ML20082L545
| ML20082L545 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 04/18/1995 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20082L529 | List: |
| References | |
| PROC-950418, NUDOCS 9504210203 | |
| Download: ML20082L545 (133) | |
Text
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l Revision 3 l
4 INSERVICE INSPECTION PLAN l FOR l
BRAIDWOOD NUCLEAR GENERATING STATION .
UNITS 1 AND 2 .
l 1ST INSPECTION INTERVAL Commercial Service Date Unit 1: July 29,1988 Commercial Service Date Unit 2: October 17,1988 Eraidwood Nuclear Station R.R. #1 Box 84 Braceville, Blinois 60407 Comed P.O. Box 767 Chicago, Blinois 60690 l l
zeenc\inserv.wpf l ADOC 0 0 56 O PDR l
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Revision 3 BRAIDWOOD NUCLEAR STATION, IRfITS 1 AND 2, INSERVICE INSFECTIcel PLAN, 1ST INTERVAL REVISION
SUMMARY
SHEET Effective Page(s) Rev.
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1-1 3 t 2-1 to 2-8 3 3-1 to 3-3 3 <
3-4 to 3-5 2 ,
3-6 2 3-7 to 3-8 2 3-3 3 3-10 2 3-11 to 3-12 2 1 3-13 2 ,
3-14 2 3-15 to 3-21 3 3-22 2 3-23 2 1 3-24 to 3-25 4 3-26 3 l 3-27 2 -
3-28 to 3-30 2 3-31 2 3-32 3 3-33 to 3-35 3 3-36 1 ,
3-37 to 3-38 0 1 3-39 to 3-40 0 l 3-41 to 3-42 0 1 4-1 to 4-2 3 4-3 to 4-18 2 4-19 1 4-20 to 4-21 2 4-22 to 4-26 0 5-1 3 5-2 to 5-30 2 5-31 3 6-1 to 6-3 2 6-4 to 6-5 1 7-1 to 7-2 1 7-3 1 7-4 to 7-7 2 7-8 to 7-9 1 8-1 3 9-1 to 9-7 3 10-1 3 l
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TABLE OF CCtfTENTS Section Description Page >
1.0 Introduction and Plan Description 1-1 9
2.0 Piping P&ID's and Isometric Drawings 2-1 l 1
3.0 Relief Requests 3-1 l 4.0 Technical Approach and Positions 4-1 '
5.0 ISI Plan Summary 5-1 ,
6.0 Program Plan for Component Support 6-1 7.0 Program Plan for Snubbers 7-1 f 8.0 Augmented Inservice Inspection Requirements 8-1 i
9.0 Exempt Component Summary 9-1 J 10.0 References I 10-1 ;
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LIST OF TABIJ:S l Table Description Page 2.0-1 List of piping and Instrumentation Diagrams 2-2 ,
(sorted by System name) !
2.0-2 List of Piping Isometrics 2-3 2.0-3 Inservice Inspection Line List Class 2-6 1 and 2 (
3.0-1 Relief Requests Summaries 3-2 3 4.0-1 Technical Approach and Position Summaries 4-2 5.0-1 ISI Plan Summary 5-2 5.0-2 Exam Method Abbreviations 5-31 I 9.0-1 Exempt Component Listing 9-2 l
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Revision 3
1.0 INTRODUCTION
AND PLAN DESCRIPTION I This Inservice Inspection (ISI) Plan for non-destructive examination of ISI Class 1, 2, and 3 components was developed in accordance with the inservice inspection requirements in ASME Section XI, Subsections IWA, IWB, IWC, IWD, and IWF. All references to the " Code" in this Plan are taken from the ASME Boiler and Pressure Vessel Code,Section XI, Division I, " Rules for Inservice .
Incpection of Nuclear Power Plant Components", the 1983 Edition through and including the Summer 1983 Addenda. Per 10CFR50.55a (B) (2) (iv), Class 2 portions of Emergency ~Uoro Cooling Systems (ECCS), Containment Heat Removal Systems, and Residual *deat Removal Systems were selected / exempted based upon the 1974 Edition through the Summer 1975 Addenda of ASME Section XI.
This Inservice Inspection Plan will be effective from July 29, 1988 to July 28, 1998 and from October 17, 1988 to October 16, 1998 for Braidwood Units 1 1 and 2, respectively. ;
1 This plan covers non-destructive examination of piping welds and components j only. Inservice Testing (pumps and valves) at Braidwood Nuclear Station Units 1&2 is defined in a separate plan.
The key features of this plan are; the relief requests, the explanation of I technical approach and positions, and the definition of scope for the Inservice Inspection Plan. The details of the Inservice Inspection Plan are :
addressed in the ISI data base available at Braidwood Station. Component detail drawings, piping isometric drawings, a component listing of each bolt, weld, vessel internal, etc., administrative procedures, NDE procedures,"and other records required to define and execute the Inservice Inspection Plan are all retained at Braidwood Station.
The construction code (s) used for general categories of components include:
- ASME See III 1974 Edition through Summer 1975 Addenda
, (piping systems)
+ ASME Sec. III 1974 Edition through Summer 1974 Addenda j (snubbers and component support) 1 This ISI Plan does not address the following examination requirement:
Subject Controller Program
+ Snubbers Technical Specification 3/4.7.8 l
+
Steam Generator Tubing Technical Specification 3/4.4.5 System Classification Class 1 piping including in the Reactor Coolant Pressure is defined by 10CFR50.2(v). Class 2 and 3 piping was classified by Sargent and Lundy Engineering utilizing the guidelines of U.S. NRC Regulatory Guide 1.26. All ASME piping is installed to the requirements of the ASME Boiler and Pressure Vessel Code 1974 Edition and Addenda through the Summer 1975 Addenda,Section III, " Nuclear Power Plant Components".
Specific information with regards to individual component construction codes or materials can be found in the Byron /Braidwood Final Safety Analysis Report.
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A Revision 3 2.0 PIPING P&ID'S AND ISCMETRIC DRANINGS The Braidwood Unit 1 and Unit'2 Piping and Instrumentation Diagrams (P&ID's) and Piping Isometric Drawings corresponding to systems included in the ISI Program have been identified in Table 2.0-1 and Table 770-2 (sorted by System name).
A controlled :et of P&ID's and Piping Isometric Drawings are maintair.eu at Braidwood Station under Piping Specification 2907, Vendor Code P212, and are available upon request.
Table 2.0-3 is a listing of all ASME class 1 and 2 Piping Systens which require non-destructive examination and visual examination in accordance with the requirements of ASME Section XI Subsections IWA, IWB, and IWC. Class 3 visual examination requirements will be found in tables 5.0-1 in accordance with Subsection IWD.
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Table 2.0-1 l List of P&ID's l System Name P&ID Number Chemical Feed & Volume Control Unit 1: M-060-1B, M-060-02, M-064-01, M-064-02, M-064-05. I' Unit 2: M-135-1A, M-135-1B, M-135-02, M-138-01, M-138-02 i M-138-05 i
Containment Spray Unit 1: M-04 6-1A, M-04 6-1C, i M-061-04 Unit 2: M-129-1A, M-129-1C, ,
M-136-04 i Feedwater Unit 1: M-036-1A, M-036-1B, M-036-1C, M-036-1D Unit 2: M-121-1A, M-121-1B, M-121-1C, M-121-1D Main Steam Unit 1: M-035-01, M-035-02, Unit 2: M-120-01, M-120-2A, M-120-2B Residual Heat Removal Unit 1: M-062-01, M-062-04, Unit 2: M-136-04, M-137-01, Reactor Coolant & Pressurizer Unit 1: M-060-1A, M-060-1B, M-060-02, M-060-03, M-060-04, M-060- )
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Unit 2: M-135-1A, M-135-1B, 1 M-135-02, M-135-03, M-135-04 Safety Injection Unit 1: M-061-02, M-061-03, M-061-04, M-061-05, M-061-06 Unit 2: M-136-02, M-136-03, M-136-04, M-136-05, M-136-06 2-2
Revision 3 Table 2.0-2 (sheet 1 of 3)
List of Isometric Drawings Specification 2907, Vendor Code P212 System Containment Spray Feedwater (Cont.)
1/2 CS-01 1/2 W-10 1/2 CS-02 1/2 W-11 1/2 CS-03 1/2 W-12 1/2 CS-04 1/2 CS-05 Main Steam i 1/2 CS-06 1/2 CS-07 1/2 MS-01 1/2 MS-02 Chemical & Volume Control 1/2 MS-03 1/2 MS-04 1/2 CV-01 1/2 MS-05 1/2 CV-02 1/2 MS-06 1/2 CV-03 1/2 MS-07 ';
1/2 CV-04 1/2 MS-08 1/2 CV-05 1/2 CV-06 Reactor Coolant 1/2 CV-07 1/2 CV-08 1/2 RC-01 l 1/2 CV-09 1/2 RC-02 !
1/2 CV-10 1/2 RC-03 1/2 CV-11 1/2 RC-04 1/2 CV-12 1/2 RC-05 1/2 CV-13 1/2 RC-06 1/2 CV-14 1/2 RC-07 1/2 CV-15 1/2 RC-08 1/2 CV-16 1/2 RC-09 1/2 RC-10 Feedwater 1/2 RC-11 1/2 RC-12 1/2 W-01 1/2 RC-13 1/2 W-02 1/2 RC-14 1/2 W-03 1/2 RC-15 1/2 W-04 1/2 RC-16 1/2 W-05 1/2 RC-17 1/2 W-06 1/2 RC-18 1/2 W-07 1/2 RC-19 1/2 W-08 1/2 RC-20 1/2 W-09 1/2 RC-21 2-3
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Revision 3 Table 2.0-2 (sheet 2 of 3)
Systems Reactor Coolant (Cont.) Safety Injectf.on (C$nt.)
1/2 RC-22 1/2 SI-05 1/2 RC-23 1/2 SI-06 !
1/2 RC-24 1/2 SI-07 1/2 RC-25 1/2 SI-08 i 1/2 RC-26 1/2 SI-09 1/2 RC-27 1/2 SI-10 1/2 RC-28 1/2 SI-11 l 1/2 RC-29 1/2 SI-12 1/2 RC-30 1/2 SI-13 1/2 RC-31 1/2 SI-14 1/2 RC-32 1/2 SI-15 1/2 RC-33 1/2 SI-16 1/2 RC-34 1/2 SI-17 !
1/2 RC-35 1/2 SI-18 l 1/2 RC-36 1/2 SI-19 ;
1/2 RC-37 1/2 SI-20 1/2 RC-38 1/2 SI-21 !
1/2 RC-39 1/2 SI-22 1/2 RC-40 1/2 SI-23 1/2 RC-41 1/2 SI-24 -
1/2 RC-42 1/2 SI-25 ;
1/2 SI-26 !
Residual Heat 1/2 SI-27 l 1/2 SI-28 !
1/2 RH-01 1/2 SI-29 l 1/2 RH-02 1/2 SI-30 '
1/2 RH-03 1/2 SI-31 1/2 RH-04 1/2 SI-32 1/2 RH-05 1/2 RH-06 Pressurizer 1/2 RH-07 1/2 RH-08 1/2 PZR-01 1/2 RH-09 Reactor Vessel Safety Injection 1/2 RV-01 1/2 SI-01 1/2 RV-02 1/2 SI-02 1/2 RV-03 1/2 SI-03 1/2 SI-04 I
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Revision 3 Table 2.0-2 (sheet 3 of 3)
System Class 2 ECCS Systems RHR Heat Exchanger Chemical & Volume Control 1/2 RHX-01 1A-CV-03 1A-CV-3A Steam Generator 2A-CV-16 2A-CV-17 l 1/2 SG-01 2A-CV-18 i 1/2 SG-02 1/2 SG-03 Containment Spray ;
1/2 SG-04 '
1/2 CS-01 Reactor Coolant Pump 1/2 CS-02 ,
1/2 CS-03 +
1/2 RCP-01 1/2 CS-04 1/2 CS-05 Containment Spray Pump 1/2 CS-06 ;
1/2 CS-07 ;
1/2 CSP-01 i Residual Heat Removal l CV & RHR Pumps 1 1/2 RH-03 1/2 RHP-01 1/2 RH-04 1/2 RH-05 !
1/2 RH-06 I 1/2 RH-07 ;
1/2 RH-08 1/2 RH-09 l 1
Safety Injection lA-SI-13 .
lA-SI-14 i 2A-SI-21 2A-SI-22 2A-SI-23 2A-SI-40 2SI-12 1/2 SI-25 :
1/2 SI-32 !
1 3I-33 2 SI-28 2 SI-30 .
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Table 2.0-3 (sheet 1 of 3)
Class 1 and 2 lines which may be subject to Non-Destructive examination.
Chemical Feed & Volume Control .
I CVA3AB-2" CVA6AB-2" CV?.0DB-3" CV14GA-1.5" l CVA3B-2" CVA7AA-2" CV14FA-2" CV14GB-1.5" !
CVA3AA-2" CVA7AB-2" CV14 FB-2" CV14GC-1.5" CVA5AA-2" CVB7A-3" CV14FC-2" CV14GD-1.5" i CVA5AB-2" CV10DA-3" CV14 FD-2 " CV45B-2" CVA6AA-2" i
- WO3DA-16" *FW81AB-6"
- W81BC-6" *W86AD-16"
- FWO3DB-16" *W81AC-6" *FW81BD-6" *FW87CA-6" .
- WO3DC-16"
- W81AD-6" *FW86AA-16" *W87CB-6"
- WO3 DD-16" *W81BA-6"
- W 8 6AB-16" *W87CC-6" [
- W81AA-6"
- W81BB-6" *FW86AC-16" *FW87CD-6" !
- WO 3 CA-16" *FWO3CB-16" *FWO3CC-16" *WO3CD-16" i Main Steam '
- MS01AA-30.25" *MS01CD-30.75" *MSO9AC-6" *MS12AB-6" I
- MS01AB-32.75" *MS07AA-28" *MSO9AD-6" *MS12AC-6" *
- MS01AC-32.75" *MS07AB-23" *MS10AA-6" *MS12AD-6"
- MS01AD-30.25" *MS07AC-28" *MS10AB-6" *MS13AA-8" '
'MS01BA-30.25" *MS07AD-28" *MS10AC-6" *MS13AB-8" !
- MS01BB-32.75" *MSOBAA-6" *MS10AD-6" *MS13AC-8"
- MS01BC-32,75" *MSOBAB-6" *MS11AA-6" l
- MS13AD-8"
- MS01BD-30.25" *MSOBAC-6" *MS11QB-6" *MS143AA-12" l
- MS01CA-30.25" *MSOBAD-6" *MS11AC-6" *MS143AB-12" l
- MS01CB-32.75" *MSO9AA-6" *MS11AD-6" *MS143AC-12"
- MS01CC-32.75" *MSO9AB-6" *MS12AA-6" *MS143AD-12" Residual Heat Removal
- RH01AA-12" RH01CA-16" RH03AA-8" RH12A-8"
- RH01AB-12" RH01CB-16" RH03AB-8" RH14A-B" RH01BA-12" RH02AA-8" RH09AA-8" RH01BB-12" RH02AB-8" RH09AB-8"
- These lines may be subject to volumetric & surface examination.
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Table 2.0-3 (sheet 2 of 3) 1 l
Safety Injection SIO3DA-2" *SIOSCB-8" SIO8GC-1.5" *SIO9BB-10" SIO3DB-2" *SIO5CC-8" SIO8GD-1.5" *SIO9BC-10" SIO3FA-2" *SIOSCD-8" SIO8HA-2.0" *SIO9BD-10" SIO3FB-2" *SIO5DA-6" SIO8HB-2" SI18FA-2"
- 2SIO3GB-6" *SIO5DB-6" SIOBHC-2" SI18FB-2"
- 2SIO3GA-6" *SIO5DC-6" SIO8HD-2" SI18FC-2"
- SIO4A-12" *SIO5DD-6" SIOBJA-1.5" SI18FD-2"
- SIO4B-12" *SIO6BA-24" SI08JB-1.5" SI34A-8"
- SIO4C-8" *SIO6BB-24" SI0BJC-1.5" *SI82BA-12"
- SIO4D-8" S1086B-1.5" SIO8JD-1.5" *SI82BB-12" SIO5AA-8" SIOBE-3" SIO9AA-10" *SIA4A-8" SIO5AB-8" SIO8FA-3" SIO9AB-10" *SIA4B-8"
- SIO5BA-8 SIO8FB-3" SIO9AC-10"
- SIO5BB-8" SIO8GA-1.5" SIO9AD-10"
- SIOSCA-8" 2SIO8GB-1.5" *SIO9BA-10" Reactor Coolant & Pressurizer :
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- RC01AA-29" RC09CB-2" RC14AA-2" *RC21BB-8"
- RC01AB-29" RC09CC-2" RC14AB-2" *RC21BC-8"
- RC01AC-29" RC09CD-2" RC14AC-2" *RC21BD-8"
- RC01AD-29" RC09DA-2" RC14AD-2" RC22AA-1.5" .
- RCO2AA-31" RC09DB-2" RC16AA-2" RC22AB-1.5"
- RCO2AB-31" RC09DC-2" RC16AB-2" RC22AC-1.5"
- RCO2AC-31" RC09DD-2" RC16AC-2" RC22AD-1.5"
- RCO2AD-31" RC09EA-3" RC16AD-2" *RC24AA-4"
- RC03AA-27.5" RC09EB-3" RC17AA-2" RC24AB-4"
- RC03AB-27.5" RC09EC-3" RC17AB-2" RC26A-2"
- RC03AC-27.5" RC09ED-3" RC17AC-2" RC28A-3"
- RC03AD-27.5" RC09FA-3" RC17AD-2" *RC29AA-10"
- RCO4AA-12" RC09FB-3" RC17BA-2" *RC29AB-10"
- RC04AB-12" RC09FC-3" RC17BB-2" *RC29AC-10"
- RC05AA-6" RC09FD-3" RC17BC-2" *RC29AD-10"
- RC05AB-6" RC10DA-3" RC17BD-2" RC30AA-1.5" RC09BA-2" RC10DB-3" *RC21AA-8" RC09BB-2" RC30AB-1.5" RC13AA-2" *RC21AB-8" RC30AC-1.5" RC09BC-2" RC13AB-2" *RC21AC-B" RC09BD-2" RC30AD-1.5" RC13AC-2" *RC21AD-8" *RC35AA-6" RC09CA-2" RC13AD-2" *RC21BA-8" *RC35AB-f"
- - volumetric exam (ultrasonic) required 2-7 m
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Table 2.0-3 (sheet 3 of 3)
Reactor Coolant & P ssurizer RC36A-3" RC53DA-2" RC37A-3" RYO1AA-4" RC43AA-2" RY01AB-4" i
RC43AB-2" *RYO1B-6" RC43AC-2" RYO1C-4" RC43AD-2" *RYO2A-6" RC45AA-3" RYO2B-3" RC45AB-3" *RYO3AA-6" RC45AC-3" *RYO3AB-6" RC45AD-3" *RYO3AC-6" RC46AA-3" *RYO3BA-6" RC46AB-3" *RYO3BB-6" RC46AC-3" *RYO3BC-6" RC46AD-3" RYO6A-3" RC53AA-2" *RY11A-14" RC53BA-2" RYl8A-2" RC53CA-2" RY76A-2" containment Spray CS01AA-16" CS01AB-16" -
CS23AA-14" CS23AB-14" CS02AA-10" CS02A3-10" CS10AA 6" CS10AB-6" CS06AA-6" CS06AB-6"
- volumetric exam (ultrasonic) required 2-8
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Revision 3 d
3.0 RELIEF REQUESTS Pursuant to 10CFR.55a (g) (5) (iv) , Relief Requests have been included when I specific requirements in the Code are considered impractical. The enclosed .
Relief Requests are subject to change throughout the inspection interval. If testing requirements are determined to be impractical during the course of the ;
interval, additional or modified Relief Requests will be submitted in accordance with.10CFR50.55a (g) (4) (iv) . ;
A sununary of relief requests is provided in Table 3.0-1. l l
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Table 3.0-1
__ Relief moquest susunaries (sheet 1 of 2)
NR-1 Inaccessible welds due to saddle plate location in-the Main Steam System.
NR-2 Cast stainless steel to cast stainless steel welds in which ultraconic tests can only detect 25% or greater thru wall defects. J NR-3 VT-3 on class 1 large bore >4" valve body internals will be performed only when valve is disassembled for maintenance with only one valve from each design group being inspected.
NR-4 Withdrawn (Ultrasonic exams of nozzle inner radius).
NR-5 Cast stainless steel (elbow) to cast stainless steel (nozzle) in which ultrasonic tests can only detect 25% or greater thru wall defects and 6 circumferential scan lidtations. ,
NR-6 Cast stainless steel (elbow) to stainless steel pipe in which ultrasonic tests can only detect 25% or greater thru wall defects and i cirewmferential scan limitations.
NR-7 Scan Obstructions on 6 welds within the Reactor Coolant System.
NR-8 Deleted (Ultrasonic examinations on 2 316 stainless steel weldolets).
NR-9 Core support guide lug obstructions on Reactor Vessel welds (2 welds per unit). <
NR-10 Deleted (refer to technical position, Note 13)
NR-ll Deleted (refer to technical position, Note 13)
NR-12 No volumetric exam on the RHR Heat Exchanger Nozzle inner radius and a best effort ultrasonic exam on the nozzle to vessel weld. ;
NR-13 Deleted (reducer weld best effort Ultrasonic exam)
NR-14 VT-3 on Reactor Coolant Pump internals to be performed when pump is disassembled for maintenance only (1 pump per interval).
NR-15 Pump lug obstructions on the RHR and CV pumps.
NR-16 Weld obstructions in che Main Steam and Feedwater systems. I NR-17 Deleted (refer to technical position Note 13). l NR-18 Weld obstruction on 2 Reactor Coolant System welds.
NR-19 Deleted (pipe branch connection weld best effort ultrasonic exam, weld
- 2SI-02-45).
NR-20 No Ultrasonic Examination on the Unit One Steam Generator Nozzle Inner Radius Sections. l l
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Table 2.0-1 l Relief Request & w e ies (sheet 2 of 27 l NR-21 Alternate Hydrostatic Pressure Test Requirements for ASME Class 1, 2 and 3 repaired or replaced components. !
NR-22 Alternate rules for 10 year Hydrostatic Pressure Testing for Class 3 '
systems. ;
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Revision 2
REMEF REQUEST NR-1 i
- 1. SYSTD( Main Steam >
- 2. NUMBER OF ITEMS: 40 Line Number Weld Number '
1MS07AA-28" 1MS-04-25, IMS-04-26, IMS-04-27, -
1MS-04-28, 1MS-04-29 IMS07AB-28" *1MS-06-43, *1MS-06-44, *1MS-06-45, '
- 1MS-06-46, *1MS-06-47 l
1MS07AC-28" *1MS-08-25, *1MS-08-26, *1MS-08-27, ;
- 1MS-08-28, *1MS-08-29 r IM507AD-28" *1MS-02-37, *1MS-02-38, *1MS-02-39,
2MS07AB-28" *2MS-06-34, *2MS-06-37, *2MS-06-40, '
- 2MS-06-43, *2MS-06-46 i 2MS07AC-28" *2MS-08-31, *2MS-08-34, *2MS-08-37,
- 2MS-08-40, *2.MS-08-43 i 2MS07AD-28" *2M3-02-36, *2MS-02-38, *2MS-02-42, ,
- 2MS-02-44, *2MS-02-47 :
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- 3. A.S.M.E. CODE CLASS: 2 !
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Examination Category C-F, Item C5.31 l requires a surface examination of the area described in Figures IWC-2500-9 to ;
IWC-2500-13. The required area is 100% of each circumferential weld requiring l examination. !
- 5. BASIS FOR RELIEF: The above listed welds are inaccessible to surface examinations, due to the location of saddle plates over the pressure retaining welds.
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- 6. ALTERNATE TEST METHOD: A surface examination, (magnetic particle) and visual examination (leak test) will be performed on the saddle plate fillet welds in lieu of the required surface examinations for those pressure retaining welds l selected for examination. l t
- 7. JUSTIFICATION: Performing a surface and visual examination of the saddle plate fillet welds provides an acceptable level of structural integrity for system operation. Radiograph performed on the branch pipe circumferential welds verify its structural integrity at the time of construction. Therefore, and an adequate level of plant operational safety can be assured on these welds. :
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Revision 2 RELIEF REQUEST MR-1 (cont. )
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month inspection interval. <
- These welds are not selected for inspection as required-by code. However, they are examined within the scope of augmented high energy lines as described in Note 5.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
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Revision 2 RELIEF REQUEST NR-2
- 1. SYSTEM: Reactor Coolant
- 2. NUMBER OF ITEMS: 6 Line Number Weld Number Line Number Weld Number 1RC01AA-29" 1RC-01-4 2RC01AA-29" 2RC-01-4 1RC02AA-31" 1RC-01-17 2RC02AA-31" 2RC-01-17 s 2RC01AD-29" 2RC-04-5 '
2RC02AD-31" 2RC-04-18
- 3. A.S.M.E. CODE CLASS: 1
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category B-J, Item B.9.11 requires surface and volumetric examination of the regions {
described in Figure IWB-2500-8 for piping 4 in. nominal pipe size and greater.
This examination includes essentially 100% of the weld length. In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.D. notches with a depth of 10% wall thickness.
- 5. BASIS FOR RELIEF: The above listed welds are cast stainless elbows to either cast pumps or cast valves. The optimized ultrasonic technique used for the statically cast stainless steel welds will detect large flaws (25% or greater through the wall) therefore, this sensitivity is less than that required by code.
In addition, due to the unwieldy characteristics of the contoured wedge search units and variations in the machined surfaces, the welds listed above experience axial and circumferential scanning limitations.
- 6. ALTERNATE TEST METHOD: None.
- 7. .?USTIFICATION: The structural integrity of these welds shall be assured by: ;
- a. Performing a best effort ultrasonic exam, based on state of the art techniques, of the cast stainless welds,
- b. Performing surface examinations of the required region.
- c. Performing a system leakage test each refueling outage and a hydrostatic test each interval.
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month ,
inspection interval. !
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
I l
3-6
. - - . .- - - _ _ .--~ - _ - . -.
Revision 2 .
RELIEF REQUEST NR-3 ,
- 1. SYSTEN: Reactor Coolant, Pressurizer, Residual Heat Removal, and Safety -
Injection.
- 2. NUMBER OF ITEMS: 18 (Groups).
Reactor Coolant IRC8001-A 1RC8002-A 1RC8003-A 1RC8001-B 1RC8002-B 1RC8003-B .
IRC8001-C 1RC8002-C 1RC8003-C l 1RC8001-D 1RC8002-D 1RC8003-D Pressurizer 1RY8010-A-1RY8010-8 -
1RY8010-C. ;
Residual Heat 1RH8701-A Removal 1RHB701-B ,
1RH8702-A .
1RH8702-B Safety Injection 1SI8808-A 1SI8818-A 1SI8841-A 1SI8808-B ISI8818-B 1SI8841-B 1 SIB 808-C 1SI8818-C 1SI8808-D 1SI8948-A ISIB956-A ISI8949-A ISI8948-B ISI8956-B 1SI8949-B '
1SI8948-C 1SI8956-C 1 SIB 949-C ,
1SI8948-D ISI8956-D ISI8949-D Reactor Coolant 2RC8001-A 2RC8002-A 2RC8003-A 2RC8001-B 2RC8002-B 2RC8003-B i 2RC8001-C 2RC8002-C 2RC8003-C e
, 2RC8001-D 2RC8002-D 2RC8003-D Pressurizer 2RY8010-A l 2RY8010-B l 2RY8010-C Residual Heat 2RH8701-A Removal 2RH8701-B 2RH8702-A 2RH8702-B Safety Injection 2SI8808-A 2SI8818-A 2SI8841-A l 2 SIB 808-B 2SI8818-B 2SI8841-B- '
2SI8808-C 2SI8818-C 2SI8808-D 2SI8818-D '
2SI8948-A 2SI8956-A 2 SIB 949-A 2SI8948-B 2SI8956-B 2SI8949-B 2SI8948-C 2SI8956-C 2SI8949-C 2 SIB 948-D 2SI8956-D 2 SIB 949-D
- 3. A.S.M.E. CODE CLASS: 1 !
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Subsection IWB, Table IWB-2500-1, Examination Category B-M-2, Item B12.50 requires visual examination (VT-3) of valve body internal surfaces exceeding 4 in. nominal pipe size. Examinations shall be conducted on one valve within each group of valves that are of the same construction, and nanufacturing method and that perform similar functions in the system, each inspection interval.
3-7
Revision 2 RELIEF REQUEST NR-3 (Cont. )
- 5. BASIS FOR RELIEF: The code requirement to disassemble one valve from each design group for the purpose of visual examination results in high radiation exposure and does not produce a proportionally higher potential for identification of service induced flaws or degradation._ The industrial performance of these valves have proven their excellent ability to resist service degradation or flawing. The inappropriate balance of potential flaw detection and the large impact on expenditures of manpower without substantially increasing component reliability is considered impractical.
Coupled with this is the highly negative impact on the Station's ALARA program. Therefore, the examination frequency is not appropriate for the above listed valves.
- 6. ALTERNATE TEST METHOD: Visual examination, (VT-3) will be performed on one valve from each design group in each system performing a similar function when disassembly is required for maintenance purposes but not to exceed once per inspection interval per design group. In addition, the visual examination leakage test will be performed after each refueling outage.
- 7. JUSTIFICATION: The licensee has specified valves which are free of pressure boundary welds. The above listed Clars 1 valves are free of any pressure boundary welds. Therefore, there is very little likelihood of degradation or flow propagation in these valves. Since these valves have not shown significant service induced degradation in the past, deferral of the internal inspections will not reduce the assurance that their structural integrity has been maintained.
- 8. APPLICABLE TIME PERIOD: This relief will b'e required for t'he first 120 month inspection interval. Note, this inspection is deferrable until the end of the inspection interval and application of this relief may not be required if maintenance is performed on one valve from each group during the first inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
3-8
i i
Revision 3 !
1 RELIEF REQUEST NR-4 j j
l i
l 1
i Relief withdrawn a
3-9
Revision 2 RELIEF REQUEST NR-5
- 1. SYSTEM: Reactor Coolant
- 2. NUMBER OF ITEMS: 16 Cast Stainless Steel SA-351-CF8A (Elbow) to Cast Carbon Steel SA-216 GR-WCC (Nozzle)
Line Number Weld Number Line Number Weld Number 1RC01AA-29" 1RC-01-8 2RC01AA-29" 2RC-01-8 1RCO2AA-31" 1RC-01-9 2RC02AA-31" 2RC-01-9 1RC01AB-29" 1RC-02-19 2RC01AB-29" 2RC-02-19 1RC02AB-31" 1RC-02-23 2RC02AB-31" 2RC-02-23 1RC01AC-29" 1RC-03-8 2RC01AC-29" 2RC-03-6 1RCO2AC-31" 1RC-03-9 2RCO2AC-31" 2RC-03-9 1RC01AD-29" 1RC-04-9 2RC01AD-29" 2RC-04-9 1RC02AD-31" 1RC-04-10 2RC02AD-31" 2RC-04-10
- 3. A.S.M.E. CODE CLASS: 1
- 4. A.S.H.E. CODE SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category B-F, Item B5.70 requires surface and volumetric examination of the regions described in Figure IWB-2500-8 for all Steam Generator nozzle-to-safe end welds. This examination includes essentially 100% of the weld length. 'In addition, Appendix III Supplement 7, requires that ultrasonic examination sensitivity be established using I.D. notches with a depth of 10% wall )
thickness. j
- 5. BASIS FOR RELIEF: The welds listed above are cast austenitic stainless steel l SA-351-CF8A to cast carbon steel SA-216 GR-WCC with austenitic stainless steel cladding. The techniques used to examine the cast stainless side of these ;
welds will detect large flaws (25% or greater through the wall) therefore, i this sensitivity is less than that required by code. In addition, Welds 1RC- 1 02-23, 2RC-02-23 had limited circumferential scans near the weld toe area.
This limitation is due to the transducer system's inability to maintain coupling while scanning over the weld-base metal transition.
- 6. ALTERNATE TEST METHOD: None.
- 7. JUSTIFICATION: The structural integrity of these welds shall be assured by: I
- a. Performing an ultrasonic examination from the carbon steel side.
- b. Perforndnp a best effort ultrasonic examination based on state of the ;
art techniques from the cast stainless side. I
- c. Performing surface examinations of the required regions.
- d. Performing a system leakage test each refueling outage and a system I hydrostatic test each interval.
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
3-10
Revision 2
-RELIEF REQUEST NR-6 A. SYSTEM: Reactor Coolant
- 2. NUMBER OF ITEMS: 10 Cast Stainless Steel SA-351-CF8A (Elbow) to Stainless Steel SA-376 Type 304N (Pipe)
Line Number Weld Number Line Number Weld Number 1RCO2AA-31" 1RC-01-10, 16 2RC02AA-31" 2RC-01-10, 16 1RC03AA-27.5" 1RC-01-30 2RC03AA-27.5" 2RC-01-30 2RC02AD-31" 2RC-04-17 Cast Stainless Steel SA-351-CF8 (Pump) to Stainless Steel SA-376 Type 304N (Pipe)
Line Number Weld Number Line Number Weld Number 1RC03AA-27.5" 1RC-01-18* 2RC03AA-27.5 2RC-01-18*
Cast Stainless Steel SA-351-CF8M (Valve) to
- Stainless Steel SA-376 Type 304N (Pipe)
Line Number Weld Number IRC03AA-27.5" 1RC-01-22
- 3. A.S.M.E. CODE CLASS: 1 4.
A.S.M.E. CODE SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category B-J, Item B9.11 requires surface and volumetric examination of the regions described in Figure IWB-2500-8 for piping 4 in. nominal pipe size and greater.
This examination includes essentially 100% of the weld length. In addition, Appendix III, Supplement 7, requires that ultrasonic examination sensitivity be established using I.D. notches with a depth of 10% wall thickness.
5.
BASIS FOR RELIEF: The above listed welds are cast stainless steel to wrought stainless pipe. The optimized ultrasonic technique used for the statically cast stainless steel welds will detect large flaws (25% or greater through the wall) therefore this sensitivity is less than that required by code. In addition, due to the unwieldy characteristics of the contoured wedge search units and variations in the nachined surfaces those welds listed with an asterisk, experienced axial and circumferential scanning limitations.
- 6. ALTERNATE TEST METHOD: None.
1 3-11
Revision 2 melief moquest MR-6 (Coat. )
- 7. JUSTIFICATION: The Structural integrity of these welds shall be assured by:
- a. Performing an ultrascaic exam from the non-cast side,
- b. Performing a best effort ultrasonic examination based on state of the art techniques, from the case stainless side,
- c. Performing surface examinations of the required regions,
- d. Performing a system leakage test each refueling outage and a system hydrostatic test each interval.
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month '
inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
e
]
3-12
I Revision 2 RELIEF REQUEST NR-7
- 1. SYSTEM: Reactor Coolant _
- 2. NUMBER OF ITEMS: 6 Weld # Line # Unable to be Examined Reason for limited Exam 1RC-162 1RYO1C-4" 16% Elbow inner radius, reducer geometry 1RC-16-5 1RYO1B-6" 16% Elbow inner radii 1RC-32-1 1RYO3AA-6" 16% Elbow inner radius, nozzle geometry 1RC-32-7 1RYO3AB-6" 16% Elbow inner radius, nozzle geometry 1RC-32-13 1RYO3AC-6" 16% Elbow inner radius, nozzle geometry 1RC-35-1 1RYO2A-6" 16% Elbow inner radius, nozzle geometry
- 3. A.S.M.E. CODE CLASS: 1
- 4. A.S.M.E. SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category B-J, Item B9.ll requires volumetric and surface examination of the areas described in Figure IWB-2500-8 for essentially 100% of the weld length.
- 5. BASIS FOR RELIEF: The above listed welds have interfering conditions on each side of the weld. These interferences can cause: poor coupling of the transducer, limited movement of the transeucer, redirecting of the sound beam and, in some cases, complete restriction of a particular scan. These conditions sufficiently limit the axial scans so as to leave the listed percent of weld length uninspected.
- 6. ALTERNATE TEST METHOD: All welds shall receive the required Section XI surface examination in addition to the best effort ultrasonic examination.
7 JUSTIFICATION: The estimates of weld lengths unable to be examined are extremely conservative and are actually a percent of weld length for which there is less confidence that the entire required weld volume will be examined. Based on the required surface and leakage tests the structural integrity of this weld shall be assured.
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
I i
3-13
1 l
i
)
Revision 2 i i
RELIEF REQUEST NR-8 i
~
l l
l l
J Relief Deleted I
3-14
1 Revision 3 RELIEF REQUEST NR-9
- 1. SYSTEM: Reactor Coolant
1RV-02-002 1,2,3, 2RV-02-002 (RV-002) (RV-002) 1RV-03-001 4,5,6 2RV-03-001 (RVCH-001) (RVCH-001)
- 3. A.S.M.E. CODE CLASS: 1
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Subsection IWB, Table IWB-2500-1, Examination Category B-A, Item Bl.ll and Bl.40 require volumetric examination of the regions described in Figures IWB-2500-1 and 5 respectively for welds in the reactor pressure vessel each inspection interval.
- 5. BASIS FOR RELIEF:
- a. Lower Shell Course-to-Dutchman weld RV-002 (1RV-02-002), (2RV-02-002) has six core support guide lugs welded to the interior surface of the reactor vessel approximately 3.50" above the weld. These lugs restricted the automated inspection tool from inspecting the required volume in the areas of the lugs, shown in Figure 1. All of the weld and heat affected zone received 100% coverage from at least one direction, however the required base metal was not fully inspected in the area of the core support guide lugs. Figures 2 and 3 show exactly what was inspected. Note that the dimensions used for actual coverage are for transducer position, not volume inspected.
- b. Closure Head Flange-to-Dutchman Forging Weld RVCH-001 ( 2 RV-03-001) ,
(1RV-03-001) has the flange which physically obstructs the ultrasonic transducer from performing the required scan area. Part of the three larger lifting lugs also fall in the required scan area. Figures 4 and 5 show the position of the weld and flange. A detailed diagram of the transducer position for actual and required coverage is shown in Figure
- 6. The code required surface exams will be performed on the accessible areas.
- 6. ALTERNATE TEST METHOD: None.
- 7. JUSTIFICATION: The Reactor Vessel is examined remotely using the immersion technique. Completion of the remaining portions of the above listed welds is impractical and would result in undue hardship without a compensating increase in safety. By performing the limited ultrasonic examinations and the leakage test each refueling outage, an adequate level of structural integrity can be assured for plant operation.
- 8. APPLICABLE TIME PERIOD: Relief will be required for the first 120 month inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
3-15
i Revision 3 NR-9
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Revision 2 RELIEF REQUEST NR-10 l
Relief Deleted t
0 3-22 l
_ - - _ _ _ - - - - - _ _ _ _ - - - _ _ _ _l
Revision 2 RELIEF REQr/EST NR-11 Relief Deleted 3-23 l
)
l Revision 4 RELIEF REQUEST NR-12
)
- 1. SYSTEM: Residual Heat Removal (Residual Heat Removal Heat Exchanger). 'l
- 2. NUMBER OF ITEMS: 4 Component . Attachment Restricted ,
Number Weld Number- Numbers Exam i 1RH02AA RHXN-01, RHXN-02 1 Inner Radii and Nozzle to Vessel Weld l 2RH02AA RHXN-01, RHXN-02 1- Inner Radii and .
Nozzle to Vessel j Weld i
- 3. A.S.M.E. CODE CLASS: 2 ;
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Subsection IWC, Table IWC-2500-1, ,
Examination Category C-B, Item C2.22 requires volumetric examination of the nozzle inner radius and Item C2.21 requires volumetric and surface examination' i of the Nozzle to Shell weld of the regions described in Figure IWC 2500-4(a) !
or (b), for nozzles without reinforcing plate in vessels >1/2 in, nominal thickness. Examinations shall be conducted on nozzles at-terminal ends of >
piping runs selected for examination under Examination Category C-F, each '
inspection interval. :
- 5. BASIS FOR RELIEF: The nozzles listed above contain inherent geometric constraints which limit the ability to perform meaningful ultrasonic i examinations The Residual Heat removal Heat Exchanger is approximately 7/8 in. nominal wall ,
thickness with nozzles of 14 inch diameter and approximately 3/8 in. in l nominal wall thickness. The configuration is best characterized as a fillet ;
welded nozzle using an internal reinforcement pad and, thereby is not analogous to a full penetration butt welded nozzle as shown in Figure IWC-2500-4. In addition, the inner radius of the reinforcement pad would be '
representative of the nozzle inner radius required for inspection. The inherent geometric constraints of the nozzle design prevent the performance of the required ultrasonic examinations of the nozzle inner radius, see )
attachment 1. A nozzle-to-shell UT may not achieve full.ASME coverage of the >
weld areas. !
- 6. ALTERNATE TEST METHOD: The welds listed above will receive the required !
Section XI surface examinations. Visual examination (VT-1) of the nozzle ;
inner radii shall be performed either directly or remotely to the extent ;
practical when disassembly is required for maintenance purposes not to exceed once per inspection interval. In addition, visual examination (VT-2) shall be i performed each inspection period on all pressure retaining components. A best ,
effort UT shall be performed on the RHR HX nozzle-to-shell welds on a i frequency consistent with ASME Section XI.
- 7. JUSTIFICATION: The VT-1 examination will assure early detection of ,
detrimental flaws. Therefore, in performing the proposed alternative i examinations during disassembly for maintenance, an adequate level of !
structural integrity can be assured for continued plant operation, j
- 8. APPLICABLE TIME PERIOD: This relief will be required for the first 120 month l inspection interval. ,
- 9. APPROVAL STATUS: Relief granted for IRS EXAM ONLY per SER dated 10/04/91.
3-24 ;
L s
Revision 4 str i me cutt a = W -
- T --* (NOT TO SCALE) wa-12 Attachment 1 l
VESSEL WALL :
/
T
- 0.875 l
N0ZZL E (TYR) ;
N.0.375 !
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_ REINFORCEMENT
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Wal.25 i i (SEE DETAIL A)
V4 a 4F CHAMFER TYR(4) CORNERS
+--- 2 3* ~
f t i . DIWENSIONS SHOWN ARE NOMINAL 20'- -
l r <
ti, SHELL AXIS I DETAIL A . netNr oAD 3-25
Revision 3 RELIEF REQUEST NR-13 Relief Deleted 3-26
Revision 2 RELIEF REQUEST NR-14
- 1. SYSTEM: Reactor Coolant (Reactor Coolant Pumps)
- 2. NUMBER OF ITEMS: 1 of 4; 1RC0lPA, IRC0lPB, IRC01PC, IRCDlPD.
1 of 4; 2RC0lPA, 2RC0lPB, 2RC0lPC, 2RC0lPD.
- 3. A.S.M.E. CODE CLASS: 1
- 4. A.S.M.E. SECTION XI REQUIREMENTS: Subsection IWB, Table IWB-2500-1, Examination Category B-L-2, Item B12.20 requires a visual examination (VT-3) of pump casing internal surfaces. Examinations shall be conducted on at least one pump in each group of pumps performing similar functions in the system, each inspection interval. In addition, Examination Category B-P, Item B15.60 requires a system leakage test, (IWB-5221) each refueling outage for pump pressure retaining boundaries.
- 5. BASIS FOR RELIEF: The code requirement to disassemble at least one pump from each group of pumps performing similar functions, results in high radiation exposures and does not produce a proportionally higher potential for identification of service induced flaws or degradation. The industrial performance of these pumps have proven their excellent ability to resist service degradation or flawing. The inappropriate balance of potential flaw detection and the large impact on expenditures of manpower without substantially increasing component reliability is considered impractical.
Coupled with this is the highly negative impact on the Station's ALARA program. Therefore, the examination frequency is not appropriate for '
inspection of the above listed pumps.
- 6. ALTERNATE TEST METHOD: Visual examination (VT-3) will be performed on one of these pumps when disassembly is required for maintenance purposes but will not exceed more than one inspection per interval. In addition, the visual examination, (VT-2) leakage test will be performed after each refueling outage.
- 7. JUSTIFICATION: The licensee has specified pumps which are free of pressure boundary welds. The above listed Class 1 pumps are free of any pressure boundary welds, and also free of any integrally welded attachments.
Therefore, there is very little likelihood of degradation or flaw propagation in these pumps. Since these pumps have not shown significant service induced degradation in the past, deferral of the internal inspections will not reduce the assurance that thejr structural integrity has been maintained.
- 8. APPLICABLE TIME PERIOD: This relief will be required for the first 120 month inspection interval. Note, this inspection is deferrable until the end of the inspection interval and application of this relief may not be required if maintenance is performed on a pump during the first inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
3-27
Revision 2 RELIEF REQUE U NR-15
- 1. SYSTEM: Chemical and Volume control (Centrifugal Charging Pump), Residual Heat Removal (Residual Heat Removal Pump). __
- 2. NUMBER OF ITEMS: 14 Component Number Weld Number Attachment Number icv 01PA CVP-01, CVP-01 1 CVP-03, CVP-04 1RH0lPA RHP-01, RHP-02 2 REP-03 2Cv0lPA CVP-01, CVP-02 1 CVP-03, CVP-04 2RH0lPA REP-01, RHP-02 2 RHP-03
- 3. A.S.M.E. CODE CLASS: 2
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Subsection IWC, Table IWC-2500-1, Examination Category C-C, Item C3.30 requires surface examination of the regions described in figure IWC-2500-5 for integrally welded attachments in pumps. Examinations shall be conducted on components required to be exa* mined under Examination Categories C / or C-G each inspection interval.
- 5. BASIS FOR RELIEF: The above .sted welds connect the support lugs to the pump casings. These integrally welded attachments can not be examined on one of the required sides due to the location of structural supports. Attachments 1 and 2 provide sketches of the obstructions. As can be seen from the sketches, the welds are located between the pump casing and the structural supports for the pumps. These welds can only be examined in the areas which allow sufficient clearance.
- 6. ALTERNATE TEST METHOD: A visual examination, (VT-1) will be performed on the inaccessible side of the integrally welded attachments when the surface examination is performed on the remaining portion of the attachments. In addition, the system leakage test will be performed each inspection period.
- 7. JUSTIFICATION: These welds are not full penetration welds, as such, the proposed VT-1 examination, in conjunction with the surface examination will identify developing defects. Therefore, by performing the alternate visual examination to supplement the partial surface examination, an adequate level of structural integrity can be assured for continued plant operation.
- 8. APPLICABLE TIME PERIOD: This relief will be required for the first 120 month inspection interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
3-28
Revision 2 NR-15 Attachment 1
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7 Revision 2 NR-15 Attachment 2 i
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Revision 2 ;
RELIEF REQUEST NR-16
- 1. SYSTEM: Main Steam and Feedwater ,
- 2. NUMBER OF ITEMS: 6 .__
Percentage of Weld Langth l Wald Idag Not F*aM Reason for Linuted Exam j IFW-01-01 IFW86AD-16" 16 % Branch enaeion, valve geometry ;
IFW-02-01 IFW86AA-16" 16 % Branch connection, !
valve geometry l IFW 04-29 IFWO3DC-16* 25 % Elbow inner radius, l valve geometry 1MS-0101 1MS01AD-32' 8% Nozzle, gn=== plug IMS-02 07 IMS01BD-30.25" 16 % Tee, weldolets
- IMS-04-07 IMS01BA-30.25" 26 % Weldolets, see geometry
- 3. A.S.M.E. CODE CLASS: 2 ,
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Subsection IWC, Table IWC-2500-1, ,
Examination Category C-F, Item C5.21 requires surface and volumetric ;
examination of the regions described in IWC-2500-7 for essentially 100% of the '
weld length. *
- 5. BASIS FOR RELIEF: The welds listed above have been selected for examination ;
during the first inspection interval. They have interfering condition on each side of the weld. These interferences can cause poor coupling of the ;
transducer, limited movement of the transducer, redirecting of the sound beam and in some cases, complete restriction of a particular scan. These conditions sufficiently limit the axial scans so as to leave the listed j percent of the weld uninspected.
- 6. ALTERNATE TEST METHOD: The welds listed above will receive the required Section XI surface examination in addition to a best effort ultrasonic :
examination.
- 7. JUSTIFICATION: The estimates of weld length not examined are extremely conservative and are actually a percent of weld length for which there is less !
confidence that the entire required weld volume was examined. Based on )
acceptable surface examinations, acceptable radiographic examinations and a !
best effort ultrasonic examination which inspected a majority of the weld i length, an adequate level of structural integrity is assured.
'8. APPLICABLE TIME PERIOD: This relief will be required for the first 120 month interval.
- 9. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
I l
3-31 i
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I Revision 3 RELIEF REQUEST NR-17 Relief Deleted e
3-32
, . -. . . . . - . ~ -
O }
Revision 3 l
.i i
i RELIEF REQUEST NR-18 .l
- 1. SYSTEM: Reactor Coolant.
~
- 2. NUMBER OF ITEMS: 2 Line Number Weld Number Interfering Condition 2RC29AC-10" 23I-09-17 Permanent Restraint 2RC29AD-10" 2SI-13-28 Permanent Restraint j
- 3. A.S.M.E. CODE CLASS: 1 .
r
- 4. A.S.M.E. CODE SECTION XI REQUIREMENTS: Table IWB-2500-1, Examination Category l' B-J, Item 9.11 requires volumetric and surface examination of the areas.
described in Figure IWB-2500-8 for essentially 100% of the weld length.
P
restraint making it inaccessible to both surface and volumetric examination, i see attachment 1. Weld number 2SI-09-17 is adjacent to a permanent whip i restraint making it accessible for surface examination but inaccessible for !
volumetric exad nation since it is a valve to pipe weld on the upstream side,. !
see attachment 2.
l
- 6. ALTERNATE TEST METHOD: Surface exam on accessible weld.
{
- 7. JUSTIFI' CATION: The structural integrity of this weld shall be insured b'y:
- a. Performing a surface examination of the region accessible.
1
- b. Performing a system leakage test each refueling outage and a system !
hydrostatic test each interval. ;
- 8. APPLICABLE TIME PERIOD: . Relief will be required for the first 120 month ;
inspection interval. I
- 9. . APPROVAL STATUS: Relief granted per SER dated 10/04/91. ;
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3-35
Revision 0 RELIEF REQUEST NR-19
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1 Relief Deleted i
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3-36 i
Revision 0 RELIEF REQUEST NR-20 l
l
- 1. SYSTEM: Steam Generators, Unit 1 only __
- 2. NUMBER OF ITEMS: Nine (9)
Component Number Weld Numbers Code Item 1RC01BA ISG-01-SGN-OlA (IR) B3.140 ISG-01-SGN-01B (IR) B3-140 ISG-01-SGN-02 (IR) C2.22 ,
IRC01BB ISG-02-SGN-OlA(IR) B3.140 ISG-02-SGN-01B (IR) B3.140 1RC01BC 1SG-03-SGN-01A (IR) B3.140 ISG-04-SGN-01A(IR) B3.140 1RC01BD ISG-04-SGN-01B(IR) B3.140
- 3. A.S.M.E. CODE CLASS: 1,2
- 4. A.S.M.E. CODE SECTION XI REQUIREMENT:
Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.140 and Subsection IWC, Table IWC-2500-1, Examination Category C-B, Item C2.22 require volumetric examination of the nozzle Inner Radius Section (IRS) per figure IWB-2500-7 and figure IWC-2500-4 (a) .
- 5. BASIS FOR RELIEF:
Relief is requested from examining the IRS of eight (8) primary nozzles on -
four (4) steam generators and one (1) feedwater nozzle on the "A" steam generator on the basis that compliance with the ASME Section XI requirement would result in hardship or unusual difficulty without a compensating increase in the level of plant quality and safety.
t In accordance with the requirements of ASME Section XI 1983 Edition through Swnmer 1983 Addenda, a volumetric examination of IRS of the subject nozzles is required once for each Inspection Interval. Performing the volumetric examination on the IRS of the subject nozzles will subject plant personnel to high radiation exposure. The total radiation exposure to personnel to build scaffolding, remove and reinstall insulation, prepare the surface for examinations, and examine the subject nozzles on four (4) steam generators is estimated to be at least eight (8) person-rems. This radiation exposure is not justifiable since these steam generators will be replaced during the AIR 07 refuel outage, currently scheduled for 1998. ;
The subject nozzle inner radius sections are not susceptible to thermal fatigue cracking because the primary nozzles do not experience thermal stratification or high thernal gradient during operation, and the feedwater ,
nozzles have a thermal sleeve connected to the " wrapper barrel" that minimizes thermal cycling due to cooler feedwater back splashing. Stress corrosion ,
Cracking (SCC) in the stainless steel clad of the primary nozzle inner radius sections is unlikely because of low oxygen content of the primary water. The feedwater nozzles have no stainless steel cladding and are made of low alloy steel that is not susceptible to SCC. To date, cracking in the IRS of similar steam generator nozzles has not been a problem. Recent state-of-the-art IRS examinations conducted on eight (8) steam generator primary nozzles and one (1) feedwater nozzle at another Commonwealth Edisor pressurized water reactor of the same vintage as that of Braidwood Unit I reiaforces this experience. t 3-37 ,
l P
' Revision 0 !
l ssLzar magozar us-no (cent.)
- 5. -RASIS POR RELIEF (Cont): .
This relief request will only apply to Braidwood Unit l'Eteam generators that will be replaced in 1998. The IRS of the nozzles on the new steam generators !
that will be installed in Braidwood Unit I will be subject to the applicable ;
requirements of ASME Section XI for the second ten-year Inspection Interval.
- 6. ALTERNATE TEST METHOD: ,
Periodic visual. examination (VT-2) of the nozzles' outside surface will be :
performed in accordance with the requirements of ASME Section XI, Table IWB-
-2500-1, Examination Category B-P and Table IWC-2500-1, Examination Category C-H, including applicable Code Case (s) .
Prior to the Braidwood Unit 1 steam generators replacement, if future IRS ,
examinations of Braidwood Unit 2 and Byron Unit 2 steam generator nozzles ;
(primary and feedwater nozzles) reveal flaws that exceed the applicable acceptance criteria of ASME Section XI IWB-3500, appropriate Braidwood Unit 1 steam generator nozzle IRS will be examined in accordance with the applicable ;
requirements of ASME Section XI. !
- 7. JUSTIFICATION: l Periodic IRS examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-D and Table IWC-2500-1, i
Examination Category C-B on Braidwood Unit 2 and Byron Unit 2 steam generator i nozzles (primary and feedwater nozzles).will serve to provide a reasonable -ii sample from which to assess the integrity of the steam generator primary-nozzles and feedwater nozzles. Currently, there are no plans to replace the ,
steam generators-in Braidwood Unit 2 and Byron Unit 2.
- 8. APPLICABLE TIME PERIOD. ,
This relief request will be required'for the first ten-year Inspection i Interval. ;
i
- 9.
REFERENCES:
None, j
- 10. APPROVAL STATUS: Pending NRC revice. ;
i I
3-38 !
l l
Revision 0 t
RELIEF REQUEST NR-21 CCapoMENT IDENTIFICATION !
Code Classes: 1, 2, and 3
Reference:
IWA-4400 IWA-4600 IWA-5000 Examination Categories: B-P '
C-H D-A, D-B, D-C Item Numbers B15.10 through B15.71 C7.10 through C7.8 0 D1.10, D2.10, and D3.10
Description:
Alternate Hydrostatic Pressure Test Requirements for ASME Class 1, 2 and 3 Repaired / Replaced Components
- Component Numbers: All ASME Class 1, 2, and 3 pressure components subject '
to Hydrostatic Pressure Retaining Tests per IWA-4400 and IWA-4600.
CODE REQUIREM NT IWA-4400(a) and IWA 4600 requires an elevated pressure hydrostatic test to be performed in accordance with IWA-5000 after repairs by welding or the attachment of ;
replacement items by welding on the pressure retaining boundary of ASME Class 1, 2, or 3 components, except those exempted by IWA-4400(b) or IWA-7400. .
. BASIS rom REI.IEF Elevated pressure hydrostatic tests are difficult to perform and often represent a true hardship. Some of the difficulties associated with hydrostatic testing include:
- Complicated or abnormal valve line-ups to provide system draining, filling, venting, and system / component isolation.
- Relief valves with setpoints lower than the hydrostatic test pressure must be blocked closed, or removed and blank flanged. This process requires draining, i refilling of the system prior to the test and draining, valve restoration, and refilling once more for system restoration. Improper blocking or gagging can result in damage to the relief valve.
- Valves that are not normally used for isolation are often required to provide pressure isolation for a hydrostatic test. In order to provide tight 1 isolation, time consuming valve amintenance would be required prior to a j hydrostatic test.
- The radiation exposure required to setup and perform a hydrostatic test is quite high in comparison to an operational pressure test due to time required for valve manipulation, filling and venting, valve maintenance, etc.
The difficulties encountered in performing a hydrostatic pressure test are prohibitive when weighed against the benefits. Industry experience, which is supported by Comed experience, shows that most through wall leakage is detected during system operation as opposed to during the elevated pressure tests associated with the ten-year hydrostatic test.
3-39
Revision 0 RELIxF REQUEST NR-21 (Cont.)
Little benefit is gained from the added challenge to the piping system provided by an elevated pressure hydrostatic test when compared to an operational test. The piping stress experienced by a hydrostatic test does not include the significant stresses associated with the thermal growth and dynamic loading during operation or a design basis event. Therefore, the system is more likely to leak at operating conditions, due to operational dynamic and thermal loading, than during the careful, slow pressurization associated with a hydrostatic test.
The acceptability of performing nominal operating pressure tests in lieu of hydrostatic tests is also supported by the recent approval by the Board of Nuclear Codes and Standards of ASME Code Case N-416-1, " Alternate Pressure Test Requirement for Welded Repairs or Installation of Replacement Itens by Welding for Class 1, 2, and 3 Systems,Section XI, Division 1". This Code Case allows a system leakage test at nominal operating pressure and temperature, in accordance with IWA-5000 of the 1992 Edition of Section XI, to be performed in lieu of a hydrostatic test, provided that Non Destructive Examination (NDE) of the weld (s) is performed in accordance with the methods and acceptance criteria of the applicable subsection of the 1992 Edition of Section III.
Based on the above, Braidwood Station requests relief from the ASME Section XI Class 1, 2, and 3 repair / replacement elevated pressure hydrostatic testing requirements.
PROPOSED ALTERNATE PROVISIONS As an alternate to the existing ASME Section XI requirements, Brai-dwood Station will adopt the provisions of Code Case N-416-1, as approved by the Board of Nuclear Codes and Standards, with additional NDE requirements. Listed below are the proposed alternate provisions to be performed, which is a summary of Code Case N416-1 requirements with additional NDE requirements imposed by Braidwood Station.
- A VT-2 visual examination will be performed at nominal operating pressure and temperature in conjunction with a system leakage test in accordance with IWA-5000 of the 1992 Edition of Section XI. The examination will be performed prior to or immediately upon return of the component to service.
- Non-Destructive Examination will be performed on the repair / replacement welds or welded areas with the methods and acceptance criteria of the applicable Subsection of the 1992 Edition of Section III. In addition, when NDE is required by ND-5222 for Class 3 components, an additional surface examination will be performed on the root (pass) layer. A surface examination will also be performed on Class 3 socket / fillet welds.
- The use of this relief request shall be documented on the applicable NIS-2 Form.
- If the previous version of Code Case N-416 were used to defer a Class 2 hydrostatic test, the deferred test may be eliminated when the requirements of this relief request are met. In addition, the NDE methods and acceptance criteria of the Code Edition and addenda used for the repair must be reconciled to those of the 1992 Edition of Section III.
APPLICABLE TIME PERIOD Relief is requested for the first ten-year inspection interval of the Inservice Inspection Program for Braidwood Unit 1 and Unit 2.
APPROVAL STA'ms:
Pending NRC Review.
3-40
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1
)
1 Revision 3 RELIEF REQUEST NR-22 ;
)
COMPONENT IDENTIFICATION Code Classes: 3 __
References -
IWD-5210 IWD-5223 Examination Categories: D-A D-B D-C Item Numbers: D1.10 D2.10 D3.10
Description:
Alternate Rules for 10 Year Hydrostatic Testing for ,
Pressure Class 3 Systems Component Numbers: All Class 3 Systems subject to Hydrostatic Testing per
- IWD-2500.
CODE REQUIREMENTS IND-2500 requires elevated pressure hydrostatic tests (VT-2) to be performed each 10 year inspection interval on ASME Class 3 pressure retaining components in accordance- !
with IWD-5223. .
B masIs rom RELIsr Elevated pressure hydrostatic tests are difficult to perform and often represent a
- true hardship without benefits gained. Some of the difficulties associated with 10 year system hydrostatic testing include:
Complicated or abnormal valve line-ups to provide system draining, filling, venting, and system isolation.
Relief valves with setpoints lower than the hydrostatic test pressure must be blocked closed, removed and blank flanged. This process requires draining, refilling of the system prior to the test and draining, valve. restoration, and refilling once more for system restoration. Improper blocking or gaging can result in damage to the relief valve.
Valves that are not normally used for isolation are often required to provide pressure isolation for a hydrostatic test. In order to provide tight isolation, time consuming valve maintenance would be required prior to a hydrostatic test.
The radiation exposure required to perform hydrostatic testing is quite high !
in comparison to oprational pressure testing due to time required for valve '
manipulation, filling and venting, valve maintenance, etc.
At hydrostatic test pressures required by ASME Section XI,.10% and 25% over the piping design pressure, a hydrostatic test does not induce significantly more stresses in the system than in a system operational test. Also, the system stresses >
associated with the hydrostatic test do not compare to the stress associated with i thermal growth and dynamic loading during design basis events. Therefore, little benefit is gained from the hydrostatic test over the nominal operational pressure test.
Industry experience, which Comed Stations experience supports, indicates that most #
through wall leakage is detected during system operation as opposed to hydrostatic testing at elevated pressures.
3-41
Revision O RELIEF REQUEST NR-22 (Cont.)
These arguments are also supported by ASME Code Case N-498-1, " Alternate Rules for 10 Year Hydrostatic Pressure Testing for Class 1, 2 and 3 Systems,Section XI, Division 1" and ASME Code Case N-498, " Alternate Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and 2 Systems,Section XI, Division 1". Code Case N-498-1 has been reviewed and approved by the Board of Nuclear Codes and Standards (BNCS). Code Case N-498 for Class 1 and 2 systems had previously been approved and accepted for industry use in Regulatory Guide 1.147, Revision 10.
Based on the above, Braidwood Station requests relief from the ASME Section XI Class 3 10 Year System Hydrostatic Pressure Testing requirements.
PROPOSED ALTERNATE PROVISIONS A system pressure test with a VT-2 visual examination will be performed with the Class 3 system pressurized to a test pressure equal to nominal operating pressure.
The visual examination will be conducted after the system has been pressurized to test pressure for a minimum of 10 minutes for noninsulated components or 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated components prior to examination. The system will be maintained at test pressure for the duration of the VT-2 visual examination. Hydrostatic test instrumentation requirements of IWA-5260 are not applicable as test parameter recording is performed by normal operating system instrumentation or equivalent.
The system pressure test will be conducted at or near the end of the inspection interval or during the same inspection period of each irepection interval.
The boundary subject to test pressurization and VT-2 visual examination during the system pressure test shall extend to all Class 3 components included in those portions of systems required to operate or support the safety system function up to and including the first normally closed valve (including safety or relief valve) or valve capable of automatic closure when the safety function is required.
APPLICABLE TIME PERIOD Relief is requested for the first ten-year inspection interval of the Inservice Inspection Program for Braidwood Unit 1 and Unit 2.
3-42
Revision 3 4.0 TECWIICAL AFFAQhCW AMD POSITIcets When the requirements of the Code are not easily interpreted, Braidwood Station has reviewed general licensing / regulatory requirement and industry practice to determine a practical method of implementing each esquirement. Augmented examinations and additional examinations performed 9n i.o regulatory requirements and Station commitments are identified. The use of any Code case approved and documented in Regulatory Guide 1.147 is identified in Table 4.0-1.
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4-1 I
I 1
._ _ = _ - - -
i Revision 3 Table 4.0-1 [
> 7 of Technical Approaches and Positions (sheet 1 of 1) l Note 1 The Main Steam Nozzle does not have a nozzle inner radius as described in Figure IWC-2500-4 (a) and (b) so no ultrasonie-examination is required.
Note 2 Code Case N-419 limiting exans (Code Category B-G-1) to those components selected for Categories B-B, B-J, B-L-2, and B-M-2.
Note 3 Code Case N-426 limiting exams (Code Category B-G-2) to those components selected for Categories B-B, B-J, B-L-2, and B-M-2.
Note 4 Code Case N-401 which allows the use of digitally encoded eddy current data.
Note 5 Augmented ultrasonic examinations will be performed on high energy pipe lines which have not been postulated for line breaks per Nuclear Regulatory Guide 1002 (Mainsteam & Feedwater system). l Note 6 Augmented Reactor Coolant Pump Flywheel exams will be performed per Nuclear Regulatory Guide 1.14 (ref. B/B UFSAR A1.14-3 and Technical "
Specification 3/4.4.10).
1 Note 7 Ultrasonic re-examination will be performed on the Unit 1 Loop A upper transition cone weld (1SG-01-8C) per IWB-2420.
Note 8 Augmented inspections of turbine' rotors and discs will be performed per Westinghouse recommendations.
Note 9 Ultrasonic examinations will be performed on accessible portions of the Reactor vessel head welds (upper & lower)
Note 10 Augmented ultrasonic examinations on 7 1/2% of the circumferential welds within the ECCS systems (RH, CV, CS, & SI) for the detection of IGSCC. i Note 11 Ultrasonic re-examination will be performed on a Unit 2, Reactor Coolant ,
main loop weld (2RC-01-04) per IWB-2420. i Note 12 Deleted (refer to Note 13) ,
i Note 13 Code Case N-408-2 exempts vessels with cumulative inlet and cumulative outlet pipe cross-sectional diameter which do not exceed 4" NPS.
Note 14 Augmented ultrasonic examinations will be performed on areas which may be subject to thernal stratification per NRC Bulletin 88-08. :
Note 15 Ultrasonic and surface examinations will be performed on four (4)
Residual Neat exchanger nozzle to vessel welds in which indications were detected during A2R02 and AIR 03 per IWC-2420. ;
Note 16 Code Case N-416 defers system hydrostatic test required by IWA-4400 for repair or replacement of Class 2 piping.
Note 17 Code Case N-498 allows a scheduled 10-year leakage test in lieu of a 10-year hydrostatic test.
i 6
4-2 e
l l
Revision 2 f I
FH)T1C 1 !
Subsection IWC, Table IWC-2500-1, Examination Category C-B, Item C2.22 requires volumetric examination of the nozzle inner radii of nozzles without reinforcing plate in vessel >'l/2 in. nominal thickness. The main steam nozzle was designed with an internal multiple venturi type flow restrictor with an equivalent throat t diameter of 16 in. This design is used to limit the flow in the event of a postulated steam line break. Attachment 1 provides a sketch of the nozzle. This design does not utilize a radius nozzle as described in Figures IWC-2500-4(b), and i therefore, is not considered subject to inner radii inspection. However, these
~
nozzles will receive a system leakage test during each inspection period.
(
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[
4-3 i
Revision 2 Note 1
~
Attachment 1 A
. . , ,% -- +
\
MULTIPLE VENTURI STEAM GENERATOR HEAD TYPE N0Z2LE a _
- h b
4-4 l
Revision 2 NOTE 2 Braidwood will incorporate Code Case N-419 which allows limiting the examinations required by Subsection IWB, Table IWB-2500-1, Examination Category B-G-1 to those components selected for examination under Examination Categories B-B, B-J, B-L-2 and ,
B-M-2. .
Pumps and valves shall be grouped in accordance with Examination Categories B-L-2 and B-M-2 as follows:
- 2RC01PA *1RC01PA *2RC8001A *1RC8001A 2RC01PB 1RC01PB 2RC8002A 1RC8002A 2RC01PC 1RC01PC 2RC8001B 1RC8001B 2RC01PD 1RC01PD 2RC8002B 1RC8002B 2RC8001C 1RC8001C 2RC8002C 1RC8002C 2RC8001D 1RC8001D 2RC8002D 1RC8002D
- Component selected for examination within group, i
i I
L I
4-5
Revision 2 i
NOTE 3 Braidwood will incorporate Code Case N-426 which allows limiting the examinations i required by Subsection IWB, Table IWB-2500-1, Examination Category B-G-2, to those components selected for examination under Examination Categories B B, B-J, B-L-2 and B-M-2. !
Pumps and valves shall be grouped in accordance with Examination Categories B-L-2 and B-M-2 as follows:
- 2RC01PA *1RC01PA *2RC8074A *1RC8074A *2RC8036A *1RC8036A 2RC01PB 1RC01PB 2RC80748 1RC80748 2RC80365 1RC8036B 2RC01PC 1RC01PC 2RC8074C 1RC8074C 2RC8036C 1RC8036C 2RC01PD 1RC01PD 2RC8074D 1RC8074D 2RC8036D 1RC8036D 2RC8073A 1RC8073A 2RC8037A 1RC8037A
- 2CV8378A *1CV8378A 2RC80738 1RC80738 2RC80378 1RC80378 2CV83788 1CV83788 2RC8073C 1RC8073C 2RC8037C 1RC8037C 2CV8379A 1CV8379A 2RC8073D 1RC8073D 2RC8037D 1RC8037D 2CV8379B ICV 83798 !
- 2RC8003A *1RC8003A *2RYO22 *1RYO22 I
- 2CV459 *1CV459 2RC8003B 1RC8003B 2RYO23 1RYO23 2CV460 1CV460 2RC8003C 1RC8003C 2RYO24 1RYO24 2RC8003D 1RC8003D 2RYO25 1RYO25
- 2CV8145 *1CV8145 *2RC8085 *1RC8085
- 2RY455A *1RY455A *2SI8808A *1SI8808A *2SI8841A *1SI8841A ,
2RY456 1RY456 2SI8808B ISI8808B 2SI88418 15I88418 '
2SI8808C 1SI8808C
- 2RY455B *1RY455B 2SI8808D ISI8808D *2SI8948A *1SI8948A 2RY455C 1RY455C 2SI89488 ISI8948B
- 2SI8810A *1SI8810A 2SI8948C ISI8948C
- 2RY80000-A *1RY8000-A 2SI8810B ISIS 810B 2SI8948D ISI8948D 2SI8810C 1SI8810C 2SI8956A ISI8956A
- 2RY8000-8 *1RY8000-B 2SI6810D ISI8810D 2SI8956B ISI89568 ,
2SI8956C 1SI8956C I
- 2RY8010-A *1RY8010-A *2SI8815 *1SI8815 2SI8956D ISIB956D +
2RY8010-B 1RY8010-P i
- 2SI8818A 2RY8010-C 1RY8010-C *1 SIB 818A *2SI8949A *1SI8949A 2SI88188 ISI88188 2SI89498 ISI89498 :
- 2RH8701A-1 *1RH8701A-1 2SI8818C ISI8818C 2SI8949C 1SI8949C 2RH87018-2 1RH87018-2 2SI8818D 1SI8818D 2SI8949D ISIB949D F I
2RH87025-2 1RH87025-2 2RH8702A-1 1RH8702A-1 ,
t
- Component selected for examination within group.
\
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4-6 ,
i f
I
Revision 2 NOTE 4
~
Eddy Current Insoectien of Steam Generator U-Tubes Eddy Current Inspection of Steam Generator tubes is an effective means of determining defects because it detects the presence of defect-caused variations in effective electrical conductivity and/or magnetic permeability.
The Eddy Current Program Plan will attempt to maximize the detection of defects while complying with U.S. Nuclear Regulatory Commission Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes", and Braidwood Technical Specification 3/4.4.5, " Steam Generators".
Eddy Current examination shall be performed in accordance with the requirements of Braidwood Technical Specification 3/4.4.5, " Steam Generators." <
Braidwood will incorporate Code Case N-401 which allows the use of digitally encoded oddy current data as an alternative to magnetic tape and strip chart recordings.
t I
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4-7 l
l Revision 2 NOTE 4 Eddy Current Inanection of Steam Generator U-Tubes )
Eddy Current Inspection of Steam Generator tubes is an effective means of determining defects because it detects the presence of defect-caused variations in effective electrical conductivity and/or magnetic permeability.
4 The Eddy Current Program Plan will attempt to maximize the detection of defects l while complying with U.S. Nuclear Regulatory Commission Regulatory Guide 1.83, Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes", and eraidwood Technical Specification 3/4.4.5, " Steam Generators".
Eddy Current examination shall be performed in accordance with the requirements of Braidwood Technical Specification 3/4.4.5, " Steam Generators."
Braidwood will incorporate code case N-401 which allows the use of digitally encoded ,
eddy current data as an alternative to magnetic tape and strip chart recordings.
9 4-7
Revision 2 NOTE 5 Augmented Esamination are to be performed on Class 2 and 3 High Energy Piping Systems outside Containment where breaks have not been postulated. Volumetric examination of circumferential welds shall be performed on six inch and larger diameter piping in the Main Steam (MS) and Main Feedwater (FW) Systems from the containment wall to the MS and FW torsional restraints downstream of the outboard containment isolation valves (within the valve room). A volumetric examination of longitudinal seam welds adjacent to circumferential welds shall be performed in accordance with ASME Section X1 criteria. These augmented inspections meet the requirements of NUREG-1002 Section 6.6.
b i
i 4-8 I
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Revision 2 '
NOTE 6 The Augmented Reactor Coolant Pump Flywheel examination shall be conducted as follows:
- 1. In-place ultrasonic volumetric examination of the areas of higher stress concentration of the bore and key ways will be performed each 40 month period during refueling or maintenance shutdowns coinciding with the inservice inspection schedule as required by Section XI of the ASME Code.
- 2. A surface examination of all exposed surfaces and complete ultrasonic !
volumetric examination shall be performed whenever the flywheels are removed for maintenance purposes, but not more frequently than once each 10 year interval. Those inspections shall coincide with the inservice inspection schedule as required by Section II of the ASME Code. !
The requirements for examination procedures and acceptance criteria as described in '
the Regulatory Guide will be followed.
This inspection program meets the intent of Regulatory Guide 1.14 in assuring the continued integrity of the reactor coolant pump flywheels. 6 h
)
9
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b 4-9 ;
I i
Revision 2 NOTE 7 In Section 5.2.4.4 of NUREG-1002, Supplement No. 1 dated September 1986, the Nuclear Regulatory Cossaission requested Braidwood Unit 1 to commit to augmented inservice inspections of the upper shell-to-transition cone weld (weld #6) of the loop 1 steam generator and the lower-to-upper intermediate circumferential shell weld (weld #8c) ,
of the Unit 1 pressurizer. These two welds are those welds in which subsurface i indications were found that exceeded the Section XI acceptance criteria. In the SER i
Supplement, the NRC has concluded that weld repair of the two flaws is not necessary if the following augmented inspections are completed.
- 1. The areas containing the flaws must be inspected in accordance with the inspection interval requirements of Section XI IWB-2420s that is, the subject areas are reexamined during the next three (3) inspection periods. If the reexaminations show that the indications remains essentially unchanged for these three (3) successive inspection periods, the examination schedule may revert to the original schedule.
- 2. The Loop 1 steam generator secondary side hydrotest and leak tests must be performed at temperatures greater than 165cr and 1500F, respectively.
- 3. The pressurizer primary side hydrotest and leak test must be performed at ;
temperatures greater than 1200F.
4-10
l Revision 2 NOTE 8 Augmented Inspection if Turbine Rotors and Turbine Disks.
Due to the possibility of missile generation from failure of steam turbine disks Braidwood Station plans to have augmented Non-Destructive Tests performed on the HP and LP turbine disks and rotors. Inspection techniques and frequency will be determined by Westinghouse Corporation with the turbines with the highest probability of failure being inspected during the first refueling outage and the remaining turbine disks and rotors to be inspected during the second refueling outage.
Referer.ces:
Braidwood SER 3.5.1.3.4 NUREG-1002 Supp. No. 1 FSAR 10.2.3.6 Appendix C.
4-11
l i
Revision 2 i'
NOTE 9 subsection INE, Tabli~IWB-2500-1, Examination Category B-A, Item Bl.21 requires volumetric examination of the accessible portions of head circumferential welds in ,~
the reactor vessel. Listed below are the applicable welds and descriptions of their inaccessible portions. '
- a. Lower Disk-to-Dutchman weld RV-001, (lRV-02-001), (2RV-02-001) Figure 1 has 58 instrument tubes which physically obstruct the search unit and/or eaarch unit positioning device. Figures 2 and 3 give a detailed diagram of the areas inspected and those which were obstructed by the tubes.
Note the dimensions used for the actual coverage are for transducer position.
- b. Dutchman Forging-to-Closure Head Dome Weld RVCH-002 (2RV-03-002), (1RV-03-002), has 6 lifting lugs which physically obstruct the ultrasonic transducer from performing the required scan. Figures 4 and 5 show the position of the weld and lifting lugs. A detailed diagram of the transducer position for actual and required coverage is shown in Figure ;
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4-18 l
Revision 1 NOTE 10 The NRC has expreseed'a concern dealing with intergranular stress corrosion cracking (IGSCC) in lines that contain stagnant borated water. Braidwood Station will perform augmented volumetric examinations on Class 2 ECCS systems; Containment Spray (CS), Chemical and Volume Control (C.V.), Residual Heat Removal (R.H.) and Safety l Injection (SI) that are not currently subject to volumetric examination as required 1 by code. The inspections shall include seven and one-half percent (7.5%) sampling > !
4" nominal pipe size of the total population of circumferential welds which contain stagnant borated water. Nominal pipe wall thickness and pressure / temperature exemptions do not apply.
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i 4-19
Revision 2 NOTE 11 The NUREG-1002, Supplement No. 6 dated May 1988, the Nuclear Regulatory Commission requested Braidwood Unit 2, to commit to augmented inservice inspections of an indication found during preservice inspection. The indication _was found on the loop 1, elbow side of the cast austenitic stainless steel elbow-to-loop isolation valve weld (2RC-01-04). Indications found during preservice inspection exceeded the standards of IWB-3500. However, an evaluation of the indication was performed in accordance with IWB-3640 of the 1983 Edition, Winter 1983 Addenda of Section XI.
This evaluation showed the indication to be acceptable to the standards of IWB-3640.
Braidwood has committed to perform augmented ultrasonic examination of the flawed area for the next three (3) inspection periods to confirm the fatigue growth analysis.
I I
l 4-20
Revision 2 Revision 2 NOTE 12 h
Deleted l
4-21
Revision O NOTE 13 In addition to the exempted compononts listed in Section(s) 9.0 of the Braidwood Inservice Inspection Plan, Braidwood will also be exempting the following class 2 vessels from the volumetric and surface examination requirements of IWC-2500:
Components Applicable Exemptions ,
Regenerative Heat Exchangers (4 total) 1/2CV03AA IWC-1220(c) and 1/2CVO3AB Volume Control Tanks (2) IWC-1220(b) 1/2CV0lT Reactor Coolant Filters (2) IWC-1220(c) 1/2CVO3F ,
Seal Water Return Filters (2) IWC-1220(c) 1/2CV02F '
Excess Letdown Heat Exchangers 4) IWC-1220(c) 1/2C01AA and 1/2CV01AB Letdown Reheat Heat Exchangers (2) IWC-1220(c) 1/2CVOSA Horizontal Letdown Heat Exchangers (4) IWC-1220(c) 1/2CV04AA and 1/2CV04AB The Volume Control Tanks operate at a pressure less than 275 psig and a temperature l less than 200 degrees F. The remaining above listed Class 2 vessels are exempt since the cumulative inlet and cumulative outlet pipe cross-sectional areas for these vessels do not exceed a 4" NPS cross-sectional area. This position has been clarified in later versions of ASME Section XI as well as in Code Case N-408-2. All these vessels receive a periodic pressure test (VT-2) which assures their structural and operational integrity.
I l
l 1
4-22
Revision O NOTE 14 In response to NRC Bulletin 88-08 " Thermal Stresses in Piping Connected to Reactor Coolant systems," Ultrasonic examinations utilizing IGSCC sensitivity will be performed at specific locations on the Pressurizer Auxiliary _line and the RHR discharge lines from the RCS. Examinations shall be performed at alternating refueling outages.
i 4-23
Revision 0 NOTE 15 In Safety Evaluation ~ Report Docket No. STN 50-457 dated November 21, 1991 the Nuclear Regulatory Commission requested Braidwood Unit 2 to commit to augmented inservice inspections of the Residual Heat Removal (RHR) heat _ exchanger primary nozzle to vessel welds. These four welds are those welds, in which subsurface indications were found that exceeded the acceptable criteria in ASME Code Section XI Subarticle IWB-3500. A Fracture Mechanics Analysis was performed and the flaws meet the allowable sizes as specified in the tables in paragraph IWB-3640 of ASME Code Section XI. In the SER, the NRC has concluded that weld repair of the flaws is not necessary if the following augmented inspections are completed.
- 1. The areas containing the flaws must be inspected in accordance with the inspection interval requirements of ASME Section XI, Paragraph IWC-2420.
If future inspections of these flaws indicate that there is no flaw growth, then the RHR heat exchangers are acceptable for the 40-year life of the plant.
- 2. This methodology is acceptable for Braidwood, Unit 1 and Unit 2.
4-24 1
i l
I Revision 0 l
NOT1C 16 ;
Braidwood will incorporate Code case N-416 which defers system hydrostatic tests j required by IMA-4400 for repair or replacement of Class 2 piping that cannot be isolated by existing valves or that requires securing safety or relief valves for isolation until the next regularly scheduled system hydrostatic tests (IWC 5000),
provided both of the following conditions are met.
(a) Prior to or immediately upon return to service, a visual examination (VT-2) for leakage shall be conducted during a system functional test or ;
during a system inservice test in the repaired or replaced portion of l the piping system.
(b) The repair or replacement welds shall be examined in accordance with IWA-4000 and IWA-7000 using volumetric examination methods (IWA-2230) )
for full penetration welds or surface examination methods (IWA-2220) for i partial penetration welds. 1 l
I I
l l
l 4-25
Revision O NOTE 17 Braidwood will incorporate Code Case N-498 which allows a scheduled system leakage test in lieu of a 10-year hydrostatic as required by Table IWB-2500-1, Category B-P and Table IWC-2500-1, Category C-H providing the following rules shall be used:
(a) It is the opinion of the (ASME) Committee that as an alternative to the 10-year hydrostatic pressure test required by Table IWB-2500-1, Category B-P, the following rules shall be used.
- 1. A system leakage test (IWB-5221) shall be conducted at or near the end of each inspection interval, prior to reactor startup.
- 2. The boundary subject to test pressurization during the system leakage test shall extend to all class 1 pressure retaining components within the system boundary.
- 3. Prior to performing VT-2 visual examination, the system shall be pressurized to nominal 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 systems and 10 minutes for noninsulated systems. The system shall be maintained at nominal operating pressure during the performance of the VT-2 visual examination.
- 4. Test temperatures and pressures shall not exceed limiting conditions for hydrostatic test curve as contained in the plant Technical Specifications.
- 5. The VT-2 visual examinatio'n shall include all components within the boundary identified in (2) above.
(b) It is also the opinion of the (ASME) Committee that, as an alternative to the 10-year hydrostatic pressure test required by Table IWC-2500-1, Category C-H, the following rules shall be used.
- 1. A system pressure test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of Inspection Program B.
- 2. The boundary subject to test pressurization during the system pressure test shall extend to all Class 2 components included in those portions of systems required to operate or support the safety system function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required.
- 3. Prior to performing VT-2 visual examination, the system shall be pressurized to nominal 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 insulated systems and 10 minutes for noninsulated systems. 1 The system shall be maintained at nominal operating pressure during the performance of the VT-2 visual examination. l l
- 4. The VT-2 visual examination shall include all components within I the boundary identified in (2) above. l f
l l
l 4-26
m Revision 3 5.0 IsZ FLAN StaedhRE Table 5.0-1, "ISI Plan Summary" contains the NDE program tables. These tables are presented in a tabular format consistent with Section XI examination ,
requirements tables (IWB IWC, and IWF 2500-1). The NDE program tables include the correspondit , code category, code-item, desEription and exam requirements in conformance with Inspection Program B, ASME Section XI, 1983 Edition through the Summer 1983 Addenda. Per 10CFR50.55a(b)(2)(iv), Class 2 portions of Emergency Core Cooling Systems (ECCS), Containment Heat Removal Systems, and Residual Heat Removal Systems were selected / exempted based upon the 1974 Edition through the Summer 1975 Addenda of ASME Section XI. Also included are tables which address Augmented Examinations.
A complete listing of those specific weldments and components subject to examination is maintained in the Braidwood Station ISIBASE computer program.
The following table gives the total number of components per item for each Braidwood Unit. The numbers are identical except where noted on the table.
6 i
I 5-1
Revision 2 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND ;
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION '
REMARKS B-A Pressure retaining welds in reactor vessel Bl.lO Shell welds Bl.ll Circumferential 3 VOL NR-9 All welds Bl.12 Longitudinal O Bl.20 Head welds Bl.21 Circumferential 2 VOL Note 9 Accessible Bl.22 Meridional O length of all welds Bl.30 Shell-to-Flange weld 1 VOL l-weld Bl.40 Head-to-Flange weld 1 VOL and SUR NR-9 1-weld Bl.50 Repair welds Bl.51 Beltline Region O I
o 5-2
_ _ - - _ . _ . - _ _ _ _ _ ~
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL .
APPROACH g EXAMINATION ITEM NUMBER OF EXAM RELIEF AND -
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION JtEMARKS B-B Pressure Retaining welds in vessel other than Reactor Vessel.
Pressurizer B2.lO Shell-to-Head Welds B2.11 Circumferential 2 VOL 2-welds B2.12 Longitudinal 2 VOL 1Ft of weld.
AUG Shell-to-Shell Weld 1 VOL NOTE 7 B2.20 Head Welds B2.21 Circumferential O B2.22 Meridional 0 Steam Generators B2.30 Head Welds B2.31 Circumferential 0 B2.32 Meridional 0 e B2.40 Tube sheet-to-Head Weld 4 VOL All welds 1
5-3
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Revision 2 TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH ;
EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REQUEST POSITION REMARKS B-D Full penetration welds of nozzles in vessels (Inspection Program B)
Reactor Vessel B3.90 Nozzle-Vessel Welds 8 VOL All Welds B3.100 Nozzle Inner Radius 8 VOL All Welds Pressurizer B3.110 Nozzle-Vessel Welds 6 VOL All Welds B3.120 Inner Radius 6 VOL All Welds Steam Generators B3.130 Nozzle-to-Vessel Welds O B3.140 Nozzle Inner Radius 8 VOL NR-21 All Welds Heat Exchangers (Unit 1 only)
B3.150 Nozzle-to-Vessel Welds O I B3.160 Nozzle Inner Radius. O l
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Revision 2' TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- cf astS 1 1
TECHNICAL APPROACH ,
EXAMINATION ITEM NUMBER OF EXAM RELIEF AND f CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION '
REMARKS B-F Pressure Retaining Dis-similar Metal Welds Reactor Vessel B5.10 Nozzle-to-Safe End Butt 8 VOL and SUR All Welds Welds 2 4 in NPS B5.20 Nozzle-to-Safe End Butt O welds < 4 in NPS B5.30 Nozzle-to-Safe End O Socket Welds Pressurizer B5.40 Nozzle-to-Safe End Butt 6 VOL and SUR All Welds welds 2 4 in NPS B5.50 Nozzle-to-Safe End Butt O welds < 4 in NPS l
f 5-7
_ _ _ _ _ _ _ .. ,_.________._m_____m__m m .er- . mm_-,u ,_m_ - - .__-,, m.___.m m --w- w r. w. . , , - _ , _ + . ,,w e . = _ + e__ -_ %
Revision 2 TABLE 5.0-1 Comed Ermidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUNBER OF EXAM RELIEF AND I CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B5.60 Nozzle-to-Safe End O Socket Welds Steam Generator B5.70 Nozzle-to-Safe End Butt 8 NR-5 All Welds Welds 2 4 in. NPS B5.80 Nozzle-to-Safe End Butt O Welds < 4 in. NPS B5.90 Nozzle-to-Safe End O Socket Welds Heat Exchangers B5. LOO Nozzle-to-Safe End Butt O Welds 2 4 in. NPS B5.llO Nozzle-to-Safe End Butt O Welds < 4 in.
l 5-8
._ . _ . _ . . . . _ . . __. _ _ - . . _ _ . . . . - _ _ _ _ _ _ _ . . ~. ,. . _ _ . _ _ _
Revision 2 TABLE 5.0-1 Costed Straidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REOUEST POSITION REMARKS B5.120 Nozzle-to-Safe End O Socket Welds Piping B5.130 Dissimilar Metal Butt O Welds 2 4 in.
B5.140 Dissimilar Metal Butt O Welds < 4 in.
B5.150 Dissimilar Metal socket O Welds l
5-9
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH ,
EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REQUEST POSITION REMARKS B-G-1 Pressure staining Bolting > 2in. DIA Reactor Vessel B6.10 Closure Head Nute 54 SUR All B6.20 Closure Stude, in place 54 VOL All B6.30 closure Studs, removed 54 VOL and SUR All B6.40 Threads in Flange 54 VOL All B6.50 Closure 54 VT-1 All Washers / Bushings Pressurizer B6.60 Bolts and Studs O B6.70 Flange Surface O B6.80 Nuts, Bushings, Washers O Steam Generators B6.90 Bolts and Stude- 0-B6.100 Flange Surface 0 l 5-10 u_---__-.______________ _-_-_s,-+w -,,ev- --,. . - .~.----er -
- + - - % .--- w e e . < . . , - - - . . - , , , .. - . , , . -
'I Revision 2 TABLE 5.0-1 Comed traidwood Nuclear Station Units 1 & 2 ISI PIM
SUMMARY
- CI. ASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B6.110 Nuts, Bushings, Washers O Heat Exchangers B6.120 Bolte and Studs O B6.130 Flange Surface O B6.140 Nuts, Bushings, Washers O Piping B6.150 Bolts and Studs O B6.160 Flange Surface O B6.170 Nuts, Bushings, Washers O Pumps B6.180 Bolts and Studs 24 VOL Note 2 1 Pump B6.190 Flange Surface, When 24 VT-1 Note 2 1 Pump Connection is Disassembled B6.2OO Nuts, Bushings, Washers 24 VT-1 1 Pump Note-2 l 5-11
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- cfMS 1 TECIBtICAL APPROhCH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND I CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITICII REMARKS Valves 86.210 Bolts and Stude 8 valves VOL Note 2 1 valve B6.220 Flange Surface, When 8 valves VT-1 Note 2 1 Valve Connection is Note 2 1 Valve Disassembled B6.230 Nuts, Bushings, Washers 8 valves VT-1 I
l 5-12 l
Revision 2 TABLE 5.0-1 Comed traidwood Nuclear Station Unita 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EKAM RELIEF AND l CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS C-G-2 Pressure Retainine Bolting 5 2 in. dia.
Reactor Vessel B7.10 Bolts, Stude and Nute 7 assemblies VT-1 2 RVLIS & 5 incore thermal-couples Pressurizer B7.20 Bolts, Stude and Nuts 1 Manway VT-1 Primary Manway Steam Generator B7.30 Bolts, Stude, and Nuts 8 Manways VT-1 Primary Manway Heat Exchangers B7.40 Bolts, Stude and Nute O Piping B7.50 Bolts, Stude and Nuta 28(U2-29) VT-1 Ul-16(U2-22) 5-13
Revision 2 TABLE 5.0-1 Comed Ermidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS Pumps B7.60 Bolts, Stude and Nute 4 pumps VT-1 Note 3 1 Pump B7.70 valves 72 VT-1 Note 3 21 valves CRD Housings B7.80 Bolts, Stude and Nuts O
+
r 5-14
Revision 2.
TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- rrmm_a 1 TECHNIchL APPROACH ,
EXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION ECMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B-H Integral attachments for vessels Reactor Vessels B8.lO Integrally welded O attachments Pressurizer B8.20 Integrally welded 1 SUR Support Skirt attachments Steam Generator B8.30 Integrally welded O attachments Heat Exchanger B8.40 Integrally welded O
, attachments l 1
I l 5-15 l
l
_- -_ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ . . . -~ _ _ . . _ - - - . - - _ . . _ - - _ _ , . _ _ . - . . _ - ,
Revision.2 TABIJI: 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH ,
EXAMINATION ITEM NUMBER OF EXAM RELIEF AND '
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS B-J Pressure Retaining Welds in Piping B9.10 Nominal Pipe Size > 4" Dia.
B9.11 Circumferential Welds 604(U2-554) VOL and SUR NR-2,6,7,18 25% of the Cire Weld B9.12 Longitudinal Welds 0 VOL and SUR 89.20 Nominal Pipe Size <4 in dia.
B9.21 Circumferential Welds 265(U2-232) VOL and SUR 25% of the Circ Weld B9.22 Longitudinal O B9.30 3 ranch Pipe Connections B9.31 h wninal Pipe Size 2 41n 11(U2-13) VOL and SUR 25% of the Welds B9.32 Noninal Pipe Size < 41n 56(U2-76) SUR 25% of the l
Welds 89.40 Socket Welds 1061(U2- SUR 25% of the 1079) Welds 5-16
Revision 2 TABLE 5.0-1 Comed Ermidwood Nuclear Station Unite 1 & 2 ISI PLAN SUN'fARY - CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND I CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REOUEST POSITION REMARKS C-K-1 Integral attachments for Piping, Pumps, and valves Piping B10.10 Integrally Welded 7(U2-28) SUR 3-U1 Attachments 22-U2 Pumps B10.20 Integrally Welded 0 Attachments Valves BlO.30 Integrally Welded O Attachments l
5-17
Revision 2 TABLE 5.0-1 Comed Ermidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF !
AND
. CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS E-L-1 Pressure Retaining Welds in pump casings B12.10 Pump Casing Welds 0 0-L-2 B12.20 Pump Casings 4 VT-3 NR-14 *1 Pump B-M-1 Pressure retaining welds in valve bodies B12.30 Valve body welds <4 in. O B12.40 valve body welds 14 in. O C-M-2 B12.50 Valve Body 41 VT-3 NR-3 *9 Valve
. Bodies l
- NOTE:
Those for the interval.
pump and valve groups that have not been inspected should be noted in the last 90-day outage report 5-18
Revision 2 TABLE 5.0-1 Comed Itraidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH CIAMINATION ITEM NUMBER OF EXAM RELIEF AND !.
CATEGORY NUMBER DESCRIPTION COMPONENTS REOUIREMENTS REQUEST POSITION REMARKS B-N-1 Interior of Reactor Vessel B13.10 Vessel Interior 1 VT-3 Accessible Areas B13.50 Interior Attachments All VT-1 Accessible within Beltline Region Welds B13.60 Interior Attachments All VT-3 Accessible beyond Beltline Region Welds B13.70 Core Support Structure All VT-3 Accessible B-O Welds Pressure retaining welds in control rod housing B14.10 Welds in CRD housing 45 SUR 10% Peri-pheral Welds l
5-19
Revision 2 TABLE 5.0-1.
Comed li;raidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXANINATION ITEM NUMBER OF EXAM RELIEF AND ;
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REOUEST POSITION ' REM &RKS B-P All Pressure Retaining Components Reactor Vessel B15.10 Pressure Retaining ALL VT-2 Sys Leakage Boundary Test B15.11 Pressure Retaining ALL VT-2 Note 17 Sys Hydro-Boundary static Test Pressurizer B15.20 Pressure Retaining ALL VT-2 Sys Leakage Boundary Test B15.21 Pressure Retaining ALL VT-2 Note 17 Sys Hydro-Boundary static Test Steam Generators B15.30 Pressure Retaining ALL VT-2 sys Leakage Boundary Test B15.31 Pressure Retaining ALL VT-2 Note 17 i Sys Hydro-Boundary I static Test Heat Exchanger B15.40 Pressure Retaining O VT-2 Sys Leakage Boundary Test B15.41 Pressure Retaining 0 VT-2 Sys Hydro-Boundary static Test 5-20
1 Revision 2-TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 i
ISI PLAN
SUMMARY
- CLASS 1 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND g CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION RER&RKS Q-P (Cont.) B15.50 Piping Pressure Retaining All VT-2 Sys Leakage Boundary Test B15.51 Pressure Retainirn All VT-2 Note 17 Sys Hydro-Boundary static Test Pumps B15.60 Pressure Retaining All VT-2 Sys Leakage Boundary Test B15.61 Pressure Retaining All VT-2 Note 17 Sys Hydro-Boundary static Test Valves B15.70 Pressure Retaining All VT-2 Sys Leakage Boundary Test B15.71 Pressure Retaining All VT-2 Note 17 Sys Hydro-Boundary static Test S-Q Steam Generator l
i B16.20 U-Tube Design 4 VOL Note 4 ' Tech. Spec l
4.5.0-1 5-21
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 2 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND l CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION ' REMARKS C-A Pressure Retaining Welds in Pressure Vessel C1.10 Shell Circ Welds 14 VOL Note 7, 13 4 Welds C1.20 Head Circ. Welds 6 VOL Note 13 2 Welds C1.30 Tubesheet-to-Shell 4 VOL 1 Weld Welds C-B Pressure Retaining Nozzle Welds in Vessels O
C2.10 Nozzles in Vessels i "
Nominal Thickness O C2.11 Nozzle-to-Shell Welds C2.20 Nozzle Without Reinforcing Plates in Vessel > hin. Thickness C2.21 Nozzle-to-Shell Weld 16 SUR and VOL NR-20 Note 15 5 Welds C2.22 Nozzle Inside Radius 12 VOL NR-12 Note 1 1 Weld 5-22
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PI.AN
SUMMARY
- CLASS 2 TECHNICAL APPROACH CXAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY y NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS C2.30 Nozzles with Reinforcing Plates
> in. Noeninal Thickness C2.31 Plate Welds to Nozzles O C2.32 Nozzle to Shell (or 0 Head) Welds When Inside of vessel is Accessible C2.33 Nozzle to Shell (or 0 Head) Welds When Inside of Vessel is Inaccessible l
5-23
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 2 TECHNICAL APPROACH CIAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION IREMARKS C-C Integral Attachments for Piping, Pumps and Valves Pressure vessels C3.10 Integrally Welded 2 SUR Note 13 1 Welds Attachments Piping C3.20 Integrally Welded 43(U2-37) SUR 26 Attachments Attachments Unit 1 33 Attachments Unit 2 Pumps C3.30 Integrally Welded 14 SUR NR-15 7 Attachments Attachments Valves C3.40 Integrally Welded 0 Attachments I
C-D Pressure Retaining Bolting > 2in. Diameter Pressure Vessels C4.10 Bolts and Studs 0 5-24
Revision 2' TABLE 5.0-1 Comed Braidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 2 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND i CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION ; REMARKS Piping C4.20 Bolts and Studs O Pumps C4.30 Bolts and Stude O Valves C4.40 Bolts and Studs O C-F Pressure Retaining Welds in Piping C5.10 Piping Welds $ in.
Nominal Wall Thickness C5.11 Circumferential Welds 511(U2-475) SUR U1=82 welds U2=86 welds C5.12 Longitudinal Weld 354(U2-185) SUR U1=23 welds U2=16 welds C5.20 Piping Welds > "
C5.21 Circumferential Welds 497(U2-497) SUR and VOL NR-16 U1-94 welde U2=100 welds C5.22 Longitudinal Welds 183(U2-221) SUR and VOL l U1=44 welds U2=47 welds l
5-25
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CIASS 2 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF PRD !
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS C5.30 Pipe Branch Connections
> 41n. Branch Pipe Size C5.31 Circumferential Welds 34(34-U2) .SUR NR-1 U1=11 welds U2=10 welds C-G C5.32 Longitudinal Welds 3-41 None Pressure Retaining Welds in Pumps and Viva Pumps C6.10 Pump Casing Welds O '
Valves C-H C6.20 Valves Body Welds O All Pressure Retaining Components Pressure vessels VT-2 System C7.10 Pressure Retaining ALL Pressure Test Components l 5-26
Revision 2-TABLE 5.0-1 Comed Ermidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 2 TECHNICAL APPROACH EKAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REOUEST POSITION I REMARKS C7.20 Pressure Retaining ALL VT-2 Note 17 System Hydro-Components static Test Piping C7.30 Pressure Retaining ALL VT-2 System Components Pressure Test C7.40 Pressure Retaining ALL VT-2 Note 17 System Hydro-Components static Test Pumps C7.50 Pressure Retaining ALL ,VT-2 Components System Pressure Test C7.60 Pressure Retaining ALL W-2 Components System Hydro-static Test Valves C7.70 Pressure Retaining ALL VT-2 Components System Pressure Test C7.80 Pressure Retaining ALL VT-2 Note U Components System Hydro-static Test 5-27
Revision 2 TABLE 5.0-1 Comed Eraidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 3 TECHNICAL APPROACH EXAMINATION ITEM NUMBER OF EXAM RELIEF AND !
CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS D-A Systems in Support of Reactor Shutdown Function D1.10 Pressure Retaining ALL VT-2 NR-23 Leakage Each-Components Period Hydro-static Each Interval D1.20 Integral Attachments 46(U2-35) VT-3/4 U1=35 U2=27 Component Supports D1.30 Integral Attachments O Snubbers D1.40 Integral Attachments O Spring Type Supports l D1.50 Integral Attachments O Constant Load supports I DI.60 Integral Attachments O Shock Absorbers D-B Systems in Support of ECCS Leakage Each D2.10 Pressure Retaining ALL VT-2 Period, Components Hydro-static Each Interval 5-28
Revision 2 TABLE 5.0-1 CosRd Ermidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 3 TECHNICAL APPROACH j ChAMINATION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION BEM&RKS D2.20 Integral Attachments 67(31-U2) VT-3 Ul=44 U2=15 ,
Component Supports D2.30 Integral Attachments O Snubbers D2.40 Integral Attachments O Spring Type Supports D2.50 Integral Attachments O Constant Load Supports D2.60 Integral Attachments O Shock Absorbers D-C System in Support of RHR from the Spent Fuel Pool D3.10 Pressure Retaining ALL VT-2 Components D3.20 Integral Attachments 1-U2 VT-3/4 Component Supports D3.30 Integral Attachments O Snubbers D3.40 Integral Attachments 0 . Spring Type Supports D3.30 Integral Attachments O Constant Load Supports 5-29
. . _ _ . . - , , - _ _ _ _ _ . - . .. . _ . _ . . _ . - . . . ...__ -- _ _ - , _ . _ -_ . _ . _ . _- _ _ . . . - _ . . ~ - -_ . - _ - . . _ _ _ - - _ . _ . - _ _ --
Revision 2 TABLE 5.0-1 Comed Craidwood Nuclear Station Units 1 & 2 ISI PLAN
SUMMARY
- CLASS 3 I TECHNICAL APPROACH EXAMINJ. TION ITEM NUMBER OF EXAM RELIEF AND CATEGORY NUMBER DESCRIPTION COMPONENTS REQUIREMENTS REQUEST POSITION REMARKS D-C D3.60 Integral Attachments O Shock Absorbers AUG Augmented High Energy U1=282 VOL Note 5 All Welds Main Steam & Feedwater U2=297 AUG Reactor Coolant Pump 4 VOL and SUR Note 6 Each Period Flywheel Examination AUG Turbine Rotors Note 8 AUG Locations Subject to 7 VOL Thermal Stratification Note 14 Each Perio.1 AUG ECCS IGSCC Ul-789 VOL Note 10 7 1/2% of the Examination U2-783 Welds Tables for subsection INE have not been approved for use. Components identified in subsection IWF selected for examination shall be the supports of those components that are required to be examined under IWB, IWC, and IND during the first inspection interval.
5-30
_ __ _ - . _ - . . _ _ . - . . - _ _ _ - . _ _ _ . -.- - _ ~ _ . , _ _ _ . _ - . . . . . _ . _ _ _ . , . _ _ _ _ . _ .
I Revision 3 Table 5.0-2 l Method Abbreviations !
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Exam Description SUR A Surface Examination method (e.g. a liquid penetrant or magnetic particle examination) used to detect the presence of surface ;
cracks or discontinuities (IWA-2220) !
VOL A Volumetric Examination method (e.g. radiographic examination, ultrasonic examination, or eddy current examination) used to detect the presence of discontinuities throughout the volume of i the material (IWA-2230) ;
i VT-1 Visual Examination conducted to determine the condition of a part, ;
component, or surface (IWA-2211)
VT-2 Visual Examination conducted to locate evidence of leakage from pressure retaining components, or abnormal leakage from components j with or without leakage collection systems as required during the ,
conduct of system pressure or functional test (IWA-2212)
}
VT-3 Visual Examination conducted to determine the general mechanical !
and structural condition of components and their supports (IWA- I 2213) and the operability of components or devices (IWA-2214)-
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t 5-31 '
Revision 2 i
j 6.0 ISI Plan susanary for component support l The component supports to be examined are the supports of those components required to be examined under the rules of Subsections IWB-1220, IWC-1210 and IWD-1210, during the first inspection interval. Component supports exempt from examination are the supports of those components exempted under subsections IWB-1220, IWC-1220 and IWD-1220. Additionally, the examinations are to be completed in accordance with the examination schedule established for the components required to be examined under the rules of IWB, IWC and IWD. For multiple components within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. For these reasons, component supports selected for examination, and the scheduling of examinations, are coordinated with the examinations performed under the rules of Subsections IWB, IWC and IWD, described in the Inservice Inspection Program Plan for Nondestructive Examination.
Subsection IWF, in conjunction with IWB, requires visual examinations (VT-3/VT-4) to be performed on the component supports of class 1 components and piping greater than 1 inch nominal pipe size.
Visual examinations (VT-3/VT-4) are required to be performed on the component supports of class 2 and class 3 components not exempt from examination by IWC-1220 and IWD-1220. However, as described by Note 1, this does not include the exemptions allowed for the Class 2 Auxiliary Feedwater System Components. The componwnt supports of the Class 2 portions of the Auxiliary Feedwater System will be examined even though they are exempt by IWC.
When the results of the visual examinations require corrective measures in accordance with the provisions of Article IWF-3000, the component" supports immediately adjacent to those components requiring corrective action shall be examined, except as described by Relief Request CR-2.
In addition, the examination shall be extended to include additional supports equal in number and similar in type, design, and function to those initially examined during the inspection. When these additional examinations require corrective measures in accordance with the provisions of IWF-3000, the remaining component supports within the system, of the same type, design and function, as selected for the additional examinations, shall be examined.
Component supports which require corrective measures in accordance with the provisions of IWF-3000, shall be reexamined during the next inspection period. A complete listing of component supports is maintained procedurally at the station.
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6-1
Revision 2 RELIEF REQUEST CR-1 :
Relief Deleted 6-2
Revision 1 RELIEF REQUEST CR-2
- 1. SYSTEMS: All ~
- 2. NUMBER OF ITEMS: All -
- 3. ASME CODE CLASS: 1, 2, and 3
- 4. ASME SECTION XI CODE REOUIREMENTS: Subarticle IWF-2430 states in part, "When the results of examinations require corrective measures in accordance with the provisions of IWF-3000, the component supports immediately adjacent to those requiring corrective action shall be examined."
- 5. BASIS POR RELIEF: The additional examinations required by failure to satisfy the acceptance standards of IWF-3400 are not limited by section XI to the type .
of indication identified, the type of support being examined, or the types of 1 adjacent supports. This would require that a documented visual examination be performed which, in some cases, would provide no meaningful results. These actions cause undue hardships and increase ALARA concerns without significantly affecting the overall safety of the plant.
- 6. ALTERNATE TEST METHOD: A visual examination is performed on component supports required to be examined by ASME Section XI. When corrective actions are required as a result of these examinations not satisfying the requirements of IWF-3400, the adjacent supports will be examined if they are susceptible to the same type of indication or mode of failure, regardless of support type. ;
- 7. JUSTIFICATION: Corrective actions are' required for indications failing to satisfy the acceptance standards of IWF-3400. The acceptance standards of IWF-3400 are both generic and specific in nature and therefore, in some cases, apply only to a particular type of support. It is impractical to perform an examination on an adjacent support when the indication identified is not applicable to that type of support. If the indication requiring corrective j measures can be applied to the adjacent supports, they will be examined.
Since additional examinations are required and performed in accordance with IWF-2430 and the components are reexamined per the requirements of IWF-2420 during the next inspection period, this provides a high degree of system reliability and safety while providing ALARA benefits to plant employees.
- 8. APPLICATION: This relief request applies to VT-3 and VT-4 visual examinations performsd on component supports.
- 9. IMPLEMENTATION: Visual examinations are performed during plant operations or normal plant outages.
- 10. TIME PERIOD: This request for relief applies for the first ten year interval.
- 11. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
I 6-3 ;
Revision 1 NOTE 1 Braidwood Station's Auxiliary Feedwater System consists of Class 2 and Class 3 components. The class 2 portions are 4 inches or less in size and the Class 3 portions are 6 inches or less in size.
The class 2 components are exempt from examination per IWC-1220 which states in part, "The following components shall be exempted from the inservice examination requirements of IWC-2500 . . . (c) Component connections (including nozzles in vessels and pumps), piping and associated valves, and vessels and their attachments that are 4 in, nominal pipe size and smaller." However, the component supports of Class 2 portions of the Auxiliary Feedwater system will still be examined, regardless of the exemption.
The Class 3 components of the Auxiliary Feedwater System that are 4 in. or less in size are not exempt from examination per IWD-1220.1 which states in part, " Integral attachments of supports and restraints to components that are 4 in, nominal pipe size and smaller . . . are exempt from the visual examination VT-3, exceot for PWR Auxiliary Feedwater System." Auxiliary Feedwater piping greater than 4 in. is not exempt from examination per IWD 1220.2 (a) which states that components are exempt, provided they are " located in system . . . whose function is D21 required in support in reactor residual heat removal containment heat removal, and emergency core cooling."
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l Revision 2 f
7.0 Ist plan susumary for snubbers j The Inservice Inspection requirements for snubbers are required by the Braidwood !
Technical specifications, and in general, are in excess of the section XI -j requirements. However, where conflicts exist between the Braidwood Technical-specifications and the ASME section XI Code, or where section XI requirements are ;
determined to be impractical, specific relief requests are included. ;
A visual examination is required of all Class 1, 2 and 3 safety related snubbers in' l accordance with the schedule shown in Technical specification, 3/4.7.8 snubbers may !
be grouped as accessible and inaccessible during normal operations and inspected i independently in accordance with the examination schedule. Additionally, an .
i examination shall be performed on snubbers attached to sections of systems that have '
experienced unexpected, potentially damaging transients, following such an event, !
within the time limitations of the Technical specification. The purpose of this ;
examination is to verify freedom of motion for mechanical anubbers by evaluation of ;
the snubber setting, partial stroking, or full stroking of the mechanical snubber. ;
The boundary for visual examinations is defined by subarticle IWF-1300 and Fig. IWF- ;
1300-1. This examination boundary is limited in some cases described relief request :
SR-1. ,
The visual examination acceptance criteria is listed in the Technical Specifications !
and includes verification that there are no visible indications of damage or j impaired operability, and that fasteners and attachments to the component, snubber, !
and supporting structure are functional. The standards of subarticle IWF-2400 shall be used for evaluation of indications.
subarticle IWF-2430 requires that additional examinations be performed when '
corrective measures are required in accordance with the provisions of IWF-3000.-
Relief Request SR-2 describes the limited application of these requirements of additional examinations.
Per Braidwood station's Technical Specifications, snubbers which appears inoperable as a result of the visual examination may be determined operable for the purpose of ;
establishing the next visual examination interval, provided thats j a
- 1) The cause for the rejection is clearly established and remedied for that j particular snubber and for other snubbers irrespective of type that may be i generically susceptible; and l
- 2) The affected snubber is functionally tested in the as-found condition and j determined operable per the functional testing acceptance criteria. !
Functional testing of snubbers is scheduled to coincide with scheduled refueling outages at intervals of approximately 18 months. Testing is performed utilizing the 1 sampling plans identified in the Technical specifications. The representative ;
sample is randomly chosen from the various types of snubbers installed in the plant ,
and is reviewed to ensure, as far as practical, that the samples are representative ,
of the various configurations, operating environments, range of size and capacity of '
each type of snubber. The total numoer of snubbers functionally tested le dependent upon the sample plan chosen and the number of failures. l The functional testing acceptance criteria is listed in the Technical Specifications and satisfies the inservice test requirements of subarticle IWF-5400, paragrapl:s i b(1), b(2), and b(3) for snubbers less than 50 kips. Although there currently are ,
no ASME Section XI requirements for' functional testing of snubbers 50 kips or j greater, functional testing is required by the Technical specifications for ,
utilising testing methods to measure parameters indirectly, or parameters other than ;
those specified, if those results can be correlated to the specified parameters i through established methods. !
)
7-1 t
Revision 2 7.0 ISI Plan sussnary for snubbers (cont.)
snubbers which fail the inservice test require an engineering evaluation, additional testing, and corrective measures in accordance with the Techni_ cal Specifications and ASME Section II.
An engineering evaluation is performed on the components to which the inoperable snubbers are attached. The purpose of this evaluation is to determine if the components were adversely affected by the inoperability of the snubbers in c er to ensure that the components remain capable of meeting their designed service. An engineering evaluation shall be made of each failure to meet the functional tes' ac acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in ca ef fort to determine the operability of other snubbers irrespective of type which way be subject to the same failure mode. If any snubber selected for functional test ty either fails to activate or fails to move, the cause will be evaluated and, if caused by manufacturer or design deficioney, all snubbers of the same type subject to the same failure mode. If any snubber selected for functional testing either fails to activate or fails to move, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be functional tested.
If additional samples are selected in accordance with the sampling plans, the selection of snubbers shall be based upon the angineering evaluation.
If a snubber fails the inservice functional test requirements, the snubber, or.its replacement, shall be ratested at the time of the'next scheduled Yunctional testing ,
but shall not be included in the sample plan. '
A anubber service life program is required by the Technical Specifications. The service life of various components are established by engineering information and shall be adjusted based upon test results and failure histories. The purpose of this program is to ensure that the service life is required to be operable. A complete listing of component supports is maintained procedurally at the station.
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7-2
Revision 1 ,
NOTE 1 .
Braidwood station design incorporates to use of large bore hydraulic snubbers (greater than 50 kips) on the steam generators. ASME section-XI does not address testing requirements for snubbers with a capacity of 50 kips or greater. However, i testing of these snubbers is required by the Braidwood Technical Specifications and will be performed in accordance with these specifications. Testing methods may be used to measure parameters other than those specified if those results can be correlated to the given parameters through established methods. Testing also may be performed at the subcomponent level to verify that the acceptance criteria has been i satisfied.
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h 7-3
d Revision 2 RELIEF REQUEST SR-1
- 1. SYSTEMS: All non exempt portions of safety related piping systems.
- 2. NUMBER OF ITEMS: Snubbers in the following systems: _ ,
Systems snubbers Attached to Insulated Pipe Unit 1 Unit 2 ;
I CS 0 9 CV 82 91 DO O 10 FW 4 42 MS 12 12 ,
RC 93 80 RH 24 28 !
RY 31 26 SD 15 21 ,
Total snubbers on insulated lines 264 320 i Total snubbers population 369 384 '
Table showing the number of snubbers on insulated lines vs. the total number of snubbers in the population. The above numbers will vary with time as a result of ,
snubber reduction and other plant modifications.
l
- 3. ASME CODE CLASS: 1, 2 anc 3. ,
- 4. ASME SECTION II CODE REOUIKEMENTS: The component support examination ,
boundaries are defined by Ihr-1300 a..a Figure 1300-1. Per IWF-1300e, the IWF ,
support exam boundary for snubbers which have non-integral attachments extends from the contact surface between the component and the support to the surface ,
of the building structure.
- 5. BASIS FOR RELIEF: The visual examination of an integral or non-integral pipe attachment is limited by the insulation installed on the piping. It would ;
impose a great deal of hardship in terms of manpower, time and radiation i exposure to remove insulation to visually inspect all snubber pipe clamps, I particularly if there are alternative methods that provide an equivalent means i of determining pipe clamp integrity. )
1 The majority of snubbers are located inside containment in high radiation l areas. Removing insulation on all anubber pipe clamps would require one l Health Physics Technician to survey the insulation prior to its removal and l then a two man insulator crew to remove the insulation. This would add three l people to the customary two man inspection crew, which would more than double ]
the man rem exposures for performing the surveillance.
7-4 1
_ _ _ _ _ _ _ _ _j
i Revision 2 melief megeest sR-1 (Cent. ) ,
t The CBCo SPPM VT-3/4 procedure allows remote inspections to be performaid on ;
onubbere that are out of reach for direct. inspection. Scaffolds or man baskets would have to be used to remove insulation on remove snubbers. Extra ,
scaffolds built for snubber pipe clamp insulation removal would increase .;
congestion in containment and increase the amount-of material being handled :
and surveyed during the outage. It would also produce additional DAW and !~
result in more scaffold material acquiring fixed contamination during the outage. It would pose an additional burden on the examination in terms of manpower, time, safety and radiation exposure. It also defeats the purpose of j the remote examination methods allowed in the SPPM. ,
- 6. ALTERNATE TEST METHOD: ASME Section XI Code, IWA-2240, allows for alternate examination method if they provide results that are equivalent to the ,
specified method. In lieu of removing insulation on snubber pipe clamps, the following alternate exam methods will be employed on all snubbers that are accessible for direct examination:
- a. A hands-on inspection of the pipe clamp will be performed to verify the clamp is tight.
- b. Clamp alignment with the load pin axis will be observed to verify alignment is within design tolerances.
- c. The load pin / stud will be inspected to verify its integrity. This will '
insure that parts are in place and that the pin is tight. If the' load pin is obscured by insulation, the insulation will be removed or modified to allow for this inspection.
- d. Insulation will be checked for evidence of damage due to slipped or loose clamps. >
- e. If boric acid contamination or corrosion is observed, insulation will be removed to inspect the pipe clamp.
- 7. JUSTIFICATION: This relief request is intended for non-integral attachments !
on insulated lines. The visual inspection of snubbers are performed using the Ceco SPPM VT-3/4 procedure. The inspections are performed on all safety related snubbers every 18 months i 25%. Under this procedure, support indications to be observed and documented include the following:
- cracks, pitting
- erosion, corrosion, wear
- loose, missing, damaged parts ,
- contamination, debris !
- weld degradation
- slipped clamps
- arc strikes, weld spatter, paint
- clearances, settings
- condition of spherical bearings ,
The proposed alternate exam methods listed in part 6 of the Relief Request i enhance the SPPM VT-3/4 inspection procedure. The hands on check combined with the VT-3/4 procedure will insure that snubber pipe clamps are installed and secure on snubbers that can be reached for direct examination. If there are any indications of degradation, the insulation will be removed to allow for a total clamp inspection. '
7-5 i
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Revision 2 l melief moquest SR-1 (Cont. )
l For the snubbers that have pipe clamps completely buried in insulation, the insulation will be removed for a complete inspection. Based on a review of the piping line list and snubber data bases this condition occurs primarily on PSA-1/4 anubbers attached to 1 in. or less piping covered with 2 in. thick or greater insulation. Insulation removal to inspect these pipe clamps will be accomplished several ways. When these snubbers are removed for functional testing, the insulation will have to be removed to unpin the snubber. The visual inspection of the pipe clamp will be performed at this time.
Visuals on pipe clamps will be coordinated with NDE inspections when insulation removal to access welds also exposes the pipe clamp for inspection.
Snubber exams are documented so it can be verified that 11 snubber pipe clamps completely buried in insulation receive a visual inspection with the ten year interval.
On snubbers that are inaccessible for direct examination, a remote exam will be performed per the SPPM VT-3/4 procedure. This would include snubbers in heat exchanger pits or on overhead runs of piping that are out of reach. The number of snubbers in this category represents only about 5% of the snubbers in Units 1 and 2. None of these snubbers have pipe clamps completely buried in insulation. They are not in high traffic areas where typically the most clamp slippage and other damage is experienced. Boric acid spray from valves or flanges is also less likely on the majority of these-snubbers. Because they are in relatively safe areas, an indirect inspection verifying no outward indications are present will demonstrate that the pipe clamps are secure'.
When these snubbers are functionally tested, scaffold will be built to access them. A hands on exam of the pipe clamp will be done at that time. If any of the above conditions are present, scaffold will be built to more thoroughly investigate the indication.
Previous inspection experience has not shown an increasing trend in regard to
. loose pipe clamps. A review of earlier data indicates that the incidence of loose pipe clamps is rare. The alternate methods proposed in this relief request combined with the commitment to remove insulation on those pipe clamps completely obscured by insulation will provide a complete examination of the entire snubber population in Units 1 and 2. This approach meets the requirements of an alternate inspection and will provide a high degree of confidence that the snubber pipe clamps are in place and secure. By limiting the number of people required to perform the surveillance, the proposed methods will minbmize man rem exposures. They will eliminate the need for additional scaffolding, which will lower the amount of contaminated material produced during the outage, and reduce the traffic and congestion associated with moving scaffolding in and out of containment. This will promote the efficient and cost effective execution of the visual surveillances.
- 8. APPLICATION: This request for relief applies to visual examinations (VT-3/4) of non-integral attachments to snubbers.
- 9. IMPLEMENTATION: Examinations may be performed during plant operations or normal plant shutdowns.
- 10. IIME PERIOD: This request for relief applies for the first ten year interval.
- 11. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
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RELIEF REQUEST SR-2 ;
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- 2. NUMEER OF ITEMS: Nonexampt snubbers _, ;
- 3. ASME CODE CLASS: 1, 2, and 3 I
- 4. ASME SECTION II CODE REOUIREMENTS: Subarticle IWF-2430 states in part, "When the results of examinations require corrective measures in accordance with the j provisions of IWF-3000, the component supports immediately adjacent to those t requiring corrective action shall be examined." ;
- 5. EASIS FOR RELIEF: The additional examinations required by failure to satisfy the acceptance standards of IWF-3400 are not limited by Section XI to the type i of indication identified, the type of support being examined, or the types of i adjacent supports. This would require that a documented visual examination be ;
performed which, in some cases, would provide no meaningful results. These .
actions cause undue hardships and increase ALARA concerns without i significantly affecting the overall safety of the plant. ;
- 6. ALTERNATE TEST METHODS: A visual examination is performed on all safety related snubbers, as required by the Braidwood Technical Specifications. When corrective actions are required as a result of these examinations not ;
satisfying the requirements of IWF-3400, the adjacent supports will be ;
examined if they are susceptible to the same type of indication or mode of failure, regardless of support type. ,
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- 7. JUSTIFICATION: Corrective actions are required for indications failing to !
satisfy the acceptance standards of IWF-3400 and the visual examination l acceptance criteria of the Technical Specifications. The acceptance standards :
of IWF-3400 are both generic and specific in nature and therefore, in some !
cases, apply only to a particular type of support. It is impractical to perform an examination on an adjacent support when the indication identified i is not applicable to the type of support. -l.
If the indication requiring corrective measures can be generically applied to l i
the adjacent supports, they will be examined. All safety related snubbers are required to be examined by the Technical Specifications, and include augmented ,
examinations of exempt components on the AF,'CC, CS, CV, FW,.MS, RC, RE, RH, RY, SD, SI, and SX systems. Additionally, since the inspection frequency is ,
increased based upon the number of failures identified, this approach to '
performing additional examinations, along with the additional examination required by IWF-2430 for nonexempt components, provides a high degree of system reliability and safety while providing ALARA benefits to plant employees.
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- 8. APPLICATION: This relief request applies to VT-3 and VT-4 visual examinations 3 performed on component supports. !
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- 9. IMPLEMENTATION: Visual examinations are performed during plant operations or i normal plant outages. ;
- 10. TIME PERIOD: This request for relief applies for the first ten year interval. ,
- 11. APPROVAL STATUS: Relief granted per SER dated 10/04/91.
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Revision 3 0.0 ADOSNTED INSERVICE INSFECTION REQUIREMENTS ,
The following augmented Inservice Inspection requirements are being ,
implemented at Braidwood Nuclear Stations
- a. Augmented ultrasonic examinations will'be performed on all six inch and larger circumferential Mainsteam and Feedwater welds from the ;
containment wall to the torsional restraints in the MSIV rooms. These ;
high energy lines were not postulated for high energy breaks. (ref. Note 5, sec. 4.0)
- b. Augmented ultrasonic and surface examinations will be performed on each ;
Reactor Coolant Pump flywheel once each inspection period per Regulatory !
Guide 1.14. (ref. Note 6, sec. 4.0)
- c. Augmented non-destructive examinations will be performed on the turbine !
discs and rotors per Westinghouse recommendations. (ref. Note 8, sec. ,
4.0)
- d. Augmented ultrasonic examinations will be performed on 7 1/2% sampling > !
4" nominal pipe size on class 2 ECCS systems. (ref. Note 10, sec. 4.0).
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- e. Ultrasonic examinations will be performed on predetermined locations [
within the Reactor Coolant system that may be subject to thermal-stratification per NRC Bulletin 88-08. (ref. Note 14, sec. 4.0)
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Revision 3 -
9.0 EEEMPT CX3GCOEENT LISTING ]
Table 9.0-1, " Exempt Component Listing" contains a summary of the piping, pump, valves and vessels that are exempt from the ISI NDE examination requirements. These tables list those systems / components which have been 7 exempted from examination as specified in ASME Section~II, subarticles IWB-1220, IWC-1220. The Exempt component Listing tables includes the Description '
System, Component, Exemption and Remarks Information.
All components listed herein will receive the required system leakage test as ,
specified in the appropriate table. ;
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Revleion 3-TABLE 9.0-1 comed Craidwood Nuclear Station Unit 1 & 2 CLASS 1 EXEMPT COMP 0tMNT LISTING IESCRIPTION SYSTDI CCMPONENT EXEMPTION REMARKS Piping Chemical and Volume CV lines i 1.0* NPS IWB-1220(b) l Control (CV)
Reactor Coolant (RC) RC lines i 1.0" NPS IWB-1220(b)
Residual Heat Removal RH lines i 1.0" NPS IWB-1220(b)
(RH)
Pressurizer System (RY) RY lines i 1.0" NPS IWB-1220(b)
Pumps There are no exegt class 1 pumps.
Valves . .
Chemical and Volume Valves on CV lines i 1.0* NPS IWB-1220(b) Valves on exempt class 1 piping.
Control (CV)
Reactor Coolant (RC) Valves on RC lines i 1.0" NPS IWB-1220 (b) Valves on exenpt class 1 piping.
Residual Heat Removal Valves on RH lines i 1.0* NPS IWB-1220(b) Valves on exengt class 1 piping.
(RH)
Pressurizer System (RY) Valves on RY lines i 1.0" NPS IWB-1220(b) Valves on exenpt class 1 piping.
I 9-2
-w__w,. - . -.--we.- -- -- r - - - - = e*- -- - e - e--+ - - = ~ - = -*v-'*----e .+v-- - -+ --~+vt-.e-- + -e +- +e 3, - -w* . - - . . - . ~-----.nha= .--mw- .w r-,eww-w=w.- -- -.. -- e e-
s Revision 3 TABLE 9.0-1 ConBd ErmidWood Nuclear Station Unit 1 & 2 CLASS 2 EXEMPT COMPOtWNT LISTING DESCRIPTION SYSTDS COMPONENT EXEMPTION REMARKS Piping Auxiliary Feedwater (AT) All piping IWC-1220(c) All lines i 4.0" diameter Boron Thermal Regeneration All piping IWC-1220(c) All lines i 4.0" diameter (BR)
Corponent Cooling (CC) All piping IWC-1220(b) All lines i 200* F Containment Spray (CS) All piping 'INC-1220tal or (d) All lines i 275 psig, i 200* F and/or i 4.0* .
diameter Chemical and Volume Control See Remarks IWC-1220 (b) or (c) All lines i 275 psig, i 200* F and/or 1 4.0" diameter (CV)
Fuel Pool Cooling (fr) All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or i 4.0*
diameter Fire Protection All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or i 4.0" diameter Main Feedwater (FW) IW lines i 4.0* NPS IWC-1220(c)
Instrument Air (IA) All piping INC-1220(b) or (c)
Main Steam (MS) MS lines i 4.0" NPS IWC-1220(c)
Nitrogen (NT) All piping IWC-1220 (b) or (c) All lines i 275 psig, i 200* F and/or i 4.0" diameter Off Gas (OG) All piping IWC-1220 (b) or (c) All lines 1 275 psig, i 200* F and/or i 4.0" diameter h
9-3
__ . - . _ , _ . . . _ _ _ _ _ . _ _ _ _ _ . . _ . . . _ . . - _ _ . _ . _ . . _ . - _ _ ._ ,_ . _ . - . _ . _ . - ~ . . . _ .- - _ _ _ . ,_ _ . . . _ . . _ . _ _ . . _ . _ _ . . _ _ _ _ _ . _
novision 3.
Tnata 9.0-1 Comed Braidwood Nuclear Station Unit 1 & 2 CIASS 2 EXEMPT CG4PONENT LISTING DESCRIPTION SYSTEM CCMPONEfff EXEMPTION REMARKS Piping (Cont.)
Process Radiation Mocitoring All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or i 4.0*:
diameter (PR)
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Process Sampling (PS) All piping IWC-1220(c) All lines i 4.0" diameter 4
Reactor Coolant (RC) RC lines i 4.0" diameter IWC-1220(c)
Reactor Building Drain (RE) All piping IWC-1220(b) or (c) Ali lines i 275 psig, i 200* F and/or i 4.0" diameter Reactor Building Floor Drain All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or 5 4.0" diameter (RE)
Residual Heat Removal (RH) RH lines i 4.0* diameter *1HC-1220(d)
Pressurizer (RY) RY lines i 4.0" diameter IWC-1220(c)
Service Air (SA) All lines IWC-1220 (b) or (c) All lines i 275 psig, i 200* F and/or i 4.0" dlameter Steam Generator Blowdown (SD) All lines i 4.0" diameter IWC-1220(c)
Station Heating (SH) All lines 'IWC-1220 (b) or (c) All lines 1 275 psig, i 200* F and/or i 4.0" ^
diameter Safety injection (SI) See Remarks *INC-1220 (a) or (d) All lines i 275 psig, i 200* F and/or i 4.0*
diameter Essential Service Water (SX) All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or i 4.0*
diameter Primary Containment Purge All piping IWC-1220(b) or (c) Alllinesi275psig,1l200*Fand/ori4.0"
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System (VQ) diameter 9-4
_ . . . _ - _ _ . - - - _ . _ . _ . _ . . . _ . _ , - . , . . . . _ . . . . - - . _ - , . ~ _ . - , , _ . _ . . . .. . . . . . _ . _ _ . _ . _ . _ . . _ _ _ . . _ . . .._. _. ._ _.
Revision 3-TABLE 9.0-1 Comed Braidwood Nuclear 8tation Unit 1 & 2 CLASS 2 EXEMPT CCHPOSENT LISTING DESCRIPTION SYSTEM CG4PONENT EXDtPTION REMARKS Piping (Cont.)
Make-up Domineralizers (m) All piping IWC-1220 (b) or (c) All lines i 275 psig, i 200* F and/or i 4.0*
diameter l Chilled Water System (WO) All piping IWC-1220(b) or (c) All lines i 275 psig, i 200* F and/or i 4.0*
diameter Vessels Containment Spray (CS) CS01SA IWB-1220 (b) Spray Eductors, 275 psig, 200* F CS01SB IWB-1220(b) Spray Eductors, 275 psig. 200* F C301T IWB-1220(b) Spray Additive Tank,150 psig,120' F Chemical and Volume CV01FA !*8C-1220 (bl . (c) Seal Water Injection Filter, 4.0" dia.,130' F Control (CV) CV01FB IW '?2ato), (c) Seal Water Injection Filter, 4.0" dia., 130* F CV02A IWC-1220(b) Seal Water Heat Exchanger, 85 psig, 160* F Safety Injection (SI) S101T IWC-1220(b) Refueling Water Storage Tank, 50 pelg, 150' F S104TA IWC-1220(a) Accumulator Tanks, These tanks are not required to SIO4TB IWC-1220(a) operate during normal plant operations but remain SIO4TC IWC-1220 (a) flooded at nominal operating pressure (700 psig).
SIO4TD IWC-1220(a)
Pusps and Valves Those punp and valves connected to exempt piping, except schere one side is connected to a nonexempt line.
l 9-5
, . . _ _ _ . - ~ . ~ _ ._, -- - . . _ _ - . - - . ._ . . - - , . _ . _ -. _ _ _ . _ - , _ . _. _.. _ . . _ _ _ . - - _ _ . _ . - , _ _
movision 3 TRSLE 9.0-1 Coned Ermidwood Nuclear Station Unit 1 & 2 CIASS 3 EXEMPT CGEPONENT LISTING DESCRIPTION SYSTD( CG4PONENT EXEMPTION RDEARFS Piping Boric Acid Processing (AB) All AB lines i 4.0* diameter IMD-1220.1 Auxiliary Feedwater (AT) All piping IWD-1220. 2 (b) Integral attachments of suppotts and restraints are examined. All' lines i'175 psig or i 200' r.
Boron Thermal Regeneration All BR lines i 4.0" diameter IWD-1220.1 (BR)
Component Cooling (CC) All piping . IND-1220.1 All lines.1 275 psig, i 200* T and/or i 4.0" diameter IWD-1220. 2 (b)
Chemical and Volume Control All CV lines i 4.0" diameter IWD-1220.1 (CV)
Diesel 011 Transfer (DO) All Do lines i 4.0* diameter IWD-1220. I '
Fuel Pool Cooling (ft) All piping IWD-1220.2(b) All lines i 275 psig, i 200* r and/or i 4.0*
diameter Fire Protection All piping IWD-1220.2(b) All lines i 275 psig, i 200' r Radioactive Waste Gas All lines i 4.0" NPS IWD-1220.1 Reactor Equipment Drains (RE) All RE lines i 4.0* NPS IWD-1220.1 Residual Heat Removal (RH) All RH lines i 4.0" diameter IWD-1220.1 Pressurizer (RY) All RY lines 1 4.0* diameter ' IWD-1220.1 Service Air (SA)- All SA lines i 4.0" diameter IMD-1220.1 (Diesel Gen. Starting Air) l 9-6
Revision 3 TABIa 9.0-1 i
CosEd y Braidwood Nuclear Station Unit 1 & 2 CIAss 3 EXEMPT COMPOEIENT LISTING DESCRIPTION SYSTEM CNPONENT EXEMPTION REMARKS Piping (Cont.)
l Safety Injection (SI) M1 SI lines i 4.0" diameter IWD-1220.1 Essential Service Water (SX) M1 piping IWD-1220. 2 (b) All lines i 275 psig and i 200* F Aux. Building Equip. Drains M1 WE lines i 4.0" diameter IND-1220.1 (WE)
Chilled Water System (NO) M1 piping IWD-1220. 2 (b) M1 lines i 275 psig and i 200* F l
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10.0 REFERENCES
The references used to develop this Inservice Inspection Program Plan lacludes
- a. Code of Federal Regulations, Title 10, Part 50, Paragraph 50.55a, " Code of Standards". --
- b. ASME Boiler and Pressure Vessel Code,Section XI, Division 1, " Inservice Inspection of Nuclear Power Plant Components", the 1983-Edition through and including the Summer 1983 Addenda,
- c. Regulatory Guide 1.26, Revision 3, " Quality Group Classifications and Standards for Water , Steam , and Radioactive Waste- Containment Components of Nuclear Power Plants".
- d. Code of Federal Regulations, Title 10, Part 50, Paragraph 2, " Definitions",
the definition of " Reactor Coolant Pressure Boundary".
- e. Final Safety Analysis Report, Sections: ,
5.3.3.7, Inservice Inspection of the Reactor Vessel.
5.2.4, Inservice Inspection of RCS Pressure Boundary & Supports.
5.4.2.2, Inservice Inspection of Steam Generators.
6.6, Inservice Inspection requirements for ASME Class 2 & 3 Components.
- f. Technical Specifications, Sections:
3/4.4.5, Inservice Inspection Requirements for Steam Generator Tubing.
3/4.7.8, Inservice Inspection Requirements for Snubbers.
3/4.4.10, Inservice Inspection Requirements for ASME Class 1, 2, and 3 j Components. j l
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