 Measurement of Creep Crack Growth Rates in Metals E 647 - 00 Measurement of Fatigue Crack Growth Rates E 466 – 96(2002) Conducting Force Controlled Constant Amplitude Axial Fatigue Tests of Metallic Materials E 468 - 90(1998) Presentation of Constant Amplitude Fatigue Test Results for Metallic Materials E 606 - 92(1998) Strain-Controlled Fatigue Testing E 23-02a E 1921- 03

Notched Bar Impact Testing of Metallic Materials Determination of Reference Temperature, To, for Ferritic Steels in the Transition Range E 1820 - 01 Measurement of Fracture Toughness E 208 - 95a(2000) Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels E 807 - 96 E 3-01 E45-97(2002) E 1245 - 03 E 407-99 E 1181 -02 E 112-96ε3

7.2

Metallographic Laboratory Evaluation Preparation of Metallographic Specimens Determining the Inclusion Content of Steel Determining the Inclusion or Second-Phase Constituent Content of Metals by Automatic Image Analysis Microetching Metals and Alloys Characterizing Duplex Grain Sizes Determining Average Grain Size

Equivalent Test Methods Applicable to Imported Test Data

Some candidate materials proposed for the NGNP reactor system have been extensively investigated in other nations. The existing data generated from international sources should be assessed, and, when necessary, verified using ASTM Standards. In the assessment process, equivalent international test methods and analytical procedures applied to the generation of the existing data should be filed in this subsection for future analysis and reference.

50

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ 8. TEST MATRIX DEVELOPMENT This section is the “container” of all the test matrices for this test plan. At present, only guidelines and information helpful for developing testing matrices are provided. With the progress of the project, these guidelines will be appended by test matrices that are developed. As mentioned in the Introduction of this plan, all the testing activities supported by the NGNP funds should input test matrices to the test plan coordinator. This section is intended to serve as a vehicle for filing, coordinating, compiling and documenting these matrices. The matrix filed will be given an ID in the form of AlloyStatus-Property-Organization-Date, and the matrix developer will be informed of the ID for future communications. For example, 316FR-Aged-Creep-XYZ-092305 represents a test matrix for creep testing on aged stainless steel 316FR filed on September 23, 2005 by an organization whose acronym is XYZ. The full name of the acronym can be found in Appendix A: TEST PARTICIPANTS.

8.1

Test Matrices of Alloy 617

A significant amount of data already exits for Alloy 617. To develop test matrices for the alloy, information listed as follows may be helpful: •

In the mid-1980s an ad hoc group of the Code Subgroup on Elevated Temperature Design was established to develop a draft Code Case for the use of Alloy 617 nuclear components at very high temperatures. DOE requested the Case through General Electric for high-temperature gas-cooled reactor applications. Alloy 617 was chosen because of the availability of applicable data. The draft Case, which was subsequently approved by the Subgroup on Elevated Temperature Design, covered temperatures to 950°C (1742°F). The Case, as well as identified gaps and deficiencies that would have had to be filled before it could be fully applied, is described in a paper authored by Corum and Blass (1991) [see Appendix C – General]. This paper will be reviewed before developing a test matrix for Alloy 617. The case was dropped from further consideration by the Code when the DOE program was terminated and no other need for the Case was identified. The other references listed in Appendix C- General may contain extensive existing data and/or general information about Alloy 617. They should also be reviewed.



Test matrices for tensile properties of Alloy 617 should be developed according to data needs 1) and 2) discussed in subsection 4.1.1.1 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix C - Tensile.



Test matrices for creep properties of Alloy 617 should be developed according to data needs 3) and 4) discussed in subsection 4.1.1.1 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix C – Creep.

51

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Test matrices for creep properties of Alloy 617 welds should be developed according to data needs 5) discussed in subsection 4.1.1.1 of this plan and closely coordinated with the creep test matrices for the base metal. •

Test matrices for fatigue properties of Alloy 617 should be developed according to data needs 6) discussed in subsection 4.1.1.1 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix C – Fatigue.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of Alloy 617, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix C – Crack.

8.2

Test Matrices of Hastelloy X

A significant amount of data already exits for Hastelloy X. To develop test matrices for the alloy, information listed as follows may be helpful: •

Hastelloy X was one of the alloys included in the Nuclear System Materials Handbook (NSMH) developed by DOE in support of advanced nuclear reactor system. To develop test matrices for this alloy, information in NSMH must be reviewed. Other helpful information is also listed with NSMH in Appendix D - General.



Matrices for tensile testing Hastelloy X should be developed according to data needs 1), 2) and 3) discussed in subsection 4.1.1.2 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix D – Tensile.



Matrices for creep testing of Hastelloy X should be developed according to data needs 4) and 5) discussed in subsection 4.1.1.2 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix D – Creep.



Matrices for creep testing of Hastelloy X welds should be developed according to data needs 6) discussed in subsection 4.1.1.2 of this plan and closely coordinated with the creep test matrices for the base metal. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix D – Weld.



Matrices for fatigue testing of Hastelloy X should be developed according to data needs 7) and 8) discussed in subsection 4.1.1.2 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix D – Fatigue.

52

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of Hastelloy X, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix D – Crack.

8.3 Test Matrices of Hastelloy XR Some data already exit for Hastelloy XR. To develop test matrices for the alloy, the information listed as follows may be helpful: •

Matrices for tensile testing Hastelloy XR should be developed according to data needs 1) discussed in subsection 4.1.1.3 of this plan. Literature survey is needed for existing information about tensile properties of Hastelloy XR.



Matrices for creep testing of Hastelloy XR should be developed according to data needs identification 2) and 3) in subsection 4.1.1.3 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix E – Creep.



Matrices for creep testing of Hastelloy XR welds should be developed according to data needs identification 4) in subsection 4.1.1.3 of this plan and closely coordinated with the creep test matrices for the base metal. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix E – Weld.



Matrices for fatigue testing of Hastelloy XR should be developed according to data needs identification 5) and 6) in subsection 4.1.1.3 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix E – Fatigue.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of Hastelloy XR, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix E – Crack.

8.4

Test Matrices of Alloy 800H

Because Alloy 800H is qualified for nuclear service covered in current ASME B&PV Code Subsection NH, a significant amount of data already exits. To develop test matrices for the alloy, information listed as follows may be helpful: •

Alloy 800H is one of the alloys included in the Nuclear System Materials Handbook (NSMH) developed by DOE in support of advanced nuclear reactor system. To develop test matrices for this alloy, information in NSMH must be

53

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ reviewed. Other helpful information is also listed with NSMH in Appendix F - General. •

Test matrices for creep properties of Alloy 800H should be developed according to data needs 1) and 2) discussed in subsection 4.1.1.4 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix F – Creep.



Test matrices for creep properties of Alloy 800H welds should be developed according to data needs 3) discussed in subsection 4.1.1.4 of this plan and closely coordinated with the creep test matrices for the base metal. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix F – Weld.



Test matrices for fatigue properties of Alloy 800H should be developed according to data needs 4) and 5) discussed in subsection 4.1.1.4 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix F – Fatigue.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of Alloy 800H, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix F – Crack.



To develop test matrices for creep-fatigue properties of Alloy 800H, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix F – Creep-Fatigue.

8.5

Test Matrices of Stainless Steel 316FR

Some data already exit for stainless steel 316FR. To develop test matrices for the alloy, information listed as follows may be helpful: •

Test matrices for tensile properties of 316FR should be developed according to data needs 1) discussed in subsection 4.1.1.5 of this plan. A paper that may provide helpful information, existing data, and leads to more literature about tensile properties of 316FR is listed in Appendix G – Tensile.



Test matrices for creep properties of 316FR should be developed according to data needs 2) and 3) discussed in subsection 4.1.1.4 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix G – Creep.



Test matrices for creep properties of 316FR welds should be developed according to data needs 4) discussed in subsection 4.1.1.4 of this plan and closely coordinated with the creep test matrices for the base metal. Papers

54

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ that may provide helpful information and existing data are listed in, but not limited to, Appendix G – Weld. •

Test matrices for fatigue properties of 316FR should be developed according to data needs 5) and 6) discussed in subsection 4.1.1.4 of this plan. Papers that may provide helpful information and existing data are listed in, but not limited to, Appendix G – Fatigue.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of 316FR, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix G – Crack.



To develop test matrices for creep-fatigue properties of 316FR, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix G – Creep-Fatigue.

8.6

Test Matrices of Stainless Steel 316H

Because 316H is qualified for nuclear service covered in current ASME B&PV Code Subsection NH, a significant amount of data already exists. To develop test matrices for the alloy, information listed as follows may be helpful: •

The NSMH covers the standard Type 316 stainless steel, but also has a section on 20% cold worked 316 whose chemistry appears to be 316H.



Test matrices for creep properties of 316H should be developed according to data needs 1) discussed in subsection 4.1.1.6 of this plan. A paper that may provide helpful information, existing data, and leads to more literature is listed in Appendix H – Creep and Fatigue.



Test matrices for fatigue properties of 316H should be developed according to data needs 2) discussed in subsection 4.1.1.6 of this plan. A paper that may provide helpful information, existing data, and leads to more literature is listed in Appendix H – Creep and Fatigue.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of 316H, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix H – Crack.

8.7 Test Matrices of Grade 91 Because Grade 91 is qualified for nuclear service covered in current ASME B&PV Code Subsection NH, a significant amount of data already exits. To develop test matrices for the alloy, information listed as follows may be helpful:

55

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ •

Grade 91 is one of the alloys included in the Advanced Fuel Cycle Initiative Materials Handbook (AFCIMH) developed by DOE in support of particle accelerator applications. To develop test matrices for this alloy, information in AFCIMH listed in Appendix I - General must be reviewed.



Test matrices for creep properties of Grade 91 in helium environment should be developed according to data needs 1) discussed in subsection 4.1.1.7 of this plan. A literature survey needs to be conducted for relevant information or existing data.



Test matrices for fatigue properties of Grade 91 in helium environment should be developed according to data needs 2) discussed in subsection 4.1.1.7 of this plan. A literature survey needs to be conducted for relevant information or existing data.



To develop test matrices for crack growth (creep, fatigue and creep-fatigue cracks) properties of Grade 91, a paper that may provide helpful information, existing data, and leads to more literature is listed in Appendix I – Crack.



To develop test matrices for toughness properties of Grade 91, papers that may provide helpful information and existing data are listed in, but not limited to, Appendix I – Toughness.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX A: TEST PARTICIPANTS The testing activities to generate high temperature metallic materials data for the design and construction of the NGNP reactor system will be participated in by various organizations from government, universities, and private industries. This section is dedicated to keeping a record of all the participants for coordination purposes. New participants will be added as they join in the testing activities. The test plan coordinator is responsible for keeping this section updated. Participants are listed in alphabetical order of their names. Organization & Acronym Xa Ya Za Cop. XYZ

Test Matrix ID 316FR-AgedCreep-XYZ092305

Contact Person John Dow

57

Phone & Email

Note

987-654-3210 [email protected]

This is a template

HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX B: REFERENCES [1]

Hayner, G O; Corwin, W R; Jones, R. et al., “Next Generation Nuclear Plant Materials Selection and Qualification Program Plan”, prepared for the U.S. Department of Energy, Assistant Secretary For Office of Nuclear Energy, Under DOE Idaho Operations Office, Contract DE-AC07-99ID13727, INEEL/EXT-0301128, Revision 0, November 7, 2003.

[2]

“ASME Boiler and Pressure Vessel Code, an International Code”, ASME Bolier and Pressure Vessel Committee, American Society of Mechanical Engineers, New York, NY.

[3]

Corum, J M, “High-Temperature Structural Design Methodology and Associated Materials Data Requirements for Generation IV reactor Components”, position paper for Gen IV project, Oak Ridge National Laboratory, April 2003.

[4]

Natesan, K; Purohit, A; Tam, S W and Greene, C A, “Materials Behavior in HTGR Environments”, prepared for Division of Engineering Technology, Office of Nuclear Regulatory Research, U. S. Nuclear Regulatory Commission, NRC Job Code Y6537, NUREG/CR-6824, ANL-02/37, July, 2003.

[5]

Rittenhouse, P L, “Gen IV Materials Database FY 2004 Milestone”, Gen IV project task communication, Oak Ridge National Laboratory, August 2004, Data Quality Criteria updated in November, 2004.

[6]

“Quality Assurance Requirements for Nuclear Facility Applications”, an American national standard, The American Society of mechanical Engineers, ASME NQA-12000,

[7]

“ASME Boiler and Pressure Vessel Code, an International Code, Subsection XI, Rules for Inservice Inspection of Nuclear Power Plant Components”, ASME Bolier and Pressure Vessel Committee, American Society of Mechanical Engineers, New York, NY.

[8]

Randy Nanstad, Lou Mansur, and Arthur Rowcliffe, “Plan for Qualification of Materials for Radiation Service, Generation IV Reactor Materials Program”, Oak Ridge National Laboratory, Revision 1, August 31, 2004.

[9]

Wilson, D, “Aging and Environmental Effects”, proposed activities for environmetal effects on the NGNP candidate materials, submitted to the NGNP materials management, July 2004.

[10] “Annual Book of ASTM Standards, Metals Test Methods and Analytical Procedures”, Vol. 03.01, American Society for Testing and Materials, 2004.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX C: INFORMATION ABOUT ALLOY 617 Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about Alloy 617. General: • ASME Task Force Very High-Temperature Design, “Draft Alloy 617 Code Case”, Cases of ASM Boiler and Pressure Vessel Code, July 27, 1988. • Baldwin, D H; Kimball, O F; Williams, R A, “Design Data for Reference Alloys: Inconel 617 and Alloy 800H”, Prepared by General Electric Company for the Department of Energy, Contract DE-AC03-80ET34034, April 1986. • Corum, J M and Blass, J J, “Rules for Design of Alloy 617 Nuclear Components to Very High Temperatures,” pp. 147-153, PVP – Vol. 215, Fatigue, Fracture, and Risk, Am. Soc. of Mechanical Engineers, 1991. • ASME, “Proposed Code Case and Supporting data Package Covering Ni-Cr-Co-Mo Alloy UNS 06617”, Robert W. Swindeman’s personal communication with Tom Johnson. • McCoy, H E; King, J F; “Mechanical Properties of Inconel 617 and 618”, Oak Ridge National Laboratory, ORNL/TM-9337, 1985. • Penkalla, H J; Over, H H; Schubert, F; Nickel, H, “Design Values of Inconel 617 for High Temperature Reactors Operating at Temperatures Above 800ºC”, Nuclear Engineering and Design, Vol. 83, No. 3, pp. 397-402. Dec. (II) 1984. Tensile: • Abd El-Azim, M E; Mohamed, K E; Hammad, F H, “The Deformation Characteristics Of Alloy 800H And Alloy 617”, Mechanics of Materials. Vol. 14, No. 1, pp. 33-46. Nov. 1992. • Bruch, U; Schuhmacher, D, “Tensile And Impact Properties Of Candidate Alloys For High-Temperature Gas-Cooled Reactor Applications”, Nuclear Technoloty, Vol. 66, August, 1984. • Hosier, J C; Tillack, D J, “Inconel Alloy 617-A New High-Temperature Alloy”, Materials Engineering Quarterly, August, 1992. • Kewther, M A; Hashmi, M S J; Yilbas, B S, “Tensile And Fatigue Testing Of Inconel 617 Alloy After Heat Treatment And Electrochemical Tests”, Industrial Lubrication and Tribology. Vol. 53, No. 3, pp. 112-118. May-June 2001. • Laanemae, W M; Bothe, K; Gerold, V, “High Temperature Mechanical Behaviour Of Alloy 617. II. Lifetime And Damage Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 12, pp. 847-857. Dec. 1989. • Laanemae, W M; Bothe, K; Gerold, V, “High-Temperature Mechanical Behaviour Of Alloy 617. Deformation Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 11, pp. 788-799. Nov. 1989. • Schwertel, J; Merckling, G; Hornberger, K; Schinke, B; Munz, D, “Experimental Investigations On The Ni-Base Superalloy IN617 And Their Theoretical

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Description”, MD-Vol. 26/AMD-Vol. 121, High Temperature Constitutive Modeling – Theory and Application, ASME 1991. • Strizak, J. P. et al., “High-Temperature Low-Cycle Fatigue And Tensile Properties Of Hastelloy X And Alloy 617 In Air And HTGR Helium”, 1981, 15 pp. refs. Presented at IAEA Specialists Meeting on High-Temperature Materials for Application in GasCooled Reactors”, Vienna, 4 May 1981, (CONF-810530-4) Avail: NTIS. • Whittenberger, J D, “Tensile Properties Of Haynes Alloy 230 And Inconel 617 After Long Exposures To LiF-22CaF2 And Vacuum At 1093K”, Journal of Materials Engineering and Performance. Vol. 3, No. 6, pp. 763-774. Dec. 1994. • Yates, D H; Ganesan, P; Smith, G D, “Recent Advances In The Enhancement Of Inconel Alloy 617 Properties To Meet The Needs Of The Land Based Gas Turbine Industry”, Advanced Materials and Coatings for Combustion Turbines; Pittsburgh, Pennsylvania; United States; 17-21 Oct. 1993. pp. 89-97. 1994. Creep: • Abd El-Azim, M E; Mohamed, K E; Hammad, F H, “The Deformation Characteristics Of Alloy 800H And Alloy 617”, Mechanics of Materials. Vol. 14, No. 1, pp. 33-46. Nov. 1992. • Allen, D; Keustermans J P; Gijbels S; Bicego V, “Creep Rupture And Ductility Of As-Manufactured And Service-Aged Nickel Alloy IN617 Materials And Welds”, white paper • An, S U, “ Primary Creep And Microstructure Of Inconel 617 At 1073K”, Journal of the Korean Institute of Metals and Materials. Vol. 30, No. 3, pp. 247-255. Mar. 1992. • Baldwin, D H; Kimball, O F; Williams, R A, “Design Data For Reference Alloys: Inconel 617 And Alloy 800H”, Prepared by General Electric Company for the Department of Energy, Contract DE-AC03-80ET34034, April 1986. • Bassford, T H; Schill, T V, “A Review Of Inconel Alloy 617 And Its Properties After Long-Time Exposure To Intermediate Temperatures”, Applications of Materials for Pressure Vessels and Piping. MPC-10; San Francisco, Calif ; U.S.A ; June 1979. pp. 1-12. 1979. • Cook, R H, “Creep Properties Of Inconel-617 In Air And Helium At 800 To 1000ºC”, Nuclear Technology. Vol. 66, No. 2, pp. 283-288. Aug. 1984. • Cook, R H, “Creep Properties Of Inconel 617 In Air And Helium At 800 To 1000ºC”, Nuclear Technology, 66, 283, 1984. • Czyrska-Filemonowicz, A; Ennis, P; Schuster, H; Nickel, H, “Microstructural Evolution Of Inconel Alloy 617 During Creep Deformation In Air And In Decarburization Environment”, Prace Komisji Metalurgiczno-Odlewniczej, Polska Akademia Nauk--Oddzial W Krakowie, Metalurgia , No. 39, pp. 71-93. 1990. • Ennis, P J; Quadakkers, W J; Schuster, H, “Effect Of Selective Oxidation Of Chromium On Creep Strength Of Alloy 617”, Materials Science and Technology. Vol. 8, No. 1, pp. 78-82. Jan. 1992. • Han, Y H; Dzo, M H, “The Actual True Stress Change Due To Pore Formation In Inconel 617 Creep Test”, J. Korean Inst. Met. Vol. 23, No. 5, pp. 489-494. May 1985. • Hosier, J C; Tillack, D J, “Inconel Alloy 617-A New High-Temperature Alloy”, Materials Engineering Quarterly, August, 1992.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Hosoi, Y; Abe, S, “Effect Of Helium Environment On The Creep Rupture Properties Of Inconel 617 At 1000ºC”, Metall. Trans. A. Vol. 6A, No. 6, pp. 1171-1178. June 1975. • Khasin, G A; Brazgin, I A; Ermanovich, N A; Eremeev, V I; Tarnovsky, V I, “Resistance To Deformation Of High-Temperature Alloys EI 617 And EI 427B, And Steel EI481”, IZVEST VUZ Chernaya Met, --7--, 101-105, 1969. • Kihara, S; Newkirk, J B; Ohtomo, A; Saiga, Y, “Morphological Changes Of Carbides During Creep And Their Effects On The Creep Properties Of Inconel 617 At 1000ºC”, Metall. Trans. A. Vol. 11A, No. 6, pp. 1019-1031. June 1980. • Kozyrskii, O I; Petrunin, G A; Tikhonov, L V, “Effect Of The Conditions Of Preliminary Heat Treatment On The Structure And Properties Of The Alloy EI 617 During Creep”, Probl. Prochn , No. 1, pp. 39-43. Jan. 1976. • Laanemae, W M; Bothe, K; Gerold, V, “High Temperature Mechanical Behaviour Of Alloy 617. II. Lifetime And Damage Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 12, pp. 847-857. Dec. 1989. • Laanemae, W M; Bothe, K; Gerold, V, “High-Temperature Mechanical Behaviour Of Alloy 617. Deformation Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 11, pp. 788-799. Nov. 1989. • Laanemaee, W M, “Damage Of Nickel Base Alloy 617 In The Case Of Creep And High Temperature Change Deformation”, pp. 196, 14 Feb. 1989. • Mino, K; Ohtomo, A; Saiga, Y, “Effect Of Grain Boundary Migration And Recrystallization On The Creep Strength Of Inconel 617”, Tetsu-to-Hagane. Vol. 63, No. 14, pp. 2372-2380. Dec. 1977. • Mino, K; Ohtomo, A, “Effect Of Grain Boundary Migration And Recrystallization On The Creep Strength Of Inconel 617”, Trans. Iron Steel Inst. Jpn. Vol. 18, No. 12, pp. 731-738. 1978. • Mino, K; Kitagawa, M; Ohtomo, A; Saiga, Y, “Effect Of Thermal And Mechanical History On The Creep Rate Of Inconel 617”, Trans. Iron Steel Inst Jpn. Vol. 20, No. 12, pp. 826-832. 1980. • Mino, K; Kitagawa, M; Ohtomo, A; Fukagawa, M, “Creep Rupture Properties Of Inconel 617 In A Simulated HTR [High Temperature Gas-Cooled Reactor]-Helium Atmosphere”, Tetsu-to-Hagane (J. Iron Steel Inst. Jpn.). Vol. 68, No. 3, pp. 477-485. Mar. 1982. • Ohnami, M; Imamura, R, “Effect Of Vacuum Environment On Creep Rupture Properties Of Inconel 617 At 1000ºC (Especially Based On Crack Initiation And Propagation)”, Bull. Jpn. Soc. Mech. Eng. Vol. 24, No. 95, pp. 1530-1536. Sept. 1981. • Osthoff, W; Schuster, H; Ennis, P J; Nickel, H, “Creep And Relaxation Behavior Of Inconel-617”, Nuclear Technology. Vol. 66, No. 2, pp. 296-307. Aug. 1984. • Osthoff, W; Schuster, H; Ennis, P J; Nickel, H, “The Creep And Relaxation Behavior Of Inconel 617”, Creep and Fracture of Engineering Materials and Structures, Vol. 1; Swansea; U.K ; 1-6 Apr. 1984. pp. 307-318. 1984. • Osthoff, W; Schuster, H and Ennis, P, “Creep And Relaxation Behavior Of Inconel 617”, Nuclear Technology, 66, 383, 1984.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Schneider, K; Ilschner, B, “Creep Behavior Of Materials For High-Temperature Reactor Application”, Nuclear Technology, Vol. 66, August 1984. • Schubert, F; Bruch, U; Cook, R; Diehl, H; Ennis, P J; Jakobeit, W H; Penkalla, J; Heesen, E T and Ullrich, G, “Creep Rupture Behavior Of Candidate Materials For Nuclear Process Heat Applications”, Nuclear Technology, 66, 227, 1984. Draft copy collected. • Schwertel, J; Merckling, G; Hornberger, K; Schinke, B; Munz, D, “Experimental Investigations On The Ni-Base Superalloy IN617 And Their Theoretical Description”, MD-Vol. 26/AMD-Vol. 121, High Temperature Constitutive Modeling – Theory and Application, ASME 1991. • Tanabe, T; Sakai, Y; Shikama, T; Fujitsuka, M; Yoshida, H and Watanabe, R, “Creep Rupture Properties Of Superalloys Developed For Nuclear Steelmaking”, Nuclear Technology, 66, 260, 1984. • Yamamoto, S; Fujiwara, A, “Creep Rupture Strength Of Inconel 617 Pipe-To-Pipe Interconnections Subjected To Internal Pressure”, Pressure Engineering (Japan) (Atsuryoku Gijitsu-Journal of the High Pressure Institute of Japan). Vol. 20, No. 3, pp. 121-128. Mar. 1982. Biaxial Creep: • Penkalla, H J; Schubert, F; Nickel, H, “Torsional Creep Of Alloy 617 Tubes At High Temperature”, Superalloys 1988; Champion, Pennsylvania; USA; 18-22 Sept. 1988. pp. 643-652. 1988. • Penkalla, H J; Nickel, H; Schubert, F, “Multiaxial Creep Of Tubes From Incoloy 800H And Inconel 617 Under Static And Cyclic Loading Conditions”, Nuclear Engineering and Design. Vol. 112, pp. 279-289. Mar. 1989. Fatigue: • Ali, M K; Hashmi, M S J; Yilbas, B S, “Fatigue Properties Of The Refurbished INCO-617 Alloy”, Journal of Materials Processing Technology. Vol. 118, No. 1-3, pp. 45-49. 3 Dec. 2001. • Burke, M A; Beck, C G, “The High-Temperature Low-Cycle Fatigue Behavior Of The Nickel-Base Alloy IN-617”, Metall. Trans. A. Vol. 15A, No. 4, pp. 661-670. Apr. 1984. • Gontareva, R G; Kozyrskii; O I, Tikhonov, L V, “Structural Changes In Alloy EI 617 During Thermal Fatigue”, Akad. Nauk Ukr. SSR, Metallofiz, No. 70, pp. 52-56. 1977. • Hattori, H; Kitagawa, M; Ohtomo, A, “Effect Of Grain Size On High Temperature Low-Cycle Fatigue Properties Of Inconel 617”, Tetsu-to-Hagane (Journal of the Iron and Steel Institute of Japan). Vol. 68, No. 16, pp. 2521-2520. Dec. 1982. • Hattori, H; Kitagawa, M; Ohtomo, A, “Effect Of Vacuum Environment On High Temperature Low-Cycle Fatigue Properties Of Inconel 617”, Journal of the Society of Materials Science, Japan. Vol. 32, No. 357, pp. 667-671. June 1983. • Hattori, H; Kitagawa, M; Ohtomo, A, “Effect Of Specimen Geometry On Low-Cycle Fatigue Property Of Inconel 617 At HTGR Temperatures”, Journal of the Society of Materials Science, Japan. Vol. 35, No. 391, pp. 343-349. Apr. 1986.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Kewther, M A; Hashmi, M S J; Yilbas, B S, “Tensile And Fatigue Testing Of Inconel 617 Alloy After Heat Treatment And Electrochemical Tests”, Industrial Lubrication and Tribology. Vol. 53, No. 3, pp. 112-118. May-June 2001. • Kitagawa, M; Tamura, K; Ohtomo, A, “Fracture Criterion Of Inconel 617 Under Cyclic Straining At High Temperatures”, Journal of the Society of Materials Science, Japan. Vol. 32, No. 357, pp. 662-666. June 1983. • Kufaev, V N; Ishchenko, I I; Pogrebnyak, A D; Sinaisky, B N, “Effect Of The Unsteady State Of The Loading On The Fatigue Life And Strength Of The Alloy EI617 At High Temperature”, FIZ-KHIM MEKHAN MAT. Vol. 5, No. 2, pp. 142145. Mar.-Apr. 1969. • Laanemae, W M; Bothe, K; Gerold, V, “High Temperature Mechanical Behaviour Of Alloy 617. II. Lifetime And Damage Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 12, pp. 847-857. Dec. 1989. • Laanemae, W M; Bothe, K; Gerold, V, “High-Temperature Mechanical Behaviour Of Alloy 617. Deformation Mechanisms”, Zeitschrift fur Metallkunde. Vol. 80, No. 11, pp. 788-799. Nov. 1989. • Lipscomb, W G, “Mechanical Properties And Corrosion Resistance Of Inconel Alloy 617 For Refinery Service”, W.G. Lipscomb, et. al., Corrosion 89/259, NACE, Houston, TX, 1989. • Mannan, S K; Smith, G D; Wilson, R K, “Effect Of Grain Size, Microstructure, Test Temperature, And Frequency On The Low Cycle Fatigue Properties Of Inconel Alloy 617”, Structure-Property Relationships and Correlations With the Environmental Degradation of Engineering Materials; Monterey, California; United States; 31 July-1 Aug. 1991. pp. 387-428. 1992. • Meurer, H P; Gnirss, G K H; Mergler W; Raule G; Schuster H; Ullrich G; “Investigations On The Fatigue Behavior Of High-Temperature Gas-Cooled Reactor Components”, Nuclear Technology, Vol. 66, August 1984. • Meyer-Olbersleben, F; Meyer-Olbersleben, F; Kasik, N; Ilschner, B, “ The Thermal Fatigue Behavior Of The Combustor Alloys IN 617 And Haynes 230 Before And After Welding”, Metallurgical and Materials Transactions A. Vol. 30A, No. 4, pp. 981-989A. Apr. 1999. • Mishin, E. V; Loginov, N Z; Shkanov, I N, “The Effect Of Stress Concentrations And Surface Hardening On The Fatigue Strength Of The Alloy EI 617”, TRUDY KAZANSK AVIATS INST, --102--, 31-36, 1968. • Rao, K B S; Schiffers, H; Schuster, H, “Low Cycle Fatigue Behaviour of Inconel Alloy 617”, High Temperature Alloys--Their Exploitable Potential; Petten; The Netherlands; 15-17 Oct. 1985. pp. 411-422. 1987. • Rao, K B S; Schiffers, H; Schuster, H; Nickel, H, “Influence Of Time And Temperature Dependent Processes On Strain Controlled Low Cycle Fatigue Behavior Of Alloy 617”, Metall. Trans. A. Vol. 19A, No. 2, pp. 359-371. Feb. 1988. • Schwertel, J; Merckling, G; Hornberger, K; Schinke, B; Munz, D, “Experimental Investigations On The Ni-Base Superalloy IN617 And Their Theoretical Description”, MD-Vol. 26/AMD-Vol. 121, High Temperature Constitutive Modeling – Theory and Application, ASME 1991.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Srivastava, S K; Klarstrom, D L, “The LCF Behavior Of Several Solid Solution Strengthened Alloys Used In Gas Turine Engines”, Gas Turbine and Aeroengin Congress and Exposition, June 11-14, 1990, Bressels, Belgium. • Stentz, R H; Neugebauer, R B; Merkle, R A, “Elevated Temperature, PartiaConstraint And Thermal Ratcheting Simulations Using Closed-Loop Testing Machines”, PVP-Vol. 262, High-Temperature Service and Time-Dependent Failure, ASME 1993. • Strizak, J. P. et al., “High-Temperature Low-Cycle Fatigue And Tensile Properties Of Hastelloy X And Alloy 617 In Air And HTGR Helium”, 1981, 15 pp. refs. Presented at IAEA Specialists Meeting on High-Temperature Materials for Application in GasCooled Reactors”, Vienna, 4 May 1981, (CONF-810530-4) Avail: NTIS. • Strizak, J. P. et al., “Influence Of Temperature, Environment And Thermal Aging On The Continuous Cycle Fatigue Behavior Of Hastelloy X And Inconel 617”, Oak Ridge National Lab., Tenn. Apr. 1982, 50 p., refs. (ORNL/TM-8130). Avail: NTIS. • Yates, D H; Ganesan, P; Smith, G D, “Recent Advances In The Enhancement Of Inconel Alloy 617 Properties To Meet The Needs Of The Land Based Gas Turbine Industry”, Advanced Materials and Coatings for Combustion Turbines; Pittsburgh, Pennsylvania; United States; 17-21 Oct. 1993. pp. 89-97. 1994. Biaxial Fatigue: • Meurer, H P; Hanswillemenke, H; Breitling, H; Dietz, W, “Biaxial Fatigue Tests On Thin Walled Tubes Of NiCr23Co12Mo (Inconel 617) At 950ºC. (Retroactive Coverage)”, Fatigue Under Biaxial and Multiaxial Loading; Stuttgart; Germany; 3-6 Apr. 1989. pp. 249-264. 1991 Crack Growth: • Hsu, S S, “Time-Dependent Crack Growth In A Heat-Resistant Alloy Inconel 617”, Journal of Nuclear Science and Technology (Japan). Vol. 30, No. 4, pp. 302-313. Apr. 1993. • Rodig, M; Huthmann, H; Hartnagel, W, “Creep And Fatigue Crack Growth Of The Materials X10Ni CrAlTi32 20 And NiCr22Co12Mo. (Alloy 800 And Inconel 617)”, Modern Materials (Moderne Werkstoffe); Nurnberg; Germany; 20-21 Mar. 1990. pp. 245-256. 1990. • Rodig, M; Huthmann, H; Hartnagel, W, “Fatigue And Creep Crack Growth Of Alloy 800 And Alloy 617 At High Temperatures”, Materials at High Temperatures. Vol. 10, No. 4, pp. 268-274. Nov. 1992. • Yun, H M; Ennis, P J; Nickel, H; Schuster, H, “The Effect Of High-Temperature Reactor Primary Circuit Helium On The Formation And Propagation Of Surface Cracks In Alloy 800 H And Inconel 617”, Journal of Nuclear Materials. Vol. 125, No. 3, pp. 258-272. Aug. 1984. Creep-Fatigue: • Hattori, H; Kitagawa, M; Ohtomo, A, “An Evaluation Of Creep-Fatigue/Environment Behaviors Of Inconel 617 And Hastelloy XR For HTGR

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Application”, International Conference on Creep; Tokyo; Japan; 14-18 Apr. 1986. pp. 117-122. 1986. • Meurer, H P; Breitling, H; Dietz, W, “Influence Of Hold-Time And Strain Rate On The LCF Behaviour Of Alloy 617 At 950ºC”, Low Cycle Fatigue and Elasto-Plastic Behaviour of Materials; Munich; FRG; 7-11 Sept. 1987. pp. 246-251. 1987. • Penkalla, HJ; Schubert, Florian; Nickel, Hubertus, “Description Of Creep And Fatigue Exposed Tubes Of Alloy 617”, ASME Pressure Vessels Piping Division Publication, PVP, ASME, New York, NY, (USA), Vol. 262, pp. 221-224, 1993. • Rao, K B S; Meurer, H P; Schuster, H, “Creep--Fatigue Interaction Of Inconel 617 At 950ºC In Simulated Nuclear Reactor Helium”, Materials Science and Engineering A. Vol. A104, pp. 37-51. Oct. 1988. • Shimikawa, T, "Sophisticated Creep-Fatigue Life Estimation Scheme Far Pressure Vessel Components Based On Stress Redistribution Locus Concept" PVP-Vol. 472, Elevated Temperature Design and Analysis, Nonlinear Analysis, and Plastic Components - 2004 July 25-29, 2004, PVP2004-2258. • Yukawa, S. "Re-Analysis Of Inconel 617 Fatigue Data" Committee Correspondence, ASME SG-HTD, Sept. 15, 1988. Toughness: • Bruch, U; Schuhmacher, D, “Tensile And Impact Properties Of Candidate Alloys For High-Temperature Gas-Cooled Reactor Applications”, Nuclear Technoloty, Vol. 66, August, 1984. • Kimball, O F; Lai, G Y; Reynolds, G H, “Effects Of Thermal Aging On The Microstructure And Mechanical Properties Of A Commercial Ni-Cr-Co-Mo Alloy (Inconel 617) ”, Metall. Trans. A. Vol. 7A, No. 12, pp. 1951-1952. Dec. 1976. • Krompholz, K; Grosser, E D; Ewert, K, “Determination Of J-Integral R-Curves For Hastelloy X And Inconel 617 Up To 1223 Degree K Using The Potential Drop Technique”, Z. Werkstofftech. Vol. 13, no. 7, pp. 236-244. July 1982. Others: • McCoy, H E; “Mechanical Properties Of Hastelloy X And Inconel 617 After Aging 53,000 Hours In HTGR-He”, Oak Ridge National Laboratory, ORNL/TM-9604, 1985. • Richter, F, “Thermophysical Properties Of The High Temperature Material NiCr22Co12Mo (Inconel 617)”, Materialwissenschaft und Werkstofftechnik. Vol. 19, No. 2, pp. 55-61. Feb. 1988. • Penkalla, H J; Over, H H; Schubert, F; Nickel, H, “Design Values Of Inconel 617 For High Temperature Reactors Operating At Temperatures Above 800ºC”, Nuclear Engineering and Design. Vol. 83, No. 3, pp. 397-402. Dec. (II) 1984. • Schubert, F; Breitbach, G; Nickel, H, “German Structural Design Rule KTA 3221 for Metallic HTR-Components”, High-Temperature Service and Time-Dependent Failure, PVP Vol. 262, 9-18, Am. Soc. Of Mechanical Engineers, 1993. • Yukawa, S. "Elevated Temperature Fatigue Design Curves For Ni-Cr-Co-Mo Alloy 617" JSME/ASME Joint International Conf. on Nuclear Eng., Nov. 4-7, 1991

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Ennis, P J; Mohr, K; Schuster, H, “Effect Of Carburizing Service Environments On The Mechanical Properties Of High Temperature Alloys”, Nuclear Technology, Vol. 66, August, 1984. Draft copy also collected. • Ganesan, P; Smith, G D; Yates, D H, “Performance Of Inconel Alloy 617 In Actual And Simulated Gas Turbine Environments”, Materials and Manufacturing Processes. Vol. 10, No. 5, pp. 925-938. Sept. 1995. • Mankins, W L; Hosier J C; Bassford, T H, “Microstructure And Phase Stability Of INCONEL Alloy 617”, Metallurgical Transactions, Vol. 5, December 1974. • Meyer-Olbersleben, F; Meyer-Olbersleben, F; Kasik, N; Ilschner, B, “ The Thermal Fatigue Behavior Of The Combustor Alloys IN 617 And Haynes 230 Before And After Welding”, Metallurgical and Materials Transactions A. Vol. 30A, No. 4, pp. 981-989A. Apr. 1999.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX D: INFORATION ABOUT HASTELLOY X Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about Hastelloy X. General: • “Nuclear System Materials Handbook”, Oak Ridge National Laboratory, Prepared for the U. S. Department of Energy under Contract No. DE-ACO5-840R21400. • McCoy, H E et al., “Hastelloy-X for High-Temperature Gas-Cooled Reactor Applications”, Nucl. Technol. 66, 161-74, July 1984. • McCoy, H E; King, J F, “Evaluation of Hastelloy X for Gas Cooled Reactor Applications”, Report ORNL/TM-8499 (DE83004229). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Nov.1982. • McCoy, H E J, “Mechanical Property Changes of Hastelloy X Exposed to a GasCooled Reactor Environment”, Environmental Degradation of Engineering Materials in Aggressive Environments; Blacksburg; VA; 21-23 Sept. 1981. pp. 337-347. 1981. Tensile: • Bohm, H; Closs, K D, “Effects of Strain Rate on High Temperature Mechanical Properties of Irradiated Incoloy 800 and Hastelloy X”, Radiation Effects in Breeder Reactor Structural Materials. ASTM, New York. 1977, 347-356. • Canistraro, H A; Jordan, E H; Shi, S; Favrow, L H; Reed, F A, “Elastic constants of single crystal Hastelloy X at elevated temperatures”, Journal of Engineering Materials and Technology (Transactions of the ASME). Vol. 120, no. 3, pp. 242-247. July 1998. • Discus, D L; Buckley, J D, “Effects of High-Temperature Brazing and Thermal Cycling on the Mechanical Properties of Hastelloy X”, Paper from "Proceedings of the Symposium on Welding, Bonding and Fastening," NASA, 1972. • Fujioka, J; Murase, H; Matsuda, S, “Effect of Grain Size and Cold Working on HighTemperature Strength of Hastelloy X”, Pressure Vessel Technology, Vol. 1; London; England; 19-23 May 1980. pp. 197-203. 1980. • Fujioka, J; Murase, H, “Tensile Properties of Hastelloy X and Incoloy 800 Exposed to Air and Helium at High Temperature”, Tetsu-to-Hagane (J. Iron. Steel. Inst. Jpn. Vol. 64. no. 10, pp. 1588-1597. Sept. 1978. • Kondo, Y; Fukaya, K; Kunitomi, K; Miyamoto, Y, “Tensile and Impact Properties Changes of Hastelloy X After Exposure in High-Temperature Helium Environment”, Metall. Trans. A. Vol. 19A, no. 5, pp. 1269-1275. May 1988. • Kotval, P S, “Discussion On A Paper By J S Levy, B Mastel And J L Brimhall On..-The Influence Of Thermomechanical Treatments On The Microstructure And Tensile Properties Of Hastelloy X-280”, TRANS MET SOC AIME. Vol. 242, no. 1, pp. 163164. Jan. 1968.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Levy, I S; Mastel, B; Brimhall, J L “Influence Of The Thermomechanical Treatments On The Strength High-Temperature Stability And Microstructure Of Hastelloy X280”, AIME MET SOC TRANS. Vol. 239, no. 2, pp. 204-209. Feb. 1967. • McCoy, H E et al., “Hastelloy-X For High-Temperature Gas-Cooled Reactor Applications”, Nucl. Technol. 66, 161-74, July 1984. • McCoy, H E; Strizak, J P; King, J F, “Hastelloy-X for high-temperature gas-cooled reactor applications”, Nuclear Technology. Vol. 66, no. 1, pp. 161-174. July 1984. • McCoy, H E; King, J F, “Creep and tensile properties of Alloy 800H/Hastelloy X weldments”, Report ORNL/TM-8728 (DE83017067). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Aug.1983. • McCoy, H E; King, J F, “Evaluation of Hastelloy X for gas cooled reactor applications”, Report ORNL/TM-8499 (DE83004229). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Nov.1982. • Richter, F, “Thermophysical Properties of the High Temperature Nickel-Base Alloys Hastelloy S, Hastelloy X and Hastelloy C276 Between 20-1000ºC”, Metall. Vol. 42, no. 12, pp. 1188-1191. Dec. 1988. • Strizak, J. P. et al., “High-Temperature Low-Cycle Fatigue And Tensile Properties Of Hastelloy X And Alloy 617 In Air And Htgr-Helium”, Presented at IAEA Specialists Meeting on High-Temperature Materials for Application in Gas-Cooled Reactors, Vienna, 4 May 1981, (CONF-810530-4) Avail: NTIS. • Thiele, B A; Schubert, F; Derz, H; Pott, G, “Influence of Test Temperature on Post Irradiation, High Temperature Tensile and Creep Properties of X8CrNiMoNb16 16, X10NiCrAlTi32 20 (Alloy 800) and NiCr22Fe18Mo (Hastelloy X)”, Journal of Nuclear Materials. Vol. 171, no. 1, pp. 94-102. Apr. 1990. Creep: • Arata, Y; Susei, S; Shimizu, S; Satoh, K; Nagai, H, “Creep properties of EB [electron beam] welded joint on Hastelloy X”, Journal of the Japan Welding Society. Vol. 49, no. 11, pp. 755-760. Nov. 1980. • Bohm, H; Closs, K D, “Effects of Strain Rate on High Temperature Mechanical Properties of Irradiated Incoloy 800 and Hastelloy X”, Radiation Effects in Breeder Reactor Structural Materials. ASTM, New York. 1977, 347-356. • Fujioka, J; Murase, H; Matsuda, S, “Effect of Grain Size and Cold Working on HighTemperature Strength of Hastelloy X”, Pressure Vessel Technology, Vol. 1; London; England; 19-23 May 1980. pp. 197-203. 1980. • Fujioka, J; Murase, H; Matsuda, S, “Effect of Grain Size on Creep and Creep Rupture Properties of Hastelloy X”, J. Jpn. Inst. Met. Vol. 43, no. 11, pp. 1078-1085. Nov. 1979. • Hada, K; Mutoh, Y, “Study on the Creep Constitutive Equation of Hastelloy X. I.-Generation of a Creep Constitutive Equation and Its Sensitivity Analysis”, Bulletin of the JSME. Vol. 26, no. 221, pp. 1839-1848. Nov. 1983. • Inouye, H; Rittenhouse, P L, “Relationship Between Carburization And ZeroApplied-Stress Creep Dilation In Alloy 800 H And Hastelloy X”, Oak Ridge National Lab., TN., 1981, 12 p., (CONF-810530-6), NTIS. From IAEA specialists' meeting on

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ high-temperature metallic materials for application in gas-cooled reactors; Vienna, Austria, 4 May 1981. • Inouye, H, “Carburization and Dimensional Stability of Hastelloy X and Alloy 800H in HTGR Helium”, 1979 Winter Meeting; San Francisco; Calif ; 11-15 Nov. 1979. pp. 297-298. 1979 & Trans, Am. Nucl. Soc., Vol. 33, p. 297-98 November 1979. • Kikuchi, A; Nakanishi, T; Shin, S; Takatsu, T; Saitoh, T, “Creep Property of the Hastelloy X Weld Metals Obtained by Filler Metals With Minor Elements”, Quarterly Journal of the Japan Welding Society. Vol. 4, no. 2, pp. 371-377. May 1986. • Kiuchi, K; Kondo, T, “Metallurgical and Environmental Factors Influencing Creep Behavior of Hastelloy-X”, Japan Atomic Energy Research Inst., Tokai, Ibaraki. pp. 77-90 of Proceedings of the second U.S.-Japan seminar on HTGR safety technology, 2. Material properties and design methods; seismic research. Tokai, Ibaraki; JAERI (Jun 1979). From 2. U.S.-Japan seminar on HTGR safety technology; Fuji, Japan, 24 Nov 1978. • Kiyoshige, M; Murase, H; Fujioka, J; Shimizu, S; Satoh, K, “Creep Properties of Hastelloy X \Heats\ and Their Application to the Structural Design”, Trans. Iron Steel Inst. Jpn. Vol. 18, no. 7, pp. 397-403. 1978. • Kiyoshige, M; Murase, K; Fujioka, J; Shimizu, S; Satoh, K, “Creep Properties of (NiBase Superalloy) Hastelloy X and Their Application to Structural Design”, J. Soc. Mater. Sci. Jpn. Vol. 26, no. 282, pp. 248-254. Mar. 1977. • Kondo, Y; Matsuo, T; Shinoda, T; Tanaka, R, “Effect of Grain Size on the High Temperature Creep Properties of Hastelloy X”, Tetsu-to-Hagane (J. Iron Steel Inst. Jpn.). Vol. 67, no. 10, pp. 1805-1814. Aug. 1981. • Kurata, Y; Ogawa, Y, “Internal Stress During High-Temperature Creep of Special Grade Hastelloy X Alloys”, Journal of Nuclear Materials. Vol. 158, pp. 42-48. Aug.Sept. 1988. • Kurata, Y; Ogawa, Y; Kondo, T, “Creep and Rupture Behavior of a Special Grade Hastelloy-X in Simulated HTGR Helium”, Nuclear Technology. Vol. 66, no. 2, pp. 250-259. Aug. 1984. • Kurihara, R; Ueda, S, “Study of Internal Pressure Creep Strength of Hastelloy X Cylindrical Specimen Containing an Axial Surface Notch”, International Journal of Pressure Vessels and Piping. Vol. 30, no. 1, pp. 37-56. 1987. • Lee, K S, “Creep Rupture Properties of Hastelloy-X and Incoloy-800H in a Simulated HTGR Helium Environment Containing High Levels of Moisture”, Nuclear Technology. Vol. 66, no. 2, pp. 241-249. Aug. 1984. • Levy, I S; Mastel, B; Brimhall, J L “Influence Of The Thermomechanical Treatments On The Strength High-Temperature Stability And Microstructure Of Hastelloy X280”, AIME MET SOC TRANS. Vol. 239, no. 2, pp. 204-209. Feb. 1967. • McCoy, H E; King, J F, “Creep and tensile properties of Alloy 800H/Hastelloy X weldments”, Report ORNL/TM-8728 (DE83017067). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Aug.1983. • McCoy, H E, “Creep Behavior of Hastelloy X, 2.25Cr -- 1Mo Steel and Other Alloys in Simulated HTGR Helium”, Report of Oak Ridge National Laboratory, ORNL/TM6822, June 1979.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Murase, H; Fujioka, J; Kita, K; Shimizu, S, “Creep and Creep Rupture Properties of Hastelloy X in Air and He at 900 C”, J. At. Energ. Soc. Jpn. Vol. 18, no. 10, pp. 641645. 1976. • Nakanishi, T; Kawakami, H, “Creep Properties of Hastelloy-X in Impure Helium Environments”, Nuclear Technology. Vol. 66, no. 2, pp. 273-282. Aug. 1984. • Nakanishi, T; Kawakami, H, “Creep Properties Of Hastelloy X In A Carburizing Helium Environment”, (Fuji Electric Corporate Research and Development, Ltd., Yokosuka, Japan), 1982, NTIS, BNL-NUREG--51674-Vol. 2). From 3. Japan/US HTGR safety technology seminar; Upton, NY, 2 Jun 1982. • Nakanishi, T; Matsumoto, N, “Creep Property Of Hastelloy X And Incoloy 800 In A Helium Environment”, Fuji Electric Co. Ltd., Yokosuka, Kanagawa (Japan). pp. 115124 of Proceedings of the second U.S.- Japan seminar on HTGR safety technology, 2. Material properties and design methods; seismic research. Tokai, Ibaraki; JAERI (Jun 1979). From 2. U.S.-Japan seminar on HTGR safety technology; Fuji, Japan, 24 Nov 1978. • Nakanishi, T; Matsumoto, N; Kawata, O, “Study on the Effect of He Environment and the Grain Size on the Creep Behaviour of Hastelloy X Alloy”, J. Jpn. Inst. Met. Vol. 41, no. 3, pp. 263-269. Mar. 1977. • “Studies on the Structural Behaviour of Hastelloy X Cylinders With Welded Joints Under Internal Pressure at High Temperature. III.--Creep Behaviour Under Internal Pressure of a Hastelloy X Cylinder With an EB [Electron Beam] Butt Welded Joint”, Journal of the Japan Welding Society. Vol. 51, no. 1, pp. 21-26. July 1982. • Satoh, K; Toyoda, M; Matsui, S; Mori, E; Shimizu, S, “Creep behaviour evaluation of Hastelloy X welded joints”, Nuclear Technology. Vol. 55, no. 2, pp. 479-486. Nov. 1981. • Shimizu, S; Ltd, K H I, “Effects Of High Temperature Environment On Creep Properties Of Hastelloy X”, Kawasaki Heavy Industries Ltd., Kobe (Japan) pp. 106114 of Proceedings of the second U.S.-Japan seminar on HTGR safety technology, 2. Material properties and design methods; seismic research. Tokai, Ibaraki; JAERI (Jun 1979). From 2. U.S.-Japan seminar on HTGR safety technology; Fuji, Japan, 24 Nov 1978. • Tamura, M; Kondo, T, “Surface Reactions Associated With Creep Deformation Of Hastelloy-X In Simulated Htgr Environment”, Japan Atomic Energy Research Institute. Presented at the 102nd ISIJ meeting, Lecture No. S1264, November 198l. • Thiele, B A; Schubert, F; Derz, H; Pott, G, “Influence of Test Temperature on Post Irradiation, High Temperature Tensile and Creep Properties of X8CrNiMoNb16 16, X10NiCrAlTi32 20 (Alloy 800) and NiCr22Fe18Mo (Hastelloy X)”, Journal of Nuclear Materials. Vol. 171, no. 1, pp. 94-102. Apr. 1990. • Wassilew, C; Ehrlich, K; Bergmann, H J, “Analysis of the In-Reactor Creep and Rupture Life Behavior of Stabilized Austenitic Stainless Steels and the Nickel-Base Alloy Hastelloy-X”, Influence of Radiation on Material Properties: 13th International Symposium. II; Seattle, Washington; USA; 23-25 June 1986. pp. 30-53. 1987. • Yoshioka, Y; Okabe, N; Saito, D; Fujiyama, K, “Effects of microstructural change and stress on creep strength of Hastelloy X”, Journal of the Society of Materials Science, Japan. Vol. 43, no. 484, pp. 36-40. Jan. 1994.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Yoshioka, Y; Saito, D; Fujiyama, K; Okabe, N; Nakamura, S, “Effect of Long Term Exposure on Mechanical Properties and Precipitation Behaviors of Hastelloy X for Gas Turbine Combustor”, Tetsu-to-Hagane (Journal of the Iron and Steel Institute of Japan). Vol. 79, no. 11, pp. 1293-1298. Nov. 1993. Weld: • Blue, C A; Blue, R A; Lin, R Y; Lei, J F; Williams, W D, “Joining of Hastelloy X to Inconel 718 using an infrared process”, Journal of Materials Processing Technology. Vol. 58, no. 1, pp. 32-38. 1 Mar. 1996. • Isobe, N; Sakurai, S, “Micro-crack growth behavior in weldments of the Ni-base superalloy Hastelloy-X under biaxial low-cycle fatigue and high temperature”, Journal of the Society of Materials Science, Japan. Vol. 51, no. 2, pp. 221-226. Feb. 2002. • King, J F; Mccoy, H E, “Weldability evaluations and weldment properties of Hastelloy X”, Report CONF-810530-5. Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; 1981. • Li, Z; Gobbi, S L; Richter, K H, “Autogenous welding of Hastelloy X to Mar-M 247 by laser”, Journal of Materials Processing Technology. Vol. 70, no. 1-3, pp. 285-292. Oct. 1997. • McCoy, H E; Strizak, J P; King, J F, “Hastelloy-X for high-temperature gas-cooled reactor applications”, Nuclear Technology. Vol. 66, no. 1, pp. 161-174. July 1984. • Satoh, K; Toyoda, M; Matsui, S; Mori, E; Shimizu, S, “A study on creep behaviour evaluation for welded joint [TIG and EB welded Hastelloy X]”, Fundamental and Practical Approaches to the Reliability of Welded Structures, Proceedings, 4th International JWS Symposium, Osaka, vol.1, Paper III-6; 24-26 Nov. 1982. pp. 255260. 1982. • Shimizu, S; Mutoh, Y, “Weldability and weld performance of a special grade Hastelloy-X modified for high-temperature gas-cooled reactors”, Nuclear Technology. Vol. 66, no. 1, pp. 44-53. July 1984. • Shin, S; Takatsu, T; Saitoh, R; Nakanishi, T; Kukuchi, A, “The weldability of Hastelloy X filler metals with addition of minor elements”, Quarterly Journal of the Japan Welding Society. Vol. 4, no. 2, pp. 365-371. May 1986. • Suzumura, A; Onzawa, T; Tamura, H, “Solid state diffusion weldability of high temperature alloy A286 and Hastelloy X”, Transactions of the Japan Welding Society. Vol. 14, no. 2, pp. 110-116. Oct. 1983. • Toyoda, T; Hyodo, T; Endo, T, “TLP bonding of Hastelloy X with ion plating filler”, Tetsu-to-Hagane (Journal of the Iron and Steel Institute of Japan). Vol. 82, no. 6, pp. 509-513. June 1996. • Toyoda, T; Endo, T O, “Transition liquid phase bonding of a Hastelloy X and the bond strength at 1173K”, Journal of Materials Science. Vol. 31, no. 9, pp. 2461-2467. 1 May 1996. • Udoguchi, T; Kobatake, K; Satoh, K; Indo, H; [Indow, H; Nakanishi, T; Kurumaji, T, “Studies on structural behaviour of Hastelloy X cylinder with welded joint under internal pressure at elevated temperature. Report 3: Internal pressure creep behaviour

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ for Hastelloy X cylinder with EB [electron beam] butt welded joint”, Journal of the Japan Welding Society. Vol. 51, no. 1, pp. 21-26. Jan. 1982. • Udoguchi, T; Kikuchi, A; Nakanishi, T; Kobatake, K; Indow, H; Kurumaji, T, “Studies on structural behaviour of Hastelloy X cylinder with welded joint under internal pressure at elevated temperature. Report 2: Improvement of the creep property of Hastelloy X cylinder with TIG welded joint”, Journal of the Japan Welding Society. Vol. 50, no. 12, pp. 1158-1164. Dec. 1981. • Udoguchi, T; Indo, H; Isomura, K; Kobatake, K; Nakanishi, T; Kurumaji, T, “Studies on structural behaviour of Hastelloy X cylinder with welded joint under internal pressure at elevated temperature. Report 1: Creep strength of Hastelloy X TIGwelded cylinder under internal pressure at elevated temperature”, Journal of the Japan Welding Society. Vol. 50, no. 9, pp. 904-910. Sept. 1981. Fatigue: • Bohm, H; Closs, K D, “Effects of Strain Rate on High Temperature Mechanical Properties of Irradiated Incoloy 800 and Hastelloy X”, Radiation Effects in Breeder Reactor Structural Materials. ASTM, New York. 1977, 347-356. • Carden, A E; Slade, T B, “High-Temperature Low-Cycle Fatigue Experiments On Hastelloy X”, Paper From Fatigue At High Temperature, Astm, Philadelphia, Pa. 1969. • Chen, L J; Liaw, P K; Wang, G Y; McDaniels, R L; Liaw, K; Thompson, S A; Blust, J W; Browning, P F; Bhattacharya, A K; Aurrecoechea, J M; Seeley, R R; Klarstrom, D L, “Cyclic Deformation Behavior And Damage Mechanisms Of HASTELLOY X Superalloy Under Fatigue And Creep-Fatigue Loading”, Fatigue: A David L. Davidson Symposium as held at the 2002 TMS Annual Meeting; Seattle, WA; USA; 17-21 Feb. 2002. pp. 191-200. 2002. • Discus, D L; Buckley, J D, “Effects of High-Temperature Brazing and Thermal Cycling on the Mechanical Properties of Hastelloy X”, Paper from "Proceedings of the Symposium on Welding, Bonding and Fastening," NASA, 1972. • Fujioka, J; Murase, H; Matsuda, S, “Effect of Grain Size and Cold Working on HighTemperature Strength of Hastelloy X”, Pressure Vessel Technology, Vol. 1; London; England; 19-23 May 1980. pp. 197-203. 1980. • Huang, J S; Pelloux, R M, “Low-Cycle Fatigue Crack Propagation in Hastelloy-X at 25 and 760ºC”, Metall. Trans. A. Vol. 11A, no. 6, pp. 899-904. June 1980. • McCoy, H E J, “Mechanical Property Changes of Hastelloy X Exposed to a GasCooled Reactor Environment”, Environmental Degradation of Engineering Materials in Aggressive Environments; Blacksburg; Va ; 21-23 Sept. 1981. pp. 337-347. 1981. • Miner, R V; Castelli, M G, “Hardening Mechanisms in a Dynamic Strain Aging Alloy, HASTELLOY X, During Isothermal and Thermomechanical Cyclic Deformation”, Metallurgical Transactions A (USA). Vol. 23A, no. 2, pp. 551-561. Feb. 1992. • Nishiguchi, I; Muto, Y; Tsuji, H, “(JAERI-M-83-224) Design Fatigue Curve for Hastelloy-X”, Japan Atomic Energy Research Institute, pp. 44, NTIS, (US Sales Only), File Number DE85701031, Dec. 1983.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Seaver, D W, “Low Cycle Fatigue Properties of Hastelloy X at 704ºC (1300 deg F)”, Methods for Predicting Material Life in Fatigue [Proc. Conf.], New York, U.S.A., Dec. 1979, pp. 133-144, 1979. • Shimizu, S; Ikemoto, Y, “High Temperature Low-Cycle Fatigue Strength of Hastelloy X. (English translation: BISI 19511)”, J. Soc. Mater. Sci. Jpn. Vol. 26, no. 282, pp. 255-261. 1977. • Strizak, J P et al., “Influence Of Temperature, Environment And Thermal Aging On The Continuous Cycle Fatigue Behavior Of Hastelloy X And Inconel 617”, Oak Ridge National Lab Report, ORNL/TM-8130, Avail: NTIS, Apr. 1982. • Strizak, J. P. et al., “High-Temperature Low-Cycle Fatigue And Tensile Properties Of Hastelloy X And Alloy 617 In Air And Htgr-Helium”, Presented at IAEA Specialists Meeting on High-Temperature Materials for Application in Gas-Cooled Reactors, Vienna, 4 May 1981, (CONF-810530-4) Avail: NTIS. • Suzuki, H; Iseki, T; Shoda, Y, “High-Temperature Low-Cycle Fatigue Tests on Hastelloy X”, J. Nucl. Sci. Tech. Vol. 14, no. 5, pp. 381-386. May 1977. • Tsuji, H and Kondo, T, “Effect of Strain Rate on High-Temperature Low-Cycle Fatigue Behavior of Hastelloy X and Hastelloy XR”, (Japanese, English summary). Sep. 1987 43 p (DE88-752302), Scientific and Technical Information Office, NASA, P.O. Box 8757, BWI Airport, MD 21240, Avail: NTIS (US Sales Only), 1987. • Tsuji, H and Kondo, T, “Low-Cycle Fatigue Of Hastelloy X And Hastelloy XR In Simulated Vhtr Helium Coolant Environment At Elevated Temperature”, (Japan Atomic Energy Research Inst., Tokyo), p. 33, NTIS, (US Sales Only), File Number DE85701032, Feb. 1984. • Uga, T, “Creep Strain Growth Behaviour of Hastelloy X Above 800ºC Under Thermal-Stress Cyclings”, International Journal of Pressure Vessels and Piping. Vol. 57, no. 3, pp. 305-309. 1994. Crack: • Isobe, N; Sakurai, S, “Micro-crack growth behavior in weldments of the Ni-base superalloy Hastelloy-X under biaxial low-cycle fatigue and high temperature”, Journal of the Society of Materials Science, Japan. Vol. 51, no. 2, pp. 221-226. Feb. 2002. • Kim, K S; Van Stone, R H, “Characterization of elevated temperature crack growth in Hastelloy-X using integral parameters”, Journal of Engineering Materials and Technology (Transactions of the ASME). Vol. 117, no. 3, pp. 299-304. July 1995. • Lu, Y L; Chen, L J; Liaw, P K; Wang, G Y; Benson, M L; Thompson, S A; Blust, J W; Browning, P F; Bhattacharya, A K; Aurrecoechea, J M, “Elevated-temperature crack-growth behavior of nickel-base HASTELLOY X alloy”, Materials Lifetime Science and Engineering as held at the 2003 TMS Annual Meeting; San Diego, CA; USA; 2-6 Mar. 2003. pp. 33-42. 2003. • Marchand, N J; Pelloux, R M; Ilschner, B, “Non-Isothermal Fatigue Crack Growth in Hastelloy-X”, Fatigue Fract. Eng. Mater. Struct. Vol. 10, no. 1, pp. 59-74. 1987. • Sakurai, S; Umezawa, S, “Fatigue Crack Growth Threshold for Distributed-SmallCracks in Hastelloy X at High Temperature”, Journal of the Society of Materials Science, Japan. Vol. 40, no. 455, pp. 1035-1041. Aug. 1991.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Suzuki, H; Ichikawa, Y; Sakurai, S; Kaneko, R I, “In-Situ Observation by ``SEM Servo'' on Fatigue Crack Initiation and Growth Behaviors in Hastelloy-X at Elevated Temperature”, Asian Pacific Conference on Fracture and Strength '93. APCFS '93 ; Tsuchiura; Japan; 26-28 July 1993. pp. 425-430. 1993. • Weerasooriya, T and Strizak, J P, “Crack Propagation In Hastelloy X”, Oak Ridge National Lab., Tenn. Metals and Ceramics Div. (ORNL/TM-7255) Avail: NTIS, 1981. • Weerasooriya, T “Fatigue crack propagation in Hastelloy X weld metal”, Report ORNL/TM-6999. Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Nov.1979.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX E: INFORMATION ABOUT HASTELLOY XR Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about Hastelloy XR. Creep: • Hasegawa, K; Okazaki, S; Matsuda, S, “Environmental Influences on High Temperature Fatigue Strength and Creep Strength of Hastelloy XR”, Journal of the Society of Materials Science, Japan. Vol. 37, no. 413, pp. 185-190. Feb. 1988. • Kaji, Y; Tsuji, H; Nishi, H; Muto, Y; Penkalla, H; Schubert, F, “Multiaxial creep behavior of nickel base heat-resistant alloys Hastelloy XR and Ni-Cr-W superalloy at elevated temperatures”, Journal of Nuclear Science and Technology (Japan). Vol. 39, no. 8, pp. 923-928. Aug. 2002. • Kaji, Y; Kikuchi, K; Yokobori, A T J, “Estimation of creep fracture life for Hastelloy XR by Q* parameter”, Engineering Fracture Mechanics. Vol. 50, no. 4, pp. 519-528. Mar. 1995. • Kikuchi, K; Kaji, Y, “Relaxation Behaviour of Hastelloy XR in LCF Tests”, Engineering Fracture Mechanics. Vol. 40, no. 4-5, pp. 749-755. 1991. • Kurata, Y; Tanabe, T; Mutoh, I; Tsuji, H; Hiraga, K; Shindo, M; Suzuki, T, “Creep properties of base metal and welded joint of Hastelloy XR produced for hightemperature engineering test reactor in simulated primary coolant helium”, Journal of Nuclear Science and Technology (Japan). Vol. 36, no. 12, pp. 1160-1166. Dec. 1999. • Kurata, Y; Nakajima, H, “Creep properties of 20% cold-worked Hastelloy XR”, Journal of Nuclear Materials. Vol. 228, no. 2, pp. 176-183. 1 Mar. 1996. • Kurata, Y; Ogawa, Y; Suzuki, T; Shindo, M; Nakajima, H; Kondo, T, “Creep Behaviour of Hastelloy XR in Simulated High-Temperature Gas-Cooled Reactor Helium”, Report of Japan Atomic Energy Research Institute, 95-037, pp. 1-42, June 1995. • Kurata, Y; Ogawa, Y; Suzuki, T; Shindo, M; Nakajima, H; Kondo, T, “Long term creep properties of hastelloy XR in simulated high-temperature gas-cooled reactor helium”, Journal of Nuclear Science and Technology (Japan). Vol. 32, no. 11, pp. 1108-1117. Nov. 1995. • Kurata, Y; Nakajima, H, “Temperature dependence of creep properties of coldworked Hastelloy XR”, Journal of Nuclear Science and Technology (Japan). Vol. 32, no. 6, pp. 539-546. June 1995. • Muto, Y; Hada, K; Koikegami, H; Ohno, N, “High Temperature Strength of Hastelloy XR Under Biaxial Stress States”, Nihon-Genshiryoku-Gakkai Shi (Journal of the Atomic Energy Society of Japan). Vol. 33, no. 5, pp. 475-481. May 1991. • Nakasone, Y; Tanabe, T; Ohba, T; Yagi, K; Tsuji, H; Nakajima, H, “Creep Damage of Hastelloy XR at Very High Temperatures in Simulated Helium Gas”, Creep: Characterization, Damage and Life Assessments, Lake Buena Vista; Florida; USA; 18-21 May 1992. pp. 551-555. 1992.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Ogawa, Y; Kurata, Y; Monma, Y; Yoshizu, H; Suzuki, T; Kondo, T, “Relation between fracture mode and carbide of Hastelloy XR creep-tested in heliulm M23C6 /M6C environment”, Tetsu-to-Hagane (Journal of the Iron and Steel Institute of Japan). Vol. 83, no. 5, pp. 329-334. May 1997. • Ogawa, Y; Kurata, Y; Suzuki, T; Nakajima, H; Kondo, T, “Creep and rupture behavior of Hastelloy XR in simulated HTGR helium”, Nihon-Genshiryoku-Gakkai Shi (Journal of the Atomic Energy Society of Japan). Vol. 36, no. 10, pp. 967-975. 1994. • Ohno, N; Murakami, S; Hattori, M; Mutoh, Y, “Creep of Hastelloy XR in steady and nonsteady multiaxial stress states at 900 degree C.”, TRANS. JAPAN SOC. MECH. ENG. (SER. A)., vol. 53, no. 491, pp. 1191-1196, 1987. • Schafer, L, “Comment on the Paper: H. Tsuji, T. Tanabe, Y. Nakasone and H. Nakajima, Applicability of Creep Damage Rules to a Nickel-Base Heat-Resistant Alloy Hastelloy XR \J. Nucl. Mater. 199 (1992) 43”, Journal of Nuclear Materials. Vol. 203, no. 2, pp. 186. Aug. 1993. • Tanabe, T; Abe, F; Sakai, Y; Okada, M, “The Effect of Boron Addition on the Creep Rupture Properties of Hastelloy XR in an Impure Helium Environment”, Trans. Iron Steel Inst. Jpn. Vol. 26, no. 11, pp. 968-976. Nov. 1986. • Tsuji, H; Tanabe, T; Nakasone, Y; Nakajima, H, “Creep Properties With Short Period Excessive Loadings on a Nickel-Base Heat-Resistant Alloy Hastelloy XR”, Journal of Nuclear Science and Technology (Japan). Vol. 31, no. 4, pp. 274-278. Apr. 1994. • Tsuji, H, “Reply to the Comment by L. Schafer on the Paper "Applicability of Creep Damage Rules to a Nickel-Base Heat- Resistant Alloy Hastelloy XR", Journal of Nuclear Materials. Vol. 203, no. 2, pp. 187. Aug. 1993. • Tsuji, H; Nakajima, H; Tanabe, T; Nakasone, Y, “Creep Properties Under Varying Stress/Temperature Conditions on Nickel-Base Heat-Resistant Alloy Hastelloy XR”, Journal of Nuclear Science and Technology (Japan). Vol. 30, no. 8, pp. 768-776. Aug. 1993. • Tsuji, H; Nakajima, H; Tanabe, T; Nakasone, Y, “Applicability of Creep Damage Rules to a Nickel-Base Heat-Resistant Alloy Hastelloy XR”, Journal of Nuclear Materials. Vol. 199, no. 1, pp. 43-49. Dec. 1992. Weld: • Kurata, Y; Tanabe, T; Mutoh, I; Tsuji, H; Hiraga, K; Shindo, M; Suzuki, T, “Creep properties of base metal and welded joint of Hastelloy XR produced for hightemperature engineering test reactor in simulated primary coolant helium”, Journal of Nuclear Science and Technology (Japan). Vol. 36, no. 12, pp. 1160-1166. Dec. 1999. • Shimizu, S; Satoh, K; Mutoh, Y; Ogawa, Y, “Evaluation of weldability of Hastelloy XR”, Journal of the Society of Materials Science, Japan. Vol. 32, no. 363, pp. 13071313. Dec. 1983. • Tanabe, T; Kurata, Y; Mutoh, I; Tsuji, H; Hiraga, K; Shindo, M, “Creep damage in welded joints of a Ni-base heat-resistant alloy Hastelloy XR”, Materials Science and Engineering A. Vol. A234-236, pp. 1087-1090. 30 Aug. 1997.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Fatigue: • Hasegawa, K; Okazaki, S; Matsuda, S, “Environmental Influences on High Temperature Fatigue Strength and Creep Strength of Hastelloy XR”, Journal of the Society of Materials Science, Japan. Vol. 37, no. 413, pp. 185-190. Feb. 1988. • Hattori, H; Kitagawa, M; Ohtomo, A, “Effect of Strain Wave Form on High Temperature Low-Cycle Fatigue Properties of Hastelloy XR in Vacuum Environment”, Journal of the Society of Materials Science, Japan. Vol. 35, no. 391, pp. 427-433. Apr. 1986. • Nagato, K; Murakami, T; Hashimoto, T, “High Temperature Low-Cycle Fatigue Strength Of Hastelloy-XR”, K. Nagato, T. Murakami, and T. Hashimoto, June 1989. (CONF-8806156-: IAEA specialists meeting on high-temperature metallic materials for gas-cooled reactors, Cracow (Poland), 20-23 June 1988). In High-temperature metallic materials for gas-cooled reactors: Proceedings of a specialists meeting held in Cracow, 20-23 June 1988. Order Number DE90635511. Source: NTIS. • Tsuji, H and Kondo T, “Effect Of Strain Rate On High-Temperature Low-Cycle Fatigue Behavior Of Hastelloy X And Hastelloy XR”, (Japanese, English summary)., Sep. 1987 43 p (DE88-752302) Avail: NTIS (US Sales Only. • Tsuji, H; Kondo T, “Low-Cycle Fatigue of Hastelloy X and Hastelloy XR in Simulated VHTR Helium Coolant Environment at Elevated Temperature”, (Japanese), Japan Atomic Energy Research Inst., Tokyo, Feb. 1984, 33 p., NTIS, (US Sales Only), File Number DE85701032. Crack: • Kaji, Y; Kikuchi, K; Sugae, H; Yokobori, T, “Creep crack growth estimated by Q* parameter for Hastelloy XR”, TRANS. JAPAN SOC. MECH. ENG. (SER. A)., vol. 58, no. 548, pp. 515-519, 1992. Creep-Fatigue • Hattori, H; Kitagawa, M; Ohtomo, A, “An Evaluation of Creep--Fatigue/Environment Behaviors of Inconel 617 and Hastelloy XR for HTGR Application”, International Conference on Creep; Tokyo; Japan; 14-18 Apr. 1986. pp. 117-122. 1986. • Muto, Y; Hada, K; Koikegami, H; Ohno, N, “High Temperature Strength of Hastelloy XR Under Biaxial Stress States”, Nihon-Genshiryoku-Gakkai Shi (Journal of the Atomic Energy Society of Japan). Vol. 33, no. 5, pp. 475-481. May 1991. • Tsuji, H; Nakajima, H, “Creep-fatigue damage evaluation of a nickel-base heatresistant alloy Hastelloy XR in simulated HTGR helium gas environment”, Journal of Nuclear Materials. Vol. 208, no. 3, pp. 293-299. Feb. 1994. • Tsuji, H; Nakajima, H, “Creep-fatigue interaction property of a nickel-base heatresistant alloy Hastelloy XR in simulated HTGR helium gas environment”, Report of Japan Atomic Energy Research Institute (Japan), JAERI-M 93-187, pp. 20, Oct. 1993.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX F: INFORMATION ABOUT ALLOY 800H Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about Alloy 800H. General: • “Nuclear System Materials Handbook”, Oak Ridge National Laboratory, Prepared for the U. S. Department of Energy under Contract No. DE-ACO5-840R21400. • “Incoloy Alloy 800H (Fe-Ni-Cr Alloy for High-Temperature Uses)”, Alloy Digest, no. SS-347, pp. 2 p. Jan. 1978. Tensile: • Abd El-Azim, M E, “Correlation between tensile and creep data in alloy 800H at 850ºC”, Journal of Nuclear Materials. Vol. 231, no. 1-2, pp. 146-150. 1 July 1996. • Abd El-Azim, ME; Mohamed, KE; Hammad, FH, “Deformation Characteristics of Alloy 800H and Alloy 617”, Mechanics of Materials [MECH MATER.], vol. 14, no. 1, pp. 33-46, 1992. • “Carlson alloy C 800/800H”, Alloy Digest. Vol. Ni-474, pp. 2. Jan. 1995. • Bassford, T H; Rahoi, D W, “Proposed Changes to Incoloy Alloy 800 and 800H Specifications”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 151-161, 1978. • “Cabot Alloys No. 800 and 800H”, Cabot Corp., pp. 3, 1983. • “Data sheets on the elevated-temperature properties of iron based 21Cr-32Ni-Ti-Al superalloy for corrosion-resisting and heat-resisting superalloy plates (NCF 800HP)”, NRIM Creep Data Sheet (Japan), vol. 27B, pp. 40, 2000. • “Data sheets on the elevated-temperature stress relaxation properties of iron based 21Cr-32Ni-Ti-Al alloy for corrosion-resisting and heat-resisting superalloy bar (NCF 800H-B), NRIM Creep Data Sheet (Japan), vol. 47, pp. 25, 31 Mar. 1999. • “Data Sheets on the Elevated-Temperature Properties of Iron Base--21Cr--32Ni--Ti-Al Alloy for Heat Exchanger Seamless Tubes (NCF 800H TB)”, NRIM Creep Data Sheets , no. 26A, pp. 14. 1983. • “Properties of Heat and Corrosion Resistant High Alloy Steel Tubes: Tempaloy 800H”, Nippon Kokan Tech. Rep , no. 92, pp. 36-46. Jan. 1982. • Kamemura, Y; Minegishi, I; Murase, S; Tamura, M; Settai, Y; Matushita, A, “Properties of heat and corrosion resisting high alloy steel tubes - Tempaloy 800H”, Nippon Kokan Technical Report-Overseas , pp. 47-55. 1982. • Lai, G Y; Kimball, O F, “Aging Behavior of Alloy 800H and Associated Mechanical Property Changes”, pp. 28, Nov. 1978. • Lucks, I, “Influence of Niobium Additions on Mechanical Properties and Corrosion of INCOLOY 800H”, pp. 65, Feb. 1980. • McCoy, H E; King, J F, “Creep and tensile properties of Alloy 800H/Hastelloy X weldments”, Report ORNL/TM-8728 (DE83017067). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Aug.1983. 20pp.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Monma, Y; Sakamoto, M; Miyazaki, A; Nagai, H; Yokoi, S, “Assessment of Elevated-Temperature Property Data for Alloy 800H”, Trans. Natl. Res. Inst. Met. (Jpn.). Vol. 26, no. 3, pp. 215-229. Sept. 1984. Creep: • Abd El-Azim, M E, “Correlation between tensile and creep data in alloy 800H at 850ºC”, Journal of Nuclear Materials. Vol. 231, no. 1-2, pp. 146-150. 1 July 1996. • Abd El-Azim, ME; Mohamed, KE; Hammad, FH, “Deformation Characteristics of Alloy 800H and Alloy 617”, Mechanics of Materials [MECH MATER.], vol. 14, no. 1, pp. 33-46, 1992. • Ahn, Y S; Bae, C H, “A study on the creep cavitation of cold deformed Incoloy 800H”, Journal of the Korean Institute of Metals and Materials. Vol. 33, no. 8, pp. 1110-1116. 1995. • Ahn, Y S; Jeong, H Y; Chung, W S, “Influence of Cold Deformation on the Creep Properties of Incolloy 800H”, Journal of the Korean Institute of Metals and Materials. Vol. 32, no. 1, pp. 120-127. Jan. 1994. • “RA 800H (Austenitic Heat-Resistant Alloy)”, Alloy Dig. Vol. SS-373, pp. 2. July 1980 • “Uranus 800H (oxidation and heat-resistant alloy)”, Alloy Digest. Vol. SS-650, pp. 2. Aug. 1996. • “Jessop JS800/800H (Austenitic Iron--Nickel--Chromium Alloy)”, Alloy Digest. Vol. SS-448, pp. 2. June 1984. • Bassford, T H; Rahoi, D W, “Proposed Changes to Incoloy Alloy 800 and 800H Specifications”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 151-161, 1978. • Booker, M K, “An Analytical Representation of the Creep and Creep-Rupture Behaviour of Alloy 800H”, ASME MPC-7. Characterisation of Materials for Service at Elevated Temperatures. Proceedings, Symposium, Montreal, 25-29 June 1978. Publ. New York, NY 10017; American Society of Mechanical Engineers; 1978. pp. 1-27. 21 Fig., 6 Tab., 33 Ref. • Booker, M K, “Analytical Representation of the Creep and Creep-Rupture Behaviour of Alloy 800H”, Report Conf-780609-8. Publ: Oak Ridge, Tenn.; Oak Ridge National Laboratory, 1978. • “Cabot Alloys No. 800 and 800H”, Cabot Corp., pp. 3, 1983. • Chow, J G Y; Soo, P; Epel, L, “Creep and Fatigue Properties of Incoloy 800H in a High-Temperature Gas-Cooled Reactor (HTGR) Helium Environment”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 331-336, 1978. • Coppolecchia, V; Bryant, J; Hofmann, F; Drefahl, K, “Loss of Creep Ductility in Alloy 800H With High Levels of Titanium and Aluminum”, Performance of High Temperature Materials in Fluidized Bed Combustion Systems and Process Industries; Cincinnati, Ohio; USA; 10-15 Oct. 1987. pp. 201-209. 1987. • Degischer, H P; Aigner, H; Lahodny, H; Spiradek, K, “Qualification of Stationary Creep of the Carbide Precipitating Alloy 800H”, High Temperature Alloys--Their Exploitable Potential; Petten; The Netherlands; 15-17 Oct. 1985. pp. 487-498. 1987.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Diglio, M R; Straube, H; Spiradek, K; Degischer, H P, “Improving the Creep Resistance of Alloy 800H”, High Temperature Materials for Powder Engineering 1990. I; Liege; Belgium; 24-27 Sept. 1990. pp. 545-554. 1990 • Drossler, E; Danzer, R; Hochreiter, E, “Stress--Rupture Tests on Incoloy 800H”, Berg- und Huttenmannische Monatshefte. Vol. 131, no. 3, pp. 73-79. Mar. 1986. • Gommans, R J; Verheesen, K F; Heerings, J H, “Oxidation Cracking and Residual Creep Life of an Incoloy 800H Bottom Manifold in a Steam Reformer at 800ºC”, Creep: Characterization, Damage and Life Assessments, Lake Buena Vista; Florida; USA; 18-21 May 1992. pp. 257-263. 1992. • Guttmann, V; Timm, J, “On the Influence of the Thermal Pretreatment on Creep and Microstructure of Alloy 800H”, Zeitschrift fur Metallkunde. Vol. 81, no. 6, pp. 428432. June 1990. • Guttmann, V; Timm, J, “Corrosion and Creep of Alloy 800H Under Simulated Coal Gasification Conditions”, Werkst. Korros. Vol. 39, no. 7, pp. 322-331. July 1988. • Guttmann, V; Burgel, R, “Creep--Structural Relationship in Steel Alloy 800H at 9001000ºC”, Met. Sci. Vol. 17, no. 11, pp. 549-555. Nov. 1983. • Hunter, C P; Hurst, R C; Taplin, D M R, “Creep and Creep Crack Growth Studies on Alloy 800H”, Mechanical Behaviour of Materials--VI. Vol. 4; Kyoto; Japan; 29 July2 Aug. 1991. pp. 551-557. 1992. • Hunter, C P; Hurst, R C; Taplin, D M R, “Determination of the Controlling Multiaxial Stress Rupture Criterion in Tubular Alloy 800H Components”, High Temperature Materials for Powder Engineering 1990. I; Liege; Belgium; 24-27 Sept. 1990. pp. 619-628. 1990 • Hurst, R C, “The Influence of Multiaxiality of Stress and Environmental-Induced Degradation on the Creep Behavior of Alloy 800H Tubular Components”, Mechanical Behavior of Materials--IV, Vol. 1; Stockholm; Sweden; 15-19 Aug. 1983. pp. 345-351. 1983. • Inouye, H, “Carburization and Dimensional Stability of Hastelloy X and Alloy 800H in HTGR Helium”, 1979 Winter Meeting; San Francisco; Calif ; 11-15 Nov. 1979. pp. 297-298. 1979, or Am. Nucl. Soc., Vol. 33, p. 297-98 November 1979. • “Data sheets on the elevated-temperature properties of iron based 21Cr-32Ni-Ti-Al superalloy for corrosion-resisting and heat-resisting superalloy plates (NCF 800HP)”, • NRIM Creep Data Sheet (Japan), vol. 27B, pp. 40, 2000. • “Data sheets on the elevated-temperature stress relaxation properties of iron based 21Cr-32Ni-Ti-Al alloy for corrosion-resisting and heat-resisting superalloy bar (NCF 800H-B), NRIM Creep Data Sheet (Japan), vol. 47, pp. 25, 31 Mar. 1999. • “Data sheets on the elevated-temperature properties of iron based 21Cr-32Ni-Ti-Al alloy for heat exchanger seamless tubes (NCF 800H TB)”, NRIM Creep Data Sheets. Vol. 26B, pp. 42. 1998. • “Data Sheets on the Elevated-Temperature Properties of Iron Base--21Cr--32Ni--Ti-Al Alloy for Heat Exchanger Seamless Tubes (NCF 800H TB)”, NRIM Creep Data Sheets , no. 26A, pp. 14. 1983.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • “Data Sheets on the Elevated-Temperature Properties of Iron Base--21Cr--32Ni--Ti-Al Plates for Corrosion and Heat Resistant Applications (NCF 800H-P)”, NRIM Creep Data Sheets. Vol. 27A, pp. 14. 1983. • “Properties of Heat and Corrosion Resistant High Alloy Steel Tubes: Tempaloy 800H”, Nippon Kokan Tech. Rep , no. 92, pp. 36-46. Jan. 1982. • Konosu, S; Sakaba, N; Kaneko, T, “Effects of heat treatment on the creep properties of alloy 800H processing long service history”, Engineering Failure Analysis. Vol. 1, no. 4, pp. 267-274. Dec. 1994. • Lai, Y, “High Temperature Gaseous Corrosion and Specimen Size Effects on CreepRupture Behavior of Hastelloy Alloy X, Alloy 800H and 2.25 Cr-1 M Mo Steel: G”, General Atomic Company, San Diego, CA, Journal of Metals, Metallurgical Society of AIME. • Lai, G Y and Wolwowicz, R J, “Creep Rupture Behavior of 2-1/4cr-1 Mo Steel, Alloy 800H and Hastelloy Alloy X in a Simulated HTGR Helium Environment”, General Atomic Co., San Diego, Calif. Dec. 1979 75 p refs. (GA-A-15572) Avail. NTIS. • Lee, K S, “Creep Rupture Properties of Hastelloy-X and Incoloy-800H in a Simulated HTGR Helium Environment Containing High Levels of Moisture”, Nuclear Technology. Vol. 66, no. 2, pp. 241-249. Aug. 1984. • Lee, K S, “Creep Rupture Properties of Hastelloy-X and Incoloy-800H in a Simulated HTGR Helium Environment Containing High Levels of Moisture”, Brookhaven National Laboratory, Upton, New York, 11973, Nuclear Technology, 66, No. 2, 241249, (Aug. 1984). • Leet, D M, “Creep, Fatigue, and Environmental Response of Alloy 800H Tested at 900ºC”, Dissertation Abstracts International. Vol. 51, no. 4, pp. 162. Oct. 1990. • Lerch, B A; Kempf, B; Steiner, D; Gerold, V, “Fatigue and Creep Behaviour of Alloy 800H at Elevated Temperatures”, Strength of Metals and Alloys (ICSMA 7). Vol. 2; Montreal; Canada; 12-16 Aug. 1985. pp. 1299-1304. 1985. • McAllister, S; Hurst, R C; Chung, T E, “Modelling The Multiaxial Creep Behavior Of Alloy 800H”, S. McAllister, R.C. Hurst, T.E. Chung, Int. J. Pressure Vessels & Piping, 47(3), 1991, pp. 355-370. (ISSN 0308-0161). • McAllister, S; Hurst, R C; Chung, T E, “Modelling the Multiaxial Creep Behaviour of Alloy 800H”, International Journal of Pressure Vessels and Piping. Vol. 47, no. 3, pp. 355-370. Sept. 1991. • Monma, Y; Sakamoto, M; Miyazaki, A; Nagai, H; Yokoi, S, “Assessment of Elevated-Temperature Property Data for Alloy 800H”, Trans. Natl. Res. Inst. Met. (Jpn.). Vol. 26, no. 3, pp. 215-229. Sept. 1984. • Ohba, T; Kanemaru, O; Yagi, K; Tanaka, C, “Long-term stress relaxation properties and microstructural change of NCF 800H”, Materials Science Research International. Vol. 3, no. 1, pp. 10-15. Mar. 1997. • Ohba, T; Kanemaru, O; Yagi, K; Tanaka, C, “Long-term stress relaxation properties of NCF 800H alloy”, Journal of the Society of Materials Science, Japan. Vol. 46, no. 1, pp. 19-24. Jan. 1997.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Ohba, T; Kanemaru, O; Yagi, K; Tanaka, C, “Long-Term Stress Relaxation Behaviour of NCF 800H Alloy”, Strength of Materials. ICSMA 10; Sendai; Japan; 22-26 Aug. 1994. pp. 649-652. 1994. • Penkalla, H J; Nickel, H; Schubert, F, “Multiaxial Creep of Tubes From Incoloy 800H and Inconel 617 Under Static and Cyclic Loading Conditions”, Nuclear Engineering and Design. Vol. 112, pp. 279-289. Mar. 1989. • Portella, P D, “Monotonous and Cyclic Creep Characteristics of the 800H Alloy at 800ºC”, pp. 195, 19 July 1984. • Portella, P D; Blum, W, “Analysis of the Creep Mechanism at 800ºC of Alloy 800H. (Retroactive Coverage)”, 6th Brazilian Congress on Engineering and Materials Science (CBECIMAT 84), Rio de Janeiro, Brasil; 9-12 Dec. 1984. pp. 205-209. 1984. • Radhakrishnan, V M; Ennis, P J; Schuster, H, “Stress Relaxation Behaviour of Alloy 800H”, Transactions of the Indian Institute of Metals (India). Vol. 45, no. 5, pp. 323329. Oct. 1992. • Rees, D W A, “The Creep Behaviour of Alloy 800H Under Tension and Torsion”, International Journal of Pressure Vessels and Piping. Vol. 20, no. 2, pp. 101-126. 1985. • Rees, D W A; Hurst, R C, “Multiaxial Creep Fracture Strain Analysis for Alloy 800H”, Environmental Problems in Materials Durability. IDFC 1; Dublin; Ireland; Mar. 1983. pp. 27-37. 1984. • Roach, D B; VanEcho, J A, “Compressive Creep and Notched Creep Rupture Behavior of Hk-40 And Alloy 800H”, D. B. Roach and J. A. VanEcho, BATTELLE, Columbus Laboratories, 505 King Ave., Columbus, OH, 43201. Corrosion, 84/14, NACE, Houston, TX. $3.00 per copy. • Roach, D B; VanEcho, J A, “Compressive Creep and Notched Creep Rupture Behavior of HK-40 and Alloy 800H”, Corrosion `84; New Orleans, La ; U.S.A ; 2-6 Apr. 1984. 14 pp. 1984. • Roach, DB; van Echo, JA, “Compressive creep and notched creep-rupture behavior of HK-40 and alloy 800H”, CORROSION/84 PREPRINTS., 1984. • Roedig, M; Penkalla, H J; Franzke, K; Schubert, F; Nickel, H, “Experiments With Tubes Made of Incoloy 800H Under Uniaxial and Multiaxial Loading Conditions”, pp. 66, Feb. 1985 • Spiradek, K; Czyrska-Filemonowicz, A, “Structural Analysis of Fe--Ni--Cr Alloy 800H After High Temperature Creep”, Prace Komisji Metalurgiczno-Odlewniczej, Polska Akademia Nauk--Oddzial w Krakowie, Metalurgia , no. 39, pp. 39-47. 1990. • Spiradek, K; Czyrska-Filemonowicz, A; Ennis, P J, “The Structure of Alloy 800H After High-Temperature Creep Deformation”, Archives of Metallurgy. Vol. 35, no. 1, pp. 47-53. 1990. • Spiradek, K, “Correlation Between Microstructional and Creep Deformation of Alloy 800H”, Prace Komisji Metalurgiczno-Odlewniczej, Polska Akademia Nauk--Oddzial w Krakowie, Metalurgia , no. 39, pp. 49-68. 1990. • Spiradek, K; Degischer, H P; Aigner, H, “Metallographic Explanation of the Improvement of Creep Strength of Alloy 800H After Cold Deformation and Heat Treatment”, Praktische Metallographie. Vol. 26, no. 12, pp. 626-639. Dec. 1989.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Spiradek, K; Degischer, H P; Aigner, H, “Metallographic Explanation of the Improvement of Creep Strength of Alloy 800H After Cold Deformation and Heat Treatment”, pp. 15, June 1988. • Stroosnijder, M F; Guttmann, V; de Wit, J H W, “Corrosion and Creep Behaviour of Alloy 800H in Sulphidizing /Oxidizing/Carburizing Environments at 700ºC. II. Creep Behaviour”, Werkst. Korros. Vol. 41, no. 9, pp. 508-513. Sept. 1990. • Stroosnijder, M F; Guttmann, V; de Wit, J H W, “Corrosion and Creep Behaviour of Alloy 800H in Sulphidizing /Oxidizing/Carburizing Environments at 700ºC. I. Corrosion Behaviour in the Stress Free and the Stressed State”, Werkst. Korros. Vol. 41, no. 9, pp. 503-507. Sept. 1990 • Stroosnijder, M F; Guttmann, V; Gommans, R J N; de Wit, J H W, “Influence of Creep Deformation on the Corrosion Behaviour of a CeO2 Surface-Modified Alloy 800H in a Sulphidising--Oxidising--Carburising Environment”, Materials Science and Engineering A. Vol. A120-A121, pp. 581-587. 1 Dec. 1989. • Taylor, N G; Guttmann, V; Hurst, R C, “The Creep Ductility and Fracture of Carburised Alloy 800H at High Temperatures”, High Temperature Alloys--Their Exploitable Potential; Petten; The Netherlands; 15-17 Oct. 1985. pp. 475-485. 1987. • Watanabe, K; Tsuji, H; Nakajima, H; Tanabe, T; Hiraga, K; Sakai, Y; Shiraishi, H, “Effects Of Environment And Aging on Creep Properties of Alloy 800H”, (Japanese), Japan Atomic Energy Research Inst., Tokyo (Japan), Mar. 1990. 38 p. Order Number DE90520207, Source: NTIS, (US Sales Only) • Watanabe, K; Tsuji, H; Nakajima, H; Tanabe, T; Hirage, K; Sakai, Y; Shiraishi, H, “Effects of Environment and Aging on Creep Properties of Alloy 800H (Japanese)”, K. Watanabe, H. Tsuji, H. Nakajima, T. Tanabe, K. Hirage, Y. Sakai, H. Shiraishi, Mar. 1990, 38 p. (DE90-520207; JAERI-M-90-061) Avail: NTIS (US Sales Only). • Willems, H, “Investigation of Creep Damage in Alloy 800H Using Ultrasonic Velocity Measurements”, Nondestructive Characterization of Metals. II; Montreal; Canada; 21-23 July 1986. pp. 471-479. 1987. • Wilson, H; Korhonen, M A; Li, C Y, “Effects of Grain Boundary Sliding on the Flow Properties of Incoloy 800H”, Materials Science and Engineering A. Vol. A156, no. 1, pp. 33-41. 1 Aug. 1992. Weld: • “Uranus 800H (oxidation and heat-resistant alloy)”, Alloy Digest. Vol. SS-650, pp. 2. Aug. 1996. • Bernasovsky, P, “The soundness of welded joints in Incoloy 800H nickel alloy”, Welding International. Vol. 3, no. 9, pp. 814-816. 1989. • Bernasovsky, P, “Inhomogeneity of Welded Joint in a Nickel Alloy Type Incoloy 800H”, Zvaranie/Svarovani. Vol. 37, no. 6, pp. 177-180. June 1988. • Cook, R H, “Transition in the Creep Failure Mode of an Alloy 800H Weld at 850ºC”, High Temp. Technol. Vol. 3, no. 3, pp. 143-149. Aug. 1985 • Ernst, S C; Lai, G Y, “A new Fe-Ni-Co-Cr filler metal for joining alloy 800H”, Materials Performance. Vol. 28, no. 8, pp. 58-61. Aug. 1989.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Ernst, S. C. and Lai, G. Y. “A New Fe-Ni-Co-Cr Filler Metal for Joining Alloy 800H”, CORROSION 89/528, 891671 (CO). NACE International, P.O. Box 218340, Houston, TX 77218. • King, J F; McCoy, H E, “Weldability and mechanical property characterisation of weld clad Alloy 800H tubesheet forging”, Report ORNL/TM-9108 (DE85001737). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Sept. 1984. • Lefort, A; Bressan, J; Franzoni, U, “Creep corrosion behaviour of welds in alloy 800H and 316L and in several experimental ferritic stainless steels”, Paper presented at 119th TMS Annual Meeting and Exhibit, Anaheim, CA, USA, 18-22 Feb.1990; Report CEA-CONF-10132 (DE90518442; CONF-900206-11. Publ: F-38041 Grenoble, France; CEA Centre d'Etudes Nucleaires de Grenoble; 1990. • Lundin, C D; Qiao, C Y P, “Stress rupture evaluation of weldments and base metal in a modified 800H alloy”, Welding, Joining, Coating and Surface Modification of Advanced Materials, Proceedings, Pre-Assembly Symposium, 47th Annual Assembly of IIW, Dalian, China, Vol.1; 1-2 Sept. 1994. pp. 97-104. 1994. • Lundin, C D; Qiao, C Y P; Swindeman, R W, “Weld HAZ [heat affected zone] characterisation of modified 800H alloys”, High Temperature Service and Time Dependent Failure. Symposium during Pressure Vessels and Piping Conference, Denver, CO; 25-29 July 1993. pp. 65-74. 1993. • Lundin, C D, “Evaluation Of Haz Liquation Cracking Susceptibility And HAZ Softening Behavior in Modified 800H”, C.D. Lundin, 20 Nov., 1992. 115p. Order Number DE93004064. Source: OSTI; NTIS; GPO Dep. Book • McCoy, H E; King, J F, “Creep and tensile properties of Alloy 800H/Hastelloy X weldments”, Report ORNL/TM-8728 (DE83017067). Publ: Oak Ridge, TN 37830, USA; Oak Ridge National Laboratory; Aug.1983. 20pp. • Shi, C; Sa, Y, “Repair Welding of Long-Term Serviced Incoloy 800H Weldments”, J. Aeronaut. Mater. (China). Vol. 9, no. 1, pp. 40-46. 1989. • Sonoya, K; Tomisawa, Y, “Cracking by Elevated Temperature Embrittlement in the HAZ [Heat Affected Zone] of Alloy 800H”, Stainless Steels '91, Proceedings, International Conference, Chiba, Japan, Vol.2; 10-13 June 1991. pp. 1024-1031. 1991. • Sonoya, K; Tomisawa, Y, “Cracking by Elevated Temperature Embrittlement in the HAZ of Alloy 800H”, Transactions of the Japan Welding Society. Vol. 22, no. 1, pp. 10-15. Apr. 1991. • Sonoya, K; Tomisawa, Y, “Cracking by elevated temperature embrittlement in the HAZ [heat affected zone] of alloy 800H”, Welding International. Vol. 5, no. 6, pp. 425-429. 1991. • Sonoya, K; Tomisawa, Y, “Cracking by Elevated Temperature Embrittlement in the HAZ of Alloy 800H”, Quarterly Journal of the Japan Welding Society. Vol. 8, no. 3, pp. 109-115. Aug. 1990. • Tomisawa, Y; Sonoya, K, “Cracking failure in the HAZ [heat affected zone] of Alloy 800H pipe in the temperature range of 550-700ºC”, Advanced Technology in Welding, Materials Processing and Evaluation, Proceedings, 5th JWS International Symposium, Tokyo, Vol. II, Paper V-14; 17-19 Apr. 1990. pp. 901-906. 1990.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ Fatigue: • Leet, D M, “Creep, Fatigue, and Environmental Response of Alloy 800H Tested at 900ºC”, Dissertation Abstracts International. Vol. 51, no. 4, pp. 162. Oct. 1990. • Chow, J G Y; Soo, P; Epel, L, “Creep and Fatigue Properties of Incoloy 800H in a High-Temperature Gas-Cooled Reactor (HTGR) Helium Environment”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 331-336, 1978. • Lerch, B A; Kempf, B; Steiner, D; Gerold, V, “Fatigue and Creep Behaviour of Alloy 800H at Elevated Temperatures”, Strength of Metals and Alloys (ICSMA 7). Vol. 2; Montreal; Canada; 12-16 Aug. 1985. pp. 1299-1304. 1985. • Bressers, J; Hessler, W; Hildebrandt, U W; Willems, H, “NDT Assessment of the Residual Life of Fatigue Damaged Alloy 800H”, High Temperature Materials for Powder Engineering 1990. I; Liege; Belgium; 24-27 Sept. 1990. pp. 629-642. 1990. • Bressers, J; Schusser, U; Ilschner, B, “Environmental Effects on the Fatigue Behaviour of Alloy 800H”, Low Cycle Fatigue and Elasto-Plastic Behaviour of Materials; Munich; FRG; 7-11 Sept. 1987. pp. 365-370. 1987. • Ellis, J R, “Results of an Interlaboratory Fatigue Test Program Conducted on Alloy 800H at Room and Elevated Temperatures”, ASTM Journal of Testing and Evaluation. Vol. 15, no. 5, pp. 249-256. Sept. 1987. • Estrin, Y; Giese, A, “Steady State Behaviour of Alloy 800H Under Cyclic Deformation”, Scripta Metallurgica et Materialia (USA). Vol. 29, no. 9, pp. 12231228. 1 Nov. 1993. • “Data Sheets on Elevated-Temperature, Time-Dependent Low-Cycle Fatigue Properties of NCF 800H-B (Fe--21Cr--32Ni--Ti--Al) Alloy Bar for Corrosion and Heat-Resisting Applications”, NRIM Fatigue Data Sheets , no. 36, pp. 16. 1983. • Kaae, J L; Villagrana, R E, “The Effect of Ageing and Cold Working on the HighTemperature Low-Cycle Fatigue Behaviour of Alloy 800H. II.--Low-Cycle Fatigue Behaviour”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 181191, 1978. • Kempf, B; Bothe, K; Gerold, V, “High Temperature Fatigue Damage Mechanisms in Alloy 800H”, ECF6--Fracture Control of Engineering Structures. Vol. II ; Amsterdam; The Netherlands; 15-20 June 1986. pp. 1129-1138. 1986. • Mu, Z; Bothe, K; Gerold, V, “Effect of Compressive Hold Time on Fatigue Life and Creep-Fatigued Damage in Alloy 800H at 750ºC”, Scripta Metallurgica et Materialia. Vol. 24, no. 11, pp. 2145-2150. Nov. 1990. • Nilsson, J O; Thorvaldsson, T, “Low Cycle Fatigue Behaviour of Alloy 800H at 600ºC--Effect of Grain Size and gamma '-Precipitate Dispersion”, Fatigue Fract. Eng. Mater. Struct. Vol. 8, no. 4, pp. 373-384. 1985. • Portella, P D; Osterle, W, “Mechanical behaviour and microstructural evolution of alloy 800H under biaxial cyclic loading”, Fatigue '99: Seventh International Fatigue Congress; Beijing; China; 8-12 June 1999. pp. 911-916. 1999. • Rao, K B S; Schiffers, H; Schuster, H; Halford, G R, “Temperature and strain-rate effects on low-cycle fatigue behavior of alloy 800H”, Metallurgical and Materials Transactions A (USA). Vol. 27A, no. 2, pp. 255-267. Feb. 1996.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Rao, K B S; Schuster, H; Halford, G R, “Mechanisms of high-temperature fatigue failure in alloy 800H”, Metallurgical and Materials Transactions A (USA). Vol. 27A, no. 4, pp. 851-861. Apr. 1996. • Rao, K B S; Schuster, H; Halford, G R, “On massive carbide precipitation during high temperature low cycle fatigue in alloy 800H”, Scripta Metallurgica et Materialia (UK). Vol. 31, no. 4, pp. 381-386. 15 Aug. 1994. • Rees, D W A, “High temperature cyclic deformation processes in alloy 800H”, Thermal Stresses '99: Third International Congress on Thermal Stresses; Cracow; Poland; 13-17 June 1999. pp. 219-222. 1999. • Soo, P; Sabatini, R L, “High-Cycle Fatigue Behavior of Incoloy Alloy 800H in a Simulated HTGR Helium Environment Containing High Moisture Levels”, Nuclear Technology. Vol. 66, no. 2, pp. 324-346. Aug. 1984. • Soo, P et al., “High Cycle Fatigue Behavior of Incoloy 800H in a Simulated HighTemperature Gas-Cooled Reactor Helium Environment”, Brookhaven National Lab., Upton, NY. Jan 1980. 84 p. NTIS, GPO. File Number TI86002772. • Teranishi, H; McEvily, A J, “A Comparison of the Low-Cycle Fatigue Behaviour of Alloy 800 and Alloy 800H”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 125-132, 1978. • Villagrana, R E; Kaae, J L; Ellis, J R, “The Effect of Aging and Cold Working on the High-Temperature Low-Cycle Fatigue Behavior of Alloy 800H. II.--Continuous Cyclic Loading”, Metall. Trans A. Vol. 12A, no. 11, pp. 1849-1857. Nov. 1981. • Villagrana, R E; Kaae, J L; Ellis, J R; Gantzel, P K, “Effect of Aging and Cold Working on the High-Temperature Low-Cycle Fatigue Behavior of Alloy 800H. Pt. 1. Effect of Hardening Process on the Initial Stress-Strain Curve”, Metall. Trans. A. Vol. 9A, no. 7, pp. 927-934. July 1978. • Villagrana, R E; Kaae, J L, “The Effect of Ageing and Cold Working on the HighTemperature Low-Cycle Fatigue Behaviour of Alloy 800H. I.--The Effect of Ageing and Cold Working on the Initial Stress/Strain”, Alloy 800 \Proc. Conf.\, Petten, The Netherlands, Mar. 1978, pp. 175-179, 1978. • Walter, M; Schutze, M; Rahmel, A, “Behavior of Oxide Scales on Alloy 800H and HK40 During Thermal Cycling”, Oxidation of Metals. Vol. 40, no. 1-2, pp. 37-63. Aug. 1993. Crack: • Bressers, J; Weise, W; Hollstein, T, “High Temperature Creep and Fatigue Crack Growth in Alloy 800H”, Low Cycle Fatigue and Elasto-Plastic Behaviour of Materials; Munich; FRG; 7-11 Sept. 1987. pp. 655-660. 1987. • Foulds, J R, “Creep Crack Growth of Alloy 800H in Controlled-Impurity Helium”, Nonlinear Fracture Mechanics. Volume I--Time-Dependent Fracture; Knoxville, Tennessee; USA; 6-8 Oct. 1986. pp. 112-126. 1988. • Foulds, J R, “The Effect of Recovery and Recrystallization on the High-Temperature Fatigue Crack Growth Behavior of Alloy 800H in Controlled-Impurity Helium”, Analyzing Failures: the Problems and the Solutions, Salt Lake City; Utah; USA; 2-6 Dec. 1985. pp. 269-277. 1986.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Foulds, JR, “Effect of recovery and recrystallization on the high-temperature fatiguecrack growth behavior of alloy 800H in controlled-impurity helium”, International Conference on Fatigue, Corrosion Cracking, Fracture Mechanics and Failure Analysis, Salt Lake City, UT (USA), 2-6 Dec 1985. • Hollstein, T; Djavanroodi, F; Webster, G A; Holdsworth, S R, “High Temperature Crack Growth in Alloy 800H and a 1% CrMoV Steel--the Results of an EGF Round Robin. (Retroactive Coverage)”, ECF 7. Failure Analysis--Theory and Practice. Vol. II; Budapest; Hungary; 19-24 Sept. 1988. pp. 656-668. 1988. • Hollstein, T; Voss, B, “Experimental Determination of the High-Temperature Crack Growth Behavior of Incoloy 800H”, Nonlinear Fracture Mechanics. Volume I--TimeDependent Fracture; Knoxville, Tennessee; USA; 6-8 Oct. 1986. pp. 195-213. 1988. • Hour, K Y; Stubbins, J F, “Fatigue Crack Growth Behavior of Alloy 800H at Elevated Temperature”, Journal of Engineering Materials and Technology (Transactions of the ASME). Vol. 113, no. 3, pp. 271-279. July 1991. • Hour, K Y; Stubbins, J F, “Study of the Applicability of C* for Correlating Crack Growth Rates at Elevated Temperature in Alloy 800H”, Scr. Metall. Vol. 23, no. 6, pp. 913-918. June 1989. • Hour, K Y; Stubbins, J F, “The Effects of Hold Time and Frequency on Crack Growth in Alloy 800H at 650ºC”, Metall. Trans. A. Vol. 20A, no. 9, pp. 1727-1734. Sept. 1989. • Hunter, C P; Hurst, R C; Taplin, D M R, “Creep Crack Growth Studies on Alloy 800H Tubes Under Complex Loading Conditions”, Materials at High Temperatures. Vol. 10, no. 2, pp. 144-149. May 1992. • Hunter, C P; Hurst, R C, “Creep Crack Growth Verification Testing in Alloy 800H Tubular Components”, SMIRT 10; Anaheim, California; USA; 14-18 Aug. 1989. pp. 171-178. 1992 • Hunter, C P; Hurst, R C; Taplin, D M R, “Creep and Creep Crack Growth Studies on Alloy 800H”, Mechanical Behaviour of Materials--VI. Vol. 4; Kyoto; Japan; 29 July2 Aug. 1991. pp. 551-557. 1992. • Konosu, S; Sakaba, N; Kaneko, T, “Effects of heat treatment on the creep crack growth of alloy 800H possessing long service history”, Engineering Failure Analysis. Vol. 2, no. 3, pp. 209-214. Sept. 1995. • Oberparleiter, W; Agatonovic, P, “Creep Crack Growth And Crack Propagation of IN 800H at 830ºC”, ECF6--Fracture Control of Engineering Structures. Vol. II ; Amsterdam; The Netherlands; 15-20 June 1986. pp. 1279-1288. 1986. • Welker, M; Rahmel, A; Schutze, M, “Investigations on the Influence of Internal Nitridation on Creep Crack Growth in Alloy 800H”, Metall. Trans. A. Vol. 20A, no. 8, pp. 1553-1560. Aug. 1989. Creep-Fatigue: • Portella, P D, “Monotonous and Cyclic Creep Characteristics of the 800H Alloy at 800ºC”, pp. 195, 19 July 1984. • Kempf, B; Bothe, K; Gerold, V, “Damage Mechanisms Under Creep Fatigue Conditions in Alloy 800H”, Low Cycle Fatigue and Elasto-Plastic Behaviour of Materials; Munich; FRG; 7-11 Sept. 1987. pp. 271-276. 1987.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Kempf, B, “Damage Mechanisms in Alloy 800H at High Temperatures Under Conditions of Creep-Fatigue Load”, (German). 12 May 1987. 150p. NTIS (US Sales Only), Order Number DE88757138/JAW. • Mu, Z; Bothe, K; Gerold, V, “Damage Mechanisms in Alloy 800H Under CreepFatigue Conditions”, Fatigue and Fracture of Engineering Materials and Structures. Vol. 17, no. 5, pp. 523-537. May 1994. • Mu, Z; Gerold, V, “The Interaction of Creep Damage and Fatigue Damage and Their Temporal Development in the High Temperature Superalloy 800H”, Zeitschrift fur Metallkunde. Vol. 82, no. 8, pp. 633-639. Aug. 1991. • Nilsson, J O; Sandstrom, R, “Influence of Temperature and Microstructure on Creep-Fatigue of Alloy 800H”, High Temp. Technol. Vol. 6, no. 4, pp. 181-186. Nov. 1988. • Nilsson, J-O; Sandstroem, R, “Influence of temperature and microstructure on creepfatigue of Alloy 800H”, HIGH TEMP. TECHNOL., vol. 6, no. 4, pp. 181-186, 1988. • “Data Sheets on Elevated-Temperature, Time-Dependent Low-Cycle Fatigue Properties of NCF 800H-B (Fe--21Cr--32Ni--Ti--Al) Alloy Bar for Corrosion and Heat-Resisting Applications”, NRIM Fatigue Data Sheets, no. 36, pp. 16. 1983. • Mu, Z; Bothe, K; Gerold, V, “Effect of Compressive Hold Time on Fatigue Life and Creep-Fatigued Damage in Alloy 800H at 750ºC”, Scripta Metallurgica et Materialia. Vol. 24, no. 11, pp. 2145-2150. Nov. 1990. • Penkalla, H J; Nickel, H; Schubert, F, “Multiaxial Creep of Tubes From Incoloy 800H and Inconel 617 Under Static and Cyclic Loading Conditions”, Nuclear Engineering and Design. Vol. 112, pp. 279-289. Mar. 1989.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX G: INFORMATION ABOUT 316FR Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about 316FR. Tensile: • Kurome, Kazuya; Date, Shingo; Sukekawa, Masayuki; Takakura, Kenichi; Kawasaki, Nobuchika; Tanaka, Yoshihiko, “Material strength standard of 316FR stainless steel and modified 9Cr-1Mo steel”, ASME Pressure Vessels Piping Div Publ PVP. Vol. 391, pp. 47-54. 1999. Creep: • Kurome, Kazuya; Date, Shingo; Sukekawa, Masayuki; Takakura, Kenichi; Kawasaki, Nobuchika; Tanaka, Yoshihiko, “Material strength standard of 316FR stainless steel and modified 9Cr-1Mo steel”, ASME Pressure Vessels Piping Div Publ PVP. Vol. 391, pp. 47-54. 1999. • Hongo, H; Yamazaki, M; Watanabe, T; Kinugawa, J; Tanabe, T; Monma, Y; Nakazawa, T, “Creep deformation behavior of weld metal and heat affected zone on 316FR steel thick plate welded joint”, Journal of the Society of Materials Science, Japan. Vol. 48, no. 2, pp. 116-121. Feb. 1999. • Nakazawa, T; Kimura, H; Kimura, K; Kaguchi, H, “Advanced type stainless steel 316FR for fast breeder reactor structures”, Journal of Materials Processing Technology. Vol. 143-144, pp. 905-909. 20 Dec. 2003. Weld: • Kaguchi, H; Koto, H; Fujioka, T; Taguchi, K; Sukekawa, M, “Evaluation of fatigue properties of 316FR stainless steel welded joints at elevated temperature”, ASME Pressure Vessels Piping Div Publ PVP, ASME, New York, NY, (USA), 1996, vol. 323, no. 1, pp. 305-315. • Otani, Tomomi; Koto, Hiroyuki; Wada, HIroshI; Fujioka, Terutaka, “Evaluation method of creep-fatigue damage of 316FR stainless steel welded joint”, ASME Pressure Vessels Piping Div Publ PVP. Vol. 365, pp. 257-268. 1998. • Hongo, H; Yamazaki, M; Watanabe, T; Kinugawa, J; Tanabe, T; Monma, Y; Nakazawa, T, “Creep deformation behavior of weld metal and heat affected zone on 316FR steel thick plate welded joint”, Journal of the Society of Materials Science, Japan. Vol. 48, no. 2, pp. 116-121. Feb. 1999. Fatigue: • Kaguchi, H; Koto, H; Fujioka, T; Taguchi, K; Sukekawa, M, “Evaluation of fatigue properties of 316FR stainless steel welded joints at elevated temperature”, ASME Pressure Vessels Piping Div Publ PVP, ASME, New York, NY, (USA), 1996, vol. 323, no. 1, pp. 305-315. • Kobayashi, M; Ohno, N; Igari, T, “Thermal ratchetting analysis of cylinders subjected to axial travelling of temperature distribution--comparison with

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ experiments of 316FR steel cylinders”, Journal of the Society of Materials Science, Japan. Vol. 46, no. 8, pp. 906-913. Aug. 1997 • Abdel-Karim, Mohamed; Ohno, Nobutada; Mizuno, Mamoru, “Experiments and simulations of uniaxial ratchetting of 316FR steel at room temperature”, ASME PRESSURE VESSELS PIPING DIV PUBL PVP. Vol. 360, pp. 283-289. 1998. • Ohno, N; Kobayashi, M; Igari, T, “Ratchetting characteristics of 316FR steel at high temperature. II. Analysis of thermal ratchetting induced by spatial variation of temperature”, International Journal of Plasticity. Vol. 14, no. 4-5, pp. 373-390. 1998. • Ohno, N; Abdel-Karim, M; Kobayashi, M; Igari, T, “Ratchetting characteristics of 316FR steel at high temperature. I. Strain-controlled ratchetting experiments and simulations”, International Journal of Plasticity. Vol. 14, no. 4-5, pp. 355-372. 1998. • Kurome, Kazuya; Date, Shingo; Sukekawa, Masayuki; Takakura, Kenichi; Kawasaki, Nobuchika; Tanaka, Yoshihiko, “Material strength standard of 316FR stainless steel and modified 9Cr-1Mo steel”, ASME PRESSURE VESSELS PIPING DIV PUBL PVP. Vol. 391, pp. 47-54. 1999. • Mizuno, M; Mima, Y; Abdel-Karim, M; Ohno, N, “Uniaxial ratchetting of 316FR steel at room temperature. I. Experiments”, Journal of Engineering Materials and Technology (Transactions of the ASME). Vol. 122, no. 1, pp. 29-34. Jan. 2000. • Ohno, N; Abdel-Karim, M, “Uniaxial ratchetting of 316FR steel at room temperature. II. Constitutive modeling and simulation”, Journal of Engineering Materials and Technology (Transactions of the ASME). Vol. 122, no. 1, pp. 35-41. Jan. 2000. • Yamauchi, M; Chuman, Y; Otani, T; Takahashi, Y, “Study on high temperature lowcycle fatigue properties of 316FR stainless steel weldment by miniature specimen”, Journal of the Society of Materials Science, Japan. Vol. 49, no. 11, pp. 1224-1229. Nov. 2000. Crack: • Isobe, N; Sakurai, S; Imou, K; Yorikawa, M; Takahashi, Y, “Crack growth behaviour in 316FR stainless steel plate under tensile-bending loading and its simplified evaluation”, Materials at High Temperatures. Vol. 15, no. 2, pp. 81-86. 1998. • Sakurai, Shigeo; Satou, Yoshimi; Isobe, Nobuhiro; Takahashi, Yukio, “High temperature damage evaluation method for 316FR steel and its application to large structural component tests”, ASME Pressure Vessels Piping Div Publication PVP. Vol. 365, pp. 223-229. 1998. • Nakayama, Yasunari; Miura, Naoki; Takahashi, Yukio; Shimakawa, Takashi; Date, Shingo; Toya, Yuji, “Development of fatigue and creep crack propagation law for 316FR stainless steel in consideration of FBR operating condition”, ASME Pressure Vessels Piping Div Publication PVP. Vol. 365, pp. 191-198. 1998. • Takahashi, H; Uno, T; Tanaka, K, “Evaluation of creep-fatigue crack propagation for 316FR stainless steel welded joints”, Journal of the Society of Materials Science, Japan. Vol. 49, no. 3, pp. 322-326. Mar. 2000. Creep-Fatigue: • Chuman, Yasuharu; Yamauchi, Masafumi; Kaguchi, Hitoshi; Takahashi, Yukio, “Evaluation of creep-fatigue life of 316FR cylinder subjected to cyclic movement of

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ axial temperature distribution”, ASME Pressure Vessels Piping Div Publication PVP. Vol. 365, pp. 269-278. 1998. • Otani, Tomomi; Koto, Hiroyuki; Wada, HIroshI; Fujioka, Terutaka, “Evaluation method of creep-fatigue damage of 316FR stainless steel welded joint”, ASME Pressure Vessels Piping Div Publication PVP. Vol. 365, pp. 257-268. 1998. • Takahashi, Yukio, “Advancement of high-temperature structural design method for fast reactor components. Part I: Creep-fatigue damage evaluation method for 316FR”, ASME PRESSURE VESSELS PIPING DIV PUBL PVP. Vol. 365, pp. 159-166. 1998. • Yamaguchi, K, “Long-term creep-fatigue properties of 316FR stainless steel for fast breeder reactor”, NRIM Research Activities, pp. 104. 1998. • Takahashi, Yukio; Date, Shingo; Kaguchi, Hitoshi, “Effect of elastic follow-up on creep-fatigue life of 316FR stainless steel”, ASME PRESSURE VESSELS PIPING DIV PUBL PVP. Vol. 391, pp. 151-157. 1999. • Yamaguchi, K, “Long-term creep-fatigue properties of 316FR stainless steel for fast breeder reactor”, NRIM Research Activities, pp. 72-73. 1999. • Isobe, N; Sakurai, S; Yorikawa, M; Imou, K; Takahashi, Y, “Life prediction of 316FR stainless steel under creep-fatigue loading with elastic follow-up”, International Journal of Pressure Vessels and Piping. Vol. 77, no. 13, pp. 817-823. Nov. 2000.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX H: INFORMATION ABOUT 316H Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about 316H. Creep and Fatigue: • Albertini, C; Del Grande, A; Forlani, M; Pachera, A; Montagnani, M; Iida, K, “Residual Tensile Properties at Low and High Strain Rates of AISI 316H Predamaged by Creep, Low Cycle Fatigue, and Irradiation to 2 dpa”, Effects of Radiation on Materials. Vol. II; Andover, Massachusetts; USA; 27-30 June 1988. pp. 387-403. Crack: • Spindler, M W; Cotton, C C, “Creep-fatigue crack growth in type 316h stainless steel through a zone of tensile residual stress”, Materials at High Temperatures. Vol. 15, no. 2, pp. 117-121. 1998. • Dean, D W; Gladwin, D N, “Characterization of creep crack growth behaviour in type 316h steel using both C* and creep toughness parameters”, 9th International Conference on Creep & Fracture of Engineering Materials & Structures ; Wales, Swansea; UK; 1-4 Apr. 2001. pp. 751-761. 2001. • Mandorini, V; D'angelo, D, “Residual life assessment of AISI 316H steam piping by fatigue crack growth analysis in the heat-affected zones of butt-welds”, Remanent Life: Assessment and Extension, Proceedings, [5th] International [Eurotest] Conference, Brussels, Belgium, Paper S5-4; 19-21 Mar. 1985. 20 pp. 1985.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ APPENDIX I: INFORMATION ABOUT GRADE 91 Papers are listed in this appendix under various topics based on a very brief review of their titles or abstracts. A paper that covers multiple topics may be listed in multiple places for topics it covers to provide convenience for the reviewers. It should be stressed that the papers listed may only represent a fraction of the literature about Grade 91. General: • “AFCI Materials Handbook”, Materials Data for Particle Accelerator Applications, Revision 4, Los Alamos Program Office, October 2003. Crack: • James, L A; Carlson, K W, “The fatigue-crack growth and ductile fracture toughness behaviour of ASTM A387 Grade 91 [modified 9%Cr-1%Mo] steel”, Transactions of the ASME, Journal of Pressure Vessel Technology. Vol. 107, no. 3, pp. 271-278. Aug. 1985, or Residual-Life Assessment, Nondestructive Examination, and Nuclear Heat Exchanger Materials, Proceedings, 1985 Pressure Vessels and Piping Conference, New Orleans, PVP-Vol.98-1; 23-26 June 1985. pp. 97-107. 1985. Toughness: • Alexander, D J; Maziasz, P J; Brinkman, C R, “The Effect of Long-Term Aging on the Impact Properties of Modified 9Cr--1Mo Steel”, First International Conference on Microstructures and Mechanical Properties of Aging Materials; Chicago, Illinois; USA; 2-5 Nov. 1992. pp. 343-348. 1993, or From Fall meeting of the Metallurgical Society and American Institute of Metallurgical and Petroleum Engineers; Chicago, IL, November 1-5, 1992, NTIS; GPO Dep. Order Number DE93015453. • “Croloy 9V (A Modified 9Cr--1Mo Alloy)”, Alloy Digest. Vol. SA-402, pp. 2. Sept. 1984 • Bocquet, P; Bourges, P H; Cheviet, A, “Properties of Heavy Components of Steel Grade 91 and Their Welds”, Nuclear Engineering and Design. Vol. 144, no. 1, pp. 149-154. Oct. 1993. • Cai, G; Long, X; Svensson, L E, “Investigation of fracture and determination of fracture toughness of modified 9Cr-1Mo steel weld metals using AE technique”, Materials Science and Engineering A. Vol. 270, no. 2, pp. 260-266. 30 Sept. 1999 • Chen, C; Shiue, R K; Lan, K C, “Toughness and austenite stability of modified 9Cr1Mo welds after tempering”, Materials Science and Engineering A. Vol. 287, no. 1, pp. 10-16. 31 July 2000 • Coussement, C; Witte M,; Backer T, “Weldability and high temperature behaviour of the modified 9%Cr steel grade 91 tube and pipe base materials and weldments”, The Manufacture and Properties of Steel 91 for the Power Plant and Process Industries, Proceedings, ECSC Information Day, Dusseldorf, Germany; 5 Nov. 1992. 32 pp. 1992. • Dufrane, J J, “Plate material characterization of grade 22 and grade 91 steels used in power generation”, Materials for Advanced Power Engineering 1994; Liege; Belgium; 3-6 Oct. 1994. pp. 229-238. 1994.

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ • Gelpi, A, “Evaluation of the Modified 9Cr--1Mo Steel Forging by French Laboratories”, Steel Forgings; Williamsburg, Virginia; USA; 28-30 Nov. 1984. pp. 328-345. 1986 • Huang, F H; Hamilton, M L, “Fracture Toughness Database for HT9 and Modified 9Cr--1Mo Irradiated in Several Reactors Up to Approximately 100 dpa”, pp. 17, July 1989 • James, L A; Carlson, K W, “The fatigue-crack growth and ductile fracture toughness behaviour of ASTM A387 Grade 91 [modified 9%Cr-1%Mo] steel”, Transactions of the ASME, Journal of Pressure Vessel Technology. Vol. 107, no. 3, pp. 271-278. Aug. 1985, or L. A. James and K. W. Carlson, Hanford Engineering Development Lab., Richland, Jul. 1984, 12 p., refs, Presented at the ASME Pressure Vessel and Piping Div. Conf., New Orleans, 24 Jun. 1985, (HEDL-SA-3180-FP), Avail: NTIS. • Khare, A K; Sikka, V K, “Evaluation of Modified 9Cr--1Mo Steel Forging”, Steel Forgings; Williamsburg, Virginia; USA; 28-30 Nov. 1984. pp. 303-327. 1986 • Lee, W H; Shiue, R K; Chen, C, “Mechanical properties of modified 9Cr-1Mo steel welds with notches”, Materials Science and Engineering A. Vol. 356, no. 1-2, pp. 153-161. 15 Sept. 2003 • Matsuzaki, A; Saito, Y; Masuko, O; Oka, H; Shiga, C; Nakagawa, I, “Mechanical Properties of Modified 9Cr--1Mo Steel Produced by TMCP-Tempering Process”, Zairyo to Purosesu (Current Advances in Materials and Processes). Vol. 2, no. 6, pp. 1728-1731. 1989 • Sasaki, T; Kobayashi, K; Yamaura, T; Kasuya, T; Masuda, T, “Production and Properties of Modified 9Cr--1Mo Steel Seamless Tube for Boiler”, Kawasaki Steel Giho. Vol. 22, no. 4, pp. 257-265. 1990 • Sikka, V K, “Development of Modified 9Cr--1Mo Steel for Elevated-Temperature Service”, Topical Conference on Ferritic Alloys for Use in Nuclear Energy Technologies; Snowbird, Utah; U.S.A ; 19-23 June 1983. pp. 317-327. 1984 • Sikka, V K; Ward, C T; Thomas, K C, “Modified 9Cr--1Mo Steel--an Improved Alloy for Steam Generator Application”, Ferritic Steels for High-Temperature Applications; Warren; Pa ; 6-8 Oct. 1981. pp. 65-84. 1983 • Sireesha, M; Albert, S K; Sundaresan, S, “Microstructure and mechanical properties of weld fusion zones in modified 9Cr-1Mo steel”, Journal of Materials Engineering and Performance. Vol. 10, no. 3, pp. 320-330. June 2001 • Tsuchida, Y; Hashimoto, K; Tokuno, K, “Development and BOF manufacture of modified 9Cr-1Mo steel plates with excellent strength and toughness”, Nippon Steel Technical Report. Vol. 58, pp. 27-35. 1993 • Tsuchida, Y; Yamaba, R; Tokuno, K; Hashimoto, K; Ogawa, T; Takeda, T, “BOP [basic oxygen process] manufacturing and properties of ASTM A387 grade 91 steel plates”, New Alloys for Pressure Vessels and Piping. Symposium during Proceedings, 1990 Pressure Vessels and Piping Conference, Nashville, TN, USA; 1721 June 1990. pp. 105-114. 1990. • Yamamoto, Y; Seo, S; Matsumoto, J; Kadoya, Y; Nishimura, T; Magoshi, R, “Production and properties of modified 9Cr-1Mo steel forging (F91) for valve bodies”, Tetsu-to-Hagane (Journal of the Iron and Steel Institute of Japan). Vol. 85,

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HIGH TEMPERATURE METALLIC MATERIALS TEST PLAN FOR GENERATION IV NUCLEAR REACTORS ________________________________________________________________________ no. 7, pp. 558-563. July 1999, or Advanced Heat Resistant Steel for Power Generation; San Sebastian; Spain; 27-29 Apr. 1998. pp. 560-573. 1999.

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