XI.M9 (NUREG-2191 R0)

XI.M9 BWR VESSEL INTERNALS

Program Description

The program includes inspection and flaw evaluations in conformance with the guidelines of applicable and staff-approved Boiling Water Reactor Vessel and Internals Project (BWRVIP) documents to provide reasonable assurance of the long-term integrity and safe operation of boiling water reactor (BWR) vessel internal components. The program manages the effects of cracking due to stress corrosion cracking (SCC), intergranular stress corrosion cracking (IGSCC), or irradiation-assisted stress corrosion cracking (IASCC), cracking due to cyclic loading (including flow-induced vibration), loss of material due to wear, loss of fracture toughness due to neutron or thermal embrittlement, and loss of preload due to thermal or irradiation-enhanced stress relaxation.

The BWRVIP documents provide generic guidelines intended to present the applicable inspection recommendations to assure safety function integrity of the subject safety-related reactor pressure vessel internal components. The guidelines provide information on component description and function; evaluate susceptible locations and safety consequences of failure; provide recommendations for methods, extent, and frequency of inspection; discuss acceptable methods for evaluating the structural integrity significance of flaws detected during these examinations; and recommend repair and replacement procedures.

In addition, this program provides screening criteria to determine the susceptibility of cast austenitic stainless steel (CASS) components to thermal aging on the basis of casting method, molybdenum content, and percent ferrite, in accordance with the criteria set forth in the May 19, 2000 letter from Christopher Grimes, U.S. Nuclear Regulatory Commission (US NRC), to Mr. Douglas Walters, Nuclear Energy Institute (NEI). The susceptibility to thermal aging embrittlement of CASS components 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 percent molybdenum), 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, CF8M or other steels with 2.0 to 3.0 percent molybdenum), static-cast steels with >14 percent ferrite and centrifugal-cast steels with >20 percent ferrite are potentially susceptible to thermal embrittlement. Static-cast high-molybdenum steels with ≤14 percent ferrite and centrifugal-cast high-molybdenum steels with ≤20 percent ferrite are not susceptible. In the susceptibility 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 subsequent license renewal (SLR) applicant may use alternative staff-approved screening criteria in determining susceptibility of CASS to neutron and thermal embrittlement (e.g., screening criteria approved in the June 22, 2016, safety evaluation regarding BWRVIP-234(Archived)).

The screening criteria are applicable to all cast stainless steel (SS) primary pressure boundary and reactor vessel internal components with service conditions above 250 °C (Celsius) (482 °F (Fahrenheit)). The screening criteria for susceptibility to thermal aging embrittlement are not applicable to niobium-containing steels; such steels require evaluation on a case-by-case basis. For “potentially susceptible” components, the program considers loss of fracture toughness due to neutron embrittlement or thermal aging embrittlement.

This aging management program (AMP) addresses aging degradation of nickel alloy and SS that are used in BWR vessel internal components. When exposed to the BWR vessel environment, these materials can experience neutron embrittlement and a decrease in fracture toughness. CASS, PH martensitic SS (e.g., 15-5 and 17-4 PH steel) and martensitic SS (e.g., 403, 410, 431 steel) are also susceptible to thermal embrittlement. Effects of thermal or neutron embrittlement can cause failure of these materials in vessel internal components. In addition, nickel alloy in a BWR environment is susceptible to IGSCC.

Evaluation and Technical Basis

1. Scope of Program: The program is focused on managing the effects of cracking due to SCC, IGSCC, or IASCC, cracking due to cyclic loading (including flow-induced vibration) and loss of material due to wear. This program also manages loss of fracture toughness due to neutron or thermal embrittlement and loss of preload due to thermal or irradiation-enhanced stress relaxation. The program applies to wrought and cast reactor vessel internal components. The program contains inservice inspection (ISI) to monitor the effects of cracking on the intended function of the components, uses staff-approved BWRVIP reports as the basis for inspection, evaluation, repair and/or replacement, as needed, and evaluates the susceptibility of nickel alloy, CASS, PH martensitic SS (e.g., 15-5 and 17-4 PH steel), martensitic SS (e.g., 403, 410, 431 steel) and other SS (e.g., 304 steel) components to neutron or thermal embrittlement.

The scope of the program includes the following BWR reactor vessel (RV) and RV internal components as subject to the following staff-approved applicable BWRVIP guidelines:

Core shroud: BWRVIP-76-A(Archived) provides guidelines for inspection and evaluation; BWRVIP-02-A, Revision 2, provides guidelines for repair design criteria.

Core plate: BWRVIP-25(Archived) provides guidelines for inspection and evaluation; BWRVIP-50-A provides guidelines for repair design criteria.

Core spray: BWRVIP-18, Revision 1-A(Archived) provides guidelines for inspection and evaluation; BWRVIP-16-A and BWRVIP-19-A provide guidelines for replacement and repair design criteria, respectively.

Shroud support: BWRVIP-38 provides guidelines for inspection and evaluation; BWRVIP-52-A provides guidelines for repair design criteria.

Jet pump assembly: BWRVIP-41(Archived) and BWRVIP-138, Revision 1-A(Archived), provide guidelines for inspection and evaluation; BWRVIP-51-A provides guidelines for repair design criteria.

Low-pressure coolant injection (LPCI) coupling: BWRVIP-42-A(Archived) provides guidelines for inspection and evaluation; BWRVIP-56-A provides guidelines for repair design criteria.

Top guide: BWRVIP-26-A and BWRVIP-183(Archived) provide guidelines for inspection and evaluation; BWRVIP-50-A provides guidelines for repair design criteria. The program includes inspection of 10 percent of the top guide locations using enhanced visual technique (EVT-1) or ultrasonic testing every 12 years with at least 5 percent inspected within the first 6 years of each 12-year interval.

Reinspection Criteria:

BWR/2-5–Inspect 10 percent of the grid beam cells containing control rod drives/blades every 12 years with at least 5 percent to be performed within 6 years.

BWR/6–Inspect the rim areas containing the weld and heat affected zone from the top surface of the top guide and two cells in the same plane/axis as the weld every 6 years.

The top guide inspection locations are those that have high neutron fluence exceeding the IASCC threshold. The extent of the examination and its frequency will be based on a 10 percent sample of the total population, which includes all grid beam and beam-to-beam crevice slots.

Control rod drive (CRD) housing and lower plenum components (reactor vessel internal components): BWRVIP-47-A provides guidelines for inspection and evaluation; BWRVIP-55-A provides guidelines for repair design criteria.

Steam Dryer: BWRVIP-139-A(Archived) provides guidelines for inspection and evaluation for the steam dryer components; BWRVIP-18, Revision 1-A(Archived) provides guidelines for repair design criteria.

In addition, BWRVIP-180(Archived) provides guidelines for inspection and flaw evaluation of access hole covers and BWRVIP-217 provides guidelines for repair design criteria for these components.

2. Preventive Actions: The BWRVIP is a condition monitoring program and has no preventive actions. Maintaining high water purity reduces susceptibility to SCC or IGSCC. Reactor coolant water chemistry is monitored and maintained in accordance with the Water Chemistry program. The program description, evaluation and technical basis of water chemistry are presented in Generic Aging Lessons Learned for Subsequent License Renewal (GALL-SLR) Report AMP XI.M2, “Water Chemistry.”

3. Parameters Monitored/Inspected: The program manages the effects of aging on the intended function of the component by inspecting for cracking and loss of material in accordance with the guidelines of applicable and staff-approved BWRVIP documents and the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Section XI, Table IWB 2500-1.

Loss of fracture toughness due to neutron embrittlement in CASS materials can occur with a neutron fluence greater than 1 × 1017 n/cm2 [E>1 MeV]. Loss fracture toughness of CASS material due to thermal embrittlement is dependent on the material’s casting method, molybdenum content, and ferrite content in accordance with the criteria set forth in the May 19, 2000, letter from Christopher Grimes, U.S. Nuclear Regulatory Commission (US NRC), to Mr. Douglas Walters, Nuclear Energy Institute (NEI). A subsequent license renewal applicant may use alternative staff-approved screening criteria in determining susceptibility of CASS to neutron and thermal embrittlement (e.g., screening criteria approved in the June 22, 2016, safety evaluation regarding BWRVIP-234(Archived)). This program does not directly monitor for loss of fracture toughness that is induced by thermal aging or neutron irradiation embrittlement. 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.

Loss of fracture toughness due to neutron or thermal embrittlement cannot be identified by typical ISI activities. However, by performing visual or other inspections, applicants can identify cracks that could lead to failure of a potentially embrittled component prior to component failure. Applicants can thus indirectly manage the effects of embrittlement in the nickel alloy and SS components by identifying aging degradation (i.e., cracks), implementing early corrective actions, and monitoring and trending age-related degradation.

This program also manages loss of preload due to thermal or irradiation-enhanced stress relaxation for core plate rim holddown bolts and jet pump assembly holddown beam bolts by performing visual inspections or stress analyses for adequate structural integrity.

4. Detection of Aging Effects: The extent and schedule of the inspection and test techniques prescribed by the applicable and staff-approved BWRVIP guidelines are designed to maintain structural integrity, to detect aging effects, and to perform repair or replacement before the loss of intended function of BWR vessel internals. Vessel internal components are inspected in accordance with the requirements of ASME Code Section XI, Subsection IWB, Table IWB-2500-1, Examination Category B-N-2 for core support structures, and Examination Category B-N-1 for reactor vessel internal components. This inspection specifies visual VT-3 examination to determine the general mechanical and structural condition of the component supports by (a) verifying parameters, such as clearances, settings, and physical displacements and (b) detecting discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion. BWRVIP program requirements provide for inspection of BWR internals to manage loss of material and cracking using appropriate examination techniques such as visual examinations (e.g., EVT-1, VT-1) and volumetric examinations (e.g., ultrasonic testing).

The applicable and staff-approved BWRVIP guidelines recommend more stringent inspections, such as EVT-1 examinations or ultrasonic methods of volumetric inspection, for certain selected components and locations. The nondestructive examination (NDE) techniques appropriate for inspection of BWR vessel internals, including the uncertainties inherent in delivering and executing NDE techniques in a BWR are included in BWRVIP-03(Archived).

Loss of fracture toughness due to neutron or thermal embrittlement is indirectly managed by performing periodic visual inspections capable of detecting cracks in the components. This program also determines whether supplemental inspections are necessary in addition to the existing BWRVIP examination guidelines to manage loss of fracture toughness for nickel alloy and SS internals, including welds. If supplemental inspections are determined necessary for BWR vessel internals, the program identifies the components to be inspected and performs supplemental inspections to adequately manage loss of fracture toughness due to neutron or thermal embrittlement. This evaluation for supplemental inspections is based on neutron fluence, thermal aging susceptibility, fracture toughness, and cracking susceptibility (i.e., applied stress, operating temperature, and environmental conditions). This program further determines whether supplemental inspections are necessary to manage cracking due to IASCC for nickel alloy and SS internals, including welds. This evaluation is based on neutron fluence and cracking susceptibility. If determined necessary, the program performs the supplemental inspections on the internal components identified in the evaluation.

The inspection technique is capable of detecting the critical flaw size with adequate margin. The critical flaw size is determined based on the service loading condition and service-degraded material properties. One example of a supplemental examination is VT-1 examination of ASME Code, Section XI, IWA-2210. The initial inspection is performed either prior to or within 5 years after entering the subsequent period of extended operation.

If cracking is detected after the initial inspection, the frequency of reinspection should be justified by the applicant based on fracture toughness properties appropriate for the condition of the component. The sample size is 100 percent of the accessible component population, excluding components that may be in compression during normal operations.

5. Monitoring and Trending: Inspections are scheduled in accordance with the applicable and staff-approved BWRVIP guidelines provide timely detection of cracks. Each BWRVIP guideline recommends baseline inspections that are used as part of data collection towards trending. The BWRVIP guidelines provide recommendations for expanding the sample scope and reinspecting the components if flaws are detected. Any indication detected is evaluated in accordance with ASME Code, Section XI or the applicable BWRVIP guidelines. BWRVIP-14-A, BWRVIP-59-A, BWRVIP-60-A, BWRVIP-80-A and BWRVIP-99-A documents provide additional guidelines for evaluation of crack growth in SSs, nickel alloys, and low-alloy steels. BWRVIP-100-A(Archived) describes flaw evaluation methodologies and fracture toughness data for SS core shroud exposed to neutron irradiation.

Inspections scheduled in accordance with ASME Code, Section XI, IWB-2400 and reliable examination methods provide timely detection of cracks. The fracture toughness of precipitation-hardened (PH)-martensitic steels, martensitic SSs, and nickel alloys susceptible to thermal or neutron embrittlement need to be assessed on a case-by-case basis.

6. Acceptance Criteria: Acceptance criteria are given in the applicable staff-approved BWRVIP documents and ASME Code, Section XI. Flaws detected in the reactor vessel internals are evaluated in accordance with the procedures in the applicable staff-approved BWRVIP documents and ASME Code, Section XI.

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 Title 10 of the Code of Federal Regulations (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 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 procedures are equivalent to those requirements in ASME Code Section XI. Repair and replacement is performed in conformance with applicable staff-approved BWRVIP guidelines. Guidelines for performing weld repairs to irradiated internals are described in BWRVIP-97-A. In addition, for core shroud repairs or other IGSCC repairs, the program maintains operating tensile stresses below a threshold limit that mitigates IGSCC of X-750 material in accordance with the guidelines in BWRVIP-84, Revision 2(Archived). For top guides where cracking is observed, sample size and inspection frequencies are increased in accordance with the BWRVIP guidelines.

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: There is documentation of cracking in both the circumferential and axial core shroud welds, and in shroud supports. Extensive cracking of circumferential core shroud welds has been documented in US NRC Generic Letter (GL) 94-03 and extensive cracking in vertical core shroud welds has been documented in US NRC Information Notice (IN) 97-17. It has affected shrouds fabricated from Type 304 and Type 304L SS, which is generally considered to be more resistant to SCC. Weld regions are most susceptible to SCC, although it is not clear whether this is due to sensitization and/or impurities associated with the welds or the high residual stresses in the weld regions. This experience is reviewed in US NRC GL 94-03 and NUREG–1544; some experiences with visual inspections are discussed in US NRC IN 94-42. In addition, IASCC was observed in the core shroud beltline region and IGSCC was observed in core shroud tie rod upper supports made of X-750 alloy (BWRVIP-76-A(Archived)).

Both circumferential (US NRC IN 88-03) and radial cracking (US NRC IN 92-57) have been observed in the shroud support access hole covers that are made from Alloy 600. Instances of cracking in core spray spargers have been reviewed in US NRC Inspection and Enforcement (IEB) 80-13, and cracking in core spray pipe has been reviewed in BWRVIP-18, Revision 1-A(Archived).

Cracking of the core plate has not been reported, but the creviced regions beneath the plate are difficult to inspect. BWRVIP-06, Revision 1-A and BWRVIP-25(Archived) address the safety significance and inspection requirements for the core plate assembly. Only inspection of core plate bolts (for plants without retaining wedges) or inspection of the retaining wedges is required. US NRC IN 95-17 discusses cracking in top guides of United States and overseas BWRs. Related experience in other components is reviewed in US NRC GL 94-03 and NUREG–1544. Cracking has also been observed in the top guide of a Swedish BWR.

Instances of cracking have occurred in the jet pump assembly (US NRC IEB 80-07), hold-down beam (US NRC IN 93-101), and jet pump riser pipe elbows (US NRC IN 97-02). Cracking of dry tubes has been observed at 14 or more BWRs. The cracking is intergranular and has been observed in dry tubes without apparent sensitization, suggesting that IASCC may also play a role in the cracking.

Two control rod drive mechanism lead screw male couplings were fractured in a pressurized water reactor (PWR), designed by Babcock & Wilcox, at Oconee Nuclear Station, Unit 3. The fracture was due to thermal embrittlement of 17-4 precipitation-hardened (PH) material (US NRC IN 2007-02). While this occurred at a PWR, it also needs to be considered for BWRs.

IGSCC in the X-750 materials of a tie rod coupling and jet pump hold-down beam was observed in a domestic plant.

The program guidelines outlined in applicable staff-approved BWRVIP documents are based on an evaluation of available information, including BWR inspection data and information on the elements that cause SCC, IGSCC, or IASCC, to determine which components may be susceptible to cracking. Implementation of the program provides reasonable assurance that cracking will be adequately managed so the intended functions of the vessel internal components will be maintained consistent with the current licensing basis for the subsequent period of extended operation.

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.

EPRI. BWRVIP-02-A, Revision 2 (EPRI 1012837), “BWR Vessel and Internals Project, BWR Core Shroud Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. October 2005.

_____. BWRVIP-03, Revision 1 (EPRI 105696-R1)(Archived), “BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines.” Palo Alto, California: Electric Power Research Institute. July 1999.

_____. BWRVIP-06, Revision 1-A (EPRI 1019058), “Safety Assessment of BWR Reactor Internals.” Palo Alto, California: Electric Power Research Institute. December 2009.

_____. BWRVIP-14-A (EPRI 1016569), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals.” Palo Alto, California: Electric Power Research Institute. September 2008.

_____. BWRVIP-16-A (EPRI 1012113), “BWR Vessel and Internals Project, Internal Core Spray Piping and Sparger Replacement Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-18-A, Revision 1-A (EPRI 1025060)(Archived), “BWR Vessel and Internals Project, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. April 2012.

_____. BWRVIP-19-A (EPRI 1012114), “BWR Vessel and Internals Project, Internal Core Spray Piping and Sparger Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-25 (EPRI 107284)(Archived), “BWR Vessel and Internals Project, BWR Core Plate Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. December 2000.

_____. BWRVIP-26-A (EPRI 1009946), “BWR Vessel and Internals Project, BWR Top Guide Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. November 2004.

_____. BWRVIP-38 (EPRI TR-108823), “BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. September 1997.

_____. BWRVIP-41 (EPRI 108728)(Archived), “BWR Vessel and Internals Project, BWR Jet Pump Assembly Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. October 1997.

_____. BWRVIP-42-A (EPRI 1011470)(Archived), “BWR Vessel and Internals Project, BWR LPCI Coupling Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. February 2005.

_____. BWRVIP-44-A (EPRI 1014352), “BWR Vessel and Internals Project, Underwater Weld Repair of Nickel Alloy Reactor Vessel Internals.” Palo Alto, California: Electric Power Research Institute. August 2006.

_____. BWRVIP-45 (EPRI TR-108707), “BWR Vessel and Internals Project, Weldability of Irradiated LWR Structural Components.” Palo Alto, California: Electric Power Research Institute. June 2000.

_____. BWRVIP-47-A (EPRI 1009947), “BWR Vessel and Internals Project, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. November 2004.

_____. BWRVIP-50-A (EPRI 1012115), “BWR Vessel and Internals Project, Top Guide/Core Plate Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-51-A (EPRI 1012116), “BWR Vessel and Internals Project, Jet Pump Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-52-A (EPRI 1012119), “BWR Vessel and Internals Project, Shroud Support and Vessel Bracket Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-55-A (EPRI 1012117), “BWR Vessel and Internals Project, Lower Plenum Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-56-A (EPRI 1012118), “BWR Vessel and Internals Project, LPCI Coupling Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. September 2005.

_____. BWRVIP-59-A (EPRI 1014874), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Nickel-Base Austenitic Alloys in RPV Internals.” Palo Alto, California: Electric Power Research Institute. May 2007.

_____. BWRVIP-60-A (EPRI 1008871), “BWR Vessel and Internals Project, Evaluation of Stress Corrosion Crack Growth in Low Alloy Steel Vessel Materials in the BWR Environment.” Palo Alto, California: Electric Power Research Institute. June 2003.

_____. BWRVIP-62-A (EPRI 1021006)(Archived), “BWR Vessel and Internals Project, Technical Basis for Inspection Relief for BWR Internal Components with Hydrogen Injection.” Palo Alto, California: Electric Power Research Institute. November 2010.

_____. BWRVIP-76-A (EPRI 1019057)(Archived), “BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. October 2009.

_____. BWRVIP-80-A (EPRI 1015457), “BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Shroud Vertical Welds.” Palo Alto, California: Electric Power Research Institute. October 2007.

_____. BWRVIP-84 (EPRI 1026603)(Archived), “BWR Vessel and Internals Project, Guidelines for Selection and Use of Materials for Repairs to BWR Internal Components.” Revision 2. Palo Alto, California: Electric Power Research Institute. September 2012.

_____. BWRVIP-97-A (EPRI 1019054), “BWR Vessel and Internals Project, Guidelines for Performing Weld Repairs to Irradiated BWR Internals.” Palo Alto, California: Electric Power Research Institute. June 2009.

_____. BWRVIP-99-A (EPRI 1016566), “BWR Vessel and Internals Project, Crack Growth Rates in Irradiated Stainless Steels in BWR Internal Components.” Palo Alto, California: Electric Power Research Institute. October 2008.

_____. BWRVIP-100-A (EPRI 1013396)(Archived), “BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds.” Palo Alto, California: Electric Power Research Institute. August 2006.

_____. BWRVIP-138, Revision 1-A (EPRI 1025139)(Archived), “BWR Vessel and Internals Project, Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. October 2012.

_____. BWRVIP-139-A (EPRI 1018794)(Archived), “BWR Vessel and Internals Project, Steam Dryer Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. July 2009.

_____. BWRVIP-167NP (EPRI 3002000690)(Archived) “BWR Vessel and Internals Project Boiling Water Reactor Issue Management Tables.” Revision 1. Palo Alto, California: Electric Power Research Institute. August 2013.

_____. BWRVIP-180 (EPRI 1013402)(Archived), “BWR Vessel and Internals Project, Access Hole Cover Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. November 2007.

_____. BWRVIP-181-A (EPRI 1020997), “BWR Vessel and Internals Project, Steam Dryer Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. July 2010.

_____. BWRVIP-183 (EPRI 1013401)(Archived), “BWR Vessel and Internals Project, Top Guide Beam Inspection and Flaw Evaluation Guidelines.” Palo Alto, California: Electric Power Research Institute. December 2007.

_____. BWRVIP-190 (EPRI 1016579)(Archived), “BWR Vessel and Internals Project: BWR Water Chemistry Guidelines—2008 Revision.” Palo Alto, California: Electric Power Research Institute. October 2008.

_____. BWRVIP-217 (EPRI 1019067), “BWR Vessel and Internals Project, Access Hole Cover Repair Design Criteria.” Palo Alto, California: Electric Power Research Institute. July 2009.

_____. EPRI 3002000628(Archived), “Materials Degradation Matrix.” Revision 1. Palo Alto, California: Electric Power Research Institute. May 2013.

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 Vessels and Piping. pp. 37–44. 1997.

US NRC. “Final Safety Evaluation of the BWRVIP-234: Thermal Aging and Neutron Embrittlement Evaluation of Cast Austenitic Stainless Steel for BWR Internals.” Agencywide Documents Access and Management System (ADAMS) Accession No. ML16096A002. Washington, DC: U.S. Nuclear Regulatory Commission. June 22, 2016.

_____. Generic Letter 94-03, “Intergranular Stress Corrosion Cracking of Core Shrouds in Boiling Water Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. July 1994.

_____. IE Bulletin 80-07, “BWR Jet Pump Assembly Failure.” Washington, DC: U.S. Nuclear Regulatory Commission. April 1980.

_____. IE Bulletin 80-07, Supplement 1, “BWR Jet Pump Assembly Failure.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1980.

_____. IE Bulletin 80-13, “Cracking in Core Spray Spargers.” Washington, DC: U.S. Nuclear Regulatory Commission. May 1980.

_____. Information Notice 88-03, “Cracks in Shroud Support Access Hole Cover Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1988.

_____. Information Notice 92-57, “Radial Cracking of Shroud Support Access Hole Cover Welds.” Washington, DC: U.S. Nuclear Regulatory Commission. August 1992.

_____. Information Notice 93-101, “Jet Pump Hold-Down Beam Failure.” Washington, DC: U.S. Nuclear Regulatory Commission. December 1993.

_____. Information Notice 94-42, “Cracking in the Lower Region of the Core Shroud in Boiling Water Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. June 1994.

_____. Information Notice 95-17, “Reactor Vessel Top Guide and Core Plate Cracking.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1995.

_____. Information Notice 97-02, “Cracks Found in Jet Pump Riser Assembly Elbows at Boiling Water Reactors.” Washington, DC: U.S. Nuclear Regulatory Commission. February 1997.

_____. Information Notice 97-17, “Cracking of Vertical Welds in the Core Shroud and Degraded Repair.” Washington, DC: U.S. Nuclear Regulatory Commission. April 1997.

_____. Information Notice 2007-02, “Failure of Control Rod Drive Mechanism Lead Screw Male Coupling at Babcock and Wilcox-Designed Facility.” Washington, DC: U.S. Nuclear Regulatory Commission. March 2007.

_____. Letter from Christopher I. Grimes, U.S. Nuclear Regulatory Commission, License Renewal and Standardization Branch, to Douglas J. Walters, Nuclear Energy Institute, License Renewal Issue No. 98-0030, “Thermal Aging Embrittlement of Cast Stainless Steel Components.” ADAMS Accession No. ML003717179. May 19, 2000.

_____. NUREG–1544, “Status Report: Intergranular Stress Corrosion Cracking of BWR Core Shrouds and Other Internal Components.” Washington, DC: U.S. Nuclear Regulatory Commission. March 1996.

_____. 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.

_____. NUREG/CR–6923, “Expert Panel Report on Proactive Materials Degradation Assessment.” Washington, DC: U.S. Nuclear Regulatory Commission. March 2007.

Xu, H. and S. Fyfitch. “Fracture of Type 17-4 PH CRDM Lead Screw Male Coupling Tangs.” 11th International Conference on Environmental Degradation of Materials in Nuclear Power Systems-Water Reactors. Stevenson, Washington. American Nuclear Society. 2003.

XI.M9 (NUREG-2191 R0)

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