ML18096A679
| ML18096A679 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 04/30/1992 |
| From: | Edmonds R, Kittle W, Lake L Public Service Enterprise Group |
| To: | |
| Shared Package | |
| ML18096A678 | List: |
| References | |
| PROC-920430, NUDOCS 9205080169 | |
| Download: ML18096A679 (895) | |
Text
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SECOND TEN (10) ~IDA.lt;:tN~RVAL Public Service En~e~Pff~~-~.£<p,:.,.p~;ta***ti.*on: * * , Nuclear Services'.-'Dc:-pai:J~!jj~Ei-~:-;-*.-,~.,~* Inspection.. servicei;. r-s*:Vr:uroiip."i"-
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- 0 Peer Review Approved by
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ISI staff Engineer Matt~.cl~:i;'~.,- t .* *,"; . . . . , Site Seryices Approved by-
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Salem Station
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9205080169-92U~O~ ( PDR- ADOCK 05000311
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VOLUME A SALEM
.GENERATING STATION UNIT 2
- INSERVICE INSPECTION PROGRAM (2ND INSPECTION INTERVAL) LONG TERM** PLAN
- 10-YEAR INSERVICE INSPECTION PROGRAM PLAN FOR SALEM NUCLEAR GENERATING STATION UNIT #2 LOCATION:POST OFFICE BOX 236 HANCOCKS BRIDGE
- NEW JERSEY 08038 OPERATORS:PUBLIC SERVICE ELECTRIC AND GAS COMPANY 80 PARK AVENUE PO BOX 570 NEWARK, NEW JERSEY 07101 REACTOR SUPPLIER: WESTINGHOUSE ELECTRIC COMPANY PITfSBURGH,PENNSYLVANIA ARCHITECT/ENGINEER: PUBLIC SERVICE ELECTRIC AND GAS COMPANY NEWARK, NEW JERSEY WESTINGHOUSE ELECTRIC COMPANY PITfSBURGH, PA.
I NRC DOCKET NUMBER: 50-272 FACILITY OPERATING LICENSE: DPR-75 CAPACITY: 3411 Mwt
. COMMERCIAL OPERATION DATE: OCTOBER 13,1981
SALEM UNIT 2 ISI PROGRAM PLAN EXECUTIVE
SUMMARY
The Code of Federal Regulations 10CFR50.55(a) (g)(4) requires the ISI Program to be updated every 10 years. For Salem Unit 2 this means a new program must be in place by the end of the current refueling outage April 5, 1992. In addition to the Code of Federal Regulations the requirement to perform Inservice Inspection comes from the Technical Specifications 4.0.5 and FSAR Section 5. d, d. 3/4 \ S. Z.,,f;Y Federal Regulations also require:
- 1. The Edition of the Code used for the update shall be 1986 Edition of ASME Section XI.
- 2. Any areas where it is impractical to comply with Code Requirements shall be submitted to the NRC as relief requests. This updated program has 8 such relief requests.
ASME Code Cases that have been accepted for use by NRC and are applicable to Salem have been incorporated. In addition to Section XI requirements PSE&G has committed to
,perform the following augmented examinations:
Examinations of Reactor Coolant Pump Flywheel Steam Gen. and Reactor Coolant Pump Bolting CS Piping Major differences between the previous ISI Program: A greater portion of the Safety Injection System will receive NDE providing greater reliability. The initial selection of component supports for visual examination is now a Sampling Program on a smaller portion of piping resulting in fewer supports required to be examined. Corrective Action on piping and component supports. Although the number of supports required to be inspected has been significantly reduced through sampling, the acceptance criteria for examining supports has become more stringent. Supports which now have loose bolted connections, improper load settings, etc. are now considered unacceptable and will require additional examinations during that outage to be scheduled .
2
- A greater portion of the Auxiliary Feed System is subject to examination providing greater reliability.
Clarification on the different types of visual examinations required. This improves implementation. Clarifications on different types of pressure tests required. 10 year Hydro tests for Class 1 and Class 2 Components can be performed as a system leakage test at normal operating pressure. This provides for a much more economical test wit~out sacrificing safety. A reduction in the amount of welds requiring relief from code requirements. Impact of this update on other programs: There is no impact on the Station IST Program. The Nuclear Department Repair Program needs to be updated to the 1986 Edition of Section XI . Corrective Action required on leaking bolted connections may have a significant impact on the Maintenance Organization. Any leaking bolted connection identified during a scheduled system pressure test will require some bolting to be removed and visually inspected. A Code Case is being pursued to limit the impact of this requirement but it has not been approved yet .
- TABLE OF CONTENTS PAGE
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 CODE APPLICABILITY . . . . . . . . . . ; . . . . . . . . . . . . . . . . . . . . . . . 2-1 3.0 EXEMPTIONS . . . . . . . . . . . . . . * . . . . . . . . . . . . . . . . . . . . . . . . . 3-1
- 3. 1 IWB - Class 1 Exemptions . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 IWC - Class 2 Exemptions . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.3 IWD - Cla.Ss 3 Exemptions . . . . . . . . . . . . . . . . . . . . . . . . 3-3 3.4 IWF - Support Exemptions . . . . . . . . . . . . . . . . . . . . . . . . 3-3 4.0 ISI CLASS 1 EXAMINATION CATEGORIES ................... .
- 4-1 5.0 ISI CLASS 2 EXAMINATION CATEGORIES .................... 5-1
- 6.0 7.0 8.0 ISI CLASS 3 EXAMINATION CATEGORIES COMPONENT SUPPORT EXAMINATION CRITERIA AUGMENTED EXAMINATION CRITERIA 6-1 7-1 8-1 9.0 SYSTEM PRESSURE TESTS CRITERIA ........................ 9-1 10.0 EXAMINATION SCHEDULING CRITERIA 10-1 11.0 ISI 10 YEAR PROGRAM AND SCHEDULE 11-1 12.0 INSERVICE INSPECTION DRAWINGS . . . . . . . . . . . . . . . . . . . . . . . 12-1 13.0 REQUESTS FOR REIJEF FROM EXAMINATION 13-1 14.0 NDE ULTRASONIC (UT) CALIBRATION BLOCKS 14-1
1.0 INTRODUCTION
This Program Plan has been prepared to fulfill the Second Interval 10-Year Inservice Inspection (ISI) requirements for the Salem Nuclear Generating Station Unit 2. This Program Plan has been written to meet the requirements specified by the Code of Federal Regulations, 10CFR50.55a(g). It has also been expanded to include augmented inservice inspection requirements which comply with commitments made to the regulatory authorities and are in addition to the requirements of ASME Boiler and Pressure Vessel Code, Section XI. The scope of the Program Plan and examinations meet the requirements outlined in Section XI of the ASME Boiler and Pressure Vessel Code, "Rules for Inservice Inspection of Nuclear Power Plant Components", 1986 Edition. Accordingly, all Class 1, 2, and 3 pressure retaining components and their supports as defined by 10CFR50.55a(g) and the Salem Updated Final Safety Analysis Report (UFSAR) will be examined to comply with ASME Code Section XI requirements to the extent practical. This plan also contains the relief requests for those components where compliance with - code requirements was found to be impossible or impractical. Repairs and Replacements to systems, components and their supports within the scope of the ISI Program will be performed in accordance with National Board NR Stamp No. 36, PSE&G Nuclear Department Repair Program Manual, which has been issued to cover repairs, replacements, and modifications. The repair program manual incorporates
. the requirements of ASME Section XI and refers to
- the design specification and construction code of the component or system, as listed in the Salem UFSAR as well as detailed specifications available at the site.
All Section XI, ASME Code Class 1, 2, and 3 water, steam, air and other fluid systems within the scope of ASME Section XI are listed in Tables 11-A, B, and C. For the purposes of this program, Nuclear Class I, II and ill are used in place of ASME Class 1, 2 and 3. Several components (including the Reactor Coolant Pump Flywheels) will receive augmented inspections and are identified in Section 8. The Safety Injection System Accumulators and associated discharge piping which are classified as Nuclear Class III on the design drawings have been voluntary upgraded to Nuclear Class II for inservice inspection requirements only, in compliance with guidance received on Unit 1 Long Term Plan and Reg. Guide 1.26. This document does not include pump and valve testing (IWP and JWV) commitments, which have been submitted to the Nuclear Regulatory Commission (NRC) under a separate document, or metal containment examination and inspection (IWE) .
- 1-1
- The. Program Plan is
- subject to revision due to plant modifications or changes in regulatory requirements.
Revisions to this program plan will be handled as follows: Minor changes - (typographical errors, editorial changes, etc.) will be identified
- by marking up a copy of the affected pages in the Master Copy, making any necessary changes in the PC database program "ISIC" and issuing the new revised sheet(s) through the Document Distribution Group (DDG).
Component selection and scheduling changes - will be tracked by marking up the Master Copy of the inspection plan during the scheduled outage. At the end of a scheduled outage, the marked up plan will be reviewed and all required changes as well as all outage exam results will be entered into "ISIC". The revised Program Plan database portion of the manual will then be submitted for a controlled distribution through DDG. Major program plan changes - (addition and deletion of components, relief requests, changes to Code requirements, etc.) will all be tracked and maintained in a separate section within the manual. This section will receive a controlled distribution as the changes are made. Upon the completion of an inspection period, the entire manual will be revised to incorporate changes in this section. The revised manual will receive the same review and approval process of the
. original manual and be distributed to all controlled copy holders.
In addition, major manual revisions will also be submitted to the NRC at the end of each period. l. The "ISIC" PC database will always be equivalent to the Master Copy and shall always reflect all program changes as they are made with the exception of component selection and scheduling changes will be changed in the database at the end of the affected outage. Upon completion of the second 10 year inspection interval, this ISI program plan will be reviewed and revised as necessary to meet the requirements of the latest approved edition of Section XI listed in 10CFR50.55a which is in effect 12 months prior to the start of the next inspection interval .
- 1-2
2.0 CODE APPLICABILITY 2.1 Operating License Date/Code Commitment The date of issuance of the operating license for Salem Nuclear Generating Station Unit 2 by the Nuclear Regulatory Commission (NRC) was October 13, 1981. The Facility operating license number is DPR-75. In accordance with 10CFR50.55a(g)(4)(ii), Inservice Inspection of components subject to examination during the second 10 year inspection interval at Salem Unit 2 will comply with the requirements of the specified Code of record referenced
- by 10CFR50.55a(b) on the date 12 months prior to the start of the second 10 year inspection interval. Based on this requirement the applicable Code is Section XI of the ASME Boiler and Pressure Vessel Code, Division 1, 1986 Edition.
2.2 Inspection Intervals The beginning of the first 10 year inspection interval for Salem Nuclear Generating Station Unit 2 was October 13, 1981. The start of the second 10 year inspection interval will commence at the end of the 6th refueling outage currently scheduled to be completed mid April, 1992. 2.3 Program Plan Scope 10CFR50.55a(g) addresses ASME Class I, II and III components and their supports. This corresponds to ASME Section XI Subsections IWA, IWB, IWC and IWF, respectively. Subsection IWE of ASME Section XI, Requirements for Class MC Components of Light-Water Cooled Power Plants, is not addressed in 10CFR50.55a(g) and therefore its requirements are not included as part of this program plan. The Salem Nuclear Generating Station Unit 2 Inservice Inspection Program complies with the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1986 Edition. References in this document to the Code, Examination Categories, Item Number, etc. refer to Section XI unless otherwise noted. Code allowed alternate examination and additional instructions identified in the following Code Cases have been incorporated: a.) ASME Code Cases Code Case No.
- N-401 N-402 2-1 Eddy Current Examinations Eddy Current Calibration Standard Material
- N-427 N-435-1 N-437 Code Cases in Inspection Plans Alternative Examination Requirements for Vessels with Wall 2 in. or less Use of Digital Readout and Digital Measurement Devices for Performing Pressure Tests N-457 Qualification Specimen Notch Location for Ultrasonic Examination of Bolts and Studs N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds N-461 Alternative Rules for Piping Calibration Block Thickness Section XI, Division 1 N-491 IWF Sampling Plan N-495 Hydrostatic Tetsing of Relief Valves
- b.)
N-498 NRC Regulatory Guides Alternate Rules for 10-Year Hydrostatic Pressure Testing for Class* 1 and 2 Systems The following NRC Regulatory Guides apply. In certain instances exceptions may have been taken to portions of a particular guide. These exceptions are identified in Appendix 3A of the Salem UFSAR. 1.14 Reactor Coolant Pump Flywheel Integrity - Rev. 1, 8/75 1.65 Materials and Inspections for Reactor Vessel Closure Studs (June* 1973) 1.83 Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes - Rev. 1 modified, 7/75
- 1.147 2-2 Inservice Inspection Code Case Acceptability - ASME Section XI, Division 1 (Latest Rev. in effect)
- c.) Licensin~ Commitments The following UFSAR commitments apply:
UFSAR Reference Application Section 5.2.1.3/4 Compliance wI 10CFR50 UFSAR Reference Application Section 5.2.8 Inservice Inspection Program d.) Requests for Relief from Examination (See Section 13 of this document.) e.) Augmented .Examinations (See Section 8 of this document) 2.4 Request for Relief from Examination In accordance with 10CFR50.55a(g)(5)(iii), where it is determined that conformance to the requirements of the Code is impractical, within the limitations of design, geometry, and materials of constructil;m of a component, specific relief from examination will be submitted to the Commission (NRC) with the necessary information and justification to support the determination(s). Applicable requests for relief for the second 10 year inspection interval are located in Section 13 of this program plan .
- 2-3
- 3.0 EXEMPTIONS Components (or parts of components) may be exempted from the volumetric, surface, or visual examination requirements of Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1.
3.1 IWB - Class 1 Exemptions IWB-1220(a) Components that are connected to the reactor coolant system and part of the reactor coolant pressure boundary and that are of such size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under normal plant operating conditions is within the capacity of makeup systems which are operable from on-site emergency power. IWB-1220(b) Component connections, piping, and associated valves of NPS 1 inch and smaller, except steam generator tubing. IWB-1220(c) Reactor vessel head ~nnections and associated piping, NPS 2 inch and smaller, made inaccessible by control rod drive penetrations. 3.2 IWC - Class 2 Exemptions
- 3.2.1 Components Within Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR) Systems (or Portions of Systems)
IWC-1221(a) Vessels, p1pmg, pumps, valves, and other components NPS 4 inch and smaller in all systems except high pressure safety injection systems. IWC-122l(b) Vessels, piping, pumps, valves, and other components NPS 1-1/2 inches and smaller in high pressure safety injection systems. IWC-122l(c) Component connections NPS 4 inch and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size in all systems except high pressure safety injection systems. IWC-122l(d) Component connections NPS 1-1/2 inches and smaller (including nozzles, sockets fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size in high pressure safety injection systems. 3-1
- IWC-1221(e) Vessels, piping, pumps, valves, other components, and. component connections of any size in statically pressurized, passive (i.e., no pumps) safety injection systems.
IWC-1221 (t) Piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions. 3.2.2 Components Within Systems (or Portions of Systems) Other Than RHR, ECC, and CHR Systems IWC-1222(a) Vessels, p1pmg, pumps, valves, and other components NPS 4 inch and smaller. IWC-1222(b) Component connections NPS 4 inch and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size. IWC-1222(c) Vessels, piping, pumps, valves, other components, and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275;sig and at a temperature equal to or less than 200 F. IWC-1222(d) Piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions.
- 3. 3 IWD - Class 3 Exemptions IWD-1220.1 Integral attachments of supports and restraints to:
( 1) piping NPS 4 inch and smaller; (2) vessels, pumps, and .valves and their connections in piping 1 NPS 4 and smaller, within the system boundaries of Examination Categories D-A, D-B, and D-C of Table IWD-2500-1 shall be exempt from the visual examination
- In piping is defined as having a cumulative inlet and a cumulative outlet pipe crou-scctional area neither of which exceeds the nominal OD cross-sectional area of the designed size.
3-2
- VT-3, except for PWR Auxiliary Feedwater Systems.
IWD-1220.2 Integral attachments of supports and restraints not exempt under IWD-1220.1 may be exempted from the visual examination VT-3 of Table IWD-2500-1 provided: (a) the components are located in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling; and (b) the components operate at a pressure of 275 psig or less and at a temperature of 200° For less. 3.4 IWF - Support Exemptions It is PUBLIC SERVICE ELECTRIC AND GAS COMPANY's position to use Code Case N-491Paragraph1230 and exempt those component supports from the examination requirements of Table IWF-2500-1 that are connected to components and items exempted from examination under IWB-1220, IWC-1220, and IWD-1220. In addition, portions of supports that are inaccessible by being buried, encased in concrete, or fire barrier insulation or enclosed by a guard pipe are also exempt from the examination requirements of Table IWF-2500-1 .
- 3-3
- 4.0 ISI CLASS 1 EXAMINATION CATEGORIES Included in this section are the requirements for the Class 1 examination categories in accordance with Section XI.
- The examination categories are used for organization purposes in preparing the inservice inspection program plan for the second 10 year inspection interval at Salem Nuclear Generating Station Unit 2.
4.1 Listed below are the Class 1 Code Examination Categories and a description of
*each item:
EXAMINATION CATEGORY EXAMINATION AREA B-A Pressure Retaining Welds in Reactor Vessel B-B Pressure Retaining Welds in Vessel Other Than Reactor Vessels B-D Full Penetration Welds of Nozzles in Vessels
- B-E
- B-F B-G-1 Pressure Retaining Partial Penetration Welds in Vessels Pressure Retaining Dissimilar Metal Welds Pressure Retaining Bolting, Greater than 2 inches m Diameter B-G-2 Pressure Retaining Bolting, 2 inches or Less in Diameter
-B-H Internal A~chment Welds for Vessels B-J Pressure Retaining Welds in Piping B-K-1 Integral Attachments for Piping, Pumps, and Valves B-L-1 Pressure Retaining Welds in Pump Casings B-L-2 Pump Casing B-M-1 Pressure Retaining Welds in Valve Bodies B-M-2 Valve Bodies
- B-N-1 Interior of Reactor Vessel 4-1
EXAMINATION CATEGORY EXAMINATION AREA B-N-2 Integrally Welded Core Support Structures and Interior Attachments to Reactor Vessel B-N-3 Removable Core Support Structures B-0 Pressure Retaining Welds in Control Rod Housings B-P All Pressure Retaining Components B-Q Steam Generator Tubing 4.2 Examination selection criteria used on Class I components during the second interval at Salem Nuclear Generating Station Unit 2 are presented in Table 11-1. The listing and schedule of components subject to examination during the inspection interval are located in Table 11 ~A .
- 4-2
- 5.0 ISI CLASS 2 EXAMINATION CATEGORIES Included in this section are the requirements for the Class 2 examination categories in accordance with Section XI.
The examination categories are used for organization purposes in preparing the examination program plan for the second 10 year inspection interval at Salem Nuclear Generating Station Unit 2.
- 5. 1 Listed below are the Class 2 Code Examination Categories and a description of each item:
EXAMINATION CATEGORY EXAMINATION AREA C-A Pressure Retaining Welds in Pressure Vessels C-B Pressure Retaining Nozzle Welds in Vessels c-c Integral Attachments for Vessels, Piping, Pumps, and Valves
- C-D
. C-F-1 Pressure Retaining Bolting Greater than 2 inches in Diameter Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping
. C-F-2 Pressure Retaining Welds in Carbon Steel or Low Alloy Steel Piping C-G Pressure Retaining Welds in Pumps and Valves C-H All Pressure Retaining Components .
- 5. 2 Examination selection criteria used on Class 2 components during the second inspection interval at Salem Nuclear Generating Station Unit 2 are presented in Table 11-1.
The listing and schedule of components subject to examination during the second ten year inspection interval are located in Table 11-B.
- 5-1
- 6.0 ISi CLASS 3 EXAMINATION CATEGORIES Included in this section are the requirements for the Class 3 examination categories in accordance with Section XI.
The examination categories are used for organization purposes in preparing the examination program plan for the second IO year inspection interval at Salem Nuclear Generating Station Unit 2. 6.1 Listed below are the Class 3 Code Examination Categories and a description of each item: EXAMINATION CATEGORY EXAMINATION AREA D-A Systems in support of Reactor Shutdown Function D-B Systems in Support of Emergency Core Cooling, Containment Heat Removal, Atmosphere Cleanup, and Reactor Residual Heat Removal D-C Systems in Support of Residual Heat Removal from Spent Fuel Storage Pool 6.2 Examination selection criteria used on Class 3 components during the second inspection interval at Salem Nuclear Generating Station Unit 2 are presented in Table 11-1. The listing and schedule of components subject to examination during the second 10 year inspection interval are located in Table 11-C .
- 6-1
- 7.0 COMPONENT SUPPORT EXAMINATION CRITERIA Included in this section are the examination requirements for Inservice Inspection of Class 1, 2, and 3 component supports. Component supports are those metal supports designed to transmit loads from the component to the load carrying building or foundation structures. This includes the attachment portion of intervening elements to pressure retaining components, integral and non-integral attachments of pressure retaining components, and integral and non-integral supports including the connection to the building structure.
Code Case N-491 and Subsection IWF-5000 are the basis for the development of this portion of the Program Plan. Table IWF-2500-1 of the 1986 Edition of Section XI redundantly categorizes the component support examination requirements e.g.: Categories F-A, F-B, and F-C have essentially identical examination requirements and acceptance standards. The Program Plan combines these three examination categories into one, designated as "F-A, Supports" as found in the Code Case N-491. The component supports selected for examination are.the supports of non-exempt, Class
. 1, 2, and 3 portions of systems required to be examined under IWB, IWC, IWD, and IWF during the inspection interval.
Inservice testing of mechanical and hydraulic snubbers will be performed under the provisions listed in the Salem Nuclear Generating Station Unit 2 Technical Specifications as identified on Relief Request# RR-Al in Section 13 of this program plan. The examination method for the component supports is designated as VT-3 (visual examination). The VT-3 visual examination consists of a determination of general mechanical and structural conditions of components and their supports, such as the verification of clearances, settings, physical displacements, loose or missing parts, debris, corroion, wear, erosion, or the loss of integrity at bolted or welded connections. The listing and schedule of the component supports subject to examination listed in Tables 11-A, 11-B or 11-C respectively, are for Class 1, 2, or 3 .
- 7-1
8.0 AUGMENTED EXAMINATION CRITERIA In addition to the requirements specified in Section XI, Salem Nuclear Generating Station Unit 2 has committed to meet the examination requirements listed in this section. The following augmented requirements apply: Subject Basis Comments
- 1. Reactor Coolant Tech. Spec. UT, MT, Per Reg.
Flywheel 4.4.10.1 Guide 1.14
- 2. Steam Generator NRC IE Bulletin 82.02 -
Manway and RC Pump Main Flange Bolting
- 3. . CS piping with wall a) Request for Additional Total welds examined thickness less than Information (TAC NO. will be 7.5% of total 3/8" 66932) dated Nov. 9, welds in this system 1988 to Salem Unit 1 per code requirements
- Plan, 2nd Interval Update b) Previously identified indications in the Containment Spray System
- 8-1
9.0 SYSTEM PRESSURE TESTS CRITERIA This section identifies the criteria for pressure testing systems subject to visual examination (VT-2) requirements of Section XI. Systems and components within the prescribed boundaries are VT-2 tested in accordance with the requirements of IWA-5000 of Section XI. Pressure testing and examinations are *conducted in conjunction with the following operations: Leakage testin& is conducted following opening and closing of a component in the system after pressurization to nominal operating pressure. Functional testin& is conducted to verify operability in systems, or components, not required to operate during normal plant operation, while under system operating pressure. Inservice testin& is conducted to perform visual examination VT-2 while the system is in service under operating pressure. Hydrostatic testin& is conducted during a plant shutdown at pressure above nominal operating pressure or system pressure for which over-pressure protection is provided . Pneumatic testin& is conducted in lieu of a hydrostatic pressure test for components within the scope of Class 2 and 3 requirements. The table forniat provided below identifies the Code classifications requiring VT-2 examination and the corresponding ex~ination category from Section XI. CLASS EXAMINATION REOU!REMENTS TEST TYPE 1 Table IWB-2500-1 Hydrostatic Testing Leakage Testing 2&3 Table IWC-2500-1 Hydrostatic Testing
- Table IWD-2500-1 Inservice Testing Functional Testing
- 9-1
- 10.0 EXAMINATION SCHEDULING CRITERIA Scheduling of nondestructive and visual examinations for the ISI program is based upon the percentage requirements of Section XI, Inspection Program B as detailed in Table 2412-1 in sections IWB, IWC, and IWD, and Table 2410-2 of Code Case N-491.
The duration of the second inspection interval is 10 years (120 months) following the completion of the first ten year inspection interval which ends mid April, 1992. The inspection interval is divided into three equal periods of 40 months each. These inspection periods are used for determination of component examination percentage criteria. The examinations are scheduled to coincide with plant's refueling outages (RFO). A standard RFO is tentatively scheduled to occur at the end of a fuel cycle which is approximately 12-18 months in length. A grace period of plus or minus one year.is granted by Section XI so that the inspection periods can correspond with refueling outages. With a RFO scheduled at 12-18 months intervals and a duration of approximately 60-90 days (estimated for scheduling only), seven RFO's are expected to occur during the second inspection interval. The ISi program divides the examinations into the seven outages in order to calculate percentage requirements . Section XI permits some component examinations to be deferred until the end of the inspection interval. Other component examinations are scheduled to meet the percentage requirements as follows: 1st Period - 16% minimum examinations completed with credit taken for no more than 34% 2nd Period - 50 % minimum examinations completed with credit taken for no more than 67% (these are cumulative percentages). 3rd Period - 100% of the ISi program examinations shall be completed by the end of this period .
- 10-1
- 11. 0 ISI 10 YEAR PROGRAM AND SCHEDULE Included in this section is the listing and proposed schedule of components subject to the examination during the second 10 year inspection interval.
- 11. 1 The Inservice Inspection Program is categorized in conjunction with Section XI examination categories for Class 1, 2, 3 and IWF components. The following items describe the information presented in the ISI Program:
A. Examination Category - An itemized listing of examination areas, items, and components required for Class 1, 2, and 3 examination categories (e.g., B-A, C-A, D-A and F-A). B. Item Number - A division within an examination category which separates the specific examination requirements (e.g., Bl.10, C2.20 and D3.30). C. Description - Identifies the components subject to examination. D. NDE Method - Identifies the nondestructive examination method required for each component. The NDE abbreviations in the Program are as follows:
- Surface (SUR)
MT PT Volumetric (VOL) UT
, Magnetic Particle Testing Dye Penetrant Testing Ultrasonic Testing RT Radiography Testing Visual (VIS)
VT Visual Examination (Surface Conditions) VT-2 Visual Examination (Leak Testing) VT Visual Examination (General Conditions) E. Schedule - Identifies the inspection period which the component will be subject to examination during the inspection interval.
- 11. 2 The following is a listing of specific system acronyms used at SALEM NUCLEAR GENERATING STATION UNIT 2 and are incorporated throughout the Inservice Inspection Program:
- 11-1
- ACRONYM AF DESCRIPTION AUX. FEEDWATER FEEDPUMP & PIPING PUMP TURBINES AUX.
BF STEAM GENERATOR MAIN FEEDWATER BIT BORON INJECTION TANK BR BORIC ACID RECOVERY CA CONTROL AIR cc COMPONENT COOLING WATER cs CONTAINMENT SPRAY SYSTEM CV CHEMICAL VOLUME AND CONTROL SYSTEM DG DIESEL GENERATORS AND AUXILIARIES DR DEMINERALIZED WATER - RESTRICTED FO DIESEL OIL STORAGE TANK & TRANSFER PUMPS FP
- FIRE PROTECTION SYSTEM FW MAIN FW PUMP & PIPING GB STEAM GENERATOR BLOWDOWN SYSTEM HT HEAD TANK MS MAIN STEAM SYSTEM PR PRESSURE RELIEF SYSTEM PS PRESSURIZINGc SYSTEM PZR PRESSURIZER RC . REACTOR VESSEL INTERNAL & LIFTING FACILITIES REACTOR COOLANT SYSTEM
- RCP REACTOR COOLANT PUMPS, PUMPS PRESSURIZER
~OTORS & LIFT 11-2
- ACRONYM.
RH SA DESCRIPTION RESIDUAL HEAT REMOVAL STATION AIR SYSTEM SF SPENT FUEL POOL. COOLING & FILTERING REFUELING CANAL & FUEL TRANSFER TUBE SPENT FUEL POOL SJ SAFETY INJECTION SYSTEM SW SERVICE WATER PUMPS & PIPING WL WASTE GAS TANK & TRANSFER PUMP WASTE LIQUID 11.3 Class 1 and 2 components selected for examination during the second inspection interval to the extent practical meet the percentage requirements of Section XI, _ Tables IWB-2500-1 and IWC-2500-1, respectively. They will be scheduled in accordance with program "B" of Subarticle IWB and IWC-2400 .
- Guidelines for weld selection in piping systems (Category B-J for Class 1 and C-F-1 and C-F-2 for Class 2) are listed in Table 11-1.
- 11-3
- TABLE 11-1 EXAMINATION SELECTION CRITERIA I. CLASS 1 EXAMINATION SELECTION CRITERIA In accordance with 10CFR50.55a(b)(2)(ii) the extent of examinations for Code Class 1 piping welds may be determined by the requirements of Tables IWB-2500 and IWB-2600, Category B-J of the 1974 Edition through and including the Summer 1975 Addenda of Section XI, or other requirements the NRC may adopt. During the first 10-year inspection interval, the initial examination selection of Class 1 welds was performed in compliance with the 1974 Summer 1975 Addenda, ASME Section XI IWB-2412 inspection program Class 1 Category B-J. Piping weld examination selection criteria for the second - ten year Inspection Interval at SALEM NUCLEAR GENERATING STATION- UNIT 2 will be based in accordance to the requirements of the 1974 Summer 1975 Edition of ASME Section XI as permitted by 10CFR50.55a. The selected welds of the first ten year Inspection Interval will be subject to re-examination during the second ten year Inspection Interval to the extent practical to meet the percentage requirements of Table IWB-2500.
The 25 % number for Code Category B-J welds is a target value to help ensure that the selected welds are uniformly distributed. The actual nuµiber of welds selected per Code Item No. may be slightly higher or lower than the target value provided that 25 % of the total number of welds in the reactor coolant piping system are selected. This total does not include welds exempted by IWB-1220. - Class 1 components other than piping welds will be selected based upon the requirements of ASME Section XI 1986 Edition, Table IWB-2500-1. As a clarification of Code Category B-G-1, Footnote 3, and B-G-2, Footnote 2, of Table IWB-2500-1-for heat exchanger, piping, pumps and valves, examinations will be limited to components selected for examination categories B-B, B-J, B-L-2, and B-M-2, respectively. Class 1 components and applicable welds selected for the second interval are listed in Table 11-A. II. CLASS 2 EXAMINATION SELECTION CRITERIA Class 2 piping weld examination selection criteria for the initial weld selection of the first inspection interval was defined by Article IWC-2411 of ASME Section XI 1974 Edition Summer 1975 addenda. This random selection was divided among the number of components of the same size and geometry in each of the multiple stream systems performing the same function. 11-4
Additionally, the required examinations were scheduled for completion by the end of the postulated service life of the plant, with an equal portion being performed in each ten year inspection interval. As required by IWC-2520 Categories C-F and C-G, piping system circumferential welds selected for examination are only those at structural discontinuities and includes 100 % of those in systems recirculating reactor coolant and 50 % of those systems containing other than reactor coolant. Class 2 examination selection criteria for the second ten year inspection interval at SALEM NUCLEAR GENERATING STATION UNIT 2 will be selected in accordance to the new requirements of Table IWC-2500-1 of the 1986 Edition of Section XI. Class 2 examination selection criteria for Code Category C-F-1 and C-F-2 shall include the following:
- 1. The welds examined shall include 7.5 %, but not less than 28 welds of the welds not exempted by IWC-1220; however, the welds not required to be examined per
- Table IWC-2500-1 shall be included in the total weld count to which the 7.5%
sampling is applied with the following exception. The Containment Spray system does not contain any piping which would be required to be examined per Table IWC-2500-1 because of wall thickness. Due to previous indications in this system and utilizing the guidance given in a Request for Additional Information (TAC NO. 66932) dated Nov. 9, 1988 received during our Unit 1 update, we will take the following action with the Containment Spray System. The total Class II weld population associated with this system, will not be included in the total non-exempt Class II count, however, we will perform an augmented examination program on this system which will consist of selecting and examining 7 .5 % of the total Class II weld population in this system utilizing the selection guidance given in Table IWC-2500-1 and our response to the Unit 1 Request for Additional Information referenced above.
- 2. The examinations will be distributed on a prorated basis to the number of non-exempt welds of a system by line size distributed among terminal ends and structural discontinuities to the degree practical; *
- 3. The welds initially selected for examination will be re-examined during the third and fourth inspection intervals over the service lifetime of the piping component.
According to ASME Section XI, Table IWC-2500-1, Examination Category C-C, Item No. C3.20, Integrally Welded Attachments to Class 2 piping where the attachment base material thickness is 3/4 inch or greater, a surface examination is required. However, this surface examination is limited to attachments of those components which are required to be examined under Examination Category C-F.
- Class 2 components and applicable welds selected for the second interval are listed in
- Table 11-B.
11-5
III . CLASS 3 EXAMINATION SELECTION CRITERIA The first 10-year interval was in accordance with the requirements of ASME Section XI 1974 Edition Summer 1975 Addenda, Subsection IWD-2400 and was later updated to the 1977 Edition 1978 Addenda. During the second 10-year inspection interval, examination selection of Class 3 integral attachment welds shall be in compliance with the requirements of Table IWD-2500-1 of ASME Section XI 1986 Edition. Inservice inspection will be conducted on integral attachment welds for supports selected for examination by Table 2500-1 of Code Case N-491. Class 3 integral attachment welds selected for the second interval are listed in Table 11-C of this document. IV. IWF SUPPORT EXAMINATION SELECTION CRITERIA IWF-2510 of Code Case N-491 states that component and piping supports to be examined will be those supports of those components that are required to be examined under IWB-2500, IWC-2500, and IWD-2500 by volumetric, surface or visual (VT-1 or VT-3) examination methods. Piping supports to be examined will be the supports of piping not exempted under IWB-1220, IWC-1220, and IWD-1220 .
- For ASME Section XI, Class 1, 2, and 3 piping, the following criteria is used to select the piping supports for examination in the second ten-year Inservice Inspection Program.
Supports attached to valves are considered piping supports. Class 1 The total percentage sample (25 %) will be comprised of supports from each Class 1 system where the individual sample size is proportional to the total number of non-exempt supports of each type and function within each system. Class 2 The total percentage sample (15 %) will be comprised of supports from each Class 2 system where the individual sample size is proportional to the total number of \ non-exempt supports of each type and function within each system. Class 3 The total percentage sample (10%) will be comprised of supports from each Class 3 system where the individual sample size is proportional to the total number of non-exempt supports of each type and function within each system .
- Component and piping supports selected for the second interval are listed in Tables 11-A, 11-B, and 11-C respectively for Class 1, 2, and 3.
11-6
- 12.0 INSERVICE INSPECTION DRAWINGS Included in this section is a listing of the drawings of systems and components subject to examination inspection and testing in the SALEM NUCLEAR GENERATING STATION UNIT 2 Inservice Inspection Program.
12.1 *Classification of systems for Class 1, 2, and 3 boundaries are documented on controlled Boundary Drawings listed in Table 12-1 and Table 12-2. Table 12-1 shows the Section XI boundary subject to in-service and hydrostatic test and visual examination of the pressure boundary (VT-2). Table 12-2 shows that portion of the Section XI piping system and components subject to either surface examination, surface and volumetric examination, or visual examination of integral attachments (VT-3). These drawings were developed as separate layers of controlled Piping and Instrumentation Drawings (P&ID's). The classification boundaries are established based upon SALEM NUCLEAR GENERATING STATION UNIT 2 interpretation of the requirements and provisions established by 10CFR50 SALEM UFSAR, and Public Service Piping Specifications No. S-C-MPOO-MGS-0001. 12.2 Inservice Inspection Sketches are supplied in order to generally depict weld, component and support locations. Each weld and component is identified with a unique identification number. Table 12-3 lists the Inservice Inspection Sketches applicable to this submittal .
- 12-1
- TABLE 12-1 INSERVICE INSPECTION BOUNDARY DRAWINGS (Piping and components subject to in-service and hydrostatic tests and visual examination of the pressure retaining boundary (VT-2).)
P&ID NO. SHEET # 'rITLE 671301-A-1352 SHT 1 OF 3 REACTOR COOLANT SYSTEM 671301-A-1352 SHT 2 OF 3 REACTOR COOLANT SYSTEM 671301-A-1352 SHT 3 OF 3 REACTOR COOLANT SYSTEM 671302-A~1352 SHT 3 OF 3 STEAM GENERATOR FEED & CONDENSATE 671303-A-1352 SHT 1 OF 6 MAIN, REHEAT & TURBINE BY-PASS STEAM 671303-A-1352 SHT 2 OF 6 MAIN, REHEAT & TURBINE BY-PASS STEAM 671317-A-1352 SHT 2 OF 2 COMPRESSED AIR 671222-A-1352 SHT 2 OF 5 FIRE PROTECTION 671325-A-1352 SHT 1 OF 2 STEAM GENERATOR DRAINS & SLOWDOWN 671325-A-1352 ' .SHT 2 OF 2 STEAM GENERATOR DRAINS & SLOWDOWN 671328-A-1352 SHT 1 OF 3 CHEMICAL & VOLUME CONTROL OPERATION
- 671328-A-1352 671328-A-1352 671331-A-1352 671331-A-1352 SHT 2 OF 3 SHT 3 OF 3 SHT 1 OF 3 SHT 2 OF 3 CHEMICAL CHEMICAL VOLUME CONTROL OPERATION VOLUME CONTROL OPERATION COMPONENT COOLING COMPONENT COOLING 671331-A-1352 SHT 3 OF 3 COMPONENT COOLING 671332-A-1352 SHT 1 OF 2 RESIDUAL HEAT REMOVAL 671332-A-1352 SHT 2 OF 2 RESIDUAL HEAT REMOVAL 671333-A-1352 SHT 1 OF 1 SPENT FUEL COOLING 671334-A-1352 SHT 1 OF 4 SAFETY INJECTION 671334-A-1352 SHT 2 OF 4 SAFETY INJECTION 671334-A-1352 SHT 3 OF 4 SAFETY INJECTION*
671334-A-1352 SHT 4 OF 4 SAFETY INJECTION 671335-A-1352 SHT 1 OF 1 CONTAINMENT SPRAY 671336-A-1352 SHT 1 OF 1 AUXILIARY FEEDWATER
- 12-2
- TABLE 12-1 (CONT'D)
INSERVICE INSPECTION BOUNDARY DRAWINGS (Piping and components subject to in-service and hydrostatic tests and visual examination of the pressure retaining boundary (VT-2) * ) P&ID NO. SHEET I TITLE 671339-A-1352 SHT 3 OF 3 WASTE DISPOSAL LIQUID 671342-A-1352 SHT 1 OF 6 SERVICE WATER NUCLEAR AREA 671342-A-13_52 SHT 2 OF 6 SERVICE WATER NUCLEAR AREA 671342-A-1352 SHT 3 OF 6 SERVICE WATER NUCLEAR AREA 671342-A-13.52 SHT 4 OF 6 SERVICE WATER NUCLEAR AREA 671342-A-1352 SHT _5 OF 6 SERVICE WATER NUCLEAR AREA 671342-A-1352 SHT 6 OF 6 SERVICE WATER NUCLEAR AREA 671344-A-1352 SHT 1 OF 5 SAMPLING 671344-A-1352 SHT 3 OF 5 SAMPLING 671347-A-1352 SHT 1 OF 3 CONTROL AIR 671347-A-1352 SHT 3 OF 3 CONTROL AIR
- 12-3
- TABLE 12-2 INSERVICE INSPECTION BOUNDARY DRAWINGS (Nuclear Class I and II piping and components subject to surface or surface and volumetric examination, and Nuclear Class III piping with integral attachments subject visual examination.)
I P&ID NO. SHEET # TITLE I 670301-A-1352 SHT 1 OF 3 REACTOR COOLANT SYSTEM 670301-A-1352 SHT 2 OF 3 REACTOR COOLANT SYSTEM 670301-A-1352 SHT 3 OF 3 REACTOR COOLANT SYSTEM 670302-A-1352 SHT 3 OF 3 STEAM GENERATOR FEED & CONDENSATE 670303-A-1352 SHT 1 OF 6 MAIN, REHEAT & TURBINE BY-PASS STEAM 670303-A-1352 SHT 2 OF 6 MAIN, REHEAT & TURBINE BY-PASS STEAM 670328-A-1352 SHT 2 OF 3 CHEMICAL & VOLUME CONTROL OPERATION 670328-A-1352 SHT 3 OF 3 CHEMICAL & VOLUME CONTROL OPERATION 670331-A-1352 SHT 1 OF 3 COMPONENT COOLING 670331-A-1352 SHT 2 OF 3 COMPONENT COOLING 670331-A-1352 SHT 3 OF 3 COMPONENT COOLING
- 670332-A-1352 670332-A-1352 670334-A-1352 670334-A-1352 SHT 1 OF 2 SHT 2 OF 2 SHT 1 OF 4 SHT 2 OF 4 RESIDUAL HEAT REMOVAL RESIDUAL HEAT REMOVAL SAFETY INJECTION SAFETY INJECTION 670334-A-1352 SHT 3 OF 4 SAFETY INJECTION 670334-A-1352 SHT 4 OF 4 SAFETY INJECTION 670335-A-1352 SHT 1 OF 1 CONTAINMENT SPRAY 670336-A-1352 SHT 1 OF 1 AUXILIARY FEEDWATER 670342-A-1352 SHT 1 OF 6 SERVICE WATER 670342-A-1352 SHT 2 OF 6 SERVICE WATER 670342-A-1352 SHT 3 OF 6 SERVICE WATER 670342-A-1352 SHT 4 OF 6 SERVICE WATER 670342-A-1352 SHT 5 OF 6 SERVICE WATER 670342-A-1352 SHT 6 OF 6 SERVICE WATER
- 12-4
- TABLE 12-3 INSERVICE INSPECTION SKETCHES SKETCH NUMBER TITLE A-1 REACTOR COOLANT SYSTEM REACTOR PRESSURE VESSEL A-2 REACTOR COOLANT SYSTEM RPV INNER RADIUS & SAFE-END A-2A REACTOR COOLANT SYSTEM RPV SAFE-END TO PIPE WELDS A-3 NOZZLES A-4 REACTOR PRESSURE VESSEL CLOSURE HEAD A-5 REACTOR PRESSURE VESSEL CONTROL ROD DRIVE HOUSINGS A-6 REACTOR COOLANT SYSTEM PRESSURIZER A-7 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 24 A-8 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 23 A-9 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 22 A-10 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 21 A-12 CHEMICAL & VOLUME CONTROL SYSTEM LETDOWN A-13 CHEMICAL & VOLUME CONTROL SYSTEM CHARGING A-15 2 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM CHARGING A-17 CHEMICAL & VOLUME CONTROL SYSTEM AUXILIARY SPRAY A-18 REACTOR COOLANT SYSTEM PRESSURE RELIEF A-19 REACTOR COOLANT SYSTEM PRESSURE RELIEF A-20 REACTOR COOLANT SYSTEM PRESSURE RELIEF A-24 REACTOR COO}'...ANT SYSTEM PRESSURE RELIEF A-25 REACTOR COOLANT SYSTEM PRESSURE RELIEF A-26 3 SHEETS REACTOR COOLANT SYSTEM PRESSURE SURGE A-28 REACTOR COOLANT SYSTEM PRESSURE SPRAY A-29 REACTOR COOLANT SYSTEM PRESSURE SPRAY A-30 REACTOR COOLANT SYSTEM PRESSURE CROSSOVER LEG A-31 REACTOR COOLANT SYSTEM PRESSURE CROSSOVER LEG A-32 REACTOR COOLANT SYSTEM PRESSURE CROSSOVER LEG A-33 REACTOR COOLANT SYSTEM PRESSURE CROSSOVER LEG A-34 REACTOR COOLANT SYSTEM PRESSURE HOT LEG A-35 REACTOR COOLANT SYSTEM PRESSURE HOT LEG A-36 REACTOR COOLANT SYSTEM PRESSURE HOT LEG A-37 REACTOR COOLANT SYSTEM PRESSURE HOT LEG A-38 REACTOR COOLANT SYSTEM PRESSURE COLD LEG A-39 REACTOR COOLANT SYSTEM PRESSURE COLD LEG A-40 REACTOR COOLANT SYSTEM PRESSURE COLD LEG A-47 REACTOR COOLANT SYSTEM PRESSURE COLD LEG A-SO REACTOR COOLANT SYSTEM PRESSURE DRAIN LINE A-53 REACTOR COOLANT SYSTEM PRESSURE DRAIN LINE A-56 REACTOR COOLANT SYSTEM PRESSURE DRAIN LINE A-61 RESIDUAL HEAT REMOVAL SYSTEM A-62 RESIDUAL HEAT REMOVAL SYSTEM A-63 RESIDUAL HEAT REMOVAL SYSTEM A-64 SAFETY INJECTION SYSTEM A-65 SAFETY INJECTION SYSTEM ACCUMULATOR DISCHARGE A-66 SAFETY INJECTION SYSTEM ACCUMULATOR DISCHARGE A-67 2 SHEETS SAFETY INJECTION SYSTEM ACCUMULATOR DISCHARGE A-69 2 SHEETS SAFETY INJECTION SYSTEM A-71 SAFETY INJECTION SYSTEM \
A-73 SAFETY INJECTION SYSTEM A-74 SAFETY INJECTION SYSTEM A-76 SAFETY INJECTION SYSTEM A-77 RESIDUAL HEAT REMOVAL SYSTEM A-79 SAFETY INJECTION SYSTEM A-80 SAFETY INJECTION SYSTEM 12-5
- TABLE 12-3 (CONT'D)
INSERVICE INSPECTION SKETCHES SKETCH NUMBER TITLE A-81 SAFETY INJECTION SYSTEM A-84 SAFETY INJECTION SYSTEM A-85 SAFETY INJECTION SYSTEM A-86 SAFETY INJECTION SYSTEM A-87 SAFETY INJECTION SYSTEM A-89 2 SHEETS SAFETY INJECTION SYSTEM A-91 SAFETY INJECTION SYSTEM A-92 SAFETY INJECTION SYSTEM A-93 SAFETY INJECTION SYSTEM A-94 2 SHEETS SAFETY INJECTION SYSTEM A-96 3 SHEETS SAFETY INJECTION SYSTEM A-98 SAFETY INJECTION SYSTEM A-99 3 SHEETS SAFETY INJECTION SYSTEM A-101 3 SHEETS SAFETY INJECTION SYSTEM A-103 2 SHEETS SAFETY INJECTION SYSTEM A-105 2 SHEETS SAFETY INJECTION SYSTEM A-107 2 SHEETS SAFETY INJECTION SYSTEM A-109 REACTOR COOLANT SYSTEM REACTOR COOLANT PUMPS B-1 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 24 B-2 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 23 B-3 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 22 B-4 REACTOR COOLANT SYSTEM STEAM GENERATOR NO. 21 B-5 CHEMICAL AND VOLUME CONTROL TANK ., B-6 EXCESS LETDOWN HEAT EXCHANGER B-7 HEAD TANKS B-8 LETDOWN HEAT EXCHANGER B-9 REACTOR COOLANT FILTER B-10 REGENERATIVE HEAT EXCHANGER B-11 RHR HEAT EXCHANGER 21 B-12 RHR HEAT EXCHANGER 22 B-13 BOILER FEED SYSTEM STEAM GENERATOR FEED B-14 BOILER FEED SYSTEM STEAM GENERATOR FEED B-16 BOILER FEED SYSTEM STEAM GENERATOR FEED B-17 BOILER FEED SYSTEM STEAM GENERATOR FEED B-18 2 SHEETS CONTAINMENT SPRAY SYSTEM B-20 3 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-23 CHEMICAL & VOLUME CONTROL SYSTEM B-24 CHEMICAL & VOLUME CONTROL SYSTEM B-25 CHEMICAL & VOLUME CONTROL SYSTEM B-26 MAIN STEAM SYSTEM MAIN STEAM HEADER B-27 MAIN STEAM SYSTEM MAIN STEAM HEADER B-28 MAIN STEAM SYSTEM MAIN STEAM HEADER B-29 MAIN STEAM SYSTEM MAIN STEAM HEADER B-30 2 SHEETS MAIN ~TEAM SYSTEM B-31 2 SHEETS MAIN STEAM SYSTEM B-32 2 SHEETS MAIN STEAM SYSTEM B-33 2 SHEETS MAIN STEAM SYSTEM B-38 MAIN STEAM SYSTEM MAIN STEAM RELIEF B-39 MAIN STEAM SYSTEM MAIN STEAM RELIEF B-40 MAIN STEAM SYSTEM MAIN STEAM RELIEF
- 12-6
- TABLE 12-3 (CONT'D)
INSERVICE INSPECTION SKETCHES SKETCH NUMBER TITLE B-41 MAIN STREAM SYSTEM MAIN STEAM RELIEF B-42 MAIN STREAM SYSTEM MAIN STEAM RELIEF B-43 2 SHEETS MAIN STREAM SYSTEM AUXILIARY MAIN STEAM B-45 MAIN STREAM SYSTEM AUXILIARY MAIN.STEAM B-46 2 SHEETS REACTOR COOLANT SYSTEM CONTAINMENT PRESSURIZER RELIEF B-49 REACTOR COOLANT SYSTEM CONTAINMENT PRESSURIZER RELIEF B-50 REACTOR COOLANT SYSTEM PRESSURIZER RELIEF B-51 REACTOR COOLANT SYSTEM PRESSURIZER RELIEF B-52 REACTOR COOLANT SYSTEM PRESSURIZER RELIEF B-53 REACTOR COOLANT SYSTEM PRESSURIZER RELIEF B-54 RESIDUAL HEAT REMOVAL SYSTEM B-55 RESIDUAL HEAT REMOVAL SYSTEM B-56 RESIDUAL HEAT REMOVAL SYSTEM B-58 RESIDUAL HEAT REMOVAL SYSTEM B-59 RESIDUAL HEAT REMOVAL SYSTEM B-60 RESIDUAL HEAT REMOVAL SYSTEM B-61 RESIDUAL HEAT REMOVAL SYSTEM B-63 RESIDUAL HEAT REMOVAL SYSTEM B-65 SAFETY INJECTION SYSTEM B-67 SAFETY INJECTION SYSTEM B-68 RESIDUAL HEAT REMOVAL SYSTEM B-70 SAFETY INJECTION SYSTEM B-72 SAFETY INJECTION SYSTEM B-74 RESIDUAL HEAT REMOVAL SYSTEM B-75 RESIDUAL HEAT REMOVAL SYSTEM B-77 SAFETY INJECTION SYSTEM B-79 2 SHEETS CONTAINMENT SPRAY SYSTEM B-80 SAFETY INJECTION SYSTEM B-84 2 SHEETS SAFETY INJECTION SYSTEM B-85 REACTOR COOLANT .SYSTEM B-86 3 SHEETS SAFETY INJECTION SYSTEM B-87 2 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-88 2 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-89 2 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-90 CHEMICAL & VOLUME CONTROL SYSTEM B-91 CHEMICAL & VOLUME CONTROL SYSTEM B-92 CHEMICAL & VOLUME CONTROL SYSTEM B-93 CHEMICAL & VOLUME CONTROL SYSTEM B-94 CHEMICAL & VOLUME CONTROL SYSTEM B-95 2 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-96 2 SHEETS CONTAINMENT SPRAY SYSTEM B-97
- CHEMICAL & VOLUME CONTROL SYSTEM B-98 CHEMICAL & VOLUME CONTROL SYSTEM B-99 CHEMICAL & VOLUME CONTROL SYSTEM B-100 CHEMICAL & VOLUME CONTROL SYSTEM B-101 CHEMICAL & VOLUME CONTROL SYSTEM B-102 3 SHEETS CHEMICAL & VOLUME CONTROL SYSTEM B-113 COMPONENT COOLING SYSTEM B-114 COMPONENT COOLING SYSTEM
- 12-7
13.0 RELIEF FROM EXAMINATION REQUEST Relief requests are included where specific requirements of Section XI are determined to be impractical. Individual relief requests are included in Section 13.2 as identified in the index. Section 13.2 is subject to change throughout the inspection interval, as examination requirements at SALEM NUCLEAR GENERATING STATION UNIT 2 are determined to be impractical. Additional, or modifications to existing relief requests will be submitted for NRC approval in accordance with 10CFR50.55a(g)(5)(iii). 13.l Relief Request Format Relief request will be in accordance with NRC guidelines. Each relief request will include the following:
- a. Component
Description:
- b. ASME XI Code Class:
- c. ASME XI Examination Requirements
- d. Basis for Relief:
- e. Alternate Examination:
13.2 Relief Request Index The following relief requests are included in this section: II REQmTNO. GENERAL DESCRIPTION II RR-Al Snubber Testing RR-A2 Welds Subject to NDE RR-A3 Reactor Vessel Nozzle-to-Vessel Welds RR-A4 Internal Surfaces of Pumps and Valves RR-AS Reactor Coolant Pump Casing Weld RR-A6 Reactor Coolant Pump Casing Bolts RR-A7 Component Supports RR-Bl Welds Subject to NDE RR-Cl Integral Attachments 13-1
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-Al Mechanical and Hydraulic Snubbers ASME CODE CLASS: ASME Section XI Class 1, 2, and 3 ASME SECTION XI EXAMINATION REQUIREMENTS: Article IWF-2000 required VT-3 inservice visual t?Xaminations of mechanical and hydraulic snubbers. Article IWF-5000 requires inservice tests to be performed on mechanical and hydraulic snubbers. The number of snubbers subject to these requirements is 135 at Salem Generating Station Unit 2.
- BASIS FOR RELIEF:
Performance of examinations and testing to the requirements of the Technical Specification meet
- some of the code requirements. However, this approach differs in the areas of examination scheduling, re-examinations, and functional testing requirements. Testing to the Technical Specification results in an increase in the overall level of plant quality and safety.
These mechanical and hydraulic snubbers were constructed and installed in accordance with the requirements of the Salem UFSAR. Documentation of fabrication and installation examinations is stored at the plant site. Subsequent to the plant going into operation these snubbers have been and continue to be visually inspected and functionally tested in accordance with Plant Technical Specifications. ALTERNATIVE EXAMINATION: Visual examinations and functional testing will be performed on Code Class 1, 2, and 3 (and other) snubbers in compliance with Salem Generation Technical Specification 3/4. 7.9 and its
- associated bases .
- 13-2
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-A2 Pressure vessel, component, and piping welds and integrally welded attachments subject to surface and/or volumetric examination. ASME CODE CLASS: ASME Section XI Class 1 welds ASME EXAMINATION REQUIREMENTS: In accordance with the examination requirements of Subsection IWB, Table IWB-2500-1 of examination categories: B-A, B-H, B- J, B-K-1, B-L-1, and B-M-1 when an examination (surface or volumetric) is performed, essentially 100% of the weld length is required to be inspected. BASIS FOR RELIEF:
- The ASME Code requires surface and volumetric examination of various Class 1 vessel and piping welds as defined in Paragraph IWB-2500.
During the inservice examinations performed at SALEM NUCLEAR GENERATING STATION UNIT 2 it has been the position of PUBLIC SERVICE ELECTRIC AND GAS COMPANY that examinations which could not be performed completely, (i.e., performed from both sides of the weld or because of component configuration or restrictions from permanent structures) the examination would be performed to the greatest extent possible and whatever limitation that existed be documented. It has also been PUBLIC SERVICE ELECTRIC AND GAS COMPANY's position that when there was a "removable" structure, (i.e., hanger, support) these items were removed, when practical, providing greater access to the component being examined. The basis for this position* is as follows: The Code of Federal Regulations, 10CFR50, "Domestic Licensing of Production and Utilization Facilities" defines those requirements which must be adhered to by the utility in order to be issued and maintain an operating license. This document states in Paragraph 55a(g) that inservice inspections shall meet the requirements of the ASME Boiler & Pressure Vessel Code. Throughout 10CFR50 numerous references are made allowing performance of these ~* inspections "to the extent practical". 13-3
/>
lI
- Some of the specific paragraphs are:
Paragraph 50.55a(g)4 states that throughout the service life of a nuclear power facility components which are classified as ASME Code Class 1, 2 and 3 shall meet the requirements of Section XI of the ASME B&PV Code, "to the extent practical within the limitations of design, geometry and materials of construction of the components". Paragraph 50.55a(2)i of this document .states that inspection of systems and components is performed "to the maximum extent practical, in accordance with generally recognized codes and standards" i.e., ASME Boiler & Pressure Code. Paragraph 50.55g(l) states that components (including supports) shall meet the requirements of (ASME Code) "to the extent practical". The ASME Code itself provides allowance for "limited" examinations. The 1986 Edition, Section V, Article 4, Paragraph T-441.3.2.4 states: "Whenever feasible, the scanning of the examination volume shall be carried out from both sides of the weld on the same surface. Where configuration of adjacent parts of the component are such that the scanning from both sides is not feasible, this fact shall be included in the report of the examination" .
- It is PUBLIC SERVICE ELECTRIC AND GAS COMPANY's practice to utilize approved technical procedures written in accordance with the applicable Section/Paragraph of the ASME Code in regard to the area/volume to be examined and the specified requirements of the examination. Recognizing that because of component design, construction, etc. there are cases when examinations can only be performed "to the greatest extent possible". In those cases, plant procedures require the documentation of the location of and cause of the limitation. PUBLIC SERVICE ELECTRIC AND GAS COMPANY has on permanent file a record of each component or weld where a limitation precluded a 100% examination during the first inspection interval.
ALTERNATIVE EXAMINATION: Weld selection for this interval has been made to minimize known limitations identified during either pre-service examination (PSn or the 1st inspection interval. When utilizing Code Case N-460 it is determined that a selected component or weld cannot be examined essentially 100% due to component geometry or physical access, the limitation
- will be recorded on the examination data sheet and included in the Owner's Summary Report as to the extent the component or weld was examined and filed with the regulatory agency. Where possible, an alternate weld or component in the same examination category and item number will be selected and examined. Attached is a listing of all known component and piping limitations where no alternate component or piping weld is available for selection .
- 13-4
Page No. 1 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE jJ_ CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 001300 B-A Bl.22 2-RPV-1443A LIMITED EXAMINATION DUE TO NONE MERIODIONAL WELD AT 270 DEG THE LOWER HEAD PENETRTAIONS 001900 B-A Bl.21 2-RPV-3443 NO EXAMINATIONS DUE TO THE NONE LOWER HEAD DISC TO PEEL LOWER HEAD PENETRATIONS SEGMENTS 002700 B-A Bl.30 2-RPV-7442 LIMITED EXAMINATION DUE TO NONE INSIDE TAPER 002800 B-A Bl.40 2-RPVCH-6446A LIMITED EXAMINATION DUE TO NONE HEAD TO FLANGE FLANGE CONFIGURATION 002900 B-D B3.90 29-RCN-1240 @ 22 DEG LIMITED EXAMINATION DUE TO NONE OUTLET NOZZLE AT 22 DEG. INTEGRAL EXTENSION 003200 B-D B3.90 29-RCN-1230 @ 158 DEG LIMITED EXAMINATION DUE TO NONE OUTLET NOZZLE AT 158 DEG INTEGRAL EXTENSION 003300 B-D B3.90 29-RCN-1220 @ 203 DEG LIMITED EXAMINATION DUE TO NONE OUTLET NOZZLE AT 203 DEG INTEGRAL EXTENSION 003600 B-D B3.90 29-RCN-1210 @ 339 DEG LIMITED EXAMINATION DUE TO NONE OUTLET NOZZLE AT 339 DEG INTEGRAL EXTENSION 011830 B-F B5.40 4-PR-1200-1 NO UT FROM THE UPSTREAM NONE NOZZLE TO SAFE-END SIDE DUE TO THE NOZZLE CONFIGURATION. lIMITED UT FROM THE DOWNSTREAM SIDE DUE TO THE SAFE-END CONFIGURATION.
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Page No. 2 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 011850 B-F B5.40 4-PS-1231-33 LIMITED OT FROM THE NONE SAFE END TO NOZZLE UPSTREAM SIDE DUE TO THE SAFE-END CONFIGURATION. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. 021100 B-D B3.140 29-STG-1230-IRS LIMITED EXAMINATION DUE TO NONE HOT LEG NOZZLE 323 STG BOLT BRACKET CONFIGURATIONS. 070000 B-F B5.130 31-RC-1240-1 NO OT FROM THE UPSTREAM NONE NOZZLE TO ELBOW SIDE DUE TO THE NOZZLE CONFIGURATION. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE ACOUSTIC PROPERTIES FO THE ELBOW. 072300 B-F B3.3 31-RC-1230-1 NO OT FROM THE UPSTREAM NONE NOZZLE TO ELBOW SIDE DUE TO THE NOZZLE CONFIGURATION. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW. 074600 B-F B3.3 31-RC-1220-1 NO OT FROM THE UPSTREAM NONE NOZZLE TO ELBOW SIDE DUE TO THE NOZZLE CONFIGURATION. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW.
Page No. 3 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 076800 B-F B5.130 31-RC-1210-1 NO OT FROM THE UPSTREAM NONE NOZZLE TO ELBOW SIDE DUE TO THE NOZZLE CONFIGURATION. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW. 080000 B-F B5.130 29-RC-1240-5 NO OT FROM THE UPSTREAM NONE ELBOW TO NOZZLE SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. 081200 B-F B5.130 29-RC-1230-5 NO OT FROM THE UPSTREAM NONE ELBOW TO NOZZLE SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. 082200 B-F B5.130 29-RC-1220-5 NO OT FROM THE UPSTREAM NONE ELBOW TO NOZZLE SIDE DUE TO THE ACOUSTIC PROPERTIES OF THE ELBOW. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION 083300 B-F B5.130 29-RC-1210-5 NO OT FROM THE UPSTREAM NONE ELBOW TO NOZZLE SIDE DUE .TO THE ACOUSTIC PROPERTIES OF THE ELBOW. NO OT FROM THE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION.
Page No. 4 I . \ COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 039200 B-K-1 Bl0.10 2-CV-1275-27PS-172 NO OT DUE TO THE PIPE AND NONE PIPE SUPPORT PIPE SUPPORT CONFIGURATIONS. PT PERFORMED AS A SUPPLEMENTAL EXAMINATION. LIMITED PT DUE TO INACCESSIBILITY. 061700 B-K-1 Bl0.10 4-PS-1231-llPS-1 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 061800 B-K-1 Bl0.10 4-PS-1231-llPS-2 LIMITED EXAINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 061900 B-K-1 Bl0.10 4-PS-1231-llPS-3 LIMITED EXAINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 062000 B-K-1 Bl0.10 4-PS-1231-llPS-4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 066000 B-K-1 Bl0.10 4-PS-1211-12PS-1 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 066100 B-K-1 Bl0.10 4-PS-1211-12PS-2 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 066150 B-K-1 Bl0.10 4-PS-1211-12PS-3 LIMITED EXAINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY 066175 B-K-1 Bl0.10 4-PS-1211-12PS-4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILILTY
Page No. 5 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE It CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 165790 B-K-1 Bl0.10 8-SJ-1262-6PS-4 THRU 7 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. THE PIPE SUPPORTS ARE INSIDE THE PENETRATION. 168240 B-K-1 Bl0.10 8-SJ-1252-9PS-4 THRU 7 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. THE PIPE SUPPORTS ARE INSIDE THE PENETRATION. 177200 B-K-1 Bl0.10 4-SJ-1294-SPL-1 THRU 4 NO EXAMINATION DUE STO NONE PIPE LUG INACCESSIBILITY. THE LUGS ARE INSIDE THE PENETRATION 179750 B-K-1 ( Bl0.10 4-SJ-1282-23PL-1-THRU 4 NO EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. THE LUGS ARE INSIDE THE PENETRATION. 181800 B-K-1 Bl0.10 4-SJ-1272-23PL-1 THRU 4 NO EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. THE LUGS ARE INSIDE THE PENETRATION. 250100 B-G-1 B6.180 24-PMP-BOLTS 1-24 NO EXAMINATION OF BOLTS 1 NONE AND 15 DUE TO INACCESSIBILITY 250200 B-G-1 B6.180 23-PMP-BOLTS 1-24 NO EXAMINATION OF BOLTS 3, NONE 13, 14, AND 23 DUE TO INACCESSIBILITY. 250300 B-G-1 B6.180 22-PMP-BOLTS 1-24 NO EXAMINATION OF BOLTS 7, NONE 21, AND 22 DUE TO INACCESSIBILITY.
Page No. 6 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 250400 B-G-1 B6.180 21-PMP-BOLTS 1-24 NO EXAMINATION OF BOLTS 1 NONE AND 5 DUE TO INACCESSIBILITY 275358 c-c C3.10 2-RHE-IVS-1&2 NO EXAMINATION DUE TO NONE VESSEL SUPPORT INACCESSIBILITY. 275359 c-c C3.10 2-RHE-5VS-1&2 NO EXAMINATION DUE TO NONE VESSEL SUPPORT INACCESSIBILITY 275360 c-c C3.10 2-RHE-9VS-1&2 NO EXAMINATION DUE TO NONE VESSEL SUPPORT INACCESSIBILITY 275400 C-B c2.21 21-RHRHEX-OUT NO UTOW OR UT45T DUE TO NONE THE WELD CROWN CONFIGURATION. LIMITED EXAMINATION FROM THE VESSEL SHELL DUE TO THE ADJACENT FLANGES AND VESSEL SUPPORTS 275410 C-B C2.21 21-RHRHEX-IN NO UTOW OR UT45T DUE TO NONE THE WELD CROWN CONFIGURATION. LIMITED EXAINATION FROM THE VESSEL SHELL DUE TO THE ADJACENT FLANGES AND VESSEL SUPPORTS.
Page No. 7 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 275420 C-B C2.2 22-RHRHEX-IN NO EXAMINATION FROM THE NONE UPSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. LIMITED EXAMINATION FRO THE VESSAEL SHELL DUE TO THE ADJACENT FLANGES AND VESSEL SUPPORTS. 275430 C-B C2.21 22-RHRHEX-OUT NO EXAMINATION FROM THE NONE DOWNSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. LIMITED EXAMINATION FROM THE VESSEL SHELL DUE TO THE ADJACENT FLANGES AND VESSEL SUPPORTS~ 330040 c-c C3.20 14-BF-2241-2PS-.2 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. THE SUPPORT WELD IS INSIDE THE PENETRATION. 330060 c-c C3.20 14-BF-2241-3PS-1 THRU 4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. 330130 c-c C3.20 14-BF-2241-7PL-1&2 LIMITED EXAMINATION DUE TO NONE PIPE LUG ADJACENTE PERMANENTLY INSTALLED RESTRAINT RING. 330140 c-c C3.20 14-BF-2241-BPL-1 THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY
Page No. 8 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE fl CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 330160 c-c C3.20 14-BF-2241-8PL-ll THRU 18 NO EXAMINATION FROM THE NONE PIPE LUG UPSTREAM SIDE DUE TO THE PIPE SUPPORT. LIMITED EXAMINATION DUE TO INACCESSIBILITY. 330300 c-c C3.20 14-BF-2231-3PL-1&2 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY 330360 c-c C3.20 14-BF-2231-6PL-1 THRU 4 WELD IS WITHIN AN NONE PIPE LUG ENCAPSULATION AND IS INACCESSIBLE. 330420 c-c C3.20 14-BF-2231-lOPS-1 THRU 4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. 330660 c-c C3.20 14-BF-2221-PS-2 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. SUPPORT IS INSIDE PENETRATION. 330690 c-c C3.20 14-BF-2221-SPS-1 THRU 4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. 330860 c-c C3.20 14-BF-2221-14PS NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY CAUSED BY PIPE RESTRAINT. 330950 c-c C3.20 14-BF-2211-3PS-1 NO EXAMINATION DUE TO NONE PIPE SUPPORT' INACCESSIBILITY. PIPE SUPPORT IS INSIDE PENETRATION. 330960 c-c C3.20 14-BF-2211-3PS-2 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT ADJACENT PIPE LUGS.
Page No. 9 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 331160 c-c CJ.20 14-BF-2211.;..14PS NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY CAUSED BY PIPE RESTRAINT. 381010 c-c CJ.20 32-MS-2241-lPS-1 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. WELD IS INSIDE SLEEVE. 381020 c-c CJ.20 32-MS-2241-lPS-2 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. 381048 C-F-2 C5.52 32-MS-2241-2LD-O LIMITED EXAMINATION DUE TO NONE LONGITUDINAL INTERFERENCE. 381110 c-c CJ.20 32-MS-2231-lPS-1 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. WELD IS INSIDE SLEEVE. 381152 C-F-2 C5.52 32-MS-2231-JLU-1 LIMITED EXAMINATION DUE TO NONE LONGITUDINAL - THE SUPPORT STRUCTURE INTERFERENCE. 381154 C-F-2 C5.52 32-MS-2231-JLU-U LIMITED EXAMINATION DUE TO NONE LONGITUDINAL THE WHIP RESTRAINT AND SUPPORT STRUCTURE INTERFERENCE. 381210 c-c CJ.20 32-MS-2221-lPS-1 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. WELD IS INSIDE SLEEVE. ' 381230 c-c CJ. 2_0 32-MS-2221-lPS-3 THRU 12 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY.
Page No. 10 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION 'AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 381248 C-F-2 C5.52 32-MS-2221-2LD-O LIMITED EXAMINATION DUE TO NONE LONGITUDINAL THE WHIP RESTRAINT INTERFERENCE. 381310 c-c C3.20 32-MS-2211-lPS-1 NO EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. WELD IS INSIDE SLEEVE. 381348 C-F-_2 C5.52 32-MS-2211-2LD-O LIMITED EXAMINATION DUE TO NONE LONGITUDINAL THE WHIP RESTRAINT INTERFERENCE. 381420 c-c C3.20 30-MS-2241-2PL-1 THRU 3 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. 381480 c-c C3.20 30-MS-2241-5PL-3 THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. .* 381530 c-c C3.20 30-MS-2241-8PL-1&2 NO EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. 381620 c-c C3.20 30-MS-2231-2PL-1 THRU 3 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. 381670 c-c C3.20 30-MS-2231-SPL-3 THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. 381695 c-c C3.20 30-MS-2231-8PL-1&2 NO EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. PIPE LUGS
'ARE INSIDE FLOOR PENETRATION.
381700 c-c C3.20 30-MS-2231-8PL-3 THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY.
Page No. 11 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION J81800 C-F-2 C2.1 JO-MS-2221-1 NO EXAMINATION FROM THE NONE NOZZLE TO REDUCER UPSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. LIMITED EXAMINATION FROM THE DOWNSTREAM SIDE DUE TO GAMMA PLUG WELD. J81820 C-C CJ.20 JO-MS-2221-2PL-1 THRU J NO EXAMINATION FROMTHE NONE PIPE LUG UPSTREAM SIDE DUE TO THE PIPE HANGER CONFIGURATION. J81880 C-C CJ.20 JO-MS-2221-SPL-J THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. J81920 C-E-1 CJ. 20 JO-MS-2221-7PL-1&2 NO EXAIMINATIO DUE TO NONE PIPE LUG INACCESSIBILITY. PIPE LUGS ARE INSIDE FLOOR PENETRATION. J82000 C-F-2 C5.51 JO-MS-2211-1 NO EXAMINATION FROM THE NONE NOZZLE TO REDUCER UPSTREAM SIDE DUE TO THE NOZZLE CONFIGURATION. LIMITED EXAMINATION FROM THE DOWNSTREAM SIDE DUE TO GAMMA PLUG WELD. J82020 C-C CJ.20 JO-MS-2211-2PL-1 THRU J LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. J82070 C-C CJ. 20 JO-MS-2211-SPL-J THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. J82120 C-C CJ.20 JO-MS-2211-8PL-1&2 NO EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY.
Page No. 12 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 382130 c-c C3.20 30-MS-2211_.8PL-3 THRU 10 LIMITED EXAMINATION DUE TO NONE PIPE LUG INACCESSIBILITY. 384655 c-c C3.20 6-MS-2231-6PS-1 THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384730 c-c CJ.20 6-MS-2231-12PS-1 THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384765 c-c C3.20 6-MS-2231-15PS-l THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384775 c-c C3.20 6-MS-2231-16PS-1&2 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384780 c-c C3.20 6-MS-2231-16PS-3&4 WELD IS WITHIN AN NONE PIPE SUPPORT EXCAPSULATION AND IN INACCESSIBLE. 384795 c-c C3.20 6-MS-2231-17PS-l .THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER 384805 c-c C3.20 6-MS-2231-18PS-1 THRU 16 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384850 c~c C3.20 6-MS-2231-21PS WELD IS WITHIN AN NONE PIPE SUPPORT ENCAPSULATION AND IS INACCESSIBLE.
Page No. 13 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 384895 c-c C3.20 6-MS-2231-25PS 1 THRU 16 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384935 c-c C3.20 6-MS-2231-29PS-l THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 384957 c-c C3.20 6-MS-2231-32PS-l THRU 4 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 385515 c-c C3.20 6-MS-2211-lJAPS-l THRU 8 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 385600 c-c C3.20 6-MS-2211-21PS-l THRU 8 LIMITED EXAMINATION DUE TO NONE - PIPE SUPPORT THE PROXIMITY OF THE PIPE SUPPORTS TO EACH OTHER. 573055 c-c C3.20 12-RH-2252-5PL-1 THRU 6 NO EXAMINATION ON LUGS 2, NONE PIPE LUG. 3, 4, AND 5 DUE TO INACCESSIBILITY. LUGS ARE WITHIN THE PENETRATION. 573225 c-c C3.20 12-RH-2252-22PS-1 LIMITED EXAMINATION FROM NONE PIPE SUPPORT THE DOWNSTREAM SIDE DUE TO THE ADJACENT PIPE RESTRAINT. 573387 c-c C3.20 12-RH-2252-38PS-3 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT THE PENETRATION CONFIGURATION.
Page No. 14 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE ll CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 577085 c-c C3.20 8-RH-2215-8PS-1&2 LIMITED EXAMINATION DUE TO NONE PIPE SUPPORT INACCESSIBILITY. 010900 B-B 82.11 *2-PZR-CIRC DUH LIMITED EXAMINATION FROM NONE SHELL D TO UPPER HEAD THE UPSTREAM AND DOWNSTREAM SIDES DUE TO THE PRESSURIZER SUPPORT RINGS AND THE WELD PADS. 021000 B-D B3.140 29-STG-1240-IRS INSPECTION LIMITED DUE TO NONE INLET NOZZLE IRS PERMANENTLY WELDED INSULATION SUPPORT. 002800 B-A Bl.40 2-RPVCH-6446A
- FLANGE CONFIGURATION NONE HEAD TO FLANGE LIMLTS EXAMINATION OF THE CODE REQUIRED VOLUME.
010100 B-B B2.12 2-PZR-CIRC-LHA CLOSE PROXIMITY OF NONE LOWER HEAD TO SHELL A INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE
' REQUIRED VOLUME.
020600 B-D B3.140 31-STG-1240-IRS CLOSE PROXIMITY OF NONE OUTLET NOZZLE IRS INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 020800 B-D B3.2 31-STG-1220-IRS INSULATION SUPPORT LIMITS NONE INNER RADIUS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME.
Page No. 15 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE l CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 020900 B-D B3.2 31-STG-1210-IRS CLOSE PROXIMITY OF NONE OUTLET NOZZLE IRS INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE ,, REQUIRED VOLUME. 021000 B-D BJ.2 29-STG-1240-IRS CLOSE PROXIMITY OF NONE INLET NOZZLE IRS INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 021200 B-D BJ.2 29-STG-1220-IRS INSULATION SUPPORT LIMITS NONE INNER RADIUS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 021300 B-D B3.2 29-STG-1210PIRS CLOSE PROXIMITY OF NONE INLET NOZZLE IRS INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 275040 c-c Cl.3 2-CVCT-2VS-1&2 INACCESSIBILITY OF WELD NONE VESSEL SUPPORT LIMITS THE EFFECTIVE EXAMINATION OF CODE' VOLUME. 275210 C-A Cl. l 2-LHEX-1 FLANGE CONFIGURATION AND NONE FLANGE TO SHELL THE CLOSE PROXIMITY OF 4 11 SENSING LINES LIMITS THE EFFECTIVE EXAMINATION OF CODE VOLUME.
Page No. 16 COMPONENT AND PIPING LIMITATIONS
SUMMARY
ASME ITEM EXAMINATION AREA/ ALTERNATE ll CATEGORY NO. EXAMINATION REQUIREMENT BASIS FOR RELIEF EXAMINATION 275230 C-A Cl.1 2-RCF-1 ADJACENT FLANGES AND NONE UPPER HEAD TO FLANGE SUPPORTS LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 380050 C-E-1 C2.5 34-MS-2241-2PS-1&2 ADJACENT PIPE SUPPORT NONE PIPE SUPPORT LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME. 573383 C-3-1 C2.5 12-SJ-2252-38PS-1&2 PIPING ORIENTATION LIMITS NONE. PIPE SUPPORT EXAMINATION OF CODE REQUIRED VOLUME. 020900 B-D B3.2 31-STG-1210-IRS CLOSE PROXIMITY OF NONE OUTLET NOZZLE IRS INSULATION SUPPORT BRACKET LIMITS EFFECTIVE EXAMINATION OF CODE REQUIRED VOLUME.
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-A3 Reactor Vessel Nozzle-to-Vessel Welds: Outlet Reactor Nozzle Inlet Reactor Nozzle ASME CODE CLASS: ASME Section XI, Class 1, Full Penetration Welds of Nozzles in Vessels, Code Category B-D, Item Number B3.90 and B3.100 _ ASME EXAMINATION REQUIREMENTS: Relief is being requested from the requirement to perform a volumetric examination of a minimum of 25 % but not more than 50 % (credited) of the reactor vessel nozzle-to-vessel welds and inner radii by the end of the first inspection period of the second inspection interval .
- BASIS FOR RELIEF:
During the second refueling outage (2RFO) at SALEM NUCLEAR GENERATING STATION UNIT 2 the 29" outlet nozzle-to-vessel welds were examined using Southwest Research automated remote reactor inspection device. The two welds examined met the 25 % requirement of Table IWB-2500-1. No recordable indications were detected. During the last refueling outage (6RFO) the core barrel was removed from the SALEM NUCLEAR GENERATING STATION UNIT 2 reactor vessel and all nozzle-to-vessel welds were examined, using the automated inspection device, as required by Table IWB-2500-1, including the 29" outlet reactqr vessel nozzle-to-vessel welds. As in 2RFO no 'reportable indications were detected in the 29" outlet nozzles. ALTERNATIVE EXAMINATION: It is requested to defer the re-examination of all the reactor vessel nozzle-to-vessel welds to the last inspection period of the interval .
- 13-5
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-A4 Pressure Retaining Pump and Valves ASME CODE CLASS: ASME Section XI, Class 1, Pumps and Valves ASME EXAMINATION REQUIREMENTS: Relief is being requested from the visual (VT-3) examination requirements of Table IWB-2500-1 Examination Categories B-L-2 and B-M-2 Items Nos. B12.20 and Bl2.50. BASIS FOR RELIEF: The internal visual examination of various pumps and valves throughout the second ten year interval requires an undue hardship to disassemble and reassemble components that are functioning in a satisfactory manner and may cause additional undue problems in the operation of a particular component. Salem Nuclear Generating Station Unit 2 feels that this requirement is impractical since several components are disassembled and reassembled as a result of repair, rework or replacement. Later editions of Section XI now require this examination to coincide with maintenance activities. ALTERNATIVE EXAMINATION: It is Public Service Electric and Gas Company's position that whenever an applicable pump or valve is disassembled, a visual (VT-3) examination of the internal surfaces* will be performed in the course of any maintenance, repair or replacement.
- 13-6
RELIEF REQUEST RR-A5 COMPONENT DESCRIPTION: Reactor Coolant Pump Casing Weld ASME CODE CLASS: Class 1 ASME EXAMINATION REQUIREMENTS: Volumetric examination of the pump casing weld. Examination may be limited to welds in at least one pump in each group of pumps performing similar functions. BASIS FOR RELIEF: NRC Safety Evaluation dated August.12, 1981 granted relief from the volumetric requirements of the Code Item B5.6 Category B-L-1 such that only surface examination on the external surface of the weld is required to be performed. As such, PSE&G has been evaluating new techniques for volumetrically examining the pump casing welds. Such a technique exists using the Miniature Linear Accelerator (MINAC) which was. built under an EPRI sponsored program. This equipment has been made available to other utilities, and currently constitutes the only method available for the volumetric examination of reactor coolant pump casing welds. This examination has been performed at several sites. The MINAC examination was performed at Ginna in the spring of 1981, at Point Beach Unit 1 in the fall of 1981, at Turkey Point Unit 3 early in 1982, and at H.B. Robinson Unit 2 later in 1982. No problems with welds were found at any of the sites. - The successful performance of this volumetric examination using the MINAC demonstrates that the method is capable of satisfying ASME Section XI examination requirements. Based on the following information, however, PSE&G does not plan to use this technique. The volumetric examination method is radiographic and is performed by placing the MINAC inside the pump -casing and placing film on the outside of the pump. To perform the examination, the pump must be completely disassembled. This disassembly is far beyond that performed for normal maintenance. In addition, insulation must be removed from the exterior of the pump casting. The performance of the examination has shown that there is a relatively high radiation exposure associated with it. The total exposure associated with insulation removal, disassembly examination, and reassembly of the pump has averaged about 40 man-rem .
- The pumps casing examinations are also not justified fr~m a cost/benefit perspective. The pump disassembly, examination and reassembly is estimated to cost $750,000.
13-7
- Based on the preceding factors, PSE&G does not consider it justifiable to disassemble these pumps solely for the purpose of performing these examinations. Therefore, relief is required from performing volumetric examination of the pump casing weld.
ALTERNATIVE EXAMINATION: A surface examination on the external surface of the weld~
- 13-8
- COMPONENT DESCRIPTION:
Reactor Coolant Pump Casing Bolts RELIEF REQUEST RR-A6 ASME CODE CLASS: Class 1 ASME EXAMINATION REQUIREMENTS: Volumetric examination of all reactor coolant pump casing bolts is required per Examination Category B-G-1. BASIS FOR RELIEF: There are as many as four (4) bolts on each of four (4) reactor coolant pumps which are inaccesible to examination due to interference from structural members and pump oil piping. ALTERNATIVE EXAMINATION:
- If a pump(s) is required to be disassembled for maintenance, the bolts will be volumetrically examined at that time .
- 13-9
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-A7 Component Supports ASME CODE CLASS: ASME Section XI, Class 1, 2, and 3 ASME EXAMINATION REQUIREMENTS: Subsection IWF of the 1986 Edition of Section XI requires examination of 100 % of the non-exempt piping supports and at least one component support where similar components in design, size, function, and service exist during each inspection interval. BASIS FOR RELIEF: Approval to use Code Case N-491 is requested. Code Case N-491 uses statistical sampling to select component and piping supports for examinations.
- Statistical sampling recognizes that a 100 % inspection is not necessary to ensure overall system integrity. These statistical principles are implemented through sampling plans which are recognized and permitted throughout Subsections IWB and IWC where 100% of applicable pipe and component welds are not required to be examined. Similarly, through Code Case N-491 and the 1990 addendum to ASME Section XI, component and piping support integrity providing to overall system integrity can be verified adequately by examination of a percentage of the total support population.
During the first inspection interval PSE&G conducted a comprehensive inspection program of all applicable safety related component supports. Any operational or transient condition which would cause the degradation of a support would be expected to affect more than one support, thus making sampling an appropriate method to ensure overalf component support integrity of a system. Implementation of the Code Case N-491 Sampling Plan will significantly reduce manpower requirements to support scaffold erection and insulation removal as well as performance of examinations. Based on the sampling percentages permitted by Code Case N-491, it is expected that occupational radiation exposure involving Class 1, 2, and 3 piping systems could be reduced by as much as 75 %, 85 %, and 90 %, respectively, over that expected if a 100 % inspection were
- performed .
13-10
- ALTERNATIVE EXAMINATION:
Examination of ASME Class 1, 2, and 3 supports will be performed in accordance with Code Case N-491.
- 13-11
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-Bl Pressure Vessel, Component, and Piping Welds and integrally welded attachments subject to surface and/or volumetric examination. ASME CODE CLASS: ASME Code Class 2 Welds ASME EXAMINATION REQUIREMENTS: In accordance with Subsection IWC, Table IWC-2500-1 of examination categories C-A, C-F-1, and C-F-2 when an examination of a weld (surface or volumetric) is performed essentially 100% of the weld length is required to be inspected. BASIS FOR RELIEF: The ASME Code requires surface and volumetric examination of various Class 2 vessel and
- piping welds as defined in Paragraph IWC-2500 .
During the inservice examinations performed at SALEM NUCLEAR GENERATING STATION UNIT 2 it has been the position of PUBLIC SERVICE ELECTRIC AND GAS COMPANY that examinations which could not be performed completely, (i.e., performed from both sides of the weld or because of component configuration or restrictions from permanent structures) the examination would be performed to the greatest extent possible and whatever limitation that existed be documented. It has also been PUBLIC SERVICE ELECTRIC AND GAS COMPANY's position that when there was a "removable" structure, (i.e. hanger, support) these items were removed, when practical, providing greater access to the component being examined. The basis for this position is as follows: The Code of Federal Regulations, 10CFR50, "Domestic Licensing of Production and Utilization Facilities" defines those requirements which must be adhered to by the utility in order to be issued and maintain an operating license. This document states in Paragraph 55a(g) that inservice inspections shall meet the requirements of the ASME Boiler & Pressure Vessel Code. Throughout 10CFR50 numerous references are made allowing performance of these inspections "to the extent practical". Some of the specific paragraphs are:
- 13-12
- Paragraph 50.55a(g)4 states that throughout the service life of a nuclear power facility components which are classified as ASME Code Class 1, 2, and 3 shall meet the requirements of Section XI of the ASME B&PV Code, "to the extent practical within the limitations of design, geometry and materials of construction of the components".
Paragraph 50.55a(2)i of this document states that* inspection of systems and components is performed "to the maximum extent practical, in accordance with generally recognized codes and standards" i.e. ASME Boiler & Pressure Code. Paragraph 50.55g(l) states that components (including supports) shall meet the requirements of (ASME Code) "to the extent practical". The ASME Code itself provides allowance for "limited" examinations. The 1986 Edition, Section V, Article 4, Paragraph T-441.3.2.4 states: "Whenever feasible, the scanning of the examination volume shall be carried out from both sides of the weld on the same surface. Where configuration of adjacent parts of the component are such that scanning from both sides is not feasible, this fact shall be included in the report of the examination". It is PUBLIC SERVICE ELECTRIC AND GAS COMPANY's practice to utilize approved technical procedures written in accordance with the applicable Section/Paragraph of the ASME
- - Code in regard to the area/volume to be examined and the specified requirements of the examination. Recognizing that because of component design, construction, etc. there are cases where examinations can only be performed "to the greatest extent practical". In those cases, plant procedures require the documentation of the location of and cause of the limitation.
PUBLIC SERVICE ELECTRIC AND GAS COMPANY has on permanent file a record of each component or weld where a limitation precluded a 100 % examination during the first ten year inspection interval. ALTERNATIVE EXAMINATION: When utilizing Code Case N-460 it is determined that a selected component or weld cannot be examined essentially 100% due to part geometry or physical access the limitation will be recorded on the examination data sheet and included in the Owner's summary report as to the extent the component or weld was examined and filed with the regulatory agency .
- 13-13
- COMPONENT DESCRIPTION:
RELIEF REQUEST RR-Cl Integral Attachments to Pressure Retaining Components ASME CODE CLASS: ASME Section XI, Class 3 ASME EXAMINATION REQUIREMENTS: In accordance with subsection IWD, Table IWD-2500-1 of examination categories D-A, D-B and D-C all non-exempt integral attachments are required to receive a VT-3 visual examination. BASIS FOR RELIEF: The integral attachments listed below are buried within wall sleeves making them totally inaccessible for visual examination.
- Support#
SWPA-19 SWPA-35 SWPA-42 2-SWWA-6K2 Support# SWPA-24 SWPA-38 2P-SWA-WS10 2-SWWA-6K3 Support# SWPA-26 SWPA-41 2-SWWA-6K2 2-SWWA-6P2 2-SWWA-6P3 A2-SWG-319 A2-SWG-320 A2-SWG-321 A2-SWG-322 ALTERNATIVE EXAMINATION: The structural integrity of the lines associated with these integral attachments will be verified during the VT-2 examination which will be conducted during system pressure tests. In addition, adjacent non-exempt pipe supports on these lines are included as part of the IWF subsection of this program and are subject to being examined during the interval as part of the sample selection process .
- 13-14
- 14.0 NDE ULTRASONIC (UT) CALIBRATION BLOCKS Piping (UT) Calibration Blocks used during the second ten-year interval are listed as follows. PSE&G reserves the right to change segments of piping used dependent upon pipe size, (e.g. small diameter piping ~ 4 inches may be 360 degree sections).
Calibration reflectors will be installed per the requirements of Appendix Ill and VI. As an alternative other calibration block designs may be used per the provisions of IWA-2240 .
- 14-1
- Table 14-1 ULTRASONIC CALIBRATION STANDARDS Calibration Standard 12-SS-160-1.283-21-SAM 10-SS-160-1.119-22-SAM
, 8-SS-XX-.860-23-SAM 8-SS-10-.140-24-SAM 6-SS-160-. 764-25-SAM 6-SS-40-.287-26-SAM 4-SS-XXS-.689-27-SAM 4-SS-160-.553-28-SAM R 3-SS-160-. 451-30-SAM 12-SS-40-.377-31-SAM 8-SS-80-.484-32-SAM 8-SS-20-.268-33-SAM 16-CS-160-1. 610-34-SAM 14-CS-80-. 760-35-SAM
- 14-CS-120-1.14-36-SAM
- 2.3122-SS-37-SAM 2-SS-160-.330-39-SAM 5-CSCL-42-SAM 3-SS-XX-. 600-43-SAM 8-SS-40-. 330-44-SAM 6-SS-10-.140-45-SAM 8-CS-160-. 906-46-SAM 32-CS-XX-1. 618-47-SAM 6-SS-160-.553-28-SAM 3-CS-80-.432-49-SAM 7-CSCL-50-SAM PL-3-CS-51-SAM 1 l-CSCL-53-SAM 9-CSCL-54-SAM 9.563-7-8-CS-59-SAM 7-1.125-8-CS-60-SAM 1lXl1-CSCL-62-SAM 6-SS-XX-1.5-64-SAM PL- l .5-CS-65-SAM 12-SS-SOS-.500-66-SAM 6-SS-80-.432-68-SAM 6-CS-160-. 718-69-SAM 4.5-. 75-8-CS-70-SAM 4.5-SS-XX-1.0-71-SAM IR-CSCL-73-SAM 14-2
- Calibration Standard PL-CSCL-3.0-76-SAM 14-SS-140-1.25-77'-SAM 14-SS-160-1.40-78-SAM 14-SS-40-.438-79-SAM - I 6-SS-XX-1.5-64-SAM '
4.125-2.563-8-MSIV-81-SAM 2.563-8-12MSIV-8259-SAM CRD-SS/IN-. 625-83-SAM IR-CSCL-84-SAM PL-SS-. 750-85-SAM 9.5-SS-X-. 750-86-SAM 1.125-8-S-CS-87-SAM PL-CS-4.5-88-SAM
- 1. 625-8-S-CS-89-SAM 1.250-8-S-SS-90-SAM 1.125-8-S-SS-91-SAM 1.625-8-N-CS-92-SAM 1.250-8-N-SS-93-SAM 1.125-8-N-SS-94-SAM
- 27 .5 lD-CCSS-2. 75-95-SAM 14-SS-40-.250-96-SAM PL-SS-. 250-97-SAM
- 1. 625-8-S-CS-98-SAM 1.375-8-S-CS-99-SAM 1.250-8-B-CS-100-SAM R 1.125-8-N-CS-101-SAM 1.125-8-S-CS-103-SAM 1.375-8-N-CS-104-SAM 1.250-8-N-CS-105-SAM 1.250-8-S-CS-106-SAM 1.125-8-B-CS-107-SAM 1.250-8-SCS-108-SAM VF/S-CSCL-109-SAM N/S-CSCL-110-SAM
- 14-3
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DATE: 04/24/92 SALEH NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION
SUMMARY
FOR THE SECOND INTERVAL SECTION XI
SUMMARY
NUMBER NUMBER OF COMPONENTS OF COMPONENTS SCHEDULED I COMPLETED I PERCENTAGE OF REQUIREMENT CATEGORY COMPONENTS REQUIRED FIRST PERIOD SECOND PERIOD THIRD PERICO B-A 28 28 7 0 25% 9 0 57% 12 0 100% B-B 13 5 2 0 40% 0 0 40% 3 0 100% B-D 34 38 10 0 26% 10 0 52% 18 0 100% B-E 4 4 0 0 0% 0 0 0% 4 0 100% B-F 22 22 7 0 31% 7 0 63% 8 0 100% B-G-1 13 20 6 0 30% 5 0 55% 9 0 100% B-G-2 80 36 22 0 61% 8 0 83% 6 0 100% B-H 1 1 0 0 0% 1 0 100% 0 0 100% B-J 1298 326 107 0 32% 82 58% 136 0 100% B-K-1 30 27 9 0 33% 9 0 66% 9 0 100% B-L-1 4 1 0 0 0% 0 0 0% 0 100% B-L-2 4 0 0 0% 0 0 0% 0 100% B-H-2 30 10 4 0 40% 3 0 70% 3 0 100% B-N-1 1 1 0 0 0% 0 0 0% 0 100% B-N-2 2 2 0 0 0% 0 0 0% 2 0 100% B-N-3 0 0 0% 0 0 0% 0 100% B-0 6 8 0 12% 0 25% 6 0 100% B-P 24 73 24 0 32% 24 0 65% 25 0 100% C-A 66 24 8 0 . 33% 8 0 66% 8 0 100% C-B 24 8 3 0 37% 3 0 75% 2 0 100% c-c 160 74 25 0 33% 24 0 66% 25 0 100% C-F-1 1189 93
- 28 0 30% 20 0 51% 45 0 100%
C-F-2 407 31 8 .0 25% 11 0 61% 12 0 100% C-H 66 132 33 0 25% 33 0 50% 66 0 100% D-A 54 30 9 0 30% 9 0 60% 12 0 100% D-B 166 106 33 0 31% 34 0 63% 39 0 100% D-C 2 4 0 25% 0 50% 2 0 100% F-A 1727 349 116 0 33% 120 0 67% 113 0 100% RG .1.14 5 12 4 0 33% 4 0 66% 4 0 100% TOTAL 5461 *** 467 0 31% 426 60% 573 0 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8
- CLASS 1 SECTION XI
SUMMARY
CATEGORY B-A - PRESSURE RETAINING IJELDS IN REACTOR VESSEL ASHE TABLE IIJB
# OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOO SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B1.11 CIRCUMFERENTIAL SHELL IJELDS VOLUMETRIC REACTOR PRESSURE 3 3 0 0 0 0 3 0 VESSEL B1.12 LONGITUDINAL SHELL IJELDS VOLUMETRIC REACTOR PRESSURE 9 9 0 0 0 0 9 0 VESSEL B1 .21 CIRCUMFERENTIAL HEAD IJELDS VOLUMETRIC REACTOR PRESSURE 2 0 0 0 0 0 VESSEL Bottom Hd. on B1.22 MERIDIONAL HEAD IJELDS VOLUMETRIC REACTOR PRESSURE 12 12 6 0 6 0 0 0 VESSEL Bottom Hd. on 81.30 SHELL-TO-FLANGE IJELD VOLUMETRIC REACTOR PRESSURE 2 0 0 0 0 VESSEL B1.40 HEAD-TO-FLANGE IJELD VOLUMETRIC REACTOR PRESSURE 0 0 0 0 0 SURFACE VESSEL 81.51 REPAIR IJELDS-BELTLINE REGION N/A CATEGORY TOTAL 282870 9 0 12 0 25% 57% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\,/ER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL 8 CODE EDITION: CLASS 1 SECTIO~ XI
SUMMARY
TABLE HJB CATEGORY B-B - PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B2.11 PRESSURIZER-CIRCUMFERENTIAL VOLUMETRIC PRESSURIZER 5 2 0 0 0 0 2 0 SHELL-TO-HEAD WELDS B2.12 PRESSURIZER-LONGITUDINAL N/A SHELL-TO-HEAD WELDS 82.21 PRESSURIZER-CIRCUMFERENTIAL N/A HEAD WELDS B2.22 PRESSURIZER-MERIDIONAL HEAD N/A WELDS 82.31 STEAM GENERATORS (PRIMARY N/A SIDE)-CIRCUMFERENTIAL HEAD WELDS 82.32 STEAM GENERATORS (PRIMARY N/A SIDE)-MERIDIONAL HEAD WELDS B2.40 STEAM GENERATORS (PRIMARY VOLUMETRIC STEAM GENERATORS 4 0 0 0 0 0 SIDE)-TUBESHEET-TO-HEAD WELD 82.51 HEAT EXCHANGERS (PRIMARY N/A SIDE)-HEAD-CIRCUMFERENTIAL HEAD WELDS 82.52 HEAT EXCHANGERS (PRIMARY N/A SIDE)-HEAD-MERIDIONAL HEAD WELDS B2.60 HEAT EXCHANGERS (PRIMARY N/A SIDE)-SHELL-TUBESHEET-TO-HEAD WELDS 82.70 HEAT EXCHANGERS (PRIMARY N/A SIDE)-SHELL-LONGITUDINAL WELDS B2.80 HEAT EXCHANGERS (PRIMARY N/A SIDE)-SHELL-TUBESHEET-TO-SHELL WELDS
DATE: 04/24/92 SALEH NUCLEAR PO\IER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-B - PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIP~ION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS CATEGORY TOTAL 9 3 0 0 0 2 0 33% 33% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-D - FULL PENETRATION WELDS OF NOZZLES IN VESSELS (INSPECTION PROGRAM A> ASHE # OF COMPONENTS SEC. XI # OF NO.. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 83.10 REACTOR N/A VESSEL-NOZZLE-TO-VESSEL WELDS 83.100 REACTOR VESSEL-NOZZLE INSIDE N/A RADIUS SECT I ON
- 83. 110 PRESSURIZER-NOZZLE-TO-VESSEL N/A WELDS 83.120 PRESSURIZER-NOZZLE INSIDE VOLUMETRIC PRESSURIZER 6 6 2 0 2 0 2 0 RADIUS SECT! ON 83.130 STEAM GENERATORS (PRIMARY N/A SIDE)-NOZZLE-TO-VESSEL WELDS 83.140 STEAM GENERATORS (PRIMARY VOLUMETRIC STEAM GENERATORS 8 8 2 0 2 0 4 0 SIDE)-NOZZLE INSIDE RADIUS SECTION 83.150 HEAT EXCHANGERS (PRIMARY N/A SIDE)-NOZZLE-TO-VESSEL WELDS 83.160 HEAT EXCHANGERS (PRIMARY N/A SIDE)-NOZZLE INSIDE RADIUS SECTION 83.20 REACTOR VESSEL-NOZZLE INSIDE N/A
-:1 RADIUS SECTION 83.30 PRESSURIZER-NOZZLE-TO-VESSEL WELDS N/A I
83.40 PRESSURIZER-NOZZLE INSIDE N/A RADIUS ~ECTION 83.50 STEAM GENERATORS (PRIMARY N/A SIDE)-NOZZLE-TO-VESSEL WELDS 83.60 STEAM GENERATORS (PRIMARY N/A SIDE)-NOZZLE INSIDE RADIUS SECTION .
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IIJB CATEGORY B-D - FULL PENETRATION IJELDS OF NOZZLES IN VESSELS (INSPECTION PROGRAM A) ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOO SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 83.70 HEAT EXCHANGERS (PRIMARY N/A SIDE)-NOZZLE-TO-VESSEL IJELDS B3.80 HEAT EXCHANGERS (PRIMARY N/A SIDE)-NOZZLE INSIDE RADIUS SECTION 83.90 REACTOR VOLUMETRIC REACTOR PRESSURE 8 12 2 0 2 0 8 0 All Nozzles, VESSEL-NOZZLE-TO-VESSEL IJELDS VESSEL to 50% lst period, Remai by end of Interval CATEGORY TOTAL 22 26 6 0 6 0 14 0 23% 46% 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-E - PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 84.10 PARTIAL PENETRATION WELDS N/A B4. 11 VESSEL NOZZLES N/A 84.12 CONTROL ROD DRIVE NOZZLES VISUAL REACTOR PRESSURE 0 0 0 0 0 VESSEL 84.13 INSTRUMENTATION NOZZLES VISUAL REACTOR PRESSURE 0 0 0 0 0 VESSEL 84.20 PRESSURIZER-HEATER PENETRATION VISUAL PRESSURIZER 0 0 0 0 o* WELDS CATEGORY TOTAL 3 3 0 0 0 0 3 0 0% 0% 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-F - PRESSURE RETAINING D"ISSIMILAR METAL WELDS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS BS.10 REACTOR VESSEL-NOZZLE-TO-SAFE VOLUMETRIC REACTOR PRESSURE 8 8 2 0 2 0 4 0 All welds. May END BUTT WELDS >= 4 INCHES SURFACE VESSEL coincide with NOMINAL PIPE SIZE Cat. B-D BS .100 HEAT EXCHANGERS-NOZZLE-TO-SAFE N/A END BUTT WELDS >= 4 INCHES NOMINAL PIPE SIZE BS.110 HEAT EXCHANGERS-NOZZLE-TO-SAFE N/A END BUTT WELDS < 4 INCHES NOMINAL PIPE SIZE BS.120 HEAT EXCHANGERS-NOZZLE-TO-SAFE N/A END SOCKET WELDS BS.130 PIPING-DISSIMILAR METAL BUTT VOLUMETRIC PRESSURIZING SYSTEM 0 0 0 0 0 All welds WELDS >= 4 INCHES NOMINAL PIPE SURFACE SIZE VOLUMETRIC REACTOR COOLANT 8 8 2 0 3 0 3 0 SURFACE SYSTEM ITEM TOTAL 9 9 3 0 3 0 3 0 33% 66% 100% BS.140 PIPING-DISSIMILAR METAL BUTT N/A WELDS < 4 INCHES NOMINAL PIPE SIZE BS. 1SO PIPING-DISSIMILAR METAL SOCKET N/A WELDS BS.20 REACTOR VESSEL-NOZZLE-TO-SAFE N/A END BUTT WELDS < 4 INCHES NOMINAL PIPE SIZE BS.30 REACTOR VESSEL-NOZZLE-TO-SAFE N/A END SOCKET WELDS
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-F - PRESSURE RETAINING DISSIMILAR METAL WELDS ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOO SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS BS.40 PRESSURIZER-NOZZLE-TO-SAFE END VOLUMETRIC PRESSURIZER s s 2 0 2 0 0 BUTT WELDS >= 4 INCHES NOMINAL SURFACE PIPE SIZE BS.SO PRESSURIZER-NOZZLE-TO-SAFE END N/A BUTT WELDS < 4 INCHES NOMINAL PIPE SIZE BS.60 PRESSURIZER-NOZZLE-TO-SAFE END N/A SOCKET WELDS BS.70 STEAM GENERATOR-NOZZLE-TO-SAFE N/A END BUTT WELDS >= 4 INCHES NOMINAL PIPE SIZE BS.BO STEAM GENERATOR-NOZZLE-TO-SAFE N/A END BUTT WELDS < 4 INCHES NOMINAL PIPE SIZE BS.90 STEAM GENERATOR-NOZZLE-TO-SAFE N/A END SOCKET WELDS CATEGORY TOTAL 222270 7 0 8 0 31% 63% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\IER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CLASS 1 SECTION XI
SUMMARY
TABLE lllB CATEGORY B-G PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN DIAMETER ASME SEC. XI
# OF COMPONENTS
# OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B6.10 REACTOR VESSEL-CLOSURE HEAD SURFACE REACTOR PRESSURE 3 0 0 0 All nuts, NUTS VESSEL Deferral Permissable B6.100 STEAM GENERATORS-FLANGE N/A SURFACE, WHEN CONNECTION DISASSEMBLED B6.110 STEAM GENERATORS-NUTS, N/A BUSHINGS, AND WASHERS B6.120 HEAT EXCHANGERS-BOLTS AND N/A STUDS B6.130 HEAT EXCHANGERS-FLANGE N/A SURFACE, WHEN CONNECTION DISASSEMBLED B6.140 HEAT EXCHANGERS-NUTS, N/A BUSHINGS, AND WASHERS B6.150 PIPING-BOLTS AND STUDS N/A 86.160 PIPING-FLANGE SURFACE, WHEN N/A CONNECTION DISASSEMBLED B6.170 PIPING-NUTS, BUSHINGS, AND N/A WASHERS B6.180 PUMPS-BOL.TS AND STUDS VOLUMETRIC REACTOR COOLANT 4 4 0 0 2 0 All bolts & s VISUAL PUMPS Limit to comp sched. by Cat B-L-1 Defer Permis B6.190 PUMPS-FLANGE SURFACE, WHEN VISUAL REACTOR COOLANT 4 4 0 0 0 3 0 All flange su CONNECTION DISASSEMBLED PUMPS Limit to c~
sched. by Cat B-L-1 Defer Permis
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B*G-1
- PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN DIAMETER ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCR I PT! ON EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B6.20 REACTOR VESSEL-CLOSURE STUDS, REACTOR PRESSURE 0 0 0 0 0 0 0 All studs, IN PLACE VESSEL Deferral Permissable B6.200 PUMPS-NUTS, BUSHINGS, AND N/A WASHERS B6.210 VALVES-BOLTS AND STUDS N/A B6.220 VALVES-FLANGE SURFACE, WHEN N/A CONNECTION DISASSEMBLED B6.230 VALVES-NUTS, BUSHINGS, AND N/A WASHERS B6.30 REACTOR VESSEL-CLOSURE STUDS, VOLUMETRIC REACTOR PRESSURE 3 0 0 0 All studs, WHEN REMOVED SURFACE VESSEL Deferral Permissable B6.40 REACTOR VESSEL-THREADS IN VOLUMETRIC REACTOR PRESSURE 3 0 0 0 All threads i FLANGE VESSEL flange, Defer Permissable B6.50 REACTOR VESSEL-CLOSURE VISUAL REACTOR PRESSURE 3 0 0 0 All washers &
WASHERS,.BUSHINGS VESSEL Bushings, Deferral Permissable B6.60 PRESSURIZER-BOLTS AND STUDS N/A B6.70 PRESSURIZER-FLANGE SURFACE, N/A WHEN CONNECTION DISASSEMBLED B6.80 PRESSUR.IZER*NUTS, BUSHINGS, N/A AND WASHERS B6.90 STEAM GENERATORS-BOLTS AND N/A STUDS CATEGORY TOTAL 13 20 6 0 9 0 5 0 30X 55% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE I\.IB CATEGORY B-G PRESSURE RETAINING BOLTING, 2 INCHES AND LESS IN DIAMETER ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B7.10 REACTOR VESSEL-BOLTS, STUDS, N/A AND NUTS B7.20 PRESSURIZER-BOLTS, STUDS, AND N/A NUTS B7.30 STEAM GENERATORS-BOLTS, STUDS, VISUAL STEAM GENERATORS 8 8 8 0 0 0 0 0 AND NUTS B7.40 HEAT EXCHANGERS-BOLTS, STUDS, N/A AND NUTS B7.50 PIPING-BOLTS, STUDS, AND NUTS VISUAL PRESSURE RELIEF 3 0 0 0 0 0 All bolts, st SYSTEM and nuts Limi COfll> sched. b Cat. B-J, B-L B-M-1 VISUAL SAFETY INJECTION 12 3 2 0 0 0 0 SYSTEM ITEM TOTAL 15 4 3 0 0 0 0 75% 100% 100% B7.60 PUMPS-BOLTS, STUDS, AND NUTS N/A B7.70 VALVES-BOLTS, STUDS, AND NUTS VISUAL VALVES 56 23 11 0 6 0 6 0 All bolts, st and nuts Limi COfll> sched. b Cat. B-J, B-L B-M-1 B7.80 CRD HOUSINGS-BOLTS, STUDS, AND N/A NUTS CATEGORY TOTAL 79 35 22 0 7 0 6 0 62% 82% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\IER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B*H - INTEGRAL ATTACHMENTS FOR VESSELS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS
~8.10 REACTOR VESSEL-INTEGRALLY N/A WELDED ATTACHMENTS B8.20 PRESSURIZER-INTEGRALLY WELDED N/A ATTACHMENTS B8.30 STEAM GENERATORS-INTEGRALLY N/A WELDED ATTACHMENTS B8.40 HEAT EXCHANGERS-INTEGRALLY N/A WELDED ATTACHMENTS CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CATEGORY B-J - PRESSURE RETAINING WELDS IN PIPING ASME SEC. XI CLASS 1 SECTION XI
SUMMARY
TABLE IWB
# OF
# OF COMPONENTS NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPT.ION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B9.11 CIRCUMFERENTIAL PIPE WELDS >= VOLUMETRIC CHEMICAL AND VOLUME 6 2 0 0 0 0 At least 25%
4 IN. NOMINAL PIPE SIZE SURFACE CONTROL SYSTEM the welds VOLUMETRIC PRESSURE RELIEF 36 9 4 0 3 0 2 0 SURFACE SYSTEM VOLUMETRIC PRESSURIZING SYSTEM 63 16 5 0 5 0 6 0 SURFACE VOLUMETRIC REACTOR COOLANT 40 7 2 0 0 4 0 SURFACE SYSTEM VOLUMETRIC REACTOR PRESSURE 8 5 2 0 0 SURFACE VESSEL VOLUMETRIC RESIDUAL HEAT 55 14 7 0 2 0 5 0 SURFACE REMOVAL SYSTEM VOLUMETRIC SAFETY INJECTION 331 82 29 0 23 0 30 0 SURFACE SYSTEM ITEM" TOTAL 539 135 50 0 36 48 0 37% 64% 100% B9.12 LONGITUDINAL PIPE WELDS >= 4 N/A IN. NOMINAL PIPE SIZE B9.21 CIRCUMFERENTIAL PIPE WELDS < 4 SURFACE CHEMICAL AND VOLUME 67 17 2 0 0 0 15 0 At least 25% IN. NOMINAL PIPE SIZE CONTROL SYSTEM the welds SURFACE PRESSURE RELIEF 33 11 5 0 3 0 3 0 SYSTEM VOLUMETRIC REACTOR COOLANT 10 2 0 0 0 0 SURFACE SYSTEM SURFACE SAFETY INJECTION 28 7 2 0 0 0 5 0 SYSTEM
- ITEM TOTAL 138 37 10 27%
0 3 0 35% 24 0 100%
DATE: 04/24/92 SALEM NUCLEAR PO\lER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IW CATEGORY B-J - PRESSURE RETAINING WELDS IN PIPING ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOO SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B9.22 LONGITUDINAL PIPE WELDS < 4 N/A IN. NOMINAL PIPE SIZE 89.31 BRANCH CONNECTION WELDS >= 4 VOLUMETRIC REACTOR COOLANT 12 4 3 0 0 0 0 At least 25% IN. NOMINAL PIPE SIZE SURFACE SYSTEM the welds I I' VOLUMETRIC RESIDUAL HEAT 0 0 0 0 0 SURFACE REMOVAL SYSTEM ITEM TOTAL 13 5 3 0 0 0 2 0 60% 60% 100% B9.32 BRANCH CONNECTION WELDS < 4 SURFACE PRESSURIZING SYSTEM 2 0 0 0 0 0 IN. NOMINAL PIPE SIZE
- SURFACE SURFACE REACTOR COOLANT SYSTEM RESIDUAL HEAT REMOVAL SYSTEM 26 6 0 0 0
2 0 0 0 4 0 0 0 SURFACE SAFETY INJECTION 9 2 0 0 0 0 SYSTEM ITEM TOTAL 38 10 3 o* 2 0 5 0 30% 50% 100% 89.40 SOCKET WELDS SURFACE CHEMICAL AND VOLUME 44 11 0 2 0 8 0 At least 25% CONTROL SYSTEM the welds SURFACE REACTOR COOLANT 20 4 0 0 0 3 0 SYSTEM VOLUMETRIC SAFETY INJECTION 474 120 37 .. 0 39 0 44 0 SURFACE SYSTEM ITEM TOTAL 538 135 39 0 41 0 55 0
- CATEGORY TOTAL 1266 322 105 28%
0 82 59% 134 100% 0
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-K INTEGRAL ATTACHMENTS FOR PIPING, PUMPS, AND VALVES ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 810.10 PIPING-INTEGRALLY WELDED SURFACE CHEMICAL AND VOLUME 0 0 0 0 0 25% of piping ATTACHMENTS CONTROL SYSTEM int. attach. SURFACE PRESSURE RELIEF 0 0 0 0 0 SYSTEM SURFACE PRESSURIZING SYSTEM 0 0 0 0 0 SURFACE RESIDUAL HEAT 0 0 0 0 0 REMOVAL SYSTEM SURFACE SAFETY INJECTION 22 19 6 0 7 0 6 0 SYSTEM ITEM TOTAL 26 23 7 0 9 0 7 0 30% 69% 100% 810.20 PUMPS-INTEGRALLY WELDED N/A ATTACHMENTS 810.30 VALVES-INTEGRALLY WELDED N/A ATTACHMENTS CATEGORY TOTAL 26 23 7 0 9 0 7 0 30% 69" 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CLASS 1 SECTION XI
SUMMARY
CATEGORY B*L PRESSURE RETAINING WELDS IN PUMP CASING ASME SEC. XI TABLE IWB
# OF COMPONENTS
# OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 812.10 PUMPS-PUMP CASING WELDS VOLUMETRIC REACTOR COOLANT 4 0 0 0 0 0 PUMPS CATEGORY TOTAL 4 0 0 0 0 1 0 0% 0% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\.IER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-L PUMP CASINGS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 812.20 PUMPS-PUMP CASINGS N/A CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR PO\IER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-M PRESSURE RETAINING WELDS IN VALVE BODIES ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B12.30 VALVES-VALVE BODY WELDS < 4 N/A INCHES NOMINAL PIPE SIZE B12.40 VALVES-VALVE BODY WELDS >= 4 N/A INCHES NOMINAL PIPE SIZE CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE HIB CATEGORY B-M VALVE BODIES ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER
- 2ND PER 3RD PER COMMENTS B12.50 VALVES-VALVE BODIES EXCEEDING VISUAL PRESSURE RELi E'F 3 0 0 0 0 0 One valve eac 4 INCHES NOMINAL PIPE SIZE SYSTEM group of valv if diss. for reasons Defer Permis VISUAL RESIDUAL HEAT 5 3 0 0 0 REMOVAL SYSTEM VISUAL SAFETY INJECTION 22 6 2 0 2 0 2 0 SYSTEM ITEM TOTAL 30 10 4 0 3 0 3 0 40% 70% 100%
- CATEGORY TOTAL 30 10 4 0 40%
3 0 70% 3 0 100%
DATE: 04/24/92 SALEM NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-N INTERIOR OF REACTOR VESSEL ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B13.10 REACTOR VESSEL-VESSEL INTERIOR N/A CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE 11.IB CATEGORY B-N INTEGRALLY WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSEL ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B13.20 REACTOR VESSEL CBWR)-INTERIOR N/A ATTACHMENTS WITHIN BELTLINE REGION B13.30 REACTOR VESSEL CBWR)-INTERIOR N/A ATTACHMENTS BEYOND BELTLINE REGION B13.50 REACTOR VESSEL CPWR)-INTERIOR VISUAL REACTOR PRESSURE 0 0 0 0 0 ATTACHMENTS WITHIN BELTLINE VESSEL REGION B13.60 REACTOR VESSEL CPWR)-INTERIOR VISUAL REACTOR PRESSURE 0 0 0 0 0 ATTACHMENTS BEYOND BELTLINE VESSEL REGION
- CATEGORY TOTAL 2 2 0 0 0%
0 0% 0 2 0 100%
DATE: 04/24/92 SALEH NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE 1\.IB CATEGORY B-N REMOVABLE CORE SUPPORT STRUCTURES ASHE # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B13.40 REACTOR VESSEL CB\.IR)-CORE N/A SUPPORT STRUCTURE B13.70 REACTOR VESSEL (P\.IR)*CORE N/A SUPPORT STRUCTURE CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE BIB CATEGORY B PRESSURE RETAINING WELDS IN CONTROL ROO HOUSINGS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B14.10 REACTOR VESSEL-WELDS IN SURFACE REACTOR PRESSURE 6 8 0 0 6 0 CONTROL ROO DRIVE HOUSINGS VESSEL CATEGORY TOTAL 6 8 0 0 6 0 12% 25% 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN /OR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE BIB CATEGORY B-P - ALL PRESSURE RETAINING COMPONENTS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B15.10 REACTOR VESSEL-SYSTEM LEAKAGE N/A TEST B15.11 REACTOR VESSEL-SYSTEM N/A HYDROSTATIC TEST B15.20 PRESSURIZER-SYSTEM LEAKAGE VISUAL PRESSURIZER 5 2 0 2 0 0 TEST B15.21 PRESSURIZER-SYSTEM HYDROSTATIC VISUAL PRESSURIZER 0 0 0 0 0 TEST B15.30 STEAM GENERATORS-SYSTEM VISUAL STEAM GENERATORS 5 2 0 2 0 0 LEAKAGE TEST B15.31 STEAM GENERATORS-SYSTEM VISUAL STEAM GENERATORS 0 0 0 0 0 HYDROSTATIC TEST B15.40 HEAT EXCHANGERS-SYSTEM LEAKAGE N/A TEST B15.41 HEAT EXCHANGERS-SYSTEM N/A HYDROSTATIC TEST B15.50 PIPING-SYSTEM LEAKAGE TEST VISUAL CHEMICAL AND VOLUME 5 2 0 2 0 0 Each refuel in CONTROL SYSTEM outage VISUAL REACTOR COOLANT 5 2 0 2 0 0 SYSTEM VISUAL RESIDUAL HEAT 5 2 0 2 0 0 REMOVAL SYSTEM VISUAL SAFETY INJECTION 5 2 0 2 0 0 SYSTEM ITEM TOTAL 4 20 8 0 8 0 4 0 40% 80% 100% B15.51 PIPING-SYSTEM HYDROSTATIC TEST VISUAL CHEMICAL AND VOLUME 0 0 0 0 0 One test at o CONTROL SYSTEM near encl of inspection interval
DATE: 04/24/92 SALEM NUCLEAR POIJER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL 8 COOE EDITION: CLASS 1 SECTION XI
SUMMARY
TABLE BIB CATEGORY B-P - ALL PRESSURE RETAINING COMPONENTS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 815.51 PIPING-SYSTEM HYDROSTATIC TEST VISUAL REACTOR COOLANT 0 0 0 0 0 One test at o SYSTEM near end of inspection interval VISUAL RESIDUAL HEAT 0 0 0 0 0 REMOVAL SYSTEM VISUAL SAFETY INJECTION 0 0 0 0 0 SYSTEM ITEM TOTAL 4 4 0 0 0 0 4 0 0% 0% 100% 815.60 PUMPS-SYSTEM LEAKAGE TEST VISUAL REACTOR COOLANT 5 2 0 2 0 0 Each refeul in PUMPS outage 815.61 PUMPS-SYSTEM HYDROSTATIC TEST VISUAL REACTOR COOLANT 0 0 0 0 0 One test at o PUMPS near end of inspection interval 815.70 VALVES-SYSTEM LEAKAGE TEST VISUAL CHEMICAL AND VOLUME 3 16 6 0 6 0 4 0 Each refuel in CONTROL SYSTEM outage VISUAL REACTOR COOLANT 5 2 0 2 0 0 SYSTEM ITEM TOTAL 4 21 8 0 8 0 5 0 38% 76% 100% 815. 71 VALVES-SYSTEM HYDROSTATIC TEST VISUAL CHEMICAL AND VOLUME 3 3 0 0 0 0 3 0 One test at o CONTROL SYSTEM near end of inspection interval VISUAL REACTOR COOLANT 0 0 0 0 0 SYSTEM ITEM TOTAL 4 4 0 0 0 0 4 0 0% 0% 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 1 SECTION XI
SUMMARY
TABLE IWB CATEGORY B-P - ALL PRESSURE RETAINING COMPONENTS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS CATEGORY TOTAL 22 67 22 0 22 0 23 0 32% 65% 100%
' I
DATE: D4/24/92 SALEM NUCLEAR PO\JER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CATEGORY B-Q - STEAM GENERATOR TUBING ASME SEC. XI CLASS 1 SECTION XI
SUMMARY
TABLE HIB
# OF COMPONENTS
# OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B16.10 STEAM GENERATOR TUBING IN N/A STRAIGHT TUBE DESIGN B16.20 STEAM GENERATOR TUBING IN N/A U-TUBE DESIGN CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR PO\.lER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 2 SECTION XI
SUMMARY
TABLE IWC CATEGORY C-A - PRESSURE RETAINING WELDS IN PRESSURE VESSELS ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C1.10 SHELL CIRCUMFERENTIAL WELDS VOLUMETRIC BORON INJECTION TANK 2 2 0 0 0 0 Welds at gros struct. disco only. Limit simi l. vess. arnoung vess. VOLUMETRIC EXCESS LETDO\.lN HEAT 0 0 0 0 0 EXCHANGERS VOLUMETRIC LETDO\.lN HEAT 0 0 0 0 0 EXCHANGERS VOLUMETRIC REACTOR COOLANT 0 0 0 0 0 FILTER REGENERATIVE HEAT 0 0 0 0 0 0 0 EXCHANGER VOLUMETRIC RESIDUAL HEAT 4 2 2 0 0 0 0 0 REMOVAL HEAT EXCHANGER VOLUMETRIC STEAM GENERATORS 20 3 3 0 0 0 0 0 ITEM TOTAL 30 10 6 0 0 3 0 60% 70% 100% C1.20 HEAD CIRCUMFERENTIAL WELDS N/A C1 .30 TUBESHEET-TO-SHELL WELDS VOLUMETRIC REGENERATIVE HEAT 6 2 0 0 2 0 0 0 EXCHANGER VOLUMETRIC STEAM GENERATORS 4 0 0 0 0 0 ITEM TOTAL 10 3 0 2 0 0 0 33% 100% 100%
- CATEGORY TOTAL 40 13 7 0 53%
3 0 76% 3 0 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CLASS 2 SECTION XI
SUMMARY
CATEGORY C-B - PRESSURE RETAINING NOZZLE. WELDS IN VESSELS ASHE SEC. XI TABLE !WC
# OF COMPONENTS
# OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C2.11 NOZZLE-TO-SHELL (OR HEAD) WELD N/A
<= 1/2 IN. NOMINAL THICKNESS C2.21 NOZZLE-TO-SHELL (OR HEAD) WELD N/A
> 1/2 IN. NOMINAL THICKNESS WITHOUT REINFORCING PLATE C2.22 NOZZLE INSIDE RADIUS SECTION > VOLUMETRIC BORON INJECTION TANK 2 2 0 0 0 0 1/2 IN. NOMINAL THICKNESS WITHOUT REINFORCING PLATE VOLUMETRIC STEAM GENERATORS 8 2 0 0 2 0 0 0 ITEM TOTAL 10 4 0 2 0 0 25% 75% 100%
- C2.31 C2.32 REINFORCING PLATE WELDS TO NOZZLE AND VESSEL > 1/2 IN.
NOMINAL THICKNESS NOZZLE-TO-SHELL (OR HEAD) N/A N/A WELDS WHEN INSIDE OF VESSEL IS ACCESSIBLE > 1/2 IN. NOMINAL C2.33 NOZZLE-TO-SHELL (OR HEAD) N/A WELDS WHEN INSIDE OF WELD IS INACCESSIBLE > 1/2 IN. NOMINAL CATEGORY TOTAL 10 4 0 2 0 0 25% 75% 100%
DATE: 04/24/92 SALEM NUCLEAR POIJER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8
- CLASS 2 SECTION XI
SUMMARY
TABLE !WC CATEGORY C-C - INTEGRAL ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES ASHE SEC. XI
# OF COMPONENTS
# OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOO SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C3.10 PRESSURE VESSELS-INTEGRALLY SURFACE BORON INJECTION TANK 4 4 0 0 2 0 2 0 25% of -int.
WELDED ATTACHMENTS attach. Limit 1 sim. vess. among vessel SURFACE CHEMICAL AND VOLUME 4 4 4 0 0 0 0 0 CONTROL TANK SURFACE REACTQR COOLANT 4 4 0 0 0 0 4 0 FILTER REGENERATIVE HEAT 3 0 0 0 0 0 0 0 EXCHANGER SURFACE RESIDUAL HEAT 4 2 0 0 2 0 0 0 REMOVAL HEAT EXCHANGER ITEM TOTAL 19 14 4 0 4 0 6 0 28% 57% 100% C3.20 PIPING-INTEGRALLY WELDED N/A ATTACHMENTS C3.30 PUMPS-INTEGRALLY WELDED N/A ATTACHMENTS C3.40 VALVES-INTEGRALLY WELDED SURFACE FEEDWATER SYSTEM 0 0 0 0 0 25% of int. ATTACHMENTS attachments CATEGORY TOTAL 20 15 5 0 4 0 6 0 33% 60% 100%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8
- CLASS 2 SECTION XI
SUMMARY
TABLE !WC CATEGORY C-D - PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN DIAMETER ASME # OF COMPONENTS SEC. XI #OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C4.10 PRESSURE VESSELS-BOLTS AND N/A STUDS C4.20 PIPING-BOLTS AND STUDS N/A C4.30 PUMPS-BOLTS AND STUDS N/A C4.40 VALVES-BOLTS AND STUDS N/A CATEGORY TOTAL 0 0 0 0 0 0 0 0 0% 0% 0%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE:
- REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CATEGORY C-F PRESSURE RETAINING WELDS IN PIPING CLASS 2 SECTION XI
SUMMARY
TABLE !WC CODE EDITION: 8 ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C5.11 CIRCUMFERENTIAL PIPE WELDS <= VOLUMETRIC CHEMICAL AND VOLUME 73 0 0 0 0 0 0 0 Multiple stre 1/2 IN. NOMINAL WALL THICKNESS SURFACE CONTROL SYSTEM VOLUMETRIC CONTAINMENT SPRAY 123 15 0 0 13 0 SURFACE SYSTEM VOLUMETRIC PRESSURIZER RELIEF 75 6 2 0 0 0 4 0 SURFACE SYSTEM VOLUMETRIC REACTOR COOLANT 5 0 0 0 0 0 0 0 SURFACE SYSTEM VOLUMETRIC RESIDUAL HEAT 352 27 14 0 5 0 8 0 SURFACE REMOVAL SYSTEM VOLUMETRIC SAFETY INJECTION 139 11 0 0 0 10 0 SURFACE SYSTEM ITEM TOTAL 767 59 18 0 6 0 35 0 30% 40% 100% C5.12 LONGITUDINAL PIPE WELDS <= 1/2 N/A IN. NOMINAL.WALL THICKNESS C5.21 CIRCUMFERENTIAL PIPE WELDS > VOLUMETRIC CHEMICAL AND VOLUME 374 33 10 0 13 0 10 0 Multiple stre 1/2 IN. NOMINAL WALL THICKNESS SURFACE CONTROL SYSTEM C5.22 LONGITUDINAL PIPE .WELDS > 1/2 N/A IN. NOMINAL WALL THICKNESS C5.30 CIRCUMFERENTIAL PIPE BRANCH SURFACE CHEMICAL AND VOLUME 40 0 0 0 0 0 Multiple stre CONNECHON WELDS > 4 IN. CONTROL SYSTEM NOMINAL BRANCH PIPE SIZE C5.41 LONGITUDINAL PIPE BRANCH SURFACE CHEMICAL AND VOLUME 4 0 0 0 0 0 0 0 Multiple stre CONNECTION WELDS > 4 IN. CONTROL SYSTEM NOMINAL BRANCH PIPE SIZE SURFACE PRESSURIZER RELIEF 4 0 0 0 0 0 0 0 SYSTEM ITEM TOTAL 8 0 0 0 0 0 0 0
******X ******X ******%
DATE: 04/24/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: 8 CLASS 2 SECTION XI
SUMMARY
TABLE llJC CATEGORY C*F CIRCUMFERENTIAL PIPE BRANCH CONNECTIONS >= 211 IN STAINLESS STEEL PIPING ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C5.42 N/A CATEGORY TOTAL 1189 93 28 0 20 0 45 0 30% 51% 100%
DATE: 03/28/92 SALEM NUCLEAR POWER STATION UNIT 2 PAGE: 3 REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL CODE EDITION: w8 CLASS 2 SECTION XI
SUMMARY
TABLE IWC CATEGORY C-F PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL PIPING ASME # OF COMPONENTS SEC. XI # OF NO. SCHEDULED/COMPLETED ITEM# ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C5.51 CIRCUMFERENTIAL PIPE WELDS >= VOLUMETRIC COMPONENT COOLING 17 0 0 0 0 0 0 0 3/8 IN. NOMINAL WALL THICKNESS SURFACE SYSTEM FOR PIPING > NPS 4 VOLUMETRIC FEEDWATER SYSTEM 77 8 6 0 0 0 2 0 SURFACE VOLUMETRIC MAIN STEAM SYSTEM 219 19 4 0 8 0 7 0 SURFACE ITEM TOTAL 313 27 10 0 8 0 9 0 37°!. 66% 100% C5.52 LONGITUDINAL PIPE WELDS >= 3/8 N/A IN. NOMINAL WALL THICKNESS FOR PIPING > NPS 4 CS.61 CIRCUMFERENTIAL PIPE WELDS > N/A 1/5 IN. NOMINAL WALL THICKNESS FOR PIPING >= NPS 2 AND <= NPS C5.62 LONGITUDINAL PIPE WELDS > 1/5 N/A IN. NOMINAL WALL THICKNESS FOR PIPING >= NPS 2 AND <= NPS 4 C5.70 SOCKET WELDS N/A C5.81 CIRCUMFERENTIAL PIPE WELD VOLUMETRIC MAIN STEAM SYSTEM 30 2 0 0 0 0 BRANCH CONNECTIONS OF BRANCH SURFACE PIPING >= NPS 2 C5.82 LONGITUDINAL PIPE WELD BRANCH N/A CONNECTIONS OF BRANCH PIPING
>= NPS 2 CATEGORY TOTAL 343 29 10 0 9 0 10 0 34% 65% 100%
DATE: 04/24/92 SALEM NUCLEAR PO\lER STATION UNIT 2 PAGE: REVISION: 0 INSERVICE INSPECTION PLAN FOR THE SECOND INTERVAL COOE EDITION: 8
- SEC. XI CLASS 2 SECTION XI
SUMMARY
CATEGORY C-G - PRESSURE RETAINING WELDS IN PUMPS AND VALVES ASHE TABLE IWC
# OF NO.
# OF COMPONENTS SCHEDULED/COMPLETED ITEM # ITEM DESCRIPTION EXAM METHOD SYSTEM DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C6.10 PUMPS-PUMP CASING WELDS}}