2191 R0 XI.M12

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XI.M12 THERMAL AGING EMBRITTLEMENT OF CAST AUSTENITIC STAINLESS STEEL (CASS)

Program Description

The reactor coolant system components are inspected in accordance with the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI. This inspection is augmented to detect the effects of loss of fracture toughness due to thermal aging embrittlement of cast austenitic stainless steel (CASS) piping components except for valve bodies. This aging management program (AMP) includes determination of the potential significance of thermal aging embrittlement of CASS components based on casting method, molybdenum content, and percent ferrite. For components for which thermal aging embrittlement is “potentially significant” as defined below, aging management is accomplished through either (a) qualified visual inspections, such as enhanced visual examination (EVT-1); (b) a qualified ultrasonic testing (UT) methodology; or (c) a component-specific flaw tolerance evaluation in accordance with the ASME Code, Section XI. Additional inspection or evaluations to demonstrate that the material has adequate fracture toughness are not required for components for which thermal aging embrittlement in not significant. The scope of the program includes all primary pressure boundary components constructed from CASS with service conditions above 250 °C (Celsius) (482 °F (Fahrenheit)).

For pump casings, as an alternative to the screening and other actions described above, no further actions are needed if applicants demonstrate that the original flaw tolerance evaluation performed as part of Code Case N-481 implementation remains bounding and applicable for the subsequent license renewal (SLR) period or the evaluation is revised to be applicable for 80 years. For valve bodies, based on the results of the assessment documented in the letter dated May 19, 2000, from Christopher Grimes, U.S. Nuclear Regulatory Commission (US NRC), to Douglas Walters, Nuclear Energy Institute (May 19, 2000 US NRC letter), screening for significance of thermal aging embrittlement is not required. The existing ASME Code, Section XI inspection requirements are adequate for valve bodies.

Reactor vessel internal (RVI) fabricated from CASS are not within the scope of this AMP. GALL-SLR Report AMP XI.M9, “BWR Vessel Internals” contains aging management guidance for CASS RVI components of boiling water reactors (BWRs). GALL-SLR Report AMP XI.M16A, “PWR Vessel Internals” contains aging management guidance for CASS RVI components of pressurized water reactors (PWRs).

Evaluation and Technical Basis

1. Scope of Program: This program manages loss of fracture toughness in ASME Code Class 1 piping components made from CASS. The program includes screening criteria to determine which CASS components have the potential for significant loss of fracture toughness due to thermal aging embrittlement and require augmented inspection. The screening criteria are applicable to all primary pressure boundary components constructed from CASS with service conditions above 250 °C [482 °F]. The screening criteria for the significance of thermal aging embrittlement are not applicable to niobium-containing steels; such steels require evaluation on a case-by-case basis.

Based on the criteria set forth in the May 19, 2000, NRC letter, the potential significance of thermal aging embrittlement of CASS materials is determined in terms of casting method, molybdenum content, and ferrite content. For low-molybdenum content steels {SA-351 Grades CF3, CF3A, CF8, CF8A or other steels with ≤ 0.5 weight percent [wt.%] Mo}, only static-cast steels with >20 percent ferrite are potentially susceptible to thermal embrittlement. Static-cast low-molybdenum steels with ≤20 percent ferrite and all centrifugal-cast low-molybdenum steels are not susceptible. For high-molybdenum content steels (SA-351 Grades CF3M, CF3MA, and CF8M or other steels with 2.0 to 3.0 wt.% Mo), static-cast steels with >14 percent ferrite and centrifugal-cast steels with >20 percent ferrite thermal embrittlement can be potentially significant, (i.e., screens in). For static-cast high-molybdenum steels with ≤14 percent ferrite and centrifugal-cast high-molybdenum steels with ≤20 percent ferrite, thermal aging embrittlement is not significant, (i.e., screens out). The thermal embrittlement screening criteria of CASS with different molybdenum and ferrite contents are summarized in Table XI.M12-1, “Thermal Embrittlement Screening Criteria.”

In the significance screening method, ferrite content is calculated by using the Hull’s equivalent factors (described in NUREG/CR–4513, Revision 1) or a staff-approved method for calculating delta ferrite in CASS materials. A fracture toughness value of 255 kilo-joules per square meter (kJ/m²) [1,450 inch-pounds per square inch] at a crack extension of 2.5 millimeters [0.1 inch] is used to differentiate between CASS materials for which thermal aging embrittlement is not significant and those for which thermal aging embrittlement is potentially significant. Extensive research data indicate that for CASS materials without the potential for significant thermal aging embrittlement, the saturated lower-bound fracture toughness is greater than 255 kJ/m² (NUREG/CR–4513, Revision 1).

Table XI.M12-1. Thermal Embrittlement Screening Criteria
Molybdenum (Mo) Content	Fe Content	Casting Method	Potentially Significant (Screens In)	Not Significant (Screens Out)
Low or ≤ 0.5 wt.%	>20% ferrite	Static	X	—
Low or ≤ 0.5 wt.%	≤20% ferrite	Static	—	X
Low or ≤ 0.5 wt.%	Any	Centrifugal	—	X
High or 2.0-3.0 wt.%	>14% ferrite	Static	X	—
High or 2.0-3.0 wt.%	>20% ferrite	Centrifugal	X	—
High or 2.0-3.0 wt.%	≤14% ferrite	Static	—	X
High or 2.0-3.0 wt.%	≤20% ferrite	Centrifugal	—	X

For valve bodies, screening for significance of thermal aging embrittlement is not needed (and thus there are no AMR items). For valve bodies greater than 4 inches nominal pipe size (NPS), the existing ASME Code, Section XI inspection requirements are adequate. ASME Code, Section XI, Subsection IWB requires only surface examination of valve bodies less than 4 inches NPS. For these valve bodies less than 4 inches NPS, the adequacy of inservice inspection (ISI) according to ASME Code, Section XI has been demonstrated by an US NRC-performed bounding integrity analysis (May 19, 2000 letter). For pump casings, as an alternative to screening for significance of thermal aging, no further actions are needed if applicants demonstrate that the original flaw tolerance evaluation performed as part of Code Case N-481 implementation remains bounding and applicable for the SLR period, or the evaluation is revised to be applicable to 80 years.

2. Preventive Actions: This program is a condition monitoring program and does not mitigate thermal aging embrittlement.

3. Parameters Monitored or Inspected: The program monitors the effects of loss of fracture toughness on the intended function of the component by identifying the CASS materials that are susceptible to thermal aging embrittlement.

The program does not directly monitor for loss of fracture toughness that is induced by thermal aging; instead, the impact of loss of fracture toughness on component integrity is indirectly managed by using visual or volumetric examination techniques to monitor for cracking in the components.

4. Detection of Aging Effects: For valve bodies, and other “not susceptible” CASS piping components, no additional inspection or evaluations are needed to demonstrate that the material has adequate fracture toughness.

For piping components for which thermal aging embrittlement is “potentially significant,” the AMP provides for qualified inspections of the base metal, such as EVT-1 or a qualified UT methodology, with the scope of the inspection covering the portions determined to be limiting from the standpoint of applied stress, operating time, and environmental considerations. Examination methods that meet the criteria of the ASME Code, Section XI, Appendix VIII are acceptable. Alternatively, a plant-specific or component-specific flaw tolerance evaluation, using specific geometry, stress information, material properties, and ASME Code, Section XI can be used to demonstrate that the thermally-embrittled material has adequate toughness. For CASS piping, UT may be performed in accordance with the methodology of Code Case N-824, as conditioned by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a.

5. Monitoring and Trending: Inspection schedules in accordance with ASME Code, Section XI, IWB-2400 or IWC-2400, reliable examination methods, and qualified inspection personnel provide timely and reliable detection of cracks. If flaws are detected, the period of acceptability is determined from analysis of the flaw, depending on the crack growth rate and mechanism.

6. Acceptance Criteria: Flaws detected in CASS components are evaluated in accordance with the applicable procedures of ASME Code, Section XI. The most recent version of the ASME Code, Section XI incorporated by reference in 10 CFR 50.55a (2007 edition through 2008 addenda), does not contain any evaluation procedures applicable to CASS with ferrite content ≥ 20 percent. (Nonmandatory Appendix C to the ASME Code, Section XI states that flaw evaluation methods for CASS with ≥ 20 percent ferrite are currently in the course of preparation.) Therefore, methods used for evaluations of flaws detected in CASS piping or components containing ≥ 20 percent ferrite, and methods used for flaw tolerance evaluations of such components, must be approved by the NRC staff on a case-by-case basis until such methods are incorporated into editions of the ASME Code, Section XI or code cases that are incorporated by reference in 10 CFR 50.55a, or in US NRC-approved code cases, as documented in the latest revision to Regulatory Guide (RG) 1.147. NUREG/CR–4513, Revision 1 provides methods for predicting the fracture toughness of thermally aged CASS materials with delta ferrite content up to 25 percent.

7. Corrective Actions: Results that do not meet the acceptance criteria are addressed in the applicant’s corrective action program under those specific portions of the quality assurance (QA) program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the corrective actions element of this AMP for both safety-related and nonsafety-related structures and components (SCs) within the scope of this program.

Repair and replacement are performed in accordance with ASME Code, Section XI, IWA-4000.

8. Confirmation Process: The confirmation process is addressed through those specific portions of the QA program that are used to meet Criterion XVI, “Corrective Action,” of 10 CFR Part 50, Appendix B. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the confirmation process element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

9. Administrative Controls: Administrative controls are addressed through the QA program that is used to meet the requirements of 10 CFR Part 50, Appendix B, associated with managing the effects of aging. Appendix A of the GALL-SLR Report describes how an applicant may apply its 10 CFR Part 50, Appendix B, QA program to fulfill the administrative controls element of this AMP for both safety-related and nonsafety-related SCs within the scope of this program.

10. Operating Experience: The AMP was developed by using research data obtained on both laboratory-aged and service-aged materials. Based on this information, the effects of thermal aging embrittlement on the intended function of CASS components will be effectively managed.

The program is informed and enhanced when necessary through the systematic and ongoing review of both plant-specific and industry operating experience including research and development such that the effectiveness of the AMP is evaluated consistent with the discussion in Appendix B of the GALL-SLR Report.

References

10 CFR Part 50, Appendix B, “Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

10 CFR 50.55a, “Codes and Standards.” Washington, DC: U.S. Nuclear Regulatory Commission. 2016.

ASME. ASME Code Section XI, “Rules for Inservice Inspection of Nuclear Power Plant Components.” New York, New York: The American Society of Mechanical Engineers. 2008.

_____. ASME Code Section XI, Division 1, Code Case N-824, “Ultrasonic Examination of Cast Austenitic Piping Welds From the Outside Surface.” New York, New York: The American Society of Mechanical Engineers. 2012.

_____. ASME Code Section XI, Division 1, Code Case N-481, “Alternative Examination Requirements for Cast Austenitic Pump Casings.” New York, New York: The American Society of Mechanical Engineers. Approval Date March 5, 1990.

EPRI. BWRVIP-03, Revision 6 (EPRI 105696-R6)(Archived), “BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines.” Palo Alto, California: Electric Power Research Institute. December 2003.

_____. MRP-228, “The Materials Reliability Program: Inspection Standard for PWR Internals.”(Archived) Palo Alto, California: Electric Power Research Institute. 2009.

Grimes, Christopher I., U.S. Nuclear Regulatory Commission, License Renewal and Standardization Branch, letter to Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No. 98-0030, “Thermal Aging Embrittlement of Cast Stainless Steel Components.” Agencywide Documents Access and Management System (ADAMS) Accession No. ML003717179. Washington, DC: U.S. Nuclear Regulatory Commission. May 19, 2000.

Lee, S., P.T. Kuo, K. Wichman, and O. Chopra. “Flaw Evaluation of Thermally-Aged Cast Stainless Steel in Light-Water Reactor Applications.” International Journal of Pressure Vessel and Piping. pp 37–44. 1997.

Maxin, Mark J., letter to Rick Libra (BWRVIP Chairman), Safety Evaluation for Electric Power Research Institute (EPRI) Boiling Water Reactor Vessel and Internals project (BWRVIP) Report TR-105696-R6(Archived) (BWRVIP-03), Revision 6, “BWR Vessel and Internals Examination Guidelines (TAC No MC2293).” June 2008. ADAMS Accession No. ML081500814.

US NRC. NUREG/CR–4513, “Estimation of Fracture Toughness of Cast Stainless Steels During Thermal Aging in LWR Systems.” Revision 1. Washington, DC: U.S. Nuclear Regulatory Commission. August 1994.

_____. Regulatory Guide 1.147, Revision 17, “Inservice Inspection Code Case Acceptability.” Washington, DC: U.S. Nuclear Regulatory Commission. August 2014.

Cast Austenitic Stainless Steel

American Society of Mechanical Engineers

Aging Management Program

Enhanced Visual Test

Ultrasonic Test - A nondestructive test method using ultrasonic waves to determine the thickness of a component and to detect cracks or other degradation.

Subsequent License Renewal - The extension of the original license from 40 years to 60 years is known as License Renewal. Extensions of the license beyond 60 years are known as Subsequent License Renewal.

United States of America

Nuclear Regulatory Commission - The US federal government agency that regulates nuclear power plants

US NRC regulatory guide

Aging Management Review

Quality Assurance

Generic Aging Lessons Learned - The US NRC GALL Report is NUREG-1801. The US NRC Subsequent License Renewal GALL is NUREG-2191

Electric Power Research Institute

Boiling Water Reactor Vessel Internals Program - The BWRVIP is an industry program for General Electric boiling water reactors that has developed procedures for performing inspections of reactor vessel internals.

Boiling Water Reactor

Pressurized Water Reactor

Agency wide Documents Access and Management System - The official recordkeeping system, through which the U.S. Nuclear Regulatory Commission (US NRC) provides access to the "libraries" or collections of publicly available documents