ML20148E524
| ML20148E524 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 05/21/1997 |
| From: | SOUTHERN NUCLEAR OPERATING CO. |
| To: | |
| Shared Package | |
| ML20148E520 | List: |
| References | |
| NUDOCS 9706030147 | |
| Download: ML20148E524 (168) | |
Text
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SOUTHERN NUCLEAR OPERATING COMPANY INSERVICE INSPECTION PROGRAM SECOND 10-YEAR INTERVAL VOGTLE ELECTRIC GENERATING PLANT I
UNITS 1 AND 2 i
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SNC - GENERAL OFFICE SNC - VOOTLE PLANT PREP'D REV'D APPV. APPV. APPV. APPV.
BY BY BY VOGTLE MGR. PLANT REV DATE DESCRIPTION (ITS) (ITS) (ITS) PROJECT ENGRG. GEN-(NMS) SUPP. hjGR; j 0 5/9/97 BASIC ISSUE *
- M I"akfd( h [hilD
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- Multiple preparers and reviewers. Preparation /redew sheets on file.
l 9706030147 970529 PDR ADOCK 05000424 O PDR
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{ VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM
!( LIST OF EFFECTIVE PAGES Distribution, Rev. 0 Table of Contents, Rev 0
Introduction:
Page 1-1, Rev. O Page 1-2, Rev. O Page 1-3, Rev. O Page 1-4, Rev. O Page 1-5, Rev. O Page 1-6, Rev. O Page 1-7, Rev. O Class 1:
Page 2-1, Rev. O Page 2-2, Rev. O Page 2-3, Rev. O Page 2-4, Rev. O Page 2-5, Rev. O O Page 2-6, Rev. O Page 2-7, Rev. O Page 2-8, Rev. O Page 2-9, Rev. O Page 2-10, Rev. O Page 2-11, Rev. O Page 2-12, Rev. O Page 2-13, Rev. O Page 2-14, Rev. O Class 2:
Page 3-1, Rev. O Page 3-2, Rev. O Page 3-3, Rev. O Page 3-4, Rev. O Page 3-5, Rev. O Page 3-6, Rev. O Page 3-7, Rev. O Rev.O
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4 VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM i
LIST OF EFFECTIVE PAGES
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(Continued) i Class 3:
i Page 4-1, Rev. 0 Page 4-2, Rev. O
- Page 4-3, Rev. 0 Page 4-4, Rev. 0 Component Supports
- 1 L Page 5-1, Rev. 0 Page 5-2, Rev. 0 )'
! Classes MC and CC:
, Page 6-1, Rev. 0 l
l Requests for Relief.-
Page 7-1, Rev. O
]
l-j Page 7-2, Rev. 0 l
Page 7-3, Rev. O
- Page 7-4, Rev. O q Page 7-5, Rev 0
- V Page 7-6, Rev. O
- Page 7-7, Rev. O Page 7-8, Rev. O 1 Page 7-9, Rev. O
{ Page 7-10, Rev. 0 4 Page 7-11, Rev. O j Page 7-12 Rev. 0 j Page 7-13, Rev. O i l Page 7-14, Rev. O !
I Page 7-15, Rev. O j Page 7-16, Rev. 0
} Page 7-17, Rev. 0
- Page 7-18, Rev. O
! Page 7-19, Rev. O Page 7-20, Rev. O Page 7-21, Rev. O Page 'i-22, Rev. O Page 7-23, Rev. 0 ,
Page 7-24, Rev. O Page 7-25, Rev. 0 O
Rev.0
VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM LIST OF EFFECTIVE PAGES O -
(Continued)
Requests for Relief (continued): .
Page 7-26, Rev. 0 Page 7-27, Rev. 0 -
Page 7-28, Rev. O Page 7-29, Rev. O Page 7-30, Rev. O Page 7-31, Rev. O Page 7-32, Rev. O Page 7-33, Rev. O Page 7-34, Rev. O Page 7-35, Rev. O Page 7-36, Rev. O Page 7-37, Rev. O I Page 7-38, Rev. O Page 7-39, Rev. O Page 7-40, Rev. O Page 7-41, Rev. O Page 7-42, Rev. O i Page 7-43, Rev. 0 O- Page 7-44, Rev. O Page 7-45, Rev. O Page 7-46, Rev. O Page 7-47, Rev. O Page 7-48, Rev. O Page 7-49, Rev. O Page 7-50, Rev. O Page 7-51, Rev. O l Page 7-52, Rev. O Page 7-53, Rev. 0 :
Page 7-54, Rev. O Page 7-55, Rev. O j Page 7-56, Rev. 0 -
l Page 7-57, Rev. O j Page 7-58, Rev. 0 l Page 7-59, Rev. 0 :
Page 7-60, Rev. O I Page 7-61, Rev. 0 l Page 7-62, Rev. O Page 7-63, Rev. O Page 7-64, Rev. O Page 7-65, Rev. O 1
Rev.0
VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM LIST OF EFFECTIVE PAGES
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(Continued) l Requests for Relief (continued):
Page 7-66, Rev. 0 ,
i Page 7-67, Rev. O Page 7-68, Rev. 0 Page 7-69, Rev. 0 l Page 7 70, Rev. 0 -
l Page 7-71, Rev. O Page 7-72, Rev. 0 l Page 7-73, Rev. O
- Page 7-74, Rev. O I
Page 7-75, Rev. O Page 7-76, Rev. O Page 7-77, Rev. 0 l Page 7-78, Rev. O
- Page 7-79, Rev. O
! Page 7-80, Rev. 0
- Page 7-81, Rev. O Page 7-82, Rev. O Page 7-83, Rev. 0 O
' Page 7-84, Rev. O Page 7-85, Rev. O Page 7-86, Rev. O Page 7-87, Rev. 0 )
Page 7-88, Rev. O Page 7-89, Rev. O !
Page 7-90, Rev. O Page 7-91, Rev. O Page 7-92, Rev. O !
j Page 7-93, Rev. O l Page 7-94, Rev. 0 Page 7-95, Rev. O Page 7-96, Rev. O Page 7-97, Rev. O Page 7-98, Rev. O __
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Page 7-99, Rev. O Page 7-100, Rev. O Page 7-101, Rev. O Page 7-102, Rev. O j Page 7-103, Rev. O Page 7-104, Rev. O I
Page 7-105, Rev. O Rev.0 l
l l VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM l
LIST OF EFFECTIVE PAGES (Continued)
Requests for Relief (continued):
f Page 7-106, Rev. O Page 7-107, Rev. O Page 7-108, Rev. 0 l Page 7-109, Rev. 0 l Page 7-110, Rev. O Page 7-111, Rev. O Page 7-112, Rev. 0 l Page 7-113, Rev. O i Page 7-114, Rev. 0 l Page 7-115, Rev 0 l Page 7-116, Rev. 0 1
Page 7-117, Rev. O I Page 7-118, Rev. 0 l Page 7-119, Rev. O Page 7-120, Rev. O Page 7-121, Rev. O ;
Page 7-122, Rev. O. l l Page 7-123, Rev. O !
!. Page 7-124, Rev. 0 '
Page 7-125, Rev. O Page 7-126, Rev. O l
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VEGP-1 AND 2 INSERVICE INSPECTION PROGRAM DISTRIBIJrION LIST l
MANUAL HOLDER MANUAL NUMBER l
, Authorized Nuclear Inservice Inspector,- 1 l VEOP l
J. J. Churchwell 2 i
D. R. Cordes 3 D. R. Graham 4 l
ITS Shelf Copy 5 1
-i ITS Shelf Copy ~ 6 '
Manager, SNC Vogtle Project Nuclear 7 i Maintenance and Support Manager, SCS Vogtle Project Support 8 Supervisor, SNC Document Control, 9 VEGP l Supervisor, SNC Engineering Support 10 (Performance Group), VEGP l Functional Maintenance Team Leader, SNC 11 VEGP Rev.0 4
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VEGP-1 AND 2INSERVICEINSPECTIONPROGRAM TABLE OF CONTENTS I
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j SECTION PAGE i
j Introduction 1-1
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! Class 1 Systems and Components 2-1 1
1 Class 2 Systems and Components 3-1 i Class 3 Systems and Components 4-1 Class 1,2, and 3 Component Supports 5-1 -
Classes MC and CC 6-1 !
Requests for Relief 7-1 Rev.O
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i VOGTLE ELECTRIC GENERATING PLANT !
ls . . UNITS 1 AND 2 i
i INSERVICE INSPECTION PROGRAM -
SECOND TEN-YEARINTERVAL l
4 l.0 . INTRODUCTION
- i 1.1 General i i ;
j This inservice inspection (ISI) program was developed by Southern Nuclear Operating ;
i . Company (SNC) as the licensee and operator of the Georgia Power Company-owned
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Vogtle Electric Generating Plant (VEGP), Units I and 2. It was developed to comply ;
with the requirements of the American Society ofMechanical Engineers (ASME) Boiler .
' and Pressure Vessel Code,Section XI,1989 Edition, to the extent practicable. Inspection !
! Program "B" will be used as defined by IWA-2400 of ASME Section XI. Where the l Code requirements are impractical, requests for relief have been developed to provide for !
l alternative examinations and/or tests. In certain instances, requests for relief have been t'eveloped proposing alternatives to the Code requirements which would provide an -
i acceptable level of quality and safety. Where applicable, the use of requests for relief and !
j Code Cases are documented in tables which accompany each individual Code class ,
i discussed herein. Requests for relief are numbered RR-1, RR-2, etc. Some of the t
[ ' requests for relief may be generic in nature and may not be applicable to a particular Code ;
i category. For the convenience of the Nuclear Regulatory Commission (NRC) reviewer, !
j the following are provided relative to the enclosed requests for relief:
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(1) In requests for reliefwhich are similar to those which were submitted for the first ten-year inspection interval, the previous requests for relief number and the date ofNRC
- correspondence which provided the safety evaluation for the previous requests for relief !
l are specifically identified, and ;
' (2) For requests for relief requesting use of ASME Section XI Code Cases which have not !
been previously approved by the NRC, a copy of the respective Code Case is provided as l part of the request for relief package. j The contents of this document are subject to change (with approval) during the course of inservice inspections.
l This program document is applicable for the second ten-year inspection interval. The . !
second ten-year inspection interval begins on May 31,1997 and ends on May 30, 2007. l In order to use the same edition of the Code for both VEGP units, SNC requests relief l i
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1.1 General (continued) from the NRC to update VEGP-2 early. Therefore, relief has been requested (see RR-1)
O to atiew this eerir undate aed for 1he remeinder of vieet iife. unen aggrovai ef the request for relief, the second ten-year inspection interval for VEGP-2 will commence on May 31,1997 and end on May 30,2007 -(the same as for VEGP-1). i Code Cases or Requests for Relief that may be applied for specific Code Categories and
' Items are listed in the tables which accompany each Code Class section in this program document. Code Cases that are generic in nature and which would be applied on a case-by-case basis are not listed in those tables; but, are identified in inservice inspection plans
! as required by IWA-2420(a)(6). Code Cases applied at VEGP during the second ten-year
- inspection interval will be those initially selected from NRC Regulatory Guide 1.147, l Revision 11, October 1994, unless relief has been received by SNC to use a Code Case not listed in the aforementioned NRC Regulatory Guide. As new or revised Code Cases are approved for use through subsequent revisions to NRC Regulatory Guide 1.147, such l Code Cases may be used at VEGP when deemed appropriate by SNC.
l 1.2 Scoops This document is a description of the Inservice Inspection Program for Class 1,2, and 3 l components. Class MC and CC components will be addressed in a separate program
' l document as discussed in section 1.12.
1.3 Component Upgrading Plant components have been reviewed to determine the appropriate classification for inservice inspection. The classification is given in the Line Designation Lists and Equipment Designation Lists which are to be located in the Inservice Inspection Plan documents for VEGP-1 and 2. It must be noted, however, that the classification of components as ISI Class 1,2, or 3 for insenice inspection does not imply that the I
components were designed or constructed in accordance with the same ASME classification requirements. The component design codes remain as stated in the VEGP Final Safety Analysis Report (FSAR) and updates thereto.
1.4 Responsibility i
l Southern Nuclear Operating Congany bears the overall responsibility for the performance
! of the insenice inspections. Certain nondestructive examinations will be performed by a l qualified examination agency. The results of such examinations will be reported to SNC 3
for final evaluations and disposition. SNC has arrangements with an Authorized
- Inspection Agency to provide inspection services at VEGP.
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1 1.5 Records j -(~ Records and documentation ofinformation and examination results, which provide the basis for evaluation and which facilitate comparison with results from subsequent ,
4 mspections, will be available for the active life of the plant. !
L j 1.6 Methods of Examination '
4 The method of examination planned for each area is delineated in subsequent sections.
j Personnel performing nondestructive examinations will be trained in accordance with the American Society for Nondestructive Testing (ASNT) and the ASME Code. Pursuant to the requirements of the 1989 Edition of ASME Section XI, ASNT document SNT-TC-(
1 A,1984 Edition, is the governing document for quahfying personnel to perform I nondestructive inspections, examinations, or testing at VEGP.
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' 1.6.1 Eddy Current i
j Eddy current examinations will be performed on the steam generator tubing in accordance j with the requirements of VEGP Improved Technical Specification 5.5.9 and NRC j
Regulatory Guide 1.83, Revision 1. The alternative techniques used for this examination,
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as permitted by paragraph IWA-2240, either satisfy or exceed the requirements of
). Appendix IV of ASME Section XI and/or Article 8, Appendix I of ASME Section V. l
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1.6.2 Liquid Penetrant Dye penetrant examinations will be performed whenever a surface examination is required
! on non-magnetic components.
1 i 1.6.3 Mannetic Particle 1 Magnetic particle tests will typically be used when surface examine. tion of carbon steel l j components is required. .
i I j 1.6.4 Radion_ra_ohv .
{ Radiography may be used as an alternative method to ultrasonic examinations. !
j 1.6.5 Ultrasonic Examination Ultrasonic examinations will be conducted in accordance with the provisions of Article I-2000 of ASME Section XI. The reactor pressure vessel will be examined in accordance with the requirements of NRC Regulatory Guide 1.150, Revision 1, to the extent practical. 1 0
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1.6.6 Visual Examinations 4
O A visual examination (VT) wilfbe used to provide evidence ofleakage or to provide a ;
i report on the general condition of a component.
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- a. The VT-1 visual examination shall be conducted to determine the condition of :
the part, component, or surface examined , including such conditions as ;
i cracks, wear, corrosion, erosion, or physical damage on the surfaces of the l part or components.
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- b. The VT-2 visual examination shall be conducted to locate evidence ofleakage from pressure-retaining components, or abnormal leakage from components j with or without leakage collection systems as required during the conduct of system pressure or functional test.-
, c. The VT-3 visual examination shall be conducted to determine the general i j '
mechanical and structural condition ofcomponents and their supports, such as the verification of clearances, settings, physical displacements, loose or missing parts, debris, corrosion, wear, erosion, or loss ofintegrity at bolted or j welded connections. Further, the VT-3 visual examination shallinclude examinations for conditions that could affect operability or functional
- adequacy of snubbers, and constant load and spring-type supports. ,
1.7 Evaluation ofExamination Results l l
- Examination results are evaluated per IWA-3000, IWB-3000, IWC-3000, and IWF-3000 i of ASME Section XI. Article IWD-3000, " Acceptance Standards" for flaw indications l
, are in the course of preparation and, as yet, are not available for use. Therefore, the rules ;
- of Article IWB-3000 may be used for Class 3 components.
1.8 Renair and Reolacement Procedure i
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The ASME Section XI repair / replacement requirements are controlled and implemented by plant Maintenance Department procedures. These requirements are documented in ;
l Section 3 to the VEGP Welding Manual (GEN-25). Repairs and replacements for Code Class 1,2, and 3 and their supports will be performed to the requirements of the 1989 Edition of ASME Section XI except where requests for relief have been submitted to and approved by the NRC. ;
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l.9 Limitations of Examinations O Certain limitations to the nondestmetive examination of welds due to geometric configuration or inaccessibility were identified during the Preservice Inspection and during Inservice Inspection activities conducted during the first ten-year inspection interval.
During ISI activities, the required examination will be accomplished to the maximum extent possible and limitations will be documented in requests for reliefwhere examination coverage is not at least ninety percent (90%) or greater. By ASME Section XI Code Case N-460 and as accepted by the NRC, a full Code examination involves examination coverage of ninety percent (90%) or greater. Known requests for relief for components having examination coverage less than 90%, including those for the reactor pressure vessel, are contained in the Requests for Relief section of this program document.
The inservice inspection program outlined in the attached tabulations have been developed as a result of a design review, the Preservice Inspections, and Inservice Inspections performed during the first ten-year inspection interval.
1.10 Auzmented Examinations The NRC has required that certain augmented examinations be performed to assure structural reliability. The areas ofinterest and the examinations to be performed are discussed below:
1.10.1 Reactor Coolant Pump Flywheels The Reactor Coolant Pump Flywheels shall be examined per the recommendations of Regulatory Position C.4.b ofNRC Regulatory Guide 1.14, Revision 1, August 1975, as required by VEGP Improved Technical Specification 5.5.7.
1.10.2 Main Steam and Feedwater Piping Examinations The four main steam lines and feedwater lines from the containment penetration flued head outboard welds up to the first five-way restraint shall be examined as required by VEGP Improved Technical Specification 5.5.16. The extent of the inservice examinations completed during each ten-year inspection interval shall provide one hundred percent (100%) volumetric examination of circumferential and longitudinal piping welds to the extent practical. This augmented inservice inspection is consistent with the requirements ofNRC Branch Technical Position MEB 3-1, " Postulated Break and Leakage Locations in Fluid System Piping Outside Containment," November 1975, and Section 6.6 in the updated FSAR.
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i 1.10.3 Snubbers O s==66 r i=>t iied e r tv-rei t d r t - . - ii ===66 r i=>t ii d o# o"- rety-related systems, whose failure would adversely affect a safety-related system shall be examined. Snubber examination and functional testing will be implemented in accordance with the VEGP Snubber Program as discussed in Request for ReliefRR-29. The VEGP Snubber Program may be found in section 13.7.2 of the VEGP Technical Requirements Manual.
1.11 Requests for Relief During the course of the preservice inspection and subsequent inservice inspection activities which were conducted during the first ten-year inspection interval, examination areas were identified where total compliance with the requirements of the ASME Code were not achieved. Requests for reliefwere prepared for each of these areas. In addition, requests for reliefwere added following the preservice inspection and during inservice inspection activities in order to use ASME Section XI Code Cases which were advantageous to use. The requests for relief address the area ofrelief, ASME Code, examination requirements, Code Item Number and Category (if applicable), basis for relief alternate examination (if any), and implementation schedule (if applicable). Each request for reliefis identified by a unique number and is contained in Section 7 of this document.
Requests for relief that pertain to a particular category and code item number are listed in the Class 1,2,3, and Support tables opposite the appropriate item number.
O i.12 Ciass MC and CC Comoonent,
.Qnly ASME Code Class 1, 2, and 3 components and their supports are included in the scope of this prognun document.
Southern Nuclear Operating Company acknowledges that as a result of a September 9, 1996 rulemaking, the NRC is requiring that licensees revise their ISI programs to incorporate by reference the 1992 Edition with 1992 Addenda ofASME Section XI, Subsection IWE, " Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants", and Subsection IWL, " Requirements for Class CC Concrete Components ofLight-Water Cooled Power Plants", with specific modifications and a limitation which are in addition to the Code requirements. Subsection IWE provides criteria for the visual examinat:on of the steel liner of the containment, pressure-retaining bolting, and seals and gaskets. A volumetric examination may also be required on the steel liner if degradation problems are observed during the visual examinations. Subsection IWL provides for the visual examination of concrete pressure-retaining shells, shell components, and for the examination of unbonded post-tensioning systems. In accordance with the provisions of 10 CFR 50.55a(g)(6)(ii)(B)(4), VEGP will continue implementation of the existing, previously approved, post-tensioning system surveillance program. Expedited examinations as addressed by the rulemaking are to be completed by September 9,2001.
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L 1.12 Class MC and CC Components (Continued) l In order to comply with the miemaking, a program for the examination and testing of !
2 ASME Code Class MC and CC components is in the process of being developed for
- VEGP-1 and 2. Once developed, the program will be documented separately from this l program document which is for Class 1,2, and 3 components and their supports. The use i of separate program documents is preferred administratively because Class MC and CC
- components and Class 1,2, and 3 components and their supports are to be examined to different editions and/or addenda ofASME Section XI. As noted herein, Class MC and CC components will be examined to a program developed using the requirements of the i
1992 Edition of ASME Section XI with 1992 Addenda while the program developed for
- Class 1,2, and 3 components and their supports will be examined to the requirements of the 1989 Edition of ASME Section XI, to the extent practicable. Any requests for relief
- identified during the development of the program document for Class MC and CC components will be submitted to the NRC pursuant to 10 CFR 50.55a for review and
, approval. Once developed, the program document for Class MC and CC components is not required to be submitted to the NRC, but will be maintained at the VEGP plant site for j review by the NRC upon request. Maintenance of the program document at the plant site i
for review upon request is consistent with the requirements of the rulemaking.
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) Any repairs or replacements to Class MC and CC components will be performed in 2
accordance with the requirements of the 1992 Edition with 1992 Addenda ofASME Section XI except where relief from the Code requirements might be necessary. As a lV j
q result of communications between the NRC and the Nuclear Energy Institute (NEI) concermng the rulemaking, including repair and replacement requirements, Georgia Power Company, the former licensee for VEGP, submitted letter LCV-0968 dated February 17, j 1997 to the NRC requesting that it be allowed a one year extension (from the effective date of the rule) in order to develop and implement a repair / replacement program for
- Class MC and CC components. The development and implementation of a repair / replacement program to address these particular components is expected to be completed by September 9,1997.
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2.0 CLASS 1 SYSTEMS AND COMP _ONENTS 2.1 Purpose The purpose of this section is to define an inservice inspection program for Class 1
- systems and components to meet the requirements of the ASME Boiler and Pressure f
Vessel Code,Section XI,1989 Edition, to the extent practicable.
i 2.2 Insocction Schedule As much as practicable, Class 1 systems and components shall be scheduled for
[ examinatiora using Inspection Program "B" as defined in ASME Section XI, IWB-2412.
2.3 Insoection Scope ,
Areas subject to inservice inspection are shown in the following table by examination category.
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T/ E1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoory Description Areafs) to Be Examined Examination Examinations B1.11 B-A Reador Pressure Vessel Circumferential Shell Welds Volumetric RR-3 s
B1.12 B-A Reador Pressure Vessel Longitudinal Shell Weld Volumetric RR-3 B1.21 B-A Reador Pressure Vessel Circumferential Head Welds Volumetric RR-3 and Closure Head B1.22 B-A Reactor Pressure Vessel Meriodional Head Welds Volumetric No and Closure Head B.1.30 B-A Reador Pressure Vessel Shell-to Flange Weld Volumetric No B1.40 B-A Reador Pressure Vessel Head-to-Flange Weld Surface and RR-4 Closure Head Volumetric B1.51 B-A Reactor Pressure Vessel Beltline Region Repair Volumetnc Not applicable to Welds either Vogtle unit t
B2.11 B-B Pressurizer Shell-to-Head Welds Volumetric RR-7 ,
B2.12 B-B Pressurizer She!I-to-Head Welds Volumetric No B2.20 B-B Pressurizer Head Welds Volumetric Not applicable to either Vogtle unit B2.30 B-B Steam Generator Head *!eMr Volumetric Not app:icable to (Primary Side) eitherVogtle unit I
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TA' 1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoory Description Area (s) to Be Examined Examination Examinations B2.40 B-B Steam Generator Tubesheet-to-Head Welds Volumetric RR-6 (Primary Side)
B2.50 B-B Heat Exchangers Head Welds Volumetric Not applicable to (Primary Side) either Vogtle unit B2.60 B-B Heat Exchangers Tubesheet-to-Shell Volumetric Not applicable to (Primary Side) (or Head) Welds either Vogtle unit B2.70 B-B Heat Exchangers Tubesheet-to-Shell Volumetric Not applicable to (Primary Side) (or Head) Welds either Vogtle unit !
B2.80 B-B Heat Exchangers Tubesheet-to-Shell Volumetric Not applicable to (Primary Side) (or Head) Welds eitherVogtle unit i
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I o ISI PROGRAM FOR ASME CLASS 1 COMPONENTS T 3, o VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative -
Item No. Cateaory Description Area (s) to Be Examined Exammation Exammations '
B3.90 B-D Reactor Pressure Nozzle-to-Vessel Welds Volumetric RR-2 Vessel B3.100 ~ B-D Reactor Pressure Nozzle inside Radius Volumetric RR-2 Vessel Section B3.110 B-D Pressurizer Nozzle-to-Vessel Welds Volumetric RR-7 B3.120 B-D Pressurizer Nozzle inside Radius Volumetric RR-7 i Section B3.130 B-D Steam Generators Nozzle-to-Vessel Welds Volumetric Not applicable to (Primary Side) estherVogtle unit.
B3.140 B-D Steam Generators Nozzle inside Radius Volumetric RR-6 (Primary Side) Section B3.150 B-D Heat Exchangers Nozzle-to-Vessel Welds Volumetric Not applicable to (Primary Side) either Vogtle unit.
B3.160 B-D Heat Exchangers Nozzle inside Radius Volumetric Not applicable to (Primary Side) Section either Vogtle unit. ,
B4.11 B-E Vessel Nozzles Extemal Surfaces of Visual, VT-2 RR-23, RR-27 Partial Penetration Welds .
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TA 1 ;
ISI PROGRAM FOR ASME CLASS 1 COMPONENTS ;
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. _Cateoory Descnotion Areafs) to Be Examined Examination Examinations i
B4.12 .B-E Control Rod Drive Extemal Surfaces of Visual, VT-2 RR-23, RR-27 [
i Nozzles Partial Penetration Welds i
B4.13 B-E instrumentation Extemal Surfaces of Visual, VT-2 RR-23, RR-27 .
Nozzles Partial Penetration Welds B4.20 B-E F <emrizer Extemal Surfaces of Visual, VT-2 RR-23 RR-27 Hester Penetration Welds ;
i B5.10 B-F Reactor Pressure Vessel Nozzle-to-Safe End Butt Surface and RR-2 1 Welds (Nominal Pipe Volumetric Size 14 in.) :
I 135.20 B-F Reactor Pressure Vessel Nozzle-to-Safe End Butt Surface Not apptele to i Welds (Nommel Pipe either Vogtle unit Size <4 in.)
i 05.30 B-F Reactor Pressure Vessel Nozzle-to-Safe End Surface Not applicable to Socket Welds either vogtle unit ;
E,5.40 B-F Pressurizer Nozzle-to-Safe End Butt Surface and RR-7 t Welds (Nommel Pipe Volumetric !
Size 14 in.) ,
t B5.50 B-F Pressurizer Nozzle-to-Safe End Butt Surface Not applicable to Welds (Nominal Pipe either Vogtle unit
- Size <4 in.) j B5.60 B-F Pressurizer Nozzle-to-Safe End Surface Not applicable to }
Socket Welds either Vogtle unit i I
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t
. .. . _ . m .. _ _. _ _ _ -_ -. . _ .... _ _ _ .. ____.__.___.m_._..-
O TAQ , O ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIO GENERATING PLANT, UNITS 1/J4D 2 Examination System or Component Method of Altemative item No. Cateoorv Descnotion Area (s) to Be Examined Examination Exammatens B5.70 B-F Steam Generators Nozzle-to-Safe End Butt Surface and No (Primary Side) Welds (Nominal Pipe Volumetric Size 14 in.)
B5.80 B-F Steam Generators Nozzle-to-Safe End Butt Surface Not applicable to (Primary Side) Welds (Nominal Pipe either Vogtle unit Size <4 in.)
B5.90 B-F Steam Generators Nozzle-to-Safe End Surface Not applicable to (Primary Side) Socket Welds eMher Vogtle unM B5.100 B-F Heat Exchangers Nozzle-to-Safe End Butt Surface and Not applicable to (Primary Side) Welds (Nommel Pipe Volumetric edherVogtle unit Size 14 in.)
B5.110 B-F Heat Exchangers Nozzle-to-Safe End Butt Surface Not apphcable to (Primary Side) Welds (Nominal Pipe esther Vogtle unit Size <4 in.)
B5.120 B-F Heat Exchangers Nozzle-to-Safe End Surface Not applicable to (Primary Side) Socket Welds eRherVogtle unit B5.130 B-F Piping Dissmular Butt V* elds Surface and Not applicable to (Nominal Pipe Size Volumetric eitherVogtle unit 14in.)
B5.140 B-F Piping Dissimdar Butt Welds Surface Not apphcable to (Nominal Pipe Size either Vogtle unit
<4 in.)
B5.150 B-F Piping Dissimilar Metal Socket Surface Not applicable to Welds eitherVogtle unit 24 Rev.0
TA 1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateaory Description Areafs) to Be Examined Examensbon Examinations B6.,10 B-G-1 Reactor Pressure Vessel Closure Head Nuts Surface RR-5 B6.20 B-G-1 Reactor Pressure Vessel Closure Studs,in Place Volumetric Not applicable to either Vogtle unit .
B6.30 B-G-1 Reactor Pressure Vessel Closure Studs, When Removed - Surface and No Volumetric B6.40 B-G-1 Reactor Pressure Vessel Threads in Flange Volumetric No B6.50 B-G-1 Reactor Pressure Vessel Closure Washers, Bushings Visual, VT-1 No B6.60 B-G-1 Pressurizer Bolts and Studs Volumetnc Not applicable to either Vogtle unit B6.70 B-G-1 Pressurizer Flange Surface, When Visual, VT-1 Not apphcable to Connecteon Disas%mbled either Vogtle unit 96.80 B-G-1 Pressurizer Nuts, Bushings, and Visual, VT-1 Not applicable to Washers either Vogtle unit B6.90 B-G-1 Steam Generator Bolts and Studs Volumetric Not applicable to either Vogtle unit B6.100 B-G-1 Steam Generator Flange Surface, When Visual, VT-1 Not applicable to Connection Disassembled either Vogtle unit B6.110 B-G-1 Steam Generator Nuts, Bushings, and Visual, VT-1 Not applicable to Washers either Vogtle unit B6.120 B-G-1 Heat Exchangers Bolts and Studs Volumetric Not applicable to either Vogtle unit 2-7 Rev.O
. _ . _ _ _ . _ _ . _ _ _ _ -. . . . _ _ . _ . . . _ . _ _ _ _ _ -... .__. ._ _ _- . _ - ....m . _ .__ _ ,
-TA 1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Akemative I Item No. Cateoorv Description Areafs) to Be Examined Examination Exami. nations B6.130 B-G-1 Heat Exchangem Flange Surface, When Visual, VT-1 Not apphcable to Connection Disassembled other Vogtle unR B6.140 B-G-1 Heat Exchangers Nuts, Bushings, and Visual, VT-1 Not apphcable to Washers other Vogtle unit B6.150 B-G-1 Piping Bots and Studs Volumetric Not apphcable to ;
either vogue unit t B6.160 B-G-1 Piping Flange Surface, When Visual, VT-1 Not apphcable to Connechon Disassembled other Vogtle unR B8.170 B-G-1 Piping Nuts, Bushings, and Visual, VT-1 Not apphcable to Washers edher Vogtle unM B6.180 B-G-1 Pumps Bots and Studs Volumetnc No B6.190 B-G-1 Pumps Flange Gurface, When Visual, VT-1 No Connection Disassembled B6.200 B-G-1 Pumps Nuts Bushings, and Visual, VT-1 No Washers B6.210 B-G-1 Valves BoRs and Studs Volumetric Not apphcable to either Vogtle unM B6.220 B-G-1 Valves Flange Surface, When Visual, VT-1 Not apphcable to Connedion Disassembled edher Vogtle unit i B6.230 B-G-1 Valves Nuts, Bushings, and Visual, VT-1 Not apphcable to i
Washers edher Vogtle unit I
f
?
2-8 Rev.O t
. . - . - . ~ . - - - . - . - - , . . - . . . . . _ - - - - - . ~ - - . ~ . - - - . . - . . - . . . . . - . .
. . . ~. ~ ~ ~ ..
[
TA 1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 .
1 t
Examination System or Component Method of Allemative P ltem No. Cateoory Descriobon Areafs)to Be Examined Examination Examinsbons i
B7.10 B G-2 ' Reactor Pressure Vessel Bolts, Studs, and Nuts Visual, VT-1 Not applicable to either Vogtle unit
.3 B7.20 B-G-2 Pressurizer Bolts, Studs, and Nuts Visual, VT-1 No B7.30 B-G-2 Steam Generators Bolts, Studs, and Nuts Visual, VT-1 No B7.40 B-G-2 Heat Exchangers Bolts, Studs, and Nuts Visual, VT-1 Not applicable to !
either Vogtle unit ;
i
! B7.50 B-G-2 Piping Bolts Siuos, aisd Nuts Visual, VT-1 No j
- B7.60 B-G-2 Pumps Bolts, Ciuds, and Nuts Visual, VT-1 No
}
B7.70 B-G-2 Valves Bolts, Studs, and Nuts Visual, VT-1 No B7.00 B-G-2 CRD Housings Bolts, Studs, and Nuts Visual, VT-1 Not applicable to '
1 (when Disassemtsec, eitherVogtle unit ,
B8.10 B-H Reactor Premore Vessel integrally Welded Volumetnc or RR-4, RR-20 Attachments Surface as !
Applicable !
B8.20 B-H Pressurizer integrally Welded Volumetric or RR-8, RR-20 I Attachments Surface, as :
Apphcable j 4 i B8.30 B-H Steam Generators integrally Wolded Volumetnc or Not apphcable to j Attachments Surface, as eitherVogtle unit ;
t 2-9 Rev.O i
i I
)
_ _.- _ . . _ . . . . . . _ . . . _ . . . . . . _ . . - . _ _ . _ _ . _ . _ . _ . _ _ . . _ . . _ . . . _ . - _ _ , . _ . ~ . . . . . . . . . _ _ . . _ . , . . - _ . , . . . _ . , - , . . . _ . . . . . . _ . . . . , _ . .
r i
O O TALm 1 O-ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoory Description Ares (s) to Be Examined Examinatum Examinshons B8.40 B-H Heat Exchangers integrally Welded Volumetnc or Not applicable to Attachments Surface, as eitherVogtle unit Applicable B9.11 B-J Piping Circumferential Welds Surface and RR-10, RR-12, (Nominal Pipe Size Volumetric 14in.)
B9.12 B-J Piping Longitudinal Welds Surface and Not applicable to (Nominal Pipe Size Volumetnc either Vogtle unit 14in.)
89.21 B-J Piping Circumferential We!ds Surface No (Nominal Pipe Size
<4 in.)
B9.22 B-J Piping Longitudinal Welds Surface Not appl 6 cable to (Nommel Pipe Size eitherVogtle unit
<4 in.)
B9.31 B-J Piping Branch Pipe Connection Surface and RR-11 Welds (Nommal Pipe Volumetric Size 14 in.)
B9.32 B-J Piping Branch Pipe Connection Surface No Welds (Nominal Pipe Size <4 in.)
B9.40 B-J Piping Socket Welds Surface No 2-10 Rev.0
l O TA1 O-ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoofy Descnotion Area (s) to Be Examined Exammahon Examenations B10.10 B-K-1 Piping Integrally Welded Volumetnc or Not applicable to Attachments Surface, as either Vogtle unit Apphcable B10.20 B-K-1 Pumps Integrally We:Jed Volumetric or Not apphcable to Attachments Surface, as either Vogtle unit l
APPicable i B10.30 B-K-1 Valves Integrally Weined Volumetric or Not apphcable to Attachments Surface, as either Vogtle unit Applicable B12.10 B-L-1 Pumps Pump Casmg Welds Volumetrx: Not apphcable to eithervogtle unit B12.20 B-L-2 Pumps Pump Casing Vaual, VT-3 No B12.30 B-M-1 Valves Valve Body Welds Surface Not apphcable to (Nominal Pipe Size either Vogtle unit
<4 in.)
B12.40 B-M-1 Valves Valve Body Welds Volumetric . Not apphcable to (Nominal Pipe Size either Vogtle unit
>4 in.)
B12.50 B-M-2 Valves Valve Body Vsual, VT-3 No (Nominal Pipe Size
>4 in.)
2-11 Rev.O
.... .~ , .-.-..-.~.-.~~._..~ - -- -. -. .-- - - - . . ~ .
- - - - - - - . . . - -c . . . - .
TA 1 ISI PROGRAM FOR ASME Cl>SS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 i Examination System or Component Method of Altemative ;
Item No. Cateoorv Descriobon Areafs) to Be Examined Examination Examinations .
B13.10 B-N-1 Reactor Pressure Vessel VesselInterior Visual, VT-3 No i B13.20 B-N-2 Reactor Pressure Vessel interiorAttachments Visual, Vi-1 Not applicable to -
(BWR) Within Beltline Region either Vogtle unit B13.30 B-N-2 Reador Pressure Vessel interiorAttachments Vaual, VT-3 Not applicable to (BWR) Beyond Beltline Region either Vogtle unit t B13.40 B-N-2 Reactor Pressure Vessel Core Support Structure Visual, VT-3 Not applicable to (BWR) either Vogtle unit ;
B13.50 B-N-2 Reactor Pressure Vessel InteriorAttachments Visual, VT-1 No (PWR) WIthin Belthne Region j B13.60 B-N-2 Reactor Pressure Vessel interiorAttachments Visual, VT-3 No +
(PWR) Beyond Belthne Region !
l B13.70 B-N-3 Reactor Pressure Vessel Core Support S ncture Visual, VT-3 No ;
(PWR) (Removed) {
B14.10 B-0 Reactor Pressure Vessel Welds in Control Rod Volumetric or No Drive Housmo Surface f B15.10 B-P Reactor Pressure Vessel Pressure Retaining System Leakage RR-27 Boundary Test; Visual,
- VT-2 -
B15.11 B-P Reactor Pressure Vessel Pressure Retaining System Hydro- RR-23, RR-27 '
Boundary static Test; Vsual, VT-2 i
2-12 Rev.O I
I T 1 ISI PROGRAM FOR ASME CLASS 1 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative l Item No. Cateoory Description Area (s) to Be Examined Examina* ion ExaminatNm3 l
B15.20 D-P Pressurizer Pressure Retaining System Leaka0e RR-27 Boundary Test; Visual, VT-2 B15.21 B-P Pressurizer Pressure Retaining System Hydro- RR-23, RR-27 Boundasy static Test; Visual, VT-2 B15.30 B-P Steam Generators Pressure Retaining System Leak- RR-27 Boundary age Test; Visual, VT-2 B15.31 B-P Steam Generators Pressure Retaining System Hydro- RR-23, RR-27 Boundary static Test; Visual, VT-2 B15.40 B-P Heat Exchangers Pressure Retaining System Leak- Not applicable to Boundary age Test; evtherVogtle unit Visual, VT-2 B15.41 B-P Heat Exchangers Pressure Retaining System Hydro- Not applicable to Boundary static Test; Visual, either Vogtle unit VT-2 B15.50 B-P Piping Pressure Retaining System Leak- RR-27 Boundary age Test; Visual, VT-2 2-13 Rev.O
_,_._____-____._.._.__.________.______...__._,m._..
N l
1 I i
\
$ 3.0 CLASS 2 SYSTEMS AND COMPONENTS 3.1 Purnose i
The purpose of this section is to define an inservice inspection program for Class 2 systems and !
components to meet the requirements of the 1989 Edition of the ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition, to the extent practicable. Contingent upon NRC review and approval ofRequest for ReliefRR-17, SNC will not only include the Class 2 pressure- retaining austenitic -
l stainless steel piping welds greater than 4 inches NPS and less than 3/8 inches nominal wall thickness in ,
the total weld count to which the 7.5% sampling rate is applied as required by Examination Category C- _
F-1, but to also include these welds as part of the weld population from which the 7.5% sampl: of
) welds receiving nondestructive examination (NDE) is selected. The examinations will be distributed l among the Class 2 systems prorated, to the degree practical, on the number ofnonexempt austenitic ;
j stainless steel or high alloy welds in each system. Within a system, the examinations will be distributed j among terminal ends and structural discontinuities prorated, to the degree practicable, on the number of i j nonexempt terminal ends and structural discontinuities in that system; and within each system, !
examinations shall be distributed between line sizes prorated to the degree practicable. Structural )
! discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such
]
as elbows, tees, reducers, flanges, etc., conforming to ANSI B16.9), and pipe branch connections and '
j fittings. High pressure safety injection is not addressed in RR-17 as these welds will be selected for I examination and examined per the requirements of the 1989 Edition of the ASME Code, Table-2500-1, ;
!( Category C-F-1. High pressure safety injection is defined as the piping from the centrifugal charging i pumps to the Class I cold leg injection lines. Branch lines are included out to the first isolation valve.
Pump suction piping (Residual Heat Removal, reactor water storage tank, etc.) is treated as an ;
Emergency Core Cooling System (ECCS), but not included as high pressure safety injection. These j ECCS welds are included in RR-17.
3.2 Inspection Schedule !
l As much as practicable, Class 2 systems and components shall be scheduled for examination using I Inspection Program "B" as defined in ASME Section XI, IWC-2412.
l ,
a j 3.3 Insocction Scope i
?
Areas subject to inservice inspection are shown in the following table by examination category. ,
t 3-1 Rev.O
O O O -
TABLE 2 ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 '
Examination System or Component Method of Altemative item No. Cateoory Description Area (s) to Be Examined Examination Examinations C1.10 C-A Pressure Vessels Shell Circumferential Welds Volumetric RR-21 C1.20 C-A Pressure Vessels Head Circumferential Welds Volumetric RR-14, RR-21 C1.30 C-A Pressure Vessels Tubesheet-to-Shell Weld Volumetric RR-21 C2.11 C-B Pressure Vessels Nozzle-to-Shell (or Head) Surface No Weld Where Vessel Nominal Thickness < 1/2 inch .
C2.21 C-B Pressure Vessels Nozzle-to-Shell (or Head) Surface and RR-14 Weld for Vessels Wdhout Volumetric ,
Reinforcing Plates Where Vessel NominalThickness
>1/2 inch C2.22 C-B Pressure Vessels Nozzle inside Radius Volumetric RR-14 Section forVessels Wdhout .
Reinforcing Plates Where Vessel Nominal Thickness ,
>1/2 inch C2.31 C-B Pressure Vessels Reinforemg Plate Welds to Surface RR-14 Nozzle and Vessel Where Vessel Nominal Thickness >1/2 inch 3-2 Rev.O ;
I 9
L l
I
?
- 0) (
% Q/
TABLE 2 ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 ,
Examination System or Component Method of Altemative it*m No. Cateoory Description Area (s) to Be Examined Examination Examinations C2.32 ~C-B Pressure Vessels Nozzle-to-Shell (or Head) Volumetric No Welds for Vessels With Re-enforcing Plates When inside of Vesselis Accessible and Vessel NominalThickness
>1/2 inch ;
C2.33 C-B Pressure Vessels Nozzle-to-Shell (or Head) Visual, RR-27 Weld forVessel With VT-2 :
Reinforcing Plates When inside of Vesselin inaccessible and Vessel Nominal Thickness i > 1/2 inch.
C3.10 C-C Pressure Vessels Integrally Welded Surface RR-20, RR-21 Attachments !
I C3.20 C-C Piping Integrally Welded Surface RR-20 Attachments I
C3.30 C-C Pumps integrally Welded Surface RR-20 Attachments C3.40 C-C Valves Integrally Welded Surface Not applicable to Attachments either Vogtle unit ;
C4.10 C-D Pressure Vessels Bolts and Studs Volumetric Not applicable to either Vogtle unit 3-3 Rev.O i
i
i O O O TABLE 2 ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 :
Examination System or Component Method of Altemative It*m No. Cateoorv Descnotion Area (s) to Be Examined Examination Examinahons C4.20 C-D Piping Bolts and Studs Volumetric Not apphcable to eitherVogtle unit ;
C4.30 C-D Pumps Bolts and Studs Volumetric No C4.40 C-D Valves Bolts and Studs Volumetric Not apphcable to :
either Vogtle unit !
RR-12, RR-17, !
C5.11 C-F-1 SS Piping Circumferential Welds Susface and t 3/8 inch Nominal Wall Volumetric RR-18 y Thecaness for Piping
> NPS 4 j C5.12 C-F-1 SS Popmg Longitudinal Welds Surface and RR-19 .
t 3/8 inch Nominal Wall Volumetric l Theckness for Piping -
> NPS 4 '
SS Piping Surface and No i C5.21 C-F-1 Circumferential Welds
> 1/5 inch Nominal Wall Volumetric l Th6ckness for Piping ;
1NPS 2 and < NPS 4 :
i C5.22 C-F-1 SS Piping Longitudinal Welds Surface and RR-19 i
> 1/5 inch Nominal Wall Volumetric Theckness for Piping !
1NPS 2 and < NPS 4 C5.30 C-F-1 SS Piping Socket Welds Surface RR-18 i i
3-4 Rev.0 ,
_ . , _ ~ _ _ _ . . _ _ __..____..______.._..___m____.__._______.._.__________________________m_.___ __ _._ m__m___ _ _ _ _ _ _ _ _ _ _ ______________________2m__.____ _ ___.__ ____. .. .
O O TABLE 2 O
ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoorv Descnotion Areafs) to Be Examined Examination Examinations C5.41 .C-F-1 SS Piping Circumferential Welds Surface RR-18 Pipe Branch Connections (Nominal Pipe Size 12)
C5.42 C-F-1 SS Piping Longitudinal Welds Surface RR-19 Pipe Branch Connections (Nominal Pipe Size 12)
C5.51 C-F-2 CS Piping Circumferential Welds Surface and RR-16 E 3/8 inch NominalWall Volumetnc .:
Thickness for Piping ,
> NPS 4 C5.52 C-F-2 CS Pipeng Longitudmal Welds Surface and RR-19 .!
E3/8 inch Nommal Wall Volumetric Thickness for Piping
> NPS 4 i
C5.61 C-F-2 CS Piping Circumferential Welds Surface and No
> 1/5 inch Nominal Wall Volumetric i for Piping t NPS 2 and
- 1. NPS 4 i
C5.62 C-F-2 CS Piping Longitudinal Welds Surface and RR-19
> 1/5 inch Nominal Wall Volumetric Thickness for Piping i t NPS 2 and 5 NPS 4 C5.70 C-F-2 CS Piping Socket Welds Surface No 3-5 Rev.0 ,
I
. _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ - _ _ _ _ _ _ _ _ _ _ ____ _ _ _ _ _ _ _ - _ _ _ - _ _ __ _ _ _ _ . _ _ _ - . _ - _ - ._ - . . _ - - __ .__ -- .._.____ L
TABLE 2 ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 '
Examination System or Component Method of Altemative !
Itcm & Cateaory Description Areafs)to Be Examined Examination Examinations i C5.21 C-F-2 CS Piping Circumferential Welds Pipe Surface No ;
Branch Connections .
(Nominal Pipe Size ;> 2) l C5.02 C-F-2 CS Piping Longitcdinal Welds Pipe Surface RR-19 Branch Connections I (Nominal Pipe Size > 2)
C6.10 C-G Pumps Pump Casing Welds Surface RR-14 C61:0 C-G Valves Valve Body Welds Surface Not applicable to eitherVogtle unit ,
i C7.10 C-H Pressure Vessels Pressure Retaining System Pres- RR-27 !
Components sure Test; Visual, VT-2 l
C7.20 C-H Pressure Vessels Pressure Retaining System Hydro- RR-23, RR-27 (;
Components static Test; Visual, VT-2 ,
r C7.30 C-H Piping Pressure Retaining System Pres- No Components sure Test; Visual, VT-2 !
C7AO C-H Piping Pressure Retaining System Hydro- RR-23. RR-27
- Components static Test; Visual, VT-2 P
P 3-6 Rev.0
_ ..___-.._..m ._ . . . _ . . _ .__ _ . _ _ _ . _ _ _ . _ _ . _ . _ . - .m . . . _ _ _ . - . . ._ _ _ . . _ . _ . . _ _ _ . _ _ _ . - _ _ _ __ ...- .
O O O TABLE 2 ISI PROGRAM FOR ASME CLASS 2 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative liem No. Cateaory Description Area (s) to Be Examitted Examination Examinations C7.50 C-H Pumps Pressure Retaining System Pres- RR-27 Components 'sure Test; Visual, VT-2 C7.60 C-H Pumps oressure Retaining System Hydro- RR-23, RR-27 Components static Test; _
Visual, VT-2 C7.70 C-H Valves Pressure Retaining System Pres- RR-27 >
Components sure Test; Visual, VT-2 C7.80 C-H Valves Pressure Retaining System Hydro- RR-23, RR-27 Components static Test; Visual, VT-2 i
i 3-7 Rev.0 i
i k
_ _ _ _ . . _ _.______________________.______.._____...____________________.____.__.__m_______. _ ___ __ _____________m. _ _ _ _ _ _ _ _ . _ _ - . _ . - __
4.0 CLASS 3 SYSTEMS AND CO_MPONENTS 4.1 Purnose The purpose of this section is to define an inservice inspection program for Class 3 systems and components to meet the requirements of the ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition, to the extent practicable.
4.2 Inspection Schedule As much as practicable, Class 3 systems and components shall be scheduled for examination using Inspection Program "B" as defined in AShE Section XI, IWD-2412.
4.3 Inspection Scop _c j Areas subject to inservice inspection are shown in the following table by examination category. ;
2 1
I 1 i
!O i
t i
41 Rev.0 0
. . . . - . - . - . . _ - - . .. .. .-- - . -.-.-.~ . - - . .. . . - - - . ..
O O O TABLE 3 -
ISI PROGRAM FOR ASME CLASS 3 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative item No. Cateoorv Description Areafs) to Be Examined Examination Examinations DI.10 D-A Systems in Support of Pressure Retaining Visual, VT-2 RR-23, RR-27 Reactor Shutdown Function Components D1.20 D-A Systems in Support of integral Attachment- Visual, VT-3 RR-20 Reactor Shutdown Function Component Supports and Restraints D1.33 D-A Systems in Support of Integral Attachment- Visual, VT-3 RR-20 ,
Reactor Shutdown Function Mechanical and Hydraulic Snubbers D1.40 D-A Systems in Support of Integral Attachment- Visual, VT-3 RR-20 '
Reactor Shutdown Function Spring Type Supports D1.50 D-A Systems in Support of Integral Attachment- Visual, VT-3 Not applicable to Reactor Shutdown Function Constant Load Type either Vogtle unit Supports D1.60 D-A Systems in Support of Integral Attachment - Visual, VT-3 Not applicable to t Reactor Shutdown Function Shock Absorbers either Vogtle unit !
t t
i r
4-2 Rev.O
O
~
~
TABLE 3 .!
ISI PROGRAhi FOR ASME CLASS 3 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination . System or Component Method of Altemative '
itim No. Cateoorv Descnotion Areefs) to Be Examined Examinaten Exammations D2.10 D-B Systems in Support of Pressure Retaining Visual, VT-2 RR-23 RR-27 ECC, CHR, Atmosphere Components i Clear,up, and Rudor RHR D2.20 D-B Systems in Support of integral Attachment- Visual, VT-3 RR-20 ECG, CHR, Atmosphere Component Supports and Cleanup, and Reactor RHR Restraints ,
i i D2.30 D-B Systemsin Support of integral Attachment- Visual, VT-3 RR-20 '
- ECC, CHR, Atmosphere Mechanical and Hydraulic l Cleanup, and Reactor RHR Snubbers D2.40 D-B Systems in Support of Integral AttMment- Visual, VT-3 RR-20 t
- ECC, CHR, Atmosphere Spring Type Supports ,
Cleanup, and Reactor RHR D2.50 D-B Systems in Support of . Integral Attachment. Visual, VT-3 Not appbcable to ECC, CHR, Atmosphere Constant Load Type either Vogtle unit Cleanup, and Reactor RHR Supports !
i D2.60 D-B Systems in Support of Integral Attachment- Visual, VT-3 Not apphcable to i ECC, CHR, Atmosphere Shock Absorbers either Vogtle unit f Cleanup, and Reactor RHR b
i I
i i
l 4-3 Rei. 0 i i
I
. . . . . . . _ . . . . . - . - . . - . . . . . . - . - . - - - - . . . . . - - . . -.- .._- - . . - . -.. . - - - _ . ~ -- .
O 1 TABLE 3 ISI PROGRAM FOR ASME CLASS 3 COMPONENTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 i
Examination System or Component Method of Altemative ,
It m No. Cateaorv Description Area (s) to Be Examined Examination Examinations ,
D3.10 D-C Systems in Support of Pressure Retaining Visual, VT-2 RR-23, RR-27 RHR from Spent Fuel Components Storage Pool D3 20 D-C Systems in Support of Integral Attachment- Visual, VT-3 RR-20 RHR from Spent Fuel Component Supports and '
Storage Pool Restraints D3.30 D-C Systems in Support of integral Attachment- Visual, VT-3 Not applicable to RHR from Spent Fuel Mechanical and Hydraulic eitherVogtle unit Storage Pool Snubbers D3.40 D-C Systems in Support of integral Attachment- Vsual, VT-3 Not applicable to RHR from Spent Fuel Spring Type Supports either Vogtle unit Storage Pool D3.50 D-C Systems in Support of Integral Attachment- Vmual, VT-3 Not applicable to RHR from Spent Fuel Constant Load Type either Vogtle unit Storage Pool Supports D3.60 D-C Systems in Support of Integral Attachment- Visual, VT-3 Not applicable to RHR from Spent Fuel Shock Absorbers either Vogtle unit Storage Pool 4-4 Rev.0 i
. . . . . - - - . . - - . - - - - . - _ - - - - - - . . _ _ . - . - . - _ . - - - _ - _ _ _ - - . - - - - - _ _ . _ . - - - _ - . . - - _ _ _ _ _ - - - _- --_._n - - - - - .-__ ,- - - - . . ~ . - -,-m-e . - s
5.0 CLASS 1. 2. AND 3 COMPONENT SUPPORTS 5.1 Purpose The purpose of this section is to define an inservice inspection program for Class 1,2, and 3 component supports to meet the requirements of the ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition, to the extent practicable. SNC intends to use ASME Section XI Code Case N-491 as an alternative for determining the scope of component supports subject to examination for Class 1, 2, and 3 component supports under subsection IWF. The code case was approved by the AShE on March 14, 1991 and is contained in NRC Regulatory Guide 1.147, Revision 10.
5.2 Insoection Schedule As much as practicable, Class 1,2, and 3 component supports shall be scheduled for examination using Inspection Program "B" as defined in ASME Section XI Code Case N-491, Table-2400-2.
l 5.3 Inspection Scope Areas subject to inservice inspection are shown in the following Table-2500-1 of ASME Section XI Code Case N-491.
O v
l i
5-1 Rev.0 0
_ . _ _ . _ . . _ _ _ . . _ . . . . _ . . . . . _ - _ . . _ . . _ _ - _ . _. . _ . . _ . __ . _ ... .. ~ _ . . . _ _ . _ . .
m O Of- U TABLE 4 ISI PROGRAM FOR ASME CLASS 1,2 AND 3 COMPONENT SUPPORTS VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 Examination System or Component Method of Altemative '
It~m No. Cateaory Description Area (s) to Be Examined Examination Examinations F1.10 F-A Supports Mechanical Connections to Visual, VT-3 N-491 Pressure Retaining Components and Building Structure F1.20 F-A Supports Weld Connections to Visual, VT-3 N-491 !
Building Structure i
F1.30 F-A Supports Weld and Mechanical Visual, VT-3 N-491 Connections at intermediate Joints in Multiconnected integral and Nonintegral Supports F1.40 F-A Supports Clearances of Guides and Visual, VT-3 N-491 Stops, Alignment of Supports, Assembly of Support items F1.50 F-A Supports Spring Supports and Visual, VT-3 N-491 Constant Load Supports F1.60 F-A Supports Sliding Surfaces Visual, VT-3 N-491 F1.70 F-A Supports Hot or Cold Position of Visual, VT-3 N-491 Spring Supports and Constant Load Supports 5-2 Rev.O
l
)
I i
6.0 Class MC and CC Components C)
As noted in section 1.12, only ASME' Code Class 1,2, and 3 components and their supports are j included in the scope of this program document.
Southern Nuclear Operating Company acknowledges that as a result of a September 9,1996 rulemaking, the NRC is requiring that licensees revise their ISI programs to incorporate by reference the 4
4 1992 Edition with 1992 Addenda of ASME Section XI, Subsections IWE and IWL which pertain to ASME Section XI Code Classes MC and CC, respectively, with specific modifications and a limitation i which are in addition to the Code requirements. In order to comply with the rulemaking, a program for l the examination and testing of ASME Code Class MC and CC components is in the process of being i developed for VEGP-1 and 2. Once developed, the program document for Class MC and CC components will be maintained at the VEGP plant site for review by the NRC upon request.
Maintenance of the program document at the plant site for review upon request is consistent with the j requirements of the rulemaking.
W a
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l 6-1 Rev.O J
7.0 Requests for Relief 4
Requests for relief have been prepared from information determined in support of Preservice j inspection activities and Inservice Inspection activities conducted during the first ten year ,
l inspection interval at VEGP-1 and 2. The subject requests for relief apply to both VEGP units i unless specifically denoted otherwise. Please refer to the following for a detailed listing of -
requests for relief. The actual requests for relief follow the listing. Please note that in cases where the use of an ASME Section XI code cas'eis requested, a copy of the code case is provided solely for the convenience of the NRC reviewer,
! Requests for Relief Summary RR No. Code Category (s) Subject 1st Int. RR No.
3 RR-1 N/A VEGP-210 Year Update N/A i
l RR-2 B-D, B-F RPV Nozzle Schedule N/A J (Code Case N-521)
) RR-3 B-A RPV Mechanized Limitations RR-2, RR-3, RR-5 RR-4 B-A,B-H RPV Closure Head Limitations (manual) RR-7, RR-52 RR-5 B-G-1 RPV Closure Head Nuts N/A RR-6 B-B Steam Generator (Class 1) Welds and RR-19, RR-42 Inner Radii Limitations RR-7 B-B, B-D, B-F Pressurizer Weld Limitations RR-12, RR-14, RR-15 RR-8 B-H Pressurizer Support Limitations RR-10, RR-16 RR-9 B-J Primary Loop Piping Calibration Block RR-22, RR-23 and Scan Limitations RR-10 B-J Primary Loop Piping Weld Limitations RR-17, RR-24 RR-11 B-J Primary Loop Piping Branch Connection RR-21 Limitations RR-12 B-J 10" Safety Injection Weld Limitations RR-26 C-F-1 RR-13 B-J Piping (Class 1) Weld Limitations N/A RR-14 C-A, C-B, C-G Vessel (Class 2) Weld Limitations RR-28, RR-29, RR-30. RR-32 O . -
7-1 Rev.O
. _ . _ _ _ - _ _ _ . _ _ ___.__ _ _ . _ _ _ _ _ _ _ . _ _ _ . . . _ _ _ . _ _ _ . . _ _ _ _ _ - - .. . . . _ . .__m Requests for R: lief Summtry (continutd)
RR No. Code Category (s) Subject 1st Int. RR No.
RR-15 C-F-1 Piping (Class 2) Weld Limitations RR-35, RR-36,
, RR-37
! RR-16 C-F-2 Piping (Class 2) Weld Limitations RR-34 l l
RR-17 C-F-1 Piping (Class 2) Weld Selection Criteria N/A RR-18 C-F-1 NSCW Piping Classification N/A RR-19 C-F-1, C-F-2 Piping Longitudinal Welds N/A l (Code Case N-524)
RR-20 B-H, C-C, D-A, D- Integral Attachment Examinations RR-61 B, D-C (Code Case N-509)
RR-21 C-A, C-C Vessel (Class 2) Exemptions RR-62 1 (Code Case N-408-2)
RR-22 ASME 1" and Under Piping Exemption N/A (Code Case N-544) ,
RR-23 B-E,B-P,C-H, Hydrostatic Testing RR-60 D-A, D-B, D-C (Code Case N-498-1)
RR-24 ASME Hydrostatic Testing RR-59 '
RR-25 ASME Bolted Connections N/A (Code Case N-566)
RR-26 ASME Insulation at Bolted Connections N/A )
RR-27 ASME VT-2 Personnel Requirements N/A
]
RR-28 ASME NIS-2 Requirements for Snubbers N/A (Code Case N-508-1)
RR-29 ASME Inservice Testing and Examination RR-43 Requirements for Snubbers RR-30 ASME NIS-1 and NIS-2 Reporting Requirements N/A (Code Case N-532) l i
7-2 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS I AND 2
,. SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-1 I. System / Component (s) for Which Reliefis Requested:
This request for relief applies to the ASME Class 1,2, and 3 tables in this ISI Program and their supports.
II. Code Requirement: .
Part 50.55a(g)(4)(ii) of Title 10 of the Code of Federal Regulations (CFR) states: " Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (b) of this section 12 months prior to the start of the 120-month inspection interval, subject to the limitations and modifications listed in paragraph (b) of this section".
Paragraph IWA-2413, " Successive Inspection Intervals", of the 1989 Edition of ASME Section XI states: "The inspection plan for each successive. inspection interval shall comply with the Edition and Addenda of this Division that has been adopted by the regulatory authority 12 months prior to the start of the inspection interval, or subsequent Editions and Addenda that have been adopted by the regulatory authority. Specific portions or such subsequent Editions or v Addenda may be used provided all related requirements are met".
III. Code Requirement from Which Reliefis Requested:
Reliefis requested to start the VEGP-2 Second Ten-Year Interval (and subsequent intervals, for the remainder of the plant life) ahead of schedule such that the two VEGP units will be under the same edition and/or addenda of the Code, i.e., the 1989 Edition of ASME Section XI. Further, this relief would provide an exemption to the requirements of 10 CFR 50.55a(g)(4)(ii) for updating the VEGP-2 ISI program at its normal update time, i.e., May 20,1999, fo* its Second Ten-Year Interval.
IV. Basis for Relief:
By letter LCV-0861 dated August 16,1996, Georgia Power Company (the former licensee for VEGP and sister company to Southem Nuclear Operating Company, the current licensee and operator of VEGP) requested NRC concurrence to update the VEGP-2 ISI program for the Second Ten-Year Interval approximately two years ahead of schedule.
O i-3 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL w) REOUEST FOR' RELIEF NO. IG-1 (continued)
IV. Basis for Relief (continued):
The early update of the VEGP-2 ISI program would be coincident with the update for VEGP-1 and would allow the two VEGP units to be performed to the same edition and/or addenda of the Code. Otherwise, a different edition and/or addenda of the Code would have to be used for the ISI programs for each unit at VEGP because of the differences in commercial operation dates on which the ISI program interval dates are based. Specifically, the ISI program for VEGP-1 would be performed to the 1989 Edition of ASME Section XI while that for VEGP-2 would continue to be performed to the 1983 Edition 'f ASME Section XI with Addenda through Summer 1983 until its ISI program was updated in 1999. The commercial operation dates for VEGP-1 and 2 are May 31,1987 and May 20,1989, respectively. In its August 16,1996 letter to the NRC, Georgia Power Company indicated that a formal request for relief would be submitted requesting permission to update the VEGP-2 ISI program concurrent to that for VEGP-1 and requesting exemption from having to update VEGP-2 at its normal update time.
By concurrently updating the VEGP-1 and 2 ISI programs, a more comprehensive examination will be achieved which will enhance the possibility .of detecting a generic problem and will also D
(d reduce costs involved with maintaining two separate ISI programs should the use of different Code editions / addenda be required. In addition, use of the same edition and/or addenda of the l Code for the two units would help prevent possible errors associated with maintaining two I separate programs with difTerent requirements.
The practice of early updates has been used by Georgia Power Company since 1985 with the updates at llatch Nuclear Plant, Unit 2, for the Second Ten-Year Interval and most recently for the Third Ten-Year Interval update.
In its November 27,1996 response to the Georgia Power Company letter, the NRC indicated its concurrence with the concurrent update of the VEGP-2 ISI program with that for VEGP-1.
Further, it was indicated that the applicable ASME Code, i.e., the 1989 Edition, for the ISI program during the adjusted interval for VEGP-2 was acceptable in accordance with 10 CFR 50.55a(g)(4).
V. Alternate Examinations: )
l Alternate examinations are not applicable to this request for reliefinvolving the early update of ;
the VEGP-2 ISI Program and exemption from updating at the normal time for VEGP-2. As noted I in the Georgia Power Company letter of August 16,1996, the remaining examinations for the final two years of the current interval for the VEGP-2 ISI Program will be rescheduled for the i first two years of the new ten-year interval which will be concurrent with the update for VEGP-1.
~
l 7-4 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-1 l k.
(continued) 1 V. Alternate Examinations (continued):
These examinations will continue to be scheduled as required by ASME Section XI, Paragraph IWA-2413.
VI. Justification for Granting Relief:
By concurrently updating the VEGP-1 and 2 ISI programs, a more comprehensive examination
- will be achieved which will enhance the possibility of detecting a generic problem and will also
- reduce costs involved with maintaining two separate ISI programs should the use of different Code editions / addenda be required. In addition, use of the same edition and/or addenda of the Code for the two units would help prevent possible errors associated with maintaining two
, separate programs with different requirements. Examinations which remain to be performed to
- complete the First Ten-Year Interval on VEGP-2 will be rescheduled such that they are performed during the first two years of the adjusted interval on that plant unit. This will help ensure that no more than ten years will elapse between examinations.
O
- V Southern Nuclear Operating Company requests that relief and the exemption discussed herein be authorized pursuant to 10 CFR 50.55a(a)(3)(i) since an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered.
VII. Implementation Schedule:
The Code-required examinations will be performed during the Second Ten-Year Interval (and subsequent intervals, for the remainder of the plant life) which commences May 31,1997.
A i A 7-5 Rev. 0
SOUTHERN NUCLEAR OPERATING COMPANY j VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN YEAR INTERVAL
, REOUEST FOR RELIEF NO. RR-2 I. System / Component (s) for Which Reliefis Requested:
Volumetric examination of Reactor Pressure Vessel nozzle to shell (B-D) welds, nozzle inside !
radius sections (B-D), and nozzle to safe end (B-F) welds as identified in Attachment I to this request for relief. i II. Code Requirement:
ASME Section XI Table IWB-2500-1, Examination Category B-D, Items Nos. B3.90 and B3.100, Nozzle to Vessel (Shell) Welds and Nozzle Inside (Inner) Radius Section, and Examination Category B-F, Item No. B5.10, Nozzle to Safe End Welds, require volumetric examination of all welds and their inside radius section once each ten-year inspection interval. l The subject table does not allow deferral of volumetric examination of nozzle to shell welds, ,
nozzle inside (inner) radius sections, and nozzle to safe end welds to the end of the ten-year l inspection interval.
III. Code Requirement from Which Reliefis Requested:
As an attemative to the existing schedule, reliefis requested to allow the rescheduling of the
('d] examinations of those items identified in Attachment 1 to the end of the ten-year inspection interval so that all such examinations can be performed as part of the " ten-year Inservice Inspection" at which time the RPV is usually examined.
IV. Basis for Relief:
1 l
ASME Section XI Code Case N-521 (copy provided as Attachment 2) accepts the alternative j scheduling of the subject examinations provided that certain conditions are met. These l conditions include the following: (1) No inservice repairs or replacements have been performed to the RPV, (2) no existing flaws requiring successive inspections exist, and (3) the unit is not in the first inspection interval.
V. Alternate Examination:
i The required types of examinations will continue to be performed except that they will be ;
performed at the end of the ten-year interval. ,
I VI. Justification for Granting Relid:
Historically, pressurized water reactors such as VEGP-1 and 2 examine the RPV outlet nozzle to d shell welds, their inside (inner) radius sections, and associated nozzle to safe end welds during ,
7-6 Rev. 0
1 SOUTHERN NLCLEAR OPERATING COMPANY !
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 l SECOND TEN-YEAR INTERVAL j REOUEST FOR RFI IEF NO. RR-2
)
l (continued)
VI. Justification for Grantinsr Relief (continued):
the first inspection period in order to comply with the requirements of the ASME Section XI Code. These examinations are performed from the inside surfaces of the RPV using submerged ultrasonic examination techniques with an automated reactor vessel inspection tool. These examinations are generally performed while defueled with water in the refueling canal, thus typically being a critical path activity. The similar reactor vessel inlet nozzles are accessible only with the RPV core barrel removed, as during the " Ten-Year Inservice Inspection" at which time they are examined. In order to consolidate the examination of the reactor vessel inlet and outlet nozzles such that they are examined at one time, SNC proposes the use of ASME Section XI Code Case N-521 which allows rescheduling of all such RPV examinations to the end of the inspection interval. The rescheduling allows SNC significant opportunities for savings in contractor cost, critical path, radiation es.posure, and internal manpower requirements while still maintaining compliance with the examination requirements of the Code.
in order to facilitate pnssible future use of Code Case N-521, each of the subject RPV areas for which reliefis being requested were examined during VEGP-1 Outage IR6 in Spring 1996 O during which the " Ten-Year Inservice Inspection" was performed. This included re-examining those areas, e.g., the outlet nozzles and their associated components, which were examined during the first inspection period in the First Ten-Ve-Interval. This was done voluntarily by the licensee so that no more than ten years would elapse before being examined again at the end of the Second Ten-Year Interval. In the case of VEGP-2, a similar course of action is planned for l' Outage 2R6, currently scheduled for March 1998, at which time the remaining RPV examinations for its first Ten-Year interval will be completed. Each of the outlet nozzles and their associated components will be re-examined on VEGP-2 during its " Ten-Year Inservice Inspection" so that all of the RPV examinations may be performed at the end of the Second Ten-Year Interval. No rejectable indications have been observed nor have any repairs been made to either reactor pressure vessel at VEGP-1 and 2 which would preclude the use of Code Case N- l 521.
Based on the foregoing, SNC requests that this request for relief be authorized pursuant to 10 l CFR 50.55a(a)(3)(i) permitting the use of ASME Section XI Code Case N-521 at VEGP-1 and 2.
VII. Implementation Schedule: l The subject examination schedule will be implemented during the Second Ten-Year Interval which commences May 31,1997. In the case of VEGP-2, the alternative examination schedule provided by the use of Code Case N-521 would commence following the completion of RPV weld examinations during Maintenance / Refueling Outage 2R6 (March 1998).
7-7 Rev.0 l
1 O SOUTHERN NUCLEAR OPERATING COMPANY O O VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-2 (continued)
A'ITACHMENT 1 ASME Section XI Identification No. Description Category / Item No.
B-D / B3.90 11201-V6-001-WO25 VEGP-1 RPV N1 Outlet Nozzle to Shell Weld B-D / B3.90 11201-V6-001-WO28 VEGP-1 RPV N4 Outlet Nozzle to Shell Weld B-D / B3.90 11201-V6-001-WO29 VEGP-1 RPV N5 Outlet Nozzle to Shell Weld B-D / B3.90 11201-V6-001-WO32 VEGP-1 RPV N8 Outlet Nozzle to Shell Weld B-D / B3.90 21201-V6-001-WO25 - VE6P 7 RPV N1 Outlet Nozzle to Shell Weld B-D / B3.90 21201-V6-001-WO28 VEGP-2 RPV N4 Outlet Nozzle to Shell Weld B-D / B3.90 21201-V6-001-WO29 VEGP-2 RPV N5 Outlet Nozzle to Shell Weld B-D / B3.90 21201-V6-001-WO32 VEGP-2 RPV N8 Outlet Nozzle to Shell Weld B-D / B3.100 11201-V6-001-IR01 VEGP-1 RPV N1 Outlet Nozzle Inside (Inner) Radius B-D / B3.100 11201-V6-001-IR04 VEGP-1 RPV N4 Outlet Nozzle Inside (Inner) Radius B-D / B3.100 11201-V6-001-IR05 VEGP-1 RPV N5 Outlet Nozzle Inside (Inner) Radius ,
B-D / B3.100 11201-V6-001-lR08 VEGP-1 RPV N8 Outlet Nozzle Inside (Inner) Radius a
7-8 Rev.0 i
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-2 (continued)
ATTACllMENT 1 (continued)
ASME Section XI Identification No. Description Category / Item No.
B-D / B3.100 21201-V6-001-IR01 VEGP-2 RPV N1 Outlet Nozzle Inside (Inner) Radius B-D / B3.100 21201-V6-001-IR04 VEGP-2 RPV N4 Outlet Nozzle Inside (Inner) Radius B-D / B3.100 21201-V6-001-IR05 VEGP-2 RPV N5 Outlet Nozzle Inside (Inner) Radius B-D / B3.100 21201-V6-001-IR08 VEGP-2 RPV N8 Outlet Nozzle Inside (Inner) Radius B-F / B5.10 -
11201-V6-001-WO33 VEGP-1 RPV N1 Outlet Nozzle to Safe End Weld B-F / B5.10 11201-V6-001-WO36 VEGP-1 RPV N4 Outlet Nozzle to Safe End Weld B-F / B5.10 I1201-V6-001-WO37 VEGP-1 RPV N5 Outlet Nozzle to Safe End Weld B-F / B5.10 11201-V6-001-WO40 VEGP-1 RPV N8 Outlet Nozzle to Safe End Weld B-F / B5.10 21201-V6-001-WO33 VEGP-2 RPV N1 Outlet Nozzle to Safe End Weld B-F / B5.10 21201-V6-001-WO36 VEGP-2 RPV N4 Outlet Nozzle to Safe End Weld B-F / B5.10 21201-V6-001-WO37 VEGP-2 RPV N5 Outlet Nozzle to Safe End Weld B-F / B5.10 21201-V6-001-WO40 VEGP-2 RPV N8 Outlet Nozzle to Safe End Weld 7-9 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL t
G t REOUEST FOR RELIEF NO. RR-2 V -
(continued)
ATTACHMENT 2 CASE (continued)
N-521 CASES OF AhhtE 8011.9 N AND PRESSURE VI'.S$EL CODE Approval Date:Autuet 9,1993 l See NumettelIndes kr exporetoon and any reaMumstron detes.
Case N-52I 4
Alternative Rules for Deferral ofInspections of Noule-to-Vessel Welds, Inside Radius Sections, and Nonle-to-Safe End Welds of a Pressurized Water
{ Reactor (PWR) Vessel Section XI Division 1 Ingwry: What alternative rules may be used in lieu of Table IWB-25001, Examination Category B-D,
" Full Penetration Welded Noules in Vessels - In-4 spection Program B" and Examination Category B.
F. " Pressure Retarning Dissimilar Metal Welds in
- . Vessel Noules."Section XI, Drvision 1, to allow de-4 ferral of inspections of Noule-to-Vessel Welds, In-side Radius Sections, and Noule-to-Safe Ends Welds of a PWR vessel? ,
1 Reply It is the opinion of the Committee that, as an alternative to the existing requirements, inspec-
- j tions of Nonie-to-Vessel Welds Inside Radius Sec-tions, and Noule-to-Safe End Welds of a PWR ves- )
i sel may be deferred to the end of the inspection interval if the following conditions are met for the {
reactor vessel in question:
(a) No inservice repairs or replacements by weld-ing have ever been performed on any of the Nonle-to-Vessel Welds, Inside Radius Sections, or Noule.
to-Safe End Welds.
(b) None of the Nonle to-Vessel Welds. Inside Radius Sections, or Noule to Safe End Welds con-tains identified flaws or relevant conditions that cut-i rently require successive inspections in accordance with IWB 2420(b).
(c) The unit is not in tbr first inspection interval-i i
f \
4 7-10 Rev.0 4
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
- REOUEST FOR RELIEF NO. RR-3 I. System / Component (s) for Which Reliefis Requested
- Mechanized volumetric examination of pressure-retaining welds in the Reactor Pressure Vessel 4
(RPV) as identified in Attachment I to this request for relief.
II. Code Requirement:
ASME Section XI Category B-A, Table IWB-2500-1 requires volumetric examinations of pressure-retaining welds in the reactor pressure vessel. The applicable examination volume is
- shown in ASME Section XI Figures IWB-2500-1, IWB-2500-2, and IWB-2500-3 and includes essentially one hundred percent (100%) of the weld' length.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing a Full Code-Coverage volumetric examination of the RPV welds identified in Attachment I to this request for relief.
IV. Basis for Relief:
Physical obstructions, e.g., core support lugs, bottom-mounted instrumentation nozzles, etc.,
prevent movement of the mechanized sled / transducer along the required scan region of the referenced welds thereby limiting the amount of examination coverage which can be attained.
Full Code Coverage of ninety percent (90%) or greater (as defined by AS.ME Section XI Code Case N-460 and as accepted by the NRC) is not possible for the referenced welds which are in the vicinity of the physical obstructions.
! V. Alternate Examination:
No alternate examination is proposed. The affected RPV welds are being volumetrically examined to the extent practical.
VI. Justification for Granting Relief:
The volumetric examination of the referenced RPV welds are being conducted to the extent practical with a remote mechanized inside diameter vessel tool.
Relief was initially granted by the NRC during the First Ten-Year Interval for those welds in l Attachment I having requests for relief submitted for them. These included First Ten-Year !
Interval Requests for Relief RR-2, RR-3, and RR-5. NRC approval was documented in 1 7-11 Rev. O
I l
SOUTHERN NUCLEAR OPERATING COMPANY VOGTI E ELECTRIC GENERATING PLANT. UNITS 1 AND 2 l SECOND TEN-YEAR INTERVAL l REOUEST FOR RELIEF NO. RR-3 (continued)
VI. Justification for Grantina Relief (continued):
correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively. Subsequent to that time, the NRC withdrew its approval of requests for relief associated with ASME Section XI Category B-A which had been granted to operating reactor licensees through rulemaking involving 10 CFR 50.55a [ refer to 10 CFR 50.55a(g)(6)(A)(1) and (2)]. Because both VEGP units were in their First Ten-Year Interval of commercial operation, each of the RPV welds had to be examined regardless of the rulemaking due to the requirements of the 1983 Edition of ASME Section XI with Summer 1983 Addenda. Each of the required welds on VEGP-1 were examined during its First Ten-Year Interval. On VEGP-2, the remaining welds which are required to be examined are scheduled for Maintenance / Refueling Outage 2R6 j which is currently scheduled to begin in March 1998. i As a result of changes in the Code requirements which are reflected in the 1989 Edition of ASME Section XI, each Category B-A weld is required to be volumetrically examined ,
regardless of which inspection interval might be involved. Although there are phy sical 1 obstructions which limit the amount of examination coverage for those welds identified in O Attachment 1, reasonable assurance still exists that an acceptable level of quality md safety will be achieved. Further, public health and safety will be maintained since there have been no recorded catastrophic failures of reactor pressure vessels combined with the fact that Code changes require that each Category B-A weld be examined, which is a significant improvement over past Code versions. As a result, SNC requ'ests that this request for relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is impractical to perform these examinations to the extent required by the Code. ;
VII. Implementation Schedule: ,
The subject examinations will be performed to the fullest extent practical during the Second -
Ten-Year Interval which commences May 31,1997.
7-12 Rev. O
Q V (V
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v/
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-3 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
i Category / Item No. Percentage RR No.
B-A / Bl.11 11201-V6-001-WO6 VEGP-1 RPV lower shell Core support lugs 62 % RR-2 to bottom head weld B-A / Bl.11 21201-V6-001-WO6 VEGP-2 RPV lower shell Core support lugs 66 % RR-2 to bottom head weld B-A / Bl.12 11201-V6-001-W12 VEGP-1 RPV upper shell Reactor Coolant System 75 % N/A longitudinal weld nozzles B-A / Bl.12 11201-V6-001-W13 VEGP-1 RPV upper shell Reactor Coolant System 80 % N/A i longitudinal weld nozzles B-A / Bl.12 11201-V6-001-W14 VEGP-1 RPV upper shell Reactor Coolant System 85 % N/A
- longitudinal weld nozzles B-A / Bl.12 21201-V6-001-W12 VEGP-2 RPV upper shell Reactor Coolant System See Note 1 N/A longitudinal weld nozzles B-A / Bl.12 21201-V6-001-W13 VEGP-2 RPV upper shell Reactor Coolant System See Note 1 N/A longitudinal weld nozzles B-A / Bl.12 21201-V6-001-W14 VEGP-2 RPV upper shell Reactor Coolant System See Note i N/A longitudinal weld nozzles B-A / Bl.12 11201-V6-001-W18 VEGP-1 RPV lower shell Core support lugs 77% RR-3 longitudinal weld B-A / Bl.12 11201-V6-001-W19 VEGP-1 RPV lower shell Core support lugs 77 % RR-3 longitudinal weld B-A / Bl.12 11201-V6-001-W20 VEGP-1 RPV lower shell Core support lugs 77% RR-3 longitudinal weld ,
7-13 Rev.0
O SOUTHERN NUCI EAR OPERATING COMPANY O O VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-3 (continued)
ATTACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int. l Category / Item No. Percentage RR No.
B-A / Bl.21 11201-V6-001-WO7 VEGP-1 RPV bottom head Bottom Mounted 29 % RR 5 8 weld Instmmentation Lines B-A / Bl.21 21201-V6-001-WO7 VEGP-2 RPV bottom head Bottom Mounted See Note 1 RR-5 weld Instrumentation Lines t
Notes:
- 1. Approximate percentage for the First Ten-Year Interval RPV examinations to be determined during the 2R6 maintenance / refueling outage scheduled for March 1998. -
b 7-14 Rev. O
SOUTHERN NUCI FAR OPERATING COMPANY
' VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-4 I. System / Component (s) for Which Reliefis Requntul:
i Volumetric and surface examination of pressure-retaining welds and integral attachments in the Reactor Pressure Vessel Closure Head as identified in Attachment I to this request for relief.
II. Code Reautrement:
I (a) ASME Section XI, Category B-A, Item Bl.40, Table IWB-2500-1 requires volumetric ;
examinations of pressure-retaining welds in the reactor pressure vessel. The applicable i examination volume is shown in ASME Section XI, Figure IWB-2500-5 and includes essentially one hundred percent (100%) of the. weld length. !
(b) Code Case N-509 (see Request for Relief RR-20), Category B-K, Item B10.10, Table IWB-2500-1 requires surface examinations ofintegrally welded attachments on the reactor pressure vessel. The applicable examination volume is shown in ASME Section XI, Figure IWB-2500-15 and includes essentially 100% of the weld length.
l Ill Code Requirement from Which Reliefis Requested:
O b Reliefis requested from perfonning the full Code-required volumetric and surface examinations !
of the Reactor Pressure Vessel Closure Head Welds identified in Attachment 1.
l IV. Basis for Relief:
(a) Physical obstructions, e.g., RPV closure head lifting lugs and the RPV closure head torus flange configuration, limit or otherwise prohibit movement of the transducer used for the ultrasonic examination of the RPV closure head to flange weld along the required scan region thereby limiting the amount of examination coverage that can be attained. Full Code Coverage of ninety percent (90%) or greater (as defined by ASME Section XI Code Case l N-460 and as accepted by the NRC) is not possible for the referenced welds that are in the vicinity of the obstructions.
(b) The geometric configuration and location of the Control Rod Drive (CRD) braces and RPV closure head lifting device prevents a complet'e surface examination of the R.PV closure head lifling lugs.
V. Alternate Examination:
No alternate examination is proposed. The RPV closure head to flange weld will be volumetrically examined to the extent practical. Likewise, the RPV closure head lifting lugs will O be examined by surface means to the fullest extent practical.
7-15 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY yOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL O REOUEST FOR RELIEF NO. RR-4 V
(continued)
VI. JustiGcationJar Grantine Relief:
Relief was granted by the NRC for the RPV closure head torus to flange welds and closure head lifting lugs during the First Ten-Year Interval. These included First Ten-Year Interval Requests for Relief RR-7 and RR-52 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
Although there are physical obstructions which limit the amount of examination coverage for those welds identified in Attachment 1, reasonable assurance still exists that an acceptable level of quality and safety will be maintained since there have been no catastrophic failures of reactor pressure vessels. As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is impractical to perform these examinations to the extent required by the Code.
VII. Implementation Schedule:
("]
V The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
O 7-16 Rev.O
. . . ~. . - . . _ - . . . . . . . _ - _ ... .- _-
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE EISCTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-4 (continued)
ATTACHMENTI '
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-A / Bl.40 11201-V6-001-WO2 VEGP-1 RPV closure head RPV closure head titling lugs 65 % RR-7 torus to flange weld and flange configuration B-A / Bl.40 21201-V6-001-WO2 VEGP-2 RPV closure head RPV closure head lifting lugs 65 % RR-7 torus top flange weld and flange configuration B-K / B10.10 11201-V6-001-W204 VEGP-1 RPV closure head RPV closure head lifting 75 % RR-52 lifting lugs device and CRD braces B-K / B10.10 11201-V6-001-W205 VEGP-1 RPV closure head RPV closure head lifting 75 % RR-52 lifting lugs device and CRD braces B-K / B10.10 11201-V6-001-W206 VEGP-1 RPV closure head RPV closure head lifting 75 % RR-52 lifting lugs device and CRD braces B-K / B10.10 21201-V6-001-W204 VEGP-2 RPV closure head RPV closure head lifting 50 % RR-52 lifting lugs device and CRD braces B-K / B10.10 21201-V6-001-W205 VEGP-2 RPV closure head RPV closure head lifting 50 % RR-52 lifting lugs device and CRD braces B-K / B10.10 21201-V6-001-W206 VEGP-2 RPV closure head RPV closure head lifting 50% RR-52 lifting lugs device and CRD braces 7-17 Rev.O
SOUTHERN NUCI E AR OPERATING COMPANY VOGTLE EIECTRICSENERATING.fLANT, UNlTS 1 AND 2 SECOND TEN-YEAR INTEJLYAL I REOUEST FOR RELIEF NO. RR-5
- 1. System / Component (s) for Which Reliefis Reauested:
RPV Closure Head Nuts 11201-V6-001-N01 through 11201-V6-001-N54 and 21201-V6-001-N01 through 21201-V6-001-N54 II. Code Requirement:
Section XI, Table IWB-2500-1, Examination Category B-G-1, Item B6.10 requires a surface i examination of the Reactor Pressure Vessel Closure Head Nuts. l III. Code Requirement from Which Reliefis Reauested:
Reliefis requested from performing the Code-required surface examination on above identified Reactor Pressure Vessel Closure Head Nuts.
IV. Basis for Relief:
Table IWB-2500-1, Category B-G-1, Item B6.10 of the 1989 Addenda to the 1989 Edition of ASME Section XI allows for a VT-1 Visual Examination in lieu of the surface examination required by the 1989 Code.
V. Alternate Examinations: !
1 In lieu of the 1989 Edition, Code-required surfqce examination, the subject RPV Closure Head Nuts will receive a VT-1 visual examination.
VI. Justification for the Granting of Relief:
The RPV closure head nut configuration does not allow for an adequate magnetic particle (MT) examination. The MT method requires two directional coverage to detect the surface flaws. The configuration permits examination in one direction and limits the coverage in the other direction.
Sectio'n XI Code personnel in ISI Optimization performed a survey on bolting which revealed no service-induced cracking. This survey was then used as part of the technical basis for changing the Code-required examination for Category B-G-1, Item B6.10. The 1989 Addenda and subsequent editions of ASME Section XI changed the examination requirement from a surface examination to a VT-1. Since the change (visual examination) was issued by ASME, the 7-18 Rev.0
SOUTHERN NUCIIAR OPERATING COMPANY VOGTI F ELErTRIC GENERAT!NG PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL b
V REOUEST FOR RELIEF NO. RR-5 (continuert)
VI. Justification for the Granting of Relief (continued):
alternative examination should be technically acceptable for determining flaws. The proposed alternative visual examination, VT-1, will provide reasonable assurance that unallowable inservice flaws have not developed in the subject components or that they will be detected and repaired prior to return of the reactor pressure vessel to service. If relevant indications are detected, alternate surface / volumetric techniques will be performed as necessary. Thus, an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed altemative examination in lieu of the Code requirement. Therefore, this request for relief should be granted pursuant to the requirements of 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedule:
The subject examinations will be performed during the Second Tca-Year Interval which commences May 31,1997.
O e
i O !
7-19 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOCTLE ELECTRIC GENERATING PLANT. UNITS I AND 2 SECOND TEN-YEAR INTERVAL O REOUEST FOR RELIEF NO. RR-6 V
I. System /Comnonent(s) for Which Reliefis Requested:
Volumetric examination of steam generator pressure-retaining welds and nozzle inner radius sections (Class 1) as identified in Attachment I to this request for relief.
II. Code Requirement:
ASME Section XI Category B-B and ASME Section XI Category B-D, Table IWB-2500-1 require volumetric examinations of pressure-retaining welds in the steam generator and nozzle i inner radii. Applicable examination volume is shown in ASME Section XI Figures IWB-2500-6 and IWB-2500-7. !
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required volumetric examinations of the components identified in Attachment I to this request for relief.
IV. Basis for Relief:
O V (a) Physical limitation of the steam generator tube sheet obstructs and/or prohibits transducer movement along the required scan region of the channel head to tube sheet weld. Full Code coverage is not possible in the vicinity of the obstructions.
1 (b) The steam generator primary side nozzles are integrally cast as part of the channel head.
The steam generator nozzle radius section cannot be volumetrically examined from outside of the nozzle or channel head because the rough, as-cast contact surface is not suitable for ultrasonic coupling, and the geometric configuration requires an excessively long test metal distance resulting in high ultrasonic attenuation. The inside of the nozzle and channel head areas are covered with cladding in the "as: welded" condition; therefore, meaningful volumetric examination cannot be performed from the "as-welded" surface. Even with proper preparation of the inside surface for volumetric examination, an adequate examination of the area ofinterest (base metaljust below the cladding) could not be achieved due to the resulting ultrasonic response at the clad-to-base metal interface. Refer to Attachment 2 for a depiction of the nozzles at VEGP.
V. Alternate Examinatioji:
(a) No alternate examination is proposed. The volumetric examination of the referenced steam generator tube sheet welds are being conducted to the fullest extent practical.
(b) No alternate examination is proposed for steam generator primary side nozzles inner radii.
7-20 ~ Rev.O
k SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ElICTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-6 (continued)
VI. Justification for Granting Relief:
l The examinations of the steam generator tubesheet welds are being conducted to the extent practical.
Compliance with the requirements of ASME Section XI for the specific nozzle inside radius section is impractical. Due to the high radiation field present (generally greater than 10 REM / hour), visual examinations are not practical and dose acquired is contrary to the principles
, of"As Low As Reasonably Achievable"(ALARA). During the First Ten-Year Interval, visual i examinations and the Code-required pressure tests were performed and no evidence of degradation was observed.
Relief was initially granted by the NRC during the First Ten-Year Interval for those I welds / components in Attachment I having requests for relief submitted for them. These included First Ten-Year Interval Requests for Relief # RR-19 and RR-42. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 i G and 2, respectively.
(D Although there are physical obstructions which limit and/or prohibit the amount of examination coverage for those components identified in Attachment 1, reasonable assurance still exists that an acceptable level of quality and safety will be maintained since there have been no catastrophic failures of steam generator primary side nozzle inner radii nor steam generator tubesheet welds.
As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is impractical to perform the examinations as required by the Code.
VII. Implementation Schedule:
The subject examinations will be performed to the fullest extent practical during the Second Ten-
~
Year Interval which commences May 31,1997.
7-21 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RFI IEF NO. RR-6 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int. .
Category / Item No. Percentage RR No.
B-B / B2.40 I!201-B6-001-WO8 VEGP-1 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld B-B / B2.40 11201-b6-002-WO8 VEGP-1 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld B-B / B2.40 11201-B6-003-WO8 VEGP-1 Channel head to Tube sheet configuration 80 % RR-19 4 tube sheet weld B-B / B2.40 11201-B6-004-WO8 VEGP-1 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld B-B / B2.40 21201-B6-001-WO8 VEGP-2 Channel head to Tube sheet configuration 80 % RR-19 -
tube sheet weld B-B / B2.40 21201-B6-002-WO8 VEGP-2 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld
- B-B / B2.40 21201-B6-003-WO8 VEGP-2 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld B-B / B2.40 21201-B6-004-WO8 VEGP-2 Channel head to Tube sheet configuration 80 % RR-19 tube sheet weld B-D / B3.140 11201-B6-001-IR-01 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 11201-B6-001-IR-02 VEGP-1 SG inlet / outlet Nozzle configuranon 0% RR-42 nozzle inner radius t B-D / B3.140 11201-B6-002-IR-01 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 11201-B6-002-IR-02 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius 7-22 Rev.0
SOUTHERN NUCLEAR ERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR REI IEF NO. RR-6 (continued)
ATTACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-D / B3.140 11201-B6-003-IR-01 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 11201-B6-003-IR-02 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 11201-B6-004-lR-01 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 11201-B6-004-lR-02 VEGP-1 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-001-IR-01 VEGP-2 SG, inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-001-IR-02 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius
_ B-D / B3.140 21201-B6-002-IR-01 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-002-IR-02 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-003-IR-01 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-003-IR-02 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-004-IR-01 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius B-D / B3.140 21201-B6-004-IR-02 VEGP-2 SG inlet / outlet Nozzle configuration 0% RR-42 nozzle inner radius 7-23 Rev. 0
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTI E EISCTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
, REOUEST FOR RELIEF NO. RR-7 4
1
- 1. System /Comnonent(s) for Which Reliefis Reauested:
Volumetric examination of pressurizer pressure retaining welds (Class 1) as identified in Attachment I to this request for relief.
i II. Code Requirement:
ASME Section XI, Category B-B (Item No. B2.11) and B-D (Item No. B3.110), Table IWB-2500-1, requires volumetric examination of pressure-retaining welds in the pressurizer. ASME Section XI Category B-F (Item No. B5.40) requires surface and volumetric examinations of pressure-retaining welds in the pressurizer. The applicable examination volume is shown in ASME Section XI Figures IWB-2500-1,7(b) and 8.
4 III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required volumetric examinations of the !
l components identified in Attaciunent I to this request for relief.
/ IV. Basis for Relief:
(])
(a) Geometric limitation of the pressurizer skirt obstructs and/or prevents transducer movement along the required scan regica of the Category B-B Jower shell to lower head weld. Full Code Coverage is not possible in the vicinity of the obstructions. .j (b) Geometric configuration of the 16" Category B-D pressurizer surge nozzle prevents scanning from the nozzle side. In addition, the pressurizer lower head heater penetrations obstructs and/or prevents transducer movement along the required scan region from the shell side . Limited coverage can be obtained for the 16" Category B-D nozzle to shell l weld. However, no coverage can be obtained for the 16" nozzle inner radius due to the i I
transducer " set back" dimension required.
(c) Geometric configuration of the 4" and 6" relie'f and spray nozzles prevent scanning from the nozzle side for the Category B-D upper head to nozzle welds and 6" Category B-F nozzle to safe-end welds.
(d) Geometric limitations of the pressurizer supports, ID plates and instrumentation nozzles obstructs and/or prevents transducer movement along the required scan region of the Category B-B upper head to upper shell weld. Full Code Coverage is not possible in the vicinity of the obstructions.
7-25 Rev. O
)
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE EI.ECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YE AR INTERVAL O
U REOUEST FOR RELIEF NO. RR-7 (continued)
V. Alternate E===in ation:
No alternate examinations will be performed for Category B-B or B-F welds. A 100% surface examination will be perfanned on the 4" and 6" nozzle to shell welds, and approximately 85%
on the 16" nozzle to she!! weld (heater penetration obstruction).
VI. Justification for Granting Relief:
The examinations are being conducted to the extent practical.
Relief was initially granted by the NRC during the First Ten-Year Interval for those welds / components in Attachment I having requests for relief submitted for them. These included First Ten-Year Interval Requests for Relief RR-12, RR-14, and RR-15. NRC approval I was documented in correspondence dated Nc,vember 26,1991 and December 17,1991 for I VEGP-1 and 2, respectively.
Although there are physical obstructions which limit or prohibit the amount of examination O coverage for those welds / components identified in Attachment 1, reasonable assurance still exists ;
that an acceptable level of quality and safety will be achieved. As a result, SNC requests that this request for relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since it is impractical to perform the examinations to the extent required by the Code.
VII. Implementation Schedule:
The subject examinations will be performed to the extent practical during the Second Ten-Year Interval which commences May 31,1997.
7-26 Rev. O
O
- G n O J hw/
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-7 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-B / B2.11 11201-V6-002-W01 VEGP-1 Upper head to Nozzles, supports & 68 % RR-12 .
upper shell weld identification plate B-B / B2.11 11201-V6-002-WOS VEGP-1 Lower shell to Support skirt 84 % RR-12 lower head weld B-B / B2.11 21201-V6-002-W01 VEGP-2 Upper head to Nozzles, supports & 68 % RR-12 upper shell weld identification plate B-B / B2.11 21201-V6-002-WO5 VEGP-2 Lower shell to Support skirt 84 % RR-12 lower head weld B-D / B3.110 11201-V6-002-W10 VEGP-1 Upper head to 6" Nozzle configuration 50% RR-12 safety nozzle weld B-D / B3.110 11201-V6-002-W11 VEGP-1 Upper head to 6" Nozzle configuration 50% RR-12 safety nozzle welds B-D / B3.110 11201-V6-002-W12 VEGP-1 Upper head to 6" Nozzle configuration 50 % RR-12 safety nozzle welds B-D / B3.110 11201-V6-002-W13 VEGP-1 Upper head to 6" Nozzle configuration 50 % RR-12 -
safety nozzle welds B-D / B3.110 11201-V6-002-W14 VEGP-1 Upper head to 4" Nozzle configuration 50% RR-12 ,
safety nozzle weld B-D / B3.110 11201-V6-002-W16 VEGP-1 Lower head to Nozzle configuration and 15 % RR-14 ~
16" surge nozzle weld Heater penetrations B-D / B3.110 21201-V6-002-W10 VEGP-2 Upper head to 6" Nozzle configuration 50 % RR-12 safety nozzle weld B-D / B3.110 21201-V6-002-W11 VEGP-2 Upper head to 6" Nozzle configuration 50 % RR-12 safety nozzle welds 7-27 Rev. 0
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-7 (continued)
ATI'ACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-D / B3.110 21201-V6-002-W12 VEGP-2 Upper head to 6" Nozzle configuration 50% RR-12 safety noule welds B-D / B3.110 21201-V6-002-W13 VEGP-2 Upper head to 6" Nozzle configuration 50 % RR-12 safety nozzle welds B-D / B3.110 21201-V6-002-W14 VEGP-2 Upper head to 4" Nozzle configuration 50% RR-12 safety nozzle weld B-D / B3.110 21201-V6-002-W16 VEGP-2 Lower head to Nozzle configuration and 15 % RR-14 16" surge nozzle weld heater penetrations B-D / B3.120 - 11201-V6-002-IR-06 VEGP-1 16" surge nozzle -
Heaterpenetrations 0% RR-15 inner radius B-D / B3.120 21201-V6-002-IR-06 VEGP-216" surge nozzle Heater penetrations 0% RR-15 inner radius B-F / B5.40 11201-V6-002-W17 VEGP-16" safety nozzle Nozzle / safe-end 50 % RR-12 to safe end welds configuration B-F / B5.40 11201-V6-002-W18 VEGP-16 safety nozzle Nozzle / safe-end 50 % RR-12 to safe end welds configuration B-F / B5.40 11201-V6-002-W19 VEGP-16" safety nozzle Nozzle / safe-end 50 % RR-12 to safe end welds configuration B-F / B5.40 11201-V6-002-W20 VEGP-16" safety nozzle Nozzle / safe-end 50% RR-12 to safe end welds configuration B-F / B5.40 21201-V6-002-W17 VEGP-2 6" safety nozzle Nozzle / safe-end 75 % RR-12 to safe end welds configuration B-F / B5.40 21201-V6-002-WI8 VEGP-2 6" safety nozzle Nozzle / safe-end 75 % RR-12 to safe end welds configuration 7-28 Rev. O
rn p b
SOUTHERN NUCTIAR OPERATING COMPANY VOGTLE ELFCTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-7 (continued)
ATTACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-F / B5.40 21201-V6-002-W19 VEGP-2 6" safety nozzle Nozzle / safe-end 75 % RR-12 to safe end welds configuration B-F / B5.40 21201-V6-002-W20 VEGP-2 6" safety nozzle Nozzle / safe-end 75 % RR-12 to safe end welds configuration B-F / B5.40 21201-V6-002-W21 VEGP-2 4" spray nozzle to Nozzle / safe-end 87 % RR-12 safe end weld configuration B-F / B5.40 21201-V6-002-W22 VEGP-714" surge nozzle Nozzle / safe-end 74 % RR-12 to stfe end welds configuration _
7-29 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-8 I. System / Component (s) for Which Reliefis Requested:
1 Surface examination of pressurizer integrally welded attachments as identified in Attachment I to this request for relief. ;
II. Code Requirement: I i
1 Item B10.10, Category B-K, Table IWB-2500-1 of Code Case N-509 (see Request for Relief j RR-20) requires that surface examinations ofintegrally welded attachments on the pressurizer be performed. The applicable examination volume is shown in ASME Section XI, Figure IWB-2500-15.
III. Code Requirement from Which Reliefis Requested:
l Reliefis requested from performing the Code-required surface examinations of the presswizer !
welds identified in Attachment 1.
IV. Basis for Relief: -
l O
O The pressurizer heater penetrations restrict personnel access to " Area C-D" of the pressurizer support skirt shown in ASME Section XI Figure IWB-2500-13 to perform the required surface examination. " Area A-B" will receive the full code-required surface examination.
The support rack assembly which holds the spray and relief piping of the pressurizer, rests above the integrally welded attachments (support brackets) and, thereby, restricts access to the integral attachment welds. Removal of the rack assembly to perform the examination is not feasible.
V. Alternate Examination:
No alternative examinations are proposed due to restricted access to the examination areas in question.
VI. Justification for Granting Relief:
The surface examination of the pressurizer integrally welded attachments are being conducted to the fullest extent practical. As noted herein, physical access is restricted thereby preventing full Code examination coverage.
Relief was granted by the NRC for the pressurizer support skirt weld and support skirt bracket
(-
\.
attachment weld during the First Ten-Year Interval. These included First Ten-Year Interval Requests for Relief RR-10 and RR-16 for VEGP-1 and 2. NRC approval was documented in 7-30 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL f REOUEST FOR RELIEF NO. RR-8 (continued)
VI. Justification for Granting Relief (continued):
correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
Compliance with the Code would require redesign or replacement of the affected component.
Such a change in an operating plant would be completely impractical. Therefore, SNC requests that relief by authorized pursuant to 10 CFR 50,55a(g)(6)(i).
VII. Implementation Schedule:
The subject examinations will be perfonned to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
4 4
a J
l l
l 1
7-31 Rev.O j
O SOUTHERN NUCLFAR OPERATING COMPANY O O VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-8 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-K / B10.10 11201-V6-002-W35 VEGP-1 pressurizer Heater penetrations 50 % RR-10 support skirt B-K / B10.10 21201-V6-002-W35 VEGP-2 pressurizer Heater penetrations 50% RR-10 support skirt B-K / B10.10 11201-V6-002-W23 VEGP-1 pressurizer Pressurizer support rack 0% RR-16 '
through W30 support bracket attachment assembly
. welds .
B-K / B10.10 21201-V6-002-W23 VEGP-2 pressurizer Pressurizer support rack 0% ,
RR-16 through W30 support bracket attachment assembly welds i
i i
t i
7-32 Rev. O t e . _ _ _ _ _m___ . _ _ _ _ . _ _ _ _ _ _ . . . _ _ _ _ _ _ ____________________________m_____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2
,_ SECOND TEN.YE AR INTERVAL REOUEST FOR RELIEF NO. RR-9 I. System /Companent(s) for Which Reliefis Requested:
Manual and/or mechanized volumetric examination of pressure-retaining welds in centrifugally-cast stainless steel piping and static-cast stainless steel elbows made of SA 351-CF8A material in the Reactor Coolant System from the outside diameter.
II. Code Requirement:
(a) Article III-2430 of ASME Section XI requires that manual ultrasonic examination scanning shall be done at twice (+6dB) the primary reference level as a minimum.
(b) ASME Section XI, Article I, Supplement 1(b) requires that the calibration block thickness to be of a size sufficient to contain the entire examination path. Also, Article III, paragraph 111-3410 requires that basic calibration blocks be made from material of the same nominal diameter and nominal wall thickness as the pipe to be examined.
III. Code Requirement from Which Reliefis Requested:
(a) Reliefis requested fr'om the Code requirements cited above. Scanning will be performed at the primary reference level.
(b) Calibrations will be performed utilizing two calibration blocks, one with a 2.45 inch
. nominal wall thickness and one with a 3.00 inch nominal wall thickness.
i IV. Hanjs for Relief:
(a) The cast SA-351, CF8A material contains a banded microstructure that consists of a duplex grain size ranging from extremely coarse to very fine. This irregular grain structure ciauses significant attenuation and some angular variations during a typical shear-wave ultrasonic examination. Therefore, a 1.0 inch dual-element focused ultrasonic transducer, utilizing a 45 degree refracted longitudinal wave with a frequency of 1.0 megahertz is used. During ,
calibration, the primary reference level is set using side-drilled holes, with the notch brought to the Distance Amplitude Correction'(DAC) curve, as allowed by ASME Section XI, Article 111-3230. Scanning is possible only at the primary ref:rence level due to excessive noise associated with the higher gain levels and the metallurgical structure of the material. A demonstration using this technique was performed for NRC Region 11 and was determined to be a conservative method of detecting reflectors from the inside diameter (ID). (Reference NRC report numbers 50-425/85-24 and 50-425/85-25).
7-33 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE El ECTRIC GENERATING PLANT. U. NITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-9 (continued) l IV. Ihais for Relief (continued): ,
l b) Calibration / examination with the two calibration blocks of 29.0" Diameter / 2.45"
- Thickness (Block No. 331 A) and 32.0" Diameter / 3.00" Thickness (Block No. 329A),
along with the use of ASME Coc.Aase N-461, satisfy Code requirements for wall ,
thickness. Also, ASME Section V, Paragraph T-542.2.1 allows a tolerance of .1". The i applicable SNC cast stainless ultrasonic examination procedure, as well as procedures for l ultrasonically examining piping, will address the requirement for verification of actual wall ;
thickness for appropriate screen range determination. The curvature of the calibration
)
blocks aho meet the requirement of ASME Section XI Article 1, Appendix I, Supplement '
- l. ;
I V. Alternate Examtnation: l l
Ultrasonic examination scanning will be performed at the primary reference level rather than twice the reference level for the centrifugally-cast stainless steel piping and static-cast elbows in i the Reactor Coolant System. In addition, calibrations for examining the subject piping and i fittings will be performed using the method described above.
4 i VI. Justification for Grantimr Relief: .
- Relief was granted by the NRC for using the primary reference level for ultrasonic scanning for the First Ten-Year Interval. These included First Ten-Year Interval Requests for Relief RR-22 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
)
Relief was granted by the NRC for use of these calibration blocks during the First Ten-Year Interval. These included First Ten-Year Interval Requests for Relief RR-23 for VEGP-1 and 2.
j NRC approval was documented in correspondence dated November 26,1991 and December 17, !
1991 for VEGP-1 and 2, respectively. A subsequent revision of the original requests for relief !
were approved by the NRC as documented in correspondence dated March 8,1996 and August I 13,1996 for VEOP-1 and 2, respectively.
t The examinations of the centrifugally-cast stainless' steel piping and static-cast elbows in the Reactor Coolant System are being conducted to the fullest extent practical. As a result, SNC l requests that relief be authorized pursuant to 10 CFR 50.55a(a)(3)(i) since the proposed J
alternatives provide an acceptable level of quality and safety.
7-34 Rev.0
__. . = . . . - . - - _. _. .-. _ - . - . .. -
SOUTHERN NUCI EAR OPERATING COMPANY VOGTI E ELECTRIC GENERATING PLANT. UNITS I AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-9
- (continued) vil.Impi e g i
The subject examinations will be performed to the extent practical during the Second Ten-Year Interval which commences May 31,1997 I
J i
t i
j O
7-35. Rev. 0
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN YEAR INTERVAL REOUEST FOR REI IEF NO. RR-10
- 1. System /Comnonent(s) for Which Reliefis Requested:
Volumetric examination of pressure-retaining nozzle to elbow welds in Steam Generator (Class 1), Reactor Coolant Pump Inlet nozzle to elbow welds and Reactor Coolant Pump outlet nozzle to pipe welds as identified in Attachment I to this request for relief.
II. Code Requirement:
ASME Section XI, Table IWB-2500-1, Examination Category B-J, Item No. B9.11, requires volumetric and surface examination of pressure-retaining welds in piping. The applicable examination volume is shown in ASME Section XI Figure IWB-2500-8.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing a full Code-Coverage volumetric examination of the piping welds identified in Attachment I to this request for relief. ;
IV. Basis for Relief: )
O )
O Examinations for these welds are limited to the " half node" technique due to the coarse grain structure of the cast piping material (Reference RR-9). Due to geometric configuration, a meaningful ultrasonic examination cannot be performed from the nozzle side. Scanning for reflectors parallel to the weld coverage will be accomplished using a one half-node refracted longitudinal wave from one beam direction only. In addition, transverse scans will be performed on the weld and the elbow side.
l V. Alternate Examination: i No alternative examinations are proposed.
VI. Justification for Grantino Relief: , l The examinations are being conducted to the extent practical.
Relief was initially granted by the NRC during the First Ten Year Interval for those welds in Attachment I having requests for relief submitted for tl.em. These included First Ten-Year Interval Requests for Relief RR-17 and RR-24. NRC approval was documented in i correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively.
7-36 Rev. O
SDUTIIERN NUCIIAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-10
- (continued)
VI. Justification for Grantina Relief (continued):
Examinations performed during the First Ten Year Interval provided an accep'.able level of quality and safety. Therefore, reliefis requested pursuant to the requirements of 10 CFR
, 50.55a(a)(3)(i).
Vll. Implcmpntation Schedule:
1 The subject examinations will be performed to the extent practical during the Second Ten-Year
. Interval which commences May 31,1997.
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7-37 Rev.0
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r3 SOUTHERN NUCI F AR OPERATING CO?ITrANY YQGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-10 (continued)
ATTACHMENT 1 ASME Section XI IdentificatioF. No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-J / B9.11 11201-691-5 VEGP-131" Elbow to SG Nozzle configuration 50 % RR-17 Nozzle B-J / B9.11 11201-002-5 VEGP-131" Elbow to SG Nozzle configuration 50% RR-17 Nozzle B-J / B9.11 11201-003-5 VEGP-131" Elbow to SG Nozzle configuration 50% RR-17 Nozzle B-J / B9.11 11201-004-6 VEGP-131" Elbow to SG Nozzle configuration 50% RR-17 Nozzle B-J / B9.11 11201-005-1 VEGP:131"SG Nozzle to Nozzle configuration -
50% RR-17 Elbow B-J / B9.11 11201-006-1 VEGP-131" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 11201-007-1 VEGP-131" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 11201-008-1 VEGP-131" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 11201-005-8 VEGP-131" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 11201-006-8 VEGP-131" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 11201-007-8 VEGP-131" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 11201-008-8 VEGP-131" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle 7-38 Rev.0
O .O O SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL '
REOUEST FOR RFI IEF NO. RR-10 (continued)
ATTACHMENT I (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int. l Category / Item No. Percentage RR No.
B-J / B9.11 11201-009-1 VEGP-127.5" RC Pump Nozzle configuration 75 % RR-24 Nozzle to Pipe i
B-J / B9.11 11201-010-1 VEGP-127.5" RC Pump Nozzle configuration 75 % RR-24 l
Nozzle to Pipe B-J / B9.11 11201-011-1 VEGP-127.5" RC Pump Nozzle configuration 75 % RR-24 Nozzle to Pipe B-J / B9.11 11201-012-1 VEGP-127.5" RC Pump Nozzle configuration 75 % RR-24 Nozzie io Pipe B-J / B9.11 21201-001-5 VEGP-2 31" Elbow to SG Nozzle configuration 50% RR-17
- Nozzle ,
B-J / B9.11 21201-002-5 VEGP-2 31" Elbow to SG Nozzle configuration 50% RR-17 Nozzle B-J / B9.11 21201-003-5 VEGP-2 31" Elbow to SC Nozzle configuration 50 % RR-17 Nozzle B-J / B9.11 21201-004-6 VEGP-2 31" Elbow to SG Nozzle configuration 50 % RR-17 Nozzle B-J / B9.11 21201-005-1 VEGP-2 31" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 21201-006-1 VEGP-2 31" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 21201-007-1 VEGP-2 31" SG Nozzle to Nozzle configuration 50% RR-17 Elbow B-J / B9.11 21201-008-1 VEGP-2 31" SG Nozzle to Nozzle configuration 50 % RR-17 Elbow 7-39 Rev.0
l O O SOUTHERN NUCLF AR OPERATING COMPANY O
VOGTLE FI FCTRIC GENERATING PLANT. UNITS 1 AhP 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-10 (continued)
A'ITACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-J / B9.11 21201-005-8 VEGP-2 31" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 21201-006-8 VEGP-2 31" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 21201-007-8 VEGP-2 31" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 21201-008-8 VEGP-2 31" Elbow to RC Nozzle configuration 90 % RR-24 Pump Nozzle B-J / B9.11 21201-009-1 VEGP-2 27.5" RC Pump Nozzle configuration
- 75 % RR-24 Nozzle to Pipe B-J / B9.11 21201-010-1 VEGP-2 27.5" RC Pump Nozzle configuration 75 % RR-24 Nozzle to Pipe B-J / B9.11 21201-011-1 VEGP-2 27.5" RC Pump Nozzle configuration 75 % RR-24 Nozzle to Pipe B-J / B9.11 21201-012-1 VEGP-2 27.5" RC Pump Nozzle configuration 75 % RR-24 Nozzle to Pipe 7-40 Rev.O
SOUTHERN NUCI FAR OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 l SECOND TEN-YEAR INTERVAL l REQUEST FOR RELIEF NO. RR-11
- 1. System /Comnonent(s) for Which Reliefis Requested: l Volumetric examination of pressure-retaining pipe branch connection welds for nominal pipe size 4 inches and greater (Class 1) as identified in Attachment I to this request for relief.
II. Code Requirement:
ASME Section XI Table IWB-2500-1, Examination Category B-J, Item No. B9.31, requires !
volumetric and surface examinations on branch pipe connection welds. The examination areas !
are shown in ASME Section XI Figures IWB-2500-9,-10, and -11. I l
III. Code Requirement from Which Reliefis Requesinl:
1 Reliefis requested from performing a full Code Coverage volumetric examination of the branch connection welds identified in Attachment I to this request for relief.
IV. Ilasis for Rldisf: ,
(\ Examination of branch connection welds is typically difficult due to the configuration of the branch connection fitting and the weld design. For branch connection welds in the VEGP !
Reactor Coolant System (RCS) primary loop piping, these problems exist in addition to the problems of examining cast stainless steel pipe material. (See Attachment 2) l Two basic weld configurations are used for the branch connection designs on the RCS primary loop piping. The 4 inch branch connections were installed using a " set-on" weld design, while the 6 ,10 ,12 , and 16-inch branch connections were installed using a " set-in" design.
Typically, examination coverage of branch connection welds can be obtained by scanning from the main run of pipe using a 45 degree shear wave technique. In this case, due to the cast i stainless steel material used in the RCS primary loop piping, the examination is limited to a 1/2 node examination using a 45 degree refracted longitudinal (RL) wave technique developed for this piping material.
Examination of the 4-inch branch connection welds.from the main run of piping using the 1/2 node RL wave is not possible due to the geometry of the " set-on" configuration. However, partial coverage of the 6 ,10 ,12 , and 16-inch branch connection welds from the main run using the 1/2 node RL technique is possible because of the " set-in" configuration. The VEGP branch connection configuration is depicted in Attachment 2.
7-41 Rev.0 .
SOUTIIERN NUCLEAROPERATING COMPANY ;
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 i SECOND TEN-YEAR INTERVAL I O
b REOUEST FOR RELIEF NO._RR-11 1
(continued)
V. Alternate Examblation: )
It was determined that the only feasible examination from the branch connection side would be a l refracted shear wave technique from the taper of the fitting. This technique relies on the ability of the shear wave to reflect cff of the inner wall of the fitting bore. Calibration blocks were built with a 10% notch for sensitivity and 3/16 inch side-drilled holes for establishing a DAC curve.
The forged stainless steel material used in all but the 10-inch branch connections allows the use of shear wave ultrasonic techniques; however, th geometry of the fittings still presents problems in obtaining Code-required examination coverage. The fitting side of the 10-inch branch connection is cast stainless steel which precludes the use of a shear wave technique. Also, due to the geometry of the part, examination from the branch connection side using the 1/2 node RL technique is not feasible. -
VI. Justification for Granting Relief:
J The examinations are being conducted to the fullest extent practical.
Relief was initially granted by the NRC during the First Ten-Year Interval for those welds in
- Attachment I having requests for relief submitted for them. These included First Ten-Year Interval Requests for Relief RR-21 for VEGP-1 and 2. NRC approval was documented in correspondence dated N'ovember 26,1991 and December 17,1991 for VEGP-1 and 2, respectively. For the 4 inch,6 inch,12 inch and 16 inch, the " Approximate Percentages" are less than reported during the first interval. Although the same technique is utilized, circumferential scan limitations were not initially included in calculations which reduce the accumulative Code Coverage obtained by one half. Performing circumferential scans for the applicable branch connections is not possible due to weld configurations. Examinations performed during the First Ten-Year Interval provided an acceptable level of quality and safety. While coverages listed in Attaclunent I vary from those shown for the First Ten-Year Interval, the quality of the examinations remain the same. Therefore, reliefis requested pursuant to the requirements of 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedule:
The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
. ,O
. U
- 7-42 Rev. 0 I
O O SOUTHERN NUCLEAR OPERATING COMPAhT O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-11 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
u-J / B9.31 11201-009-4 4" Branch Connection Cire (1) scan (0% cw / ccw(2)) 20 % RR-21 Ax (3) scan (0% pipe / 80% branch)
B-J / B9.31 11201-012-4 4" Branch Connection Cire scan (0% cw / mv) 20% RR-21 Ax scan (0% pipe / 80% branch)
B-J / B9.31 11201-002-2 6" Branch Connection Cire scan (0% cw / ccw) 22% RR-21 Ax scan (40% pipe / 50% branch)
B-J / B9.31 11201-003-2 6" Branch Connection Cire scan (0% cw / ccw) 22 % RR-21 Ax scan (40% pipe / 50% branch)
B-J / B9.31 11201-009-6 10" Branch Connection Cire scan (50% ew / 50% ccw) 50 % RR-21 Ax scan (100% pipe / 0% branch)
B-J / B9.31 11201-010-4 10" Branch Connection Cire scan (50% cw/ 50% ccw) 50% RR-21 Ax scan (100% pipe / 0% branch)
B-J / B9.31 11201-011-5 10" Branch Connection Cire scan (50% cw/ 50% ccw) 50 % RR-21 Ax scan (100% pipe / 0% branch)
B-J / B9.31 11201-012-6 10" Branch Connection Cire scan (50% cw / 50% ccw) 50 % RR-21 l Ax scan (100% pipe / 0% brancid B-J / B9.31 11201-001-2 12" Branch Connection Cire scan (0% cw / ccw) 22 % RR-21 Ax scan (10% pipe / 50% branch) ,
B-J / B9.31 11201-0 # ~ 12" Branch Conacction Cim scan (0% cw / ccw) 22 % RR-21 Ax scan (40% pipe / 50% branch)
B-J / B9.31 11201-004-2 16" Branch Connection Cire scan (0% cw / ccw) 22 % RR-21 Ax scan (40% pipe / 50% branch)
B-J / B9.31 21201-009-4 4" Branch Connection Cire scan (0% cw/ ccw ) 20 % RR-21 Ax scan (0% pipe / 80% branch) 7-43 Rev.0
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SOUTHERN NUCI.F AR OPERATING COMPANY VOGTLE FLECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-11 (continued)
ATTACHMENT I (continued) .
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-J / B9.31 21201-012-4 4" Branch Connection Cire scan (0% cw / ccw ) 20 % RR-21 Ax scan (0% pipe / 80% branch)
B-J / B9.31 21201-002-2 6" Branch Connection Cire scan (0% cw / ccw) 22 % Unit 2 Ax scan (40% pipe / 50% branch) RR-21 B-J / B9.31 21201-003-2 6" Branch Connection Cire scan (0% cw / ccw) 22 % Unit 2 Ax scan (40% pipe / 50% branch) RR-21 B-J / B9.31 21201-009-6 10" Branch Connection Cire scan (50% cw / 50% ccw) 50% Unit 2 Ax scan (100% pipe / 0% branch) RR-21 ;
B-J / B9.31 21201 010-4 -
10" Branch Connection Circ scan (50% cw/ 50% ccw) 50% Unit 2
. Ax scan (100% pipe / 0% branch) RR-21 .
B-J / B9.31 21201-011-5 10" Branch Connection Cire scan (50% cw / 50% ccw) 50% Unit 2 Ax scan (100% pipe / 0% branch) RR-21 ;
B-J / B9.31 21201-012-6 10" Branch Connection Cire scan (50% cw / 50% ccw) 50 % Unit 2 Ax scan (100% pipe / 0% branch) RR-21 i B-J / B9.31 21201-001-2 12" Branch Connection Cire scan (0% cw / ccw) 22 % Unit 2 Ax scan (40% pipe / 50% branch) RR-21 B-J / B9.31 21201-004-3 12" Branch Connection Cire scan (0% cw / ccw) 22 % Unit 2 :
Ax scan (40% pipe / 50% branch) RR-21 B-J / B9.31 21201-004-2 16" Branch Connection Cire scan (0% cw/ ccw) 22 % Unit 2 I Ax scan (40% pipe / 50% branch) RR-21 Notes:
- 1. Cire scan - Circumferential scan for reflectors parallel to the weld.
- 2. CW - Clockwise & CCW - Counter clockwise
- 3. Ax scan - Axial scan for reflectors perpendicular to the weld.
7-44 Rev.0
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. SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
.O REOUEST FOR RELIEF NO. RR-12
, G I. System / Component (s) for Which Relief is Regug31cd:
Volumetric examination of pressure-retaining circumferential welds in 10-inch diameter Safety Injection piping made from SA-376 and SA-312 grade material, as identified in Attachment I to l this request for relief. l II. Code Requirement: ,
Item No. B9.11, Category B-J, Table IWB-2500-1 and Item No. C5.11, Category C-F-1, Table IWC-2500-1 of the ASME Section XI require a surface and volumetric examination of ,
, circumferential piping welds. The required examination volume is shown in ASME Section XI l Figures IWB-2500-8(c) and IWC-2500-7(a). ASME Section'XI, Article 111-4420 requires that the examinations shall be performed using sufriciently long examination beam path to provide coverage of the required examination volume in two beam-path directions. The examination shall be performed from two sides of the weld where practicable, or from one side of the weld as a minimum. ASME Section XI, Article 111-4430 requires that the angle beam examination for i reflectors transverse to the weld shall be performed on the weld crown on a single scan path to examine the weld root by one-half v-path in two directions along the weld. Article Ill (3 Supplement 4 also requires an additional examination volume for reflectors transverse to the V weld to include 1/2" on each side of the weld. ASME Section XI, Article 111-2430 requires that manual scanning shall be done at twice (+6dB) the primary reference as a minimum.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required volumetric examinations of the welds identified in Attachment 1.
IV. Basis for Relief:
The SA-376 and SA-312 materials exhibit severe angular variations and significant attenuation problems during a typical shear-wave ultrasonic examination. This was de: ermined to be caused by severely banded microstructure. A meaningful ultrasonic examination could not be accomplished on the SA-376 and SA-312 materials using conventional shear-wave techniques.
In addition, physical limitations exist due to the geometric configuration of some joints. l l
V. Alternate Examination:
A refracted-longitudinal (L) wave, which is a 1/2-node examination technique, was found to be the best technique during preservice inspection activities. During calibration for the refracted L-p wave examination, the primary reference level will be set using side-drilled holes. Scanning was b
7-46 Rev. O I l
- . - _ ~ - - - - - . _ _ . .. _ _ .
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 '
SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-12 (continued) ,
V. Alternate Eununinntion (continued):
4 possible only at the primary reference level due to the excessive noise associated with the metallurgical structure of the material. On pipe to valve configurations, the weld was examined i
on the pipe side using the half-node technique. Scanning from the valve side was not possible i due to valve geometry. On pipe to tee configurations, the weld was examined on the pipe side using the half node technique. Only partial scanning from the tee was possible due to geometry.
VI. Justification for Grantina Relief: l l
Relief was granted by the NRC for the First Ten-Year Interval. These included First Ten-Year Interval Requests for Relief RR-26 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively. Subsequent revisions to the requests for relief were approved by the NRC as documented in correspondence dated January 6,1993 and August 13,1996 for VEGP-1 and 2, respectively. Examinations conducted during the First Ten-Year Interval provided an acceptable level of quality and safety.
L The examinations of the Class 1 and 2 Safety Injection System piping are being conducted to the fullest extent practical in the same manner as for the First Ten-Year Interval. As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(a)(3)(i) since the proposed alternative examinations provide an acceptable level of quality and safety.
VII. Implementation Schedule:
The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
l O -
7-47 Rev. O
O O O SOUTHERN NUCII AR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RFI IEF NO. RR-12 (continued)
A'ITACIIMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. _
Percentage RR No.
B-J / B9.11 11204-124-1 VEGP-1 10* Valve to Pipe Geometry of valve 50% RR-26 l Weld (100% pipe / 0% valve) ;
B-J / B9.11 11204-124-7 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.1 i 11204-124-8 VEGP-1 10" Valve to Pipe Geometry of valve 50% RR-26 ,
Weld (100% pipe / 0% valve)
B-J / B9.11 11204-124-11 VEGP-1 10" Pipe to Tee Geometry of tee 88% RR-26 ,
Weld B-J / B9.11 11204-124-12 VEGP-1 10" Tee to Pipe Gebmetry of tee 88% RR-26
. . Weld B-J / B9.11 11204-124-15 VEGP-1 10" Pipe to Valve Geometry of valve 50 % RR-26 :
Weld (100% pipe / 0% valve)
B-J / B9.11 11204-124-16 VEGP-1 10" Valve to Pipe Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve) ,
B-J / B9.11 1126 . 5-1 VEGP-1 10" Valve to Pipe Geometry ofvalve 50 % RR-26 i Weld (100% pipe / 0% valve)
B-J / B9.11 11204-125-7 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 ; '
Weld (100% pipe / 0% valve)
VEGP-1 10" Valve to Pipe Geometry of valve RR-26 B-J / B9.11 11204-125-8 50 %
Weld (100% pipe / 0% valve) i B-J / B9.11 11204-125-15 VEGP-1 10" Pipe to Valve Geometry ofvalve 50% RR-26 Weld (100% pipe / 0% valve) !
B-J / B9.11 11204-125-16 VEGP-1 10" Valve to Pipe Geometry of valve 50 % RR-26 ,
Weld (100% pipe / 0% valve) i 7-48 Rev.0 l
- _ _ . - - - _ . _ - - _ - -__.-._-_-_--.-__-_____---__._.-_-.---__.--a ,--v , - ,
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UhTFS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-12 (continued)
ATTACHMENT 1 (continued)
! ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-J / B9.11 11204-126-1 VEGP-1 10" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve) ,
B-J / B9.11 11204-126-7 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-126-8 VEGP-1 10" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-126-15 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-126-16 VEGP-110" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-127-1 VEGP-1 10" Pipe to Valve Geometry ofvalve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-127-7 VEGP-1 10" Pipe to Valve Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-127-8 VEGP-1 10" Valve to Pipe Geometry ofvalve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-127-19 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 11204-127-20 VEGP-1 10" Valve to Pipe Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-124-1 VEGP-210" Valve to Pipe Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve) 7-49 Rev.0
SOUTHERN NUCI FAR OPERATING COMPANY VOGTI E ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-12 (continued)
ATTACHMENT I (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Categog/ Item No. Percentage RR No.
B-J / B9.11 21204-124-7 VEGP-210" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-124-8 VEGP-210" Valve to Pipe Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-124-12 VEGP-210" Tee to Pipe Geometry of tee 88 % RR-26 Weld B-J / B9.11 21204-124-15 VEGP-210" Pipe to Valve Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve) ;
B-J / B9.11 l i204-124-16 VEGP-210" Valve to Pipe -Geometry of valve 50% RR-26 -
Weld (100% pipe / 0% valve) .,
B-J / B9.11 21204-125-1 VEGP-210" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-125-7 VEGP-210" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-125-8 VEGP-210" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-125-12 VEGP-210" Tee to Pipe Geometry of tee 88 % RR-26 Weld B-J / B9.11 21204-125-15 VEGP-210" Pipe to Valve Geometry ofvalve 50 % RR-26 Weld) (100% pipe / 0% valve)
B-J / B9.11 21204-125-16 VEGP-210" Valve to Pipe Geometry ofvalve 50 % RR-26 Weld (100% pipe / 0% valve) 7-50 Rev.O
O SOUTHERN NUCLEAR OPERATING COMPANY O O VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RFI IEF NO. RR-12 (continued)
A'ITACHMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
B-J / B9.11 21204-126-1 VEGP-210" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-126-7 VEGP-210" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-126-8 VEGP-210" Valve to Pipe Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-126-12 VEGP-210" Tee to Pipe Geometry of tee 88 % RR-26 Weld B-J / B9.11 21204-126-15 VEGP-210" Pipe to Valve Geometry of valve '
50% RR-26
. Weld (100% pipe / 0% valve)
B-J / B9.11 21204-126-16 VEGP-210" Valve to Pipe Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-127-1 VEGP-210" Valve to Pipe Geometry of valve 50 % RR-26 Weld _
(100% pipe / 0% valve)
B-J / B9.11 21204-127-7 VEGP-210" Pipe to Valve Geometry of valve 50 % RR-26 i Weld (100% pipe / 0% valve)
B-J / B9.11 21204-127-8 VEGP-210" Valve to Pipe Geometry ofvalve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-127-19 VEGP-210" Pipe to Valve Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
B-J / B9.11 21204-127-20 VEGP-210" Valve to Pipe Geometry ofvalve 50 % RR-26 Weld (100% pipe / 0% valve) 7-51 Rev.O
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-12 (continued)
ATTACHMENT I (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
C-F-1 / C5.11 11204-120-6 VEGP-1 10" Pipe to Valve Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
C-F-1 / C5.11 11204-121-6 VEGP-1 10" Pipe to Valve Geometry of valve 50 % RR-26 Weld (100% pipe / 0% valve)
C-F-1 / C5.11 11204-122-6 VEGP-1 10" Pipe to Valve Geometry ofvalve 50 % RR-26 Weld (100% pipe / 0% valve)
C-F-1/ C5.11 11204-123-6 VEGP-1 10" Pipe to Valve Geometry of valve 50% RR-26 Weld (100% pipe / 0% valve)
C-F-1/ C5.11 21204-123-6 VEGP-210" Pipe to Valve Geometry ofvalve 50 % RR-26
. Weld (100% pipe / 0% valve) ,
7-52 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTI E ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
- REOl'EST FOR RELIEF NO. RR-13 I. System / Components (s) for Which Reliefis Requested
Volumetric examination of pressure-retaining piping welds in Class 1 systems as identified in Attachment I to this request for relief.
II. Rcqujrrment from Which Reliefis Requested:
.ASME Section XI Category B-J, Table IWB-2500-1 requires surface and volumetric examination of pressure-retaining welds in Class I piping. Applicable examination volumes are shown in ASME Section XI Figure IWB-2500-8 and includes essentially 100% of the weld length. In addition, ASME Section XI Appendix 111, Supplement 4, requires that the angle beam examination for reflectors transverse to the weld be performed on the weld crown and 1/2 inch of the base material on each side of the weld.
4 III. Code Requirement from Which Reliefis Requested:
- Reliefis requested from performing a full Code Coverage volumetric examination of the Class 1 piping welds identified in Attachment I to this request for relief.
", \ IV. liasis for Relief:
Physid limitations due to geometric configuration of the welded areas restrict coverage of examination volume as required by Figure IWB-2500-8. For the First Ten-Year Interval, the transverse scans were only required on the weld crown per Appendix 1I1 of the 1983 Edition of 3 ASME Section XI with Addenda through the Summer of 1983. This requirement could usually be met for most configurations. The new requirement to perform the transverse scans on the weld crown and 1/2 inch of base material on each side of the weld will not be determined until examinations are performed during the Second Ten-Year Interval.
- V. Ahernate Examination
i l No alternate examination is proposed.
i VI. Justification for Granting Relief:
The volumetric examinations of the Class 1 piping welds are being conducted to the fullest extent practical. As noted herein, physical access is restricted thereby preventing full Code examination coverage.
O 7-53' Rev.0
=
[
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
( REOUEST FOR RELIEF NO. RR-13 (continued VI. JustlGcation for Granting Relief (continued):
In order to examine 100% of the weld volume, that systems would require extensive modifications. The resulting increase in plant safety would not compensate for the burden that would result from imposition of the requirements.
As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(a)(3)(ii) since imposition of the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety were they to be imposed.
VII. Implementation Schedule:
The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
[
O 7-54 Rev.O
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-13 (continued)
ATTACIIMENT I ASME Section XI Identification No. Description Limitation Approximate Ist Int.
Category / Item No. Percentage RR No.
B-J (See Note 1) (See Note 1) (See Note 1) (See Note 1) (See Note 1)
Notes:
- 1. To be determined during examinations conducted during the Second Ten-Year Interval. Additions to this request for relief will be resubmitted.
k L
i i
l 7-55 Rev.O i i
l SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE EISCTRIC GENERATING PLANT. UNITS 1 AND_2 1
SECOND TEN-YEAR INTERVAL REOUEST FOR REL IEF NO. RR-14 l
l
- 1. System / Component (s) for Which Reliefis Requested:
Volumetric examination of pressure-retaining welds and nozzle inner radius sections of vessels and surface examinations of nozzle reinforcement plates (Class 2) as identified in Attachment 1 i to this request for relief. l II. Code Reunirement:
ASME Section XI, Category C-A, Item No. C1.20, Table IWC-2500-1, requires volumetric examination of pressure-retaining welds of Class 2 pressure vessels. Applicable examination !
volumes are shown in Figures IWC-2500-1 and 2. ASME Section XI, Category C-B, Item Nos.
C2.21 and C2.22, Table IWC-2500-1, requires volumetric examinations ofpressure-retaining welds of Class 2 pressure vessels and nozzle inner radii and surface examinations of nozzle ;
reinforcement plate welds. The applicable examination volume is shown in Figure IWC-2500-4. ;
ASME Section XI, Table IWC-2500-1, Examination Category C-G, Item C6.10, requires surface )
examination of pressure-retaining welds in Class 2 pumps. The examination area is shown in l Figure IWC-2500-8.
/ III. Code Requirement from Which Reliefis Reaucsigd:
Reliefis requested from performing the Code-required volumetric examinations of the components identified in Attachment I to this request for relief.
IV. Basis for Relief:
(a) The configuration of the steam generator (SG) steam outlet nozzle is such that no inner radius exists. The nozzle is manufactured from a forging that is a solid block of steel.
Seven (7) holes, each 81/2-inch diameter, have been drilled through this forging to provide an outlet for the steam. Thus, this nozzle does not have a conventional inner radius.
(b) Geometric configuration of the SG main steam, auxiliary feedwater and main feedwater nozzles presents physical limitations that prevent complete coverage during ultrasonic examination. Scanning from the nozzle side is not feasible.
(c) The configuration of the VEGP Residual Heaf Removal (RHR) heat exchanger (Hx) nozzles differs from that shown in ASME Section XI Figure IWC-2500-4(c). Although the icinforcing plate welded to the vessel has a rounded configuration in the flow it is not a true nozzle inner radius when compared with t,he configuration in Figure IWC-2500-4(c) Please refer to Attachment 2 for a figure depicting the VEGP configuration. It is not possible to perform an inner radius ultrasonic examination since the interface between the reinforcing i % plate and the RHR heat exchanger vessel wall prohibits volumetric examination. Although
! the reinforcement plate is welded to the inside.
7-56 Rev. O
SOUTIIERN NUCLEAR OPERATING COMPANY VOCTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-14 (continu'e d)
IV. Basis for Relief (continued):
diameter of the heat exchanger wall, the reinforcing plate-to-vessel welds are inaccessible.
Therefore, it is impractical to perform a surface examination on the reinforcing plate-to-vessel welds.
(d) For the safety injection (SI) pumps, access limitations due to geometric configuration of the welded areas. Flanges and supports restrict coverage of required examination volume and areas. Full Code Coverage is not possible in the vicinity of the obstructions.
(e) For the VEGP-1 Boron Injection Tank (BIT), the nozzle and shell configuration presents physical limitations that prevent complete coverage during ultrasonic examinations.
~
V. Alternate Examination:
No alternate examination is proposed. The affected Class 2 vessel welds are being examined to fullest extent practical.
VI. Justification for Granting Relief:
The examinations are being conducted to the fullest extent practical. ;
Relief was initielly granted by the NRC during the first ten-year interval for those welds / components in Attachment I having requests for relief submitted for them. These i included First Ten-Year Interval Requests for Relief RR-28, RR-29, RR-30 and RR-32. NRC l approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, respectively. It was determined during the first ten-year interval review process that geometric configurations and interferences make the volumetric examinations of these welds impractical to perform to the extentrequired by the Code. The subject components would require extensive modifications in order to obtain complete compliance with the specific requirements of ASME Section XI. The increase in plant safety would not compensate for the burden placed on the licensee that would result from imposition of the requirement.
Although there are physical obstructions which limit the amount of examination coverage for those components identified in Attachment 1, reasonable assurance still exists that an acceptable I level of quality and safety will be maintained since there have been no catastrophic failures of l Class 2 pressure vessels. SNC requests that relief be authorized pursuant to 10 CFR ;
50.55a(g)(6)(i) since imposing the Code requirements is impractical. l n
v 7 Rev.O
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-14 (continued)
VII. Impicmentation Scheduls: ,
The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
O V
O 7-58 Rev.O
O O SOUTHERN NUCIFAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL .
REOUEST FOR RELIEF NO. RR-14 (continued)
ATTACHMENT I ASME Section XI Identification No. Description Limitation Approximate Ist Int.
Category / Item No. Percentage RR No.
C-A / C1.20 11204-V6-001-WO2 VEGP-1 BIT Head to Shell Vessel supports and taper 38% RR-30 Weld (76% head / 0% shell)
C-A / Cl.20 11204-V6-001-WO3 VEGP-1 BIT Head to Shell Vessel taper 50 % RR-30 . ,
Weld (100% head / 0% shell)
C-B / C2.21 11201-B6-001-W18 VEGP-1 SG Main Steam Nozzle configuration 50% RR-29 outlet nozzle to head weld (100% shell / 0% nozzle) ,
C-B / C2.21 11201-B6-002-W19 VEGP-1 SG Feedwater Nozzle configuration 50% RR-29 nozzle to shell weld (100% shell / 0% nozzle)
C-B / C2.21 11201-B6402-W26 VEGP-1 SG Auxiliary Nozzle configuration 50 % RR-29 -
Feedwater nozzle to shell (100% shell / 0% nozzle)
C-B / C2.21 11204-V6-001-W01 VEGP-1 BIT Nozzle to Nozzle configuration 50 % RR-30 Head Weld (100% shell / 0% nozzle)
C-B / C2.21 11204-V6-001-WO4 VEGP-1 BIT Nozzle to Nozzle configuration 50% RR-30 Head Weld (100% shell / 0% nozzle)
C-B / C2.21 21201-B6-001-W18 VEGP-2 SG Main Steam Nozzle configuration 50 % RR-29 outlet nozzle to head weld (100% shell / 0% nozzle) i C-B / C2.21 21201-B6-002-W19 VEGP-2 SG Feedwater Nozzle configuration 50% RR-29 nozzle to shell weld (100% shell / 0% nozzle)
C-B / C2.21 21201-B6-002-W26 VEGP-2 SG Auxiliary Nozzle configuration 50 % RR-29 ;
feedwater nozzle to shell (100% shell / 0% nozzle) ,
C-B / C2.22 11201-B6-001-IR03 VEGP-1 SG Main Steam Nozzle configuration 0% RR-28 '
outlet nozzle inner radii (no inner radius)
C-B / C2.22 11205-E6-001-IR01 VEGP-1 RHR Hx Inner Nozzle configuration 0% RR-32 i
& IR02 radii (0% shell / 0% nozzle) f 7-59 Rev.0 .
i
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-14 (continued)
ATTACHMENT 1 (continued) l ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
C-B / C2.22 11205-E6-002-IR01 VEGP-1 RHR Hx Inner Nozzle configuration 0% RR-32
& IR02 radii (0% shell / 0% nozzle) t C-B / C2.22 21201-B6-001-IR03 VEGP-2 SG Main Steam Nozzle configuration 0% RR-28 outlet nozzle inner radii (no inner radius)
C-B / C2.22 21205-E6-001-IR01 VEGP-2 RHR Hx Inner Nozzle configuration 0% RR-32
& IR02 radii (0% shell / 0% nozzle) .
C-B / C2.22 21205-E6-002-IR01 VEGP-2 RHR Hx Inner Nozzle configuration 0% RR-32
& IR02 radii (0% shell/ 0% nozzle)
- C-B / C2.31 11205-E6-001-WO9, VEGP-4 RHR Hx Nozzle configuration 0% RR-32 thru W12 reinforcement plate (Inaccessible)
C-B / C2.31 11205-E6-002-WO9, VEGP-1 RHR Hx Nozzle configuration 0% RR-32 thru W12 reinforcement plate (Inaccessible)
C-B / C2.31 21205-E6-001-WO9, VEGP-2 RHR Hx Nozzle configuration 0% RR-32 thru W12 reinforcement plate (Inaccessible)
C-B / C2.31 21205-E6-002-WO9, VEGP-2 RHR Hx Nozzle configuration 0% RR-32 thru W12 reinforcement plate (Inaccessible)
C-G / C6.10 11204-P6-003-WO2 VEGP-1 SI Pump Casing Pump supports 66 % RR-30 to Suction Nozzle Weld (Partially inaccessible)
C-G / C6.10 21204-P6-003-WO2 VEGP-2 SI Pump Casing Pump supports 66 % RR-30 to Suction Nozzle Weld (Partially inaccessible) 7-60 Rev.0
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECrMIC GENERATING PLANT, UNITS 1 AND 2
! SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-14 l (continued)
ATTACHMENT 2 W
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7-61 Rev. O l
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 ;
SECOND TEN-YEAR INTERVAL l
REOUEST FOR RELIEF NO. RR-15 l l
I. System /Comnonents(s) for Which Reliefis Requested:
Volumetric examination of pressure-retaining piping welds in Class 2 systems as identified in Attachment I to this request for relief.
1 i
II. Requirement from which Reliefis Requested:
ASME Section XI, Category C-F-1, Table IWC-2500-1 requires surface and volumetric examination of pressure-retaining welds in Class 2 piping. The applicable examination volumes !
are shown in Fig. IWC-2500-7 and includes essentially 100% of the weld length. In addition, i ASME Section XI, Appendix III, Supplement 4 requires that the angle beam examination for ;
reflectors transverse to the weld be performed on the weld crown and 1/2 inch of the base material on each side of the weld.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing a Full Code Coverage volumetric examination of the Class 2 piping welds identified in Attachment I to this request for relief.
IV. Basis for Relief:
Physical limitations due to geometric configuration of the welded areas restrict coverage of examination volume as required by Figure IWC-2500-7. For the first ten-year interval, the transverse scans were only required on the weld cro>vn per Appendix III of the 1983 Edition of ASME Section XI with Addenda through Summer 1983. This requirement could usually be met for most configurations. The new requirement to perform the transverse scans on the weld crown and 1/2 inch of base material on each side of the weld will not be determined until examinations 1 are performed during the Second Ten-Year Interval.
V. Alternate Ermmination:
No alternate examination is proposed.
VI. Justification for Granting Relief:
The volumetric examinations of the Class 2 piping welds are being conducted to the fullest extent practical. As noted herein, physical access is restricted thereby preventing full Code examination coverage.
O 7-62 Rev.O
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-15 !
(continued) !
VI. Justification for Granting Relief (continued):
- Relief was granted by the NRC for the Class 2 piping welds during the First Ten-Year Interval.
These included First Ten-Year Interval Requests for Relief RR-34, RR-35, RR-36 and RR-37 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991
, and December 17,1991 for VEGP-1 and 2, respectively. It was determined that in order to examine 100% of the weld volume, that systems would require extensive modifications. The resulting increase in plant safety would not compensate for the burden that would result from imposition of the requirement. While Code coverage during the second ten-year interval may vary due to the imposition of the new Code requirement, the level of quality will not change from that obtained during the first ten-year interval.
As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(a)(3)(ii) since imposition of the Code requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety were they to be imposed.
VII. Imniementation Schedule:
,O) .
V The subject examinations will be performed to the fullest extent practical during the Second Ten-Year Interval which commences May 31,1997.
(
l l
1 1
l
[
\
7-63 Rev.0 3
C O 0 SOUTHERN NUCLEAR OPERATING COMPANY
. VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-15 ,
(continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No. ,
C-F-1 (See Note 1) (See Note 1) (See Note 1) (See Note 1) (See Note 1)
Notes:
- 1. ' To be determined during examinations conducted during the Second Ten-Year Interval. Additions to this request for relief will be resubmitted.
b L
j 7-64 Rev.0
. . . - . . . . - - _ - - - - . ..- --_- . - .. . - - ..- - - ~- . - . . - - - -
SOUTHERN NUCLEAR OPERATING COMPANY VOGTI E ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN YEAR INTERVAL REOUEST FOR RELIEF NO. RR-16 I. System / Components (s) for Which Reliefis Requested:
4 Volumetric examination of pressure-retaining piping welds in Class 2 systems as identified in
] Attachment I to this request for relief.
t Requirement from Which Reliefis Requested:
J ASME Section XI, Category C-F-2, Item C5.51, Table IWC-2500-1 requires surface and volumetric examination of pressure-retaining welds in Class 2 piping. The applicable i examination volumes are shown in Figure IWC-2500-7 and include essentially 100% of the weld i length.
III. Code Requirement from WDich Reliefis Requested:
Reliefis requested from performing a Full Code Coverage volumetric examination of the Class 2 l piping welds identified in Attachment I to this request for relief.
IV. Basis for Relief:
Physical limitations due to geometric configuration of the welded areas restrict coverage of the ,
examination volume required by Figure IWC-2500-7. I
- V. Alternate Examination
i No attemate examination is proposed.
VI. Justification for Granting Relief:
] ,
! The volumetric examinations of the Class 2 piping welds are being conducted to the fullest 1 extent practical. As noted herein, physical access is restricted thereby preventing full Code !
examination coverage.
i Relief was granted by the NRC for the Class 2 piping welds during the First Ten-Year Interval. l These included First Ten-Year Interval Requests for Relief RR-34, RR-35, RR-36 and RR-37 for VEGP-1 and 2. NRC approval was documented in correspondence dated November 26,1991 and December 17,1991 for VEGP-1 and 2, rest iectively.
As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55a(g)(6)(i) since imposition of Code required examinations is impractical.
O .
7-65 Rev. O
i SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN. YEAR INTERVAL REOUEST FOR RELIEF NO. RR-16 4
(continued) ,
VII. Implementation Schedule:
- The subject examinations will be performed to the fullest extent practical during the Second Ten- ;
j Year Interval which commences May 31,1997.
- 1 1 I i
4
- 1 2 I O
1 l
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O O SOUTHERN NUCLEAR OPERATING COMPANY O '
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-16 (continued)
ATTACHMENT 1 ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
C-F-2 11301-001-5 VEGP-129.5" Pipe to l 0% from valve side,64% 32 % RR-3 4 Valve from pipe side C-F-2 11301-001-6 VEGP-129.5" Valve to 0% from valve side,100% 50% RR-34 Pipe from pipe side C-F-2 11301-001-8 VEGP-129.5" Pipe to 0% from valve side,82% 41 % RR-34 Valve from pipe side C-F-2 11301-002-5 VEGP-129.5" Pipe to 0% from valve side,82% 41 % RR-34 Valve from pipe side :
C-F-2
- 11301-002-6 VEGP-129.5" Valve tb 0% from valve side,100% 50% -RR-34 Pipe , from pipe side C-F-2 11301-002-8 VEGP-129.5" Pipe to 0% from valve side,100% 50 % RR-34 Valve from pipe side C-F-2 11301-003-5 VEGP-129.5" Pipe to 0% from valve side,64% 32 % RR-34 Valve from pipe side C-F-2 11301-003-6 VEGP-129.5" Valve to 0% from valve side,64% 32 % RR-34 Pipe from pipe side C-F-2 11301-003-8 VEGP-129.5" Pipe to 0% from valve side,64% 32 % RR-34 Valve from pipe side C-F-2 11301-004-5 VEGP-129.5" Pipe to 0% from valve side,88% 44 % RR-34 :
Valve from pipe side C-F-2 11301-004-6 VEGP-129.5" Valve to 0% from valve side,82% 41 % RR-34 Pipe from pipe side C-F-2 11301-004-8 VEGP-129.5" Pipe to 0% from valve side,82% 41 % RR-34 ,
Valve from pipe side i 7-67 Rev.0
O O SOUTHERN NUCLEAR OPERATING COMPANY O
VOGTLE FLECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR (continued)
ATTACIIMENT 1 (continued)
ASME Section XI Identification No. Description Limitation Approximate 1st Int.
Category / Item No. Percentage RR No.
C-F-2 11302-108-14 VEGP-16" Pipe to 22% from penetration side, 86 % RR-34 Penetration 86% from pipe side C-F-2 21301-004-10 VEGP-2 29.5" Pipe to 0% from valve side,85% 85 % RR-34 ,
Valve from pipe side !
C-F-2 21301-004-13 VEGP-2 29.5" Pipe to 0% from valve side,85% 85 % RR-34 Valve from pipe side C-F-2 21305-062-7 VEGP-216" Valve to Pipe 0% from valve side,86% 86 % RR-34 from pipe side 7-68 Rev.O i
_.._____..._.__..__ __. .____ ______.______ ____________ __ _ e w
- . . _. = - . . . . . . . . . . - . - .. ..
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 a
SECOND TEN. YEAR INTERVAL j
- Q
/O REOUEST FOR RELIEF NO. RR-17 s
e I. System / Component (s) for Which Reliefis Requested:
Each of the ASME Class 2 pressure-retaining welds in austenitic stainless steel or high alloy piping included within the scope of the ISI Program with ASME Examination Category C-F-1, Item Number C5.11.
l II. Code Requirement:
Table IWC-2500-1, Examination Category C-F-1, item C5.11 in the 1989 Edition of ASME Section XI requires a volumetric and surface examination of the piping welds. These examinations are to be performed on 7.5%, but not less than 28 welds, of all austenitic stainless steel or high alloy welds not exempted by IWC-1220. Welds below 3/8 inch nominal wall thickness are not required to be nondestructively examined per Examination Category C-F-1.
These welds, however, shall be included in the totalveld count to which the 7.5% sampling rate is applied.
III. Code Requirement from Which Reliefideauested:
O Reliefis requested.from distributing the nondestructive examinations (NDE) of 7.5% of the Class 2 austenitic or high alloy welds not exempted by IWC-1220 among only the non-exempt welds contained in Examination Category C-F-1.
1 I
IV. Basis for Relief:
The majority of the Class 2 pressure-retaining austenitic stainless steel piping welds in this scope at VEGP-1 and 2 are greater than 4 inches nominal pipe size (NPS) and less than 3/8 inches nominal wall thickness. These welds are not exempted by IWC-1220 but because of their wall thickness are not required to have NDE performed per Examination Category C-F-1. Therefore, if only the non-exempted welds greater than 4 inch 6 NPS and greater than or equal to 3/8 inches nominal wall thickness are considered for NDE, the selection process may exclude entire systems !
from NDE. I V. Alternate Examination:
Southern Nuclear Operating Company proposes to not only include Class 2 pressure-retaining austenitic stainless steel piping welds greater than 4 inches NPS and less than 3/8 inches nominal wall thickness in the total weld count to which the 7.5% sampling rate is applied as b
7-69 ~ Rev. O
.~ - - . _ - - _ _ - _ _ - - - -
SOUTHERN NUCLEAR OPERATING COMPANY 1 YOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-17
- (continued)
V. Alternate Ernmination (continued):
required by Examination Category C-F-1, but to also include these welds as part of the weld population from which the 7.5% sample of welds receiving NDE is selected. As a result, NDE will be performed on a 7.5% sample of all non-exempt piping welds greater than 4 inches NPS.
The examinations will be distributed among the Class 2 systems prorated, to the degree practical, i
on the number of nonexempt austenitic stainless steel or high alloy welds in each system. Within a system, the examinations will be distributed among terminal ends and structural discontinuities
. prorated, to the degree practical, on the number of non-exempt terminal ends and structural
. discontinuities in that system; and within each system, examinations shall be distributed between l line sizes prorated to the degree practicable. Structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, i etc., conforming to ANSI B16.9), and pipe branch connections and fittings. Each weld selected
- will receive a surface and volumetric examination with the exception of socket welds and pipe 1
branch connections of branch piping which will receive a surface examination only.
VI. Justification for Granting Relief:
im
- In the First Ten-Year Interval, the " Multiple Stream Concept" was used in the selection of Class
- 2 piping welds in the Residual Heat Removal (RHR), Emergency Core Cooling System (ECCS),
l and the Containment Heat Removal (CHR) systems per the 1974 Edition of ASME Section XI with Addenda through Summer 1975 as required by 10 CFR 50.55a(b)(2)(iv). Georgia Power Company (GPC), the former licensee of VEGP, chose not to apply the pressure and temperature exemption on RHR, ECCS, and CHR systems and r.equired that 7.5% of the welds in these j systems not exempted by IWC-1220(a) or IWC-1220(c) be volumetrically examined once each ten year interval. GPC also committed to volumetrically examine, once each ten year interval, a 7.5% sample of the welds in these systems which only required a surface examination by the Code. (In addition, GPC committed to volumetrically examine 7.5% of the welds four inches and smaller but greater than or equal to two inches on high pressure safety injection systems once each ten year interval). Therefore, the scope of the Class 2 pressure-retaining austenitic stainless steel welds for VEGP-1 and 2 in the First Ten-Year Interval was similar to the scope defined by the 1989 Edition of the ASME Section XI. SNC is simply requesting a continuation of the philosophy that was used during the first interval for the selection of the subject Class 2 welds, except that in lieu of the " Multiple Stream Concept", a straight 7.5% sample will be taken.
These piping welds greater than 4 inches NPS and less than 3/8 inches nominal wall thickness make up the majority of the VEGP-1 and 2 Class 2 pressure-retaining austenitic stainless steel O
7-70 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND_2 SECOND TEN-YEAR INTERVAL
[ REOUEST FOR RELIEF NO. RR-17 (continued)
. VI. Justification for Granting Relief (continued):
piping welds. Including 7.5% of these welds in the NDE scope will provide for a better representative sample of the total number of Class 2 austenitic stainless steel piping welds found i in both units as well as providing an acceptable level of quality and safety. Therefore, it is requested that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedule: '
i The proposed alternative examinations will be performed during the Second Ten-Year Interval j which commences May 31,1997.
l l
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I O -
7-71 Rev.O
___ _ _ - __ _ __ . . _ _ _ _ . .m .
SOUTHERN NUCLEAR' OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR REI IEF NO. RR-18 I. System / Component (s) for Which Reliefis Requested:
)
This request for reliefis applicable to all ASME Class 2 Nuclear Service Cooling Water (NSCW) pressure-retaining welds included within the scope of the ISI Program. These welds are included in VEGP-1 and 2 systems 1202.1501, and 1515. Specifically, these welds are to be examined in accordance with the requirements of Table IWC-2500-1, Examination Category C-F-1, Items C5.11, C5.30, and C5.41.
II. Code Requirement:
Table IWC-2500-1, Examination Category C-F-1, Item C5.11, in the 1989 Edition of ASME Section XI require a volumetric and surface examination of the piping welds and Items C5.30 and C5.41 require a surface examination of piping welds. These examinations are to be l perfonned on 7.5%, but not less than 28 welds, of all austenitic stainless steel and high alloy welds not exempted by IWC-1220. Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Category C-F-1. These welds, however, should be included in the total weld count to which the 7.5% sampling rate is applied.
III. Code Requirement from Which Reliefis Requested:
Relief is requested from counting the Class 2 piping welds contained in VEGP-1 and 2 NSCW Systems 1202,1501, and 1515 as part of the total weld count from which the 7.5% sample of welds are to be nondestructively examined.
IV. Basis for Relief:
The NSCW system is a cooling water system that performs a Class 3 function as defined in NRC Regulatory Guide 1.26, Revision 3," Quality Group Classifications and Standards for Water ,
Steam , and Radiological-Waste-Containing Components of Nuclear Power Plants", and therefore per Table 2500-1, Examination Categories D-A, D-B, and D-C, no nondestructive examinations are required to be performed on the pressure-retaining welds. There is a small portion of this piping that is classified ASME Olass 2 as it provides a containment isolation function as well the Class 3 cooling water function. These Class 2 pressure-retaining welds are greater than 4 inches nominal pipe size (NPS) and less than 3/8 inches nominal wall thickness and do not require nondestructive examination per ASME Section XI Table IWC-2500-1, Examination Category C-F-1.
O 7-72 Rev.0
1 SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN.YE AR INTERVAL REOUEST FOR RELIEF NO. RR-18 )
(continued)
V. Alternate Eunmination: I System pressure testing will be performed on the Class 2 portions of the NSCW system piping as described in IWA-5000 and IWD-5000 or in accordance with ASME Section XI Code Case N-498-1. In addition, the associated component supports will be examined per the requirements of ASME Section XI Code Case N-491. The Class 2 pressure-retaining welds of NSCW will be eliminated from the total population of ASME Category C-F-1 welds from which the 7.5%
sample of welds am to be nondestructively examine.d. To insure containment integrity, Primary Reactor ContMmnent Leakage Testing as required by 10 CFR 50, Appendix J will continue to be l performed.
VI. Justification for Granting Relief: l The NSCW system is a cooling water system that performs a Class 3 function as defined in NRC Regulatory Guide 1.26, Revision 3, and therefore, per Table IWD-2500-1, Examination Categories D-A, D-B, and D-C, no nondestructive examinations are required to be performed on the pressure-retaining welds. To satisfy the Class 3 function Inservice Inspection (ISI) requirements, this piping receives system pressure test as described in IWA-5000 and IWD-5000 or in accordance with ASME Section XI Code Case N-498-1 (See Request for Relief RR-16) and the component supports are examined per the requirements of ASME Section XI Code Case N-491. There is a small portion of this piping that is classified ASME Class 2 as it provides a containment isolation function as well the Clasq 3 cooling water function. (Note: To insure its containment integrity, the VEGP-1 and 2 containments receive 10 CFR 50, Appendix J Primary Reactor Containment Leakage Tests.) The piping welds in this portion of the NSCW system receive the required Class 3 examinations described above and because they are classified Class 2, are required to be counted as part of the total population of ASME Category C-F-1 welds from which the 7.5% sample of welds are to be nondestructively examined. These NSCW welds, however, are not required to be nondestructively examined per Examination Category C-F-1 because of a NPS greater than 4 inches and a nominal wall thickness less than 3/8 inches.
Therefore, the Class 2 portion of the NSCW system is required to receive the same level ofISI as the Class 3 portion and performs primarily a cooling water function. The net impact of the Examination Category C-F-1 requirement is that it increases the quantity of pressure-retaining welds to be examined and results in those examinations being required on Class 2 systems other than NSCW. Because the primary function of these welds is a Class 3 cooling water function, SNC believes that the ISI requirements for Class 3 are adequate for providing an acceptable level of quality and safety. Also, in the First Ten-Year Interval, SNC performed nondestructive examinations on 7.5% of the Class 2 NSCW pressure-retaining welds and did not find any significant problems with the welds examined. Therefore, it is requested that the proposed A alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
b .
7-73 Rev. O
SOUTHERN NUCLEAR OPERATING COMPM L3GTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEA'R INTERVAI, REOUEST FOR RELIEF NO. RR-18 (continued)
VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, 1997.
O -
l 7-74 Rev. 0 l
SOUTHERN NUCLEAR OPERATING COMPANY !
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 '
SECOND TEN YEAR INTERVAL 9 REOUEST FOR RFI IEF NO. RR-19 (G
I. System / Component (s) for Which Reliefis Requested:
ASME Class 2 piping included within the scope of the ISI Program.
Examination Category C-F-1, Items C5.12, C5.22, C5.42 Examination Category C-F-2, Item C5.52, C5.62, C5.82 II. Code Requirement:
l Table IWC-2500-I of the 1989 Edition of ASME Section XI, Examination Category C-F-1, Items C5.12 and C5.22 and Examination Category C-F-2, items C5.52 and C5.62 require a ;
surface and volumetric examination as defined by Figure IWC-2500-7. Examination Category C-F-1, Item C5.42 and Examination Category C-F-2, Item C5.82 require a surface examination. l III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required surface and volumetric examination of the above identified piping longitudinal welds
/'~T U IV. Basis for Relief:
I Code Case N-524 which was approved August 9,1993 by ASME addresses the alternative '
requirements for surface and volumetric examination requirements oflongitudinal piping welds.
By implementing the provisions of this ASME Section XI code case, personnel radiation i exposure, outage examination time, and costs can be significantly reduced at VEGP. A copy of Code Case N-524 is provide as Attachment I to this request for relief.
V. Alternate Examination:
Southern Nuclear Operating Company will comply with the requirements of ASME Section XI, Code Case N-524 as follows:
(a) When only a surface examination is required, examination oflongitudinal piping welds is not r.: quired beyond those portions of the welds within the examination boundaries of the intersecting circumferential welds, and (b) When both surface and volumetric examination are required, examination oflongitudinal piping welds is not required beyond those portions of the welds within the examination boundaries ofintersecting circumferential welds provided the following requirements are met:
O v
1-75 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR kELIEF NO. RR-19 (G)
(L0nlin' icd)
V. Alternate Examination (continued):
(1) Where longitudinal welds are specified and locations are known, examination requirements shall be met for both transverse and parallel fla vs at the intersection of the welds and for that length oflongitudinal weld within the circumferential weld examination volume, and (2) Where longitudinal welds are specified but locations are unknown, or the existence of longitudinal welds is uncertain, the examination requirements shall be met for both transverse and parallel flaws within the entire examination volume ofintersecting circumferential welds.
VI. Justification for Granting Relief:
The proposed alternative testing requirements h, ave b' een evaluated by the ASME Code Committee and have been deemed acceptable for determining the pressure boundary integrity of the affected components. The proposed alternative requirements, in accordance with the Code O Case, will provide reasonable assurance that unallowable inservice flaws have not developed in
\
the subject welds or that they will be detected and repaired prior to return of the reactor vessel to service. Dms an acceptable level of quality and safety will have been achieved and public health and safety will not be endangered by allowing the proposed alternative examination in lieu of the Code requirements. Therefore, it is requested that the proposed alternative examinations be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedule:
The proposed alternative examinations, as noted in the Code Case, will be performed during the Second Ten-Year Interval which commences May 31,1997.
7-76 Rev.0
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECrMIC GENERATING PLANT, UNITS 1 AND 2 4
p SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-19 (continued) a ATTACHMENT 1 CASE i
N-524 CASF.51W WtF. hut 11R AND PRESSL'RE \ ESSEL CODE Approval Dete: August 9,190s See Numerwalindex for expwehon and any reeWurmanon dates.
Case N-514 7
Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI. Division 1
- i Inquuy
- What alternative requirements may be ap-plied to the surface and volumetric examination of longitudinal piping welds specified in Table IWB-1 25001. Examination Category BJ, Table IWC-2500
- 1. Examination Categories C F 1 and C F 2 (Exam.
s ination Category C-F prior to Winter 1983 Adden.
5 v) da), and Table IWC-2520. Examination Category C.
G (1974 Edition Summer 1975 Addenda)?
' Reply It is the opmion of the Committee that the following shall apply-(a) When only a surface examinacion is required, i exammation of longitudinal piping welds is not re.
' quired beyond those portions of the welds within the examination boundaries of intersecting circumfer-ential welds.
(b) When both surface and volumetric examina-tions are required, examination oflongitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of inter-secting circumferential welds provided the following requirements are met.
j (D Where longitudinal welds are specified and l locations are known, examination requirements sha!!
be met for both transverse and parallel flaws at the intersection of the welds and for that length of lon-gitudinal weld within the circumferential weld ex-amination volume; (2) Where longitudinal welds are specified but locations are unknown, or the existence of longitu-dinal welds is uncertain, the examination require-ments shall be met for both transverse and parallel flaws within the entire examination volume of inter.
secting circumferential welds.
m i
J 7-77 Rev. 0 1
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-20 I. System / Component (s) for Which Reliefis Requested:
Each of the ASME Class 1,2, and 3 Integrally Welded Attachments included within the scope of the ISI Program. Specifically, these include the following:
ASME Examination Category B-H, Items B8.10 and B8.20 for Vessels, ASME Examination Category C-C, Items C3.10, C3.20, and C3.30 for Vessels, Piping, Pumps, and Valves, ASME Examination Category D-A, Items Dl.20, DI.30, and Dl.40 for Systems In Support Of Reactor Shutdown, ASME Examination Category D-B, Items D2.20, D2.30, and D2.40 for Systems In Support Of ECC, CHR, Atmosphere Cleanup, and Reactor RHR, and l
ASME Examination Category D-C, Item D3.20 for Systems In Support of RHR and Spent Fuel Storage Pool. ,
It should be noted that certain code item numbers related to integrally welded attachments are not included in the above listings. These are not included because either the current plant design does not include integrally welded attachments which met the item number description or the integrally welded attachments are similar in design to other item number descriptions and have been included therein. Should a change in the plant design occur or existing integrally welded i attachments be reevaluated and identified with an item number not listed above, this request for relief would be applicable to the examination requirements for those integrally welded attachments.
II. Code Requirement:
Table IWB-2500-1, Examination Category B-H, Items B8.10 and B8.20 in the 1989 Edition of ASME Section XI require a volumetric or surface examination of the integrally welded attachments that meet certain conditions as noted in the subject table. Table IWC-2500-1, Examination Category C-C, Items C3.10, C3.20, and C3.30 of that same Code edition require a surface examination of the integrally welded attachments that meet certain conditions as noted in the subject table. Table IWD-2500-1, Examination Category D-A, Items Dl.20, DI.30, D1.40; Examination Category D-B, Items D2.20, D2.30, D2.40; and Examination Category D-C, Item D3.20 of that same Code edition require a visual examination of the integrally welded attachments that meet certain conditions as noted in the subject table.
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE EI ECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-20 (continued)
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required volumetric, surface, or visual examination on those Integral Attachments required by the above referenced tables.
IV, Basis for Relief:
On November 25,1992, ASME issued Code Case N-509 (copy provide as Attachment I to this request for relief) which approved a set of alternative rules for the selection and examination of Class 1,2, and 3 Integrally Welded Attachments under ASME Section XI, Division 1. This Code Case has not been formally endorsed by inclusion in NRC Regulatory Guide 1.147 but has been authorized by the NRC for use previously at VEGP and other plants. ASME Section XI l Code Case N-509 was approved by the NRC for use at VEGP during the First Ten-Year Interval !
as addressed in NRC correspondence d '.ted March 8,1996 and August 13,1996 for VEGP-1 and l 2,respectively. (
Reference:
First Ten-Year Interval Requests for Relief RR-61 for VEGP-1 and l 2.) l V. Alternate Ernmination:
SNC proposes that the following examinations be p'erformed in lieu of the Code-required volumetric, surface, or visual examination on those Integrally Welded Attachments required by Table IWB-2500-1, IWC-2500-1, and IWD-2500-1 in the 1989 Edition of ASME Section XI:
Surface Examinntions:
Those integrally welded attachments as specifically noted in ASME Code Case N-509, ASME Examination Category B-K, Integral Attachments for Class 1 Vessels, Piping, Pumps, and Valves, and ASME Examination Category C-C, Integral Attachments for Class 2 Vessels, Piping, Pumps, and Valves.
I Visual Examinntions: .
ASME Examination Category D.A, Integral Attachments for Class 3 Vessels, Piping, ;
Pumps, and Valves.
O !
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s SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 4
SECOND TEN-YEAR INTERVAL h
O REOUEST FOR RELIEF NO. RR-20 (continued) 4
- VI. Justification for Granting Relief:
Code Case N-509 provides an alternative sampling which will retain an acceptable level of quality and safety for Class 1,2, and 3 Integrally Welded Attachments. Since approval was granted by ASME, the alternative requirements should be technically acceptable for determining
- flaws and authorized pursuant to 10 CFR 50.55 a(a)(3)(i) provided a minimum of 10% of the 3 total number ofintegral attachments in all Class 1,2, and 3 systems are examined. A VEGP-1
- study was performed previously that compared the number ofintegrally welded attachment
{ examinations required under the present ASME Section XI scope (based on the requirements of the 1983 Edition of ASME Section XI with Summer 1983 Addenda which was the code of I l reference for the First Ten-Year Interval) versus the number ofintegrally welded attachment examinations required under ASME Code Case N-509. The study is shown in Attachment 2 to this request for relief, and indicates that at least 10% of the present ASME Section XI Integrally Welded Attachment scope for piping will be examined when ASME Code Case N-509 is implemented. Since Units 1 and 2 are similar at VEGP, the results of the VEGP-1 study will be typical for VEGP-2. By implementing the alternative examinations, cost savings, personnel radiation dose, and outage time can be realized by SNC for VEGP-1 and 2.
) VII. Implementation Schedule:
The proposed alternative examinations will be performed during the Second Ten-Year Interval which commences May 31,1997.
I 7-80 Rev.O i
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECI KIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL p REOUEST FOR RELIEF NO. RR-20 d .
(continued)
ATTACHMENT 1 i
l CASE 4
N-509
, CASES OF ASME BO!! ER AND PaES$t!RE YESSEL CODE 1
Approwel Date: November 25.1H2 See Numere indeir for esowntoon and any restfirmstron dates.
1 Case N 509 1.1 Exemption Criteria Alternative Rules for the Selection and Examinaion (a) The exemption criteria provided in IWB 1220, of Class 1. L and 3 Integrally Sielded AttachmentsSection XI, Division 1 IWC 1220, and IWD-1220 may be applied to Cass 1,2, and 3 components respectively, with integrally 1
welded attachments, required to be examined in +.
Ingwyr What alternative s ,quirements to those of cordance with Table 2500-1, IWB. !WC, and 1C may be used to select and (b) Cass 1,2, and 3 integrally welded attachment E i examine integrally welded attachments?
craminations performed as a result of component
.. support deformation cannot be credited under the Reply: It is the opinion of the Committee that the '* 9" * * *"*3 '
following rules may be used to select and examine r IWC-2412, and IWD-2411 or IWD-2412, respec-integrally welded attachments: I *I7'
] (a) This Case is limited to Examination Categories B-H, B K 1, C-C, D A, D B, and D-C. 1.2 Inspection Schedule 4
(b) Class 1. 2. and 3 component supports shall be selected for examination in accordance with IWF of Cass 1,2, or 3 integrally welded attachments se-the 1989 Edition with the 1990 Addenda. lected for examination by sample selection criteria in E (c) Except for the selection of component supports accordance with Table 2500-1. Examination Cate.
for examination, all references to Section XI within gories B K, C-C. and D-A, shall meet the require-this Case shall be from the edition and addenda spec. ments of IWB-2411 or IWB 2412, IWC-2411 or sfied in the Owner's Inservice Inspection Program. !WC 2412, or IWD 2411 or IWD 2412, repectively.
i IJ Additional and Successive Examinations 1.0 SCOPE (a) Cass 1,2 and 3 additional and successive ex-
- amination requirements ofIWB 2430 and IWB-2420 These requirements apply to examination and for Cass 1. IWC 2430 and IWC-2420 for Cass 2 and sample selection of Class 1,2, and 3 integrally welded 3 as applicable, shall be applied to integrally welded attachments of vessels, piping, pumps, and valves attachments whose examinations reveal flaws or rel-listed in Table 25001 as follows: j evant conditions that exceed the acceptance stan-(a) Table 2500-1 Examination Category B K shall dards of IWB-3000, IWC 3000, and IWD-3000, re-i be used for Class 1 integrally welded attachments in spectively. l Examination Categories B H and B-K-1 of IWB. (b) When integrally welded attachments are ex-(b) Table 2500-1. Examination Category C-C shall amined as a result of identified component support l
I be used for Cass 2 integrally welded attachments in defonnation and the results of these examinations Examination Category C-C of IWC.
exceed the applicable acceptance standards listed (c) Table 25001. Examination Category D-A shall above, additional or successive examinations shall be l
be used for Class 3 integrally welded attachments in performed when determined necessary based on an Examination Categories D A, D-B, and D-C ofIWD. evaluation by the Owner.
l i
V 7-81 Rev.O
. - _., . s_ , . . . .- - - ~ ~ ~ _ -_ . n. . - - _
u - ._ -. . _ _ = . + . - -
%J \ q) zp in : <
O TABLE 2500-1 0-EXAMINAil0N CATEGORIES CD o3 h a ERAMINAll0N CATEGORY 8 K INTEGRAL ATTACHkENIS FOR CLASS 1 VES$E LS, PIPING, PUMPS, AND VALVES (#1 ,
{--
Examination e M ltem Nequerementsf E manunation Acceptance Entent of Frequency el S g fee. Parts Esamined* Fig No. Method Standard Esam6mation" E manunatiose p 81010 Pressure Vessels IW8-2500- 13, Surfme' IW8-3516 100% of required areas of each Each identified occurrence and g '
M Intryatly Wended 14,and 15 welded attacament emh inspection interwaf* g Attactonents h k O g "
8is 20 Piping IWS 250013, Surfme IWS 3516 100% of required areas of exh Each identtred occurrence and Integrasty Weided 14, and 15 welded attachment each 6nspection interval * **e k @ w r Attacionents $ O d C O h M N Sie 30 Pump
- IWS 250013, Surfme IW8-3516 100% of required areas of each Each identifled occurrence and so Q Q Z M i M M
= g Z mr :c 14, and -15 welded attactenent each inspection interwat* O O lategrally VI/elded -
0e mm .n.ents
$- O. % d' %
81040 Values IW8 2500 13, Surfme IWS 3516 100% of required areas of emh Iach 6dentified occurrence and $ N N lategretty Weided 14,and 15 welded attachment emh inspection interval
- M Attachments 3 ^
Q. h N M O q M O i mo rt s. e a 2' q [
til E meninateen es limited to these Inteyalty welded attacionents that swet the following cen$tions: g M m w v., anmh ni es en u.e omside sur me .f e e prenu, retainin, term (b) the attachment provides conyonent support as defined in NF-t i10; and e ,
d a
g Q p a
g g
kl the auxhnernt weld joins the attachment either directly to the sesrtue of the temponent er to an integrally cast er forged attachrnent to the cer*ponent g fD g h (2) The estent of the esaminateen includrs essent6affy 100% of the length of the attachment weld at emb attahment sutsett le eaamir:atson-
<3) Setecard sarrytes of 6ntegracy welded attactanents shall be enamined emh inspection interval.
r*
e1 h g M d
(4) la tie case af nwatpie vessels af struitar dessyg, function and serwke, only one integrally welded attachment of enty one of the muhele vessels shall be selected for eaarrunation. @ tJ t O ;
(5) f or poing, purryg and valves, a sanele of 10% of the welded attachments associated w6th the component supports selected for esarnenataan under the 1990 Addenda, BWF-2510 shalt t's O L be emarruned.
tu Esamination is retymred whenever convenerit support member deformat6en te g , brokeg bent, er pulled out partsi is identified during operatten, refueang, sna4ntenance, emanunate inservke 6nspectasq, or testing. (p) k (P) f or the tesdeguratiet shown in Fig. IW8-2500-14, a ve4 metric enamenation of valuene A 8-C-D from 56de 18 Cl of the circumferen+.ial welds may be perforened in fleu of the surfme ,,,,
esammation of surfues A 0 and 8-C.
> +
Z C
to N
3 i
i
. ., _ =.mm. -- .-_ _. ._.--m. _ _ . . . . _m . . _. ..__ _ . . _ _ _ . . _ . _ . .
g f"% a
\
0 TAsLE 25ee-1(cestT*o)
EXAllIIIIATISIl CATECORIES O
dm O EXAMINATION CATES 80tf C-C, INTEGRAL ATTACNetENTS F9it CLASS 2 WESSELS, PIPING. PtaMPS, AND WALVES Esassinaties h C*
Iteam Requiremeett/ fiandmaties Acceptance M
Estent of Freguency of F j No. Parts Esamined' flg. Bee. heettied Standard Enamaeatleau g,asi,p. g p C310 Pressere Wessels IwC 2500 5 Swfme integrally Weeded Attalweenes IWC 3532 100% of required areas of each Each identified occurrence and welded attahment enti inspection intervat* { t M
O Z
Q C3.20 Pipleg lateyalty Weided twC 2500 5 Swfme IWC-35t2 100% of required areas of each Exh identified occurrence and b Attachments metded attachment enk inspection 6aternat' R b N
C3 30 Pomps O n O M p lesegrally Welded tWC 2500 5 Swise IWC-3512 1001d of required areas of each Enh identified occurrence and l C g Z '
melded attachment exh inspection interval
- Se g
= At-s g s = m i:i o C3 40 Valves Integrany welded IWC 2900 5 Swfme IWC-3512 100% of required areas of each Enh identifled occurrence and 1 - 8:
O e > "tf M
Attachnents melded attachment each inspection interwaf*
$O h b rT)
NOTES:
E.
= 2 1 .
+
it) E mam6nathi is timited to these Integrafty melded attactunents that meet the folisming canditions: E tai the attachnient is en the outside surface of Ihe p essure retaenisg cenipenent; E
- fn %
(td tne attachntent provides cornponent seppert as defined in NF il10; and (cl the attKhment meld joins tfie attachrnent either directly to the surface of the tenyonent er to an IntegreHy cast er forged attachment to the e h N g g t
- 12) The catent of tie esaminatweiincludes essentially 100% of the tength of the attactenent tuend at each attachment subject te enarpenation.
f1 V % % ,
Q 133 Selected sanyies of lategrapy welded attac1=nents shall be esamined each impection betervat, @
tg it) in the case of sieltiple vessets of sirnitar desigg functies and serwke, endy one integrally wrhied attacturent of enty one of the snuttipte tessets shall be setected for emandnation. g y 1 4 53 For piping, pm and wah% a sample of 10% of Wie melded attactypents assectated with the conyonent supports selected for examinahen under the Ig90 AddeA IWF-25te shall be esamined. *1 [
l 9 a Examinatten is required weienever ces'ponent support eneniber defonnation te e, hrokeg bent, or pulled out parts) is identified during operation, refueling, maintenacce, esamination, imerwice impatieg er teiteg. @ y) n a-*
q p a
m x !
1 O f e
en 8-o W co a O
C i
_ _ . _ _ _ _ . . . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ . _ . _ _ _ _____m_____ . _ _ _ _ _ _ _ _ _ _ _ _ ______._____m_ _ _ _ __ _ ._. . _ - - - . - - . s --
_. . ~ . _ - - . _. _ ~ _ ~ ~ . . _ _ . . . . _ . .- . . . - -- - _ - _ _ .
, n $
L,/ j z $
& =g o TABLE 2500-1 (CONT'B) m EXAMINATION CATEGORIES EXAMfMATIOft CATEGORY D-A, INTEGRAL ATTACitMENTS FOR CLASS 3 VFSSELS, PIPfNS, PUMPS, Afl0 VALVES O$
g O
Emandsstise Itene Regelreeseets/ Esandeatfee Acceptance Estent of Fregsencp el j F tuj g
Ne, Parts Emandsed* Fig. fee. Method Standard Eraadnationu Emandsaties**
3 M p f
01.10 Pressere Wessels latera*1y Wekted IWD-2500 t Visual, VT-I IWD 3000 100% of mruired areas of emh Emh hientilied occurnace and wehted attactmvent enh inspecteen interval
> , O $ g h' Attalvnents Q w Q
08.20 Piping IntegraHy Welded LWD-2500 I ' visent, VT t IWO.3000 100% of requ%d areas of each f ach hientified occurrence and N O -
O O Attachrraents welded attufwnrat exh 6nspection irecewat k *Tj E'$
g gg h p Dt.30 Penops r o N Z M 0g lategrally Wefded SWD-25001 Visual. VI-l IWD-3000 100% of required areas of each Each identined occurrence and enh impection intervat y 7 % N welded attachment o
- g Q DI.40 A-s Vaters lategrally Welded IWD-2500 t Visual, VT 1 IWD 3000 100% of required areas of each ; Each identified occurrence and wehled attachment each trupection Interval e
Q O
- .""*q ,
Attachments 'E]
j N E %
e-.
If0f ES:
I10 Enarnination ls limited to these inteyalty evelded attachments that smeet the lettowing conditions:
.e m E
o i
tal the attahment is en the outside tanface of the pressure retaining component, %
(bl the attactunent provedrs c. . -. siepert as defined in NF Il10, and N y tc) the attactwnent weld joins the attachment either derectfy to the surtme of the cornpenent er to an inteyalty cast er forged attattunent to the carryonent-y ,, Q Cl The entent of the esamination bicludes essentlaffy 100% of the length of the attachment neW at each attachment subject to esaminatten.
p O E33 Selected sampics of integrafly wedded attachmeg diall be esamined each inspection laterval Allinterafty welded attactenents setuted for enaaninatten shall be sub6ett to corrosieg si detennined Dy the Omner, such as the integral'7*/ ded attachments of the Service Water or f mergency Service Water systents. la the case of neitiple vessels of similar desigg n
O funcaten and service, the Integrally welded attr # ets el enty one of the multiple westels shall be selected for esaminatten. For integrally welded attachrnents af pipina -
valves e 10% sample shall be selected ler emarna, dost This percentage saeple shall be pmportional to the total number of nonenempt integrally wehled attutwnents
.F4 W W ; tic h N p piping, pumpg and valvet located within each systern subject to Wiese esaminattens.
- g a=d (e) E maminatlee is required nhenever convenent support member defonnattaa te g, broken, bent, or pulled out parts) is identdied during operation, refueEng snanntenance, enarmnetum, l Imevvice Impettleeg er testing.
g p
i u '
l N
a 4
O I
f
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECi KIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
,]V REOUEST FOR RELIEF NO. RR-20 (continued)
ATTACHMENT 2 VOGTLE ELECTRIC GENERATING PLANT ASME CODE CASE N-509 STUDY UNIT 1 Intentally Weided Attachment Examinations Reauired For Picine ClassI Class 2 Class 3 M Present Scope 0 103 329 432 N-509 Scope 0 9 37 46 Exams Saved 0 94 292 386 NOTES p 1. ASME Code Case N-509 was approved on November 25,1992 but has not yet been Q included in NRC Regulatory Guide 1.147,
- 2. Class L,2, and 3 component supports shall be selected for examination in accordance with IWF of the 1989 Edition w/1990 Addenda of Section XI to the ASME Boiler and Pressure Vessel Code.
- 3. The 1989 Edition w/1990 Addenda of Section XI to the ASME Boiler and Pressure Vessel Code is essentially AShE Code Case N-491 contained in NRC Regulatory Craide 1.147, Revision 10.
- 4. Except for selection ofcomponent supports for examination, all references to ASME Section XI within the code case shall be from the edition and addenda specified in the owners ISI program.
- 5. Table 2500-1 in ASME Code Case N-509 uses Examination Category D-K for Class 1 integrally welded attachments in place ofExamination Categories B-H and B-K-1 ofIWB and Examination Category C-C for Class 2 integrally welded attachments in the place of Examination Category C-C ofIWC. It also uses Examination Category D-A for Class 3 integrally welded attachments in the place of Examination Categories D-A, D-B, and D-C ofIWD.
- 6. The base metal design thickness exemption is lost for Class 1 and 2 integrally welded attachments. '
7-85 Rev.O j
SOUTHERN NUCI FAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS I AND 2 SECOND TEN-YE AR INTERVAL O
d REQUEST FOR RELIEF NO. RR-21 I. System / Component (s) for Which Reliefis Requested:
Reliefis requested from the surface and volumetric examination requirements of vessels and their connections in piping 4" nominal pipe size (NPS) and smaller in:
- 1. Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) systems or portions thereof, except high pressure safety injection systems.
Specifically, the following components are involved:
Component VEGP-1 Tag No. VEGP-2 Tag No.
Regenerative Heat Exchanger 1-1208-E6-001 2-1208-E6-001 Excess Letdown Heat Exchanger 1-1208-E6-002 2-1208-E6-002 Letdown Heat Exchanger 1-1208-E6-003 2-1208-E6-003 Letdown Reheat Heat Exchanger 1-1208-E6-007 2-1208-E6-007 Suction Dampener 1-1208-V4-001 2-1208-V4-001 Discharge Dampener 1-1208-V4-OL
- 2-1208-V4-002 II. Code Requirement:
Table IWC-2500-1, Examination Category C-A* (Items C1.10, C1.20, and C1.30) and Examination Category C-C (Item C3.10) of the 1989 Edition of ASME Section XI require a surface and volumetric examination.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested to exclude the components cited above from the required surface and volumetric examinations as allowed in IWC-1220," Components Exempt From Examination,"in the 1989 Addenda (including subsequent editions and addenda) of ASME Section XI and ASME Code Case N-408-2 which has been endorsed for general use in Revision 9 of Regulatory Guide 1.147. A copy of the aforementioned code case is provided as Attachment 1 to this request for relief.
O 7-86 Rev.0
SOUTIIERN NUCI EAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-? EAR INTERVAL REOUEST FOR RELIEF NO. RR-21 j (continued) u i
- IV. Basis for Relief
4 Subarticle IWC-1220 of the 1989 Addenda of ASME Section XI allowed the exemption of selected components from the surface and volumetric examination requirements ofIWC-1220.
The 1996 Addenda of ASME Section XI also includes these exemptions in IWC-1220. The e
NRC granted these exemptions to VEGP in the first interval through correspondence dated March 8,1996 and August 13,1996 for VEGP-1 and 2, respectively. (Reference First Ten-Year Interval Request for Relief RR-62 for VEGP-1 and 2.)
V. Alternate Examination:
These exemptions exclude the applicable vessels from the surface and volumetric examinations
- required by IWC-2500. The remainder of the Code-required examinations (i.e., pressure tests)
- would be performed to assure that an acceptable level of safety and quality is maintained for the i applicable components.
VI. Justification for Grantina Relief:
O j V These exemptions will be allowed when the newer Addenda and Editions of the Code are authorized in 10 CFR 50.55a. SNC sees no benefit in performing examinations on components which the Code has determined can be exempted. The other requirements in the Code are therefore acceptable to assure an acceptable level of safety or quality. It is impractical to perform ,
, examinations which do not provide a compensating increase in the level of safety or quality.
1 These added exemptions would apply to several components which are in high dose rate areas.
The most significant of these components is the regenerative heat exchanger. A conservative
! whole body dose in the range of one to two Rem is a reasonable estimate for examining the regenerative heat exchanger. The dose rate surveys for the regenerative heat exchanger indicate a contact dose rate of two to three Rem / hour and a dose rate at eighteen inches away from the heat .
i exchanger of one to one-and-one-half (1 to 1-1/2) Rem / hour. The estimated stay time to perform I
- the Code-required examinations on the regenerative heat exchanger is one hour. Such exposure
! is contrary to the principles of ALARA to perfonn examinations on components without a 4
- compensating increase in the level of safety or quality. For the reasons discussed above, SNC
{ has determined that implementation of the Code requirements is impractical. Therefore, SNC requests that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). !
{ VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, O 1997.
7-87 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELEUI KIC GENERATING PLANTS UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL p REOUEST FOR RELIEF NO. RR-21 l
(continued)
ATTACHMENT 1 CASE N-408-2 CASES OF AS%fE 80tLER AND PRESSL'RE 4 ESSEL CODE Approval Date: July 24.1989 l See Numerocal onces for esorrevan I and ersy reaMurmstron dates Case N-40g-2 Alternative Rules for Examination of Class 2 Piping ically pressunzed. passive (i.e.. no pumps) safety in-Section XI. Division 1 jection systems' of pressunzed water reactor plants;
- 1 (6) piping and other components of any size be-Ingwy When determining the components subject yond the last shutoff valve in open ended portions of j
to exammauon and establishing exammanon require- systems that do not contain water durmg normal plant i operatmg conditions.'
ments for Class 2 piping under Secuan XI. Division I what alternative exemptions to those stated in IWC- (61 The following components (or parts of compo.
nents) of systems (or portions of systems) other than 1220 and what alternative exammation requirements RHR. ECC. and CHR systems are exempt from the l to those stated in IWC 2500. Category C F. may be ;
used? volumetnc ond surface examination requirements of IWC 2500; !
r Reply: It is the opmion of the Committee that the (D piping NPS 4 and smaller; j
I (11 vessels, pumps. and valves and their connec-following alternauve rules may be used for determmmg !
tions in piping NPS 4 and smaller; components subject to examination and for establishing examinauon requirements for Class 2 piping under (3) vessels, piping, pumps, valves, other compo-Secuan XI. Division 1. rients. and component connecuons of any size in sys-tems or portions of systems that operate (when the (a) The following components (or parts of compo-nents) of RHR. ECC and CHR systems (or poruons system function is required) at a pressure equal to or of systems)' are exempt from the volumetne and sur- le.s than 175 psig and at a temperature equal to or less than 200'F; face exammation requirements of IWC 2500:
(4) piping and other components of any size be.
(/) piping NPS 4 and smaller in all systems except high pressure safety injecuon systems of pressunzed yond the last shutoff valve in open ended portions of water reactor plants: systems that do not contain water during normal plant I operating conditions.'
(2) vessels, pumps, an,d valves and their connec-2 (c) For welds in austenttic stamless steel or high uons in pipmg NPS 4 and smaller in all systems except I alloy piping, the requirements of Table 1. Examinanon high pressure safety injection systems of pressurtzed water reactor plants; Category C F 1. Preuvre Retaining Welds in Austen-itic Stamless Steel or High Alloy Piping. shall be used (J) piping NPS 14 and smaller in high pressure as an alternative to the requirements of Table IWC-safety injecuan systems of preuunted water reactor 25001.
3 pisnts:
(41 vessels, pumps, and valves and their connec- (d) For welds in carbon or low alloy steel piping.
the requirements of Table 2. Examination Category 1 tions in pipmg NPS 1% and smaller in high pressure !
r C F 2. Pressure Retammg Welds in Carbon or Low safety injection systems of preuunzed water reactor plants: Alloy Steel Piping, shall be used as an alternative to
, the requirements of Table IWC 25001.
/J) vessels, pipmg. pumps, valves other compo-nents, and component connecuons of any size in stat.
' 5 sticanh pressurized. passne sarety miecten systems of pressur-
' RHR ECC. and CHR svarems are Rendual Heas Removat Emer.ued ame nacter plants are typicany called bi such name as gency Core Coosmg. and Conismmerit Heat Remosal svisems, re-spectnety ascumuistor tank and associated sessem, safety inset 1 ion tank and associated essiem. and sore Goodmg tank and associated system
/s peer as de6ned as havmg a cumuisteve mm and a cumulauve outlet pipe cross-serimnal area nestner af which escoeds the nommal
- Nrmal plant operstmp conditions mclude reactor startup. oper-OD crossanctionas area of the designated use. aien as power. hos standbv. and reactor cooldown to cold shutdown
% senditent but do not include tesi conditens 7-88 Rev. 0 a
4
x
]
\
TABLE 1 Zo >
e ftem EXAMI.6 ' 'M CATEGORY C-F-1. PRESSURE RETAINING WELDS IN AUSTENITIC STAINLESS STEEL OR HIGH ALLOY PIPING Esammation o$
A Reemrements/ Enammaten Acceptance No. Parts Enammed' tog No* hlethod Standard' Estent of Enarmnaten* Emannaten*
h3E O C5 le P9i ng welds 2 *,in. . " wah thKkness c O for papeng > NPS 4 C5 Il Cwcumferentiat weld M O IWC- 2500- 7 Salace and IWC 3514 100% of each weld regarsng Enh espection wohsenetric ,a enammation 05 12 tongitudad weld IWC 2500-F Saface and IWC 3514 2 5e - at the intersectmg eterval tej [
Exh especten wohametric tacumferentiai weto enterwat Q Es grj g
C5 20 P9eng wehfs e P, m. noend wat thicknen e e M for P9'a9 2 NPS 2 and g NPS 4 N Q O g.I Z .r Z C5 21 Circumferential weed IWC 2500-P y e Surface and IWC 3514 g O a x 100% of each weed requirme Enh especten g g ] O C woksmetric examenation intervat C5 22 : on,tud.nd weed IWC 2500-7 g y Surface and IWC 3514 2 St - at the intersecting Each inspecten volumetric cercumferential weld enterwd la q qi g C5 30 Soc 6et welds IWC.2500 F Salace IWC-3514 e 100% of each weld reemeng Each inspecteon
$ g enamenaten interwat g
, y q C5 40 Pye branch connections of branch p9mg 2 > M M . g
. ne5 2 y Z ea(8 C5 41 Circeamferent.al weed IWC-2500 9 to Surface . IWC-3514 100% of eacts we d requaing Each especten
.o m ] te] O
- 13. encluseve p Ng C542 t onestueinal weed WC-2500- 8 2 Surfme IWC.nle enamnaten interval h g M g M and -1) 2 5t - at the intersecting Each especten C' ft N g cwcumferenteal wetd enterwas E *II
=0i E 5; < . 2 til Requwements for emanunaten of weeds in piping 5 NPS 4 apply to PWft high pressure safety insectaan systems in accordance with the esempton cratena of Elvs Case. g g-(2)
The weeds selected for esammation shad include F.5%. but not less than 20 welds, of all austenstic stemless steel or begh 490y welds not esempted by thes Case p,
e* (Some welds not g o. {
e.empw. t,, this Caw are noi ugu red to be nondesin.ct,ves, f.anuned pu o.m.nat.on Cate,o,, C.r.l. These weeds. however. shan be mch.ded a the ional wetd coont ie i.tuch the 7 5% samphne rate is apphed ) The emaenmatens shat be d strebuted as foNews:
n <Z h g b
tal the enamenatment shaN 4e ei.stributed among the Class 2 systems prorated, to the degree praticable, en the number of noneuempt austemtsc steentest steet ce tugh> ancy welds h M en that on eachsystem);
system is e., af a system contaens 30% of the nonesempt welds, then 30% of the nondestructeve enammateons requwed by Enamenation Category C41 should be performed w n=6 t' O tbl etthen a systeen. the esaminatsons shaR be destritputed afnDng teringnal ends lSee Note f 3)) and structural descontinueters (see Note (43) prorated, to the degree paar t<able.
on the number of nonenempt ternunal ends and structurat descantinenties en that system; and ect w.th.e enh systent esaminatens shaE be destributed between ime saaes prorated to the degree practicable. @ >
t)) Trenunal ends are the entremetset of papeng runs that connect to structu.es. components tsuch as vessets, pumps, waives) or pope anchors, each of wiuch acts as a setid sestrama or ,
provedes at least two degrees of translatsonal restramt to pipeg thermat emparrsion.
(4) Structural discontenustees enchede pipe weto kmts to vessel nearles, valve bodies, pump casmos, pipe Ideengs (such as esbows. .
tees., .reducers flanges confernung etc to Ah51816 96 k
and pope branch connecteens and fettings.
45)
The welds selected for emarmnatoon shaR be reenamened during subsequent enspection intervals owM the service biet ne of the papeg component.
(6) figure numbers and aceptance standards eefev to those in Secten XI. Water Ig83 Addenda.
N eb O
d SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL l REOUEST FOR RELIEF NO. RR-22 I. Syalem/ Component (s)for Which Reliefis Requested:
1 Each of the ASME Class 1,2, and 3 small items 1-inch and smaller nominal pipe size (NPS) subject to repair or replacement within the scope of the ISI Program.
j II. Code Requirement:
IWA-7400 (a) (4) exempts NPS 1 inch and smaller replacement items and their installation from i the requirements ofIWA-7000; whereas, the repairs ofitems 1-inch NPS and smaller are not similarly exempted from the requirements ofIWA-4000.
j III. Code Requirement for Which ReliefIs Requested:
- Reliefis requested from the requirements ofIWA-4'000 and IWA-7000 on piping, valves, and fittings 1 inch NPS and smaller, j EV. liasis for Relief
ASME Code Case N-544 was issued on August 24,1995 and provides an alternative to the
(
requirements ofIWA-4000 and IWA-7000 when a repair or replacement is performed on piping, l valves and fittings 1-inch NPS and smaller. A copy of the code case is provided as Attachment I to this request for relief.
V. Alternate to the Code Requirements:
Southern Nuclear Operating Company will coraply with the requirements of Code Case N-544 in lieu ofIWA-4000 and IWA-7000. -
VI. Justification for the Granting of Relief:
The 1989 Edition of ASME Section XI Code provides an exemption for replacement items 1-inch NPS and smaller from the requirements ofIWA-7000, but repairs to such items cre not similarly exempted. Therefore, a repair to an item is subject to more restrictive requirements '
than replacing the item. Code Case N-544 allows application of the alternative requirements for replacement to all repair and replacement activities. The ASME Code Committee evaluated the proposed alternatives contained in the code case and determined that they are acceptable for repair and repl acement activities on piping, valves, and fittings 1-inch NPS and smaller. In addition the provisions of this code case were added to IWA-4131 and IWA-4132 in the 1995 Addenda to ASME Section XI. SNC has determined that implementation of the code case will !
not affect the level of quality and safety, nor decrease the margin of public health and safety.
O ]
1 7-90 Rev. O !
4
j i
SOUTHERN NUCLEAR OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL I REOUEST FOR RELIEF NO. RR-22 (coniinued)
VI. Justification for the Grantine of Relief (continued):
While the cost savings associated with Code Case N-544 have not been quantified as a Cost l Beneficial Licensing Action item, its implementation is consistent with the intent to eliminate i
, nonbeneficial work activities and their associated costs. Therefore, it is requested that the
- proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedults This relief for request is applicable to the Second Ten-Year Interval which commences May 31,
. 1997.
s 2
j p.
V I
I i
i A
L) 7-91 Rev. 0
)
i SOUTTIERN NUCLEAR OPERATING COMPANY s VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 !
SECOND TE_N-YEAR INTERVAL i
/ REOUEST FOR RELIEF NO. RR-22 l
! U] -
(continued) !
i ATTACEMENT 1 l
CASE N-544 CASES OF 45%tt BOILER AND PRE 55LRE VE55EI. CODE Approval Date: Aveust Je.1995 See Numerreal laces for esperation ano any resNormaroon antes. 1 l
Case N.544 reactor coolant makeup systems operable from onsite Repair / Replacement of Small Items emergency power.
Section XI. Division I 1400 Class 2 and 3 pipmg. tubing (except heat eschanger tubing. and sleeves and welded plugs used Inquerv When performing repair or replacement of f r heat exchanger tubing). val +es. and Asungs. NPS piping. valves, and 6ttmgs. NPS I and smaller, what I and smaller and, associated supports.
=
alternative requirements may be used in lieu of [WA.
4000 for repair and [WA.7000, for replacement con.
2000 ACTIVITIES PERFORSIED TO ducted in accordance with Section XI Editions and SUN!alER 1978 ADDENDA THROUGH Adder.da through the 1990 Addenda, or IWA-4000 for 1990 ADDENDA OF SECTION XI repab and replacement conducted in accordance etth A the 1990 Addenda and later? 2100 For the items idenuned in .1300 and .1400 g gg , , g, ,,
l Reptvt la is the opinion of the Committee that repair pair and replacement requirements of IWA 4000 and and replacement of piping. valves, and 6tungs. NPS IWA.7000' I and smaller may be performed in accordance with 2200 Items to be used for replacement shall be the following requirements:
constructed in accordance with IWA-7210. except that possession of a Ceruncate of Authorization and an agreement wuh an Authorized Inspection Agency are 1000 GENERAL REPA!R AND neither required not prohibned.
REPLACES!ENT REQUIRENIENTS .
2300 items to be used for repair and replacement 1100 Repair and replacement acuvines performed in shall be obtained. installed, and documented in accord.
accordance with Section XI Ediuon and Addenda from ance wnh the provisions of IWA 1400(n) and the the 1977 EJiuon with the Summer 1978 Addends technical requirements of IWA.7210. For Section !!!
through the 1959 Ediuon with the 1990 Addends on items. the requirements of NA.3700/NCA.3800 need items idenuhed in .1300 and -1400 shall satisfy the not be met, provided the Owner's Quality Assurance requirements of .2000. Program provides measures to assure that matenal used I
1200 Repair and replacement activines performed in I ' " *** *"' I"
- 'k " *## # *"#' Y 8pPlictele matenal specificanon and matenal require.
accordance with Section XI Editions and Addenda from l the 1989 Edinon with the 1991 Addenda through the ments of Section Ill. A Repair / Replacement Plan, pres.
1992 Edinon with the 1994 Addenda on items identi6ed sure testirig, services of an Authorized Inspection in .1300 and .1400 shall satisfy the requirements Agency. an C mpleuon d W bnns m m nW d .3000. t ' "''li " t "h'** 5"****
t 1300 Class I piping. tubing (except heat exchanger CP8 5 2H pehmed and huntd in subing, and sleeves and welded plugs used for heat **""*"***" ^ "' ^
l etchanger tubing). valves, ntungs, and associated sup.
"" ^ # '"
l pons, no larger than the smaller of (a) or (b) below: in ace rdance with .2300. 'A Repair / Replacement Plan.
l W NPS h pressure tesung, services of an Authon:ed Inspection (b) A size and design such that, in the event of I'"#I' "" *""b'" "
postulated failure dunng normal plant operaung condi. 9"#
tions, the reactor can be shut down and cooled in an 2500 If an item to be repaired or replaced does not orderly manner, assuming makeup is provided by normal sausfy the requiremems of Secuon XI the Owner shall i
7-92 Rev.O i
l l
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELEC1 MIC GENERATING PIANT. UNITS 1 AND 2
/7 SECOND TEN-YEAR INTERVAL U REOUEST FOR RELIEF NO. RR-22
~
(continued)
ATTACHMENT 1 (continued)
CASE (continued N-544 CASES OF 4 SALE BOILER OD PRESStlRE VESSEL CODE determine de cau>e of unxceptabilitv and evaluate the NCA 3300 need not be met, proudeJ the Omner's sunability of the repair or replacement pnor to retumint the item to seruce. If the speci6cauon for the onginal Quahty Assurance Program proudes measures to assure item is determmed to be de6cient, appropnate correcuve that matenal used for replacement is furmshed in accord-provisions shall be included in the specincauon for the ance with the apphcable matenal speci6 cation and >
repair or the replacement item. rnatenal requirements of Secuon Ill. A Repair / Replace.
ment Plan. pressure testing. services of an Authonzed 2600 For sabes whose performance parameters could I" " " " 3'"* " # ""**
be affected by the repair or replacement activity, a n t required for ?' ""d #"E"
installauon of these items.
prescrCice test shall be made in accordance with Subsec. 3400 rion IWV pnor to returmng the sabe to service. Repair shall be performed and documented m accordance mich IWA 1400(n). IWA 4I20. IWA 4200 2700 For purnps ahose performance parameters could " * # #
be affected by the repair or replacement activity, a and later. Weld 611er matenal shall be obtamed in'"Y" new set of reference values shall be determined in '"#***"" ^ 'E 'E'*****"* ""*
accordance with Subsection IWP dunng the 6rst inser. pm13um tuh4 58 CH an Au 2e "Spech q ~
vice test performed after the pump is put into service. ^8ency. And e mpletion of fi!S-2 forms are not re-quired.
2800 Use of these alternative requirements, including specifying the size of Class I items to which these 3500 If an item to be repaired or replaced does not requirements will be appbed, shall be documented by satisfy the requirements of Secuon XI. the Owner shall the Owner and is subject to review by the inspector. determme the cause of unacceptability and evaluate the
,,g the item to rervice, if the speci6 canon for the ongmal 3000 ACTIVITIES PERFORMED TO 1991 item is determmed to be de6eient. appropnate correcove ADDENDA THROI'GH 1994 ADDENDA OF SECTION XI provisions shall be included in the speci6 cation for the repair or the replacement item.
3100 For the items identined in .1300 and .1400. 3600 For wahes whose performance parameters could the following requirements are an alternanve to the repair and replacement requirements of IWA-4000. be affected by the repair or replacement activity, a preservice test shall be made in accordance with Subsec.
3200 items to be used for replacement shall be "" E"" " * *
"I constructed in accordance with IWA.4170. escept that 3700 possession of a Ceni6cate of Authorttauon and an For pumps whose performance parameters could agreement with an Authonzed Inspecuen Agency att be affected by the repair or replacement acuvity. a neither required nor prohibited. new set of reference values shall be determined in accordance with Subsection IWP dunng the 6rst inser-3300 trems to be used for repair and replacement vice test perf rmed after the pump is put into service.
shall be obtained. mstalled, and documented in accord- 3800 ance with the provisions of IWA 1400(n), and IWA. Use of these alternative requirements. including 4132 for the 1993 Addenda and later, and IWA-4170.
specifying the size of Class I items to which these For Secuon Ill ne: the regstrements of NA 37001 requirernents will be applied. shall be documented by the Owner and is subject to review by the Inspector, p
t I
7-93 Rev.0
i 1
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL a REOUEST FOR RELIEF NO. RR-23
- 1. System /Componenus) for Which Reliefis Reguntrd:
Class 1,2, and 3 systems subject to hydrostatic testing.
II. Code Requirement:
Tables IWB-2500-1, IWC-2500-1, and IWD-2500-1 of the 1989 Edition of ASME Section XI require system hydrostatic and leakage testing as shown below. The Code requires system hydrostatic testing once per ten-year interval at or near the end of the interval.
ASME Examination Category B-E, Items.B4.11, B4.12, B4.13, and B4.20, l 1
ASME Examination Category B-P, Items B15.11, B15.21, B15.31, B15.51, B15.61, and B15.71, -
i ASME Examination Category C-H, Items C7.20, C7.40, C7.60, and C7.80,
]
i ASME Examination Category D-A, Item Dl.10, I s ASME Examination Category D-B, Item D2.10, and ASME Examination Category D-C, Item D3.10.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing the Code-required hydrostatic tests. Alternative examinations are proposed.
IV. rsasis for Relief:
ASME Section XI Code Case N-498-1 which is provided as Attachment I to this request for l relief was issued on May 11,1994. This code case has not been formally endorsed for inclusion in NRC Regulatory Guide 1.147 but has been authorized by the NRC for use previously at VEGP and other plants. ASME Section XI Code Case N-498-1 was approved by the NRC for use at VEGP during the First Ten-Year Interval as addressed in NRC correspondence dated March 8, 1996 and August 13,1996 for VEGP-1 and 2, respe.ctively. (
Reference:
First Ten-Year Interval Requests for Relief No. RR-60 for VEGP-1 and 2.)
O' 7-94 Rev.0
SQUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
[ REOUFST FOR RELIEF NO. RR-23 (continued)
V. Alternative Examination: ,
Southem Nuclear Operating Company proposes to perform an alternative examination delineated in Code Case N-498-1 as an option to performing Code-required hydrostatic tests. Code Case N-498-1 requires that a VT-2 visual examination be performed in conjunction with a system pressure test at nominal operating pressure.
VI. Jrtification for Granting Relief:
i l
Southern Muclear Operating Company has determined that hydrostatic tests represent a hardship with little benefit. Hardships are generally encountered with the performance of hydrostatic testing performed in accordance with the Code. For example, since hydrostatic test pressure would be higher than nominal opecating pressure, hydrostatic pressure testing frequently requires i significant efTort to set up and pe: form. The need to use special equipment and the need for individual valve lineups can cause the testing to impact maintenance / refueling outage schedules.
Piping components are designed for a number ofloadings postulated to occur under the various
- e modes of plant operation. ASME Section XI hydrostatic testing only subject the piping components to a small increase in pressure over the design pressure and, therefore, does not present a significant change to pressure boundary conditions. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leakage detection during the examination of j components under pressure rather than as a measure to determine the structural integrity of the components.
The ASME Working Group on Pressure Testing concluded that no additional benefit is gained 4
by conducting the existing system hydrostatic tests in place of the alternate rules which require a leak test at nominal operating pressure. The conclusion of the group was that ASME Section XI hydrostatic testing does not verify structural integrity and, in fact, the slightly higher test pressure currently called for in the Code could result in operational difficulties as well as extended maintenance / refueling outages and increased costs. ' Industry experience has demonstrated that leaks are not discovered as a result of hydrostatic test pressures propagating a preexisting flaw
- through wall. This experience indicates that leaks in most cases are being found when the system is at normal operating pressure. This is largely due to the fact that hydrostatic pressure
- testing is infrequently performed while system leakage tests at nominal operating pressures are conducted a minimum of once each refueling outage for Class I systems, and each 40-month inspection period for Class 2 and 3 systems. In addition, leaks may be identified during system walkdowns by plant operators.
O 7-95 Rev.0
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL l REOUEST FOR RELIEF NO. RR-23 v -
(continued)
VI. Justification for Granting Relief (continued):
The use of Code Case N-498," Alternative Rules for 10-Year System Hydrostatic Testing for Class 1 and 2 Systems", was previously approved by the NRC in Regulatory Guide 1.147, Revision 11. The alternative rules for Code Class 1 and 2 in Code Case N-498-1 are unchanged from N-498. Code Case N-498-1 added an alternative to the 10-year system hydrostatic tests required for Class 3 Systems by Table IWD-2500-1, Categories D-A, D-B, or D-C to the Class 1 and 2 alternatives included in Code Case N-498. Code Case N-498 was found to be acceptable because the alternative provided adequate assurance and because compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality or safety.
Southern Nuclear Operating Company has determined that the alternative rules of ASME Code Case N-498-1 provide reasonable assurance of the structural integrity of the Code system.
Consequently, an acceptable level of quality and safety will be achieved and public health and safety will be maintained by allowing the proposed alternative examination as an option to the Code requirements. Therefore,it is requested that the proposed alternative examinations be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, 1997.
O V
7-96 Rev.O
FOUTHERN NUCLEAR OPERATING COMPANY !
VOGTLE ELEUIKIC GENERATING PLANT. UNITS 1 AND 2 ,
A SECOND TEN-YEAR INTERVAL '
l U REOUEST FOR RELIEF NO RR-23
~
I (continued)
ATTACHMENT 1 cas N-498-1 i C ASES OF ASME aOILER AND PaESSt'RE VI3SEL CODE Approval oste: May 11.1994 See NumericalIndex for esDiration and any reaffirmation dates.
\
1 Case N-4981 (2) The boundary subject to test pressunzation Alternative Rules for 10-Year System liydrostatic during the system pressure test shall extend to all Testing for Class 1,2, and 3 SystemsSection XI, Division 1 Class 2 components included in those portions of sys-
)
tems required to operate or support the safety system function up to and including the first normally closed Inquuy What alternative rules may be used in lieu valve. including a safety or relief valve, or valve ca- i of those required by Section XI. Division 1. Table j pable of automatic closure when the safety function i IWB-2500-1. Category B P. Table IWC 25001, Cat- is required.
egory C-H. and Table IWD 2500-1, Categones D-A. (J) Prior to performing the VT 2 visual exami-D B. and D-C. as applicable, for the 10-year system j hydrostaue test? nation, the system shall be pressudzed to nominal '
operatmg pressure for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for in.
sulated systems and 10 minutes for noninsulated sys.
tems. The system shall be maintained at nominal op-ersting pressure during performance of the VT-2 Reply:
visual examination.
c (a) It is the opmion of the Committee that as an !
( (4) The VT 2 visual examination shall include l alternative to the 10-year system hydrostatic test re.
C quired by Table IWB-2500-1, Category B-P, the fol.
all components within the boundary identified in j (b)(2) above.
lowing rules shall be used.
(5) Test instrutnentation requirements of IWA-(1) A system leakage test (IWB-5221) shall be 5260 are not applicable-conducted at or near the end of each inspection in. (c) It is the opinion of the Committee that, as an terval, pnot to reactor startup. alternative to the 10-year system hydrostatic test re-l (2) The boundary subject to test pressurization quired by Table IWD-2500-1, Categones D A. D-B, during the system leakage test shall extend to all i or D-C (D-B for the 1989 Edition with the 1991 and Class 1 pressure retaining components within the sys. subsequent Addenda), as applicable, tne following tem boundary. ,
rules shall be used. !
(J) Pr.d to perfr3 ting the VT 2 visual exami. ,
(1) A system pressute test shall be conducted at nation, the system shC be pressurized to nominal or near the end of each inspection interval or during operstmg pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated the same inspection period of each inspection inter-systems and 10 minutes for noninsulated systems, val of Inspection Program B.
The system shall be maintained at nominal operating (2) The boundary subject to test pressurization pressure dunng performance of the VT 2 visual ca. during the system preswre test shall extend to all amination. j Class 3 components included in those portions of sys-(4) Test temperatures and pressures shall not tems required to operate or support the safety system j
exceed limiting conditions for the hydrostatic test function up to and including the first normally closed curve as contained in the plant Technical Specifica. valve, including a safety or relief valve, or valve ca-tions.
pable of automatic closure when the safety function (5) The VT 2 visual examination shall include is required.
all components within the boundary identified in (J) Prior to performing the VT 2 visual exami.
(a)(2) above.
nation. the system shall be pressunzed to nominal (6) Test instrumentation requirements of IWA-) operstmg pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated 5260 are not applicable. ;
systems and 10 minutes for noninsulated systems.
(b) It is the opmion of the Committee that, as an The system shall be matntained at nominal operating i
alternative to the 10-year system hydrostatic test re. pressure dunng performance of the VT-2 visual ex-quared by Table IWC-25001. Category C H, the fol- ammation.
lowing rules shall be used.
NJ The VT 2 visual examination shall include fl) A svstem pressure test sha!! be conducted at all components within the boundary identified in !
f or near the end of each inspection mterval or dunng (c)(2) above.
(- the same inspection period of each inspection inter-val of Inspection Program B.
(5) Test mstrumentation requirements of IWA-5260 are not apphcable.
l l
7-97 Rev.0 ,
l i
SOUTIIERN NUCLEAR OPERATING COMPANY ;
) VOGTLE ELECTRIC GENEliATING PLANT. UNITS 1 AND 2 j SECOND TEN-YEAR INTERVAL '
REOUEST FOR RELIFF NO. RR-24 I
- 1. System / Component (s) for Which Reliefis Requested:
> I ASME Class 1,2, and 3 piping and components. l II. Code Requirement:
Paragraph IWA-4700(a) of the 1989 Edition of ASME Section XI requires that a system hydrostatic test be performed in accordance with IWA-5000 after a welded repair on a pressure-l retaining boundary.
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from performing this Code-required post-repair / replacement hydrostatic pressure test on Class 1,2, and 3 welds. Alternative examinations are proposed.
IV. Basis for Relief:
1
- ASME Section XI Code Case N-416-1 which is provided as Attachment I to this request for )
relief was issued on February 15,1994. This code case has not been formally endorsed by inclusion in NRC Regulatory Guide 1.147 but has been authorized by the NRC for use previously at VEGP and other plants. ASME Section XI Code Case N-416-1 was approved by the NRC for use at VEGP during the First Ten-Year Interval as addressed in NRC correspondence dated March 8,1996 and August 13,1996 for VEGP-1 and 2, respectively. ;
(
Reference:
First Ten-Year Interval Requests for Relief No. RR-59 for VEGP-1 and 2.)
V. Alternate Examination:
Southern Nuclear Operating Company proposes to perform alternative examinations delineated 1 in ASME Code Case N-416-1, with augmented exams for Class 3 piping and components, in lieu cf Code-required hydrostatic tests. These alternative examinations are as follows:
- 1. Perform nondestructive examinations in accordance with the methods and acceptance criteria of the applicable subsection of the 1992 Edition of ASME, Section 111.
1
- 2. Perform a VT-2 visur! examination of the welds in conjunction with the system leakage test in accordance with IWA-5000 of the 1992 Edition of ASME Section XI.
- 3. Perform surface examinations on the root pass layer of butt and socket welds on the pressure-retaining boundary of Class 3 piping and components.
O 1 7-98 Rev. 0
1 SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE El FCTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-24 (continued)
V. Alter inte Ernmination (continued):
- 4. The nondestructive examinations and pressure tests shall be documented on an Owner's Report for Repairs or Replacements, Form NIS-2.
VI. Justification for Granting Relief:
I Southern Nuclear Operating Company has determined that hydrostatically testing post-repair / installation welds represents a hardship with little benefit. Hardships are generall,,
encountered with the performance of hydrostatic testing performed in accordance with the Code.
For example, since hydrostatic test pressure would be higher than nominal operating pressure, '
hydrostatic pressure testing frequently requires significant effort to set up and perform. The need to use special equipment and the need for individual valve lineups can cause the testing to impact maintenance / refueling outage schedules.
Piping components are designed for a number ofloadings that would be postulated to occur under the various modes of plant operation. ASME Section XI hydrostatic testing only subjects the piping components to a small increase in pressure over the design pressure and, therefore, does not present a significant change to pressure boundary conditions. Accordingly, hydrostatic pressure testing is primarily regarded as a means to enhance leakage detection during the examination of components under pressure, rather than solely as a measure to determine the structural integrity of the components.
The ASME Subcommittee Working Group on Pressure Testing concluded that no additional benefit is gained by conducting the existing system hydrostatic tests in place of the attemate rules which require a leak test at nominal operating pressure. The conclusion of the group was that hydrostatic testing does not necessarily verify structural integrity and, in fact, the slightly higher test pressures currently called for in the Code could result in operational difficulties as well as extended outages and increased costs.
Industry experience has demonstrated that leaks are not discovered as a result of hydrostatic test pressures propagating a pre-existing flaw through-wall. Thie experience indicates that leaks in most cases are being found when the system is at normal operating pressure. This is mainly due to the fact that hydrostatic pressure testing is infrequently performed, while system leakage tests at normal operating pressures are conducted a minimum of once each maintenance / refueling outage for Class 1 systems, and each 40-month inspection period for Class 2 and 3 systems. In addition, leaks may be identified during system walkdowns by plant operators.
Q V
Southern Nuclear Operating Company has determined that the nondestructive examinations and their associated acceptance criteria provide assurance of the structural integrity of the weld. The i
7-99 Rev.0
SOUTilERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN. YEAR INTERVAL REQUEST FOR RELIEF NO. RR-24 (continued)
VI. Justification for Granting Relief (continued):
proposed alternative examinations will provide reasonable assurance that unallowable flaws are not present in the subject welds. Consequently, an acceptable level of quality and safety will be
- achieved and public health and safety will not be endangered by allowing the proposed 1 alternative examination in lieu of the Code requirement. Therefore, it is requested that the proposed alternative examinations be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).
VII. Implementation Schedule:
j This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, 1997.
4 i
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e i
J O
V 7-100 Rev.O
.. ~ - . - . . _- - - . - . _ . . . - . _
SOUTHERN NUCLEAR OPERATING COMPANY I
_VOGTLE ELECI RIC GENERATING PLANT. UNITS 1 AND 2
[, SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-24
~
(continued)
- ATTACHMENT 1
~
l
. CASE I N-416-1 CASES OF ASME BOtLER AND PRESSL RE VESSELCODE 1
Approval Date: Febewarv 15.1994 s, -,m.. ,., .. ,. ,.n OAdany reaffantetsen 68ttS.
Case N-416-1 placement items by welding, a system leakage test may Alternative Pressure Test Requirement for Welded !
be used provided the following requirements are met. '
Repairs or installation of Replacement Iterns by (a) NDE shall be performed in accordance wuh the Welding. Class I,2 and 3 methods and acceptance entena of the appheable Sub-4 Section XI. Division 1 section of the 1992 Edition of Section Ill.
(b) Pnor to or immediately upon teturn to service, a visual esamination (VT-2) shall be performed in con-Ingwry: What ahernative pressure test may be per-junction with a system leakage test, using the 1992 Edi-
[ formed in lieu of the hydrostatic pressure test requued tion of Section XI. in accordance with para. IWA-5000
-(/ by paru. IWA-4000 for welded repairs or installation of at nominal operating pressure and temperature, replacement items by welding? --
(c) Use of this Case shall be documented on an NIS-2 Form.
! Reply: It is the opinion of the Commmee that in lieu If the previous version of this case were used to defer of performmg the hydrostatic pressure test required by a Class 2 hydrostatic test, the defened test may be elim-para. IWA-4000 for welded repairs or installation of re- inated when the requirements of this revision are met.
4 i
i l
/
G 7-101 Rev.0 .
SOUTIIERN NUCLEAR OPERATING COMPANY 4
VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL O
V REOUEST FOR RELIEF NO. RR-25 i I. System / Component (s) for Which Reliefis Requestal:
This request for relief provides alternative corrective actions that may be used in lieu of the corrective actions associated with leakage at Class 1,2, and 3 bolted connections as prescribed by IWA-5250, " Corrective Measures".
II. Code Reauirement:
Subparagraph IWA-5250(a)(2) of the 1989 Edition of ASME Section XI states "ifleakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100".
i III, Code Requirement from Which Reliefis Requested:
i i Reliefis requested from being required to remove bolting and performing VT-3 examinations
- and as an alternative it is requested that ASME Section XI Code Case N-566 be allowed to be used. A copy of the code case is provided as Attachment I to this request for relief.
IV. Basis for Relief:
Some of the problems associated with the current requirements ofIWA-5250(a)(2) are summarized as follows:
. i
- 1. IWA-3100 does not provide an acceptance standard for a VT-3 bolt examination, !
- 2. The requirement calls for bolt removal without regard to the size of the leakage,
- 3. The requirement increases the radiological dose to workers for leaks that are often not a challenge to operational nor structural limits,
- 4. Bolts sometimes cannot be removed without damaging the bolt or cannot be removed due to the component configuration,
- 5. It is not a requirement of the Code that the owner must stop the leakage and inspection of the bolting is not necessarily going to stop the leak,
- 6. Removing one bolt at a time, if allowed by system conditions, may actually increase the leakage, and
- 7. In many cases, implementation of the requirement would cause the plant an unnecessary transient or delay startup,
. 1 I
The 1983 Edition through Summer 1983 Addenda of ASME Section XI was applicable for the First Ten-Year Interval at VEGP. IWA-5250 of the 1983 Edition through Summer 1983 l Addenda of ASME Section XI did not require bolting removal, VT-3 visual examination for l corrosion, and evaluation in accordance with IWA-3100. Therefore, this request for relief was i not needed at VEGP in the First Ten-Year Interval.
7-102 Rev.O j l
i SOUTHERN NUCLEAR OPERATING COMPANY y_OGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 l s SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. HR-25 (continued) l V. Alternate Examination:
Southern Nuclear Operating Company will comply with the requirements of Code Case N-566 in lieu ofIWA-5250(a)(2).
VI. Justification for Granting Relief:
The ASME Code Committee evaluated the proposed alternatives contained in Code Case N-566 and determined that they are acceptable for corrective action for leakage at bolted connections.
Code Case N-566 resolves the implementation problems associated with IWA-5250(a)(2) and allows two alternatives as corrective actions for leaks at bolted connections. One option is to stop the leak and review thejoint for integrity. This review will consist of cleaning thejoint after 1
the leakage is stopped and documenting an inspection of thejoint for corrosion or other signs of degradation. The second option is that, if the leakage is not stopped, thejoint must be evaluated ,
in accordance with IWB-3142.4 of ASME Section XI. This is an engineering evaluation that i
must consider the number and condition of bolts, the leakage medium, the bolt and component material, the system function, and the ability to monitor leakage.
O D The implementation of Code Case N-566 will not affect the level of quality and safety, nor decrease the margin of public health and safety. Therefore, it is requested that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, 1997.
~
O '
7-103 Rev.0
_ _ _ _ . _____ _ _-_. _ ~_ _ . _ _ _ . _ _ _ . _ _ . _ _ _ _ _
SOUTHERN NUCLEAR OPERATING COMPANY ,
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 I
} SJKQ',fD TEN-YEAR INTERVAL 1
g 3,00 UKS.T FOR RELIEF NO. RR-25 V -
(continued) 1 l
i 1 ATTACHMENT 1 1
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CASE N-566 Aspresal Dess Assesst & W i )
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- 5.cdea XI, Divumes 1 i
1 % W bst . J m to the zugnimassa of ,,
!WQ-322XaX2) may be used when laakass is dammad l at baked connections?
i it ply: It is ths opemas d the f*W ther, as
! an altanadve to abs regausmans of IWA-525000CZ),
3 nos of abs fbBounsg meanmanom shnH be est for -
4
- 1mahnge at beland - - 8 am (s) The kahmes shun be sempped. and she helmag sad maascal shun be noviewed for jaiar
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7-104 Rev.0-
.m SOUTIIERN NUCLEAR OPERATING COMPANY VOCTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
/ REOUEST FOR RELIEF NO. RR-26
(
I. System / Component (s) for Which Reliefis Requested:
Reliefis requested from Subparagraph IWA-5242(a) of the 1989 Edition of ASME Section XI,
- which requires removal ofinsulation from pressure-retaining bolted connections during system pressure testing, when systems are at the pressure and temperature requirements ofIWA-5000, IWB-5000, IWC-5000, and IWD-5000. This includes the following systems which are borated 4 for the purposes of reactivity control
- Reactor Coolant System, e Chemical and Volume Control System, .
- Residual Heat Removal System, e Safety Injection System, and
- Nuclear Sampling System - Liquid.
II. Cadt Requirement:
Subparagraph IWA-5242(a) of the 1989 Edition of ASME Section XI states in part "For systems p borated for the purposes of controlling reactivity, insulation shall be removed from pressure i
Q retaining bolted connections for visual examination VT-2."
III. Code Requirement from Which Reliefis Requested:
Reliefis requested from removing insulation from pressure-retaining bolted connections and performing VT-2 visual examinations for the purpose of detecting boric acid residue when systems are at the pressure and temperature requirements ofIWA-5000, IWB-5000, IWC-5000, and IWD-5000.
IV. Basis for Relief:
Subparagraph IWA-5242(a) specifies that insulation must be removed from pressure-retaining bolted connections for VT-2 visual examination during the performance of system pressure testing. This is applicable to the following systems:
1
- Reactor Coolant System (System consists of Class 1 and 2 components),
e Chemical and Volume Control System (System. consists of Class 1,2, and 3 components),
e Residual Heat Removal System (System consists of Class 1 and 2 components),
e Safety Injection System (System consists of Class 1 and 2 components), and
. Nuclear Sampling System - Liquid (System consists of Class 2 components).
7-105 Rev.O
I SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL i
,O REOUEST FOR RELIEF NO. RR-26 i p 1 (continued) l IV. Basis for Relief (continued): l Class 1 comnonents: l Table IWB-2500-1, Examination Category B-P requires a system leakage test (IWB-5221) and corresponding VT-2 visual examination on Class I components each refueling outage prior to plant startup. This system leakage test is performed in Mode 3 when the Reactor Coolant System is at Nominal Operating Pressure (= 2235 psig) and Nominal Operating Temperature (= 550 F to s
650 F). The majority of the Class I components are in the Reactor Coolant System however i
some portions of the Class i boundary extend to include portions of Safety Injection, Chemical and Volume Control, and Residual Heat Removal systems. All Class I components are in
! containment. The removal and installation ofinsulation during the performance of system pressure testing inside containment presents the following hazards:
1 l e Increased potential for personnel heat stress since the containment ambient temperature is =
100 F, I e Increased potential for personnel burn injuries due to installation ofinsulation in proximity of extremely hot components, 2 e e Increased personnel safety hazard since ladders would have to be used to inspect many of the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4,
- Increased radiation exposure to personnel since temporary shielding is removed prior to l entering Mode 4, l
- Increaaed potential for debris to be in containment which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during
- accident (LOCA) conditions. All debris is required to be removed from containment prior )
to entering Mode 4, and l j e Increased potential for impacting outage duration due to the amount of manpower required ;
to support insulation removal and examinations during Mode 3 following refueling outage activities.
Class 2 components:
Table IWC-2500-1, Examination Category C-li requires a system pressure test (IWC-5221)
_ during system functional anhystem inservice tests and corresponding VT-2 visually 7-106 Rev. 0
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 4
SECOND TEN-YEAR INTERVAL O
V REOUEST FOR RELIEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 2 components (continued):
examination on Class 2 components once each inspection period (40 months). The following discusses the applicable systems and the basis for relief for each Class 2 system:
- 1. Basis for Relief for the Reactor Coolant System (RCS):
The Class 2 portions of RCS are located adjacent to the Class 1 boundary and are classified as Class 2 based on line size and isolation valve criteria. The system inservice tests for i
these Class 2 portions of RCS are VT-2 examined in Modes 1,2, and 3 and therefore the same basis for relief as provided for Class 1 applies. These Class 2 pressure boundaries are located in containment.
- 2. Basis for Relief for the Chemical and Volume Control System (CVCS):
3 Q
V For those portions of CVCS which are located inside containment (Charging, Letdown, Excess Letdown, Alternate Pressurizer Spray, Reactor Coolant Pump Seal Leakoff, etc.) the same basis for relief as provided for Class I above applies except that the system operating i temperatures are less iesulting in less potential for bum-related injuries. The VT-2 examinations are performed in Modes 1,2, and 3.
For those portions of CVCS which are located outside containment, radiation levels, high component temperatures, and availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized. CVCS is inservice during power operation and, as such, many, if not all,
, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, as previously addressed, conditions
- may be present which may not allow insulation removal except during refueling outages.
- 3. Basis for Relief for the Residual Heat Removal System (RHR)
RHR is placed inservice during a shutdown prior to refueling activities in Mode 4 when the Reactor Coolant System is = 350 F and = 350 psig. RHR remains inservice in Modes 5 and 6 during the refheling outage and remains inservice in Mode 4 during startup following the refueling outage. The VT-2 examinations are performed in either Mode 4 or Mode 5 when the Reactor Coolant System is = 350 psig.
7-107 Rev.O
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
(
V) REOUEST FOR REl:IEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 2 components (continued):
- 3. Basis for Relief for the Residual Heat Removal System (RHR) (continued):
For those portions of RHR which are located inside containment the same basis for relief as provided for Class I above applies except that the system operating temperatures are less resulting in less potential for burn-related injuries. It is impractical to attempt to limit VT-2 examinations to Mode 5 in order to avoid the complications of performing RHR pressure test in Mode 4. The VT-2 examinations are performed in Modes 4 or 5.
For those portions of RHR which are located outside containment radiation levels, high component temperatures, availability of personnel, and increased thermal loads on chilled water room cooling systems may preclude removing insulation while the pressure-retaining bolted connections are pressurized. It is significantly more prudent to uninsulate, VT-2 examine, and reinsulate the pressure retaining bolted connections in RHR when the system
('
\
is not pressurized during non-outage times or during refueling outages when the system is not at the required pressure.
- 4. Basis for Relief for the Safety Injection System (SI):
The system pressure tests performed on SI are either system functional tests or system inservice tests as follows:
Some of the system functional tests are performed during Modes 1,2, and 3 when Reactor Coolant System pressure is greater than SI pump discharge pressurc. VT-2 examinations are performed on portions of SI during various activities and tests which require a SI pump to be in operation. The scope of these VT-2 examinations includes components both inside and outside containment. The performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the SI pumps.
~
The remainder of the system functional tests are performed during Mode 6 and defueled conditions with the reactor vessel head removed. VT-2 examinations are performed on portions of SI which are pressurized during check valve flow testing activities which involve injection of water into the reactor pressure vessel. The scope of these VT-2 examinations includes components both inside and outside containment. The performance o of the VT-2 examinations during these activities is generally performed in less than one Q)
- 7-108 Rev. 0
SOUTIIERN NUCLEAR
- OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-26 (continued)
]
IV. Basis for Relief! continued):
Class 2 component A niinued):
- 4. Basis for Relief for the Safety Injection System (SI) (continued):
hour to minimize run time on the applicable pumps and to minimize the impact on critical path testing during refueling outages.
4 Some of the system inservice tests are performed during Modes 1,2, or 3 on portions of SI
, which are pressurized by the SI accumulator tanks. The SI accumulator tanks are generally depressurized during refueling outages. T,he scope of these VT-2 examinations includes components which are only inside containment.
The remainder of the system inservice tests are performed on portions of SI which are
- pressurized by the static head of the refueling water storage tank. The VT-2 examinations I on these portions of SI are generally performed during power operation (Mode 1) but may ;
O be performed in other Modes if tank levels are adequate. The scope of these VT-2 l examinations includes components which are only outside containment.
For those portions of SI which are located outside containment radiation levels and I availability of personnel during non-outage times or during system functional testing may l preclude removing insulation while the pressure-retaining bolted connections are pressurized.
For those portions of SI which are located inside containment and VT-2 examined during Modes 1,2, and 3 the same basis for relief as provided for Class 1 above applies except that
, the system operating temperatures are less resulting in less potential for bum-related l injuries-Fr r those portions of SI which are located.inside containment and VT-2 examined during Mode 6 and defueled conditions with the reactor vessel head removed containment radiation levels and availability of personnel during system functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
- 5. Basis for Relief for the Nuclear Sampling System - Liquid:
The liquid portions of the Nuclear Sampling System are used for providing samples for p analysis purposes of the Reactor Coolant System, Chemical and Volume Control System, d and Residual Heat Removal System. This system is located both inside and outside l
7-109 Rev.O
SOUTIIERN NUCLEAR OPERATING COMPANY
- VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN. YEAR INTERVAL 3
, [d REOUEST FOR RELIEF NO. RR-26 (continued l
IV. Basis for Relief (continued):
I Class 2 components (continued):
- 5. Basis for Relief for the Nuclear Sampling System - Liquid (continued):
1 containment and is subject to the same system pressure tests as the systems for which it
{ used to provide samples. Therefore, the same basis for relief as discussed above is l
applicable to the liquid portions of the Nuclear Sampling System.
l
- Class 3 components
i
, Table IWD-2500-1, Examination Category D-A requires a system inservice test (IWD-5221) and corresponding VT-2 visually examination on Class 3 components once each inspection period l (40 months). Subparagraph IWA-5242(a)is applicable to the boric acid storage tank and boric acid transfer portions of the Chemical and Volume Control System. System inservice test are
- performed as follows
, n(.) Some of the system inservice tests are performed on portions of CVCS which are pressurized by l
the static head of the boric acid storage tank. The scope of these VT-2 examinations includes I components which are only outside containment. The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. These systedi inservice tests are generally performed during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
The remainder of the system inservice tests are performed on portions of CVCS which are pressurized when a boric acid transfer pump is operating. The scope of these VT-2 examinations includes components which are only outside containment. The VT-2 examinations on these portions of CVCS are generally performed during power operation (Mode 1) with a boric acid transfer pump running with system valves aligned in a recirculation flowpath which precludes injecting high concentrations of borated water into CVCS and ultimately into the RCS. The boric acid transfer pumps are operated as necessary to perform system functions and necessary testing and, as such, are not continuously in operation. Since these pumps are not continuously in operation availability of personnel during non-outage times may preclude removing insulation l while the pressure-retaining bolted connections are pressurized.
l k l 1
7-110 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 ;
i SECOND TEN-YEAR INTERVAL i
'O REOUEST FOR RELIEF NO. RR-26 O {
(continued) j 1
J IV. Basis for Relief (continued):
Class 3 comnoncats (continued):
The 1983 Edition through Summer 1983 Addenda of ASME Section XI was applicable for the First Ten-Year Interval at VEGP. IWA-5242 of the 1983 Edition through Summer 1983 Addenda of ASME Section XI did not require insulation removal; therefore, this request for relief was not needed at VEGP during the First Ten-Year Interval.
j V. Alternate Examinatin:
- Insulation will be removed and pressure-retaining bolted connections will be VT-2 examined 1
when systems may or may not be pressurized. Insulated Class 1 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each refueling outage. Similarly, insulated Class 2 and 3 pressure-retaining bolted connections will be uninsulated and VT-2 examined once each inspection period.
O In addition to the above, the piping and components associated with these Class 1,2, and 3 V systems will be VT-2 examined at their required frequencies and under the conditions specified
- in IWA-5000, IWB-5000, IWC-5000, and IWD.5000 with the exception of the removal of insulation from bolted connections.
VI. Justification for Granting Relief:
Evidence ofleakage through pressure-retaining bolted connections which are in systems which are borated for the purpose of controlling reactivity is readily detectable by visual observation when systems are not at operating temperature and pressure. The boric acid concentrations are sufficiently high such that boric acid residues will be present ifleakage has occurred at the pressure-retaining bolted connection. In addition the ASME Section XI Code Committee has issued Code C ise N-533 (copy provided as Attachment I to this request for relief) which allows i as an alternative for Class I pressure-retaining bolted connections that insulation may be removed and VT-2 examined when the connection is not pressurized. For the reasons discussed i
above, SNC has determined that implementatioh of the proposed alternatives to the Code
- requirements provides an acceptable level of quality and safety and therefore requests that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
Vll. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, O 1997.
7-111 Rev. O
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECI MIC GENERATING PLANT UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR PELIEF NO. RR-26 v .
(continued)
ATTACHMENT 1 CASE N-533 5
CASE 3 OF ASME 80 flea AND PRESStlRE VE5SEL CODE Approval Date: Marsh 14,1995 See Numerical Mden kr espiration and any reaWirmation dates.
Case N 333 Alternative Requirements for VT 2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted ConnectionsSection XI, Division 1 Inquiry: What alternative requirements may be used
[ in lieu of those of IWA 3242(a) to remove insulation from Class ! pressure retauung bolted connections to perforto a VT 2 visual examination?
Reply: It is the opinion of the Committee that, as an alternative to the requirements of IWA 3242(a) to remove insulation from Class 1 pressure retaining bolted connections to perform a VT-2 visual exami-nation, the following requirements shall be met.
(a) A system pressure test and VT-2 visual exam-ination shall be performed each refueling outage 1 without removal of insulation.
(b) Each refueling outage the insulation shall be removed from the bolted connection, and a VT 2 vis-i ual examination sha!! be performed. The connection is not required to be preuurized. Any evidence of
!cakage shall be evaluated la anordance witn IWA- I 1
5250.
l l
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7-112 Rev.0 1
l l
SOUTHERN NUCLEAR OPERATING COMPANY ;
VOGTLE El ECTRIC GENERATING PL ANT. UNITS 1 AND_2 l SECOND TEN YEAR INTERVAL '
REOUEST FOR RELIEF NO. RR-27 l I. l System /Comnonent(s)_for Which Reliefis Requesisd: i l
This request for relief provides an alternative for qualification of VT-2 examination personnel. ,
II. Code Requirement:
i l
Subarticle IWA-2300 of the 1989 Edition of ASME Section XI contains the requirements for I qualifications of nondestructive examination personnel.
l III. Code Requirement from Which Reliefis Requested:
Reliefis requested from the requirements ofIWA-2300 for qualification of VT-2 examination j personnel and as an alternative it is proposed that ASME Code Case N-546 be used. A copy of !
the code case is provide as Attachment I to this request for relief. l
)
IV. Basis for Relief: I i Code Case N-546 provides an alternative to IWA-2300 for qualification of VT-2 examination O personnel. SNC reviewed Code Case N-546 and determined its implementation will l IU substantially reduce the burdens required by IWA-2300 for qualification of VT-2 examination personnel.
1 V. Alternate Ernmination:
- Southern Nuclear Operating Company will comply with the requirements of Code Case N-546 with the exception of using IWA-2321 of the 1995 Edition for visual acuity. SNC is requesting l to use IWA-2321 of the 1989 Edition as an alternative.
i VI, Justification for Grantine Relief:
1 The ASME Code Committee evaluated and approved Code Case N-546 as an acceptable
! alternative for qualification of VT-2 examination personnel. Code Case N-546 contains the following requirements for VT-2 examination personnel:
- 1. At least 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> plant walkdown experience, such as that gained by licensed and non-
, licensed operators, local leak rate personnel, system engineers, and inspection and nondestructive examination personnel.
J l 2. At least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of training on ASME Section XI requirements and plant specific procedures for VT-2 visual examination.
A 7-113 Rev.0
_ _ _ . . ~ . - - . . . . . . - . - - - - . _ . . . - . _ . - . -
4 SOUTIIERN NUCLEAR OPERATING COMPANY VOCTLE El FCTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL >
O G REQUEST FOR RELIEF NO. RR-27 (continued)
VI Justification for Grantina Relief (continued):
- 3. Vision test requirements ofIWA-2321,1995 Edition.
An exception is requested to the requirements ofIWA-2321 of the 1995 Edition for the vision test requirements. It is requested that the requirements ofIWA-2321 of the 1989 Edition be used j in lieu of the 1995 Edition. The Second Ten-Year Intervals for VEGP-1 and 2 are required by 10 CFR 50.55a to be in compliance with the requirements of the 1989 Edition. The use of the 1989 )
Edition is sufficient to assure the visual acuity of ex~ amination personnel and, as such, there is no !
need to impose the 1995 Edition on those examination personnel qualified to the requirements of
- The implementation of Code Case N-546, including the exception to IWA-2321 of the 1995
- Edition, will not affect the level of quality and safety, nor decrease the margin of public health l and safety. Therefore, it is requested that the proposed alternative be authorized pursuant to 10 4
CFR 50.55a(a)(3)(i).
O VII. Implementation Schedule:
'd
- This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, 1997.
i d
1 1
- O 1
7-114 Rev. 0 e-
I SOUTHERN NUCLEAR OPERATING COMPANY l VOGTLE ELEUI MIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-27 (continued)
ATTACHMENT 1 i l
4 4
I
- CASE :
N-546 CASES OF ASME BOILER AND PRESSURE YESSEL CODE Approval Date: August 24,1996 See hlumersal Inder for ensretion and any reamemoroon cates.
i' Case N 546 accordance with the referenced standard (i.e ANSI Alternadve Requirements for Quali8cadon of N45.2.6. ASNT SNT-TC-IA. or ASNT CP 189) pro.
VT 2 Examination Perunnel vided the examinauon personnel are qualined in accord-Seedon XI. Division 1 ance with the following requirements.
(a) At least 40 hrs plan walkdown expenence, such I as that gamed by licensed and nonlicensed operators.
Inquiry. What attemauve to the requirements of IWA-2300 may be used for qualineauon of VT 2 visual Ie led rate Penonnel, system engineen, and inspec. l examination penonnel? tion and nondestrucuve camminanon penonnel.
O *'
(6) At least 4 hn of training on Secuon XI require- j j
ments and plant specine procedures for VT-2 visual Asply It is the opmion of the Comnunce that VT. examination.
2 visual examination personnel need not be qualified (c) Vision test requirements of IWA 2321.1995 nor cemned to comparable levels of competence in Edition.
)
i i
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i
/m
(
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l 7-115 Rev. O
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 -
SECOND TEN-YEAR INTERVAL P REOUEST FOR RELIEF NO. RR-23 I. System / Component (s) for Which Reliefis Requested:
This request for relief proposes an alternative to the requirements ofIWA-7000 for replacement activities which involve " Rotation of Serviced Snubbers and Pressure Relief Valves for the Purpose of Testing".
II. Code Requirement:
Article IWA-7000 of the 1989 Edition of ASME Section XI contains the requirements for replacement activities including the completion of Form NIS-2," Owner's Report for Repairs or Replacements".
III. Code Requirement from Which Reliefis Reauested:
Reliefis requested from the requirements ofIWA-7000 and as an alternative it is proposed that ASME Section XI Code Case N-508-1 be used.
IV. Basis for Relief:
ASME Code Case N-508-1 provides an alternative to IWA-7000. SNC has reviewed the code case and has determined its implementation will substantially reduce nonbeneficial work activities required by IWA-7000.
V. Alternate Examination:
Southern Nuclear Operating Company will comply with the requirements of Code Case N-508-1 in lieu ofIWA-7000.
VI. Justification for Grantine kelief:
The ASME Code Committee evaluated the proposed alternatives contained in Code Case N-508-I and determined that they are acceptable for replacement activities involving snubbers and relief valves that are rotated from stock and installed on components for the purpose of testing. The implementation of Code Case N-508-1 will not affect the level of quality and safety, nor decrease the margin of public health and safety. While the cost sevings associated with Code Case N-508-I have not been quantified as a Cost Beneficial Licensing Action item, its implementation is consistent with the intent to eliminate nonbeneficial work activities and their associated costs.
Therefore, it is requested that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
O 7-116 Rev. 0 .
i SOUTilERN NUCLEAR OPERATING COMPANY VGGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 1 SECOND TEN-YEAR INTERVAL 3
~
REQUEST FOR RELIEF NO. RR-28 (G
(continued) j VII. Implementation Schedule:
I I
This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, l 1997.
I ;
1 f
i J
0 -
1 O
l 7 117 Rev. O
i SOUTITERN NUCLEAR OPERATING COMPANY VOGTLE ELEC FMIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL d REOUEST FOR RELIEF NO. RR-28 (continued) 4 ATTACBMENT 1 4
CASE N-508-1 l CASESOF As\tE BOILER A%D PRES $LRE VESSELCOoE i
J f 4
Approval oste: May 11,1994 1
See Numencalindes for exowntoon and any reartirmsten dates, 1
Case N 5081 (b) Items being removed shall have no evidence of Rotation of Serviced Snubbers and Pressure Relief failure at the time of removal; Valves for the Purpose of Testing (c) Items being rotated shall be removed and Section XI. Division 1 installed only by mechanical means:
(d) Items being installed shall previously have Inquay: What alternative rules to those stated in been in service; 1
IWA-4000 (IWA 7000 for Editions and Addenda (c) Preservice inspections and pressure tests shall prior to the 1991 Addenda) may be used when, for
> pI t the purpose of testing, snubbers and pressure relief be performed as required by IWA-4000 (IWA-7000 for Editions and Addenda prior to the 1991 Adden-
- V valves are rotated from stock and installed on com- a)-
J ponents (including piping systems) within the Section (f) "he Owner shall maintain a method ot trackias
! XI boundary? the items to ensure traceability of inservice inspec-
' tion and testing records; Rep & It is the opmion of the Committee that, as (g) Use of an NIS-2 form is not required except as j an alternative to the provtsions of TWA-4000 (IWA. provided in (i) below, 7000 for Editions and Addetda prior to the 1991 (h) Testing of removed snubbers and pressure re-Addenda) and for the purpose of testing snubbers lief valves, including required sample expansions, 1 and relief valves may be rotated from stock and in. shall be pet4ntmed in accordance with the Owner's stalled on components (including piping systems) test program:
within the Section XI boundary provided the follow. (i) Repair or replacement of removed items,when ing requirements are met: required, shall be performed in accordance with !
(a) Items being removed and installed shall be of IWA-4000 (IWA 4000 or IWA-7000 for Editions and the same design and construction: Addenda prior to the 1991 Addenda).
4 4
4 0
7-118 Rev. O i
- I
SOUTIIERN NUCLEAR OPERATING COMPANi
- VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2
- SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-29 I. System /Comnonent(s) for Which Reliefis Requested:
f An alternative to the requirements of Article IWF-5000 of the 1989 Edition of ASME Section XI for the inservice examination and testing of hydraulic and mechanical snubbers is proposed.
t II. Code Requirement:
i Article IWF-5000 of the 1989 Edition of ASME Section XI provides the rules and requirements for inservice inspection of snubbers including the " schedule and frequency for inservice testing and examining of snubbers and the qualification for personnel required to witness, perform, and/or evaluate the inspection program."
j III. Code Requirement from Which Reliefis Requested:
Reliefis requested from the requirements ofIWF-5000. As an alternative, it is proposed that the required inservice examination and testing of snubbers be performed in accordance with Section
. 13.7.2 of the VEGP Technical Requiremen'.s Manual (TRM). The details for the performance of I
inservice examinations and tests per the 1989 Edition of ASME Section XI are provided in the "O
~
V first addenda to ASME/ ANSI OM-1987, Part 4. The VEGP TRM was accepted by the NRC on December 10,1996, as part of the Improved Technical Specifications (ITS). Section 13.7.2 of the VEGP TRM is essentially the same as IWF-5000 except for the schedule for inservice examination of snubbers as described in OM-1987, Part 4, paragraph 3.2.3(b). OM-1987, Part 4 requires all snubbers to be visually examined every 18 months. The VEGP TRM incorporates the provisions of NRC Generic Letter 90-09, dated December 11,1990, and allows the frequency of snubber examinations to be extended up to every 48 months, contingent upon the results of examinations. !
.. l IV. Basis for Relici:
In lieu of using Article IWF-5000, the on-going examination and testing program will continue to be performed in accordance with the VEGP TRM. The existing examination and testing program is designed to demonstrate the functional integrity of the snubbers and is, at least, equivalent to the requirements of Article IWF-5000 and the provisions of NRC Generic Letter l 90-09. Continued use of the existing snubber examination and testing program w2 substantially l
reduce nonbeneficial work activities and radiation exposure if the requirements ofIWF-5000 '
were imposed.
Personnel who are required to witness, perform, and/or evaluate the inspection program, l including visual examinations, will be qualified in accordance with administrative procedures. I q The functional testing of snubbers will be perfo'rmed by trained personnel using a detailed l b procedure.
l 7-119 Rev.0 i
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-29 (continued)
V. Alternate Examination: ;
As noted above, the examination and testing program described in Section 13.7.2 of the VEGP TRM will continue to be performed. SNC will comply with those existing requirements provided in the aforementioned section of the VEGP TRM in lieu ofIWF-5000.
VI. Justification for Granting Relief:
l The snubber inservice examination and testing program as described in Section 13.7.2 of the 1 VEGP TRM is unchanged from that which was previously approved for use in the VEGP Standard Technical Specifications with First Tep-Year Interval Request fcr Relief RR-43
(
Reference:
RR-43 approved by the NRC correspondence dated November 26,1991 for VEGP-1 i
and December 17,1991 for VEGP-2). The subsequent adoption and approval of the VEGP !
Improved Technical Specifications (ITS) included the relocation of the snubber program, unchanged, to the VEGP TRM. The proposed alternative provides a level of quality and safety equivalent to the requirements of Article IWF-5000 of the 1989 Edition of ASME Section XI. '
As a result, SNC requests that relief be authorized pursuant to 10 CFR 50.55ah)(3)(i).
,Q V VII. Implementation Schedule:
l This request for reliefis applicable to the Second Ten-Year Interval which commences May 31, l 1997.
l l
l O
v 1 7-120 Rev.O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 i
SECOND TEN-YEAR INTERVAL Q REOUEST FOR RELIEF NO. RR-30 V
I. Sub_iect for Which Reliefis Requested:
Administrative change to the documentation and reporting requirements ofIWA-6000 for Class 1,2, and 3 pressure-retaining components and their supports.
II. Code Requirement:
Article IWA-6000 of the 1989 Edition of ASME Section XI requires the preparation and submittal ofinservice inspection summary reports within 90 days of the completion of the inservice inspection conducted during each refueling outage. Form NIS-1," Owner's Report for Inservice Inspection," and Form NIS-2, " Owner's Report for Repairs or Replacements," are i integral portions of the summary reports.
1 III. Code Requirement for Which ReliefIs Requested:
4 i
Relief from preparing the NIS 1 and NIS-2 forms end submitting the inservic. i..spection summary report within the 90-day time limit is requested.
IV. Basis for Relief:
U ASME Code Case N-532 provides an alternative to the 1989 ASME Code Section XI repair and replacement documentation and regulatory repqrting requirements. A copy of the code case is provided as Attachment I to this request for relief. SNC reviewed Code Case N-532 and determined that its implementation will substantially reduce the administrative burden created by IWA-6000.
V. Alternate to the Code Requirements:
Southern Nuclear Operating Company will comply with the requirements of Code Case N-532, with the following clarification regarding reporting of corrective measures. Code Case N-532, paragraph 2.0(c), requires an abstract for repairs, replacements, and corrective measures required due to an item containing a flaw or relevant condition exceeding the acceptance criteria of ASME Code Section XI. According to Section XI, the term " corrective measures" has two applications.
One application involves repair and replacement activities on pressure-retaining components (e.g., metal removal and welding). The other application involvetnaintenance-type activities, such as tightening of bolting, replacing gaskets / packing, cleaning surface corrosion products, and adjusting component supports. For Code Case N 532 reporting, SNC considers " corrective measures" to involve only repair and replacement activities.
In U -
7-121 Rev.O
SOUTHERN NUCLEAR. OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-30
- (continued) i
- VI. Justification for the Granting of Relief
Previously, voluminous NIS-1 Owner's Data Reports for Inservice Inspection were submitted which included the NIS-2 Owner's Reports for Repairs and Replacements. The process of i preparing, reviewing, and submitting the reports within the 90-day time requirement often resulted in difficulties for the licensee's staff. Implementation of Code Case N-532 allows the submittal of abstracts versus the reports required by the ASME Code Section XI. In addition, the implementation of that code case is consistent with the NRC's philosophy found in NRC letter l SECY-94-093 dated May 10,1995. Per SECY-94-093, the NRC is to take a proactive role through its representatives in the ASME Code to modify reporting requirements and to eliminate the need to submit inservice inspection reports following each refueling outage.
The ASME Code Committee evaluated the proposed alternative reporting requirements and determined that the requirements of Code Case N-532 are acceptable for replacing the existing
- documentation and reporting requirements. Since Code Case N-532 only affects documentation j and reporting requirements, its implementation will not affect the level of quality and safety, nor decrease the margin of public health and safety. While the cost savings associated with Code Os i Case N-532 have not been quantified as a Cost Beneficial Licensing Action item, its implementation is consistent with the intent to eliminate work activities which are not beneficial and costs therefor. Therefore, it is requested that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
VII. Imnlementation Schedule:
The proposed alternative reporting requirements will be implemented during the Second Ten-Year Interval which commences May 31,1997.
7-122 Rev. 0
I SOUTIIERN NUCLEAR OPERATING COMPANY I VOGTLE ELECFRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
' REOUEST FOR RELIEF N,Q. RR-30
- (continued)
ATTACHMENT 1 CASE 3
N-532 CASES OF 45ME aOILEa AND PRE 15t;RE VE.5SEL CODE l
Approval oste: cesamter 12.1994 See Numene Indes for enourshon and any reaffirmaroon dates.
Case N-531 (c) The completed Form NIS-2A shall be main.
Alternative Requirements to Repair and tained by the Owner.
Replacement Documentation Requirements and (f) The Owner shall maintain an index of Repair /
Inservice Summary Report Preparation and
. Replacement Plans in accordance with IWA-63d3 Submission as Required by IWA4000 and IWA
- 6000' The index shall identify the identification number required by (b) above and the inspection interval and Section XI. Division 1 period durmg which each repair or replacement was completed.
' Inquuy What alternatives may be used to the re.
quirements of IWA4910(d) and IWA-6210(e) for 2.0 OWNER'S ACTI\TIY REPORT completion ot Form NIS 2 following repair or re. PREPARATION AND SUBMITTAL placemen, and IWA-6210(c) and (d). !WA-6220 IWA-M.30(b). (c), and (d) and IWA-6240(b) for An OWNER'S ACTIVITY REPORT FORM preparation and submittal of the inservice summarv OAR-1 shall be prepared and certified upon com-teport and Form NIS-1? pletion of each refueling outage. Each Form OAR.
I prepared during an inspection period shall be sub-Reply. It is the opinion of the Committee that as mitted following the end of the inspection period.
Each Form OAR 1 shall contain the following:
an alternative to the requirements of IWA4910(d). (a) Abstract of applicable examinations and tests IWA 6210(c),(d), and (e). !WA 6220. IWA 6230(b).
with the information and format of Table 1.
(c), and (d), and IWA 6240(b) the following provi- ib) A listmg of itemts) with flaws or relevant con-sions may be used. This Case shall be utilized at least ditions that required evaluation to determine ac-until the end of the inspection penod in which it was invoked. ceptabilitv for continued service, whether or not the flaw or relevant condition was discovered during a scheduled examination or test. The listing shall pro-1.0 CERTIFICATION OF THE REPAIR OR vide the information in the format of Table 2.
REPI.ACEMENT (c) Abstract for repairs. replacements and correc.
(a) The Owner's Repair / Replacement Program tive measures performed, which were required due shallidentify use of this Case.
to an item containing a flaw or relevant condition (b) A Repair; Replacement Plan shall be prepared that exceeded IWB-3000. !WC 3000. IWD-3000, in accordance with IWA4140'. and shall be given a IWE 3000. IWF-3000. or IWL 3000 acceptance cri-unique identification number.
teria; even though the discovery of the flaw or rele.
(c) Upon completion of all required activities as-vant condition that necessitated the repair, replace-sociated with the Repair / Replacement Plan, the ment or corrective measure, may not have resulted Owner shall prepare a REPAIR / REPLACEMENT from an exammation or test required by this Division.
CERTIFICATION RECORD. FORM NIS-2A. If ac'ceptance criteria for a particular item is not (d) Form NIS 2A shall be presented to the specified in this Division the provtsions of IWAt inspector for certification.
3100(b) shall be used to determine which repairs, replacements, and corrective measures are required O wil references to Iw 4-4000 and tw A 6000 useon this Case reie, to the 199 Edmon. to be included m the abstract.The abstract shall pro-vide the information in the format of Tab!c 3.
7-123 Rev. O
._ . - . _ . . _ - - _ _ . - - . . . . _ _. _ _ - ~ . -_
SOUTHERN NUCLEAR OPERATING COMPANY
< VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL (n,
REOUEST FOR RELIEF NO. RR-30 4 (continued) ,
ATTACHMENT 1 (continued) t 4
CASE (continued) ,
N-532 CASES OF ASME BOILER AND PRESSL'RE VESSEL CODE 4
k 4
b FORM NIS.2A REPAIR / REPLACEMENT CERTIFICATION RECORD l
1 QWNER'S CERTIFICATE OF CONFORMANCE
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4 Sv s.gneg t,ws ce,tencate no the, me msoecto, no, h.s empiove, menes any me,,sas,. eso, esses e, wnoted. concenung the .ctanhos oe.e, bed a me Aener/mesiacement osan. Fu,me,mo,e. nenne, me m.ooeto, no, h.e emptove, snes be know m i en. .nenn., fo, e , se,s.nei mio,, o, o.ooe,t ge e, io.. of ..n. .ns g ..nnecie. h o m.-
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- h 7-124 Rev.O
1 SOUTRERN NUCLEA R OPERATING COMPANY ;
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 1 1
SECOND TEN-YEAR INTERVAL Cr REOUEST FOR RELIEF NO. RR.30 (continued) l ATTACHMENT 1 (continued) !
CASE (continued)
N-532 CASES OF AS\1E 80lLER AND PRE 55tRE VESSELCODE J
l FORM OAR.1 OWNER'S ACTMTV REPORT a.
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O i (a 1 7 125 Rev. O
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RELIEF NO. RR-30 (continued) 1 ATTACHMENT 1 (continued)
CASE (continued)
N-532 CASES OF ASME BOILER AND PRESSL'RE VESSELCODE l
1 l
l TABLE 1 ABSTRACT OF EXAMINATIONS AND TESTS Total Total Total Total Examinauons Exammauons Exammauons Emanunauens Credited N To Enaminauen Requered for Credited for credited N Date for The Category The Interval This Pened For The Pened interval Remarks
. l e TABLE 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Flaw or Relevant Condauon Found l Examinauen item item Charactenrauon Dunng Scheduled Section XI l Category Nunter Desenpuon (1WA.3300) Esarnmauon or Test (Yes or No) l l
TABLE 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE l naw or Reevant Condition Found .
Repair, During scheduled l Replacemeni, Section XI Repair /
or Correctrve item Description Examination or Date Replacement Code Class Measum Desenpuan of work Test (Yes/No) Complete Pian Number 7-126 Rev.O