ML20205P234
| ML20205P234 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 05/01/1999 |
| From: | Castronora M VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20205P192 | List: |
| References | |
| PROC-990501, NUDOCS 9904200085 | |
| Download: ML20205P234 (120) | |
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VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1
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INSERVICE INSPECTION PROGRAM I
SUMMARY
OF SIGNIFICANT CHANGES BETWEEN THE 2"d & 3"'
INTERVALS 1.
A new Relief Request was added (SPT-7)to provide an alternative to the Code which requires that insulation be removed for performance of pressure tests on systems which are borated for the purpose of controlling reactivity.
This Relief Request has been submitted and approved by the NRC for NAPS Unit 2, and for SPS Unit's 1&2.
2.
A new Relief Request was added (SPT-5) to provide an alternative to the Code which requires that bolting be removed and examined when leakage is evident at a pressure retaining bolted connection during a system pressure test.
3.
Class 2 piping weld selection for systems other than CHR, ECC and RHR are now based on current Code requirements, and not selected based on the 74 Summer 75 Code requirements.
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4.
Class 2 piping welds are now divided into two Code l
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Categories, C-F-1 for Stainless Steel piping, and C-F-2 for carbon steel piping.
Previously there was no distinction in the Code between stainless and carbon steel piping.
Selection is now based on these categories.
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A new Relief Request was added (CS-1)to provide an alternative to the Code requirements which require visual examinations of snubbers be performed in accordance with ASME/ ANSI OM -1987, Part 4.
l 6.
All other changes are procedural in nature.
The Code changes affect performance of NDE, acceptance and evaluation Criteria, and Repair / Replacement.
The affected procedures are being reviewed and revised as necessary.
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l 9904200085 990400 PDR ADOCK 05000338 NMg G
PDR L
NIA POWER p\\
iv' INSERVICEINSPECTION PROGRAM l
CHANGE / ISSUE APPROVAL FORM i
STATION:
North Anna UNIT NO.:
1 INTERVAL No.:
3 SUBMITTAL TITLE:
Inservice Inspection Program CHANGE NO.:
SECTION NO.:
REVISION: 0 (1F APFLICABLE)
================================__.___===============
DESCRIPTION OF CHANGES:
Initial Issue (ATTACH PAGE(S) SHOWING EXACT WORDING OF PROPOSED CHANGE)
REASON FOR CHANGE:
End of Second Inservice Inspection Interval.
Updated ISI Program required per 10 CFR 50.55(a).
ORIGINATED BY:
</tzd7 DATE:
3[s[49
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DEPARTMENT:
Qawww. m u e us.
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SUPERVISOR - MATERIALS /ISI ENGINEERING (CORPORATE)
REVIEWED AND
/
APPROVED:
b/
gru DATE:
3h /97 SUPERVISOR - TESTING and INSPECTION (STATION)
[A( & _
REVIEWED AND A
/i DATE:
J /d 'ff APPROVED:
x--------------------------------------
ANII (STATION)
REVIEWED:
//c/' [u DATE:
-____----______----- g ___________---___------_-__-__-----______--
STATION NUCLEAR SAFETY AND OPERATING COMMITTEE (SNSOC)
REVIEWED AND APPROVED:
. 7-DATE:
3 "//~[f I
-===========y=================================:-===
============-
AFTER APPROVAL, RETURN TO:
Joyce A. Lawler, IN-3NW l
Virginia Power, 5000 Dominion Blvd., Glen Allen, VA 23060 (J
DISTRIBUTION NO.:
DISTRIBUTION DATE:
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i VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 INSERVICE INSPECTION PROGRAM FOR THE THIRD INSPECTION INTERVAL VOLUME 1 INSERVICE INSPECTION PROGRAM FOR
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COMPONENTS AND COMPONENT SUPPORTS REVISION 0 May 1, 1999 INTERVAL 3 May 1, 1999 - April 30, 2009 ni-o*
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l VIRGINIA ELECTRIC AND POWER COMPANY l
NORTH ANNA POWER STATION UNIT 1 l
INSERVICE INSPECTION PROGRAM THIRD INSPECTION INTERVAL May 1, 1999 - April 30, 2009
%)
REVISION O DISTRIBUTION RECORD DISTRIBUTION DATE DISTRIBUTION DATE l
NO.
ENTERED INITIAL NO.
ENTERED INITIAL 1
21 2
22 3
23 4
24 5
25 6
26 7
27 8
28 9
29 O
10 30 i
t 11 31 12 32 13 33 14 34 15 35 16 36 17 37 18 38 19 39 20 40 l
Future distributions to this program plan will be accompanied by two copies of the distribution memo, one of which is to be returned per instructions thereon, and the other one is to be filed sequentially behind this page giving you a record of the contents of each distribution.
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N1-000 5/99 Rev. O ii l
r TABLE OF CONTENTS
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'w.
SECTION TITLE PAGE Revision i
Assignment 0
ii Distribution Record 0
iii Table of Contents 0
iv Abstract 0
v List of Drawings 0
viii Relief Request Status 0
1 Inservice Inspection Program -
0 General 2
Inservice Inspection Program -
O for Components 3
Inservice Inspection Program -
0 g-~s, for Component Supports
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4 Miscellaneous Interim Relief 0
Requests and Documentation 5
Partial Coverage Relief Requests 0
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Rev. 0
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i ABSTRACT i
[~D VIRGINIA ELECTRIC AND POWER COMPANY I
\\m l NORTH ANNA POWER STATION UNIT 1 INSERVICE INSPECTION PROGRAM j
THIRD INSPECTION INTERVAL j
May 1, 1999 TO April 30, 2009 In accordaace with 10 CFR 50.55a, the North Anna Unit 1 (NAPS-1)
Inservice Inspection (ISI) Program was updated to meet the requirements of ASME Section XI, 1989 Edition.
Steam generator inspections will continue 'to be performed under Plant Technical Specifications.
This updated program is for the NAPS-1 third ten year inspection interval scheduled to commence May 1, 1999 and be completed April 30, 2009.
These dates reflect extension of the first intervu; by 201 days documented in our letter to the NRC dated 8/25/88, Serial No.88-486, and extension of the second interval by 127 days approved per NRC Letter No.98-167, dated 3/6/98.
In cases where the requirements of Section XI have been determined to be impractical, requests for relief have been i
developed per 10 CFR 50.55a(g) (5).
This document provides an overview and summary of the NAPS-1 ISI Program for Subsections IWA, IWB, IWC, IWD, and, IWF.
The boundaries of the ISI Program, component classifications, and the employment of exemptions in IWB-1220, IWC-1220, IWD-1220, and
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IWF-1230 are shown on the ISI Classification Boundary Drawings l
(CBDs).
The codings, symbols and text used on the CBDs are s-detailed on 11715-CBM-LLS-3, Legends and Symbols Drawing.
Section 1 of this document, provides the general information and format regarding this program.
Applicable Code Cases and drawings are also listed.
Section 2 of this document provides an ISI Program implementation overview for ISI Class 1, 2,
and 3 components.
Relief requests for components are also provided.
Section 3 of this document provides a program implementation overview for ISI Class 1, 2,
and 3 component supports.
Relief requests for component supports are provided.
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LIST OF DRAWINGS 7-s lv UNIT 1 CBM DRAWING 11715-CBM-L&S-3 Legend and Symbols UNIT 1 CBB DRAWINGS 11715-CBB-006A-3 Air Cooling and Purging System 11715-CBB-040C-3 Air' Conditioned Chilled Water System l
11715-CBB-040D-3 Air Conditioning condenser Water System 11715-CBB-102B-3 Interior Fire Protection & Hose Rack System UNIT 1 CBM DRAWINGS 11715-CBM-070A-3 Main Steam System 11715-CBM-070B-3 Main Steam System 11715-CBM-072A-3 Auxiliary Steam and Air Removal System
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11715-CBM-074A-3 Feedwater System t
11715-CBM-078A-3 Service Water System 11715-CBM-078B-3 Service Water System 11715-CBM-078C-3 Service Water System 11715-CBM-078G-3 Service Water System 11715-CBM-078H-3 Service Water System I
11715-CBM-078J-3 Service Water System l
11715-CBM-078K-3 Service Water System 11715-CBM-078L-3 Service Water Chemical Addition System 11715-CBM-079A-3 Component Cooling Water System 11715-CBM-079B-3 Component Cooling Water System 11715-CBM-079C-3 Component Cooling Water System p>
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LIST OF DRAWINGS CONTINUED i (
UNIT 1 CBM DRAWINGS 11715-CBM-079D-3 Component Cooling Water System 11715-CBM-079E-3 Component Cooling Water System 11715-CBM-082A-3 Compressed Air System 11715-CBM-082F-3 Compressed Air System 11715-CBM-082N-3 Compressed Air System 11715-CBM-088A-3 Fuel Pit CLNG & Refueling Pur. System i
11715-CBM-089B-3 Sampling System l
11715-CBM-089D-3 Sampling System 11715-CBM-090A-3 Vent and Drains System j
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11715-CBM-090C-3 Vent and Drains System 11715-CBM-091A-3 Cont Quench & Recirc Spray Sub System
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11715-CBM-091B-3 Cont Quench & Recirc Spray Sub System 11715-CBM-092A-3 Leakage Monitoring System 11715-CBM-092A-3 Containment Vacuum System 11715-CBM-093A-3 Reactor Coolant System 11715-CBM-093B-3 Reactor Coolant System 11715-CBM-094A-3 Residual Heat Removal System 11715-CBM-095A-3 Chemical and Volume Control System 11715-CBM-095B-3 Chemical and volume Control System 11715-CBM-095C-3 Chemical and Volume control System 11715-CBM-096A-3 Safety Injection System 11715-CBM-096B-3 Safety Injection System 11715-CBM-098A-3 Steam Generator Blowdown System
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f LIST OF DRAWINGS CONTINUED I
l UNIT 1 CBM DRAWINGS l
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l 11715-CBM-102A-3 Chemical Feed System 11715-CBM-105B-3 Seconda / Plant Gas Supply System 11715-CBM-106A-3 Containment Atmosphere cleanup System 13075-CBM-093C-3 Reactor Coolant System 13075-CBM-102C-3 Chemical Feed System Additionally, several Unit 1 drawings contain Unit 2 Components. These follow:
11715-CBB-040C-3 Air Conditioned Chilled Water System 11715-CBB-040D-3 Air Conditioning Condenser Water System 11715-CBM-072A-3 Auxiliary Steam and Air Removal System 11715-CBM-078A-3 Service Water System 11715-CBM-078C-3 Service Water System 11715-CBM-078G-3 Service Water System 11715-CBM-078H-3 Service Water System 11715-CBM-078J-3 Service Water System 11715-CBM-078K-3 Service Water System 11715-CBM-079A-3 Component Cooling Water System 11715-CBM-088A-3 Fuel Pit CLNG & Refueling Pur. System 11715-CBM-093B-3 Reactor Coolant System 11715-CBM-10LJ-3 Secondary Plant Gas Supply System 11715-CBM-106A-3 Containment Atmosphere Cleanup System i
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N1-000 5/99 ON Rev. O vii
I RELIEF' REQUEST STATUS i
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Relief No.
Status VP Ltr No.
NRC Ltr. No NDE-1 Pending NDE-2 Pending NDE-3 Pending NDE-4 Pending NDE-5 Pending NDE-6 Pending NDE-7 Pending NDE-8 Pending NDE-9 Pending NDE-10 Pending NDE-ll Pending NDE-12 Pending NDE-13 Pending NDE-14 Pending SPT-1 Pending q
SPT-2 Pending i
SPT-3 Pending SPT-4 Pending SPT-5 Pending I
SPT-6 Pending SPT-7 Pending f-g SPT-8 Pending N
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O VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 INSERVICE INSPECTION PROGRAM O
GENERAL THIRD INSPECTION INTERVAL N1-001 Rev. O 5/99 i
1-1
I SECTION 1' TABLE OF CONTE TS SECTION TITLE PAGE 1.O INSERVICE INSPECTION PROGRAM - GENERAL 1-3 1.1
' GENERAL INFORMATION 1-3 1.1.1 Preservice Examinations 1-2 1.1.2 Initial Inservice Inspection Interval 1-3 l
1.1.3 Second Ins'ervice Inspection Interval 1-3 1.1.4 Third Inservice Inspection Interval 1-3 1.1.5 Inservice Inspection Program 1-4 l
Description l
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I 1.0-INSERVICE INSPECTION PROGRAM - GENERAL O
1.1 GENERAL INFORMATION North Anna Power Station Unit 1 (NAPS-1) is located on Lake Anna in Louisa County, Virginia.
The plant employs a Pressurized Water Reactor (PWR) and associated Nuclear Steam Supply System components provided by Westinghouse Electric Corporation.
1.1.1 Preservice Examinations Preservice examinations at NAPS-1 were per#ormed j
utilizing the requirements of ASME Section XI, i
1974 Edition with Addenda throug,h Summer 1975.
1.1.2 Initial Inservice Inspeccion Interval NAPS-1 commercial operation commenced on June 6, 1978.
Accordingly, the initial ten-year inspec-tion interval started on 6/6/78.
This interval, however, was extended 201 days to 12/24/88 as documented by letter to the NRC dated 8/25/88, serial no.88-486.
Examinations during the first inspection interval were also performed per ASME Section XI, 1974 Edition with Addenda through summer 1975.
Certain system pressure O
testing requirements were completed per ASME Section XI, 1977 Edition with Addenda through i
summer 1979.
I 1.1.3 Second Inservice Inspection Interval The second inservice inspection interval commenced on December 24, 1988 and was completed on April 30, 2000.
The end of interval date reflects an extension of 1 year and 127 days as approved by NRC Letter 98-167, dated 3/6/98.
Examinations were performed per ASME Section XI, 1983 Edition through the summer 1983. Addenda.
j 1.1.4 Third Inservice Inspection Interval As mandated by the Code of Federal Regulations, Title 10, Part 50, Section 50.55a (10 CFR 50.55a), the NAPS-1 Inservice Inspection (ISI)
Program has been updated to the 1989 Edition of ASME Section XI.
This is the latest edition and addenda of Section XI incorporated into 10 CFR 50.55a as of April 30, 1998.
The third inservice inspection interval is scheduled to j
N1-001 Rev. O S/99 1-3 l
commence on May 1, 1999 and be completed on
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April 30, 2009.
1.1.5 Inservice Inspection Program Description The ISI Program contained herein addresses the inservice inspection and testing of ISI Class 1, i
2, and 3 components and,their associated component supports.
The applicable requirements in Subsections IWA, IWB, IWC, IWD, IWF and the Mandatory Appendices of the 1989 Edition of ASME I
Section XI have been incorporated into our corporate and site ISI programs and procedures.
This document is not intended to provide specific information on the implementation of the ISI Program.
The intent of this document is to provide specific information on the scope of the NAPS-1 ISI program (i.e.,
its boundary and compliance with Section XI) and identify those Section XI requirements which are deemed impractical.
Requests for relief for these impractical requirements have been developed per 10 CFR 50. 55a (g) (5).
1.1.5.1 Exclusion of Subsections IWE and IWL (O)
Subsections IWE and IWL, Inservice Inspection of Class MC and CC components and Inservice Testing of Class CC concrete components in Nuclear Power Plants are not included in this program.
1.1.5.2 Exclusion of Subsections IWP and IWV Subsections IWP and IWV, Inservice Testing of Pumps and Inservice Testing of Valves in Nuclear Power Plants are not included in this program.
The IWP and INV program are submitted separately.
1.1.5.3 Exclusion of Categories B-T and B-J ASME Section XI, Subsection IWB, Categories B-F, Pressure Retaining Dissimilar Metal Welds, and B-J, Pressure Retaining Welds in Piping, are not currently included in this Third Interval ISI Program.
The welds within
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N1-001 Fev. 0 5/99 i
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I' the scope of these categories are to be l
selected and examined per a Risk
(
Informed ISI Program for North Anna Unit 1.
Virginia Power has submitted a
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request to implement this alternative via Virginia Power Letter Serial No.98-696, dated 12/22/98.
This letter included a request f'or a two year delay l
in the implementation of the t
alternative, during which time a Risk Informed ISI Program for North Anna Unit 1 will be developed.
1.1.5.4 Inservice Inspection Program Plan The inservice inspection program plan I
required by IWA-2420, detailing the components and component supports selected for examination during the I
interval will be provided as a separate document.
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.4 1.1.5.5 Inspection Program Employed l
The ISI Program for North Anna Unit 1 will utilize the interval format of
- p Inspection Program B, as shown in IWA-l g 2432.
1.1.5.6 Weld Reference System The alternative to the weld reference I
system required by IWA-2600 is detailed in relief requests NDE-4 and NDE-5.
1.1.5.7 Classification of Components A.
Classification of Components-Design i
North Anna Unit 1 was issued construction permit No. CPPR-77 in February 1971.
The station design incorporates the codes and standards that were in effect when the equipment was purchased.
The codes and standards used for the design, fabrication, erection, and testing of safety related components are commensurate with the importance of the safety functions to be performed.
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N1-001 Rev. 0 5/99 f
1-5 I
r The group classifications tabulated in the " Standard Format and Content l
of Safety Analysis Reports for Nuclear Power Reactors", issued in February 1971, and in Safety Guide No. 26, published in March 1972, incorporated, in most cases, later i
editions of codes than those in effect when the majority of safety-related equipment was designed.
Some of the. equipment that would fall under a " group" as defined in Safety Guide No. 26 was designed to different codes or different editions of the same code.
For example, for different components that would be in the same group, one may be designed to ASME III-1968, one to ASME III-1971, and one to ASME VIII-1968.
Therefore, pressure-containing l
components of safety-related i
L systems do not necessarily fall under the group classification i
I listed above.
(
B.
Classification of Components-j Inservice Inspection The classification of components for Inservice Inspection was per-formed by employing.the guidelines of 10 CFR 50.55a, USNRC Regulatory Guide 1.26 Revision 3, ANSI N18.2-1973, ANSI N18.2a-1975, and NUREG-0800, Standard Review Plan 3.2.2-Revision 1.
The ISI Classification of each component included in NAPS-1 program is shown on the ISI Classification Boundary Drawings (CBB's and CBM's) listed on pages v through vii. These drawings detail the line codings and component markings which provide the ISI Classification of all items included within the ISI Program boundaries.
In general, the ISI classification of piping is g.
N1-001 Rev. 0 5/99 1-6
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accomplished by line " codings" or i
symbols shown on 11715-CBM-L&S-3.
ISI classification of components which graphically cannot be coded, such as pressure vessels, pumps, l
l strainers, and tanks, etc. include L
text (e.g.,
ISI Class'1A, 3E, etc.)
I t
shown either inside of the component graphics or adjacent to 1
the component mark number.
Classification changes are i
typically shown at the seat of a l
valve, with the ISI Class shown on l
either side of the classification
" break line".
l 1.1.5.8 Components and Component Supports Ex-empt from Examination The application of the exemptions allowed per IWB-1220, IWC-1220, IWD-1220, and IWF-1230 are also detailed by the codings and component markings of 11715-CBM-L&S-3 discussed in Section-1.1.4.5.B.
The IWF exemptions are f
those exemptions allowed by Code Case N-491.
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1.1.5.9 Requests For Relief-Components and Component Supports Where the requirements of ASME Section l
XI have been determined to be impractical, requests for relief have been developed in accordance with 10 CFR 50.55a(g)(5;.
The impractical requirements are detailed in five sets of relief requests: a NDE series for components and nondestructive examination areas, a SPT series for system pressure testing areas, a CS series for component supports, a series for interim relief requests and a series for documenting partial coverage exams.
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1.1.5.10 ASME Section XI Code Cases Incorporated Into Program As allowed by 10 CFR 50.55a(c) (3) and USNRC Regulatory Guide 1.147, Revision 11, the following Code Cases are being incorporated into the NAPS-1 ISI Program:
Case N-401-1 Eddy Current Examination,Section XI, Division 1.
Case N-402-1 Eddy Current Calibration l
Standards,Section XI, l
Division 1.
Case N-416-1 Alternative Pressure Test l
Requirement For Welded.
Repairs or Installation of Replacement Items by l
Welding, Class 1, 2, and.
l 3,
Section XI, Division l
1.
Case N-435-1 Alternative Examination Requirements for Vessels with Wall thickness 2 in.
or less,Section XI, Division 1.
Case N-437 Use of Digital Readout and Digital Measurement
' Devices for Performing Pressure Tests,Section XI, Division 1.
Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1.
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Case N-461 Alternative Rules for l
Piping Calibration Block Thickness,Section XI, Division 1.
I Case N-481 Alternate Examination Requirements for Cast Austenitic Pump Casings,Section XI, Divisiva 1.
(Reference WCAP-13045 for O
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L the evaluation to demon-strate the safety and, serviceability of the l
pump casing.)
Case N-489 Alternative Rules for Level III NDE l
Qualification Examinations,Section XI, Divisions 1, 2,
and 3.
Case N-491 Alternative Rules for Examination of Class 1,2, and 3 and MC Component i
Supports of Light-Water i
Cooled Power Plants,Section XI, Division 1.
Case N-498-1 Alternative Rules for 10-year Hydrostatic Pressure Testing for Class 1, 2
and 3 Systems, Section l
XI, Division 1.
Case N-508-1 Rotation of Serviced Snubbers and Pressure
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L Relief Valves for the Purpose of Testing,Section XI, Division 1.
Case N-509 Alternative' Rules-for the Selection and Examination l
of Class 1, 2, and 3 Integrally Welded i
Attachments,Section XI, Division 1.
I Case N-522 Pressure Testing of Containment Penetration Piping,Section XI, Division 1.
Case N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping,Section XI, Division 1.
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Case N-535 Alternative Requirements l /'
for Inservice Inspection
'M Intervals,Section XI, Division 1.
Case N-566 Corrective Action for Leakage Identified at Bolted Connections, Sec-l tion, XI, Division 1.
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Code Cases identified with an asterisk
(*) are not currently approved by Regulatory Guide 1.147 for use.
Relief Requests have been included requesting approval to implement these Code Cases.
These relief requests are included in the section of this program titled j
l Inservice Inspection Program for Components.
1.1.5.11 Augmented Inservice Inspection Program for ISI Class 1, 2,
and 3 Components Augmented Inservice Inspection activities which pertain to components l
within the scope of the ASME Section XI l
s program and specific components j.
scheduled to be examined per the l
augmented program will be included as
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part of the Inservice Inspection Program Plan discussed in Section 1.1.5.2.
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N1-001 Rev. O S/99 1-10
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VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 i
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INSERVICE INSPECTION PROGRAM l
FOR COMPONENTS l
THIRD INSPECTION INTERVAL
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O N1-002 Rev. 0 5/99
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l SNCTION 2 TABLE OF CONTENTS
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l SECTION TITLE PAGE 2.0 INSERVICE INSPECTION PROGRAM FOR COMPONENTS 2-3 2.1 PROGRAM DESCRIPTION 2-3 2.2 PROGRAM
SUMMARY
2-3 2.3 RELIEF REQUESTS 2-4 I
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' O N1-002 Rev. 0 5/99 2-2
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2.0 INSERVICE INSPECTION PROGRAM FOR COMPONENTS 2.1 PROGRAM DESCRIPTION 2.1.1 The Inservice Inspection Program for Class 1, 2
and 3 components meets the requirements of Section XI of the ASME Boiler and Pressure i
Vessel Code, 1989 Edition.
Where these requirements are determined to be impractical, j
or where alternatives are identified in accordance with 10 CFR 50.55a (a) (3), specific requests for relief have been written and included in this section.
2.1.2 Repairs and replacements shall be in accordance with Station Administrative Procedures, which assure compliance with ASME Section XI, Division 1,
1989 Edition.
2.2 PROGRAM
SUMMARY
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2.2.1 The Inservice Inspection Program for North Anna Unit 1 utilizes the tables IWB-2500-1, IWC-2500-1, and IWD-2500-1 of the 1989 Edition of the ASME Section XI Code.
Components selected for examination are identified in the Inservice Inspection Program Plan.
2.2.2 Categories B-E, B-P, C-H, D-A, D-B, and D-C constitute the scheduled 10-year interval pressure testing requirements.
Additionally, j
Code Case N-498-1, see below, will be followed.
The following guidance will be utilized I
programmatically concerning pressure testing, a) Pressure tests which meet the pressure, temperature, hold time, and VT-2 examination requirements of IWA-5000, IWB-5000, IWC-5000, IWD-5000, or Code Case N-498-1 as applicable will be treated as complete for inspection scheduling purposes, b) Pressure tests required for ASME Section XI repairs or replacements will be conducted in accordance with the station's repair replace-ment program as documented at the station.
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2.3 RELIEF REQUESTS l
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ksl 2.3.1 Nondestructive examination (NDE) relief requests are s
presented in a numeric sequence with a NDE prefix.
Pressure tests relief requests are presented in a i
numeric sequence with a SPT prefix.
l 2.3.2 The following is a summary the third interval relief requests.
RELIEF PAGE
SUMMARY
REQUEST No.
NDE-1 2-9 Relief is requested from meeting the requirements l
of Category C-G, Item Number C6.10 of the 1989 l
Edition of ASME Section XI.
Several welds on the Outside Recirculation Spray Pump casings are inaccessible, and it is therefore not practical to perform the required 100% examination on these welds.
i NDE-2 2-11 Relief is requested from meeting the requirements l
of Category C-G, Item Number C6.10 of the 1989 Edition of ASME Section XI.
Several welds on the Safety Injection Pump casings are inaccessible, and it is therefore not practical to perform the p.)
required 100% examination on these welds.
(
NDE-3 2-13 Relief is requested from meeting the requirements of Appendix I of the 1989 Edition of ASME Section XI.
Calibration blocks were initially developed prior to the issuance and adoption of ASME Section XI, and it is therefore not practical to j
meet the current code requirements for calibration blocks.
NDE-4 2-15 Relief is requested from meeting the requirements of IWA-2600 of the 1989 Edition of ASME Section XI.
The original construction code, ANSI B31.7, 1969 Edition, did not establish a weld reference system.
It is not practical to fully comply with the current requirements regarding weld reference systems.
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(~s NDE-5 2-16 Relief is requested from meeting the requirements
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of IWA-2600 of the 1989 Edition of ASME Section XI. The automated tooling used to inspect the Reactor Vessel and Reactor Vessel Nozzles uses its own reference system.
It is not practical to l
fully comply with the current requirements regarding weld reference systems.
NDE-6 2-17 Relief is requested from meeting the requirements of Category B-D, Item Numbers B3.110 and B3.120 of the 1989 Edition of ASME Section XI.
Due to the physical configuration of the pressurizer, attached electrical equipment and local dose rates, it is not practical to perform the code required UT examinations.
l l
NDE-7 2-19 Relief is requested from meeting the requirements of Category B-D, Item Numbers B3.150 and B3.160 of the 1989 Edition of ASME Section XI.
Due to nozzle joint design, it is not practical to comply with the code required UT examinations of the Regen. Heat Exchanger nozzle-to-sessel welds and nozzle inside radius welds.
l NDE-8 2-25 Relief is requested from meeting the requirements of Category C-F-2, Item Number C5.81 of the 1989
[
\\
Edition of ASME Section XI.
Branch connections i
\\-
on the main steam relief header utilize reinforcement pads which cover the branch connection welds.
It is not practical to perform the code required surface examination of the branch welds as they are encased by the l
reinforcement pads.
i NDE-9 2-26 Relief is requested from meeting the requirements l
of IWA-5250 of the 1989 Edition of ASME Section
(
XI.
It is not practical to follow the code rules for repair / replacement of the identified portions of the service water system when repair / replacement is deemed necessary'during plant operations.
Temporary repairs and augmented inspections will be performed until full repair / replacement can be performed during a scheduled outage.
l t
V N1-002 Rev. 0 5/99 2-5 e
Relief is requested from the meeting the
,~
NDE-10 2-43
(}'
requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Tables INB-
~
2500-1, IWC-2500-1 and IWD-2500-1 Examination Categories B-H, B-K-1, C-C, D-A, D-B, and D-C, examination of Integrally Welded Attachments, as outlined in II above. Code Case N-509 will be implemented for the examination of all Class 1, 2
and 3 integral attachments with conditions.
liDE-11 2-45 Relief is requested from meeting the examination requirements for Class 1 and Class 2 longitudinal j
piping welds contained in~ASME Section XI,
}
Division 1, 1989 Edition, Table IWB-2500-1, 1
Category B-J, Table IWC-2500-1, Categories C-F-1 and C-F-2.
Code Case N-524 will be implemented for the examination of Class 1 and 2 longitudinal pipinq welds.
NDE-12 2-46 Reliet is requested from meeting the requirements of ASME Section XI, Division 1, 1989 Edition, Article IWA-7000 when removing snubbers or safety / relief valves for testing.
As an alternative Code Case N-508-1 will be implemented when removing snubbers or safety / relief valves for testing.
(~
(,
NDE-13 2-47 Request to utilize Code Case N-535, which allows an alternative to the Code allowed interval extension.
Relief is requested from meeting the NDE-14 2-48 requirements of Table IWB-2500-1, Category B-G-1, Item No. B6.10 which requires a surface examination of the reactor vessel closure head nuts.
Examination will be performed in accordance with the 1989 Addenda, a VT-1 examination along with supplementary exams.
SPT-1 2-50 Relief is requested from meeting the requirements of Category B-E, Item Number B4.13, and Category B-P, Item Numbers B15.10 and B15.11 of the 1989 Edition of ASME Section XI.
It is not practical to perform the code required VT-2 examinations on Reactor Vessel partial penetration welds during a hydrostatic test, and the bottom of the Reactor Vessel during a~ system leakage test.
O N1-002 Rev. 0 5/99 2-6
("')s Relief is requested from meeting the requirements SPT-2 2-52
(,
to perform a hydrostatic pressure test of Class 1,
Class 2 and Class 3 components in accordance with ASME Section XI, Division 1, Paragraph IWA-4000 for welded repairs or installation of e
replacement items by welding.
Code Case N-416-1 will be implemented with additional requirements for the performance of hydrostatic pressure tests of components after repair or replacement by welding.
SPT-3 2-54 Relief is requested from meeting the 10 year System Hydrostatic Testing requirements of ASME Section XI, Division 1, 1989 Edition, Table IWB-2500-1 Category B-P, Table IWC-2500-1 Category C-H, and Table IWD-2500-1 Categories D-A, D-B and D-C.
Code Case N-498-1 will be implemented for the performance of 10 year system hydrostatic tests.
SPT-4 2-55 Relief is requested from meeting the requirements of Category B-P, Item Numbers B15.51 and B15.71 of the 1989 Edition of ASME Section XI.
It is not practical to perform the code required VT-2 examination on the identified Class 1 vent and drain, sample, and instrumentation piping and
(}
connections.
SPT-5 2-58 Relief is requested from meeting the requirements of ASME Section XI, Division 1, 1989 Edition, paragraph IWA-5250(a)(2) as outlined in paragraph II above.
Code Case N-566 will be implemented as an alternative when leakage is found at a bolted connection during a system pressure test.
SPT-6 2-59 Relief is requested from meeting-the requirements of IWA-52 50 (a) (2 ) which requires that if leakage occurs at a bolted connection, the bolting shall be removed, given a VT-3 examination and evaluated in accordance with IWA-3100.
The proposed alternative is to perform a VT-2 examination, and evaluate the need to remove bolting.
V N1-002 Rev. 0 5/99 2-7
/'S SPT-7 2-61 Relief is requested from meeting the pressure g
( )
test requirements of Table IWC-2500-1,
'~
Examination Category C-H, Items C7.30, C7.40, C7.60 and C7.70 for penetration piping where the balance of the system is outside the scope of Section XI.
As an alternative, the Appendix J testing of these penetrations is performed as allowed in Code Case N-522.
SPT-8 2-66 Relief is requested from meeting the requirements of Subparagraph IWA-5242(a)to remove insulation from pressure retaining bolted connections to perform a VT-2 examination.
This applies to bolted connections in systems borated for the purpose of controlling reactivity.
As an alternative the insulation will be removed and the area examined during refueling outages.
lO) v 9
N1-002 Rev. 0 5/99 2-8
Virginia Electric and Power Company North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-1 I.
Identification of Components:
Systems : Outside Recirculation Spray Pumps (1-RS-P-2A and 1-RS-P-2B).
Components:
Pump casing welds identified below:
Component Drawing No.
Weld No.
1-RS-P-2A 11715-WMKS-RS-P-2A SW-1 1-RS-P-2A 11715-WMKS-RS-P-2A SW-2 1-RS-P-2A 11715-WMKS-RS-P-2A SW-3 1-RS-P-2A 11715-WMKS-RS-P-2A LS-6 1-RS-P-2A 11715-WMKS-RS-P-2A LS-7 1-RS-P-2A 11715-WMKS-RS-P-2A LS-8 1-RS-P-2A 11715-WMKS-RS-P-2A LS-9 (Partial 1-RS-P-2A 11715-WMKS-RS-P-2A LS-10 Access)
AIQ 1-RS-P-2B 11715-WMKS-RS-P-2B SW-1 1-RS-P-2B 11715-WMKS-RS-P-2B SW-2 1-RS-P-2B 11715-WMKS-RS-P-2B SW-3 1-RS-P-2B 11715-WMKS-RS-P-2B LS-6 1-RS-P-2B 11715-WMKS-RS-P-2B LS-7 1-3S-P-2B 11715-WMKS-RS-P-2B LS-8 1-RS-P-2B 11715-WMKS-RS-P-2B LS-9 (Partial 1-RS-P-2B 11715-WMKS-RS-P-2B LS-10 Access)
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Category C-G, Item Number C6.10, requires that a surface examination be performed on 100% of the welds each inspection interval. Code Case N-460, Alternate Examination Coverage for Class 1 and Class 2 welds, allows a reduction in coverage, if it is less than 10%.
The examination can be limited to one pump in the case of multiple pumps of similar design, size, function, and service in a system.
The examination may be performed from either the inside or outside surface of the component.
fm N1-002 Rev. 0 5/99 2-9
RELIEF REQUEST.NDE-1
(
CONTINUED III. Basis for Relief:
Each of the two outside recirculation spray pump casings have a total of five circumferential welds and five longitudinal welds.
Three of the circumferential welds (SW-1, SW-2 and SW-3), and three of the longitudinal welds (LS-6, LS-7 and LS-8) are completely encased in concrete and are not accessible for eximination from the outside diameter (O.D.).
Of the remaining two longitudinal welds, one weld is partially encased in concrete (LS-9) and one weld is partially covered by a vibration plate (LS-10).
Partial O.D.
examinations can be performed on both of these longitudinal welds.
Two circumferential welds are com-pletely accessible for examinations from the O.D.
Surface examinations from the Inside Diameter (I.D.) are not a practicable alternative.
Access to the inside of the pump casings is limited by physical size (24 inch outside diameter), the pump shaft, and the pump shaft support obstructions.
Disassembly of the pump and removal of the pump shaft would cause a burden without a compensating i
l increase in quality and safety.
i gr g IV. Proposed Alternate Examination:
U A surface examination of the accessible portions of the j
circumferential and longitudinal welds will be performed to l
the extent and frequency described in Table IWC-2500-1.
A j
remote visual examination (VT-1) of the I.D.
of the pump casing welds will be performed only if the pump is disassembled for maintenance, and the pump shaft is removed.
l The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised.
OU N1-002 Rev. 0 5/99 2-10
Virginia Electric and Power Company North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-2 I.
Identification of Components:
Systems: Safety Injection Pumps (1-SI-P-1A and 1-SI-P-1B).
Components: Pump casing welds identified below:
Component Drawing No.
Weld No.
1-SI-P-1A 11715-WMKS-SI-P-1A 1
1-SI-P-1A 11715-WMKS-SI-P-1A 2
1-SI-P-1A 11715-WMKS-SI-P-1A 3
1-SI-P-1A 11715-WMKS-SI-P-1A LS-1 1-SI-P-1A 11715-WMKS-SI-P-1A LS-2 1-SI-P-1A 11715-WMKS-SI-P-1A LS-3 1-SI-P-1A 11715-WMKS-SI-P-1A LS-4 (Partial 1-SI-P-1A 11715-WMKS-SI-P-1A LS-5 Access) 1-SI-P-1B 11715-WMKS-SI-P-1B 1
(q l
1-SI-P-1B 11715-WMKS-SI-P-1B 2
i 1-SI-P-1B 11715-WMKS-SI-P-1B 3
1-SI-P-1B 11715-WMKS-SI-P-1B LS-1 1-SI-P-1B 11715-WMKS-SI-P-1B LS-2 i
1-SI-P-1B 11715-WMKS-SI-P-1B LS-3 l
1-SI-P-1B 11715-WMKS-SI-P-1B LS-4 (Partial 1-SI-P-1B 11715-WMKS-SI-P-1B LS-5 Access) l l
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Category C-G, Item Number C6.10, requires that a surface examination be performed on 100% of the welds each inspection interval. Code Case N-460, Alternate Examination Coverage for Class 1 and Class 2 welds, allows a reduction in coverage, if it is less than 10%.
The examination can be limited to one pump in the case of multiple pumps of similar design, size, function, and service in a system.
The examination may be performed from either the inside or outside surface of the component.
Od N1-002 Rev. 0 5/99 2-11
RELIEF REQUEST NDE-2
"~
CONTINUED III. Basis for Relief:
Each of the two-outside-recirculation spray pump casings have a total of five circumferential welds and five longitudinal welds. Three of the circumferential welds (1, 2 and 3), and three of the longitudinal welds (LS-1, LS-2 and LS-3) are completely encased in concrete and are not accessible for examination from the outside diameter (O.D.).
Of the remaining two longitudinal welds, one weld is partially encased in concrete (LS-4) sand one weld is partially. covered by a vibration plate (LS-5).
. Partial O.D.
]
examinations can be performed on both of these longitudinal welds.
Two circumferential welds are completely accessible for-examinations from the O.D.
Surface examinations from the Inside Diameter (I.D.) are not a practicable alternative.
Access to the inside of the pump casings is limited by physical size (24 inch outside diameter), the pump shaft,, and the pump shaft support obstructions.
^
l Disassembly of the pump and removal of the pump shaft would cause a burden without a compensating increase in quality and safety.
j IV. Proposed Alternate Examination:
()
A surface examination of the accessible portions of the I
circumferential and longitudinal welds will be performed to
)
the extent and frequency described in Table IWC-2500-1.
A J
remote visual' examination (VT-1) of the I.D.
of the pump casing welds will be performed only if the pump is disassembled for maintenance, and the pump shaft is removed.
I The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised.
1 i
N1-002 Rev. 0 5/99 2-12
Virginia Electric and Power Company
/
North Anna Power Station Unit 1 k--)/
Third Ten Year Interval RELIEF REQUEST NDE-3 I.
Identification of Components:
Ultrasonic calibration blocks.
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Appendix I specifies requirements for ultrasonic calibration blocks for vessels, piping, bolts and studs.
Specifically, calibration blocks must meet the requirements of ASME Section V (vessels > 2" thick) or ASME Section XI, Appendix III (piping and vessels s 2") as supplemented by Table I-2000-1.
The existing calibration blocks at North Anna Unit 1 were designed and fabricated before the guide-
-lines of ASME Section XI were developed and approved.
The blocks for piping and vessels s 2" do not meet the
[~')
recommended design specified by Figure III-3230-2 for V
-thicknesses < 1" in that the notches are not staggered.
Also, the notches in some of the piping blocks are located one (1)
"t" from the end of the block instead of 1%" as specified by Figure III-3230-1. The vessel calibration blocks used for the reactor vessel head-to-flange weld, steam generator primary side tubesheet-to-head weld, and pressurizer welds are partially clad instead of fully clad as shown by Figure T-441.1 of ASME Section V, Article 4.
III. Basis for Relief:
North Anna Power Station was constructed prior to the issuance and adoption of ASME Section XI.
Therefore, ultrasonic calibration blocks were fabricated before the guidelines of ASME Section XI were developed and approved.
Meeting the requirements of the 1989 Code would require us to manufacture new calibration blocks.
The fabrication of new ultrasonic calibration blocks would'cause a burden without a compensating. increase in quality and safety.
Satisfactory ultrasonic system calibration can be performed I
with the existing calibration blocks.
Use of the existing calibration blocks also allows correlation of ultrasonic data from previous interval examinations as required by IWA-1400(h). The location of the notches in the piping O
N1-002 Rev. 0 5/99 2-13
r i
1 I
l RELIEF' REQUEST NDE-3 CONTINUED calibration blocks provides adequate signal separation for sweep calibration. Distance-amplitude calibration down to the clad-to-base metal interface, as delineated by Nonmandatory Appendix B to Section V, Article 4, can be performed from the unclad portion of the clad side of the existing vessel calibration blocks.
IV. Proposed Alternate Requirements:
The existing ultrasonic calibration blocks will be used for the third inspection interval examinations in lieu of current code requirements.
In addition, Code Case N-461, Alternative Rules for Piping Calibration Block Thickness, will be used as necessary.
The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.
("N I
l l
'L,10-002 Rev. 0 5/99 2-14
Virginia Electric and Power Company
()N North Anna Power Station Unit 1
\\_
Third Ten Year Interval RELIEF REQUEST NDE-4 I.
Icentification of Components:
ISI Class 1 and 2 piping, vessel, and component welds.
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWA-2600, weld reference system.
III. Basis for Relief:
The original construction code used at North Anna Power Station, ANSI B31.7, 1969 Edition, did not establish a weld reference system.
Immediate establishment of a weld reference system cannot be practically attained within the scope and schedule of existing outages.
To attempt to do so would cause a burden without a compensating increase in m
quality and safety.
IV. Proposed Alternate Requirements:
North Anna Unit 1 updated its weld isometrics during interval two. These isometrics provide a detailed identification of weld location.
Additionally, welds examined during the second interval were marked to indicate a zero point and direction of examination per second interval Relief Request NDE-13.
Second interval Relief Request NDE-13 was approved by NRC Letter No.92-255, Dated i
l 4/7/92, with the condition that each Class 1.and 2 piping l
weld examined receives all of the required reference markings as the inservice examination are performed. It is our intention to continue to use the markings made in the second interval, and to continue to mark welds in the third interval in an identical manner.
Welds which contain recordable indications shall be appropriately marked to j
ensure future location of the indication.
The reference system and marks will be permanently fixed on the weld.
The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.
\\
N1-002 Rev. 0 5/99 2-15
Virginia Electric and P,ower Company O
North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-5 I.
Identification of Components:
l Pressure Retaining Welds in the Reactor Vessel and Vessel Nozzle area examined by the automated vessel tool inspection device.
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWA-2600, weld reference system.
III. Basis for Relief:
The automated tool establishes its reference point using an existing zero reference in the reactor vessel.
This point allows the device to repeat examination locations without-the necessity of any other reference systems.
It accomplishes this by the use of an electronic encoder l
system, which provides for sufficient repeatability.
IV. Proposed Alternate Requirements:
The automated vessel tool examinations will continue to establish its reference system based upon the existing zero reference.
No other system is planned or deemed necessary.
The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.
I I
N1-002 Rev. 0 5/99 2-16 j
Virginia Electric and Power Company North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-6 I.
Identification of Components:
Pressurizer (1-RC-E-2) nozzle-to-vessel weld 9 and nozzle inner radius section 9NIR.
These welds are shown on drawing 11715-WMKS-RC-E-2.
II. Code Requirements:
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, requires in Category B-D, (Item Nos. B3.110 and B3.120) that the pressurizer surge line nozzle-to-vessel weld and nozzle inside radius section be volumetrically examined.
III. Basis for Relief:
The North Anna Unit 1 pressurizer surge line nozzle is
(~
surrounded by 78 heater penetrations.
Engineering-recommends that the heater cables be disconnected prior to the removal of insulation.
This recommendation is due to the possibility of damage to the heater element connections if the insulation is removed while the cables are connected.
Most of the heater connections are soldered in place.
Removal of these connections equires removal of.the brazing using a jewelers torch.
Based upon a previous survey of the applicable area, the dose rate is 150MR to 800MR in the general area and 300 to 2000 MR contact.
Based upon estimates provided by site Electrical Maintenance, Insulation Removal, and ISI/NDE, it would require six man hours to disconnect and reconnect the heater cables, four man hours to remove and reinstall the reflective insulation and seven man hours to prepare and examine the nozzle-to-vessel weld and nozzle inside radius section.
The resulting dose estimate for these examinations is 8.5 man-Rem.
Shielding is considered impractical because the shielding material would preclude accessibility to the examination surface.
Based upon a review of the fabrication drawings, the estimated percentage of the required volume that could be N1-002 Rev. 0 5/99 2-17
RELIEF REQUEST NDE-6 CONTINUED examined on the pressurizer surge 1ine nozzle-to-vessel weld 9 is as follows:
EXAMINATION ANGLE PERCENTAGE EXAMINED 45 Degrees 60%
60 Degrees 40%
0 Degrees 80%
l The examination coverage of the nozzle inside radius section 9NIR would be somewhat larger, however we feel that the confined access to the nozzle as a result of the pressurizer skirt, surge line piping and heater penetrations, and area dose rates would result in only a "best effort" examination in either case.
To attempt to perform the examination, which would be a l
" limited best effort" would cause a burden without a compensating increase in quality and safety.
A.similar relief was granted for use during interval two by NRC Letter No'.92-255, Dated 4/7/92.
IV. Proposed Alte*a
- e Requirements:
A visual (VT-2) aamination of the pressurizer surge line nozzle-to-vessel weld will be performed during the normally scheduled system leakage test each refueling.
In addition:
1.
Technical Specifications require that the Reactor Coolant System Leak Rate be limited to one gallon per minute unidentified leakage.
This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and 2.
The containment atmosphere particulate radioactivity is checked every.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The proposed alternatives stated above will ensure that the overall level of plant quality and safety will not be compromised.
O N1-002 Rev, 0 5/99 2-18
Virginia Electric and Power Company O
North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-7 I.
IDENTIFICATION OF COMPONENTS System:
Chemical and Volume Control (CH)
Component:
Regenerative Heat Exchanger (1-CH-E-3)
Connecting Circumferential Piping Welds Welds / Components Description Code Item #
l Class 3
tubesheet-to-shell B2.80 1
7 tubesheet-to-shell B2.80 1
11 tubesheet-to-shell B2.80 1
4 circumferential head B2.51 1
l 8
circumferential head B2.51 1
l 12 circumferential head B2.51 1
13 nozzle-to-vessel B3.150 1
l 14 nozzle-to-vessel B3.150 1
15 nozzle-to-vessel B3.150 1
gg 16 nozzle-to-vessel B3.150 1
t 1
17 nozzle-to-vessel B3.150 1
18 nozzle-to-vessel B3.150 1
i 13NIR nozzle inside radius B3.160 1
14NIR nozzle inside radius B3.160 1
15NIR nozzle inside radius B3.160 1
16NTR nozzle inside radius B3.160 1
d 17NIR nozzle inside radius B3.160 1
18NIE nozzle inside radius B3.160 1
WS-1 welded attachment B10.10 1
WS-2 welded attachment B10.10 1
WS-3 welded attachment B10.10 1
1 circumferential head C1.20 2
5 circumferential head C1.20 2
9 circumferential head C1.20 2
2 tubesheet-to-shell C1.30 2
6 tubesheet-to-shell C1.30 2
10 tubesheet-to-shell C1.30 2
II. CODE REQUIREMENTS j
Examination Categories B-B, B-D, B-K, and C-A require that volumetric and surface examinations be performed as indicated by the Code item numbers above.
U N1-002 Rev. 0 5/99 2-19
RELIEF REQUEST NDE-7 O
CONTINUED
%J III. BASIS FOR RELIEF The regenerative heat exchanger (1-CH-E-3) provides preheat for the normal charging water going into the reactor coolant system (RCS).
The preheat is derived from normal letdown water coming from the RCS.
Charging and letdown constitute the normal chemical and volume control within the RCS.
The heat exchanger itself is actually three heat exchangers in series, interconnected with piping.
This fact was previously utilized in limiting examinations to one.of the heat exchangers as allowed by the Code.
The heat exchanger has an outside shell diameter of 9.55 inches.
The shells were manufactured with ASTM A351 CF8 type material.
The heads were manufactured with ASTM A240 TP304 material.
The 3 inch nozzle i
necks were manufactured with ASTM A182 F304 material.
In the I
past the regenerative heat exchanger was entirely classified ASME Class 2 for inservice inspection activities.
However, a l
reanalysis changed the classification of the letdown side of the heat exchanger to ASME Class 1.
This action significantly increases the examination requirements associated with this heat exchanger.
Nozzles, which were previously exempt under Class 2 requirements are now required to be examined.
Additionally, all-Class 1 nozzles are required to be examined, 1
and the examinations are not limited to one heat exchanger.
\\-
The nozzle-to-vessel welds and nozzle inside radius sections for this vessel were not designed for ultrasonic examination from the outside diameter of the vessel.
The small diameter of the vessel and nozzles along with the cast stainless steel vessel shell prevents a meaningful ultrasonic examination of these components.
The Code required volumetric examinatica on the vessel head circumferential welds is limited due to the weld crown, radius of the closure caps, and the nozzles.
The Code required volumetric examination of the tubesheet welds is limited by the weld crown and is obstructed by a support clamp.
This clamp must be mechanically removed prior to the welds' examination. Additionally weld 11 is partially obscured by the six integral attachments, which are themselves butted up against a clamp. It is estimated that between 21 and 42 percent of the circumferential welds could be examined, and 42 percent of the tubesheet welds could be examined, if the clamps _are removed.
Weld 11 would be significantly less due I.
to the integral attachment location.
Previous partial l
examinations completed on these welds have identified no problems.
O N1-002 Rev. 0 5/99 2-20 i!.
I RELIEF REQUEST NDE-7
[)
CONTINUED j
%_/
An ALARA evaluation has been conducted on each activity associated with these examinations.
A table is provided j
documenting these results.
It is estimated that more than 14
)
man-rem will be required to complete these examinations over the interval.
This estimate assumes optimum inspection and preparation times.
If difficulties are encountered, a j
corresponding increase in dose would be expected.
Shielding is not considered practical since the source of radiation is j
the component receiving the examinations.
Considering the examination limitations previously discussed, expending this much dose is deemed impractical.
This relief request was originally submitted for the second interval by letter Serial No.93-018, dated February 16, 1993.
The relief was granted with conditions by NRC Letter No.96-525 Dated 10/01/96 and its associated safety evaluation report.
IV. ALTERNATE REQUIREMENTS Technical Specifications require that the RCE Leak Rate be limited to 1 gallon per minute unidentified leakage.
This value is calculated every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in accordance with rs Technical Specification requirements. Additionally, the
( )
containment atmosphere particulate radioactivity is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Technical Specification requirements.
As a result, new leakage is rapidly identified and located during operation. Leakage identified from these components can be easily isolated by two upstream valves with manual operation from within the control room.
The valves also receive an automatic control signal to close on inventory loss based on pressurizer level.
However, these valves could not be used as the Class 1 boundary valves due to their nonsafety-related actuation.
Correspondingly, as a result of the reclassification to Class 1.
these components will receive a system leakage test prior to start up after each refueling outage.
During this system leakage test the components will receive a visual (VT-2) examination.
The support structures will receive a
visual (VT-3) examination to the extent required by the Code without insulation removal.
Your evaluation of our original relief request, dated August 7,
- 1995, added alternative requirements.
The appropriate portion of the Technical Evaluation Report follows:
i N1-002 Rev. 0 5/99 2-21
I I
RELIEF REQUEST JNDE-7
()
CONTINUED Based on the statement by the licensee that previous partial examinations have been completed on these welds, it is concluded that a best effort volumetric examination of the lower RHX, in addition to system radiation monitoring and the Code required visual examinations, would provide a reasonable assurance of the system's inservice structural integrity.
I Therefore, pursuant to 10 CFR 50.55a (g) (6) (i), it is recommended that relief be granted provided the lower RHX vessel receives Code-rsquired volumetric examinations to the extent possible, l
As such, Code required volumetric examinations will be performed on the lower RHX vessel to the extent possible by removing the insulation necessary to examine the welds in the as found arrangement.
l l
l l
l l
N1-002 Rev. 0 5/99 2-22
RELIEF REQUEST NDE-7
[h CONTINUED V)
NORTH ANNA UNIT 1 1-CH-E-3 EXAMINATIONS MAN-REM ESTIMATE Work Task Man-Hours (hrs)
Dose Rate (R/hr)
Man-Rem Insulation Remove / Install 5.3 0.500 2.650 t
Scaffolding Install / Remove 2.0 0.300 0.600 Clamp l
Remove / Install 2.0 1.000 2.000 l
Weld Prep.
1.0 0.650 0.650 HP Coverage 6.25 0.020 0.125 Nozzle-to-Vessel Inspection (UT) 4.0 0.850 3.400 Nozzle Inside Radius jeg Inspection (UT) 3.0 0.850 2.550 U
Circumferential/Tubesheet l
Inspection (UT) 2.0 0.850 1.700 s
Welded Attachment Inspection (PT) 0.5 0.850 0.425 Total Estimate 26.05 14.100
=
l t
I O
I N1-002 Rev. O 5/99 i
{
2-23 l
RELIEF REQUEST NDE-7
(~)J CONTINUED R-6 Terminal End Welds (Lower Heat Exchanger Only)
Work Task Man-Hours (hrs)
Dose Rate (R/hr)
Man-Rem Insulation Remove / Install 1.80 0.500 0.900 l
]
Install / Remove 0.00 0.300 0.000 Clamp Remove / Install 2.00 1.000 2.000 Weld Prep.
0.33 0.650 0.215 HP Coverage 2.25 0.020 0.045 Nozzle-to-Vessel Inspection (UT) 1.20 0.850 1.020 Nozzle Inside Radius Inspection (UT) 0.90 0.850 0.765 i
Circumferential/Tubesheet Inspection (UT) 0.60 0.850 0.510 Welded Attachment Inspection (PT) 0.15 0.850 0.128 Total Estimate 9.23 5.582
=
fs t
)
's /
N1-002 Rev. 0 5/99 2-24
r Virginia Electric and Power Company North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST NDE-8 I.
Identification Of Components Branch connection welds - Main steam relief headers.
Drawing No.
Weld Nos.
11715-WMKS-0101A-1 SW-52 to SW-56 11715-WMKS-0101A-2 SW-15 to SW-17 and SW-40W to SW-41W 11715-WMKS-0101A-3 SW-32W to SW-35W and SW-18W See Branch-reinforcing pad detail and Section B-B of each
)
- drawing, j
l II.
Code Requirements y
Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Category C-F-2, Item No. C5.81 requires q
i circumferential pipe branch connection welds be examined using a surface method.
III. Basis For Relief The design of the main steam relief header branch connection
[
welds calls for the use of a reinforcement pad.
These pads l
are fillet welded and completely encase the branch i
connection welds.
l IV. Proposed Alternate Requirements i
l A surface examination of the reinforcement pad's fillet l
welds associated with one branch connection weld will be l-performed during the interval.
i f
O N1-002:
Rev. 0 5/99 2-25
RELIEF REQUEST NDE-9 CONTINUED uJ "IWA-5250 Corrective Measures (a)
The source of leakage detected during the conduct of a system pressure test shall be located and evaluated by the owner for corrective measures as follows:
(3) repairs or replacements of components shall be performed in accordance with IWA-4000 or IWA-7000, respectively."
Articles IWA-4000 and IWD-4000 of ASME Section XI Code repair requirements would require removal of the flaw and subsequent weld repair.
Code repairs for through-wall leaks require the line to be j
isolated and drained.
Taking a train of service water out of service in some instances is a major evolution and requires entering a Technical Specification action statement.
Welds and piping with through-wall flaws caused by MIC can typically be shown to have adequate structural integrity to remain in service.
This type of through-wall flaw is unpredictable but not normally subject to catastrophic failure.
It is impractical to require a Code rs repair within the time specified by the Technical
()
Specification Limiting Condition for Operation every time a through-wall flaw is identified.
Generic Letter 90-05, " Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2,
and 3 Piping",
provides guidance for submitting relief requests to allow continued operation with a through-wall flaw.
Submitting a relief request for each instance of through-wall leakage caused by MIC is an administrative burden and causes additional unnecessary reviews for the NRC.
Implementing GL 90-05 each time a through-wall flaw is identified is also impractical.
This relief request (NDE-9, Revision 0) establishes a plan for continued operation with through-wall flaws in stainless steel piping in the Service Water System based upon the guidance of GL 90-05 to the extent it is believed practical.
III. Basis For Relief This relief request is submitted in a format laid out in NRC GL 90-05.
The following information and justification is provided in accordance with the guidelines of Part B and C of Enclosure 1 to GL 90-05.
A N1-002 Rev. 0 5/99 2-27 t
RELIEF REQUEST NDE-9
~
CONTINUED Scope, Limitations and Specific Considerations Scope The scope consists of welds and stainless steel piping, pipe class 153A and 163, with evidence of possible through-wall leaks in the Service Water System at North Anna Power Station Units 1 and 2.
Limitations Based on radiographic examinations and laboratory examinations of removed portions of piping from replacements, North Anna Power Station is experiencing MIC in its stainless steel piping.
The MIC-caused flaws originate on the inner diameter of the pipe.
The Service Water System is common to both Units. As long as one Unit is in Mode 1, 2,
3, or 4 both trains of service water must be operable.
If both Units are in Mode 5 or 6 then one train of Service Water must be operable.
The intent of this relief is to permit continued operation with the identified through-wall flaws until repairs are accomplished in a scheduled service water outage.
Specific Considerations I
System interactions, i.e. corcequences of flooding and
'~
spray, will be eval"ated.
Piping with through-wall leakage that could affect plant safety related equipment will be declared inoperable and the appropriate Technical Specification action statement entered.
Welds containing evidence of through wall flaws will be either repaired or assessed for structural integrity within 14 days of detection.
Butt welds and piping remaining in j
service after leakage is detected for longer than 14 days and accessible to volumetric examination by Ultrasonic (UT) or Radiographic (RT) examination methods, will be evaluated i
for structural integrity for all design loading conditions, including dead weight, pressure, thermal expansion and seismic (DBE) loads.
The methods used in the structural integrity analysis will consist of area reinforcement, fracture mechanics, and limit load analysis. These methods are detailed in Attachment 1.
The welds that are found to be unacceptable will be declared inoperable and the appropriate Technical Specification action statement entered.
A 3/4" hole will be postulated for any location with a through-wall flaw that can not be characterized volumetrically by ultrasonics (UT) or radiography (RT),
N1-002 Rev. 0 5/99 2-28
RELIEF REQUEST FDE-9
~N CONTINUED
)
(d socket welds and welds that are inaccessible for UT or RT.
Laboratory examination of cut sections of MIC degraded socket weld samples indicate that flaws are enveloped within the 3/4" size. A leaking socket weld location will be analyzed by treating the cross section as equivalent to the cross section of the attached pipe with a 3/4" hole.
The methods used in the structural integrity analysis will consist of area reinforcement, fracture mechanics, and limit load analyses.
These methods are detailed in Attachment 1.
A through-wall flaw size is postulated in order to perform a structural analysis.
Additional monitoring is performed for a period of twa (2) months on socket welds to assure the degradation mechanism is behaving in a manner expected for a MIC flaw.
The structural integrity for all welds identified with evidence of through-wall leakage will be monitored by the following methods:
Weekly visual monitoring of through-wall flaws from the time of identification until repair of the weld or completion of the structural integrity analysis.
Repair jN of the weld or completion of the structural analysis is
'~
(
required within 14 days of detecting a through wall flaw.
If the welds are determined to be structurally acceptable, then the visual monitoring frequency will be decreased to once every six weeks.
Weekly visual monitoring of through-wall flaws in socket welds and butt welds inaccessible to RT or UT examination for a period of two (2) months.
If there is no significant change in the leakage rate, the monitoring frequency will be decreased to once every six weeks until the welds are repaired.
A significant change is defined as a 0.5 gpm increase in the leakage rate from the initial observed leakage condition for each weld.
Should any location reach the threshold of a significant change the weld will be reassessed for structural integrity and flood / spraying consequences.
A total leakage rate limit of 1.0 gpm for a single supply or return line to an individual component will be established. If this total leakage rate limit is exceeded then an evaluation will be performed to determine if design service water flow is available to l
affected components.
Inadequate service water supply will cause the associated service water lines and fg equipment to be declared inoperable and appropriate
(_)
l N1-002 Rev. 0 5/99 2-29
l RELIEF REQUEST NDE-9 CONTINUED
)
action will be taken according to Technical Specifications.
l The temporary non-code repair will be to leave the welds as they are found, subject to monitoring and meeting the criteria for consequences and for structural integrity as described above.
Evaluation Flaw Detection During Plant Operation and Impracticality Determination The Service Water System is a common system for both Units at North Anna Power Station. Both trains are required to be operational or the appropriate Technical Specification action statement is entered.
The through-wall flaws on Service Water lines in service are anticipated based on the North Anna Power Station history of MIC.
Virginia Electric and i
Power Company requests to evaluate the flaws and leave acceptable through-wall flaws in service in order to perform Code repairs in controlled conditions during scheduled service water outages.
)
Flaw characterization of MIC has traditionally been performed by radiography and has wide acceptance by the industry as an appropriate technique.
Virginia Power has also developed an ultrasonic technique, which is as effective as radiography for detection and length sizing of MIC indications.
This ultrasonic technique additionally has l
depth sizing capabilities.
The acceptability of ultrasonics as an alternative to radiography is documented in a report titled " Ultrasonic Examination Technique for Detection of Microbiologically Induced Corrosion in North Anna Stainless Steel Service Water Piping" dated December 29, 1997, reference Virginia Power Letter No. 98-037A, dated 11/12/98.
Virginia Power proposes to use either radiography or ultrasonics or both to characterize flaws.
Root Cause Determination and Flaw Characterization The Service Water System at North Anna Power Station has previously experienced MIC.
Radiograph examinations of L
service water welds having evidence of through-wall leakage I
revealed small voids surrounded by exfoliation, which is l
typical of MIC. No other type of inservice defects were l
identified by the radiographs near the areas with through-wall leaks.
Additionally, a visual examination performed by a Virginia Electric and Power Company staff f}.
V N1-002 Rev. 0 5/99 2-30
RELIEF REQUEST NDE-9
[~%
CONTINUED metallurgist of a sample of piping segments removed to repair the leaking welds confirmed the presence of MIC.
Flaw Evaluation Flaw evaluation for welds with through-wall leakage will be performed as described in Attachment 1.
The flaws in butt welds that can be characterized by RT or UT examination j
methods will be evaluated by three types of analyses, area reinforcement, limit load analyses, and fracture mechanics using the guidance from NRC Generic Letter 90-05.
The flaws in welds that cannot be characterized by RT or UT examination, i.e.,
socket welds and inaccessible butt welds,
)
will be evaluated by the same analysis by assuming a 3/4" hole for each point of leakage within a weld with through-wall leakage.
Socket welds will be analyzed by treating the cross section at the socket weld as equivalent pipe cross section.
This relief request was previously approved for use at North Anna Units 1 and 2 in the Second ISI Interval, reference NRC Letter No.98-720, dated 12/3/98.
/~g V.
Augmented Inspection An augmented inspection program will monitor a sample of ten (10) butt welds in the Service Water System using RT or UT examination methods.
RT or UT examination methods will be performed every three (3) months.
The frequency of radiography will be assessed periodically and may be adjusted for each location based on the results of the radiographs.
VI.
Alternate Provisions As an alternative to performing Code repairs in accordance with IWA-5250 (a) (2) to through-wall flaws in the Service Water System the through-wall flaws will be left as is.
The through-wall flaws will be monitored for leakage and must meet the criteria for flooding and spray.ng consequences.
Structural integrity must be determined as described herein to remain in service beyond 14 days from detection. If structural integrity is determined, operation in this mode will continue until the subject welds are replaced.
All welds identified with through-wall flaws will be replaced within 18 months from the time of discovery.
i The structural integrity of the Service Water System will be monitored'by the following methods until the repairs required by IWA 5250 (a) (2) are completed.
1 O
v N1-002 Rev. 0 5/99 2-31
RELIEF REQUEST NDE-9
[
CONTINUED Weekly visual monitoring c through-wall flaws from the time of identification unt.
repair or completion of Structural integrity analysis.
If the welds are determined to be structurally acceptable then the visual monitoring frequency will be decreased to once every six weeks.
All welds identified as having through-wall flaws and I
not repaired within 14 days of detection,-will be i
assessed for structural integrity within 14 days of detection.
Butt welds will be examined by RT or UT methods, if accessible, to characterize the flaws.
l Socket welds and butt welds inaccessible for RT or UT examinations will be assessed for structural integrity l
by assuming a conservative %" diameter hole.
Welds I
determined to be structurally adequate will be included in the above monitoring program.
Identification of a l
structurally inadequate weld will result in the associated piping to be declared inoperable and the appropriate Technical Specification action statement to be taken.
Weekly visual monitoring of through-wall flaws in socket s.
welds and butt welds inaccessible to RT or UT examination for a period of two (2) months.
If there is no significant change in the leakage rate, the monitoring' frequency will be decreased to once every six weeks until the welds are repaired.
A significant change is defined as a 0.5 gpm increase in the leakage rate from the initial observed leakage condition for each weld.
Should any location reach the threshold of a significant change, the weld will be reassessed for structural integrity and flood / spraying consequences A walkdown of the accessible stainless steel portions of the Service Water System will be performed every six weeks. This inspection frequency is based on a 12 week Maintenance Rule schedule, which reduces the time equipment is out of service for testing and maintenance.
Two one week windows, six weeks apart, are available within this schedule to perform service water repairs.
Performing inspections within this schedule will allow repairs to be made in a timely manner and reduce the time safety related equipment is out of service. The frequency of the walkdowns will be assessed after a year and adjusted based on the results of the inspections.
Ab N1-002 Rev, 0 l
5/99 2-32
RELIEF REQUEST NDE-9 CONTINUED l
A total leakage rate-limit of 1.0 gpm for a single supply or return line for an individual component will be established.
If this total leakage rate limit is exceeded then an evaluati.on will be performed to l
determine if design service water flow is available to i
affected components.
Inadequate service water supply will cause the associated service water lines and equipment to be declared inoperable and appropriate action will be taken according to Technical l
Specifications.
i An augmented inspection program will monitor a sample of l
ten (10) butt welds in the Service Water System using RT or UT examination methods.
Examinations will be performed every three (3) months.
The frequency of radiography will be assessed'after a year and may be adjusted for each location based on the results of the radiographs.
l The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.
Attachments:
- 1. Flaw Evaluation Methods and Results
- 2. Drawing /Line Number Designations 1
References:
- 1. USAS B31.1 Power Piping 1967 Edition 1
2.
EPRI Report NP-6301-D, " Ductile Fracture Handbook"
- 3. Nuclear Regulatory Commission Generic Letter 90-05 " Guidance for Performing Temporary Non-Code Repair of ASME Code Class 1, 2,
and 3 Piping I O
. V N1-002 Rev. 0 5/99 2-33
RELIEF REQUEST NDE-9
(
CONTINUED
\\
Flaw Evaluation Methods and Results Introduction Butt welds identified as having possible through-wall leaks will be volumetrically examined by radiographic (RT) or ultrasonic (UT) methods, if accessible.
Flaws in butt welde that are inaccessible for examination will be postulated as a 3/4" hole for each area identified with a through-wall flaw.
All butt welds will be analyzed for structural integrity by three methods, area reinforcement, limit load analysis, and linear elastic fracture mechanics evaluation.
Flaw size in socket welds identified as having possible through-wall.
leaks cannot be characterized by.
nondestructive examination. A conservatively large hole, 3/4", will be postulated for each area identified with a through-wall flaw.-
The postulated flaw will be analyzed for structural integrity by treating the cross section as equivalent to the cross section of the attached pipe.
Area Reinforcement Analysis tO The area reinforcement analysis is used to determine if adequate reinforcing exists such that ductile' tearing would not occur.
The guidelines of ANSI B31.1 paragraph 104.3.(d) 2 (reference 1) are used to determine the Code required reinforcing area.
The actual reinforcing area is calculated and is checked against the required reinforcement area.
The Code required reinforcement area in square inches is defined as:
- 1. 07 (tm) (dl) l Where tm is the code minimum wall, and dl is the outside I
diameter.
i The Code reinforcement area required is provided by the available material around the flaw in the reinforcing zone.
Limit Load Analysis The structural integrity of the piping in the degraded
.O condition will be established by calculating the minimum V
N1-002 Rev. 0 5/99 2-34
4 O'
RELIEF REQUEST
'DE-9 CONTINUED margin of safety based upon a Limit Load Analysis.
These methods are documented in EPRI report NP-6301-D (Ductile Fracture Handbook) (reference 2).
MR = 4 MY' + MZ' + T*
The limit load analysis of the postt _ aced flawed sections will be performed with a material flow stress representing the midpoint of
~
the ultimate strength and yield point stress.
The flawed sections will be subjected to deadweight, thermal, and seismic DBE loading.
The allowable limit load is given by, Ma = 2 sf Rm2 t-(2cos(b)-sin (Q)) in-lbf sf
= flow stress = 0.5 (Sy + Su) psi Sy = yield stress psi 1
Su = ultimate stress psi Rm
= mean radius of the pipe f = e D e( R,' e P}+F
+
2 4
- af
- R. e t Ri
= internal radius of the pipe P
= pressure psig F
= axial load in lbs t
= pipe thickness = inches D
= Outside diameter inches e = half angle of the crack (radians) = crack length 2 Rm MR = Resultant Moment NY = Bending Moment MZ = Bending Moment T
Torsion The calculated factor of safety is, FS
Ma (MR)
The minimum factor of safety of 1.4 is required to be qualified for continued operation.
O N1-002 Rev. 0 5/99 2-35
_/~']
RELIEF REQ'. TEST NDE-9
\\m)
CONTINUED Fracture Mechanics Evaluation A
linear elastic fracture mechanics analysis will be performed for circumferential through-wall crack using the guidance provided in NRC Generic Letter 90-05.
The structural integrity of the piping in the degraded condition was established by calculating the minimum margin of safety -based upon a
Fracture Mechanics evaluation.
This method is documented in EPRI report NP-6301-D (Ductile Fracture Handbook) (reference 2).
A through-wall circumferential crack will be postulated for every area containing MIC. The cracks will be subjected to a design pressure loading in addition to the deadweight, normal operating thermal and seismic DBE loadings.
For the purpose of this evaluation a generic allowable stress intensity factor of KIC = 135 ksi (in.) ** will be used for the stainless steel material per NRC GL 90-05.
The applied stress intensity factor for bending, KIB, is found by:
KIB = [sb-(p Rm-Q)0.5] Fb The applied stress intensity factor for internal pressure, KIP, is found by:
KIP = sm-(p Rm-Q)0.5 Fm The applied stress intensity factor for axial tension, KIT is found by:
i KIT = st-(p Rm-Q)0.5 Ft i
The stress intensity factor for residual stresses, KIR is found by:
KIR = S-(p Rm-Q)0.5 Ft Total applied stress intensity KT includes a 1.4 safety factor and is calculated by:
KT = 1.4-(KIB + KIP + KIT) + KIR The allowable stress intensity factor is taken from Generic Letter 90-05.
/~'
KALL = 135 ksiGin us l
N1-002 Rev. 0 5/99
)
2-36
RELIEF REQUEST NDE-9 g
CONTINUED Stress Intensity Factor Ratio is defined as:
SR = KT l
KALL The stress intensity factor ratio shall be less than 1.0 for continued operation.
O O
N1-002 Rev. 0 5/99 2-37 l
RELIEF REQUEST NDE-9
(
CONTINUED ATTACHMENT 2 List of drawings containing the Service Water System and the associated line number designations for pipe class 153A and 163.
DRAWING 11715-CBB-040D-3, SHEET 1 1-1/4"-WS-G37-153A-Q3 4"-WS-G01-163-Q3
)
3/4"-WS-F83-163-Q3 4"-WS-F99-163-Q3 3/4"-WS-F82-163-Q3
' AWING 11715-CBB-040D-3, SHEET 2 m
1-1/4"-WS-H78-153A-Q3 1-1/4"-WS-H81-153A-03 DRAWING 11715-CBM-078A-3, SHEET 1 3/4"-WS-G31-163-Q3 3/4"-WS-F67-163-Q3 3/4"-WS-H77-163-Q3 4"-WS-F64-163-Q3 4"-WS-H76-163-Q3 3/4"-WS-F80-163-03 3/4"-WS-G30-163-03 3/4"-WS-F68-163-Q3 3/4"-WS-339-163-03 4"-WS-F65-163-Q3 f-'g 3/4"-WS-G36-163-Q3 3/4"-WS-F69-163-03
( f 3/4"-WS-930-163-Q3 8"-WS-93-163-Q3 1/2"-WS-F61-163-03 8"-WS-94-163-03 2"-WS-D72-163-Q3 4"-WS-F62-163-03 3/4"-WS-F78-163-Q3 3/4"-WS-F66-163-Q3
'4"-WS-F63-163-Q3 3/4"-WS-F79-163-03 4"-WS-G35-163-03 3/4"-WS-F81-163-Q3 DRAWING 11715-CBM-078A-3, SHEET 4 8"-WS-515-163-03 8"-WS-516-163'-03 8"-WS-513-163-03 8"-WS-514-163-03 4"-WS-H48-163-Q3 3/4"-WS-F64-163-03 3/4"-WS-H52-163-Q3 3/4"-WS-H53-163-Q3 4"-WS-H50-163-Q3 3/4"-WS-H66-163-03 3/4"-WS-H54-163-Q3 4"-WS-H51-163-Q3 3/4"-WS-H67-163-Q3 3/4"-WS-H55-163-Q3 3/4"-WS-G32-163-03 3/4"-WS-G33-163-03 8"-WS-115-163-03 8"-WS-116-163-Q3 8"-WS-113-163-Q3 4"-WS-46-163-Q3 4"-WS-47-163-Q3 2"-WS-926-163-Q3 O
N1-002.
Rev, 0 5/99 2-38
RELIEF REQUEST NDE-9 CONTINUED DRAWING 11715-CBM-078A-3, SHEET 4, (Continued) 3/4"-WS-H60-163-Q3 3/4"-WS-H61-163-Q3
~
3/4"-WS-H62-163-Q3 3/4"-WS-H63-163-Q3 4"-WS-56-163-03 4"-WS-57-163-Q3 4"-WS-H49-163-Q3 3/4"-WS-H65-163-Q3 8"-WS-114-163-Q3 DRAWING 11715-CBM-078B-3, SHEET 1 1-1/2"-WS-346-163-Q3 1"-WS-347-163-Q3 1"-WS-G06-163-Q3 1-1/2"-WS-347-163-Q3 1/2"-WS-348-163-Q3 1"-WS-349-163-Q3 1"-WS-G05-163-03 1-1/2"-WS-349-163-Q3 1-1/2"-WS-350-163-C3 1"-WS-351-163-Q3 1"-WS-G04-1C;-03 1-1/2"-WS-351-163-Q3 1-1/2 "-WS-3 52-163 -Q3 1"-WS-353-163-Q3 1"-WS-G03-163-Q3 1-1/2"-WS-353-163-03 DRAWING 11715-CBM-078B-3, SHEET 3 1-1/2"-WS-724-163-03 1"-WS-725-163-Q3 1"-WS-933-163-Q3 1-1/2"-WS-725-163-03 1-1/2"-WS-'726-163-Q3 1"-WS-727-163-Q3 Os 1"-WS-934-163-Q3 1-1/2"-WS-728-163-03 1"-WS-729-163-Q3 1"-WS-935-163-03 1-1/2"-WS-729-163-Q3 1-1/2"-WS-730-163-Q3 1"-WS-731-163-Q3 1"-WS-936-163-Q3 1-1/2"-WS-731-163-Q3 1-1/2"-WS-727-163-Q3 DRAWING 11715-CBM-078C-3, SHEET 2 4"-WS-46-163-03 1"-WS-85-163-03 3/4"-WS-A47-163-03 1"-WS-485-163-Q3 1"-WS-81-163-Q3 1"-WS-487-163-03 1"-WS-77-163-Q3 1"-WS-489-163-03 1"-WS-477-163-Q3 1"-WS-491-163-Q3 1"-WS-481-163-03 3"-WS-75-163-Q3 3"-WS-73-163-03 3"-WS-76-163-03 4"-WS-47-163-03 3/4"-WS-79-163-Q3 3/4"-WS-A49-163-03 3/4"-WS-381-163-Q3 1"-WS-82-163-Q3 2"-WS-80-163-03 1"-WS-78-163-Q3 3/4"-WS-83-163-Q3 1"-WS-478-163-03 3/8"-WS-383-163-Q3 O
N1-002 Rev. 0 5/99 2-39
RELIEN REQ'UEST NDE-9
[
CONTINUED 11715-CBM-078C-3, SHEET 2 (Cont'd) 1"-WS-482-163-Q3 3/8"-WS-382-163-Q3 3"-WS-74-163-03 2"-WS-376-163-Q3 4"-WS-56-163-Q3 3/3"-WS-397-163-Q3 3/4"-WS-A48-163-03 3/8"-WS-398-163-03 1"-WS-90-163-Q3 2"-WS-377-163-03 3/4"-WS-C06-163-Q3 3/8"-WS-C01-163-03 1"-WS-92-163-03 3/8"-WS-399-163-03 1"-WS-86-163-Q3 3/4"-WS-378-163-03 3/4"-WS-C01-163-Q3 2"-WS-84-163-Q3 1"-WS-488-163-Q3 3/4"-WS-379-163-03 l
3/4"-WS-913-163-Q3 1"-WS-88-163-Q3 l
1"-WS-490-163-Q3 3/4"-WS-380-163-Q3 1"-WS-78-163-03 3/4"-WS-83-163-Q3 1"-WS-492-163-Q3 3/8"-WS-383-163-03 4"-WS-57-163-Q3 3/4"-WS-A50-163-Q3 3/4"-WS-773-163-Q3 1"-WS-89-163-Q3 2"-WS-772-163-Q3 3/4"-WS-400 163-03 3/8"-WS-910-163-Q3 3/4"-WS-774-163-Q3 L
3/8"-WS-914-163-Q3 2'~-WS-775-163-03 2"-WS-776-163-Q3 3/8"-WS-913-163-03 3/8"-WS-916-163-Q3 3/8"-WS-915-163-Q3 2"-WS-777-163-Q3 3/4"-WS-779-163-03 O
1"-WS-486-163-Q3 3/4"-WS-909-163-03 3/8"-WS-912-163-03 DRAWING 11715-CBM-078G-3, SHEET 1 2"-WS-D46-153A-03 2"-WS-C88-153A-03 1"-WS-D50-153A-Q3 2"-WS-C81-153A-03 3/4"-WS-D39-153A-03 2"-WS-C87-153A-03 L
1"-WS-D30-153A-Q3 2"-WS-63-163-Q3 3/4"-WS-D41-153A-03 3/4"-WS-D61-163-Q3 3/4"-WS-D55-163-03 2"-WS-62-163-03 1"-WS-D31-153A-03 2"-WS-C80-153A-Q3 3/4"-WS-D43-153A-03 3/4"-WS-C67-153A-03 3"-WS-73-163-03 3/4"-WS-C73-153A-Q3 2"-WS-54-163-Q3 3/4"-WS-C76-153A-03 E
2"-WS-52-163-03 3/4"-WS-C70-153A-Q3 3/4"-WS-D67-163-03 2"-WS-C86-153A-03 2"-WS-50-163-03 2"-WS-79-153A-03 3"-WS-74-163-Q3 2"-WS-C85-153A-03 N1-002 Rev. 0 5/99 2 l
RELIEF REQUEST NDE-9
()
CONTINUED
-11715-CBM-078G-3, SHEET 1 (Continued) 2"-WS-55-163-Q3 3/4"-WS-D62-163-Q3 3/4"-WS-D54-163-Q3 2"-WS-60-163-Q3 2"-WS-53-163-03 3"-WS-75-163-03 3/4"-WS-D68-163-Q3 3/4"-WS-D69-163-Q3 2"-WS-51-163-03 2"-WS-61-163-03 3/4"-WS-D56-163-Q3 3/4"-WS-D70-163-03 2"-WS-C83-153A-Q3 3"-WS-76-163-Q3 2"-WS-C89-153A-Q3 2"-WS-C78-153A-Q3 3/4"-WS-D60-163-Q3 3/4"-WS-C66-153A-Q3 2"-WS-64-163-03 3/4"-WS-C72-153A-03 2"-WS-65-163-Q3 3/4"-WS-C75-153A-03 2"-WS-C82-153A-Q3 3/4"-WS-C69-153A-Q3 3/4"-WS-C68-153A-Q3 2"-WS-C84-153A-Q3 3/4"-WS-C74-153A-Q3 3/4"-WS-C77-153A-Q3 3/4"-WS-C7.1-153A-03 DRAWING 11715-CBM-078G-3, SHEET 2 2"-WS-D46-153A-03 3/4"-WS-D65-163-Q3 3/4"-WS-D71-153A-Q3 3/4"-WS-D66-163-03 1"-WS-D47-153A-Q3 3/4"-WS-934-153A-03 3/4"-WS-D33-153A-03 2"-WS-948-153A-03
(~)s
(
3/4"-WS-D51-163-Q3 3/4"-WS-940-153A-03 1"-WS-D48-153A-Q3 1-1/2"-WS-976-153A-Q3 3/4"-WS-D35-153A-03 1-1/2"-WS-982-153A'-03 3/4"-WS-D52-163-Q3 3/4"-WS-943-153A-Q3 l
1"-WS-D49-153A-Q3 3/4"-WS-937-153A-Q3 3/4"-WS-D37-153A-03 2"-WS-954-153A-03 3/4"-WS-D53-163-03 2"-WS-947-153A-03 3"-WS-73-163-03 1-1/2"-WS-977-153A-03 j
3/4"-WS-D63-163-03 1-1/2"-WS-963-153A-03 3"-WS-74-163-03 2"-WS-953-153A-Q3 2"-WS-454-163-Q3 3/4"-WS-D58-163-03 2"-WS-455-163-Q3 2"-WS-462-163-03 2"-WS-452-163-03 2"-WS-463-163-03 2"-WS-453-163-03 2"-WS-946-153A-Q3 2"-WS-450-163-Q3 1-1/2"-WS-979-153A-Q3 2"-WS-451-163-03 1-1/2"-WS-978-153A-Q3 2"-WS-949-153A-03 2"-WS-952-153A-03 1-1/2"-WS-981-153A-03 2"-WS-945-153A-03 1-1/2"-WS-965-153A-Q3 1-1/2"-WS-980-153A-Q3 2"-WS-955-153A-Q3 1-1/2"-WS-961-153A-Q3 1
l N1-002 Rev. O I
5/99 2-41 I
RELIEF REQUEST NDE-9
[V')
CONTINUED DRAWING 11715-CBM-078G-3, SHEET 2 (Continued) 3/4"-WS-D57-163-Q3 2"-WS-460-163-03 2"-WS-464-163-Q3 2"-WS-945-153A-03 2"-WS-465-163-03 2"-WS-944-153A-Q3 3"-WS-75-163-03 1-1/2"-WS-975-153A-Q3 3"-WS-76-163-03 1-1/2"-WS-971-153A-03 3/4"-WS-D59-163-Q3 2"-WS-950-153A-Q3 2"-WS-461-163-03 2"-WS-951-153A-03 DRAWING 11715-CBM-078H-3, SHEET 1 j
i 1"-WS-9-163 1"-WS-10-163 j
1 3/4"-WS-E03-163 3/4"-WS-F45-163 3/4"-WS-F44-163 2"-WS-E64-163 1"-WS-G07-163 1/2"-WS-E68-163 3/4"-WS-E08-163 1"-WS-D97-163 1"-WS-947-163 2"-WS-D91-163 1"-WS-G50-163 1"-WS-H48-163 1"-WS-G52-163
[G i
i
()h N1-002 Rev. 0 5/99 2-42
l Virginia Electric and Power Company North Anna Power Station Unit 1 Third. Ten Year Interval
\\
RELIEF REQUEST NDE-10 I.
Identification of Components l
Class 1, 2, and 3 Integrally Welded Attachments
)
II. Code RequirementsSection XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Tables IWB-2500-1, IWC-2500-1 and IWD-2500-1, l
Examination Categories B-H, B-K-1,-C-C, D-A, D-B, and D-C, examination of Integrally Welded Attachments.
l III. Basis for Relief Code Case N-509, Alternate Rules for the Selection and Examination of Class 1, 2,
and three Integrally Welded Attachments,Section XI, Division 1, is not currently approved by Regulatory Guide 1.147 for use.
10 CFR i
- 50. 55 (a) (3 ) footnote 6 notes that the use of other Code Cases may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to 10CFR e
50.55 (a) (3).
As such, Code Case N-509 is requested for use on North Anna Unit 1 in the second inspection interval.
i l
The current Code requires a certain size base material design thickness before examination is required on Code Class 1 or 2 integrally welded attachments.
This size limitation is apparently arbitrary.with no technical basis.
The current Code also has no inspection requirements for Class 1 integrally welded attachments for piping, pumps and valves (B-K-1) for the third inspection interval Inspection Program B (North Anna's).
Additionally, there is no selection criteria for Class 3 nonexempt integrally welded ar.tachments, requiring 100% examination.
These deficiencies have been corrected in Code Case N-509.
.This relief request was submitted and approved for the Second Inservice Inspection Interval at NAPS Unit 1.
The relief was granted with conditions per NRC Letter No.95-446 Dated 8/18/95. Relief was granted provided that at least 10% of all integral attachments were scheduled for examination.
IV. Proposed Alternate Requirements Code Case N-509 will be used in its entirety.
Further, at
()
least 10% of all' integral attachments will be scheduled for N1-002 Rev. 0 5/99 2-43
RELIEF REQUEST NDE-10 CONTINUED examination.
Code Case references to the 90 Addendum of ASME Section XI are the same as the provisions in Code Case N-491, which has been implemented by our support programs.
I U
O N1-002 Rev, 0 5/99 2-44
Virginia Electric and Power Company
[~h North Anna Power Station Unit 1
\\ s/
Third Ten Year Interval RELIEF REQUEST NDE-ll j
I.
Identification of Components Class 1 and Class 2 longitudinal piping welds.
II.
Code Requirements j
Examination of Class 1 and Class 2 longitudinal piping welds in accordance with ASME Section XI, Division 1, 1989 Edition, Table IWB-2500-1, Category B-J, Table IWC-2500-1, Categories C-F-1 and C-F-2.
1 III. Basis for Relief Section XI of the ASME Boiler and Pressure Vessel Code, 1989 1
Edition, currently requires one pipe diameter in length, but no more than 12 inches, be examined for Class 1 longitudinal piping welds.
Class 2 longitudinal piping welds are required to be examined for a length of 2.5t, where "t"
is j
the thickness of the weld.
These lengths of weld are
(~)g
(,
measured from the intersection of the circumferential v.Lld and longitudinal weld.
Code Case N-524 significantly I
reduces the required examination volume or surface area.
it does this by limiting the examination requirements of the intersecting circumferential weld. This would include the weld and one-half (1/2) inch on both sides of the weld crown for surface examinations, and one-quarter (1/4) inch on both sides of the weld crown for volumetric examinations in the lower one-third (1/3) weld volume.
Code Case N-524 directs the examination effort at the high risk area associated with the weld intersections.
It eliminates low risk areas on the longitudinal weld from examination, significantly reducing examination time and radiation exposure to examination personnel. Compliance with the existing ASME Section XI requirements, in lieu of the Code Case, results in unnecessary personnel radiation exposure to complete the required examinations without a compensating increase in the level of quality or safety.
Code Case N-524 was previously approved for use at North Anna for the second ten-year interval per NRC Letter 94-602, dated 10/03/94.
IV.
Proposed Alternate Requirements ID)
Code Case N-524 will be implemented in its entirety.
N1-002 Rev. 0 5/99 2-45
l Virginia Elect'ric and Power Company O
North Anna Power Station Unit 1 Third Ten Year Interval
)
RELIEF REQUEST NDE-12 l
I I.
Identification of Components
)
Snubbers and Safety / Relief Valves within the scope of ASME Section XI.
II.
Code Requirements Replacement of items shall meet the requirements of ASME Section XI, Division 1, 1989 Edition, Article IWA-7000.
III. Code Requirements from Which Relief Is Requested Relief is requested from meeting the requirements of ASME Section XI, Division 1, 1989 Edition, Article IWA-7000 when removing snubbers or safety / relief valves for testing.
IV.
Basis for Relief gs Currently, when a snubber or relief valve is removed for
( )
testing, the activity is required to be treated like an ASME
)
Section XI, Article IWA-7000 replacement activity.
This entails the use of replacement programs, review by an Authorized Nuclear Inservice Inspector, and preparation / maintenance of a NIS-2 form.
This activity 1
requires the degraded system to remain out of service for the time that the item is removed, tested and reinstalled.
The use of Code Case N-508-1, when a snubber or pressure relief valve is removed from a system for the purpose of testing, would allow a spare component to be put into place of the removed component, and to test the removed component at a later time.
This minimizes the time that a degraded system remains out of service.
The use of Code Case N-508-1 will not negate the existing snubber service life monitoring program.
The use of Code Case N-508-1 will provide an acceptable level of quality and safety, and assurance of snubber reliability will be maintained.
V.
Proposed Alternate Requirements Code Case N-508-1 will be implemented in its entirety.
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4
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N1-002 Rev. 0 5/99 2-46
l Virginia Electric and Power Company i
l
[~h North Anna Power Station Unit 1
(_)
Third Ten Year Interval RELIEF REQUEST NDE-13 I.
Identification of Components All components within the scope of the North Anna Unit 1 ASME Section XI ISI Program.
II.
Code Requirements ASME Section XI, Division 1, 1989 Edition, IWA-2430(d), for components inspected under Program B (IWA-2432), each of the inspection intervals may be extended or decreased by as much as one year.
Adjustments shall not cause successive intervals to be altered by more than 1 year from the original pattern of intervals.
III. Basis for Relief Code Case N-535, Alternative Requirements for Inservice Inspection Intervals, is not currently approved for use by Regulatory Guide 1.147, Inservice Inspection Code Case i
f-ws 4
Acceptability ASME Section XI Division 1.
Footnote 6 of 10 CFR 50.55a states that other Code Cases may be authorized upon request pursuant to 10 CFR 50.55a(a) (3).
In accordance with these provisions, permission is requested to utilize Code Case N-535 in North Anna's ASME Section XI Program.
Code Case N-535 interprets interval inspection scheduling requirement.s and allows both interval and period extensions of up to cne year to enable an inspection to coincide with a plant outage.
Application of this Code Case would provide additional scheduling flexibility in completing interval inspections.
The Code Case allows examinations to be performed to satisfy requirements of the extended interval in conjunction with examinations performed to satisfy requirements of the successive interval.
However, an examination performed to satisfy requirements of either the extended interval or the successive interval shall not be credited to both intervals.
The use of this Code case has been previously approved for North Anna Unit's 1 & 2 by NRC Letter 98-340, Dated 5/29/98.
IV.
Proposed Alternate Requirements Code Case N-535 may be implemented in its entirety.
,_,T t
N1-002 Rev. 0 5/99 l
1 2-47 L
l Virginia Electric and Ppwer Company f
North Anna Power Station Unit 1
,L(
Third Ten Year Interval RELIEF REQUEST NDE-14 l
I.
IDENTIFICATION OF COMPONENT l
Mark No.
Component No. _D_rawing No.
Class N-01 thru N-58 1-RC-R-1 11715-WMKS-RC-R-1.4 1
i The above are reactor vessel closure head nuts.
II.
CODE REQUIREMENT The 1989 Edition of ASME Section XI Table IWB-2500-1, examination category B-G-1, item number B6.10 requires a surface examination of the reactor vessel closure head nuts.
III.
BASIS FOR RELIEF The 1989 Addenda of ASME Section XI Table IWB-2500-1, examination category B-G-1, item number B6.10 has 0
changed the requirement from a surface examination-to a visual (Vr-1) examination.
Extensive cleaning of these nuts is required for a surface examination to be performed.
This extensive cleaning results in additional costs and the inefficient use of available manpower resources.
1 Due to design factors, the stripping areas of the female threads in a nut are approximately 1.3 times the areas of the mating male threads (see ASME Bl.1, Unified Inch Screw Threads).
Accordingly, if a defect were to be developed during service, they would' occur in the threads of the bolt or stud before developing in the nut's threads because of higher stresses in the male threads.
When reactor vessel head fasteners are tightened for closure or loosened for opening, the studs are tensioned and the nuts are run on the threads with.no load as the load is taken by the stud or bolt through the tensioning device.
This relief was previously granted for the second
' interval at North Anna Unit 1 per NRC Letter No.97-683, Dated 11/14/97.
v N1-002 Rev. 0 5/99 l
2-48
l '.
}'
RELIEF REQUEST NDE-14 CONTINUED.
IV.
PROPOSED ALTERNATE REQUIREMENT It is proposed that the requirements of the 1989 Addenda of ASME Section XI Table IWB-2500-1, exami-nation category B-G-1, item number B6.10 (visual, VT-1).be substituted for the Code surface examina-tion.
In addition:
l 1
l.
A visual (VT-2) examination will be performed during the normally scheduled system leakage test each refueling outage; L
2.
Technical Specifications require that the reactor coolant system leak rate be limited to one gallon per minute unidentified leakage.
This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; and 3.
The containment atmosphere particulate radioactivity is checked every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The proposed alternative examinations stated above will ensure that the overall level of plant quality and safety will not be compromised.
O N1-'002 Rev. 0 5/99 2-49
i Virginia Electric and Power Company North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST SPT-1 I.
IDENTIFICATION OF COMPONENTS System:
Reactor Coolant (RC) l Components:
Partial Penetration Welds @ Bottom of Reactor Vessel Bottom of Reactor Vessel ISI Class :
1
)
II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition, Category B-E, Item No. 4.13, requires reactor vessel partial penetration welds to have a visual (VT-2) examination during the system hydrostatic test of IWB-5222, j
in addition, Category B-P, Item Nos. B15.10, and B15.11 l
require a visual (VT-2) examination of the bottom of the reactor vessel during the' system leakage test of IWB-5221 and during the system hydrostatic test of IWB-5222,
]
f-s respectively'.
j III. BASIS FOR RELIEF
~
In order to meet the Section XI pressure and temperature requirements for the system leakage and system hydrostatic tests of the reactor vessel, reactor containment at North Anna Unit 2 is required to be at a subatmospheric pressure.
Station administrative procedures require that self contained breathing apparatus be worn for containment entries under these conditions.
This requirement significantly complicates the visual (VT-2) examination of the bottom of the reactor vessel during testing.
Access to the bottom of the reactor vessel requires that the examiner descend several levels by ladder and navigate a small i
entrance. leading to the reactor vessel.
In addition to these physical constraints, the examiner must contend with extreme environmental conditions:
elevated air temperatures due to reactor coolant at temperatures above 500 degrees F and limited air circulation in the vessel cubicle.
In addition, the examiner is limited to the approximate 30 minute capacity of the breathing apparatus for containment entry, the VT-2 examination, and containment exit.
N1-002 Rev. 0 5/99 2-50
i RELIEF REQUEST SPT-1
/}
Continued v
IV. ALTERNATE TESTING Technical Specifications require that the Reactor Coolant System Leak Rate be limited to 1 gallon per minute unidentified leakage.
This value is calculated at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Additionally the containment atmosphere particulate radioactivity is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
The incore sump room has a level alarm in the control room requiring operator action.
These actions would identify any integrity concerns associated with this area.
A VT-2 examination will be conducted when containment is at atmospheric conditions each refueling for evidence of boric acid corrosion.
J l
/^\\
N1-002 nuv, 0 5/99 l
2-51
Virginia Electric and Power Company fi North Anna Power Station Unit 1
\\m,l-Third Ten Year Interval RELIEF REQUEST SPT-2 I.
IDEN*IFICATION OF COMPONENTS j
Clas-2, and 3 components II. IMPRACTICAL CODE REQUIREMENTS l
IWA-4000 and IWA-5214 require a hydrostatic pressure test in accordance with IWB-5000, IWC-5000, or IWD-5000 following welded repairs or installation of replacement items by welding.
l III. BASIS FOR RELIEF In past situations our utility has been required to defer or ask relief from ASME Section XI hydrostatic tests following repair or replacement activities due to various reasons, i
which identified a basis of impracticality.
This has ranged from boundary valve isolation problems to incorporation of the steam generators in the test boundary.
These situations have necessitated in some cases immediate communication with O
and approval from the NRC, so that start-up delays or LCO conditions could be avoided.
Test deferrals, like those associated with the steam generator, eventually must be conducted, and are considered inordinately burdensome, l
considering that the.10-year hydrostatic tests have been eliminated for Class 1 and 2 systems with the approval in l
Regulatory Guide 1.147, Code Case N-498, " Alternative Rules l
-for 10-year Hydrostatic Pressure Testing for Class 1 and 2 j
Systems,Section XI, Division 1."
The Code has recognized i
that alternative rules should be available to hydrostatic testing to allow an option to the owner.
They have developed l
Code Case N-416-1, " Alternative Pressure Test Requirement for Welded Repairs or Installation of Replacement Items by Welding."
Accordingly, it is considered impractical to maintain only the hydrostatic option.
IV. ALTERNATE TESTING In situations following welded repairs or installation of replacement items by welding, when the hydrostatic test required by.IWA-4000 or IWA-5214 is not performed, the alternative pressure test-outlined in Code Case 416-1 shall be applied.-Additional surface examinations will be performed on the root (pass) layer of butt and socket welds on the pressure
\\
N1-002 Rmt. 0 5/99 2-52
['.
l l
RELIEF REQUEST SPT-2 Continued v
(
retaining boundary of Class 3 components when the surface l
examination method is used in accordance with the applicable l
Subsection of ASME Section III.
I Use of this alternative will be documented on the NIS-2 Form.
i V
O N1-002 Rev. 0 5/99 2-53
E 1
I 1
l t
Virginia Electric and P,,ower Company
/"T North Anna Power Station Unit 1 l
(_)
Third Ten Year Interval RELIEF REQUEST SPT-3 l
I.
IDENTIFICATION OF COMPONENTS Class 1, 2,
and 3 components II. IMPRACTICAL CODE REQUIREMENTS Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, Categgry B-P Item numbers B15.11, B15.21, B15.31, i
B15.41, B15.51, B15.61, B15.71, Category C-H Item numbers C7.20, C7.40, C7.60, and C7.80. Category D-A Item number D1.10, Category D-B Item number D2.10, Category D-C Item number D3.10 require a hydrostatic pressure test in l
accordance with IWB-5000, IWC-5000, or IWD-5000.
III. BASIS FOR RELIEF Code Case N-498-1 now provides an alternative to the Class 3, as well as the Class 1 and 2 (N-498) hydrostatic test Code requirements. Hydrostatic test are historically difficult to s
perform. They frequently require extending test duration's f
(
)
while non-safety related issues such as maintenance boundary valve isolation problems are resolved. This increases the cumulative exposure.of test personnel and system tag-out duration. By taking advantage of Code Case N-498-1,
" Alternative Rules for 10-year Hydrostatic Pressure Testing for Class 1, 2,
and 3 Systems,Section XI, Division 1.",
increased testing flexibility is provided, allowing testing to be performed at nominal operating pressure. In turn, this significantly reduces the number of test blocks, system tag-outs, and corresponding boundary valves required to complete testing. This flexibility is accomplished while maintaining an acceptable level of safety and quality as determine by the ASME Code consensus process.
IV. ALTERNATE TESTING As an alternative to the 10-year hydrostatic pressure testing requirements of ASME Section XI, Code Case N-498-1 will be used in its entirety.
j i
t I
/
I N1-002 Rev. 0 5/99 2-54
Virginia Electric and Power Company
/'
North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST SPT-4 I.
IDENTIFICATION OF COMPONENTS Approximately 30, small diameter (g 1 inch), Class 1, reactor coolant system (RCS) pressure boundary vent and drain, sample, and instrumentation connections.
II.
IMPRACTICABLE CODE REQUIREMENTS Section XI Examination Category B-P, Items B15.51 and B15.71 require system hydrostatic testing and associated VT-2 visual examination of all Class 1 pressure retaing piping and valves.
III. ISI BASIS FOR RELIEF REQUEST l
These piping segments are equiped with valves that provide for double isolation of the reactor coolant system (RCS) pressure boundary. These valves are maintained closed during i
normal operation and the piping outboard of the first l
/"N isolation valve is, therefore, not normally pressurized. The j (,)
proposed alternative provides an acceptable level of safety j
and quality based on the following:
l 1.
ASME Section XI Code, paragraph IWA-4400, provides the l
requirements for hydrostatic pressure testing of piping i
and components af ter repa' irs by welding tc the pressure boundary. INA-4700 (b) (5 ) excludes component i
connections, piping, and associated vavles that are 1 inch nominal pipe size and snslier from the hydrostatic test. Visual examination of these < 1 inch diameter RCS vent / drain / sampling connections once each 10-year interval is unwarranted considering that a repair weld l
on-the same connections is exempted by the ASME XI l
Code.
i l
2.
The non-isolable portion of the RCS vent and drain l
. connections will be pressurized and visually examined as required. Only the isolable portion of these small diameter vent and drain connections will not be l
pressurized.
l 3.
All piping connections are typically socket-welded and i
the welds received a surface examination after installation. The piping and valves are nominally heavy wall (schedule 160 pipe and 1500# valve bodies). The vents, drains, and sample lines are not subject to high N1-002 Rev. 0 l
5/99 2-55 l.
I
\\
l l
RELIEF REQUEST SPT-4 l
/~T Continued (j
l stresses or cyclic loads, and the design ratings are significantly greater than RCS operating or design pressure.
The Technical Specifications (TS) require RCS leakage monitoring (TS 4.4.6.2.1) during normal operation. Should any of the TS limits be exceeded, then appropriate corrective actions, which may include shutting the plant down, are required to identify the source of the leakage and restore the RCS boundary integrity.
During 1998 North Anna Unit 1 refueling outage similar piping segments were pressurized by removing a flange and connecting a test rig. A majority of these piping segments are located in close proximity to the RCS main loop piping thus requiring personnel entry into high radiation areas within the containment. The dose associated with this testing was 1.5 man-Rem.
1 V.
ALTERNATE PROVISIONS As an alternative to the Code reguired hydrostatic test of the subject Class 1 reactor coolant system pressure boundary connections the following is proposed:
1.
The RCS vent, drain, instrumentation, and sample connections will be visually examined for leakage and any evidence of past leakage, with the isolation valves in the normally closed position each refueling outage during the ASME XI Class 1 System Leakage Test (IWB-5221).
2.
The RCS vent, drain, instrumentation, and sample connections will also be visually examined with the isolation vavles in the normally closed position during the 10-year ISI pressure test (IWB-5222 and Code Case N-498-1). This examination will be performed with the RCS at nominal operating pressure and at near operating temperature after sat.isfying the requried 4-hour hold time.
t In addition during modes 1 through 4 the RCS will be monitored for leakage for at following frequency for TS requirements:
l l
1.
Every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, during steady state operation, the reactor coolant system leak rate will be monitored to assure the limit of one gallon per minute unidentified leakage is maintained.
U rn~002 Rev. 0 5/99 2-56 e
L l
RELIEF REQUEST SPT-4 I~
Continued 2.
Every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the containment atmosphere particulate radioactivity will be monitored.
The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be compromised.
t VI IMPLEMENTATION SCHEDULE This alternative to Code requirements will be followed upon receiving NRC approval for the remainder of the second ten-
-year inspection interval.
The Unit 2 second ten-year inspection interval will-end December 14, 2000.
By a letter dated September 3, 1998 the NRR approved a similar relief request for Edwin I.
Hatch Plant, Units 1 and 2.
O.
N1-002 Rev. 0 5/99 2-57
~
l Virginia Electric and Power Company i
r'-
North Anna Power Station Unit 1
(_,}/
Third Ten Year Interval i
RELIEF REQUEST SPT-5 l
I.
Identification of Components Pressure retaining bolted connections within the scope of ASME Section XI.
II.
Code Requirements Per ASME Section XI, Division 1, 1989 Edition, paragraph IWA-52 50 (a) ( 2 ), when leakage is detected during the conduct of a system pressure test at a bolted connection, the bolting shall be removed and subjected to a VT-3 examination to detect evidence of corrosion, and evaluated in accordance with paragraph IWA-3100.
III. Basis for Relief Using Code Case N-566 in lieu of Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWA-fg 5250 (a) (2) allows greater flexibility and prudent decision k-)
making.
Leaking conditions at a bolted connection may be an f
important variable in the degradation of fasteners.
However, leakage is not the only variable, and some cases may not be the degradation mechanism.
Other variables to be considered are:
bolting materials; leaking medium; duration of the leak; and orientation of the leak (not all bolts may be wetted).
These variables are important to consider before disassembling a bolted connection for a visual VT-3 examination.
Removal of bolting at a mechanical connection may not be the most prudent decision and may cause undue hardship without a compensating increase in the level of quality or safety.
IV.
Proposed Alternate Requirements Code Case N-566 shall be implemented in it's entirety.
NOTE:
If this relief request is approved, then Relief Request SPT-6 shall be considered as withdrawn.
If this relief request is not approved, then please consider Relief Request SPT-6 as an alternative, and this Relief Request shall be considered as withdrawn.
I N1-002 Rev. O 5/99 2-58
i Virginia Electric and Power Company C[
North Anna Power Station. Unit 1 Third Ten Year Interval RELIEF REQUEST SPT-6 I.
Identification of Component-Class 1, 2 and 3' pressure retaining bolting II.
Impractical Code requirements IWA-5250 (a) (2) states, "if leakage occurs at a bolted connection, the bolting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA-3100;"
III. Basis for Relief Leaking conditions at a bolted connection may be an important variable in the degradation of fasteners.
However, leakage is not the only variable, and some cases may not be the degradation mechanism.
Other variables to be considered are: bolting materials, leaking medium, duration of the leak, and orientation of the leak (not all bolts may be wetted).
These variables are important to consider
(
before disassembling a bolted connection for a visual VT-3 examination.
Removal of bolting at a mechanical connection j
may not be the most prudent decision and may cause undue hardship without a compensating increase in the level of quality or safety.
Virginia Power proposes an alternative to the requirements of_IWA-5250 (a) (2) that will provide'an j
equivalent level of quality and safety at Class 1, 2,
and 3 i
bolted connections.
'IV.
Alternate Requirements Leakage discovered at a bolted-connection by visual VT-2 examination during a system pressure test will be evaluated to determine the susceptibility of the bolting to corrosion and potential failure.
This evaluation will consider the following variables:
1.
Location of leakage; 2.
History of leakage; 3.
Fastener materials; 4.
Evidence of corrosion with the connection assembled; 5.
Corrosiveness of the process fluid; 6.
History and studies of similar fastener material in a similar environment; and 7.
Other components in the vicinity that may be degraded due to the leakage.
N1-002 Rev. O 5/99 2-59
i RELIEF REQUEST SPT-6 Continued
)
l When evaluation of the above variables is concluded and the evaluation determines that the leaking condition has not degraded the fasteners, then no further action is necessary.
If evaluation of the variables above indicates the need for further evaluation, or no evaluation is performed, then the bolt l
closest to the source of leakage will be removed.
The bolt will l'
receive a visual VT-3 examination, and be evaluated in accordance l
with IWA-3100 (a).
This visual VT-3 examination may be deferred to the next outage of sufficient duration if the evaluation l
supports continued service.
When the removed bolting shows evidence of rejectable degradation, all remaining bolts shall be i
removed and receive a visual VT-3 examination and evaluation in accordance with IWA-3100 (a).
l l
NOTE:
If Relief Request SPT-5 has been approved then this
~
Relief Request is withdrawn.
O N1-002 Rev, 0 5/99 2-60
I
)
1 Virginia Electric and Power Company 7
l t
North Anna Power Station Unit 1
(
Third Ten Year Interval RELIEF REQUEST SPT-7 I.
IDENTIFICATION OF COMPONENTS Class 2 piping that penetrates the containment vessel where f
i the piping and isolation valves are part of the containment system but the balance of the piping is outside of the scope of Section XI.
l Drawing Test boundary Penetration Number 11715-SPM-096A-3 1"-SI-13-602-02 between penetration 20 20 l
SH. 1 and 2-SI-47 11715-SPM-096B-3 1"-SI-13-602-Q2 between 2-SI-136 and 20 y
SH. 1 penetration 20 I
l 11715-SPM-106A-3 h"-HC-13-ICN6-Q2 between 1-HC-TV-101B 31 i
SH. 1 and 2"-HC-2-154-Q2, and 2%"-HC-36-154-l Q2 between 1-HC-TV-105B, and 1-HC-14 l
13075-SPM-102C-3 3"-SGD-16-601-02 32 O.
SH. 1 11715-SPM-090C-3 2"-DG-35-153A-Q2 33 SH. 1 11715-SPB-102B-3 4" line between 1-FP-275 and 1-FP-272 34 SH. 1 11715-SPM-090A-3 2"-DG-35-153A-Q2 38 SH. 1 11715-SPM-082F-3 2"-ASC-20-153A-02 42 SH. 1 11715-SPM-082N-3 1"-ARC-1-153A-Q2 between 1-RM-TV-100D 43 SH.'3 and 1-IA-149 j
11715-SPM-082N-3 1"-ARC-2-153A-Q2 between 1-RM-TV-100B 44 l
SH. 3 and 1-RM-TV-100C i
l 11715-SPM-093B-3 3"-RC-52-153A-02 between 1-RC-TV-1519A 45 SH. 2 and 1-RC-149 11715-SPM-082A-3 2"-ACC-21-153A-Q2 between penetration 47 SH. 1 47 and 1-IA-55 t
11715-SPM-082N-3 2"-ACC-21-153A-Q2 between penetration 47 SH. 1 47 and 1-IA-TV-102A
.,s N1-002 Rev. O S/99 2-61 l
\\
RELIEF REQUEST SPT-7 Continued 11715-SPM-090C-3 1 %"-VG-6-153A-02 between 1-VG-TV-100B 48 SH. 1 and 1-VG-TV-100A 11715-SPM-096B-3 1"-SI-98-601-Q2 between 1-SI-HCV-1936 50 SH. 1 and 1-SI-TV-101 11715-SPM-096B-3 1"-SI-117-1501-Q2 between 1-SI-106 and 53 SH. 1 1-SI-TV-100 11715-SPM-090C-3 2"-VA-57--154-Q2 between 1-DA-41 and 1-54 SH. 1 DA-39 l
i 11715-SPM-092A-3 h"-LM-3-ICN6-02 from 1-LM-TV-100E to 55D SH. 1 penetration SSD 11715-SPM-092A-3 h"-LM-10-ICN6-02 from 1-LM-TV-100G to 57A SH. 1 penetration 57A 11715-SPM-072A-3 4"-SS-234-ICN9-02 57D SH. 2 11715-SPM-089D-3 6"-AJA-10-151-02 89 SH. 1 11715-SPB-006A-3 36"-AR-90-151-Q2 between 1-HV-MOV-101, 91 (N
SH. 1 1-HV-MOV-100C and, 1-HV-MOV-100D 11715-SPB-006A-3 36"-AR-91-151-02 between 1-HV-MOV-102, 91 SH. 1 1-HV-MOV-100A and, 1-HV-MOV-100B 11715-SPM-092A-3 2"-CV-10-154-02 between 1-CV-TV-150D 92 SH. 2 and penetration 92 11715-SPM-106A-3 2"-HC-33-154-Q2, 2%"-HC-34-154-Q2 from 92 SH. 4 1-HC-TV -104B 11715-SPM-092A-3 2"-CV-9-154-Q2 between 1-CV-TV-150B and 93 SH. 2 penetration 93 11715-SPM-106A-3 2"-HC-37-154-Q2, 2%"-HC-38-154-02 from 93 SH. 4 1-HC-TV -106B 11715-SPM-092A-3 8"-CV-11-154-Q2 between 1-CV-4 and 1-94 SH. 2 CV-TV-100 11715-SPM-092A-3 4"-LM-3-ICN6-02 from 1-LM-TV-100A to 97B SH. 1 penetration 97B 11715-SPM-106A-3 h"-HC-57-ICN9-Q2 between 1-HC-TV-100A 98A SH. 1 and 1-HC-TV-100B 11715-SPM-106A-3 h"-HC-67-ICN9-Q2 between 1-HC-TV-108A 98B SH. 3 and 1-HC-TV-108B b\\j N1-002 Rev. 0 5/99 2-62
RELIEF REQUEST SPT-7
/~'\\
Continued
Y 13075-SPM-102C-3 3"-SGD-17-601-02 100 SH. 1 11715-SPM-088A-3 6"-RP-51-152-02 and 6"-RP-43-153A-Q2 103 SH. 3 11715-SPM-088A-3 6"-RP-23-152-Q2 104 SH. 3 11715-SPM-092A-3 4"-LM-1-ICN6-02 from 1-LM-TV-100C to 105A SH. 1 penetration 105A 11715-SPM-106A-3 4"-HC-454-ICN6-Q2 between 1-HC-TV-102A 105B SH. 2 and 1-HC-TV-102B 11715-SPM-092A-3 4"-LM-7-ICN6-Q2 from 1-LM-TV-101A to 105C SH. 1 penetration 105C 11715-SPM-092A-3 4"-LM-35-ICN6-Q2 from 1-LM-TV-101C to 105D SH. 1 penetration 105D 13075-SPM-102C-3 3"-SGD-18-601-Q2 108 SH. 1 11715-SPM-106A-3 h"-HC-11-ICN6-Q2 between 1-HC-TV-103B 109 SH. 2 and 2"-HC-3-154-Q2, and 2%"-HC-40-154-(
Q2 between 2"-HC-3-154-Q2 and 1-HC-TV-107B, and 2"-HC-3-154-02 11715-SPM-090C-3 2"-DA-46-153A-Q2, %"-DA-47-153A-Q2, 111D SH. 1 DA-48-153A-Q2, and 2"-DA-49-153A-02 II.
CODE REQUIREMENTS Table IWC-2500-1, Examination Category C-H, Items C7.30 and C7.70 requires a system pressure test each inspection period and Items C7.40 and C7.60 requires a system hydrostatic test each inspection interval.
III.
BASIS FOR RELIEF The sole safety function of the piping and associated valves listed is to provide con *ainment isolation. The components listed are part of the containment system.
Containment penetrations are classified as Class 2 per ANSI 18.2, " Nuclear Safety Criteris for the Design of Stationary Pressurized Water Reactor Plants", section 2.3.1.2 (1). For the subject penetrations the connecting piping beyond the containment isolation valves serves no safety function and is classified as nonclass by the classification criteria O
N1-002 Rev. 0 5/99 2-63
O RELIE'F REQUEST SPT-7 Continued used by Virginia Electric and Power Company for North Anna Unit 1.
The ASME Section XI pressure testing requirements have verified leak-tight integrity by an over pressure test every ten years and a nominal operating test every inspection period. The 10-year hydrostatic tests were considered inordinately burdensome for the marginal benefit in safety they assure and have been eliminated by Code Case N-498, " Alternative Rules for 10-year Hydrostatic Pressure Testing for Class 1 and 2 Systems,Section XI, Division 1",
which has been approved by Regulatory Guide 1.147, Inserv-ice Inspection Code Case Acceptability ASME Section XI Division 1. The 10-Year hydrostatic test is now conducted at nominal operating pressure.
The subject penetrations are Type C pressure tested to a.
peak containment internal pressure of greater than or equal to 44.1 psig. This test is performed to satisfy Technical Specification Surveillance Requirement 4.6.1.2 which requires all containment penetrations to be leak rate tested as required by 10 CFR 50, Appendix J, Option B, as modified-by approved exemptions, and in accordance with the O
guideline contained in Regulatory Guide 1.163, dated September 1995.
The testing frequency of 10 CFR 50, Appendix J, Option B is performance based and can vary from 2 years to 5 years or three refueling cycles.
This frequency will not coincide with the inspecticn period frequency required in Table IWC-2500-1 for system pressure tests and therefore, the ASME Code in effect at North Anna will require additional leak tightness testing.
The ASME Section XI Code has acknowledged that testing of these components beyond the requirements of Appendix J is not necessary and issued Code Case N-522, " Pressure Testing of Containment Penetration Piping", to define its position.
NUREG-1493, " Performance-Based Containment Leak-Test Program", concluded that prescriptive leak rate testing could be replaced with performance based requirements with only a marginal and acceptable impact on safety. The total cost of Type B (electrical penetrations) and Type C testing all containment penetrations (approximately 90 penetrations) was estimated to be $87,500 per outage for North Anna as reported in NUREG-1493. NUREG-1493 estimates that 5% of'the total cost of Type B & C testing could be saved if the acceptance criteria were relaxed. Performing
-ASME Section XI pressure testing beyond the requirements of N1-002 Rev. 0 5/99 2-64
RELIEF REQUEST SPT-7 Continued 10 CFR 50, Appendix J, Option B testing would cause Virginia Electric & Power Company to incur additional cost with a marginal gain in safety.
This relief was previously approved for North Anna Power Station Unit 2 reference NRC Letter No.97-290, dated 4/30/97 and Surry Power Station Unit's 1 and 2, reference NRC Letter No.96-216, dated 4/16/96.
IV.
PROPOSED ALTERNATIVE As an alternative to the testing frequency and pressures required by Table IWC-2500-1, Examination Category C-H, Items C7.30, C7.40, C7.60, and C7.70, the subject penetrations and associated piping and valves, will be pressure tested at peak containment calculated pressures to the requirements of 10 CFR Appendix J, as allowed by Code Case N-522. Testing will be performed in accordance with Technical Specification Surveillance Requirement 4.6.1.2 which requires all containment penetrations to be leak rate tested as required by 10 CFR 50, Appendix J, Option B, as modified by approved exemptions, and in accordance with the guideline contained in Regulatory Guide 1.163, dated September 1995.
O Methods for the detection and location of leakage at containment isolation valves and the pipe segments between the containment isolation valves will be identified in procedures.
All subject penetrations will be Type C tested at least once every 60 months.
\\
N1-002 Rev. 0 5/99 2-65
Virginia Electric and P,ower Company
/ \\
North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST SPT-8 I.
IDENTIFICATION OF' COMPONENTS System:
Reactor Coolant (RC)
Charging (CH)
Safety injection (SI)
Components:
Pressure retaining bolted connections located 1
inside containment which are normally tested in subatmospheric conditions.
ISI Class:
1 and 2 II.
CODE REQUIREMENTS Subparagraph IWA-5242(a); For systems borated for the purpose of controlling reactivity, insulation shall be removed from pressure retaining bolted connections for visual examination VT-2.
III. BASIS FOR RELIEF A majority of these systems are located inside of containment. In cases where there is no intermediate isolation from the RC system, the system pressure test is performed when the RC system is greater than 500 degrees F, and the containment is subatmospheric.
Removing and reinstalling insulation under these conditions is difficult to perform and deemed impractical when compared to the remainder of the Pressure Testing Program.
Further, in our response to IE Bulletin 82-02 we agreed to examine pressure retaining bolting on lines 4 inch NPS and larger each refueling.
In addition, in our response to NRC Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants, we committed to performing a visual examination of all accessible portions of systems containing boric acid anytime the unidentified leak rate exceeds a predetermined value, at every cold shutdown prior to decontaminating the containment, and during every reactor startup from Mode 5.
This requirement is a burden with no compensating increase in safety This relief was previously approved for Surry Power Station Unit's 1 and 2, reference NRC Letter No.96-216, dated N1-002 Rev. 0 5/99 2-66
]
m i
i
)
l RELIEF REQUEST SPT-8
(
Continued 4/16/96, and North Anna Power Station Unit 2, reference NRC Letter No.92-730, dated 11/5/92.
IV.
PROPOSED ALTERNATIVE As an alternative to the Code requirements, pressure j
retaining bolted connections on Class 1 systems tested during subatmospheric conditions, and within the scope of Section XI, will be visually examined each refueling outage at zero or static pressure.
The examination will be performed with the insulation removed.
Pressure retaining bolted connections on Class 2 systems tested during subatmospheric conditions, and within the scope of Section XI, will be visually examined each inspection period at zero or static pressure.
The examination will be performed with the insulation removed.
The Code required pressure testing will continue to be performed with a VT-2 examination performed with the insulation in place.
' O 1
i N1-002 Rev. O 5/99 2-67
O VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 INSERVICE INSPECTION PROGRAM PLAN O
FOR COMPONENT SUPPORTS THIRD INSPECTION INTERVAL L
3-1
SECTION 3 TABLE OF CONTENTS SECTION TITLE PAGE 3.O INSERVICE INSPECTION PROGRAM 3-3 FOR COMPONENT SUPPORTS 3.1 PROGRAM DESCRIPTION 3-3 3.2 PROGRAM EXCLUSIONS 3-3 3,3 COMPONENT SUPPORT RELIEF REQUESTS 3-5 l
1 I
r.,
N1-003 Rev.0
/
)
5/99
/
3-2
3.0 INSERVICE INSPECTION PROGRAM FOR COMPONENT SUPPORTS Ok/
3.1 PROGRAM DESCRIPTION m
The Inservice Inspection program for Class 1, 2 and 3 component supports meets the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, and the alternative provisions found in Code Case N-491, " Alternative Rules for Examination of Class 1,
2, 3,
and MC Component Supports of Light-water Cooled Power Plants,Section XI, Division 1."
3.2 PROGRAM EXCLUSIONS This program does not cover the functional testing of snubbers.
Snubbers will be tested in accordance with Technical Specifications.
3.3 COMPONENT SUPPORT RELIEF REQUESTS Component support relief requests are presented in a j
numeric sequence with a "CS" prefix.
{
l l
I (v)
N1-003 Rev.0
[j--
5/99 s
3-3
RELIEF REQUEST STATUS
{
Relief No.
Status V.P.
Ltr. No.
NRC Ltr. No.
CS-1 Pending O
N1-003 Rev.0 O
5/99
~
3-4
I Virginia Electric and Power Company O
North Anna Power Station Unit 1 Third Ten Year Interval RELIEF REQUEST CS-1 I.
Identification of Components:
Class 1, 2, and 3 Snubbers II.
Code Requirements:
)
I Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWF-5000 (with regard to visual examination and j
repairs / replacements of snubbers) requires that the first i
l
. addenda to ASME/ ANSI OM-1987, Part 4 (published in 1988) be used.
l l
III. Code Requirement From Which Relief Is Requested:
l Relief is requested from fully performinc the Code-required (Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, IWF-5000) examination on the above identified s
IV.
Basis for Relief:
Differences exist between the referenced standard require-ments and the North Anna Technical Specification 4.7.10.
The first addenda to ASME/ ANSI OM-1987, Part 4 (published in l
1988) contains requirements, which were removed from techni-t cal specifications by Generic Letter 90-09, " Alternative Requirements For Snubber Visual Inspection Intervals and j
Corrective Actions."
This document was issued on December 11, 1990 to reduce the burden placed upon utilities by the then excessively restrictive inspection schedule contained in technical specifications.
~9 referenced standard in ASME Section XI again requires this excessively restrictive inspection schedule for snubbers.
Later Code, ASME OMc Code-1994 Addenda to ASME OM Code-1990 Edition, subsection ISTD (replacing OM Part 4), Table ISTD 6.5.2-1, now contains the newer visual examination schedule similar to that in the North Anna Technical Specification 4.7.10.
Therefore, requiring the use of the referenced standard's inspection schedule is considered to impose a burden without a compen-l sating increase in quality and safety.
tu-003 Rev.0 O
5/99 3-5 l
g RELIEF REQUEST CS-1 CONTINUED The referenced standard additionally contains requirements eliminated by the later ISTD Code (1994 addenda) with regard i
to visual failure evaluation groupings.
As this is not perceived by the industry as a significant technical issue, performance of this requirement would be an unnecessary administrative burden, and therefore impractical to perform.
i The remaining aspects of the referenced standard contain requirements, which are essentially the same as those al-ready delineated in the North Anna Technical Specification 4.7.10, and the ASME Section XI Repair / Replacement program.
Requiring the use of the referenced standard in these re-maining areas is an unnecessary.
This Relief Request has been previously submitted and approved for-Surry Power Station Unit's 1 & 2, reference NRC Letter No.95-303, Dated 6/8/95.
V.
Proposed Alternate Requirements:
In lieu of the Code requirements specified in IWF-5000 for snubber visual examination, the requirements of North Anna Technical Specification 4.7.10 with regard to snubber visual
(
examination shall be followed.
The required examination shall be performed by personnel certified visual (VT-3) as required by the Code.
Repairs and replacements shall be performed as required by the general requirements (IWA-4000 and IWA-7000) of ASME Section XI.
The proposed alternative stated above will ensure that the overall level of plant quality and safety will not be com-promised.
(N N1-003 Rev 0 t
5/99 3-6
O VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 O
I'NTERIM RELIEF REQUESTS V
AND MISCELLANEOUS DOCUMENTATION THIRD INSPECTION INTERVAL N1-004 Rev 4-1
SECTION 4 TABLE OF CONTENTS 4.0 INTERIM RELIEF REQUESTS AND DOCUMENTATION 4.1 RELIEF REQUESTS (None at this time) 4.2 MISCELLANEOUS DOCUMENTATION (None at this time) l l
l l
O l
l 1
N1-004 Rev,0 5/99 4-2
O VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 O
PARTIAL COVERAGE RELIEF REQUESTS THIRD INSPECTION INTERVAL i
l l.
i 1
I N1-005 Rev. O 5/99 5-1 I
SECTION 5 TABLE OF CONTENTS 5.O PARTIAL COVERAGE RELIEF REQUESTS 5.1 RELIEF REQUESTS (None at this time)
O N1-005 Rev. 0 O
5/99 5-2
1 l
VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNIT 1 O
COoE cases uTIt1zEo IN THE THIRD INSPECTION INTERVAL O
'N1-006 Rev. 0 5/99 6-1
SECTION 6 TABLE OF CONTENTS 6.0 CODE CASE UTILIZED IN THE THIRD INSERVICE INSPECTION INTERVAL Case N-401-1 Eddy Current Examination,Section XI, Division 1.
-Case N-402-1 Eddy Current Calibration Standards,Section XI, Division 1.
I i
Case N-416-1 Alternative Pressure Test Requirement For Welded Repairs or Installation of Replacement Items by l
Welding, Class 1, 2,
and 3,Section XI, Division i
1.
Case N-435-1 Alternative Examination Requirements for Vessels with Wall thickness 2 in. or less,Section XI, l
Division 1.
l Case N-437 Use of Digital Readout and Digital Measurement Devices for Performing Pressure Tests,Section XI, Division 1.
l Case N-460 Alternative Examination Coverage for Class 1 and j
Class 2 Welds,Section XI, Division 1.
)
Case N-461 Alternative Rules for Piping Calibration Block Thickness,Section XI, Division 1.
Case N-481 Alternate Examination Requirements for Cast Austenitic Pump Casings,Section XI, Division 1.
(Reference WCAP-13045 for the evaluation to demonstrate the safety and serviceability of the pump casing.)
Case N-489 Alternative Rules for Level III NDE Qualification Examinations,Section XI, Divisions 1,,
2, and 3.
Case U-491 Alternative Rules for Examination of Class 1,2, and 3 and MC Component Supports of Light-Water l
Cooled Power Plants,Section XI, Division 1.
N1-006 Rev. 0 5/99 6-2
F l
l l
Case N-498-1 Alternative Rules for 10-year Hydrostatic i
/~N
{
e Pressure Testing for Class 1, 2 and 3 Systems, q
Section XI, Division 1.
Case N-508-1 Rotation of Serviced Snubbers and Pressure Relief Valves for the Purpose of Testing,Section XI, Division 1.
Case N-509 Alternative Rules for the Selection and Examination of Class 1, 2,
and 3 Integrally Welded Attachments,Section XI, Division 1.
Case N-522 Pressure Testing of Containment Penetration Piping,Section XI, Division 1.
Case N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping, i
Section XI, Division 1.
Case N-535 Alternative Requirements for Inservice Inspection j
Intervals,Section XI, Division 1.
Case N-566 Corrective Action for Leakage Identified at Bolted Connections, Section, XI, Division 1.
(~})
N1-006 Rev. O I-sT 5/99
! V 6-3
CASE N-401-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approvel Date: May 4,1988 See NumericalIndex for expiration and any reaffirmation dates.
Case N-4011 Eddy Current Examination Section XI, Division 1 Inquiry: Is it permissible, as an alternative to the Section XI, Division 1. Appendix IV, requirement for magnetic tape and strip chart recording of eddy current test data, to use digitized collection and storage of data as the permanent record for eddy current examina-tions, and to what extent do the calibration require-ments of IV-3000 apply?
Reply: It is the opinion of the Committee that, when demonstrated equivalent to the Authorized Nuclear Inservice Inspector, digitized collection and storage of eddy current test data may be used as an alternative to magnetic tape and strip chart recording of eddy current test data required by Section XI Division 1.
[
Appendix IV. Periodic calibration in accordance with
\\
IV-3000 is required only for the analog elements of the digital eddy current data collection and storage equipment.
t'v' 737
CASE I
N-402-1
(
CASES OF ASME BOILER AND PRESSURE VESSEL CODE 1%
Approval Date: March 14,1991 See NumericalIndex for expiration and any reaffirmation dates.
1 1
Case N-4021 dards on tubing of the same nominal chemical com-Eddy Current Calibration Standards position and product form may be used in lieu of Section XI, Division 1 standards on tubing of the same nominal chemical l
composition and heat treatment subject to the fol-Inquiry: When conducting eddy current examina-lowing conditions:
tions of steam generator tubing in accordance with (a) Standards fabricated from UNS Alloy N06600 i
Section XI, Division 1, IWA 2233, may eddy current shall be fabricated from a length of tubing of the calibration standards on tabing of the same nominal same material specification and same nominal size chemical composition and product form be used in as that to be examined in the vessel.
lieu of standards on tubing of the same nominal (b) Standards fabricated from materials other than chemical composition and heat treatment?
UNS Alloy N06600 and which are heat treated dif-ferently from the tubing to be examined may be used only when signal responses from the calibration dis-continuities described in Section XI, Division 1, Ap-Reply: It is the opinion of the Committee that, pendix IV are demonstrated to the satisfaction of the when conducting eddy current examinations of steam Inspector to be equivalent in both the calibration generator tubing in accordance with Section XI, Di-standard and tubing of the same heat treatment as vision 1, IWA-2233, eddy current calibration stan-the tubing to be examined.
\\
l
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G i
739 I
CASE N-416-1
[
CASES OF ASME BOILER AND PRESSURE VESSEL CODE i
Approval Date: February 15,1994 See Numeric Index for expiration and any reaffirmation dates.
(a) NDE shall be performed in accordance with Case N 4161 Alternative Pressure Test Requirement for the methods and acceptance criteria of the applicable Welded Repairs or Installation of Replacement Subsection of the 1992 Edition of Section Ill.
Items by Welding, Class 1,2 and 3 (b) Prior to or inunediately upon return to service-.
Section XI, Division 1 a visual examination (VT-2) shall be perfor aed in conjunction with a system leakage test. aing the 1992 Inquiry: What altemative pressure test may be per-Edition of Section XI, in ac;ordance with para.
formed in lieu of the hydrostatic pressure test required IWA 5000, at nominal operating pressure and temper-by para. IWA-4000 for welded repairs or installation
- ature, of replacement items by welding?
(c) Use of this Case shall be documented on an NIS-2 form.
Reply: It is the opinion of the Committee that in if the previous version of this case was used to heu of performing the hydrostauc pressure test re-defer a Class 2 hydrostatic test, the deferred test may quired by para. IWA-4000 for welded repairs or instal-be e'iminated when the requirements of this revision lation of replacement items by welding, a system leak-are met.
age test may be used provided the following requirements are met.
p
!v)
)
v 547
CASE N-435-1 CASES OF ASME BOliER AND PRESSURE VESSEL CODE Approval Date: July 30, 1986 See Numeric Index for expiration and any reaffirmation dates.
Case N 4351 Alternative Examination Requireruuts for Vessels With Wall Thickneas 2 in. or LessSection XI, Division 1 Inquiry: What alternadve examination requirements may be used for examination of ferride or austeninc vessels with wall thickness of 2 in. or less under Section XI, Division 1. Table IWB-2500-1, Category B-B, or Table IWC-2500-1 Category C-A?
Reply: it is the opinion of the Committee that for ferritic or austenitic vessels with wall thickness of 2 in. or less referenced in the Inquiry, the following alternative examination requirements may be applied.
I (a) Acceptance standards for the examination shall be those specified for vessels in IWB-3000 for Class I vessels and IWC-3000 for Class 2 vessels.
J (b) For welds in vessels with nominal wall thickness of '/ in. or less, surface examination may be applied
's 5
in lieu of volumetric examination.
(c) For welds in vessels with nominal wall thickness greater than '/ in. and less than or equal to 2 in.,
3 nitrasonic examination may be performed using the rules of the Winter 1985 Addenda of Section XI.
Division 1 Appendix Ill.
>R i
v 555
l CASE N-437 CASES OF ASME BOILER AND PRESSURE VESSEL CODE f
Approval Date: July 30,1986 See Numeric Indes for expiration and any restfirmation dates.
i Case N-437 Use of Digital Readout and Digital Measurement Devices for Performing Pressure TestsSection XI, Division 1 Inquiry: What are the requirements for digital pres-sure measuring instruments used for pressure tests un-i der IWA-5260?
Replyt it is the opinion of the Committee that digital j
pressure measuring instruments may be used for pres-sure tests under IWA-5260 provided the following re-quirements are met:
(a) the calibration requirements of IWA-5262; l
(b) the accuracy is within 0.5% over the calibrated range of the instrument; and (c) the intended maximum test pressure does not exceed 70% of the calibrated range of the instrument.
/
l l
O 785
f CASE N-460 CASES OF ASME BOtLER AND PRESSURE VESSEL CODE
)
[T Approval Date: July 27,1988
\\
See Numeric Index for expiration and any restfirmation dates.
Case N 460 Alternative Examination Coverage ict Class I and Class 2 WeldsSection XI, Division 1 Inquiry: What alternative rules may L used fsr Section XI, Division 1, examination of Class I wefds (IWB-2500) or Class 2 welds (IWC-2500) when the entire examination volume or area cannot be examined due to interference by another component or part ge-ometry?
Reply It is the opinion of the Committee that when the entire examination volume or area cannot be ex-a.nine:i due to interference by another component or part geometry, a reduction in examination coverage on any Class I or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10%. The applicable examination records shall identify both the cause and percentage of reduced examination coverage.
s 581
CASE N-461 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: November 30,1988 See Numeric Index for expiration and any reaffirmation deles.
Case N 461 Alternative Rules for Piping Calibration Block ThicknessSection XI, Division 1 Inquiry: When selecting calibration blocks in accordance with Section XI, Division 1 III 3410, what alternative wall thicknesses or pipe schedules may be used?
j f
Reply It is the opinion of the Committee that any l
calibration block thickness may be used that is within 225% of the pipe wall thickness to be examined.
l C
l l
l O
711 1
CASE N-481 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Ap.wval Date: March 5,1990 See Numeric Index for expiration and any teamtmation dates.
Case N 481 (c) Perform a VT-3 visual examination of the internal Alternative Examl== tion Requirements for Cast surfaces whenever a pump is disassembled for mainte-(
Austenitic Pump Casings nance.
Section XI, Division 1 (d) Perform a evaluation to demonstrate the safety and serviceability of the pump casing. The evaluation Inquiry When conducting examination of cast aus-shall include the following:
tenitic pump casings in accordance with Section XI, (1) evaluating material propenies; including frac-Division 1, what examinations may be performed in ture toughness values; lieu of the volumetric examinations specified in Table (2) performing a stress analysis of the pump casing; 2500-1, Examination Category B-L-1, item (3) reviewing the operating history of the pump; (4) selecting locations for postulating flaws; (5) postulating one-quarter thickness reference flaw Reply: It is the opinion of the Committee that the with a length six times its depth; following requirements shall be met in lieu of per-(6) establishing the stability of the selected flaw forming the volumetric examination specified in Table under the goveming stress conditions; IWB-2500-1. Examination Category B-L-1, item (7) considering thermal aging embrittlement and B12.10:
(a) Perform a VT-2 visual examination of the exterior any other processes that may degrade the properties of all pumps during the hydrostatic pressure test required of the pump casing during service.
by Table IWB-2500-1, Category B-P.
(c) A repon of this evaluation shall be submitted (b) Perform a VT-1 visual examination of the external to the regulatory and enforcement authorities having surfaces of the weld of one pump casing, jurisdiction at the plant site for review.
O 651
1 i
CASE l
N-489 CASES OF ASME 7, OILER AND PRESStlRE VESSEL CODE j
Approval Date: August 14, 1990 See Numeric Index for expiration l
and any reaffirmation dates.
i Case N 489 (b) Level III personel shall be recertified using only l
Alternate Rules for Level III NDE Qualification the written Method and Specific Examinations and a Examinations Level II Practical Examination for each method for Section XI, Divisions 1,2, and 3 which they seek recenification.
(c) As an alternate to (b) above, Level 111 NDE Inquiry: What alternative rules to those of Section XI, Divisions 1, 2, and 3, may be used for qualifica-Personnel may be reeertified using only written Method tion examinations for certification and recertification and Specific Examinations, provided the following con-ditions are satisfied:
of Level III NDE personnel?
(1) ne Level III candidate was previously certified Reply: It is the opinion of the Committee that the n accordance with (a) above, or recertified in accordance following alternative rules may be used for Section
- '* M
- XI Divisions 1, 2, and 3 qualification examinations (2) ne Level Ill candidate is not being recertified for certification and recertification of Level III NDE due to interrupted service as defined in the employer's personnel.
(a) Level III NDE personnel shall be certified using wrnten pracuce.
the written Basic, Method, and Specific Examinations (3) ne Level Ill candidate is not being certified specified in the applicable division of Section XI. The by a new employer.
Level !!! candidate shall successfully complete a Level (d) These alternative rules for Level III NDE qualifi-Il Practical Examination for each NDE method for cation examinations shall be documented in the employ-
]
p) which they seek certification.
er's written practice.
j Q) 1 I
l (v3 663
e CASE N-491 CASES OF ASME BO!!IR AND PRESSURE VESSEL CODE
\\
Approvel Date: March 14,1M1 See NumericalIndex for expiration anti any restfirmation dates.
Case N 491
-3000 STANDARDS FOR EXAMINATION
{
Alternative Rules for Examination of Class 1,2,3, EVALUATIONS and MC Component Supports of Light Wat'er
-3100 Evaluation of Examination Results Cooled Power Plants 3110 Presewice Examination Section XI, Division 1 3111 General I
-3112 Acceptance Inquiry: What alternative examination require.
3120 Inservice Examinations ments to those stated in Section XI, Division 1, Sub-3121 General section IWF may be used when determining the com.
-3122 Acceptance ponent supports subject to examination and
-3200 Supplemental Examinations establishing requirements for component supports?
-3400 Acceptance Standards 3410 Acceptance Standards - Component Reply: It is the opinion of the Committee that the Support Structural Integrity following alternative rules may be used for determin-TAB W ing component supports subject to examination and 2410-1 Inspection Program A for establishing examination requirements for Class
-2410-2 Inspection Program B 1,2,3, and MC component supports under Subsec-
-2500-1 Examination Categories js tion IWF, Section X1, Dmston 1.
TABLE OF CONTENTS
-1000 SCOPE AND RESPONSIBILITY 1100 SCOPE 1000 SCOPE AND RESPONSIBI(ITY 1100 Scope This Case provides alternative rules for inservice 1200 Component Supports Subject to Exam-inspection of Class 1,2,3, and MC component sup-ination and Test ports.
1210 Examination Requirements
-1220 Snubber inspection Requirements 1230 Supports Exempt from Examination
-1200 COMPONENT SUPPORTS SUBJECT
-1300 Support Examination Boundaries TO EXAMINATION AND TEST 2000 EXAMINATION AND INSPECTION
-1210 Examination Requirements 2100 Scope
-2200 Preservice Examination The examination requirements shall apply to the
-2210 Initial Examination following:
2220 Adjustment, Repair, and Replacement (a) piping supports;
-2400 Inspection Schedule (b) supports other than piping supports.
2410 Inspection Program 2420 Successive inspections 1220 Snubber Inspection Requirements
-2430 Additional Examinations
-2500 Examination Requirements The inservice inspection requirements for snub-
-2510 Supports Selected for Examina' tion bers shall be in accordance with the requirements of
-2520 Method of Examination IWF 5000.
O m
817
CASE (c:ntinued)
N-491 O
CASES OF ASME BOILER AND PRESSURE VESSEL CODE 1230 Supports Exempt from Examination adjusted in accordance with 3000, repaired, or re-placed.
Component supports exempt from the examination (b) For systems that operate at a temperature requirements of -2000 are those connected to com-greater than 200'F during normal plant operation, ponents and items exempted from examination under the Owner shall perform an additional presemce ex-IWB-1220, IWC 1220, IWD-1220, and IWE-1220. In amination on the affected component supports dur-addition, portions of supports that are inaccessible ing r f 11 wing the subsequent system heatup and by being encased in concrete, buried underground, Idown cycle unless determined unnecessary by c
or encapsulated by guard pipe are also exempt from evaluation. This examination shall be performed dur-the examination requirements of -2000.
ing operation or at the next refueling outage.
1300 SUPPORT EXAMINATION BOUNDARIES Support examination boundaries shall be in ac-2400 INSPECTION SCHEDULE cordance with IWF-1300.
2410 Inspection Program (a) Inservice examinations shall be performed
-2000 EXAMINATION AND either during normal system operation or plant out-INSPECTION ages.
(b) The required examinations shall be completed
-2100 SCOPE m accordance with the inspection schedule provided
[
The requirements of this Case apply to the ex-in Table -2410-1 or Table -2410-2.
L amination and inspection of component supports, but (c) The ir.spection period specified in (b) above not to the inservice test requirements of IWF-5000.
may be decreased or extended by as much as one year to enable an inspection to coincide with a plant outage, within the limitations of IWA-2400.
2200 PRESERVICE EXAMINATION (d) Following completion of Program A after 40 years, successive inspection intervals shall follow the 2210 Initial Examination 10 year inspection mterval of Program B.
(a) All examinations listed in Table -25001 shall be performed completely, once, as a preservice ex-
-2420 Successive Inspections amination. These preservice examinations shall be extended to include 100% of all supports not ex.
(a) The sequence of component support exami-empted by -1230.
nations established during the first inspection inter-(b) Examinations for systems that operate at a val shall be repeated during each successive inspec-temperature greater than 200*F during normal plant tion interval, to the extent practical.
operation shall be performed during or following in-(b) When a component support must be subjected itial system heatup and cooldown. Other examina-to corrective measures in accordance with 3000, that tions may be performed prior to initial system heatup support shall be reexamined during the next inspec-and cooldown.
tion period listed in the inspection schedules of the
-inspection programs of 2410.
(c) en a na e trec e measures are n t 2220 Adjustment, Repair, and Replacement required during the next inspection period as a result (a) Prior to return of the system to service, the of the examinations required by (b) above, the in-applicable examinations listed in Table 2500-1 shall spection schedule may revert to the requirements of be performed on component supports that have been (a) above.
818
CASE N-498-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE p
Approval Date: May 11,1994 See Nurneriest index for expiration and any reaffirmation dates.
Case N-498-1 (2) The boundary subject to test pressurization Alternative Rules for 10-Year System Hydrostatic during the system pressure test shall extend to all Testing for Class 1,2, and 3 Systems Class 2 components included in those portions of sys-Section XI, Division 1 tems required to operate or support the safety system function up to and including the first normally closed Inquiry: What alternative rules may be used in lieu valve, including a safety or relief valve, or valve ca-of those required by Section XI, Division 1, Table pable of automatic closure when the safety function IWB-2500-1, Category B-P, Table IWC-2500-1, Cat-is required.
egory C-H, and Table IWD-2500-1, Categories D-A, (3) Prior to performing the VT-2 visual exami.
D B, and D C, as applicable, for the 10-year system nation, the system shall be pressurized to nominal hydrostatic test?
operating 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-erating pressure during performance of the VT 2 Reply:
visual examination.
(a) It is the opinion of the Committee that as an (4) The VT 2 visual examination shall include alternative to the 10-year system hydrostatic test re-all components within the boundary identified in quired by Table IWB-2500-1, Category B-P, the fol-(b)(2) above.
lowing rules shall be used.
(5) Test instrumentation 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 Committer that, as an
[d terval, prior to reactor startup.
alternative to the 10-year system hydrostatic test re-t (2) The boundary subject to test pressurization quired by Table IWD 2500-1, Categories D-A, D-B, during the system leakage test shall extend to all 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, the following tem boundary, rules shall be used.
(3) Prior to performing the VT-2 visual exami-(1) A system pressure test shall be conducted at nation, the system shall be pressurized to nominal or near the end of each inspection interval or during operating 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 during performance of the VT-2 visual ex.
during the system pressure test shall extend to all amination.
Class 3 components includ:d in those portions of sys-(4) Test temperatures and pressures shall not tems required to operate or support the safety system 7
exceed limiting conditions for the hydrostatic test function up to and including the first normally clor,ed 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 (3) Prior to performing the VT-2 visual exami-(a)(2) above.
nation, the system shall be pressurized to nominal (6) Test instrumentation requirements of IWA-operating 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 opinion of the Committee that, as an The system shall be maintained at nominal operating alternative to the 10-year system hydrostatic test re-pressure during performance of the VT-2 visual ex-quired by Table IWC-2500-1 Category C-H, the fol-amination.
lowing rules shall be used.
(4) The VT-2 visual examination shall include f"%
(1) A system pressure test shall be conducted at all components within the boundary identified in
\\j or near the end of each inspection interval or during (c)(2) above, the same inspection period of each inspection inter-(5) Test instrumentation requirements of IWA-val of Inspection Program B.
5260 are not applicable.
943
CASE N-508-1 CASES OF ASME BOILER AND PRESSt*RE VESSEL CODE v
Approval Date: May it 1994 See Numeric Index for expiration and any reaffirmation dates.
Case N-508-1 (a) Items being removed and installed shall be of Rotation of Serviced Snubbers and Pressure the same design and construction; Relief Valves for the Purpose of Testing (b) Items being removed shall have no evidence of Section XI, Division 1 failure at the time of removal; (c) Items being rotated shall be removed and installed only by mechanical means; Inquiry: What altemative rules to those stated in (d) Items being installed shall previously have been IWA-4000 (TWA-7000 for Editions and Addenda prior in service; to the 1991 Addenda) may be used when, for the (c) Preservice inspections and pressure tests shall be purpose of testing, snubbers and pressure relief valves performed as required by IWA-4000 (IWA-7000 for are rotated from stock and installed on components Editions and Addenda prior to the 1991 Addenda);
(includmg piping systems) within the Section XI
(/) The Owner shall maintain a method of tracking boundary?
the items to ensure traceability of inservice inspection and testing records; (g) Use of an NIS-2 form is not required except as Reply: It is the opinion of the Committee that, as provided in (i) below; an alternative to the provisions of IWA-4000 (h) Testing of removed snubbers and pressure relief (IWA-7000 for Editions and Addenda prior to the valves, including required sample expansions, shall be 1991 Addenda) and for the purpose of testing, snub-performed in accordance with the Owner's test program; y/
bers and relief valves may be rotated from stock and (i) Repair or replacement of removed items, when installed on components (including pipmg systems) required, shall be performed in accordane with within the Section XI boundary provided the following IWA-4000 (IWA-4000 or IWA-7000 for Editions and requirements are met:
Addenda prior to the 1991 Addenda).
i\\m 739
CASE N-509 CASES OF AShtE BOILER AND PRESStlRE VESSEL CODE l
Approval Date: November 25,1992 See Numeric Index for expiration and any reaffirmation dates.
Case N 509 (c) Table 2500-1, Examination Category D-A shall Alternative Rules for the Selection and be used for Class 3 integrally welded attachments in
{
Examination of Class 1, 2, and 3 Integrally Examination Categories D-A, D-B. and D-C of IWD.
Attachments 1.1 Exemption Criteria Section XI, Division 1 (a) The exemption enteria provided in IWB-1220.
Inquiry: What attemative requirements to those of IWC-1220, and IWD-1220 may be apphed to Class 1.
IWB, IWC, and ID may be used to select and examine 2, and 3 components respectively, with integrally welded attachments, required to be examined in accordance integrally welded attachments?
with Table 2500-1.
(b) Class 1, 2, and 3 integrally welded attachment Reply: It is the opinion of the Committee that the examinations performed as a result of component sup-followmg rules may be used to select and examine Port deformation cannot be credited under the require-integrally welded attachments:
ments of IWB-2411 or IWB-2412, IWC-2411 or (a) This Case is limited to Examination Categories IWC-2412, and IWD-2411 or IWD-2412, respectively.
B H, B-K-1, C-C, D-A, D-B, and D-C.
(b) Class 1, 2, and 3 component supports shall be 1.2 Inspection Schedule selected for examination in accordance with IWF of Class 1,2, or 3 integrally welded attachments selected j
the 1989 Edition with the 1990 Addenda.
f r examination by sample selection criteria in accord-m (c) Except for the selection of component supports ance with Table 2500-1, Examination Categories B K, f
(
)
for examination, all references to Section XI within C-C, and D-A, shall meet the requirements of IWB-2411 this Case shall be from the edition and addenda specified or IWB-2412, IWC-2411 or IWC-2412, or IWD-241I m the Owner,s inservice Inspection Program.
or IWD-2412, respectively.
1.3 Additional and Successive Examinations (a) Class 1,2. and 3 additional and successive exami-nation requirements of IWB-2430 and IWB-2420 for Class 1. IWC-2430 and IWC 2420 for Class 2 and 3 1.0 SCOPE as applicable, shall be applied to integrally welded These requirements apply to examinat on and sample attachments whose examinations reveal flaws or relevant selection of Class 1,2, and 3 integrally welded attach-conditions that exceed the acceptance standards ments of sessels, piping, pumps, and valves hsted m IWB-3000, IWC-3000, and IWD-3000 respectively.
Table 25001 as follows:
(b) When integrally welded attachinent are exanuned ta) Table 2500-1. Examination Category B K shall as a result of identified component support deformation be used for Class I integrally welded attachments m and the results of and the results of these examinations Exammatmn Catepones B H and B K-1 of IWB exceed the applicable acceptance standards listed above, ih) Table 2500-1. Exemmation Category C-C shall additional or successive exammations shall be performed be u ed for Class 2 integrally welded attachments in wh:n determined necessary based on an evaluation by Examination Category C C of IWC.
the Owner.
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I CASE N-522 CASES OF ASME BOILER AND PRESSURE VESSEL CODE
(}
V Approval Date: December 9,1993 See Numeric Index for expiration and any restfirmation dates.
Case N-522 Pressure Testing of Containment Penetration Piping Section XI, Division 1 Inquiry What alternative to the mies of Table IWC-2500-1. Category C-H may be used for pressure testing piping that penetrates a containment vessel, when the piping and isolation valves that are pan of the containment system are Class 2 but the balance of the piping system is outside the scope of Section XI?
Reply: It is the opinion of the Committee that 10 CFR 50, Appendix J, may be used as an altemative to the mies in Table IWC-2500-1, Category C-H for pressure testing piping that penetrates a containment vessel, when the piping and isolation valves that are part of the containment system are Class 2 but the balance of the piping system is outside the scope of O
Section XI.
(-
U 797
CASE N-524 CASES OF ASME BOILER AND PRESSLJRE VESSEL CODE A
{G Approval Date: August 9,1993 See Numeric Index for expiration and any reaffirmation dates.
Case N 524 Alternative Examination Requirements for Longitudinal Weids in Class 1 and 2 Piping Section XI, Division 1 Inquiry: What alternative requirements may be ap-plied to the surface and volumetric exanunation of longitudinal piping welds specified in Table IWB-2500-1, Examination Category B-J, Table IWC-2500-1 Examination Categories C-F-1 and C-F-2 (Examination Category C-F prior to Winter 1983 Addenda), and Table IWC-2520. Examination Cate-gory C-G (1974 Edition, Summer 1975 Addenda)?
Reply: It is the opinion of the Committee that the following shall apply:
(a) When only a surface examination is required, examination of longitudinal piping welds is not required l
A beyond those portions of the welds within the examina-
' ('j tion boundaries of intersecting circumferential welds.
(b) When both surface and volumetric examinations are required, examination of longitudinal piping welds l
is not required beyond those portions of the welds within the examination boundaries of intersecting cir-cumferential welds provided the following requirements I
are met.
(/) Where longitudinal welds are specified and locations are known, examination requirements shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of longitudi-nal weld within the circumferential weld examination volume; (2) Where longitudinal welds are speciSed but locations are unknown, or the existence oflongitudinal welds is uncertain, the examination requirements shall be met for both transverse and parallel flaws within the entire examination volume of intersecting circumfer-I ential welds.
/~S (L) '
801 1
CASE N-535 CASES OF ASME BOILER AND PRESSURE VESSEL CODE k,
Approval Date: December 12,1994 See Numeric Index for expiration and any reaffirmation dates.
Case N-535 end dates nor the inservice inspection program for the Alternative Requirements for laservice inspection successive interval need be revised.
Intervals (b) Examinations may be performed to satisfy the Section XI, Division 1 requirements of the extended interval in conjunction with examinations performed to satisfy the requirements Inquiry: What alternative to the requirements of of the successive interval. However, an examination IWA-2430(d) may be used for scheduling inservice performed to satisfy requirements of either the extended inspections for components inspected under Program B?
interval or the successive interval shall not be credited to both intervals.
Reply: It is the opinion of the Committee that the (c) That portion of an inspection interval described J
following requirements may be used in lieu of as an inspection period may be reduced or extended j
IWA-2430(d) for scheduling inservice inspections for by as much as one year to enable an inspection to components inspected under inspection Program B:
coincide with a plant outage. This adjustment shall (a) Each inspection interval may be reduced or ex-not alter the requirements for scheduling inspection tended by as much as one year. Adjustments shall not intervals.
cause successive intervals to be altered by more than (d) The inspection interval for which an examination one year from the original pattem intervals. If an was performed shall be indentified on examination inspection interval is extended, neither the start and records.
(O
/N) 837
y CASE N-566
[]
CASES OF ASME BOILER AND PRESSURE VESSEL CODE O
Approval Date: August 9,1996 See Numeric Index for expiration and any reaffirmation dates.
Case N-566 Corrective Action for Leakage Identified at Bolted ConnectionsSection XI, Division 1 Inquiry: What alternative to the requirements of IWA-5250(a)(2) may be used when leakage is detected at bolted connections?
Reply: It is the opinion of the Committee that, as an alternative to the requirements of IWA-5250(a)(2),
one of the following requirements shall be met for leakage at bolted connections:
(a) The leakage shall be stopped, and the bolting and component material shall be reviewed for joint integrity.
(b) If the leakage is not stopped, the joint shall be evaluated in accordance with IWB-3142.4 for joint integrity. This evaluation shall include consideration of
(,/
the number and condition of bolts, leaking medium, bolt and component material, system function, and leakage monitoring.
bG 929