ML16134A487
| ML16134A487 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 11/13/1996 |
| From: | Moulton J NRC (Affiliation Not Assigned) |
| To: | NRC (Affiliation Not Assigned) |
| References | |
| TAC-M96277, TAC-M96278, TAC-M96279, NUDOCS 9611180030 | |
| Download: ML16134A487 (76) | |
Text
.
,a 8 REG UNITED STATES C
0NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-l001 November 13, 1996 ORGANIZATION: Duke Power Company
SUBJECT:
SUMMARY
OF MEETING BETWEEN THE U.S. NUCLEAR REGULATORY COMMISSION AND DUKE POWER REPRESENTATIVES CONCERNING THEIR PROPOSALS FOR IMPROVING THE FORMAT AND CONTENT FOR OLRP-1001 (TAC NOS. M96277, 96278, 96279)
On November 5, 1996, the Nuclear Regulatory Commission (NRC) staff met with representatives of the Duke Power Company to discuss their proposal for correcting the format and content deficiencies in their first license renewal technical information topical report (OLRP-1001) identified by the staff on September 18, 1996. The list of meeting attendees is contained in.
The meeting started with Mike Tuckman of Duke power explaining the goal Duke expects to achieve in their license renewal topical report review process.
Specifically, Mr. Tuckman stated that Duke is expecting to get a sense of what it will take to seek a renewed license and establish enough predictability and stability in the staff's review process such that Duke can make a decision regarding whether to seek a renewed license by early 1998.
Duke representatives then presented the staff with their proposed generic format and content specification document for their first submittal of OLRP 1001 that they believe addresses the staff's concerns expressed in September.
Additionally, Duke presented specific examples of the implementation of their format and content specification document. Duke requested that the staff review these documents and provide them feedback as to whether or not they are nearer to meeting the staff's expectations regarding format and content such that staff reviews could begin. Duke's specification document and implementation examples are contained in Attachment 2.
Tim Martin, of the NRC staff, thanked Duke for their efforts in preparing the specification document and implementation examples and stressed the importance of such efforts in assisting the staff with developing the necessary review guidance for license renewal application. Mr. Martin stressed that many of the level of detail areas are very new to the staff and as such the staff does not yet have all the answers regarding specifically what it needs to have in an application to support its reviews and therefore efforts with BGE, NEI, and Duke play an important role in the ultimate guidance that will be adopted.
Mr. Martin stated that the staff's license renewal review resources are strained but committed to get back to Duke shortly regarding what review FPR A
rO 0
November 13, 1996
-2 schedule the staff could support, possibly similar to the review being conducted currently with BGE Duke stated that they would be able to support site visits by the staff and periodic senior management meetings to assess the progress of reaching the correct format and con ent i t i ne e rs.
o
- t, Project Manager L ense Renewal Project Directorate ivision of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos.: 50-269, 50-270, and 287 Attachments:
- 1. Attendance List
- 2. Meeting Handout cc w/attachments:
See service list R. L. Gill, Duke Power
-2 schedule the staff could support, possibly similar to the review being conducted currently with BGE. Duke stated that they would be able to support site visits by the staff and periodic senior management meetings to assess the progress of reaching the correct format and content in their license reports.
John P. Moulton, Project Manager Original signed by License Renewal Project Directorate Division of Reactor Program Management Office of Nuclear Reactor Regulation Docket Nos.:
50-269, 50-270, and 287 Attachments:
- 1. Attendance List
- 2. Meeting Handout cc w/attachments:
See service list R. L. Gill, Duke Power Dlimla ltalmt:DISTRIMITIM Mi attaclmments Central File/Docket File Frae~a, 0-12 GIS PUBLIC RZimsiumn, 0-12 GI PDLR R/F G~olahan,.0-8 E2 NCase, 0-11 H21 AThadani, 0-12 G18 DLaBarge, 0-14 921 ISheron, 0-7 D26 DISTRII via -mit artin, 011 21 JNitchelt, EDO (JAN)
Jcreig, RES (JWCI)
ACUS, T-2E MWayf tld, RES (NEN2)
Nora, RES (JPV) iStech*,-15 BIB ANurphy, RES (AJN1)
Resainan, ER (RU)
TSpeis* T10 F12 Narammer, URse CML)
-sStromnidmr, NRN(JRS2)
EJ'T0 RCorrees, NR(RPC)
Gwix 666 Jllooe/EiOr, 0-1S B18 PDLR Staff Soroggitis,,OS (SCO)
GOcs 0-15 B18 LShao, RES (LCS1)
Fherny, RES CFCCI)
Dmthews, 0-10-15 Giagchi, NRR (XI)
Peterson, URR (SUP)
RJohnsn, T-10 EIO RFrahm, Jr.,
NR (RKF)
DOCUMENTAMAE: A:NTGnovg96.duk (JMo0'lton/LLM Disk)
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Attachment lI NRC MEETING WITH_DUKE POWERCMPN FORMAT AND CONTENT (OLR -1001)
November 5. 1996 NAME OGANIZATION
- 1.
John P. Moulton NRC/NRR/DRP /PDLR
- 2.
P.T.-Kuo -
NRC/NRR -/DRPM/PDLR 3..
Stv ofAn NC/R/RMP
- 4.
ScttFlnr
-NRC/NRR/DRPMPDLR
- 5.
Dbbie Ramsey DBue PoweCM/anv
- 6.
Robert Gill Dk oe oon
- 1.
Gre Rison Duke Power Company
- 8.
Mike Tuckman
-Duke Power Company
- 9.
Doug Wal-ters NE I
- 10.
Tricia Heroux for EPRI
- 11.
Barth Doroshulk BGE
- 12.
David Matthews NRC /NRR/DRPM
- 13.
Tim Marti NC /NRR/DRPM
- 14.
Richard Johnson NRC/RES/EMMEB
- 15.
- 16.
- 18.
- 19.
- 20.
- 21.
- 22.
- 23.
- 24.
- 25.
- 26.
- 28.
- 29.
- 30.
- 31.
- 32.
- 33.
- 34.
- 35.
- 36.
- 37.
- 38.
4.3.
40.1_
Oconee Nuclear Station Duke Power Company cc:
Mr. Ed Burchfield Mr. Paul R. Newton Compliance Duke Power Company, PB05E Duke Power Company 422 South Church Street Oconee Nuclear Site Charlotte, North Carolina 28242-0001 P. 0. Box 1439 Seneca, South Carolina 29679 J. Michael McGarry, III, Esquire Winston and Strawn Ms. Karen E. Long 1400 L Street, NW.
Assistant Attorney Genera]
Washington, DC 20005 North Carolina Department of Justice Mr. Robert B. Borsum P. 0. Box 629 Babcock & Wilcox Raleigh, North Carolina 27602 Nuclear Power Division Suite 525 Mr. G. A. Copp 1700 Rockville Pike Licensing - ECO50 Rockville, Maryland 20852 Duke Power Company 526 South Church Street Manager, LIS Charlotte, North Carolina 28242-0001 NUS Corporation 2650 McCormick Drive, 3rd Floor Dayne H. Brown, Director Clearwater, Florida 34619-1035 Division of Radiation Protection North Carolina Department of Senior Resident Inspector Environment, Health and U. S. Nuclear Regulatory Commission Natural Resources Route 2, Box 610 P. 0. Box 27687 Seneca, South Carolina 29678 Raleigh, North Carolina 27611-7687 Regional Administrator, Region II Mr. J. W. Hampton U. S. Nuclear Regulatory Commission Vice President, Oconee Site 101 Marietta Street, NW. Suite 2900 Duke Power Company Atlanta, Georgia 30323 P. 0. Box 1439 Seneca, South Carolina 27679 Max Batavia, Chief Bureau of RadiologKicalaHealth SouthNCarolina Department of Health and Environmental Control 2600 Bull StreetB RColumbia, South Carolina 29201 County SupervisSroof Oconee County CWalhalla, South Carolina 29621 Rule/Guidance/Submittal/Onsite Documentation Comparison Table Duke 10 CFR Part 54 NEI 95-10 Technical Onsite Information Documentation2 Submittal'
§54.4 Systems,
§3.0 Identification
§2.2jdentification Yes structures and of SSC's Within the of Systems, components within Scope of License Structures and the scope of license Renewal and Their Components Within renewal Intended Functions the Scope of License Renewal
§54.21(a)(1) and
§4.1 Identification
§§2.3 - 2.7 Yes (a)(2) Identify and of Structures and Reactor List Structures and Components Building Components Subject to an Aging RCS subject to review; Management Mechanical Describe and justify Review Electrical the methods used Structures and Structural Components
§54.21(a)(3) Aging
§4.2 Aging
§§ 3.3 - 3.7 Yes Management Management Reactor Review and Reviews Building Demonstration
- Mechanical
- Electrical
- Structures and Structural Components
§54.21 (c) Time-
§5.0 Time-Limited
§ 1.4 Time-Limited Yes Limited Aging Aging Analyses Aging Analyses and Analyses and including exemptions, list.
Exemptions; List Exemptions
§3.3-3.7 Time and evaluate for the Limited Aging period of extended
- Analyses, operation evaluations Based on guidance contained in NEI 95-10, §6.2 as well as other guidance and feedback provided.
.2 Available pursuant to the requirements contained in §54.37(a) and guidance contained in NEI 95-10, §§ 3.3, 4.4, and 5.3.
SPECIFICATION for the FORMAT and CONTENT of the OCONEE NUCLEAR STATION LICENSE RENEWAL - TECHNICAL INFORMATION TOPICAL REPORT OLRP-1001 Revision B October 23, 1996
Specification for the Format & Content of OLRP-1001 October 23, 1996 Table of Contents
- 1. INTRODUCTION 1
1.1 Purpose 1
1.2 Organization 1
1.3 CLB Changes During NRC Review of Application 1
1.4 Time-Limited Aging Analysis Review 1
1.4.1 Identification of Time-Limited Aging Analyses 1
1.4.2 Exemptions 2
1.5 References 2
- 2. INTEGRATED PLANT ASSESSMENT - STR UCTURE/COMPONENT IDENTIFICATION___________________
4 2.1 Introduction 4
2.2 Identification of Systems, Structures, and Components within the Scope of License Renewal4 2.3 Reactor Building 5
2.4 Reactor Coolant System 8
2.4.1 Reactor.Vessel 11 2.4.2 Reactor Vessel Internals 11 2.4.3 Reactor Coolant System Piping I1I 2.4.4 Once Through Steam Generators I I 2.4.5 Pressurizer 12 2.4.6 Reactor Coolant Pump Casing 12 2.4.7 Class 1 Component Supports 12 2.4.8 CRDM Motor Tube Housing 12 2.5 Mechanical Components 14 2.6 Electrical / Instrumentation & Control Components 19 2.7 Structures and Structural Components 23
- 3. INTEGRATED PLANT ASSESSMENT AGING MANAGEMENT REVIEW_
29 3.1 Introduction 29_________________
3.2 Aging Management Process Overview 29 3.3 Reactor Building 30 3.3.1 Concrete Components 31 3.3.2 Steel Components (Group 1) 32 3.3.3 Steel Components (Group 2) 32 3.3.4 Post-Tensioning System 32 3.3.5 Other Reactor Building Interior Components 32 3.3.6 Conclusion 33 3.4 Reactor Coolant System
-36 3.4.1 Reactor Vessel 37 1
Specification for the Format & Content of OLRP-1001 October 23, 1996 3.4.2 Reactor Vessel Internals 38 3.4.3 Reactor Coolant System Piping 38 3.4.4 Once Through Steam Generators 38 3.4.5 Pressurizer 39 3.4.6 Reactor Coolant Pump Casing 39 3.4.7 Class I Component Supports 39 3.4.8 CRDM Motor Tube Housing 39 3.4.9 Time-Limited Aging Analyses 40 3.4.10 Conclusion 40 3.5 Mechanical Components 43 3.5.1 Mechanical Components with Internal Service Environment of Borated Water 44 3.5.2 Mechanical Components with Internal Service Environment of Treated Water 44 3.5.3 Mechanical Components with Internal Service Environment of Raw Water 45 3.5.4 Mechanical Components with Internal Service Environment of Air or Gas 45 3.5.5 Mechanical Components with Internal Service Environment of Oil 45 3.5.6 Tanks 45 3.5.7 Heat Exchangers 46 3.5.8 Conclusion 46 3.6 Electrical / Instrumentation & Control Components 47 3.6.1 Cable (Type) 48 3.6.2 Cable (Type) 48 3.6.3 Cable (Type) 49 3.6.4 Cable (Type) 49 3.6.5 Cable (Type) 49 3.6.6 Cable (Type) 49 3.6.7 Cable (Type) 50 3.6.8 Cable (Type) 50 3.6.9 Cable (Type) 50 3.6.10 Cable Connections 50 3.6.11 Electrical Penetration Assemblies 51_I 3.6.12 Isolators 51 3.6.13 Transformers 51 3.6.14 Wire 52 3.6.15 Time-Limited Aging Analyses 52 3.6.16 Conclusion 52 3.7 Structures and Structural Components 54 3.7.1 230 kV Switchyard 55 3.7.2 Auxiliary Building 55 3.7.3 Earthen Dams / Dikes 55 3.7.4 Intake Structure 56 3.7.5 Keowee 56 3.7.6 Radwaste Facility 56 3.7.7 Spent Fuel Pools 57 3.7.8 Standby Shutdown Facility 57 3.7.9 Switchgear Enclosures 57 3.7.10 Trenches 57 3.7.11 Turbine Building 58 3.7.12 Unit Vents 58 3.7.13 Conclusion 58 11
0 0
Specification for the Format & Content of OLRP-1001 October 23, 1996
- 4. CONCLUSION 59 Ill
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
- 2. INTEGRATED PLANT ASSESSMENT 2
STRUCTURE/COMPONENT IDENTIFICATION 3
2.1 INTRODUCTION
4 Part 54, §54.21(a)(1), requires that for those systems, structures, and components within 5
the scope of this part, as delineated in §54.4, those structures and components subject to 6
an aging management review be identified and listed. Part 54, §54.21(a)(2), further 7
requires that the methods used to identify and list these structures and components be 8
described and justified.
9 10 The purpose of Chapter 2 is to provide the information required by these two sections of 11 Part 54. Section 2.2 will describe the overall plant scoping methodology which will be 12 utilized on Oconee. This scoping methodology is consistent with the guidance provided 13 in NEI 95-10, Chapter 3. The system, structure, and component scoping and Integrated 14 Plant Assessment will be divided along engineering discipline lines traditional to Duke 15 Power (e.g. Civil/Structural, Electrical, and Mechanical). The Reactor Coolant System 16 and the Reactor Building, as important elements in the radioactive release line-of-defense, 17 receive special focus and are handled individually in Sections 2.4 and 2.3 respectively.
18 Sections 2.3 through 2.7 will utilize the results of the scoping methodology to meet the 19 requirements of §§54.21(a)(1) and (a)(2) and are further described in the respective 20 sections that follow.
21 2.2 IDENTIFICATION OF SYSTEMS, STRUCTURES, AND COMPONENTS 22 WITHIN THE SCOPE OF LICENSE RENEWAL 23 A description of the methodology used to identify the Oconee systems, structures, and 24 components within the scope of license renewal and their intended functions will be 25 provided in this section of OLRP-1001. The methodology is consistent with the guidance 26 provided in NEI 95-10, §3.1.
27 28 References for this section of OLRP-1001 will be indicated by '[Reference #]' in the text, 29 listed at the end of the section, and available in the Public Document Room. Specific 30 sections of referenced material will be cited, as appropriate. Reliance upon and reference 31 to proprietary documents should be avoided, if possible.
32 4
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
2.3 REACTOR BUILDING 2
Based on the review to identify structures within the scope of license renewal as 3
described in Section 2.2, the Reactor Building has been identified as a structure within 4
the scope of license renewal.
5 6
A general description of the Oconee Reactor Buildings will be provided and a reference 7
to the Oconee UFSAR, Section 3.8 will be included in Section 2.3 of OLRP-1001. The 8
information provided in Section 2.3 of OLRP-1001 should be of sufficient detail to allow 9
a reader to gain an understanding what specific Reactor Building component groups are 10 subject to aging management review in Section 3.3 of OLRP-1001.
11 12 The evaluation boundary covered by Section 2.3 of OLRP-1001 will be shown by 13 providing simplified drawings of the Reactor Building and penetrations. The guidance 14 contained in NEI 95-10, §4.1 will be utilized to identify the components of the Reactor 15 Building and their intended functions that are subject to aging management review.
16 17 The basis for the components groupings of the Reactor Building that are subject to an 18 aging management review will be described in Section 2.3 of OLRP-1001. The current 19 basis for Reactor Building component groupings is by materials of construction, 20 component level function, and aging management program.
21 22 A table will be provided in Section 2.3 of OLRP-1001 which lists the Reactor Building 23 components and their intended functions. Table 2.3-1 of this specification is a working 24 draft of this table. Table 2.3-2 of.this specification is a working draft of a table provided 25 as an aide to the reader to identify where the component aging management review is 26 located in OLRP-1001. Identification of individual components of the Reactor Building 27 will be available in either the Oconee UFSAR or onsite documentation.
28 29 As an aid to the reader, other sections of OLRP-1001 where interfacing structures and 30 components are being addressed will be provided.
31 32 References for this section of OLRP-1001 will be indicated by '[Reference #]' in the text, 33 listed at the end of the section, and available in the Public Document Room. Specific 34 sections of referenced material will be cited, as appropriate. Reliance upon and reference 35 to proprietary documents should be avoided, if possible.
5
0 Specification for the Format & Content of OLRP-1001 October 23, 1996 1
Table 2.3-1 2
Reactor Building Components and 3
Their Intended Functions 4
(working draft) 5 6
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function eFunctions (Identified in the note below)
Concrete Components Cylinder Wall 2
3 4k5 6
Dome 2
3 5
6 Equipment Foundations Floor 2
3 5
6 7
Foundation Slab 2
3 5
6 Masonry Brick Walls Primary Shield Walls Removable Missile Shields Secondary Shield Walls Steel Components (Group 1)
Anchorage/Embedments/Attachments Electrical Penetrations Emergency Personnel Hatch Equipment Hatch Fuel Transfer Tube Liner Plate Mechanical Penetrations Steel Components (Group 2)
Cable Tray & Conduit Cable Tray & Conduit Supports Class 2 & 3 Component Supports Controlled Leakage Doors Crane Rails & Girders Electrical Racks, Panels & Cabinets Equipment Supports HVAC Duct Supports Instrument Racks Panels & Frames Jet Barriers Missile Shields Pipe Whip Restraints ISump Screens 6
Specification for the Format & Content of OLRP-1001 October 23, 1996 Table 2.3-1 (continued)
Reactor Building Components and Their Intended Functions (working draft)
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
M E M
O
=Functions (Identified in the note below)
Post Tensioning System Other Reactor Building Interior Components Lead Shielding Supports Fire Stops Shield Wall Tendons 2
3 Reactor Building Component Intended Functions:
4
- 1.
Provides essentially leaktight barrier to prevent uncontrolled release of radioactivity.
5
- 2.
Provides structural and/or functional support to safety-related SSCs. More specifically for the post-tensioning 6
system, this function means to impose compressive forces on the concrete containment structure to resist the 7
internal pressure resulting from a design basis accident with no loss of integrity.
8
- 3. Provides shelter/protection to safety-related SSCs (including radiation protection).
9
- 4.
Provides rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plant.
10
- 5. Serves as external missile barrier.
11
- 6.
Provides structural and/or functional support to non-safety related SSCs where failure of this structural component 12 could directly prevent satisfactory accomplishment of any of the required safety-related functions.
13
- 7.
Provides heat sink during design basis accidents or station blackout.
14 15 16 7
0 Specification for the Format & Content of OLRP-1001 October 23, 1996 1
2.7 STRUCTURES AND STRUCTURAL COMPONENTS 2
A description of the methodology used to identify the Oconee structures within the scope 3
of license renewal and their intended functions will be described in Section 2.2 of OLRP 4
1001.
5 6
Except for the Reactor Building, Oconee structures that are within the scope of license 7
renewal will be addressed in this Section 2.7 of OLRP-1001. The Reactor Building will 8
be addressed separately in Section 2.3 of OLRP-1001 because of the importance of the 9
structure in overall design of the plant. The information provided in Section 2.7 of OLRP 10 1001 should be of sufficient detail to allow a reader to gain an understanding of the 11 specific structures and structural component groups that are subject to aging management 12 review in Section 3.7 of OLRP-1001.
13 14 For each Oconee structure that has been identified as being within the scope of license 15 renewal, the structural components or groups of structural components contained therein 16 will be identified. The basis for the groupings of the Oconee structural components that 17 are subject to an aging management review within each structure will be described in 18 Section 2.7 of OLRP-1001. The current basis for grouping structural components is that 19 they will be grouped by material of construction, intended functions, and aging 20 management programs.
21 22 A table will be provided in Section 2.7 of OLRP-1001 which will list each Oconee 23 structure, including their structural components, that are subject to aging management 24 review. Table 2.7-1 of this specification is a working draft of this table. Further 25 identification of individual structures and structural components will be available in 26 documentation maintained onsite.
27 28 References for this section of OLRP-1001 will be indicated by '[Reference #]' in the text, 29 listed at the end of the section, and available in the Public Document Room. Specific 30 sections of referenced material will be cited, as appropriate. Reliance upon and reference 31 to proprietary documents should be avoided, if possible.
32 33 23
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
2 Table 2.7-1 3
Oconee Structures and Structural Components 4
Subject to Aging Management Review 5
(working draft) 6 7
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
Functions (Identified in the note below) 230 kV Switchyard (includes Relay House, Structure and Transmission Towers)
Anchorage Embedments Equipment Foundations Equipment Supports Masonry Block Walls Reinforced Concrete Structural Steel Auxiliary Building (excluding Spent Fuel Pools, which are listed separately)
Anchorage Battery Racks Cable Tray & Conduit Cable Tray & Conduit Supports Compressible Joints & Seals Control Boards Control Room Ceiling Controlled Leakage Doors Electrical Racks, Panels & Cabinets Embedments Equipment foundations Equipment Suppoits Expansion Anchors Fire Doors Fire Stops Fire Walls Flood Curbs (steel)
Flood Doors Hatches HVAC Duct Supports Instrument Racks, Panels & Frames Lead Shielding Masonry Block Walls Masonry Brick Walls Platform Supports 24
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
Table 2.7-1 (continued)
Structural Components and Their Intended Functions (working draft)
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
Functions (Identified in the note below)
Reinforced Concrete Sump Liners Sump Screens Vibration Isolators Earthen Dams/ Dikes Intake Canal Dike Keowee River Dam Little River Dam Underwater Weir Intake Structure Cable Tray & Conduit Electrical Racks, Panels & Cabinets Instrument Racks, Panels & Frames Trash Racks & Screens Anchorage Embedments Equipment Foundations Expansion Anchors Reinforced Concrete Compressible Joints & Seals Keowee (includes Breaker Vault, Intake Structure, Power House, Service Bay Structure, and Spillway)
Anchorage Battery Racks Cable Tray & Conduit Cable Tray & conduit Supports Class I Pipe Supports Compressible Joints & Seals Control Boards Control Room Ceiling Crane Rails & Girders Electrical Racks, Panels & Cabinets Embedments Equipment Foundations Expansion Anchors HVAC Duct Supports Instrument Racks, Panels & Frames 25
Specification for the Format & Content of OLRP-1001 October 23, 1996 Table 2.7-1 (continued)
Structural Components and Their Intended Functions (working draft)
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
Functions (Identified in the note below)
Masonry Block Walls Reinforced Concrete Structural Steel Radwaste Facility Concrete Curb Spent Fuel Pools Spent Fuel Racks Spent Fuel Pool Liner Standby Shutdown Facility Anchorage Battery Racks Cable Tray & Conduit Cable Tray & Conduit Supports Control Boards Control Room Ceiling Electrical Racks, Panels & Cabinets Embedments Equipment Foundations Equipment Supports Expansion Anchors Fire Doors Fire Walls Flood Curbs (Concrete)
Flood Curbs (Steel)
Flood Doors Hatches HVAC Duct Supports Instrument Racks, Panels, & Frames Reinforced Concrete Switchgear Enclosures Cable Tray & Conduit Cable Tray & Conduit Supports Electrical racks, Panels & Cabinets Missile Shields Anchorage Embedments 26
Specification for the Format & Content of OLRP-1001 October 23, 1996 Table 2.7-1 (continued)
Structural Components and Their Intended Functions (working draft)
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
Functions (Identified in the note below)
Equipment Foundations Expansion Anchors Masonry Block Walls Pipe Piles Reinforced Concrete Compressible Joints & Seals Trenches Reinforced Concrete Turbine Building Anchorage Battery Racks Cable Tray & Conduit Cable Tray & Conduit Supports Compressible Joints & Seals Electrical Racks, Panels & Cabinets Embedments Equipment Foundations Equipment Supports Expansion Anchors Fire Doors Fire Walls Flood Curbs (concrete)
Flood Doors Foundation Dowels Grating (QA4)
HVAC Duct Supports Impulse Line Supports Instrument Racks, Panels & Frames Masonry Block Walls Reinforced Concrete Stairs (QA4)
Structural Steel Unit Vents Anchoragre-lo 2
Structural Component Intended Functions:
3 To be developed 27
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
28
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
- 3. INTEGRATED PLANT ASSESSMENT AGING MANAGEMENT 2
REVIEW 3
3.1 INTRODUCTION
4 Part 54 requires that for each structure and component requiring an aging management 5
review, the applicant must demonstrate that the effects of aging will be adequately 6
managed so that the intended function(s) will be maintained consistent with the current 7
licensing basis for the period of extended operation. §54.21 (a)(3). Part 54 further 8
requires that identified TLAA need to be evaluated for the period of extended operation.
9
§5 4.21 (c). The structures and components that had been identified as requiring aging 10 management reviews in Chapter 2 have been evaluated to determine if the effects of aging 111 will be adequately managed for the period of extended operation.
12 13 The purpose of Chapter 3 is to provide the demonstration required by §54.21(a)(3).
14 15 In addition, TLAA that were identified in Section 1.4 are evaluated concurrently with the 16 aging effects evaluation on each respective structure, component, or component grouping.
17 The results, as required by §5 4.21(c), are provided in the applicable sections of Chapter 3.
18 3.2 AGING MANAGEMENT PROCESS OVERVIEW 19 A brief overview of the remaining sections will be provided in Section 3.2 of OLRP-1001 20 as an aid to the reader.
21 29
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
3.3 REACTOR BUILDING 2
For each Reactor Building component grouping identified in Section 2.3, the materials of 3
construction, design codes or standards used, and service environment will be described 4
in Section 3.3 of OLRP-1001. For each component grouping the applicable aging effects 5
will be identified, the aging management programs to be credited will be identified, and 6
the effectiveness to these programs will be demonstrated by providing objective 7
evidence. Upon completion of the review, a conclusion regarding aging management will 8
be provided for each component grouping. In addition, in the event that a time-limited 9
aging analysis has been identified, it will be evaluated concurrent with the applicable 10 component group. The following paragraphs provide more information as to how this 11 review will be conducted.
12 13 Applicable Aging Effects 14 The applicable aging effects will be determined by reviewing the materials of 15 construction and service environment. A brief description of how this determination of 16 applicable aging effects was made will be provided in Section 3.3 of OLRP-1001. NRC 17 generic communications will be reviewed in order to validate the identification of aging 18 effects applicable to Oconee. The NRC Bulletins, Generic Letters, and Information 19 Notices that have been determined to validate the applicable aging effects for the Oconee 20 Reactor Buildings will be listed in Section 3.3 of OLRP-1001. This method is consistent 21 with the guidance provided in NEI 95-10, §4.2.1.1.
22 23 Aging Management Programs 24 The specific programs or activities that will be credited to manage the identified aging 25 effects of concern will be described. Consistent with NEI 95-10, §4.2.1.2, the 26 description of each credited program will include:
27
- the scope of the credited program; 28
- the aging effects which will be detected before there is a loss of the structure's or 29 component's intended function; 30 the acceptance criteria contained in the program against which th e need for corrective 31 action is evaluated (identification of specific acceptance criteria is optional);
32
- corrective actions that will be taken when these acceptance criteria are not met; 33 any monitoring or trending which will provide adequate predictability and timely 34 corrective or mitigative actions; and 35 the administrative controls on the credited program.
36 If any of the above elements is not contained in the credited program, an explanation will 37 be provided in Section 3.3 of OLRP-1001.
38 39 Demonstration of Aging Management 40 For existing programs which are being credited to manage the effects of aging, objective 41 evidence from recent plant specific experience will be provided in order to demonstrate 42 that there is reasonable assurance that the credited program will manage the applicable 43 aging effects that have been identified.
30
Specification for the Format & Content of OLRP-1001 October 23, 1996 2
For new inspection programs which are intended to be credited to manage the applicable 3
effects of aging, the guidance contained in NEI 95-10, §4.3 will be followed in their 4
development. Descriptions of any new programs will be included in Section 3.3 of 5
OLRP-1001 and submitted for NRC review.and approval.
6 7
A conclusion regarding the management of applicable aging effects by programs that 8
have been demonstrated to be effective, as well as the technical basis for this conclusion, 9
will be provided for each component grouping.
10 11 Time-Limited Aging Analysis 12 A brief description of each time-limited aging analysis that has been identified and listed 13 in Section 1.4 and is applicable to the Reactor Building will be provided. This 14 description will explain why the analysis is considered to be a time-limited aging analysis 15 and make reference to appropriate docketed documents. In addition, the results of the 16 evaluation will be provided along with a conclusion covering acceptability for the period 17 of extended operation. The time-limited aging analyses associated with the Reactor 18 Building structure are presented concurrent with the aging management review of the 19 affected component. The guidance of NEI 95-10, §5.1 has been followed.
20 21 References 22 References for this section of OLRP-1001 will be indicated by '[Reference #]' in the text, 23 listed at the end of the section, and available in the Public Document Room. Specific 24 sections of referenced material will be cited, as appropriate. Reliance upon and reference 25 to proprietary documents should be avoided, if possible.
26 27 Format 28 The following sections indicate the format in which the aging management review 29 information for the Reactor Building components will be presented in OLRP-1001.
30 3.3.1 CONCRETE COMPONENTS 31 3.3.1.1 AGING MANAGEMENT REVIEW 32 3.3.1.1.1 APPLICABLE AGING EFFECTS 33 3.3.1.1.2 AGING MANAGEMENT DEMONSTRATION 34 3.3.1.2 TLAA 35 None 31
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
3.3.2 STEEL COMPONENTS (GROUP 1) 2 3.3.2.1 AGING MANAGEMENT REVIEW 3
3.3.2.1.1 APPLICABLE AGING EFFECTS 4
3.3.2.1.2 AGING MANAGEMENT DEMONSTRATION 5
3.3.2.2 TLAA 6
3.3.3 STEEL COMPONENTS (GROUP 2) 7 3.3.3.1 AGING MANAGEMENT REVIEW 8
3.3.3.1.1 APPLICABLE AGING EFFECTS 9
3.3.3.1.2 AGING MANAGEMENT DEMONSTRATION 10 3.3.3.2 TLAA 11 3.3.4 POST-TENSIONING SYSTEM 12 3.3.4.1 AGING MANAGEMENT REVIEW 13 3.3.4.1.1 APPLICABLE AGING EFFECTS 14 3.3.4.1.2 AGING MANAGEMENT DEMONSTRATION 15 3.3.4.2 TLAA 16 Loss of prestress 17 3.3.5 OTHER REACTOR BUILDING INTERIOR COMPONENTS 18 3.3.5.1 AGING MANAGEMENT REVIEW 19 3.3.5.1.1 APPLICABLE AGING EFFECTS 20 3.3.5.1.2 AGING MANAGEMENT DEMONSTRATION 21 3.3.5.2 TLAA 22 None 32
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
3.3.6
SUMMARY
AND CONCLUSION 2
This is a conclusion for the entire chapter that will summarize the results of each of the 3
component grouping aging management reviews described within the chapter. The 4
contents of this conclusion section will be reflected in the final conclusion for the report 5
contained in Chapter 4.
6 7
Aging management programs credited during the review of the Reactor Building will be 8
identified. These programs will manage the identified applicable aging effects such that 9
the structure/component intended functions previously identified will be maintained 10 consistent with the CLB during the period of extended operation. Reference will be made 11 to Table 3.3-1.
12 13 Summary descriptions of the programs listed, as well as the identification of the 14 applicable administrative control program(s), will be provided in a supplement to the 15 Oconee UFSAR at the time of application.
16 33
Specification for the Format & Content of OLRP-1001 October 23, 1996 1
Table 3.3-1 2
Aging Management Results for 3
Reactor Building Components 4
(working draft) 5 Components Subject to Applicable Aging Management Aging Management Aging Programs Review Effects Concrete Components None None required Cylinder Wall Dome Equipment Foundations Floor Foundation Slab Masonry Brick Walls Reinforced Concrete (Primary, secondary shields)
Removable Missile Shields Steel Components (Group 1)
Loss of material due to corrosion Reactor Building Civil Anchors/Embedments/
for the liner, hatches, and Inspection Program for the Attachments penetrations if the coatings Integrated Leak rate Test Electrical Penetrations are not maintained; Emergency Personnel Hatch for the liner below the Reactor Building Integrated Equipment Hatch concrete floor if the Leak Rate Test (defense-in Fuel Transfer Tubes expansion joint sealants are depth)
Liner not maintained; and Mechanical Penetrations for the liner behind welded Reactor Building Local Leak Personnel Hatch attachments if the cavity Rate Test (defense-in-depth) formed between the attachment and the liner is not sealed 34
Specification for the Format & Content of OLRP-1001 October 23, 1996 Table 3.3-1 (continued)
Aging Management Results for Reactor Building Components (working draft)
Components Subject to Applicable Aging Management Aging Management Aging Programs Review Effects Steel Components (Group 2)
To be determined To be determined Cable Tray & Conduit Cable Tray & Conduit Supports Class 2 & 3 Pipe Supports Controlled Leakage Doors Crane Rails and Girders Electrical Racks, Panels &
Cabinets Equipment Supports Grating (QA4)
HVAC Duct Supports Impulse Line Supports Instrument Racks, Panels &
frames Jet Barriers Missile Shields Pipe Whip Restraints Platform Supports Sump Screens Post-Tensioning System Loss of Material at tendon Reactor Building Tendon Tendon Anchorage anchorage Surveillance Tendon Wires Other Reactor Building Interior Components Lead Shielding Supports Fire Stops Shield Wall Tendons 2
3 4
5 6
7 35
Example License Renewal Technical Information Submittal November 4, 1996 1
- 3. INTEGRATED PLANT ASSESSMENT 2
AGING MANAGEMENT REVIEW 3
3.1 INTRODUCTION
4 Part 54, §54.21(a)(3), requires that for each structure and component requiring an aging 5
management review, the applicant must demonstrate that the effects of aging will be adequately 6
managed so that the intended function(s) will be maintained consistent with the current licensing 7
basis for the period of extended operation. Part 54, §54.21(c), further requires that identified 8
TLAA need to be evaluated for the period of extended operation. The structures and components 9
that had been identified as requiring aging management reviews in Chapter 2 have been 10 evaluated to determine if the effects of aging will be adequately managed for the period of 11 extended operation.
12 13 The purpose of Chapter 3 is to provide the demonstration required by §54.21(a)(3).
14 In addition, TLAA that were identified in Section 1.4 of this report are evaluated concurrently 15 with the aging effects evaluation on each respective structure, component, or component 16 grouping. The results, as required by §54.21(c), are provided in the applicable sections of 17 Chapter 3.
18 3.2 AGING MANAGEMENT REVIEW PROCESS OVERVIEW 19 (This section is under development and will be provided later.)
1
Example License Renewal Technical Information Submittal November 4, 1996 1
3.3 REACTOR BUILDING AGING MANAGEMENT REVIEW 2
Reactor Building component groups that are within the scope of license renewal and their 3
intended functions that require aging management reviews were identified and listed in Section 4
2.3. The approach used to perform the aging management review of these components is 5
consistent with that provided in NEI 95-10, Section 4.2 [Reference 1]. The aging management 6
review consists of identifying the applicable aging effects for each of the identified Reactor 7
Building component groups and then demonstrating the ability of programs and activities to 8
manage those effects.
9 10 The applicable aging effects that can challenge the intended functions have been identified for 11 each of the Reactor Building component groups by reviewing the materials of construction' and 12 the service environment for each component group. In order to validate the identified aging 13 effects, a review of NRC generic communications relative to Reactor Building components was 14 performed.
15 16 Identified aging effects were evaluated against the existing plant programs at Oconee. The 17 demonstration process consists of evaluating the applicable existing programs with the guidance 18 of NEI 95-10, Section 4.2 [Reference 1] and reviewing actual Oconee specific results obtained 19 from the implementation of these existing programs. These actual results provide objective 20 evidence that these existing programs are effective in managing the identified effects of aging for 21 the scope of components that are subject to aging management reviews.
22 23 If an aging effect was not found to be adequately managed by existing programs for the period of 24 extended operation, then program enhancements were identified to manage these aging effects.
25 While new programs may be required as a result of other aging management reviews, no new 26 aging management programs were identified as a result of this aging management review of 27 Oconee Reactor Building components.
28 29 The aging management review is considered complete when the credited programs provide 30 reasonable assurance that the applicable aging effects are managed so that the intended 31 function(s) will be maintained consistent with the CLB for the period of extended operation. The 32 process described in this section is intended to meet the requirements of §54.21(a)(3) and to 33 permit the staff to make the finding identified in §54.29(a).
34 35 In addition to the above aging management reviews, the time limited aging analyses associated 36 with the Reactor Building that were identified in Table 1.4-1 of this report have been evaluated, 37 and the results are presented with their respective Reactor Building components within this 38 section of the report.
39 40 (The example material being submitted includes only certain Reactor Building components. The 41 remaining components of the Reactor Building are still be evaluated and will be provided later.)
2
Example License Renewal Technical Information Submittal November 4, 1996 1
3.3.1 CONCRETE COMPONENTS 2
Reactor Building concrete components are exposed to different service environments depending 3
on their location. The Reactor Building concrete foundation slab and the portion of the external 4
cylinder wall below grade are exposed to backfill and groundwater. The groundwater chemistry 5
plays a major role in the determination of the degradation of the below grade components.
6 External surfaces of the Reactor Building dome and cylinder wall above grade are exposed to the 7
external atmospheric environment. The cylinder wall above grade enclosed by adjacent buildings 8
is exposed to controlled environment which protects it from external weather and temperature 9
changes.
10 11 The top of the concrete floor is exposed to the internal environment of the Reactor Building.
12 High temperature, humidity and radiation play a role in the potential degradation of the concrete 13 components located within this environment.
14 15 The codes and standards used for the Reactor Building design and fabrication including 16 applicable edition are given in the Oconee UFSAR, Section 3.8.1 [Reference 2]. The Reactor 17 Building concrete component design complies with the ACI 3.18-63 [Reference 3]. The concrete 18 and reinforcing design parameters are identified in the Oconee UFSAR Chapter 3.8, 19
[Reference 2].
20 3.3.1.1 Aging Management Review 21 3.3.1.1.1 APPLICABLE AGING EFFECTS 22 The aging effects that could potentially result in loss of the Oconee Reactor Building concrete 23 component intended functions are: loss of material, cracking, and change in material properties 24
[Reference 4]. Each aging effect has been assessed for the Oconee Reactor Building concrete 25 components. In addition, a review of NRC generic communications was performed to validate 26 the applicable aging effects. The results of these evaluations are provided in the following 27 paragraphs and are summarized in Table 3.3-1..
28 3.3.1.1.1.1 Loss OF MATERIAL ASSESSMENT 29 Loss of material includes scaling, spalling, pitting, and erosion in concrete components which are 30 described and illustrated in ACI 201.1 [Reference 5]. Aging mechanisms and stressors that can 31 lead to loss of material include freeze-thaw, abrasion and cavitation, elevated temperature, 32 aggressive chemicals, or corrosion of embedded steel (rebar) [Reference 4].
33 34 Loss of material has been assessed and determined to be not applicable for the Oconee Reactor 35 Building concrete components. The Oconee concrete components are designed in accordance 36 with ACI 318-63 [Reference 6] and constructed'in accordance with ACI 301 using ingredients 37 conforming to ACI and ASTM standards which provide a good quality, dense, low permeability 38 concrete [Reference 7]. In addition, the components are located in environments, both above 3
Example License Renewal Technical Information Submittal November 4, 1996 1
and below grade, where they are not exposed to continuously flowing water, nor to temperatures 2
which exceed the thresholds for degradation identified in ACI 318-63 [Reference 6].
3 Groundwater chemical concentrations have been tested at Oconee and it has been determined that 4
they do not exceed the minimum threshold limits for degradation to occur which are described in 5
the NUMARC PWR Containment Industry Report [Reference 4].
6 3.3.1.1.1.2 CRACKING ASSESSMENT 7
Cracking is manifested in concrete components as a complete or incomplete separation of the 8
concrete into two or more parts. Aging mechanisms and stressors that can lead to cracking 9
include freeze-thaw, reaction with aggregates, shrinkage, settlement, elevated temperature or 10 fatigue [Reference 5].
11 12 Cracking has been assessed and determined to be not applicable for the Oconee Reactor Building 13 concrete components. The concrete components are designed in accordance with ACI 318-63 14
[Reference 6] and constructed in accordance with ACI 301 [Reference 7] using ingredients 15 conforming to ACI and ASTM standards which provide a good quality, dense, low permeability 16 concrete. During initial construction, concrete component constituents were carefully selected to 17 mitigate aggregate reactions. Tests were performed on the aggregate contained in the Oconee 18 Reactor Building which indicated that they were not considered potentially reactive. Shrinkage 19 in concrete is not a possible aging effect after 20 years [Reference 8]. The Oconee Reactor 20 Buildings are founded on bed rock thus precluding settlement. The concrete components are not 21 exposed to temperatures which exceed the thresholds for degradation due to elevated temperature 22 identified in ACI 318-63 [Reference 6]. Finally, a review of the loadings that concrete 23 containments experience during normal operation indicates that the periodic Type A Integrated 24 Leak Rate tests are the major sources of load changes. However, the number of cycles are 25 generally low for a 60-year operating life with only low to moderate stress levels. Cracking is 26 not a significant aging effect [Reference 4].
27 28 3.3.1.1.1.3 CHANGE IN MATERIAL PROPERTIES ASSESSMENT 29 The change in material properties aging effect is manifested in concrete components as increased 30 permeability, increased porosity, reduction in pH, reduction in tensile strength, reduction in 31 compressive strength, reduction in modulus of elasticity, and reduction in bond strength. Aging 32 mechanisms and stressors which can lead to change in material properties include leaching of 33 calcium hydroxide, elevated temperature, aggressive chemical attack, or irradiation 34 embrittlement.
35 36 Change in material properties has been assessed and determined to be not applicable for the 37 Oconee Reactor Building concrete components for the following reasons. The Oconee Reactor 38 Building concrete components are designed in accordance with ACI 318-63 [Reference 6] and 39 constructed in accordance with ACI 301 [Reference 7] using ingredients conforming to ACI and 40 ASTM standards which provide a good quality, dense, low permeability concrete. A dense 4
Example License Renewal Technical Information Submittal November 4, 1996 1
concrete with a suitable cement content that has been well cured is less susceptible to calcium 2
hydroxide leaching from percolating water because of its low permeability and low absorption 3
rate.
4 5
The concrete components are not exposed to temperatures which exceed the thresholds for 6
degradation identified in ACI 318-63 [Reference 6], nor are they exposed to chemical 7
concentrations which exceed the minimum threshold limits where degradation may occur.
8 Change in material properties due to irradiation embrittlement is not an applicable aging effect 9
because the concrete components will not experience sufficient irradiation to cause 10 embrittlement.
11 3.3.1.1.1.4 INDUSTRY EXPERIENCE 12 In order to validate the set of applicable aging effects and to assure no additional aging effects 13 beyond those discussed herein, a review of industry experience was performed. This review 14 included a survey of NRC generic communications and NRC NUREG's. No concrete 15 component aging effects were identified in NRC generic communications. The following 16 NUREG's were reviewed:
17 18
- NUREG/CR-6424, Report on Aging of Nuclear Power Plant Reinforced Concrete Structures 19
- NUREG/CR-4652, Concrete Component Aging and its Significance Relative to Life 20 Extension of Nuclear Power Plants 21
- NUREG-1522, Assessment of Inservice Conditions of Safety-Related Nuclear Plant 23 Structures 24
- NUREG-1557, Summary of Technical Information and Agreements from Nuclear 25 Management and Resources Council Industry Reports Addressing License Renewal 26 27 These studies conclude that the concrete deterioration has generally been minor due to the high 28 quality of the original construction. Most instances related to degradation occurred early in the 29 life of the structure and have been corrected. Degradation was due to either improper material 30 selection, construction/design deficiencies or environmental effects.
31 32 Based upon the review of industry information, LER's and NRC Generic Communications, no 33 additional aging effects beyond those evaluated in the preceding sections were identified for the 34 Reactor Building concrete components.
35 3.3.1.1.2 AGING MANAGEMENT DEMONSTRATION 36 From review of the Oconee Reactor Building design parameters, concrete component specific 37 materials of construction, service environment and aging effects that could possibly affect the 38 concrete components, no applicable aging effects were identified for Reactor Building concrete 39 components that could result in the loss of the intended functions previously identified in table 5
Example License Renewal Technical Information Submittal November 4, 1996 1
3.3-1; therefore, no aging management programs are necessary to manage the Reactor Building 2
concrete components for the period of extended operation.
3 3.3.1.2 TLAA 4
No time-limited aging analyses associated with the Reactor Building concrete components were 5
identified.
6 3.3.2 STEEL COMPONENT (GROUP 1) AGING MANAGEMENT REVIEW 7
The steel components of the Reactor Building are exposed to various service environments 8
depending on their location. The liner plate and other steel components are exposed to the 9
internal environment of the Reactor Building. High temperature, humidity and radiation within 10 the interior of the Reactor Building play a role in the degradation of the components located 11 within this environment.
12 13 The codes and standards used for the Reactor Building design and fabrication including 14 applicable edition are given in the Oconee UFSAR, Section 3.8.1 [Reference 2]. The Reactor 15 Building steel component design complies with the American Society of Mechanical Engineers 16 (ASME Section III - 1965) for the pressure boundary, the American Instituteof Steel 17 Construction (AISC, sixth edition) for the structural steel, and the American Welding Society 18 (AWS).
19 3.3.2.1 Aging Management Review 20 3.3.2.1.1 APPLICABLE AGING EFFECTS 21 The aging effects that could potentially result in loss of the Reactor Building steel component 22 intended functions are loss of material, cracking, and change in material properties [Reference 4].
23 Each aging effect has been assessed for the Oconee Reactor Building steel components. In 24 addition, a review of NRC generic communications was performed to validate the applicable 25 aging effects. The results of these evaluations are provided in the following paragraphs.
26 3.3.2.1.1.1 Loss OF MATERIAL ASSESSMENT 27 Loss of material in the Reactor Building steel components may be caused by corrosion of the 28 steel. Exposed steel components are coated for corrosion protection, therefore, loss of material 29 due to corrosion is not an applicable aging effect as long as the coatings are maintained. Coatings 30 for the liner, attachments to the liner, penetrations and hatches are identified in the Oconee 31 UFSAR, Table 3-12 [Reference 2]. For the liner behind miscellaneous welded attachments, loss 32 of material due to corrosion is an applicable aging effect if the cavity formed between the 33 attachment and the liner is not sealed to protect against moisture intrusion. For steel components 34 encased in concrete, loss of material due to corrosion is not an applicable aging effect because 35 the adjacent concrete provides an alkaline environment that is an effective inhibitor of corrosion.
36 The only exception to this conclusion may be the liner below the floor if the expansion joint 37 sealant is not maintained.
6
Example License Renewal Technical Information Submittal November 4, 1996 2
3 3.3.2.1.1.2 CRACKING ASSESSMENT 4
In general, cracking of steel components may be caused by stress corrosion and fatigue.
5 However, cracking due to stress corrosion is not an applicable aging effect for the steel 6
components of the Oconee Reactor Building because the conditions necessary (corrosive 7
environment, susceptible material, and tensile stresses) for stress corrosion cracking do not exist.
8 Cracking due to fatigue is not an applicable aging effect for the steel components of the Oconee 9
Reactor Building. This aging effect has been identified as a time-limited aging analysis and has 10 been evaluated for the period of extended operation. The results of this evaluation are presented 11 in Section 3.3.2.2.1 of this report. This fatigue evaluation for the liner and penetrations for 60 12 years has determined that the original analysis remains valid for the period of extended operation.
13 Furthermore, the design and operation of the steel components will not exceed 2 x106 loadings 14 as specified by the American Institute of Steel Construction [Reference 9].
15 16 3.3.2.1.1.3 CHANGE IN MATERIALS PROPERTIES ASSESSMENT 17 The change in material properties aging effect is manifested in steel components as a reduction or 18 increase in yield strength, reduction in modulus of elasticity, reduction in ultimate tensile 19 ductility, and an increase in ductile-to-brittle transition temperature. Aging mechanisms and 20 stressors which can lead to change in material properties include elevated temperature and 21 irradiation embrittlement. Change in material properties due to elevated temperature is not an 22 applicable aging effect because the steel components are not exposed to temperatures above the 23 threshold where material property changes would occur. Temperatures are monitored for various 24 locations throughout the Oconee Reactor Buildings and are below 130 0 F. Structural steel has a 25 high temperature threshold before significant strength reductions occur. Temperatures as high as 26 7000 F must be reached before small reductions in material properties occur [Reference 4].
27 28 Change in material properties due to irradiation embrittlement is not an applicable aging effect 29 because the steel components of the Oconee Reactor Building will not experience irradiation 30 above the threshold necessary to cause embrittlement. The primary shield wall and the concrete 31 pedestal under the reactor vessel provide shielding. In addition, the distance between the steel 32 components and reactor vessel provides a further reduction in the irradiation levels at the steel 33 components of the Oconee Reactor Building [Reference 4].
34 3.3.2.1.1.4 INDUSTRY EXPERIENCE 35 In order to validate the set of applicable aging effects and to assure no additional aging effects 36 exist beyond those discussed herein, a review of industry experience was performed. This review 37 included a search of NRC generic communications and NUREG's. The following documents 38 were identified in this search:
39 7
Example License Renewal Technical Information Submittal November 4, 1996 1
- NUREG/CR-6424, Report on Aging of Nuclear Power Plant Reinforced Concrete Structures 2
- ORNL/NRC/LTR-95/29, Degradation Assessment Methodology for Application to Steel 3
Containments and Liners of Reinforced Concrete Structures in Nuclear Power Plants 4
- LER's associated with corrosion, cracking and change in material properties 5
- IEB 80-08, "Examination of Containment Liner Penetration Welds" 6
- IN 86-99, "Degradation of Steel Containments" 7
- NUREG-1540, BWR Steel Containment Corrosion 9
- NUREG-1557, Summary of Technical Information and Agreements from Nuclear 10 Management and Resources Council Industry Reports Addressing License Renewal 11 12 As a result of the review of industry data and NRC generic communications, no additional aging 13 effects beyond those discussed in this section have been observed. Therefore, the applicable 14 aging effect for the period of extended operation for the steel components of the Reactor 15 Building is loss of material due to corrosion for the liner, hatches, and penetrations if the coatings 16 are not maintained; for the liner below the concrete floor if the expansion joint sealants are not 17 maintained; and for the liner behind welded attachments if the cavity formed between the 18 attachment and the liner is not sealed.
19 3.3.2.1.2 AGING MANAGEMENT DEMONSTRATION 20 Three existing programs are being credited in license renewal to manage the potential loss of 21 material due to corrosion of the Reactor Building steel components for the period of extended 22 operation. The primary aging management program credited is the Reactor Building Civil 23 Inspection for Integrated Leak Rate Test. Supplementing this program and providing defense in 24 depth are two leak rate test programs: the Reactor Building Type A Integrated Leak Rate Test 25 and the Reactor Building Type B Local Leak Rate Test. All three of these programs are 26 described further in the following paragraphs.
27 28 3.3.2.1.2.1 REACTOR BUILDING CIVIL INSPECTION FOR INTEGRATED LEAK RATE TEST 29 Appendix J to 10 CFR Part 50 [Reference 10] contains the requirement to perform a Reactor 30 Building visual inspection of the accessible interior and exterior surfaces which includes the 31 interior surface of the containment liner, including attachments to the liner, the steel penetrations 32 and the equipment and personnel hatches. This inspection must be performed prior to each 33 Integrated Leak Rate Test. Appendix J,Section V., provides limited guidance for implementing 34 an effective inspection. However, additional inspection guidance has been provided in ASME 35 Section XI, Subsection IWE [Reference 11]. In addition, licensees are required to implement the 36 inservice examinations specified for the first period of the first inspection interval in Subsection 37 IWE of the 1992 Edition with the 1992 Addenda in conjunction with the modifications in 38
§50.55a(b)(2)(ix) by September 9, 2001 [Reference 12].
39 8
Example License Renewal Technical Information Submittal November 4, 1996 1
In response to the original requirement of Appendix J,Section V., the Oconee Reactor Building 2
Civil Inspection Program was established when Oconee was initially licensed. The existing 3
Oconee Reactor Building Civil Inspection for Integrated Leak Rate Test has been evaluated 4
against the aging management program elements identified in NEI 95-10, Section 4.2.1.3 5
[Reference 1]. This inspection program provides a visual examination of accessible interior and 6
exterior surfaces of the containment structure and its components. Any corrosion that could 7
affect the structural integrity and leaktightness of the Reactor Building is required to be 8
documented. Accessible interior and exterior concrete surfaces, liner plate, mechanical 9
penetrations, electrical penetrations, equipment hatch, personnel hatches, tendon anchorages, and 10 tendon gallery are required to be inspected.
11 12 The aging effects that will be detected by this inspection program include loss of material due to 13 corrosion of the liner, airlocks, penetrations, hatch, and attachments; and cracked or defective 14 moisture barriers. Acceptance criteria are provided for metallic surfaces and moisture barriers in 15 the plant procedures which implement this inspection requirement.
16 17 Repairs to affected surfaces are made to restore the component where necessary such that it can 18 continue to perform its intended function. Previously identified deterioration is monitored and 19 confirmation of corrective action is provided.
20 21 The Oconee Reactor Building Civil Inspection for Integrated Leak Rate Test is implemented by 22 written procedures that are maintained in accordance with administrative controls which are 23 summarized in the Oconee UFSAR, Section 13.5 [Reference 2]. As result of NRC promulgation 24 of IWE and other reviews, existing procedures are being reviewed to determine what additional 25 enhancements would be appropriate to provide further assurance that the steel components of the 26 Reactor Building are properly inspected and timely actions are taken in order to maintain the 27 identified component intended functions under CLB design conditions.
28 29 A total of 18 Reactor Building visual inspections have been performed at Oconee through 1996.
30 Results of the Reactor Building visual inspections have revealed only minor degradation.
31 Observations include some minor local coating failures including peeling or flaking. A small 32 local area of corrosion on the liner was observed at the interface of the base slab and the liner in a 33 few areas, particularly under the equipment hatch. Minor local corrosion on a few isolated 34 welded attachments to the liner and on a few penetrations has been observed. Coating 35 degradation and minor local corrosion on steel components were repaired by an existing coatings 36 maintenance procedure applicable to steel surfaces inside the Reactor Building. The observed 37 aging effects were relatively minor and had no impact on the essentially leaktight barrier and 38 other intended functions, previously identified in Table 2.3-1, of the Reactor Building.
39 40 In addition to the Reactor Building visual inspection, the passive long-lived components that are 41 part of the essentially leaktight barrier have been successfully leakrate tested numerous times 42 under existing procedures and indicate no degradation of the essentially leaktight barrier.
9
Example License Renewal Technical Information Submittal November 4, 1996 1
NUREG-1540 [Reference 13] states that inspection mandated by Appendix J to 10 CFR 50, 2
though basically visual, has been reasonably effective in identifying containment problems 3
known to date. The existing Oconee Reactor Building Civil Inspection for Integrated Leak Rate 4
Test program is able to detect and manage loss of material due to corrosion in accessible steel 5
components so that the intended functions will be maintained consistent with the CLB.
6 7
In summary, the above evaluation and Oconee operating experience have shown that loss of 8
material due to corrosion is an applicable aging effect:
9 10 (1) for the liner, hatches, and penetrations if the coatings are not maintained; 11 (2) for the liner below the concrete floor if the expansion joint sealants are not 12 maintained; and 13 (3) for the liner behind welded attachments if the cavity formed between the attachment 14 and the liner is not sealed.
15 16 Plant operating experience has shown that periodic visual inspections can identify areas where 17 loss of material is occurring and corrective actions can be taken to repair the affected area such 18 that the essentially leaktight barrier and structural support will be maintained during the period of 19 extended operation. Furthermore, additional assurance that the leaktight barrier and structural 20 support will be maintained will be provided as the requirements of ASME Section XI, 21 Subsection IWE are implemented at Oconee, as required by §50.55a(b)(2). For these reasons, 22 Duke Power has determined that the Reactor Building Civil Inspection Program for the 23 Integrated Leak Rate Test, in conjunction with the enhancements of ASME Section XI, 24 Subsection IWE which will be implemented by September 9, 2001, provide reasonable 25 assurance for managing the loss of material of the above locations of the steel components of the 26 Reactor Building, such that the essentially leaktight barrier, structural support, and heat sink 27 intended functions (previously identified in Table 2.3-1) of the Reactor Building will be 28 maintained during the period of extended operation.
29 30 3.3.2.1.2.2 REACTOR BUILDING TYPE A INTEGRATED LEAK RATE TEST 31 The Type A Integrated Leak Rate Test (ILRT) provides defense in depth to detect severe 32 corrosion of the Reactor Building steel components that causes a breech of the pressure 33 boundary. The Type A ILRT has been evaluated. against the effective aging management program 34 elements identified in NEI 95-10, Section 4.2.1.3 [Reference 1].
35 36 The Type A ILRT measures the leak rate of the Reactor Building containment under conditions 37 as prescribed in 10 CFR 50, Appendix J.[Reference 10] Pressure boundary components 38 including the liner, penetrations, and hatches are tested to detect loss of material, cracking or 39 severe corrosion that breach the pressure boundary. Acceptable leakage rates are established in 40 Oconee Technical Specifications [Reference 14] and corrective actions are taken in accordance 41 with the requirements of 10 CFR 50, Appendix J [Reference 10].
10
Example License Renewal Technical Information Submittal November 4, 1996 1
2 A total of 21 Type A ILRTs have been performed for Oconee through 1996. Results have shown 3
that all steel components have successfully passed the Type A ILRT. The Oconee Type A ILRT 4
is implemented by written procedures that are maintained in accordance with administrative 5
controls which are summarized in the Oconee UFSAR, Section 13.5 [Reference 2].
6 3.3.2.1.2.3 REACTOR BUILDING TYPE B LOCAL LEAK RATE TEST 7
The Type B Integrated Leak Rate Test (LRT) provides defense in depth to.detect degradation of 8
penetrations and hatches, as required by 10 CFR 50, Appendix J[Reference 10]. The Type B 9
LLRT has been evaluated against the effective aging management program elements identified in 10 NEI 95-10 [Reference 1].
11 12 The Type B TLRT measures the leak rate of the pressure boundary components under conditions 13 as prescribed in 10 CFR 50, Appendix J.[Reference 10] Pressure boundary components 14 including penetrations and hatches are inspected and tested for loss of material, cracking or 15 severe corrosion that breach the pressure boundary. Acceptable leakage rates are established and 16 corrective actions are taken in accordance with the requirements of 10 CFR 50, Appendix J 17
[Reference 10].
18 19 The Oconee Type B ILRT is implemented by written procedures that are maintained in 20 accordance with administrative controls which are summarized in the Oconee UFSAR, Section 21 13.5 [Reference 2].
22 23 Numerous Type B LLRTs have been performed at Oconee in over 20 years of operation. Results 24 of previous Type B tests have shown few failures. When test failures have occurred, they are 25 attributed to failure of non-metallic components (gaskets, O-rings). Results have shown no test 26 failures of steel components during the Type b LLRT.
27 28 Based on the above evaluation, the Oconee Type A ILRT and Type B LLRT, compliment the 29 Oconee Reactor Building Civil Inspection for Integrated Leak Rate Test and provide additional 30 assurance that the steel components of the Reactor Building that form the essentially leaktight 31 barrier will be maintained during the period'of extended operation.
32 3.3.2.2 Time-Limited Aging Analysis 33 3.3.2.2.1 LINER PLATE 34 The interior surface of the Reactor Building is lined with welded steel plate to provide an 35 essentially leak tight barrier. At all penetrations, the liner plate is thickened to reduce stress 36 concentrations. Design criteria are applied to the liner to assure that the specified leak rate is not 37 exceeded under design basis accident conditions. The following fatigue loads, as described in 38 the Oconee UFSAR, Section 3.8.1.5.3 [Reference 2], were considered in the design of the liner 11
Example License Renewal Technical Information Submittal November 4, 1996 1
plate and are considered to be time-limited aging analyses (TLAA) for the purposes of license 2
renewal:
3 4
(a) Thermal cycling due to annual outdoor temperature variations. Number of cycles for this 5
loading is 40 cycles for the plant life of 40 years.
6 7
(b) Thermal cycling due to Reactor Building interior temperature varying during the startup 8
and shutdown of the Reactor Coolant System. The number of cycles for this loading is 9
assumed to be 500 cycles.
10 11 (c) Thermal cycling due to the loss-of-coolant accident will be assumed to be one cycle.
12 13 (d) Thermal load cycles in the piping systems are somewhat isolated from the liner plate 14 penetrations by concentric sleeves between the pipe and the liner plate. The attachment 15 sleeve is designed in accordance with ASME Section III considerations. All penetrations 16 are reviewed for a conservative number of cycles to be expected during the plant life.
17 18 Each of the above four TLAA have been evaluated for continued operation for up to 60 years.
19 For item (a), an increase in the number of thermal cycles due to annual outdoor temperature 20 variations from 40 to 60 cycles is considered to be insignificant in comparison to the assumed 21 500 thermal cycles due to Reactor Building interior temperature varying during heatup and 22 cooldown of the Reactor Coolant System. Thus, this TLAA is considered to be valid for the 23 period of extended operation as it is enveloped with item (b) above.
24 25 For item (b), with respect to the assumed 500 thermal cycles due to startup and shutdown of the 26 Reactor Coolant System, a more limiting number of thermal cycles is contained in the Oconee 27 UFSAR, Section 5.2 [Reference 2] for actual plant operation. Oconee UFSAR, Table 5.2 28
[Reference 2] indicates a design limit of 360 heatup cycles and 360 cooldown cycles for the 29 Reactor Coolant System. The projected number of cycles for each Oconee unit through 60 years 30 of operation has been determined to be less than the original 360 cycle design limits. This TLAA 31 is considered to be valid for the period of extended operation because actual operating cycle 32 values fall well within the assumed 500 thermal cycles due to startup and shutdown of the 33 Reactor Coolant System.
34 35 For item (c), the assumed value for thermal cycling due to loss-of-coolant accident remains valid.
36 None have occurred and none are expected to occur. This TLAA is considered to be valid for the 37 period of extended operation.
38 39 Finally for item (d), the design of the Reactor Building penetrations has been reviewed. The 40 designs meet the general requirements of ASME Code [Reference 15]. The only high 41 temperature lines penetrating the Reactor Building wall and liner plate are the feedwater and 42 main steam lines. The design number of thermal load cycles in these two systems is bounded by 12
Example License Renewal Technical Information Submittal November 4, 1996 1
the number of design heatup and cooldown cycles of the Reactor Coolant System. The projected 2
number of cycles for each Oconee unit through 60 years of operation has been determined to be 3
less than these original design limits. Thus, based on a review of the existing fatigue analysis, 4
this TLAA is considered to be valid for the period of extended operation.
5 6
In conclusion, the existing analyses addressing thermal fatigue of the Reactor Building liner plate 7
and penetrations are considered to be valid for the period of extended operation. The evaluation 8
of these TLAA will be contained in the FSAR Supplement at the time of application, as required 9
by §54.21(d).
10 3.3.2.2.2 POLAR CRANE 11 This information is being developed and will be provided later.
12 13 3.3.3 STEEL COMPONENTS (GROUP 2) AGING MANAGEMENT REVIEW 14 This information is being developed and will be provided later.
15 16 3.3.4 POST-TENSIONING SYSTEM AGING MANAGEMENT REVIEW 17 The Oconee Reactor Building design incorporates a post-tensioning system that provides 18 prestress forces to counteract forces resulting from the design loads. The post-tensioning system 19 components are exposed to external atmosphere environment or bulkfill grease environment with 20 the exception of the tendon end caps enclosed by adjacent buildings which are exposed to 21 controlled environments which protect them from external weather and temperature changes. The 22 post-tensioning system components of the Reactor Building within the scope of license renewal 23 are listed in Table 3.3-1.
24 25 The codes and standards used for the Reactor Building design and fabrication including 26 applicable edition are given in the Oconee UFSAR, Section 3.8.1 [Reference 2]. The design of 27 the Reactor Building post-tensioning buttresses and anchorage zone complies with ACI 318-63 28
[Reference 6].
29 3.3.4.1 Aging Management Review 30 3.3.4.1.1 APPLICABLE AGING EFFECTS 31 The aging effect that could potentially result in loss of the ability of the post-tensioning system to 32 impose compressive forces on the concrete containment structure is loss of material
- 33.
[Reference 4]. Loss of material in the post-tensioning system components is due to corrosion.
34 The effects of corrosion must be considered for both the tendon wires within the grease-filled 35 conduits and for the anchorage providing the tendon wire terminations. Most corrosion-related 36 failures of prestressing tendons have been attributed to pitting, stress corrosion cracking, 37 hydrogen embrittlement, or some combination of these aging effects [Reference 16].
Stressed 13
Example License Renewal Technical Information Submittal November 4, 1996 1
components of the post-tensioning system are normally well protected against corrosion. The 2
tendon and the anchorage are enclosed within the ducts and end caps that are filled with bulkfill 3
grease. Potential grease leakage could occur and would be most likely at the tendon anchorage.
4
[Reference 4]. Loss of material due to corrosion may be an applicable aging effect for the tendon 5
anchorage if the grease is hot maintained.
6 7
8 3.3.4.1.2 INDUSTRY EXPERIENCE 9
In order to validate the set of applicable aging effects and to assure no additional aging effects 10 beyond those discussed herein, a review of industry experience was performed. This review 11 included a search of NRC generic communications and NRC NUREG's. The following 12 documents were identified during this investigation:
13 14
- NUREG/CR-4652, Concrete Component Aging and its Significance Relative to Life 15 Extension of Nuclear Power Plants 16
- NUREG/CR-6424, Report on Aging of Nuclear Power Plant Reinforced Concrete 17 Structures 18
- LER's associated with changes in liftoff force and corrosion.
19
- IN 85-10, "Post-Tensioned Containment Tendon Anchor Head Failure" 20
- IN 91-80, "Failure of Anchor Head Threads on Post-Tensioning System during 21 Surveillance Inspection" 22
- NUREG-1557, Summary of Technical Information and Agreements from Nuclear 23 Management and Resources Council Industry Reports Addressing License Renewal 24 25 The documents identified tendon loss of prestress and corrosion as applicable aging effects. Loss 26 of prestress is a TLAA and is evaluated in Section 3.3.4.2. Of the over 30 million tendons used 27 throughout the western world (to 1978), the number of corrosion incidents (200 in completed 28 permanent structures) is small. All of the corrosion-related incidents were related to either ill 29 conceived detailing, poor construction, or contaminants causing corrosive 30 environments[Reference 17].
31 32 As a result of the review of industry data and NRC generic communications, no additional aging 33 effects beyond those discussed in this section have been observed. Therefore, loss of material 34 due to corrosion was determined to be an applicable aging effect for the period of extended 35 operation for the tendon anchorage if the grease is not maintained. Moisture intrusion and 36 leakage of the anti-corrosion grease may cause the corrosion. Material loss at the tendon 37 anchorage can ultimately lead to tendon failure if the corrosion progresses to the point of 38 cracking of the tendon anchorage. Loss of prestress is a TLAA and is evaluated in Section 39 3.3.4.2.
40 3.3.4.1.3 AGING MANAGEMENT DEMONSTRATION 14
Example License Renewal Technical Information Submittal November 4, 1996 1
In the past, Oconee Technical Specification 4.4.2 provided specific surveillance requirements for 2
the Reactor Building tendons. Oconee has submitted a proposed technical specification change 3
[Reference 18] that would implement the guidance contained in NRC Regulatory Guide 1.35 4
[Reference 19]. The proposed change to the Oconee Technical Specifications is expected to be 5
approved in early 1997. In addition, all licensees are required to implement the inservice 6
examinations which correspond to the number of years of operation which are specified in 7
Subsection IWL of the 1992 Edition with the 1992 Addenda in conjunction with the 8
modifications specified in §50.55a(b)(2)(ix) by September 9, 2001 [Reference 12].
9 3.3.4.1.3.1 REACTOR BUILDING TENDON SURVEILLANCE PROGRAM 10 The Oconee Technical Specifications have in the past and will in the future require surveillance, 11 testing, and trending of the post-tensioning system. The previous program consisted of periodic 12 inspections of nine pre-designated tendons for each unit - three hoop, three vertical and three 13 dome tendons. In the future, a representative sample of tendons will be selected, pursuant to the 14 guidance of Regulatory Guide 1.35 [Reference 19] for testing. The program focuses on the 15 condition of the tendon wires, grease, anchorage and adjacent concrete, trending of prestress 16 force, and tensile testing of wire samples. The program assesses the condition and functional 17 capability of the system, and therefore, verifies the adequacy of the system and provides an 18 opportunity to take proper corrective action should adverse conditions be detected.
19 20 The Reactor Building Tendon Surveillance Program has been evaluated for the aging 21 management program elements identified in NEI 95-10, Section 4.2 [Reference 1]. The purpose 22 of the Reactor Building Tendon Surveillance Program is to demonstrate the integrity of the 23 prestressing system including the tendons, tendon end anchorage hardware, general and adjacent 24 concrete, and the corrosion protection (grease) system.
25 26 Following NRC approval of the proposed change to the Oconee Technical Specifications, the 27 guidance of Regulator Guide 1.35 [Reference 19] will be followed to identify which tendons will 28 be selected for inspection. Components inspected include tendon anchorage. and end caps, 29 general exterior accessible concrete surfaces on dome and cylinder and concrete surface adjacent 30 to the surveillance tendons, tendon wires and grease. The program is intended to identify loss of 31 material, including section loss due to wire breakage or corrosion, leakage of corrosion 32 protection grease, contamination of grease, loss of prestress, corrosion of anchorages, and 33 cracking of adjacent concrete.
34 35 Specific acceptance criteria for monitoring prestress force loss (including meeting requirements 36 for minimum required force levels), condition of protective grease, corrosion on wire and 37 anchorages, and tensile testing of wire samples are provided. Previously identified conditions are 38 monitored and the effectiveness of corrective actions is confirmed.
39 15
Example License Renewal Technical Information Submittal November 4, 1996 1
The program is implemented by written procedures that are maintained in accordance with 2
administrative controls which are summarized in the Oconee UFSAR, Section 13.5 3
[Reference 2].
4 5
The results of the Reactor Building Tendon Surveillance Program are documented in Reactor 6
Building Tendon Surveillance reports submitted to the NRC. A total of 18 tendon surveillances 7
have been performed at Oconee through 1996 (six surveillances on each unit). Reactor Building 8
tendon surveillances have revealed only minor pitting corrosion-on bearing plates (this condition 9
existed at the time of installation and no deterioration has occurred since installation) and minor 10 grease leakage. Coating degradation and minor local corrosion of the bearing plates, which are 11 considered to be cosmetic and not an applicable aging effect for license renewal, were repaired 12 by an existing maintenance procedure.
13 14 A total of 54 tendon wires have been visually examined and found in excellent condition with no 15 corrosion observed. A total of 162 wire segments have been tensile tested to date with no 16 significant change in ultimate. strength or elongation of the wire as compared to results obtained 17 during initial acceptance tests. No moisture has been detected and no change in grease coloring 18 or condition has been noted. Tendon grease leakage has been identified but no evidence exists to 19 show that the bulkfill grease has any detrimental effect on the concrete.
20 21 In summary, operating experience has shown that loss of material due to corrosion is the only 22 applicable aging effect for the post-tensioning system components. Plant operating experience 23 has shown that periodic inspections of tendons can identify corrosion on wire and anchorage and 24 the condition of protective grease. In addition, tensile testing of wire samples has been 25 performed and prestress force levels have been determined. Corrective actions are taken to repair 26 affected areas such that the post-tensioning system can maintain the required prestress during the 27 period of extended operation. Furthermore, additional assurance will be provided by the 28 incorporation of the random sample of tendons per the guidance contained in Regulatory Guide 29 1.35 [Reference 19]. For these reasons, Duke Power has determined that the Oconee Reactor 30 Building Tendon Surveillance Program, in conjunction with the enhancements of random 31 sampling of tendons, provide reasonable assurance for managing the aging effects of the post 32 tensioning system such that the structural support function of the Reactor Building by the post 33 tensioning system, as previously identified in Table 2.3-1 will be maintained during the period 34 of extended operation.
35 3.3.4.2 Time-Limited Aging Analysis 36 Loss of prestress in the post-tensioning system is due to material strain occurring under constant 37 stress. Loss of prestress over time is accounted for in the design and is a time-limited aging 38 analysis requiring review for license renewal.
39 40 In accordance with ACI 318-63 [Reference 6], the design of the Oconee Reactor Building post 41 tensioning system provides for prestress losses caused by the following:
16
Example License Renewal Technical Information Submittal November 4, 1996 1
2
- Elastic shortening of concrete 3
- Creep of concrete 4
- Shrinkage of concrete 5
- Relaxation of prestressing steel stress 6
- Frictional loss due to curvature in the tendons and contact with tendon conduit.
7 8
No allowance is provided for seating of the anchor since no slippage occurs in the anchor during 9
transfer of the tendon load into the structure [Reference 2].
10 11 By assuming an appropriate initial jacking (tensile loading) stress and using appropriate prestress 12 loss parameters, the magnitude of the design losses and the final effective prestress at the end of 13 40 years for typical dome, vertical, and hoop tendons was calculated at the time of initial 14 licensing. This analysis is presently summarized in the Oconee UFSAR, Section 3.8.1.5.2 15
[Reference 2].
16 17 This analysis is currently being revised to cover an additional 20 years of operation beyond the 18 initial term. The evaluation of this TLAA is contingent on:
19 20 (1) the NRC review of the Oconee methodology for determining the most accurate 21 minimum required lift-off force for each tendon group [Reference 20]; and 22 (2) the results of the next scheduled tendon surveillance (approximately May 1997).
23 24 Following completion of the above activities, NRC will be advised when the evaluation of this 25 TLAA for the period of extended operation will be provided.
26 27 References to appropriate correspondence will be provided.
28 3.3.5 OTHER REACTOR BUILDING INTERIOR COMPONENTS 29 This information is under development and will be provided at a later date.
- 30.
3.3.6
SUMMARY
AND CONCLUSIONS 31 The approach for demonstrating the management of aging effects applicable to the Reactor 32 Building was to determine the applicable aging effects by reviewing materials of construction, 33 operating environment, industry experience and NRC generic communications and then to 34 determine the ability of existing programs to manage the applicable aging effects. Table 3.3-1 35 provides a summary of the applicable aging effects and credited aging management programs for 36 each of the component groups of the Oconee Reactor Building. The following sections provide 37 summary discussions of the aging management review results.
17
Example License Renewal Technical Information Submittal November 4, 1996 1
3.3.6.1 Concrete Components 2
The Oconee Reactor Building reinforced concrete components were designed in accordance with 3
ACI 318-63 and constructed in accordance with ACI 301, using ingredients conforming to ACI 4
and ASTM standards which provide a good quality, dense, low permeability concrete that 5
preclude aging effects. In addition, resistance to surface deterioration is enhanced by the 6
application of prestress to the concrete sections. The prestressed concrete design places the 7
cylinder and dome concrete in compression for all normal loading conditions over the current and 8
extended period of operation. The compression minimizes the number and width of shrinkage, 9
temperature, or load induced cracks. Environmental impacts on Reactor Building concrete 10 components have been assessed and no applicable aging effects were identified even for the 11 concrete components located below grade and in contact with backfill and groundwater. No 12 time-limited aging analyses associated with the Reactor Building components were identified.
13 3.3.6.2 Steel Components (Group 1) 14 The design of the steel components used in the Oconee Reactor Buildings minimized the 15 potential for occurrence of detrimental aging effects. Operating experience has shown that loss of 16 material due to corrosion is an applicable aging effect:
17 18 (1) for the liner, hatches, and penetrations if the coatings are not maintained; 19 (2) for the liner below the concrete floor if the expansion joint sealants are not 20 maintained; and 21 (3) for the liner behind welded attachments if the cavity formed between the attachment 22 and the liner is not sealed.
23 24 Plant operating experience has shown that periodic visual inspections can identify areas where 25 loss of material is occurring and corrective actions can be taken to repair the affected area such 26 that the essentially leaktight barrier, structural support, and heat sink, as previously defined in 27 Table 2.3-1, will be maintained during the period of extended operation. Furthermore, 28 additional assurance that the essentially leaktight barrier and structural support will be 29 maintained will be provided when the requirements of ASME Section XI, Subsection IWE are 30 implemented at Oconee, as required by §50.55a(b)(2). For these reasons, Duke Power has 31 determined that the Reactor Building Civil Inspection Program for the Integrated Leak Rate Test, 32 in conjunction with the enhancements of ASME Section XI, Subsection IWE which will be 33 implemented by September 9, 2001, provide reasonable assurance for managing the loss of 34 material of the above locations of the steel components of the Reactor Building, such that the 35 essentially leaktight barrier, structural support, and heat sink (as previously identified in table 36 2.3-1) of the Reactor Building is maintained during the period of extended operation.
37 38 Fatigue loads as described in the Oconee UFSAR, Section 3.8.1.5.3 were considered in the 39 design of the liner plate and are considered to be time-limited aging analyses for the purposes of 40 license renewal. These existing analyses have been evaluated and determined to be valid for the 41 period of extended operation.
18
Example License Renewal Technical Information Submittal November 4, 1996 1
2 The time-limited aging analysis associated with the polar crane will be provided later.
3 3.3.6.3 Steel Components (Group 2) 4 This information is in preparation and will be provided later.
5 6
3.3.6.4 Post-Tensioning System 7
Evaluation and operating experience has shown that loss of material due to corrosion at the 8
tendon anchorage if grease is not maintained is the only applicable aging effect for the post 9
tensioning system components. Plant operating experience has shown that periodic inspections 10 of tendons can identify corrosion on wire and anchorage and the condition of protective grease.
11 In addition, tensile testing of wire samples has been performed and prestress force levels have 12 been determined. Corrective actions are taken to repair affected areas such that the post 13 tensioning system can maintain the required prestress during the period of extended operation.
14 Furthermore, additional assurance will be provided by the incorporation of the random sample of 15 tendons per the guidance contained in Regulatory Guide 1.35 [Reference 19]. For these reasons, 16 Duke Power has determined that the Oconee Reactor Building Tendon Surveillance Program, in 17 conjunction with the enhancements of random sampling of tendons, provide reasonable 18 assurance for managing the aging effects of the post-tensioning system, such that the structural 19 support of the Reactor Building, as previously identified in Table 2.3-1, is maintained during 20 the period of extended operation.
21 22 Summary of the time-limited aging analysis for the post-tensioning system will be provided later.
23 3.3.6.5 Other Reactor Building Interior Components 24 This information is in preparation and will be provided later.
25 3.3.6.6 Conclusions 26 The evaluations presented in Section 3.3 demonstrate that the effects of aging on the components 27 of the Reactor Building will be adequately managed such that the Reactor Building intended 28 functions, as previously identified in Section 2.3 (please see page 7) will be maintained under all 29 loadings for the period of extended operation. These evaluations meet the requirements of 10 30 CFR Part 54, §54.21(a)(3).
31 19
Example License Renewal Technical Information Submittal November 4, 1996 3.
3.7 REFERENCES
1.
Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule, Nuclear Energy Institute, NEI 95-10, Revision 0, March 1996.
- 2.
Oconee Nuclear Station, Updated Final Safety Analysis, as revised.
- 3.
Building Code Requirements for Reinforced Concrete, ACI 318-63, American Concrete Institute, Detroit, Michigan.
- 4.
Pressurized Water Reactors Containment Structure License Renewal Report, NUMARC Report Number 90-01, Nuclear Management and Resource Council, Revision 1, September 1991.
- 5.
Guide for Making a Condition Survey of Concrete in Service, ACI 201. 1R-92, American Concrete Institute, Detroit, Michigan.
- 6.
Building Code Requirements for Reinforced Concrete, ACI 318-63, American Concrete Institute, Detroit, Michigan.
- 7.
Specifications for Structural Concrete for Buildings, ACI 301, American Concrete Institute, Detroit, Michigan.
- 8.
Prediction of Creep, Shrinkage, and Temperature Effects in Concrete Structures, ACI 209R-82, American Concrete Institute, Detroit, Michigan.
- 9.
Specification for the Design, Fabrication, and Erection of Structural Steel for Buildings, American Institute of Steel Construction, 1963.
- 10.
Appendix J to 10 CFR Part 50 - Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors.
- 11.
"Requirements for Class MC and Metallic Liners of Class CC Components of Light Water Cooled Power Plants," Subsection IWE, ASME Code Section XI, 1992 Edition with the 1992 Addenda.
- 12.
61 FR 41303, August 8, 1996, Codes and Standards for Nuclear Power Plants; Subsection IWE and Subsection IWL, Final Rule.
- 13.
NUREG-1540, BWR Steel Containment Corrosion, April 1996.
20
Example License Renewal Technical Information Submittal November 4, 1996
- 14.
Oconee Nuclear Station Technical Specifications, as revised by amendments to Facility Operating Licenses DPR-38, DPR-47, and DPR-55.
- 15.
ASME Boiler and Pressure Vessel Code,Section III, "Nuclear Vessels, " 1965.
- 16.
Greiss, J.C., Corrosion of Steel Tendons in Concrete Pressure Vessels - Review of Recent Literature and Experimental Investigations, NUREG/CR-0092, ORNL/NUREG-37, Oak Ridge, Tennessee, June 1978.
17 NUREG/CR-4652, ORNL/TM-10059, Concrete Component Aging and its Significance to Life Extension of Nuclear Power Plants, Oak ridge National Laboratory, Oak Ridge, TN, September 1986.
- 18.
J. W. Hampton, Oconee Nuclear Station, letter dated October 30, 1996 to NRC, "Proposed Amendment to the Reactor Building Structural Integrity Technical Specifications"
- 19.
Regulatory Guide 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Revision 3, July 1990.
- 20.
Hampton, J. W. letter dated March 14, 1996 to NRC, Response to Request for Additional Information concerning Reactor Building Post-Tensioning Systems, Sixth Surveillance.
21
Example License Renewal Technical Information Submittal November 4, 1996 1
Table 3.3-1 2
Aging Management Results for 3
Reactor Building Components 4
(working draft) 5 Components Subject to Applicable Aging Management Aging Management Aging Programs Review Effects Concrete Components None None required Cylinder Wall Dome Equipment Foundations Floor Foundation Slab Masonry Brick Walls Reinforced Concrete (Primary, secondary shields)
Removable Missile Shields Steel Components (Group 1)
Loss of material due to corrosion Reactor Building Civil Anchors/Embedments/
for the liner, hatches, and Inspection Program for the Attachments penetrations if the coatings Integrated Leak rate Test Electrical Penetrations are not maintained; Emergency Personnel Hatch for the liner below the Reactor Building Integrated Equipment Hatch concrete floor if the Leak Rate Test (defense-in Fuel Transfer Tubes expansion joint sealants are depth)
Liner not maintained; and Mechanical Penetrations for the liner behind welded Reactor Building Local Leak Personnel Hatch attachments if the cavity Rate Test (defense-i n-depth) formed between the attachment and the liner is not sealed 22
Example License Renewal Technical Information Submittal November 4, 1996 1
Table 3.3-1 -(continued)
Aging Management Results for Reactor Building Components (working draft)
Components Subject to Applicable Aging Management Aging Management Aging Programs Review Effects Steel Components (Group 2)
To be determined To be determined Cable Tray & Conduit Cable Tray & Conduit Supports Class 2 & 3 Pipe Supports Controlled Leakage Doors Crane Rails and Girders Electrical Racks, Panels &
Cabinets Equipment Supports Grating (QA4)
HVAC Duct Supports Impulse Line Supports Instrument Racks, Panels &
frames Jet Barriers Missile Shields Pipe Whip Restraints Platform Supports Sump Screens Post-Tensioning System Loss of Material at tendon Reactor Building Tendon Tendon Anchorage anchorage if grease is not Surveillance Program Tendon Wires maintained.
Other Reactor Building To be determined To be determined Interior Components Lead Shielding Supports Fire Stops Shield Wall Tendons 2
3 4
5 23
Example License Renewal Technical Information Submittal November 4, 1996 1
- 2. INTEGRATED PLANT ASSESSMENT 2
STRUCTURE/COMPONENT IDENTIFICATION 3
2.1 INTRODUCTION
4 The License Renewal Rule in 10 CFR Part 54, §54.21(a)(1) [Reference 1] requires that 5
for those systems, structures, and components within the scope of this part, as delineated 6
in §54.4, those structures and components subject to an aging management review be 7
identified and listed. Part 54, §54.21(a)(2), further requires that the methods used to 8
identify and list these structures and components be described and justified. The 9
requirements of Part 54 have been met by using methods that are consistent with the 10 guidance provided in NEI 95-10, Section 4.1 [Reference 2].
11 12 The purpose of Chapter 2 is to provide the information required by the above two sections 13 of Part 54. Section 2.2 describes the overall plant scoping methodology which has been 14 utilized on Oconee. This scoping methodology is consistent with the guidance provided 15 in NEI 95-10, Chapter 3 [Reference 2]. The system, structure, and component scoping 16 and Integrated Plant Assessment have been divided along engineering discipline lines 17 traditional to Duke Power (e.g. Civil/Structural, Electrical, and Mechanical). The Reactor 18 Building and the Reactor Coolant System, as important elements in the radioactive 19 release line-of-defense, receive special focus and are handled individually in Sections 2.3 20 and 2.4 respectively. Sections 2.3 through 2.7 will utilize the results of the scoping 21 methodology to meet the requirements of §§54.21(a)(1) and (a)(2) and are further 22 described in the respective sections that follow.
23 24 2.2 IDENTIFICATION OF SYSTEMS, STRUCTURES, AND COMPONENTS 25 WITHIN THE SCOPE OF LICENSE RENEWAL 26 The methodology used to identify the Oconee systems, structures, and components within 27 the scope of license renewal is provided in this section and is consistent with the guidance 28 provided in NEI 95-10, §3.1 [Reference 2]. Section 2.2.1 discusses the review 29 performed to identify systems, structures, and components that meet the criteria contained 30 in §§54.4(a)(1) and (a)(2). The review performed to identify the systems, structures, and 31 components that meet the criteria contained in §54.4(a)(3) is described in Section 2.2.2.
32 33 A list of Oconee structures was created by reviewing the Oconee FSAR[Reference 3],
34 Oconee general arrangement drawings, and other Oconee specific documents. The 35 identification of safety-related structures and those nonsafety-related structures whose 36 failure prevents safety-related systems, structures, or components from fulfilling their 37 safety-related functions was based upon the classification of each Oconee structure as 38 documented in the Oconee UFSAR [Reference 3], as further described in Section 2.2.1.1.
39 The identification of those structures relied on to demonstrate compliance with the
Example License Renewal Technical Information Submittal November 4, 1996 1
regulations listed in §54.4(a)(3) was made by reviewing Oconee specific documents 2
associated with the listed regulations, as further described in Section 2.2.2.
3 4
Mechanical systems scoping will be provided later 5
6 Electrical systems scoping will be provided later.
7 2.2.1 REVIEW To CRITERIA CONTAINED IN §§54.4(a)(1) AND (a)(2) 8 2.2.1.1 Structures 9
Oconee structures are designated as either Class 1, 2, or 3. Oconee Class 1 structures are 10 those which prevent uncontrolled release of radioactivity and are designed to withstand 11 all loadings without loss of function, UFSAR, Chapter 3.2 [Reference 3]. This Oconee 12 classification is consistent with the intent of §54.4(a)(1). Therefore, Oconee Class 1 13 structures are considered to be within the scope of license renewal and meet the criteria 14 contained in §54.4(a)(1).
15 16 Class 2 structures are those whose limited damage would not result in a release in 17 radioactivity and would permit a controlled plant shutdown but could interrupt power 18 generation. Class 2 structures do not perform a nuclear safety function but their failure 19 could reduce the function of a nuclear safety system to an unacceptable level. UFSAR, 20 Chapter 3.2 [Reference 3]. This Oconee classification is consistent with the intent of 21
§54.4(a)(2). Therefore, for Oconee, Class 2 structures are considered to be within the 22 scope of license renewal and meet the criteria contained in §54.4(a)(2).
23 24 Class 3 structures are those structures whose failure could inconvenience operation but 25 are not essential to power generation, orderly shutdown or maintenance of the reactor in a 26 safe shutdown. Oconee Class 3 structures are not considered to meet either §54.4(a)(1) or 27
§54.4(a)(2).
28 2.2.1.2 Mechanical Systems 29 Mechanical systems scoping will be provided later 30 2.2.1.3 Electrical Systems 31 Electrical systems scoping will be provided later.
32 33 2.2.2 REVIEw To CRITERIA CONTAINED IN §54.4 (a)(3) 34 Oconee structures and systems were also evaluated to determine whether they are 35 required to demonstrate comltiance with the regulated events identified in §54.4 (a)(3).
.36 Each of the following sections includes a brief discussion of the regulated event and the 37 associated Oconee documents which identify the Oconee structures and mechanical 2
Example License Renewal Technical Information Submittal November 4, 1996 1
systems relied on in the safety analyses or plant evaluation to demonstrate compliance 2
with the regulated event.
3 2.2.2.1 Fire Protection 4
Part 54, §54.4(3) requires that all systems, structures, and components relied on in safety 5
analyses or plant evaluations to demonstrate compliance with the regulations for fire 6
protection (§50.48) be included within the scope of license renewal. The Oconee Fire 7
Protection Program is based on compliance with General Design Criteria (GDC 3) 8
[Reference 4]. Oconee conforms to GDC 3 by compliance with BTP 9.5-1, Appendix 9
A[Reference 5]. In addition to the BTP, Oconee was also required to comply with the 10 provisions of 10 CFR 50, Appendix R (Sections III.G, III.J, and III.0)[Reference 6]. The 11 following list of documents were reviewed to determine which Oconee structures and 12 mechanical systems are relied upon to meet the requirements of 10 CFR 50 Appendix R.
13 14 U. S. Nuclear Regulatory Commission, "Fire Protection Safety Evaluation Report by 15 the Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission in 16 the Matter of Duke Power Company Oconee Nuclear Station, Units 1, 2, & 3," dated 17 August 11, 1978.
18 19 U. S. Nuclear Regulatory Commission, "Exemption from the Fire Protection 20 Requirements of Section III.G of 10 CFR Part 50, Appendix R," dated August 21, 21 1989.
22 23 Based upon a review of these documents, as well as Oconee specific documents 24 maintained onsite, the structures and mechanical systems which are required to 25 demonstrate compliance with the fire protection requirements in §54.4(a)(3) and their 26 intended functions were identified.
27 28 2.2.2.2 Environmental Qualification 29 Part 54, §54.4(3) requires that all systems, structures, and components relied on in safety 30 analyses or plant evaluations to demonstrate compliance with the regulation for 31 environmental qualification (§50.49) be included within the scope of license renewal. A 32 master list of all equipment contained in the Environmental Qualification Program has 33 been compiled for each of the Oconee units.
34 35 Based upon a review of these documents, the structures and mechanical systems which 36 are required to demonstrate compliance with the environmental qualification 37 requirements in §54.4(a)(3) and their intended functions were identified.
38 3
Example License Renewal Technical Information Submittal November 4, 1996 1
2.2.2.3 Pressurized Thermal Shock (PTS) 2 Part 54, §54.4(3) requires that all systems, structures, and components relied on in safety 3
analyses or plant evaluations to demonstrate compliance with the regulation for 4
pressurized thermal shock (§50.61) be included within the scope of license renewal.
5 Pressurized thermal shock is a phenomenon limited to the Reactor Coolant System and 6
the reactor vessel. Pressurized thermal shock has also been identified as a time-limited 7
aging analysis applicable to the Reactor Coolant System which is evaluated in Section 8
3.4.9 of OLRP-1001.
9 10 The only mechanical system required to demonstrate compliance with §50.61 is the 11 Reactor Coolant System. No Oconee structures are required to demonstrate compliance 12 with §50.61.
13 14 2.2.2.4 Anticipated Transient Without Scram (ATWS) 15 Part 54, §54.4(3) requires that all systems, structures, and components relied on in safety 16 analyses or plant evaluations to demonstrate compliance with the regulation for 17 anticipated transient without scram (§50.62) be included within the scope of license 18 renewal. At Oconee, the Diverse Scram System and the ATWS Mitigation System 19 Actuation Circuitry were installed to address this regulation. The design of these systems 20 were reviewed to identify structures and mechanical systems relied upon to demonstrate 21 compliance with §50.62.
22 23 Based on this review, the structures and mechanical systems required to demonstrate 24 compliance with the anticipated transient without scram requirements in 10 CFR 25 54.4(a)(3) and their intended functions were identified.
26 27 2.2.2.5 Station Blackout 28 Part 54, §54.4(3) requires that all systems, structures, and components relied on in safety 29 analyses or plant evaluations to demonstrate compliance with the regulation for station 30 blackout (§50.63) be included within the scope of license renewal. The documents listed 31 below, as well as Oconee specific documents maintained onsite, were reviewed to 32 identify which structures and mechanical systems are relied upon to meet the 33 requirements of station blackout:
34 35
- Tucker, Hal B., Letter to U. S. NRC Document Control Desk, "Oconee 36 Nuclear Site Station Blackout," dated April 17, 1989.
37 38 U. S. Nuclear Regulatory Commission, "Safety Evaluation for Station 39 Blackout (10 CFR 50.63) - Oconee Nuclear Station, Units 1, 2, and 3,"
40 dated March 10, 1992.
41 4
Example License Renewal Technical Information Submittal November 4, 1996 1
Summary of June 4, 1992, Meeting on Station Blackout Response for the 2
Oconee Nuclear Station, dated June 12, 1992.
3 4
Hampton, J. W., Letter to U. S. NRC Document Control Desk, "Oconee 5
Nuclear Site Revised Response to 10 CFR 50.63 Requirements for 6
Station Blackout," dated July 1, 1992.
7 8
U. S. Nuclear Regulatory Commission, "Supplemental Safety Evaluation 9
for Station Blackout (10 CFR 50.63) - Oconee Nuclear Station, Units 1, 10 2, and 3," dated December 3, 1992.
11 12 Based upon a review of these documents, the structures and mechanical systems required 13 to demonstrate compliance with the station blackout requirements in §54.4(a)(3) and their 14 intended functions were identified.
15 2.2.3 RESULTS 16 The results of the determination of structures and systems within the scope of license 17 renewal and their intended functions are maintained in documentation available in an 18 auditable and retrievable form, in accordance with the requirements of §54.37(a). The 19 Reactor Building and the Reactor Coolant System, as important elements in the 20 radioactive release line-of-defense, receive special focus and are handled individually in 21 Sections 2.3 and 2.4 respectively. Mechanical, electrical and structural components are 22 evaluated in Sections 2.5, 2.6 and 2.7, respectively.
23 24 The Oconee Reactor Buildings have been determined to be within the scope of license 25 renewal and subject to aging management review as further described in Sections 2.3 and 26 3.3 of OLRP-1001. All other Oconee structures which have been determined to be within 27 the scope of license renewal and subject to aging management review are listed and 28 described in Sections 2.7 and 3.7 of OLRP-1001. The intended functions of the Reactor 29 Building and other Oconee structures are listed in Sections 2.3 and 2.7, respectively.
30 31 The Reactor Coolant System has been determined to be within the scope of license 32 renewal and subject to aging management review, as further described in Sections 2.4 and 33 3.4 of OLRP-1001. All other Oconee mechanical systems, or portions thereof, which 34 have been determined to be within the scope of license renewal and subject to aging 35 management review are listed and described in Sections 2.5 and 3.5 of OLRP-1001. The 36 intended functions of the Reactor Coolant System and other Oconee mechanical systems 37 are listed in Sections 2.4 and 2.5, respectively.
5
Example License Renewal Technical Information Submittal November 4, 1996 2
2.
2.4 REFERENCES
3
- 1.
Requirements for Renewal of Operating Licenses for Nuclear Power Plants, 10 CFR Part 54.
- 2.
Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule, NEI 95-10, Revision 0, Nuclear Energy Institute, March 1996.
- 3.
Oconee Nuclear Station, Updated Final Safety Analysis Report, as revised.
- 4.
Appendix A to 10 CFR Part 50 - General Design Criteria for Nuclear Power Plants, Criteria 3, Fire Protection.
- 5.
Guidelines for Fire Protection for Nuclear Power Plants, Appendix A to Branch Technical Position APCSB 9.5-1, August 23, 1976.
- 6.
Appendix R to 10 CFR Part 50 - Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979.
6
Example License Renewal Technical Information Submittal November 4, 1996 1
2.3 REACTOR BUILDING 2
The Oconee Reactor Buildings were determined to be within the scope of the License 3
Renewal Rule based on the review that was performed as previously described in Section 4
2.2 of this report. The following intended functions of the Reactor Building were 5
determined by reviewing information contained in the Oconee UFSAR, Section 3.8 6
[Reference 1] and Oconee engineering documents, as well as NEI 95-10, Section 4.1, 7
Table 4.1-1 [Reference 2]:
8 9
- 1. Provides essentially leaktight barrier to prevent uncontrolled release of 10 radioactivity.
11
- 2. Provides structural and/or functional support to safety related systems, 12 structures, and components. More specifically for the post-tensioning 13 systems, this function means to impose compressive forces on the concrete 14 containment structure to resist the internal pressure resulting from a design 15 basis accident with no loss of structural integrity.
16
- 3. Provides shelter/protection to safety related systems, structures, and 17 components (including radiation shielding).
18
- 4. Provides rated fire barrier to confine or retard a fire from spreading to or from 19 adjacent areas of the plant.
20
- 5. Serves as an external missile barrier.
21
- 6. Provides structural and/or functional support to non-safety related systems, 22 structures, and components where failure of this structure could directly 23 prevent satisfactory accomplishment of any of the required safety related 24 functions.
25
- 7. Provides heat sink during design basis accidents or station blackout.
26 27 Each Oconee Reactor Building is a composite structure consisting of a post-tensioned, 28 reinforced concrete structure with cylindrical wall, a flat foundation slab, and a shallow 29 dome roof. An illustration of the Oconee prestressed concrete Reactor Building is shown 30 in Figure 2.3-1. The Reactor Building completely encloses the reactor and the associated 31 Reactor Coolant System along with other vital electrical, mechanical and structural 32 components. The cylinder wall integrity is provided by a post-tensioning system 33 consisting of horizontal and vertical tendons in the cylinder wall. The dome integrity is 34 provided by three sets of tendons, each set oriented 120 degrees from the other. The 35 concrete foundation slab is conventionally reinforced. The entire structure is internally 36 lined with a steel liner plate to assure a high degree of leaktightness. The principle 37 dimensions of the Reactor Buildings are provided in the UFSAR, Chapter 3.8 and Figure 38 3-19.[Reference 1].
39 40 The Reactor Building structure is subdivided into component groupings in preparation for 41 the aging management review. Many structural components are not typically associated 42 with unique equipment identifiers and thus are not individually identified during the 43 identification of components subject to aging management review. Specific structural 7
Example License Renewal Technical Information Submittal November 4, 1996 1
component identifiers are not needed because the aging management review process and 2
any resulting programmatic oversight will be performed across an entire component 3
grouping. With materials of construction and service environment the same, the aging 4
effects would manifest similarly across the entire component grouping.
5 6
A list of Reactor Building structural components was developed based upon a review of 7
NUMARC PWR Containment Industry Report [Reference 3], NEI 95-10, Table 2 8
[Reference 2], and Oconee specific documents. The functions of each of the Reactor 9
Building structural components were determined by reviewing information contained in 10 the Oconee UFSAR [Reference 1], Oconee engineering documents, as well as NEI 95 11 10, Section 4.1, Table 4.l-1[Reference 2]. These functions are the same as the Reactor 12 Building functions and serve to provide specific component-level focus which will be 13 needed for the component aging management review. Components which do not perform 14 an intended function are not within the scope of license renewal. The list of Reactor 15 Building components groups within the scope of 10 CFR Part 54 and their intended 16 functions is provided in Table 2.3-1. Individual Reactor Building components are 17 identified in Oconee specific documents maintained onsite.
18 19 The Oconee Reactor Building structural components within the scope of 10 CFR Part 54 20 were reviewed to determine those components which are subject to an aging management 21 review in accordance with §54.21(a)(1). The aging management review of a structural 22 component is directly related to whether the component performs an intended function 23 without moving parts or without a change in configuration or properties (i.e., passive) and 24 whether they are subject to replacement based on a qualified life or specified time period 25 (i.e., long-lived). Consistent with the guidance provided in NEI 95-10 [Reference 2], the 26 Reactor Building structural components within the scope of the Rule are long-lived and 27 passive and will require an aging management review.
28 29 The Reactor Building structural components have been grouped based on material of 30 construction and component-level function. These component groups are described in the 31 following sections.
32 33 2.3.1 CONCRETE COMPONENTS 34 2.3.1.1 Dome and Cylinder Walls 35 The reinforced concrete dome and cylinder walls are prestressed by a post-tensioning 36 system, as shown in Figure 2.3-2. The combined strength provided by the concrete, 37 conventional reinforcing steel, and the post-tensioning system is utilized to satisfy the 38 design loads. Although these three material components act together as one composite 39 system, the post-tensioning system is addressed as a separate component because it is 40 installed and stressed after the reinforced concrete components are complete and because 41 of the unique tendon surveillance program.
42 8
Example License Renewal Technical Information Submittal November 4, 1996 1
Conventional reinforcing is provided near the surface of the cylinder walls and dome 2
primarily to resist local moment and shear loads at discontinuities and for temperature 3
and shrinkage crack control. The conventional reinforcing is accounted for in the strength 4
design of the concrete sections for the internal shear forces and moments resulting from 5
the design loadings.
6 7
The concrete sections are thickened and the conventional reinforcing steel is increased at 8
the structural discontinuities to account for the increased stresses in those local areas.
9 Primary structural discontinuities occur at the base of the cylinder and at the transition of 10 the cylinder walls and dome to the ring girder. The ring girder serves as the anchorage 11 area for the upper end of the vertical tendons and for both ends of the dome tendons. Six 12 vertical buttresses are provided along the exterior face of the cylinder to serve as the 13 anchorage points for the hoop tendons. The hoop tendons extend for 120 degrees of arc.
14 Supplementary reinforcing steel is provided at tendon anchorage zones to account for the 15 local forces at the anchorages. The concrete cylinder walls are also thickened and 16 additional reinforcing is provided locally at the equipment hatch to account for the flow 17 of forces in the walls around the relatively large diameter opening required for the hatch.
18 Additionally, the concrete dome is coated with silicone rubber on the exterior to protect 19 the dome from weathering conditions.
20 2.3.1.2 Floor 21 A reinforced concrete floor is provided in the Reactor Building above the embedded 22 portion of the liner plate to protect the liner plate from punctures and corrosion.
23 2.3.1.3 Foundation Slab 24 The conventionally reinforced concrete foundation slab serves as the structural foundation 25 support for the Reactor Building. The vertical tendons extend through the foundation 26 slab thickness and are anchored on the underside of the slab. A reinforced concrete 27 enclosure, the lower tendon access gallery, shown in Figure 2.3-1, is provided at the 28 underside of the foundation slab for access to the lower vertical tendon anchorages for 29 tendon installation and surveillance purposes. The lower tendon.access gallery and the 30 foundation slab are constructed of separate concrete pours with horizontal and vertical 31 isolation joints provided. The lower tendon access gallery does not support the intended 32 functions of the Reactor Building and is therefore not within the scope of the Rule.
33 2.3.1.4 Add Remaining Concrete Components 34 35 2.3.2 STEEL COMPONENTS (GROUP 1) 36 2.3.2.1 Liner Plate 37 The interior of the Reactor Building is lined with steel liner plates that are welded 38 together. The liner plate covers the dome, the cylinder wall and also runs between the 9
Example License Renewal Technical Information Submittal November 4, 1996 I
floor and the foundation slab to form an essentially leaktight barrier. The Oconee Reactor 2
Building liner plate is ASTM A36 or A516 plate attached to the concrete by means of an 3
angle grid system of ASTM A36 material stitch welded to the liner plate and embedded in 4
the concrete. The liner plate is anchored in both the longitudinal and hoop direction. The 5
anchor spacing and welds were designed to preclude failure of an individual anchor. The 6
frequent anchoring is designed to prevent significant distortion of the liner plate during 7
accident conditions and to ensure that the liner maintains its essentially leak tight 8
integrity. All penetrations are continuously welded to the liner plate before the concrete 9
in which they are embedded is placed. The entire length of every seam was leak tested 10 following fabrication. Radiographs were taken for at least one foot in each fifty feet of 11 welding completed by each welder during fabrication. The liner plate is coated on the 12 inside with inorganic zinc primer and Phenoline 305 for corrosion protection. There is no 13 coating on the side in contact with the concrete. At all penetrations, the liner plate is 14 thickened to reduce stresses in accordance with the ASME Code, 1965 [Reference 4].
15 The liner was designed as a free standing vessel for erection loads including use of the 16 liner as the internal form for the concrete. The liner plate is thickened at large attachments 17 such as the polar crane brackets to accommodate strength and welding requirements for 18 the attachment and anchorage. The general liner configuration is shown in Figure 2.3-2.
19 20 The ASME Code [Reference 4] is used as the basis for establishing allowable liner plate 21 strains and stresses. The ASME Code requires that the liner material be prevented from 22 experiencing significant distortion due to thermal loads and that stresses be considered 23 from a fatigue standpoint. In accordance with the ASME Code, the liner plate is 24 restrained against significant distortion by continuous angle anchors, never exceeds the 25 temperature limitation of 700oF and also satisfies the criteria for limiting strains on the 26 basis of fatigue considerations.
27 2.3.2.2 Anchors/Embedments/Attachments 28 Anchors/embedments are steel commodities, such as angles and anchor studs, that are 29 welded to the liner and serve to anchor the liner to the Reactor Building concrete shell.
30 The liner anchors are shown in Section 1-1 of Figure 2.3-2. In addition, other 31 anchors/embedments are provided that serve to transfer loads into the concrete cylinder 32 wall or foundation mat from attachments to the liner. Figure 2.3-3 provides a detail of the 33 anchors for major equipment. In these cases a thickened insert plate is welded to the liner 34 and is used as the point of attachment for the anchorages. The polar crane bracket as 35 shown in Figure 2.3-2 is anchored to the concrete shell by a welded plate assembly that is 36 embedded in the concrete.
37 38 The anchors/embedments serve to maintain the essentially leaktight barrier by preserving 39 the integrity of the liner. The structural integrity of the liner insert part of other 40 anchorages, such as shown in Figure 2.3-3, is also necessary to maintain the essentially 41 leaktight barrier of the liner. The load carrying capacity of these anchorages is also 42 required to assure that the supported equipment, such as the polar crane or the steam 43 generators, can continue to perform safely as required.
10
Example License Renewal Technical Information Submittal November 4, 1996 2
Attachments to the liner that are integral with the liner and concrete structure (i.e.,
3 attachment has corresponding anchor in concrete), include those equipment or system 4
supports that are connected to the inside face of the liner and thus exposed to the interior 5
of the Reactor Building. The polar crane brackets are examples of attachments to the 6
liner. These attachments are shown in Figure 2.3-2. The polar crane brackets consist of 7
welded carbon steel plate construction of the same material and fabrication, and were 8
inspected using similar requirements for the liner. Other miscellaneous attachments to 9
the liner include structural steel attachments which are welded directly to the liner to 10 support SSCs.
11 2.3.2.3 Personnel Hatch 12 Two hatches are provided into each Reactor Building for personnel access and egress 13 (See Figure 2.3-4). The larger personnel hatch is used as the primary access point into the 14 Reactor Building. The smaller personnel hatch is used for emergency egress.
15 16 The personnel hatch consists of a double-door, welded steel assembly. The hatch is 17 designed to withstand all Reactor Building design conditions with either or both doors 18 closed and locked. The doors open toward. the center of the Reactor Building which 19 prevents unseating of the door during Reactor Building pressurization. The personnel 20 hatch.may be individually pressurized to demonstrate leaktightness. Quick acting 21 equalizing valves connect the personnel lock with the interior and exterior of the Reactor 22 Building vessel for the purposes of equalizing pressure between the two systems when 23 entering or leaving the Reactor Building. The equalizing valves are active components of 24 the airlocks and do not require an aging management review. Functionality of the 25 equalizing valves is verified periodically when the airlocks are pressurized and tested for 26 leakage.
27 28 The two personnel hatch doors are interlocked to prevent both being opened 29 simultaneously and to ensure that Reactor Building containment integrity is always 30 maintained by one door being completely closed before the other door can be opened.
31 The interlocking system has the capability to be bypassed allowing the doors to be left 32 open during plant cold shutdown. The interlock system is also an active component of 33 the personnel hatch and is not within the scope of this report. Serviceability of the 34 interlock system is verified during periodic personnel hatch leakage testing as well as 35 during the periodic maintenance.
36 37 Each personnel hatch door is provided with flexible seals. The exterior door is provided 38 with double seals to allow for local leakage testing between the seals. The seals are 39 replaced when warranted by their condition. The seals are not long-lived components and 40 therefore do not require an aging management review.
41 42 Hatches are designed and fabricated in accordance with the ASME Code [Reference 4].
43 The hatches conform to ASME Section III requirements for Class B vessels.
11
Example License Renewal Technical Information Submittal November 4, 1996 1
2 The plate materials that comprise the personnel hatch pressure vessel components are 3
painted carbon steel complying with ASME material specification A-516, Grade 70, 4
made to ASTM A300 specification, for fine grained materials with ductile material 5
properties suitable for low temperature use.
6 2.3.2.4 Equipment Hatch 7
A single equipment hatch as shown in Figure 2.3-5 is provided for each of the Reactor 8
Buildings. The equipment hatch design and fabrication conform to the ASME Code 9
[Reference 4] for Class B vessels. As with the personnel hatches, the equipment hatch is 10 fabricated using A-516 Grade 70, painted carbon steel made to ASTM A300 S11 specification.
12 13 The equipment hatch is furnished with a double sealed flange and bolted, dished head.
14 The barrel portion of the equipment hatch is thicker than required based on permissible 15 stresses. The space between the double seals on the equipment hatch flange can be 16 pressurized for local leakage testing. As with the personnel hatches, the flexible seals are 17 tested and replaced when warranted by condition. The seals are not long-lived, passive 18 components and do not require an aging management review.
19 2.3.2.5 Mechanical Penetrations 20 All penetrations through the Reactor Building pressure boundary are designed to maintain 21 the essentially leaktight barrier to prevent uncontrolled release of radioactivity. In 22 addition to supporting the essentially leaktight barrier function, each penetration performs 23 service related functions depending on the particular type of penetration. Penetrations 24 may also serve as support points for systems such as piping passing through the Reactor 25 Building boundary.
26 27 Penetration plate and sleeve material is ASTM A516 Grade 70 material. The plate 28 material is also fabricated to firebox quality and ASTM A300. The penetrations' 29 materials and fabrication meet the requirements of ASA N 6.2-1965 [Reference 5].
30 31 Mechanical penetrations provide the means for passage of process piping transmitting 32 liquids or gases across the Reactor Building boundary. A typical mechanical piping 33 penetration is shown in Figure 2.3-6 which shows a single barrier piping penetration with 34 a single closure between the process pipe and the Reactor Building liner. The 35 penetrations are solidly anchored to the Reactor Building wall or foundation slab 36 precluding any requirements for expansion bellows. In accordance with the design 37 requirement of ASME Section III, piping penetration reinforcing plates and the weldment 38 of the pipe closure to it were stress relieved UFSAR, Section 3.8.1.4.5[Reference 1].
39 40 The mechanical penetration boundaries for this report include the entire penetration 41 assembly exclusive of the process piping within the penetration. The welds of the 42 assembly to the process pipe are part of the process piping as defined by ASME Section 12
Example License Renewal Technical Information Submittal November 4, 1996 1
XI and are not included in the scope. Penetrations are designed to maintain the adjacent 2
concrete within an acceptable temperature range. The Reactor Building evaluation 3
boundary is shown on Figure 2.3-6.
4 5
Spare penetrations consist of a sleeve with welded end cap closure(s) or bolted blind 6
flange plate(s) with gaskets at both ends of the penetration sleeve. Spare penetrations can 7
readily be converted during an outage into additional permanent mechanical or electrical 8
penetrations, if required, during the life of the plant. The entire spare penetration 9
assembly is included in the boundary of this report.
10 11 The Reactor Building sump penetrations provide passage of the Low Pressure Injection 12 System (LPI) and Reactor Building spray piping across the Reactor Building boundary.
13 The LPI system serves to remove heat from the Reactor Building in the event of an 14 accident. The sump penetration detail is shown in Figure 2.3-7. During normal operation, 15 the inside end of the sump piping is open. The Reactor Building evaluation boundary is 16 shown in Figure 2.3-7 is at the weld to the inside end of the piping excluding the piping.
17 2.3.2.6 Electrical Penetrations 18 Electrical penetrations provide the means for electrical and instrumentation conductors to 19 cross the Reactor Building boundary while maintaining the essentially leaktight barrier.
20 An electrical penetration through the Reactor Building is shown in Figure 2.3-8. The 21 scope of the evaluation in this section includes all metallic components of the electrical 22 penetration that are part of the Reactor Building essentially leaktight barrier. The inside 23 steel header plate for the electrical terminals are included in the scope. The wiring, 24 sealing compound, fixtures to hold the sealing compound, and seal welds of the fixtures 25 to the header plate are addressed in Environmental Qualification reports. The associated 26 electrical wiring and sealing materials addressed in the Environmental Qualification 27 reports will be evaluated separately in Section 3.6. of this report. The Reactor Building 28 evaluation boundary is shown in Figure 2.3-8.
29 2.3.2.7 Fuel Transfer Tube 30 Two fuel transfer tubes penetrate the Reactor Building and link the refueling canal in the 31 Reactor Building with the fuel transfer canal in the fuel handling building. They serve as 32 the underwater pathway for moving the fuel assemblies into and out of the Reactor 33 Buildings as part of the refueling operations occurring during plant shutdown. As part of 34 the Reactor Buildings, the tubes must assure the essentially leaktight barrier function for 35 the design basis conditions.
36 37 The fuel transfer tube arrangement through the Reactor Building is shown in Figure 2.3-9.
38 As shown in the figure, the closure between the transfer tube and the sleeve that is 39 integrally welded to the Reactor Building liner, consists of a circular plate shop welded to 40 the tube and a short segment of pipe to mate with the sleeve. During normal operation, a 41 blind flange is in place on the fuel transfer tube penetration and serves as part of the 42 Reactor Building essentially leaktight barrier. Closure on the Fuel Handling Building end 13
Example License Renewal Technical Information Submittal November 4, 1996 I
outside of the Reactor Building consists of a gate valve supported from the end of the 2
transfer tube.
3 4
The boundary of this report includes the closure detail between the Reactor Building liner 5
and the transfer tube as shown in Figure 2.3-9. The transfer tube, blind flange, and gate 6
valve are part of the spent fuel pool system and are not within the scope of this report.
7 The transfer tube, blind flange, gate valve and other spent fuel pool cooling system 8
components are addressed in Sections 2.5 and 3.5 of this report.
9 2.3.3 STEEL COMPONENTS (GROUP 2) 10 This information is in preparation and will be provided later.
11 12 2.3.4 POST-TENSIONING SYSTEM 13 For information, an elevation section of the Oconee Reactor Building is provided in 14 Figure 2.3-2 which shows the orientation of the tendons. A section of a typical post 15 tensioned tendon assembly is also shown in Figure 2.3-10.
16 17 The Reactor Building cylinder wall is prestressed by 176 vertical tendons anchored at the 18 top surface of the upper ring girder at the top of the concrete cylinder and at the bottom of 19 the foundation slab and six groups of 105 hoop tendons plus two additional tendons 20 enclosing 1200 of arc for a total of 632 tendons anchored at the six vertical buttresses.
21 The dome is prestressed by three groups of 54 tendons oriented at 1200 to each other for a 22 total of 162 tendons anchored at the vertical face of the upper ring girder. Each tendon 23 consists of 90 wires bundled together. The design of the tendon system provides for the 24 loss of any three adjacent tendons in any of the groups without significantly affecting the 25 load carrying capacity of the Reactor Building. Conduits and bearing plates are cast into 26 the concrete shell to receive the tendons which are installed after construction of the 27 reinforced concrete is complete. The tendons are continuous from anchorage to 28 anchorage, being deflected around penetrations.
29 30 A tendon assembly consisting of the buttonheaded tendon wires (Birkenmeier 31 Brandestinin Ros Vogt or BBRV system), anchorage and conduit is shown in Figure 2.3 32
- 10. The BBRV system uses parallel wires with cold-formed buttonheads at the ends 33 which bear upon a perforated steel anchor head, thus providing a positive mechanical 34 means for transferring the prestress force into the concrete shell. Extensive prototypical
- 35.
static, dynamic, and low-temperature testing have been performed on the BBRV 36 anchorage system to assure that the ultimate capacity of the tendons can be developed.
37 This testing is described in detail in the Oconee UFSAR [Reference 1]. The testing and 38 evaluation of tendon prestress as a function of time is a TLAA and is discussed in Section 39 3.3 of this report.
40 14
Example License Renewal Technical Information Submittal November 4, 1996 1
The post-tensioning system is the primary means of satisfying the controlling design 2
loads, although the conventional mild steel reinforcing is taken into account when 3
checking representative sections of the structure internal forces and moments resulting 4
from the load combinations. The tendon stress remains in the elastic range for the 5
controlling design load combinations.
6 7
2.3.5 ADD OTHER REACTOR BUILDING INTERIOR COMPONENTS 8
This information is in preparation and will be provided later.
9 10 11 12 2.
3.6 REFERENCES
- 1.
Oconee Nuclear Station Updated Final Safety Analysis Report, as revised.
- 2.
Industry Guideline for Implementing the Requirements of 10 CFR Part 54 - The License Renewal Rule, NEI 95-10, Revision 0, Nuclear Energy Institute, March 1996.
- 3.
Pressurized Water Reactor Containment Structures License Renewal Industry Report, NUMARC Report Number 90-01, Nuclear Management and Resources Council, Revision 1, September 1991.
- 4.
ASME Boiler and Pressure Vessel Code,Section III, "Nuclear Vessels, " 1965.
- 5.
ASA N6.2-1965, "Safety Standard for the Design, Fabrication and Maintenance of Steel Containment Structures for Stationary Nuclear Power Reactors."
15
Example License Renewal Technical Information Submittal November 4, 1996 1
2 Table 2.3-1 3
Reactor Building Components and 4
Their Intended Functions 5
(working draft) 6 Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function Secondarye ShieldtWall (Identreddnth not below)
Concrete Components Cylinder Wall 2
3 4t5 6
7 Dome 2
3 5
6 7
Equipment Foundations Floor 2
3 5
6 7
Foundation Slab 2 13 5
6 7
Masonry Brick Walls Primary Shield Walls Removable Missile Shields Secondary Shield Walls Steel Components (Group 1)
Anchorage/Embedments/Attachments 1
Electrical Penetrations 1
Emergency Personnel Hatch I7 Equipment Hatch I_7 Fuel Transfer Tube1 Liner Plate I7 Mechanical Penetrations 12 Personnel Hatch I7 Steel Components (Group 2)
Cable Tray & Conduit Cable Tray & Conduit Supports Class 2 & 3 Component Supports Controlled Leakage Doors Crane Rails & Girders Electrical Racks, Panels & Cabinets Equipment Supports HVAC Duct Supports Instrument Racks Panels & Frames Jet Barriers Missile Shields Pipe Whip Restraints Sumnp Screens 16
Example License Renewal Technical Information Submittal November 4, 1996 1
Table 2.3-1 (continued)
Reactor Building Components and Their Intended Functions (working draft)
Key:
Structural function numbers identified in table correspond to functions list following table.
Shaded cells indicate that the component is not required to perform an intended function.
(Identre lnth not below)
Post Tensioning System Tendon Wires 2
Tendon Anchorage 2
Other Reactor Building Interior Components Lead Shielding Supports Fire Stops Shield Wall Tendons 2
3 Reactor Building Component Intended Functions:
4
- 1.
Provides essentially leaktight barrier to prevent uncontrolled release of radioactivity.
5
- 2.
Provides structural and/or functional support to safety-related SSCs. More specifically for the post-tensioning 6
system, this function means to impose compressive forces on the concrete containment structure to resist the 7
internal pressure resulting from a design basis accident with no loss of structural integrity.
8
- 3.
Provides shelter/protection to safety-related SSCs (including radiation protection).
9
- 4.
Provides rated fire barrier to confine or retard a fire from spreading to or from adjacent areas of the plant.
10
- 5. Serves as external missile barrier.
11
- 6.
Provides structural and/or functional support to non-safety related SSCs where failure of this structural component 12 could directly prevent satisfactory accomplishment of any of the required safety-related functions.
13
- 7.
Provides heat sink during design basis accidents or station blackout.
14 15 16 17 18 19 20 21 22 17
Example for License Renewal TeIcal Information Submittal October 23, 1996 1
Figure 2.3-1 2
Oconee Prestressed Concrete Reactor Building Personnel Buttress (Typ.)
Hatch Fuel Transfer Penetration Dome Equipment Hatch Steel Brackets to Support Crane Rail I
II Buttress I
I 11 II Cylinder Wall Grade Floor
.77777 Foundation Slub Lower Ien(on Access Gallery ock 18
0 0
Example for License Renewal Technical Information Submittal October 23, 1996 1
Figure 2.3-3 2
Attachments - Anchorages Across Liner 3
Cadweld Connector Leak Chase System Liner Plate 54 Thickened Plate 20
Example for License Renewal Technical Information Submittal October 23, 1996 Figure 2.3-4 2
Personnel Hatch 3
oncrete wall Inside Reactor Building aner Door Thickened Liner Plate Airlock Length Gussc( Anchors 21
Example for License Renewal Tec ical Information Submittal October 23, 1996 Figure 2.3-5 2
Equipment Hatch 3
Concrete Wall Thickened Wall Inside Containment Equipment Hatch Thickened Liner Plate Liner Plate 22
Example for License Renewal Tec ical Information Submittal October 23, 1996 Figure 2.3-6 2
Mechanical Penetrations - Single Barrier 3
Concrete WaU Report Boundary (excludes weld)
Anchors Pipe Cap, Dished Head, or Closure Plate (Out of Scope)
Ot ofSScope)
Sleeve Full Penetration Weld Lin(outr Scope)
Full Penetration Weld Liner (in Scope)
Full Penetration Weld 23
Example for License Renewal Te ical Information Submittal October 23, 1996 Figure 2.3-7 2
-Jerrode (Out of Scope)
Pe
-- p or AWE (Out of(Scope)
Us&. <. S.
eLcfbdC (In Scope) tofScope) 24
Example for License Renewal Tecnical Information Submittal October 23, 1996 Figure 2.3-8 2
Electrical Penetration 3
Report Boundary (includes weld)
Concrete WaU O-Ring Seal a A bSteel Header Plate (In Scope)
Outside Reactor uilding I so L
riside Reactor Building Liner Plate Full Penetration Weld 25
Example for License Renewal T nical Information Submittal October 23, 1996 Figure 2.3-9 2
Fuel Transfer Tube Penetration 3
Concrete Wall Test Connections Report Boundary Sec losure Plate (In Scope)
Fuel Transfer Tube (Out of Scope)
Outside Containment
.Inside Containment Full Penetration Weld()
(Considerd Process Pipe Weld)
(Out of Scope)
- 0 Liner Full Penetration Weld (In Scope) 26
Example for License Renewal Te ical Information Submittal October 23, 1996 Figure 2.3-10 2
Typical Post-Tensioned Tendon Assembly 3
T urnpet WASHER NUT; INSTALL COMPOSITC W.ASHCR AFTER FIELD HEADING B' END GCAt.WCtD.
TRANSITION FUNNEL I-WASHER.SHOP INST u 5a
-f cc uSc.essi g
Itte BEA^RING PLATE
- 0.
TYP. E.A. END 6
Field End Shop End 27