ML15261A140
| ML15261A140 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/19/1997 |
| From: | Hoffman S NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| NUDOCS 9705210258 | |
| Download: ML15261A140 (27) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 19, 1997 ORGANIZATION:
Duke Power Company
SUBJECT:
SUMMARY
OF MEETING WITH DUKE POWER COMPANY ON LICENSE RENEWAL ACTIVITIES FOR OCONEE NUCLEAR STATION UNITS 1, 2, AND 3 On May 8, 1997, representatives of Duke Power Company (Duke) met with the Nuclear Regulatory Commission staff to discuss the status of Duke's license renewal activities for Oconee Nuclear Station, Units 1, 2, and 3. A list of meeting attendees is provided in Attachment 1. The Duke handouts are contained in Attachments 2, 3, 4, and 5.
Duke informed the staff that the goal of their program is to complete the Oconee license renewal application and be prepared for its submittal as early as July 1998. Duke is assessing the technical, regulatory, environmental, financial, and political aspects of license renewal before making the decision to apply.
The intent is to submit the technical information required to support a renewal application in advance of an actual license renewal application. The information is being submitted in the form of a topical report consisting of 5 major sections. Attachment 3-is the table of contents for the topical report, "Oconee Nuclear Station, License Renewal - Technical Information Topical Report," OLRP-1001. The first section on the reactor building was submitted in July 1996 and revised in March 1997.
In-house license renewal basis documents are being prepared as input to OLRP-1001. Attachment 4 contains excerpts from the basis document for the reactor building. Duke plans to.
submit the remaining sections (structures, electrical, mechanical, and reactor coolant system) by September 1997. In addition, Duke plans to incorporate into their application by reference topical reports prepared by the Babcock and Wilcox Owners Group and Westinghouse.Owners Group (Attachment.5). Duke is, requesting staff review and approval by January 1998, to the extent possible.
Preparation of the environmental report for license renewal has also begun.
Duke is scheduled to submit vertical slice evaluations for selected structures, components, or time-limited aging analyses for each section prior to submittal of 'the complete section to demonstrate implementation of Duke's process. Staff comments regarding the format and content would be considered in developing the complete sections. The vertical slice evaluations are scheduled to be submitted for staff review between May and July 1997 with the complete sections scheduled for submittal between July and September.1997.
Duke has requested periodic technical meetings with the staff beginning in May 1997 to discuss the submittals.
The Oconee application materials are being created using the guidance contained in the Nuclear Energy Institute (NEI) guideline, NEI 95-10, Revision
- 0. The-staff's draft license renewal regulatory guide, DG-1047, proposes to 9705210258 970519 PDR ADOCK 05000269 P.
-2 endorse NEI 95-10, Revision 0. Duke indicated that as discussed at the April 22, 1997, NEI-NRC senior management meeting, maintaining DG-1047 as draft for trial use will not adversely impact the Oconee license renewal activities.
Duke requested that a technical review meeting be scheduled for the end of May or early June 1997 to discuss the structural and electrical vertical slice submittals.
Stephen T. Hoffman, Senior Project Manager License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor-Regulation Docket Nos. 50-269, 50-270 and 50-287 cc:
See next page R. L. Gill, Duke Power
Oconee Nuclear Station Units 1, 2, and 3 cc:
Mr. Paul R. Newton Mr. Ed Burchfield Duke Power Company, PB05E Compliance 422 South Church Street Duke Power Company Charlotte, North Carolina 28242-0001 Oconee Nuclear Site P. 0. Box 1439 J. Michael McGarry, III, Esquire Seneca, South Carolina 29679 Winston and Strawn 1400 L Street, NW.
Ms. Karen E. Long Washington, DC 20005 Assistant Attorney General North Carolina Department of Mr. Robert B. Borsum Justice Framatome Technologies P. 0. Box 629 Suite 525 Raleigh, North Carolina 27602 1700 Rockville Pike Rockville, Maryland 20852 Mr. G. A. Copp Licensing -
EC050 Manager, LIS Duke Power Company NUS Corporation 526 South Church Street 2650 McCormick Drive, 3rd Floor Charlotte, North Carolina 28242-0001 Clearwater, Florida 34619-1035 Dayne H. Brown, Director Senior Resident Inspector Division of Radiation Protection U.S. Nuclear Regulatory Commission North Carolina Department of Route 2, Box 610 Environment, Health and Seneca, South Carolina 29678 Natural Resources P. 0. Box 27687 Regional Administrator, Region II Raleigh, North Carolina 27611-7687 U. S. Nuclear Regulatory Commission Atlanta Federal Center Mr. J. W. Hampton 61 Forsyth Street, S.W., Suite 23T85 Vice President, Oconee Site Atlanta, Georgia 30303 Duke Power Company P. 0. Box 1439 Max Batavia, Chief Seneca, South Carolina 27679 Bureau of Radiological Health South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, South Carolina 29201 County Supervisor of Oconee County Walhalla, South Carolina 29621
Meeting Summary HARD COPY Docket File j PUBLIC PDLR R/F OEDO RIV Coordinator, 0-17G21 E-MAIL:
S. Collins/F. Miraglia (SJC1/fJM)
R. Zimmerman (RPZ)
M. Slossom (MMS)
S. Weiss (SHW)
S. Hoffman (STH)
S. Meador (SAM)
D. Matthews (DBM)
C. Craig (CMC1)
R. Wessman (RHW)
J. Strosnider (JRS2)
S. Droggitis (SCD)
S. Peterson (SRP)
G. Lainas (GCL)
B. Morris (BMM)
J. Moore/E. Holler (JEM/EJH)
G. Mizuno (GSM)
G. Holahan (GMH)
B. Sheron (BWS)
M. Mayfield (MEM2)
A. Murphy (AjG1)
H. Brammer (HLB)
L. Shao (LCS1)
G. Bagchi (GXBl)
R. Johnson (REJ)
C. Grimes (CIG)
PDLR Staff
ATTENDANCE LIST NRC MEETING WITH DUKE POWER COMPANY MAY 8, 1997 NAME ORGANIZATION
- 1. Steve Hoffman NRC/NRR/PDLR
- 2. Robert Gill Duke Power
- 3. Greg Robison
-Duke Power
- 4. Bob Borsum FTI
- 5. Claudia Craig NRC/NRR/PGEB
- 6. P. T. Kuo NRC/NRR/PDLR
- 7. Paul Shemanski NRC/NRR/PDLR
- 8. D. B. Matthews NRC/NRR/PGEB
- 9. Marylee Slosson NRC/NRR/DRPM Attachment I
Oconee License Renewal Project Status Report Duke
/ NRC Meeting May 8, 1997 Oconee License Renewal Project Goal The goal of the Oconee License Renewal Prject is to complete the Oconee license renewal application and to be prepared for its submittal in 1998.
Key Assumptidn-License Renewal will be predictable and cost effective.
Project Challenges a Establishing a predictable framework for License Renewal.
a Producing a technical product that has use today and tomorrow for Oconee.
a Producing an Application for Renewal License that drives us toward maturity.
ATTACHMENT 2
K, 0
How We Address the Challenges We are addressing the project challenges by M Working with the NRC License Renewal Project staff a Working within the industry structure (BWOG, WOG, EPRI and NEI) n Working with the Oconee and NGO engineering staffs a Working with the Duke environmental specialists, business strategists and comnmunication specialists o...t u 4
Near Term Key Near term we need to know the standard by which we will be measured in order for us to be prepared to submit an application for a renewal license.
OmsLame mml h5 Duke Commitment Completing the preparation of the Oconee License Renewal Application is a key initiative in Duke's corporate strategic direction of worldclass utility operations and becoming a leading power generation company worldwide.
from 1997 Duke Power Company Busimes Unit Plan 0-.
U-a.d Fq2 2
Project Organization a 7 full time engineers on the Ocnee Lices R~ewul Project Team
- maid Aiqfmw froms
-Envhasmental
-Strategic business planning______________
-Oconee eneng
-Nuclear General Office engn~eerng
- cmact Supprt fiom Duke Engineening & Services
- Framatome Technologies Areas ofproject focus
- Technical
- Regullatory
- Environmental
- Financial 0 Political Duke License Renewal Process Design LWWWns Aplication Bauis &
Renewal for UFAP Operating Basis Renewal Supplement Pragrmnsent License 3
F Duke License Renewal Process a Developing a series of comprehensive in house license renewal basis documents.
m Use the basis documents as input to a licensing topical (OLRP-1001).
a OLRP-1001 forms the technical portion of the Oconee license rnewal application.
License Renewal Basis Documents n Structured format to address:
Cbezom + Aging Effect + Prepam +Dem
-Resa m Aimed at the plant engineenng staff.
a Living Engineering Document a GSI-166 & 168 are being addressed a Consistent with the guidance provided in Chapters 4 & 5, NEI 95-10, Revision 0 Application for Renewal License a Formal Application
- Exhibit A - Technical Information
- Exhibit B - UFSAR Supplement
- Exhibit C-Technical Specification changes
- Exhibit D -Environmental Report om u.e A.m al 12 4
Technical Information n OLRP-1001 License Renewal, Technikal Information - Topical m Based on sound technical reviews documented in Oconee engineering documents previously discussed a Guidance from Chapter 6, NE1 95-10, Rev 0 applies a Submittal # 2 was made March 12, 1997 m Subsequent submittals to be made over next several monds months Ocas Laee Rme 13 Technical Information a Review/Approval of OLRP-1001 to the extent possible byJanuary 1998 a The Application for Renewal License will incorporate by reference the approved portions of OLRP-1001 and will address any open topics (i.e. Renewai Applicant Action Items) a Table of Contents of OLRP-1001 in Handout UFSAR Supplement a Descriptions of audited aging management programs required to be incled a TLAA evalutions required to be inchded a Coordination with ongoing UFSAR improvement efforts s continuing a Example level of detail provided via NEI to NRCin January 1997 n Discussions recommended a Standard for level of detail in UFSAR Supplement needs to be in place byJanuary 1998 Ons Uesl5 h
IS 5
Technical Specification Changes M Convezsion from conee CIS to M1 in ptgess m Oconee 113 submittalexpected October 1997 n NRC aplpeced innmid1998 0 Tech Spec changes requred by Ic-ense
-lsv if -ny,___________________
am not antkipated to be requested until after the 113 is smlrezdat Oconee 0 Some credited programs wil be in Oconee M1 Chapter 5 N Some former CIS proyans may also be credkcd Technical Specification-Changes 0 113 Chapter 5 Programs
> Cmapmn C~tk c Tnwt limit
> Tendon Surweet_________________________
> SG Tube I0CMpetoS
> Saoc-d-yW-awO vziy
> Veation, Faff Temng
> W DmJ RAe (M Tnim%
> PnsmzyQdM &== 0~c ~zn
- Former CIS Progrms
> RG3 axmnity Environmental Report a First naee* nwith NRC held on March 5, 1997
- Draft czonee Environmental Report for License Renewal in preparaton
- Key Topics include;
- HLWTramPoMtiea SAMDAx
- Akemaimatothl
~ction
- Second meeting with NRC anticipated forjune n NRC draft guidance ezpected! this summer U Need to establish the degree of acceptability of dhe Oconee_
Environmental Report byJanruary 1998
- a.
Uii P*
6
Priority Items for Duke a RevieApprow OLRP 1001 S~xi-12 C~ActwMPAVm TLAA
- TLA Wdesiitoo s 0-ax Gimq, Piiy I- (-e bamkxg) m RevienVApuv PFw= OLRP. 2001 S~t~
Sbucusmi IPA sMdTL4A EkcuicalIPAand TLAA hiecink EPA ad TLAA KCS PA MW TLAA m Revfrwdraft Euimerjl Report a Recwdraft I.FSAR &*~pkawr 0-L-
3.J P.W.
1 Future Interactions
-Technical meetings with review and feedback on__________________
exumples 0 July
- ngenM Meeting - assess Progs-s a July.- Septembe
-FormalSuDMitta Of OLRP-1001 Sections for revle/appm-val I'
Future Interactions
- Seper
- December 1997
-Technica metng, RAhs and responses, Complete ftWVie to the --rn Possible 0 Januzyl1998
-Managemnent meeting. assess prosress and futur plans 0 January. June 1998
- Continue Technical REviews wvith NRC__
- Continue Duke internal preparations for submittal
- July1998
-Earliest Expected Submittal Date 7
Regulatory Guide Development a Oconee IPA and TLAA Reviews are being created using the guidance provided in NEI 95-10, Revision 0.
a The development of a Regulatory Guide endorsing NEI 95-10, Revision 1 on the priority established during the NEI/NRC Management Meeting on April 22,1997 will not adversely impact the Oconee license renewal activities.
Om L-Amd RPq Meeting Our Challenges x Byjanuary 1998, we need to know the standard by which we will be measured, as best we can.
m We can know this through:
- processing submittals
- technical meetings
- RAIs and responses
- Draft technical SRP
- ER draft Reg Guide and draft SRP
- completed evaluations Key Future Dates 1998 Timeframe for Submittal 2000 Staff Decision on Oconee Application 2001 Deregulation Timeframe 2003 Comnission Decision on Application 4th Oconee ISI Interval Begins 2033 Renewal EOL o
u...
R..
P..
4*
8
OCONEE NUCLEAR STATION LICENSE RENEWAL - TECHNICAL INFORMATION TOPICAL REPORT OLRP-1001 Revision 1 February 1097 ATTACHMENT 3.
uconee Nuclear Station License Renewal - Technical Information Topical Report OLRP-1001 Table of Contents
- 1. INTRODUCTION.........................................1-1 1.1 PURPOSE.....
1.2 ORGANIZATION......
....... 2-1 1.3 CLB CHANGES DURING NRC REVIEW OF APPLICATION
........... 3-1 1.4 TIME-LIMITED AGING ANALYSIS REVIEW...........................................1.4-1 1.4.1 IDENTIFICATION OF TIME-LIMITED AGING ANALYSES...................................
1.4-1 1.4.2 EXEMPTIONS...................................
..................... 1.4-2 1.
4.3 REFERENCES
1.4-2
- 2. INTEGRATED PLANT ASSESSMENT - STRUCTURE/COMPONEP'T IDENTIFICATION.....................................
- ..**..**..**..************......................2.1-1
2.1 INTRODUCTION
.................. 2.1-1 2.2 METHODOLOGY TO IDENTIFY SYSTEMS, STRUCTURES, AND COMPONENTS WITHIN THE SCOPE OF LICENSE RENEWAL.....................2.2-1 2.2.1 REVIEW TO CRITERIA CONTAINED IN §§54.4(a)(1) AND (a)(2).......... 2.2-1 2.2.1.1 Structures.......................
2.2-1 2.2.1.2 M echanical Systems...................
........................................ 2.2-2 2.2.1.3 Electrical Systems.....................................-..............
2.2-2 2.2.2 REVIEW To CRITERIA CONTAINED IN §54.4 (a)(3)..............--.................2.2-2 2.2.2.1 Fire Protection.....................................
-................. 2.2-2 2.2.2.2 Environmental Qualification.............................
2.2-3.
2.2.2.3 Pressurized Thermal Shock (PTS).........................
2.2-3 2.2.2.4 Anticipated Transient Without Scram (ATWS).................
2.2-3 2.2.2.5 Station Blackout.....................................2.2-4 2.2.3..
RESULTS 2.2.3 RESULTS 2.2-4 2.
2.4 REFERENCES
2.2-6 2.3 REACTOR BUILDING...................2.3-1 2.3.1 CONCRETE COMPONENTS...............
.2.3-3 2.3.1.1 Dome and Cylinder Walls.
2.3-3 2.3.1.2 Floor
......... 2.3-3 2.3 -3 2.3.1.3 Foundation Slab......................................2.3-3 2.3.2 STEEL COMPONENTS..........
................. 2.3-4 2.3.2.1 Liner Plate
........ 2.3-4 Revision I February 1997
%-ce iNuclear atatlon License Renewal -Technical Information Topical Report OLRP-1001 2.3.2.2 Anchors/Embedments/Attachments........................2.3-4 2.3.2.3 Personnel Hatch....................
2.3-5 2.3.2.4 Equipment Hatch....................................2.3-6 2.3.2.5 Mechanical Penetrations...............................2.3-6 2.3.2.6 Electrical Penetrations.................................2.3-7 2.3.2.7 Fuel Transfer Tube....................
............... 2.3-8 2.3.3 POST-TENSIONING SYSTEM.................................2.3-8 2.
3.4 REFERENCES
2.3-9 2.4 REACTOR COOLANT SYSTEM 2.4-1 2.5 MECHANICAL COMPONENTS 2.5-1 2.6 ELECTRICAL INSTRUMENTATION & CONTROL COMPONENTS........26-1 2.7 STRUCTURES AND STRUCTURAL COMPONENTS...............2.7-1
- 3. INTEGRATED PLANT ASSESSMENT -
AGING MANAGEMENT REVIEW.
3.1
3.1 INTRODUCTION
3.1-1 3.2 AGING MANAGEMENT REVIEW PROCESS OVERVIEW....................3.2-1 3.3 REACTOR BUILDING 3.3-1 3.3.1 CONCRETE COMPONENTS AGIG MANAGEMENT REVIEW...................3.3-1 3.3.1.1 Aging Management ReviewO CONTROL...COMPONE....
... 3.....3-2 3.3.1.2 TLAA.....................
.....3.3-5 3.3.2 STEEL COMPONENTS AGING MANAGEMENT REVIEW
................. 3.3-5 3.3.2.1 Aging M anagement Review...
3 -5 3.3.2.2 Time-Limited Aging Anaysis 3.3-11 3.3.3 POST-TENSIONING SYSTEM AGING MANAGEMENT REVIEW PROCESS.OVERVI.3.3-13 3.3.3.1 Aging Management Review............................3.3-13 3.3.3.2 Time-Limited Aging Analysis.............................3.3-17 3.3.4
SUMMARY
AND CONCLUSIONS...................................3.3 3.3.4.1 Concrete Components....................................3.3-18 3.34.2 Steel Components....................................
3.3-19 3.3.4.3 Post-Tensioning System...............................
3.3-19 3.3.4.4 Conclusions.......................................
3.3-20 3.
3.5 REFERENCES
3-22 3.4 REACTOR COOLANT SYSTEM GE...................W........
.... 3.4-1 3.5 MECHANICAL COMPONENTS...............................
........... 3.5-1 Revision I February 1997.
License Renewal - Technical Information Topical Report OLRP-1001 3.6 ELECTRICAL / INSTRUMENTATION & CONTROL COMPONENTS........3.6-1 3.7 STRUCTURES AND STRUCTURAL COMPONENTS..........
3.7-1
- 4. CO N CLUSIO N...........................................................................................................
4-1 Ill Revision I February 1997
Duke Power Company Oconee Nuclear Station Units 1, 2, & 3 Oconee Reactor Building Containment Aging Management Review for License Renewal Q A Con8i~or\\,
OSS-0274.00-00-0003 Revision 1 March 31, 1997 ATTACHMENT 4
OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page ii VERIFICATION OF SPECIFICATION Station and Unit Number: Oconee Units 1, 2, & 3 Title of Specification: Oconee Reactor Building Containment Aging Management Review for License Renewal Specification Number: OSS-0274.00-00-0003 Revision: 1 This document specifies items related to License Renewal. In accordance with established procedures, its quality has been assured. Signatures certify that the above specification was originated, checked, approved, and inspected (or waived) as noted. Document issuance is done in accordance with QA Condition 1 issuance standards.
Signature also certifies that a review for determining potential impact to work performed per previous revisions was conducted for this revision.
Previous work impacted by this revision:
Yes, See Attachment No Prepared by:-
Date: 3,8
?
Debra V. Ramsey Checked by:
ate ________
Robert V. Hester Checked by:
Date:.
7 MarkJ.
i Approved by:
Date:.- / 77 Gregory D on Inspections Date Date Date Date Date Date
OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page iii REVISION DESCRIPTION SHEET Revision Number People Involved Page(s) or Sections Revised and Description 0
Preparer:
D. V. Ramsey Original Checker:
L. M. Llibre Checker:
M. J. Ferlisi Approver G. D. Robison I
Preparer D. V. Ramsey Changed Reactor Building to Containment, Checker:
R. V. Hester where appropriate, throughout the Spec. to Checker:
M. J. Ferlisi emphasize that the focus is on the Containment Approver:
G. D. Robison portion of the building.
Page 7 Reword sentences on lines 8 and 9 Page 12 Checked function 7 for all reinforced.
concrete comps. in Table 4.2; changed function 2 to read "...with no loss of structural integrity.";
changed function 3 to radiation shielding; included design basis accidents in function 7.
Page 19 Checked function 7 for all reinforced concrete comps. in Table 7.1; changed function 2 to read "... with no loss of structural integrity.";
changed function 3 to radiation shielding; included design basis accidents in function 7.
Page 33 Added statement from NRC Commission briefing on shrinkage cracks.
Page 36 Changed function 2 to read "... witn no loss of structural integrity."; changed function 3 to radiation shielding; included design basis accidents in function 7.
Page 47 Deleted ref. to 7.2.2.3.4 and added info on irradiation embrittlement.
Pages 52-57 Added new text on ASME Subsection IWE.
Page 58 Reword last sentence of first paragraph.
Editorial change to sentences on lines 20 and
- 21.
Page 59 Editorial change to last sentence in first paragraph.
Page 62 Editorial change to function description.
Page 65 Added sentence on potential grease leakage on line 18.
Pages 67-72 Added new text on ASME Subsection IWL.
Page 73 Included design basis accidents in function description.
OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page iv TABLE OF CONTENTS 2
DEFINITIONS AND TERMS..................................................................
vi 3
1.0 INTRODUCTION
1 4
2.0 PURPOSE......................................................3 5
3.0 REACTOR BUILDING DESCRIPTION.......................................................................................................
4 6
4.0 REACTOR BUILDING SCOPING - 10 CFR 54.4.............................................................................................
6 7
4.1 SAFETY-RELATED SCOPING RESULTS 7
8 4.2 REGULATED EVENTS SCOPING RESULTS....
7 9
4.2.1 Fire Protection 7
10 4.2.2 Environmental Qualification 8
11 4.2.3 Pressurized Thermal Shock (PTS) 8 12 4.2.4 Anticipated Transient W ithout Scram (ATW S).....................................
8 13 4.2.5 Station B lackout (SBO )................................................................................................................................
8 14 4.3 REACTOR BUILDING SCOPING RESULTS AND INTENDED FUNCTIONS 9
15 4.4 REACTOR BUILDING CONTAINMENT COMPONENT SCOPING AND INTENDED FUNCTIONS....................................11 16 5.0 STRUCTURAL COMPONENTS SUBJECT TO AMR.............................................................................
14 17 6.0 AGING MANAGEMENTREVIEW (AMR) METHODOLOGY FOR CONTAINMENT STRUCTURAL 18 COM PONENTS....................................................................................................................
15 19 7.0 AGING MANAGEMENT REVIEW RESULTS FOR CONTAINMENT STRUCTURAL 20 COMPONENTS..........................
17 21 7.1 IDENTIFICATION OF COMMODITY SETS 17 22 7.2 CONCRETE COMPONENTS AGING MANAGEMENT REVIEw...............................................18 23 7.2.1 Concrete Component Descriptions 20 24 7.2.2 Concrete Component Aging Effects......................................................................................................
21 25 7.2.2.1 Loss of M aterial A ssessm ent...................................................................................................................21 26 7.2.2.1.1 Freeze-thaw......................................................................22 27 7.2.2.1.2 A brasion and C avitation..........................................................
22 28 7.2.2.1.3 Elevated Tem perature 23 29 7.2.2.1.4 Aggressive Chem ical.......................................................................................................................
24 30 7.2.2.1.5 Corrosion of Embedded Steel/Rebar...................................................................................
25 31 7.2.2.2 Cracking Assessment 27 32 7.2.2.2.1 Freeze-thaw.........................................................................27 33 7.2.2.2.2 Reaction with Aggregates...................................................
27 34 7.2.2.2.3 Shrinkage 28 35 7.2.2.2.4 Settlement 29 36 7.2.2.2.5 Elevated Temperature 29 37 7.2.2.2.6 Fatigue 29 38 7.2.2.3 Change in Material Properties Assessment..........
29 39 7.2.2.3.1 Leaching of Calcium Hydroxide..................................
30 40 7.2.2.3.2 Elevated Temperature 30 41 7.2.2.3.3 Aggressive Chemical Attack.............................................30 42 7.2.2.3.4 Irradiation Embrittlement...................
31
OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page v 1
7.2.2.4 Industry Experience............................................................................................
31 2
7.2.2-5 Summary of Containment Concrete Component Aging Effects.............................................
33 H1 3
7.2.3 Management of Aging Effects for Concrete Comoonents......
34 4
7.3 STEEL COMPONENTS AGING MANAGEMENT REVIEW................................................................
35 5
7.3.1 Steel Component Description...................................................................................
37 6
7.3.2 Steel Component Aging Effects..................................................................................
43 7
7.3.2.1 Loss of Material Assessment.................................................................................
43 8
7.3.2.2 Cracking Assessment.........................................................................
- ................ 45 9
7.3.2.2.1 Stress Corrosion...............................................................................................
45 10 7.3.2.2.2 Fatigue......................................................................................................
46 11 7.3.2.3 Change in Material Properties Assessment...................................................................
47 12 7.3.2.3.1 Elevated Temperature......................................................................................
47 13 7.3.2.3.2 Irradiation Embrittlement...................................................................................
47 14 7.3.2.4 Industry Experience.........................................................................................
48 15 7.3.2.5 Summary of Steel Component Aging Effects......................
5 16 7.3.3 Management of Aging Effects for Steel Components..........................................................
51 17 7.3.3.1 ASME Code Section MI, Subsection IWE Inservice Examination..........................................
52 18 7.3.3.2 Reactor Building Integrated Leak Rate Test..................................................................
57 19 7.3.3.3 Reactor Building Local Leak Rate Test.......................................................................
58 20 7.3..).4 Summary of Management of Aging Effects for Steel Components..........................................
59 21 7.3.4 Time-Limited Aging Analysis...................................................................................
60 22 7.4 POST-TENSIONING SYSTEM AGING MANAGEMENT REvIEw...................................
- .. ****I'**........62 23 7.4.1 Post-Tensioning System Component Description..............................................................
63 24 7.4.2 Post-Tensioning System Component Aging Effects..........................................................
64 25 7.4.2.1 Loss of Material Assessment.................................................................
I................ 64 26 7.4.2.2 Industry Experience..................................
65 27 7.4.2.3 Summary of Post-Tensioning System Aging Effects.........................................................
67 28 7.4.3 Management of Aging Effects for Post-Tensioning System....................................................
67 29 7.4.3.1 ASME Section XU, Subsection 1WL Inservice, Examination.................................................
68 30 7.4.3.2 Summary of Managemn ofAing Effects for Post-Tensioning System...................................
71 31 7.4.4 Time-Limited Aging Analysis..................................................................................
72 32 8.0
SUMMARY
AND CONCLUSIONS..............................................................................
73 33
8.1 INTRODUCTION
TO
SUMMARY
AND CONCLUSIONS........................................................................
73 34 8.2 REINFORCED CONCRETE COMPO~NNS....................................
73 36 8.4 POST-TENSIONING SYSTEM COMPONENTS.............................................................................
74 37
8.5 CONCLUSION
S............................................................................................................
76 38 FIGURE 1 - OCONEE PRESTRESSED CONCRETE CONTAINMENT............................................................
78 39 FIGURE 2 - CONTAINMENT-TENDON LOCATIONS...............................................................
............ 79 40 FIGURE 3 - ATrACwHMENTs - ANCHORAGES ACROSS LINER................................................................
80 41 FIGURE 4 - PERSONNEL HATCH............................................................................................
81 42 FIGURE 5-EQUIPMENT HATCH............................................................................................
82 43 FIGURE 6 - MECHANICAL PENETRATION - SINGLE BARRIER...............................................................
83 44 FIGURE 7-SumpPENETRATION...........................................................................................
84 45 FIGURE 8 - ELECTRICAL PENETRATION...................................................................................
85 46 FIGURE 9 - FUEL TRANSFER TUBE PENETRATION...................:........................................................
86 47 FIGURE 10 -TYPICAL POST-TENSIONED TENDON ASSEMBLY..............................................................
87 48 REFERENCES............................................................................................................
88
7 OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page vi 1
DEFINITIONS AND TERMS 2
3 Where the term "Reactor Building Containment" or "Containmnent Building" is used in this specification or any of 4
the references, either term is synonymous with the term Containment.
5 6
ACI........................ American Concrete Institute 7
AISC...................... American Institute of Steel Construction 8
AISI....................... American Iron and Steel Institute 9
AMR...................... Aging Management Review 10 AMSAC................. Anticipated Transient without Scram Mitigation System Actuation Circuitry 11 ASME.................... American Society of Mechanical Engineers 12 ASTM.................... American Society of Testing and Materials 13 ATWS....................Anticipated Transient without Scram 14 AWS...................... American Welding Society 15 B&PV...................Boiler
& Pressure Vessel 16 BBRV...................Birkenmeier Brandestinin Ros Vogt 17 BTP.......................Branch Technical Position 18 CFR.......... Code of Federal Regulations 19 CLB.......................Current Licensing Basis 20 DAC.................... Dominant Area of Concern 21 DSS....................... Diverse Scram System 22 ECGB.........
Civil Engineering Geosciences Branch 23 EQ......................... Environmental Qualification 24 EQML...................Environmental Qualification Master List 25 GDC...................... General Design Criteria 26 ILRT......................Integrated Leak Rate Test 27 IPA........................Integrated Plant Assessment 28 IR...........................Industry Report 29 ISI.......................... Inservice Inspection 30 LER.......................
Licensee Event Reports 31 LLRT.....................Local Leak Rate Test 32 LTR....................... Letter 33 LRA....................... License Renewal Application 34 MIC.......................Microbiologically Induced Corrosion 35 NEI........................Nuclear Energy Institute (formerly NUMARC) 36 NPRDS..................Nuclear Plant Reliability Data System 37 NRC....................... U. S. Nuclear Regulatory Commission 38 NSSS.....................Nuclear Steam Supply System 39 NUMARC.............Nuclear Management and Resources Council (now NEI) 40 OLT...:...................Oconee List 41 ONS.......................Oconee Nuclear Station 42 ORNL....................Oak Ridge National Laboratory 43 OSS.......................
Oconee Station Specification 44 ppm........................parts per million 45 IS........................Pressurized Thermal Shock 46 PWR..........
Pressurized Water Reactor 47 RG.........................Regulatory Guide 48 SBO.......................Station Blackout 49 SER....................... Safety Evaluation Report 50 SCC...................... Stress Corrosion Cracking 51 SSC...........Systems Structures.Components 52 TLAA................... Time-Limited Aging Analysis 53 UFSAR.................. Updated Final Safety Analysis Report
OSS-0274.00-00-0003 March 31, 1997 Revision 1 Page I 1
1.0 INTRODUCTION
2 3
The nuclear plant license renewal rule, 10 CFR 54, Requirements for Renewal of Operating 4
Licenses for Nuclear Power Plants [Ref. 1 ], describes the license renewal process and provides 5
the requirements for the contents of a license renewal application.. This rule requires that each 6
licensee determine the systems, structures and components (SSC) within the scope of the rule and 7
complete an Integrated Plant Assessment (IPA), focusing on the effects of aging on long-lived, 8
passive structures and components. The EPA, as described in 10 CFR 54.21, is an assessment by
- 9 the licensee which demonstrates that the effects of aging of the, long-lived, passive structures and 10 components within the scope of this rule will be managed during the period of extended 11I operation.
12 13 The SSC scoping and IPA for Oconee has been divided along enginee.ring discipline lines 14 traditional to Duke Power.
Dividing the Oconee license renewal EPA efforts among the 15 disciplines (Civil/Structural, Electrical and Mechanical) will facilitate the technical reviews 16 consistent with the current Oconee technical information set. The Reactor Coolant System and 17 the Reactor Building Containment, as important elements in the radioactive release line-of 18 defense, receive special focus and are handled individually.
The license renewal basis 19 specifications covering the Oconee IPA, are:
20 Specification Number Specification Title OSS-0274.00-0O-0001 Oconee Mechanical System Scoping Specification for License Renewal OSS-0274.00-00-0002 Oconee Mechanical Component Screening Specification for License Renewal OSS-0274.00-00-0003 Oconee Reactor Building Containment Aging Management Review for License Renewal OSS-0274.0O-0O-0004 Oconee Reactor Coolant System.Aging Management Review for License Renewal OSS-0274.00-00-0005 Oconee Mechanical Component Aging Management j
Review Specification for License Renewal OSS-0274.0O-00-0006 Oconee Integrated Plant Assessment (IPA) of Electrical' Components for License Renewal OSS-0274.00-00-0007 Oconee Structures and Structural Component Aging Management Review for License Renewal 21 22 In addition, the rule also focuses on time-limited aging analyses (TLAA). As defined in 10 CFR 23 54.3, these analyses are typically the boundary conditions and assumptions within the current 24 licensing basis specifically linked to 40 years of operation. Note that some TLAA resolutions are 25 contained in the above documents covering the Oconee IPA results and a clear-directory to each 26 applicable TLAA resolution is contained in OSS-0274.00-O0-0012.
In addition. to. the above 27 documents, the license renewal basis specifications covering the Oconee ThAA are:
28
OSS-0274.00-00-0003 March 31, 1997 Revision I Page 2 Specification Number Specification Title OSS-0274.00-00-0008 Time-Limited Aging Analyses. (TLAA) of Electrical Components for License Renewal OSS-0274.00-00-0011 Identification and Evaluation of Exemptions to 10 CFR for Oconee Nuclear Station OSS-0274.00-00-0012 Identification and Evaluation of Time Limited Aging Analyses Per 10 CFR 54.21(c) 2 3
4 As part of the overall IPA, this Oconee license renewal basis document provides both the 5
methodology and the technical results associated with the Reactor Building Containment IPA.
6 This methodology was developed in accordance with the guidance provided in NEI 95-10, NEI 7
95-10 (Revision 0) Industry Guideline for Implementing the Requirements of 10 CPR Part 54 8
The License Renewal Rule [Ref. 2]. TLAA associated with components of the Reactor Building 9
Containment within the scope of this document are addressed within their pertinent chapters.
10 Liner plate and penetration fatigue are addressed in Section 7.3.4. Tendon loss of prestress is 11 addressed in Section 7.4.4.
12 13 The results of the IPA and TLAA efforts are the technical bases upon which the Oconee license 14 renewal application (LRA) is built. The Oconee LRA information will be reviewed by the 15 Nuclear Regulatory Commission staff to assure compliance with 10 CFR 54. The license 16 renewal basis information in this document serves as the engineering input to this renewal 17 licensing basis. At the conclusion of the license renewal process, the information contained in 18 these license renewal basis specifications should be reviewed against and merged with the 19 Oconee engineering information, as appropriate.
I*
Priority List of Owners Group (GLRP and WOG) Items Requiring NRC Action to Support Oconee Application for a Renewal License in 1998 No.
Description Date Required
- 1.
BAW-2244 Pressurizer Topical - Review pressurizer drawings 7/1/97 and issue Safety Evaluation
- 2.
RCS Flow Nozzle Inspection Plan-Feedback from NRC 6/4/97 regarding acceptable level-of-detail
- 3.
BAW-225 1 Reactor Vessel Topical. Review status of responses 6/4/97 to RAIs 1 through 17 (Open or Closed) and Appendix D
- 4.
BAW 1543, Revision 4, Supplement 2, Master Integrated 6/30/97 Reactor Vessel Surveillance Program, (submitted August 1996)
Issue Safety Evaluation
- 5.
Supplemental Examinations--Small Bore Piping and Small Bore ASAP Nozzles-NRC and ASME initiate discussions_
- 6.
BAW-2251 Reactor Vessel Topical, GLRP responses to June 1997 RAIs 18 through 26.
- 7.
GL 97-01-Alloy 600 RPV closure head penetrations, B&WOG August 1997
/ Licensees provide responses. Staff action?
- 8.
BAW-2241P Fluence and Uncertainty Methodologies Topical
-Complete RAls July 1997
-Complete Final Safety Evaluation December 1997
- 9.
BAW-2251 Reactor Vessel Topical Safety Evaluation December 1997
- 10.
BAW-2248 Reactor Vessel Internals Topical-GLRP anticipates submittal in early June 97
-Issue RAls August 97
-Draft Safety Evaluation December 1997
-Final Safety Evaluation Early 1998
- 11.
BAW-2245, Acceptance of.-27 oF initial RTNDT for Linde 80 Pending ASTM welds.
Review/Approval
- 12.
WCAP-14222 RCS Equipment Supports, issue Safety Evaluation July 97 ATTACHMENT 5 5~/
5:7
T endorse NEI 95-10, Revision 0. Duke indicated that as discussed at the April 22,.1997, NEI-NRC senior managementrmeeting, maintaining DG-1047 as draft for trial use will not adversely impact.the Oconie lice nse renewal activities.
Duke requested that a technical review meeting be scheduled for the end' ofMa'y or early June 1997 to discuss the structural. and. electrical vertical sl'ic'e submittals.
Original sig red by:
Stephen T. Hoffman, Senipr Project Manager License Renewal Project Directorate Division of Reactor Program-Management Office of Nuclear Reactor Regulation Docket Nos. 50-269, 50-270 and 50-287 cc:
See next page R. L. Gill, Duke Power NAME: A:DUKE0508.MTS (S.HOFFMAN DISK)
To receive a copy of this document, Indicate in the box: "C" = Copy without attachment/enclosure "E"
Copy with attachment/enclosure "N"
No copy OFFICE SPM:PDLR 1E AD:PDLR I
NAME SHoffmanravl' PTKuo DATE 05/16/97 05/4/9" OFFICIAL RECORD -COPY