ML20246E777
| ML20246E777 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 04/25/1989 |
| From: | Burse B, Casey K, Shawn Smith TENNESSEE VALLEY AUTHORITY |
| To: | |
| Shared Package | |
| ML20246E773 | List: |
| References | |
| SI-114.1, NUDOCS 8905120018 | |
| Download: ML20246E777 (404) | |
Text
{{#Wiki_filter:-- _ _. _ __ - _. . _ _ .- _ i TENNESSEE VALLEY AUTHORITY i SEQUOYAH NUCLEAR PLANT h [. SURVEILLANCE INSTRUCTION [. SI-114.1 ASME SECTION XI I INSERVICE INSPECTION PROGRAM T Revision 14
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h, UNIT QA REVIEW REQUIRED 00 1h 2 D N/A D YESg NO O I~ t RESPONSIBLE SECTION: NDE Section XI Programs ,y PREPARED / REVISED BY: K._E. Casey ^ SUBMITTED BY: M.3 RespdqvitGe Section Supervisor QR: N'- / V ' , I y. SITEQACONbRRENCE: c - APPROVED BY: Appropr(ate Respons1 Die Manager DATE APPROVED: F d7S e f EFFECTIVE DATE: dM , h Y> 7 The last page of this instruction is number: 393 8905120018 890505 PDR ADOCK 05000327 Q PDC g{
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SEQUOYAH NUCLEAR PLANT l PLANT INSTRUCTION REVISION LOG [t-
- t SURVEILLANCE INSTRUCTION SI-114.1 .,
REVISION Date Pages REASON FOR REVISION (INCLUDE COMMIT-Approved [ LEVEL Affected MENTS AND ALL ICF FORM NUMBERS) 07/29/82 0 ALL
'2 1 01/04/84 ALL Revise:
~ 50-65, f 89-191 f'l Added:
65A-650, h t 88A,888, j; 191A - 2 02/23/84 19101 3 04/02/84 61-650 I Punch 11st Table of ! Contents kt (pp 1-2) 46-65, ' 65A-65D. ;- t 88,91,107 116,177, ' 178,190, 191V1, 1!UHl. ' 191X1 4 06/19/84 >!/ ' 121.122, 128-131,
- 141,142., ,
144-147. 159,191F, 191H,191L 191Q,191T i 191U,1912 ' 191G1,
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191I1 191J1, f 191R1, I 191T1, ' 5 09/14/84 - 0399C/1ds -
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SEQUOYAH NUCLEAR PLANT [
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PLANT INSTRUCTION REVISION LOG
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t-SURVEILLANCE INSTRUCTION SI-114.1 REVISION Date Pages REASON FOR REVISION (INCLUDE COMMIT- $ LEVEL Approved Affected HENTS AND ALL ICF FORM NUMBERS) ; i 6 05/28/85 196 i ALL i-Specific ." Changed '
- Pages: '
1-5,9,16, To reflect organizational changes
- 18,20,35, general program revision, and page 36,38-50, renumbering. ,
54,58, - 60-65, 67-71,73, i: 75,77-82, 86-96,124 170,217, s. 269-273, t~ 274-298, Specific reasons: ' 305,318 To comply with recommendations of QA , 361,383, Discrepancy Report SQ-DR-86-05-119R, 388,389, NSRS Investigation Report No. + 392,393, I-85-738-NPS, CATS item No. 84-116, 396,399, recommendations of NRC generic letter 400,411 85-02, and NRC commitments. See CCTS ' 417,421, item Nos.: NCO-85-0284-004, 4 13 NCO-85-0284-006, NCO-86-0119-001, C Added: NCO-86-0119-002, NCO-86-0119-003, L 273a,273b NCO-86-0119-004, NCO-86-0119-005, ; 298a,298b NCO-86-0199-006 and SPTS Item Nos. "' 7 10/14/86 298c,298d 860752 and 860712E. l l l \ I t 0399C/1ds 4 ,
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SEQUOYAH NUCLEAR PLANT l L PLANT INSTRUCTION REVISION LOG e l SURVEILLANCE INSTRUCTION SI-il4.1 REVISION Date Pages REASON FOR REVISION (INCLUDE COMMIT-LEVEL Approved Affected
- MENTS AND ALL ICF FORM NUMBERS)
Revised to comply with requirements of -I 2-4,47, NQAM Part 1. Section 2.16, and AI-12, - 50,63,64, Part I;. incorporate use of CAQR (see 69,71,79, SPTS Nos. 8611270 and 8701190); and to 81,82,87, ic correct typographical errors in Appendix f 8 2/11/87 90-93.269 A. Table A. 1 Punchlist Table of Revised to incorporate changes as needed ' Contents, to correct deficiencies as shown in CAQR p Page 2 of Nos. CHS870006 and CHS870010; to add - 2, 22-28, punchlist; to incorporate ICF No. #87-708, 43-47,51, PORC reviewed and approved 6/2/87. [ 52,59,62, also SAL No. 947; This revision consti-See i. 89-95,97- :- tutes a biennial review in accordance A 99,130- with AI-4, Appendix.H. Pages 130 through r 446,Added 530 renumbered to accommodate added draw- ? 9 08/07/87 447-S30 inos and revised Appendix G tables. 100-129, Corrected in accordance with AI-4, : 08/13/87 1 315 Section 9.1. j/ General revision, added Appendix H for support examination boundaries, editorial changes, organ-izational changes, revised drawings for ASME code ', class and prisim, added detailed equipment support drawings, revised Appendix G for support update i
, and clarification, revised NOI form, incorporated t
; the long term solution to CAQR.No. CHS 870006, incorporated changes in SPTS Nos. 871007A and [.
p 870906A, incorporated changes in ICF No. 87-1801, : Revised Appendix C for clarification on valve t examinations, revised Appendix A for scheduled component examiliations, added ASME code class 3
- equipment supports description in program, and "
10 12/23/87 All deleted Punchlist. ', {
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PLANT INSTRUCTION REVISION LOG ,- I o
- k i;l SURVEILLANCE INSTRUCTION SI-114.1 REVISION Date Pages REASON FOR REVISION (INCLUDE COMMITMENTS AND LEVEL Approved Affected ALL ICF FORM NUMBERS) '
Revised to incorporate changes in ICF No. 88-0695, . 88-0221, 88-0065, and 87-2410. Added mainsteam , supports 1-MSH-440 and 1-MSH-322. Revised f support information on I-AF0H-342, 1-CCH-430, 1-CCH-1006, 1-CCH-1043, 1-ERCWH-149, - t, , 1-FDH-203(MS), 1-FDH-203(VS-2), 1-RCH-80, j. 1-RCH-251, 1-RHRH-7, 1-SIH-106 and 1-VHIH-035. ! 3,66,71, Resorted tre entire Appendix G to Alpha-Numeric ? 72,73,78, on support numbers. Text moved only on pages . 95,231, 73, 332, 334-336, 341-344, 346, 348-354, 356, r 332-375, 357, 361, 362, 366, 367, 369-372, 374 and 375. u 11 08/08/88 379 Incorporated SPTS No. 880601H. . 28,44,51, 62,66,68- Revised Appendix G, Appendix H, and associated 70,95,219 pages to change reference for component support 377,378 movement and acceptance range information to the ;; 331-375, Scan Plan. Added supports 1-RHRH-422A and - 12 10/26/88 382,383 1-RHRH-449A. Incorporated SPTS No. 880504C. 88, Added support 1-RCH-130 due to Work Plan No, { e 13 01/25/89 210,361 WP475-01. 02/01/89 69,70 Interfiled ICF 89-0100.
- General revision to make organizational changes; address long-term commitments (C.1 NC0870038004) L to implement accelerated field weld program; [
incorporate ICF 89-0100; add Appendixes I and J. ' Add sections, drawings, and revised Appendix G A for RCS Mair Loop supports and ERCH Screen Wash L Pump supports. Add augmented sections for NRC Bulletins 88-08, 88-09, 88-11. Add revised 1-388; drawings and miscellaneous changes. Incorpor-Added: ated SPTS No. 890404C (Quality Notice to 14 389-393 NOAM, Part II, Section 5.1, dated 3/21/89. 9 i'
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TO: Those Listed on Attached Distribution List ' FROM: Document Control and Records Management Section, Sequoyah Nuclear Plant ,s DATE:
SUBJECT:
TRANSMITTAL OF SEQUOYAH INSTRUCTIONS Please sign below and return to acknowledge receipt of the following material and affirm that it has been placed in the appropriate manuals, your table of contents has been updated, and the superseded material /information removed and discarded. The items listed below must be filed by ' and this transmittal returned to Document Control within 10 working days of the filing deadline. 9 SI-114.1 [ ASME SECTION XI h INSERVICE INSPECTION PROGRAM Remove and Destroy Insert E Page Date Revision Page Date Revision ; l-388 Various/Various 1-393 14
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Filed by i Date Manual Number [. l Address change or c.orrection: Return to: Joan Record, DCRMS, O&PS-2, Sequoyah Nuclear Plant
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5.' SQNP 51-114.1 : Page 1 of 1 ! Rev. 14
. Owner: Tennessee Valley Authority' !f
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Address of Corporate Office: Chattanooga Office Complex 1101 Market Street ; Chattanooga, Tennessee 37402-2801 Name and Address'of Nuclear Power Plant: Sequoyah Nuclear Plant ', P.O. Box 2000 L Daisy, Tennessee 37319 f
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Applicable Nuclear Power Units: Sequoyah Nuclear Plant, Unit 1 l-i Commercial Operation Date: July 1. 1981 Iy e l' i-T
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S 14.1 -l Page 1 of 2 [Li Rev. 14 ; 7 _ TABLE OF-CONTENTS Page'No. b
'p 0.1 Prerequisites i ?;
0.2 Precautions i 1.0 Statement of Applicability k 1 2.0 Purpose 1 - s-" 3.0 Inspection Intervals and Inspection Periods 1 -
- .ic 4.0 Codes of Record 2 a 5.0 Method of Implementation and Responsibilities 3 {
6.0. Abbreviations.and Definitions 5 P, 7.0 Components Subject to Examination - ASME Code Class 1 6 . 8.0 Components Subject to Examination - ASME Code Class 2 36 .R 9.0 Components Subject to Examination - ASME Code Class 3 [K 52 6 10.0 Authorized Inspector U 64 - 11.0 Examination Methods 64 I 12.0 Qualifications of Nondestructive Examination Personnel 68 13.0 Acceptance Criteria 68 14.0 Repairs i, p 68 p 15.0 Replacements 68
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16.0 Records and Reports. . 68
- 1. 7. 0 Notification of Indication 72 18.0 Calibration Blocks 76 I..
. p, 19.0 Requests for Relief 77
'20.0 Augmented Inspections 77 ji 21.0 References 83 q
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SQNP SI-114.1 ,. Page 2 of 2 ; Rev. 14 1. o TABLE OF CONTENT _S Page No. 9 L: Appendix A - Tables 86 ! .- Appendix B - Drawings 110 Appendix C - Valve Tables 297 4 Appendix D - Data Sheets 305 t Appendix E - Notification of Indication Form 306 [ i Appendix F - Requests for Relief 307 [. i.- 1 Appendix G - Support Load Criteria 341 '(
!E Appendix H - Guidelines for Determining Piping Component Support Examination Boundaries 379 [j.' ,
l', Appendix I - Field Corrected Drawing Transmittal 392 i' Appendix J - Commitment List for SI-114.1 393 . L yi r i. i. t I i p. I f
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SONP r 51-114.1 ' Page i of i !. Rev. 14 b p 0.1 PREREQUISITES - b' O.1.1 When craft support of minor o7 similar maintenance (examples: ! scaffolding, insulation removal (pressurizer loop seal insulation ' may not be removed using a generic MR), buffing of welds using ' scotchbrite pads, and cleaning bolts) is required to facilitate b ' performance of this SI, a single work request (WR) may be used. This WR shall be processed in accordance with Standard Practice SQM2 and the system code.for S1-114.1 (system code 114) will be referenced on the WR.
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0.1.2 Contact Radiation Control (RC) for radiation work permit -l (RWP)/ALARA preplanning requirements. . O.l.3 Forty-five days prior to a refueling outage, the NDE Section XI 5 Programs of NDE Engineering shall submit a list of snubbers within y their scan plan to the Technical Support Programs Section.. This ' list will be reviewed by the Technical Support Section within two . weeks of receipt to determine if benefits can be gained by the k site quality organization performing the examinations required by : SI-162.1 during the performance of this procedure. If benefits are identified, the Technical Support Programs and Site Quality ;* , Organization will establish dates for the required SI-162.1 h briefings and other coordinations that may be necessary. The : meetings will be held at least two weeks prior to the start of the F outage. 0.2 PRECAUTIONS [ 0.2.1 Safety belts should be worn when working from scaffolding or ladders in accordance with HCI-M2. f. 0.2.2 Protective clothing, such as long sleeve shirts or gloves. _ should be worn except in RWP areas when working around hot pipes and f equipment. 0.2.3 Care should be exercised when climbing on plant structures and piping to ensure firm footing. Flexing of instrument lines and electrical conduit, for example, could cause equipment damage as well as bodily injury resulting from a fall. ,f i 0.2.4 Efforts should be made to ensure proper planning to reduce delays g and radiation exposure in performance of the work. F' O.2.5 Read and observe all applicable precautions as indicated in SQN ! Administrative Instructions (AI)-26 " Requirements for an Open i j Reactor Coolant System" and AI-8 " Access To Containment". t i 0963o/COC4 _f_ ; I 9
SQNP , SI-ll4.1 . Page 1 of 85 j' Rev. 14 l. 1.0 STATEMENT OF APPLICABILITY '; This program outlines requirements for performing the first 10-year interval inservice nondestructive examinations of the Sequoyah Nuclear
-I Plant, unit 1, ASME Code Class 1, 2, and 3 (Equivalent) components (and
- their supports) containing water, steam, or radioactive material (other '
than radioactive waste management systems). The program has been organized to fulfill inservice examination requirements of the Nuclear Quality Assurance Manual, Part II, Section 5.1 and comply as practical , with the requirements of Section XI of the ASME Boiler and Pressure , Vessel Code. In addition, this program implements applicable portions - of the Sequoyah Technical Specifications. The Inservice Inspection g Program _.(ISI) satisfies the requirements of Surveillance Requirements - 4.4.3.2.4, 4.4.5.0, and 4.4.10, and partially satisfies the requirements '- c of Surveillance Requirement 4.0.5 and 4.7.9. Inspection frequencies are . in accordance with Appendix A of this instruction and are generally , scheduled to coincide with outage periods. l9 The requirements of this program are applicable beginning at the date of [ commercial operation of the unit. 2.0 PURPOSE , y_ The ISI program shall be used for planning inspections and nondestructive ( examinations of the Sequoyah ASME Class 1, 2, and 3 components for the - first interval. Personnel responsible for performance of the , examinations should familiarize themselves with the requirements of this program prior to performing the examinations. The examinations required by this program will establish acceptance of components for continued service. 5 3.0 INSPECTION INTERVALS AND INSPECTION PERIODS b The inservice examinations required by ASME Section XI shall be per-formed during each 10-year interval of service (inspection interval). ' The inspection intervals represent calendar years after the unit has ' been placed into commercial service. The commercial operation date for unit I was July 1, 1981. The first 10-year interval may be decreased or extended by as much as one year. If the unit is out of service continuously for six months or more, the inspection . interval may be , extended for an equivalent period. This SI may be performed in any mode and is applicable for all operational modes. The inspection interval shall be separated into three inspection r periods. Except for examinations that may be deferred to the end of the l inspection interval, the required examinations shall be performed in i ! accordance with the following schedule. 'l l I
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1 Minimum Examinations Maximum Examinations b l Inspection Periof Completed, Percent Completed, Percent [! g, 3 Years 16 34 7 Years 50 67 :. 10 Years 100 100 \ [' L The examinations deferred to the end of the inspection interval shall be [- completed by the end of the inspection interval. ' On~ August 26, 1985 Sequoyah Unit I went off line and it stayed off line , until November 10, 1988. In accordance with IWA-2400(c) TVA will extend L the first interval by 1173 days. This will make the first interval end [L on September 15,1994. The period duration will be as follows: 1) ist period 7-1-81 thru 6-30-84, 2) 2nd period 7-1-84 thru 9-15-91, and 3) 3rd period 9-16-91 thru 9-15-94. 4.0 CODES OF RECORD AND CODE CASES The Sequoyah unit.1 operating. license (low power) was issued on February 29, 1980. Thus, as a minimum the ISI program shall not be it prepared to a Section XI Code Edition and Addenda prior to the 1974 . Edition Summer 1975 Addenda. F
.In accordance with 10 CFR Part 50.55a(g)(4)(iv), this program was I prepared to meet the requirements of.the 1977 Edition, Summer 1978 ,
Addenda, of Section XI of the ASME Boiler and Pressure Vessel Code. Steam Generator Tubing Examination requirements are in accordance with
- Regulatory Guide 1.83, Rev.1, and Technical Specification and surveill-ante requirement .4.4.5. In accordance with 10 CFR Part 50.55a(b)(2),
the extent of examination for piping welds Examination Categories B-J ' and C-F is in-accordance with the 1974 Edition, Summer 1975 Addenda of + ASME, Section $1 (Examination Categories B-J, C-F and C-G). Extent of L examination is defined as criteria for the selection of Class 1 and " Class 2 components for examination and as criteria for determining which Class 2 components may be exempt from examination. The' extent of examination specifi.es the length of weld to be examined. The qualifications on NDE 2300 examination personnel shall be as per paragraphs IHA-1600, IWA, and Table IHA-1600-1 of ASME Section XI 1983 , Edition with Winter 1985 Addenda. , Preservice examinations were conducted in ;ccordance with the 1974 ! Edition, Summer 1975 Addenda, of ASME Section XI (S1-114). . r TVA will use Code cases N-209, N-234, N-235, N-307-1, N-308, N-401, and :. N-416 which have been approved by NRC as per Regulatory Guide 1.147 as ' outlined on'the applicable NDE procedure. Code Cases N-341 and N-356 were authorized for TVA use by NRC on letter from Gary L Zech to S. A. 4 White dated Jan. 25, 1988. t 0963o/COC4 ' R.6 >
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m _ . . .. . .. -- . - . _ . ~ . . _ _ _ - ~_ - -u-1 SQNP SI-114.1 1 Page 3 of 85 L Rev. 14 L V E. TVA will utilize Regulatory Guide 1.150, Rev. 1. " Ultrasonic Testing of . ?, Reactor Vessel Welds During Preservice and Inservice Examinations" for j its examinations, as outlined.in the applicable NDE procedure. [; P 5.0 METHOD OF IMPLEMENTATION AND RESPONSIBILITIES ( Nuclear Quality Assurance (NQA) shall prepare scan plans using component l' listings supplied by NDE'Section XI Programs for the inservice 2 inspection and a list of steam generator tubes to be examined supplied _ F by NDE Section XI Programs with the assistance of the Steam Generator
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L Group for inservice inspection. The plans should include as a minimum , references to components to be examined, methods of examinations, 'I examination procedures, and calibration standards. Prior to performing E examinations on a system or component, the scan plans shall be approved t; by an NDE Level III individual from NQA and made available by NQA to the k Plant Manager for information. When inservice examinations are E implemented by instructions other than this program (e.g., MI's), copies [, of the instruction data sheets shall be submitted to NQA by the performing organization such that these examinations may be included in the reports discussed in Section 16.0. . During implementation phases (usually outage periods), it may become l;> necessary to change the scan plan. Scan plan changes can be initiated . by the NQA, NDE Section XI Programs, or by other personnel involved with c the implementation of the scan plan. All changes shall be coordinated )* with NDE Section XI programs engineer in charge of the specific plant and unit for concurrence of the change and with the appropriate plant - planning and scheduling personnel for facilitating the use of supporting ^ craft personnel. Revisions to the scan plan shall be controlled in the same manner as the original. However, interim working copies may be ,. hand written to allow examinations to be performed before a formal : revision is issued. These changes shall be made only by the NDE Section - XI Programs Engineer with the concurrence of the NDE Level III. Both ' individuals shall sign and date ajl.1, such changes. y, t' When major portions of existing piping are replaced or whole new systems E are added, a system walkdown should be performed to identify the piping . configuration, welds, and components that will be included in the inspection program. Walkdowns of individually repaired or replaced components are not a requirement prior to performing the examination. If variations in piping configurations are discovered or modification or E repairs and replacements to piping are made during the service life of ,; the unit, these changes shall be marked on field controlled copies of ;. drawings. This information shall be communicated to NDE Section XI L Programs, utilizing the Appendix I form, which shall be responsible for ? revising the drawings and incorporating the revised drawings into this instruction. The NQA scan plan shall also be revised to reflect these changes. ' The in-service examinations w'ill be performed by either QA personnel or " outside contractors. Contract preparation, administration, and super-vision will be the responsibility of NQA. Inspection plans submitted by i outside contractors shall be reviewed and approved by NQA prior to use. ' l 0963o/COC4 V
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.y All specific NDE procedures used during the inspection program shall be reviewed and app ~ roved in accordance with NQAM Part II, Section 6.3. NDE o procedures will be implemented in accordance with TI-SI.
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t NQA personnel will be responsible for notifying the plant manager of unacceptable indications found during in-service inspection as soon as s practical. Whenever an unacceptable indication is discovered., this
- procedure (Section 17.0) and the form in Appendix E shall be utilized. '
In those cases where an outside contractor is furnishing inservice examination services, the contractor will normally initiate the form in ,, l' / Appendix E under the supervision of the NQA representative. See Section- 17.0 of this program. ; L' ~ . When the examination is completed the NQA representative shall compare the completed exam results to the requirements of the scan plan, , identify any limitations or impractical examinations, and provide , notification to NDE Section XI Programs for possible action in l' accordance with Section 19.0 of this instruction. J't An NDE Level II or III individual shall evaluate the NDE results in accordance with ASME Section XI, IWA-3000. The results shall be compared with recorded results of the preservice NDE and previous fl . in-service NDE results. Following each refueling outage when inservice p examinations are performed, the data package cover sheet in Appendix D C shall be completed by the NQA representative and submitted to the plant ' with the final report discussed in Section 16.0.
- ISI program preparation is'the responsibility of NDE Section XI Programs. Any revisions initiated by other groups shall be submitted to NDE Section XI Programs for approval prior to incorporating the revisions into this procram.
NDE Section X Programs maintains a computerized data base system for status and Section XI credit of completed ISI examinations. [ NDE Section XI Programs shall be responsible for preparing and revising the color coded flow diagrams to show the ASME Section XI Code Classes 1, 2 and 3 (Equivalent). These classifications have been prepared in accordance with i~ 10 CFR 50.2(v), Regulatory Guide 1.26 R3, ASME Section XI and guidelines in 10 CFR 50.55a. .
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It must be understood that these drawings are to be used for ASME ; Section XI activities only. . NDE Section XI Programs shall be responsible for preparing and revising the Piping Component Support Examination Boundaries (Appendix H). l 0963o/COC4 i
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SQNP l S1-114.1 j' l Page 5 of 85 -I Rev. 14 r 1 6.0 i-. ABBREVIATIONS AND DEFINITIONS L;
!' j 6.1 AIA--Authorized Inspection Agency. bJ 6.2 AI--Authorized Inspector (may denote an ANI or ANII). f L
6.3 ANI--Authorized Nuclear Inspector. f, f 6.4 ANII--Authorized Nuclear Inservice Inspector. 6.5 Type A Examination Boundary: The type of boundary to be used for p those supports that are attached to building floor, walls, C ceiling, or embedded plate. jT 6.6 Type B Examination Boundarv: The type of boundary to De used for f. I those supports that are attached to another existi:'; support. L 6.7 Type C Examination Boundary: The type of boundary to be used for b g those supports that are attached to existing steel. 6.8 Tvpe D Examination Boundary: The type of boundary to be used for those supports that are attached to an intervening element. [ 6.9 NOI--Notification of Indication. [ r. 6.10 CAQR--Condition Adverse to Quality Report. 6.11 USQD--Unreviewed Safety Question Determination. 6.12 NQAM--Nuclear Quality Assurance Manual. I 6.13 DCR--Design Change Request. i, ' 6.14 MR--Maintenance Request. ? 6.15 WR--Hork Request. 6.16 CAQ--Condition Adverse to Quality. 6.17 Components--Denotes items in a power plant such as vessels, piping systems, pumps, valves, and component supports. ; i 6.18 Examination--Denotes the performance of all visual observation and ? nondestructive testing such as radiography, ultrasonic, eddy current, liquid penetrant, and magnetic particle methods. [ S 6.19 Inspection--Denotes verifying the performance of examinations and tests by a representative of an Authorized Inspection Agency. 1 1 0963o/COC4 ! L=1_=_-_-__~
SONP SI-114.1 Page 6 of 85 L Rev. 14 , i 6.20 Normal Operation--Normal plant operation conditions include ! reactor startup, operation at power, hot standby, and reactor [~ cooldown to cold shutdown conditions. Test conditions are f excluded. 6.21 Pressure-Retaining Material-Applies to items such as vessel i heads, nozzles, pipes, tubes, fittings, valve bodies, bonnets, ,, disks, pump castings, covers, and boltings which join pressure-retaining items. 6.22 Repair--Those operations involving welding, heat treatment, or s defect removal which are required to restore an item to a safe and r satisfactory operating condition. ' 1 6.23 Replacement--Replacements include spare and renewal components, or parts of a component (e.g., valve body bonnet, disc, bolting). It also includes the addition of components such as valves and system changes such as rerouting of piping. For the purpose of this . q procedure, the term replacement shall apply where attachment to the pressure boundary is by welding or mechanical means. - 7.0 COMPONENTS SUBJECT TO EXAMINATION--ASME CODE CLASS 1 (EOUIVALENT) The ASME Code Class 1 (Equivalent) components to be examined during the I inspection interval are outlined in the following paragraphs. The a entire length of each weld described will be examined for the first - ' 10-year ISI unless otherwise noted. When a portion of a weld length is to be examined during an inspection period, the areas examined each inspection period shall be documented on the examination data sheets. Components that are exempted from examination are discussed in I Section7.7ofthisprogram. Table A in Appendix A supplies additional information such as reference ' drawing numbers and ASME Section XI Table IWB-2500-1, examination r categories. 7.1 Reactor Vessel 7.1.1 Reactor Vessel Seam Welds 7.1.1.1 Circumferential Shell Welds '; 1 There are four circumferential welds (each [' approximately 50 feet in length) .in the vessel cylindrical shell, three of which are located r behind the thermal shield. The entire length of ' each of these welds will be ultrasonically examined during the third inspection period using remote inspection devices from the vessel I.D. with the core internals removed. 09630/COC4 I l i u , - .- ... -. . .. - : 2
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, The vessel shell sections are machined forgings $
fabricated of A-508. Class 2, manganese-molybdenum steel and are clad with weld deposited !!
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austenitic stainless steel. 7.1.1.2 Longitudinal Shell Welds f L There are no longitudinal shell welds associated [j with the reactor vessel. ' 7.1.1.3 Closure Head Circumferential Weld ' The entire length of the closure head cap weld I t (approximately 41 feet in length) will be bl manually ultrasonically examined from the head . outside diameter. The length of weld to be k examined each inspection period is included in i-Table A of Appendix A. C' y1 During the conduct of the unit I preservice ( examination a flaw indication in the closure head y weld was observed (weld 09-10). Flaw evaluation 3 and acceptability was based on Code Case N-209 p (see nonconforming report 6P for additional ...l Information concerning flaw evaluation and f. i location). In addition to the examinations F above, the flaw indication volume shall be ' examined during each inspection period. This provides examination of the flaw indication volume for three successive inspection periods. , Should these successive examinations reveal that . the flaw indication has remained essentially T
- i. unchanged, then the examination frequency may revert to that of Examination Category B-A of 6 Table IWB-2500-1 of ASME Section XI. If the P successive examinations reveal that the flaw indication size.has increased, then technical justification shall be presented to NRC for
. allowing continued operations. '
The closure head ring is fabricated of A-508, , [ Class 2, manganese-molybdenum steel. The closure ; head hemispherical section is fabricated of A-533, Gr. B, Class 1, manganese-molybdenum j p steel. Both sections are clad with weld deposited austenitic stainless steel. [(f i 0963o/COC4 7 iI
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? <_ . , l p SQNP- r. 1 SI-114.1 , Page 8 of.85 t Rev. 14 F. ; E F, 4 7.1.1.4 Lower Head Circumferential Weld [? The entire length of the lower head circumfer- ', ential weld (approximately 38 feet in length) g will be ultrasonically examined during the third ) , inspection period using remote inspection devices - from the. vessel inside diameter with the core R internals removed (see Request for Relief ISI-5). :< r. The lower head sections are fabricated of A-533, p) l Gr. B, Class 1, manganese-molybdenum steel, and are clad with weld deposited austenitic stainless - steel. (n ', l F 7.1.1.5 Closure Head Meridional Weld r .. . The closure head does not include any meridional 4, welds. ,'! 7.1.1.6 Lower Head Meridional Welds b
.l.
There are six meridional welds (each approximately F 4 feet in length) located in the lower head. The 7, entire length of each of these welds will be " ultrasonically examined during the third inspec- p tion period using remote inspection devices from 3 the vessel inside-diameter with the core internals removed. The lower head section material is identified in section 7.1.1.4 of this program. 7.1.1.7 Shell-To-Flance Weld :.
,, is a
!. The entire length of the shell-to-flange weld R (approximately 50 feet in length) will be f ultrasonically examined from the vessel inside '
i diameter using remote inspection devices during the third inspection period. The vessel flange section is fabricated of A-508, Class 2, manganese-molybdenum steel and is clad internally and on the gasket face with weld , deposited austenttic stainless steel. L P 7.1.1.8 Closure Head-To-Flance Weld i
?
The entire length of the head-to-flange weld [ (approximately 45 feet in length) will be , manually ultrasonically examined from the head outside diameter. The length of weld to be examined each inspection period is included in Table.A of Appendix A. 0963o/COC4 [l
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- E SQNP SI-ll4.1 Page 9 of 85 i Rev. 14 ;
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. The closure head flange section is fabricated of i A-508, Class 2, manganese-molybdenum steel and is ,-
clad internally and on the gasket face with weld E deposited austenitic stainless steel. F! j t, 7.1.1.9 Repair Welds Base metal weld repairs in the beltline region Z where the repair depth exceeds 10 percent nominal a of the vessel wall shall be ultrasonically ' examined. There are no base metal repair welds in ' the beltline region of the unit i vessel. . 7.1.2 Reactor Vessel Nozzle-To-Vessel Welds U t. There are four inlet nozzles (27.500 inch I.D.) and four outlet nozzles (29.937-inch I.D.). The eight nozzle-to-
;j vessel welds will be ultrasonically examined from the ;
inside diameter using remote inspection devices, j The four outlet no7zle-to-vessel welds snall be ! l ultrasonically examined from the nozzle bore when the - upper internals are removed during the first inspection i period. The inlet nozzles are not accessible until the ', core barrel is removed. The four inlet nozzle-to-vessel i welds shall be ultrasonically examined from the nozzle 1 bore and from the vessel shell inside diameter during the * ' third inspection period when the core barrel has been ! removed. The outlet nozzle-to-vessel welds shall also be ! ultrasonically examined from the vessel shell inside , ; diameter during the third inspection period when the core 0i
-barrel has been removed.
i'
- i~
The nozzle forgings are fabricated of A-508 Class 2, ; ' manganese-molybdenum steel and are clad with weld ; deposited austenitic stainless steel. v , 7.1.3 Reactor Vessel Nozzle Inside Radius Section The eight nozzle inside radius sections (including the i outlet nozzle integral extensions) shall be ultrasonically . , examined at the same time as the examination of the [l nozzle-to-vessel welds conducted from the nozzle bore (see F' Section 7.1.2). L P
' Nozzle forging material is identified in Section 7.1.2 of Ll this program. [
-9 0963o/C0C4 k )
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SQNP *f bof85 : Rev. 14 !-{ Pl Fj 7.1.4 Reactor Vessel Partial Penetration Welds ~ E'
. b The vessel includes 4 upper head injection nozzles, 1 vent [
pipe nozzle, 78 control rod drive nozzles, and 58 instru-mentation nozzles with partial penetration welds. F< Approximately 25 percent of each group of nozzles shall be - visually examined from the vessel outside diameter in I. accordance with visual examination method VT-2 (see r section 11.1). This 25 percent shall incluce 1 upper head injection nozzle, 1 vent pipe nozzle, 20 control rod drive t nozzles, and 15 instrumentation nozzles. Examination of , these nozzles during the inspection interval shall be . distributed among the inspection periods in accordance- 4 with Table A of Appendix A. f The NQA shall indicate in scan plans the nozzles requiring examination to be examined in accordance with SI-146 or c SI-250. The examination results may be recorded on SI-146 s or'SI-250 data sheets. 7.1.5 Reactor Vessel Nozzle-To-Safe End Welds k The inlet and outlet nozzle-to-safe end welds shall be i ultrasonically examined from the inside diameter using ? remote inspection devices. The ultrasonic examination f' shall be performed at the same time as.the examination of ' the nozzle-to-vessel welds conducted from the nozzle bore * (see Section 7.1.2). All of the nozzle-to-safe end welds shall also be liquid penetrant examined during the inspection interval f
<oincident with the ultrasonic examinations The Sequoyah .
iNuclear Plant Unit 2 Report--Evaluation of Cracking in F Reactor Vessel Nozzle Stainless Steel Buttering, states */ that Unit i nozzle-to-safe end welds RC-9SE and RC-32SE l shall be examined during the first inspection interval , (see memorandum from D. W. Wilson to H. L. Abercrombie, dated August 1, 1986 (L18 860730 899). The remaining nozzles will be examined during the normal inservice inspection intervals. Each nozzle safe end weld is a stainless steel type 304 m weld build up. ' 7.1.6 Reactor Vessel Pressure Retaining Bolting larger Than Two r
- Inches In Diameter During each ~ refueling outage all closure studs, nuts, and '
washers are removed. All of the 54 closure studs, nuts, ' washers, and ligaments between the vessel flange stud i I 0963o/COC4 w e e
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-Page 11 of 85 i Rev..f4 ;r ho_les shall be examined during the inspection interval in-accordance with Table A of Appendix A. The closure nuts E shall be magnetic particle examined. The closure studs .
C. shall be ultrasonically and magnetic particle examined. The ligaments between the vessel flange stud holes shall .c be ultrasonically examined. (This provides for ultrasonic- : examination of the threads in base material.) The closure ._ washers'shall be visually examined in accordance with . visual examination method VT-1 (see section 11.1). The ( bolting may be examined in place under tension, when the .s connection is disassembled, or when the bolting is removed. ' a Provisions for this examination are included in MI-1.2. E 14here needed the examinations of the studs should be L identified with the particular stud in accordance with the L unique identification system for the studs'and nuts. : C The 7-inch diameter studs, nuts, and washers are ,, fabricated of SA-540, Gr. 823 Nickel, chrome-molybdenum .- steel. . 7.1.7 Reactor Vessel Pressure Retainino Boltina Two Inches And " Smaller In Diameter There is no pressure retaining bolting two inches or smaller in diameter. ' 7.1.8 Reactor Vessel Integrally Welded Attachments I There are no integrally welded vessel supports. 7.1.9 -Reactor Vessel Interior { The vessel interior shall be visually examined in l accordance with visual examination method VT-3 (see L Section 11.1). These examinations shall include the space above and below the reactor core that is made accessible - for examination by removal of components during normal ~ refueling outages. The examinations shall be performed at ; the first refueling outage and subsequent refueling ., outages at approximately three year intervals. ,'
. i-7.1.10 Reactor Vessel Removable Core Support Structures The visually accessible attachment welds and visually l'
-accessible surfaces of the core support structure shall be b visually examined in accordance with visual examination bi method VT-3'(see Section 11.1). This examination may be j deferred to the third inspection period. The structure i shall be removed from the reactor vessel for examination.
4
\
1 i-0963o/COC4 :
SQNP SI-114.1 Page 12 of 85 Rev. 14 f The removal of the core support structure is implemented by VI-1.4. , 7.1.11 Reactor Vessel Control Rod Drive Housinos There are 78 control rod drive housings penetrating the closure head. Each housing includes a pressure retaining , dissimilar metal butt weld and canopy seal weld. The canopy seal welds are exempted from volumetric or surface . examination in accordance with ASME Section XI, - IHB-1220(a). , There are 20 peripheral control rod drive housings. Two b (10 percent) of the peripheral housing butt welds shall be ultrasonically examined during the inspection interval in , accordance with Table A of Appendix A. The housings
- consist of a 6" 0.D. adapter of A-182, 304SS and a 4" 0.D.
body of SB-167. 7.1.12 Reactor Vessel Auxiliary Head Adapters Each of the four auxiliary head adapters includes a '- pressure retaining dissimilar metal weld. The four dissimilar metal welds shall be ultrasonically and liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A. The adapters consist of a SA-182, 30455 upper portion and a SB-166 lower portion. . 7.2 Pressurizer , 7.2.1 -Pressurizer Circumferential Shell-To-Head Welds There are two circumferential sheil-to-head welds, each approximately 24 feet in length. The entire length of each weld shall be ultrasonically examined during the inspection interval in accordance with Table A of Appendix A. All vessel shell and head sections are fabricated of SA-533, Class 2, manganese-molybdenum steel and are clad , with austenitic stainless steel. ' 7.2.2 Pressurizer Longitudinal Shell-To-Head Welds
-There is one longitudinal weld intersecting each k circumferential shell-to-head weld. One foot of one :
longitudinal weld shall be ultrasonically examined during the inspection periods in accordance with Table A of Appendix A. The one foot of weld examined during each i 0963o/C0C4 e & =_e_-_ - -
= -- - - .~ - -- ~-
SONP S1-114.1 Page 13 of 85 i Rev, 14 . 1 examination shall include the length of weld as measured i from the point of intersection of the longitudinal weld b' with the circumferential head-to-shell weld. 'c The vessel shell section material is identified in Section $- 7.2.1. 7.2.3 Pressurizer Circumferential and Meridional Head Helds r There are no pressurizer circumferential or meridional ; head welds. -
. i.
7.2.4 Pressurizer Nozzle-To-Vessel Welds and Nozzle Inside Radius Section The pressurizer includes three 6-inch safety valve [i nozzles, one 6-inch relief valve nozzle, one 4-inch e aozzle, and one 14-inch surge nozzle. All of the t nozzle-to-vessel welds, including nozzle inside radius - section, shall be ultrasonically examined during the inspection interval in accordance with Table A of ., Appendix A. The nozzles are fabricated of SA-508, Class 2, - manganese-molybdenum steel. 7.2.5 Pressurizer Heater Penetration Welds There are 78 heater penetration welds located in the pressurizer lower head. Approximately 25 percent of the ' heater penetration welds (20 welds) shall be visually
-examined during the inspection interval in accordance with i, visual examination method VT-2 (see Section 11.1),
L Examination of these penetrations during the inspection ; interval shall be distributed among the inspection periods ' in accordance with Table A of Appendix A. The NQA shall designate in scan plans the heater penetrations requiring examination such that the ' organization performing the examinations may document , examination results on SI-146 or SI-250 data sheets. , t 7.2.6 (. Pressurizer Nozzle-To-Safe End Welds 1 t : Each of the six nozzles identified.in section 7.2.4 , includes a welded forging safe end. All of the ?, nozzle-to-safe end welds shall be ultrasonically and ' liquid penetrant examined during the inspection interval I in accordance with Table A of Appendix A. 4
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i SONP SI-114.1 Page 14 of 85 Rev. 14 i
- 5. I Safe end connections are'SA-182, Gr. F-316L forgings. -
L' 7.2.7 Pressurizer Pressure Retaining Boltino larger Than Two [. Inches In Diameter F F There is no pressure retaining bolting larger than two '
.l inches in diameter. '
7.2.8 Pressurizer Pressure Retainino Bolting Two Inches And ' Smaller In Diameter ' All of the pressurizer manway bolts shall be visually-examined in accordance with' visual examination method VT-1 d 'J (see Section 11.1). The examinations during the inspection interval shall be in accordance with Table A of Appendix A. I The bolts may be examinec in place under tension or when the-bolts are removed. It is preferable to perform the ~ examinations when the bolts are removed if possible. .. Removal of the manway cover is performed in accordance !, with MI-4.5 and provides for examination of bolting. The manway includes 16 bolts at 1.88 inches in diameter. f 7.2.9 Pressurizer Integrally Welded Support Attachments f-The entire length of the pressurizer support skirt-to- 'l vessel weld (approximately 23 feet in length) shall be ! surface examined (magnetic particle) during the inspection / interval in accordance with Table A of Appendix A. - The support skirt is approximately 1.5 inches thick and is
-fabricated of SA-516, Gr. 70, carbon steel plate.
, i.
7.3 Steam Generators (4) 7.3.1 Steam Generator Primary Side Circumferential And Meridional Head Welds . There are no steam generator primary side circumferential or meridional head welds. .
" ~
7.3.2 Steam Generator Primary Tubesheet-To-Head Weld Each steam generator includes a tubesheet-to-head weld h (approximately 36 feet in length). The entire length of ld
'each weld shall be ultrasonically examined during the I' inspection interval. The entire length of a tubesheet-to- 6i head weld shall be examined during the first and second
- inspection periods. The entire length of the two remaining welds shall be examined during the third I
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SONP SI-114.1 Page 15 of 85 Rev. 14 i 1rtspection period. See Table A of Appendix A for examinations during the inspection interval. l The tube plate is a SA-508 Class 2, steel forging, clad - on the primary side with NiCrFe alloy. The hemispherical chamber is a SA-216, Gr. WCC casting, clad with austenitic ', stainless steel. 7.3.3 Steam Generator Primarv Nozzle-To-Vessel Welds And Nozzle Inside Radius Section The steam generator primary nozzles are an integral part
~
of the vessel. The inside radii of all nozzles cannot be ultrasonically examined and achieve meaningful results due , to limitations of examining integrally cast material (see request for relief ISI-6). I 5 The nozzles are fabricated to SA-216, Gr. WCC. 7.3.4 Steam Generator Primarv Nozzle-To-Safe End Welds Each steam generator includes two nozzles with buttered safe ends. Each nozzle-to-safe end weld from each generator shall be ultrasonically and liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A. The nozzles have buttered 308L safe ends. 7.3.5 Steam Generator Primary Pressure Retainina Boltino Larcer Than Two Inches In Diameter . Thereisnopressureretainingbolting4argerthantwo - inches in diameter. 7.3.6 Steam Generator Primarv Pressure Retainina Boltino Two i Inches And Smaller In Diameter Each steam generator includes two manways. All the manway i bolts from each steam generator manway shall be visually examined in accordance with visual examination method VT-1 (see Section 11.1). The examinations shall be distributed during the inspection interval in accordance with Table A of Appendix A. , ,
-The bolts may be examined in place under tension or when the bolts are removed. It is preferable to perform the examinations when the bolts are removed if possible.
Removal of bolting is performed in accordance with MI-3.1 and provides for examination of bolting. 0963o/COC4 - - . . - _ - . ~ . n - - . .- a .- - -
SONP l SI-114.1 . Page 16 of 85 ! Rev. 14 'l F l : Each manway includes 16 bolts at 1.88 inches in diameter. (. 7.3.7 Steam Generator Primary Inteorally Welded Support ,, Attachments - There are no integrally welded vessel supports. The four main support pads are secured to the steam generator field support system by high strength bolts. 7.3.8 Steam Generator Tubino !- Each steam generator tube bundle consists of 3,388 NiCrFe , alloy (Inconel SB-163) U-tubes of 0.875 0.D. by 0.050 % average wall thickness. During the inspection interval, steam generator tubing ., shall undergo eddy current examination. This examination e, is done in accordance with Sequoyah technical specifica- i tions and satisfies Surveillance Requirement 4.4.5.0. The ' initial and additional samples (as required) lists identifying the steam generator tubes to be examined shall - be created and supplied by NDE Section XI Programs with , the assistance of the Steam Generator Support group for ,, inclusion in the scan plan. 7.3.8.1 Steam Generator Sample Selection And Inspection Each steam generator shall De determined operable during the shutdown by selecting and inspecting r at least the minimum number of steam generators 5 specified in Table 1 of Appendix A. - 7.3.8.2 Steam Generator Tube Sample Selection and Inspection [c.2) i The steam generator tube minimum sample size inspection result classification, and the corresponding action required shall be as specified in Table 2 of Appendix A. The inservice inspection of steam generator tubes ... shall be performed at the frequencies specified ,; in Section 7.3.8.3, and the inspected tubes shall .. be verified acceptable per the acceptance ? criteria of section 7.3.8.4. The full length of , the tubes selected for inspection will be p examined (TVA commitment to NRC L44 850617 801). - Note that the hot leg inspection sample and the + cold leg inspection sample do not necessarily involve the same tube, i.e., it does not preclure making separate entries from the hot and cold leg 0963o/COC4 9
' W 'O p e -ee - no e aqu e.m e 4 n.,
b .j SONP SI-114.1 ! Page 17 of 85 ; Rev. 14 (; u I.
. sides and selecting different tubes on.the hot P leg and colo leg sides to meet the minimum .
i. sample. The tubes selected for each inservice . Inspection shall include at least 3% of the total- l~ number of tubes in all steam generators 1 the tubes selected for these inspections shall be' selected on a random basis except: y
- a. Where experience in similar~ plants with similar water chemistry indicates critical ;;
areas to be inspected, then at least 50% of- j the tubes inspected shall be from these ,
~
critical areas. i;
- b. The first sample of tubes selected for each,
- inspection (subsequent to the preservice !
inspection) of each steam generator shall 4 include. r
- 1. All nonplugged tubes that previously had 1 detectable wall penetrations >20%. t
- 2. Tubes in those areas where experience.has "
indicated potential problems. ,t
- 3. A tube inspection (pursuant.to Section I; 7.3.8.4.a.8) shall be performed on each .,
selected tube. If any selected tube does , not permit the passage of the eddy current probe for a tube inspection, this L. shall be evaluated and recorded and an P, adacent tube shall be selected and I i, subjected to a tube inspection. I'
- c. The tubes selected as the second and third samples (if required by Table 2 of Appendix A) during each. inservice inspection may be sub- '
jected to a partial tube inspection provided: r.
- 1. The tubes selected for these samples -
include the tubes from those areas of the "! - tube sheet array where tubes with E imperfections were previously found.
- 2. The inspections include those portions of N the tubes where imperfections were k'
previously found. 0963o/COC4 ' !' s ._ J =_a l -__-_:__z_u- --. _--e. e_-__ :c :m. - - - - - _ rs s a--- -m- -e-~~.- - -- -~~
I SQNP S1-114.1 Page 18 of 85 Rev. 14 i The results of each sample inspection shall be classified into one of the following three , categories: t Category Inspection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none - of the inspected tubes are defective. C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the -; total tubes inspected are degraded tubes. C-3 More than 10% of the total tubes
- inspected are degraded tubes or more than 1% of the inspected tubes are .
defective. NOTE: In all inspections, previously degraded tubes must exhibit significant ( 10%) , further wall penetrations to be included in the above percentage calculations. 7.3.8.3 Inspection Frequencies The above required inservice inspections of steam generators shall be performed at frequencies indicated in the following paragraphs and in such
- a manner that the maximum allowable time between i eddy current inspections on individual steam +
generators is 72 months.
- a. The first inservice inspection shall be per-formed af ter, 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months ,
If two after the previous inspection. - consecutive inspections following service under AVT conditions, not including the preservice inspection, result in all ' i inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the 0963o/COC4 i
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SONP SI-il4.1 Page 19 of 85 Rev. 14 ,
, inspection interval may be extended to a ,
maximum of once per 40 months. -
- b. If the results of the inservice inspection of I a steam generator conducted in accordance with Table 2 of Appendix A at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once '
per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of section : 7.3.8.3.a: the interval may then be extended - to a maximum of once per 40 months.
.i
- c. Additional, unscheduled inservice inspections '
shall be performed on each steam generator in accordance with the first sample inspection ;. specified in Table 2 of Appendix A during the j shutdown subsequent to any of the following conditions.
- 1. Primary-to-secondary tubes leaks (not including leaks originating from tube-to- -
tube sheet welds) in excess of the limits of Technical Specification 3.4.6.2. k
- 2. A seismic occurrence greater than the '
Operating Basis Earthquake.
- 3. A loss-of-coolant accident requiring actuation of the engineered safeguards.
., 4. A main steam line or feedwater line break. \
7.3.8.4 Acceptance Criteria I
- a. As used in Section 7.3.8: '
- l. Imperfection means an exception to the dimensions, finish or contour of a tube '
from that required by fabrication drawings . or specifications. Eddy-current testing - indications below 20% of the nominal tube % wall thickness, if detectable, may be ( considered as imperfections. p
- 2. Degradation means a service-induced cracking, wastage, wear or general corrosion occurring on either inside or outside of a tube.
09630/COC4 i
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SQNP SI-114.1 , Page 20 of 85 i i Rev. 14 , j
~
- 3. Degraded Tube means a tube containing 4 imperfections 20% of the nominal wall -
thickness caused by degradation.
- 4. Percent Degradation means the percentage . t of the tube wall thickness affected or ,
1 removed by degradation. .
- 5. Defect means an imperfection of such J severity that it exceeds the plugging limit. A tube containing a defect is defective. .
- 6. Pluccina Limit means the imperfection depth at or beyond which the tube shall be removed from service because it may - '
become unserviceable prior to the next ' inspection and is equal to 40% of the nominal tube wall thickness. ,.
- 7. Unserviceable describes the condition of '
a tube if it leaks or contains a defect large enough to affect its structural ' integrity in the event of an Operating . Basis Earthquake, a loss-of-coolant , accident, or a steam line or feedwater ' line break as specified in Section 7.3.8.3.c.
- 8. Tube Inspection means an inspection of the steam generator tube from the point .
of entry (hot leg side) completely around L the U-bend to the point of exit (cold leg . side) 1.e., tube end to tube end. Entries may be made from either the hot or cold leg sides and separate entries on the hot. leg and cold leg sides on different tubes are allowed.
- 9. Preservice Insoection means a tube ,
inspection of each steam generator tube . e performed by eddy current techniques e prior to service to establish a baseline l l condition of the tubing. This inspection shall be performed prior to initial power 3 operation using the equipment and i techniques expected to be used during ; subsequent inservice inspections. l 1 0963o/COC4 ax_ : - - - .:- . : t? - -- --* - - - ~ ~ * - ~ - -- ~ ~ - - - - ~~*
SONP SI-114.1 Page 21 of 85 Rev. 14 ,
. b. The steam generator shall be determined i' operable after completing the corresponding' ,
actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 2. 7.4 Piping All ASME Code Class 1 piping systems to be examined are fabricated i of stainless steel. The reactor coolant main loop piping straight lengths are centrifugal cast and the elbows are static cast. The ;. upper head injection auxiliary head adapter is included in Section 7 1.1.3. Specific material specifications for each piping system are included in weld map isometrics in Appendix B (see Request for . < Relief ISI-3). ,i The following Class 1 piping systems are subject to examination: Reactor Coolant : Chemical and Volume Control Residual Heat Removal i Safety Injection
- Upper Head Injection
- 7.4.1 Pipino Dissimilar Metal Welds if There are no Class 1 dissimilar metal welds.
7.4.2 Piping Pressure Retainino Bolting larger Than 2 Inches In Diameter
-1
-There is no piping pressure retaining bolting larger than !
!2 inches in diameter. ,
7.4.3 Piping Pressure Retaining Bolting 2 Inches and Smaller in i! Diameter
]
The f'ollowing sections define the number of bolted piping flange connections in each system. All of the bolts or { studs and nuts in each flange connection shall be visually 1 examined during the inspection interval in accordance with , visual examination method VT-1 (See Section 11.1).. The :- examinations shall be distributed during the inspection h interval in accordance with Table A of Appendix A. ;<
- )
- The bolting may be examined in place under tension or when r the bolting is removed. -
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SQNP SI-114.1 Page 22 of 85 Rev. 14 ,
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- 7. 4_. 3.1 Reactor Coolant System Piping Bolting The Reactor Coolant System piping includes bolted flange connections. ,
7.4.3.2 Chemical and Volume Control System Piping Bolting The Chemical and Volume Control System piping - includes bolted flange connections (from seal water injection). ! 7.4.3.3 Residual Heat Removal System Piping Bolting j, . The Residual Heat Removal System piping does not .. include any bolted connections. {, 7.4.3.4 Safety Iniection System Pipino Bolting ( The Safety Injection System piping includes ' bolted flange connections. ' 7.4.3.5 Upper Head Injection System Pipina Boltina The Upper Head Injection System piping includes 9 bolted clamp / seal (grayloc) connectors. 7.4.4 Circumferential and Longitudinal Pipe Welds ' The entire length of each circumferential pipe weld
.alected for examination shall be ultrasonically and/or liquid penetrant examined as practical. All Class 1 T piping is seamless. :
Circumferential pipe welds four inches and greater nominal : pipe size selected for examination shall be ultrasonically : and liquid penetrant examined. Circumferential pipe welds less .than four-inch nomical pipe size selected for examination shall be liquid penetrant examined. The examinations performed during the inspection interval L shall include approximately 25 percent of the 40-year '; sample of circumferential welds. The examinations shall be distributed during the inspection interval in f-accordance with Table A of Appendix A. p.
~7.4.4.1 Reactor Coolant System Main loop'Pipina .
Circumferential Welds F' The Reactor Coolant System main loop piping includes circumferential pipe welds 4 inches and
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l :, o l_ L SQNP SI-ll4.1 [ Page 23 of 85 ! Rev. 14 i _ greater nominal pipe size. The welds selected for examination shall be ultrasonically and " liquid penetrant examined during the inspection C interval in accordance with Table A of Appendix A (see Request for Relief ISI-7). v 7.4.4.2 Reactor Coolant System Piping Circumferential Welds . Reactor Coolant System piping includes circum- . L ferential pipe welds 4 inches and greater nominal . l pipe size. The welds selected for examination ' shall be ultrasonically and liquid penetrant f' examined during the inspection interval in :. accordance with Table A of Appendix A. There are ' pipe welds less than 4-inch nominal pipe size. - The welds selected for examination shall be liquid penetrant examined during the inspection {' interval in accordance with Table A of Appendix A. 7.4.4.3 Chemical and Volume Control System Piping , l Circumferential Welds 3 The Chemical and Volume Control System piping S (including seal water injection) includes ' circumferential welds less than 4-inch nominal ' pipe size. The welds selected for examination shall be liquid penetrant examined during the inspection interval in accordance with Table A of . Appendix A. There are no Class 1 pipe welds 4 inches and greater in the CVCS system.
.J.4.4.4 Residual Heat Removal System Piping
- Circumferential Welds ;
The Residual Heat Removal System piping includes ' circumferential. welds 4 inches and greater nominal pipe size. The welds selected for ,i examination shall be ultrasonically and liquid penetrant examined during the inspection interval , in accordance with Table A of Appendix A. There ' r are no Class 1 pipe welds less than 4-inch '
)
F nominal pipe size in the RHR system. 7.4.4.5 Safety Iniection System Piping Circumferential I< Welds. U The Safety Injection System piping includes circumferential pipe welds 4 inches and greater nominal pipe size. The welds selected for 1
-23 0963o/COC4 -!
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l SONP 51-114.1 , Page 24 of 85 l Rev. 14 e examination shall be ultrasonically and liquid . penetrant examined during the inspection interval . , in accordance with Table A of Appendix A. There :~ are pipe welds less than 4-inch nominal pipe size. The welds selected for examination shall be . liquid penetrant examined during the inspection 4 interval in accordance with Table A of Appendix A. / 7.4.4.6 Upper Head Iniection System Piping Circumferential Welds The Upper Head Injection System piping includes h circumferential pipe welds 4 inches and greater ,l nominal pipe size. The welds selected for ' examination shall be ultrasonically and liquid ' penetrant examined during the inspection interval in accordance with Table A of' Appendix A. There ; are no piping welds less than 4-inch nominal pipe - size. . 7.4.5 Branch Piping Connection Welds The entire length of each branch pipe connection weld 2 selected for examination shall be examined. Branch pipe
. connection welds exceeding 2 inches nominal pipe size i selected for examination shall be ultrasonically and '
liquid penetrant examined. Branch pipe connection welds 2 inches nominal pipe size and smaller selected for ' examination shall be liquid penetrant examined. The examinations performed during the inspection interval e Tshall include approximately 25 percent of the branch pipe [
' connection welds. The examinations shall be distributed t
during the inspection interval in accordance with Table A ' of Appendix A. " NOTE: In the case of branch pipe connections 2 inches nominal pipe size and smaller, the welds of all the systems affected have been added and shall be evenly distributed over the four inspection intervals. . 7.4.5.1 Reactor Coolant Svstem Main Loop Branch Pipe f Connection Helds i.,
' i The Reactor Coolant System Main Loop piping includes a branch pipe connection weld exceeding (
2 inches nominal pipe size. This weld shall be ' ultrasonically and liquid penetrant examined during the inspection interval in accordance with t 24 0963o/COC4
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n7 SQNP SI-114.1 Page 25 of 85 Rev. 14 _ Table A of Appendix A. There are branch pipe connection welds 2 inches nominal pipe size and - smaller. The welds selected for examination shall be liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A. NOTE: Each of these branch pipa connections 2 inches nominal pipe size and smaller has a special boss. This is a special boss for use with temperature elements with outside diameter (OD) of 0.875". The special boss has an OD - 21/2" .01 with and inside diameter (ID) of 0.9" reamed to provide 0.015" - 0.020" clearance with 0.010" misalignment. Assuming the worst possible case (the weld to pipe fails) the largest hole is that for a 2" pipe (any schedule). Therefore, we shall treat it as a 2" branch pipe for ASME Section XI purposes. See contract 68C60-91934, (N2M-2-8) drawing nos. 206C470, 206C471
, and installation detail N2M-2-50.
7.4.5.2 Reactor Coolant System Branch Pipe Connection Welds The Reactor Coolant System piping includes branch pipe connection welds exceeding 2 inches nomiral
. pipe size. The welds selected for examination shall be ultrasonically and liquid penetrant
.'. examined during the inspection interval in accordance with Table A of Appendix A. There are branch pipe connection welds 2 inches nominal pipe size and ' smaller. The welds selected for examination shall be liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A.
, 7.4.5.3 Chemical and Volume Control System Branch Pipe .
.; Connection Helds The Chemical and Volume Control System piping A includes branch pipe connection welds exceeding 2 inches nominal pipe size. The welds selected
~.-
( for examination shall be ultrasonically and
- liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A.
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SQNP SI-114.1 l Page 26 of 85 i Rev. 14 > I
. There is a branch pipe connection weld 2 inches i nominal pipe size and smaller. The welds -
selected for examination shall be licuid T penetrant examined during the insper. tion interval in accordance with Table A of Appendix A. ' 7.4.5.4 Residual Heat Removal System Branch Pipe , Connection Welds The Residual Heat Removal System piping includes branch pipe connection welds exceeding 2 inches ;' nominal pipe size. The welds selected for
~
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examination shall be ultrasonically and liquid penetrant examined during the inspection interval N in accordance with Table A of Appendix A. There ! are branch pipe connection welds 2 inches nominal . pig size and smaller in the RHR System. The ; wel o selected for examination shall be liquid- J penetrant examir,ed during the inspection interval }l in accordance with Table A of Appendix A. 7.4.5.5 Safety injection System B. ranch PiDe Connection Welds .! The Safety Injection System piping includes ;[ branch pipe connection welds exceeding 2 inches , nominal pipe size. The welds selected for - examination shall be ultrasonically and liquid penetrant examined during the inspection interval in accordance with Table A of Appendix A. There
- are branch pipe connection welds 2 inches nominal
; pipe size and smaller. The welds selected for
- examination shall be liquid penetrant examined ,
during the inspection interval in accordance with + Table A of Appendix A. 7.4.5.6 Upper Head Injection System Branch Pipe Connection Welds The Upper Head Injection System piping includes - branch pipe connection welds 2 inches nominal
- pipe size and smaller. The welds selected for examination shall be liquid penetrant examined ! l during the inspection interval in accordance with i Table A of Appendix A. There are no branch pipe .
connection welds exceeding 2 inches nominal pipe c size in the UHI system. 4 0963o/COC4 ., i A. 3 -
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t. 7 SONP , 51-114.1 1-Page 27 of 85 l Rev. 14 !. p .. 7.4.6 Pipino Socket Welds 'p
~
y. The entire length of each socket. weld selected for ' examination shall be liquid penetrant examined. . The examinations performed during the inspection interval , shall include.approximately 25 percent of the socket welds. The examinations shall be distributed during the ! inspection interval in accordance with Table A of ' Appendix A. P 7.4.6.1 Reactor Coolant System Piping Socket Welds '
, a The Reactor Coolant System piping includes socket I, welds. The welds selected for examination shall k be examined during the inspection interval in !.-
accordance with Table A of Appendix A. ! 7.4.6.2 Chemical and Volume Control System Pipino Socket U Welds k j." The Chemical and Volume Control System piping includes socket welds. The welds selected for c; examination shall'be examined during the l inspection interval in accordance with Table A cf ' Appendix A. 7.4.6.3 Residual Heat Removal System Piping Socket Welds The Residual Heat Removal System piping includes socket welds. The welds selected for examination
- shall be examined during the inspection interval ;[:.
L in accordance with Table A of Appendix A. M, P 7.4.6.4 Safety Iniection Svstem Pipino Socket Welds # The Safety Injection System piping includes , socket welds. The welds selected for examination shall be examined during the inspection interval in accordance with Table A of Appendix A. "ll 7.4.6.5 Upper Head Injection System Pipina Socket Welds l' i The Upper Head Injection System piping includes ji4 socket welds. The welds selected for examination i. shall be examined during the inspection interval [.' in accordance with Table A of Appendix A. p F
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I SQNP ;. ! SI-114.1 i =I Page 28 of 85 ; ! Rev. 14 !. i 7.4.7 Piping and Valve Integrally-Welded Support Members ' f Integrally-welded support members include the support ' attachments of piping required to be examined by , Examination Category B-J. Included are those supports f. which have attachment welds to the valve and piping ' pressure retaining boundary and those attachments whose
- support base material design thickness is S/8 inch and greater. The entire length of each support attachment ;
weld selected for examination shall be surface examined. -
,l The examinations performed during the inspection interval b.
shall include 100 percent of the integrally-welded support '
.i members. The examinations shall be distributed during the Inspection interval in accordance-with Table A of Appendix U A. F 7.4.7.1 Reactor Coolant System Piping and Valve jf Integrally-Welded Support Members The Reactor Coolant System piping includes I integrally-welded support members. All of these i shall be examined during the inspection interval [
in accordance with Table A of Appendix A. p 7.4.7.2 Chemical and Volume Control System Pipino and Valve Integrally-Welded Support Members I)l The Chemical and Volume Control System piping . does not include any integrally-welded support ' members. f. i7473
,. . . Residual Heat Removal System Pipina and Valve Integrally-Welded Support Members
[ The Residual Heat Removal System piping includes integrally-welded support members. All of these , i supports shall be examined during the inspection , interval in accordance with Table A of Appendix A. ' 7.4.7.4 Safety Injection System Piping and Valve
- b '
Integrallv-Welded Support Members [ r i The Safety Injection System piping includes y' 1 integrally-welded support
- members. All of these d I shall be examined during the inspection interval "4 in accordance with Table A of Appendix A. !
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SI-114.1 i Page 29 of 85 Li Rev. 14 ; I 7.4.7.5 Upper Head Injection System Piping and Valve Integrally-Welded Supports : r The Upper Head Injection System piping includes - integrally-welded support members. All of these shall be examined during the inspection interval in accordance with Table A of Appendix A. 'l i 7.4.8 Piping and Valve Component Supports
- All piping and valve component supports of piping required to be examined by Examination Category B-J shall be l
~
visually examined during the inspection interval in accordance with visual examination methods VT-3 and VT-4 as applicable (see section 11.1). This examination includes integrally-welded and nonintegrally-welded component supports. Component supports extend from the , i piping and valves to and including the attachment to the i supporting structure. The setting of snubbers, shock absorbers, and spring-type hangers shall be verified in accordance with the accept-ance criteria shown in the Scan Plan or the applicable work instruction. The setting of snubbers and shock ' absorbers is performed in accordance with SNP SI-162.2 or the MI-6.13 series. The examinations shall be distributed during the inspec- ) tion interval in accordance with Table A of Appendix A. 7.4.8.1 Reactor Coolant System Piping and Valve Component l Supports The Reactor Coolant System piping includes component supports. All of these shall be ; examined during the inspection interval in accordance'with. Table A of Appendlx A. 7.4.8.2 Chemical and Volume Control System Piping and Valve Comoonent Supports F The Chemical and Volume Control System piping f includes component supports. All of these shall Lj be examined during the inspection interval in F; accordance with Table A of Appendix A. lj l 29- 0963o/COC4
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'7 4.8.3 Residual Heat Removal System Pipina and Valve $'
. Component Supports . ;;
The Residual . Heat Removal System piping includes l' component supports. All-of these shall be examined during the inspection interval in '. accordance with Table A of Appendix A. 7.4.8.4 Safety Injection System Pipina and Valve ~ Component Supports T The Safety Injection System piping includes " component supports. All of these shall be examined during the inspection interval in ' accordance with Table A of Appendix A. 7.4.8.5 Upper Head Injection System and Valve Piping Component Supports - The Upper Head Injection System piping includes component supports. All of these shall be : examined during the inspection. interval in accordance with Table A of Appendix A. 7.4.8.6 Reactor Coolant System Main Loop Pipino and Valve . Component Support , The Reactor Coolant System Main Loop Piping includes component supports. All of these shall be examined during the inspection interval in ; accordance with Table A of Appendix A. , l 7.5 Reactor! Coolant Pumps (4) - RCP .. 7.5.1- RCP Pressure Retaining Boltino Larger Than Two Inches In Diameter ' The main flange on each pump includes 24 bolts of 4-1/2 inches in diameter and 30-1/2 inches in length. Bolting may be examined either (a) in place under tension, , (b) when the connection is disassembled, or (c) when the [_ bolting is removed. L All the bolts from each pump shall be ultrasonically
" examined during the inspection interval . All threads in ,
the base material and flange ligaments between threaded l' bolt holes shall be visually examined in accordance with visual examination method VT-1 (see section 11.1) when the connection is disassembled.
-30_ 0963o/COC4 ;
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SONP SI-114.1 " Page 31 of 85 ! Rev. 14 L 6 Whep the bolts are removed each shall be magnetic particle I examined. This exam may be deferred until the end of the , f' i nte rva l .. Provisions for this examination are included in MI-2.2. O f .' The main flangt bolts are fabricated of 4340 steel, heat treated to A-540, GR. 24. - 7.5.2 RCP Pressure Retainino Boltino 2 Inches and Smaller in p Diameter ;, f Each RCP includes two sets of pressure retaining bolting 2 - inches and smaller in diameter. The bolting sets include 1",l the number 1, and 2 seal assembly bolting. -The number 1 seal assembly bolting from each pump shall be visually ' examined in accordance with visual examination method VT-1 (see Section 11.1). The number 2 seal assembly bolting is . Class 2 and does not require examination (see memo to '. C. R. Brimer from V. A. Blanco dated January 10, 1987, I No. B25870109045). .. t The bolting may be examined in place under tension or when .. removed. It is preferable to perform the examination: ? when the bolts are removed if the connection (s) is ' disassembled. Removal of bolting is performed in ' accordance with MI-2.1 and-provides for examination of bolting. All of the bolting from one RCP shall be examined during . the first inspection period, and all of the bolting from a
-different pump shall be examined during the second
'J nspection period. . All of the bolting from the remaining
[ two pumps shall be examined during the third inspection ~ period. ! The Number 1, and 2 seal. housing bolting includes 12 - socket head cap screws at 2 inches in diameter,12 socket . , head cap screws at 1 inch in diameter, respectively. 7.5.3 RCP Integrally-Welded Support Members c There are no integrally-welded support components associated with the RCP. L 7.5.4 ' RCP Component Supports ;-
?*
Each RCP includes three integrally cast pump feet bolted to the support system. All conponent supports from each pump shall be visually examinec during the inspection
, i 0963o/C0C4
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SQNP g SI-114.1 ' Page 32 of 85 { Rev. 14 !! r
. interval in accordance with visual examination method VT-3 (see Section 11.1). Support components extend from the
[ RCP to and-including the attachment to the supporting structure. ,, All of the supports from one RCP shall be examined during a the first inspection period, and all of the supports from f a different pump shall be examined during the second inspection period. All of the supports from the remaining i two pumps shall be examined during the third inspection period. ju 7.5.5 RCP Casing Helds I:! i t Each RCP casing includes a two-piece welded type 304SST '.' casting. The casing welds cannot be ultrasonically .. examined and achieve meaningful results due to limitations ~ of examining integrally cast material. The entire length t1 of one RCP casing weld shall be liquid penetrant examined ~ during the inspection interval. This examination may be deferred to the third inspection period. (See Request for : Relief ISI-8.) 7.5.6 RCP Casing I If a pump is disassembled for maintenance during the ., inspection interval, the internal pressure boundary surfaces shall be visually examined in accordance with visual examination method VT-1 (see Section 11.1). Disassembly of RCP's is performed in accordance with i.
.MI-2.2 and provides for these visual examinations. i.
Ifduringtheinspectionintervalapumpfromeitherunit ? 1 or unit 2 is not disassembled for maintenance, a pump 4 from one unit shall be examined from the exterior by ultrasonic thickness measurements. (See Request for ' Relief ISI-1.) . 7.5.7 RCP Fivwheel 4 At approximately 3-year intervals each pump flywheel shall . undergo an in-place ultrasonic examination of areas of higher stress concentration at the bore and keyway. A complete in-place ultrasonic examination shall be > S conducted at approximately 10-year intervals along with a b
' surface examination of exposed surfaces. (See Regulatory i~
Guide 1.14.) This examination is done in accordance with Sequoyah technical specifications and satisfies Surveillance Requirement 4.4.10.
-32 0963o/COC4 - c
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SONP. < 51-114.1 ,
, Page 33 of 85 [
Rev. 14 i;.. U The flywheel consists of 2 plates, approximately 5 inches and 8 inches thick, bolted together. Each plate is (!' fabricated from vacuum degassed A-533, GR. B, Class 1 *i steel. ! 7.6 Valves Systems including ASME Code Class 1 valves subject to examination h are identified in Section 7.4. A tabulation of valves is : presented in Appendix C. p 7.6.1 Valve Pressure Retaining Boltino Larger Than 2 Inches in '. Diameter a There are no valver with pressure retaining bolting larger . than 2 inches in diameter. '
/,
M 7.6.2 Valve Pressure Retaining Boltina 2 Inches and Smaller in : Diameter
~
The following sections define the number of valves with bolted bonnet connections in each system. All of the bolts or studs and nuts in each connection shall be l visually examined during the inspection interval in J. accordance with visual examination method VT-1 (see l Section 11.1). The examinations shall be distributed , during the inspection interval in accordance with Table A ' of Appendix A. , ', The bolting may be examined in place under tension or when the bolting is removed. It is preferable to examine the
; bolting when removed if possible. Valve disassembly is l-
' performed in accordance with MI-6.15 and provides for examination of bolting. [-
7.6.2.1 Reactor Coolant System Valve Bolting TheReactorCooiantSystemincludesvalveswith l bolted bonnet connections. All of these shall be e L examined during the' inspection interval in " accordance with Table A of Appendix A. , 7.6.2.2 Chemical and Volume Control System Valve Boltino [ The Chemical and Volume Control System includes valves with bolted bonnet connections. All of TJ these shall be examined during the inspection
- interval in accordance with Table A of Appendix A. *
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SONP I SI-114.1 l L~ Page 34 of 85 Rev. 14 f[ ll 01
- 7. ft. 2. 3 Residual Heat Removal System Valve Bolting
- f. I ri n ("
r The Residual Heat Removal System includes valves Oj g with bolted bonnet connections. All of these
- shall be examined during the inspection interval ,'. l in accordance with Table A'of Appendix A. 0 'I
- . i; 7.6.2.4 Safety Injection System Valve Bolting '
The Safety Injection System includes valves with (* bolted bonnet connections. All of these shall be y examined during the inspection interval in l L - accordance with Table A of Appendix A. t"' 7.6.2.5 Upper Head Injection System Valve Boltino {. l The Upper Head Injection System includes valves y with bolted bonnet connections. All of these ;. shall be examined during the inspection interval d in accordance with Table A of Appendix A- ~ 7.6.3 Valve Integrally-Welded Support Members Examination of valve integrally-welded support members is [ included in Section 7.4.7. > n 7.6.4 Valve Component Supports Examination of valve component supports is included in , Section 7.4.8. T 7.6.5 -Valve Body Welds ' There are'no valves with body welds. n. 7.6.6 Valve Bodies The internal pressure boundary surfaces of valve bodies 'A exceeding 4-inch nominal pipe size shall be visually examined in accordance with visual examination method VT-1 ; (see Section 11.1). Examinations are limited to one valve q. within each group of valves that are of the same construc- ' tional design (i.e., globe, gate, or check valve), !L manufacturing method, and that are performing similar 1-
- functions in the system. p u .-
When it becomes necessary to disassemble any valve, S subject to internal surface visual examination, for normal
- maintenance purposes, the interior surface of the valve
- body will be visually examined and the results recorded.
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SQNP - SI-114.1 i Page 35 of 85 i , Rev. 14 i. ) 5 1 Disassembly of valves solely for' visual examination will EC not be performed (See ASME Section XI, Table IHB-2500, . T Examination Category B-M-2; also see Request for. Relief' "
,l ISI-2). '
y A tabulation of valves by groupings is presented in L Table 1 of Appendix C. Disassembly of valves is performed J in accordance with MI-ll.4'and provides for examination of . valve internal pressure boundary surfaces. l-1 7.7 Exempted Components Gomponents exempted from examination include component connections, piping., and associated valves and their support that are one inch - nominal pipe size and smaller, except for steam generator tubing; ' components connected to and part of the reactor coolant pressure ' boundary (defined.in 10 CFR 50, Section 50.2(V): revised : January 1, 1975) but exempted from Class 1 requirements by , regulations of the regulatory authority having jurisdiction at the ~ plant site; reactor vessel head connections and associated piping, ; 2 in. nominal pipe size and smaller, made inaccessible by control rod drive penetrations. # ' 7.8 Successive Examinations l' Areas of flaw indications evaluated in accordance with IHB 3122.4 and TI-51 that qualify for continued service shall be reexamined during the next three inspection periods listed in the inspection schedules. If the re-examinations reveal that the flaw indications - remain essentially unchanged for three successive inspections, then the component examination schedule may revert to the original schedule. Exceptions to this requirement may occur when the "' examination schedule is dictated by the augmented inspection requirements (see Section 20.0). ~ 7.9 System Pressure Tests All ASME Code Class 1 pressure-r'etaining components shall be subjected to a system leakage test in accordance 4ith IWB-5221 of ASME Section XI each refueling outage. This is performed in - accordance with 5I-146. During this test, attention must also be 3 given to the cono-seal in the CRDM's (reference memorandum from ' J. F. Murdock to D. L. Williams dated August 22, 1988; 846 880822 F 582). In addition, all Class 1 pressure-retaining components p'. shall undergo a system hydrostatic pressure test .in accordance ; with IWB-5222 of ASME Section XI at the end of the inspection t, interval. This is performed in accordance with SI-250. The F components shall be visually examined during the pressure tests in 5 accordance with visual examination method VT-2 (see Section 11.1). I All Class I system pressure tests shall be performed in accordance with Program Procedure 1502.03. r! 0963o/C0C4 i
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I' SQNP SI-ll4.1 - Page 36 of 85 i Rev. 14 ! I'; 8.0 M COMPONENTS SUBJECT TO EXAMINATION - ASME CODE CLASS 2 (EOUIVALENT) i' The ASME Code Class 2 components to be examined during the inspection / interval are outlined in the following paragraphs. Extent of ,: examination piping welds will be in accordance with paragraph IWC-2411 and Table IWC-2520 of the 1974 Edition, Summer 1975 Addenda, ASME i Section XI. The ISI Programs Section shall select areas to be examined . or the NDE Branch of NQA may assist in selecting areas to be examined.
/
Components that are exempted from examination in accordance with [ IWC-1220 of ASME Section XI are discussed in Section 8.12 of this program. p1
~
Where examinations specify a percentage of the total length of weld to be examined, the areas (s) examined shall be documented in the examina-tion report. Where a percentage of weld length is not referenced, the ;
- i entire weld length shall be examined. (See Request for Relief ISI-3.) ,
Table B in Appendix A supplies additional information such as reference I. drawing numbers and ASME Section XI Table-IWC-2500-1 examination categories. 8.1 Steam Generators (4) - 8.1.1 Steam Generator Secondary Side Circumferential Shell Welds . There are three circumferential shell welds at gross structural discontinuities on each generator. The entire length of these three shell welds from one steam generator shall be ultrasonically examined during the inspection interval. The number of welds to be examined during each Inspection period shall be in accordance with Table B of Sppendix A. One of the three welds on each steam generator is
- partially inaccessible for examination due to the upper steam generator support arrangement (weld nos. SGW-01, D2, D3, and D4; see Request for Relief ISI-4). The weld selected for examination shall be ultrasonically examined on a best-effort basis. '.
R' The vessel shell sections are fabricated of SA-533, GR A, ; Class 1, steel plate. - J 8.1.2 Steam Generator Secondary Side Circumferential Head Welds Each steam generator includes a circumferential head-to- yI shell weld. The entire length of one head-to-shell weld ' shall be ultrasonically examined during the inspection interval in accordance with Table B of Appendix A. The I r 0963o/COC4 I
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wejd selected for examination may be from the generator ?l selected for examination by Section 8.1.1. ti l The vessel head section is fabricated of SA-533, GR. A, - l Class 1, steel plate. - i 8.1.3 Steam Generator Secondary Side Tubesheet-To-Shell Weld h '
\
Each steam generator includes a tubesheet-to-shell weld. ...J The entire length of one tubesheet-to-shell weld shall be Li ultrasonically examined during the inspection interval in ; j accordance with Table B of Appendix A. The weld selected ?, 4 for examination may be from the generator selected for ~. ! examination by section 8.1.1 The tube plate is a SA-508, Class 2, steel forging. 8.1.4 Steam Generator Secondary Side Nozzle-To-Vessel Welds Each steam generator includes one feedwater nozzle (3.62 inches nominal vessel thickness) and one main steam nozzle (3.62 inches nominal vessel thickness). All of the nozzle-to-vessel welds from each generator shall be , ultrasonically and magnetic particle examined during the , inspection interval in accordance with Table B of Appendix A. The nozzle inside radius sections on each nozzle will be ultrasonically examined once during the inspection o ' interval. i The nozzles are fabricated of SA-508, Class 2, steel. 8.1.5 -Steam Generator Secondary Side Integrallv-Welded Support Attachments s. There are no integrally welded vessel support attachments. . 8.1.6 Steam Generator Secondary Side Comoonent Supports
. There are no component supports (including mechanical and hydraulic supports) which are in contact with the vessel.
8.1.7 Steam Generator Secondary Side Pressure Retaining Bolting Exceeding 2 Inches in Diameter There is no steam generator secondary side bolting (
- exceeding 2 inches in diameter. t E
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.Page 38 of 85
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1-8.2 Residual Heat Removal Heat Exchangers (2) - RHRHX . 8.2.1- RHRHX Circumferential Welds
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8.2.1.1 RHRHX Shell Circumferential Weld c.,, N There is one circumferential shell weld located .j' at a gross structural discontinuity on each- '
?
RHRHX. The entire length of this shell weld from : one. heat exchanger shall.be ultrasonically N examined during each inspection. interval ~in. , accordance with Table B of Appendi.* A. The RHRHX shell section is fabricated from SS, SA-182, F304. d l[ 8.2.1.2 RHRHX Head Circumferential Weld ' There'is one circumferential head-to-shell weld k per RHRHX. -The entire length-of one head-to- '- shell weld shall be ultrasonically examined s' during the inspection interval in accordance with 7 Table B of Appendix A. The weld selected for e examination may be from the heat exchanger
- selected for examination by section 8.2.1.1. .7
^
The channel cylinder section (shell) and channel head are one inch thick fabricated from SS, . SA-240, TP-304.
~'
-8.2.1.3 RHRHX Tubesheet to Shell Weld .
~
1 t
- i. There are no RHRHX tubesheet-to-shell welds. S 8.2.2 RHRHX Nozzle-to-Vessel Welds [
1: The channel cylinder section of each RHRHX includes one i inlet nozzle (14-inch I.D.) and one outlet nozzle (14-inch I.D.) over 1/2-inch nominal thickness. A total of four - nozzle-to-vessel welds from the two RHRHX will be . . ultrasonically and liquid penetrant examined during the
- inspection interval. -The nozzle inside radius sections .
will be ultrasonically examined during the inspection 3 interval. cc
'The nozzles are 2.5 inches thick, fabricated from SS, h SA-240, TP-304. k' 0963o/COC4 ;
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1 SQNP SI-114.1 Page 39 of 85 [ Rev. 14 ' b ! 8.2.3 RHRHX Integrally-Welded Support Attachments I i There are two integrally-welded support attachments . (1-inch) on each RHRHX whose base material exceeds - 1/2-inch nominal thickness. A total of two support pad-to-vessel welds from the RHRHX will be liquid ^ penetrant examined during each inspection interval in ' accordance with Table B of Appendix A. The welds selected ' for examination may be conducted on one heat exchanger and shall cover 100 percent of the required area of each , support attachment. ' The support pad is fabricated from SS, SA-240, TP-304. . 8.2.4 RHRHX Component Supports , There are two component supports on each RHRHX which are ' in contact with the vessel. All component supports from ' each heat exchanger shall be visually examined during the inspection interval in accordance with visual examination method VT-3 (see Section 11.1). This examination includes integrally-welded and nonintegrally-welded component supports. Component . supports extend from the heat exchanger to and including ; the attachment to the supporting structure. , The examinations shall be distributed during the inspection interval in accordance with Table B of Appendix A. S-
-There are no mechanical (snubbers) and/or hydraulic (shock '
tabsorbers) supports. which are in contact with the vessel.
- 8.2.5 RHRHX Pressure Retainino Boltino Exceedino 2 Inches in Diameter There is no RHRHX bolting exceeding 2 inches in diameter.
8.3 Regenerative Heat Exchancer (One) - RHX 8.3.1
'P RHX Circumferential Helds '
8.3.1.1 RHX Shell Circumferential Welds t
?
The regenerative heat exchanger is composed of three heat exchangers interconnected with piping. '; There are twelve circumferential welds located in " at structural discontinuities on the heat exchanger. There are six channel cylinder to l' 0963o/C0C4 i 9 5 : = =:=: w w - : 8, , >= :- . :~-~-~ c -~n s ~w v - ~r~ - -
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1 SQNP SI-114.1 l' . , Page 40 of 85 II Rev. 14 !! channel head welds and there are six tubesheet-to- : shell. welds. See Request for Relief 151-12. [] The channel cylinder sections are fabricated to b SS, SA-351, CF8. The channel flanges are i-fabricated to SS, SA-182, F304. :, 8.3.1.2 RHX Head Circumferential Welds i' There are six channel cylinder to channel head ' welds. Relief from performing these examinations i is requested in ISI-12. Y The. channel heads are fabricated to SS, SA-240, 304L. c 8.3.1.3 RHX Tubesheet to Shell Weld r There are six.RHX tubesheet-to-shell welds. [t Relief from performing these examinations is .' requested in 151-12. The tubesheet is fabricated to SA-182, F304 SS. . 8.3.2 RHX Nozzle-to-Vessel Welds f There are not any nozzles greater than four inch diameter. - The nozzles are fabricated to SA-479, 304 SCH 160 material. 8.3.3 RHX Integrally-Welded Supports
-There are not any integrally-welded support attachments on 'f
'the heat exchanger. There are three stops 5/8" x 5/8" x ;t 1-3/4" welded to the heat exchanger. See Request for F Relief 151-12. '
8.3.4 R_HX Component Supports . I There are two component supports on the regenerative heat exchanger. These supports are not required to be examined ;, ) because of footnote 4, examination category C-E, Table. . IWC-2500-1 of ASME Section XI. ;- 8.3.5 RHX Pressure Retaining Bolting Exceedino Two Inches j_nn I
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Diameter -
- t; There is not any pressure retaining bolting associated y i
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SI-114.1 . Page 41 of 85 Rev. 14 I r 8.4 Excess Letdown Heat Exchanger (One) - ELHX i' 8.4.1 ELHX Circumferential Welds i B.4.1.1 ELHX Shell Circumferential Welds ;ll t There are not any shell circumferential weldt ,, ! associated with the ELHX. 8.4.1.2 I ELHX Head Circumferential Welds J There is one circumferential weld, the channel ' ' '
~
flange to channel head, located at a structural ' discontinuity on the head side of the ELHX. 4 { Relief from performing this examination is i requested in ISI-12.
.l The channel flange and the channel head are j fabricated to.SA182, F316 and SA-240, TP316, "
respectively. 8.4.1.3 ELHX Tubesheet to Shell Weld There are not any ELHX tubesheet to shell welds. 8.4.2 ELHX Nozzle to Vessel Weld ' There are not any nozzles greater than four-inch diameter.
- The nozzles are fabricated to SA-312, 316.
8.4.3 ELHX Inteorally-Weided Supports
~
[There are no integrally-welded supports'to the head side m. of the ELHX. n 8.4.4 ELHX Component Supports ', There are no component s'upports to the head side of the ' ELHX. ' 8.4.5 ELHX Pressure Retaining Bolting Exceeding Two Inches in .. Diameter ' I
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The channel flange bolting associated with the ELHX is fi less than two inches in diameter. h The flange includes 12 studs with nuts at 1-5/3 inch U diameter. The studs and nuts are fabricated to SA-193, GRB7 and SA-194, GR2H, respectively. 1 09630/COC4 t ._ . m. . - . . , , . . s . .m... ..
SONP SI-114.1 : Page 42 of 85 Rev. 14 [ t 8.5 Boron Iniection Tank (One) - BIT lt
. N 8.5.1 BIT Circumferential Welds (Shell and Head) l l .
There are two circumferential head-to-shel'. welds located t at structural discontinuities on the BIT. These welds. E shall be ultrasonically examined during the inspection ' t interval in accordance with Table B of Appendix A. The ' examination shall cover 100 percent of the weld length. ' t The head and shell are SA-264 material consisting of . SA-516, GR70 steel backing'outside with 1/8 inch SA-240, " TP304L cladding inside. 8.5.2 BIT Nozzle-to-Vessel Welds There are two nozzles, one located on each head with a six-inch inside diameter whose nominal thickness (2.00 f in.) is greater than 1/2 inch. These nozzle to ve'ssel ' welds shall be ultrasonically _y and surface examined during - the inspection interval in accordance with Table B of ;L Appendix A. The examination shall cover 100 percent of the i weld length. j 1-The nozzles are fabricated to SA-350, LF2. y v 8.5.3 BIT Integra11v-Welded Supports There are four integrally-welded support attachment pads welded to the shell, whose base material design thickness
- is 5/8 inch and therefore requires liquid penetrant examination in accordance with Table IWC-2500-1 of ASME iSection XI.
f i 8.5.4 BIT Component Supports ' There.are four component. supports associated with the boron injection tank. All of these supports shall be visually examined (VT-3) during the inspections interval in accordance with Table B of Appendix A. - 8.5.5 BIT Pressure Retainino Boltina Exceedino Two Inches j,n, Diameter I t.. i_ There are 16 manway cover studs at 2-1/2 inches in '
' diameter. All 16 studs shall be ultrasonically examined r during the inspection interval in accordance with Table B :.
of Appendix A. The studs may be examined in place under ? tension or when they are removed. It is preferable to perform the examinations when the studs are removed if possible. 0963o/COC4 i.____ " ..- .h i .'
t SONP S1-114.1 i, Page 43 of 85_ ! Rev. 14 ! ( The studs and nuts are fabricated to SA-193, GRB7 and is SA-194, GR2H, respectively. . j': 8.6 UHI Water Accumulator (One) WA h The UHI water accumulator is exempted from. examination because of c IWC-1220(a) of ASME Section XI. p 8.7 UHI Surae Tank (One) - ST The UHI' surge tank is exempted from examination because of IWC-1220(a) of ASME Section XI. !
~ ,
8.8 Seal Water Injection Filters (2) SWIF F-8.8.1 SWIF Shell and' Head Circumferential Welds Each SWIF has 2 circumferential shell welds, one top flange-to-shell and one bottom head-to-shell weld. According to Table IWC-2500-1 Category C-A footnote 3, we will; ultrasonically examine 2 welds of either the same i-filter or 2 different ones of each filter in accordance c, with Table B of Appendix A. These. exams should be , coordinated with the plant so as to perform the exams when the filters are drained..or flushed.to reduce. radiation exposure to the personnel. z., 8.8.2 SWIF Integrally Welded Supports Each SWIF has 3 integrally welded support legs. The legs are 2-1/2" x 2-1/2" x 3/8" angle iron of SA 304 stainless ,,
-steel and are exempt from examination due to their s
!. thickness being less than 1/2". %
r 8.8.3 SWIF Component Support F Each,5WIF component support consists of 3 integrally - welded support'1egs. The supports are exempt by footnote 4 to Category C-E of Table IWC-2500-1 since the piping to the filters are not required to be examined under . categories C-F or C-G. .t c. 8.9 Piping [. . Material specifications for each piping system are included in I,5 weld map isometrics in Appendix B. The following Class 2 piping i systems are subject to examination: * . Residual Heat Removal Safety Injection 0963o/COC4 !
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SONP ; SI-114.1 i Page 44 of 85 I Rev. 14 [: f Main Steam 1-Feedwater .
>pl j Containment Spray b Upper Head Injection ?
t: 8.9.1 Pipino and Valve Inteorally-Welded Support Members 7, Integrally-welded support members include the support i attachments of piping required to be examined by Examination Category C-F. Included are those supports ',9 which have attachment welds to the valve and piping pressure retaining boundary, and those attachments whose support base material desion thickness exceeds 3/4 inch. t. The entire length of each support attachment weld selected " for examination shall be surface examined. 4 r The examinations performed during the inspection interval
..[ .
shall include 100 percent of the integrally-welded support L. members. The examinations shall be distributed during the ;. inspection interval in accordance with Table B of ;.. Appendix A. *
- 8. 9.1.1 Residual Heat Removal System Pipino and Valve '
Integrallv-Welded Support Members
- 2 The Residual Heat Removal System piping does not ,
include any integrally-welded support members. '
- 3. 9.1.2 Safety Injection System Pipina and Valve '
Inteora ll y-Welded Support Members ' The Safety Injection System piping includes - integrally-welded support members. All of these C shall be examined during the inspection interval F in accordance with Table B of Appendix A. .
- 8. 9.1. 3 Main Steam Svstem Piping and Valve Inteorallv-
- Welded Support Memoers .
The Main Steam System piping includes integrally- E welded support members. All of these shall be . p examined during the inspection interval in P accordance with Table B of Appendix A. l, 8.9.1.4 Feedwater System Pipino and Valve E Inteorallv-Welded Support Members ? The Feedwater System piping includes integrally- [ l welded support members. All of these shall be ! i
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h-I SQNP I L p SI-114.1 j Page 45 of 85 n Rev. 14 [ l E ;' t ,. _ examined during the inspection interval in~ if accordance with Table B of Appendix A. , !g i 8.9.1.5 Containment Spray System Piping and Valve I. Intecrally-Welded Support Members
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The Containment Spray System piping does not i,- include any integrally-welded support members. 8.9.1.6 Upper Head Iniection System Piping and_,7 .ve l' Integrally-Welded Supports - T-i The upper head injection system piping includes "- integrally-welded supports. All of these shall [ be examined during the inspection interval in ~ accordance with Table B of Appendix A. J 8.9.2 Pipino and Valve Component Supports
- All piping and valve component supports shall be visually k examined during the inspection interval in accordance with visual examination methods VT-3 and VT-4 (see Section F
11.1). This examination includes integrally-welded and nonintegrally-welded component supports. Component , supports extend from the piping and valves to and i including the attachment to the supporting structure. ;, The setting of snubbers, shock absorbers and spring-type hangers shall be verified in accordance with the c. acceptance criteria thown in the Scan Plan or the ;, applicable work int u action. The setting of snubbers and D
-shock absorbers is performed in accordance with the F iSI-162.2 or the MI 6.13 series. U w
The examinations shall be distributed during the inspec- .- tion interval in accordance with Table B of Appendix A. [f 8.9.2.1 Residual Heat Re'moval System Pipinc and Valve Component Supports The Residual Heat Renoval System piping includes h.' component supports. All of these shall be *
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examined during the kspt tion interval in is accordance with Table 8 r. Appendix A. ' i 0963o/COC4 l
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SONP S1-114.1 l. Page 46 of 85 F Rev. 14 t
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8.9 2.2 Safety Injection System Piping and Valve .i . Component Supports . ,'3 E The Safety Injection System piping includes- I! component supports. All of these shall be ;1-examined during the inspection interval in r-accordance with Table B of Appendix A. Q U 8.9.2.3 Main Steam System Piping and Valve Component Supports ,: 4 t. The Main Steam System piping includes component y
~
supports. All of these shall be examined during 6 the inspection interval in accordance with
- Table B of Appendix A. .
8.9.2.4 Feedwater System Piping and Valve Component Supports ,' The Feedwater System piping includes component v supports. All of these shall be examined during / the inspection interval in accordance with Table B of Appendix A. < l 8.9.2.5 Containment Spray System Pipina and Valve h Component Supports I;. The Containment Spray System piping includes component supports. All of these shall be ,, examined during the inspection interval in , accordance with Table B of Appendix A. [. c
.9.2.6 Upper Head Injection System Piping and Valve p Component Supports p The upper head injection system piping includes ;
component supports. All of these shall be j examined during the inspection interval in ,- accordance with Table B of Appendix A. 8.9.3 Pressure-Retainino Boltina ./ R There is no Class 2 Pressure-Retaining Bolting larger than two inches in diameter. f.. d 8.9.4 'Circumferential and longitudinal Pipe Welds W
~,
N The entire length of each weld selected for examination E shall be ultrasonically and/or surfaced examined. Selection of welds for examination is based on Table , [
-46 0963o/COC4 E. - . . . - - . - -
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SQNP SI-Il4.1 , Page 47 of 85 i Rev. 14 L f: IWC-2520, Paragraph IWC-1220, and Paragraph IWC-2411 . F (Summer 1975 Addenda). All of the welds selected shall be L examined during the inspection interval and distributed in L accordance with Table B of Appendix A. Circumferential and longitudinal piping welds in piping f with a nominal wall thickness of 1/2 inch or less shall be ? surface examined. Circumferential and longitudinal piping welds in piping with a nominal wall thickness greater than 1/2 inch shall be Ultrasonically and surfaced examined. ', i The areas subject to examination include circumferential " pipe welds, at structural discontinuities within 3 pipe i, diameters of the centerline of rigid pipe anchors, anchors , at the penetration of primary containment or at rigidly anchored components, and longitudinal weld joints in pipe fittings.
'.3 .
8.9.4.1 Residual Heat Removal Piping ' The Residual Heat Removal Piping System includes 5 Class 2 circumferential and longitudinal piping welds with a nominal wall thickness greater than ! 1/2 inch subject to examination. The welds selected for examination shall be ultrasonically ' and surfaced examined each inspection interval in . accordance with Table B of Appendix A. The residual heat removal piping system includes the - circumferential and longitudinal pipe welds with - a nominal wall thickness of 1/2 inch or less. ; The welds selected for examination shall be surface examined each inspection interval in
- accordance with Table B of Appendix A. (7
[ 8.9.4.2 Safety Injection Pipino j The Safety Injection Piping System includes Class 2 circumferential piping welds with a nominal wall thickness greater than 1/2 inch subject to examination. The welds selected for - examination shall be ultrasonically and surface . examined each inspection interval in accordance with Table B of Appendix A. The safety injection '.; - piping system includes the circumferential and F i longitudinal pipe welds with a nominal wall If thickness of 1/2 inch or less. The welds N! selected for examination shall be surfaced f examined each inspection interval in accordance
- l with Table B of Appendix A.
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SONP l SI-114.1 ,4 Page 48 of 85 ! i Rev. 14 i; 8._9. 4 . 3 Main Steam Pipinq rc The Main Steam Piping System includes Class 2 4 circumferential and longitudinal piping welds with a nominal wall thickness greater than 1/2 6 inch subject to examination. The piping size is l' i 32" 0.D. x 1.088 wall unless noted. The - circumferential and longitudinal piping welds selected for examination shall be ultrasonically and surfaced examined each inspection interval in
- accordance with Table B of Appendix A. ;.
8.9,4.4 Feedwater Pipina
- The Feedwater ?iping System includes Class 2 '
circumferential piping welds with a nominal wall ? thickness greater than 1/2 inch subject to exam-Ination. The circumferential and longitudinal . piping welds selected for examination shall be ultrasonically and surfaced examined each inspection interval in accordance with Table B of ' Appendix A. 8.9.4.5 Containment Spray Piping h The Containment Spray Piping System includes Class 2 circumferential and longitudinal pipe welds with a nominal wall thickness of 1/2 inch or less, subject to examination. The circumfer-ential and longitudinal pipe welds selected for ,
~
examination shall be surface examined each
- inspection interval in accordance with Table B of *.
Appendix A. The nominal wall thickness for the
- containment spray system is 0.406 inches.
8.9.4.6 Upper Head Injection Piping The upper head injection piping system includes Class 2 circumferential welds and no Class'2 longitudinal piping welds with a nominal wall - thickness greater than 1/2 inch subject to . ~l examination. The circumferential welds selected ' for examination shall be ultrasonically and ' surface examined during the inspection interval l. in accordance with Table B of Appendix A. ;) 8.9.5 Branch Pipino Connection Welds I* There are no Class 2 Branch Pipe Connection Welds. 0963o/COC4 _ _ _ - . l__ ' _ _ - - _ - - - . _ _ _ _ . .._-_. . __..:. . - - - _ - . ~_ .N -- a. -
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l SQNP ' I SI-114.1 :! Page 49 of 85 E Rev. 14 [., z
.j 8.10 Pumps m
,g' L 8.10.1 Residual Heat Removal Pumps (2) - RHRP i. -
8.10.1.1 RHRP Integrally-Welded Supports '!. 1 There are no integrally-welded supports , associated with the RHRP. J 8.10.1.2 RHRP Support Components i Each RHRP includes three support components bolted to the pump feet which are integrally forged with the pump. All component supports from each pump shall be visually examined during
- the inspection interval in accordance with visual !
examination method VT-3 (see Section 11.1). .i Support components extenti from the RHRP to and l' l including the attachment to the supporting . structure. S{' i Two of the supports from one RHRP shall be examined during the first inspection period. The /1 final (3rd) support from the same RHRP shall be
- examined during the second inspection period. I All of the supports from the remaining RHRP shall ,
be examined during the third inspection period.
- 8.10.1.3 RHRP Supports - Mechanical or Hydraulic ' '
There are no mechanical or hydraulic supports i. associated with the RHRP. r, ;- l
' 8.10.1.4 RHRP Pressure Retaining Bolting F:
4 i L The stuffing box extension to pump casing connection bolting is not greater than two inches . in diameter. 1 The connection includes 24 studs at 1-1/4 inches ' in diameter with nuts and washers. The studs are , , fabricated to SA-193, GR. B7, and the nuts to ": SA-194, GR. 2H. !, L
, 8.10.1.5 RHRP Casing Welds f The RHRP does not include any casing welds. The A!
casing is a one piece forging fabricated to l SA-182 F304. l l 0963o/COC4
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~SQNP bof85 Rev. 14 [.;
c'. l 8.10.2 CVCS Centrifugal ~ Charging Pumps (2) CCP 4: f
~
Examination of these pumps, and their associated component supports, pressure retaining bolting, and casing welds are k.) t exempted under previsions outlined in table IWC-2500-1 and accompanying footnotes, examination categories C-C, C-G,
'9 '
which limit the' exam requirements to components in piping .c examined under examination category C-F. U 8.10.3L CVCS Positive Displacement Pump (1) Pop / Reciprocating Charging Pump - Examination'of these pumps, and their associated component supports, pressure retaining bolting, and casing welds are ' exempted under previsions outlined in table IWC-2500-1 and , accompanying footnotes, examination categories C-C,: C-G, ,' which limit the exam requirements to components in piping examineu under examination category C-F. q 8.10.4 Safety Iniection Pumps (2) SIP ", Examination of these pumps, and their associated component . supports, pressure retaining bolting, and casing welds are -- exempted under previsions outlined in table IWC-2500-1 and j accompanying footnotes, examination categories C-C, C-G,
- which limit the. exam requirements to components in.pipirg 1 examined under examination category C-F. '
8.11 Valves Systems including ASME Code Class 2 valves subject to examination are iderrtified in Section 8.9. A tabulation of valves is presented in Appendix C. /[ 8.11.1 Valve Inteorallv-Welded Supports ,' 2xamination of valve integrally-welded support members is included in Section 8.9.1. . 8.11.2 Valve Component Supports Examination of valve component supports is included in 'i Section 8.9.2. ; 8.11.3 Valve Pressure-Retaining Boltina T-
~
i. There is no Class 2 pressure-retaining bolting greater e than 2 inche's in diameter. 0963o/C0C4 [
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SQNP , SI-114.1 Page 51 of 85 [ Rev. 14 L 8.11.4 Valve Body Welds '" I There are no Class 2 valves with body welds. 0 8.12 System Pressure-Test y' All Class 2 pressure retaining components in systems or portions / of systems which are not required to operate during normal reactor ' operation but for which periodic system or component functional tests are required (excluding open-ended portions of systems), f i.all be subjected to a system functional leakage test with each ,- inspection period in accordance with IWC-5221 of ASME Section XI. ' This is performed in accordance with the SI-267 series. Those
- Class 2 pressure retaining components in systems or portions of E systems not subject to the functional leakage tests-(excluding '
open-ended portions of systems), shall be subjected to a system . hydrostatic test with each inspection interval in accordance with ; IWC-5222 of ASME Section XI. This is performed in accordance with . the SI-265 series. The components shall be visually examined A during the pressure tests in accordance with visual examination method VT-2 (see Section 11.1). All Class 2 pressure tests shall ' be performed in accordance with Program Procedure 1502.03. 8.13 Exempted Components 5- . 8.13.1 Exempted Components (Except Pipino Welds) - Components exempted from examination include: (a) l ' components of systems or portions of systems that during , normal plant optrating conditions are not required to , operate or perform a system function but remain flooded =.
-under static conditions at a pressure of at least 80% of c,
'the pressure that the component or system will be J subjected to when required to operate; or (b) components of systems or portions of systems, other than Residual '.
Heat Removal Systems and Emergency Core Cooling Systems, . that.are not required to operate above a pressure of 275 psig or above a temperature of 200*F; or (c) component , connections (including nozzles in vessels and pumps), , piping and associated valves and vessels (and their supports) that are 4 in. nominal pipe size and smaller. . 8.13.2 Exempted Components (Piping Welds Only) ' Piping exempted from examination include: (a) Piping t.
' systems where both the design pressure and temperature are y equal to or less than 275 psig and 200*F, respectively; *
(b) piping systems or portions of systems other than 1 emergency core cooling systems which do not function during normal reactor operation; (c) piping that is 4-inch nominal pipe size and smaller.
-51 09630/COC4 !
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SONP SI-114.1 i-Page 52 of 85 L. Rev. 14 Cl
?i 8.14 Successive Examinations Components with flaw indications evaluated in accordance with IWC-3000 and TI-51 that qualify for continued service shall be [T, reexamined during the next inspection period listed in the inspection schedule. If the reexamination reveals that the flaw ,,
indications remain essentially unchanged the component examination r schedule may revert to the original schedule. Exceptions to this '. requirement may occur when the examination schedule is dictated by ; the augmented inspection requirements (see Section 20.0). . 9.0 COMPONENTS SUBJECT TO EXAMINATION - ASME CODE CLASS 3 (E0VIVALENT) '5 ASME Code Class 3 (Equivalent) pressure retaining components identified ., in sections 9.1 and 9.2 shall be visually examined, VT-2, in conjunction ? with the system pressure test in accordance with IWA-5000, IWD-2000, and , IWD-5000. 1. The component supports and restraints within the boundaries identified in sections 9.1 and 9.2 for components exceeding 4-inch nominal pipe size shall be visually examined, VT-3, during each inspection period. Mechanical and hydraulic snubbers, spring loaded and constant weight U supports within the boundaries identified in sections 9.1 and 9.2 for components exceeding 4-inch nominal pipe size shall be visually , examined, VT-4 during each inspection period. 9.1 Pipino and Valve Compo9ent Supports All piping and valve component supports shall be visually examined . during each inspection period in accordance with visual examination ' methods-VT-3 and VT-4 (see Section 11.1). This examination - includes, integrally-welded and nonintegrally-welded component , supports. Component supports extend from the piping and valves to ' and including the attachment to the supporting structure. The . setting of snubbers, shock absorbers and spring-type hangers shall be verified 1.n accordance with the acceptance criteria shown in
- the Scan Plan or the applicable work instruction. The examina- ,
tions (1007.) shall be conducted during each inspection period during the inspection interval in accordance with Table C of . Appendix A. . t Examinations performed during any one inspection that reveal 'i indications exceeding allowable standards of IND-3000 (IWB-3000) h, (and for which 1007. of the supports for a given category are not c being examined) in a component of an examination category h-examinations shall be extended to include an additional number of Q' components within the same category, approximately equal to the ~~ ' number of components examined during the inspection. In the event 0963o/C0C4 E m .- .. . _ .. .. , . . , _ , m
SQNP j S1-114.1 l Page 53 of 85 l Rev. 14 [(
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further indications in excess of the allowable standards are .'. revealed, all of the remaining components within the same ; examination category shall be examined to the extent specified in rf Table IWD-2500-1. .' 9.1.1 Auxiliary Feedwater System Piping and Valve Component Supports t The auxiliary feedwater system piping includes component supports. All of these shall be examined during each , inspection period in accordance with Table C of Appendix A. ., 9:1.2 Chemical and Volume Control System Piping and Valve Component Supports The chemical and volume control system piping component - supports are not required to be examined because there is ! no examination category for this piping. (The change in support numbers during the second period was due to new ' class boundaries.) 9.1.3 Component Cooling System Pipino and Valve Component - Supports The component cooling system piping includes component " supports. All of these shall be examined during each inspection period in accordance with Table C of Appendix A. " 9.1.4 Containment Spray System Piping and Valve Comoonent Supports
-The containment spray system piping has no component l' isupports. (The change in support numbers during the second period was due to new class boundaries.)
l 9.1.5 Essential Raw Coolina Water System Piping and Valve Comoonent Supoorts . The essential raw cooling water system piping includes component supports. All of these shall be examined during each inspection period in accordance with Taole C of Appendix A. U, 9.1.6 Fuel Pool Cooling System Piping and Valve Component Supports
*)
yl The fuel pool cooling system piping includes component
- l1 supports. All of these shall be examined during each .
I inspection period in accordance with Table C of Appendix A. * { \ c) 0963o/COC4 { p w. .e o ao ee- **.m -"
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SONP i 5I-114.1 : Page 54 of 85 f. : Rev. 14 7 Y 9.1.7 Residual Heat Removal System Piping and Valve Component I-Supports -
,h The residual heat removal system piping has no component. ..
supports. (The change in' support numbers during the second period was due to new code class boundaries.) , 9.1.8 Safety Iniection System Piping and Valve Component Supports [ The safety injection system piping has no component supports. (The change in support numbers during the
- 2.
second period was due to new code class boundaries.) '
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9.2 Equipment Component Supports { The equipment component support and the associated ,. ,
- integrally-welded support attachment will be (VT-3) examined / J together. :
l 9.2.1 Containment Spray Exchanger (2)-CSH - 9.2.1.1 CSH Integrally-Welded Support Attachment u. There are four integrally-welded support attach- i ments on each CSH. All of these attachments - shall be (VT-3) examined during the inspection r interval in accordance with Table C of Appendix A. ' 9.2.1.2 CSH Component Support , There is one component support on each CSH. These ,
; supports shall be (VT-3) examined during the ; '
inspection. interval in accordance with Table C of ' i Appendix A. ..! eJ 9.2.2 Nonregenerative letdown Heat Exchanger (l)-NRLHX 7
,9 9.2.2.1 NRLHX Integrally-Welded Support Attachment
- l There is one integrally-welded support attachment "
on the NRLHX. This attachment shall be (VT-3)
,.j
. I examined during the inspection interval in '
accordance with Table C of Appendix A. '
. I 9.2.2.2 NRLHX Component Support '
a.. There is one component support on the NRLHX. This 'i . support shall be (VT-3) examined during the [ inspection interval in accordance with Table C of Appendix A. I
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' 9.2.3 Gas Stripper and Boric Acid Evaporator (1)-GSBAE
(,;. 9.2.3.1 GSBAE Evaporator Condensor-EC-k t,
. 9.2.3.1.1 EC Intearally-Welded Support ' Attachment There are no integrally-welded support :
attachments on the EC. ' +if.. 9.2.3.1.2 EC Component Support E, E. There are two component supports on the-g EC. . These supports shall be (VT-3). l '. examined during the inspection interval !. in accordance with Table C of : Appendix A. '.'
'9...
2'3 2 GSBAE Distillate Cooler (1)-DC ht 9.2.3.2.I'DC Inte0 rally-Welded Support Attachment [.' There are two integrally-welded support $ attachments on the OC. All of these P attachments shall'be (VT-3) examined I.- during the inspection interval in p accordance with Table C of Appendix A. , 9.2.3.2.2 DC Component Support
- There are'two component supports on the '
DC. These supports shall be (VT-3)
; -examined during the inspection interval- y
.in accordance with Table C of t/
Appendix A. f
' E 9.2.3.3 GSBAE Vent Cooler (1)-VC I' 9.2.3.3.1 VC Intecrallv-Welded Support Attachment There are no integrally-welded support attachments on the VC. ,
9.2.3.3.2 VC Component Support h - I .- There are two componen.t supports on the g. VC. These supports shall be (VT-3) i' examined during the inspection interval 2 in accordance with Table C of t Appendix A.
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5I-114.'l Page 56 of 85 Rev. 14
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9.1.3.4 GSBAE Support Frame (1)-SF k 9.2.3.4.1 SF Integrally-Welded Support Attachment c 1 There are no integrally-welded support attachments on the SF. .'< v. 9.2.3.4.2 SF Component Support O The SF consists of one component support. This support shall be.(VT-3) [ C examined during the inspection interval i in accordance with Table C of i Appendix A. 9.2.4 component coolina Surce Tank (2)-CCST I, 9.2.4.1 CCST Integrallv-Welded Support Attachment ['; The entire length of each CCST support skirt-to-- tank weld shall be (VT-3) examined during the .l inspection interval in accordance with Table C of F Appendix A. v. ( 9.2.4.2 CCST Component Support ( There is one component support on each CCST. .. These supports shall be (VT-3) examined during the inspection interval in accordance with- ' Table C of Appendix A. 9.2.5 " Component Coolina System Positive Displacement Charoino
' Pump Oii Cooler (1)-PDOC 9.2.5.1 PDOC Integrally-Welded Support Attachment ('
There are no integrally-welded support attachments on the PDOC. 9.2.5.2 PDOC Component Support There are two component supports on the PDOC. < These supports shall be (VT-3) examined during i, the inspection interval in accordance with Table C of Appendix A. 5 g<
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SQNP S1-114.1 Page 57 of 85 - Rev. 14 V 9.2.6 Component Cooling Water Gross Failed Fuel Detector (l)-GFFD y' 9.2.6.1 GFFD Integrally-Welded Support Attachment , There are no integrally-welded support . attachments on the GFFD. f. 9.2.6.2 GFFD Comoonent Support { l There is one component support on the GFFD. This - support shall be (VT-3) examined during the inspection interval in accordance with Table C of - Appendix A. f. . 9.2.7 Centrifugal Charging Pumo Gear Oil Cooler (2)-CCPGOC ((- 9.2.7.1 CCPGOC Integrally-Welded Support Attachment There are no integrally-welded support attachments on the CCPGOC. I 9.2.7.2 CCPGOC Comoonent Support fl 1 There are two component supports on each CCPGOC. 5 These supports shall be (VT-3) examined during f, the inspection interval in accordance with , Table C of Appendix A. 9.2.8 Centrifugal Charging Pump Oil Cooler (2)-CCPOC . 9.2.8.1 CCPOC Integrally-Welded Support Attachment
- There are no integrally-welded support '
attachments on the CCPOC. ; 9.2.8.2 CCPOC Comoonent Support - There are two component supports on each CCPOC. These supports shall be (VT-3) examined during - the inspection interval in accordance with . Table C of Appendix A. ' i: 9.2.9 Waste Gas Compressor Heat Exchanger (1)-WGCHX : 9.2.9.1 WGCHX Integrally-Welded Support Attachment .
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There are no integrally-welded support attachments on the WGCHX. '! i I I 0963o/COC4 [ l
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,There are two component supports on the WGCHX- . !'
These supports shall be (VT-3) examined during '; the inspection interval in accordance.with Table C of Appendix A. i'_ l
.9.2.10 Centrifugal Charcing Pump Mechanical Seal Cooler (4)-CCPMSC. i" '
9.2.10.I 'CCPMSC Integrally-Welded Support Attachment .h N There are no integrally-Welded support i ' attachments on the CCPMSC. ( 9.2.10.2 CCPMSC C6mponent Support i R. There is one component supports on each CCPMSC. 'g.. These supports shall be (VT-3) examined during- i the inspection interval in accordance with Table C of Appendix A. h L 9.2.11 RHR Pump Seal Cooler (2)-RHRSC .b t 9.2.11.1 RHRSC Intecrally-Welded Support Attachment h 1 l There are no integrally-welded support attachments'on the RHRSC. 9.2.11.2 RHRSC Component Support , There is,one component supports on each RHRSC. ,:
; These supports shall be (VT-3) examined during J,1.
the inspection interval in accordance with '.' Table C of Appendix A. .R4 3 9.2.12 Safety Iniection Pump Oil Cooler (2)-SIPOC " 9.2.12.1 SIPOC Integrally-Welded Support Attachment - There are no integrally-welded support attachments on the SIPOC. . I L. 9.2.12.2 SIPOC Component Support ^~ yf There are two component supports.on each SIP 0C. $l i: These supports shall be (VT-3) examined during the inspection interval in accordance with f,- Table C of Appendix A. 0963o/COC4 .. [ '
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a: SQNP . 3 SI-114.1 L Paga 59 of.85 f Rev. 14 6 b .. g 9.2.13 Safety Injection Pump-Seal Cooler (4)-SIPSC 8k 9.2.13.1 SIPSC Integrally-Welded Support Attachment ( f.[. i There are no integrally-welded support 3, . attachments on.the SIPSC. [' f,. .. 9.2.13.2 SIPSC Component Support i e g There is one component support onLeach SIPSC. [ These supports shall-be (VT-3) examined during .L; the inspection interval in accordance with . Table C of Appendix-A. ? 9.2.14 Component Cooling System Seal Water Heat Excjanger(l)-SWHX P 9.2.14.1 SWHX Integrally-Welded Support Attachment - . There is one integrally-welded support attachment. :" on the SWHX. This support shall be (VT-3) P examined during the inspection interval'in n accordance with Table C of Appendix'A. i' 9.2.14.2 SWHX Component Support There is one component-support on the SWHX. This
. support shall be (VT-3) examined during 'the . ;!
inspection interval in accordance with._ Table C of : Appendix A. ' 9.2.15 Component Coolino System Thermal Barrier Booster
- Pump ( 2 )-TBBP 1
N w s 9.2.15.1 TBBP Integrally-Welded Support Attachment U il There are no integrally-welded support l attachments on the TBBP. J 9.2.15.2 TBBP Component Support
- There are two component supports on each TBBP. .
f These supports shall be (VT-3) examined during : the inspection interval in accordance with Table C of Appendix A. {h
'c 9.2.16' Turbine Driven Auxiliary Feedwater Pump (1)-TDAFP 9.2.16.1 TDAFP Integrallv-Welded Support Attachment i There are no integrally-welded support attachments on the TDAFP. >
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SQNP : 6bof85 - Rev. 14- bj U. f 9.2.16.2 TDAFP Component Support 5.,.
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There are two component supports on the TDAFP. - These supports shall be (VT-3) examined during E. l the inspection interval in accordance with t Table C of Appendix A. 3 A- l 9.2'17 Motor Driven Auxiliary Feedwater Pump (2)-MDAFP r 9.2.17.1 MDAFP Integrally-Welded Support Attachment - There are no integrally-welded support ',' attachments on the MDAFP. :. 9.2.17.2 MDAFP Component Support- - s
.There are two' component supports on each MDAFP.
These supports shall be (VT-3) examined during y the inspection interval in accordance with-Table C.of Appendix A. ,l t. 9.2.18. Component Coolina System Water Pumps (3)-CCSWP : f 9.2.18.1 CCSWP Integrally-Welded Support Attachment
'f Ther-e are no integrally-weldsd support attachments on the CCSWP. /
9.2.18.2 CCSWP Component Support - i There is one component support on each CCSWP.
.' i
[ T These supports shall be (VT-3) examined during ;,
- the inspection Interval in accordance with "
Table C of Appendix A. h.- v 9.2.19 Component Cooling Heat Exchanger (2)-CCHX E 9.2.19.1 CCHX Integrally-Welded Support Attachment There are two integrally-welded support attach- [> ments on each CCHX, All of these attachments. . L shall be (VT-3) examined during the inspection
- Interval in accordance with Table C of Appendix A. j
,9.2.19.2 CCHX Component Support f a.
There are three component supports on each CCHX. [? , These supports shall be (VT-3) examined during i the inspection interval in accordance with Table C of Appendix A. 0963o/C0C4 L g Mfb4aulu* e & I c.&-+ ~~ - -+ _ e_ , - , - _y__.__ _ __$_ _.__ _y_L._3_3 mr3 r_q_. _L_:;7__y_qgy y;;
I SQNP { SI-ll4.1 Page 61 of 85 . Rev. 14 ;, 4 i. 9.2.20 Essential Raw Cooling Water System Strainer Support (2)-ERCWS , ,1, i l 9.2.20.1 ERCWS Integrally-Welded Support Attachment There is one integrally-welded support attachment f (consisting of 24 lugs) on each ERCHS. All of ' these attachments shall be (VT-3) examined during j the inspection interval in accordance with Table C of Appendix A. 9.2.20.2 ERCHS Component Support s There is one component support on each ERCWS. r These supports shall be (VT-3) examined during , the inspection interval in accordance with , Table C of Appendix A. 9.2.21 Essential Raw Cooling Water System Pump Station Pump ,' Supports (8>-ERCWP 9.2.21.1 ERCWP Intearallv-Welded Support Attachment There are no integrallywelded support attachments on the ERCWP. , 9.2.21.2 ERCWP Component Supports - There is one component support on each ERCWP. These supports shall be (VT-3) examined during the inspection interval in accordance with
- Table C of Appendix A.
, s.
9.2.22 Fuel Pool Coolino And Cleanino System - Spent Fuel Pit i Pumps (3)-SFPP 9.2.22.1 SFPP Intearally-Welded Support Attachment There are no integrallywelded support attachments
- on the SFPP.
9.2.22.2 SFPP Component Support
- There is one component support on each SFPP. ,
These supports shall be (VT-3) examined during 7 the inspection interval in accordance with 4 Table C of Appendix A. ,' 0963o/COC4
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SQNP SI-114.1 : Page 62 of 85 ! Rev. 14 } A 9.2.23 Fuel Pool Cooling And Cleanino System - Spent Fuel Pit . Hest Exchancer(2)-SFPHX :
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9.2.23.1 SFPHX Integrally-Welded Support Attachment U There are two integrally-welded support. attach- , ments on each SFPHX. All of these attachments - shall be (VT-3) examined during the inspection .' interval in accordance with Table C of Appendix A. ; 9.2.23.2 SFPHX Component Support There are two component supports on each SFPHX. ' These supports shall be (VT-3) examined during /, the inspection interval in accordance with Table C of Appendix A. .
. i 9.2.24 RHR Heat Exchanger Secondary Side Supports (2) - RHRHSXH '-
9.2.24.1 RHRHSXH Integrally-Welded Support Attachment ' There are two integrally-welded attachments on- I each RHRHSXH. All of these attachments shall be b (VT-3) examined during the inspection interval in }; accordance with Table C of Appendix A. y y 9.2.24.2 RHRHSXH Component Support e There are two component supports on each RHRHSXH. These supports shall be (VT-3) examined during the inspection interval in accordance with Table C of. Appendix A. L' 9.2.25 ERCW Screen Wash Pump Support (4) - ERCWSWPH 9.2.25.1 ERCHSWPH Integrally-Welded Support Attachment There are no int'egrally-welded support attach-ments on the ERCWSWPH. 9.2.25.2 ERCHSWPH Component Support There is one component support on each ERCWSWPH. i These supports shall be (VT-3) examined during the inspection interval in accordance with Table
}} 1 C of Appendix A. 2 9.3 System Pressure Tests
- 5 All ASME Code Class 3 (Equivalent) system pressure tests shall be performed in accordance with Program Procedure 1502.03.
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l SONP. -) SI-114.1 . Page 63 of 85 ' Rev. 14_ r i i; o L 9.3.1 System Inservice Tests
' . < , i Pressure retaining components-within the boundary of f systems or portions of systems required to operate in l l
' support of normal plant safety functions of shutting down t and maintaining the reactor in the cold shutdown
- condition, and pressure retaining piping, pumps, and ;{
valves within the boundary or systems or portions of Nj l systems required to operate in support of residual heat 1 removal from spent fuel storage pool, shall be visually ( examined in accordance with visual examination method VT-2 (see section 11.1) while the applicable systems are in i' i service under operating pressure in accordance with IND-5221 of ASME Section XI. This is performed in , accordance with the SI-267 series. ,. These examinations and tests shall be performed during each inspection period. ' e 9.3.2 System Functional Tests *
~
Pressure retaining components within the boundary of systems or portions of systems required to operate in , support of the postaccident safety functions of emergency core cooling, containment heat removal and atmospheric , cleanup, and long-term residual heat removal from the reactor vessel shall be visually examined in accordance r with visual examination method VT-2 (see Section 11.1) during a system functional test conducted to verify operability in systems in accordance with IHD-5222 of ASME Section XI and in accordance with the SI-267 series. I. System Hydrostatic Tests (.'. 9.3.3 >. Pressure retaining components identified in Sections 9.3.1 I and 9.3.2 shall be visually examined in accordance with J visual examination method VT-2 (see Section 11.1) during a system hydrostatic test conducted in accordance with IND-5223 of ASME Section XI and the SI-265 series. These ' examinations and tests shall be performed over each
- inspection interval. .
9.4 Component Supports and Restraints , Component supports and restraints within the boundaries of the . systems' identified in Sections 9.3.1 and 9.3.2 for components # exceeding 4-inch nominal pipe size shall be visually examined each - inspection period in accordance with visual examination method
- VT-3 (see Section 11.1). l l
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! SONP . I SI-il4.1 , , Page 64 of 85 i: 1 Rev.-14 L g, j ' 9.5 Snubbers and Hangers ?
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Mechanical and hydraulic snubbers, spring loaded, and constant- % weight support hangers for components exceeding 4-inch nominal p' j
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pipe size shall be visually examined each inspection period in > accordance with visual examination method VT-4 (see Section 11.1.) !! > u ! 10.0 AUTHORIZED INSPECTOR TVA shall employ an Authorized Inspector (s) in accordance with ASME h. Section XI for inservice examinations, repairs, and replacements of ASME k Code Class 1, 2 and 3 components at Sequoyah Nuclear Plant. The f" i inspector shall verify, assure, or witness that code requirements have L" ' been met. He shall have the prerogative and authorization to require M requalification of any operator or procedure when he has reason to )- believe the. requirements are not being met. TVA shall provide access a for the AI in accordance with'IHA-2140 of ASME Section XI. - y> TVA's interface with the Authorized Inspector for inservice inspection, E repairs, and replacements is defined in NQAM Part II, Section 2.3 and .' Part II, Section 5.1, Program Procedure 1402.02, MI 6-21 and in AI-19 - Parts IV and VI. L i' 11.0 EXAMINATION METHODS V
+; ,
11.1 Visual Examination < (l Visual examinations that require clean surfaces or decontamination - for. valid interpretation of results shall be preceded by . appropriate cleaning processes. O 11.1.1 -Vi sual Examination VT ' i : The VT-1 visual examination shall be conducted to determine '," the condition of the part, component, or surface examined, 7 including such conditions as cracks, wear, corrosion. ?' erosion, or physical damage on the surfaces of the part or ' components. 11.1.1.1 Direct Visual Examination VT-1 " Direct VT-1 visual examination may be conducted I. when access is sufficient to place the eye 24 ; inches (610 mm) of the surface to be examined and S at an angle not iess than 30 degrees to the' h surface. Mirrors may be used to improve the -( angle of vision. Lighting, natural or artificial 5.- ' shall be sufficient to resolve a 1/32 inch (0.8 mm) black line on an 187. neutral gray card. - I 096'3o/C0C4 , j
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SQNP S1-114.1 Page 65 of 85 Rev. 14 11.1.1.2 Remote Visual Examination VT-1 Remote VT-1 visual examination may be substituted for direct examination. Remote examination may use aids such as telescopes, borescopes, fiber optics, cameras, or other suitable instruments provided such systems have a resolution capability at least equivalent to that attainable by direct visual examination. 11.1.2 Visual Examination VT-2 The VT-2 visual examination shall be conducted to locate evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage collection systems as required during the conduct of system pressure or functional test. The visual examination, VT-2; may be conducted without the removal of insulation by examining the accessible and exposed surfaces and joints of the insulation. Essentially vertical surfaces of insulation need only be examined at the lowest elevation where leakage may be detectable. Essentially horizontal surfaces of insulation shall be examined at each insulation joint. For components whose external insulation surfaces are inaccessible for direct examination, only the examination-of surrounding area, including floor areas.or equipment surfaces located underneath the components, for evidence of leakage, or other areas to which such leakage may be channeled, shall be required. l Discoloration or residue on surfaces examined shall be given particular attention to detect evidence of boric acid accumulations from borated reactor coolant leakage. The visual examination shall be conducted during the system leakage tests conducted after refueling outages and L I prior to startup. l " L VT-2 visual examination shall be conducted in accordance i with ASME Section XI. IWA-5240 and the applicable NDE l procedure. 11.1.3 Visual Examination VT-3 l The VT-3 visual examination shall be conducted to determine l the general mechanical and structural conditions of components and their supports such as the presence of 09'63o/C0C4 AT?d_M*"M1E T .'.t NE T I" .W" I . ."#f N - P"_ 7SW' 'D.# "*#N 0 9. "*N " 4O"
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SQNP
' S1-114.1 t-
.Page 66 of 85 Rev. 14 loose parts, debris, or abnormal corrosion products, wear,
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erosion, corrosion, and the loss of integrity at bolted.or . welded connections. VT-3 may require, as applicable to determine structural integrity, the measurement of clearances, detection of physical displacement, structural adequacy of supporting [- elements, connections between load carrying structural members, and tightness of bolting. For component supports and component interiors, the visual examination may be performed remotely with or without optical aids to verify the structural integrity of the component.. 11.1.4 Visual Examination VT-4 The VT-4 visual examination shall be conducted to determine conditions relating to the operability of components or devices such as mechanical and hydraulic snubbers, component supports, pumps, valves, and spring loaded and constant weight hangers. VT-4 shall confirm functional adequacy, verification of the settings, or freedom of motion. This examination may 4 require (1) disassembly of components or devices and (2) operability test. 11.2 Surface Examination 11.2.1 Macnetic Particle Examination
! Magnetic particle examination (MT) shall be conducted in accordance with Article 7, Section V of the ASME Code.
11.2.2 Liould Penetrant Examination Liquid penetrant examination (PT) shall be conducted in accordance with Article 6, Section V of the ASME Code, 11.3 Volumetric Examination . 1 11.3.1 Radiographic Examination Radiographic techniques, employing penetrating radiation
'such 'as X-rays, gamma rays, or thermalized neutrons, may be utilized with appropriate image recording devices such as photographic film or papers, electrostatic systems, direct-image orthicons, or image converters. For radio-graphic examinations employing either X-ray equipment or 09'63o/COC4 L____ _ _ _ _ _ _ _ _ . _ _ _ _ . . . _ _ _ _-_______._____...______.________.__.___-.___-.-..__. .--m..-_.--..-r.--..:.-*..--.---n-,mee_* ewe,*-- ~=-==--Pw**
SQNP SI-114.1 Page 67 of 85 Rev. 14 radioactive' isotopes and photographic films, the procedure shsil be as specified in Article 2, Section V, of the ASME . Code. 11.3.2 Ultrasonic Examination (a) Ultrasonic examination of Class 1 and Class 2 vessel welds in ferritic material greater than 2 inch (51 mm) in thickness shall be conducted in accordance with l Article 4 of Section V of the ASME Code and TI-51. (b) Ultrasonic examination of Class 1 and Class 2 ferritic steel piping systems shall be conducted in accordance with Appendix III, amended as follows: (1) For examination of welds, reflectors that produce a response greater than 50% of the reference l level shall be recorded. (c) If the requirements of (a) and (b) above are not applicable, the ultrasonic examination shall be conducted in accordance with the applicable requirements of Article 5 of Section V amended as follows: .! (1) For examination on welds, reflectors that produce a response' greater than 50% of the reference level shall be recorded. (2) For examination of welds, all reflectors which produce a response greater than 100% of the reference level shall be investigated to the extent that the operator can determine the shape, identity, and location of all such reflectors in l terms of the acceptance-rejection standards of IWA-3100(b). (3) Thesizeofrefiectorsshallbemeasuredbetween points which give amplitudes equal to 100% of the 4 reference level. I 11.3.3 Eddy Current Examination Eddy current examination of heat exchanger tubing shall be conducted in accordance with the provisions of Appendix IV
'of Section XI of the ASME Boiler and Pressure Vessel Code, l
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SQNP S1-114.1
, Page 68 of 85 Rev. 14 12.0 QUALIFICATIONS OF NONDESTRUCTIVE EXAMINATION PERSONNEL Personnel performing nondestructive examination operations shall be qualified in accordance with IWA-2300 of ASME Section XI as required in QMP 102.4.
13.0 ACCEPTANCE CRITERIA All acceptance standards for ASME Code Classes 1, 2 and 3 (Equivalent) components shall be in accordance with IWA-3000, INB-3000, IWC-3000, or IWD-3000 of ASME Section XI, except where ASME Section III examinations are employed to satisfy ASME Section XI requirements. i 14.0 REPAIRS All ASME Section XI components and their supports (ASME Classes 1, 2, and 3) shall be repaired in accordance with the Repair and Replacement Program included in NQAM Part II, Section 2.3 and implemented by Program Procedure 1402.02, MI 6.21 and AI-19 (Parts IV and VI). 15.0 REPLACEMENTS Replacement of ASME Section XI components (ASME Classts 1, 2, and 3) shall be in accordance with the Repair and Replacement Prcgrt? fn NQAM Part II, Section 2.3 and implemented by Program Procedure 1402.02 (SQNP) HI-6.21 and AI-19 (Parts IV and VI). ASME Section XI repairs and replacements may be coordinated as necessary with the ISI Programs Section. 16.0 RECORDS AND REPORTS 16.1 Recordi o of and Report of Examinations A detailed report of all examinations shall be prepared by the performing or responsible organization and should contain but not be limited to the following information: Title Page Table of Contents . I. Introduction - The introduction should include the following information: Plant, unit number, preservice or inservice inspection and cycle number, systems, components
~and vessels examinations were performed on, organization examinations were performed by, dates examinations were performed, ASME Section XI Code of Record.
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SQNP SI-114.1 Page 69 of 85 Rev. 14 II. Summary - The summary should include a brief description of-the overall inspection: Program, performance, personnel, equipment, procedures, evaluations, and results. III. Discussion - The discussion should discuss the governing documents (ASME Code, Technical Specifications, etc.), inspection schedule, materials, calibration standards, calibration performance, reporting, recording, interpretation, and brief evaluation. IV. Evaluation - Evaluation is based on the indication's location, metal path, general shape, and any tests that could be applied, such as damping. The evaluation section also should contain a listing of each examination performed and the evaluated results. V. Summary of Notifications - The summary of notifications shall give a short summary of each notification report along with the indication discrepancy and its location. It should also contain the final disposition and the date of completion. VI. Scan Plan - The Scan Plan shall give a detailed description of all areas subject to examination during the inspection. It shall contain the following information: Examination Area, Code Category, Held Size and/or Number, Reference Drawing, Examination Method, Procedure, Calibration Block, and any reference details pertaining to the exam area, such as the weld number, meridional welds, pump studs, etc. VII. -Weld and Hanger Maps - The Held and Hanger Maps are the
' reference drawings for the inspection. The weld maps are isometric drawings showing the location of both field and shop welds on each vessel, component, and piping system subject to examination. The hanger maps are also isometrics showing the location of hangers, snubbers and supports for each vessel, component, or piping system subject to examination.
VIII. Log by System - The log is the daily status of the inspection section representative of the areas subject to examination during the inspection. This log keeps an up-to-date status of work complete and incomplete. IX. ' Personnel Certifications X. Equipment Certifications XI. NDE Procedures 0963o/COC4
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, Page 70 of 85 Rev. 14 XII. Calibration Block Drawings -
XIII. Calibration Sheets XIV. Examination Data Sheets For eddy current examination of heat exchanger tubing, the report shall include a record indicating the tube (s) examined (this may be marked on a tube sheet sketch or drawing), the extent to which each tube was examined, the location and depth of each reported indication, and the identification or the operator (s) and data evaluator (s) who conducted each examination or part thereof, and magnetic tape and strip charts. All procedures and equipment shall be identified sufficiently to permit duplication of the examination at a later date. This shall include initial calibration data for the equipment and any significant changes. All required and pertinent information will be recorded on the appropriate data sheets by the performing organization. When portions of the inspection work are contracted, a detailed report will be submitted to TVA by the contractor with all pertinent and required information. TVA will retain the original copies of all raw data taken. The NQA shall review and submit the final report to the Plant Superintendent for review. These final reports shall be flied at
.the plant site with the data sheets of Appendix 0 of this program as discussed in Section 5.0 of this program. Data Sheet 1 in Appendix D will be completed and used as a cover sheet for the
, final report and to document the review process.
16.1.1 epair and Replacement Reports Repairs and Replacements summary reports shall be prepared in accordance with Program Procedure 1402.02, MI 6.21 and . AI-19 Parts IV and VI. 16.2 ISI Report for Class I and 2 Components The NQA shall prepare an ISI Summary Report for Class 1 and 2 components (with the exception of Steam Generator Tubing examinations - see paragraph 20.5) to be submitted within 90 days after the completion of the inservice inspection to the Nuclear Regulatory Commission Region II Office in accordance with IWA-6220, ASME Section XI.
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.The ISI' Report shall have a cover sheet providing the following l Informatiorr: .:
(1) Date of document completion. ! (2) Name and address of owner. I (3) Name and address generating plant. (4)- Name or number assigned to the nuclear power unit by TVA. ;$ (5) Commercial operation date for unit. b p
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All reports shall provide the following information as a minimum. ' (a.) Numbers assigned to the components by the. state. 4 (b.) National Board Number assigned to the components by the manufacturer. . (c.) Names of the components and descriptions including size, capacity, material, location, and drawings.to aid ' identification. 'J l (d.) Name and address of principal manufacturer.(e.g., l Westinghouse, GE, etc.) and the principal contract number - which will. identify the subcontractors. Manufacturer's component identification numbers. , (e.) Date of completion of the examination. (f.) Name of ANII who witnessed or otherwise verified the i, examinations and his employer and business address, when , required. j (g.) Abstract of examinations, conditions observed; and corrective measures recommended or taken. , (h.) Signature of ANII, when required. . (1.) Owner's Data Report for inservice inspection, Form NIS-1, as , shown in Appendix II of ASME Section XI and Form NIS-2 as ! shown in Program Procedure 1402.02. [ 1 The NQA shall submit the inservice examination report to the plant i manager for retention as a quality assurance record in accordance : , with NQAM, Part III, Section 4.1. The NQA shall maintain a copy . l of the report for information, and shall submit applicable
- summaries of the report with a cover sheet as described above to the NRC via the Director, ONP Nuclear Safety and Licensing.
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We , 51-114.1- ' Page 71 of 85 +
, Rev. 14- '
t i. i - u The ISI Report shall have a cover sheet providing the 'y foll s. information: G
. h (1) Date of document completion. I (2) Name and address of owner. [
_i n Name ano address generating plant. (3) TW\ . (4) Name or number assigned to the nuclear power unit by ' (5) Commercial operation date for unit. i imum, q All reports shall provide the following information as a m n : tt (a.) Numbers assigned to the components by the.s a e. nts by the (b.) National Board Number assigned to the compone manufacturer. including size, s (c.) Names of the components and descriptions identification. (e.g., i ' (d.) Name and address of principal Manufacturer's manufacture which will identify the subcontractors.
' component identification numbers.
(e.) Date of completion of the examination.
' ified the . h # (f.) Name of ANII who witnessed or otherwise y required. d; and corrective > ;. (g.) Abstract of examinations, conditions observe -
measures recommended or taken. (h.) Signature of ANII, when required.
' tion, Form NIS-1, as (i.) Owner's Data Report 1402.02.
for inservice inspecsho t- shown in Program Procedure to the plant x f't w The NQA shall submit the inservice examination reportre P[fh manager'for retention as a quality assuranceThe NQA ble shall maintain M with NQAM, Part III, Section 4.1. 2*g' of the report for information, and shall submit d above toapplica summaries of the report with a cover s 09630/COC4 f
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SQNP SI-ll4.1 : P Page 72 of 85 Rev. 14~ [ U 16.3 Records for ASME Code' Class 1, 2 and 3 (Eauivalent) i .. The plant (master files) shall assure the availability and storage ( 17 of the following records:
-[
(1) Examination Plans and Schedules b
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(2) Reports of Examination Results (3) Examination Methods and-Procedures. 7' (4) Reports of Evaluations of Results' l' (5) Reports of Corrective Actions, Repairs, and Replacements I 16.4 Records of System Pressure Tests > Records of the visual examinations conducted during system leakage , or hydrostatic tests shall consist of an itemization of the~ number and location of leaks found in a system and the corrective actions i taken. 16.5 Augmented Examination Reports Augmented examination special reports shall be submitted to the ,' Nuclear Regulatory Commission Region II Office within the time - period specified for each report. For specific details on - records, reports'and reporting see Section 20.0, Augmented n Inspections, of this program. The status of every augmented examination requested by the ISI j> Programs Section and implemented by the NQA shall be maintained in t , a computerized data base.
}'
16.6 Status of' Completed Work
, h' NDE Section XI Programs maintains a computer based status listing .
of the completed work taken as credit for examinations performed. ;. over the period and interval. This listing at a minimum e identifies the component examined, ASME class, exam method, and :, outage period when the work was performed. This listing may also 4 be used to identify components to be inspected in the future. ;,y 17.0 NOTIFICATION OF INDICATION 2. I f;; ) Plant management shall be formally notified of the presence of ,P i j unacceptable indications found during in-service inspections and '.* { , detected during the performance of nondestructive examinations y{ , (excluding VT-2, visual examinations performed during system pressure ' tests). Unacceptable indications are defined by the applicable NDE :l t 0963o/COC4 i ,. s 4 ; I
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4 SONP 51-114.1 [ Page 73 of 85 Rev. 14 []
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. 3 procedure. Formal notification shall consist of completing and -
l submitting to the Plant Manager or his designee the " Notification of . P1 Indication" (NOI) form in Appendix E of this program. The NOI form shall .. only be used to report unacceptable indications of components within the [ scope of Section XI and which have been scheduled for examination. ' St.iduled examinations refer only to those components that are .} in-service inspection samples and exclude those examinations performed ..1 following repair or replacement of components. Any other discrepancies-should be reported in accordance with the appropriate plant procedure
;f j (e.g., PRD, WR, etc.). -
NOIs initiated prior to the requirement for Condition Adverse to Quality Report (CAQR) and the implementation of Revision 8 of this instruction
)'.
do not require " rollover" to the CAQR form. ( I Part I of the NOI form shall be completed and signed by the NDE Level II : or III examiner detecting the indication. The examiner detecting the ' I indication can initiate a CAQR at this time or defer the decision to the
- Individual responsible for the disposition in Part II of the NOI form.
The NOI form and the CAQR are to be processed together. If the indication is detected by an outside contractor, the contractor's field supervisor or his designee shall review and sign the NOI form. The NQA representative shall assign a sequential number, review, and sign the L NOI form. A NOI log shall be maintained on a plant / unit I cycle basis i by the NQA for each NOI issued. This log shall. contain as a minimum: NOI No., component, I.D., date issued examination report No., unsatisfactory condition, and MR/HR No, and work instruction and/or DCR v numbers as applicable. The original shall be sent to the plant manager
'or his designee and a copy to the NDE Section XI Programs. ;
In the case of NOIs for which no Condition Adverse to Quality Report b (CAQR) has been identified, the Nuclear Site Directors Organization . (plant managei or his designee) shall be responsible for determining : which organization (Construction Modifications, Plant Maintenance, etc.) shall be responsible for preparing a disposition in Part II of the NOI I form and performing the associated corrective action. If the organiza-tion assigned responsibility for disposition is unable to determine a ;. satisfactory disposition, then the NOI form shall be sent to NE for disposition. .- The organization assigned responsibility for the disposition shall ~
/
evaluate the NOI for the need of a CAQR in accordance with the NQAM . Part I, Section 2.16 and plant instruction AI-12 (Part III). ' The j: CAQR shall be processed in accordance with AI-12 (Part III).' If a i CAQR is needed, enter the identifying number of the CAQR in the space shown on the HOI form. The NOI form and the CAQR are to be processed y!
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3 together. If a CAQR is not required but the organization responsibit f
' AI-12 (Part I) to be used until AI-12 (Part III) is approved.
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SQNP , SI-114.1 L
, Page 74 of 85 l1 Rev. 14 c
L( for the disposition determines that a CAQ exists, the CAQ shall be proressed in accbrdance with NQAM Part I, Section 2.16 (AI-12 [Part [i III]).' i The generated CAQ shall-be identified, documented, evaluated, [ corrected, tracked, trended and reported to the appropriate level of j management by the organization responsible for the disposition using one y of the administrative programs specified in AI-12 (Part III).' C C The individual responsible for preparation of.the disposition shall sign ( ' and date Part II of the NOI form and ensure that the disposition agrees : with the corrective action proposed in the CAQR if one was initiated. The cognizant supervisor or his designee of the appropriate organization e, shall review and approve the disposition and sign and date Part II of the NCI form. Copies of the form shall be distributed to the plant manager or his designee and the NDE Section XI Programs. The original [. shall be returned to the NQA representative. One copy shall be filed .. with the examination report. A copy of the dispositloned NOI should be attached to the work generating document used in correcting the unacceptable indication, i Dispositions to correct the condition under the plant maintenance ' program shall be processed in accordance with NQAM, Part II, Section 2.1
,7 (see SQM-2). Dispositions other than restoring to original requirements i' shall be processed as modifications in accordance with NQAM, Part II, Section 3.2 (see AI-19, Parts IV and VI). Repair and replacement -
activities, including coordination with the Authorized Inspection Agency ', (AIA), shall be performed in accordance with the requirements of NQAM, ' Part II, Section 2.3 (MI 6.21 and AI-19 Parts IV and VI). The . organization responsible for corrective action shall include preservice ; examination requirements in the repair or replacement work instruction k described in NQAM, Part II, Section 2.3 (see MI-6.21 and AI-19 Parts IV and VI). In some instances the NOI may be dispositioned based on ' additional information available to the individual responsible for the 1 disposition (e g., design drawings, drawing notes, specifications, y etc.) In this case the NOI should be categorized as "other," and a ( documented justification will be included with the disposition. A USQD is not required to disposition NOIs categorized as "other." The i disposition shall include re-examination if work was performed (e.g., tightening of loose bolts, etc.). Re-examination may be limited to area of component where work was performed. ,
,{
r. Upon completion of corrective action the NQA representative shall verify ,' ( completion of corrective action, enter the work instruction and/or DCR , numbers on the NOI form, enter the examination report number if re-exam- t' l ination was performed, and sign and date the NOI form, Part III. The C signed NOI form shall remain with the original examination report for k use as a quality assurance record. If re-examination was performed, a 9 copy of the signed NOI form shall also remain with the re-examination report. Copies of the NOI form shall also be distributed to the plant { manager and NDE Section XI Programs. '
' AI-12 (Part I) to be used until AI-12 (Part III) is approved. ;
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SQNP ! S1-114.1 c' i Page 75 of 85 Rev. 14 r 17.1 Additional Sample Selection for CC-1 and CC-2 NOIs After a NOI has been dispositioned and returned to the NQA repre- ' sentative, a copy shall be forwarded to NDE Section XI Programs ;; for evaluation to determine if additional examinations shall be '- required. If it is determined that additional examinations are r-required, these examinations shall be performed during the same - outage as the initial examinations. A sample is defined as those items (welds, areas, or parts) as described or intended in a - particular examination category and item number and within the same system. For component support samples, the additional > examinations may also be limited to component supports within the F; same examination method (VT-3, and VT-4). The initial sample is N l the sample scheduled for examination at a particular outage for . Section XI credit. For steam generator tubing, additional samples f shall be per Section 7.3.8. 17.1.1 Evaluation For First Additional Sample A first additional sample shall be selected for those examinations performed in the initial sample that reveal indications exceeding the corresponding acceptance ' standards of IWX-3000 of ASME Section XI. The first additional sample shall include approximately the same .. number of items examined in the initial sample. The first additional sample shall be selected from the ten year interval saeple. The items selected should be from those , i items that have the longest service time from its previous , I inservice examination. If a system contains a sample within the same examination .!
-category and item number as the initial sample which had '
Lthe indication and is not scheduled for examination this
- outage, this system sample shall be evaluated to determine if it should be examined this cutage to provide a representative sample from each system within a particular exami. nation category and. item number. This evaluation should consider, but is not limited to, the type of ,
indication found in the initial sample and similar system design and operating parameters. The NDE Section XI Programs Engineer shall submit the ? additional sample to the NQA Data Management 5 representative for addition to the scan plan. . 17.1.2 ' Evaluation for Second Additional Sample h j ( A second additional sample shall be selected for those . examinations performed in the first additional sample that reveal indications exceeding the corresponding acceptance ] i l 1 ' 0963o/C0C4 l l I l . 8
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SQNP SI-114.1 'I Page 76 of 85 Rev. 14 .. E b. standards of IWt 3000 of ASME Section XI. The second k additional sample shall include all the remaining' items of I, the ten year interval sample not examined in the 1:itial Ei or first additional sample during this outage. If no I.: l Items remain in the ten year interval sample (e.g., (l Accelerated Field Weld Program) and indications were found t in the first additional sample, a notification of sample p' - results shall be made to NE as described in Section L 17.1.3. The NDE Section XI Programs Engineer shall submit the second additional sample to the NQA Data Management ;t representative for addition to the scan plan. L 17.1.3 Notification of Sample Results to Nuclear Engineering f, After completion of the second additional sample examina . tions, ASME Section XI code requirements for additional I
- examinations are complete. If examinations performed in the second additional sample reveal indications exceeding ,
the corresponding acceptance standards of IWX-3000 of ASME P Section XI, Nuclear Engineering shall be notified. Nuclear Engineering shall be notified to evaluate the - indications and make recommendation (s) for further action, if needed, within this and/or other systems. This notification shall be sent to the Engineering ' Specialist, ISI, in the onsite Materials Technology Group ; by the NDE Section XI Programs Department for coordination ~ with the applicable NE staff. Included in the notification should be a summary of the indications found, number of :. examinations and number of indications in each sample, " type of examination performed, examination category and bj
-item number, copies of the NOIs, and any other pertinent F Information. $
18.0 CALIBRATION BLOCKS Calibration blocks will be used for ultrasonic examinations (a calibration tube will be used for eddy current examination of steam generator tubing). The blocks will be fabricated to the general ,' requirements of ASME Section V and ASME Section XI. The blocks shall be 1 fabricated of the material to be examined or equivalent P numbers. Mill , r test reports shall be obtained and retained by the NQA for all f calibration blocks. The blocks shall employ drilled holes and/or i-notches for calibration reflectors (See Request for Relief 151-2). i
. The NQA shall' maintain as-built calibration block drawings. Copies of 5 the original drawings and any revisions shall be submitted to NDE J Section XI Programs. The calibration blocks shall be stored at the plant site and maintained by the plant NQA personnel, t
0963o/C0C4 I i
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, Page'77 of 85 p Rev. 14
{.. 19.0 REOUESTS FOR RELIEF ~" i. L
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Where TVA' has determined that certain code requirements or examinations . h are impractical, TVA will submit written requests for relief to NRC with p.
'information to support the determinations and any proposed alternate ,
examinations. The impractical code requirements or examinations shall r be identified in this program, and references to particular requests for ) relief shall be. included. !? When impractical examination requirements are identified in the field, .- NQA shall notify NDE Section XI Programs such that the information may be included in this program and requests for relief may be prepared if' p.; necessary. p The Requests for Relief are listed in Appendix F. ,
- r. '
20.0 AUGMENTED INSPECTIONS b h Augmented inspections 'are performed in addition to ASME Section XI. code T requirements. The augmented inspections may be required by the NRC or self-imposed by TVA. '- Augmented examinations of components that require reporting (verbal or ( written) to the NRC shall be the responsibility of the organization -- designated in the particular. examination. These reports shall be F submitted to the Nuclear Regulatory Commission Region II Office within - the time period specified for each report. Each augmented exam shall ' state the required reporting time and the document requiring the " information to be included in the report. i 20.1 Feedwater Nozzle-to-Pipe Welds and Adjacent Pipe and Nozzle Areas - The regi.irements of NRC IE Bulletin 79-13 were satisfied during [ the unit 1, cycle 1 outage. Due to the safety-related ramifi- i cations of the steam generator nozzle transition section cracking i problem, TVA will perform an augmented inspection (reference memorandum L29 831222 836, L29 840105 856, and memorandum from David Goetcheus to John T. Lewis) of four feedwater nozzle during
- 1
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each refueling outage. The augmented examination of the steam generator nozzle transition section shall include nozzle-to- '. 1 transition piece welds and transition piece-to-pipe welds. These
- Mi welds shall be ultrasonically examined and ultrasonic sensiti-vities should be equivalent to those required by NRC-IE Bulletin T[
83-02. The examination data sheets shall be included in the ;. 90-day ISI report discussed in section 16.2 of this program. B: f. 20.2 RPV Nozzle Safe Ends [t i The augmented examination requirements of the RPV nozzle-to-safe 4 end welds are included in the final report - Sequoyah Nuclear ! J
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0963o/COC4 .I m
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< SI-114.1 p Page 78 of 85 e Rev. 14 (( Plant'- Evaluation of cracking in reactor vessel nozzle stainless W. steel buttering. The examinations for unit I will be monitored at -) the normal inservice inspection intervals for dissimilar metal. P welds as required by Section XI of the ASME Code. This augmented S. . examination does not require a special report. The examination y data sheets shall be included in the 90 day ISI report discussed h in section 16.2 of this program. '
._^
20.3 _ Reactor Coolant Pumo Flywheel a The augmented examination requirements of the reactor coolant pump $ flywheel arc included in Regulatory Position C.4.b of Regulatory i;, Guide 1.14; (1) an inplace ultrasonic examination of the areas of n higher stress concentration at the bore and keyway at approximately [ three-year intervals during the refueling or maintenance shutdown coinciding with the inservice-inspection. schedule as required by (. Section XI of the ASME Code, and (2) a surface examination of all ' exposed surfaces and complete ' ultrasonic examination at approxi- G mately 10-year intervals during the plant shutdown coinciding with j the inservice inspection schedule as required by Section XI of the ; i ASME Code. ' This augmented examination does not require a special report ' unless .the examination reveals a flaw. If the examination and evaluation indicate an increase in flaw size or growth rate ;, greater than predicted for the service life of the flywheel, the '/ results of the examination and evaluation should be submitted to ,' the NRC for evaluation. Refer to Regulatory Guide 1.14 for information to'be included. The examination data sheets shall be included in the 90 day ISI report discussed in section 16.2 of I, this program. ' After Miy 1, 1989, the in place examinations shall be recorded f. using the following ID's" i l.i RCP-X-FLYWH-BORE h RCP-X-FLYWH-KEY ,
-j 1
For the 10 year exam the ID's will be: .I . RCP-X-FLYWH-SUR
. U RCP-X-FLYWH-VOL L1 20.4 Pressurizer Relief Line 0 n'
The augmented examination requirements of the pressurizer relief line are included in the Technical Specifications 4.0.5 and $I Sl 4.4.3.2.4. The pressurizer relief line and repair welds shall be examined in accordance with TVA procedure N-UT-25 using improved ultrasonic detection and evaluation procedures prior to entering V 4
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l SQNP SI-114.1 l Page 79 of 85 Il Rev. 14 - < I; Mode 4 whenever the plant has been in cold shutdown for 72 hours LL or more if the examination has not been performed in the previous [ ,, l six months. This procedure has been qualified in accordance with . NRC IE Bulletin 83-02 by demonstrating proficiency in detecting i Intergranular Stress-Crack Corrosion (IGSCC). All future examination of the pressurizer relief line and repair welds will .
'. sj be performed using this procedure or a procedure demonstrated to ,
i equivalent or superior for detecting IGSCC. .. In the event these six-month period examinations find the piping fI free of unacceptable indications for three successive inspections, *-{ the inspection interval shall be extended to 36 month intervals - (12 months to coincide with a scheduled refueling outage). The report shall be submitted with the Final Inservice Inspection Report. [.; 20.5 Steam Generator Tubing The augmented examination requirements of the steam generator .'. tubing are included in Technical Specifications 4.0.5, 4.4.5.0 and section 7.3.8 of this program. Following each inservice inspec-tion of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the NRC within 15 days. . Plant management shall report this information to the Nuclear
- Regulatory Commission Region II Office within the time period specified. See Plant Instruction AI-18 for reporting instructions. '
The complete results of the steam generator tube inservice inspection shall be submitted to the Nuclear Regulatory Commission in a special report pursuant to Technical Specification 6.9.2 within 42 months following completion of the inspection. The NDE 7 Engineering Section shall prepare this special report and submit P the report of the Nuclear Regulatory Commission Region II Office [, within the stated time period. This special report shall include: -
- 1. Number and extent of tubes. inspected.
- 2. Location and percent of wall-thickness penetration for each indication of an imperfection. i
- 3. Identification of tubes plugged. .i l
Results of steam generator tube inspections which fall into Technical Specification Category C-3 require prompt notification [ i, of the Huclear Regulatory Commission pursuant to Technical S Specification 6.9.1 prior to resumption of plant operation. The written followup of this report shall provide a description of investigations conducted to determine cause of the tube degradation and corrective measures to prevent recurrence. l i 0963o/C0C4 [ h __r = _ 2: . _ ::~ r-
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1 i, SQNP-bof85 Rev.' 14. 'k e ('. Refer-to Technical Specifications 4.0.5 and 4.4.5.0 and section r 7.3.8 and 16.0 of this program for information to be included. . 20.6 RPV Nozzle Cladding E All vessel nozzles cladding shall be ultrasonically examined 'at 2 p, the end of-each 10-year inspection interval, using techniques at
- least as sensitive as those used to conduct the supplemental 't '
examinations pe-formed prior to fuel loading. The results of this 1, examination shall be reported to NRC. This examination is done in accordance.with Sequoyah technical specifications and satisfies - Surveillance Requirement.4.4.10. This augmented examination does. not require a special report. This examination data sheets shall .h'
. ,1, be included in the.90 day ISI report discussed in section 16.2 of this program. :
Nozzle forging material and cladding is identified in section u
. 7.1.2 of this program. "
20.7 RPV Closure Head Circumferential Weld (WO9-10) 1 e See subsection 7.1.1.3 of this program'for augmented examination ! requirements. This augmented examination does not require a special report. The examination data sheets shall be included in f.. the 90 day ISI report discussed in section 16.2 of this program. 20.8 Steam Generator Feed Rino and J-Tubes The carbon steel J-tube nozzles have been replaced with Inconel 600 : J-tube damage. nozzles. This material is resistant to. erosion corrosion .,. Inspection of the J-tube nozzles is not necessary. A O ultrasonic thickness checks of the feed ring header and tee shall $ be perfdrmed on a frequency of every 10 years. This augmented. b examination is self-imposed by TVA (SOI 860515 828) and does not - i,; require a special report. The examination data sheets shall be k included in the 90-day 'ISI report discussed in section 16.2 of this program ,
~
20.9 Control Rod Guide Tube Flexures Due to the potential of intergranular stress corrosion cracking.of guide tube flexures, the following control rod guide tube flexure [ U. augmented examination plan shall be initiated. h The control rod guide tube flexures shall be visually examined in 2 accordance with visual examination method VT-3 (see section 11.1) it-during each refueling outage. These examinations shall be performed in accordance with SMI-0-68-15. At each scheduled O refueling outage the tops of all guide tubes shall be visually { examined to assure that all flexures are in their proper ,. o
.E !
0963o/COC4 i l . ':=" -
, - -. . - .-. . ~ . . . ~ . - . . - , . , . . - . . . . . . . , i l t i SQNP' 51-114.1 ; :) Page 81 of 85 [! Rev. 14 { .j
- 1 orientation. If any flexure heads are discovered to be broken from their~ stems, the heads should be retrieved and flexureless .L, inserts installed. If flexureless inserts are not installed after
'a break has occurred, a safety evaluation will be required to bj 4 confirm continued safe plant operation. A copy of the VT data [.
sheets shall be forwarded to the appropriate cognizant engineer in the Mechanical Maintenance Section for enclosure in SMI-0-68-15. [g This augmented examination is self imposed by TVA and does not I require a report. The examination data sheets shall be included in the 90-day ISI report discussed in section 16.2 of this program. 4.. j
- y. i Any revisions or waivers of this examination plan shall be i initiated by the Mechanical Maintenance Supervisor. b
.! h 20.10 Rod Control Cluster Assembly (RCCA) Cladding Wear Observations J As recommended in a Westinghouse memorandum to J. R. Ratliff from N. R. Metcalf, dated February 9,1984 (Westinghouse letter @
No. 84TV*-G-027) visual examinations of the RCCA's will be , performed during the unit 1, cycle 10 outage period. L L The RCCA's will be visually examined for'the existence of wear marks on the cladding of the RCCA absorber rodlets which exceed 6 the current Westinghouse acceptance criteria for wear depth. k-Examination may be performed through the use of a remote camera. l.~
.s A random sample of 5 (approximate 107.) of the 53 RCCA's will be ,
examined. Determination which RCCA's to be examined and the exact method used will be determined at a later date and based on the i available RCCA usage data. Resolution of the camera's used should j'e allow for the use of an analytical model to infer the depth of wear ind.ications from the observed wear scar widths as measured by the use of photography. { The wear indications are predominantly on the portion of the < cladding surface facing the center or hub of the RCCA at . elevations along the rodlet length corresponding-to the guide , plate surfaces. Note that, by design, some wear is expected due to contact between the RCCA rodlets and the guide surfaces. . t Measurement and wear indication data should be sent to :( Westinghouse for evaluation. This augmented examination is [ self-imposed by TVA (L22 840S01 803) and does not require a y special report. The examination data sheets shall be included in U the 90-day ISI report discussed in section 16.2 of this program. @ r l L 31 0963o/C0C4 [, ' s t ,
~ SQNP SI-114.1 ! Page 82 of 85 i Rev. 14 ! i 20.11 Accelerated Field Weld Program [C.1] !
. . b In accordance with the letter from R. L. Gridley to f B. J. Youngblood dated January 30, 1987 (L44 870130 804) .TVA ,
committed to implement an accelerated field weld program for Sequoyah Unit 1. This augmented and accelerated program requires r, completion of 1007 of ASME Classes 1 and 2 piping, piping support, and component (reactor vessel, steam generator, pressurizer, and c reactor coolant pumps) support field welds that are in the first ' 10-year program in the two.(2) consecutive refueling outages , following restart. The accelerated field weld program shall be i-completed during the unit I cycle 5 refueling outage. The . examination data sheets and results shall be included in the 90-day ISI report discussed in section 16.2 of this program. ; 20.12 Thimble Tube Guide ; Due to the potential thinning of thimble tube guides as reported p. on NRC Bulletin 88-09 and Information Notice No. 87-44 supplement ' 1, TVA will perform an augmented examination of the thimble tubes using Eddy Current Examination (ET) during the cycle 4 refueling ' outage. See memorandum from R. W. Fortenberry to S. J. Smith dated July 21, 1988 (S57 880721 821) and memorandum from J. F. Murdock to J. B. Hosmer dated June 3,1988 (L29 880531 914). . Future examinations shall be scheduled according to the results of
- the cycle 4 exams. Results of the examinations and actions taken shall be sent to the NRC within 30 days of completion of the inspections. t 20.13 Examination of Piping Connected to the Reactor Coolant System Due to Thermal Stresses ';
Due to the potential thermal stresses on unisoluble piping i attached to the RCS (see NRC Bulletin 88-08) TVA has examined the following areas: 1) the four 1.5 inch high head injection lines,
!)
I
- 2) the 2 inch pressurizer spray line from the chargi::g path, 3) the 3 inch alternate charging path, and 4) the 3 inch normal charging path. All of the welds up to the first valve have been PT examined, and the 3 inch line has been UT examined as well. -
These exams were conducted during cycle 3 shutdown period. The SON Project Engineer has prepared a report to Licensing, who 2 in turn prepared a report to the NRC. (See memorandum B29 880906 l 008, 508 880830 843, L44 880824 802, and B25 880819 014.) hI S
-82 09630/COC4 l<
a t
I SONP I SI-114.1 ~t Page 83 of 85 Rev. 14-20.14 Examination of Pressurizer Surge Line Due to Occurrence of Thermal Stratification f. ..
. 4, y;
Due the the thermal stratification of the pressurizer surge line (see NRC Bulletin 88-11) TVA will perform a visual inspection on 4 the pressurizer surge line at the first available cold shutdown after receipt of Bulletin 88-11 which exceeds seven days. This .[ examination should determine any gross discernable distress or 7 r structural damage in the entire pressurizer surge line, including l '- piping, pipe supports, pipe whip restraints, and anchor bolts. f. This examination is scheduled and will be performed per SMI-0-68-4. h The results of the examination are to be-coordinated in accordance with SMI-0-68-4. MTB-MES will submit the results of the examina- . tion to NLRA for submittal to the NRC within 30 days af ter completion of the examinations, 1, . j*
21.0 REFERENCES
u 21.1 Source Documents u 21..l.1 ASME Boiler and Pressure Vessel Code - Section XI 1974 Edition Summer 1975 addenda, 1977 Edition, Summer 1978 ,, addenda.
- u 21.1.2 ASME Boiler and Pressure Vessel Code - Section V. ,
y 21.1.3 Sequoyah Nuclear Plant Final Safety Analysis Report. ' 21.1.4 ' Nuclear Quality Assurance Manual, Part II, Sections 2.3, ; 3.2, 5.1, and 6.3, and Part I, Section 2.16. .1 F 21.1.5 i.Sequoyah Nuclear Plant Technical Specifications Unit 1. g 21.1.6 Code of Federal Regulation, Title 10, Part 50.55a. . 21.1.7 U.S.. Nuclear Regulatory Commission Regulatory Guides 1.14, 1.26,1.50, and 1.83. ' 21.2 Other Documents ', . 21.2.1 Westinghouse Technical Manual - Pressurizer, TM 1440-C225, I-Contract No. 68C60-91934, N2M-2-6. . 21.2.2 Westinghouse Technical Manual - Vertical Steam Generators. [
' TM 1440-C224, Contract No. 68C31-91934, N2M-2-4. y m: ,
21.2.3 Westinghouse Instruction Manual - Auxiliary Heat [t Exchangers, Contract No. 68C60-91934, N2M-2-25.
.1
-83 0963o/C0C4
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'Page.84 of 85-Rev, 14 :'
q
-21.2.4 Westinghouse Instruction Book r Reactor Coolant Pump, [g .,
Contract No. 68C60-91934, N2M-2-5.
. :q
- 21.2.5 Ingersoll-Rand Instruction Manual - Residual Heat Removal h' Pumps Contract No. 68C60-91934, N2M-2-30. L L
21.2.6 Sequoyah Nuclear Plant Technical Instruction TI-51. L . L k 21.2.7 Program Procedures, Quality Methods Procedures and i Technical Standards: 0202.14 or QMP 102.4, 1403 - T.S. .. 09.02.03 (N76A10), 1502.07 or QMP 110.5, 1502.01 (N83A3), 's 1502.06 (SEQ 80M7), 1502.03 (SEQ 82E1),.1402.02 (SQ82MI). L Administrative Instructions AI-12 (Part I), AI-12 ,. .. e u , (Part III), and AI-19 (Part IV). '; ;' 21.2.8 Memorandum from D. W. Wilson to H. L. 'Abercrombie, dated ? August 1, 1986 "Sequoyah Nuclear Plant, Units 1 and 2, -
" Indications in-Class 1 Welds" (L18 860730 899). '
o 21.2.9 Instruction Manual - 173-inch I.D. Reactor Pressure Vessel 1
- Rotterdam Dockyard Company, Contract No. 68C60-91934,
.N2M-2-3. (
21.2.10 Sequoyah Nuclear Plant Instruction SI-1. !
.a 21.2.11 Memorandum from V. A. Blanco to C. R. Brimer dated
~ January 10, 1987 (B25 870109 045).
l~ 21.2.12 Memorandum from G. C. Zech to S. A. White dated January 25, 1988 allowing use of code cases.N-341 and N-356. P 21.2.13!Nemorandum from P. R. Wallace to W. T. Cottle End C. C. Mason dated May 11, 1984 (L22 840501 803), " Nuclear Plant h'
?:
Operating Experience Review - Westinghouse letter . 84VT*-G-027, RCCA Cladding Hear Observations." ,* 21.2.14 Letter from R. L. Gridley to B. J. Youngblood dated January 30, 1987 (L44 870130 804). . 21.2.15 NRC Bulletin 88-09 and information Notice 87-44 supplement 1, " Thimble Tube Thinning in Westinghouse Reactors." .h
.'s 21.2.16 Memorandum from R. W. Fortemberry to S. J. Smith dated h..
- July 21, 1988 (S57 880721 821). *n n.
21.2.17 Memorandum from J. F. Murdock to J. B. Hosmer dated June 3, 1988 (L29 880531 914). d l hJ 0963o/COC4 i,
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SQNP.. j-SI-114.1 ie Page 85 of 85 Rev. 14 . J[ f.. 21.2.18 NRC-Bulletin' 88-08, " Thermal Stresses on Piping Connected T to Reactor Coolant. Systems." f.. 21.2.19 NRC Bulletin 88-11, " Pressurizer Surge Line Thermal . .!
. Strati fi ca tion.'" (.
21.2.20 Memorandum from J. F. Murdock to D. L. Williams dated I August 22, 1988 (B46 880822 582)
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- LIST OF DRAWINGS - UNIT 1 f; j '
Reactor Vessel " Drawing No. Appendix B $ Title Page No. !. '
'f*f, CHM-2341-B Reactor Vessel Stud Locations and Details 4-5 Y CHM-2343-B Reactor Vessel Seam Welds 6 CHM-2358-A r.
Reactor Vessel Closure Head 8 CHM-2359-A Control Rod Drive Housing 9-10 CHM-2360-A [ Reactor Vessel Inlet Nozzles 11 0 CHM-2361-A Reactor Vessel Outlet Nozzles 12 ?. ' CHM-2651-C CRD, UPI and Vent Pipe Penetrations 16-17 ISI-0014-A Auxiliary Head Adapter t 19 MSG-0004-C Reactor Vessel Bottom Head Penetrations 67 ,L l N. Presurizer ,,, Drawing No. Appendix B .' Title Page No. ;c CHM-2362-A Pressurizer Support Skirt Weld 13 U' CHM-2363-A Pressurizer Seam Welds 14 P MSG-0002-B Pressurizer and Steam Generator Cladding Patches and Manway Bolting 65 y MSG-0006-A -Pressurizer Heater Penetrations 69 ! Steam Generators Appendix B Drawing No. Litle Page No. 76M1 Vertical Steam Generators Tube Sheet Arrangement 70 CHM-2345-B Steam Generator - 7 F MSG-0002-B Pressurizer and Steam Generators Cladding Patches 65 i;, MSG-0005-A team Generitor/Feedvater Tranrition Spool Piece 68 e Heat Exchangers - Appendix B Dr_ awing No. Title Page No. ? CHM-2404-A Residual Heat Removal Heat Exchanger Channel Welds and Support 15 ISI-0066-A Regenerative Heat Exchanger 20-22 ISI-0067-A Excess Letdown Heat Exchanger ISI-0231-B Centrifugal Charging Pump 011 & Gear Cooler 23 36 i[K, ISI-0235-A Safety Injection Pump 011 Cooler 40 ' ISI-0232-8 i Waste' Gas Comoressor Heat Exchanger 37 e
'ISI-0230-A Gross failed fuel Detector Heat Exchanger
- 35
ISI-0229-A Positive Displacement Pump 011 Cooler 34 : ISI-0237-A Seal Water Heat Exchanger 42 ISI-0233-B Centrifugal Charging Pump Mechanical Seal Cooler 38 ISI-0234-C RHR Pump Seal Cooler 39 i
-110 0035o/C0C4
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, Page 2 of 187 l Rev. 14 '.
APPENDIX B I _ LIST OF DRAWINGS - UNIT 1 C Heat Exchangers (Continued) !-l ai Appendix B [l Drawing No. Title Page No. ,C' ISI-0236-A Safety Injection Pump Seal Cooler 41 ISI-0215-A Containment Spray Heat Exchanger 26 ,. ISI-0226-8 Gas Stripper & Boric Acid Evaporator Package 29-32 ,, ISI-0216-A Non-Regenerative Letdown Heat Exchanger 27 i:L ISI-0284-B Component Cooling Heat Exchanger 48 ISI-0285-A Essential Raw Cooling Water Strainer 49 ISI-0287-B Spent Fuel Pit Heat Exchanger 51 ISI-0290-B RHR Heat Exchanger Secondary Side Supports 53 . i Piping and Valve Held Maps Appendix B R Drawing No. Title Page No. [ CHM-2333-B Reactor Coolant Piping (Main Loops) 71-72 CHM-2333-C Safety Injection System 73-80 . ISI-0116-C Safety Injection System 140-141 : ., CHM-2334-C Reactor Coolant System 81-87 CHM-233S-C Chemical and Volume Control System 88-89 CHM-2336-C Residual Heat Removal System 90-93 CHM-2337-C Upper Head Injection System 94-96 CHM-2338-C Seal Water Injection (Chemical and Volume ;. Control System) 97-100 CHM-2339-C Feedwater System 101-102 4, CHM-2340-C Main Steam System. 103-104 > CHM-2422-C Containment Spray. System 105 e V, Class 1 and 2 Piping and Valve f Support Orawings Appendix B Drawing No. Title Pace No. ' CHM-2432-C Reactor Coolant System 106-112 Chemical and Volume Control System CHM-2433-C 113-114 , CHM-2434-C Seal Water Injection (Chemical and Volume Control) 115-118 . CHM-2435-C Residual Heat Removal System 119-122 9i CHM-2436-C Safety Injection System 123-129 fI CHM-2437-C CHM-2438-C Upper-Head Injection System Main Steam Line System . 130-132 133-134 (' CHM-2439-C Feedwater System 135-136
- CHM-2440-C Containment Spray System 137 ISI-0303-C Reactor Coolant System Support Locations (Main Loop) 55 ISI-0322-C Reactor Coolant System (Main Piping) Crossover Leg Restraint 57 0035o/COC4 i
-111-I
___________i
SONP SI-114.1 Page 3 of 187 Rev. 14 ,, APPENDIX B i LIST OF DRAWINGS - UNIT 1 -
~
i Class 3 Pipina and Valve Support Drawings Appendix B ! Drawing No. Title ' _ Pace No. I I SI-Oll 3-C Auxiliary Feedwater System 138-139 ISI-0123-C Essential Raw Cooling Water System 142-168 I SI-0126-C Component Cooling Water System 169-184 ISI-0127-C Fuel Pool Cooling System 185-187 b Pumos Appendix B . Drawing No. Title Page No.
- CHM-2675-8 Reactor Coolant Pump Main Flange and Lower ,;
Seal House Bolt Pattern 18 ! MSG-0003-8 Reactor Coolant Pump Casing 66 ISI-0325-B !i Reactor Coolant Pump Support 58 r ISI-0305-B Reactor Coolant Pump Motor Flywheel Examination 56 ISI-0099-A Residual Heat Removal Pump Supports 25 . ISI-0353-B RHR Pump Support Locations 59 ISI-0238-A Thermal Barrier Booster Pump 43 - 151-0256-B Turbine Driven Auxiliary Feedwater Pump 44 ISI-0262-B Motor Driven Auxiliary Feedwater Pump 45 ' ISI-0281-A Component Cooling Hater Pumps 46 ISI-0283-A Essential Raw Cooling Hater System Support Detall 47 ISI-0286-B ERCW Pump Support 50 ISI-0288-8 Spent Fuel Pump Support 52 " ISI-0296-8 ERCri Screen . Wash Pump Supports 54 i L : Tanks i Appendix B l Drawing No. Title _,Page No. l ISI-0069-A Boron Injection Tank 24 ISI-0225-A Seal Water Injection Filter 28 15I-0227-8 Component Cooling Surge Tank 33
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