ML20127N432
ML20127N432 | |
Person / Time | |
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Site: | Big Rock Point File:Consumers Energy icon.png |
Issue date: | 01/11/1993 |
From: | CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
To: | |
Shared Package | |
ML20127L241 | List: |
References | |
PROC-930111, NUDOCS 9301290198 | |
Download: ML20127N432 (384) | |
Text
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'l CONSUMERS POWER COMPANY l III,i ._ 212' West Michigan Avenue -{
W T Jackson, Michigan 49201 '
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l Third Interval .l Inservice Inspection Program )
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- BIG ROCK POINT NUCLEAR PTiANT-10269 US-31 North Charlevoix, Michigan 49720 COMMERCIAL SERVICE DATE: 12/08/62 Authorig_ed e Inspection Acency Factory Mutual Engineering District office Bingham Office Pk., Suite 141 30150 Telegraph Road Bingham, Michigan _48025 l
9301290198 930111 PDR ADOCM 05000155 G PM
Dig Rock Point Nuclear Plant Third Interval Inservice Inspootion Program Reviews and Approvals Reviews Bohwg --
l24- fr- a it 13-9L-Initiator Date f bate JdW" Engineering Supt.
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Quality Review Form No. \%2.-92, hporoval
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e- g TABLE QE GONTENTE
. im Section 1- Introduction A. Historical Background B. Upgrading criteria:-
C. References D. General E. Inspection Intervals F. Discontinuity Dispositioning G. Reporting Section 2 Outline of the Third 10 year Program A. Vessels B. Piping C. Extent of the Program Section 3 Basis Statements A. Exemptions B. Exclusions C. Deferrals D. Augmentation E. Category B-M-2 F. Additional Examinations G. System Hydrostatic Tests (IWA-5211(d))
H. System Pressure Tests (IWB)
I. System Pressure Tests (IWC)
J. System Pressure Tests (IND)
K. Repairs and Replacements L. Containment Penetrations
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Section 4 Technical Specifications and Code Use Section 5 Relief Requests Section 6 Verification of Section XI Compliance A. Introduction-B. Determination of Compliance C. Number of Components D. Interval Compliance E. Midinterval Requirement Changes F.' Verification of Compliance Table (s)
Section 7 Ultrasonic Calibration Block Listing S3ction 8 Piping and Instrument Diagrams Section 9 Class 1 and 2 Isometrics Section 10 Clau. 1 and 2 Master Plan Exam Table O
I iO SECTION 1 INTRODUCTION O
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- 1. INTRODUCTION
.This document is a summary of the' plan-for Inservice Inspection (ISI) to be performed over the third 10-year interval on Class 1, 2 and class.3 components and systems-(and their supports) of Consumers Power Company's(CPC) Big Rock Point Nuclear Power Plant:
Unit 1.
A. HISTORICAL BACKGROUND j The initiation of Commercial Service for Big Rock Point.was on December 8, 1962. The start of the first Inservice Inspection 10- !
year Interval was on January 1, 1972 and was invoked to comply I with Section 9.0 of the Technical Specifications of the Operating.
License DPR-6 for the Big Rock Point Nuclear Plant.
For the first 80 months of the interval, Big Rock Point utilized a 10-year ISI Plan developed by Southwest Research Institute, requiring surveillance of-Class 1 components. The selection of areas and methods of exanination for the class 1 components was in accordance with the 1971 ASME B&PV Code,Section XI, with Addenda through Summer 1971, except for the Reactor Depressurization System (RDS) which was in accordance with the 1974 Edition of the ASME B&PV Code,Section XI.
In February 1976, the NRC amended Paragraph 55a(g) of 10 CFR 50 to require nuclear plants to upgrade their Technical 7-~
- Specifications in the areas of the ISI requirements and the functional testing of pumps and valves. By amending Paragraph 55a(g) and by invoking Regulatory Guide 1.26, the NRC required nuclear plants to upgrade their systems to include not only Class 1 systems, but also Class 2 and Class 3 systems in their ISI-programs.
B. UPGRADING CRITERIA The construction of this plan was based on the following documents:
- 1. Big Rock Point Nuclear plant's Piping and Instrument Diagrams and Plant Q-List.
- 2.Section XI of the ASME Boiler and Pressure Vessel Code, " Rules for Inservice Inspection of Nuclear Power Plant Components,"
1986 Edition (no addenda), for ASME Class 1, 2, 3 Systems.
- 4. USNRC " Rules and Regulation- i 10, Chapter 1,- Code of Federal Regulations -'Energ; 50.55a.
Section 9.0 of the Technical Specificationr of the Operating O
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License DPR-6 for the Big Rock Point Plant.
-s components were scheduled'for examination in accordance with.the above stated rules and regulations. Examinations are conducted in
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\ . accordance with.the ASME Boiler and Pressure Vessel-Code.
C. REFERENSES
- 2. Operating License (DPR-6), Technical Specifications-for.the Big Rock Point Plant, Docket No 50-15), Section-9.0, Appendix A.
- 3. ASME Boiler and Pressure Vessel Code,Section XI, 1986 Edition (no addenda).
- 4. Consumers Power Nuclear Operations Dec ._tment Standards NODS-QO3, " Inspection".
- 5. Consumers Power Nuclear Operations Department Standard NODS-P02, " Test Control".
D. GENERAL
- 1. This Inservice Inspection plan for the four 10-year Inservice Intervals (Reference Section I Sub. E)1 has been developed, reviewed and approved by Consumers Power Company .
/~' for use at the Big Rock Point Nuclear Power Plant Unit 1.
( This plan incorporates all applicable relief requests and periodic surveillance requirements of Reference C-2 and C-3 for the 40 year service lifetime.
The start of the first 10-year interval was January 1, 1972. The inspection intervals set in this plan coincide with Inspection Program B as defined in Paragraph IWA-2430 of the Section XI code.
- 2. Responsibility for the maintenance of this plan and the development of subsequent plans rests with the Big Rock Point Engineering Department, Inservice Inspection Section.
- 3. In view of the fact that-the design of the Big Rock Point Nuclear Power Plant was completed prior to the issuance-of'Section XI of the ASME B&PV Code,-the inspection access requirements of IS-142,1971 Edition, were not 'available tx) impact the plant design parameters.. The Technical Specification / Relief Request Sections (4 and 5) of this.
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plan detail specific Code requirements which cannot be met..
- 4. Examination methods performed are intended to be representative of past ISI practice or of preservice .
methods utilized. In either case, it should be recognized that either UT or RT are acceptable volumetric exams and Unique weld O either PT or MT are acceptable surface exams.
joint parameters may, of course; dictate a more restrictive i.
selection criteria; og, high background radiation will j} preclude RT, stainless steel will preclude MT, etc. It is
\_/ intended that the process which selects exam nothod for examinations _under this plan treat UT and RT as interchangeable and PT and MT as interchangeable with consideration given to past practice in light of the reproducibility or results.
- 5. This submittal for the third inspection interval covers Class 1,2 and 3 components and systems.only. The pump and valve program for the third inspection interval is a seperate submittal.
E. INSPECTION INTERVALS The following table delineates the inspection inte::vals for the Big Rock Point Plant:
1st Interval From 01/01/72 to 12/31/81 1st Period 01/01/72 to 4/30/75 2nd Period 05/01/75 to 8/31/79 3rd Period 09/01/79 to 3/30/84 2nd Interval From 01/01/82 to 12/31/91 1st Period 01/01/82 to 11/30/85
( 2nd Period 3rd Period 12/01/85 to 08/06/89 08/07/89 to 12/31/91 3rd Interval From 01/01/92 to 12/31/2001 1st Period 01/01/92 to 03/31/95 2nd Period -04/01/95 to 07/31/98 3rd Period 08/01/98 to 12/31/2001 In compliance with ASME XI IWA-2400 (c) 1977 Edition Summer 1978 Addenda, the inspection interval may be extended for power units that are out of service continuously, for six months or more. The interval for which the outage occurred may be extended for a period equivalent to the outage duration.
Consumers Power Company has utilized the IWA-2400 (c) extension twice in the first interval; once, during the reactor depressurization system preservice inspection in 1976 (ISI No 4) for 177 days, and in 1979 (ISI No 6) for 277 days while repairs were made to the F-2 control rod drive J-welds.
This extension allowed for first interval exam requirements to be completed during the 1983 Refueling Outage. The start of the second interval remained unaffected by these extensions per letter form the NRC dated June 10, 1983.
I F. DISCONTINUITY D,ISPOSITIONING Discontinuities will be evaluated using the techniques outlined in Section XI 1986 Edition. Disposition of discontinuities will be in accordance with standard company procedures (ie, QA-16, Etc).
G. REPORTING Reports concerning ISI activities will be filed in accordance with IWA-6000,Section XI.
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SECTION 2 OUTLINE OF THE TIIIRD
- l. lo-YEAR PROGRAM I
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C 2. OUTLINE QI TRE TJ11RD J3-YI,AB PROGRAM
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A. Vessels and Pumos The examination areas in the reactor pressure vessel and closure head, emergency condenser, cican-up heat exchangers, steam drum, clean-up domineralizer tank, and pumps are identified by name.
B. Pinina
- a. The first character set in each code consists of a number designation which represents the nominal diameter of the pipe under consideration.
- b. The second character set in each code consists of '
alphabetic codes which designate the main piping system under consideration. Below is a table showing the codes-and their respective systems.
ECS - Emergency Condenser System LPS - Liquid Poison System SCS - Shutdown Cooling System RCS - Reactor Clean-up System MRS - Main Recirculation System CSS - Core Spray System RDC - Redundant Core Spray System
(~N CRD - Control Rod Drive System
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MSS FWS Main Steam System Feedwater System RDS - Reactor Depressurization System PIS - Post Incident Cooling System RCW - Reactor Cooling Water System SWS - Service Water System
- c. The third character set consists of a three-digit line number. The first digit indicates the safety class, while the last two digits indicate the line sequence number l
l within the system.
- d. The fourth character set gives the specific weld number.
! Circumferential welds are numbered sequentially in the 6irection of flow from the upstream end of the line (ie, 20-MRS-121-17 identifies the seventeenth circumferential weld in the 20 inch Main Recirculation-System, Class 1, line sequence Number 121) .
- c. A fifth character set (an identifying letter code following the basic weld number) is used for components other than-circumferential welds. If more than one component of one type appears at one location, a sixth character. set is added. This consists of a sequential number, assigned clockwise (looking downstream) from the zero reference (Lo) assigned to the circumferential weld, or in sequence in the direction of flow. Two or more circumferential rows of hanger lugs would be sequenced both circumferentially and
in the the firstdirection of flow hanger lug in that order (ie, HL-5 would be in khe second circumferential row c c.
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(_s) four per row) .
Longitudinal welds are identified by LU or LD as u,otream or downstream, respectively of the intersected circumferential weld. If a pipe has two longitudinal welds, they are numbered as above, clockwise from Lo. (ie, 17-LU-2 is a longitudinal wold upstream of weld 17, the second weld clockwise from Lo).
On an elbow, LU-l would be on the insido radius of the elbow and LU-2 would be on the outside radius.
- f. For branch connections larger than one inch, the following notation is used. The notation for these branch connections is in three parts.
(20-MRS-121-17) (1) (6-SCS-102)
(*) (1) (O)
(*) contains the system, loop and weld number upstream (1) Indicates the branch connection (0) Contains the branch line identifier
- g. Some examples of_ weld numbering are as follows:
6-SCS-101-6 The sixth circumferential weld from the designated system boundary in direction
( of flow in the 6 inch, Class 1 shutdown system, line sequence Number 101.
?Q-MRS-121-17/ The 6 inch shutdown system branch 6-SCS-102 connection weld into the 20 inch main recirculation system downstream from circumferential weld 17.
17-MRS-111-7-LD-2 The second longitudinal weld clockwise in the direction of flow from Lo, downstream from circumferential weld 7 in the main recirculation system, line sequence Number 111, 6-ECS-101-2PL Welded pipe lug downstream from weld 2 in the Emergency Condenser System, line sequence number 101 .
- c. Extent Of The Procram This submittal covers the third lo-year inspection interval. The Big Rock Point Plant utilizes Program B as set forth in IWA-2430,Section XI.
Selected' portions of the major components and/or systems to be examined in accordance with Section XI are as follows:
- 1. Class I
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A.-Reactor Pressure Vessel- !
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- B. Reactor Pressure Vessel. Closure Head C.-Emergency Condenser.
D. Clean-up Heat Exchangers 1 E. Steam Drum !
F. Clean-up Domineralizer. Tank G. Piping
- 1. Emergency Condenser System - ECS
- 2. Liquid Poison-System ~ LPS
- 3. Shutdown Cooling System - SCS !
- 4. Reactor Clean-up System - RCS
- 5. Main Steam System - MSS
- 6. Core Spray System - CSS- q '
- 7. Redundant Core Spray System - RDC
- 8. Control Rod Drive System - CRD
- 9. Feedwater System - FWS ,
- 10. Reactor Depressurization System - RDS
- 11. Main Recirculation System - MRS H. Pumps I. Valves =
- 2. Class II A. Piping i,
- 1. Control Rod Drive - CRD !
- 2. Core Spray System - CSS I
- 3. Feedwater System - FWS
- 4. Main Steam System - MSS
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-5. Post Incident Cooling System - PIS
- 6. Shutdown Cooling System - SCS B. Valves
- 3. Class III A. Piping
- 1. Reactor Cooling Water - RCW
- 2. Service Water System - SWS O
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O SECTION 3 BASIS STATEMENTS l
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( 3. BAHJJ STATEMENTS The following sections delineate the basis used by Consumers Power company for exemptions, exclusions, or deferral'of examinations or other modifications of the requirements of Section XI. All exemptions, exclusjons, deferrals or modifications are specifically allowed by Section XI, or relief has been granted by the regulatory authority, or the Regulatory Authority has superseded section XI by augmenting a more conservative program than that required by Section XI.
A. EXEMPTIONS All Class 1, 2 and 3 pressure retaining components-Cand their supports) are subject to examination. However,Section XI provides rules for exempting components from examination (ie, IWB-1220, IWB-2500, IWC-1220,IWC-2500, IWD-1220, IWD-2500) and Federal law allows the Regulatory Authority (NRC) to grant relief from specific portions of the code upon demonstrated-need.
Some examples of exemptions are as follows:
- 1. All Class 2 lines and nozzles less than or equal to 4" nominal pipe size are exempt by IWC-1221(a) and IWC-1222(a).
- 2. Class 1 piping less than or equal to 1" nominal diameter is exempt per IWD-1220(b) (1) .
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B. Exclusiong The following are excluded from examination per the ASME B&PV Code Section XI 1986 Edition, Table (s) IWB-2500-1 and IWC-2500-1:
- 1. Examination Category S-H, Integral Attachments for Vessels (Examine 1st and 2nd interval-only).
- 2. Examination Category B-K-1, Integral Attachments for piping, pumps and valves (Examine 1st and 2nd interval only).
- 3. Examination Category C-F-2, Pressure Retaining Welds in Carbon-or Low Alloy Steel less than 3/8" nominal wall thickness for piping greater then NPS 4".
C. DEFERRALS Section XI of the code provides a degree of latitude in the scheduling of major component examination in that certain category examinations may be deferred to the end of the inspection interval. Examples of major component examinations are mechanized UT of the reactor vessel shell welds and core support structure examinations.
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/T Tj- components Deferrable 12 End-21 Interval Section XI Category Component Comments B-A RPV Shell Welds B-A RPV Head Welds Bottom Head Only B-L-1 Pump Casing Wolds Relief Requested B-L-2 Pump Casing Relief Roquested B-M-1 Valve Body Wolds B-M-2 Valvo Bodies B-N-2 Core Support structures B-O Welds in CRD Housing D. AUGMENTATION Plant Technical Specifications may at times require more frequent examination scheduling than doesSection XI as is the-case with snubber testing.Section XI requirements are-superseded by Technical Specifications-and these examinations are not subject to Paragraphs IWB-2400, Section-XI.
The following Big Rock _ Point Technical Specifications sections-apply to the Augmented Program and will be performed throughout the life of the plant.
/"'s Other augmented examinations are determined by Big Rock Point (j
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ISI which include the Control Rod Drive F-2 J-weld, Steam Drum Relief Valves and IGSCC susceptible welds which are also identified under miscellaneous surveilJance items.
MISCELLANEOUR SURVEILLANE ITFMS Equipment Method Frequency
- 1. Control Rod Drive F-2 Volumetric Once per period J-weld
- 2. Steam Drum Relief Surface One -(1) - cach valves refueling
- 3. Snubbers Visual & -Two (2) per outage Functional
- 4. IGSCC Susceptible Ultrasonics Per IGSCC welds scheduling table P. CATEGORY R-H-1 Section XI' Allows a reduction in the number of Category-B-M-2' examinations in that only one valve in each group of' valves needs to be examined per interval. Valve grouping is determined by valve size, function, manufacturing method and
(~'g manufacturer. Identified below are the Big Rock Point Valve V
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groups.
Class'1 Valves Exceeding-(O_) 4" Nominal-Pipe Size Valve Size / Type Valve Number System Isometric 5" Chapman Gate MO-N002A 5-MRS-131 *- A-68 Valves MO-N002B 5-MRS-132
- A-69 -l 6" Anchor Darling CV-4180 6-RDS-102 A-83 l Gate Valves CV-4181 6-RDS-103 .A-83 l CV-4182 6-RDS-104 A-83 1 CV-4183 6-RDS-105 A-83 i 6" Powell Gate MO-7052 6-ECS-102 A-21 Valves MO-7062 6-ECS-101 A-20 6" Target Rock SV-4984 6-RDS-102 A-83 Depressuring Valves SV-4985 6-RDS-103 A-83 SV-4986 6-RDS-104 A SV-4987 6-RDS-105 A-83 30-FWS-101
- A-81 10" Edwards Globe VFW-9 Valve 10" Anchor Darling VFW-305 10-FWS-101 B Check Valve 10-FWS-101
- A-81 10" Edwards Check VFW-304 1 Valve 8" Powell Gate MO-7056 8-SCS-101 A-29 Valves MO-7057 8-SCS-101 A-29 MO-7058 8-SCS-102 A-30 MO-7059 8-SCS-102 A-30 12" Powell Gate MO-7050 12-MSS-105 A-47 Valve 20" Chapman Gate MO-N001A 20-MRS-121
- A-66 Valves MO-N001B 20-MRS-122
- A-67 20" Allis-Chalmers MO-N006A 20-MRS-121
- A-66 Butterfly Valves MO-N006D 20-MRS-122
- A-67 24" Chapman Gate MO-N003A 24-HRS-121 * 'A-66 Valves MO-N003B 24-MRS-122
- A NOTE: (*) Reference Relief Request Section (RR-A13)
F. ADDITIONAL EXAMINATICNS Examinations performed during any inspection that reveal indications exceeding the allowable indication-standards of Section XI, IWB-3000 shall be subject to Big Rock Point ISI Evaluation per Section XI IWB-IWC-IWD-IWF-2400 Subsection 2430.
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Consumers Power Company - Big Rock Point Position Due to the fact that the Big Rock Point Nuclear' plant was-constructed prior to the issuance of-Section XI and Preservice Examination requirements, additional surface examinations will be performed as follows.
Surface examinations performed during any inspection that '
reveal indications exceeding the-allowable indication standards of Section XI, but are interpreted as fabrication defects-(ie:not service induced) will not be considered as requiring additional examinations in accordance-with Section XI, Subsection 2430 requirements.
G. SYSTEM HYDROSTATIC TESIji,(IW?.-5211(d))
The hydrostatic testing requirements of IWB-2500, IWC-2500 and IWD-2500 are required to be performed each Inspection Interval. Class 1 system hydrostatic testing of the Reactor Coolant System is performed each refueling outage. Class 2 and 3 system hydrostatic tests will be performed at or near-the end of each inspection interval.
H. SYSTEM PRESSURE TESTS JIWBL As required by Tables IWA-5210-1 and IWB-2500-1, the system pressure tests described in IWA-5211 are conducted as follows:
CLASS I
After each refueling outage, the Class 1 pressure boundary is at 1.1 subjected to aoperating times normal visual exam (VT-2)The pressure. withpressure the system (s)ing retain boundary corresponds to the Reactor Coolant System boundary with all Class 1 valves in the normal position required for normal reactor startup. Also, the visual exam (VT-2) extands to and includes the second closed valve at the-class 1 pressure boundary extremity.
I. SYSTE7, PRESSURE TESTS (IWCL As required by Tables IWA-5210-1 and IWC-2500-1, the system pressure tests described in IWA-5211 are conducted as follows.
CLASS 11 During each Inspection Period, the Clacs 2 pressure boundary is subjected to a visual exam (VT-2) with the system (s in service under operating pressure. The pressure retainin)g boundary extends up to and includes the first mally closed valve or valve capable of automatic closurs tc; serform the safety-related system function.
System EyratioJ1l a Test (IWA-5211(b))
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During each Inspection Period, the Class 2 pressure boundary
(^'\ of those systems or nortions of systems-not required to
- \j operate during normal reactor operation are subjected to a while at the pressure obtained during-the visual system exam (VT-2)t functional tests. The pressure retaining or componen boundary extends up to and includes the first normally closed valve or valve capable of' automatic closure when the safety function is required.
J. SYSTEM EBESSURE IESTE (IWD)_
As required by Tables IWA-5210-1 and IWD-2500-1, the system pressure tests described in IWA-5211 are conducted as follows.
JMSERVICE TEST (IWA-5211(c))
During each Inspection Period, the Class 3 pressure boundary is subjected to a visual exam (VT-2) with the system (s) in service under operating pressure. The prm sure retaining bouncary extends up to and includes the first normally closed valve or valve capable of Automatic closure to perform the safety-related system function.
SYSTEM FUNCTION % TIST U NA-5211(bi)
During each Inspection Period, the Class 3 pressure boundary
/~ of those systems or portions of systems not required to
'( ,T) operate during normal reactor operation is subjected to a visual exam (VT-2) while at the pressure obtained during the system or component functional tests. The pressure retaining boundary extends up to and includes the first normally closed valve or valve capable of automatic closure when the safety function is required.
( K. REPAIRS AND REPLACEMENTS Repairs and replacements at Big Rock Point Nuclear Plant are performed in accordance with the ASME Section XI Kepair and Replacement Program. As required by IWA-4000 and'IWA-7000, this program delineates the essential ~ requirements of the complete repair cycle including weld repairs, procurement and installation of replacements.
The program consists of administrative procedures which describe overall departmental responsibilities and interfaces, the Authorized Nuclear Inspector's involvement and documentation requirements. Also, Maintenanca and Quality Assurance departmental procedures implement ccntrols for' special processes essential to the repair program such as flaw l removal, weld repair, post wold heat treatment and non-destructive examination.
The Repair and Replacement Program complies with the requirements of IWB, IWC, IWD and IWF 4000 and 7000 of ASME Section XI.
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A L. CQ11TAINMENT PENETRATIONS
Reference:
- 1) Big Rock Point Plant Technical Specifications, Section 3, Part 3.7(c)
- 2) Big Rock Point Procedure TV-01 Each reactor shutdown for refueling, but in no case at intervals greater than two years, the following shall be visually examined for evidence of corrosion, cracking or deterioration:
A. All electrical and accessible piping penetration nipple welds.
B. All accessible piping wolds to nipples.
C. All expansion joints and welds in expansion joints.
D. Potting compound in all electrical penetrations.
Insulation at piping penetration welds shall be ::emoved to permit visual examination.
The probable cause of any significant corrosion, cracking or deterioration revealed by visual examination shall be determined, and evaluated in terms of likolihood or recurrence and probable offect upon other containment sphere penetratica components. An individual component leak detection test shall be performed with air at a pressure not less than 11.5 psig.
Ox on the faulty component prior to its repair or modification.
The faulty component , and.other components if necessary, shall be repaired or modified and an individual component leak detection test performed with air at a pressure not less than 11.5 prig upon each repaired or modified component. All components so repaired or modified shall be visually re-examined at appropriate intervals, but not less frequently than'once every six months, until the adequacy of annual visual inspection is reestablished to the operators satisfaction.
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I BECTION 4 TECHNICAL SPECIFICATIONS AND CODE USE I
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- 4. TECHNIC S BPECIFICATIONS-BMD 29DR PAj:
The following;pages are sections extracted form Amendment No.108- -
to operating License DPR-6, Technical Specifications,_which_
pertain to Inservice Inspection. dated March-.16,:1990. -
The. Technical: Specifications are included-herei.n--solely for reference _ purposes as these Specifications form the basistfor. *
'this master plan.
Federal law requires-the use of a composite-of_various editions / addenda:of Section XI, ASME Code.-In-order ~to define-the?
code use at Big Rock Point with respect to the detailed application of the code, a section defining the manner iniwhich
- Federal law was appliad which'also delineates various other .
elections and-interpretations of the code which were also; applied!
to this. plan.
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N OPERATING LICENSE DPR-6 TECHNICAL SPECIFICATIONS FOR'THE BIG ROCK POINT PLANT CONSUMERS POWER COMPANY DOCKET NO S0-155 0
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i 9.0 INSERVICE' INSPECTION AND TESTING
,o - 9.1 APPLICABilH1
+ 1 Applies to inservice inspection and testing of the reactor vessel and other ASME Code Class 1, Class 2 and Class 3 system-components.
9.2 OBJECTIV1 To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.
9.3 SPECIFICATIOR
- a. Inservice Inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i), and where provisions of Sections 11.4.1.4, 4.1.5 and 11,4.3.4 take precedence,
- b. Sufficient records of each inspection shall be kept to allow comparison and evaluation of future tests. (See also Sections 6.9.3 and 6.10.2.g.)
- c. The inservice inspection program shall be reevaluated as required by 10 CFR 50, Section 50.55a(g)(5) to consider incorporation of new inspection techniques that have been proven practical, and the conclusions of the evaluation shall be used as appropriate to update the inspection program.
- d. A surveillance program to moniter radiation induced changes in the mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.1.l(h) of these Technical Specifications,
- e. The Inservice Inspection Program for piping identified in NRC /
Generic letter 88-01 shall be performed in accordance with the /
Big Rock Point IGSCC Program approved by the NRC. /
9.4 BAS _lS The inspection program implementsSection XI of the ASME Boiler and Pressure Vessel Ccde to the maximum extent practical. It is recognized that plant design and construction were completed approximately seven years prior to the development of Section XI and it is, therefore, not possible to comply fully with the code.
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Amendment No.M7,108 !
October 19, 1992
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BIG ROCK POINT PLANT 1 CODE USAGE 1
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APPLICABLE EDITION 9E, ADjig RQ1hER k. EBTaHERE VESSEL RQDE SECTIQ}i 31 Purituant to Paragraph 50.55a(g) of 10 CFR Part 50, the inservice exataination recuirements applicable to non-destructive ext.mination and system pressureare testing at the based Consumers upon the rulesPower set forth Co.npany, Big Rock Point Plant, in the 1986 Edition of Section XI of the ASME Boiler and Pressure Vessel Code (no addenda).
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' CONSUMERS POWEB COMPANY.
INSERVICE JESPECTION SQDE CASE HILLIZATIQH Consumers Power Company elects per 10 CFR 50.55a(a)(2)(ii) and ]
Footnote 6 to 10 CFR 50.55a to utilize the following ASME Code !
Cases as an integral part of the Inservice Inspection Program for the Big Rock Point Nuclear Power Plant. 1 Code case N-416 Alternato Rules for Testing of Repair or Replacement of Class 2 Piping Code case N-445 Use of latter edition of SNT-TC-1A for Qualification of Nondestructive Examination Personnel Code Case N-446 Recertification of Visual Exam Personnel Code Case N-457 Qualification Specimen Notch Location for Ultrasonic Examination of Bolts and Studs Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds Code Case N-461 Altcrnative Rules for Piping Calibration Block Thickness Additional code cases may be used without specific NRC approval as long as they are referenced in Regulatory Guide 1988 1.147Rules (seeand Federal Register, Vol 53, No 87, Thursday, May 5, Regulations). All others must be formally submitted to the NRC for review and approval.
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SECTION 5 RELIEF REQUESTS .9
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- 5. BELIXE REQVIATA This coction pots forth the requests for relief submitted to the Nuclear Reguletory Commission por 10 CPR 50.55a.Those 1985 roller requests pursuant woro previously to December approvedfor 19, 1983 requests November the secon1,d interval.
The following tablo delineates the areas which cannot bo implomonted at the Big Rock Point Nuclear Plant.
The notes following the table provido information concerning the basis for each relief request and concerning tho implomontations of the Reference Codo.
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Fase 1 of 2 Felief Re w q Efg Rock Point Nucles Plant Relief Area Section Request of II Rel'ef Requested Testing substituted Inteese ten Felief Category Dmetion C e ant Ntsubar Hydrostatic Test 40 yr Flan Pressure Core Region Circumferential Inaccessible for Voltssetric KR- Al 31.11 B-A Boundary and longitudinal Shell Welds Examination B1.12 lortgitudinal Above Bettline Circumferential WId Hydrostatic Test 40 yr Plan Pressure Bottom Fead Inaccessible for Volteetric RR-A2 Bl.22 B-A Meridional E Ids hamination Boundary Inaccessible fer $urface visual hastinaticm 40 yr Plan RR-A3 B3.90 5-D Pressure 70-in. Rectre Pressure T*'.i g Boundary Nozzles: 796-1A and/or Voluoctric h erina-B3.100 796-1B tice B5.10 B-F 8-in. Shutdewn Unioading Nozzles: 795-15 Remote Visual 40 yr Plan Pressure Welds: 795-1A, 795-1B, Insecessible for Surface RR-A4 B3.90 B-D Hydrostatic Test Boundary 795-1C, 795-1D, and volumetric Examination B3.100 B-F 795-1E, 796-6 B5.?O Pachanized UT 100% of 40 yr Flan 14-in. Steam Cutlet Nczzles Inaccessible for Surface RR-A5 B5.10 B-F Pressure wid voluw 795-11A to 11T hasination Boundery Vis.zal h asination i
Hydrostatic Testics All welds located within Examim upper portions 40 yr Flan 32.51 B-B Pressure Steam Drum and Hydrostatic Tes_s RR-A6 a 150-degree are either KR-A7 B2.70 Boundary side of the bottee Iceg-33.150 B-D itudinal centerline B3.160 MIO591-0077A-BT01 i- _ _ _ _
Page 2 of 2 Relief R* quest Big h k Point Nuclear Plant i
Relief Area Section Request of 11
- Number Relief Category nsnction Ceepenant Relief Requested Testing Substituted Implementation RR-A8 B3.160 B-D Pressure Clean-up regenerative Volumetric of Nozzle inside None 40 yr Plan Boundary heat exchanger (s) radius sections cannot be j examined due to geometry.
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RR-A09 B3.160 B-D Pressure Clean-up Demin Tank Nozzles volue-tric Examinatim of Remainder of Category 60 yr Plan Boundary nozzle inner radius B-D Testing and f sections Hydrostatic Testing j' 3R-A10 B5.130' B-F Pressure 14 eld 7 on line 6-SCS-101 Inaccessible for surface Hydrostatic Test 40 yr Plan l Boundary and volueetric Exam KR-J11 B12.10. B-L-1 Pressure Clean-up and Main Recirculation Volunetric Examination Surface Examination 40 yr Flan Boundary Peeps and Pydrostatic Tests FR-A12 B12.20 B-L-2 Pressure . Main Recirculation Ptmap(s) Visual Examination of Visual Examination only 40 yr Plan Foundary internal surfaces if rump is disassembled i-l
, RR-A13 1* 1 50 5-M-2 Pressure Vaive internal surfaces visual Examination of Visual Examination 40 yr Plan Boundary internal surfaces on 6 out only if maintenance is of 17 grcups or valves required i
- RR-A14 F3.50 ' F-C Coeponent Piping Direct visual (VI-3) for Resete Visual (VI-3) 40 yr Flan Supports. conditices that could affect operability or functional adequacy I
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y CLASS 1 COMPONENTS A. Rosctor Vossol
- 1. Requenta Ler En119L. BB-h11. Shell Heldal Cirgmatcrantial and Lon9119dinal in Csrn Ecolon and longitmlinni @nyn HRlillnn circumfersntial EeIL CatcG9IY D-L Items Di.11 And ILI.d2 C9 sin Rcanirament One circumferential and one longitudinal wold in the boltlino region are to be volumotrically examined over essentially 100% of their longths each inspection interval following the first interval of operation.
Cada Polief EnqusEt Rollof is requested from the codo requirement to volumetrically examino the sho11 volds in the core region and the portions of the longths of longitudinal wolds above the bolt 11no circumferential wold that are inaccessible.
Pronosed alternativo Examinati2D Each refueling outage, a hydrostatic test (prostartup hydro) is performed at 1.1 timos the oporating pressure.
Dasis 19I Reauesting Rollof The reactor vessel is closely surrounded by concreto, so examination from the outsido is not possible. The innor wall of the reactor Vessel is inaccessible in-the coro region due to the prosence of a thermal shield, that is immovable. Design clearance betwoon the thermal shield and reactor vosso1 vall is 1.65 inches. The thermal shield extends oo.50 inches below the bolt 11no circumferential wold.
The surrounding concreto shield and plant design mako external access to longitudinal and circumforential reactor vessel wolds physically impossible.
The concreto shield surrounds Cactually onclosos) the reactor piping and varios in thackness from 6 to 10 foot where piping penetrations are located. Removal of this concreto shield would be an enormous task in a high radiation area (500 to 1500 mr/hr). The concreto shield and exterior insulation on the reactor vossol would not permit surface or remoto visual examination of longitudinal and circumferential wolds from outsido the vossol.
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b The non-romovable thermal shiold and stainions cladding O insido the vosuol fully cover the lower longitudinal wolds and circumferential wolds and thereforo cannot be seen via visual examination from insido the vossol. It is expected that about 30 to 50% of the longitudinal wolds above the boltiino circumforential wolds (#749-1A and
- 749-1B examination. on ISO-A1) will be Those wolds accessible have for volumetrio about 60% of their longth obscured by internal compononts (coro spray sparger, baffle plates, thormal shio2d top ring, and thormal shield).
The prostartup hydrostatic test is more conservativo than the prestartup leak test at operating pressure required by the Code (IWB-5221) . Inspection of the six riser nozzlos and vescol-to-flango wold will provido some indication of reactor vessel integrity, along with results of vessel coupon testing. During operation, a failure of any wold in the reactor vencel would be readily detectable by level indication, dow cell indication,in indication adequate time for safe shutdown. makeup water flow and/or temperatu IllACMM12D The Big Rock Point reactor vessel was designed and constructed to the rul9s of ASME Codo Sections I and VIII, Nuclear Codo casos 1270N, 1271N, and 1273N, and General Electric Company Specifications. The materials O used in construction woro SA-302, Grado B plato, and SA-336 forging material. In terms of Charpy impact energy, those materials are essentially equrialent to the SA-533, Grado B Class 1, and SA-508, closn 2, which are currently being used. Additionally, the primary stresses in the boltline region of the vossol are about i0t of those allowed by Section III of the Codo.
Imposition of the Codo requiremontn would nccessitato removing portions of the concreto biological shield and the permanently installed insulation to perform the required examination of the wolds listed from the vossol exterior. The vessel internals precludo volumotric examination of the bolt 11no wold volume from the vessel interior.
Adhering to all Category B-A Codo requirements for these wolds is impractical duo to existing plant design and geometry. The volumetric examinations that will be performed on the upper longitudinal wolds will give some knowledge of vessel integrity. The recommendations of the SEP report should be followed. And where practical, the generic safety items described should be addressed.
These recommendations includo insuring (a) that the vessel wolds are examined to the maximum extent practical O
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J and most(b) that the most up-to-date experienced examiners are examination methods used to perform the and examinations. Also, visual examination for gross leakage should be performed during each system pressure test.
Reference CE Drawings E-201-801-5, E-201-794-8 and E-201-794 GE Drawing 141F797 Dechtel Drawing 0740G20128 O
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i } 2. Renucat for Relief ER-A22 Ecnael RacG9ry Dotton h item Heat.
RLl2 ligridIonal Eeldh CQd2 Etquirement one meridional vold is to be volumotrically examined onco during each inspection interval. For each wold, the extent of examination includes the accessiblo portion up to 100% of wold length.
CQ!1c Relief Renneni Relief is requested from Codo requirements to volumetrically examine the meridional wolds in the reactor vossol bottom head.
Er_QP2ntd Al_tCIM1112 EKaM1Dat19A A hydro test will be performed as identified in RR-A1.
DM.in LQI Recuesting Religf No access or penetcations are available to afford access to the mor2dional wolds (793-2A, B, C, D) in the reactor vessel bottom head. Access is byaconcretoshieldwallandsteinicssstoolbrevented lanket insuistion that surrounds the vessel The meridional wclds are furt..'t obstructed by 32 3 4-inch CRD hydraulic n) linos and incore detector lines loca ed just below the high-density aggregate trays. It is considered-at this time that adequato acccca la not available to allow for external preparation and examination of these wolds from the exterior and not feasible due to core support plato structures and stainless stcol cladding for interior examination.
Referenga CE Drawings E-201-794 and L-201-793 Bochtel Drawing 0740G20128 v
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1R-A3m Primary Nozzlo-to-Vossel H,! 3. RogR,enta far EnligInnida 11MA SectionA And
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' HClnh 1191312 End Ecidh CArcaprY D-D2 Itama B3.90 and lisizin-in-SAfn D3.100t And CAtos9Iy D-L. Itom D5 10 C9dA E92Rircacnt Gitegory D: The nozzio-to-vessel wolds and nozzio inside radDius sections are to be volumetrically examined onco during each inspection interval.
Catocory D-Es All dissimilar notal wolds are to be examined by volumetric and nurfaco nondestructivo cxamination (NDE) techniques each inspection interval.
CRds Eclief Renunata Relief is requestod from all examination of the following nozzless
- a. 20-in. recirc nozzles 796-1A and 796-1B (except accessible portions >>f nozzle-safe end (D-F) wolds).
- b. 8-in. shutdown unloading nozzio 795-15.
EI2D9&fd biternativ_q Examinat.iRD r'
Visual examination of accessible surfaces for leakage
( j) during system pressure testing for all sub oct nozzlos plus codo Category B-F volumetric examinat ons.
DAgin fpr Pequestina Re.U.91
- a. Appreximately two-thirds of the exteriors and all of the interiors of the 20-inch recirculation nozzles aro-inaccessible for volumetric examination due to interforences. The concrete chield wall and high-density aggregato traya prevent full access to the two recirculation Diffuser plates provent nozzlos from the outsido.
volumotric inspection of the nozzles from t1e inside, coomotric configuration prevents meaningful volumatric examination of the Category B-D areas. During the inlet baffle repair in 1979.
a visual inspection (romoto) was completed on those nozzlos with no indications noted,
- b. A direct manual volumetric examination is not possible on the 8-inch shutdown unloading nozzle due to inaccessibility of the nozzle. Interforence with the coro spray sparger provents use of the mechanized ultrasonic device. During the baffle repair in 1979, this nozzio was visually (remoto) inspected with no indications noted.
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Reference CE Drawings E-230-791-2, E-201-794-8 and E-2 1-795-5 Dochtel Drawing 0740020128 GE Drawings 212E456, 104R175 l'
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- 4. Recuesta for RolleL. RR-A4 2 2-lnch EgActor Vessel jiozzlest Category B-D,. Items B3.90 and B3.100. add gategpry D-L. Item BS.20 C2d2 Reauirementa catean insido n D-Q1 radius The nozzlo-to-vossel sections wolds andexamined, are to be volumotrically nozzio i Catenary D-El All dissimilar estal wolds are to be examined by surfaco nondestructivo examination (NDE) techniques.
Code Relief Reauests Roliof is requested from Codo surface and volumatric examinations of the following 3-inch nozzlos as describod:
Nozzio Examination Area A. 795-1A Nozzle-to-Vossel Wold Inside Radius Section Nozzle-to-Safo End B. 795-1B* Nozzle-to-vossel wold Inside Radius Section Nozzlo-to-Safe End C. 795-1C Inside Radius Section Nozzlo-to-Safo End D. 795-1D Inside Radius Section Nozzle-to-Safo End E. 795-1E Inside Radius Section Nozzle-to-Safo End F. 796-6 Nozzle-to-Vossel Wold Inside Radius Sectjon Nozzlo-to-Sa.fo End
- Proviously identified as nozzlo 795-6.
Proncsed Alternativo Examination Romoto visual examinations of accessible internal surfaces and hydrostatic testing are proposed.
Basis Lqr Recuestina Relief Generally, access is limited on the outsido by O
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'() concretewalls,highdensity$!!idesofthesenozzlesisgate system lines. Access to the 11mited by shleid water interference high radiation, trays, and various thermal sleeves, and other vessel infernals. The nozzle-to-vessel wolds of three nozzles (795-1C, -1D, and -1E) are accessible and will be examined.
The inside radii of six nozzles (795-1A,-1B, -10,ls-1D, are cannot be examined because materia During
-1B available not and 796-6) for calibration block fabrication.
fabrication of the Big Rock reactor vessel 3-inch nozzles, a uniquo composition of 70% manganese and 30%
nickel was used as a brazing foil. The brazing foil was wrapped around a stainless steal sloove which was inserted The nozzle was then heat treated to11.to the nozzle.
2200'F. This process 1s-no longer used in nozzle f abrication. Ef forts to obtain the brazing foil material have been futile. The nozzle fabrication process and material are required for calibration block fabrication.
piscussion Attempting to gain access to and code examine those nozzles would be impractical in terms of personnel exposure, downtimo, and cost. Also, the absence of adequate calibration blocks with the unique metallurgical configuration of the nozzles precludes meaningful UT examination of inside radii. The proposed alternative O examinations and other Codo examinations on nozzles should provide some information on the nozzios' integrity.
Referenss CE Drawings E-201-795 and E-230-791 Bochtel Drawing 0740G20128 J.,
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- 5. Rent est ' Relief BB- Priman Egaglg .te-Safa X Steam Out Nozzles,_ Cateaory R-L a We Ms___ on B5.10 3
gadg Reaulrement Nozzle-to-safe end dissimilar metal welds areLto be examined by volumetric and surface nondestructive examination (NDE) . techniques.
Code Relief EnmLOJilt Relief is requested from the code surface <
examinations on the 14-in. steam outlet nozzles,.795-lla to -lif.
Pronosed Alternativa Examination Mechanized UT, visual VT-1, and system hydrostatic- ,
test. .
Basia 19I Recuestina Relief These nozzles are inaccessible'for surface examination enclose these from outsideDuring nozzles.- due torefueling, concrete thoscinozzles' walls that are. submerged in the shielding water, making dye . -
penetrant or magnetic particle-testing impossible.-
(Total. core unloading and 1 to 2'R/hr fields also make O. surface examination from inside the nozzles impossible.)
A mechanized ultrasonic examination is performed underwater along'with the prestartup hydrostatic test.
These nozzles are also visually inspected, also underwater, per code requirements.
Discussion __
The external surfaces of-the subiect welds-are inaccessible for surface NDE. The welds are to be-volumetrically examined according-to code requirements.
The code, however, only requires the inner one-third of
.the weld volume (CDEF on Figure IWB-2500-8)' to be volumetrically examined.-For these. welds' theLentire- ;'
volume (ABDEFC on Figure.IWB-2500-8) should be so:
examined. This alternate-procedure should'give an ' '
indication ofwelds-will welds..The the. externalalso surfaceccondition be given internal VT of these examinations and visually examinedEfor leaks during-system-pressure testing. The code surface examinationsLin; this case are impracti,al.
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_ Reference CE Drawing F-230-791-2 Dechtel Drawing 0740G20128 GE Drawings F-201-794 and E-201-795 I
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("\ Heat Exchangers and Other Pressure Vessels
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- 1. Stgan prum Belief Reauestal .
l ER-A RR-A7m h slan HoldHz CatsacrX Egzzle-to-Shell WoldstD-D2 ItemsB-D.
Catacory R2J.1 and D2D)M It2ME m and D3cl60 gado Reauirentnia I
Catengry D-R1 one circumferential wold and 1 foot of a longitudinal weld in the heat exchanger shell which intersects the examined circumferential weld are to be volumetrically examined.
ns SarY M* All de radius nozzle-to-vessel sections welds and nozzle are to be volumetrically examined.
Coilg Re1ie( Ep_qqqat Relief is requested from any code requirements associated with wolds or portions of wolds in the steem drum that are located within a 150-degree arc either side of the bottom longitudinal conterline. These welds includet .
(a) All above located portions of the circumferential shell welds,
) (b) The longitudinal shell wolds, (c) one-half of the full penetration-weld nozzlear (1) downcomers B-1,- 2,-3,-4 (2) risers C-1,-2,-3,-4,-5,-6 (3) foodwater E-1,-2 (4) condensate G-1,-2 (5) gauge glass J-2 (6) level M-2,-4 Proposed biternative Examinati2D The examinations in the upper portions will disclose the condition of those areas representative of the most severe service conditions to which the steam drum is exposed, and therefore, will disclose any incipient general degradation, The hydrostatic examinations performed each refueling outage (prestartup hydro) and the nozzle-to-safe end examinations of the risors and
O downcomers will provide additional indication of the steam drum structural integrity.
Jiagig LO.I Reauestina Relief Access to the upper areas of the steam drum, within approximately a 30-degree arc oither side of the top longitudinal contor-line, is good without extensive scaffolding. The general field la relatively low, approximately 0.06 R/hr. The combination of good access and low general field will permit shell wold examinations without excessive exposure. Access to the lower portions of the steam drum, within approximately 150 degrees of is arc either side of the bottom longitudinal conterlino,in poor without extensivo scaffolding. The general field this area is approximately 10 times (0.6 R/hr) that of the upper portion, For shell welds, no examinations will be conducted in the lower portions due to the combination of high fields and poor access; i.e., approximately 20 towold 25 man-hours will
- pre paration, be required (scaffolding, insulation examination) to por-form each examina, tion in tao lower areas. With six of the ten longitudinal and ,
circumferential welds being located entirely in this '
an additional 170-240 man-rem exposure will be region,d require (28-40 man-rem per vold) to perform a complete Category B-B examination over the interval. However, by O restricting inspection to the up?ar portions of the steam drum, the code requirements can ao fulfilled on four of the ten wolds without accumulation the excess exposure.
Performing the category B-D examinations on nozzles in the lower portions of the steam drum will requiro 505-850 man-rom exposure over the interval nozzle for 17 nozzles). These figur(3os 5-50 areman-rom based on porthe general field in the area of the nozzles. Contact readings on the nozzlo-to-shell weld area are on the order of three timos that of the general field.
DiscussiQD '
For allresult the above welds personnel adhering to code requirements would in excessiv,e exposure. Adequate knowledge of vessel integrity should be gained by performing the proposed code examinations and visually l examining the stesr. drum during prestartup hydrostatic tests. The Codo requires 100% of one circumferential wold (the entire circumferoaco) and a 1-ft longth of a longitudinal weld to be examined. The total length of shell welds to be examined in the proposed upper portion of thethe steamsteam drum drum plus should i ft (1)(equal or 2) the circumferenco ofequal the total lo l
all shell wolds in the upper portion, whichever is less.
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CE Drawing E-230-101-9 ;
Master Plan Isometrics A-14 and A-15 t
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()' 2. Renntat for Rollef ER_-AE. cican_Un Ecnenerative gcAt ExchanngIn ,LEliX1 Unita h L L D1 Catec19IX D-L ltgn B3.160 -
Cnie RenuirsAont All nozzio insido volumotrically examined radius sections onco are to during each be inspection intervs1.
Cpj.q Relief Raquant B, C, D), relief is For the clean-up PJIX3 (Units A, requested from the code requirement to volumotrically examine the insido radius sections on the following nozzloss (a) shn11 inlets (A-7, D-7) gn) shc11 outlets ( B-8, C-8)
Eroposed MigIngfdY2 D(aminatioD Nc,no .
DAG1E 191 RRauRQ.11Dg Ro1ief Those nozzio insido radius Thesections cannot be UT O cxamined due to geometry.
on the ahn11 inlets and outlets will be examined RHX nozzle-to-shell These wels in accordance with the codo requirements. examinations will provido adeq the general internal condition of the nozzle and shell.
The insido radiusattempted unsuccessfully section examinations havo boon and will, thorofore, not be scheduled.
Discussion Because of the geometry of those nozzlos, UT of t.no insido radius sections returns inadequate results. Those examinations have been attempted and the results were not meaningful. Other examinations on those heat exchan$erainc examinations during prestartup hydrostatic testing should give adequato indication of the heat exchangers' integrity.
Rofer.9Aq_q Southwestern Engineering Drawings EM-65149, EM-86924 and DM-86366 r)
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- 4. Reauest f,pr Relief ER-A9 clean-Us nemineraliser Tank Nozzles Cateaory S-Qt item B3.160 cods neouirement For heat exchangers, all nozzle insido radius sections are to be volumatrically examined.
Code Belief Reauest Rollof is requested from the category D-D examinations of the fivo nozzio insido radius sections on the cleanup domineralizer tank.
Proposed Altpynativo Examination The prestartup hydro and the remainder of the category B-D testing will provido adequato. indication of tank structural integrity.
Basis ior Be_gpestina Rollef The cleanup domineralizer tank nozzleo are fabricated by welding square-ended pipo nipples into the tank shell.
The portion of the pipa nippio corresponding to a nozzlo insido radius section has a radius of essentially toro and, therefore, cannot bo UT inspected.
b D1D.211GaiSD Decause of nozzle geometry,UT of the insido radius returns inadequate results. Also, the category B-D and nozzle-to-vessel wold examinations will bo performed,ius re11of is only requested from the required inside rad section examinations. Since the nozzles have a geometry that precludes good UT results, relief is justified.
Reference Infilco Drawings Y-30-4858-3 and M28-4-2 l
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k- D. Piping Pressuro Boundary 1
- 1. IMougnt for Relint RB-Alh Shutdevn Cngling SYllt0E Idno2n.dla- r Matn1 Fittina-12-IdSn iLnlh CALESQIX ll-L. Item p5.130 Cpdn Egguirement Wolds are to be examined by volumotric and surfaco techniques.
Codo Enlief RenucAt Rollof is roguested from the Category D-F requirements to perform volumotric and surfaco examinations on vold 6-SCS-101-7.
Proposed Alternative Examination A hydrostatic test (prontartup hydro) is performed before each startup at 1.1 times the operating pressure.
This is more conservativo than the prestartup leak test at operating pressure as required by the Codo (IWD-5221).
Basis Lqr Regnenting E911ni Pipo wold is not physically accessible for NDE due to
,_s plant design. Access is procluded due to a straight,
(
\/
) uninterrupted vertical drop of 35 foot. No safe access from overhead is available.
Discunsion Attempting to examino this wold would create excessive risk to Personnel safety. Visual examination during pressure tests should provide information as to the integrity of the subject wold.
Referenca Master Plan Isometric A-28 1
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' " E. pump pressure Boundary Cateaory
- 1. neouest ist Relic.t BR-A1L. EMD Caaing Helds k-L-12 lith B12.10 Cssic Reguirement Volumotric examinatione of Class 1 pump casing wolds are to be performed each inspection interval. in Exanf nations are limited to wolds in at least one pum?
each group of pumps performing similar funutions in the system. The examinations may be doforred to the end of two interval.
WAq R.0112C ReGReat Relief is requested from t'.;o Codo required volumatric examinations of pump casing volds in the cleanup and main recirculation pumps.
l Pr.nnnned Alternattyg Examination f
Surface accessible examinations areas. are to be conducted onVisual ex l during system hydrostatic tests.
Basis Int Requesting Eplief i O Approximately 600 man-hours would be required and a total exposure in excess of 1000 man-rem would be have to be performed in full face mask or under supplie air conditions. Shiciding would reduce the total exposure by only 25%.
The man-hour estimato is based only on the onsite work performed by maintenance, operations and NDT personnel and or doesn't the includo engincoring man-hours exponded timo, by preoutage Radiation job planning, Protection personnel providing direct coverage.
The pump casing was fabricated from cast stainless stool (ASTM A-351, Grado CF8H). This alloy corresponds roughly to Type 316 stainless stool (Hi/Cr ratio modified material, and to facilitato casting). The manufacturing process lead to extraord thickness,inarily largo grain sizes. Largo grain cast stainless steels are considernd non-inspectable with ultasonics because this material is highly attenuativo. The presence of deltaDelta ferrite (typica antergranular stress corrosion cracking CIGSCC).
forrito also improvoc resistanco to pitting corrosion.
Holther Byron-Jackson (pump manufacturer) or Southw
of primary coolant pump casing vold.
External placomo.it of radiographic testing (RT) film is precluded due to the base plate configuration of the pump.
Tho RCP has a doublo voluto configuration and lacks according to the manufacturer (Byron-Jackson), ion of the accons for internal visual or surface examinat pump casing. The RCP design corresponds to the Typo C pump illustrated in Figures HB-3441.3-2 and 3423-2 (1980 Edition, ASME P&PV Codo,Section III) . The " splitter" provents placement of RT film cassottes or an exposure devico insido the pump, thus precluding radiographic testing.
DJananten A radiographic examination of the RCP casing wolds appears technically difficult for the Byron-Jackson Such typo pumps, even if the pump is disassembled. examinations are timo consu and dollars, At Point Beach, radiographic examination of wolds on one RCP casing and visual examination of the pump insido pressuro rotsining surfaces woro performed using HINAC and a manipulator. This examination required about 25 days (including pump disassembly and reassembly). It resulted in a total accumulated radiation O exposure of 36 man-rom and a cost of about $700,000.
Radiographing through two wall thicknesses to examino a wold in one wall, as would be necessary for the Byron-Jackson type pump casings.
The MINAC has boon used at Ginna, Turkey Point, Point Beach, and Robinson. No notable indications woro found in any of the pumps examined.
At this timo, disassembly of pumps solely for making this volumetric examination does not appear warranted in view of the radiation exposure and lack of a viable examination technique. However, an examination should be done by the most feasible means if a pump at Big Rock Point requires disassembly for maintenanco. At this stage of technology development, a surface examination of at least a portion of those wolds, whero practical, would be most appropriato. Tho inspection interval considered in this report ends in January 2002. No pumps woro disaccombled for maintenance or examination during the last interval. Since the examination technology is only it isinterval.
currently being actively developed, ion reasonabic to give relief for the present inspect C
t, Referenct !
Dyron Jackson Drawing 1F-4614-3 i
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- 2. Beauest for Ediaf EE-Allt Pamp Internal Eurfaca
_Examinationt gatecory B-L-1A Item B12.20 Cada Esquiremont Visual examinations of Class 1 pump casing internal surfaces are to be performed each inspection interval.
Examinations are limited to at least one pump in each group system.
of pumps performing similar functions in theThe examinatio the interval.
C2dg Relief Recuest Relief is requestod from the requiremont to visually examine the casing internal surfaces of the main recirculation pumps.
Proposed hitornative ExaminatioD A pump interior will be examined to the extent practical should it bo disassembled for any other reason.
Ragig igI Recuestina Relief Visual examination of the internal surfaces (main
/'N reci'eculation pumps 1 and 2l is not possible due to (j configuration. During the Big Rock Point 1982 refueling outage, access engineering was conducted on the main recirculation pump internal surface by Southwest Roscarch '
Instituto. Due to internal configuration, visual examination of the internal surfaces is not practical.
(Sco also the previous item,E.1).
Discussion The visual examination specified is to determine whether unanticipated sovoro degradation of the casing is occurring due to phenomena such as crosion or corrosion.
At this time, disassembly of pumps solely for making this visual examination does However, not appear an warranted examination in view of the radiation exposure.
should be dono if a pump at Big Rock Point requires disassembly for maintenanco.
Reference Byron Jackson Drawing 1F-4614-3 O
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O F. Valvo Pressurg _ Boundary
- 1. RegttefLta 19I Relief BB-ML giggg 1 yalyg Jnternal Examinationt Gateaory D-H-L ltga 1112.50 CDd2 Reauirement one valvo in each group of valvosgate, of the orsame check, constructional design, and e.g., globo, manufacturer, that performs manufacturing method,the system shall be visually similar functions inexamined during each inspection interval. This exami-nation may be performed on the samo valvo selected for the category B-M-1 examination.
The examinations may be performed at or near the end of the inspection interval.
ggdg 891ket Reauest Following is a list of 26 Class 1 valves in 12 groups at Big Rock Point. Pc411of is requested from the visualsurfaces of valvo examination of the interna) noted categories. .
Class 1 Romarks Valvos
- V 1, 1-Ingh y31y33 (Chapman-Gatel_
MO-N002A Rollef requested. Examino MO-N002B 2D12 if maintenance required.
2 4-Ingn yalven _(Anchor-Darlina Gato)_
CV-4180 Examino one valvo in this CV-4181 group each inspection CV-4182 interval.
CV-4183
- 3. 4-Ingh Yalves IEgys11-Gate)_
MO-7052 Examine one valvo in this MO-7062 group each inspection interval
- 4. s-Ingh Valves iTarget Rock)
SV-4984 Examino one valvo in this SV-4985 group each inspection SV-4986 interval.
SV-4987 O
5 1-IRGh Valven .[Powell-GAttl H0-7056 Examine one valve in this -
MO-7057 group each inspection MO-7058 interval. !
MO-7059 t
- 6. lQ-l[LQh yajygg (Edwards-Globei VIM-9 Relief requested.
Examine only if maintenance is required.
- 7. lQ-lagh Valves iEdwarda-Checkt
- VIV-304 Relief requested. Examine '
9 Dig if maintenance is required.
- 8. 10-Inch yalves-1 Anchor Darlina-checkt V"4-3 0 5 Examine-each inspection- :
interval.
9 12-Inch Valves IPowell-Gate)
MO-7050 Examine each inspection interval. ,
- 10. 22-lagh y3Lyng (Chapman-Gatel MO-N001A Relief-requested. Examine-MO-N001B only if maintenance is required.
- 11. 22-Inch Valves (Allis-Chalmers-Butterfivl Relief' requested. Examine MO-N006A only if maintenance is ;
MO-H006B required. ,
12 11-Inch Valveg fchaomRD-Gatet MO-N003A Relief requested. Examine.
- MO-N003B. DDlY if maintenance is required.-
I O .
OV EISPmLtd Altgrnative Examination A hydrostatic test (prestartup hydro) conducted at 1.1 times the operating pressure will serve as an alternate test. The inspection of other valvos and the results will also give an indication of the condition o disassemble those valvos for maintenance, the licensee has committed to visual examinations.
Dnald 1Qr Regunn. ting Bellof The main recirculation pump discharge and butterfly valvos are not fully isolatable from the reactor and are, thoroforo, not inspectable, Examination of the above valvos requires completo draining of the reactor vessel which is not practical and requires 100% core off-loading, which creates tremendous innintenance and exposuro problems. Examining the 24-inch valvos Examining the feodwatoa valvas would radiation fields. lead to 40-50 man-rom exposure nach during the interval.
Qingpsolon The visual examination specified is to datormine whether unanticipated severo degradation of valvo bodios <
is occurring due to phenomens such as crosion orflowever, . t O corrosion, the degree necessary to inspect the internal To do pressurer to personnel and radioactive waste generation.this dis of the internal surfaces is impractical.
Of the 12 valvos required to be examined, six areinformation-on the scheduled for examination. Thorofore,ill be available. In overall condition of class 1 valves waddition,Hence, any valve for where be examined if disassembled for maintenance.the current i For those valve groups in which no valva is disassembled and examined, system pressure tes J Reference A-29, A-30, A-47, Master Plan Isometrics A-20, A-21,A-83 and D-7 A-66, A-67, A-68, A-69, A-81,
G. Component Supports
- 1. Regunat Inr Rollef RR-A14,_ ggmoonent Eupports, Pinina gatgggry F-Ct Item EL1Q Cpdq Recuirementa Visual examination ?f piping supports are to be performed each inspection interval. The examinations include verification of the settings of snubbers, shock absorbers and spring-type hangers.
CgAq Eclief Requent Relief is requested from direct visual examination (for operability) of the following pipe supports:
3-LPS-102-23-PR 3-LPS-102-21-PR-1 3-LPS-102-21-PR-2 6-SCS-101-6-PS 8-SCS-101-8-PR 8-SCS-101-10-PR 1.5-MSS-117-12-PR 1.5-MSS-117-14-PR 1.5-MSS-117-16-PR-1 1.5-MSS-117-18-PR
..g 14-MRS-103-3-PR-1
,7 5-MRS-132-7-PR-1 5-MRS-132-7-PR-2 Pronosed Alternative Examination Examinations shall be performed remotely.
Basis ign Reauestina Relief The supports listed above are located 50 feet above the lower area deckdue is curved areatoof containment.
containment The floor inand/or configuration this the use of scaffolding is not poss.t.ble. Access from the top is also not feasible. These supports are located in high radiation areas which precludes direct examination due to the time that would be required to perform the examination.
Discussion IWA-2213(c) identifies that for component supports and component interiors, the visual examination may be performed andVT-3 remotely,he relief for that examination examinations is in as identified not required. For t IWA-2213(b) " Conditions that could affect operability or functional adequacy", the inaccessibility of those x
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supports justifies remoto examination.
Reference Master Plan Isolnotrica A-25, A-26, A-28, A-29, A-54, A-55, A-62 .ind A-69 1
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BECTION 6 VERIFICATION OF SECTION XI COMPLIANCE LO O
YIRIFICATION QZ COMPLIANCE A. INTRODUCTION The following tables document compliance, for the third interval, with the examination distribution requirements of Section XI of the ASME B&PV Code. The tables identify the total number of components by category which are subject to distribution and the total number of components.
B. DfrERMINATION QE ColiPLIANCE
- 1. First Eatigd The minimum and maximum number of components required to be examined during the first period has been determined by applying the minimum and maximum percentages cited in the code, Table (s) IWB-2412-1, IWC-2412-1 and IWD-2412-1.
The minimum number of components to be examined is 16% of the total components. The maximum number of components which can be examined is 34% of the total number of components.
- 2. EtqpJ1d Eetind The minimum number of components to be examined is 50% of
[) the total components. The maximum number of components A/ which can be examined is 67% of the total number of components.
- 3. Third Period The minimum and maximum number of-components to be examined is obtained based solely on the number of examinations required to complete the cumulative total.
C. NUMBER QE COMPONENTS The number of components subject to examination per peried will vary throughout the life of the plant due to code changes relief requests and line walking. Compliance per period is based solely on the number of components subject to the examination for that period.
D. INTERVAL COMPLIANCE Third period cumulative percentage totals which equal or exceed 100% verifies compliance with the distribution requirements of Section XI.
In the event that the number of examinations subject to distribution in a category decreases, and it is determined that it is impossible to achieve 100% without re-examining components, the component will not be re-examined. For example:
} During theare first period (le, 5
1.
category examined of 20 components in a 25%).
- 2. During the second period, it is determined that only 15 components actually exist in that category _and that 5 additional components are examined (ie, 33-1/3% for this interval third and 58.3% cumulative).
- 3. The 5 remaining components are examined during the last period (ie, a cumulative total of 91.7%).
Under this situation, which could result from a number of reasons, no further examining is required provided-adequate-documentation substantiates the anomaly.
C E. MIDINTERVAL REQUIREMENT _HANGES 10 CFR 50.55a(g) requires periodic updates of ISI programs-to the currently approved version of Section XI.
Implementation of these changes in this plan in midinterval may require examinations of areas not previously subject to examination. No attempt is made to " catch up" those examinations.
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VERIFICATION OF COMPLIANCE THIRD INTERVAL I
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VERIFICATION OF COMPLIANCE '3 THIRD INTERVAL
. PERCENT:OF
-l( ASME- ASME -TOTAL-. . / EXAMS- # EXAMS REQUIRED
-- =
ITEM CAT. COMPONENTS (a) -REQUIRED (b) SCHEDULED. _ EXAMS g l
Bl .11 - B-A~ 2 1 1 100.00 Bl.12 -B-A 4 4 4' 100.00 Bl.21 B-A 2 2 2 100.00 Bl . 2 2 - B-A 8 1 1 100.00 Bl.30 B-A 1 1 .1 =100.00.-
Bl.40 B-A 1 1 1 .100.00 TOTAL B-A 18 10 10 '100.00 B2.51 B-B 21 13 13 100.00 B2.60 B-B 5 2 2: 100.00 B2.70 B-B 8 1 1 100.00 B2.80 B-B 5 2 2 .100.00 TOTAL B-B 39 18 18- 100.00 B3.100 B-D 20 20 20 100.00' B3.150 B-D. 45 28 28 100.00 B3.160- B-D 45 19 19 100.00 B3.90 B-D 20 20- 20 100.00 TOTAL B-D 130- 87' 87 -100.00 B4.11 B-E 14 4 14 '100.00 B4.12 B-E 32 8 10' 100.00 TOTAL B-E 46 .12 24 100.00 F
O PAGE.11
VERIFICATION OF COMPLIANCE THIRD _ INTERVAL 2 PERCENT OF ASME '. ASME TOTAL- # EXAMS -# EXAMS- REQUIRED
-, ITEM ' CAT. COMPONENTS (a) REQUIRED (b) SCHEDULED- EXAMS, '
B5.10 B-F. 9 9 -9 100;00' B5.100 B-F 10 10 10' 100.00' B5.130' B-F 2 1 .1 100.00 B5.140 B-F 12 12 12 100,00 B5.150 B-F 16- 16 16 100.00 B5.20 B-F 6 6. 6 1100.00 TOTAL B-F 55 54 -54 100.00 B6.10 B-G-1 1 1 1 100.00 B6.120 B-G-1 2 2 2 100.00 B6.130 B-G-1 2 0 0 100.00 B6.140 B-G-1 2 2 2 100.00-B6.20 B-G-1 1 1 1~ 100.00-..
1 1 1 100.00
( B6.40 B-G-1 B6.50 B-G-1 1 1 1 -100.00-TOTAL B-G-1 10 8' 8 100.00:
B7.10 B-G-2 10 10. 10 100.00 B7.40 B-G-2 3 3- 3 100.00-B7.50 B-G-2 25 25 25 :100.00 B7.60 B-G-2 3 3 3 100.00:
B7.70 B-G-2 71- 71 :71 100.00:l [
B7.80 B-G-2 32 32 32 100.00" TOTAL B-G-2 143 143 143 100.00' B8.10 B-H 20- 0 N/A B8.20- B-H 18 _0 O N/A TOTAL B-H 38 0 0- N/A-O PAGE-2
VERIFICATION- QZ_C.QMPJJAH93 THIRD INTERVAL PERCENT OF ASME ASME TOTAL / EXAMS / EXAMS REQUIRED
/'_s) ITEM CAT. COMPONENTS (a) REQUIRED (b) SCHEDULED EXAMS
%.)
D9.11 B-J 352 90 124 100.00 B9.12 B-J 42 11 14 100.00 B9.21 B-J 347 87 100 100.00 1 B9.31 B-J 9 3 7 100.00 14 13 B9.32 B-J 4 100.00 B9.40 B-J 384 98 102 100.00 TOTAL B-J 1148 293 360 100.00 B10.10 B-K-1 41 0 0 N/A B10.20 B-K-1 3 0 0 N/A TOTAL B-K-1 44 0 0 N/A B12.10 B-L-1 8 7 7 100.00-
/ 'i B12.20 B-L-2 3 1 1 100.00 l .G B12.50 B-M-2 26 6 7 100.00 B13.10 B-N-1 1 1 1 100.00 B13.20 B-N-2 3 3 3 100.00 B13.40 B-N-2 1 1 1 100.00 TOTAL B-N-2 4 4 4 100.00 B14.10 B-O 16 2 2 100.00 C1.10 C-A 2 2 2 100.00 C2.21 C-B 1 1 1 100.00 C2.22 C-B 1 1 1 100.00 TOTAL C-B 2 2 2 100.00 s_,
PAGE 3
VERIFICATION OF COMPJJ2EEE TRIRD INTERVAL PERCENT OF ASME ASME TOTAL _
/ EXAMS / EXAMS REQUIRED-ITEM CAT. COMPONENTS (a) REQUIRED (b) SCHEDULED EXAMS
(
C3.10 C-C 4 4 4 100.00 C3.20 C-C 4 4 4 100.00 TOTAL C-C 8 8 8 100.00 !
l C5.81 C-F-2 3 1 1 100.00 C5.51 C-F-2 255 28 33 100.00 j TOTAL C-F-2 102 29 34 100.00 F1.10 F-A 51 51 51 100.00 F1.40 F-A 20 20 20 100.00 TOTAL F-A 71 71 71 100.00 F2.10 F-B 4 4 4 100.00_
F2.40 F-B 33 33 33 100.00 TOTAL F-B 37 37 37 100.00
(
F3.10 F-C 224 224 224 100.00 F3.40 F-C 67 67 67 100.00 F3.50 F-C 157 157 157 100.00 TOTAL F-C 448 448 448 100.00 (a) The total components listed above represent the total number of components subject to distribution after the application cf exemptions.
(b) Th2 total components listed above represent the total number oi components subject to distribution after the application of relief requests and exclusions.
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EXAMINATION CATEGORY B-A SYSTEM 'SECTION XI 'LINE NUMBER total: 4WDF,R COMPONENTS SCHEDULED REMARKS ITEM NO. -OF COMPONENTS TO BE INSPECTED LRVG Bl.11 Reactor Vessel 2 1 RELIEF REQUEST RR-Al RVG Bl.12 Reactor Vessel 4 4 RELIEF REQUEST.RR-Al RVG Bl.21 Reactor Vessel 2 2 RVG Bl.22 Reactor Vessel 8 1 RELIEF REQUEST RR-A2 RVG Bl.30 Reactor Vessel 1 1 RVG Bl.40 Reactor Vessel 1 1 PAGE 1 of.31 l
EXAMINATION CATEGORY B-B SYSTEM SECTION'XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED ECS 92.51 Emergency Condenser 4 2 ,
RCS B2.51 Clean-up Heat Exchangers 10 4 RSD B2.51 Steam Drum 4 4 RELIEF REQUEST RR-A6 RCS B2.51 Demin. Tank 3 3 RCS B2.60 Clean-up Heat Exchangers- 5 2 RSD B2.70 Steam Drum 6 1 RCS B2.70 Demin. Tank 2 1 RCS B2.80 Clean-up Heat Exchangers 5 2 t
F PAGE.2 of 31
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EXAMINATION CATEGORY B-D SYSTEM SECTION XI- LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED RVi .B3.100 Reactor Vessel 20 20 RELIEF REQUEST RR-A3 "CS 4 .B3.150' Emergency Condenser 4 4 RCS- B3.150 . Clean-up Heat Exchangers 4 4 RSD B3.150 Steam Drum. 32 15 RELIEF REQUEST RR-A7 RCS B3.150 Demin.. Tank 5 5 ECS B3.160 Emergency Condenser 4 4 RCS B3.160 Clean-up Heat Exchangers 4 O RELIEF REQUEST RR-A8 RSD B3.160 Steam Drum 32 15 RELIEF REQUEST RR-A7 RCS B3.160 Demin. Tank 5 0 RELIEF REQUEST. PJt-A9 RVG B3.90 Reactor vessel' 20 20 RELIEF REQUEST.RR-A3 l
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EXAMINATION CATEGORY B-E SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS .
ITEM NO. OF COMPONENTS TO BE INSPECTED RCS B4.11 Clean-up Heat Exchangers 14 14 ,
RVG B4.12 Reactor Vessel 32 10 l
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d EXAMINATION CATEGORY B-F SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REM. N ITEM NO. OF COMPONENTS TO BE INSPECTED RVG B5.10 Reactor Vessel 9 9 RELIEF REQ. RR-A3 & A-5 RSD B5.100 Steam Drum 10 10 SCS .B5.130 6-SCS-101 1 O RELIEF REQUEST RR-A10<
SCS B5.130 6-SCS-102 1 1 LPS B5.140 3-LPS-102 1 1 LPS B5.140 3-LPS-103 1 1 RCS B5.140 3-RCS-101 2 2 RCS B5.140 3-RCS-102 1 1 RSD B5.140 3-MSS-107 6 6 PIS B5.140 3-CSS-101 1 1 RCS B5.150 2-RCS-101 1 1 RVG B5.150 1.5-MSS-117 5 5 RDS B5.150 1.5-RDS-112 2 2 RDS -B5.150 1.5-RDS-113 2 2 RDS B5.150 1.5-RDS-114 2 2 RDS B5.150 1.5-RDS-115 2 2 RDS B5.150 2-RDS-107A 2 2 RVG B5.20 Reactor Vessel 6 6 RELIEF REQUEST RR-A4 PAGE'S of 31
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EXAMINATION CATEGORY B-C-1 SYSTEM SECTION XI LINE 'JUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED RVG B6.10 Reactor Vessel 1 1
-RSD B6.120 Steam Drum 2 2 RSD B6.130 Steam Drum 2 O EXAMINE IF DISASSEMBLED RSD B6.140 Steam Drum 2 2 RVG B6.20 Reactor Vessel 1 1 RVG B6.40 Reactor Vessel 1 1 RVG B6.50 Reactor vessel 1 1 PAGE 6 of 31
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EXAMINATION CATEGORY B-G-2 SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED
'RVG B7.10 Reactor Vessel 10 10 ECS B7.40 Emergency condenser 2 2 RCS B7.40 Demin. Tank 1 1 MSS B7.50 1.5-MSS-117 2 2 MSS B7.50. 3-MSS-107 12 12 RCS B7.50 3-RCS-101 3 3 RCS B7.50 3-RCS-102 1 1 PIS B7.50 4-CSS-101 1 1 PIS B7.50 4-RDC-101 2 2 RDS B7.50 6-RDS-102 1 1 RDS B7.50 6-RDS-103 1 1 RDS B7.50 6-RDS-104 1 1
.RDS B7.50 6-RDS-105 1 1 PCS. B7.60 Main Recirc Pump No.2 1 1 PCS B7.60 ' Main Recirc Pump No.1 1 1 RCS B7.60- Reactor Clean-up Pump 1 1 MSS B7.70 1.5-MSS-1111 2 2
. MSS B7.70 1.5-MSS-112 2 .:2 MSS B7.70 1.5-MSS-117- 1 1 MSS B7.70 1.5-MSS-110 2 2
. MSS B7.70- 1.5-MSS-124 1 1 RCS 87.70' -1.5-RCS-111 ~ 2 2' RCS B7.70 1.5-RCS-110 2 2 CRD- B7.70 2-CRD-101 2 2
'CRD B7.70 2-CRD-102 1 1
'CRD B7.70 '2-CRD-111 1 1 LPS B7.70 2-LPS-101 1. 1 LPS B7.70 2-LPS-102 2= 2 MSS B7.70~ 2-MSS-121 1 1 MSS- B7.70 2-MSS-131- 1 1 MSS B7.70 '2-MSS-134 1 1 RCS B7.70. 2-RCS-104 3 3 RCS B7.70- 2-RCS-105 2 2 PAGE 7 of'31 r Md
EXAMINATION CATEGORY B-G-2 SYSTEM SECTION II LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARFS ITEM NO. OF COMPONENTS TO BE INSPECTED RCS B7.70 2-RCS-106 2 2 RCS B7.70 2-RCS-107 4 4 RDS B7.70 2-RDS-107A 1 1 LPS B7.70 3-LPS-102 1 1 LPS B7.70. 3-LPS-103 2 2 MSS B7.70 3-MSS-107 7 7 RCS B7.70 3-RCS-101 4 4 RCS --B7.70 3-RCS-102 3 3 PIS B7.70 4-CSS-101 1 1
'PIS B7.70 4-RDC-101 1 1 PCS B7.70 5-MRS-131 1 1 PCS B7.70 5-MRS-132 1 1 RDS B7.70 6-RDS-102 2 2' RDS' B7.70 6-RDS-103 2 2 RDS B7.70 6-RDS-104 2 2 RDS B7.70 6-RDS-105 2 2 FWS B7.70 ~10-FWS-101 1 1 PCS B7.70 20-MRS-121 2 2 PCS B7.70 20-MRS-122 2 2 PCS B7.70 24-HRS-121 1- 1 PCS B7.70 24-MRS-122 1 1 RVG B7.80 Reactor vessel 32 32 PAGE 8 of 31
EXAMINATION CATEGORY B-H SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED RVG B8.10 Reactor Vesse) 20 0
-RSD B8.20 Steam Drum 18 0 The above items do not require examination per Table IWB-2500-1,1986 Edition of Section XI
, PAGE 9of 31
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EXAMINATION CATEGORY B-J SYSTEM SECTION X1 LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED
'FWS- B9.11 10-FWS-101 7 2 MSS B9.11 '12-MSS-105 15 3 RDS B9.11' 12-RDS-101 16 8 PCS. 'B9.11 14-MRS-101 4 2 PCS B9.11 14-MRS-102 5 2 PCS B9.11 14-MRS-103 4 2 PCS B9.11 14-MRS-104 4 2 PCS. B9.11 14-MRS-105 4 2 PCS B9.11 14-MRS-106 4 2 PCS B9.11 17-hES-111 6 3 PCS B9.11 17-MRS-112 1 PCS B9.11 '17-MRS-113 6 1 PCS B9.11 17-MRS-114- 6 2
- PCS B9.11 20-MRS-121 .1 1 PCS B9.11 20-MRS-122 11 1 PCS .B9.11 24-MRS-121 9 1 PCS B9.11 24-MRS-122 9 1 PIS B9.'11 4-CSS-101 18 5 ECS B9.11 4-ECS-103- 15 4 ECS B9.11 4-ECS-104 16 4 PCS' B9.11 4-MRS-141 12 9 MSS B9.11' 4-MSS-111 1 O MSS- 89.11 4-MSS-112 1 0 PIS B9.11 4-RDC-101 31 15 PCS- B9.11 5-MRS-131 9 ?
PCS B9.11 5-MRS-132 10 10-ECS B9.11 6-ECS-101 13 4 ECS R9.11 6-ECS-102 14 5 -
RDS B9.11 6-RDS-102 3 0 RDS B9.11 6-RDS-103~ 3 0 RDS B9.11 6-RDS-104 3 0 RDS .B9.11 6-RDS-105 4 2 SCS B9.11 6-SCS-101 6 0 SCS B9.11 6-SCS-102 17 8
'PAGE 10 of.31'
, - _ ._ _ ~_ . _ _ _ _ _ _ _ _ _ - . . - - _ - _ - . - _ _ ,
,.~ . <m EXAMINATION CATEGORY B-J SYSTEM SECTION II LINE MUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED FWS B9.11 8-FWS-102 3 1 FWS B9.11 8-FWS-103 5 1
' MSS B9.11 8-MSS-101 7 1 MSS B9.11 8-MSS-102 3 1 ,
MSS B9.11 8-MSS-103 3 1 MSS B9.11 8-MSS-104 7 O SCS B9.11 8-SCS-101 12 5 SCS .B9.11 8-SCS-102 9 3 PCS B9.12 14-MRS-101 6 2-PCS- B9.12 14-MRS-102 8 1 PCS B9.12 14-MRS-133 6 1 PCS B9.12 14-MRS-104 6 2 PCS B9.12 14-MRS-105 6 2
'PCS B9.1'2 14-MRS-106 6 2 PCS ~ B9.12 17-MRS-111 .2 2 PCS B9.12 .17-MRS-114 2 2 MSS B9.21 .1.5-MSS-106 3 1 MSS B9.21 1.5-MSS-111 16 3 MSS B9.21 1.5-MSS-112 16 4 MSS B9.21 1.5-MSS-113' 4 2 MSS ~B9.21 1.5-MSS-114 11 4 MSS 'B9.21 1.5-MSS-115 4 1 MSS B9.21 1.5-MSS-116 8 '2 CRD B9.21' 2-CRD-101 3 1 CRD B9.21 2-CRD-102 1 1 LPS 'B9.21 2-LPS-102- 2. 1
. MSS B9.21 2-MSS-121 1 1 MSS B9.21' 2-MSS-131 1 1 MSS B9.21 2-MSS-134 1 O MSS B9.21 2-MSS-136 1 1 RCS B9.21 .2-RCS-101 7 3 RCS B9.21 '2-RCS-104 2 1 RCS. B9.21 2-RCS-106 4 1 PIS 'B9.21 3-CSS-101 4 0 PAGE 11. of . 31 ~
EXAMINATION CATEGORY B-J SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMFONENTS SCHEDULED REMARFJS ITEM NO. 'OF COMPONENTS TO BE INSPECTED LPS B9.21 3-LPS-102 45 12 LPS B9.21 3-LPS-103 10 3 RCS B9.21 '3-RCS-101 96 29 RCS B9.21 .3-RCS-102 57 15 RCS B9.21 3-RCS-103 4 1 RCS B9.21 3-RCS-108 14 2 RCS B9.21 3-RCS-121 15 5 RCS B9.21 3-RCS-122 15 5 FWS B9.21 3/4-FWS-101 2 O PCS B9.31 20-MRS-121 3 3
~PCS B:9.31 20-MRS-122 2 2
.PCS 'B9.31 24-MRS-121 2 1 PCS B9.31 24-MRS-122 2 1 MSS B9.32 12-MSS-105 1 1-RDS B9.32 12-RDS-101 1 1 MSS B9.32 1.5-MSS-106 1 1 RDS 'B9.32 1.5-RDS-112 1 1 RDS B9.32 1.5-ROS-113 1 1 RDS B9.32 1.5-RDS-114 1 1 RDS B9.32 1.5-RDS-115 1 1 LPS B9.32 2-LPS-101 1 1 RDS B9.32 2-RDS-107A 1 1 ECS .B9.32 6-ECS-102 1 1 RDS B9.32 6-RDS-102 1 1 RDS B9.32 6-RDS-103 1 1 RDS B9.32 6-RDS-104 1 1 RDS .B9.32 6-RDS-105 1 O MSS B9.40 1.5-MSS-110- 22 9 t
PAGE 12 of 31
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v v EXAMINATION CATEGORY B-J SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPCNENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO bE INSPECTED MSS B9.40 1.5-MSS-117 36 10 MSS B9.40 1.5-MSS-122 6 2 MSS B9.40 1.5-MSS-123 6 0 MSS B9.40 1.5-MSS-124 13 4 MSS B9.40 1.5-MSS-132 6 O MSS B9.40 1.5-MSS-133 6 2 RCS B9.40 1.5-RCS-110 4 1 RCS B9.40 1.5-RCS-111 4 2 RDS B9.40 1.5-RDS-112 5 0 RDS B9.40 1.5-RDS-113 S O RDS' B9.40 1.5-RDS-114 5 4 ,
RDS B9.40- 1,5-RDS-115 5 1 CRD B9.40 ' 2-CRD-101 44 12 CRD B9.40 2-CRD-102 5 1 CRD B9.40 2-CRD-111 31 11 LPS B9.40 2-LPS-101 19 6 LPS B9.40 2-LPS-102 2 1-MSS B9.40 2-MSS-121 23 4 MSS B9.40 2-MSS-122 8 4 MSS B9.40 2-MSS-123 8 O-MSS B9.40 2-MSS-124 7 0 MSS B9.40 2-MSS-131 17 2 MSS B9.40 2-MSS-132 8 0 MSS B9.40 2-MSS-133 8 1 MSS B9.40 2-MSS-134 9 O MSS B9.40 2-MSS-136 3 1 RCS B9.40 2-RCS-104 5 2 RCS B9.40 2-RCS-105 4: 2 RCS B9.40 2-RCS-106' -13 4 RCS- B9.40 2-RCS-107 14 4
-RCS B9.40 '2-RCS-110 1 1 RCS B9.40 2-RCS-111 l' O RDS :B9.40 2-RDS-107A 17 -- 8-RDS B9.40 2-PDS-112 3 1-RDS- B9.40 - 2-RDS-113 3 1 RDS 'B9.40 2-RDS-114 3 0' PAGE 13 of 31
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-EXAMINATION CATEGORY B-J SYSTEM SECTION XI LINE NUMBEh TOTAL' NUMBER COMPONENTS SCHEDULED FEMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED RDS B9.40 2-RDS-115 3 1 FWS D9.40 3/4-FWS-101 2 1 ,
PAGE 14 of 31
_.--. , _ . . _ =. _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ . _ _ _ . ._ .
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EXAMINATION CATEGORY B-K-1 SYSTEM SECTION XI LINE-NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED MSS B10.10- 12-MSS-105 1 0 PDS B10.10 12-RDS-101 3 0 PCS B10.10 14-MRS-101 2 0 .
PCS 'B10.10 14-MRS-102 3 0 PCS B10.10 14-MRS-103 4 0 PCS B10.10 14-MRS-104 3 0 PCS B10.10 MRS-105 4 0 PCS B10.10- 14-MRS-106 2 0 PCS B10.10 17-MRS-111. 4 0
-PCS B10.10 17-MRS-112 2 O PCS :B10.10 17-MRS-113 2 0
-PCS B10.10 17-MRS-114 4 0 PCS B10.10 20-MRS-121 2 O PCS B10.10- 20-MRS-122 2 O ECS B10.10 6-ECS-102 1 O MSS B10.10 8-MSS-101 1. O MSS B10.10 8-MSS-104 1 0
'. RCS .B10.20- Reactor Clean-up Pump 3 O The above items do not require examination per Table IWB-2500-1,1986 Edition of Section II PAGE 15 of'31 f'
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%J %/ \J EXAMINATION CATEGORY B-L-1 SYSTEM SECTION XI LINE NUMBER TOTAL hUMBER COMPONENTS SCHEDULED PIMARKS ITEM NO. OF COMPONENTS TO LE INSPECTED PCS B12.10 Main Recirc Pump 2 1 RELIEF REQUEST RR-All RCS B12.10 Reactor Clean-up Pump 6 6 -RELIEF REQUEST RR-All PAGE 16 of 31
O O O EXAMINATION CATEGORY B-L-2 TOTAL NUMBER COMPONENTS SCHEDULED REMARKS SYSTEM SECTION XI LINE NUMBER OF COMPONENTS TO BE INSPECTED ITEM NO.
1 1 RCS B12.20 Reactor Clean-up Pump Main Recirc Pump 2 O RELIEF REQUEST RR-A12 PCS B12.20 PAGE 17 of 31
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EXAMINATION CATEGORY B-M-2 SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM No. OF COMPONENTS TO BE INSPECTED
' MSS B12.50 12-MSS-105 1 1 PCS B12.50 20-HRS-121 2 O RELIEF REQUEST RR-A13 PCS B12.50 20-MRS-122 2 O RELIEF REQUEST RR-A13 PCS B12.50 24-MRS-12) 1 O RELIEF REQUEST PJt-A13 PCS B12.50 24-MRS-122 1 O RELIEF REQUEST RR-A13
'PCS B12.50 5-MRS-131 1 O RELIEF REQUEST RR-A13-PCS B12.50 5-MRS-132 1 O RELIEF REQUEST RR-A13.
Eva B12.50 '6-ECS-101 1 1 ECS B12.50 6-ECS-102 1' O RDS B12.50 6-RDS-102 2 1 RDS B12.50 6-RDS- 103 2 2 RDS 'B12.50 6-RDS-104 2 O RDS B12.50 6-RDS-105 2 O SCS B12.50 8-ECS-101 2 O SCS B12.50 8-SCS-102 2 1 FWS B12.50 10- FWS-101 3 1 RELIEF REQUEST RR-A13'
's PAGE 18 of 31
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. EXAMINATION CATEGORY B-N-1 SYSTEM: SECTION XI LINE NUMBER TOTAL NUMBER ' COMPONENTS SCHEDULED REMAAKS ITEM NO. OF COMPONENTS TO BE INSPECTED f
PCS B13.10 Reactor Vessel 1 1 b
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- PAGE 19 of 31
EXAMINATION CATEGORY B-N-2 TOTAL NUMBER COMPONENTS SCHEDULED REMARKS SYSTEM SECTION XI LINE NUMBER OF COMPONENTS TO BE INSTECTED ITEM NO.
3 3 PCS B13.20 Reactor Vessel 1 1
PCS B13.40 Reactor Vessel i
i PAGE 20 of 31
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EXAMINATION CATEcARY C-A TOTAL NUMBER COMPONENTS SCHEDULED REMARKS SYSTEM SECTION II LINE NUMBER OF COMPONENTS TO BE INSPECTED ITEM NO. _,
2 2 CRD C1.10 Scram Dump Tank l
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PAGE 22 of 31
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EIAMINATION CATEGORY C-F-2 SYSTEM SECTION XI LINE NUMB *lR TOTAL NUMBER COMPONENTS SCHEDULED REMARKS
< ITEM No. OF COMPONENTS TO BE INSPECTED l.
FWS C5.51 10-FWS-201 45 14
- MSS C5.51 '10-MSS-204 2 O MSS C5.51 12-MSS-201 15 9 CRD C5.51 6-CRD-201' 18 4
- FWS C5.51- 6-FWS-204 4 2 FWS C5.51 6-FWS-205 8 4 FWS C5.51 8-FWS-201 7 0 FWS .C5.81 10-FWS-201-4/2-FWS-202- 1 O FWS . C5. 81.. 6-FWS-204-2/2.5-FWS-207 1 0 FWS C5.81 6-FWS-205-2/2.5-FWS-208 1 1 The total number of components identified below are those welds excluded by Category C-F-2 which are included for total weld count as required by IWC-2500 C-F-2 note (2).
CSS N/A 6-CSS-201 45 O j' PIS N/A 6-PIS-201 18 0 4
PIS N/A 8-PIS-202 11 'O.
PIS N/A 8-PIS-203 3 0 SCS N/A 8-SCS-201 5 0 t
- SCS .N/A 6-SCS-201 10 0 l; LSCS N/A 6-SCS-202 6 0 l SCS N/A 6-SCS-203 22 -O
- . SCS
~SCS N/A N/A 8-SCS-203 6-SCS-204 23 5 0 0
3
.FWS N/A 10-FW3-201 2 O MSS N/A 12-MSS-201 3 0 1
PAGE 25 of. 31
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- j. F-A l-
'; ' SYSTEM SECTION II LINE NUMBER TOTAL NUMBER COMifmr.ais SCHEDULED REMARKS ITEM NO. OF COMPOhhard M BE INSPECTED l
SWS F1.10 10-Shs-301 4 4 RCW F1.10 6-RCW-302 15 15 l RCW F1.10 6-RCW-303 18 18 RCW F1.10- 8-RCW-302 3 3 RCW F1.10- .B-RCW-311 3 3
' ' RCW F1.10 .- 8-RCW-312 5 5
- RCW F1.10 8-RCW-313 1 1
'SWS :F1.10 8-SWS-301 1 1 SWS F1.10 SWS-302 1 1 RDS' F1.40 12-RDS-101 2 2 SWS F1.40 8-SWS-302 8 8 PCS F1.40 Main Recirc Pump 10 10 i
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O O EIAMINATION CATEGORY I F-B j -- !
SYSTEM SECTION II LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REFJLIUCS ITEM NO. OF COMPONENTS M BE INSPECTED MSS F2.10 12-MSS-105 1 1 MSS F2.10 1.5-MSS-111 1 1 7 I MSS. F2.10 .1.5-MSS-112 1 1 RCS -F2.10' 2-RCS-106 1 1 FWS -F2.40 10-FWS-201 3 3 -
SWS- F2.40~ 10-SWS-302 2 2 [
MSS F2.40. 12-MSS-105 1 1 j MSS F2.40 12-MSS-201 2 2
.. MSS F2.40 1.5-MSS-111 1 1
' MSS 'F2.40 1.5-MSS-112 I 1 ,
.RCS F2340 2-RCS-106' 1 1 RCW F2.40 2-RCW-305 1 1 SWS F2.40 6-SWS-303' 3 3
.SCS F2.40 8-SCS-201. I 1
- SCS .F2.40. 8-SCS-203 1 1
< SWS F2.40' 8-SWS-301 9 9 ,
SWS F2.40 8-SWS-302 1 1 ;
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RCS F2.40 Reactor Clean-up Pump 2 2 CRD F2.40 Scram Dump Tank- 4 4 -
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PAGE 27 of 31 >
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i-FIAMINATION CATEGORY F-C 1
r-SYSTEM- SECTION XI LINE NUMBER TOTAL NUMBER' COMPONENTS SCREDULED REMARKS
' ITEM NO. OF COMPONENTS TO BE INSPECTED FWS F3.10 10-FWS-101 1 1 FWS' - F3.10 '10-FWS-201 10 10 RDS- F3.10 10-RDS-103D 2 2 MSS F3.10 12-MSS-105 2 2 MSS F3.10 12-MSS-201 2 2 RDS F3.10 12-RDS-101 8 8
.PCS F3.10 14-MRS-104 4 4 PCS F3.10 '14-MRS-105 5 5 PCS - F3.10 14-MRS-106- 3 3 PCS F3.10 17-MRS-111 5 5 PCS F3.10 17-MRS-112 3 3 PCS F3.10 17-MRS-113 3 3 FCS F3.10 17-MRS-114 5 5 MSS F3.10 1.5-MSS-110 2 2 MSS F3.10 1.5-MSS-117 10 10 PCS - F3.IO 20-MRS-121 4 4 PCS F3.10 20-MRS-122 5- 5 PCS F3.10 24-MRS-121 3 3 i: PCS F3.10 24-MRS-122 3 3
- .CRD F3.10 2-CRD-101 6- '6
-CRD F3.10 2-CRD-111 7 7 LPS F3.10 2-LPS-101 1 1
{ RCS ' F3.10 2-RCS-105 1 1 F '
RCS. F3.10. . 2-RCS-106 1 1 RCS' F3.10 2-RCS-107 1 1 RDS F3.10 2-RDS-107A 6 6
- j. LPS'. F3.10 3-LPS-102 8 8 RCS F3.10 3-RCS-101 17 17 ,1
- . .RCS F3.10 3-RCS-102 6 6 'I 7
PAGE 28 of 31
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EXAMINATION CATEGORY F-C TOTAL NUMBER COMPONENTS SCHEDULED REMARKS SYSTEM SECTION XI LINE NUMBER OF COMPONENTS TO BE INSPECTED ITEM NO.
1 1 RCS F3.10 3-RCS-103 2 2 RCS F3.10 3-RCS-108 1 1 RCS F3.10 3-RCS-121 1 1 RCS F3.10 3-RCS-122 6 6 PIS F3.10 4-CSS-101 3 3 ECS F3.10 4-ECS-103 2 2 ECS F3.10 4-ECS-104 1 1 PCS F3.10 4-MRS-141 4 4 PIS F3.10 4-RDC-101 2 2 PCS F3.10 5-MRS-131 2 2 PCS F3.10 5-MRS-132 6-CRD-201 9 9 CRD F3.10 3 6-ECS-101 3 ECS F3.10 2 2 ECS F3.10 6-ECS-102 1 1 FWS F3.10 6-FWS-204 2 2 FWS F3.10 6-FWS-205 )
2 2 FIS F3.10 6-PIS-201 l 1 1 SCS F3.10 6-SCS-101 2 2 SCS F3.10 6-SCS-102 2 2 SCS F3.10 6-SCS-201 1 1 SCS F3.10 6-SCS-202 i
3 3
! SCS F3.10 6-SCS-203 2 2 SCS F3.10 6-SCS-204 1 1 FWS F3.10 8-FWS-103 1 1 MSS F3.10 8-MSS-101 1 1 MSS F3.10 8-MSS-104 2 2 PIS F3.10 8-PIS-202 3 3 SCS F3.10 8-SCS-101 2 2 SCS F3.10 8-SCS-102 10-FWS-201 10 10 FWS F3.40 12-MSS-201 2 2 MSS F3.40 1 1 MSS F3.40 1.5-MSS-110 PAGE 29 of 31
e Om O EIAMINATION CATEGORY F-C SYSTEM SECTION II LINE NUMBER TOTAL NUMBER COMPr5ENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO PE INSPECTED CRD F3.40 2-CRD-101 6 6 CRD F3.40 2-CRD-111 7 7 RCS F3.40 2-RCS-105 1 1 LPS F3.40 3-LPS-102 1 1 RCS F3.40 3-RCS-101 6 6 RCS F3.40 3-RCS-102 2 2 RCS F3.40 3-RCS-103 1 1 ECS F3.40 4-ECS-103 1 1 PIS F3.40 4-RDC-101 2 2 CRD F3.40. 6-CED-201 1 1 PIS F3.40- 6-CSS-201 8 8 WS F3.40 6-WS-204 1 1
-FWS F3.40 6-FWS-205 2 2 PIS F3.40 6-PIS-201 2 2 SCS F3.40 6-SCS-201 2 2 SCS F3.40 6-SCS-202 1 1 SCS F3.40 6-SCS-203 3 3 SCS F3.40 6-S G-204 2 2 PIS F3.40 8-PIS-202 2 2 SCS F3.40 8-SCS-101- 1 1 SCS F3.40 8-SCS-102 2 2 FWS F3.50 .10-FWS-101 1 1 RDS F3.50 10-RDS-103D 2 2 MSS F3.50 12-MSS-105 2 2 RDS F3.50 12-RDS-101 8 8 PCS F3.50 14-MRS-101 3 3 PCS F3.50 14-MRS-102 4 4 PCS F3.50 14-MRS-103 5 5 Relief Request RR-A14 PCS F3.50 14-MRS-104 4 4 PCS F3.50 14-MRS-105: 5 5 PCS F3.50 14-MAS-106 3 3 PCS F3.50 17-MRS-111 5 5 PCS _.F3.50 17-MRS-112 3 3 PAGE 30 of 31
O EXAMINATION CATEGORY F-C ,
SYSTEM SECTION XI LINE NUMBER TOTAL NUMBER COMPONENTS SCHEDULED REMARKS ITEM NO. OF COMPONENTS TO BE INSPECTED PCS F3.50 17-MRS-113 '3 3 PCS F3.50- 17-MRS-114 5 5 MSS 'F3.50 1.5-MSS-110 1 1
.. MSS F3.50 1.5-MSS-117 10 10 Relief Request RR-A14 ;
PCS F3.50 20-MRS-121 4 4
- PCS F3.50 20-MRS-122 5 5 l PCS F3.50 24-MRS-121 3 3 PCS ~ F3.50' 24-MRS-122 3 3 7PS- F3.50' 2-LPS-101 1 1 RCS. F3.50 2-RCS-106 1 1 RCS F3.50 2-RCS-107 5 5 l LPS F3.50 3-LPS-102 7 7 Relief Request RR-A14 RCS F3.50 3-RCS-101' 11 11 6 RCS "F3.50 3-RCS-102 4 4 i RCS F3.50 3-RCS-108 2 2 i RCS F3.50 3-RCS-121 1 1 l RCS F3.50 3-RCS-122 1 1 PIS F3.50 4-CSS-101 6 6 ECS F3.50 4-ECS-103 2 2 I ECS F3.50- 4-ECS-104 2 2
.{
PCS' ~ F3.50 4-MRS-141 1 1 i PIS F3.50 4-RDC-101 2 2 i
'PCS F3.50 5-MRS-131 2 2 PCS -F3.50 5-MRS-132 2 2 . Relief Request RR-A14 ECS F3.50' 6-ECS-101 3 3 i ECS F3.50 6-ECS-102. 2. 2 SCS F3.50. 6-SCS-101 l 1 1 Relief Req = test RR-A14 SCS. F3.50 6-SCS-102 2 2 FWS' F3.50 8-FWS-103 1- 1 MSS F3.50 8-MSS-101 1 1 !
MSS .F3.50 S-MSS-104 1 1 SCS. F3.5C 8-SCS-101 2 2 Relief Request RR-A14 i PAGE 31 of 31
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t SECTION 7 ULTRADONIC CALIBRATION BLOCK L78 TING O
4 O
fa ULTRASONIC CALIBRATION BLOCKS SCHED NOTES /
No. SPECIFICATIONS MTL.
IDENTIFICATION DIMENSTIONS NO.
140 A-106 GrB CS Seamless CS (SLCS)
BR-1 Piping 18.0" O.D. x 1.625* SLCS 1/8* Hole is Void ,
6.45" O.D. x 0.726* 160 A-106 j BR-2 Piping 120 A-376-304 SS 1/8* Bole is Void BR-3 Piping 6.55" O.D. x 0.568" SLCS 6.44" O.D. r O.720* 160 A-106 BR-4 Piping 160 L-106 SLCS BR-5 Piping 8.625" O.D. x 0.942" 120 A-106 SLCS BR-6 Piping 8.84* O.D. x 0.700* SLCS 4.5" O.D. x 0.431" 120 A-106 CR-7 Piping 120 3-376 SS BR-8 Piping 4.5" O.D. x 0.437- 80 A-376 SS BR-9 Piping 3.5" O.D. x 0.30" 120 A-106 SLCS BR-10 Piping 3.5" O.D. x 0.436" 80 A-106 SLCS 2R-11 Piping 3.5" O.D. x 0.318" 120 A-376 SS BR-12 Piping 1.9" O.D. x 0.270* A-515 GR 65 CS SD CR-13 Plate 3.O* x 4.37" x 24.468" A-376 TP304 SS BR-14 Plate 4.0" x 1.5* x 14" SS "J" Weld Tool 3.875* O.D. x 0.168* A-182 TP316 BR-15B Pipe CS CS 10.8* O.D. x 0.833" 100 A-106 Gr B BR-16 Piping A-193 Gr B-7 CS UT Standard Bolt CR-17A Stud Bolt 2.25' O.D. x 13" A-193 Gr B-16 CS ACS UT Stud Bolt BR-18A Stud Bolt 1.95* O.D. x 14.64" 160 SA-336 Type 316 SS Centrifuga11y Cast CRP-1 Piping _ 14.0" O.D. x 1.365" SA-283 Gr D CS Head-Vessel Flange BRP-2 Plate 5.0" x 8.0" CCSS 12.0" O.D. x 0.858" 100 A-351-CF8M ERP-3 Piping A-351 CF8M CSS BRP-4 Plate 1.5" x 4.O* CCSS 12.75* O.D. x 1.50" 160 A-351 CFSM BRP-5 Piping SA-540, B24 CS RPV Nut BRP-6 Nut 6.5* Across Flats x 0.475" A-351-CF8M CSS Lug CSS BRP-7 Plate 1.312" x 2.5" x 6.5" SA-285 Gr D CS Stud Nut Reference Block CRP-8 Nut 6.5* O.D.
CRP-9 120 A-106 Gr B CS BRP-10 Piping 14" O.D. x 1.094* CSCL Nozzle 120 A-508 CL II Nozzle-to-Safe EnF(NSI-CSCL-12)
BRP-11 ~ NS/IR A-508 CL II CSCL BRP-12 Nozzle NSE A-182 Type 304 SS Safe End-to-Nozzle (SEN-SS-13) i CRP-13 Nozzle A-194 Gr 2H CS Nut 2.0" ID MPR Pump l
BRP-14 Nut 3.125* Across Flats A-193 Gr B16 CS MR Pump Stud CRP-15 Stud 12 Threads /in. x 2.O* O.D. A-194 Gr 2H CS Steam Erum Manway Nut CRP-16 Nut 3.125" Across Flats 2.O* I.D 80 A-312 Type 316 SS BRP-17 Piping 1.5* O.D x 0.20" CS Piping 2.0" O.D. x 0.344" 160 A-IO6 GR B BRP-18 PAGE 1 30-Apr-91
i ULTRASONIC CALIBRATION BLOCTJS SCHED NOTES NO. SPECIFICATIONS MTL.
IDENTIFICATION DIMENSTIONS NO.
A-376 Type 316 SS 80 Piping 5.0" O.D. x 0.375" 120 A-106 Gr B CS BRP-19 12" O.D. x 1.0" CS Flange-to-Nozzle SD Relief BRP-20 Piping A-105 Gr 2 BRP-21 Nozzle 5.25* O.D. x 1 125" A-240 Type 316 SS Lug Plate 0.750" x 6" x 10" A-194 Cr 2H CS Nut to SD Manwa*T BRP-22 3.375* Across Flats 2.25" I.D. 160 A-376 Type 304 SS 3" Pipe to Socket Weld BRP-23 Nut 3" Pipe to Socket Weld Piping 3* O.D. to Socket Weld A-376 Type 304 SS BRP-24A 3" O.D. to Socket Weld Mock-up of Penetration 2 BRP 24B Piping 80 Dwg SKE-62679-5-01 CSCL Mock-up of Penetration 2 BRP-25 Piping 20" O.D. x 1.250" 120 Dwg SKE-62679-5-01 SS BRP-26 Piping 6.625 0.D. x 0.562- 100 A-376 Type 304 SS BRP-27 Piping 14" O.D. x 0.750" 120 A-106 Gr B CS BRP-28 Piping 8" O.D. x 0.719" 120 A-106 Gr B CS BRP-29 Piping 6* O.D. x 0.562" 120 A-106 Gr B CS Piping 10" x 0.844* A-182 Type 304 SS Branch Connection BRP-30 BRP-31 Piping 3* O.D. x 0.750" A-516 CSCL BRP-32 Plate 2.125* Thick 80 SA-182 F316 SS BRP-33 Piping 4.50" O.D. x 0.337" SA-240 T316 SS BRP-34 Piping 24.00* O.D. x 1.25" SA-240 T316 SS BRP-35 Piping 17.00* O.D. x 1.25" 120 SA-403 T316 SS BRP-36 Piping 6.625* O.D. x 0.562" SA-193 Br B7 CS 2.25" O.D. SS BRP-37 Stud Bolt 4.50" O.D. x 0.438" 120 SA-376 T316 Piping SA-336 F316 SS BRP-38 18.00" O.D. x 1.70" Steam Drum Piping SA-516 w/304 Clad CSCL BRP-39 4.656" x 6.O' x 16.0" Scram Dump Tank Plate SA-516 Gr.70 CS BRP-40 BRP-41 Plate 1.5* x 6.0" x 8.O* SA-193 Br B7 CS Reactor vessel Stud 4.75" O.D. Rx vessel Specimen BRP-42 Stud Bolt 6.0" x 15.52" x 5.256' Thick CS w/304 Int. Clad CS 3" Inst. Nozzle (795-1A,C,D,Z) 88000002 Shell Block A-182 Type 304 SS 88000004 3" Safe End 3.5" O.D. x 2.9O" I.D. x 20.0" 173072001 Plate 3" x 2' x 12.25'
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The piping and instrument drawings contained in this section are marked to indicate which class 1 components are subject to NDE and pressure testing.
Yellow designates that the system is subject to NDE and pressure testing requirements.
Pink designates that the system requires only pressure testing (excluded from NDE).
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