ML20087C539

Rev 1 to Procedure SI-114.1, Inservice Insp Program

ML20087C539
Person / Time
Site:	Sequoyah
Issue date:	01/04/1984
From:	Lehberger N TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20087C503	List: ML20087C501 ML20087C539
References
SI-114.1, NUDOCS 8403130100
Download: ML20087C539 (425)

Text

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{ Sequoyah Nuclear Plant

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DISTRIBUTION

'i Cover Page 1 1C Plant Master File Pwr Plant Superintendent

!; asst Pwr Plant Supt., (Oper.)

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Asst Pwr Plant Supt., (Maint.)

Asst Pwr Plant Supt., (H&S)

r- Administrative Supervisor j SURVEILLANCE -INSTRUCTION IC Maintenance Supervisor (M) =

SI-114.1 IC Assistant Maintenance Sul._ .isor (M]

T Maintenance Supervisor (E)

Assistant Maintenance Supervisor (E]

d IN-SERVICE INSPECTION PROGRAM Maintenance Supervisor (I)

FOR TENNESSEE. VALLEY AUTHORITY IU Engineering Supervisor m SEQUOYAH NUCLEAR PLANT 3C Operations Supervisor si Quality Assurance Supervisor lu - s Health Physics Supervisor -

f. Public Safety Services Supv.

1 Chief Storekeeper s Preop Test Program Coordinator

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, . g; , . t.. a.e c; - w.. - a;% . .c- e Field Services Supervisor '

Chemical Engineer (Engineering) a.l Radiochem Laboratory-Instrument Shop

- Reactor Engineer (Engi2eering)

. instrument Engineer (Maint. I) .

1 Mechanical Engineer (Engineering)

~

Plant Services Supervisor Tu Prepared By: Norman Lehberger Training Center Coordinator Public Safety Services - SQNP

- Shift Engineer's Office Revised By: ,

Norman Lehbergtf? 1c Unit Control Room

[f,e Health Physics Laboratory Submitted By: u) IC Ncir Document Control Unit-C

/ Superveso Pwr Plant Superintendent, WBNP IU_ Pwr Plant Superintendent, BFNP Concurrence: e b , IU Pwr Plant Superintendent, BENP NEB,-K

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, Resident NRC Inspector - SNP

• T NSRS,-K -

PORC Reviewi /-#-Bd IU Supv. ISI Group- 1520-CSTa-C Date 1C Unit Control Room #2 s j

, # 1U Compliance Section Staff Supv.

Approved By: // Reactor Systems Group-C

,e r Plt Superintendent IU Supv. NDE Sect.-C 1520 CST 2-C 1H R mr: Proframs & Prneadures.1520'CS'

{J . Date Approved: 1 1J4 lu 'Supv. Inspection Sect., 1520 CST 27 IU NRC-101 Marietta St. JW Suite 3100'.At v _1U Stationary Equipment Group-C GA I

The' current revision' level of this instruction is: 1 .

30 Reason for current revision (include all temporary change numbers) Revision to incorporate slan plan data, 00AM channes and make corrections.

s L The last page of this instruction is number 421 .

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8403130100 840307 PDR ADOCK 05000327 G PDR _

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SURVEILLANCE INSTRUCTION Cover Page 2 '

History of Revisions -

SI,114.1

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.g i :- Rev. Level Date Revised Pages -

Rev. Level Date Revised Pages i 0 7/20/82 All I. *1 JAN 4M All ,

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' Reactor Vessel' Repair-Welds Beltline Region

.; Piping and Valve Integrally-Welded Support Hember Thickness

_ Class B Systems-Functional Leakage SI's E.

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Page No. L 1.0 Statement of Applic-hility I 2.0 -Purpose / r 3.0' Inspection Intervals and Inspection Periods J 4.0 Codes of Record . 2. 5.0 Method of Implementation and Responsibilities 2 6.0 Abbreviations and Definitions y 7.0 Components Subject to Examination - TVA Safety Class A f 8.0 Components Subject to Examination - TVA Safety Class B 3/ 9.0 Components Subject to Examination - TVA Safety Class C fo and D ~ 10.0 Authorized Inspector f/ . 11.0 Exami$1ationMethods g/ 12.0 Qualifications of Nondestructive Examination Personnel ff 13.0 Acceptance Criteria ff 14.0 Repairs , ff 15.0 Replacements fr - 16.0 Records and Reports- . ff 17.0 Notification of Indication go 18.0 Calibration Blocks // . m 19.0 Requests for Relief gj 20.0 Augmented Inspections , g/ 21.0 References 8f ~ . 4 n -- - c , .. ~ , .- ~' SQNP i SI-114.1 Page 2 of 2 Rev 1 TABLE OF CONTENTS - Page No. Appendix'A - Tables and Drawings 44 Appendix B - Calibration Block Drawings / f '# Appendix C - Data Sheets /76 m .; . Appendix D - Notification of Indication Form /97 ,._ Appendix E - Requests for Relief /9r J - Appendix F - Augmented Inspections 12.o ~ Appendix G - Scan Plan Data 2 30 s r ![ r- . {. . f i ,. - 9 l I si. l' ,r b-Wh-  % t l- L l

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i e- w e 3 SQhT Y - SI-114.1 T Pege'l of 65 Rev 1 ie m 1.0 STATEtENT OF APPLICABILITY 'M This program outlines requirements for performing the first '10 year' interval. inservice nondestructive examinations of the t Sequoyah Nuclear Plant, unit 1, ASFE Code Class 1, 2, and 3 i components (and their supports)lcontaining water, steam, or - radioactive material (other than radioactive waste management ( systems). The program has been organized to fulfill inservice examination' requirements of'the Sequoyah Operational Quality ' Assurance Manual, Part II, Section 5.1 and comply as practical with the requirements of Section XI of the AStE Boiler and q7 Pressure-Vessel Code. In addition, this program implements applicable portions of the Sequoyah Technical Specifications. The Inservice Inspection Program (ISI) satisfies the requirements of T~ Surveillance Requirements 4.4.3.2.4, 4.4.5.0, and 4.4.10, and L partially satisfies the requirements of Surveillance Requirement 4~0.5. ~The requirements of this program are applicable beginning at the date of commercial' operation of the unit. [ Specifics concerning performance of nondestructive examinations are not a part of this_ program, but arc included in nondestructive p examination procedures (DPM N80E3 and SNP TI-51). n - L ' 2.0 PURPOSE The ISI program shall.be used for planning inspections and i; . nondestructive examinations of the Sequoyah AStE Class 1, 2, and 3 components for the first interval. Personnel responsible for , performance of the examinations should familiarize themselves with

• he requirements of this program prior to performing the exami-nations.

The examinations required by.this program will establish accip-j[a tance of components for continued service. 3.0 INSPECTION INTERVALS AND INSPECTION PERIODS f - The inservice examinations required by ASPE Section XI shall be performed during each 10-year interval of service (inspection

SQSP -. - ST-114.1-Page 6 of 65 - Riv-l' . evaluation and acceptability was based on Code Case N-209 (see nonconforming report 6P for i additional information concerning flaw evaluation and location). In addition to the f examinations 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 - unchanged, then the examination frequency may revert to that of Examination Category B-A of " Table IWB-2500-1 of ASME Section XI. If the successive examinations reveal that the flaw indication size has ' increased, then technical ju::tification shall be presented to NRC for allowing continued operations. os

~ The closure head ring is fabricated of A-508, Class 2, manganese-molybdenum steel. The y closure head hemispherical section is fabricated

! of A-533, Gr. B, Class 1, manganese-molybdenum I~ steel. Both sections are clad with veld depos-ited austenitic stainless steel. .(* 7.1.1.4 Lower Head Circumferential Weld The entire length of the lower head circumfer-ential weld (approximately 38 feet in length) will be ultrasonically examined during the third v inspection period using remote inspection i devices from the vessel inside diameter with the I - core internals removed (see Request-for Relief 7 ISI-5). . ) L~ The. lower head sections are fabricated of A-533, . Gr. B, Class 1, manganese-molybdenum steel, and j 2 are clad with weld deposited austenitic l [_ stainless st. eel. l 7.1.1.5 . . Closure Head Meridional Weld D The closure head does not include any meridional welds. . _ . 7_.1.1. 6 Lower Head Meridional Welds 1 There are six meridional welds (each approxi-j~ -mately 4 feet in length) located in the lower head. The entire length of each of these welds l , will be ultrasonically exa6ined during the third ) inspection period using remote inspection {' devices from the vessel inside diameter with the p, core internals removed. The lower head section t. ~ . -6 '  ! i; ) l g . SQNP SI-.114.1 . Pigs 7 of 65 Rev 1 , material is identified in section 7.1.1.4 of this program. l 7.1.1.7 Shell-To-n ange Weld ' e \ i

The entire length of the shell-to-flange weld (approximately 50 feet in length) will be i ultrasonically examined from the vessel inside diameter using remote inspection devices during the third inspection period.

Y The vessel flange section is fabricated of A-508, Class 2, manganese-melybdenum steel and is clad internally and on the gasket face with j. weld deposited austenitic stain 1::ss steel. 7.1.1.8 Closure Head-To-Flange Weld y l 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 i examined each inspection period is included in . Table A of Appendix A. The closure head flange section is fabricated of A-508, Class 2, manganese-molybdenum steel and is clad internally and on the gasket face with weld deposited austenitic stainless steel. a 7.1.1.9 Repair Welds Base metal weld repairs in the beltline region where the repair depth exceeds 10 percent f nominal of the vessel wall shall be ultrasonically examined. 7.1.2 Reactor Vessel Nozzle-To-Vessel Welds k There are four inlet nozzles (27.500 inch I.D.) and four outlet nozzles (29.937-inch I.D.). The eight

[3 nozzle-to-vessel welds will be ultrasonically examined

-q from the inside diameter using remote inspection devices. - The four outlet nozzle-to-vessel welds shall be ' ~' ultrasonically examined from the nozzle bore when the upper internals are removed during the first I, inspection period. The inlet nozzles are not accessible until the core barrel is removed. The four inlet nozzle-to-vessel welds shall be ultrasonically examined from the nozzle bore and from the vessel L.i shell inside diameter during the third inspection . period when the core barrel has been removed. The ( outlet nozzle-to-vessel welds shall also be . i [ SQNP ~" SI-114.1 Pags 8 of 65 Rev 1-W ultrasonically' examined from the vessel shell inside - diameter during the third inspection period when the

• ! core barrel has been removed.

[ 'The-nozzle forgings are fabricated of A-508, Class 2, manganese-molybdenum steel and are clad with weld ideposited austenitic stainless steel. ?. 7.1.3 Reactor Vessel Nozzle Inside Radius Section H The eight nozzle insidd radius sections (including the outlet nozzle integral extensions) shall be ultrason-ically examined at the same time as the examination of

a the nozzle-to-vessel welds conducted from the nozzle bore (see Section 7.1.2). -

Nozzle forging material is identified in Section 7.1.2 T of this program. o .7.1.4 Reactor Vessel Partial Penetration Welds The vessel includes 4 upper head injection nozzles, 1 vent pipe nozzle, 78 centrol rod drive nozzles, and 58 instru-

i. -mentation nozzles with partial penetration welds.

J Approximately 25 percent of each group of nozzles shall be ~ visually examined from the vessel outside diameter in accordance with visual examination method VT-2 (see L. section 11.1). This 25 percent shall include 1 upper head 1' . injection nozzle, 1 vent pipe nozzle', 20 control rod drive nozzles, and 15 instrumentation nozzles. Examinttion of these nozzles during the inspection interval shall be j' distributed among the inspection periods in accordance with Table A of Appendix A. ~ ~ [L The NDE Section shall designate in scan plans the nozzles-requiring examination to be performed in accordance with SI-146 or SI-250. The examination F results may be recorded cn SI-146 or SI-250 data sheets. J-

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.7.1.5 Reactor Vessel Nozzle-To-Safe End Welds 1 . The inlet and outlet nozzle-to-safe end welds shall be lM ultrasonically examined from the inside' diameter using remote inspection devices. The ultrasonic examination I b., shall be' performed at the same time-as the examination

C of the nozzle-to-vessel welds conducted from the nozzle bore (see Section 7.1.2).

i All of the nozzle-to-safe end welds shall also be liquid penetrant examined during.the inspection interval coincident with the ultrasonic examinations. The'Sequoyah Nuclear Plant Unit 2 Report--Evaluation ~~ ~ of Cracking in Reactor Vessel Nozzle Stainless Steel Buttering, Unit 1. nozzle-to-safe end welds RC-9SE and .g

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_g. . ihu^ SQNP { , SI-114.1 Page 9 of 65 Rev 1 ,

r. RC-32SE shall be examined during the first inspection i interval. The remaining nonzles will be examined b

during the-normal inservice inspection intervals. Each nozzle safe end weld is a stainless steel type 304 weld build up. , ~- 7.1.6 Reactor Vessel Pr.ssure Retainino. Bolting Larger Than Two Inches In Diameter .- During each refueling outage all closure studs, nuts,

. -and washers are removed. All of the 54 closure studs

" and nuts shall be examined during the inspection y interval in accordance with Table A of Appendix A. i~ , The closure nuts shall be magnetic particle examined ,- and the closure studs shall be ultrasonically and . magnetic particle examined. The closure studs may be ~ ultrasonically examined in place under tension, when the closure head is removed, or when the studs are removed. I Provisions for this examination are included-in L MI-1.2. i~ 'All of the 54' ligaments between the vessal flange stud L holes shall be ultrasonically-examined during the inspection interval in accordance with-Table A of p' .' Appendix A. '(This provides for ultrasonic examination ! of the threads.in base material.) The.54 closure washers shall be visually examined-in accordance with visual examination method VT-1 (see-section 11.1) during the inspection interval in accordance with Table A of Appendix A. , {L The 7-inch diameter studs, nuts, and washers are fabricated of SA-540, Gr. B23 Nickel, chrome-molybdenum' steel. - t 7.1.7 Reactor Vessel Pressure. Retaining Bolting Tso Inches , , . . And Smaller In Diameter %- There is no pressure retaining bolting two inches or smaller in' diameter. 2 1.1.8~ Reactor' Vessel Integrally Welded Attachments t There are no integrally welded vessel supports. $. 7 7.1.9 Reactor Vessel Interior , 1 The vessel interior shall be visually examined in accordance with visual examination method VT-3 (see Section 11.1). These examinations shall include the space above and below the reactor core that is made E . SQNP ~ SI-114.1 Page 10 of 65 Rav 1 accessible for examination by removal of components during normal refueling outages. The examinations shall be performed at the first refueling outage and r.. subsequent refueling outages at approximately three l year intervals. These examinations are implemented by SMI 0-68-12. 7.1.10 Reactor Vessel Removable Core Support Structures The visually accessible attachment welds and visually ~ accessible surfaces of~the core support structure shall be visually examined in accordance with visual examination method VT-3 (see Section 11.1). Inis examinatiran may be deferred to the chird inspection ) period. The structure shall be removed from the reactor vessel for examination. These examinations are implemented by MI-1.4. 7.1.11 Reactor Vessel Control Rod Drive Housings - There are 78 control rod drive housings penetrating i the closure head. Each housing includes a pressure retaining dissimilar metal butt weld ad canopy seal _ weld. The canopy seal welds are exempted from volumetric or surface examination in accordance with I'w3-1220 (a ) '. , There are 20 peripheral control rod drive housings. Two-(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 we.ld. 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, 304SS upper portion '~ and a SB-166 lower portion. 7.2 Pressurizer 7.2.1 Pressurizer Circumferential Shell-To-Head Welds L There are two circumferential shell-to-head welds, each approximately 24 feet in length. The entire length of each veld shall be ultrasonically examined during the inspectior. interval in accordance with Table A of Appen' dix A. G e . 1 8- - SONP. N ' SI-114.1 g Pags 11 of 65 Rev 1 . 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

~ ^ .circumferential shell-to-head weld. One foot of each 1 . longitudinal weld shall be ultrasonically examined -during the inspection periods in accordance with p Table A of Appendix A.' The one foot of weld examined F during each examination shall include the length of wcld as measured from the point of intersection of the longitudinal weld with the circumferential head-to-shell weld. L . . .The vessel shell section material is identified in F Section 7.2.1. 7.2.3 Pressurizer Circumferential and Meridional Head Welds r There are no pressurizer circumferential or meridional ti L head welds. 7.2.4 Pressurizer Nozzle-To-Vessel Welds and Nozzle Inside L Radius Section The pressurizer includes three 6-inch safety valve . ~ l nozzles, one 6-inch ~ relief valve nozzle, one 4-inch-nozzle, and oneL14-inch' surge nozzle. .All of the - nozzle-to-vessel welds, including nozzle.inside radius 'sectica,' shall.be ultrasonically examined during the . inspection interval in'accordance with Table A of Appendix-A. , c The nozzles are fabricated of.SA-508, Class 2, manganese-molybdenum steel.

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7.2.5 Pressurizer Heater Penetration Welds. [h , 's-; 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 iI accordance with visual examination method VT-2 (see L. Section 11.1). Examination of these penetrations ~ during the inspection interval shall be distributed b among the; inspection periods in accordance with 1, Table A of Appendix A. i The NDE Section shall designate in scan plans the !p heater penetrations requiring examination such that l gn the organization performing the examinations may .. document examination results on SI-146 or SI-250 data l} sheets. lL

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. t' H SQNP JT . , SI-114.1 cf - Page.12 of.65' Rev i s 7.2.6 . Pressurizer Nozzle-To-Safe End Welds L . 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.in accordance with Table A of Appendix A. - Safe end'connt tions are'SA-182, Gr. F-316L forgings. a 7.2.7 Pressurizer Pressure Retaining Bolting Large'r Than Two Inches In Diameter There is no pressure retaining bolting larger than two inches in diameter. k-7.2.8 Pressurizer Pressure Retaining Bolting Two Inches -1 And Smaller In Diameter All of the pressurizer manway bolts shall be visually _, examined in accordance-with visual examination method 5 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 t- 'are removed. -It is preferrable to perform 1the 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. , -7.2.9- Pressurizer Integrally Welded Supi ,rt Attachments (?. The entire length of the pressurizer support skirt-to-vessel. weld (approximately 23~ feet in' length) shall be surface' examined (magnetic particle) during the h'^ inspection interval in accordance with Table A of Appendix A. , , The support skirt.is approximately 1.5 inches thick L and is fabricated of SA-516, Gr. 70, carbon steel plate. U.' I;, ' 7.3 Steam Generators (4)' - 3+ 7'.3.1 ' Steam Generator Primary Side Circumferential (' And Meridional Head Welds There are no steam generator primary side circumfer- ~ l '. ential or meridional head welds. l t n . _ . . - _ _ , SQNP ._ -i' . SI-114.1 !~ Paga 13 of 65 Rav 1 n 7.3.2 Steam Generator Primary Tubesheet-To-Head Weld j Each steam generator includes a tubesheet-to-head weld (approximately 36 feet in length). The entire length of each weld shall be ultrasonically examined during the inspection interval. The entire length of a tubesheet-to-head weld shall be examined during the ~ first and second inspection periods. The entire

1. length of the two remaining welds shall be examined during the third inspection period.

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 schieve i meaningful results due to limitations of axamining integrally cast material (see request for relief (( ~ ISI-6). .i The nozzles are fabricated to SA-216, Gr. WCC. 7.3.4 Steam Generat'or Primary 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. .t The nozzles have buttered 308L safe ends. - 7.3.5 Steam Generator Primary Pressure Retaining Bolting Larger Than Two Inches In Diameter ,{:# , There is no pressure retaining bolting larger than two inches in dica.ter. f 7.3.6 Steam Generator Primary Pressure Retaining Bolting Two Inches And Smaller In Diameter

- Each steam generator includes two manways. All the

-1 manway 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 A inspection interval in accordance with Table A of Appendix A. , { - - , , - - - . , - - . - . -- , r -. , . - - - - - - - , - . . SQNP r +; SI-114.1 [_ Page 14 of 6! J Rav 1 h t The bolts may be examined in place under tension or when'the bolts are removed. It is preferrable to i perform the examinations when the bolts are removed if g possible. Removal of bolting is performed in accordance with MI-3.1 :.nd provides for examination of -bolting. 7 Eac'h manway. includes 16 bolts at 1.88 inches in diameter. 7.3.7 Steam Generator Primarv Integrally Welded Support Attachments .There are no integrally welded vessel supports. The four main support pads are secured to the steam L generator ficid support system by high strength bolts. . ,~ 7.3.8 Steam Generator Tubing Each steam generator tube bundle consists of 3,388 NiCrFe alloy (Inconel SB-163) U-tubes of 0.875 0.D. by J 0.050 average wall thickness. During the inspection interval, steam generator tabing sl.all undergo eddy current examination. This ~

examination is done in accordance with Sequoyah

, . technical specifications and satisfies Surveillance t- .- Requirement 4.4.5.0. , 7 . 3. 8.1 - Steam Generator Sample Selection And Inspection + Each steam generator shall be determined operable during the shutdown by selecting and inspecting at least the minimum number of ql steam generators specified in Table 1. " T 7.3.8.2 Steam Generator Tube Sample Selection and Inspection ,

• The steam generator tube minimum sample size inspection result classification, and the

. corresponding action required shall be as M' -specified in Table SG-2. The inservice

inspection of steam generator tubes shall be -

@ performed at the frequencies specified in y Section 7.3.6.3, and the inspected tubes shall be verified acceptable'per the J acceptance criteria of section 7.3.8.4. The y tubes selected for each inservice inspection shall include at.least 3% of the total number L of tubes in all steam ' generators; the tubes . selected for these inspections shall be " selected on a random basis except: m r = SQNP

r. - SI-114.1 i Pcgs 15 of 65 Rev 1

, a. Where experience in similar plants with .l'- similar water chemistry indicates critical areas to be inspected, then at 7" least 50% of the tubes inspected shall be from these critical areas.

b. The first sample of tubes selected for each inspection (subsequent to the

, preservice inspection) cf each steam generator shall include: ] '

1. All nonplugged tubes that previously had detectable wall penetrations

.r >20%.

2. Tubes in those areas where experience has indicated potential r problems.

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3. A tube inspection (pursuant to Section 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 J probe for a tube inspection, this 4

shall be evaluated and recorded and an adjacent tube shall be selected and subjected to a tube inspection.

c. The tubes selected as the second and

, third samples (if required by Table 2) L l- during each inservice inspection may be subjected to a partial tube inspection provided: ,

1. , 1. The tubes selected for these samples include the tubes from those areas of the tube sheet array where-tubes

"~ with imperfections were previously found. v ' . 2. The inspections include those L portions of the tubes where I imperfections were previously found. }_ The results of each sample inspection zhall be classified into one of the following three categories: . Categorv , Inspection Results C-1. Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes j . are defective. L r-k SQ3P -

• SI-114,1 Pags 16 of 65

- Rrv 1 C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or 3 . between 5% and 10% of the total Li- 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 A degraded tubes must exhibit significant (>10%) further wall ' l[' 'j: penetrations to be included in the above percentage calculations. .

- 7.3.8.'3 inspection Frequencies

.The above required inservice inspections of

,. steam generator shall be performed at the following frequencies:

a. The first inservice inspection shall be

~- performed after 6 Effective Full Power Months but within 24 calendar months of initial criticality. Subsequent

- inservice inspections sha,ll be performed

'l - at intervals.of not less than 12 nor more . ^ than 24 calendar menths after the __ previous inspection. If two consecutive inspections following service under.AVT ccaditions, not including the preservice inspection, result in all inspection ~ - results falling into the C-1 catego'yr or ! if two consecutive inspections demonstrate'that previously observed ~ E- degradation has not continued and no additional degradation-has occurred, the inspection interval may be extended to a

7. .

maximum of once per 40 months.

b. If the results of the inservice inspec-tion of a steam generator conducted in accordance with Table 2 at 40-month

2. 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 L' .

9 SQNP. p , SI-114.1 t Page 17 of 65 Rev 1 a- apply until the subsequent inspections r satisfy the criteria of section 7.3.8.3.a; the interval _ may then be exter.ded to a ,,. maximum of once per 40 months. n

c. -Additional, unscheduled inservice

> inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 2 during the shutdown subsequent to any of'the'following conditions. i

1. Primary-to-secondary tubes leaks (not a ,

including leaks originating from tube-to-tube sheet welds) in excess of the limits of Technical Specification - 3.4.6.2. r [~ ' 2. A seismic occurrence greater than the Operating Basis Earthquake; -t'^

3. A loss-of-coolant' accident requiring actuation of the engineered safeguards.

4. A main steam line or feedwater line ,

. break. y i 7.3.8.4 Acceptance Criteria e

r. a. As used in'Section 7.3.8:

] -

1. Imperfection means an exception to the dimensions, finish or contcur of,a

tube from that required by fabrication i drawings or specifications. Eddy-current testing indications below 20%

of the nominal tube wall thickness, if 1, detectable, may be considered as imperfections. . 2. Degradation means a service-induced ~ cracking, wastage, wear or general corrosion occurring on either inside - or outside of a tube. ~

3. Degraded Tube means a tube containing

-/ imperfections >20% of the nominal wall L thickness caused by degradation.

4. Percent Degradation means the percent-

age of the tube wall thickness affected or removed by degradation.

j '* - , ., _. .. .-.,c_ _ - m - . . - ,.. , p_, _,,._._._ -- - - , _ , , , _ _ _ SQNP . SI-114.1 h P. age 18 of 65 Rev 1 I" 5. Defect means an imperfection of such ! severity that it exceeds the plugging limit. A tube containing a defect is p defective.

6. Plugging 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. . tw 7. Unserviceable describes the condition h-' of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of .f; . . 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. 4

8. Tube Inspection means an inspection of f the steam generator tube from the

} point of entry (hot leg side) com-pletely around the U-bend to the top

. support of the cold leg.

L - 9. Preservice Inspection means a tube inspection of each steam generator , T tube performed by eddy current - techniques prior to service to establish a baseline condition of the

r tubing. This inspection shall be

, performed prior to initial power operation using the equipment and

,_ techniques expected to be used during

subsequent inservice inspections.

n[.:.

b. The steam generator shall be determined ir operable after completing the corre-b sponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks)

. required by Table 2. 7.4 ~ Piping - . All Class A piping systems to be examined are fabricated of stainless steel. The reactor coolant main loor piping straight lengths are centrifugal cast and the elbows are static cast. t- The upper head injection auxiliary head adapter is included in Section'7.1.1.3. Specific material specifications for each l- piping system are included in weld map isometrics in Appendix A -(see Request for Relief ISI-3). i , SQNP ' p' , SI-114.1 jh ' Page 19 of 65 , Rev-1

7 e the following Class A piping systems are subject to examination

' Reactor Coolant - Chemical and Volume Control. D- . Residual 1 Heat Removal Safety: Injection Upper Head. Injection.. r p 7.4.1 Piping Dissimilar Metal Welds' , .c - .There!are no Class A dissimilar metal welds. E i- - -.7.4.2 Piping' Pressure Retaining Bolting Larger Than 2 Inches _ > In Diameter- .1.t . i e -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 Diameter ll'M* l The following sections define the number of bolted w ' piping flange connections in each system. All of.the bolts or studs-and nuts in each flange connection shall be visually-examined during the inspection [l - - interval in accordance with visual examination method VT-1:(See Section 11.1). The examinations shall be t distributed'during the inspection interval in

}; accordance with Table A of Appendix A.

The bolting may be examined in place'under tension or t when the bolting is removed. _

s. ..

7.4.3.1 Reactor-Coolant System Piping Bolting lL . . L The Reactor Coolant System piping includes eight bolted flange connections. 'h. 7.4.3.2 Chemical and Volume Control System Piping Bolting _ - . The-Chemical _and Volume Control System piping includes four bolted flange connections (from  ; . seal water injection). ~

c 7.4.3.3 Residual Heat Removal' System Piping Bolting -

'm The Residual Heat Removal System piping does not include any bolted connections. . 7.4.3.4 Safety Injection System' Piping Bolting l W' , ' The. Safety Injection System piping includes I four bolted flange connections. S D: L .__.._ _ ___ _._i.~-. _ _ -_ ,. . . =- ,- . _ . . . . _ ~ .. . ..- SQNP E  !? . SI-114.1 ., ~Page 20 of 65 Rev 1 % 7.4'.3.5 Upper Head Injection System Piping Bolting i The Upper Head Injection System piping ' includes twelve bolted grayloc connectors. 7.4.4 ,Circumferential and Longitudinal Pipe Welds '~ x The entire length of each circumferential pipe weld selected for examination shall be ultrasonically and/or liquid penetrant examined as practical. All ~ Class A piping is seamless. Circumferential pipe welds four inches and greater . nominal pipe size shall>be ultrasonically and liquid penetrant examined. Circumferential pipe welds less than four-inch nominal pipe size shall be liquid penetrant examined. . The examinations performed during the inspection interval shall include approximately 25 percent of the L 40-year sample of circumferential welds. The examina-tions shall be distributed during the inspection interval in accordance with Table A of Appendix A. 3 7.4.4.1 Reactor Coolant System Main Loop Piping e Circumferential Welds r-The Reactor Coolant System main loop piping includes'63 circumferential pipe welds -4 inches and greater nominal pipe size. . These welds shall be ultrasonically and liquid penetrant examined during the inspection interval (see Request for Relief p .ISI-7). . ~ ~ 7.4.4.2 Reactor Coolant System Piping Circumferential Welds !' Reactor Cooiant System piping includes 64 ~ circumferential pipe welds 4 inches and - greater nominal pipe size. These welds shall p_ be ultrasonically and liquid penetrant '-- examined during the inspection interval. There are'107 pipe welds less than 4-inch

l. -nominal pipe size that shall be liquid LL penetrant examined during the inspection, -

interval. 1 '7. 4.4.3 Chemical and Volume Control System Piping i] Circumferential Welds , The Chemical and Volume Control System piping (including seal water injection) includes 58 circumferential welds less than 4-inch Im U L.__. . ~- SONP' SI-114.1 Page 21 of 65 Rev 1 nominal pipe size that shall be liquid penetrant examined during the inspection interval. 7.4.4.4 Residual Heat Removal System Piping Circumferential Welds ~ The Residual Heat Removal System piping includes 51 circumferential welds 4 inches and greater nominal pipe size. These welds shall be ultrasonically and liquid penetrant examined during the inspection interval. There are no Class A pipe welds less than , 4-inch nominal pipe size in the RHR system. 7.4.4.5 Safety Iniection System Piping . Circumferential Welds w_ i'l . The Safety Injection System piping includes 105 circumferential pipe welds 4 inches and 7 greater nominal pipe size. These welds shall be ultrasonically and liquid penetrant examined during the inspection interval. _There are 24 pipe welds less than 4-inch {_. nominal pipe size that shall be liquid i penetrant examined during the inspection interval. 7.4.4.6 Upper Head Injection System Piping Circumferential Welds The Upper Head Injection System piping includes'122 circumferential pipe wlds 4 inches and greater nominal pipe size., These welds shall be ultrasonically and liquid penetrant examined during the inspection interval. 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 selected for examination'shall be examined. Branch pipe connection welds exceeding 2 inches nominal pipe - .{ size shall be ultrasonically and liquid penetrant

L. . examined. Branch pipe connection welds 2 inches i nominal pipe size and smaller shall be liquid ,

penetrant examined. 1 The examinations performed during the inspection  ; interval shall include approximately 25 percent of the i branch pipe connection welds. The examinations shall  ; be distributed during the inspection interval in  : accordance with Table A of Appendix A. I' l l j SQNP p , SI-114.1

- Pega 22 of 65 Rev.1

- 7.4.5.1 Reactor Coolant System Main Loop Branch Pipe Connection Welds , The Reactor Coolant System Main Loop piping includes I branch pipe connection weld exceeding 2 inches nominal pipe size. These welds shall be ultrasonically and liquid r penetrant examined during the inspection interval. There are 13 branch pipe connectica welde 2 inches nominal pipe size

u. and smaller that shall be liquid penetrant examined during the inspection interval.

7.4.5.2 Reactor Coolant System Branch Pipe Connection Welds The Reactor Coolant System piping includes p- 6 branch pipe connection welds exceeding 2 inches nominal pipe size. These welds shall be ultrasonically and liquid penetrant examined during the inspection interval. There are 7 branch pipe connection welds 2 inches nominal pipe size and smaller that shall be liquid penetrant examined during the inspection interval. ]- 7.4.5.3 Chemical and Volume Control System Branch - Pipe Connection Welds The Chemical and Volume Control System piping .g includes 4 branch pipe connection- welds 3 exceeding 2 inches nominal pipe size. These welds shall be ultrasonically and liquid penetrant examined during the inspection interval. There are no branch pipe connection. welds 2 inches nominal pipe size and smaller that shall be liquid penetrant 'T 3xamined _du'!ing the inspection interval. q . ,; - 7.4'.5.4 Residual Heat Removal System Branch Pipe , ,, Connection Welds - l} s L. The Residual Keat Removal System piping includes 3 branch pipe-connection welds = - exceeding 2 incher nominal pipe size. These. ![u welds shall be ultrasonical)y and liquid penetrant examined during the inspection

-- - interval. There are no brauch pipe

]. s . connection welds 2 inches nominal pipe size and smaller in the RHR System. .,L., , ~ a ~ %~ ~ ~ ._ x < . . ~s , ', f y -,k... , .- - . .SQNP ~ M-116.1 Pege 23 of 65 RV 1 7.4.5.5 Safety Injection System Branch Pipe Connection Welds The Safety Injection System piping includes 5 branch pipe connection welds exceeding 2 inches nomical pipe size. These welds shall be ultrasonically and liquid penetrant -- examined during the inspection interval. There are 4 branch pipe connection welds b 2 inches nominal pipe size and smaller that , shall be liquid penetrant examined during the inspection-interval. 7.4.5.6 Upper Head Injection System Branch Pipe rf Connection Welds The Upper Head Injection System piping 4 includes 2 branch pipe connection welds 2 L inches nominal pipe size and smaller that shall be liquid penetrant examined during the g inspection interval. . S 7.'4.6 Piping Socket Welds

• The entire length of.each socket weld selected for 4

examination shall be liquid penetrant examined. The examinations performed during the inspection s interval shall 1 :1ude approximately 25 percent of the socket welds. ine' examinations shall be distributed during the faspection interval in accordance with Table A of Appendix A. 7.4.6.1 Reactor Coolant System Piping Socket Welds - N-The Reactor Coolant System piping includes 261 socket welds. Of these welds, 66 welds will be examined during the inspection 1 j interval. s. . 7.4.6.2 Chemical and Volume Control System Piping , Socket Welds u The Chemical and Volume Control System piping ? includes 228 socket welds. Of these welds, I I~ . 57 welds will be examined during the inspec-tion interval. 1; . f 7.4.6.3- Residual Heat Removal System Piping Socket Welds k The Residual' Heat Removal System piping includes

iS socket welds. Of these welds, 5 welds will be examined during the inspection interval.

C y . b - M . SQNP P' . SI-114.1 i Paga 24. of 65 Rev 1 J_. - 7.4.6.4 .. Safety Injection System Piping Socket Welds i~ The Safety Injection Sy: ten piping includes 192 socket welds. Of these welds, 48 welds I will be examined during the inspection interval. H 7.4.6.5 Upper Head Injection System Piping Socket . Welds [ . The Upper Head Injection System piping includes 16 socket welds. Of thes'e welds, 4 welds will be examined during the inspection interval. ~ 7.4.7 Piping and Valve _ Integrally-Welded Support Members Integrally-welded rupport members include the support attachments of piping required to be examined by Examination _ Category B-J. Included are those supports which have attachment welds to the valve and piping i ', _ . ' pressure retaining boundary and those attacnmente whose support base n.aterial design thickness is $/8 inch i and greater. The entire length of each support attachment , weld selected for examination shall be surface examined.  ! The examinations performed during the inspection interval shall include 100 percent of the integrally-welded support members. The examinations shall be ~'g distributed during the inspection interval in-- , accordance with Table A of Appendix A. t 7.4.7.1 Reactor Coolant System Piping and Valve - Integrally-Welded Support Members ,~ The Reactor Coolant System piping includes .1 12 integrally-welded support members. All of m .these shall be examined during the inspection , e interval. - 1 '

• 7.4.7.2 Chemical and Volume Control System Pip pa '

p . and Valve Integrally-Welded Support tleasoers x The Chemical and Volume Control Systes piping ' ' includes 1 integrally-welded support .nember. i '/ .This support shall be examined during the h inspection interval. l l7.4.7.3 Residual Heat Removal System Pipigg and Valve k; .. Integrally-Welded Support Members; The Residual Heat Removil System piping includes .1. integrally-welded support member. This . support shall be examined during the inspection I interval. p .: p , l c j SONP rr . SI-114.1 [ Page 25 of 65 Rev 1 m 7.4.7.4 Safety Injection System Piping and Valve l Integrally-Welded Support Members

a. The Safety Injection System piping includes 9 integrally-welded support members. All of these shall be examined during the inspection

.~ interval.- 7.4.7.5 Upper Head Injection System Piping and Valve Integrally-Welded Supports );~ - The Upper Head Injection System piping includes 2 integrally-welded support member. ,a This support shall be examined during the t inspection interval. 7.4.8 Piping and Valve Component Suppor Q ( All piping and valve component supports of piping required to exsmined by Examination' Category B-J.shall !H be visually examined during the inspection interval in accordance with visual examination methods VT-3 and f- 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. The setting of snubbers and shock absorbers is performed in accordance with SI-162.; . The examinations shall be distributed during the, inspection interval in accordance with Table A of t Appendix A. U- 7.4'.8.1 Reactor Coolant System Piping and Valve y Component Supports L, ' The Reactor Coolant System piping includes [!: 127 component supports. All of these shall be examined during the inspection interval. !' 7.4.8.2 Chemical and Volume Control System Piping y and Valve Component Supports The Chemical and Volume Control System piping 7 16 includes 107 component supports. All of ~ these shall be examined' during the inspection 3- interval. h, 'I lf . . - C o SQNP , SI-114.1 , .Pags 26 of 65 Rev 1 - 7.4.8.3 Residual Heat Removal System Piping and Valve Component Supports ,. The Residual Heat Removal System piping includes 7 component supports. All of these shall.be examined during the inspection interval. 7.4.8.4 Safety Injection System Piping and Valve Component Supports

The Safety Injection System piping includes 56 component supports. All of these shall be examined during the inspection interval.

7.4.8.5 Upper Head Injection System and Valve Piping . Component Supports I r, '. The Upper Head Injection' System piping includes 86 component supports. All of these shall be examined during the inspection i interval. 7.5 Reactor Coolant Pumus (4) - RCP . u L 7.5.1 RCP Pressure Retaining Bolting Larger Than Two Inches In Diameter . The main flange on each pump includes 24 bolts at 4-1/2 inches in diameter and 30-1/2 inches in length, s All of the bolts from each pump shall be ultrason- - ically examined during the inspection interval. _The bolts may be examined in place under tension or when removed. All of the bolts from one pump shall be L examined during the first inspection period, and'all !' of the bolts from a different pump shall be examined during the second inspection period. All of the e bolting from the two remaining pumps shall be examined - (; during the third-inspection period. , , Removed, the bolts shall be ultrasonically and j' magnetic particle examined. This examination needs to be performed only once during the inspection interval and may be deferred to the third inspection period. L Provisions for this examination are included in . MI-2.2. c' When a main flange connection is disassembled, the .) threads in the base material and flange ligaments between threaded bolt holes shall be visually examined in accordance with visual examinition method VT-1 (see Section 11.1). Provisions for this examination are - included in MI-2.2. j' , The main flange bolts are fabricated of 4340 steel, (. heat treated to A-540, Gr B24. L_ SQNP x SI-114.1 Page 27 of 65 Rav 1 .c 7.5.2' RCP Pressure Retaining Bolting 2 Inches and Smaller in Diameter b ," Each RCP includes two sets of pressure retaining bolting 2 inches and smaller in diameter. The bolting sets are the number 1 and 2 seal assembly bolting. The number 1 and 2 seal assembly bolting from each pump shall be visually examined in accordance with J visual examination method VT-1-(see Section 11.1). g - The bolting may be examined in place under tension or when removed. It is preferable to perform the ' examinations 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. . N~ All of the bolting from one RCP shall be examined , during theLfirst inspection period, and all of the bolting from a different pump shall be examined during ,- the second inspection 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, and 12 ( socket head cap screws at 1 inch in diameter. 7.5.3 RCP Integrally-Welded Support Members There are no integrally-welded support components t' associated with the RCP. . 7.5.4' RCP Coniponent Supports . . Each RCP includes three integrally cast pump feet bolted to.the support system. All component supports [i. from each. pump shall,be visually examined during the

n.

inspection interval.In accordance with visual examination method VT-3 (see Section 11.1).' Support components extend froni the RCP to and including the fjy ' attachment to the supporting structure. c' i

P All of'the supports from one RCP shall be examined C during the_first inspection period, and all of the

~ supports from a different pump shall be examined p' .during.the second inspection period. All of the 'p" supports.from the remaining two pumps shall be examined during the. third inspection period. E 4 9: $b-g - l, , SQNF r- , SI-114.1

Page 28 of 65 Rev 1 7.5.5 RCP Casing Welds 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 of one RCP casing weld f shall be liquid penetrant examined during the { inspection interval. This examination may be deferred to the third inspection period. (See Request for 7- Relief ISI-8.) ]' 7.5.6 RCP Casing

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).

t. Disassembly of RCP's is performed in acccrdance with MI-2.2 and provides for there visual examinations.

If during'the inspection interval a pump from either unit 1 or unit 2 is not disassembled for maintenance, a pump from one unit shall be examined from the exterior by ultrasonic thickness measurements. (See Request for Relief ISI-1.) .7.5.7 RCP Flvwheel 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 conducted at approximately 10-year intervals -! along with a surface examination of exposed surfaces. (See Regulatory Guide 1.14.) This examination is done in accordance with Sequoyah technical specifications j and satisfies Surveillance Requirement 4.4.10. . L. The flywheel consists of 2 plates, approximately 5 inches and 8 inches thick, bolted together. Each j plate is fabricated from vacuum degassed A-533, GR. B, Class 1, steel. 7.6 Valves Systems including class A valves subject to examinatien are j7 ~ identified in Section 7.4.

L 7.6.1 Valve Pressure Retaining Bolting. Larger Than 2 Inches in Diameter

'~ ' There are no valves with pressure retaining bolting larger than 2 inches in diameter. L-B s - ,. SQNP SI-114.1 Pzge 4'9 of 65 Rev 1 p 7.6.2 Valve Pressure Retaining Bolting 2 Inches and Smaller l 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 visually 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 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 performed in accordance with MI-6.15 and provides for examination of bolting. 7.6.2.1 Reactor Coolant System Valve Bolting -- The Reactor Coolant System includes 14 valves i with bolted bonnet connections. 7.6.2.2 Chemical and Volume Control System Valve - Bolting l The Chemical and Volume Control System includes 6 valves with bolted bonnet. ~ connections. 7.6.2.3 Residual Eaat Pemoval System Valve Bolting ' ^ The Residu.d Heat Removal System includes 2 valves with bolted bonnet conenctions.. 7.6.2.4 Safety Injection System Valve Bolting ~ The Safety Injection System includes 23 i valves with bolted bonnet connections. (. 7.6.2.5 Upper Head Injection System Valve' Bolting L Ahe Upper Head Injection System includes .. 8 valves with bolted bonnet connections. w }- 7.6.3 Valve Integrally-Welded Support Members j Examination of valve integrally-welded support members (, is included in Section 7.4.7. . 7.6.4 Valve Component Supports "~ Examination of valve component supports is included in Section 7.4.8. ) s. s [L ,,. -o--, . , - -- SQNP j . SI-114.1 Page 30 of 65 Revil 7.6.5 Valve Body Welds -t !! There are no valves with body welds. T~ 7.6.6 Valve Bodies The internal pressure boundary surfaces of valve bodies r- exceeding 4-inch nominal pipe size shall be visually

i. examined in accordance with visual examination method VT-1 (see Section 11.1). Examinations are limited to

, _ _ one valve within each group of valves that are of the same constructional design (i.e., globe, gate, or

d. t - check valve), manufacturing method, and that are performing similar functions in the system.

P [ When it becomes necessary to disassemble any valve, subject to internal surface visual examination, for e normal maintenance purposes, the interior surface of the valve body will be visually examined and the results recorded. Disassembly of valves solely for _ visual examination will not be performed (See ASME Section XI, Table IWB-2500, Examination Category B-M-2; also see Request for Relief ISI-2). A tabulation cf valves by groupings is presented in '~

Table D of Appendix A. Disassembly of valves is performed in accordance with MI-11.4 and provides for

r- examination of valve internal pressure boundary surfaces.

.t 7.7 Exempted Components Components 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 par't 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 System Pressure Tests All Class A pressure-retaining components shall be subjected to a system leakage test in accordance with IWB-5221 of ASME g- Section XI each refueling outage. This is performed in t accordance with SI-146. In addition, all Class A pressure- '~ retaining components shall undergo a system hydrostatic ,_ pressure test in accordance with IWB-5222 of ASME Section XI

{ at the end of the inspection interval. This is performed L L- in accordance with SI-250. The components shall be visually examined during the pressure tests in accordance with visual examination method VT-2 (see Section 11.1).

+j-- SQNP I~ , SI-114.1 Page 31 of 65 Rev 1 , -8.0 . COMPONENTS SUBJECT TO EXAMINATION - TVA SAFETY CLASS B i t/ The Class B-(ASME Class'2) components to be examined for the ISI 2 are outlined in.the following paragraphs. Extent of examination piping welds will bt .a accordance with paragraph IWC-2411 and Table IWC-2520 of the 1974 Edition, Summer 1975 Addenda, ASME Section XI. The ISI Frograms Section shall select areas to be examined or the H NDE Section 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 8.13 of this program. \-. - - Where examinations specify a percentage of the total length of weld to. be examined, the areas (s) examined shall'be documented in the examination I' report. Where a percentage of weld length is not referenced, the entire weld length shall be examined. (See Request for Relief ISI-3.) r, Table B in Appendix A supplies additional information such as reference drawing numbers and ASME Section XI Table-IWC-2500-1 examination categories. - ,a 8.1 Steam Generators (4) 8.1.1 Steam' Generator Secondary Side Circumferential Shell ,( Welds ~There are three circumferential shell welds at gross a 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 Appendix 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. s SGW-DI,- D2, D3, and D4; see Request for Relief ISI-4). h# The weld selected for examination shall be ultrason- -ically examined on a best-effort basis. y The vessel'shell sections are fabricated of SA-533, . L -. GR. A, Class 1, steel plate. y :8.1.2 . Steam Generator Secondary Side Circumferential Head J. Welds , N Each steam generator includes a circumferential head-to-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 , weld selected for examination may be from the generator k selected for examination by Section 8.1.1. r -. .-~ O. g , SQNP SI-114.1 -- T Pega 33 of 65 Rev 1 ,-. The vessel head section is fabricated of SA-533, GR. A, Class 1, steel plate. 8.1.3 Steam Generator Secondary Side Tubesheet-To-Shell Weld Each steam generator includes a tubesheet-to-shell weld. The entire length of one tubesheet-to-shell r; . weld shall be ultrasonically examined during the i inspection interval in accordance with Table B of ' Appendix A. The weld selected for examination may be from the generator selected for examination by section [ 8.1.1 e 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. 4 ,' The nozzles are fabricated of SA-508, Class 2, steel. 8.1.5 Steam Generator Secondary Side Integrally-Welded Support Attachments [ ,. There are no integrally welded vessel support i attachments. t - 8.1.6 Steam Generator Secondary Side Component Supports L There are no component supports (including mechanical and hydraulic supports) which are in contact with the m vessel. ., ~ 8.1.7 Steam Generator Secondary Side Pressure Betaining jp. Bolting Exceeding 2 Inches in Diameter There is no steam generator secondary side bolting exceeding 2 inches in diameter. _. 8.2 Residual Heat Removal Heat Exchangers (2) - RHRHX

p 8.2.1 RHRHX Circumferential Welds 4~-

8.2.1.1 RHRHX Shell Circumferential Weld There.is one circumferential shell weld located at a gross structural discontinuity ~~ m_ ~-- I DOCUMENT CONTROL AUDIT CERTIFICATION E# -_ _ _ , AUDITED / REVIEWED BY: _ _ ////// <. DATE: _ i ~ A. , -PLANT: / 1 INSTRUCTION NUMBER: _ _fL_ //M, 'Date- ~ O. rient Noctriea. ~ ~ .. :.j O niesins vasec.) .g/J - - 1' wrons Revision;Numder O - 3.a. xerox. Copy-O ' t O No rese Number - , 0 Instruction Number ni . ins O Instruction Numder wrons V COMMENTS: _ ~ a.f w/ ais/ _ TVA 7004C (DNP-8-82) .....:.... ,-..~. L...~.~ z. . : l l l -l l i I -. . .. .- .-. . - -. - .- -._. ~ , SQNP. g SI-114-1 Page 34 of 65 Rev 1 on each RHRHX. The entire length of this shell weld from one heat exchanger shall be ultrasonically examined during each inspec-tion interval in accordance with Table B of Appendix A. The RHRHX shell section is fabricated from

~ SS, SA-182-F304.

8.2.1.2 RHRHX Head Circumferential Weld 'There is one circumferential head-to-shell weld per RHRHX. 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 weld selected for examination may be from th,e  ; heat exchanger selected for examination by section 8.~2.1.1. The' channel cylinder section (shell) and . channel he.ad are one inch thick fabricated from SS, SA-240, TP-304. 8.2.1.3 RFRHX Tubesheet to Shell Weld l There are no RHRHX tubesheet-to-shell welds. 8.2.2 RERHX Nozzle-to-Vessel Welds . The channel cylinder section of each RHRHX includes one 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 q' ' RHRHX will be ultrasonically and liquid penetran't examined during the inspection interval. 3_ The nozzles are 2.5 inches thick, fabricated from SS, { SA-240, TP-304. ,. 8.2.3, RHRHX Integrallv-Welded Support Attachments .[ , . There are two integrally-welded support attachments (1-inch)'on each RHRHX whose base material exceeds l/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. 'l 4 -The support pad is fabricated from SS, SA-240, TP-304. f p . w i J T" SQNP SI-114.1 Psga 35 of 61 Rrv 1 ,, 8.2.4 RHRHX Comconent Supports i -' There are two component supports on each RHREX 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 struci.ure. The examinations shall be distributed ~during the inspection interval in accordance with Table B of , Appendix A. There are no mechanical (snubbers) and/or hydraulic (shock absorbers) supports which are in contact with the vessel. 8.2.5 RHRHX Pressure Retaining Bc ting Exceeding 2 Inches in Diameter , There is no RHRHX bolting exceeding 2 inches in diameter. 8.3 Regenerative Heat Exchanger (One) - REX 8.3.1 RHX Circumferential Welds 8.3.1.1 RHX Shell Circumferential Welds 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. channel head welds and there are

L .six tubesheet-to-shell welds.

The channel cylinder sections are fabricated i to SS, SA-351, CFS. The channel flanges are " fabricated to SS, SA-182, F304. j 8.3.1.2 RHX Head Circumferential Welds

c. .

There are six. channel cylinder to channel head welds. These welds shall be ultra- .[' sonically examined during the inspection interval in accordance with Table B of Appendix A. The examin5tions shall cover 100 percent of the weld length. The channel heads are fabricated to SS, j SA-240, 304L. a_ -- SQNT j- - SI-114.1 Pege 36 of 65 ,. 8.3.1.3 RHX-Tu$UhbettoShellWeld There are six RHX tubesheet-to-shell welds. These welds shall be ultrasonically examined during the inspection interval in accordance with Table B of Appendix A. The examina-tions shall cover 100 percent of the weld length. The tubesheet is fabricated to SA-182, F304 . SS. - 8.3.2 RHX Nozzle-to-Vessel Welds There are not any nozzles greater than four inch diameter. The nozzles are fabricated to SA-479, 304 ' SCH 160 material. l-8.3.3 RHX Integra11v-Welded Supports ~ ._ There are not any integrally-welded support attach-t ments on the heat exchanger. There are_three stops 5/8" x 5/8" x 1-3/4" welded to the heat exchanger. These stops (lugs) shall be surface examined during " the inspection interval. 8.3.4 RHX Component Supports There are two component supports on.the regenerative heat exche 'er. Both supports shall be visually , examined (VT-3) during the inspection interval in

accordance with Table B of Appendix A.

8.3.5 RHX Pressure Retaining Bolting Exceeding Two Inches [ in Diameter ~ There is not any pressure retaining bolting , associated with the heat exchanger. a ' 8.4 Letdown Heat Exchanger (One) - LHX

) 8.4.1 LHX Circumferential Welds p,.

8.4.1.1 LHX Shell Circumfereutial Welds There are not any shell circumferential welds associated with the LHX. ! 8.4.1.2 LHX Head Circumferential Welds L: There are two circumferential welds located at structural discontinuities on the head a side .of the LHX. ' These welds shall be ultra-sonically examined during the inspe: tion 1 L. i.. , SQNP SI-114.1 Paga 37 of 65 Rev 1

r. interval in accordance with Table B of Appendix A. The examination shall cover 100 percent of the weld length.

The LE head section is fabricated to SS, SA-240, TP 304. [' 8.4.1.3 LHX Tubesheet to Shell Weld c There are not any LHX tubesheet to shell welda. ^ 8.4.2 LHX Nozzle to Vessel Weld [ 'There are not any nozzles greater than four-inch i diameter. The nozzles are fabricated to SS, SA-312, 304. r [ 8.4.3 LHX Integrally-Welded Supports There are two integrally-welded support attachments on the LHX whose base material design thickness is 1/2 inch and therefore requires no examination in ,_ accordance with Table IWC-2500-1 of ASME Section XI. 8.4.4 LHX Component Supports ~ .There are two component supports on the letdown heat .. exchanger. Both supports shall be visually examined (VT-3) during the inspections interval in accordance c- with Table B of Appendix A. ~ ' 8.4.5 LHX Pressure Retaining Bolting Exceeding Two Inches

r. in Diameter .

L The channel flauge bolting associated with the LHX is less than two inches in diameter.

r- ._

1 The' flange includes'24 studs with nuts at 7/8 inch-diameter. The studs and nuts are fabricated to , , SA-193, GRB7 and SA-194, GR2H, respectively. l 8.5. Excess Letdown Heat Exchanger (One) - ELHX 8.5.1 ELHX Circumferential Welds H ~ 8.5.1.1 ELHX Shell Circumferential Welds l L There are not any shell circumferential welds associated with t!;e ELHX. , L_ g-  : E i g7 j b m ,. SQNP

} - SI-114,1 Pcga 38 of 65 Rev 1 7i '

8.5.1.2 ELHX Head Circumferential Welds There are one circumferencial weld, the channel flange to channel head, located at a structural discentinuity on the head side of the ELHX. This weld shall be ultrasonically examined during the inspection interval in accordance with Table B of Appendix A. The examination shall cover 100 percent of the weld length. F The channel flange and the channel head are fabricated to SA182, F316 and SA-240, TP316, respectively. 8.5.1.3 ELHX Tubesheet to Shell Weld . i There are not any ELHX tubesheet to shell welds. ~ 8.5.2 EIHX Nozzle to Vessel Weld There are not any nozzles greater than four-inch diameter. The nozzles are fabricated to SA-312, 316. , "i 8.5.3 ELHX Integrally-Welded Supports 4 There are no integrally-welded supports to the head side of the ELHX. - 8.5.4 ELHX Component' Supports ~ Taere are no component supports to-the head side of the ELHX. 8.5.5 ELHX Pressure Retaining Bolting Exceeding Two Inches in Diameter i [- The channel flange bolting associated with the ELHX is less than two inches in diameter. The flange includes 12 studs with nuts at 1-5/8 inch diameter. The studs and nuts are fabricated to , SA-193, GRB7 and SA-194, GR2H, respectively. .l 8.6 Boron Injection Tank (One) - BIT 8.6.1 BIT Ciretaferential Welds (Shell and Headl ( There are two circumferential head-to-shell welds located at structural discontinuities on the BIT. These welds shall be ultrasonically examined during the inspection interval in accordance with Table 2 of ~ e Appendix A. The examination shall cover 100 percent of the weld length. . { 'h c- . SQNP , SI-114.1 Page 39 of 65 Rev 1 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. 8.6.2 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 in.) is greater than 1/2 inch. These razzle to i vessel welds shall be ultrasonically and surface _ examined during the inspection interval in accordance with Table B of Appendix A. The examination shall . cover 100 percent of the weld length. The nozzles are fabricated to SA-350, LF2. 8.6.3 BIT Integrallv-Welded Supports There are four integrally-welded support attachment pads welded to the shell, whose base material desi.gn tnickness is 3/8 inch and therefore requires no examinaticu in accordance with Table IWC-2500-1 of ASMI Section XI. 8.6.4 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. 8.6.5 BIT Pressure Retaining Bolting Exceeding Two Inches in Diameter There are 16 canway cover studs at 2-1/2 inches 'in diameter. All 16 studs shall be ultrasonically examined during the inspection interval in accordance f with Table B of Appendix A. The studs may be

L examined in place under tension or when they are removed. It.is preferrable to perform the examinations when the studs are removed if possible.

.j' . " The studs and nuts are fabricated to SA-193, GRB7 and , SA-194, GR2H, respectively. . 2-- 8.7 UHI Water Accumulator (Or :) WA 8.7.1 WA Head and Shell Circumferential Wells There are two circumferential welds located at struc-tural discontinuities. These two head to shell welds shall be ultrasonically examined during the inspection interval in accordance with Table B of Appendix A. The examinations shall cover 100 percent of the weld length, i % .6 SQNP ip . SI-114.1 Page 40 of 65 __ Rev 1 The head and shall are fabricated to SA-516-71-70. 8.7.2 WA Noz le to Vessel Weld ~ -The WA , s three 12-inch nczzles whose nominal thick-ness:is over 1/2 inch, one on the top head and two on , the bottom head. These nozzles to vessel welds shall J'~ be surface and ultrasonically examined during the inspection interval in accordance with Table B of Appendix A. , The nozzles are fabricated to SA-350, I.F2.

- 8.7.3 WA Integrally-Welded Scoports M

The water accoulator has an integrally welded support skirt attached to the bottom head. The 'f ~ support skirt-to-vessel weld hshall be surface examined during the inspection interval in accordance . with Table B of Appendix A. The examination shall - cover 100 percent of the weld length. The integrally. welded portion of the suppor skirt is , fabricated to SA-516, GR70. J 8.7.4 WA Component Supports The water accumulator has a component support skirt attached to the bottom head. The component support shall be visually examined (VT-3) during the .- inspection interval in ac'cordance with Table B of .', Appendix A. - 8.7.5 WA Pressure Retaining Bolting Exceeding Two Inches in Diameter There are 16 manway cover studs at 2-1/2 inches in t diameter. All-16 studs shall be ultrasonically 1 examined during the inspection interval in accordance . with Table B of Appendix A. 'The studs may be 4 examined in place under tension or when they are . removed. -It is preferrable to perform the examina- '* tions when the studs are removed if possible. j, The studs and nuts are fabricated to SA-193, GRB7 , L. , and SA-194, CL2H, respectively. 8.8 UHI Surge Tank (One) - ST ( 8.8.1 ST Head and,Shell Circumferential Welds There are two circumferential welds located at structural discontinuities. These two head to shell , welds shall be ultrasonically examined during the 1 .Li

i- _

SQNP v2 ' - SI-114.1 P:gs 41 of 65 Rsv 1 inspection interval in accordance with Table B of Appendix A. The examinations shall cover 100 percent of'the weld length. The head and shell are fabricated to SA-240, TP304. 8.8.2 ST Nozzle to Vessel Weld There are not any nozzles greater than four-inch ,4 diameter-associated with the surge tank. , ' The nozzles are fabricated to SA-479, TP304, sud SA-182,'F304. 8.8.3 ST Integrally-Welded Supports - There are two saddle type supports pads integrally welded to the shell. These pad to vessel welds shall be surface examined during the inspection interval in accordance with Table B of Appendix A. The examina-L- tion shall cover 100 percent of the weld length. The support pads are fabricated to SA-240, TP304. 8.8.4 ST Component Supports

• There.are two component supports associated with the surge tank. Both these component supports shall be .

visually examined (VT-3) during the inspections interval in accordance with Table B of Appendix A. 8.8.5 ST Pressure Retaining Bolting Exceeding Two Inches ~ in Diameter There are not any pressure retaining bolting' exceeding two inches in diameter. There are 24 1-inch studs and nuts associated with the surge tanks two handholes. . The studs and nuts'are fabricated to SA-193, GRB7 and SA-194, GR2H. 8.9 Piping Material specifications for cach piping system are included '- in weld map isometrics in Appendix A. The following Class B piping systems are subject to examination: Residual Heat Removal Safety Injection , I Main Steam Feedwater Containment Spray Upper Head Injection 1 SQSIP .- SI-114.1 Pags 42 of 65 Rv1 .. 8.9.1 Piping and Valve Integrally-Welded Support Members Integrally-welded support members include the support attachments of piping required to be examined by Examination Category C-F. Included are those supports which have attachment welds to the valve and piping pressure retaining boundary, and those attachments whose support base material design thickness exceeds 3/4 inch. The entire length of each support attachment weld selected for examination shall be surface examined. The exainations performed during the inspection interval shall include 100 percent of the integrally-welded support members. The examinations shall be i distributed during the inspection interval in accordance with Table B of Appendix A. ~ 8.9.1.1 Residual Heat Removal System Piping and Valve Integrally-Welded Support Members The Residual Heat Removal System piping includes 21 integrally-welded support. members. All of these shall be examined during the inspection interval. 8.9.1.2 Safety Injection System Piping and Valve Integrally-Welded Support Members The Safety' Injection System piping includes 10 integrally-welded support members. All of these shall be examined.during the inspection interval. 8.9.1.3 Main Steam System Piping and Valve Integrally Welded Support Members / The Main Steam System piping includes 11 integrally-welded support members. All of these shall be examined during the inspection interval. , 8.9.1.4 Feedwater System Piping and Valve Integrally- '- Welded Support Members [ The Feedwater System piping . includes 14-integrally-welded support members. All of 4 these shall be examined during the inspectica interval. \' 8.9.1.5 Containment Spray System Piping and Valve Integrally-Welded Support Members , The Containment Spray System piping does not include any integrally-welded support members.

1. .

o.- , - - . . . - - -- - - . , , -, aqu SI-114.1 - Paga 43 of 65 Rav 1 8.9.1.6 Upper Head Injection System Piping and Valve . Integrally-Welded Supports The upper head injection system piping includes 4 integrally-welded supports. All of these shall be examined during the inspection interval. 8.9.2 Piping and Valve Component Supports . All piping and valve component supports shall be visually examined during the inspection interval in accordance with visual examinatica methods VT-3 and VI-4 (see Section 11.1). This examination includes integrally-welded and nonintegrally-welded compocent e supports. Component supports extend from the piping and valves to and including the attachment to the r supporting structure. j' The setting of snubbers, shock absorbers and spring-type hangers shall be verified. The setting of l snubbers and shock absorbers is performed in accordance with SI-162. ^ The examinations shall be distributed during the inspection interval in accordance.with Table B of Appendix A. 8.9.2.1 Residual Heat Removal System Piping and Valve Component Supports The Residual Heat Removal System piping includes 44 component supports. All of these shall be examined during the inspection

• ft interval.

8.9.2.2 Safety Injection System Piping and Valve Component Supports ,{ L The Safety Injection System piping includes 67 component supports. All of these shall be examined during the inspection interval. . 8.9.2.3 Main Steam System Piping and Valve Component Supports The Main Steam System piping includes 54 component supports. All of these shall be p .[ examined during the inspection interval. m SQNP

7. - , SI-114.1' (r;

Page 44 of~65 1- _Rav 1 > 8.9.2.4 Feedwater System Piping and Valve Component . Supports y The'Feedwater System piping includes 22 jU - component supports. All of these shall be i examined during the inspection interval. 8.9.5.5 Contninment Spray System' Piping and Valve Component Supports

r. The Containment Spray System piping includes

j ': 5 component supports. All of these shall be

,. examined during.the inspection interval. 8.9.2.6 Upper Head Injection System Piping and Valve Component Supports ,_. The' upper head injection system piping d_ includes 7 component supports. All of these shall be examined during the inspection ! interval. 8.9.3 Pressure-Retaining Bolting There is no Class B Pressure-Retaining Bolting larger .than two inches in diameter. ~ 8.9.4 Circumferential and Longitudinal. Pipe Welds The' entire length of each weld selected for exami-nation-shall be_ ultrasonically and/or surfaced examined. Selection of welds for examination is based ~ on Table'IWC-2520, Paragraph IWC-1220, and Paragraph .r IWC-2411 (Summer 1975 Addenda).. All of the welds selected shall be examined during the inspection  : C

• interval ~and distributed in accordance with Table B of Appendix A.

~ _ ,

dj - Circumferential:and fongitudinal piping welds in piping.with a nominal wall thickness of 1/2 inch or H 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. - m iL: The areas subject to examination include circumfer-

ential pipe welds 9 at structural discontinuities 4 ip within 3 pipe diameters of.the centerline of rigid

',_. pipe anchors, anchors at the penetration of primary

containment or at rigidly anchored components, and f longitudinal weld joints in pipe' fittings.

T 3 .. ' . . 4 _gg_ , l Tq , -",s - e , , - - , , - - n-., ..-,-,,-~,.,,an. - .. . , - - - , - , , , , ,, , ,,-n., ..--.e n -c, - . ~ .. J SI-114.1 hL ~. ' Page 45 of 65 Rav 1 f- - 8.9.4.1 Residual Heat Removal Piping The Residual Heat Removal Piping System

u. includes 36 ' Class 3 circumferential and 9 Class B longitudinal' piping welds with a nominal wall thickness greater than 1/2 inch subject to examination. There are 105 F circumferential and 40 longitudinal pipe welds with a nominal wall thickness of 1/2 inch or .less' subject to examination. Nine H circumferential and three longitudinal piping i-welds with a nominal wall thickness greater than 1/2 inch shall be ultrasonically and

_ surfaced examined each inspection interval. There are 27 circumferential and 10 longitudinal pipe welds with a ncminal wall . thickness of 1/2 inch or less that'shall be tr - . surface examined each inspection interval. I See_ Table B of Appendix A for scheduled inspections. + J- 8.9.4.2 Safety Injection Piping The Safety Injection Piping System includes i! 31 Class B circumferential and 1 Class B lengitudinal piping welds with a nominal wall thickness greater than 1/2 inch subject to examination. There are 82 circumferential < and 14 longitudinal pipe welds with'a nominal l wall thickness of 1/2 inch or less subject _to r examination. Eight circumferential piping welds with a nominal wall thickness greater 'than 1/2 inch shall be ultrasonically and ' surface examined each' inspection interval.

l. There are 21 circumferential and 4 longi-m tudinal pipe welds with a nominal wall thickness of 1/2 inch or less that shall be n - surfaced examined each inspection interval.

h' See Table B of Appendix A for for schedule inspections. v h 8.9.4.3 Main Steam Piping - The Main Steam Piping System includes 46 ' Class B circumferential and 11 Class B L longitudinal piping welds with a nominal wall thickness greater than 1/2 inch subject to p examination. The piping size is 32" 0.D. 1 x 1.088 wall unless noted. Twelve circumferential and three longitudinal piping welds shall be ultrasonically and surfaced u I h h', . . , , .~ - - - - - - SQNP SI-Il4.1 - 1 i ~ Page 46 of 65 Rav 1 ~ examined each inspection interval. TVA  ! intends to terminate main steam Class B on each main steam loop after the check valve (1-623, 1-624, 1-625, 1-626) following the MSIV (1-4, 1-11, 1-22, 1-29) for ISI purposes. See Table B of Appendix A for ' scheduled inspections. 1 ', 8.9.4.4 Feedwater Piping The Feedweter' Piping System includes 34 f- -Class B circumferen*.ial welds and 1-long-tudinal weld with a nominal wall thickness - ' greater than 1/2 inch subject to examination. 'The one longitudinal weld is on Loop 4 at the -reducing elbow (18" x 16"). There are - 9 circumferential and the one longitudinal '~'. -piping welds that shall be ultrasonically and i surfaced examined each inspection interval. -See Table B of Appendix A for scheduled r inspections. i' 8.9.4.5 -Containment Spray Piping [ The Containment Spray Piping System includes 10 circumferential and 3 longitudinal pipe welds with a nominal wall thickness of 1/2 'T inch or less, subject to examination. Three i circumferential and 1 longitudinal pipe welds shall be surface examined each inspection ' interval. 'The nominal wall thickness for the mI containment spray system is 0.406 inches. See Table B of Appendix A for scheduled. , inspections. L 8.9.5 Branch Piping Connection Welds ~ - There are no Class B Branch Pipe Connection Welds. 8.10 Pumps j ' p[ 8.10.1 Residual Heat Removal Pumps (2) - RHRP _ r ~ 8.10.1.1 RHRP Integrally-Welded Supports '. ) o-C _There are no integrally-' welded supports -associated with the RHRP. i .l - e Si O )_ s. -46 . - Il - -. .. . _ _ _ _ - ._ _ . _ _ __ ___ _ ._ _ _ ~ . SQNP-r . - SI-114.1 E Page 47 of 65 Rsv 1 . 8.10.1.2 RHRP Support Components 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). Support components extend. - from~the RHRP to and including the attachment

c. to the supporting structure.

). Two of the supports from one RHRP shall be examined during the first inspection period. 7 The final (3rd) support from the same RHRP shall be examined during the second inspection period. All of the supports from e the remaining RHRP shall be examined during h the third inspection period. . -8.10.1.3 RHRP Supports - Mechanical and Hydraulic ) 1- There are no mechanical and hydraulic supports associated with the RHRP. cl' 8.10.1.4 RHRP Pressure Retaining Bolting lhe, stuffing box extension to pump casing connection bolting is not greater than two - inches in diameter. j The connection includes 24 studs at 1-1/4' t - inches in diameter with nuts and washers. .The studs'are fabricated to SA-193, GR. B7, ' and the nuts to SA-194, GR. 2H. (I t 8.10'.1.5 RHRP Casing Welds rz

d. The RHRP does not include any casing welds.

'* ~ .The casing is a one piece forging fabricated . J to SA-182 F304. { ' 8.11' Valves - f 4It 8.11.'1'. Valve Integrally-Welded Supports "' Wh. . Examination of' valve integrally-welded support members j9 is included in Section 8.9.1. i 'D .y.- + - - -.m._ - n _gy__ , y; .- I - , , ,-- ---c , . - _ _ . _ - , . . _ . - . _ - _ _ _ . , . . _ - . . - _ _ _ _ . . . _ _ _ . - - _ . ~ _ . . . . . .- __ i SQNP SI-114.1 L Paga 48 of 65 Rev l' [ 8.11.2 Valve Con;ponent Supports ,1 Examination of valve component supports is included in Section 8.9.2. . l 8.11.3 Valve Pressure-P.ataining Bolting There is no Class B pressure-retaining bolting greater than 2 inches in diameter. ~ 8.11.4 Valve Body Welds There cre no Class B valves with body velds. 8.12 Pressure Retaining Components All Class B pressure retaining components in systems or i~' portions of systems which are not required to operate during .i normal reactor operation but for which periodic system or component functional tests are required (excluding open-ended r portions of systems), shall be subjected to a system i functional leakage test with each inspection period in accordance with IWC-5221 of ASME Section XI. This is ,. performed in accordance with SI's to be developed later.

i. Those Class B pressure retaining components in systems or

~ portions of 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 i, interval in accordance with IWC-5222 of ASME Section XI.

This is performed in accordance with SI-265.3. The components shall be visually examined during the pressure ~ ~ tests in accordance with visual examination method VT-2 (see Section 11.1). 'c i 8.13 Exempted Components .i 8.13.1 Exempted Components (Except Piping Welds). .c L Components exempted from examination inc1ude: (1) Components in systems where both the design

pressure and temperature are equal to or less than 275 psig and 200 F, respectively; (2) components in systems or portions of systems, other than emergency

,y co e cooling systems, which do not function during 'l.- normal reactor operation; (3) components which L perform an emergency core cooling function for hich the control of the chemistry of the contained fluid i is verified by periodic sampling and test, and L. (4) component connections, piping, and associated . valves, and vessels (and their supports), that are 4-inch nominal pipe size and smaller. D amiP SQNP. SI-114.1 Pega 49 of 65 r Rev l' !( ..S.13.2 Exempted Components (Piping k' elds Only) i-Piping exempted from examination include: . . . (a) Piping systems where both the design pressure L and temperature are equal to or less than 275 psig and-200*F, respectively; (b) piping systems or portions of systems other than emergency core f cooling systems which do not function during i normal' reactor operation; (c) piping that is 4-inch nominal pipe size and smaller; p (d) components which perform an emergency core , cooling function provided the control of the chemistry of the contained fluid is verified by ' periodic sa'mpling and test. r' - y: e em 0 I p ,

c

'r s 4 W L,i . _ . ..y. . -- . - - - ~ .~ SQNPs -c V, " [' . SI-114.1- { s , -% Page'50 of 65 V Rav 1 j7 j)0 COMP 0ENTS SUBJECT TO EXAMINATION ~ - TVA SAFETY CI. ASS C AND D i ' .. Class C and:D (ASE C1' ass 3) components shall be subjected to the ~ following examinations,and tests. ~ 9.1 Syster Pressure Tests . - 9.1.1 System Inservice Tests ._ . Pressure retaining components within the boundary of , systems'or portions of systems required to operate in support of normal plant safety functions of shutting 9 down and maintaining the reactor in the cold shutdown f~ ~ condition, and pressure retaining piping, pumps, and i valves.within the boundary or systems or portions of systems required to' operate in support of residual ~ y m-heat = removal from spent fuel storage pool, shall be visually examined in accordance with visual exami-1 nation method VT-2 (see section 11.1) while the applicable systems,are in-service under operating pressure in accordance with IWD-5221 of ASME (- Section XI. ~ These examinations and tests shall be performed during each inspection period. f 9.1.2 System Functional Tests b 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 r- removal from the reactor vessel shall be visually I examined in accordance with visual examination method ? VT-2.(see Section 11.1) during a system functional . .. test conducted _to verify operability in systems in accordance'with-IWD-5222 of ASME Seetion XI. ]x . 9.1.3 System Hydrostatic Tests y t i_, Pressure retaining components identifie'd in Sections 9.1.1 and 9.1.2 shall be visually examined in -r accordance' with visual examination method VT-2 (see d ~ Section 11.1) during a system hydrostatic test conducted in accordance with IWD-5223 of ASME Section XI. These examinations and tests shall be {,e performed over each inspection interval. e n, . C ~ ,.- ;t i . g - -SQNP I .:, SI-114.1 j' Page 51 of 65 RW l'. I M '9.2 ~ Component Supports and Restraints _ p l C Component supports and restraints within the boundaries of _ the systems identified in Sections 9.1.1 and 9.1.2 for p~ components exceeding 4-inch nominal pipe size shall be

~

visually examined each. inspection period in accordance with virual examination method VT-3 (see Section 11.1). ff . 9.3. Snubbers and Hangers W . Mechanical and hydraulic snubbers, spring loaded, and F . constant weight support hangers'for components exceeding. , 4-inch' nominal pipe size shall be visually examined each , inspection period in accordance with visual examination-j. method VT-4 (see Section 11.1.) L Id.0 AUTHORIZED INSPECTOR r '

TVA shall. employ ~an' Authorized Inspector (s) in accordance with ASME Section XI for inservice examinations, repairs, and replacements of TVA' Class A, B and C components at Sequoyah

b. Nuclear Plant. The inspector shall verify,-assure, or witness L that code requirements have been met. He'shall have the  !

prerogative and authorization to require requalification of any [' operator or procedure when he has reason to believe the !. requirements are not being met. -TVA shall provide access for the AI in accordance with IWA-2140 of ASME Section XI. TVA's interface with the Authorized Inspector for inservice ,x.- . . inspection, repairs, and replacements is defined in 0QAM Part II,

p Section 2.3 and Part II, Section.5.1

r.

M n- - 11.0 ' EXAMINATION METHODS [~ 1 11.1 Visual Examination - Visual examinations that-require clean surfaces or ~ ' decontamination for valid. interpretation of results shall j.. be preceded by appropriate cleaning processes. , 11.1.1 Visual Examination'VT-1 "- The VT-1 visual examination shall be' conducted to determine the condition of the part, component, or 'p surface examineC, including such conditions as c cracks, wear, corrosien, erosion, or physical ~ damage on the surfaces of the part or components. <l 11.1.1.1 Direct Visual Examination VT-1 g Direct VT-1 visual examination may be 4 A conducted when access is sufficient to place the eye 24 inches (610-mm) of the

li a.

~ lb i p^ i; . '--- ~.. . ._ _ _ . . . _ . , , . . , _ _ _ _ . _ - _ , _ _ _ _ . . _ . , _ _ . . _ _ . _ . . _ . _ _ _ . ~ f" SQNP h ' SI-114.1 l t Page 52 of 65 Rev { surface to oe examined and at an angle t ' + not less than 30 degrees to the surface. Hirrors may be used to improve the angle r- of visica. Lighting, natural.or - P artificial shall be sufficient to resolve

a 1/32 inch (0.8 mm) black line on an 18%

g neutral gray card. E 11.1.1.2 Remote Visual Examination VT-1 Re. mote VT-1 visual dxamination may be fp- rubstituted for direct examination. ' Remote examination may use aids-such as telescopes, borescopes, fiber optics,

• I 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 rh . { .The VT-2 visual examination shall be conducted to locate evidence of leakage frem pressure retaining , p components, or abnormal leakage from components a uith or without leakage collection systems as required during the conduct of system pressure or functional test. ) L The visual examination,- VT-2; may be conducted 7 ^ without the removal of insulation by examining the y accessible and exposed surfaces and joints of the . W ,, insulation. Essentially vertical surfaces of insulation need only be examined at the lowest c elevation where leakage'may be detectable. . + Essentially horizontal surfaces of' insulation shall .be examined at each' insulation joint. ll For components'whose external insulation surfaces

k ,are inaccessible for direct examination,.only the

,- examination of surrounding area,. including floor , areas or equipment surfaces located underneath the

L components,.for evidence of leakage, or~other areas to which such leakage may be channeled, shall be jp required.

I Discoloration or residue on su'rfaces examined shall -be given particular' attention to detect evidence of p7 boric acid accumulations from borated reactor i coolant: leakage. ~ The visual examination shall e conducted during . Q. the system leakage tests conducted after refueling , . outages and prior to startup. dydr -- SI-114.1 Pege 53 of 65 Rev 1 [~ VT-2 visual examination shall be conducted in 1 accordance with ASME Section XI, IWA-5240. 11.1.3 Visual Examination VT-3 The VT-3 visual examination shall be conducted to determine the general mechanical and structural i conditions of components and their supports such as the presence of loose parts, debris, or abnormal . corrosion products,. wear, 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, t 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, t , valves, and spring loaded and constant weight hangers. r VT-4 shall confirm functional adequacy, A verification of the settings, or freedom of motion. This examination may require (1) disassembly of I_ . components or devices and (2) operability test. L- . 11.2 Surface Examination. L. 11.2.1 Magnetic Particle Examination 1 Magnetic particle examination (MT) shall be 1. ~ conducted.in accordance with Article 7,Section V of the ASME Code. O f' L; ~ w we ^ SQNP ,, SI-114.1 + Page 54 of 65 Rav 1 r 11.2.2 Licuid Penetrant Examination Liquid penetrant examination (PT) shall be ~ conducted in accordance with Article 6,Section V of the ASME Code. 11.3 Volumetric Examination - 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 - radiographic examinations employing either X-ray equipment or radioactive isotopes and photographic ' films, the procedure shall 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 Article 4 of Section V of-the ASME Code. (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, reflectori that .L.. produce a response greater than 50% of the reference level shall be recorded. [- (c) If the requir'ements of (a) and (b) above are not applicable, the ultrasonic examination r shall be condu~cted in accordance with the j applicable requirements of Article 5 of Section V amended as follows: [ (1) ,For examination on welds, reflectors that J_ 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 ,1, SQNP p' x SI-114.1' 6 - Page 55 of 65 Rev 1 p ' ~ operator can determine the shape, identity, and location of all such [F ' reflectors in terms of the acceptance-rejection standards of IWA-3100(b). (3) The size of reflectors shall be measured between points which give amplitudes )n ' + equal to 100% of the reference level. 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 ASE ' Boiler and Pressure Vessel Code. .g 12.0 -QUALIFICATIONS OF NONDESTRUCTIVE EXAMINATION PERSONNEL = Personnel performing nondestructive examination operations shall be qualified in accordance with-IWA-2300 of ASME Section XI (DPM N75C01). =13.0 ACCEPTANCE CRITERIA x All acceptance standards for Class A, B, C, and D components shall be in accordance1with IWA-3000, IWB-3000, IWC-3000, or IWD-3000 of ASME Section XI, except where ASME Section III examinations are employed to satisfy ASME Section XI requirements.

_~

_114.0L REPAIRS y - -- 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 N-0QAM Part II,.Section 2.3 and L -implemented by DPM SQ82M1. t E 15.0 REPLACEMENTS C  ; Replacement of ASME Section XI components (ASME Classes 1, 2, and 3).shall be in accordance with the Repair and Replacement w Program in N-0QAM Part II,-Section'2.3 and implemented by [+c -DPM SQ82M1. b ASME Section XI repairs and replacements may be coordinated as-p -necessary with the Chemical, Metallurgy and Standards Group of the Mechanical Branch. f (, -16.0 RECORDS'AND REPORTS U' 16.'1 Recording of and Report of Examinations ' E A detailed report of all examinations'shall be prepared by b the performing ~or responsible organization and should ' ~ -.contain but not be. limited to'the following information: [. m - .~55- ' - 1.. ._ . _ . _ . _ _ _ _ _ . _ _ _ _ . , , . . . - . . _ . . . _ . . _ . _ - _ . _ _ _ _ . . . _ _ . _ 7 _ - .

y. SQNP' f . SI-114,1 J. ~ Page 56 of 65 Rev 1 6 -Title Page Table of Contents Introduction - The introduction should include the

~ I. following information: Plant, unit number, ^ preservice or inservice' inspection and cycle number, systems, ccmponents and vessels examinations w'ere performed on, organization examinations were performed by, dates examinations were performed, + ASME Section XI Code'of Record. 3-1~ II. . Summary - The summary should include a brief y description of the overall inspection: Program, i~ performance, personnel, equipment, procedures, evaluations, and results. - 7 III. Discussion - The discussion should discuss the . governing. documents (ASME Code, Technical Specifications, etc.), inspection schedule, e materials, calibration standards, calibration . @ performance, reporting,-recording, interpretation, e and brief evaluation. - f ?. 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 f evaluation section also should contain a listing of each examination performed and the evaluated T. results. + V. Summary of Notifications - The summary of ~[- - notifications shall'give a short summary of each notification report along with the indication.,

J. ,

discrepancy and its location. It should also A contain the final disposition and the date of completion. . c. 'VI. Scan Plan - The Scan Plan shall give a detaiied description of all a.reas subject to examidation during the inspection. It shall contain the following information: Examination Area, Code 7{ Category, Weld Size and/or Number, Reference

_ Drawing,. Examination Method, Procedure, Calibration d1 Block, and any reference details pertaining to the exam area, such as the weld number, meridional

~ U welds, pump' studs, etc. y, VII. Weld and. Hanger Maps --The Weld 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 { L . dyar SI-114.1 Page 57.of 65 Rev 1 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 . XII. ' Calibration Block Drawings XIII. Calibration Sheets XIV. Examination Reports - Reports for inservice - inspection shall be pr~epared in accordance with IWA-6220 of ASME Section XI. For eddy current examination of heat exchanger tubing, the report shall include a record' indicating the tube (s) exam-ined (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. I 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.QEB NDE Section shall review and submit the final report to the Plant Superintendent for review. These final reports shall b'e filed at the plant site with the data sheets of Appendix C of this program as discussed in .Section 5.0 of this program. Data Sheet 1 in-Appendix C will.be completed and used as a cover sheet for the final a report 'and to. document the review process. O g ' - f -SQNP 'SI-114.1 b ' Page 58 of 65 I " ' Rev l L . 16.1.1 Repair and Replacement Reports The plant.shall. prepare a summary of repairs and replacements for all Class A and B cocponents. The report .shall include the applicable requir .~nts of IWA-6220 of ASME Section XI and shall be subu d to the Chemical, Metallurgy and Standa'rds Group within 45 days.after each refueling outage. The report shall include' repairs conducted during each refueling outage and all repairs conducted since the end of .the preceding refueling outage. ,The Chemical, Metallurgy'and Standards Group shall review the report to assure:it--contains applicable information required in IWA-6220 of ASME Section XI. After reviewing the summary report, the Chemical, U Metallurgy and Standards Group shall forward it to the Quality Engineering Branch for submittal to the NRC as part'of the inservice inspection report s described in N-00.di, Part II, Section' 5.1 and.16.2 .of this program, within.90 days from the'end of each

. . . refueling outage.

16.2 ISI Report for Class A and B Components b i The QEB NDE Section shall prepare an ISI Sumnary _ -Report for Class'A and B components to be submitted within 90 days after the completion of the inservice inspection with the Nuclear Regulatory Commission Region II Office in accordance with IWA-6220, ASME Section XI. ^ - f_. p L' V- The ISI Report sho'uld have a cover sheet providing the [ -following information: .(1) Date. .  :(2) Name of owner and address of corporate offices. (3). Name-and address'of nuclear generating plant in which r 'the nuclear power unit.is located. o . [. (4) Name and' number assigned to the nue, lear. power unit by

l. -

TVA. .1 .(5) . Commercial operation date for unit.

.All reports shall have a summary providing the following I' information

[ . ., ,^ , (1) National Board Number assigned by the manufacturer to the boiler,' pressure vessel, or component. i - la 5 .SQNP- . SX-114.1 4 Page 59 of.65 Rev 1 (2) Names of the components or parts of the components for which this.is a record,. including such information regarding size, capacity, material, location, and drawings as may aid accurate identification. (3): Name.of the manufacturer of the components or parts for which this is a record, including the manufacturer's component or part nunbers and such .information regarding the manufacturer's corporate office or manufacturing plant locations as may aid in gaining access to the manufacturer's records regarding the components or parts that the manufacturer is maintaining. .(4) Date of completion of the inservice' inspection. ~ (5) _Name or names of the Inspector (s) when required. (6)' Name and mailing address of the employer (s) of the Inspector (s). (7) Abstract of examination performed, conditions observed, corrective measures recommended and taken. (8) Signature of Inspector, when required. All ISI Reports shall have an owner's data report for . inservice. inspection, Form NIS-1 as shown in Appendix II of ASME Section XI. ^ 16.' 3 Records for Class A, B, C. and D Components ~ The1following records shall be available for review: -(1) Examination Plans ~(2) Examination Results'and Reports - (3) Examination Methods and Procedures (4) Evaluation'of Results-(5) Corrective Actions and Repairs 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 taken. m L SONP SI-ll4.1 Page 60 of 65 - Rev. I 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 Inspecticas, of this program. 17.0 NOTIFICATION OF INDICATION Plant management shall be formally notified of the presence of unacceptable indications detected during the performance of nondestructive examinations. Unacceptable indications are defined ', by the applicable NDE procedure. Formal notification shall consist of completing and submitting to the Plant Superintendent the " Notification of Indication" form in Appendix D of this. program. -Part I of the " Notification of Indication" shall be completed and signed by the NDE Level II or III examiner detecting the indication. The NDE Section Representative of the Quality Engineering Branch shall assign a sequential number review and sign the form. If the indication is detected by an outside contractor, the contractor's field supervisor shall review and sign.the form. The original shall be sent to the plant superintendent and a copy to the ISI Programs Section. The plant superintendent or his assistant shall be responsiable for determining which organization (Field Services, Plant Maintenance) shall be responsible for recommending a disposition in Part II of the form and performing the associated repair. If .the organization assigned responsibility for disposition is unable to determine a satisfactory disposition then the form should be sent to the Mechanical Branch for disposition. The individual responsible for preparation of the disposition shall sign and date Part II of the form. The cognizant supervisor of the appropriate organization shall review and approve the disposi- ' tion and sign and date Part II of the form. Copies of the form shall be distributed to the plant superintendent and the ISI Programs Section. The original shall be returned to the NDE Section Represeptative. One copy shall be filed with the Lexamination report. .If the organization assigned responsibility for repair is within NUC PR, they shall initiate corrective action mea;ures in accordance with the requirements of N-0QAM, Part III, Section 7.2, and shall prepare instructions to repair the indication in accordance with the recommended disposition on the Notification of Indication form and N-0QAM, Part II, Section 2.1. Dispositions other than restoring to original requirements shall be processed as modifications in accordance with N-0QAM, Part II, Section 3.0. Repair and replacement, , activities,. including coordination with the Authorized Inspection Agency (AIA), shall be performed in accordance with the requirements of N-0QAM, Part II, Section 2.3. Dispositions to accept the condition as-is, a USQD shall be prepared by the appropriate organization in accordance with established procedures. 7 w . gq3p. d -SI-114.1- -Page 61 of 65 Rev 1 Upon ecmpletion ofIthe action required to repair the indication,- the NDE Section Representative shall verify completion of -corrective action, enter the work instruction and/or DCR numbers ~ - on the Notification-of-Indication form, enter the examination . report number if re-examination was ' performed, and sign and date ' tbe form, Part III. 'The. signed form shall remain with the - examination report.for use as a quality assurance record. ^ If-s >. re-examitation_was performed, a copy _of the-signed form shall also' remain with the re-examination report. . Copies of the form shall also be distributed to those listed in the second . paragraph above. 18.0 CALIBRATION BLOCKS Calibration blocks will be used for ultrasonic examinations (a calibration tube will be'used for. eddy current examination of ~ g steam generator tubing). The blocks will be fabricated to the - general requirements of ASME Section V and.ASME Section XI. The -blocks shall be fabricated of the: material to be examined or-equivalent P numbers. Mill test reports shall be~obtained and retained by the Quality Engineering Branch for all calibration blocks. The blocks shall. employ drilled holes and/or notches for calibration reflectors'(See. Request for-Relief ISI-2). The NDE Section shall maintain as-built calibration block - drawings. Copies ~of the original drawings.and any revisions shall be submitted to_th'e ISI Programs Section. The calibration blocks shall be stored ^at the plant' site and maintained by the plant QA organization. 119.0 REQUESTS-FOR RELIEF-- , Vaere TVA has determined that certain code requirements or , . examinations are impractical, TVA will submit written requests-for relief to NRC with information to suppert the determinations - and any proposed alternate examinations. The impractical code . requi'rements_or examinations shall be identified in this program, . Land references to.particular' requests for relief shall be. included. . When impractical examination: requirements'are identified in the field, the NDE'Section shall notify the ISI Programs Section such Lt hat the information may be included in this program and requests _ for ' relief may'be prepared -if necessary. The Requests for Relief are listed in Appendix E. The current Requests _for Relief are ISI-1 through ISI-11. Those requiring alternate inspection are ISI-1, ISI-3, ISI-5',- ISI-8, ISI-9, , ISI-10., and ISI-11. -20.0 ' AUGMENTED INSPECTIONS Augmented inspections are performed in addition to ASME LSectionn XI code requirements. The augmented inspections may be required by the NRC or self-imposed by TVA. _ u-< SQNP SI-114 01 Page 62 of 65 Rev 1 20 ~.1 Feedwater' System Pinng and Supports In order to satisfy the requirements of NRC-IE -Bulletin 79-13, the following augmented examinations shall be perfor.aed. The augmented examination of the feedwater nozzle-to pipe welds (includes nozzle '.o-transition piece welds and -transition piece-to pipe welds) and of adjacent pipe and nozzle areas are. included in DPM No. SEQ 80M7. The requirements-are a radiographic examination supplemented by ultrasonic examination on completion of the hot functional testing and before fuel loading. During the first refueling outage, perform a volumetric examination of the feedwater nozzle-to pipe welds, the feedwater piping welds to the first support, the feedwater line-to-containment ' penetration wel_ds and an area of at le ast one pipe diameter of the main feedwater line downstream at the auxiliary feedwater to main feedwater connection. Also perform a visual inspection of all feedwater system piping supports and snubbers in containment to verify op,erability and conformance to design. l' In the event cracking is identified during examination of . the nozzle-to pipe welds, all feedwater lines up to the first piping support or snubber outboard'of the nozzle shall-be volumetrically examined as stated in tne first paragraph. All unacceptable code discontinuities shall be subject' to repair unless justification for continued operation is provided. Any cracking or other unacceptable code discontinuities. identified shall be reported to the Director of the NRC Regional Office within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of identification. Plant -management shall report this information to the Nuclear - ' Regulatory Commission Region II Office per Plant Instruc- ..e etion AI-18. A written report of the results of examina-tion shall be submitted by the QEB NEE Section within 30 days of completion of the examination. Refer to DPM No. SEQ 80M7 for'information to be included in this special report. 20.2' RPV Nozzle Safe Ends The augmented examination requirements of the RPV nozzle-to-safe end welds are included in the final report - Sequoyah Nuclear Plant - Eiraluation of cracking in reactor vessel. nozzle stainless steel buttering. The examinations for unit I will be monitored at the normal inservice inspection intervals for dissimilar metal welds as required by Section XI of the ASME Code. This' augmented examination does not require a special report. The examination data sheets Aall be included in the 90 day ISI report discussed in .ection 16.2 of this program. 6 SQNP

f, SI-114.1 Page 63 of 65 Rev 1

.20.3:-Reactor Coolant Pump Flywheel The augmented examination requirements of the reactor coolant pump flywheel are included in Regulatory Position C.4.b of Regulatory' Guide 1.14; (1) an inplace ultrasonic examination of.the areas of 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) s surface examination of all. exposed surfaces and complete ultrasonic examination -at approximately-10-year' intervals during the plant ~ shutdown coinciding with the inservice inspection schedule ' as reguired by: Section XI of the ASME Code. This eugmented 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 te 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 this-program. 20.4 Pressurizer' Relief Line The augmented examination requirements of the' pressurizer

m. relief;1ine are. included in'the Technical Specifications

4.0.5 and 4.4.3.2.4. The pressurizer' relief line and

. . repair welds shall be examined using improved ultrasonic ' detection and evaluation procedures prior to entering Mode 4 whenever the plant has been in cold shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more if the examination has not been performed in the previous six months. 4 In the event these six-month period examinations find the piping free of unacceptable indications for three 1 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 augmer.ted examination requirements of the steam -generator tubing are included in Technical Specifications 4.0.5,14.4.5.0 and section 7.3.8 of this program. Following each inservice inspection of steam generator tubes, the number of tubes plugged in each steam , generator shall be reported to the NRC within 15 days. r . syv SI-114.1 Page 64 of 65 Rev 1 Plant management shall report this information to the Nuclear Regulatory Commission Region II office within the time period specifit.a. 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 12 months following completion of the inspection. The Chemical, Metallurgy, and - Standards Group Staff Specialist shall prepare this special report and submit the report of the Nuclear Regulatory Commission Region II Office within the stated time period. This.special report shall include:

1. Nuaber and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

Results of steam generator tube inspections which fall into Technical Specification Category C-3 require prompt notification of the Nuclear Regulatory Commission pursuant to Technical Specification 6.9.1 prior to resumption of , plant cperation. 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. Refer to Technical Specifications 4.0.5 and 4.4.5.0 and section 7.3.8 of this program for information to be included. 20.6' RPV Nozzle Cladding All vessel nozzles cladding ~shall be ultrasonically examined at the end of each 10 year inspection interval, using techniques at least as sensitive as those used to conduct the supplemental examinations performed prior to fuel loading. The results of this 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 examina-tion does not require a special report. This examination data sheets shall be included-in the 90 day ISI report discussed in section 16.2 of this program. Nczzle forging material and cladding is identified in section 7.1.2 of this program. ~ l L , SQNP SI-114.1' Page 65 of.65 . . . 'Rev 1 -20.7 -RPV Closure Head Circumferential Weld (WO9-10) 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'the 90. day ISI report discussed in section 16.2'of this program. 21.0JREFERENCES L21.1 'ASME Boiler and Prissure. Vessel Code - Section XI --1974 Edition Summer 1975 addenda, Summer 1976 addenda; .1977 Edition, Summer'1978 addenda. 21;2 ASME Boiler and Pressure-Vessel Code - Section V. 21.3' Instruction Manual - 173-inch I.D. Reactor Pressure t Vessel -Rotterdam Dockyard Company, Contract No. 68C60-91934,.N2M-2-3. 21.4 Sequoya'h Nuclear' Plant Standard Practice SQA41. _ . L21.5 1Sequoyah Nuclear Plant Final Safety-Analysis Report. L21.6 Westingho.use Technical. Manual - Pressurizer, TM 1440-C225, Contract No. 68C60-91934, N2M-2-6. 21.7_ Westinghouse Technical Manual - Vertical Steam Generators,

D1 1440-C224,~ ' Contract No. 68C31-91934, N2M-2-4.

21.8 - Westinghouse ' Instruction. Manual Auxiliary Heat Exchangers,. Contract.No. 68C60-91934', N2M-2-25. .21.9. Westinghouse : Instruction Book - Reactor Coolant Pump,, = . Contract .No. 68C60-91934, N2M-2-5. .21.10 Ingersoll-Rand Instruction Manual'- Residual Heat Removal ~ , Pumps, Contract No. 68C60-91934, N2M-2-30. ~21.11 Sequoyah Nuclear Plant Operational Quality Assurince Manual,'Part II,' Sections 2.3, 3.0, 5.1, and 6.3. -21.12 Sequoyah Nuclear Plant Technical Instruction TI-51. 121.13 Sequoyah Nuclear Plant Technical Specifications Unit 1. 'e -21.14' Division Procedure Manual: -N75C01, N76A10, :;80E3, N83A3, 'SEQS0M7,. SEQ 82E1, and SQ82M1. ~ 7SQNP . . S1-114.1 page 1 of.2 tRev i . APPENDIX A TABLE 1

Hinimam Number of Steam Generators To Be Inspected During Inservice' Inspection Only a

No. of Steam Generators per Unit Four . First Inservice s Inspection Two Second & Subsequent Inservice Inspections Onc8 4 Table Notation: w w ~

1. Each of the other two steam generators not inspected during the first inservice inspections shall be inspected during the second and third inspections. The fourth and subsequent inspections shall follow the instructions described below:

1 4 1 - The inservice inspection may be. limited to one steam generator on a rotat.ing schedule enecmpassing 12% of tM tubes if the results of the first or previous inspections indicate that all steam generators are performing in i a like ma'nner. Note that under some circumstances, the operating conditions in one or more steart generators may be found to be more severe than those in other steam generators. Under such circumstances the s.n ple sequence shall be modified to inspect the.most severe conditions. e d i i 4 5 e -7 SQNP SI-ll4.1-

Page 2 of 2 Rev I APPENDIX A

, TAlli.E 2 Steam Generator Tube inspection Ist Sample Inspection 2nd Sample Inspection 3rd Sample Inspection Srmple Size Result Action Required Result Action Required Result Action Hequired A minimum of S C-1. None N/A N/A N/A N/A Tubes per S.G. C-2 Plug defective tubes C-1 None N/A N/A and inspect additional Plug defective tubes C-1 None 2S tubes in this S.G. C-2 and inspect addi- C-2 Plug defective tional 4S tubes in tubes this S.G. C-3' Perform action-Perform action for for C-3 result >s C-3 C-3 result of first of fi rst sa'nple y sample U/A N/A C-3 Inspect all tubes All other this S.G., plug ' S.G.s'are None N/A N/A defective tubes and C-1 inspect 2S tubes in Some S.G.s each other S.G. C-2 but no Perform action for N/A N/A additional C-2 result of Prompt notification S.G. are second sample to NRC pursuant C-3 to specification Additional Inspect all tubes 6.9.1 S.G. is C-3 in each S.G. and N/A N/A plug defective Lubes. Prompt notification to l NRC pursuant to specification 6.9.1 S = 12*/, n Where n is the number of steam generators inspected during an inspection. SQNP . SI-114.1 Page :1 L of 12 ~ Rev 1 APPENDIX A TAlli.E A - SEQUOYAll INSERVICE INSPECTION PROGRAN j CU.SS A COMPONENTS Program Section XI. 40 Yr. Sample 1st Insp. Interval Reference Exam Exam No. Lengtli No. Lengtli - Inspection Periods Reference Component Section- Hetliod Category Welds /of Weld Welds /of Weld 3 yrs. T yrs. 10 yrs. Dwg. No. A. Reactor Vessel

1. Circumferential SlielI Welds. 7.1.1.1 UT B-A 4/50-ft. 4/50 ft. 0 '0 4 Clif t-2343-

2. Closure Ilead D. Circ Weld 7.1.1.3 UT ,

B-A -41 ft. 41 ft. 13.fL. 14 fL. 14 fL. CIIN-2358-l 3. Lower IIcad - Circ Weld 7.1.1.4 UT B-A 38 It. 38 ft. 0 0- 38 fL. Clif t-2343-

4. Iower llead Heridional Welds 7.1.1.6 UT B-A 6 6 0 0 6 CIIII-2343-4

5. Sliell-to-Flange . .

WeId 7.1.1.7 UT B-A 50 ft. 50 ft. O ft. O ft. 50 ft. Cllti-2343-

6. Closure llead-to-Flange Weld 7.1.1.8 UT B-A 45 fL. 45 ft. 15 ft. 15 ft. 15 f t . Clif t-2358-l 7. Nozzle-to- Clifi-2343-

, Vessel Welds 7.1.2 UT B-D 8 8 4 0 4 Cllti-233'i- , , Clift-2360-

8. Inside Radius Cllti-2361-Sections 7.1.3 UT B-D 8 8 4 0 4 1S1-0008-

-9. Vessel Pene- Clit!-2651-trations and HSG-0004-Atta climents '7.1.4 VT '2 n-E 37 37 , 12 12 13 131-0014- .SQNP SI-il4.1 Page 2 of 12 ' APPENDIX A TAT >LE A (Continued) . Rev 1 SEQUOYAll INSERVICE INSPECTION PROGRAM ' CLASS A COMPONENTS-Program Section XI '40 Yr. Sample 1st Insp. Interval Reference Exam Exam No. Length No. Length ~ Inspection Periods . Reference Component Section Method Category ' Welds /or' Weld Welds /of Weld 3 yrs. 7 yrs. :10. yrs. 1)wg. No. A. Reactor Vessel (cont'd)

10. Nozzle-to-Safe CIIM-2333-

.End Welds -7.1.5 UT,PT B-F 8 8 4 0 4 ISI-0008-

11. Closure Studs and Nuts 7.1.6 UT,t!T B-G-1 54 54 18 ^18 18 Cllti-2341 -

12. I.igaments Between ,

Threaded Stud lloles 7.1.6 UT B-G-1 54 54 18 18 18 CliH-2341-

13. Closure Washers 7.1.6 VT-1 B-G-1 54 54 18 18 18 Clitl-2341-4

14. Vessel Interior 7.1.9 VT-3 B-N-1 1 1 1 1 I

15. Removable Core Support Structure 7.1.10 VT-3 B-N-3 See Program Section 7.1.10 *

16. Control Rod Clit!-2651-Drive llousings 7.1.11 UT B-0 20 2 0 1 1 Clitl-2359-

17. Auxiliary IIcad 7.1.12 UT,PT B-F 4 4 1 1 2 Cllti-2337-Cllti-2651-ISI-0001-ISI-0014-

SQNP SI-Il4.1 Page 3 of 12 APPENDIX A' Rev'l TABI.E A (Coeitinued) SEQUOYAll INSERVICE INSPECTION PROGRAtt - Cl. ASS A Cotli'ONENTS Program Section XI 40 Yr. Sample 1st Insp. Interval Reference Exam Exam No. Length No. Length Inspection Periods Reference Component Section Method Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. !0 yrs. Dwg. No. B. Pressurizer

1. Circumferential Shell-to-llcad 4

Welds 7.2.1 ,UT B-B 2/24 ft. 2/24 ft. 12 ft. 12 ft. 12 ft. Cllti-2363-/

2. Longitudinal Shell-to-llcad Welds 7.2.2 UT B-B 2/1 ft. 2/1 ft. 1* 1* 1* Cilfl-2363-/

3. Nozzle-to-Vessel and Inside Radius Section 7.2.4 ITF B-D 6 6 2 2 2 Clit!-2363-/

4

4. IIeater Penetrations 7.2.5 VT-2 B-E 78 20 6 7 7 tlSG-0006-/

5. Nozzle-to-Safe End Welds , 7.2.6 UT,PT B-F 6 6 2 2 2 Cllti-2363-/

6. Pressure Retain-ing Bolting 7.2.8 VT-1 B-G-2 1 ?!wy/16 bolts 16 bolts 16 0 0

7. Integrally Welded Vessel Supports 7.2.9 MT B-Il 23 ft. 23 ft. 7 ft. 8 ft. 8 ft. Clit!-2362-t

• The longitudinal weld selected for examination is that weld intersecting the circumferential shell-to-head weld examined.

'~ SQNP ' SI-Il4.1 Page 4 of 12 APPENDIX A Rev 1 TABI.E A (Continued) , SEQUOYAll INSERVICE INSPECTION PROGRAtt CLASS A COilPONENTS Program . Section XI 40 Yr. _ Sample 1st Insp. Interval" _ Reference . Exam Exam No. Length No. Length Inspection Periods Re fe rence Component. Section tiethod Category -Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwn. No. C. Steam Generators

1. Prima ry IIcad-to-Tulm Sheet Weld 7.3.2 UT B-B 4/36 fL. 4/36 ft. I 1 2 Cilti-2345-

2. Primary Nozzle-to-Vessel and Inside Radius Section 7.3.3 See Program Section 7.3.3 (Request for Relief ISI-6)

3. Primary Nozzle- ISI-0008-to-Safe End , Cllti-2333-Welds 7.3.4 IIT,PT B-F 8 8 2 3 3 C1111-2345-iM 4. Pressure Retain-ing Bolting 7.3.6 VT B-G-2 4 Gens./2Mwy 8 Hwy 2 2 4  !!SG-0002-4

5. Tubing 7.3.8 ET B-Q .4 Gens /3% 4 Gens /3% 2 1 1 76tli D. Pressure Retaining Ilolting

1. Reactor Coolant System 7.4.3.1 VT B-G-2 8 Sets /8 Bolts 8 Sets 2 3 3 CIlli-2334--

2. Chemical and Volume Control .

3 System (SWI) 7.4.3.2 VT B-G-2 4 Sets /4 Bolts 4 Sets 1 1 2 Cllti-2338-'

3. Residual lleat Removal System 7.4.3.3 VT B-G-2 N/A See Program Section 7.4.3.3 4

.SQNP SI-ll4,1 Page 5 of 12 ' APPENDIX A 'Rev 1 TABLE A (ConLinued) SEQUOYAll INSERVICE INSPECTION PROGRAN CLASS A COMPONENTS Program .SectiodXI 140 Yr.' Sample 'Ist Insp. Interval Reference Exam Exam No. Length ' No. Length Inspection Periods Reference Component Section Method Ca tego ry _ Welds /of-Weld Welds /of Weld- 3 yrs. 7 yrs. 10. yrs. Dwg. No. D. Pressure Retaining Bolling (cont'd)

4. Safety Injection System. 7.4.3.4 VT-1 B-G-2 4 Sets /4 Bolts 4 Sets 1 1 2 CitH-2333-(

5. Upper IIcad Injection System 7.4.3.5 VT-1 B-G-2 12 Sets /2 Bolts 12 Sets 4 4 4 CilH-2337-t -

E. Piping

1. Circumferential Welds A. Reactor Coolant .

NJ System Main Loops b' Circs }4" 7.4.4.1 UT/PT B-J '3 6 16 5 5 6 CIIM-2333-1 B. Reactor Coolant ~ Systcm Cires }4"

Nom. hize 7.4.4.2 .UT/PT B-J 64 16 5 5 6 Clitt-2334-t Cires (4" Nom. Size 7.4.4.2 PT- B-J 107 27 9 9 9 CllH-2334-t C. Chemical and Volume Control System Cires }4" Nom. Size 7.4.4.3 N/A ,

Cires (4" Nom. Size 7.4.4.3 PT B-J 58 15 5 5 5 CllH-2335-( SQNP SI-Il4.1 .. Page 6 of 12 APPENDIX A j Rev 1 TABLE A (Continued) SEQl10YAH INSERVICE INSPECTION PROGRAN Cl. ASS A COMPONENTS Program 'Section XI . 40 Yr. Sample 1st 1 Insp. Interval Reference Exam _ Exam . No. Length No. Length laspection Periods Reference Component Section Method . Category Welds /of Weld Welds /of Weld. 3 yrs. 7 yrs. 10 yrs. Dwg. No. E. Piping'(cont'd) D. Residual IIcat Removal System Cires }4" Nom. Size 7.4.4.4 UT/PT B-J 51 13 4 4 5- Clit!-2336-( E. Safety Injection System Cires 24" Nom. Size 7.4.4.5 UT/PT B-J 105 27 9 9 9 Cliff-2333-C Circs (4" Nom. Size 7.4.4 35 PT_ B-J 24 6 2 2 2 Cliti-2333-C . N to F. Upper IIcad Injection System Cires 14" 7.4.4.6 UT/PT iB-J 122 31 11 10 10

2. Branch Pipe Connection Welds A. Reactor Coolant System -

flain Loops Welds >2" Nom. Size 7.4.5.1 UT/PT B-J 1 1 O O 1 CIIti-2333-l? Welds $2" . Nom. Size 7.4.5.1 PT B-J 13 4 1 2 1 Clit!-7333-1:

k. n

SQNP. SI-il4,l' Page 7 of 12 APPENDIX A Rev 1 TABLE A (Continued) SEQUOYAll INSERVICE . INSPECTION PROGRAN Cl. ASS A C0flPONENTS Program Section XI 40 Yr. Sample 1st Insp. Interval Reference; Exam Exam No. Length No. Length Inspection Periods Reference . Component Section Method Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. E. Piping (cont'd)

2. Branch Pipe Connection Welds (cont'd)

B. Reactor Coolant System Welds >2" Nom. Size .7.4.5.2 UT/PT B-J 6 2 0 1 1 CIIH-2334-Welds (2" Nom. Size 7.4.5.2 PT B-J 7 2 0 1 1 Cllti-2334-C. Chemical and Volume Control System Velds >2" 9 Nom. Size 7.4.5.3 UT/PT B-J 4 1 0 0 1 Cllti-2335-t Welds {2" Nom. Size' 7.4.5.3 PT B-J - - - - - Clift-2335-D. Residual !! cat Removal System Welds >2" Nom. Size 7.4.5.4 UT/PT B-J 3 1 0 0 1 Cilfl-2336-Welds (2" Nom. Size 7.4.5.4 PT B-J - - - - - Cllti-2336-E. Safety Injection System Welds >2" , Nom. Size 7.4.5.5 UT/PT B-J 5 2 0 1 1 Clit!-2333-Welds {2" Nom. Size 7.4.5.5 PT B-J 4 1 0 0 1 Clit!-2333-4 , .SQNP SI-114,1 Page 8'of 12 APPENDIX A Rev 1 . TABLE A (Continued) SEQUOYAll INSERVICE INSPECTION PROGRAtl - CLASS A C0tlPONENTS Program 'Section XI 40 Yr. Sample 1st Insp. Interval Reference Exam Exam No. Length No. I.cngtli Inspection Periods Reference Component Section Metliod Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. E. Piping (cont'd) F. . Upper llead Injection System Welds >2" Nom. Size 7.4.5.6 UT/PT B-J - - - - -- Cllti-2337-t Welds {2" - Nom. Size 7.4.5.6 PT B-J 2 1 0 0 1 Cliff-2337-t

3. Socket Welds A. Reactor Coolant

,g. System 7.4.6.1 PT B-J 261 66 22 22 22 Clit!-2334-( '\ B. Chemical and Volume Control Syste'm 7.4.6.2' PT B-J 228 57 19 19 19 Cllti-2335-t C. Residual IIca t Removal , System 7.4.6.3 PT B-J

• 18 5 1 2 2 Clit!-2336-t D. Safety Injection System 7.4.6.4 PT B-J 192 48 16 16 16 Cllti-2333-f E. Upper ifead Injection System 7.4.6.5 PT B-J 16 4 1 1 2 Clit!-2337-0

SQNP SI-il4bl Page 9 of 12 APPENDIX A Rev.1 TABLE A (Continued). SEQUOYAll 1NSERVICE INSPECTION PROGRAM CLASS A C0tlPONENTS Program Sect.ioli XI 40 Yr. Sample 1st Insp. Interval Reference ~ Exam Exam No. Length No. Lengtli Inspection Periods Reference Component. Section Method Category. Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. 4 '. Piping i 'and' Valve Integrally-Welded. /. , Support ticiabers - A. Reactor , Coolant.. . ' s Systeas' 7.4.7.'l PT B-K-1 16 16 5 5 6 Clit!-2432-1 11 . Cliemical and Volume Control Cllti-2433-( System 7.4.7.2 ' PT ~ B-K-1 7 7 2 2 3 C11!1-2434-t . 4 Residual

• C. ~

lleat Removal System 7.4.7.3' PT B-K-1 4 4 1 1 2 Clit!-2435-( D. Sa fety Injection Systchi 7.4.7.4 PT B-K-1 12 12 4 4 4 Clitt-2436-t E. Ilpper llead

• Injection 7.4.7.5 PT . B-K-1 2 2 0 1 1 Cllti-2437-(

5. Piping and Valve Coml>onent Supports A. Reactor Coolant System 7.4.8.1 VT 3 B-K-2 127 127 42 42 43 Clit!-2432-(

VT-4 i SQNP SI-ll4.1 Page 10 of 12 APPENDIX A Rev 1 TAUI.E A (Continued) SEQUOYAH INSERVICE INSPECTION PROGRAN CLASS A COMPONENTS Program Section XI 40 Yr. Sample 1st Insp. Interval Reference Exam Exam No. Length No. Length Inspection Periods Reference Component Section Method Category Welds /of Weld. Welds /of Weld. 3 yrs. 7 yrs. 10 yrs. Dwg. No. B. Chemical and Volume . Control . System 7.4.8.2 VT-3, B-K-2 107 107 36 35 36~ CllH-2433-C VT-4 Cllti-2434-C C. Residual llea t Removal System 7.4.8.3 VT-3, B-K-2 7 7 2 2 3 CIIM-2435-C VT-4 D. Sa fety , Injection s System 7.4.8.4 VT-3, B-K-2 85 85 28 28 29 CilH-2436-C VT-4 E. Upper Head Injection System 7.4.8.5 VT-3, B-K-2 85 85 28 28 29 Cati-2437-C VT-4 F. Reactor Coolant Pumps ,

1. Pressure-Retain-ing Bolting 7.5.1 UT B-G-1 96 96 24 24 48 CllH-2675-B

2. Pressure-Retain- UT,PT, ing Bolting 7.5.1 or HT B-G-1 96 96 24 24 48 Clitt-2675-Il

3. Pressure-Retain-ing Bolting 7.5.2 VT B-G-1 4 Sets

• 4 Sets

• 1 Set

• 1 Set

• 2 Sets *Citti-2675-B

, SQNP SI-114,11 'Page 11 of'12 APPENDIX A -' 'Rev 1 . TABLE A (Continued) SEQUOYAll INSERVICE INSPECTION PROGRAN , , CLASS A COMPONENTS Progr'am JSection XI 40 Yr.' Sample 1st Insp. Interval Re ference . Exam Exam No. Length No. Length - Inspection . Periods . Reference f Component Section' ' Method Category -Welds /of. Weld- Welds /of Weld- 3 yrs. 7 yrs. 10 yrs. Cwg. No. '4. .Comporient: v- Supports '.5.4 7 VT B-K-2. 4 Pumps /3 ft. 12 ft. 3* -3* . 6*c MSG-0003-1 .s per pump * (1 pump)(1 pump)(2 pumps) ~

5. Ca' sing ' Welds 7.5.5 PT B-L-1~ 4 >

1. O C- 1 IISG-0003-I

6. Casing. 7.5.6 VT B-L-2 . 4 See Program Section 7.5.'6 Iy;G-0003-1 i

7. Flywheci 7.5.7 UT- N/A 4' See Progrha Section 7.5.7 N/A

,r, .)G. Valves:

1. Pressure Retain- ~

-} ing: Bolting 7'.6.2 ST B-G-2 See Program "ection 7.6.2 ' A. Reactor . Coolant System 7.6.2.1 VT-1 B-G-2 17 17 5 6 6 Clitl-2334-1 B. Chemical and Volume Control System 7.6.2.2 VT-1 B-C-2 6 6 2 2 2 Clift-2335-t C. Residual lleat Removat System 7.6.2.3 VT-1 B-G-2 2 2 0 1 1 Cllti-2336-t D. Safety Injection System 7.6.2.4 VT-1 'B-G-2 23 23 7 8 8 CilH-2333-t

• See Program Reference Section for explanation of inspection sample.

3 , a SQNP. SI-Il4*l APPENDIX A-Page 12 of 12 TABLE A (Continued)' Rev l ~ . SEQUOYAll INSERVICE INSPECTION PROGRAM ' CLASS A CONPONENTS, Program .Section XI- 40 Yr. Sample 1st Insp. Interval

Reference Exam Exam No. . Length No.. Length Inspection Periods Reference Component -Section . Method Category Welds /of. Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg.-No.

~ E. Upper Head. Injection . . System 7.6.2.5 VT-1 B-G-2 8 8 2 3 3 CilH-2337-C

2. Integrally-Welded ~

Support Members 7.6.3 UT/PT B-K See Program Section 7.4.7

3. Component Supports 7.6.4 VT-3,. B-K-2 See Program Section 7.4.8 VT-4

4. Valve Bodies 2 >4" N.P.S. 7.6.6 VT-4 B-H-2 See Table D of Appendix A

n. Exempted Components 7.7 VT-2 B-P See Program Section 7.7

1. Reactor Vessel - Pressure Retaining Boundary

2. Pressurize,r - Pressure Retaining Boundary.

i

3. Steam Generators - Pressure Retaining Boundarv i

4. lleat Exchangers - Pressure _ Retaining Boundary

5. Piping - Pressure Retaining Boundary 6.

Pumps - Pressure Setaining Boundary .

7. Valves - Pressure Retaining Boundary 1

4 W' e . SQNP SI-114.1 'Page 1 of 9 APPENDIX A Rev 1 tai!I.E ' B SEQUOYAll INSERVICE INSPECTION PROGRAM CLASS B C0tlPONENTS Program Section XI 40 ' Yr. Sample 1st insp. Interval Reference Exam Exam. No. Length No. Length Inspection Periods Reference Component Section .Hethod Ca tego ry Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. A. Steam Generators,

1. Circumferential Shell Welds 8.1.2 UT C-A 4 Gens./3 Welds (12) 3 1 1 1 Clift-2345-

2. Circumferential .

Ilead Welds 8.1. 2 - UT C-A 4 Gens./1 Weld (4) 1 Wld/46 ft. 15' 15' 16' CliH-2345- ~

3. Tubesheet-to-Shell Weld 8.1.3 UT - C-A 4 Gens./1 Weld (4) 1 Wld/36 ft. 12' 12' 12' Cl!M-2345-

4. Nozzle-to-Vessel "9 Welds 8.1.4 UT,NT C-B 4 Gens /2Noz (8) 2 1 0 1 CIR1-2345-O B. lieat Exchangers

1. Residual IIcat Removal iteat Exchangers (2)

a. HliRilX' Circum-ferential Shell Weld 8.2.1.1 UT C-A 211t.Ex./lWeld (2) 1 1 0 0 C1111-2404-

b. RllRilX Circum- .

ferential IIcad Weld 8.2.1.2 UT C-A 211t.Ex./1 Weld (2) 1 1 0 0 Clift-2404-

c. RilRilX Nozzle- .

to-Vessel Weld' 8.2.2 UT,PT C-B 21't . Ex. / 2Noz . (4) 2 1 0 1 CliM-2404-I SQNP APPFNDIX A S1-114.1 ,. TABLE B (Continued) Page 2 of 9 dcQUOYAll INSERVICE INSPECTION PROGRAN Rev 1 CLASS 11 COMPONENTS Program Section XI 40 Yr. Sample 1st insp. Interval Reference Exam Exam No. Length No. Length Inspection Periods Reference Component .Section Method Category Welds /of Weld Welds /of Weld 3 yrs. 7. yrs. 10 yrs. Dwg..No. B,lleat Exchangers (cont'd)

d. RilRIIX Integrally Welded Supports 8.2.3 PT C-C 2ilt.Ex./2 Welds (4) 2 1 0 1 CllH-240 c .- RilRllX Component Supports 8.2.4 VT C-E 21it.Ex./2 Sprts. (4) 2 1 0 1 CliM-240

2. Regenerative IIeat Exchanger (1)~RilX 9 a. RIIX Circum-ferential licad Weld 8.3.1.2 UT C-A 1llt.Ex./6 Welds 6 2 2 2 151-0066-

b. RilX Circum-ferential Tut,esheet-to-Shell' Weld 8.3.1.3 UT C-A IIIt.Ex./6 Welds 6 2 2 2 151-0066-

c. RilX Int.- ,

Wid. Support Lugs 8.3.3 ST C-C ISet/3 Lugs 1 1 0 0 ISI-0066-

d. RIIX Component Supports 8.3.4 VT-3 ,C-E 2 2 1 1 0 ISI-0066-o i

SQNP SI-11461 APPENDIX A' Page 3 of 9 TABLE D (Continued) .Rev 1 SEQUOYAH INSERVICE INSPECTION PROGRAM . CLASS Il COrlPONENTS , Program Section XI 40.Yr. Sample 1st Insp. Interval Reference Exam Exam No. Length No. . Length Inspection Periods Reference Component Section Method Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. B. Heat Exchangers (cont'd)

3. Letdown lleat Exchanger.(1) LilX

a. LIIX Cireum-ferential licad Weld 8.4.1.2 UT C-A 111L.Ex./2Wel ds 2 1 1 0 ISI-0068-

b. LilX Component Supports 8.4.4 VT-3 C-3 2 2 1 1 0 ISI-0068-

4. Excess Letdown N lleat Exchanger *

(1) ELilX

a. ELilX Circum-ferential llead Weld 8.5.1.2 UT C-A 11It.Ex./1 Weld 1 1 0 0 ISI-0067-C. Tanks

1. Boron Injection Tank (1) BIT

a. BIT Circum-

• ferential Shell/licad Wld 8.6.1 UT C-A ITank/2 Welds 2 1 0 1 ISI-0074 .

b. BIT Nozzle to Vessel W1ds 8.6.2 UT,ST C-B 1 Tank /2Noz 2 1 1 0 ISI-0074 .

SQNP SI-114.1 . APPENDIX A Page 4 of 9 TABLE B (Continued) Rev 1 SEQ 110YAH INSERVICE INSPECTION PROCRAM CLASS B C0rlPONENTS Program Section,XI- 40 Yr. Sample 1st Insp. Interval Reference -Exam Exam No. Length No. Length Inspection Periods Reference Component- Section Method Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. C. Tcnks (cont'd)-

• c, BIT Component Supports 8.6.3 VT-3 ~ C-E 4 4 1 1 2 ISt-0074-A

d. BIT Pressure Retalning ,

Ilolting >2" Dia. 8.6.4 UT C-D IMwy/16 Studs IMwy 1 0 0 ISI-0074-A

2. UllI Water Accumulator (1) WA og a. WA Circum-W ferential Shell/Itcad Wld 8.7.1 UT C-A ITank/2 Welds 2 1 1 0 ISI-0070-A

b. WA Nozzle to Vessel Welds 8.7.2 UT,ST C-B 1 Tank /3Noz 3 1 1 1 ISI-0070-A

c. WA Int.-

Weld Supports 8.7.3 S _T C-C 1 Support Skirt 35' 12' 11' 12' ISI-0070-A

d. WA Component Supports 8.7.4 VT-3 C-E 1 Cmpnt Support 1 0 1 0 ISI-0070-A Skirt 2

L - - .n_a SQNP SI-114,1 APPENDIX A Page 5 of 9 TAllLE 11 (Continued) Rev l- SEQUOYAll INSERVICE INSPECTION PROGRAN CLASS 11 COMPONENTS Program . Section XI 40 Yr. Sample 1st Insp. Interval ' Reference Exam Exam No. Length No. Length Inspection Periods Reference Component Section Method Category. Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. C, Tanks (cont'd)

e. WA Pressure Retaining Bolting

>2" Dia. 8.7.5 UT C-D lifwy/16 Studs IHwy 0 1 0 ISI-0070-

3. UIII Surge Tank (1) ST

a. ST Circum-ferential Shell/IIcad Wld 8.8.1 UT C-A ITank/2Wlds 2 0 1 1 ISI-0071 .

• 9

• b. ST Int.-

Weld Supports (PADS) 8.8.3 ST. C-C ITank/2 Pads 2 1 1 0 ISI-0071 .

c. ST Componen.

Supports 8.8.4 VT-3 C-E 2 2 1 1 0 ISI-0071 , D. Piping

1. Integrally-Welded. Supports

a. Residual lleat Removal .

System 8.9.1.1 PT C-C . 21 21 7 7 7 CllH-2435-t

b. Safety Injec-tion System 8.9.1.2 PT C-C 10 10 3 3 4 Clit!-2436-4

'i m SQNP SI-114.1 APPENulX A Page 6 of 9 ' TABLE 11 (Continued) ' Rev 1 SEQUOYAH INSERVICE INSPECTION PROGRAN' CLASS B COMPONENTS Program Sec' tion XI 40 Yr. Sample 1st Insp. Inthrval Reference . Exam . Exam No. I.cngth No. Lengtle . Inspection Periods . Reference Component Section' Method . Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. ;l0 yrs. Dwg. No. .D. Piping (cont'd) c., Main Steam . System 8.9.1.3 NT C-C 11 11 3 4 4 CllH- 1438-f

d. Feedwater System 8.9.1.4 MT C-C 14 14 4 5 5 CitH-2439-0

e. Containment Spray System 8.9.1.5 PT C-C N/A See Program Section 8.3.1.5 CllH-2440-C

f. Upper llead Injection 8.9.1.6 PT - C-C 4 4 1 1 2

%p 2. Component Supports '7

a. Residual lleat Removal ,

System 8.9.2.1 VT-3, C-E 44 44 14 15 15 Clift-2435-C , VT-4

b. Safety -

Injection System 8.9.2.2 VT-3, C-E 67 67 22 22 23 CilH-2436-C VT-4

c. Main Steam System 8.9.2.3 VT-3, C-E 54 54 18 18 18 CIIH-2438-C VT-4 ,

9 e G SQNP' SI-ll4.1; APPENDIX A Page 7 of 9 TABLE D (Continued) Rev 1 SEQUOYAll INSERVICE INSPECTION PROGRAM CLASS B COMPONENTS Program Section XI 40 Yr. Sample. 1st Insp. Interval Reference Exam Exam No._ Length Nc. Length Inspection Periods Reference Component Section Method Category Welds /of Weld Welois/of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. D. Piping (cont'd)

d. Feedwater System 8.9.2.4 VT-3, C-E 22 22 7 7 8 CilH-2439-f VT-4 ,

e. Containment Spray .

System 8.9.2.5 VT-3, C-E 5 5 1 2 2 CHH-2440-t VT-4 .

f. Upper IIcad Injection System 8.9.2.6 VT-3, C-E 7 7 2 2 3 N VT-4 W

3. Circumferential and Longitudinal A. Residual lleat Removal System

a. RilR Cires >\" Nom. Wall Thickness 8.9.4.1 UT,PT C-F,C-C 36 9 3 3 3 CIIM-2336-C

b. RilR Long. >\" Nom. Wall Thickness 8.9.4.1 UT,PT C-F,C-G 9 3 1 1 1 CIIH-2336-L

c. RilR Cires { " Nom. Wall Thickness 8.9.4.1 PT C-F,C-G 105 27 9 9 9 CIIM-2336-0

d. Rl!R Long. f\" Nom. Wall Thickness 8.9.4.1 PT C-F,C-G 40 10 3 4 3 CilH-2336-C

.SQNP. SI-114.1 APPENDIX'A Page 8 of 9 TABLE B (Continued). Rev 1 SEQUOYAll INSERVICE INSPcCTION PROGRAM CLASS B CONPONENTS Program Section XI 40 Yr. Sample 1st Insp. Interval Reference ' Exam Exam No. Length No. Length Inspection Periods Reference -Component Section . Method Category . Welds /of Weld Weliis/of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. D. Piping (cont'd) B. Safety Injection System

a. SIS Cires') " Nom. Wall Thickness 8.9.4.2 UT,PT C-F,C-G 31 8 3 3 2 CIIrl-2333-4

b. SIS Cires.f " Nom. Wall Thickness 8.9.4.2 PT C-F,C-G 82 21 7 7 7 Cllti-2333-i

c. SIS Long. f " Nom. Wall Thickness 8.9.4.2- PT C-F,C-G 14 4 1 2 1 CilN-2333-t C.  ?!ain Steam System

a.  !!S Cires >h" Nom. Wall Thickness "O

s sa 8.9.4.3 UT,tlT C-F,C-G 46 12 4 4 4. C1111-2340-t

b. tlS Long. >h" Nom. Wall Thickness 8.9.4.3 UT, tit C-F,C-G 11 3 1 1 1 Citti-2340-t l

D. Feedwater System

a. FW Cires > " Nom. Wall Thickness 8.9.4.4 UT,tlT C-F,C-G 34 9 3 3 3 Clit!-2339-t

b. FW l.ong. >4" Nom. Wall Thickness 8.9.4.4 UT, TIT C-F,C-G 1 1 0 0 1 C1111-2339-i E. Containment Spray System

a. CS Cires { " Nom. Wall Thickness 8.9.4.5 PT C-F,C-G 10 3 1 1 1 C1111-2422-1

b. CS Long. ( " Nom. Wall Thickness

-8.9.4.5 PT C-F,C-G 3 1 0 0 1 C!It!-2422-t s ..5' ~ L ,.SQhP SI-114.1 APPENDIX A Page 9 of.9 TAlli.E 11 (Continued) Rev 1 SEQUOYAll INSERVICE INSPECTION PROGBAf! CIASS 11 C0fflVNENTS - Program Section XI 40 Yr. Sample. 1st Insp. Interval Reference Exam ~ Exam No. Length No. Length Inspection Periods Reference Component Section -Method Category Welds /of Weld Welds /of Weld 3 yrs. 7 yrs. 10 yrs. Dwg. No. D. Piping (cont'd) F.' Upperhead Injection System

a. UllI Cires >!". Nom. Wall Thickness

8.9.4.6 ITf,PT C-F,C-G' 26' 7 2 2 3-E. Pumps

1. Residual lleat Regaoval (2) RIIRP

a. RilRP Support Components 8.9.1.2 VT-3 C-C , C- E 2 2- 1 0 1 4

F. Exempted . i -Q Components _ 8.12 VT-2 C-Il See Program Section 8.12 9 i r...,.,,, __.. ~ k i SQNP SI-114.1 Page 1 of 3 Re r 1 APPENDIX A TABLE C LIST OF DRAWINGS - UNIT 1 Reactor Vessel [ Appendix A . Drawing No. Title Page No. .CH-M-2333-B. Reactor Coolant Piping /19- 12o CH-M-2337-C _ Upper Head Injection System /4/ - /41 CH-M-2341-B _ Reactor Vessel Stud Locations and Details y2- 43 CH-M-2343-B Reactor Vessel Seam Welds yy CH-M-2358-A ' Reactor Vessel Closure Head 55' CH-M-2359-A Control-Rod Drive Housing f4-17 .CH-M-2360-A . Reactor-Vessel Inlet Nozzles TT CH-M-2651-C CRD, UPI and Vent Pipe Penetrations 17-/80 101 CH-M-2361-A- Reactor Vessel Outlet Nozzles ISI-0014-A- Auxiliary Head Adapter . 10 1 ISI-0016-A

• Reactor Vessel Clad Patches (PSI only) 103 MSG-0001-B Closure Head Cladding Patches (PSI only) 104 MSG-0004-C' Reactor Vessel-Bottom Head Penetrations for Presurizer Appendix A

Drawing No. Title Page No.

CH-M-2362-A Pressurizer Support Skirt Weld - /08 CH-M-2663-A Pressurizer Seam Welds 107 MSG-0002-B; Pressurizer and Steam Generator Cladding Patches for n MSG-0006-A Pressurizer Heater Penetrations 109 Steam Generators Appendix A Drawing No. Title Page No. Reactor Coolant Piping Ilf" # ##

CH-M-2333-B

'CH-M-2345-B- Steam Generator llo 76M1 - Vertical Steam Generators (Tube Sheet Arrangement) 111 MSG-0002-B -Pressurizer and Steam Generators cladding Patches . MSG-0005-A' Steam Generator /Feedwater Transition Spool Piece ill 9 79 F SQMP _ SI-114.1 Page 2 of 3 -Rev 1 APPENDIX A TABLE'C (Continued) LIST OF DRAVINGS - UNIT 1 Heat Exchangers Appendix A Drawing No. Title , Page No. CH-M-2404-A Residual Heat Removal Heat Exchanger Channel Il3 Welds ISI-0066-A Regenerative Heat Exchange- f;y- f f f ISI-0067-A Excess Letdown Heat Excha. //7 ISI-0068-A . Letdown Heat Exchanger fit Piping and Valve Weld Maps Appendix A Drawing No. Title Page No._ CH-M-2333-B Reactor Coolant Piping (Main Loops) lli- /2 O I1!-11 7 ~ CH-M 2333-C~ Safety Injection System ~ CH-M-2334-C Reactor Coolant System 11f - 13 4 CH-M-2333-C Chemical add Volume Control System ~ / J r- I ? f CH-M-2326-C Residual Heat Removal System /37-/40 CH-M-2337-C Upper Head Injection System / # / -14/ ' - CH-M-2338-C Seal Water Injection (Chemical and Volume 144 - /Y7 Control System) CH-M-2339-C Feedwater System l u t - /4f-CH-M-2340-C Main Steam System' I fo - 15~l - CH-M-2422-C Containment Spray . System / 3'l Piping and Valve Hanger Maps Appendix A . Drawing No. Title Page No. CH-M-2432-C Reactor. coolant' System iri - /rr ' CH-M-2433-C- Chemical and Volume Control System /60 - /4/- CH-M-2434-C Seal Water Injection (Chemical and Volume //A - /4r Control) , CH-M-2435-C_ Residual Heat Removal System /44 - /(7 CH-M-2436-CJ Safety-Injection System /6f- 174 CH-M-2437-C Upper Head Injection System - /77-/71 CH-M-2438-C Main Steam Line' System /to - /t/

CH-H-2439-C Feedwater System / r2 -lT3

- CH-M-2440-C Containment Spray System fr4 r . To SQNP SI-114.1 Page 3 of 3 Rev 1 APPENDIX A TABLE C (Continued) LIST OF DRAWINGS - UNIT 1 Pumps Appendix A ' Drawing No. Title Page No. CH-M-26/5-B Reactor Coolant Pump Main Flange and Lower /ff' Seal House Bolt Pattern MSG-0003-B' Reactor Coolant Pump Casing /T4 ISI-0099A Residual Heat Removal Pump Supports /r7 Tanks Appendix A Drawing No. Title Page No. , ISI-0069-A- Boron Injection Tank (Unit 1) /71 ISI-0070-A Upperhead Injection Water Accumulator /P?- /90 ISI-0071-A Accumulator Surge Tank /f/ ae I e P e . g -> O* 9 e i f/ .. =

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• SEQUOYAll NUCLEAR PLANT UNITS #18 #2 -

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2. AZIMUTHS ARE TAKEN OFF OF AZIMUTH O* a 300 / REACTOR CENTERLINE.

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REFERENCE DRAWINGS V-8X20WDS6X2-E NOTES

l. THE THREE PUMP FEET ARE BOLTED TO A COMMON SUPPORT,

2. PUMP SUPPORT NUMBERS RH R PH-I A-A RHRPH-18-B RHRPH-2A-A RHRPH-28-8 1 '

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( TYP. ) BIT-I+ 1 r BIT-2 - CLASS 0

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10. SA 403 WP 316 l y.] -

ii G" Scil 160 i I n a '/ l t n  : i s i i l100* 1 BITH-2 BITil-3 I i

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( _ _ __ - N O., DATE REVISIONS CK'D APO ENESSEE v'AttW AUHf0RHY BIT-G-> BIT-5 Divistof t or HUCitAR FOWIR g 8#~IO4

                                                                                  ,      ,[                     -i                                               SEQUOYAH NUCLEAR PLANT UNIT #l
                                                                                                -SIS-179                                                              BORON INJECTION TANK                                            .

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Page 98 of 100 , s t A sw..A (k e / M Rev 1

• REFERENCE DRAWINGS ...}

                                                                    .                              HWE-51498 i

3 1/8" THK. UHWA-I [UHWAH-l -_._

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NOZZLE A O UHWh-7

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l N ANWAY / UHWA-8 N / N / N.) N 3 f .- 5'/8" THK. - MATER!AL SPECIFICATIONS l SAFE ENDS -~~' urn:EssCE VAU EY AUTl!OhllY CLASS B l SA 182 F 304 _ _ . _ . . . _ _ . __ _ 8 HEADS AND SHELL l 1/8" THK. SEQUOYAH NUCLEAR PLANT SA 516-71-70 NOTES UPPENEA 308 CLADDING INJEC JON I, NOZZLE A IS AT 270* WATER ACCUMULATOR NOZZLES NOZZLE B IS AT O W SA 350 LF 2 NOZZLE C IS AT 180' MANWAY IS AT 9O'

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                     -;                                        SI-114.1                                                                  -

Page 100 of 100- (g~~ Rent 1 REFERENCE DRAWINGS NW 5751 MATERIAL SPECIFICATIONS , CLASS B SHELL SA 240 TP 304 II'8" THK. HEADS I" THK AST-1 AST-2 I

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O gNAME PLATE-C3 h J-ASTH-1 ASTH-2 NO~. DATE REVISIONS CK'O APP. TENNESSEE VALLEY AUTHORITY esyt13ON OF NUC1 EAR POWER j SEQUOYAH NUCLEAR PLANT UNITS #I B #2 ACCUMULATOR SURGE TANK xma. NTi -. mars 7-14-82___ 33 ------ - ------ wi, i e, a im ggygi . --___- --_---. ISI-OO71-A

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                                                                                                                    . Rev 1                                          'Page 1 of 2 APPENDIX A TAMI.E D Valves. Subject to the Requirements of Examination Category _ B-N-2 of Talste - lW8-2500-1 of Section XI Code-' Valve Piping Valve Valve. Valve Group TVA Dwg No. Vendor                                                         Material          Valve     . Forging /.

Valve No. Class Cat. System Size- Type Act No. (Weld flap) Dwg No. Venelor Spec. Function Casting _ FCV-63-67 A .8-PAS SIS 10". Cate 'Ho I - Cllti-2333-C 88926 Velan ASTM A182 Forging FCV-63-80 A. B-PAS SIS 10" Cate no. I Cllti-2333-C 88926 Velan ASTil A182 Forging.

FCV-63-98 A 8-PAS. SIS 10" Cate 110 1 Clift-2333-C 88926 Velan ASTil A182 Forging FCV-63-il8 A. B-PAS SIS 10" Cate tK) 1 Citti-2333-C 88926 .Velan ASTrl A182 Forging 63-560 A AC-Act SIS 10" Ck SA '2 - Cllti-2333-C 94-12892- Da rl ing ASTN A5168 PSiv 63-561 .A. AC-Act SIS 10" Ck SA 2 Clift-2333-C 94-12892 Ila rling ASTII A5168 l'Siv 63-562- A AC-Act SIS 10" Ck SA 2' Cllti-2333-C 94-12892 Darling ASTil A5168 PSiv 63-563 'A AC-Act SIS -10" Ck SA. 2 Cliti-2333-C 94-12892- Darling AST!! A5168 PSIV 63-622 A AC-Act SIS 20" Ck SA 2 Cilfl-2333-C ' 94-12892 Darling ASTH A5168 PSIV

       -'      63-623       A           AC-Act SIS               10"   Ck    SA'    2          Cllti-2333-C 94-1289              Darling             ASTil ASI68 PSIV

{} 63-624- A AC-Act SIS 10" Ck SA 2 Cllti-2333-C 94-12892 liarling AS1tl A5168 PSIV 63-625 A AC-Act SIS 10" Ck SA 2 Clitl-2333-C 94-12892 Darling ATnt A5168 PSIV 63-640 A AC-Act SIS /RIIR 8" Ck. SA 2 Cllti-2336-C 94-12892 ' Da rl isig ASTil A5168 l'S I V 63-643 A AC-Act SIS /RilR 8"- Ck SA 2 Clitl-2336-C 94-12892 Ilarling ASTil A516: p3gy 63-558 A AC-Act SIS 6" Ck SA 3 Clift-2333-C 78704 Velan ASTM A182 PSIV Forging 63-559 A g AC-Act SIS 6" Ck SA. 3 Cllti-2333-C 7870'e Velan ASTil A182 PSIV For-ing 63-632 A .AC-Act'Els 6" Ck SA 3 Cllli-2333-C 78704 Velan A.STtl AIS2 PSIV 15 ging 63-633 A AC-Act SIS 6" Ck SA '3 Clifl~-2333-C 78704 Velan ASTPI A182 PSIV Forging 63-634 A AC-Act SIS 6" Ck SA 3 Cllti-2333-C 78104 Velan ASTil A182 _ PSIV Forging 63-635 A AC-Act SIS 6" Ck SA 3 Cllti-2333-C 78704 Velan ASTtt A182 PSIV Forging 63-641 A -AC-Act SIS /RilR 6". Ck SA 3 Cllti-2336-C 78704 Velan ASTM A182 PSIV Forging 63-644 A AC-Act SIS /RilR 6" Ck SAJ 3 Clit!-2336-C . 78704 Velan ASTM A182 PSIV Forging 68-563 A C RCS 6" Rel SA . 4 CilN-2334-C 1151688 Crosby ASTN A182/ See A351 2 Note 2 I 1 _j.

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                                                                                     .                        SONP                  Page 7 of 2 S1 ,114.I'                                :j APPENDIX A                        Rev.I
                                                                              . tan!.E D Valves Subject to the Requirements o'f Examination                                    .

Category R-N-2 of Table IWH-2500-1 of Section XI CodefValve Piping Valve balve Valve Group TVA Dwg No. Venelor Material Valve Forging /-

         ' Valve No. Class ' Cat. System Size Type Act- No.                (Welal Hap) Dwg_No. Venelo r             Spec. Function Casting' 68-564      A       C       RCS  ' 6" ~     Re1    SA-   4       Cliff-2334-C : 1851688   Crosby           ASTN AI82/             See A3512                  Note 2' 68-565      A       C       RCS    6"       Rel    SA ' '4     'CilH-2334-C 1151688       Crosby           ASTil A182/            See A~1512                 Notez FCV 74-1    A'     .A-Act 'HIIR    14"      Gate NO     l5 ,     Clitl-2336-C E-1-144831 Copes-Vulcan ASTH'A182 PSIV.             Forging FCV-74-2    A       A-Act EllR     14"      Cate 110     5       Clitl-2336-C E-1-144831 Cupes-Volcan ASTil A382 PSIV             forging 87-558      A       AC-Act UllI    8"       Ck    ,SA    6      . CIIH-2337-C H-148885 - Copes-Vulcan ASTH AI82 PSIV              Forging 87-559      A       AC-Act Ull!    8"       Ck     SA   16       Cllti-2337-C 11-148885   Copes-Vulcan AS1H AI82 IGIV             Forging 3         87-560      A       AC-Act IIIII   8"       Ck     SA    6       Cllti-2337-C ' n-14SH85  Copes-Vulcan' ASTM AI82 PSIV            Forging L->       87-561      A       AC-Act Ull!    8"-     .Ck     SA    6       Clitl-233 7-C   H-148885 Copes-Vulcan ASTri A182 PSIV            Forging 87-562      A       AC-Act Ull!    ]2"      Ck     SA    6.      Cllti-2337-C R-148885    Copes-Vulcan ASTil AI82 PSiv            Forging 87-563'     A       AC-Act Ulli    12"      Ck     SA    6       CilH-2337-C 11-148284    Copes-Vulcan ASTri A182 PSIV            Forging
         . NOTES:

8 Seal plate manufactured to ASTH A240 F304. , 2 Nozzle manufactured to ASTH A382 F316 - Forging Body manufactured to ASTH A351 CR8H - Casting f

                                                                                                       .                                                6 b

l '. ] f SQN7 SZ-114.1 i Page 1 of 2 Rev 1 APPENDIX B TABLE A LISTS OF CALIBRATION BLOCK DRAWINGS Appendix B Block No. Drawine No. Pare No. SQ-1 CH-M-2445-C (ToBf A DOED ' SQ-2 CH-M-2445-C SQ-3 CH-M-2445-C LJrLx) SQ-4 CH-3-2445^. SQ-5 CH-M-24'..-C SQ-6 CH-M-2460-B SQ-7 CH-M-2507-C SQ-8

• CH-M-2507-C '

SQ-9 CH-M-2507-C

                    .SQ-10                           CH-M-2553-C SQ-11                 .         Ch-M-2593-C SQ-12                           CH-M-2553-C SQ-13               -

CH-M-2571-C SQ-14 CH-M-2550-C SQ-15 CH-M-2552-C C SQ-16 CH-M-2591-C SQ-17' .CH-M-2571-C SQ-18 CH-M-2571-C , SQ-19 CH-M-2571-C SQ-20 CH-M-2589-B SQ-21 CH-M-2571-C SQ-23 .CH-M-2597-C SQ-24 CH-M-2597-C SQ CH-M-2597-C - SQ-26 CH-M-2597-C SQ-27 CH-M-2597-C . SQ-29 CH-M-2591-C SQ-30 CH-M-2591-C SQ-31 CH-M-2597-C

                   'SQ                           CH-M-2597-C SQ-35                           CH-M-2597-C SQ-36                           CH-M-2597-C-SQ-37-C                         ISI-0006-A SNP-ET                       CH-M-2446-B r

' 7-1-8-CS-3-WAB. .D 5420 032 (SWRI)

                   "105-7-8-CS-4-WAB                 D-5420 033 (SWRI) 59-1                         D-5420 037 (SWRI)

SQNP SI-114.1 Page 2 of 2 Rev 1

                . (:

I APPENDIX B i TABLE A (Continued) LISTS OF CALIBRATION BLOCK DRAWINGS Appendix B Block No. Drawing No. Page No. IR CSCL-2-WAB D-5420 031 (SWRI) I'T2 Sl A#### NS-CSCL-1-WAB. D-5420 030 (SWRI) LA Tid ) . Q-7053 1 of 2

      .                    Q-7053 2 of 2                   D-5420 035 (SWRI)

T04519

                          ~ T07216-WAT-UT-05'                      D-5420'021 (SWRI)

WAT-UT-06' 2414 SH. 7 (RDM) WBT-UT-01 D-5420 022 (SWRI) WBT-UT-02

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D-5420 026 (SWRI) WBT-UT-03 D-5420 025 (SWRI) ([- .' s

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S02P ST-114.1 Page 1 of 1 Rev 1 APPENDIX C DATA SEEET 1 SEQUOD.H NUCLEAR PLANT UNIT 1, CYCLE INSERVICE INSPECTION FINAL REPORT Reviewed by: Quality Engineering Branch, Chief

  ~

Plant Superintende'nt l74

SQNP SI-114 01 Page 1 of 1 Rev 1 APPENDIX D NOTIFICATION OF INDICATION PART I - FINDINGS NOI No. Plant / Unit Examination Report No. Component ID Drawing No. Description of Indication: (Sketch / Photograph if needed) Signature of Examiner /Ctrtif. Level Date Signature of Field Supervisor (Contractor) - Date NDE Section Represeptative Date

                               ,PART Il-DISPOSITION m

Disposition Prepared By

• Date Disposition Approved By .

Date , ,, PART III-VERIFICATION . Verification of Completed Corrective Action and/or Examination by NDE Section Representative. e Signature Date - 10 7

S03P SI-114.1 Page 1 of 22 Rev 1 (~ s. M. APPD; DIX E PIQUEST FOR RELIEF 9 4 e e 9

                        .9 I?r

I SQNP-SI-114.1 i Page 2 of 22 Rev.1 , f* R50 TEST FOR RELIEF ISI-1 Components: Reactor coolant pumps (four per unit) Class: TVA Safety Class A Function: Circulates reactor coolant Inspection Requirement: ASME Section XI, Table'IWB-2500-1, examination - category B-L-2, item No. B12.20, visual exami-

                              - nation of pump internal pressure boundary

• surface.

Basis for Relief: In absence of required maintenance, dis-assembly of a reactor coolant pump solely to perform a visual examination of internal surfaces is impractical. This would represent u an unnecessary employee exposure to high radiation and contamination areas and an excessive expense to TVA. Time required for this major task of dis-asembly, examination, and reassembly would consume at least three weeks of 24-hours per-L( day work. Radiation dose rates of the pump exterior will a.verage 100 to 300 mrem / hour, and pump internal dose rates will average 10 to 20 rem / hour. The bencfit received from this major effort is minimal considering empicyee exposure, . potential damage to safety-related equipment, and cost in dollars. In addition,,the two units at Sequoyah-Nuclear - Plant will operate under similar conditions. Therefore,-we feel that if a pump fr6m one of the units is disassembled for maintenance during a ten-year interval, the visual exami-nation performed will be representati s of the pump condition for each unit. This we,uld avod unnecessary employee exposure to the high radiation dose rates noted above. We conclude 7that if one pump is disassembled for mainte-nance during the ten-year interval, the visual examination performed satisfies examination

                                . category B-L-2 requirements for both units.
                .                Disassembly of the pump solely for visual examination is impractical.

m a

177

                       ~

SONP SI-114.11

                                                           -Page-3 of 22 Rev l'                                      i 1
            . .f '.

REOUEST FOR RELIEF ISI-1 I (Continued) ' l

                         -Alternate Insoection:   The internal surface of the reactor coolant             1 pump casing will be visually examined whenever the surfaces are made accessible when a pump.

, is disassembled for maintenince purposes. If , , during.the ten-year interval a pump from - either unit is not disassembled for mainte- . nance, a pump from one unit shall be examined- '- from the exterior. This shall be accomplished . by ultrasonic thickness measurements of the ' pump casing. L S O 5 + E-O 4* t.. ( 200

                                                                                -m,

SQ3P S1-114 01 Page 4 of 22 Rev 1 [ . REOCEST FOR RELIEF ISI-2 Comocnents: Valves exceeding four-inch nominal pipe size Class: TVA Safety Class A Function: Various functions . Inspection Reouirement: ASMI Section XI, Table IWB-25CD-1, examination category B-M-2, item No. B12.40, visual exami- .- nation of valve internal pressure boundary surface. Basis for Relief: During routine maintenance, visual exami-nations of valve body internal pressure boundary surfaces are performed and documented under existing plant administrative pro-cedures. Most Class A valves, particularly

                                -containment isolation valves, are disassembled frequently for maintenance. In addition, the two units at Sequoyah Nuclear Plant will operate under similar conditions. Il a valve
   /                              from one of the units is disassembled for
  \-                             maintenance within a ten-year interval, we feel that the. visual exanination performed would be representative of both units and would be sufficient to satisfy the examination requirements for both units for that partie-
      "                          ular valve classification as defined in exami-nation category B-M-2.

We conclude that if one valve in each group of valves of the same constructional design and manufacturer that perform similar functions is disassembled from either unit during the - ten year interval, - the visual examination performed satisfies examination cagegory B-M-2 requirements of both units. Alternate Inspection: If a valve from a particular classification has not been disassembled as the end of the inspection interval approaches, a case-by-case study will be made to determine the practi-cality of disassembling a valve from one of the units solely for visual examination (determine -if draining the vessel-would be required, etc.). If necessary, a request for relief will be issued a,t that time. (. s 59/

ia

                                       ,                    SQNP SI-ll4 01
    -                                                       Page 5 of 22 Rev'l i

(  ; REQUEST FOR RELIEF ISI-3

                        . Components:              Pressure-retaining welds in piping Class:                  :TVA Safety Classes A and B Function:                ' Pressure-retaining componen,t Inspection Recuirement: ASME Section XI, Table IWB-2500-1, examination categories B-F (item No. B5.50), B-J (item Nos. B9.10, B9.20, and B9.30), and C-F (item Nos. C5.10, C5.20, and C5.30), volumetric examination of longitudinal, circumferential, and pipe branch connection welds.

Basis for Relief: In some cases it will be impractical to inspect all welds from both sides, i.e., nonremovable hanger l interference or valve and pump . casingsadjoining the welds. These welds will be noted on the ultrasonic examination data sheets. Alternate Inspectien: In additien to the' visual examination

f. (. performed during' system leakage and hydro-

        \-    \'                                   static pressure tests, a "best effort". ultra-sonic examination and a surface examination will be performed on accessible areas of the weld (s).

0 4 D

                   -0 i

O 6 2372 i-

SQNP i S1-114.1-Page 6 of 22 Rev 1 r. REOCEST FOR REI.IEF ISI-4 Comoonent: Steam generator (four per unit) Class: TVA Safety Class B

                         -Inspection Reouirement: ASME Section NI, Table IWB-2500-1, examination category C-A, item No. C1.10, volumetric
                                                   . examination of circumferential shell welds.

Basis for Relief: One circumferential shell weld on each generator-is inaccessible due' to the upper steam generator support brackets (weld Nos. SGW-DI, -D2, -D3 and -D4). See attach'ed drawing CH-M-2345-B for veld location. Also attached drawing CH-M-2345-B for weld

• location. Also attached are drawings showing arrange =ents of the support brackets. One weld on one generator will be examined on a "best effort" basis for the baseline inspection _ intervals in.accordance'with -

IWC-2411 and Table IWC-2500-1.

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                                            .          51-114.1 Page 10 of 22 Rev 1 REQUEST FOR RELIEF ISI-5
                     -Comoonent:                 Reactor Pressure Vessel Class:                    TVA Safety Class A Inseection Recuirement: ASME Section XI, Table IWB-2500-1, examination category B-A, item No. Bl.20, 100 percent preservice baseline volumetric examination of lower head dollar weld, under conditions and eith equipment and techniques equivalent to those expected to be employed duri.ig inservice inspection.
                     -Basis for Relief:          TVA will employ autcmated remote inspection devices to examine most of the ==.ucor vessel welds. These examinations will be conducted from.the vessel inside diameter. However.

the lower head weld on each reactor pressure vessel is partically inaccessible for exami- , nation from the vessel inside diameter due to instrumentation tubes which penetrate the lower head (weld No. WO1 see attached J"- N - drawings ) . Po'rtions of the weld can be examined from one side (as permitted by T-441.4, Article 4 of Section XI) and will include 100 percent of the examination volume

                                               'in accordance with IWB-3511.1 of Section V.

These portions of the weld will be reexamined o during the inservice intervals in accordance with.the examination category B-A of Table I IVB-2500-1. Alternate Inspection: A 100 percent baseline examination of the ~ EETa~will ne conducted Tf5Hithe vessel outside diameter. This will be accomplished by performance of a manual ultrasonic examination. A remote ultrasonic examination will Le conducted frem the vessel inside diameter on all accessible areas of the weld. [ Y~' r r 4 . l ' L L 1*7 t: b- ._

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SQNP SI-114.1-Page '.3 of 22 Rev 1

 . (.

REQUEST FOR RELIEF ISI-6 Component: Steam Generator (four per unit) Class: *TVA Safety Class A Inseection Recuirement: ASME Section XI, Table Ib'B-2500-1, examination ~ category B-D, item No. B3.60, volumetric examination of nozzle inside radius section , on the primary side. Basis for Relief: Each steam generator consists of two integrally cast nozzles and two integrally cast manways. The present capability of ultrasonic testing is not sufficient to examine cast material of this thickness and achieve meaningful results. - Alternate Inspection: None m 0 0 0 2 to

              . .                                          ~                       -

SONP SI-114.1 Page 14 of 22 Rev 1 REQUEST FOR RELIET ISI-7 Component: Reactor Coolant Loop Piping Welds (2) Class: TVA Safety Class A Inspection Pecuirement: ASMI Section XI,'IWB-2500-1, examination category B-J, item No. B9.10, volumetric examination of circumferential welds. Basis for Relief: Two circumferential shell welds in the reactor coolant loop piping (RC-23S1 and '

                                       -31SI, loops 3 and 4) are located inside the reactor vessel shiel_d wall and are inacces-sible for; Abasea'nd inservice examination (see' attached-drawings). Both welds have undergone shop radiographic examinations.

Since the baseline inspection serves as a reference to future inservice inspections and bot welds will be inaccessible for inservice , inspections, the shop radiographic exami-nations coupled with the Section III hydrostatic test will provide adequate proof ( ' of integrity of the' system welds. Inservice system _ leakage and hydrostatic' testing will prove weld integrity during the life of the

                                      . plant.
           ' Alternate Incpection:     None A

O*

     .s n

!\

• 2 ll

p SQNP SI-114.1 Page'15 of 22 Rev 1 -

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L SQNP, SI-114.1 s Page 17 of 22 Rev 1 (~ REOLIST FCR RELIEF ISI-8 Component: Reactor coolant pumps (four per unit) Class ~: TVA Safety Class A ,

                     ' Inspection Recuirement: ASMI Section XI, IWB-2500-1, examination category B-L-1, item No. B12.10, volumetric examination.of pressure-retaining welds in r

pump' casing. . Basis for Relief: Each reactor coolant pump cas'ng i consists of a two-piece welded type 304 SST casting.. The present capability of ultrasonic testing is

                                                         ~

not sufficient to examinate cast material of this thickness and achieve meaningful results. Alternate Insoection: All four velds will be surface examined for the preservice baseline, and one' veld will be surfice examined during each inspection inte rval .

              - (-

f 6 8 e e 9 Y t O

              ^ .
                                                            . ;2 ' Y -

(_

SQNP SI-114.1 Page 18 of 22 Rev 1 l'. REQUEST FOR RILIEF ISI-9 Component: Uncladded vessel welds in ferritic material less than two inches in thickness. Class: TVA Safety Classes A and B Inseection Recuirement: -Ultrasonie. examination of welds, paragraph T-530 of AS>E Section V, Article 5,1977 ' l Edition, Summer 1978 Addenda as referenced in paragraph IWA-2232 (c) of AS?E fection XI ' 1977 Edition Summer 1978 Addenda. Basis for Relief:  ; Paragraph T-533.2(a) of Article 5 of AS!E Section V requires that the basic calibration block include a basic calibration hole I-drilled parallel to the contact surface. However, paragraph T-533.2(b) permits the use of other calibration reflectors provided equivalent responses to that from the basic calibration hole are demonstrated. TVA currently.uses five percent notches in (; lieu of side drilled holes. Although the use s of-the five' percent notch cannot be shown to be. equivalent in all cases to.the applicable side-drilled holes, TVA considers,that , examinations are technically acceptable based

                        -                             on the calibration requirements of paragraph III-34.30 of Appendix III to'the 1977 Edition,
                                             ;SummeryN978 Addenda of AS?E Section XI. .The

_ calibration notches for ferritic material are y ~10*.t when t is less than .312" and

                                                        .104t .009t2 for material .312"-6"' thick.

TVA* s use of ;five percent notches.is , considered equivalent'to the latest approved code examination techniques. '

                     . Alternate Inspection:        TVA proposes to continue the use of notches b

located on the I.D. and 0.D. surfaces at a nominal depth of 5%t as reference

                                              - reflectors.

f d

              .. e .

_____.__=_::..------

SQNP SI-ll4.1 Page 19 of 22 Rev 1 REGCEST FOR RELIEF 151-10

                                                            ' Comoonents:                         Reactor vessel flange to upper shell weld Class:                             TVA Safety Class A Inspection Recuirement: ASME Section XI, Table IW'-2500-1, s

y/' examination category B-A,. item No. Bl.30, volumetric from flange face Bisis for Relief: The reactor vessel flange-to-upper shell weld - is located behind the core barrel and is therefore inaccessible until the core barrel is removed. The vessel flange-to-upper shell weld is 41.9 inches below the flange face.

                                                                                               Due to the location of the vessel 3

flange-to-upper shell weld, TVA intends to address the weld as a reactor vessel shell weld. c We have reviewed the Sequoyah Reactor Vessel Stress Report entitled Analysis of the Main Closure Including Core Support Ledge (Document No. 30616-1105) purposely to ( ' determine a fa'tigue usage factor for the vessel flange to shell weld. This analysis does not provide a usage factor specifically for the weld because the analysis considers

• weld and base material to be homogeneous and equal in elasticity, strength, and fatigue properties. Instead, the analysis provides usage factors at critical locations.

The maximum fatigue usage factor in the

                                                                                '               vessel in the vacinity of the flange to shell
                    ~

weld as found in the above anlaysis is 0.00662 and this value can be conservatively used for the weld. We consider the valuve of 0.00662 to be extremely low compared to the code allowed fatigue usage factor of 1.0. We, conclude

• that the distance ( ") from the flange face to the flange-to-upper shell weld coupled with present ultrasonic techniques and the very low fatigue usage factor that the, flange-to-upper shell weld should be treated as a reactor vessel shell weld.

Alternate Inspection:

                                                                                               'A. remote ultrasonic examination of the weld
                                                       '                          ~.
                                                                                      '        Ewill be, conducted from the vessel inside

( _ ' diameter near the end of the inspection

                                                                                              . interval.

s - w T D 1 4 s

                                                ~

s

SQNP SI-114.1 Page 20 of 22 Rev 1 (s P2 QUEST FOR FILIET ISI-11 Comnenents: Feedwater Piping System and Associated Supports Class: TVA Safety Class B Inseection Recuirement: IE Bulletin 79-13, Revision 2, dated October 16, 1979, was issued to provide inspection requirements for part of the . feedwater piping system and associated supports to monitor the feedwater pipe cracking problem. Item 1.a. requires the TVA perform radiographic (RT) examination, supple =ented by ultrasonic examination as necessary to evaluate indications, of all feedwater nozzle-to-pipe welds of adjacent pipe and nozzle areas (a distance equal to at least l two wall thicknesses). Evaluation shall be in accordance with ASME Section III, Subsection NC, Article NC-5000. Radiography ( shall be perfo'rmed to the 2T penetrameter sensitivity level in lieu of Table NC-5111-1 with systems void of water. TVA met this requirement after hot functional testing for Sequoyah units 1 and 2.

      ~

j~ f; Item 1.c. requires that TVA perform visual inspection of feedwater system piping > supports and snubbers in containment to' verify operability and conformance to [ design. TVA also met this requirement after - hot functio'nal testing. k Item 2.a of actions to be taken by licenses in IE Bulletin 79-13, Revision 2, dated October 16, 1979, requires during the first refueling outage that TVA perform volumetric examination of the feedwater nozzle-to pipe welds, the feedwater piping welds to the first support, the feedwater line-to-contain-ment penetration welds, the line one pipe in diameter downstream of auxiliary feedwater to main feedwater connection. Also, item 2.c. requires that TVA perform a visual inspection of all feedwater system piping supports and

  '                                                   snubbers in containment to verify operability

( and conformaace to design. Theje requirements were imposed before the cause of the feedwater line cracking problem was known.

/7

                                                                                                               ~-

a_ SQWP SI-114.1 Page 21 of 22 Rev'l

                     . f-REQUEST FOR RELIEF ISI-11 Pasis-for Relief:                       TVA requests relief from examining the feedwater piping-welds to the first support,
                                                            '"                                     the feedwater line-to-containment penetration welds, and the main feedwater line one pipe diameter downstream of the auxiliary feedwater to main fe2dwater connection. As
                                                                                               ' required by item 2.a.           TVA proposes to
                                                                    .m                           perform a volumetric examinatic. m the TT' feedwater nozzle-to pipe welds in performing visual inspections.of all feedwater system
)                                                                                                piping supports and snubbers in containment

[. , other than required by ASME Section XI.

.4

 '"                                                                                              At the-time,IE Bulletin 79-13 was issued, the real cause of the feedwater pipe cracking

(, problem was not fully understood. As a result, TVA believes that a request for . relief from these, requirements is warranted. Our justification is as follows:

1. The Westinghouse Owners Group's efforts

                   .(-
                                                                                                 ~,
                                                                                                   ~ ~' which are ' documented in WCAP-9693, Investigation of Feedwater Line Cracking 1;,ff)                                                                                                  "in Pressurized Water Reactor Plants, dated June 1980 provided information to
           ?                                                                                             substantiate that the feedwater pipe
          ?                                              ,

cracks are fatigue failures which are caused by thermal stratification and thermal st' riping during low-flow rate feedwater injection. The effects of thermal stratification and striping are enhanced by temperature difference ('

                                                                                                                     ~

between cold feedwater and hot steam generators. .

2. TVA believes there is a basis for the

                                                 '.            f
                                      .g -

item 2.a. volumetric examinations of the feedwater nor:le-to pipe welds because 4 thermal stratification and striping are e

                                             ' ~
                                                                                  '                      known to occur at this location; however,
              'W,                                                                                        now that the'cause of feedwater pipe       -

cracking is known, TVA does not believe 'g%i- _ there is a basis for requiring volumetric

                               'T                                    ^

examinations of the remaining feedwater

3. piping in item 2.a. because the thermal and geometric condilions do not exist at I

these locations to support the mal strati-

                 >b-                                                                                     fication and striping. In addition, aring a telephone conve:rsation with NRC
                                                                                                           - September 2, 1982, TVA questioned why
                                                        ...                                                  : IE Bulletin required RT examinations y

Irlpr -

h. 5 the remaining feedwater piping. -

   "j,?$   w :.y              , ',

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                                . g.

SQNP SI-114.1 4 Page 22 of 22 Rev 1 l6

                                                                      . REQUEST FOR RELIET ISI-11 Bill Crowley, of NRC,1 stated that this requirement was for purposes of
                                                                                  " upgrading" the construction radiographs of this requirement piping from 4T sensitivity to 2T sensitivity.

Mr. - Crowley indicated that other plants-which.have performed this " upgraded" RT-a c._ _ in-compliance.with the IE Bulletin had -

                                                                                ~ found instances of flaws that required repair;. however, the feedwater piping in
'                                                                                question for Sequoyah units 1 and 2 was
                                                                                -in accordance with ANSI-B31.7, 1969 Edition with 1970 Addenda, and was RT examined in accordance with TVA process specification 3.M.2.1(d) which~ requires' 2T sensitivity and acceptance standards identifical s,ith NC-5000 of ASME
                                                                               'Section III. This was verified through discussions with personnel who performed the RT.during construction and an independent review of the site

~

                 .(
                   ~                                                            = radiograph's . -It is the,refore TVA's position that'no " upgrading" of the feedwater piping welds is required for Sequoyah units 1 and 2.              -

Based on the above justification, TVA concludes that.the subject. request for relief

   ~

does not affect safe operation of the unit.

Alternate' Inspection:

                                                                          . Perform . volumetric examinations of the
                                                                          'feedwater noznle-to-pipe welds in accordance with item 2.a. of IE Bulletin 79-13.

4 W4+ e A t t'.

                                              ~

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       , e ,

SQNp SI-114.1 Page 1 of 10 Rev 1 l* A t APPENDIN F AUG!$.NTED INSPECTIONS k . t O 6 0 ( l 2 IP

r- ~ r SQXP Page 2 of 10 L- ~

                  .'*****""'                 a SI-114.1             Rev 1 esnet mus co.enz::..vr L18          8 :/ 1 :209              806 Memorandum                                                           rexxtsste vatter acraontry To               - L. 4<.- M111s,1 Manager, Nuclear Licensing, 400 CST 2-C l'linM - :        H. J.~ Green, Director of Nuclear Power, 1750 CST 2-C unt        :

DEC 13 $32 Friancp SEC'JCYAH NUCLEAR PLANT UNIT.1 - AUGMENTED INSPECTION CF TEED'4ATER SYSTEM P: PING AND SUPPORTS - DPM NO. Seq 80M7 i / -

References:

1 'IE' Bulletin No. 79-13, Rcrisicc, 2

2. Letter frem R.'C.-Lewis, NRC, to H. C. Parris dated October 20,.1982 (A02 821022 004)

                                   .A radiographic examination was perfor=ed on the feedwater no:sle-to-pipe

~ welds',:the transitioa-piece, and the transition piece-to-pipe welds and adjacent base metal equal to twc wall thicknesses. Also, a radiographic

                                  ~ examinatien was perfor=ed on an area of base =etal of the =ain feedwater line'beginning at the centerline of the auxiliary feedwater to =ain
                                   .feedwater weld and extending downstrea= one-pipe diameter. Evaluation of the radiographs =was done in accordance with ASME Section III, Subsectien NC, A-ticle.NC-5000,-to 2-27 quality level in: lieu cf table NC-5111-1. A review of the radiographs revealed no indication of cracking'.

The construction radiographs of-the feedwater piping welds to the first " :supporti-excluding those listed-above, were reviewed for cenpliance with the. bulletin'. The radiographs' met the bulletin require =ents (i.e., 2-27

                                  . quality' level, weld quality ~and technique) except in some instances where

1he base'=etal: coverage of two-wall thicknesses was not attained on both sides-of,the weld. -Since-the thermal and-geometric-conditions do not exist-at.these locations to' support stratification and striping and the radiographs were evaluated to the 2-2T' quality level-no additional .

radiography'uas performed on'thesc uelds. f Secause of -the limited access and the configuration of-the feedwater line-ta-containment penetration welds an ultrasonic examination was perfor=ed in lieu of. radiography. The ultrasonic exar,ination was performed in acecrdance,with'TVA' procedure N-UT-18, Revision 2, which meets the requirements set forth by the ASME Boiler and Pressure Vessel Code,

                       ,           Section XI,1977 Edition,15tn=:er '1978 Addenda. The ultrasonic examination
                                 - revealed no indication of. cracking.

The radiographic and ultrasonic examination reports will be retained in per=anent' storage at.the plant site. H. J. Green

                                 -RCP:DAP JRB:3XC .

cc: NL ARMS ,1520. CST 2-C 4?

         ' ' kh This was' prepared principally ty acger Bentley.                                                                       '

i 'a ' A7AUCM.DC' j% ,

                 ,: o e _ . -_ -_ y .. . . -i M m .J.      :"  r 22.l m,-                               - - - - J u r n 'E N TT32'E

_ _ _ _ _..___._____________._________________________.____i

SQEP SI-114.1 Page 3 of 10 Rev 1 STEAM GE!;ERATOR TU3E INSERVICE ItlSPECTIO!; REPORT " SEOUOYAH !;UCLEAR PLAf;T UNIT 1 CYCLE 1 REFUELIt'G OUTAGE - SEPTE!'BER 1982 - e 9 9 0 e si ' 2 2 ?- .

                                                                                                                ,,'                                      . SQ?iP SI-114.1 Page 4 of 10

[8: Tb;TS TA3' E I Inspection Su :ary

                                . TABLE II -- Inspection Su ary S/G 1 TABLE III - Inspection Su::ary S/G 4 TABLE IV .- Tubes Recording Damage in S/G 1 TABLE.V '      -

Tubes Recording Damage in S/G 4-NF ' - FIS"LE 1 -- Osnt Gize Fr quency Distribution for S/G 4 Tubes Tested in 1931 and 1982 l I

                                                                                                                                                                                      .      C r

a f 4

  .hm 8

g 5-

  ,       en 9                                                                                                                                                                         %              ,

223 - K .. . , . . . . .. . .... . ___ - - - - - - - - - - - - - - - - - - -- .

SQNP .

                                                                .Riv 1
                                           .S1-114.1                ;g.AHLE I
                                           'Page 5 of 10
                                          ' ItiSPECTIOtl 

SUMMARY

- SEQUOYAll tNIIT I1, Steam Generators f1 s #4
                                                                                       . S/G fl            S/G 44 Totals   Inlet Outlet         Inlet  Outlet Tubes In Steam Generator                                                                                             ~

(Straight Lengths) ~ 13,552' 3388 3300 ~ 3380 3388 Total Tubes inspected (Straight Lengths) 525 267 0 220 38 i Tubes Recording Severe Damage ( Approximate Hal1 Loss 40% & Grea te r) 0 0 0 0 0

• Tubes Recording Moderate Damage g (Approximate Wa ll Loss 20% to 39%) 8 3* -

., . b' 0 5* 0 t Tubes Recording Minor Damage (Approximate Wa11 Loss Less Than 20%) 29 16 0 12 1 S/G ll:

• 2 tubes with lane blocker damage unchanged since 198L I tube recorded indication unchanged from. baseline inspection S/G f4:
• all 5 tubes recorded indications unchanged from baseline inspection 4

4 e V n v b

 ;     . c.

u. SQNP.

SI-114.1 Page 6 of 10 TABLE II Rev 1

                                                                                         .s -

INSPECTICM

SUMMARY

( Sequoyah Unit #1 - Steam Generator #1 Total Hot Lee Cold Lee

     - Tubes In. Steam Generator (Straight Lengths)                              6776        3388        3385 Tubes Inspected
       -(Straight Lengths)                               267         267           0 Tubes Recording Severe Damage (Approximate Wall Loss 405 a Greater)              0           0          -0 Tubes Recording Moderate Damage
                                    ~

(Approximate. Wall Loss 20% to 39%) 3 3* O Tubes Recording Minor Damage (

     - (Approximate-Wall Loss Less~Than 20%)

16 16 0' Tubes Inspected For Row #1.U-Bend 184 92 92 Additional Inspected By Hand Probing 6 6 0

               '2' tubes with lane blocker' damage present'in 1981 1 tube-recorded. indication at baseline and is unchanged

• 1 e

t O 21T -

                                                                                                              . . . .   ~

SQNP

                                                                  -SI-114.1 Page 7 of 10 -
                                                                   'Rev 1 TASLE III b

_I N S P S C T I C ::'

SUMMARY

! g Secucyah Unit il'- Steam Generater #4 i Total Het Lee Co'id Lee Tubes In Steam Generator (Straight Lengths) 6776 3388 3338 .

        . Tubes Inspected:

(Straight Lengths) 258 220 38 Tubes Recording Severe Damage (Approximate Wall Loss '40% & Greater) 0 0

                                                                 ,.                                       0 Tubes Recording Moderate Damage (Approximate Wall Loss 20% to 39%)
                          ~

5 5* O

           ~ Tubes Recording Minor Damage i

(Approximate Wall Loss Less.Than 20%) 13 1,2 1 ( Tubes Inspected For Row fl U-Send S. 8 0

• f.

                    .A11' indications'were recorded by baseline inspection and are unchanged.                        .

4 i

p. .

O e 224 -

                                                                                                            . . w. .:_. n . -_ _ _               -.

SQNp Rev 1 Ej. ELE IV SI-ll4.1 Page 8 of 10 TUSES RECORDING DAMAGE IN STEAM GENERATOR s1 Sequoyah Unit pl

         ' Tube i                Defect R  .O       Oriein      %         Location Notes                               _
         . INLET (MOT LEG) 1    1        OD        30 TS-1           External damage near lane bl                                                    '

device. present in 1981 inspection. 1 5 ID 20 TS-1 Internal imperfection unchanged since 1981 inspection. 1 6 ID 20 TS-1 Internal imperfection unchanged since 1981 inspection. 1 26 cn 20 2-3 Siaell volume external imperfection. Not recognizable on baseline.

           'l   78'       OD        20       3-4 Small volume external imperfection.

Not recognizable at 400 kH: channel in 1982 , inspection. Not recogni:abic cn baseline. 1 90 OD 30 TS-1 External damage near lane blocker device. Unchanged since 1981 inspection. 1- 91 OD 20 TS-1 External damage near lane blocker device. Unchanged since 1981 inspection. i -

           .1   92        OD        20       TS-1

( p- External degradation near lane blocker device, unchanged since 1981.inspectier.. 1 94 OD 20 TS-1

                    -                                        External degradation near lane blocker device, unchanged since 1981 inspection.                                            -

5 26 OD. 20 1-2 External imperfection, present on baseline.

5. 61 ID 25 .

[ 6-7 Internal flaw, present on baseline.

          .5   76        ID        20}}