Embed Size (px)
Citation preview
UNITED STATES NUCLEAR REGULATORY COMMISSION
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. 20555·0001
MEMORANDUM TO:
FROM:
SUBJECT:
Marvin D. Sykes, Senior Staff Engineer Technical Support Staff ACRS/ACNW
Mario V. Bonaca, Chairman Plant License Renewal Subcommittee
CERTIFICATION OF THE SUMMARY MINUTES OF THE MEETING OF THE ACRS SUBCOMMITTEE ON PLANT LICENSE RENEWAL, NOVEMBER 4,2003 - ROCKVILLE, MARYLAND
I hereby certify, to the best of my knowledge and belief, that the minutes of the subject meeting
on November 4,2003, are an accurate record of the proceedings for that meeting.
Mario V. Bonaca, Date Plant License Renewal Subcommittee Chairman
UNITED STATES NUCLEAR REGULATORY COMMISSION
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D.C. 20555-0001
December 4, 2003
MEMORANDUM TO: Mario V. Bonaca, Chairman Plant License Renewal SUbcorn~tte~ .
FROM: Marvin D. Sykes, Seni~~ Technical support Staff ACRS/ACNW
SUB..IECT: WORKING COpy OF THE MINUTES OF THE ACRS SUBCOMMITTEE ON PLANT LICENSE RENEWAL, NOVEMBER 4, 2003 - ROCKVILLE,MARYLAND
A working copy of the minutes for the subject meeting is attached for your review. Please
review and comment on them at your earliest convenience. If you are satisfied with these
minutes please sign, date, and return the attached certification letter in the pre-addressed
envelope attached.
Attachment: Minutes (DRAFT)
cc wlo Attachment: J. Larkins S. Bahadur S. Duraiswamy ACRS File
Draft Issued: 12/04/03
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS MINUTES OF ACRS SUBCOMMITTEE MEETING ON
GINNA PLANT LICENSE RENEWAL NOVEMBER 4, 2003
ROCKVILLE, MARYLAND
On November 4, 2003, the ACRS Subcommittee on Plant License Renewal held a meeting in Room T-2B3, 11545 Rockville Pike, Rockville, Maryland. The purpose of the meeting was to discuss the license renewal application for the Robert E. Ginna Nuclear Power Plant and the related Safety Evaluation Report with Open Items.
The meeting was open to the public. No written comments or requests to make oral statements were received from members of the public related to this meeting. Mr. Marvin Sykes was the Designated Federal Official for this meeting. The meeting was convened at 12:30 p.m. and adjourned at 4:23 p.m. on November 4, 2003.
ATrENDEES:
ACRS MEMBERS/STAFF M. Bonaca, Chairman G. Leitch, Member P. Ford, Member S. Rosen, Member G. Wallis, Member M. Sykes, ACRS Staff
NRC STAFF/PRESENTERS F. Gillespie, NRR S. Lee, NRR R. Arrighi, NRR J. Rowley, NRR Q. Nguyen, NRR S. Jones, NRR B. Elliot, NRR M. Modes, Region I S. Hoffman, NRR T. Kim, NRR C. Khan, NRR R. Auluck, NRR
APPLICANT/CONTRACT SUPPORT D. Wilson, RG&E G. Gieken, RG&E M. McGraw, RG&E R. Mecredy, RG&E J. Jackson, RG&E G. Wrobel, RG&E G. Herrick, RG&E B. Ganick. ISL
The presentation slides and handouts used during the meeting and a complete list of attendees is attached to the Office Copy of these Minutes. The presentation to the Subcommittee is summarized below.
Opening Remarks (Subcommittee Chair/Designee) Dr. Mario Bonaca, Chairman of the Subcommittee on Plant License Renewal convened the meeting by providing introductory remarks and introducing Mr. Sam Lee, Section Chief, License Renewal and Environmental Review Branch of the Office of Nuclear Reactor Regulation (NRR) to begin the staff presentation.
Staff and Applicant Introduction
Mr. Lee provided brief introductory remarks for the staff and introduced Mr. George Wrobel, Ginna License Renewal Project Manager, Rochester Gas and Electric to begin the discussion.
Presentation by Rochester Gas and Electric (RG&E)
Mr Wrobel summarized the key elements of his presentation w~lich included an overview of the plant operating history, current operational status, scoping and screening methodologies, and commitment tracking.
Plant Description and Operating History (RG&E)
The Ginna plant is a 2-loop Westinghouse PWR with a rating of 1520 MWt located on Lake Ontario in Wayne County New York. The plant was granted a provisional operating license September 18, 1969 and a full term operating license in December 10, 1984. Ginna is currently the oldest operating commercial PWR in the U.S. The plant is assigned to the Licensee Response Column of the NRC Action Matrix with no greater than green NRC performance indicators or inspections findings identified during the past 4 quarters of operation.
Mr. Wrobel described major projects completed at the plant since initial plant startup. In 1996, the applicant replaced steam generators and adopted Improved Standardized Technical Specifications. The applicant conducted core barrel bolting inspections in 1999, in response to concerns related to possible vibration induced failures. Bolts with defect-like indications were replaced. Mr. Wrobel also noted that the applicant had recently completed replacement of the reactor vessel head at Ginna. The replacement head was manufactured with Inconel690 penetrations and new control rod drive mechanisms. The applicant has performed VT-1 inspections of all 36 bottom head penetrations with no indication of operational leakage identified. Cleaning of the bottom head was also completed to facilitate future inspections.
The construction of the Ginna plant was near completion prior to the establishment of 10 CFR 50 Appendix A, General Design Criteria for Nuclear Power Plants. To address this issue, the plant was evaluated for licensing under the NRC's Systematic Evaluation Program (SEP). Mr. Wrobel indicated that this detailed review resulted in a well documented and concise licensing basis which has been a valuable tool for the scoping of structures, systems, and components for license renewal. He also stated that the SEP fostered early use of risk insights in evaluating cost beneficial plant modifications.
According to Mr. Wrobel, RG&E expects to sell the plant in 2004. Final bids for the facility are due mid-November 2003 and the sale will likely be contingent on the plant's operating license being extended for another 20 years.
Development of License Renewal Application
By letter dated July 30, 2002, RG&E submitted the license renewal application for Ginna to the NRC, seeking an additional 20 years of plant operations beyond the current license expiration date of September 18, 2009.
Mr. Wrobel described the scoping and screening process used in developing the license renewal application and noted that the application was prepared in accordance with 1\1 EI 95-10, "Industry Guideline For Implementing The Requirements of 10 CFR Part 54-The License Renewal Rule" and GALL by RG&E staff, using specialty contractor support on an as-needed basis. During the
scoping and screening process, unique plant-specific design features were identified, such as grouted containment tendons and a standby auxiliary feedwater system.
Mr Wrobel stated that the Configuration Management Program specifically designed to track commitments associated license renewal is being developed and is expected to be completed prior to the period of extended operation.
The applicant listed 33 aging management programs. 20 were consistent with GALL, 10 were consistent with minor exceptions, and 3 other plant-specific programs for periodic surveillance maintenance, thimble tube inspection, and systems monitoring were included. The applicant also mentioned that there were 31 commitments related to the Ginna license renewal that have been incorporated into the Ginna Corrective Action Tracking System (CATS). One-third to one-half of the commitments have already been completed with the remaining activities scheduled for completion prior to the period of extended operation.
NRC Staff Discussion of the Ginna SER with Open Items
Mr. Russell Arrighi, NRR Project Manager for Ginna License Renewal and Jonathan Rowley of NRR were introduced to discuss the staff SER with Open Items related to the Ginna Application. The key points of their presentation were:
Overview
Mr. Arrighi began by providing a overview of the Ginna license renewal application and briefly summarized the current status of the license renewal safety evaluation review. He noted that there were 8 open and 7 confirmatory items listed in the SER with Open Items that were to be resolved prior to approval of the final SER. Since issuance of the SER with Open Items, the applicant has provided additional responses with sufficient detail for the staff to consider closing four of the eight open items. Mr. Arrighi also stated that the applicant and staff have reached mutual agreement on resolution of two other open items.
Inspections and Audits
Michael Modes, Senior Reactor Inspector - Region I, discussed the results of the Scoping and Screening and Aging Management inspections conducted by NRC staff. The inspections were implemented in accordance with Inspection Procedure IP 7002 to determine whether the applicant has included all appropriate SSCs in the scope of license renewal and to determine whether the existing aging management programs are adequately managing current age related degradation as required by 10CFR 54.
Mr. Modes informed the ACRS that during the Scoping and Screening inspections, the staff identified several SSC's that appeared to have been omitted from scope. These items were brought to the attention of the licensee for resolution. The licensee has worked cooperatively with the staff to resolve these issues. Although all open items had not been resolved prior to the staff presentation to the ACRS, the staff concluded that the applicant's overall scoping and screening approach had been successful in identifying SSCs needing aging management review in accordance with 10 CFR 54.
He also discussed the results of the staff's Aging Management Review inspection. The staff was able to confirm that the material condition of the plant was adequately maintained and aging
management program documentation was of very good quality supported by a comprehensive computer database. The applicants use of One-Time Inspections (OTis) was also evaluated to verify that specific components were appropriately classified, confirming that aging effects should not be an issue during the period of extended operation. The items for which the OTis are credited are identified in the draft SEA.
Time Limited Aging Analyses
Mr Jonathan Rowley discussed the staff's evaluation of the applicants use of Time Limited Aging Analyses. With the exception of PTS (heatup/cooldown) which will require an update to the Technical Specifications in accordance with Appendix G and H of 10 CFR 50, the staff determined that the TLAAs had been adequately evaluated for the proposed period of extended operation.
Member Comments
In general, the Subcommittee thought the application submitted by RG&E was formatted well and followed the GALL format. The staff continues to perform effective reviews of the license renewal applications and have identified several scoping issues and aging management program inconsistencies that were overlooked by the applicant. The presentations by the staff and applicant indicated that the plant materiel condition is being maintained. No obvious improvements in the efficiency and effectiveness of the license renewal application review process have been noted as a result of GALL.
Dr. Ford noted that he was cautiously satisfied with the application and SER w/Open Items. He noted that the staff should continue to ensure that applicants appropriately incorporate risk when developing their rationale for implementing One Time Inspections (OTis) and that the use of operating experience alone is not an adequate justification for approving OTis.
Mr. Leitch expressed concern that the qualification of electrical equipment will not be verified for all within scope electrical components until just priorto the period of extended operation. He stated that delaying completion of these activities places an additional burden on the staff and believed that a number of these issues could be resolved in a more timely manner.
Mr. Rosen emphasized that applicants for license renewal, especially SEP plants, should not continue to rely solely on previously approved and sometimes unorthodox methods for accident mitigation when seeking approval for license renewal.
Dr. Bonaca expressed concern that continued reliance on the staff in the identification of SSCs omitted by applicants may be indicative of a lack of understanding of the scoping methodology or examples of poor implementation by applicants.
The Subcommittee requested that the staff and the applicant provide additional details on the following topics during future meetings:
- Major Component Replacements/Overhauls Completed or Scheduled - Status of Boric Acid Inspection Program - Status of Reactor Vessel Head (Upper/Bottom) Inspections - License Renewal Commitment Tracking System
Staff Commitments
Staff committed to provide information to the Committee on the design requirements for the Ginna TSC.
Subcommittee Decisions and Follow-up Actions
The Subcommittee will report its finding to the full Committee at the October 2, 2003 meeting.
Background Materials Provided to the Committee
1. License Renewal Application for the R. E. Ginna Nuclear Power Plant, dated July 30,2002.
2. Safety Evaluation Report with Open Items Related to the License Renewal of the R. E. Ginna Nuclear Power Plant, October 2003.
3. NRC Inspection Report 50-244/03-010, Scoping and Screening Methodology, dated August 22,2003.
4. NRC Audit of the R. E. Ginna Nuclear Power Plant Aging Management Programs, dated September 8, 2003.
*********************************************
NOTE: Additional details of this meeting can be obtained from a transcript of this meeting available in the NRC Public Document Room, One White Flint North, 11555 Rockville Pike, Rockville, MD, (301) 415-7000, downloading or view on the Internet at http://www.nrc.gov/reading-rm/doc-collections/acrs/ can be purchased from Neal R. Gross and Co., 1323 Rhode Island Avenue, NW, Washington, D.C. 20005, (202) 234-4433 (voice), (202) 387-7330 (fax), [email protected] (e-mail).
***********************************************
Federal Register/Vol. 68, No. 204/Wednesday, October 22, 2003/Notices 60421
to the Committee at the meeting. Comments submitted during or after the meeting will be accepted but may not be provided to the Committee until after that meeting. FOR FURTHER INFORMATION CONTACT: Any member of the public who desires further information concerning the meeting or wishes to submit oral or written comments should contact Tina Gargus, Special Projects Coordinator, U.S. Institute for Environmental Conflict Resolution, 130 S. Scott Avenue, Tucson, AZ 85701; phone (520) 6705299, fax (520) 670-5530, or e-mail at [email protected]. Requests to make oral comments must be in writing (or by email) to Ms. Gargus and be received no later than 5 p.m. Mountain Standard Time on Tuesday, November 4,2003. Copies of the draft meeting agenda may be obtained from Ms. Gargus at the address, phone and e-mail address listed above.
Dated: October 17, 2003.
Christopher L. Helms, Executive Director, Morris K. Udall Scholarship and Excellence in National Environmental Policy Foundation, and Federal Register Liaison Officer. [FR Doc. 03-26622 Filed 10-21-03; 8:45 amI BILLING CODE 682G-FN-P
NUCLEAR REGULATORY COMMISSION
Advisory Committee on Reactor Safeguards, Subcommittee Meeting on Safeguards and Security; Notice of Meeting
The ACRS Subcommittee on Safeguards and Security will hold a closed meeting on November 12-14, 2003, at Sandia National Laboratories. Albuquergue, New Mexico.
The entIre meeting will be closed to public attendance to protect information classified as national security information pursuant to 5 U.S.C. 552b(c)(1).
The agenda for the subject meeting shall be as follows:
Wednesday, Thursday and Friday, November 12-14, 2003-8:30 a.m. until the conclusion of business.
The Subcommittee will hear presentations from representatives of the NRC staff, NRC staff consultants and industry on pilot plant study insights and potential mitigation strategies. The purpose of this meeting is to gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee.
Further information contact: Mr. Richard K. Major, telephone: (301) 415
7366 or Dr. Richard P. Savio, telephone: (301) 415-7363 between 7:30 a.m. and 4:15 p.m. (ET).
Dated: October 16, 2003.
Sher Bahadur, Associate Director for Technical Support, ACRSIACNW. [FR Doc. 03-26633 Filed 10-21-03; 8:45 am]
BILLING CODE 7590-Gl-P
Dated: October 16, 2003.
Sher Bahadur, Associate Director for Technical Support, ACRSIACNW. [FR Doc. 03-26634 Filed 10-21-03; 8:45 am]
BILLING CODE 7590-Gl-P
NUCLEAR REGULATORY COMMISSION
. :t-Advisory Committee on Reactor NUCLEAR REGULATORY 7'Safeguards; Meeting ofthe COMMISSION
Advisory Committee on Reactor Safeguards, Subcommittee Meeting on Planning and Procedures; Notice of Meeting
The ACRS Subcommittee on Planning and Procedures will hold a meeting on November 5, 2003, Room T-2Bl, 11545 Rockville Pike, Rockville, Maryland.
The entire meeting will be open to public attendance, with the exception of a portion that may be closed pursuant to 5 U.S.C. 552b(c) (2) and (6) to discuss organizational and personnel matters that relate solely to internal personnel rules and practices of ACRS, and information the release of which would constitute a clearly unwarranted invasion of personal privacy.
The agenda for the subject meeting shall be as follows:
Wednesday, November 5, 2003-8:30 a.m.-to a.m.
The Subcommittee will discuss proposed ACRS activities and related matters. The Subcommittee will gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee.
Members of the public desiring to provide oral statements and/or written comments should notify the Designated Federal Official, Dr. Richard P. Savio (telephone: 301-415-7363) between 7:30 a.m. and 4:15 p.m. (ET) five days prior to the meeting, if possible, so that appropriate arrangements can be made. Electronic recordings will be permitted only during those portions of the meeting that are open to the public.
Further information regarding this meeting can be obtained by contacting the Designated Federal Official between 7:30 a.m. and 4:15 p.m. (ET). Persons planning to attend this meeting are urged to contact the above named individual at least two working days prior to the meeting to be advised of any potential changes in the agenda.
Subcommittee on Plant License Renewal; Notice of Meeting
The ACRS Subcommittee on Plant License Renewal will hold a meeting on November 4, 2003, Room T-2B3, 11545 Rockville Pike, Rockville, Maryland.
The entire meeting will be open to public attendance.
The agenda for the subject meeting shall be as follows:
Tuesday, November 4,2003-12:30 p.m. until 4:30 p.m.
The purpose of this meeting is to discuss R. E. Ginna nuclear plant license renewal application and the NRC staff's draft Safety Evaluation Report. The Subcommittee will hear presentations by and hold discussions with representatives of the NRC staff, Rochester Gas and Electric Corporation, and other interested persons regarding this matter. The Subcommittee will gather information, analyze relevant issues and facts, and formulate proposed positions and actions, as appropriate, for deliberation by the full Committee.
Members of the public desiring to provide oral statements and/or written comments should notify the Designated Federal Official, Mr. Marvin D. Sykes (telephone 301/415-8716), five days prior to the meeting, if possible, so that appropriate arrangements can be made. Electronic recordings will be permitted.
Further information regarding this meeting can be obtained by contacting the Designated Federal Official between 7:30 a.m. and 4:15 p.m. (ET). Persons planning to attend this meeting are urged to contact the above named individual at least two working days prior to the meeting to be advised of any potential changes to the agenda.
Dated: October 16, 2003.
Sher Bahadur, Associate Director for Technical Support, ACRSIACNW. [FR Doc. 03-26635 Filed 10-21-03; 8:45 ami
BILLING CODE 7590-Gl-P
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS PLANT LICENSE RENEWAL SUBCOMMITTEE MEETING
R.E. GINNA NUCLEAR POWER PLANT NOVEMBER 4, 2003
ROCKVILLE, MARYLAND
SCHEDULE
I. Opening Remarks
II. Staff Introduction
III. Overview of License Renewal Application • Background • Operating Experience • Unique Scoping and Screening
Methodologies, Major Exceptions to GALL, Commitment Tracking fJRet>,JC..
IV. Overview of Safety Evaluation Report • Resolution of Open Items • Staff Conclusion
SER Chap. 2: Review Results V. • of Scoping Methodologies, Inspections and ROP findings, SBO
M. Bonaca, ACRS
S"'~ LEs:.?il;' I~~NRR",,..,,,,, ...\es!I£ G. Wrobel, et al.
Rochester Gas and Electric
RJ. Arrighi, NRR
RJ. Arrighi, NRR M. Modes, Region I
12:30-12:35 p.m. .,.....,
12:35 -12:40 p.m.
12:40-~.m.
z.:oo p. t'\
z.: ()O -2.; /1>PM jprItfr-!.OO p.m.
Z:,5' - Z;20 I'M.
2,:20 -!:IO a:aa 2:45 p.m.
VII. • SER Chap. 3: Aging R.J. Arrighi, NRR ~3:30p.m.
Management Programs and '"3:10 Review Results, AMP Audit, Open Items
VIII • SER Chap. 4: Time Limited R I 4.rri~A~ NRR Aging Analyses Overview, Open :r:Items
"3:~ - ,,: 21>' IX. • Subcommittee Discussion 4:16 4i30 ~.m.
X. • Adjourn -4.3"t 6.t;~--J
Note • Presentation time should not exceed 50 percent of the total time allocated for specific
items. The remaining 50 percent of the time is reserved for discussion. • 35 copies of the presentation materials to be provided to the Subcommittee
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
SUBCOMMITTEE MEETING ON PLANT LICENSE RENEWL
NOVEMBER 9, 2003 Date
PLEASE PRINT
AFFILIATION
~oA~ ~'S.~ ~~ \. I l. , , l..Mo.(~1Wet1 Me 6i 0LlL
~( neD ::::ra...c..Il. SOY'\ l'
N / C. (
~L ~<.. I,
I( IICl
")::5 L
ADVISORY COMMITrEE ON REACTOR SAFEGUARDS
SUBCOMMITTEE MEETING ON PLANT LICENSE RENEWAL
NOVEMBER 4. 2003 Date
NRC STAFF SIGN IN FOR ACRS MEETING
PLEASE PRINT
NAME NRC ORGANIZATION
Q"YNtI NGV'ffJJ N@ !J'Rlf Rl..lj
T..e.:-Y:'''j ck~ 'RO--\' AV\.l~ <{ L .
•
G" 7 c;.... (tLttI
Ne2)O!.~UL.,,L.....,lLt,....LP~UIoI....--__ vJ Il-(L r[)5YA (SUL'~
s,1v\ L/12. I ~ VIV / ( K
=c::r 1<......:.\...:...V\A---=- _
~\1·~~ Willt'dWl ~6
&~/ Ov.A~~Lffi...,.:.:... _ rv R. a.. {0 f::-I 'Eft CB
I
fJ f<.R IDE / EM CIS ,
IV I-IZ M LfJ LP- L r; (J
ADVISORY COMMITrEE ON REACTOR SAFEGUARDS
SUBCOMMITTEE MEETING ON PLANT LICENSE RENEWAL
NOVEMBER 4, 2003 Date
NRC STAFF SIGN IN FOR ACRS MEETING
PLEASE PRINT
NRC ORGANIZATION
1JRl. / 05~{isPUS IVIlAL (.:blllf? I £[f.{J
'( (, L~)Li
tlR.«"j Of I E/~ E13 ; ;
tJ(LQ. J D€( 61"1 t=f2;;>
~~IPI2\p /£J./f;-f__
1J£t.,l.bc:;kW!f:J N R~ IJ) ftt PIRI-. E:....I-P _ N'2R rVRVf/1< \.. ~~
5~~!-/erfF~C( , CTI~ \AJ '" SoV'\
R.E. GINNA NUCLEAR POWER PLANT
License Renewal SER with Open Items
Staff Presentation to the ACRS RUSSELL ARRIGHI
Project Manager November 4, 2003
> RG&E submitted its application for Ginna by letter dated July 30, 2002
> Westinghouse pressurized water reactor, two loop close cycle, generate 1520 megawatt thermal, and 490 megawatt electrical.
> Systematic Evaluation Program Plant
> Plant located on Lake Ontario in the town of Ontario, Wayne County, New York.
NOl-1!mber +, ZOO}
1
Overview (continued)
~ Current license expires September 18, 2009
~ Request license renewal through September 18, 2029
~ Application implemented the generic aging lessons learned (GALL) process
Nol/tmber "-, 2003
NRC Review Process ~ 8 Open Items (4 resolved)
~ 7 Confirmatory Items
~ Brought into scope and subjected to AMR ~ 0 new structures ex>\",~(l
~ several new components ?:~~< fltl.i' ~~~
> 1 new AMP ~~~....
November"', 2003
2
NRC Audits and Inspections > Scoping and Screening Methodology Audit
~ December 9 - 13, 2002
> Scoping and Screening Inspection ~ June 23 - 27, 2003
> Aging Management Program Audit ~ June 23 - 25, 2003
> Aging Management Review Inspection ~ July 21 -25 and August 4 -8, 2003
NOlle.her -I, 2003
~ 2.1 - SCOPING AND SCREENING METHODOLOGY
> Describes methodology used to identify SSCs that are within the scope of the license renewal rule and subject to an AMR
> Staff audit determined that the applicant's methodology satisfies the rule
NOI'fll1Jber 4, 2003
3
~ Staff reviewed section 2.2 to determine if any systems, structures or commodities supposed to be within scope were omitted
~ No Open or Confirmatory Items
NOl,tn/her+-,2OO3
> Includes:
• Reactor systems • Engineered safety features systems • Auxiliary systems • Steam power conversion systems
3 Open and 3 Confirmatory Items (2 Resolved)
No!'tI'Nbt!r+,2003
4
Section 2.3 Scoping and Screening of Mechanical Systems (Continued)
~ SSCs initially omitted from scope. These SSCs met scoping criterion 54.4(a)(2)
• Steam heating boiler and associated components • CCW piping • Spent fuel pool make-up path
~ These systems were brought into scope, along with associated aging management information
NovelNber4,2003
> Open Item 2.3.3.2 -1 (Resolved)
• Staff identified that failure of out-of-scope CCW piping could result in the failure of the CCW system and thus the ability to cool down the plant following an accident, and therefore should be in scope of the rule.
• All CCW piping is now included in scope and Open / Item is resolved. :/
NQl~l7lber+. 2003 10
5
Open Items (continued)
> Open Item 2.3.3.3 -I (Resolved)
• Staff identified that the spent fuel pool makeup source was not in scope. The loss of spent fuel pool inventory could result in the inability to maintain level and temperature, and therefore should be included in scope of the rule.
• The primary make-up flow path (RWST) was brought into scope and Open Item is resolved.
NOl'tmbtr-l.2003 11
Open Items (continued)
> Open Item 2.3.3.6 -I
• Staff identified that the fire service water booster pump (jockey pump) was not in scope. Loss of booster pu~p could result in loss of fire water header pressure and cycling of fire pump resulting in damage, and therefore should be included in scope of the rule.
Nut'fO/ber.J, 200]
6
Confirmatory Items
> Confirmatory Item 2.3.3.2-1
• Staff identified that the make-up to the CCW surge tank was not in scope. Damage to pressure boundary could affect the ability to mitigate an accident, and therefore should be in scope of license renewal.
• Makeup flow path was initially brought into scope. Need associated aging management information.
IVotJtlHber+,2003 13
SECTION 2.4 STRUCTURES AND STRUCTURAL COMPONENTS
~ Describes structures and structural components
~ Containment
~ Other structures (13)
~ No Open or Confirmatory Items
NOl'ember +. 2003 14
7
SECTION 2.5 - ELECTRICAL SYSTEMS, INSTRUMENTATION, & CONTROL SYSTEMS
> These components were evaluated on a plantwide basis utilizing the "spaces" approach
• 8 commodity groups
• One Open Item
Nonmbtfr.J,2003 15
> Open Item 2.5-1
~ Staff identified that two cables from the offsite power path that brings power into the safety buses were not in scope that could affect the ability to bring the plant to a safe shutdown condition, and therefore should be in scope of the rule.
NOI''elNher.J..2003 16
8
SCOPING AND SCREENING SUMMARY
> The applicant's methodology meets the requirements of the rule.
> Scoping and screening results included all SSCs within the scope of license renewal and subject to anAMR.
NotJf!lIJber 4, 200] 17
LICENSE RENEWAL INSPECTIONS
> Highlights:
~ Scoping and Screening Inspection
~ AMR Inspection
~ PlantROP
NUl'f!lNber .I, 2003 18
9
License Renewal Inspection Program Implementation
> License renewal manual chapter - Me 2516
> License renewal inspection procedure - IP 71002
> Site-specific inspection plan for each applicant
> Scheduled to support NRR's review
> Resources - regional inspection personnel
NOl-Vllbe,../., 2003 19
License Renewal Inspections
> SCOPING AND SCREENING INSPECTION
> Objective: to confirm that the applicant has included all appropriate SSCs in the scope of license renewal as required by the rule.
> One week in length
> Conducted June 23 - 27, 2003 at the Ginna site
> Concluded that scoping and screening process was successful in identifying those SSCs needing aging management review
November 4, 2003 20
10
SCOPING & SCREENING INSPECTION (Continued)
~ Inspection Results • Inconsistent component identification between the fire
protection program and the fire protection aging management program.
NOlJehlber'+, 2003 21
AGING MANAGEMENT INSPECTION
> OBJECTIVE: to confirm that existing AMPs are managing current age related degradation
~ Two weeks in length
~ Conducted July 21- 25 and August 4 - 8, 2003
~ Material condition of plant was being adequately maintained and has improved over time.
~ Documentation was of very good quality supportedby a comprehensive computer database.
~ Third (optional) inspection - Not required
NOl-rHlber'+,2oo3 22
11
-------
Ginna 2Q/2003 PERFORMANCE SUMMARY
Performance Indlcatol'll
---- -----
Ur~t ModiffN: JWy 22,200'
SECTION 3 • AGING MANAGEMENT REVIEW
GALL divides systems and structures into 6 broad system/structural groups
Reactor systems (3.1)
Engineered safety features systems (3.2)
Auxiliary systems (3.3)
Steam and power conversion systems (3.4)
Containments, structures and component supports (3.5)
Electrical and instrumentation and controls (3.6)
NOVl!Hlber.J, 2003 24
12
11 common aging management programs
> 22 system/structural group-specific AMPs
~ Consistent with GALL: 20 ~ Consistent w/GALL, with some deviation: 10 ~ Non-GALL: 3
~ 1 AMP added - system specific
~ 2 Open Items (1 Resolved)
NoveNJber-J,2oo3 25
GINNA AMP AUDIT ~ Date of audit - June 23 - 25, 2003
~ Auditors - 5 project managers from license renewal
~ Audited aU the attributes of the AMPs claimed to be consistent with GALL
~ Concluded AMPs were consistent with GALL except:
• Fire Protection Program • Fire Water System Program
• These AMPs were revised and resubmitted to technical staff for review - Staff found them acceptable
Not'ember-J,2oo3 26
13
SECTION 3.1 - REACTOR SYSTEMS
• Reactor coolant (class 1)
• Reactor vessel • Reactor vessel internals
• Pressurizer • Steam generators • Reactor coolant (non class 1)
> Two Open Items (1 Resolved)
November +, 2003 27
> Open item B2.1.2S-1 (Resolved)
• Staff identified that the applicant is not planning on testing the reactor vessel surveillance capsule after it receives a fast neutron fluence equivalent to the
60-year fluence.
• Commitment to test capsule; Open Item is resolved
November 4, 2003 28
14
Thimble Tube Inspection ProgramFL 01)) I,JOIJClrO VI8IlJJ,'T"oA
> Open item B2.1.36-1 (LITO~6 a(tA.ctt-J":>o. o to 6-AI,v\"'~ S""tJ"
~ Staff identified that several of the program attributes needed further clarification.
NOf--tlf,ber+,2003 29
ALLOY 600
Reactor vessel head, CRDMs, & CRDM penetrations replaced during the September 2003 refueling outage
Replacement head was fabricated with Alloy 690TI (thermally treated) penetrations. The penetrations are welded with partial penetration J-groove Alloy 52 welds.
21/ No Alloy 82/182 welds in RCS piping - is forged stainless steel
Alloy 600 components: BMI penetrations, radial core support pads
Noutmber+,2OO3 30
v
15
~~'f.~~~'G~~~
! ~
• <'/nJF-----------------Reactor Vessel Internals Program
• Commitment by applicant to submit the reactor vessel internals program to the staff for review and approval.
• The commitment will permit the staff to review the AMP for acceptability against NRC requirements as well as those inspection activities recommended by the EPRI-MRP.
NOVlf1'Jbtr.J.,2003 31
SECTION 3.3 • AUXILIARY SYSTEMS
SECTION 3.4 • STEAM AND POWER CONVERSION SYSTEMS
16
SECTION 3.5 - CONTAINMENT, STRUCTURES, AND COMPONENT SUPPORTS
> Containment structure
> 13 other structures
> No Open or Confirmatory Items
NrJ,.'(!",ber.J,200) 33
Aggressive Limit Ginna
pH <5.5 7
Chlorides > 500 ppm 6- 8 ppm
Sulfates > 1500 ppm 20 - 40 ppm
,. Periodic testing to verify chemistry remains non-aggressive
,. Below grade soiVwater environment non-aggressive
NOl'embr:r-/.,2003 3-1
17
SECTION 3.6 - Electrical & I&C
}> 8 Component commodity groups subject to AMR
~ Bus ducts ~ Insulated cables and connections ~Etc.
}> 1 Open & 1 Confirmatory Item (Resolved) --y7ftf1-'''' o,c:- pv
}> One AMP added for Electrical Cables not Subject to EQ Used in Instrumentation Circuits
Notlt:",ber +, 2003 35
• REACTOR VESSEL NEUTRON EMBRITTLEMENT
• METAL FATIGUE
• ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT
Not'l!!lJber+.2003 36
18
TLAA (CONT)
> CONCRETE CONTAINMENT TENDON PRESTRESS
> CONTAINMENT LINER PLATE AND PENETRATION FATIGUE
> OTHER PLANT-SPECIFIC TLAAs
NOlfmber.J.,2003 37
Three analysis affected by irradiation embrittlement identified as TLAAs.
1. Reactor Vessel Upper-Shelf Energy
2. Pressurized Thermal Shock 3. PressurelTemperature Curves
NOltmber.J.,2(0) 38
19
Reactor Vessel Upper Screening Criteria Shelf Energy Glnna (FT-LBS) (USE) (FT-LBS)
Beltline Welds 50 < 50
> RV has margins of safety against fracture equivalent to those DA required by Appendix G, Section XI of the ASME Code; IJ G. J•-, t therefore acceptable. ~ ~ p
> Licensee's analysis satisfies ASME code for 54 effective full-power years (EFPY) (60 years at 90% capacity factor)
Notf!JJlber-l.200J 39
Beltline Welds RTPTS Criterion, Calculated OF RTPTS, OF
.--.. SA· 847 :5 300 ®bI
~ Materials provide adequate protection against PTS events if reference temperature (RTPTs) are within limits
~ Open Item: Applicant is to provide surveillance data, detailed calculations from the surveillance data, and the analysis of the surveillance data
N01'embtr -I, 2003 40
20
..
Plant Heatup/Cooldown (Pressureffemperature) Curves
> Current analyses valid for 28 EFPY
> Technical specifications will be updated as required by Appendices G and H of 10 CFR 50
NOlJe"lhl!r4, z003 41
SECTION 4.3 METAL FATIGUE
~ Reactor coolant system components at Ginna designed to Class 1 of the ASME Code
~ All components have projected CUF <1.0 for period of extended operations
«?-Q... ~ 2 Confirmatory Items '\:J~~ (J'¥
CUF = Cumulative Usage Factor -0 Noa:fllber.J,2003 42
21
-------~"""'"i ~g
;, ~J __.~
....
SECTION 4.4 • ENVIRONMENTAL QUALIFICATION
~ Applicant has adequately identified the TLAA for EQ components
~ Applicant's EQ Program consistent with GALL
~ Staff concluded EQ Program will continue to manage equipment in accordance with 10 CFR 50.49, and meets 10 CFR 54.21(c)(I)(i), (ii) and (iii)
~ No Open or Confirmatory Items
NOl't!mb/lr.J.,2003 43
SECTION 4.5 • CONCRETE CONTAINMENT TENDON PRESTRESS
> Prestress losses estimated for 40 and 60 years
> Applicant provided Trending Analysis
> Staff considers the applicant's actions adequate during period of extended operation
> No Open or Confirmatory Items
_
NOl't!/Nher.J.,2003
22
•
4.6 - CONTAINMENT LINER PLATE AND PENETRATION FATIGUE
> Staff finds design transients and corresponding cycles acceptable
> Staff concludes that the TLAA has been projected to the end of period of extended operation
> No Open or Confirmatory Items
Nut'l!/Hber-l,200J 45
SECTION 4.7 Other Plant-Specific TLAA,..
• Containment Liner Stress • Containment Tendon Fatigue • Containment Liner Anchorage Fatigue • Containment Tendon Bellow Fatigue • Crane Cycle Load Limits • Reactor Coolant Pump Flywheel • Thermal Aging of Cast Austenitic Stainless Steel
> AU demonstrate the TLAA have been projected to the end of the period of extended operation
> No Open or Confirmatory Items
NOlJembI!T.J, 2003
23
~.
CONTENTS • Background/History
• Unique Issues/Features
• Current Issues of Interest - Reactor Vessel Head Replacement
- Reactor Vessel Bottom Head Inspections
- Containment Recirculation Sump Screen
• License Renewal Application Development/Results
• TLAAs
• Programs
• Commitments
• Open Items
BACKGROUND/HISTORY
• Westinghouse 2-loop 1520 MWt PWR
• Initial License granted September 18, 1969 - Systematic Evaluation Program Plant _ POL-FTOL Conversion 12/10/84 ffl.o\l'~IIJAi.c.. ~ Nu..~~M
- CP-OL Recapture 8/9/91
• Steam Generator Replacement - 1996 «~~ ~"c" .
• Improved Standard Tech Specs - 1996
• Baffle-Barrel Bolt Inspection/Replacement - 1999
• Reactor Vessel Head Replacement - 2003
• Plant Sale Anticipated Mid-2004 \ l\A f\k 0 ~~~ ""'II! ,(,o o"e r""o Il::'
p,o~
SYSTEMATIC EVALUATION PROGRAM
• Program Included All Plants with Provisional Operating Licenses, began in 1977
1S()~'\• Compare Ginna to Then-Current SRP
• 92 Topics Reviewed
• Resulted in Well-Documented CLB
• Early Use of Risk Perspective \,\I.... -Iqo.. """,,,,co""'
/.,$1'-- SI
• Cost-Beneficial Plant Changes c.~\ r-3O~lSets
..UNIQUE ISSUES/FEATURES
• Earliest Operating PWR
• Standby Auxiliary Feedwater System
• Grouted Containment Tendons/Rock Anchors
• Station Blackout Recovery
• All NRC Performance Indicators Green
• All NRC Inspection Findings Green
------/Jnalle Tl'c:lne1arm.. nrcl aliI 9ll1lC1laeclf Inla ------l
10 115 KV ~HYMD 5v>.11Qtl UAf
lIt,; rJI ~hl"Gl'l 9"" <II HljjJ Vollcl;ll InVUI~~ ::l4.~ I:V TIWl!>Mt55IlJ\1 llNE: ~ =SFORMER ~ KV TWttl!>W15SWN UftEtI'e"" 7§I .-K (l:ENER'ATDR CTIiCUIT 7']
Sl7'JlX)N 1Jo\ ~~ STEPUP ST'(lroN 2114 ............... <l ~lllAA~FORMER) ,.....-A---,. . I . . f>54I(1I511 1M\. 19 - 1211.7 KIJ
W LlNK ~~~U -2..5l'l; TN')
W I
UNKn7~-i --T~I 751~~7'''0
LlNK BId!!. 52!
I"i"P'" r ::::
I GENERA.TOR ~~ '" 71114)1-\ 34 K'I SllI"*'Joard Bu,. I 76111-\ J4 KV Swit.tvvrd ~a
NC i a: Hlall VOlloae lNlulatel~ f r-.c , dl Hr;~ VCltil;d Inauatllf8 I
1 IGEnR'lroR~WTIIALlAIw 'IalIDgD CDbh.. ~ !Support CRull flnQl:.n ! TIil'lIlSRlRMER ~d Tra"sfcrmal1l ~ I ~ ...crypl""l) L~ ~ ~,,,,,']
I 51... urm lHAtl5ftlRMER 11 sr... IlUK T~5FQRMER 12A\ <J m~-4~WR '18 IIIMII-4UI1II-+111i11'1 ' 34-,.5811-41 1I1l-41 IIGV~ u....iJJ P:mJUl2B <l~ PlITDU11 <1 IJJ...J.....; Iro"l1Dl ZOo
--< K Y-< -< x.,- -(.-- ryn.....L rr-rn ~n--v':l..... _ 4D~ ~
UtlKUNK LlNK 4D~
JMM JllilIII!LBUS DUCT !Ml cucr
lJI\I( UIlK UIlK UNK
PP$u,r., f"I'$u/2 Ppw... Pf'SA/~
_J-- lJ1ilr -- -- - -- -- --,::rm:
~ IlJSI2!V1!1 52/
4Cl
r-rv-v-.-{--<
11!lJS111l/~2 52f IHUSlIA/H. ~ IlUS121A/13~ lJIJlltM'f BRfAl(ER 118 r-.c 1'''' tlC ~ DUM1m' BREAK:E:R
L
CURRENT ISSUES OF l"~"<Il~eo 6'/ INTEREST ~"e"'."
I ~~\) f. ~Q.~ p.A'<'~~ /1 f'"W 6,J~7J 6 ~t;/ ~ '16 PiE~ ~ ~&PlO~~. Reactor Vessel Head Replacement \1":;: c~or'~ ~o<4
• Reactor Vessel Bottom Head Inspections Jr- I fi'< f,"'- o~rJo1 • Containment Recirculation Sump Screen
Lower Reactor Vessel Inspection
-r... f\IlC:.f,;.oV:> ~I 61\..\~ L.-f)~ ~
• Insulation lowered to accommodate inspecti<J>n
• 36 Nozzles examined - 360 0 around
- VT-l Visual
Lower Reactor Vessel Inspection
• Results: - Carbolene zinc based paint on
bottom head and weld pad was in good condition
- No leakage through nozzle penetrations
- No indication ofboric acid or boric acid buildup from operational leakage
- Cleaning performed to facilitate future inspections
"8" Sump WIRE.MESH SCREEN (SST) ,
BSM'T. FL-R. EL. 2.35·-5~ • Committed to
inspect per NRC Bulletin 2003-01
~fl0 6t.)o,~ '::::'--1\Go5
~Pir~IA'-~ - AQ@. l
5ECTlON "'3-3 II EVP(}S5 fl.aw SUMp·B'
• Inspection raised 3 issues: - Sump screen mesh per UFSAR should filter debris> 0.25"
in diameter. Installed mesh size is 3/16" x 9/16".
- Bypass flow issue
- Foreign material found
LRA PREPARATIONS/RESULTS
-(O~v4
• Primarily In-house Matrixed Staff with spe,;~~1ty:(!!'~,i>~ Contractors .. . ~
• LRA prepared following guidance ofNEI 95-10 in SRP format (3rd plant)
• Many Additional Material/Environment Combinations not in Gall (3.X-2 Tables)
• Interim Staff Guidance Addressed in LRA
• Scoping Open Items
SELECTED TLAAs o~(L\~ ~ . ti<p.'<II·~
kJl'jfI' ~ll-~vt • Pressurized ThermalShock- RV ~~S-/
Table 4.2-1 Values ofRTpTS at EOL - Ginna RPV Beltline Materials f
v ~ ~ Ji~~fl
~b ~t.-o
.., ~"Jt:l Imer Inside Heat Surface Initial Chemistry Surfuce
Material Number Fluence E19 n/cm2
RTNDT
of Margin
of Factor
of Fluence Factor
RTNDT
of RTPTS
of
Intermediate 125S255VAI 4.85 20 341 441 1.396 61.4 115.4 Shell
Lower Shell 125P666VAI 4.85 40 341 31 1 1.396 43.3 117.3
Circumferential 61782/ 5.01 -4.8 48.32 161.92 1.403 227.1 270.6 Weld SA-847
1 RegulatOlY Guide 1.99, Rev. 2, Position 1.1 2 Regulatol)' Guide 1.99, Rev. 2, Position 2.1
The RTpTs values for the intermediate and lower shell forgings remain below the NRC screening criterion of 2700 F and the RTpTs value for the beltline circumferential weld (SA-847) remains below the NRC screening criterion of 3000 F at EOl.
~~ ,J.
A~
J~ .~Fracture Toughness - RV ~:t~~
\¢~~f
• Equivalent margins analysis for limiting beltline weld showed substantial margin per ASME Section XI, App. K for ductile flaw extension and tensile stability.
RCS Fatigue Calculation Results (60 years) Components: CUF (with maximum Fen)
- RV Shell and Lower Head Region o - RV Inlet and Outlet Nozzles 0.3922
- SI Nozzle to Cold Leg 0.2517
- RHR-to-SI Tee 0.1428
- Charging Nozzle (200 events) 0.338
- Pressurizer Heater Penetration 0.74
- Pressurizer Surge Nozzle 9.633E-06
- RCS Hot Leg Surge Nozzle 0.2022
PROGRAMS
Total- 33 Consistent with Gall- 20 Plant - Specific - 3 Consistent with Exceptions - 10 New Programs Generated - 4 Enhanced Programs - 6
,
COMMITMENTS
• 29 High Level Commitments in Draft SER
• 122 Individual Commitments
• All Included in Commitment/Action Tracking System and Assigned to Plant Staff
• LRA Program to Configuration Management Process Being Developed
