ML003729453

Summary of Meeting with the Nuclear Energy Institute (NEI) to Discuss Industry Comments on the Draft Generic Aging Lessons Learned (GALL) Report - Mechanical Systems Chapter IV, Sections A1, B1, and C1

ML003729453
Person / Time
Site:	PROJ0690
Issue date:	07/03/2000
From:	Dozier I NRC/NRR/DRIP/RLSB
To:	Walters D Nuclear Energy Institute
Dozier J, NRR/RLSB 415-1014
References
Download: ML003729453 (92)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 3, 2000 &ears ORGANIZATION:

Nuclear Energy Institute

SUMMARY

OF MEETING WITH THE NUCLEAR ENERGY INSTITUTE (NEI) TO DISCUSS INDUSTRY COMMENTS ON THE DRAFT "GENERIC AGING LESSONS LEARNED" (GALL) REPORT MECHANICAL SYSTEMS CHAPTER IV, SECTIONS Al, B1, AND C1 On June 6, 2000, representatives of NEI met with the Nuclear Regulatory Commission (NRC) staff in Rockville, Maryland, regarding the industry comments on Chapter IV Section Al, "Reactor Vessel (BWR)," Section B1, "Reactor Vessel Internals (BWR), "and Section C1, "Reactor Coolant Pressure Boundary (BWR)" of the draft GALL report, dated December 6, 1999. By letter dated May 18, 2000, NEI provided written comments for discussion at this meeting. A list of meeting attendees is enclosed.

Also, enclosed is Sections IV Al, B1, and C1 of the draft GALL report dated June 6, 2000 that was discussed at the meeting.

During this meeting, the staff was seeking clarification of NEI's comments.

The Staff also discussed some of the comments from the December 6, 1999, workshop relating to these GALL sections.

Based on the discussions, NEI indicated that the industry would consider revising its comments by taking the following actions: 1. Comment on the handling of time-limited aging analyses (TLAAs) during the review of the Standard Review Plan for License Renewal (SRP-LR).

2. Provide justification for why additional components, e.g. small bore piping, CRD components, etc., should not be included in GALL. 3. Provide NEI's position regarding the Isolation Condenser.

4. Provide justification for NEI's position with bolting on pumps and valves for stress relaxation, wear, and fatigue.

5. Provide a recommendation for how NSAC-202L-R2 might be implemented to meet the requirements of 10 CFR Part 50 Appendix B. 6. State NEI's position regarding the NRC letter, dated May 19, 2000, on thermal aging embrittlement of cast austenitic stainless steel components.

7. Provide the justification for NEI's position that the recirculation pump aging effects are not a significant issue. 8. Provide any additional comments of the draft GALL, Sections IV Al, B1, and Cl, dated June 6, 2000. Also, the NRC staff would consider clarifying the GALL report by taking the following actions: 1. Articulate how fatigue and the resolution of Generic Safety Issue 190 will be treated in GALL and the SRP-LR. 2. Incorporate the supporting documents in the Aging Management Program column more consistently.

3. Clarify the bolt stress issue. 4. Clarify the boiling water reactor vessel inspection program (BWRVIP) programs (especially BWRVIP 74) in GALL. 5. Reconsider the inclusion of unanticipated cyclic loading in GALL. 6. Articulate the GE Sil 462 requirements in GALL. 7. Reconsider the SLC component provided in Item B1.1.7. The staff also emphasized that NEI should provide any additional industry comments on an expedited basis to support the aggressive schedule of issuing the draft GALL report and SRP-LR for formal comment in August, 2000. IraI Ira Jerry Dozier, General Engineer Engineering Section License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

1. Attendance List 2. Draft GALL Chapter IV, Sections Al, B1, and Cl dated June 6, 2000 cc w/encls: See next page Also, the NRC staff would consider clarifying the GALL report by taking the following actions: 1. Articulate how fatigue and the resolution of Generic Safety Issue 190 will be treated in GALL and the SRP-LR. 2. Incorporate the supporting documents in the Aging Management Program column more consistently.

3. Clarify the bolt stress issue. 4. Clarify the boiling water reactor vessel inspection program (BWRVIP) programs (especially BWRVIP 74) in GALL. 5. Reconsider the inclusion of unanticipated cyclic loading in GALL. 6. Articulate the GE Sil 462 requirements in GALL. 7. Reconsider the SLC component provided in Item B1.1.7. The staff also emphasized that NEI should provide any additional industry comments on an expedited basis to support the aggressive schedule of issuing the draft GALL report and SRP-LR for formal comment in August, 2000. Ira Jerry Dozier, General Engineer Engineering Section License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

1. Attendance List 2. Draft GALL Chapter IV, Sections Al, B1, and Cl dated June 6, 2000 cc w/encls: See next page DISTRIBUTION:

See next page *See previous concurrence DOCUMENT NAME: G:\RLSB\DOZI E-'ING

SUMMARY

662000FI NAL.WPD OFFICE LA RLSB 'i RLSB:SC, RLSB:-BC NAME EHylton* IJDozier PTKuo CGrimes DATE 06/16100 0612q/00 0W 1100 06/300 OFFICIAL RECORD C DPY 9. Provide any additional comments of the draft GALL, Sections IV Al, B1, and Cl, dated June 6, 2000. Also, the NRC staff would consider clarifying the GALL report by taking the following actions: 1. Articulate how fatigue and the resolution of Generic Safety Issue 190 will be treated in GALL and the SRP. 2. Incorporate the supporting documents in the Aging Management Program more consistently.

3. Clarify the bolt stress issue. 4. Clarify the boiling water reactor vessel inspection program (BWRVIP) programs (especially BWRVIP 74) in GALL. 5. Generalize the ASME section references to include the section reference, such as IWB, but not include the specific table section.

6. Reconsider the inclusion of unanticipated cyclic loading in GALL. 7. Articulate the GE Sil 462 requirements in GALL. 8. Reconsider the SLC component provided in Item B1.1.7. The staff also emphasized that NEI should provide any additional industry comments on an expedited basis to support the aggressive schedule of issuing the draft GALL report and SRP for formal comment in August, 2000. Ira Jerry Dozier, General Engineer Engineering Section License Renewal and Standardization Branch Division of Regulatory Improvement Programs Office of Nuclear Reactor Regulation Project No. 690

Enclosures:

1. Attendance List 2. Draft GALL Chapter IV, Sections Al, B1, and Cl dated June 6, 2000 cc w/encls: See next page DISTRIBUTION:

See next page DOCUMENT NAME: G:\RLSB\DOZIER\MEETING

SUMMARY

662000FINAL.WPD OFFICE RLSB RLSB:SC RLSB:BC NAME IJDozier PTKuo CGrimes DATE 06/1 /00 06/1{o/00 06/ /00 06/ /00 I~ OFFICIAL RECORD COPY NUCLEAR ENERGY INSTITUTE Project No. 690 cc: Mr. Dennis Harrison U.S. Department of Energy NE-42 Washington, D.C. 20585 Mr. Richard P. Sedano, Commissioner State Liaison Officer Department of Public Service 112 State Street Drawer 20 Montipelier, Vermont 05620-2601 Mr. Douglas J. Walters Nuclear Energy Institute 1776 I Street, N.W., Suite 400 Washington, DC 20006-3708 DJW@NEI.ORG National Whistleblower Center 3238 P Street, N.W. Washington, DC 20007-2756 Mr. Robert Gill Duke Energy Corporation Mail Stop EC-12R P.O. Box 1006 Charlotte, NC 28201-1006 Mr. Charles R. Pierce Southern Nuclear Operating Co. 40 Inverness Center Parkway BIN B064 Birmingham, AL 35242 Chattooga River Watershed Coalition P. 0. Box 2006 Clayton, GA 30525 Mr. David Lochbaum Union of Concerned Scientists 1616 P. St., NW Suite 310 Washington, DC 20036-1495 Mr. Garry Young Entergy Operations, Inc. Arkansas Nuclear One 1448 SR 333 GSB-2E Russellville, Arkansas 72802 NRC MEETING WITH THE NUCLEAR ENERGY INSTITUTE ON LICENSE RENEWAL ATTENDANCE LIST JUNE 6, 2000 ORGANIZATION BOB EVANS TONY GRENCI MICHAEL SEMMLER ERACH PATEL FRED POLASKI ROBIN DYLE MATHEW SORENSON CHARLES WILLBANKS JERRY DOZIER KEITH WICHMAN P. T. KUO ROBERT HERMANN MICHAEL MCNEIL CE CARPENTER BARRY ELLIOT JOHN FAIR KEN KARWOSKI VIK SHAH OMESH CHOPRA ALLEN HISER LEE BANIC CHUCK HSU WILLIAM KOO JIM STRNISHA SAM LEE BILL SHACK NEI CONSTELLATION NUCLEAR SERVICES DUKE ENERGY PECO ENERGY PECO ENERGY SOUTHERN NUCLEAR NATIONAL WHISTLEBLOWERS NUS INFORMATION SERVICES NRC/NRR/DRIP/RLSB NRC/NRR NRC/NRR/DRIP/RLSB NRC/NRR NRC/RES NRC/NRR NRC/NRR NRC/NRR NRC/RES ARGONNE NATIONAL LABORATORIES ARGONNE NATIONAL LABORATORIES NRC/NRR NRC/NRR NRC/RES NRC/NRR/DE/EMCB NRC/NRR/DRIP/RLSB NRC/NRR/DRIP/RLSB ARGONE NATIONAL LABORATORIES Enclosure 1 NAME Q-4iLL A Al. Reactor Vessel (Boiling Water Reactor) /Jo£ K)/J Al. I Top Head Enclosure Al. 1.1 Top Head Al. 1.2 Nozzles (Vent, Top Head Spray or RCIC, and Spare) A1.1.3 Head Flange Al. 1.4 Closure Studs and Nuts Al. 1.5 Vessel Flange Leak Detection Line Al.2 Vessel Shell Al.2.l Vessel Flange Al.2.2 Upper Shell A1.2.3 Intermediate (Nozzle) Shell Al.2.4 Intermediate (Beltline)

Shell Al.2.5 Lower Shell Al.2.6 Beltline Welds , Al.2.7 Attachment Welds A1.3 Nozzles A1.3.1 Main Steam A1.3.2 Feedwater A1.3.3 High Pressure Coolant Injection (HPCI) A1.3.4 High Pressure Core Spray (HPCS) Al.3.5 Low Pressure Core Spray (LPCS) A1.3.6 CRD Return Line A1.3.7 Recirculating Water (Inlet & Outlet) Al.3.8 Low Pressure Coolant Injection (LPCI) or RHR Injection Mo IVAl-1 tLxut de DRAFT- 6/06/00 Enclosure 2

A1.3.9 Isolation Condernser:Supply Al.4 Nozzles Safe Ends A1.4.1 High Pressure Core Spray (HPCS) A1.4.2 Low Pressure Core Spray (LPCS) A1.4.3 CRD Return Line A1.4.4 Recirculating Water (Inlet & Outlet) A1.4.5 Low Pressure Coolant Injection (LPCI) or RHR Injection Mode Al.5 Penetrations A1.5.1 CRD Stub Tubes A1.5.2 Instrumentation A1.5.3 Jet Pump Instrument A1.5.4 Standby Liquid Control A1.5.5 Flux Monitor A1.5.6 Drain Line ay Al.6 Bottom Head A1.7 Control Rod Drive Mechanism A1.7.1 Housing A1.7.2 Withdrawal Line A1.8 Support Skirt and Attachment Welds DRAFT -6/06/00 IV A1-2 Al. Reactor Vessel (Boiling WaterReactor)

System, Structures, and Components The system, structures, and components included in this table comprise the boiling water reactor (BWR) pressure vessel and consist of vessel shell and flanges, attachment welds, top and bottom heads, nozzles (including safe ends) for the reactor coolant systbm (recirculating system) and connected systems such as (high- and low-pressure core spray, high- and low-pressure coolant injection, main steam and feedwater systems), penetrations for instrument lines and drains, and control rod drive mechanism housing. Support skirt and attachment welds for vessel support are also included in the table. All structures and components in the reactor vessel are classified as Group A Quality Standards.

System Interfaces The systems that interface with the reactor vessel include the reactor vessel internals (Table IV BI), reactor coolant pressure boundary (Table IV CI), and emergency core cooling system (Table V D2).DRAFT -6/06/00 TV A1-3 REACTOR VESSEL. INTERNALS.

AND REACTOR COOLANT SYSTEM Al vrcýQrT (lRine Water Reactor)Structure and Region of Environ- I Aig j ng Item Component Interest Material ment Effect JMechanism_

Compnen Iners Agn IM Agn Top Head Enclosure (with cladding)Top Head, Nozzles (Vent, Top Head Spray or RCIC, and Spare). Head Flange SA302-Gr B SA533-Gr B SA336, with stainless steel (SS) cladding 288°C Steam Crack Initiation and Growth ,.rress Corrosion Crarking (SCC), Inter granular Stress Corrosion Cracking (IGSCC)_______ I +/- I I ___________

_____________

I A DRAFT -6/06/00 IV AL.I.A thru Al.1.3 rV AI1-4 IV REACTOR VESSEL. INTERNALS, AND REACTOR- COOLANTý SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Existing Aging Management Program (AMP)Inservice inspection in conformance with ASME Section XI (edition specified in 10 CFR 50.55a), Subsection IWB, Table IWB 2500-1, examination categories B-A for head welds and B-D for full penetration nozzle-to-head welds. Prevention is by material selection in accordance with guidelines of NUREG-0313, Rev. 2, and of Regulatory Guide 1.43 for control of stainless steel weld cladding of low-alloy steels. Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in BWRVIP-29 and TR-103515 to minimize the potential of crack initiation and growth.

documents BWRVIP-03 for rpai-tnr RWRVTP- 14.-59, and -60 for evaluation of crack basis for insnection relief for internal comnonents with hvdrogen injection.1 DRAFT- 6/06/00 Further Evaluation NO ,-n.f)- An tfnr tprhnirnl IV AI-5 1S"nnnrtjn0 documents BATVIP-03 fo rpartrir nre.ý.q"rf-Ven-qel internal-n m nn in ... ...........

........ ...F.&M a M" Evaluation and Technical Basis * (1) Scope qf Program: The program is focused on managing the effects of stress corrosion-cracking (SCC) of SS cladding on the intended function of top head enclosure.

NUREG-0313, Rev 2 and Generic Letter (GL) 88-01, respectively, describe the technical basis and staff guidance regarding the problem of IGSCC in BWRs. However, SCC is not anticipated to be an issue for the top head enclosure because analytical evaluations indicate that cracks in the SS cladding will stop growing in the ferritic base metal. (2) Preventive Actions: Selection of material considered resistant to IGSCC, e.g., grades of weld metal with a maximum carbon of 0.035% and minimum 7.5% ferrite, prevent or mitigate IGSCC. and Regulatory Guide (RG) 1.43 provides assurance that production cladding complies with ASME Section III and XM guidelines to prevent underclad cracking.

Coolant water chemistry Is monitored and maintained in accordance with EPRI guidelines in BWRVIP-29 and TR 103515 to minimize the potential of crack initiation and growth. (3) Parameters Monitored/Inspected:

The AMP monitors the effects of IGSCC on the intended function of top head enclosure by detection and sizing of cracks by inservice inspection (ISI). Inspection requirements of Table IWB 2500-1, examination category B-A specifies volumetric inspection of all circumferential and meridian welds and B-D specifies for all nozzles volumetric inspection of nozzle-to-vessel welds and nozzle inside radius section. (4) Detection of Aging Effects: Aging effects degradation of the top head enclosure can not occur without crack initiation; extent and schedule of inspection assure detection of cracks befgie the loss of intended function of the top head enclosuIe.

(5) Monitoring and Trending:

Inspection schedule of ASME Section X) should provide for timely detection of cracks. Top head interior is inspected at Ist refueling outage and subsequent outages at approximately 3 y intervals.

(6) Acceptance Criteria:

Any IGSCC degradation is evaluated in accordance with IWB-3 100 by comparing IS! results with the acceptance standards of IWB-3400 and IWB-3520 for visual examination, IWB 3510 for head welds, and IW3B-3512 for full penetration nozzle welds. Visual examinations that reveal relevant conditions may be supplemented by surface and volumetric examinations (IWB-3200) for flaw characterization, analytical evaluation, corrective measures, and repairs. Continued service without repair requires analytical evaluation to demonstrate acceptability.

(7) Corrective Actions: Repair and replacement are in conformance with IWA-4000 and IWB 4000, and reexamination in accordance with requirements of IWA-2200.

Also, some plants have removed cladding in top head because of cracking.

(8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures.

review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The present AMP is effective in managing the effects of IGSCC on the intended function of top head enclosure.

REACTOR VESSEL, INTERNAIS.

AND REACTOR COOLANT SYSTEM A VEAC"r ELMollng Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism Al. 1.3 Top Head Head Flange SA302-Gr B, 288*C Cumulative Fatigue Enclosure SA533-Gr B, Steam Fatigue SA336, Damage with or without SS cladding AI. 1.4 Top Head Enclosure Closure Studs and Nuts SAl 9d Gr B7. SA540 Gr B23/24. SA320 Gr L43 (AISI 4340), SA194-Gr 7 Leaklng Oxygenated Water and/or Steam at 288°C Initiation and Growth IGSCC Iy j _______ I _______ I J _____ ______ a I DRAFT- 6/06/00 IV TV AI1-6 IV REACTOR VESSEL, I:NTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boilung Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Components have been designed or Fatigue is a time-limited aging analysis (T1AA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert #8 ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB, or Section I (Power Boilers) and Section VIII, Division I (Unfired Pressure Vessel).Inservice inspection in conformance with ASME Section XI. edition specified in 10 CFR 50.55a, Subsection IWB, Table IWB 2500-1, examination category B-G- 1, and testing category B-P for system leakage, and additional recommendations of GE Rapid Information Communication Service Information Letter (RICSIL) 055 Revision I, Supplement I. Prevention and replacement in accordance with Regulatory Guide 1.65.INO DRAFT- 6/06/00 IV A1-7 (1) Scope of Program: The program is focused on managing the effects of IGSCC on the intended function of reactor vessel closure stud bolting. (2) Preventive Actions: Design requirements of ASME Section III, Subsection NB, and additional guidance of Regulatory Guide (RG) 1.65 on material selection, preservice inspection, and protection against corrosion, prevent or mitigate IGSCC. High-strength low-alloy steels with controlled tempering procedures are used. Maximum tensile strength is limited to <1172 MPa (<170 ksi) to provide resistance to SCC, and Charpy V energy requirements of Appendix G to 10 CFR Part 50 provide adequate toughness to provide resistance to crack growth in the stud threads. Metal-plated stud bolting is avoided to prevent degradation due to corrosion or hydrogen embrittlement.

Manganese phosphate or other acceptable surface treatment, or stable lubricants are permissible.

Preservice inspection in conformance with NB-2580 of Section III of the Code requires ultrasonic examination of stud bolting over the entire surface prior to threading.

During refueling and while the head is removed, the stud bolts and holes are protected from corrosion and contamination in accordance with RICSIL 055 RI 51, (3) Parameters Monitored/Inspected:

71f AMP monitors the effects of IGSCC on the intended function of closure stud bolting by detection and sizing of cracks by inservice inspection (ISI). Inspection requirements of ASME Section XU, Table IWB 2500-1, examination category B-G-1, specify the following for all closure stud bolting: volumetric examination of studs in place, from top of the nut to bottom of the flange hole, and surface and volumetric examination of studs when removed: volumetric examination of flange threads; and visual VT- I examination of surfaces of nuts, washers, and bushings.

RICSIL Rev. I and its Supplement 1 provide additional recommendations regarding inspection and evaluation of the data. Requirements for training and qualification of personnel and performance demonstration for procedures and equipment is in conformance with Appendices VII and VIII of ASME Section XM. Inspection requirements of testing category B-P conducted according to IWA-5000 specify visual Vr-2 (IWA-5240) examination of all pressure retaining components extending to and including the second closed valve at the boundary extremity, during system leakage test (IWB-5221) and system hydrostatic test (IWB-5222).

(4) Detection of Aging Fffects: Aging effects degradation of the closure stud bolting can not occur without crack initiation, the extent and schedule of inspection assure detection of cracks before the loss of intended function of closure stud bolting. (5) Monitoring and Trending:

Inspection schedule of ASME Section XI IV REACTOR VESSEL, INTERNALSS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism

.4, Al. 1.5 Top Head Vessel Flange Stainless g Crack SCC. Enclosure Leak Detection Steel xygenated Initiation IGSCC Line ater and Growth and/or team up to 88°C A1.2.1, Vessel Shell Vessel Flange, SA302-Gr B 288°C Cumulative Fatigue AI.2.2 Upper Shell SA533-Gr B Steam Fatigue SA336 Damage with SS cladding A1.2.3 Vessel Shell Intermediate SA302-Gr B 288 0 C. Cumulative Fatigue thru (Nozzle) Shell, SA533-Gr B Fatigue A1.2.6 Intermediate with Water, Damage (Beltline) 308. 309, nax 5x10 9 Shell. Lower 308L, 309L n/cm2.s Shell, Beltline cladding Welds DRAFT -6/06/00 rV A1-8 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (BoUing Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continuedfrom previous page) and, based on operating experience, additional requirements of RICSIL 055 Rev. 1, are effective and adequate for timely detection of cracks. All BWRs are inspected in accordance with Program B lWB-2412 which requires 100% inspection every 10 y, at least 16% in 3 y and 50% in 7 y. Recommendations of RICSIL 055 include expansion of sample size and ultrasonic examination from the center drilled hole of studs in compliance with ASME Code Case N-307- 1. System leakage test is conducted prior to plant startup following each refueling outage, and hydrostatic test at or near the end of each inspection interval.

(6) Acceptance Criteria:

Any cracks in closure stud bolting are evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3515/17.

(7) Corrective Actions: Repair and replacement is in conformance with IWB-4000 and material and inspection guidance of RG 1.65. (8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

SCC has occurred in BWR pressure vessel head studs. The AMP based on ASME Section XI and industry guidelines of RICSIL 055 Revision I and its Supplement

1. provides recommendations regarding inspection techniques and evaluation, material specifications, corrosion prevention, and other aspects of reactor pressure vessel head stud cracking, and is effective in managing the effects of SCC to maintain the intended function of closuretuds and nuts during the period of license renewal.

Plant-specific aging management Plant-specific aging management program is to be Yes, program: existing programs may not be evaluated, no AMP capable of mitigating or detecting SCC of vessel flange leak detection line. Components have been designed or Fatigue is a time-limited aging analysis (T'LAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert #8. ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB, or Section I (Power Boilers) and Section VIII, Division I (Unfired Pressure Vessel).

Components have been designed or Fatigue is a time-limlted aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic 71AA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # 1. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB. or Section I (Power Boilers) and Section VIII, Division 1 (Unfired Pressure Vessel).DRAFT -6/06/00 IV A1-9 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism A 1.2.4 Vessel Shell Intermediate SA302-Gr B, 2880C, Loss of Neutron (Beltline)

Shell SA533-Gr B Oxygenated Fracture Irradiation with Water. Toughness Embrittle 308,309, 5x10 8 -ment 308L, 309L x10 9 Cladding n/cm2.s AI.2.3 Vessel Shell Intermediate SA302-Gr B, 88C, Crack SCC, thru (Nozzle) Shell, SA533-Gr B Oxygenated Initiation IGSCC AI.2.6 Intermediate with Water, and Growth (Beltline) 308, 309, 5x10 8 Shell, Lower 308L, 309L 'x10 9 Shell. Beltline Cladding /cm2.s Welds Ay DRAFT- 6/06/00 IV AI-10 IV 'REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation For a 40 y design life, pressure vessel Neutron irradiation embrittlement is a time-limited aging Yes integrity is assured by fracture analysis (TLAA) to be evaluated for the period of license TLAA toughness and material surveillance renewal for all ferritic materials that have a neutron program requirements set forth in fluence of greater than 1017 n/cm 2 (E> I MeV) at the end Appendices G and H to 10 CFR Part 50, of the license renewal term. The TflAA should evaluate the and methodology of Regulatory Guide impact of neutron embrittlement on: (a) the adjusted 1.99, Rev. 2, implemented through reference temperature, the plant's pressure temperature Generic Letters (GLs) 88-1 1 and 92-01. limits, and the need for Inservice inspection of Rev. 1, Supplement

1. to predict effects circumferential and axial reactor vessel welds, (b) the of neutron irradiation on reactor vessel Charpy upper shelf energy, and (c) the equivalent margins materials.

In addition, inservice analyses performed in accordance with 10 CFR 50, inspection of ASME Section XM. edition Appendix G. Reactor surveillance program requires that specified in I OCFR50.55a, Subsection the existing reactor vessel material surveillance program IWB, examination category B-A of all be evaluated to determine whether there is sufficient pressure retaining welds in the vessel material data and dosimetry to monitor irradiation and repair welds in beltline region, embrittlement at the end of the license renewal term and defined as the region extending for the whether operating restrictions (i.e., inlet temperature.

length of the thermal shield or effective neutron spectrum and flux) are necessary.

If surveillance length of reactor fuel elements.

NRC capsules are not removed during the license renewal term Generic Letter 98-05 covers exemptions it will be necessary to establish operating restrictions to from inspection requirements for ensure the plant is operated within the environment of the circumferential welds, surveillance capsules.

ISupporting documents BWRVIP-05, -29, -74, and -78] Inservice inspection in conformance (1) Scope of Program: The program is focused on Yes with ASME Section X. edition specified managing the effects of stress corrosion cracking (SCC) of BWRVIP In I OCFR5,55a, Codes and Standards), SS cladding on the intended function of reactor vessel Guideline Subsection IWB, Table IWB 2500-1, shell. NUREG-0313 and GL 88-0 1, respectively, describe examination categories B-N- 1 for vessel the technical basis and staff guidance regarding the interior and B-A for shell welds, problem of IGSCC in BWRs. However, SCC is not Prevention is by material selection in anticipated to be an issue for the vessel shell because accordance with guidelines of NUREG- analytical evaluations and experimental Jlta indicate that 0313, Rev. 2, and of Regulatory Guide growth of the cracks in ferritic base metal will be very 1.43 for control of stainless steel weld slow. (2) Preventive Actions: Selection of material, cladding of low-alloy steels. Coolant considered resistant to IGSCC, e.g., grades of weld metal water chemistry is monitored and with a maximum carbon of 0.035% and minimum 7.5% maintained In accordance with EPRI ferrite, prevent or mitigate IGSCC. and Regulatory Guide guidelines in BWRVIP-29 and TR- (RG) 1.43 provides assurance that production cladding 103515 to minimize the potential of complies with ASME Section ll and XI guidelines to crack initiation and growth. NRC prevent underclad cracking.

Coolant water chemistry is Generic Letter 98-05 covers exemptions monitored and maintained in accordance with EPRI from inspection requirements for guidelines in BWRVIP-29 and TR- 103515 to minimize the circumferential welds. BWRVIP-74 for potential of crack initiation and growth. Also, hydrogen reactor pressure vessel inspection and water chemistry and stringent control of conductivity is flawevaluation guidelines is under staff used to inhibit IGSCC. (3) Parameters review. Monitoredllnspected:

Inspection and flaw evaluation are lSupporting documents BWRVIP-03 for to be performed in accordance with referenced BWRVIP reactor pressure vessel internals guideline, as approved by the NRC staff. (4) Detection qf examination guidelines:

BWRVIP- 14. Aging Effects: Aging effects degradation of the reactor -59. and -60 for evaluation of crack vessel shell can not occur without crack initiation.

orowth: BWRVIP-44 for weld repair of However, because of inaccessibility, the extent and size of Ni-alloys:

BWRVIP-45 for weldabilitv of inspection may not be adequate to assure detection of irradiated structural comoonents:

cracks in the SS cladding before the loss of intended IBWRVIP-62 for technical basis for function of the reactor vessel. (5) Monitoring and inspection relief for internal components Trending:

Inspection schedule in accordance with with hydrogen ninection:

and BWRVTIP- applicable approved BWRVIP guideline.

(6) Acceptance 78 BWR integrated surveillance Criteria:

Any IGSCC degradation is evaluated in -accordance with applicable approved BWRVIP guideline.

DRAFT -6/06/00 TV A1-11I REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism A1.2.6 Vessel Shell BeltJine Welds Low-alloy 288 0 C, Loss of Neutron steel (LAS) Oxygenated Fracture Irradia weldments Water, Toughness tion with x 10 8 -Embrittle 308, 309, x10 9 ment 308L, 309L /cm 2.s cladding Al.2.7 Vessel Shell Attachment SS. 288-C, Crack SCC. Welds Inconel 182 Oxygenated Initiation IGSCC Water and Growth DRAFT- 6/06/00 IV IVAl-12 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) (7) Corrective Actions: The corrective action proposed by the BWRVIP is under staff review. (8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The present AMP is effective in managing crack initiation and growth due to SCC, however, because of inaccessibility, the extent and size of inspection may not be adequate to assure detection of cracks. Same as for the effect of Neutron Same as for the egfect of Neutron Irradiation Embrittlement Yes Irradiation Embriatlement on Item A2. 1.4 on Item A2.I.4 intermediate (beltline) shell. T1AA intermediate (beltline) shell. Inservice inspection in conformance (1) Scope of Program: The program includes preventive N1 with the guidelines of BWRVIP-48 and measures to mitigate stress corrosion cracking (SCC) and ASME Section XM. edition specified in inservice inspection (ISI) to monitor the effects of SCC on IO,05a, Codes and Standards), the intended function of the component.

NUREG-0313 Subsection IWB, Table IWB 2500-1, and GL 88-01, respectively, describe the technical basis examination categories B-N-2 for and staff guidance regarding the problem of IGSCC in integrally welded core support structure.

BWRs. (2) Preventive Actions: Mitigation is by selection Prevention is by material selection in of materials resistant to IGSCC and control of coolant accordance with guidelines of NUREG- water chemistry in accordance with EPRI guidelines in 0313, Rev. 2, Coolant water chemistry BWRVIP-29 and TR- 103515 Including stringent control of is monitored and maintained in conductivity (many BWRs now operate at <0. 15 liS/cm 2). accordance with EPRI guidelines in Hydrogen additions are effective in reducing BWRVIP-29 and TR- 103515 to minimize electrochemical potentials in the recirculating piping the potential of crack initiation and system, but are less effective in the core region. Also, the growth. susceptibility of Ni-alloys to SCC should be evaluated.

(3) Parameters Monitored/Inspected:

The AMP monitors the effects of IGSCC on the intended function of the component by detection and sizing of cracks by inservlce inspection (ISI). Inspection requirements of Table IWB 2500- 1, examination category B-N-2 specifies visual VT-3 examination of all accessible surfaces of DRAFT- 6/06/00 IV Al-13 IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Region of Interest Material Aging Effect DRAFT-6/06/00 IV Al-14 IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM A. -1 12% A 5%'^1 IVF~ VTJ ru a.5-+.Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) integral welds. (4) Detection of Aging Fffects: Degradation due to SCC can not occur without crack initiation and growth. Attachment weld inspection and flaw evaluation guidelines are provided in BWRVIP-48.

(5) Monitoring and Trending:

Inspection schedule in accordance with IWB-2400 ani BVM P-4 is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with IWB 3520 and BWRVIP-48.

(7) Conrective Actions: Repair and replacement are in conformance with IWB-3140.

(8 & 9) Conftmration Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

IGSCC has occurred BWR components.

The program addresses improvements in all three of the elements, viz., a susceptible (sensitized) material, significant tensile stress, and an aggressive environment, that cause IGSCC. Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)-190 Is to be addressed.

Insert #8. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or Section I (Power Boilers) and Section VIII, Division 1 (Unfired Pressure Vessel).

Inservice inspection in conformance

11) Scope of Proaram: The program is focused on with ASME Section XI (edition specifled.

managing the effects of crack initiation and growth due to In 10 CFR 50.55a1. Subsection IWB. unanticloated cyclic loading by inservice insoection (ISI). Table IWB 2500-1 .examination W Preventive Actions: Selection of matekl considered categories B-D for nozzle-to-vessel resistant to to enhanced crack growth is In accordance welds, and testing category B-P for with guidelines of NUREG-0313.

Rev. 2. and Regulatory system leakage. Selection of materials Guide (RGI 1.43 provides assurance that production considered resistant to enhanced crack cladding complies with ASME Section III and XM guidelines growth is in accordance with guidelines to orevent underclad cracking.

Coolant water chemistry is of NUREG-0313.

Rev. 2. Coolant water monitored and maintained in accordance with EPRI chemistry Is monitored and maintained guidelines in BWRVIP-29 and TR- 103515. [3) Parameters in accordance with EPRI guidelines in Monitored/Inspmcted:

The AMP monitors the effects of BWRVIP-29 and TR- 103515 to minimize crack initiation and growth by detection and sizing of the potential of crack initiation and cracks by inservice inspection fISn. Inspection g t requirements of Table IWB 2500-1. examination category [Supporting documents BWRVTP-74 for B-D specrfies for all nozzles volumetric inspection o0 reactor pressure vessel inspection and nozzle-to-vessel welds and nozzle inside radius section.

flaw evaluation guidelines:

BWRVIP-14.

Requirements for training and Qualification of personnel and -60 for evaluation of crack and performance demonstration for procedures and orowth- BWRVIP-62 for technical basis eQuipment is in conformance with Appendices VWI and VIII for inspection relief for internal of AME components with hydrogen jnjection:

BWRVIP-75 for technical basis for revisions to GL 88-01 inspection schedule:

and BWRVIP-78 BWR integrated surveillance proram._i DRAFT -6/06/00 IV Al-15 IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM Structure and Region of Environ- g Aging Item I Component I Interest 1 Material ment Effect JMechanism

+/- .1 £ L I"lV DRAFT -6/06/00 IV Al-16 IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Existing j Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous p=oe Section XI. or any other formal program approved by the NRC, System leakage test. IWB-522 1. is conducted prior to plant startup following each refueling outage and visual VT-2 frWA-5240) examination performed for all oressure retaining components extending to and including the second closed valve at the boundary extremity.

System hydrostatic test. IWB-5222.

is conducted at or near the end of each inspection interval and visual VT-2 examination performed for all class I components within the boundary.

(41 Detection of Anino Effects: Aging effects degradation of the reactor vessel nozzles can not occur without crack initiation:

extent and schedule of inspection assure detection of cracks before the loss of intended function of the reactor vessel nozzles.

(5) Monitoring and Trending:

Inspection schedule of ASME Section XM should provide for timely detection of cracks. All BWRs are inspected in accordance with Program B IWB-2412 which requires 100% insoection every 10 v: for reactor vessel nozzles at least 25% but not more than 50% shall be examined by the end of Ist insoection interval.

(6) Acceptance Criteria:

Any degradation is evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards ot IWB-3400 and IWB-3512.

Planar and liner flaws are sized according to IWA-3300 and IWA-3400.

Continued operation without repair require that crack growth calculation be performed according to the gruidance of GL 88-01 or other approved orocedures.

17) Corrective Actions: Repair and replacement are in conformance with IWA-4000 and IWB-4000.

and reexamination In accordance with requirements of rWA-22(d.

(8 & 91 Confirmation Process and Administrative Controls:

Site QA procedures.

review and approval processes.

and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operatino Experience:

NUREG-0619 summarizes work performed by the NRC to resolve Generic Technical Activity A- 10. "BWR Nozzle Crackinge and the industry testing and analysis Program Is described in GE NEDE-2182 1-A IV Al-17 DRAF -6/06/00 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism Al.3.2 Nozzles Feedwater, SA508-C12 Up to 288°C Cumulative Fatigue A1.3.6 CRDRL with or Oxygenated Fatigue without SS Water Damage cladding AI.3.8 Nozzles LPCI (or RHR SA508-C12 p to 288 0 C Loss of Neutron Injection xygenated Fracture Irradiation Mode] Vater, Toughness Embrittle x10 8-ment x109 a/ clm2. s Al.4.l Nozzle Safe HPCS, SS. Up to 288 0 C Crack SCC, thru Ends LPCS, SB- 166 Oxygenated Initiation IGSCC A1.4.5 CRDRL, (Inconel 182 Water and Growth Recirculating butter, and Water, Inconel 82 LPCI or RHR or 182 weld) Injection DRAFT- 6/06/00 IVAI-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL fBoitnn" Water Reactori Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Components have been designed or Fatigue is a time-limited aging analysis rLIAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # 1. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB. or Section I (Power Boilers) and Section VIII, Division I (Unfired Pressure Vessel).

The technical basis and staff guidance regarding the problem of feedwater nozzle cracking due to thermal cycling is described in NUREG-0619.

Same as for the effect of Neutron Same as for the effect of Neutron Irradiation Embrittlement Yes Irradiation Embrittlement on Item A2.1 .4 on Item A2. 1.4 intermediate (beltline) shell. TIAA intermediate fbelthine) shell. ",V Program delineated in NUREG-0313, (1) Scope of Program: The program is focused on No Rev. 2 and implemented through NRC managing the effects of IGSCC on the intended function of Generic letter (GL) 88-01 and its austenitic stainless steel (SS) piping 4 in. or larger in Supplement 1, and inservice Inspection diameter, and reactor vessel attachments and in conformance with ASME Section XI appurtenances.

Although these guidelines primarily (edition specified in 10 CFR 50.55a), address austenitic SS components, they are also applied to Subsection IWB, Table rWB 2500-I, nickel alloys. NUREG-0313 and GL 88-01, respectively, examination category describe the technical basis and staff guidance regarding B-F for pressure retaining dissimilar the problem of IGSCC in BWRs. (2) Preventive Actions: metal welds in vessel nozzles and testing Mitigation of IGSCC is by selection of material considered category B-P for system leakage. and resistant to sensitization and IGSCC. e.g., low-carbon additional recommendations of Nuclear grades of austenitic SSs and weld metal, with a maximum Services Information Letter (SIL) No. carbon of 0.035% and minimum 7.5% ferrite in weld 455, Rev. I and Supplement

1. BWRVIP metal, and by special processing such as solution heat guideline is under staff review. Coolant treatment, heat sink welding, and induction heating or water chemistry is monitored and mechanical stress improvement (SI). Inconel 82 is the only maintained in accordance with EPRI nickel base weld metal considered to be resistant to guidelines in BWRVIP-29 and TR- IGSCC. Coolant water chemistry is monitored and 103515 to minimize the potential of maintained in accordance with EPRI guidelines in crack initiation and growth. BWRVIP-29 and TR- 103515. Also, hydrogen water BWRVIP-75 technical basis for revisions chemistry and stringent control of conductivity is used to to GL 88-01 inspection schedule are inhibit IGSCC. (3) Parameters Monitored/Inspected:

The under staff review AMP monitors the effects of IGSCC on the intended DRAFT -6/06/00 IVAl-19 IV REACTOR VESSEL. INTERNALS.

AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Structure and Region of Environ- Aging I. Aging Item Component Interest I Material ment Effect Mechanism.1 J I I &DRAFT -6/06/00 IV Al-20 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Existing Aging Management Program (AMP)DRAFT- 6/06/00 Further Evaluation Evaluation and Technical Basis (continued from previous page) function of reactor vessel nozzle safe ends by detection and sizing of cracks by inservice inspection (ISI). Inspection requirements of ASME Section XI, Subsection IWB. Table IWB 2500- 1, examination category B-F specifies for all nozzle-to-safe end butt welds NPS 4 or larger, volumetric and surface examination of ID region extending 1/4 in. on either side of the weld and 1/3 wall thickness deep. and surface examination of OD surface extending 1/2 in. on either side. Only surface examination is conducted for all butt welds less than NPS 4. For all nozzle-to-safe end socket welds, surface examination is specified of OD surface extending I in. on the buttered side and 1/2 in. on the other. Requirements for training and qualification of personnel and performance demonstration for procedures and equipment is in conformance with Appendices VII and VIII of ASME Section XI. or any other formal program approved by the NRC. SIL No. 455 and Supplement I contain specific recommendations regarding ultrasonic testing (UTL methods for dissimilar metal welds, i.e., the use of 45-degree and 60-degree refracted longitudinal wave transducers for detecting IGSCC cracks in alloy 182 and low-alloy materials.

Visual VT-2 (IWA-5240}

examination is performed for all pressure retaining components during system leakage test (IWB-522 1), conducted prior to plant startup following each refueling outage, and during system hydrostatic test (IWB-5222) conducted at or near the end of each inspection interval.

Leakage detection is in conformance with Position C of Regulatory Guide 1.45 and additional guidelines of GL 88-01, Supplement

1. (4) Detection of Aging Fffects: Aging effects degradation of the nozzle safe ends can not occur witHl'ut crack initiation:

extent and schedule of inspection assure detection of cracks before the loss of intended function of the reactor vessel nozzle safe ends. (5) Monitoring and Trending:

Inspection schedule of ASME Section XI should provide for timely detection of cracks. Inspection schedule and sample size specified in Table 1 of GL 88-01 are based on the condition of each weld and are adequate for timely detection of cracks. Welds of resistant material are as a minimum examined according to an extent and frequency comparable to those of ASME Section X9, e.g., 25% are examined every 10 y, at least 12% in 6y. Inspection extent and schedule are enhanced for welds of non resistant materials, or welds that have been treated by stress improvement (SI) or reinforced by weld overlay.

(6) Acceptance Criteria:

Any IGSCC degradation is evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3514.

Planar and liner flaws are sized according to IWA-3300 and -3400. (7) Corrective Actions: Repair and reexaminations are in conformance with IWB-4000.

Continued operation without repair requires that crack growth calculation be performed according to the guidance of GL 88-01 or other approved procedures.

(8 & 9) Conftmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and IVAl-21 REACTOR VESSEL, IUTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism A1.4.3 Nozzle Safe CRDRL SS, Up to 288°C Cumulative Fatigue Ends SB-166 Oxygenated Fatigue (Inconel 182 Water Damage butter, and Inconel 82 or 182 weld]Penetrations CRD Stub Tubes, Instrumenta tion, Jet Pump Inst., Standby Liquid Control, Flux Monitor, Drain Line SS, SB- 167 LUp to 2880C, Dxygenated Water Crack Initiation and Growth SCC, IGSCC, Unantic1 =tcd~n____ I I J. & __________

DRAFT-6/06/00 IV A1.5.1 thru A1.5.6 IV Al-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) will continue to be adequate for the period of license renewal. (10) Operating Experience:

IGSCC has occurred in small- and large-diameter BWR piping safe end-to-nozzle welds (IN 82-39 & IN 84-41). The present AMP has provided effective means of ensuring structural integrity of the primary coolant pressure boundary.

Components have been designed or Fatigue is a time-limited aging analysis (T1LAA to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSl)-190 is to be addressed.

Insert # I. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF).riogram delneatea m NURE;-o313, Rev. 2 and implemented through NRC Generic letter 88-01 and its Supplement I, and inservice inspectior in conformance with ASME Section X) (edition specified in 10 CFR 50.55a), Subsection IWB, Table IWB 2500-1, examination category B-E for pressure retaining partial penetration welds and testing category B-P for system leakage Coolant water chemistry Is monitored and maintained in accordance with EPRI guidelines in BWRVIP-2, and TR 103515 to minimize the potential of crack initiation and growth. Inspceion and flaw evaluation guidelines for instrument Penetratlon (BWRVIP-4Q) and for standby liould control system/core plate AP (BWRVIP-271 are under staff review. fSupporting documents for renair desfor criteria BWRVIP-57 for instrumentation penetrations and BWRVIP-53 for standby liquid control line' BWRVIP-14, -59. and -60 for evaluation of crack growth: BWRVIP-62 for technical basis for insoection relief for internal comoonents with inlpe'tlon and for revisions to GL 88-01 inspection scheduic.(1) Scope of Program: NUREG-0313 and GL 88-01, respectively, describe the technical basis and staff guidance regarding the problem of IGSCC in BWRs. The program is i focused on managing the effects of IGSCC on the intended function of austenitic stainless steel (SS) piping 4 in. or larger In diameter, and reactor vessel attachments and appurtenances.

Although these guidelines primarily address austenitic SS components, they are also applied to nickel alloys. (2) Preventive Actions: Mitigation of IGSCC is by selection of material considered resistant to sensitization and IGSCC, e.g., low-carbon grades of austenitic SSs and weld metal, with a maximum carbon of 0.035% and minimum 7.5% ferrite in weld metal, and by special processing such as solution heat treatment, heat sink welding, and induction heating or mechanical stress improvement.

Inconel 82 is the only nickel base weld meta considered to be resistant to IGSCC. Coolant water chemistry is monitored and maintained id1fccordance with EPRI guidelines in BWRVIP-29 and TR-103515.

Also. hydrogen water chemistry and stringent control of conductivity is used to inhibit IGSCC. (3) Parameters Monitored fnspected:

The AMP monitors the effects of IGSCC on the intended function of reactor vessel penetrations by detection and sizing of cracks by inservice inspection (ISI). System leakage test, IWB-522 1, is conducted prior to plant startup following each refueling outage and visual VT-2 (IWA-5240) examination performed for all pressure retaining components extending to and including the second closed valve at the boundary extremity.

Leakage detection is in conformance with Position C of Regulatory Guide 1.45 and additional guidelines of GL 88-01, Suppl. 1. System hydrostatic test, IWB-5222, is conducted at or near the end of each inspection interval and visual VT-2 examination performed for all class I components within boundary.

Inspection requirements of examination category B-E focus on visual VT-2 examination of partial penetration welds during the hydrostatic test. (4) Detection of Aging Effects: Aging effects degradation of the reactor vessel penetrations can not occur without crack initiation:

extent and schedule of inspection assure detection of cracks before loss of intended function of the reactor vessel penetrations.

(5) Monitoring and Trending:

Inspection schedule of ASME Section Xl should provide for timely detection of No DRAFT- 6/06/00 S... .. T for insnection relief for internn]comnonentq ulth hyrimeypn jPrtfen4 and RIVRVIP-7.1; fnr fpýJinlnl h 4 fýIV Al-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM AL. REACTOR VESSEL (oilling Water Reactor)Item*I ,* ~.II ., --- -Component rvcgrzon of Interest Materict m e n tin ~1 Mechanism

--4~~qh Environ ment Aging AI.5.1 Penetrations CRD Stub SS Up to 288°C Cumulative Fatigue thru Tubes, SB- 167 Oxygenated Fatigue A1.5.6 Instrumenta-ater Damage tion, Jet Pump Inst., Standby Liquid Control, Flux Monitor, Drain Line AI.6 Bottom Head -SA302-Gr B Up to 2880C Cumulative Fatigue SA533-Gr B xygenated Fatigue with Water Damage 308, 309, 308L, 309L 1cladding AI.7.1 Control Rod Housing SS Up to 288°C Crack SCC, Drive (CRD) Oxygenated Initiation IGSCC Mechanism Water and Growth DRAFT -6/06/00 rV AI-24 I Material IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) cracks. Inspection schedule and sample size specified in Table I of GL 88-01 are based on the condition of each weld and are adequate for timely detection of cracks. Welds of resistant material are as a minimum examined according to an extent and frequency comparable to those of ASME Section XM. Inspection extent and schedule are enhanced for welds of non-resistant materials.

(6) Acceptance Criteria:

Any IGSCC degradation is evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3522.

(7) Corrective Actions: Repair and replacement are in conformance with IWA-4000 and IWB 4000, and reexamination in accordance with requirements of IWA-2200.

(8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. ( 0) Operating Experience:

The program addresses improvements in all three of the elements, viz.. a susceptible (sensitized) material, significant tensile stress, and an aggressive environment, that cause IGSCC. and has provided effective means of ensuring structural integrity of the primary coolant pressure boundary.

Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic T1AA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

InserlL ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or Section I (Power Boilers) and Section VIII, y Division I (Unfired Pressure Vessel).

Components have been designed or Fatigue is a time-limited aging analysis (TIAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic "IfAA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

Insert # I1. ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB, or Section I (Power Boilers) and Section VIII, Division 1 (Unfired Pressure Vessel).

Inservice inspection in conformance (1) Scope qf Program: The program is focused on Yes, with ASME Section XI (edition specified managing the effects of stress corrosion cracking (SCC) on BWRVIP in 10 CFR 50.55a), Subsection IWB, the intended function of CRD mechanism housing. Guideline Table IWB 2500- 1, examination (2) Preventive Actions: Mitigation of IGSCC is by selection (Element 7) categories B-0 for pressure retaining of material considered resistant to sensitization and welds In control rod housings and IGSCC. e.g.. low-carbon grades of austenitic SSs and weld testing category B-P for system leakage, metal, with a maximum carbon of 0.035% and minimum and BWRVIP-27.

Prevention is by 7.5% ferrite in weld metal, and by special processing such material selection in accordance with as solution heat treatment, heat sink welding, and guidelines of NUREG-0313.

Rev. 2. " induction heating or mechanical stress improvement.

Coolant water chemistry is monitored Inconel 82 is the only nickel base weld metal considered to and maintained in accordance with be resistant to IGSCC. Coolant water chemistry is EPRI guidelines in BWRVIP--29 and TR- monitored and maintained in accordance with EPRI 103515 to minimize the potential of guidelines in BWRVIP-29 and TR-103515.

Also, hydrogen crack initiation and growth. BWRVIP water chemistry and stringent control of guideline is under staff review. II__ _DRAFT- 6/06/00 IV Al-25 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism A1.7.1 CRD Housing SS Jpto288 0 C Cumulative Fatigue Mechanism Oxygenated Fatigue Water Damage A1.8 Support Skirt -SA533-Gr B Ambient Cumulative Fatigue & Attachment (Welds SS oz Temperature Fatigue Welds Inconel 182) Ar Damage ALL72 CMR Withdrawal Crck Stress Mechanism Lin Iniaiaon Corrosion Surfacel andxyrowth CdaGkin DRAFT -6/06/00 IV Al-26 1V REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Al. REACTOR VESSEL (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation ISupporting documents fBWRVIP-58 for (continuedfrom previous page) CRD internal access weld repair: conductivity is used to inhibit IGSCC. (3) Parameters B7WRVIP- 14. -59. and -60 for evaluation Monitored/inspected:

The AMP monitors the effects of of crack growth: BWRVIP-62 for IGSCC on the intended function of CRD mechanism technical basis for inspection relief for housing by detection and sizing of cracks by inservice internal components with hydrogen inspection (ISI). Inspection requirements of Table IWB inlection' and BWRVIP-53 for standby 2500- 1. examination category B-O specifies volumetric or liquid control line repair design crlteria.1 surface examination extending 1/2 in. each side of the CRD housing welds, including weld buttering.

(4) Detection of Aging Effects: Aging effects degradation of the CRD mechanism housing can not occur without crack initiation; the extent and schedule of inspection assure detection of cracks before the loss of intended function of the CRD housing. (5) Monitoring and Trending:

Inspection schedule in accordance with Program B IWB-2412 should provides timely detection of cracks. 10% peripheral CRD housings are examined each inspection interval.

(6) Acceptance Criteria:

Any IGSCC degradation is evaluated in accordance with IWB-3100 by comparing ISI results with the acceptance standards of IWB-3400 and JWB-3523.

Planar and liner flaws are sized according to IWA-3300 and IWA-3400.

(7) Corrective Actions: The corrective action proposed by the BWRVIP is under staff review. (8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The program addresses improvements in all three of the elements, viz., a susceptible (sensitized) material, significant tensile stress, and an aggressA environment, that cause IGSCC, and has provided effective means of ensuring structural integrity of the primary coolant pressure boundary.

Components have been designed or Fatigue is a time-limited aging analysis (TL.AA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # 1. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or other evaluations.

Components have been designed or Fatigue is a time-limited aging analysis (TLAAN to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal. "IAA life, according to the requirements of ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB. or other evaluations.

The chlorides from insulation and other Plant soecific aging management nrogram is to be Ys. sources can cause externally-initiated evaluated, no generic transgranular stress corrosion cracking AM ITGSCC) in the stainless steel lines. Plant specific aging management program should be imDlemented.

DRAFT -6/06/00 IV Al-27 BI. Reactor Vessel Internals (Boiling Water Reactor) B 1.1 Core Shroud, Shroud Head, and Core Plate BI.I.1 Core Shroud Head Bolts B1.1.2 Core Shroud (Upper, Central, Lower) B 1.1.3 Core Plate B 1.1.4 Core Plate Bolts B 1.1.5 Access Hole Cover B 1. 1.6 Shroud Support Structure B 1.1.7 Standby Liquid Control Line B 1. 1.8 LPCI Coupling B11.2 Top Guide B 1.3 Feedwater Spargers B 1.3.1 Thermal Sleeve BI.3.2 Distribution Header B 1.3.3 Discharge Nozzles B 1.4 Core Spray Lines and Spargers B 1.4.1 Core Spray Lines (Headers)

B1.4.2 Spray Ring B 1.4.3 Spray Nozzles B1.4.4 Thermal Sleeve B 1.5 Jet Pump Assemblies B1.5.1 Thermal Sleeve B 1. 5.2 Inlet Header B1.5.3 Riser Brace Arm DRAFT -6/06/00 TV BI-1 B1.5.4 Holddown Beams B11.5.5 Inlet Elbow B1.5.6 Mixing Assembly B11.5.7 Diffuser B11.5.8 Castings B1.5.9 Jet Pump Sensing Line B 1.6 Fuel Supports & CRD Assemblies B 1.6.1 Orificed Fuel Support B13.7 Instrument Housings B 1.7.1 Intermediate Range Monitor (IRM) Dry Tubes B 1.7.2 Low Power Range Monitor (LPRM) Dry Tubes B 1.7.3 Source Range Monitor (SRM) Dry Tubes DRAFT- 6/06/00 IV B 1-2 BI. Reactor Vessel Internals (Boiling Water Reactor)System, Structures, and Components The system, structures, and components included in this table comprise the boiling water reactor (BWR) reactor vessel internals and consist of control rod guide tubes, core shroud and core plate, top guide, feedwater spargers, core spray lines and spargers, jet pump assemblies, fuel supports and control rod drive (CRD) housings, and instrument housings such as the intermediate range monitor (IRM) dry tubes, low power range monitor (LPRM) dry tubes, and source range monitor (SRM) dry tubes. All structures and components in the reactor vessel are classified as Group A or B Quality Standards.

The steam separator and dryer assemblies are not part of the pressure boundary and are removed during each outage, and should be covered by the plant maintenance program.

System Interfaces The systems that interface with the reactor vessel internals include the reactor pressure vessel (Table IV A1) and reactor coolant pressure boundary (Table IV Cl).DRAFT -6/06/00 IV B 1-3 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM ]81. REACTOR VESSEL INTERNAlS Water Reactori Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B I.I.1 Core Shroud. Core Shroud Alloy 600, Z88*C, Crack Stress Shroud Head Head Bolts Stainless High-Purity Initiation and Corrosion and Core Plate Steel (SS) Water Growth Cracking (SCc) Bl. 1.11 Core Shroud, Core Shroud Alloy 600, 2880C, Cumulative Fatigue Shroud Head Head Bolts SS High-Purity Fatigue and Core Plate Water Damage B 1.1.2 Core Shroud, Core Shroud SS 288 0 C. Crack Stress Shroud Head (Upper, High-Purity Initiation and Corrosion and Core Plate Central, Water Growth Cracking Lower) (SCC). Irradiation Assisted Stress Corrosion Cracking (IASCC)DRAFT- 6/06/00 IV B 1-4 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Existing Aging Management Program (AMP) Visual inspection is performed accordil to ASME Section XM, IWB-2500, categor B-N-2. and GE Services Information Letter (SIL) 433 recommends ultrasonic (LT inspection during outages, verification of required torque on bolt during shroud head removal and attachment, and replacement of bolts with crevice design by a design which is crevice-free.

Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. BWRVIP-07 and -63 for inspection and evaluation nf rnre _-hrntmr1 ont 76 for ??? are under staff review, rSuDDooIting documents BWRVIP-03 for examination guidelines:

BWRVIP- 14. -59. and -60 for evaluation of crack growth: BWRVIP-44 for weld repair of NI-alloys:

BWRVIP-45 for weldability of irradiated structural components:

and BWRVIP-62 for technical basis for inqnrtfinn rT"1lf fnr lrfprnl ,n,'-I,,.

+o with hydrogen iniection.l I EFurther Evaluation and Technical Basis IEvaluation ig (1) Scope qf Program: The program includes preventive y measures to mitigate SCC, inservice inspection (ISI) to monitor the effects of SCC on the intended function of the components, and repair and/or replacement as needed to maintain the capability to perform the intended function.

(2) Preventive Actions: Maintaining high water purity (many BWRs now operate at <0.15 ;IS/cm 2) reduces susceptibility to SCC. Hydrogen additions are effective in reducing electrochemical potentials in the recirculation piping system, but are less effective in the core region. Noble metal additions through a catalytic action appear o increase the effectiveness of hydrogen additions in the core region, but only limited data are .avaflable at pr-e.ento demonetrat@e eafec.tivenec.

GE Services Information Letter (SIL) 433 recommends replacement of bolts with crevice-free design. (3) Parameters Monitored/Inspected:

Inspection and flaw evaluation are to be performed in accordance with referenced BWRVIP guideline, as approved by the NRC staff. (4) Detection of Aging Effects: Degradation due to SCC can not occur without crack initiation and growth, inspection schedule assures detection of cracks before the loss of Intended function of the component.

(5) Monitoring and Trending:

Schedule in accordance with applicable, approved BWRVIP guideline is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with applicable, approved BWRVIP guideline.

(7) Corrective Actions: The corrective action proposed by the BWRVIP is under staff review. (8 & 9) Confirmation Process and Administrative Controls:

Site QA proceires, review and approval processes, and administrative cohtrols are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The present AMP has been effective in managing the effects of SCC on the intended function of core shroud head bolts.Yes, BWRVIP, Guideline Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes, evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insert #1. original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG. Visual inspection (VT-3) is performed (1) Scope qf Program. The program includes preventive Yes, according to ASME Section XI. IWB- measures to mitigate SCC, inservlce Inspection OSI) to BWRVIP 2500, category B-N-2. Guidance for monitor the effects of SCC on the intended function of the Guideline enhanced VT-I and uT inspections in components, and repair and/or replacement as needed to plant specific programs.

Coolant water maintain the capability to perform the intended function.

chemistry is monitored and maintained (2) Preventive Actions: Maintaining high water purity in accordance with EPRI guidelines in (many BWRs now operate at <0.15 ;iS/cm 2) reduces TR- 103515 and BWRVIP-29 to minimize susceptibility to SCC. Hydrogen additions are effective in the potential of crack initiation and reducing electrochemical potentials in the recirculation growth. Plant programs also may piping system, but are less effective in the core region. include water chemistry measures such Noble metal additions through a catalytic action appear-to as strict controls on conductivity, increase the effectiveness of hydrogen additions in the core hydrogen addition, and use of noble region. metal additions such as palladium or DRAFT- 6/06/00 S................................ AA A 2 IVBI-5 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM BI. RFACTOR T'NTERNALS ft]dll~nz Wat.. RPnittwrl Structure and Region of .Environ-Aging Aging Item Component Interest Material ment Effect Mechanism Bl1.3, Core Shroud. Core Plate, SS 288 0 C. Crack SCC, B 1. 1.4 Shroud Head Core Plate High-Purity Initiation and IASCC and Core Plate Bolts (used in Water Growth early BWRs)DRAFT- 6/06/00 IV BI1-6 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM DlI. REACTOR VESSEL INTERNALS (BoilinE Water Reactori Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) (continued from previous page) platinum to reduce electrochemical (3) Parameters Monitoredflnspected:

Inspection and potential.

Either preventive or flaw evaluation are to be performed in accordance with restorative mechanical repairs may be referenced BWRVIP guideline, as approved by the NRC made to the shroud. Possible inspection staff. (4) Detection of Aging FOffects:

Degradation due to relief based on hydrogen injection is SCC can not occur without crack initiation and growth. currently under staff review. BWRVIP- Extensive cracking has been observed at both horizontal 07 and -63 for inspection and [NRC Generic Letter (GL) 94-03] and vertical INRC evaluation of core shrouds and BWRVIP- Information Notice (IN)97-171 welds. (5) Monitoring and 76 for ??? are under staff review. Trending:

Inspection schedule in accordance with fSupporting documents BWRVIP-03 for applicable, approved BWRVIP guideline is adequate for reactor pressure vessel internals timely detection of cracks. (6) Acceptance Criteria:

Any examination guidelines:

BWRVIP- 14. degradation is evaluated in accordance with applicable.

-59. and -60 for evaluation of crack approved BWRVIP guideline.

(7) Corrective Actions: The growth: BWRVIP-44 for weld repair of corrective action proposed'by the BWRVIP is under staff Nl-alloys:

BWRVIP-45 for weldabllity of review. (8 & 9) Cornfrmation Process and irradiated structural components:

and Administrative ControWs:

Site QA procedures, review and BWRVIP-62 for technical basis for approval processes, and administrative controls are inspection relief for internal components implemented in accordance with requirements of Appendix with hydrogen inlection.]

B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. ( 0) Operating Experience:

Cracking has occurred in a number of BWRs. It has affected shrouds fpbricated from Type 304 SS and Type 304L SS, which is generally considered to be more resistant to SCC. Weld regions are most susceptible, 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 GL 94-03 and NUREG-1544.

Some experiences with visual Inspections are discussed in IN 94-42. Visual inspection (VT-3) is performed (1) Scope of Program: The program includes preventive Yes, according to ASME Section Xl, IWB- measures to mitigate SCC, Inservice inspection (ISI) to BWRVIP 2500, category B-N-2 o B I-03 monitor the effects of SCC on the intended function of the Guideline guidelines fEVr-11. Guidance for components, and repair and/or replacement as needed to enhanced Vr-I and UT inspections in maintain the capability to perform the intended function.

plant specific programs.

Coolant water (2) Preventive Actions: Maintaining high water purity chemistry is monitored and maintained (many BWRs now operate at <0.15 uS/cm 2) reduces in accordance with EPRI guidelines in susceptibility to SCC. Hydrogen additions are effective in TR-103515 and BWRVIP-29 to minimize reducing electrochemical potentials in the recirculation the potential of crack initiation and piping system. but are less effective in the core region. growth. Plant programs also may Noble metal additions through a catalytic action appear-t include water chemistry measures such increase the effectiveness of hydrogen additions in the core as strict controls on conductivity, region. but only limid. da... t .. aaa .bie at Present to hydrogen addition, and use of noble demonc--ate their- effec*&-thme.

.(3) Parameters metal additions such as palladium or Monitored/

Inspected:

Inspection and flaw evaluation are platinum to reduce electrochemical to be performed In accordance with referenced BWRVIP potential.

Possible inspection relief guideline, as approved by the NRC staff. (4) Detection of based on hydrogen injection is currently Aging Fffects: Degradation due to SCC can not occur under staff review. BWRVIP-25 for core without crack initiation and growth. (5) Monitoring and plate inspection and flaw evaluation Trending:

Inspection schedule In accordance with guidelines is under staff review.DRAFT -6/06/00 IV B 1-7 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM 31. REACTOR VESSEL INTERNALS IBoilinE Water Reactor]Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B1.1.3 Core Shroud, Core Plate SS 288 0 C, Cumulative Fatigue Shroud Head High-Purity Fatigue and Core Plate Water Damage BI. 1.5 Core Shroud, Access Hole Alloy 600, 288°C, Crack SCC. Shroud Head Cover Alloy 82 & High-Purity Initiation and IASCC and Core Plate 182 welds Water Growth y DRAFT -6/06/00 IV B 1-8 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM 81. REACTOR VESSEL INTERNALS (Bolling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation

[Supporting documents BWRVIP-03 for (continued from previous page) reactor pressure vessel internals applicable, approved BWRVIP guideline is adequate for examination guidelines:

BWRVIP-07 and timely detection of cracks. (6) Acceptance Criteria:

Any -63 for inspection and evaluation of core degradation is evaluated in accordance with applicable, shrouds: BWRVIP-76 for ??:? BWRVIP- approved BWRVIP guideline.

(7) Corrective Actions: The 14, -59. and -60 for evaluation of crack corrective action proposed by the BWRVIP is under staff growth: BWRVIP-44 for weld repa&r of review. (8 & 9) Confirmation Process and NI-alloys:

BWRVIP-45 for weldabiltvy of Administrative Controls:

Site QA procedures, review and irradiated structural components:

and approval processes, and administrative controls are BWRVIP-62 for technical basis for implemented in accordance with requirements of Appendix inspection relief for internal components B to 10 CFR Part 50 and will continue to be adequate for with hvdroeen inlectjon.]

the period of license renewal. (10) Operating Experience:

Cracking of the core plate has not been reported, but the creviced regions beneath the plate are difficult to inspect.

NRC Information Notice (IN) 95-17 discusses cracking in top guides of the U.S. and overseas BWRs. Related experience in other components is reviewed in NRC GL 94 03 and NUREG-1544.

Components have been designed or Fatigue is a time-limited aging analysis rTLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of the Safety Issue (GSI)- 190 Is to be addressed.

Insert #1. original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG. Visual inspection (VT-3) is performed (1) Scope of Program: The program includes preventive No according to ASME Section XI, IWB- measures to mitigate SCC, inservice inspection (ISI) to 2500. category B-N-2. GE Services monitor the effects of SCC on the intenda4 function of the Information Letter (SIL) 462 Sup. 3 components, and repair and/or replacement as needed to recommends ultrasonic inspection maintain the capability to perform the intended function.

techniques.

Implementation of (2) Preventive Actions: Maintaining high water purity inspection program is plant specific. (many BWRs now operate at <0.15 gS/cm 2) reduces Coolant water chemistry is monitored susceptibility to SCC. Hydrogen additions are effective in and maintained in accordance with reducing electrochemical potentials in the recirculation EPRI guidelines in TR- 103515 and piping system, but are less effective in the core region. BWRVIP-29 to minimize the potential of Noble metal additions through a catalytic action crack initiation and growth. Plant increase the effectiveness of hydrogen additions in the core programs also may include water region. but o.ly lmited data are .....able at pr.e... to chemistry measures such as strict

• thir offecti-'en-.

Also, the susceptibility of controls on conductivity, hydrogen Ni-alloys to SCC is evaluated.

(3) Parameters Monitored/

addition, and use of noble metal Inspected:

The AMP monitors the effects of SCC on the additions such as palladium or intended function by detection and sizing of cracks by platinum to reduce electrochemical inservice inspection (PSI). Table IWB-2500, category B-N-2 potential.

specifies visual VT-3 examination of all accessible surfaces ISunDorting documents BWRVIP-03 for of core support structure.

Cracking initiates in crevice reactor pressure vessel internals regions not amenable to visual inspection.

GE Services examination guidelines:

BWRVIP-14.

Information Letter (SIL) 462 Sup. 3 recommends

-59. and -60 for evaluation of crack ultrasonic techniques for such inspections.

(4) Detection growth: BWRVIP-44 for weld repair 0o of Aging Fffects: Degradation due to SCC can not occur Ni-alloys:

BWRVIP-45 for weldability of without crack initiation and growth. Analysis may be irradiated structural components:

and required to assess the impact of observed cracking on the BWRVIP-62 for technical basis for function and integrity of the shroud. (5) Monitoring and inspection relief for internal components Trending:

Inspection schedule in accordance with IWB with hydrogen inlection.d 2400 is adequate for timely detection DRAFT-6/06/00 IV BI1-9 REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Structure and Region Of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism 9

• I Core Shroud, Shroud Head and Core Plate Shroud Support Structure (Shroud Support Cylinder, Shroud Support Plate. Shroud Support Legs)Alloy 600, Alloy 82 & 182 welds I88°C, High-Purity Water Crack Initiation and Growth 5CC. IASCC DRAFT- 6/06/00 IV BI.I.6 IV BI-10 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM BI. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Existing Further Aging Management Program LAMP) Evaluation and Technical Basis Evaluation (continued from previous page) of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with IWB-3520.

(7) Corrective Actions: Repair and replacement are in conformance with IWB-3140.

(8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

Jet pump boiling water reactors (BWRs) are designed with access holes in the shroud support plate at the bottom of the annulus between the core shroud and the reactor vessel wall. These holes are used for access during construction and are subsequently closed by welding a plate over the hole. Both circumferential (IN 88-03) and radial cracking (IN 92-57) has been observed in the access hole cover.Visual inspection (VT-3) Is performed according to ASME Section XI, IWB 2500, category B-N-2. GE Services Information Letter (SIL) 462 Sup. 3 recommends ultrasonic inspection techniques.

Implementation of inspection program is plant specific.

Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. Plant programs also may include water chemistry measures such as strict controls on conductivity, hydrogen addition, and use of noble metal additions such as palladium or platinum to reduce electrochemical potential.

BWRVIP-38 for shroud support inspection and flaw evaluation guidelines is under staff review.

documents BWRVIP-03 for reactor oressure vessel internals examination BVWRVIP-52 for shroud stinriort and rennir riE'.qldn t-rfterina RUMN"M I A --rQ!3ne -AO fwý-f-.a~BWRVIP-44 for weld renair of Ni-allovs:

BWRVTP-45 for of irradiated etriri iar~1 PnmnnnPntc' anti R~V.TRtP-A9 for technical basis for inspection relief for internal comoonents with hvdroeen-+(1) Scope Qf Program: The program includes preventive measures to mitigate SCC, inservlce inspection (ISI) to monitor the effects of SCC on the intended function of the components, and repair and/or replacement as needed to maintain the capability to perform the intended function.

(2) Preventive Actions: Maintaining high water purity (many BWRs now operate at <0.15 jiS/cm 2) reduces susceptibility to SCC. Hydrogen effective in reducing electrochemical potentials in the recirculation piping system, but are less effective in the core region. Noble metal additions through a catalytic action appear-t increase the effectiveness of hydrogen additions in the core region, but only lmited data aire avaiabl, at pr-en "do.m..tr-at-thei o-fer- c j+;t-n,_.c;Also, the susceptibility of Ni-alloys to SCC Is evaluated.

(3) Parameters Monitored/

Inspected:

Inspection and flaw evaluation are to be performed in accordance with referenced BWRVIP guideline, as approved by the NRC staff. (4) Detection of Aging Fffects: Degradation due to SCC can not occur without crack initiation and growth. (5) Monitoring and Trending:

Inspection schedule in accordance with applicable, approved BWRVIP guideline is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with applicable, approved BWRVIP guideline.

(7) Corrective Actions: The corrective action proposed by the BWRVIP is under staff review. (8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal, (10) Operating Experience:

Both circumferential (IN 88-03) and radial cracking (IN 92 57) has been observed in the Ni-alloy components.

Yes. BWRVIP Guideline DRAFT- 6/06/00 iLQjconj[Sunr)ortino documents BWRIAP-03 for actor i)ressu vessel internals otxaminatton ouidelines-RWRVIP-52 fa shrmid q"nnnrt:3nr1 vpqqpl hme-ke BUrRVIP-4.1; fnr weleinhilitv of frrn dinted for internal comnonents with hvdroLe WVB1-1II IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM ]31. REACTOR V1F+~'..q WatAV Rpnr~tiwl Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B 1.1.7 Core Shroud. Standby SS 288°C, Crack SCC. Shroud Head Liquid Control High-Purity Inltlatlon and IASCC and Core Plate Line Water Growth BI.1.8 Core Shroud, LPCI Coupling SS 288 0 C. Crack SCC. Shroud Head High-Purity Initiation and IASCC and Core Plate Water Growth DRAFT -6/06/00 IV BI-12 REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Visual inspection (VW-3) is performed according to ASME Section X1, IWB 2500, category B-N-2. Guidance for enhanced VT-I inspections and UT inspections in plant specific programs and BWRVIP-03.

Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. Plant programs also may include water chemistry measures such as strict controls on conductivity.

hydrogen addition, and use of noble metal additions to reduce electrochemical potential.

BWRVIP-27 for standby liquid control system/core plate AP inspection and flaw evaluation guidelines is under staff review. [Sunnorrtfnr donements RHRVlP-0.

fnr reactor pressure vessel internals ecaminatinn fnr S .... ... ..VAA' -r U criteria:

BWRVIP- 14. -59. and -60 for evaluation of crack frowth: BWRVIP-44 for weld renair of Ni-allovsq RVWRVTP-45 for weldabllitv of irradiated structural components:

and BWRVIP-62 for tprhnic-l hadi fnr Qn nplr.ftnn fnrI rnmnnnntc srlth hvrirnaen Visual inspection (W-3) is performed according to ASME Section XI, IWB 2500, category B-N-2. Guidance for enhanced VT-I inspections and. Ur inspections in plant specific programs.

Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. Plant I. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ i (1) Scope of Program: The program includes preventive measures to mitigate SCC, inservice inspection (ISI) to monitor the effects of SCC on the intended function of the components, and repair and/or replacement as needed to maintain the capability to perform the intended function.

(2) Preventive Actions: Maintaining high water purity (many BWRs now operate at <0.15 gS/cm 2) reduces susceptibility to SCC. Hydrogen additions are effective in reducing electrochemical potentials in the recirculation piping system, but are less effective in the core region. Noble metal additions through a catalytic action appear t increase the effectiveness of hydrogen additions in the core region, but only ,mcd data are ava.lable at pren to de.-cnetra.e their (3) Parameters Monitored/Inspected:

Inspection and flaw evaluation are to be performed in accordance with referenced BWRVIP guideline, as approved by the NRC staff. (4) Detection of Aging Effects: Degradation due to SCC can not occur without crack initiation and growth. (5) Monitoring and Trending:

Inspection schedule in accordance with applicable, approved BWRVIP guideline is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with applicable, approved BWRVIP guideline.

(7) Corrective Actions: The corrective action proposed by the BWRVIjP Is under staff review. (8 & 9) Corfirnation Process ahd Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

Cracking has occurred in a number of vessel internal components.

Weld regions are most susceptible, 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.* 4 (1) Scope of Program: The program includes preventive measures to mitigate SCC. inservice inspection (ISI) to monitor the effects of SCC on the intended function of the components, and repair and/or replacement as needed to maintain the capability to perform the intended function.

(2) Preventive Actions: Maintaining high water purity (many BWRs now operate at <0.15 ujS/cm 2) reduces susceptibility to SCC. Hydrogen additions are effective in reducing electrochemical potentials in the recirculation piping system, but are less effective in the core region.Yes, BWRVIP Guideline Yes. BWRVIP Guideline DRAFT- 6/06/00 IV P nm n tn ..........

..... .... RUM nP-.rIA fn Uý umýbý"U" W mruu t ni f -1, hvdr cftnr-lhv l1i-'fi r'-ft-l If11- ,rn l '4,.4a-'IV BI-13 REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism 31.2 Top Guide Top Guide SS 288°C. Crack SCC, High-Purity Initiation and IASCC Water Growth DRAFT- 6/06/00 1V IV Bl-14 JV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM 13. REACTOR VESSEL INTERNALS (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) (continued from previous page) programs also may include water Noble metal additions through a catalytic action to chemistry measures such as strict increase the effectiveness of hydrogen additions in the core controls on conductivity, hydrogen region, but .Jy o -p-mted data .-e available at proent to addition, and use of noble metal demongtrate thir effectivenese.

(3) Parameters additions to reduce electrochemical Monitored/Inspected:

Inspection and flaw evaluation are potential.

BWRVIP-42 for LPCI coupling to be performed in accordance with referenced BWRVIP inspection and flaw evaluation guideline, as approved by the NRC staff. (4) Detection of guidelines is under staff review. Aging Effects: Degradation due to SCC can not occur documents BWRVIP-03 for without crack Initiation and growth. (5) Monitoring and reactor pressure vessel internals Trending:

Inspection schedule in accordance with examination uldellines:

BWRVIP-56 for applicable, approved BWRVIP guideline is adequate for LPCI coupling repair design criteria:

timely detection of cracks. (6) Acceptance Criteria:

Any BWRVIP-14.

-59. and -60 for evaluation degradation is evaluated in accordance with applicable, of crack growth: BWRVIP-44 for weld approved BWRVIP guideline.

(7) Corrective Actions: The repair of Ni-allovso BWRVIP-45 for corrective action proposed by the BWRVIP is under staff weldabilyty of irradiated structural review. (8 & 9) Confirmation Process and components:

and BWRVIP-62 for Administrative Controls:

Site QA procedures, review and technical basis for inspection relief for approval processes, and administrative controls are internal components with hydrogen implemented in accordance with requirements of Appendix Injection.1 B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

Cracking has occurred in a number of vessel internal components.

Weld regions are most susceptible, 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.

Visual inspection (VT-3) is performed (1) Scope of Program: The program includes preventive Yes. according to ASME Section XI. IWB- measures to mitigate SCC. inservice inspection (ISI) to BWRVIP 2500, category B-N-2. Guidance for monitor the effects of SCC on the intended function of the Guideline enhanced VT-l inspections and LIT components, and repair and/or replacement as needed to inspections in plant specific programs.

maintain the capability to perform the intended function.

Coolant water chemistry is monitored (2) Preventive Actions: Maintaining high water purity and maintained in accordance with (many BWRs now operate at <0.15 ;iS/cm 2) reduces EPRI guidelines in TR- 103515 and susceptibility to SCC. Hydrogen additions are effective in BWRVIP-29 to minimize the potential of reducing electrochemical potentials in the recirculatlon crack initiation and growth. Plant piping system, but are less effective in the core region. programs also may include water Noble metal additions through a catalytic action appeast chemistry measures such as strict increase the effectiveness of hydrogen additions in the core controls on conductivity, hydrogen region, but on..- Umite'd dta are aual-!ble at pr-esent t addition, and use of noble metal the effect.lens..a (3) Parameters additions such as palladium or Monitored/Inspected:

Inspection and flaw evaluation are platinum to reduce electrochemical to be performed in accordance with referenced BWRVIP potential.

BWRVIP-26 for top guide guideline, as approved by the NRC staff. (4) Detection of inspection and flaw evaluation Aging Effects: Degradation due to SCC can not occur guidelines is under staff review, without crack initiation and growth. (5) Monitoring and [Supnortino documents BWRVIP-03 for Trending:

Inspection schedule in accordance with reactor pressure vessel internals applicable, approved BWRVIP guideline is adequate for examination guidelines:

BWRVIP-50 for timely detection of cracks. (6) Acceptance Criteria:

Any top guide/core plate repair design degradation is evaluated in accordance with applicable, criteria BWRVIP- 14. -59. and -60 for approved BWRVIP guideline.

(7) Corrective Actions: The Scorrective action proposed by the BWRVIP is under staff DRAFT-6/06/00 JV Bl-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM RI -A-rno vr-. rErrEm#ALS IBoffind Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B1.2 Top Guide Top Guide SS 2880C, Cumulative Fatigue High-Purity Fatigue Water Damage 31.3.1 Feedwater Thermal SS 2880C, Crack SCC, thru Spargers Sleeve, High-Purity Initiation and IASCC B 1.3.3 Distribution Water Growth Header. Discharge Nozzles DRAFT- 6/06/00 IV BI-16 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (BoUing Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) (continued from previous page) evaluation of crack growth: BWRVIP-44 review. (8 & 9) Corftrmation.Process and for weld repair of Ni-alloys:

BWRVIP-45 Administrative Controls:

Site QA procedures, review and for weldability of irradiated structural approval processes, and administrative controls are components:

and BWRVIP-62 for implemented in accordance with requirements of Appendix technical basis for inspection relief for B to 10 CFR Part 50 and will continue to be adequate for internal components with hydrogen the period of license renewal. (10) Operating Experience:

IDjIjtIon.1 The NRC Information Notice (IN) 95-17 discusses cracking in top guides of US and overseas BWRs. Related experience In other components is reviewed in NRC Generic Letter (GL) 94-03 and NUREG-1544.

Cracking has also been observed in the top guide of a Swedish BWR. Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic "1AA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insert#l1 original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a], Subsection NG.Implementation of the program delineated in NUREG-0619 including inserv=ce inspection (ISI) requirements (ultrasonic, visual and dye penetrant inspections) which depend upon specific plant design and other plant actions (monitoring, etc.). An update to NUREc-0619 with qualified UT inspection methods has been approved by the NRC staff. Coolant water chemistry Is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. Plant programs also may include water chemistry measures such as strict controls on conductivity, hydrogen addition, and use of noble metal additions such as palladium or platinum to reduce electrochemical potential.

4. _________(1) Scope of Program: The program includes preventive measures to mitigate SCC. inservice inspection (ISI) to monitor the effects of SCC on the intended function of the components.

and repair and/or replacement as needed to maintain the capability to perform the intended function.

(2) Preventive Actions: Maintaining high water purity (many BWRs now operate at <0.15 gS/cm 2) reduces susceptibility to SCC. Hydrogen additions are effective in reducing electrochemical potentials in thftecirculation piping system. but are less effective in the core region. Design features aimed at mitigating thermal fatigue cracking, which has been the primary source of degradation for these components, have been implemented as per NUREG-0619.

(3) Parameters Monitoredllnspected:

The AMP monitors the effects of SCC on the intended function by detection and sizing of cracks by inservice inspection (ISI). (4) Detection qf Aging Effects: Degradation due to SCC can not occur without crack initiation and growth. An update to NUREG-0619 with qualified UT1' inspection methods has been approved by the NRC staff. (5) Monitoring and Trending:

Inspection schedule in accordance with NUREG-0619 is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with IWB-3520.

(7) Corrective Actions: Repair and replacement are in conformance with IWB 3140. (8 & 9) Conrfrmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

NUREG-0619 summarizes work performed by the NRC to resolve Generic Technical Activity A-10, "BWR Nozzle Cracking" and the industry No DRAFT -6/06/00 WV BI-17 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM El. REACTOR VESSEL INTERNALS (EoiIln% Wnte~v Rpa,.tnr)Structure and Region of Environ- Aging Aging Item Component Interest Material merit Effect Mechanism B 1.3.1 Feedwater Thermal SS 288 0 C, Cumulative Fatigue thru Spargers Sleeve, igh-Purity Fatigue B 1.3.3 Distribution Water Damage Header, Discharge Nozzles B 1.4.1 Core Spray Core Spray SS 288-C, Crack SCC, thru Lines and Lines Pigh-Purity Initiation and IASCC B 1.4.4 Spargers (Headers), ater Growth Spray Rings, Spray Nozzles, Thermal Sleeves DRAFT- 6/06/00 IV Bl-18 TV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM 13. REACTOR VESSEL INTERNALS (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (contonued from previous page) experience with cracking in the feedwater sparger system. The industry testing and analysis program is described in GE NEDE-21821-A.

The primary source of degradation in this system has been thermal fatigue. However, the inspections intended to address thermal fatigue issues are also effective in ensuring that degradation by SCC is also effectively managed.

Components have been designed or Fatigue is a time-limited aging analysis (nFAA) to be Yes evaluated for fatigue for a 4 0 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insrt #1 original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG. Implementation of the program (1) Scope of Program: The program includes preventive NQ delineated in the NRC Inspection and measures to mitigate SCC, inservice inspection (ISI) to Enforcement Bulletin (IEB) 80-13 monitor the effects of SCC on the intended function of the including enhanced visual inspection components.

and repair and/or replacement as needed to techniques to supplement or replace maintain the capability to perform the intended function.

visual inspection (VT-3) requirement of (2) Preventive Actions: Maintaining high water purity GE Services Information Letter (SIL) (many BWRs now operate at <0.15 pS/cm 2) reduces 289. BWRVIP- 18 for core spray susceptibility to SCC. Hydrogen additions are effective in internals inspection and flaw evaluation reducing electrochemical potentials in the recirculation guidelines has been approved by the piping system, but are less effective in the core region. staff. Coolant water chemistry Is (3) Parameters Monitored/Inspected:

Inspection and monitored and maintained in flaw evaluation are to be performed in accordance with accordance with EPRI guidelines in TR- referenced BWRVIP guideline, as approved by the NRC 103515 and BWRVIP-29 to minimize the staff. (4) Detection of Aging Fffects: Deradation due to potential of crack initiation and growth. SCC can not occur without crack initiation and growth. Plant programs also may include water (5) Monitoring and Trending:

Inspection schedule in chemistry measures such as strict accordance with applicable, approved BWRVIP guideline is controls on conductivity, hydrogen adequate for timely detection of cracks. (6) Acceptance addition, and use of noble metal Criteria:

In the event cracks are identified, an evaluation additions such as palladium or is performed in accordance with applicable, approved platinum to reduce electrochemical BWRVIP guideline.

(7) Corrective Actions: Coective potential.

actions in accordance with applicable.

approved BWRVIP ISupporting documents BWRVIP-03 for 16 and BWRVIP-19 guidelines are adequate.

(8 & 9) reactor pressure vessel internals Confirmation Process and Administrative Controls:

examination guidelines:

BWRVIP- 16 and Site QA procedures, review and approval processes, and -19 for internal core spray n- ining and administrative controls are implemented in accordance sparger replacemnet and repair design with requirements of Appendix B to 10 CFR Part 50 and criteria:

BWRVIP-14.

-59. and -60 for will continue to be adequate for the period of license evaluation of crack growth: BWRVIP-44 renewal. (10) Operating Experience:

IEB 80-13 reviews for weld repair of NI-alloys:

BWRVIP-45 instances of cracking in core spray spargers.

for weldabilltv of irradiated structural components:

and BWRVIP-62 for technical basis for inspection relief for internal components with hydrogen inAe______

DRAFT-6/06/00 IV Bl-19 NRC FORM 658 U.S. NUCLEAR REGULATORY COMMISSION (9-1999) TRANSMITTAL OF MEETING HANDOUT MATERIALS FOR IMMEDIATE PLACEMENT IN THE PUBLIC DOMAIN This form is to be filled out (typed or hand-printed) by the person who announced the meeting (i.e., the person who issued the meeting notice). The completed form, and the attached copy of meeting handout materials, will be sent to the Document Control Desk on the same day of the meeting; under no circumstances will this be done later than the working day after the meeting.

Do not include proprietary materials.

DATE OF MEETING The attached document(s), which was/were handed out in this meeting, is/are to be placed 06/06/2000 in the public domain as soon as possible.

The minutes of the meeting will be issued in the near future. Following are administrative details regarding this meeting: Docket Number(s)

PROJ690 Plant/Facility Name License Renewal TAC Number(s) (if available)

Reference Meeting Notice 5/25/2000 Purpose of Meeting (copy from meeting notice) To discuss NEI comments on the draft "Generic Aging Lessons Learned" (GALL) report -Mechanical systems.

NAME OF PERSON WHO ISSUED MEETING NOTICE TITLE Jerry Dozier General Engineer OFFICE NRR DIVISION DRIP BRANCH RLSB Distribution of this form and attachments:

Docket File/Central File PUBLIC NRC FORM 658 (9-1999)PRINTED ON RECYCLED PAPER This ionn was designed using InForms REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM H1. IWTFRNAXS

[rBoillnv Water Reactori Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B1.4.1 Core Spray Core Spray SS 288 0 C, Cumulative Fatigue thru Lines and Lines High-Purity Fatigue B 1.4.4 Spargers (Headers), Water Damage Spray Rings, Spray Nozzles, Thermal Sleeves Jet Pump Assemblies Thermal Sleeve, Inlet Header, Riser Brace Arm, Holddown Beams, Inlet Elbow, Mixing Assembly, Diffuser, Castings Holddown Beams: NI Alloy (X-750), Castings:

Cast Austenitic Stainless Steel (CASS), Others: SS 288°C, High-Purity Water Crack Initiation and Growth 1 1 4 4. 4. 1 IASCC 1 1 1. 1. ______ I ______DRAFT- 6/06/00 IV 13.5.1, thru BI.5.8 IV Bl1-20 IV REACTOR VESSEL, IDTERNALS, AND REACTOR COOLANTIT SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Components have been designed or Fatigue is a time-limited aging analysis 'IAA) to be Yes evaluated for fatigue for a 4 0 y design performed for the period of license renewal, and Generic T1AA life, according to the requirements of the Safety Issue (GSD1-190 is to be addressed.

Insert#l.

original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG. Implementation of inspection and (1) Scope qf Program: The program includes preventive Yes, surveillance programs delineated in measures to mitigate SCC, performance assessment and BWRVIP NRC Inspection and Enforcement periodic inservice Inspection (ISI) to monitor the effects of Guideline Bulletin (IEB) 80-07 and GE Services SCC on the intended function of the components, and Information Letter (SIL) 330 to ensure repair and/or replacement as needed to maintain the overall functionality and integrity ofjet capability to perform the intended function.

(2) Preventive pump assemblies, and additional Actions: Maintaining high water purity (many BWRs now recommendations of GE Services operate at <0.15 VS/cm 2) reduces susceptibility to SCC. Information Letter (SIL) 605 Rev. 1 for Hydrogen additions are effective in reducing Jet pump riser pipe. Coolant water electrochemical potentials in the recirculation piping chemistry is monitored and maintained system, but are less effective in the core region. in accordance with EPRI guidelines in (3) Parameters Monitored/Inspected:

Inspection and flaw TR-103515 and BWRVIP-29 to minimize evaluation are to be performed in accordance with the potential of crack initiation and referenced BWRVIP guideline, as approved by the NRC growth. Plant programs also may staff. (4) Detection of Aging Fffects: Degradation due to include water chemistry measures such SCC can not occur without crack initiation and growth, or as strict controls on conductivity, and degradation of Jet pump operation.

(5) Monitoring and hydrogen addition.

BWRVIP-4 1 foLje Trending:

Inspection schedule in accordance with pumo assembly inspection and flaw applicable, approved BWRVIP guideline is adequate for evaluation ruuldelines and BWRVIP-28 timely detection of cracks. (6) Acceptance Criteria:

Any for assessment of let pump riser elbow degradation in Jet pump operation is evaluated in to thermal sleeve weld crackino are accordance with applicable, approved BWRVIP guideline.

under staff review, (7) Corrective Actions: The corrective action proposed by ISuppDorting documents BWRVIP-03 for the BWRVIP is under staff review. (8 & ,kCoqntrmation reactor pressure vessel internals Process and Administrative Controls:

Site QA examination ouidelines' BWRVIP-51 for procedures, review and approval processes, and let pump reoalr design criteria:

administrative controls are implemented in accordance BWRVIP- 14. -59. and -60 for evaluation with requirements of Appendix B to 10 CFR Part 50 and of crack growth: BWRVIP-44 for weld will continue to be adequate for the period of license repair of Ni-alloys:

BWRVTP-45 for renewal. (10) Operating Experience:

Instances of weldability of irradiated structural cracking have occurred in Jet pump assembly (NRC IEB components:

and BWRVIP-62 for 80-07), hold-down beam INRC Information Notice (IN) 93 technical basis for inspection relief for 101), and Jet pump riser pipe elbows (NRC IN 97-02). Internal components with hydrogen 1ciniec_ _ ____DRAFT -6/06/00 IV BI1-21 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B1.5.1 Jet Pump Thermal Holddown 288°C, Cumulative Fatigue thru Assemblies Sleeve, Inlet Beams: High-Purity Fatigue B 1.5.8 Header, Riser Ni Alloy Water Damage Brace Arm, (X-750), Holddown Others: SS Beams, Inlet Elbow. Mixing Assembly, Diffuser, Castings B13.5.4 Jet Pump Assemblies Castings CASS 288 0 C, High-Purity Water Loss of !Fracture Toughness Thermal Aging and Neutron Irradiation Embrittle ment L a n -DRAFT -6/06/00 IV B 1-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM BI. REACTOR VESSEL INTERNALS (Boiling Water Reactor)Existing Further Aging Management Program CAMP) Evaluation and Technical Basis Evaluation Components have been designed or Fatigue is a tlme-llmlted aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, .and Generic TLAA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insert #. original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG.The reactor vessel internals receive a IVT-31 arcordfne toof Sihqer'tlon IXR ARMF qprftln YT T'h1c Innner-tinn ic nnt fn j # + 4,, t, * ~t th f.Rpp1 .f InaQ mf fracture tourhness due to thermal and neutron Irradiation embrittlement.

An accentable alternative AMP consists ofthefolowing:

Determination of the susceptibility of CASS components to thermal aging embrittlement based on casting method, Mo content, and percent ferrite. For "potentially susceptible" components, based on the neutron fluence of the component, implement either a supplemental examination of the affected components as part of the applicant's 10-year inservice inspection (ISI) program during the license renewal term or a component-specific evaluation to determine the susceptibility to loss of fracture toughness.

___________________________________________________________________________________________________

I For the accetntable alternative AMP: (1) Scope of Program: The program includes determination of the susceptibility of CASS components to thermal aging based on casting method, Mo content, and percent ferrite, and for -potentially susceptible' components.

to account for the synergistic loss of fracture toughness due to neutron embrittlement and thermal aging embrittlement, implement either a supplemental examination of the affected components as part of a 10 year ISI program during the license renewal term or a component-specific evaluation to determine the susceptibility to loss of fracture toughness.

(2) Preventive Actions: The program provides no guidance on methods to mitigate thermal aging or neutron embrittlement.

(3) Parameters Monitored/

Inspected:

The program specifics depend on the neutron fluence and ferrite content of the component.

Based on the criteria in NUREG- 1705. the to thermal aning embrittlement of CASS iping Is determ-ined in terms of casting method, Mo content, and ferrite content. For low-Mci _ontent (0.5 wt.% max.) steels, only static-cast steels with >20% ferrite are potentially susceptible to thermal embrittlement, static-cast steels with <20% ferrite and all centrifugal-cast steels are not susceptible.

For high-Mo content (2.0 to 3.0 wt.%) steels, static-cast steels with > 14% ferrite and centrifugal-cast steels with >20% ferrite are potentially susceptible to thermal embrittlement, static-cast steels with _< 14% ferrite and centrifugal-cast steels with 20% ferrite are not susceptible.

Ferrite content will be calculated by Hulls equivalent factors or a method producing an equivalent level of accuracy (+/-6% deviation between measured and calculated values).

Insert #3. (4) Detection of Aging Effects: For all CASS components that have a neutron fluence of greater than 1017 n/cm 2 (E>I MeV), implement a 10-year ISI program during renewal period including supplemental inspection covering portions of the susceptible components determined to be limiting from the standpoint of thermal aging susceptibility (Ferrite and Mo contents, casting process. and operating temperature), neutron fluence, and cracking susceptibility (applied stress, operating temperature, and environmental conditions).

The inspection technique, including the reliability in detecting the features of interest (crack appearance and size) in assuring the integrity of the component, should be specified.

For example, enhancement of the visual VT-I examination to achieve a 1/2-mil (0.0005 in.) resolution, with the conditions Yes. the 2 suitable AMP should I& DRAFT-6/06/00 For the accentable alternative AMP: xrlýmql Inanee-tinn fVT-.ql ar(-nrr1ino tn rnfPcrnrv R-M-_q nf.1;"ha#-ntinri TY11 A ..............................

W" ý ý" ý ýfracture tnuohness due to thermal aging and neutron irradiation embritflement.

An arrentnble alternative AMP consists IV Bl-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM BD. REACTOR VESSEL IMTERNALS (Boiling Water Reactor) Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B 1.6.1 Fuel Supports OrifIced Fuel CASS 288 0 C, Crack Thermal & CRD Support High-Purity Initiation and Aging and Assemblies Water Growth Neutron Irradiation Embrittle ment B 1.6.1 Fuel Supports Oriflced Fuel SS, CASS 288 0 C. Cumulative Fatigue & CRD Support High-Purity Fatigue Assemblies Water Damage DRAFT -6/06/00 IV B 1-24 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT.SYSTEM B1. REACTOR VESSEL INTERNALS (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical "Basis Evaluation (continued from previous page) (lighting and surface cleanness) for the ISI bounded by those used to demonstrate the resolution of the inspection technique.

Alternatively, perform a component-specific evaluation including a mechanical loading assessment to determine the maximum tensile loading on the component during ASME Code Level A, B, C, and D conditions.

If the loading is compressive or low enough (<5 ksi) to preclude fracture, then supplemental inspection of the component is not required.

Failure to meet this criteria requires continued use of supplemental inspection program. For all CASS components that have a neutron fluence of less than I017 n/cm 2 (E> I MeV), implement the supplement inspection program if they are susceptible" to thermal aging: the existing ASME Section XI inspection requirements are considered adequate if the components are "not susceptible" to thermal aging. (5) Monitoring and Trending:

Inspection schedule in accordance with IWB-2400 should provide timely detection of cracks. (6) Acceptance Criteria:

Flaws detected in CASS components are evaluated in accordance with the applicable procedures of IWB-3600.

(7) Corrective Actions: Repair is in conformance with IWA-4000 and IWB-4000, and replacement according to IWA-7000 and IWB-7000.

(8 & 9) Conflrmation Process and Administrative Controls:

Site QA procedures, review and approval processes.

and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The AMP based on susceptibility determination, neutron fluence level, and supplemental inspectiovis effective in managing the effects of synergistic loss of fracture toughness due to neutron and thermal aging embrittlement on the intended function of CASS components.

Same as for the effect of Thermal Aging Same as for the effect of Thermal Aging and Neutron Y= and Neutron Irradiation Embrittlement on Irradiation Embrittlement on Item B1.5.8 jet pump castings.

ithe Item B1.5.8jet pump castings.

i AMP should eMa~uated Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic T1AA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insert #1. original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG.DRAFT- 6/06/00 IV BI-25 IV REACTOR VESSEL. WTRNALS, AND REACTOR COOLANT SYSTEM B1. REACTOR VESSEL INTERNALS (Bolng Water Reactor) Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism B 1.7.1 Instrument Intermediate SS 288 0 C, Crack SCC, thru Housings Range Monitor High-Purity Initiation and IASCC B 1.7.3 (IRM) Dry Water Growth Tubes, Low Power Range Monitor (LPRM) Dry Tubes, Source Range Monitor (SRM) Dry Tubes B11.7.I Instrument IRM Dry SS 288 0 C, Cumulative Fatigue thru Housings Tubes. High-Purity Fatigue B 1.7.3 LPRM Dry Water Damage Tubes, SRM Dry Tubes 13.5, Jet Pum Je 2Pump. Cac UnanUci Asemblie Sensingi~ne ihPr ntaina ~c Water Growthi Loadin DRAFT -6/06/00 IV Bl1-26 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM ]E1. VESSE~L INTERNALS (Boil/ng Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Implementation of aging management (1) Scope of Program. The program includes preventive ND program recommended in GE Services measures to mitigate SCC and periodic inservice Information Letter (SIL) 409 Rev. 1. inspection (ISI) to monitor the effects of SCC on the BWRVIP-49 for instrument penetration intended function of the components.

and repair and/or inspection and flaw evaluation replacement as needed to maintain the capability to guidelines has been approved by the perform the intended function.

(2) Preventive Actions: staff. Coolant water chemistry is Based on GE SIL 409 Rev. I replacement of existing tubes monitored and maintained in with those fabricated from more IASCC-resistant materials accordance with EPRI guidelines in TR- and crevice free design. Maintaining high water purity 103515 and BWRVIP-29 to minimize the (many BWRs now operate at <0.15 pS/cm 2) reduces potential of crack initiation and growth. susceptibility to SCC. Hydrogen additions are effective in Plant programs also may include water reducing electrochemical potentials in the recirculation chemistry measures such as strict piping system, but are less effective in the core region. controls on conductivity, and hydrogen (3) Parameters Monitored/Inspected:

Inspection and flaw addition.

evaluation are to be performed in accordance with [Supporting documents BWRVIP-03 for referenced BWRVIP guideline, as approved by the NRC reactor pressure vessel internals staff. (4) Detection qf Aging Effects: Degradation due to examination guidelines:

BWRVIP-57 for SCC can not occur without crack initiation and growth. instrument penetration repair design (5) Monitoring and Trending:

Inspection schedule in criteria BWRVIP- 14. -59. and -60 for accordance with applicable, approved BWRVIP guideline is evaluation of crack growth: BWRVIP-44 adequate for timely detection of cracks. (6) Acceptance for weld repair of Ni-alloys:

BWRVIP-45 Criteria:

Crack indications are evaluated in accordance for weldability of irradiated structural with applicable, approved BWRVIP guideline.

components:

and BWRVIP-62 for (7) Corrective Actions: Corrective actions in accordance technical basis for inspection relief for with applicable.

approved BWRVIP-57 guidelines are internal components with hydrogen a (8 & 9) Confirmation Process and kUcionA Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (1 0) Operaog Experience:

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 irradiation assisted SCC (IASCC) may also play a role in the cracking.

Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of the Safety Issue (GSI)-190 is to be addressed.

Insert #I. original licensing criteria or ASME Section III (edition specified in 10 CFR 50.55a), Subsection NG. Plant specific aging management Plant specific aging management program is to beYs program should be implemented, evaluated, no generi DRAFT-6/06/00 WV Bl1-27 C1. Reactor Coolant Pressure Boundary (Boiling Water.Reactor)

C 1. 1 Piping & Fittings C1.1.1 Main Steam C 1.1.2 Feedwater C 1. 1.3 High Pressure Coolant Injection (HPCI) System C 1.1.4 Reactor Core Isolation Cooling (RCIC) System C1.1.5 Recirculation C1.1.6 Residual Heat Removal (RHR) System C1.1.7 Low Pressure Coolant Injection (LPCI) System C1.1.8 Low Pressure Core Spray (LPCS) System C1. 1.9 High Pressure Core Spray (HPCS) System C1.1.10 Isolation Condenser C1.1.11 Lines to Reactor Water Cleanup (RWC) and Standby Liquid Control (SLC) Systems C1.1.12 Steam Line to HPCI and RCIC Pump Turbine C1.1.13 Small Bore Piping C1.1.14 Jet Pump Sensing Line C 1.2 Recirculation Pump C 1.2.1 Bowl / Casing C1.2.2 Cover C 1.2.3 Seal Flange C1.2.4 Closure Bolting C 1.3 Safety & Relief Valves C1.3.1 Valve Body C 1.3.2 Bonnet DRAFT -6/06/00 TV C1-1 C1.3.3 Seal Flange C1.3.4 Closure Bolting CI .4 Isolation Condenser C1.4.1 Tubing C 1.4.2 Tubesheet C1.4.3 Channel Head C1.4.4 Shell C1.5 Control Rod Drive (CRD) Hydraulic System C1.5.1 Piping and Fittings C1.5.2 Valve Body C 1.5.3 Pump Casing C1.5.4 Filter C 1.5.5 Accumulator C1.5.6 Scram Discharge C 1.5.7 CRD Return Line DRAFT -6/06/00 IV CI1-2 C1. Reactor Coolant Pressure Boundary (Boiling Water Reactor)System, Structures, and Components The system, structures, and components included in this table comprise the boiling water reactor (BWR) primary coolant pressure boundary and consist of the reactor coolant recirculation system and portions of other systems connected to the pressure vessel extending to the first isolation valve outside of containment or to the first anchor point. The connected systems include residual heat removal (RHR), low-pressure core spray (LPCS), high-pressure core spray (HPCS). low-pressure coolant injection (LPCI), high-pressure coolant injection (HPCI). reactor core isolation cooling (RCIC), isolation condenser (IC), reactor water cleanup (RWC), feedwater (FW), and main steam (MS) systems, and steam line to HPCI and RCIC pump turbine.

All systems. structures, and components in the reactor coolant pressure boundary are classified as Group A Quality Standards.

The aging management program for containment isolation valves is reviewed in Table V C. The pump and valve internals are considered to be active components.

They perform their intended functions with moving parts or with a change in configuration and are not subject to aging management review pursuant to 10 CFR 54.21 (a) (1) (i). System Interfaces The systems that interface with the reactor coolant pressure boundary include the reactor pressure vessel (Table IV Al), containment isolation components (Table V C), emergency core cooling system (Table V D2), main steam system (Table VIII B2). and feedwater system (Table VIII D2).DRAFT -6/06/00 TV CI1-3 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor)Structure and jRegion of jEnviron-I Aging IAging Item Component I Interest Material j ment -Effect ~Mechanism C 1. 1. 1, Piping & Main Steam, Carbon Stee 288°C Wall Erosion/ Cl. Fittings Steam Line to (CS) Steam Thinning Corrosion 1.12 HPCI and SAI06-Gr B (E/C) RCIC Pump SA333-Gr 6, Turbine SA155-Gr KCF70£ ______________

I ______________

I ____________

I A ___________

__________________

DRAFT -6/06/00 IV C 1--4 I REACTOR VESSEL, INTERNALS, AND REACTOR COOlANT SYSTEM r1-- REACTOR ClOLABT PRESSURE BOUNDARY (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Program delineated in NUREG-1344 and implemented through NRC Generic Letter 89-08: CHECWORKS Code; EPRI guidelines of NSAC-202L-R2 for effective erosion/corrosion program: and water chemistry program based on EPRI guidelines in TR- 103515 and BWRVIP-29 for water chemistry in BWRs. ISunoortini documents BWRVIP-75 for rtechnical basis for revisions to GL 88 01 inspection schedules.l Yes, Element I should be further evaluated DRAFT- 6/06/00 IV WV C 1-5 (1) Scope of Program: The AMP Includes NUMARC program delineated in Appendix A of NUREG- 1344 and implemented through NRC Generic Letter (GL) 89-08: CHECWORKS computer Code: and EPRI guidelines of NSAC-202L-R2.

The program includes the following recommendations: (a) conduct appropriate analysis and limited baseline inspection, (b) determine the extent of thinning and repair/replace components, and (c) perform follow-up inspections to confirm or quantify and take longer term corrective actions. Technical aspects of the CHECWORKS Code, including the parameters and inputs. are acceptable.

However, the EPRI guidance document NSAC-202L-R2 (April 1999) is too general to ensure applicant's flow-accelerated corrosion program will be effective in managing aging in safety-related systems.

(2) Preventive Actions: The rate of E/C is affected by piping material, geometry and hydrodynamic conditions, and operating conditions such as temperature, pH, and dissolved oxygen content. Mitigation is by selecting material considered resistant to E/C, adjusting water chemistry and operating conditions, and improving hydrodynamic conditions through design modifications.

(3) Parameters Monitored/

Inspected:

The AMP monitors the effects of E/C on the intended function of piping by measuring wall thickness by nondestructive examination and performing analytical evaluations.

The inspection program delineated in NUREG-1344 requires ultrasonic or radiographic testing of 10 most susceptible locations and 5 additional locations based on unique operating conditions or special considerations.

For each location outside the acceptance guidelines, the inspection sample is expanded based on engineering judgment.

AnalytidM models such as those incorporated into the CHECWORKS code are used to predict E/C in piping systems based on specific plant data including material and hydrodynamic and operating conditions.

The inspection data are used to calibrate and benchmark the models and code. (4) Detection of Aging Fffects: Aging degradation of piping and fittings occurs by wall thinning:

extent and schedule of inspection assure detection of wall thinning before the loss of intended function of the piping. (5) Monitoring and Trending:

Inspection schedule of NUREG-1344 and EPRI guidelines should provide for timely detection of leakage. Inspections and analytical evaluations are performed during plant outage. If analysis shows unacceptable conditions, inspection of initial sample is performed within 6 months. (6) Acceptance Criteria:

Inspection results are used to calculate number of refueling or operating cycles remaining before the component reaches Code minimum allowable wall thickness.

If calculations indicate that an area will reach Code minimum (plus 10% margin), the component must be repaired or replaced.

However, NRC staff has identified the problems in implementing E/C program that pertain to weakness or errors in (a) using predictive models, (b) calculating minimum wall thickness acceptance criteria, (c) analyzing the results of UT examinations, and (d) assessment of E/C program activities (NRC IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mlechanism C 1. 1. 1 Piping & Main Steam CS 288-C Cumulative Fatigue Fittings SA 106-Gr B Steam Fatigue SA333-Gr 6, Damage SAI55-Gr KCF70 y C1.1.2 Piping & Feedwater CS Up to 225°C Wall Erosion/ Fittings SAIO6-Gr B Oxygenated Thinning Corrosion SA333-Gr 6, Water SA155-Gr KCF70 C1.1.2 Piping & Feedwater CS Up to 2250C Cumulative Fatigue Fittings SA106-Gr B, Oxygenated Fatigue SA333-Gr 6, Water Damage SA 155-Gr KCF7O CI.1.3, Piping & High Pressure CS 2880C Cumulative Fatigue CI.1.4 Fittings Coolant SAI 06-Gr B Oxygenated Fatigue Injection SA333-Gr 6, Water or Damage (HPCI). SA155-Gr Steam Reactor Core KCF70 Isolation Cooling (RCIC)DRAFT -6/06/00 IV bC1-6 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) Information Notice IN 93-21). (7) Corrective Actions: Prior to service, repair or replace to meet the requirements of NUREG-1344.

Follow-up inspections are performed to confirm or quantify thinning and take longer term corrective actions such as adjustment of chemistry and operating parameters, or selection of materials resistant to E/C. However. NRC staff has identified weakness or errors in (a) dispositioning components after reviewing the results of the inspection analysis, and (b) repairing or replacing components that failed to meet the acceptance criteria (IN 93-21). (8 & 9) Confirrmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

Wall-thinning problems in single-phase systems have occurred in feedwater and condensate systems (NRC Bulletin No. 87-01, INs 81-28, 92-35, 95-11). and in two phase piping in extraction steam lines (INs 89-53, 97-84) and moisture separation reheater and feedwater heater drains (INs 89-53, 91-18, 93-21, 97-84). The AMP outlined in NUREG- 1344 and EPRI report and implemented through GL 89-08 has provided effective means of ensuring the structural integrity of all high energy carbon steel systems.

Components have been designed or Fatigue is a time-limited aging analysis 'rlIAA) to be Yes evaluated for fatigue for a 4 0 y design performed for the period of license renewal, and Generic TlAA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

Insert#I.

ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB. or ANSI B3 1. 1. or other evaluations based on cumulative usage factor (CUF). Same as for the effect of Same asfor the effect of Erosion/Corrosion on Item C1. 1. 1 Erosion/Corrosion on Item C1. 1. 1 Main Main Steam Line Piping and Fittings.

Element1 Steam Line Piping and Fittings.

should be further Components have been designed or Fatigue is a time-limited aging analysis CTIAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # .i ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB. or ANSI B3 1.1. or other evaluations based on cumulative usage factor (CUF). Components have been designed or Fatigue is a time-limited aging analysis rTLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic T1AA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

Insert#1.

ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB. or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF).DRAFT- 6/06/00 IV C1-7 IV Structure and Item IComponent Ci. I.5 thru CI. 1.11 Pipnpg & Fittings Material I Envion Effect Mechanism

.Stainless Steel (SS) (e.g., Types 304, 316, or 316NG); Cast Austenitic Stainless Steel (CASS): Nickel Alloy. (e.g., Alloys 600, 182, and 82)288 0 C 3xygenated W'ater or Rteam Crack Initiation and Growth Stress Corrosion Cracking (SCC), Inter granular Stress Corrosion Cracking (IGSCC)CI.1.5. Piping& RHR. CASS 288°C Loss of Thermal C1. Fittings LPCI. Oxygenated Fracture Aging 1.11 LPCS. Water or Toughness Embrittle HPCS. Steam ment Lines to IC. Lines to RWC & SLC Systems DRAFT -6/06/00 REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLAIN PRWSABTT4

/u-1V1- waT- 'R TV CI-8 Region of Interest Recirculation, Residual Heat Removal (RHRM, Low Pressure Coolant Injection (LPCO). Low Pressure Core Spray (LPCS), High Pressure Core Spray (HPCS), Isolation Condenser tIC), Lines to Reactor Water Cleanup (RWC) and Standby Liquid Control (SLC) Systems I----u ............

Environ- Aging Aging Material ment Effect Mechanism IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (Bolling Water Reactor)Eximting PtFu)cher Aging Management Program (AMP) Evaluation and Techniclc Basis Evaluation, Program delineated in NUREG-0313, Rev. 2 and NRC Generic letter (GL) 88 01 and its Supplement 1, and inservice inspection in conformance with ASME Section )I (edition specified in 10 CFR 50.55a), Subsection IWB. Table IVB 2500-1, examination categories B-J for pressure retaining welds in piping and B-F for pressure retaining dissimilar metal welds, and testing category B-P for system leakage. B IP-7 Stechnical basis for revsions to GL 88-0]insoection schedule.

BWRVIP-27 for standby Ilnuid control/core_

niate AP inso~ection and flaw evaluattion atiidelines and SWRVIP-A9.

for LPCI cotinline insnection and flaw .ual,,atlon an staff revl,'w Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- I03515 and BWRVIP-29 to minimize the potential of crack initiation and growth. i[trnr~rtlne docluments for reactor nressiire vessel internals eeamlnat9on duldelIneQ' -PQ noel Sfl for eval..atlnn of rrnrfr R

lIo,,19 control line renaer desian criteria: BWRVIP-61 for BWR vessel and hnternals induction heating stress imnnrovern nt effectiveness on crack arnurth in nnPrntfne1 nlntqs qnd RUTRXflPA9 few technical for Incn.e.tfon rplI,.f for Internal rrnnonntc with hydrogen lnlectionl The reactor coolant system comno:xnents are insnected In accordance with ASqME Secption Mi Subsc-tion n1WR Thig insnection Is not to detect the effects of loss of tonnhness d 1 1.*0 thrnal aeioa h,-4t+l..,leo*

An acentable alternative AMP consists Determination of the susceptibility of CASS piping to thermal aging embrittlement based on casting method, Mo content, and percent ferrite. For "potentially susceptible" piping, aging (1) Scope qf Program: The program focuses on managing and Implementing countermeasures to mitigate IGSCC and inservice Inspection PSI) to monitor IGSCC and its effects on the intended function of austenitic stainless steel (SS) piping 4 in. or larger in diameter, and reactor vessel attachments and appurtenances.

NUREG-0313 and GL 88-01, respectively, describe the technical basis and staff guidance regarding mitigating IGSCC in BWRs. (2) Preventive Actions: Mitigation of IGSCC is by selection of material considered resistant to sensitization and IGSCC. e.g.. low-carbon grades of austenitic SSs and weld metal, with a carbon of 0.035% and minimum 7.5% ferrite in weld metal, and by special processing such as solution heat treatment, heat sink welding, and Induction heating or mechanical stress improvement (SI). Coolant water chemistry Is monitored and maintained according to EPRI guidelines in TR- 103515 and BWRVIP 29 to minimize the potential of crack initiation and growth. Also, hydrogen water chemistry and stringent control of conductivity is used to inhibit IGSCC. (3) Parameters Monitored/nspected:

Inspection and flaw evaluation are to be performed in accordance with referenced BWRVIP guideline, as approved by the NRC staff. (4) Detection of Aging Fffects: Aging degradation of the piping can not occur without crack initiation and growth; extent, method. and schedule of inspection as delineated in GL 88-01 and updated in BWRVIP-75 is adequate and will assure timely detection of cracks before the loss of intended function of austenitic SS piping and fittings.

Inser #5. (5) Monitoring and Trending:

Inspection schedule in accordance with GL88-01 oapplicable approved BWRVIP guideline.

(6) Acceptance Criteria Any'IGSCC degradation is evaluated according to applicable approved BWRVIP guideline.

(7) Corrective Actions: Insert #. (8 & 9) Conf1rmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

IGSCC has occurred in small- and large-d'ameter BWR piping made of austenitic SSs and Nickel-base alloyg. Significant cracking has occurred in recirculation.

core spyra. and RHR systems and reactor water cleanup system piping weds.9- ---- -I For the acceotable alternative AMP: (1) Scope qf Program: The program includes determination of the susceptibility of CASS components to thermal aging based on casting method, Mo content, and percent ferrite, and for potentially susceptible components aging management is accomplished either through volumetric examination or plant/component-specific flaw tolerance evaluation.

(2) Preventive Actions: The program provides no guidance on methods to mitigate thermal aging. (3) Parameters Monitored/

Inspected:

Based on the criteria In NUREG-1705.

the susceptibility to thermal aging embritfiement of CASS piping is determined in terms of casting method, Mo content, and ferrite content.Yes. BWRVIP Guideline Yes. the a sutable be evaluated DRAFT -6/06/00 IVCI1-9 technical basis for revisions to GL 88-01 inspcction schedule.

BWRVIP-27 fo standby lJouid control/core niate AP inst>ection and flaw evaluation guidelines.and BWRVIP-42 for LPCI coupling inscction and flaw r-yaluatio guidelines are under stnfrrýdAw

[Suppgrtina documents BMWVIP-03 fo reactorressure vessel internals examination Lruidelhnes:

RUM1nP_ I A-59, --d -60 for evaluation gf gryAh& LA EIWVIP5 fs conro lin einc iteri intemals induction heating stres ir t r n ... ... ... ... ... ... ... on .. .....BWRVIP-69 for technic nI basis L-insp p ction relief for internal com rient The reactor coolant system comr)onents For~~ ... .... ace tbe le are inspected in accordance with ASM ,.. ........l ..jn n -~~ ...... .... ..... .... ..... ra ...... ..to thermal aaina embittle-*

An accentahle alternative AMP consists IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor)IStructure andfI Region of fEnviron-j Aging IMAgingsm Item IComponent Interest I MaterWIa ment IEffect jechanim* I I I -DRAFT- 6/06/00 TV CI-10 REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM r111 C"OOLANT PRESSURE BOUNDARY (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous page) management is accomplished either through enhanced volumetric examination or plant/component specific flaw tolerance evaluation.

Additional inspection or evaluations are not required for "not susceptible" piping to demonstrate that the material has adequate fracture toughness.

For pump casings and valve bodies, screening for susceptibility to thermal aging is not required.

Also, the existing ASME Section XI inspection requirements, including the alternative requirements of ASME Code Case N-481 for pump casings, are considered adequate for all pump casings and valve bodies.DRAFT -6/06/00 IV IV CI-II (continued from previous page) For low-Mo content (0.5 wt.% max) steels, only static-cast steels with >20% ferrite are potentially susceptible to thermal embrittlement, static-cast steels with <20% ferrite and all centrifual-gasls are not susceptible.

For high-Mo content (2.0 to 3.0 wt.%) steels, static-cast steels with > 14% ferrite and centrifugal-cast steels with >20% ferrite are potentially susceptible to thermal embrittlement, static-cast steels with :<14% ferrite and centrifugal-cast steels with s20% ferrite are not susceptible.

Ferrite content will be calculated by the Hull's equivalent factors or a method producing an equivalent level of accuracy (+/-t6% deviation between measured and calculated values). Insert #3. For pump casings and valve bodies, screening for susceptibility to thermal aging is not required.

(4) Detection of Aging Effects: For "not susceptible" piping. no additional inspection or evaluations are required to demonstrate that the material has adequate fracture toughness.

For "potentially susceptible" piping, because the base metal does not receive periodic inspection per ASME Section XI, volumetric examination should be performed on the base metal, with the scope of the inspection covering the portions determined to be limiting from the standpoint of applied stress, operating time, and environmental considerations.

Alternatively, a plant/component-specific flaw tolerance evaluation, using specific geometry and stress information, can be used to demonstrate that the thermally-embrittled material has adequate toughness.

Current volumetric examination methods are inadequate for reliable detection of cracks in CASS components; the performance of the equipment and techniques when developed, should be demonstrated through the program consistent with the ASME Section XI, Appendix VIII. For all pump casings and valve bodies, the existing ASME Section XG inspection requirements, including the alternative requirements of ASME Code Case N-481 for pump casings, are considered adequate.

For valve bodies less than NPS 4. the adequacy of inservice inspection according to ASME Section XI has been demonstrated by a NRC performed bounding fracture analysis.

(5) Monitoring and Trending:

Inspection schedule in accordance with IWB-2400 and reliable examination metods should provide timely detection of cracks. (6) Acceptance Criteria:

Flaws detected in CASS components are evaluated in accordance with the applicable procedures of IWB-3500.

If aging management is accomplished through plant/component-specific flaw tolerance evaluation, e.g., for potentially susceptible piping, flaw evaluation for piping with <25% ferrite is performed according to the principles associated with IWB 3640 procedures for submerged arc welds (SAW). disregarding the Code restriction of 20% ferrite in IWB 3641(b)(I).

Flaw evaluation for piping with >25% ferrite Is performed on a case-by-case basis using fracture toughness data provided by the applicant.

(7) Corrective Actions: Repair is in conformance with IWA-4000 and IWB-4000, and replacement according to IWA-7000 and IWB-7000.

(8 & 9) Confnrmation Process and I IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Bolling Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism C 1.1.5, Piping & Recirculation SS 2880C. Cumulative Fatigue C 1. Fittings Lines to RWC Dxygenated Fatigue 1.11 and SLC Water Damage Systems C I. 1.6 Piping & RHR, CS. 288 0 C Cumulative Fatigue thru Fittings LPCI. SS Oxygenated Fatigue Cl. LPCS. Water or Damage 1.10 HPCS, team IC C1.2.1 Recirculatlon Bowl/Casing.

CASS, 288°C, Cumulative Fatigue thru Pump Cover. SS Oxygenated Fatigue C 1.2.3 Seal Flange Water Damage C1.2.1, Recirculation Bowl/Casing, CASS 2880C, Loss of Thermal C1.2.2 Pump Cover (SA351 CF- Oxygenated Fracture Ilng or CF-8M) Water Toughness Embrittle ment C1.2.1 Recirculation Bowl/Casing CASS, 288 0 C. Crack SCC, Pump SS Oxygenated Initiation and IGSCC Water Growth DRAFT- 6/06/00 IV Cl-12 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Administrative ControLs:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The proposed AMP is effective in managing the effects of thermal aging on the intended function of CASS components.

Components have been designed or Fatigue is a time-limited aging analysis nLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic ITLAA life, according to the requirements of Safety Issue (GSfl-190 is to be addressed.

Insert#1.

ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB. or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF). Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # 1. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF). Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

Insert#1.l ASME Section Ill (edition specified in 10 CFR 50.55a), Subsection NB. or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF). Same as for the effect of Thermal Aging Same as for the effect of Thermal Aging Embrittlement on Yes, Embrittlement on piping and fittings in piping and fittings in various reactor cookt pressure existence of various reactor coolant pressure boundary systems Items C1. 1.5 -C1. 1. 11. a suitable boundary systems Items C1.1.5 -AMP should C1.1.11. be evaluated Guidelines of NUREG-0313, Rev. 2 and (1) Scope of Program: The program includes preventive No NRC Generic letter (GL) 88-01 and its measures to mitigate SCC and inservice inspection (ISI) to Supplement 1: inservice inspection in monitor the effects of SCC on intended function of the conformance with ASME Section XU pump. NUREG-0313 and GL 88-0 1, respectively, describe (edition specified in 10 CFR 50.55a), the technical basis and staff guidance regarding the Subsection IWB, Table IWB 2500- 1, problem of IGSCC in BWRs. (2) Preventive Actions: examination categories B-L-1 for pump Mitigation of IGSCC is by selection of material considered casing welds and B-L-2 for pump resistant to sensitization and IGSCC, e.g., low-carbon casing, and testing category B-P for grades of cast SSs and weld metal, with a maximum system leakage. Coolant water carbon of 0.035% and minimum 7.5% ferrite. Also. chemistry is monitored and maintained hydrogen water chemistry and stringent control of in accordance with EPRI guidelines in conductivity is used to inhibit IGSCC. ,oweve .High TR-103515 and BWRVIP-29 to minimize carbon grades of cast SS. e.g.. CF-8 and CF--8M may-b the potential of crack initiation and = susceptible to SCC. The aging management program growth. must therefore rely upon ISI in accordance with GL 88-01 to detect possible degradation.

(3) Parameters Monitored/Inspected:

The AMP monitors the effects of SCC on the intended function of the pump by detection and sizing of cracks by ISI. The inspection requirements of pump casing welds are delineated in GL 88-01. Inspection requirements of Table IWB 2500-1, examination category B-L-2 specifies visual VT-3 examination of internal surfaces of the pump. Inspection requirements of testing category B-P conducted according DRAFT -6/06/00 WV Cl-13 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (BoUling Water Reactor)1Structure andi Region of 1 Environ-i Aging IM 1gn Item IComponent I Intee I Material Iment Effect Mechanism________

'y C 1.2.3. Recirculation Seal Flange, Flange: SS; ir, AtL4osn Wear C 1.2.4 Pump Closure Bolting: ofaMati Bolting High xygenated Strength ater Low-Alloy d/or Steel rteam at (HSLAS) 88 0 C SA193 GrB7 DRAFTI-6/06/00 IV CI-14 IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM CI. REACTOR COOLANT PRESSURE BOUNDARY Water Reactor)Existing Further Aging Management Program (AMP) Evaluation andTechnical BasiI Evaluation Recommendations for a comprehensive bolting integrity program delineated in NUREG- 1339 on resolution of Generic Safety Issue 29 and implemented through NRC Generic Letter 9 1-17; additional details on bolting integrity outlined in EPRI NP-5769; and Inservice inspection in conformance with ASME Section XI (edition specified in 10 CFR 50.55a), Subsection IWB. Table IWB 2500- 1. examination categories B-G-1I or B-G-2 for pressure retaining bolting, and category B-P for system leakage.(1) Scope qf Program: The staff guidance of NRC Generic Letter (GL) 9 1-17 provides assurance that plant specific comprehensive bolting integrity programs have been implemented to ensure bolting reliability.

The NRC staff recommendations and guidelines for a comprehensive bolting integrity program is delineated in NUREG- 1339, and the industry's technical basis for the program is outlined in EPRI NP-5769. (2) Preventive Actions: Selection of bolting material and the use of lubricants and sealants in accordance with guidelines of EPRI NP-5769 and additional requirements of NUREG 1339, prevent or mitigate degradation and failure of all safety-related closure bolting. (3) Parameter Monitored/Inspected:

I.No DRAFT -6/06/00 (continued from previous page) to IWA-5000 specify visual VT-2 (IWA-5240) examination of all pressure retaining boundary of the pump during system leakage test (IWB-5221) and system hydrostatic test (IWB-5222).

Also, coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. (4) Detection of Aging Ecffects:

Degradation of the pump due to SCC can not occur without crack initiation and growth; extent and schedule of inspection as delineated in GL 88-01 will assure detection of cracks before the loss of intended function of the pump. (5) Monitoring and Trending:

Inspection schedule in accordance with GL 88-01 should provide timely detection of cracks. All welds are inspected each inspection period from at least one pump in each group performing similar functions in the system. Visual examination is required only when the pump is disassembled for maintenance, repair, or volumetric examination, but at least once during the period. System leakage test is conducted prior to plant startup following each refueling outage, and hydrostatic test is conducted at or near the end of each inspection interval.

(6) Acceptance Criteria:

Any SCC degradation is evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400; IWB 3518 for volumetric examination of welds and IWB-3519 for visual examination of pump internal surfaces.

Supplementary surface examination may be performed on interior and/or exterior surfaces when flaws are detected In volumetric examination.

(7) Corrective Actions: Repair is in conformance with IWA-4000 and IWV-4000 or GL 88 01. Continued operation without repair require that crack growth calculations be performed according to the guidance of GL 88-01 or other approved procedure.

(8 & 9) Confirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The comprehensive AMP outlined in NUREG-0313 and GL 88-01 addresses improvements in all elements that cause IGSCC and has provided effective means of ensuring structural integrity of the primary coolant pressure boundary.IV CI-15 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM t1l PRESSURE BOUNDARY {BollnN Water Reactor)Structure and Region Lof I nirnf Ain gng Itm Component Interes Material 4 et 4 Effect jMechanism C 1.2.4 Recirculation Closure HSLAS Loss of Stress Pump Bolting SA193 GrB7 Preload Relaxation egenated ater d/or eteam at 888 0 C ________ .88° C 1.2.4 Recirculation Closure HSLAS Cumulative Fatigue Pump Bolting SA193 GrB7 Fatigue xygenated Damage ater and/or team at 88 0 C DRAFT -6/06/00 IV Cl-16 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Same as for the effect of wear on Item Same as for the effect of wear on Item C1.2.4 Closure No Cl .2.4 Closure Bolting for Recirculation Bolting for Recirculation Pump. Pump. Components have been designed or Fatigue is a time-limited aging analysis (TLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life. according to the requirements of Safety Issue (GSI)- 190 is to be addressed.

Insert # 1. ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF).DRAFT -6/06/00 WV Cl-17 (continued from previous page) The AMP monitors the effects of aging degradation on the intended function of closure bolting by detection of coolant leakage. and by detection and sizing of cracks by inservice inspection (ISI). Inspection requirements of ASME Section XI, Table PWB 2500-1. examination category B-G-1 for bolting greater than 2 in. in diameter specify volumetric examination of studs and bolts, and visual VT- I examination of surfaces of nuts, washers, bushings, and flanges. Examination category B-G-2 for bolting 2 in. or smaller specifies only visual VT- I examination of surfaces of bolts, studs, and nuts. However, because most failures have occurred in fasteners 2 in. or smaller, based on IE Bulletin 82-02, enhanced inspection and improved techniques are recommended.

Inspection requirements of ASME Section XI testing category B-P specify visual VT-2 (IWA-5240) examination of all pressure retaining components during system leakage test (IWB-5221) and system hydrostatic test (IWB-5222).

(4) Detection qf Aging Effects: Degradation of the closure bolting due to crack initiation, loss of prestress, or attrition of the closure bolting would result in leakage. The extent and schedule of inspection assure detection of aging degradation before the loss of intended function of closure bolting.

(5) Monitoring and Trending:

Inspection schedule of ASME Section XI are effective and adequate for timely detection of cracks and leakage. (6) Acceptance Criteria:

Any cracks In closure bolting are evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400 and IWB-3515 and 3517. (7) Corrective Actions: Repair and replacement is in conformance with IWB-4000 and guidhlines and recommendations of EPRI NP-5769. (8 & 9) Confimation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The bolting integrity programs developed and implemented in accordance with commitments made in response to NRC communications on bolting events have provided effective means of ensuring bolting reliability.

IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM i- I f't0AT-0 PRESSURE BOUNDARY (Boiling Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism C1.3.1 Valves Body CS 288 0 C, Wall Eroslon/ (Check, oxygenated Thinning Corrosion Control, Hand, Water Motor Operated.

and Relief Valves) C 1.3. 1, Valves Body, CASS 288 0 C, Loss of Thermal C1.3.2 (Check, Bonnet Oxygenated Fracture Aging Control, Hand, Water Toughness Embrittle MO, and Relief ment Valves) I CI.3.1. C1.3.2 Valves (Check. Control. Hand, Motor Operated, and Relief Valves)Valve Body, Bonnet CASS. SS 2880C, :)xygenated Water Crack Initiation anc Growth IGSCC______ ____________

L J -_________DRAFT -6/06/00 rV Cl-18 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Same asfor the effect of Same asfor the effect of Erosion/Corrosion on Item C1. 1. 1 Yes, Erosion/Corrosion on Item C1. LI main main steam piping andfittings.

Element I steam piping andfittings.

should be further evaluated Same asfor the effect of Thermal Aging Same as for the effect of Thermal Aging Embrittlement on Yes. Embrittlement on piping and fittings in piping and fittings in various reactor coolant pressure existence of various reactor coolant pressure boundary systems Items C1.1.5 -0)- 1. 11. a suitable boundary systems Items C1.1.5 -AMP should C1. 1. 11. be evaluated Guidelines of NUREG-0313, Rev. 2 and NRC Generic letter (GL) 88-01 and its Supplement 1; inservlce inspection In conformance with ASME Section XI (edition specified in 10 CFR 50.55a), Subsection IWB, Table IWB 2500-1, examination categories B-M- 1 for valve body welds and B-M-2 for valve body, and testing category B-P for system leakage. Coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth.(1) Scope of Program: The program Includes preventive measures to mitigate stress corrosion cracking (SCC) and inservice Inspection (ISI) to monitor the effects of SCC on intended function of the valves. NUREG-0313 and GL 88 01, respectively, describe the technical basis and staff guidance regarding the problem of IGSCC in BWRs. (2) Preventive Actions: Mitigation of IGSCC is by selection of material considered resistant to sensitization and IGSCC, e.g., low-carbon grades of cast SSs and weld metal, with a maximum carbon of 0.035% and minimum 7.5% ferrite. Also, hydrogen water chemistry and stringent control of conductivity is used to inhibit IGSCC. ,oweve High-carbon grades of cast SS. e.g., CF-8 and CF-8M hay b a= susceptible to SCC. The aging management program must therefore rely upon ISI In accordance with GL 88-01 to detect possible degradation.

(3) Parameters Monitored/inspected:

The AMP monitors the effects of SCC on intended function of the valves by detection and sizing of cracks by ISI. For welds NPS 4 or larger, the inspection requirements follow Phose delineated in GL 88-01. Inspection requirements of Table IWB 2500 1. examination category B-M-2 specifies visual VT-3 examination of internal surfaces of the valve. Inspection requirements of testing category B-P conducted according to IWA-5000 specify visual VT-2 (IWA-5240) examination of all pressure retaining components during system leakage test (IWB-522 1) and system hydrostatic test (IWB 5222). Also, coolant water chemistry is monitored and maintained in accordance with EPRI guidelines in TR 103515 and BWRVIP-29 to minimize the potential of crack initiation and growth. (4) Detection of Aging Effects: Degradation of the valves due to SCC can not occur without crack initiation and growth; extent and schedule of inspection as delineated in GL 88-01 will assure detection of cracks before the loss of the intended function of the valves. (5) Monitoring and Trending:

Inspection schedule in accordance with GL 88-01 should provide timely detection of cracks. All welds are inspected each Inspection period from at least one valve in each group performing similar functions in the system. Visual examination is required only when the valve is disassembled for maintenance, repair, or volumetric examination, but at least once during the period. System leakage test is conducted prior to plant startup following each refueling outage, and hydrostatic test is conducted at or near the end of each inspection interval.

(6) Acceptance Criteria:

Any SCC degradation is DRAFT -6/06/00 IV CI-19 I no IV REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM *III WIflAniTA PRRSSURE BOUNDARY Maollinr Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism C1.3.3, Valves Seal Flange, Flange: Air, Atu Loso Wear C1.3.4 Closure CS, SS Leaking ofMaterial Bolting Bolting: Oxygenated HSLAS Water Lnd/or 'team at 88 0 C CI.3.1 Valves Valve Body. CS, 288°C, Cumulative Fatigue thru (Check, Bonnet. CASS, SS Oxygenated Fatigue C 1.3.3 Control, Hand, Seal Flange Water Damage Motor Operated, and oy Relief Valves) C 1.3.4 Valves Closure HSLAS Ar. Loss of Stress Bolting SA193 GrB7 Leaking Preload Relaxation Oxygenated Water dd/or Steam at 888 0 C C1.3.4 Valves Closure HSLAS Cumulative Fatigue Bolting SA193 GrB7 Fatigue xygenated Damage ater d/or team at 88 0 C C1.4.1 Is lto Tu i g Tubes .: e sie rac kS C J ru Condense r L &C = a S& te ." initiation and Unantici C 1.4.4 Channel Head, Tubesheet:

Growth pa .d S h e s .s s : c ycmm a ir Chamnnl Loading Head: CS. She.ll CS DRAFT -6/06/00 IV CI1-20 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM * ,,f.1 A?.T DDr RflT1WDARY tRnlinU Water Reactor)Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (ccntinued from previous page) evaluated in accordance with IWB-3 100 by comparing ISI results with the acceptance standards of IWB-3400; IWB 3518 for volumetric examination of welds and 3519 for visual examination of valve internal surfaces.

(7) Corrective Actions: Repair and replacement are in conformance with IWA-4000 and IWB-4000 or GL 88-01. and reexamination in accordance with requirements of IWA-2200.

Continued operation without repair require that crack growth calculations be performed according to the guidance of GL 88-01 or other approved procedure.

(8 & 9) Corfrmation Process and Administrative Controls:

Site QA procedures.

review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

The comprehensive AMP outlined in NUREG-0313 and GL 88 01 has provided effective means of ensuring structural integrity of the primary coolant pressure boundary.

Same as for the effect of wear on Item Same as for the effect of wear on Item C1.2.4 Closure No C1.2.4 Closure Bolting for Recirculation Boltingfor Recirculation Pump. Pump. Components have been designed or Fatigue is a time-limited aging analysis rTLAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TLAA life, according to the requirements of Safety Issue (GSI)- 190 is to be addressed,( Insert # I. ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB, or ANSI B3 1. 1, or other evaluations based on cumulative usage factor (CUF). Same as for the effect of wear on Item Same as for the effect of wear on Item C1 .2.4 Closure No Cl .2.4 Closure Bolting for Recirculation Bolting for Pump. Pump. Components have been designed or Fatigue is a time-limited aging analysis (TIAA) to be Yes evaluated for fatigue for a 40 y design performed for the period of license renewal, and Generic TIAA life, according to the requirements of Safety Issue (GSI)-190 is to be addressed.

Inaecfl.

ASME Section III (edition specified in 10 CFR 50.55a), Subsection NB, or ANSI B3 1.1, or other evaluations based on cumulative usage factor (CUFl. ASME Section XQ (edition s2ecified in 10 (11 cope of PrJgram: The program includes inservice Yes CFR 50.55a or CLBL. Table IWC 2500-1. Inspection in accordance with ASME Section XI. and P1= examination category C-H for pressure should be augmented with temperature and radioactivity speific retaining Class 2 components should be monitoring of the shell side water. and eddy current a e augmented by a program of temperature testing of the tubes. (21 Preventive Actions: Monitor lQn and radioactivity monitorinLg of the shell isolation condenser system performance based on the prgram side water, and eddy current testing of plant technical specifications and measurements of tubes temperature and radioactivity in the shell side water, Perform ASME Section XI inspections and eddy current DRAFT- 6/06/00 WV d1-21 REACTOR VESSEL, INTERNALS.

AND REACTOR COOLANT SYSTEM f. 11 REACTOR C'OOLANT PRESSURE BOUNDARY (Boiling Water Reactor)DRMT -6/06/00 IV te Co pnn IneetrEnviron-Agn I Aig I ItmStructure and~ Region of_ Material..

n Efc IIm C1.4.1 Iolation TubtnL Tubes: Lossf leneral, Irm Condenser Tubesheet.

a s. Material EJ+/-Ung.and QI.4.4 ChanneliHad Iubeb Crevice Shelll; CS. SS Vln~lorroslon Channel _hedl: CS_IV CI-22 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM Cl. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Watet'Reacttr)

Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation (continued from previous naae) testing. 13) Parameters Monitored/Insveeted.

The temperature monitoring is directly related to detecting leakage of the condensate return valves, the radioactivity measurement.

ASME Section XI inspections, and eddy current testing to detect tube cracking.

(41 Detection o1 Agino Effects: Cumulative fatigue damage to condenser tubes would result in degradation of component performance.

Monitoring of temperatufe would detect valve leakage: monitoring of radioactivity In shell side water and ASME inspection and eddy current testing assure detection of cumulative fatigue damage to condenser tubes before the loss of intended iunction of the component.

(5) Monitoring and Trending:

The results of temperature and radioactivity monitoring are monitored and trended. (6) Acceptance Criteria:

The monitoring.

testing and inspection results are related to cumulative fatigue damage to condenser tubes and are compared with established acceptable limits. Results of Section XM leakage tests are evaluated in accordance with IWC-3 100 and acceptance standards of FWC-3400 and FWB-3516.

(7) Corrective Actions: Root cause evaluation and appropriate corrective action Is taken when acceptable limits are exceeded or leakage is detected.

Repair Is in conformance with IWVA-4000 and replacement is in accordance with rWA-7000.

I8 & 9) Confirmation Process and Administrative Controls:

Site OA procedures.

review and approval processes.

and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. GOeratino Exerience:

Ojerating olant experience with this AMP indicates timely detection of cumulative fatigue damage to condenser tubes. Same as for the effect of SCC Same as for the effect of SCC andUnantici-pated Cyclic Yes andUnantici-pated Cyclic Leading on Loading on Items C1.4.1 -C1.4.4 isolation condenser lani Items C1.4.1 -C1.4.4 isolation condenser cL~nents.

s.cific augmztion DRAFT- 6/06/00 WV Cl1-23 IV REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM t1. REATOR CfLAN PRESSURE BOUNDARY (Boiinif Water Reactor)Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect IMechanism eU. 1.13 EPiing & Fittings Smal-Bore iping CS'288*C. Qxygenated WAa=Intiaton Thermal M~and L&Adiug I ______________

J ______________

.1 ____________

1 1 ____________

.1 ___________

DRAFT- 6/06/00 TV Cl-24 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM r-i ANvTlrr BOUNDARY (Boiling Water Reactoti Inservlce inspection in conformance with ASME Section XM (edition specified in 10 CFR 50.55a). Subsection IWB.I I Scope of Proor ,l: The program mcludes preveniuvc measures to inhibit cracking and inservlce inspection flSfl to monitor the effects of cracking on the intended function Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation and4 should be further evaluated Table IWB 2500-1. examination category of small-bore piping of reactor coolant system and B-J for pressure retaining welds in piping and testing category B-P for system leakage, and primary water chemistry is monitored and maintained in accordance with EPRI guidelines in TR- 103515 and BWRVIP-29 to minimize Sthe notential of crack initiation and grogwth,..ne1 (2- Preventive ActionswCoolant water chemistr Is mort)red and maintained according to EPRI guidelines in TR- 103515 and BWRVUP-29_tQ minimizeAhe "ptential of crack initiation and groWiL. Also. hydrogen water chemistry and stringent control of conductivity is used to inhibit IGSCC. (3) Parameters Monitored/

Inspected:

The AMP monitors the effects of cracking on the intended function of niping and flttings by detection cracks and leakage Table IWB 2500-1., SISI. Inspection reQuirements of :1mination category B-J specifles surface examination for circ mferential and lonoltudinal welds in eac'h nine or branch run less thanA4 inche nnminal nine .size f(N>SI. and category B-P st~ecifies visual"V7'7- fTWA-52401 examination of all oressure retainina nents during system leakage test flWB-522 11 and i hydrostatic test (MWB-5222).

However. inspection

.,- A-qME Section XG does not require volumetric examinal sneciflc destructive i of pipes less than NPS 4. A plant-mination or a nondestructive examinationn NDEI that permits inspcction of the inside surfaces¢ of the ninine should be conductedto ensure thathas not occurred and the comnonent intended witll he maintained durn-in the extended neriod, LA) nfArdnn IT)earadation of the A... tn ,..-aplei.,r n-suit in leakace ofcoolant.

A time insroection of a samole of locations most suscentible to cracking should be conducted to verLfy that service-induced weld crackina is 1 not occurrine the small-bore ninin0 less than NP'S 4.

nine. fltt~ns. and branch connections.

Actual insoection locations should be~h nd n risk¢-informned annroaches and nhvslcal exnorure levels, and NDE examinations and locations identified in NRC Information Notice (IN) 97-46, 1511 rnitorina and Trendina:

System leakage test is conducted prior to plant startup following each refueling outage. and hydrostatic test at or near the end of each insoection insnectton will be used I 1. The results of one-time ate the freouencv of future in nec-tionn

16) Aec'etance Criteria:

An relevant ions that may be detected during the leakage tests are evaluated in accordance with IWC-3516.(17 Corrective A~itplos:

Renair is in conformance with IWA-4000 and IWB-4000.

renlacement accordine to IWA-7000 and IWB-7000.

If destructive examination is S.rnnlnu.dM tn-nar and renlacement are in accordance withSctinn Xl rules. IR & Corn¶rination Process and Administrative Controls:

Site QA nrocedures.

revie-w and approval nrocesses, and administrative controls are imnlemented In accordance with requirements of Anoendix B to 10 CFR Part 50 and will continue to be adequate for t1e nuenind of license renewal_ (101 Oneratinoa Exr~erience; has rue-c.,rn.d In H1'CT ninin0 (IN R9-S0) and instrument lines (LER 50-249/99-003-11 due to thermal and mechanical loading.DRAFT- 6/06/00 the no e ta of........

i.. .... t..... ..welds in each pipe or branch nin less than 4 inches nominal pig& size MPS), and categoly B-P spccifies visual a, -od surfaces of the piping should be conducted to ensure that cracliLng has not occurred and the component intended function will be maintained uring the extended pcriod, (4) DetectionfAIng

&ffgcts: Degmdation of the piping, de to crackim, would result in leakage of co -time Inspcction of a sample of locations most susceptible induced weld cracking is not occurring in, Ij c arnall-bore piin ................

.in l dn pie fi ti g .n branýh connections.

Actual inspsction locations should b based on risk-Informed apploaches and phyzical accessibilitycsure levels. and NDE CXaUjjUaU=

q . tcchril ues end ofeach ins2Cction (71 Cafmwtim Actions: Repair Is in conformance yvith IWA-4 )00 and PVB-4000, re lacement according to TWA-and IW 0 If..... uctv..........n i em ed .rep& and replacement are in accordance with ASME Section M rules. (8 & 91 Confirmation and Administrativy Controls:

Site QA plocedures, review and approval processes.

and administrative controls are implemented in accordance with rcQuirements of Appcndix the eriod of license renewal (10) Qperating ExRcdence C--k4 has occurred in HPQT nining f7m Aq-AQI and instnnnent lines (LER 50-249 /99-003- 11 due to thermal and mechanical

]Qadjug, IV CI1-25 IV REACTOR VESSEL, INTERNALS.

ANID REACTOR COOLANT'SYSTEM Structure and Region of Environ- Aging Aging Item Component Interest Material ment Effect Mechanism

_______ CI.5. Control Rd Pipngsand 55 Crac trs DrivLLe (RD Fittng Initatio Corrosionl Hiydraulic (Otsd and Gwyh £rraldng C 1.. 1 Control Rod Pipingsand 5 Q~genate Crac trs CI54 DriveCJRD)

Fittunga.

d Water u Initiation Corost £I.5Z Hy~drau licFitr. to 288 and Growth (racling b~ystem rDRe~tum L=n

• 1 5 1 Control Rod CiIg~n arbon Qzygenate Cuimulativeag

• 1 5.. Drive CRD) Fltingls.

Sel dWater u Fatigue LIMe C1.5. Control Ro ValeBod 51 Q~genat Cra Stress Drive fCRD) di&Water Initiatin Corosin Hydraulct

~2RAI =d Growth Cracki~n CI5 Control Rod Pump Csng 5ý Oxygenate Crack Srs Drive (CR)d Water up Inj aiaon Corrmsio Hydrauiafto I2BLq~ an Growth Crackng DRAFr -6/06/00 V -2 IV C 1-26 IV REACTOR VESSEL. INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water Reactor) Existing Further Aging Management Program (AMP) Evaluation and Technical Basis Evaluation Leaching of chlorides from insulation Plant specific aging management program is to be Yes. jackets and other sourses can cause no generic -externally-initiated transgranular stress AMR corrosion cracking fTGSCC) in the stainless steel heat-traced lines. Plant specific agino management program should be implemented.

Same as for the effect of SCCIIGSCC on Same as for the effect of SCC /IGSCC on rj_ ino and fittings I= p2igina and fittings in Items C1.J.1 thru inl tems Cl.l.l thu- C1.l. 11. BWUMP Components have been designed or Fatigue is a time-limited aging analysis MTLAAI to be Ye evaluated for fatigue for a 40 y design Performed for the period of license renewal, and Generic TLAA life, according to the requirements o- Safety Issue QGSI1- 190 is to be addressed.

Insert #1. ASME Section III (edition specified in 10 CFR 50.55a). Subsection NB. or ANSI B3 I. I. or other evaluations based on cumulative usage factor (CUFM. Sane as for the effect of SCCIIGSCC on Same as for the effect of SCCIIGSCC on Item CI.3.1 valve No Item C1.3.1 valve bodu. ty Same as for the effect Qf SCC/IGSCC on Same as for the effect of SCCIIGSCC on Item C1.2.1 Item CQ.2.1 recirculation pum= recirculation 1um= bowli/castng.

bow1L/siag.

DRAFT -6/06/00 IV Cl-27 1V REACTOR VESSEL, INTERNALS, AND REACTOR COOLANT SYSTEM C1. REACTOR COOLANT PRESSURE BOUNDARY (Boiling Water R Structure and Region of Environ- Aging Item Component Interest Material ment Effect C1.5.5. Conotrld Accumulator.

Carbon Qzgcnat Lossof C1.5 Drive fCRD) Scrami Steel dWater u Matera s to 288' System Volume DRAFT- 6/06/00 IV Cl-28 Insert #I The staff recommendation for the closure of GSI-190 is contained in a December 26, 1999, memorandum from Ashok Thadani to William Travers. The staff recommended that licensees address the effects of the coolant environment on component fatigue life as aging management programs are formulated in support of license renewal. One method acceptable to the staff of satisfying this recommendation is to assess the impact of the reactor coolant environment on a sample of critical components.

These critical components should include, as a minimum, those components selected in NUREG/CR-6260, "Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components." The sample of critical components can be evaluated by applying environmental correction factors to the existing code fatigue analyses.

Formulas for calculating the environmental life corrections factors for carbon and low-alloy steels are contained in NUREG/CR-6583, "Effects of LWR Coolant Environments on Fatigue Design Curves for Carbon and Low-Alloy Steels." The formula for calculating the environmental life corrections factor for stainless steels is contained in NUREG/CR-5704, "Effects of LWR Coolant Environments on Fatigue Design Curves of Austenitic Stainless Steels." Insert #2 The reactor vessel internals receive a visual inspection (VT-3) according to Category B-N-3 of Subsection DIB, ASME Section XI. This inspection is not sufficient to detect the effects of changes in dimension due to void swelling.

An acceptable alternative AMP consists of the following:

py 1. Participation in industry programs to address the significance of change in dimensions due to void swelling.

2. Implementation of an inspection program should the results of the industry programs indicate the need for such inspections.

Insert #3 Components containing Nb are considered susceptible and require evaluation on a case-by-case basis. Insert #4 (1) Scope of Program: The program includes inservice inspection (ISI) to monitor the condition of components that depend on preload, and repair and/or replacement as needed to maintain DRAFT- 6/06/00 IV R-5 the capability to perform the intended function.

(2) Preventive Actions: No practical preventative actions are possible.

(3) Parameters Monitoredlnspected:

The AMP utilizes ISI to monitor the effects of stress relaxation on the intended function of the component by detection and sizing of cracks that could be formed by excessive vibration etc. that may occur if the preload is lost. Table IWB-2500, category B-N-3 specifies visual VT-3 examination of all accessible surfaces of reactor internals.

Because VT-3 inspection can only detect degradation that occurs after the loss of preload, it may be adequate if there is sufficient redundancy that loss of some bolting between inspections is accepatable.

In some cases additional inspection may be required.

(4) Detection of Aging Fffects: As part of the AMP it may be possible to identify acceptable levels of preload and demonstrate whether under the fluence of interest whether loss of acceptable preload is likely. VT-3 may not be adequate to detect tight cracks. Also, creviced regions are difficult to inspect visually.

Supplementary inspections by techniques such as ultrasonic testing (UT) or other nondestructive methods may be needed to detect cracking in inaccessible regions. (5) Monitoring and Trending:

Inspection schedule in accordance with IWB-2400 is adequate for timely detection of cracks. (6) Acceptance Criteria:

Any degradation is evaluated in accordance with IWB-3520.

(7) Corrective Actions: Repair and replacement are in conformance with IWB-3140.

(8 & 9) Conrirmation Process and Administrative Controls:

Site QA procedures, review and approval processes, and administrative controls are implemented in accordance with requirements of Appendix B to 10 CFR Part 50 and will continue to be adequate for the period of license renewal. (10) Operating Experience:

There are no reports of stress relaxation producing damage in reactor vessel internals.

Insert #5 The inspection guidance in BWRVIP-75 is under staff review. The topical (BWRVIP-75) when approved by the staff may serve to replace the inspection extent and schedule in GL 88-01. Insert #6 The guidance for weld overlay repair, stress improvement or replacement is provided in GL 88 01, Code Case N 504- 1. or ASME Section XI. Insert #7 The extent and schedule of the inspections and test techniques prescribed by the program are designed to ensure continued tube integrity and that aging effects will be discovered an repaired before there is a loss of intended function.DRAFT-6/06/00 IV R-6 Insert #8 forlginally defined as OKC-steam)

The staff recommendation for the closure of GSI-190 is contained in a December 26, 1999, memorandum from Ashok Thadani to William Travers. The staff recommended that licensees address the effects of the coolant environment on component fatigue life as aging management programs are formulated in support of license renewal. An acceptable method of satisfying this recommendation is to use the high-temperature water data to assess the environmental effects on fatigue life.DRAFT- 6/06/00 IV R-7