Text

Provides Tabulated Listing of Plant First 10-yr Inservice Insp Interval Exams & Applicable Relief Requests
	ML20080R854
Person / Time
Site:	Point Beach
Issue date:	02/22/1984
From:	Fay C; WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, John Miller; Office of Nuclear Reactor Regulation
References
	TAC-48762, NUDOCS 8402280433
	Download: ML20080R854 (40)
	v • d • e

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l lHsconsm Electnc meacouesur 231 W. MICHIGAN, P.O. BOX 2046, M:LWAUKEE. WI 53201 February 22, 1984 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S.. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. J. R. Miller, Chief Operating Reactors Branch 3 Gentlemen:

DOCKET 50-301 COMPLETION OF FIRST TEN-YEAR INSERVICE INSPECTION INTERVAL POINT BEACH NUCLEAR PLANT, UNIT 2 On July 6, 1983 the first inservice inspection interval for Point Beach Nuclear Plant, Unit 2, was completed. The Unit 2 first ten-year inservice inspection interval was extended from October 1, 1982 to July 6, 1983 to permit inspections to be concurrent with plant outages, as allowed in Article IMA-2400 of the 1974 Edition up to and including the Summer 1975 Addenda of Section XI of the ASME Boiler and Pressure Vessel Code.

Enclosures 1 and 2 are a summary of the e::aminations performed during the first ten-year inservice inspection interval for Unit 2.

Upon review of the summary of the examinations performed, it is apparent that some deviations from the ASME Section XI Code occurred.

Prior to the. third period, inservice inspection was performed in accordance with Technical Specification 15.4.2.B, included in Enclosure 3 for your reference. Third period inservice inspection work was performed at the direction of the NRC in accordance with the most restrictive provisions of either Technical Specification 15.4.2.B or the inservice inspection plan Wisconsin _ Electric submitted to the NRC on February 26, 1979. The inservice inspection plan submitted to the NRC was written to the 1974 Edition (Summer 1975 Addenda) of ASME Section XI. All of the Technical Specification and/or ASME Section XI inservice inspection period and interval requirements were met with a few exceptions. These exceptions are noted and explained in Enclosure 4, and required relief requests are noted.

ADOfK O! h d3 l'

.PDR 'I a 1

L Mr. H.'R. Denton February 22, 1984

-The first ten-year inservice inspection interval requirements for Unit 2, with relief requests, were completed 1

, during the spring 1983l refueling outage. By letter dated August 31, 1982 the NRC granted relief to some examination requirements for i

1 Unit: 1. This was amended by NRC letter dated February 17, 1983 to include applicability of these relief requests to Unit 2.

- An additional relief was requested for Units 1 and 2 by Licensee's

~

letter dated: September- 29, 1982 for-relief from frequency require-

'ments of visual examination of the' reactor vessel interior surfaces.

The NRC's. letter of. February 17 did.not grant this relief but provided an interpretation of the examination requirements. The reactor vessel intericr surfaces examination was completed for Unit 2 during the spring 1983 refueling outage but, since the first ten-year inservice inspection interval for Unit 1 was complete in December 1981 (including the' grace period) , Wisconsin Electric was unsble to. conduct (the subject visual examination on Unit 1 reactor pressure: vessel interior surfaces with respect to the third 4

period of its-first ten-year inservice. inspection interval.

However,.a camera was mounted on the par device during the Unit 1 reactor-pressure vessel nozzle examination which. viewed the internal surfaces around those nozzles. No recordable. indications were observed.- In addition, Wisconsin Electric conducted the afore-mentioned interior surfaces visual examination on the Unit 1 "

reactor pressure vessel during the fall 1983~ steam generator replacement outage when all of che fuel was off-loaded. This examination is part of the second ten-year interval inservice

. inspection' program. .

This letter is being submitted for information only.

' It provides a tabulated listing of the Unit 2 first. ten-year inservice ~ inspection interval examinations and applicable relief

- requests for.your reference. Wisconsin Electric considers the ,

Unit 2 first ten-year. inservice inspection interval to be closed. j 4

Wisconsin Electric's current action is to perform the second ten-year inservice inspection interval work in accordance with the-ten-year _ plan proposed in Licensee's letter dated October 6, 1981 which was found acceptable in Mr. Colburn's letter of October 21, 1981. The. Unit 2 ten-year plan was submitted with Licensee's letter

. dated January 19, 1983. As previously informed in Licensee's October.6, 1981 and January 19, 1983 letters, the Unit 2 second ten-year-interval inspection plan was written to the 1977 Edition (Summer 1979 Addenda) of ASME Section.XI.

Very truly yours,

~ ?

giu f

Vice Presiden -Nuclear Power C. W. Fay:

Enclosures Copy to'NRC Resident Inspector w =_ _1

l .s O

ENCLOSURE 1 10-YEAR ISI

SUMMARY

TABLES The ISI summary tables provided herin are a listing by component, category and item number of the Technical Specifi-cation and Code examination requirements as well as the status of the examinations. The following is a key to the summary tables: '

Block Descriotion (1) Component Component noun name (2) Exam area Area to be examined (3) Cat ASME Section XI category (4) Item No ASME Section XI item number (5) Meth NDE method UT Ultrasonic test PT Dye penetrant test VT Visual test MT Magnetic particle test RT Radiographic test (6) Req Exam requirement s

a. TS Technical Specification b, C ASME Section XI Code ~

(7) Comp Percent of exam area complete (8) Period 40-month period (9) 10-Yrs 10-year interval Abbreviations NA Not applicable REM Remainder

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p oa al oe l l l e LRW LMW MUW U CF FW NW

COHl'ONENT - Reactor Vessel UNIT 2 INTERVAL - FIRsT

, FIRST PERIOD SECOND PERIOD TillRD PERIOD 10 1 FEARS , ,

EXAH AREA j E3 REHAltK5 j TS TS TS TS c c c c-2 2 2 6 Nozzle Inside Nozzles N zz es N zz s " ** ** " ""** "*

25 - 36 2/3%

Radius Section BD Bl.4 UT 33 1 /3% 0% 100%

25 - ee n sure 4, #8 REM 100%

33 1/3% 33 1/3%

NA NA NA

NA (1) Of Each Group of Vessel Penetrations of Comparable

. BE Bl.5 VT 0% 0% 100% 100%

Penetrations 6 1/4 - 6 1/4 - REM 25% Size and Function 8 1/3(1) 81/3(1) (1) III (1) III Nozzle to Safe-End 25'% 25% REM 100% (1) 100% = 6 Nozzles Welds BF Bl.6 UT 33 1/3% .. 0% (2) 100% of Weld Examined (6 Nozzles) 25 - 25 - 1/3% 2/3% ,

33 1/3% 33 1/3%

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25% () = N zz s 25% REM 100%

(2) 100% of Weld Examined Bl.6 PT 25 - 33 1/34 25 - 66 2/31 0% 100% (3) PT not performed on (3) (3) SI Nozzles due to Inaccess-ibilitv. Ron Fncinenvn 24 (1) 10'0% - 6 Nozzles 25% 25% REM 100% (2) 100% of Weld Examined (3) VT not performed on 33 13/% 66 2/31 0% 100%

NA NA NA (3) NA (3) p N zz s he to Inaccess-Ibility. See Enclosure 4 #1 25% 25% REM 100%

Bl.7 (2) Bl.7 Applies to Closurt Closure Studs BGl ! 33 1/31 33 1/39 100% Studs in Place.

Bl.8 25 - 25 -

REM 100% (3) Bl.8 Applies to Closur(

33 1/3% 33 1/3%

Studs Removed NA NA NA NA (1) 100% = 48 studs Bl.8 MT 25 - 25 -

REM 100%

33 1/3% 33 1/3% ,

25% 25% REM- 100%

VT 33 1/3% 33 1/3% 33 1/31 -100% (1) 100% = 48 studs NA NA HA NA i

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cuisoNENT - Reactor vessel' UNIT 2 INTE:Nat. - FIpar FIRST PERIOD SEcOND PERIOD THIRD PERIOD IS YEARS EXAM AREA REMAHES

f TS C

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a

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I 33-1/3% 33 1/3%

25% 25% REM 100%

33 1/3% 33 1/3% 33 1/34 100% (1) 100% = 48 nuts NA NA NA NA

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NA NA NA NA 166 Ligaments BG1 Bl.9 UT 100% 33 1/3% 33 1/31. *#" * "" ""* "

2/3% Disassembled NA NA NA (1) 9 25% 25% REM 100%

Washers BG1 Bl.lt VT / / 6 washers 25 - 25 -

33 1/3% 33 1/3%

NA NA NA NA Not Applicable. All Pressure Retaining Bolting Greater Than i Bolting BG2 Bl.11 VT g 2" Diameter.

NA NA NA NA I Integrally Welded 100% 0% 200% (1) 100% = 2 Support Lugs 25 - 25 -

Support BH Bl . l. UT "

33 1/3% 33 1/3%

i 2 2 2 6 Patches 2 Patches 2 Patches 2 Patches 6

! Closure Head (1) 2 Patches (1) 2 Patches (1) 2 Patches (1) 6 Patche: (1) 6" x 6" Patch

! Cladding BIl Bl.l. PT Patches (1) Patches Patches (1) Patches (1)

i _

2 2 2 6 Patches 2 Patches 2 Patches 2 Patches 6 f Bl. l VT (1) 2 Patches (1) 2 Patches (1) 2 Patches (1) 6 Patchet (1) 6" x 6" Patches III Patches Patches II) Patches II)

Patches NA NA 6 NA 2 (1) 8 Vessel Cladding BIl Bl.1< VT .

O Patcher Patche i Patche: (1) Gen Area Inspection l (2) 2 (2) 2 (2) (2) 2 (2) (2) 6 (2)*

Patches Patches Patches patches (2) 6" x 6" Patch a ... --- .- - -

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COMPONENT - Reactor Coolant' Pump "A" UNIT 2 InrE; vat. - FIRS 7 FIRST PERIDD SECOND PERIOD TIBIRD PERIOD 10 YEA'tS , ,

EXAH AHEA ]

REMAHKS TS TS TS TS h 4 c c c c-NA NA NA NA (1) 100% = 24 Studs (2) B5.1 Applies to Studs Pressure Retaining B5.1 25- 25-

% 45 2/3" 100%

in Place Studs BG2 B5.2 UT REM 100% (3) B5.2 Applies to Studs 33 1/3% 33 1/3% pomvnd

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(3) Bolting Not Removed NA NA NA (2) (3) During This Interval NA' NA NA (2) (1) 100% = 24 Studs B5.3 VT 66 2/3% 0% 33 1/31 100% (2) As Accessible for REM 100% Other Reasons 33 1/3% 33 1/3%

NA NA NA NA 9 Nu s Pressure Retaining BG1 B5.2 MT 0% - 0% Ga pemoved NA NA NA (1) During This Interval NA NA NA (1) 133 As Access M e for B5.3 VT 56 2/3% 0% 66 2/39 Other Reasons 25- 25- (2) 100% = 24 Nuts REM 100%

33 1/3% 33 1/3%

NA NA NA NA (1) if DisasseMoled Ligaments BGl B5.3 VI 0% 0% 0% 0% (2) Pump was not Disassem-s Interval.

NA NA NA 100% (1) (2)

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" Hz Cast SS. Body. Ron Fnc!

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NA NA NA NA 166 Support Legs BK2 B5.5 VT _-- 0% 6 2/3% 100% (1) 1 % = 3 Support Legs 25- 2/3%

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ENCIOSURE 2

~

SUMMARY

OF ASME SECTION XI PRESSURE TESTS-ASME.Section XI Class Reference- Type of Test Frequency Status 1 IWB 5210(a) Hydrostatic pressure test 100% every 10 years 100% complete

. IWB 5210 (b) . ' Leakage test 100% every refueling .100% complete outage 2

2 IWC 2412(a) .Not applicable __ __

IWC 2412 (b)- Pressure test of exempt' 100% each interval. 100% complete components 1

3 IWD 2410(b) Hydrostatic. pressure test 100% every 10 years 75% complete IWD 2410(c) . Inservice pressure test 100% every period 100%' complete 1

This requirement was adopted during the third period. WE conmitted to the performance of 33 1/3% of the required Class 3. hydrostatic pressure tests for the first 10 year interval in our inservice inspection plan submitted February 26, 1979.

2 As stated.in our inservice inspection plan, all Class 2 systems are exempt in-accordance with IWC 1220.

ENCLOSURE 3

,15.4.2 In-SzavIcz INS?EC ION OF PRIMARY SYSTEM COMPCNEI:TS Apolicability Applies to in-service inspe= tion of Reactor Coolant System Components. ,

ob$ectives To provide assurance of the continuing integrity of the Reactor Coolant system.

I Soecifications A. Steam Generator Tube Inspection Requirements

1. Tube Inspection l Entry from the hot-leg side with avamination frem the point of entry l ,

completely around the U-bend to the top support of the cold-leg is l considered a tube inspection.

2. Sample Selection and Testing Selection and testing of steam generator tubes shall be made on the following basis:

(a) One steam generator of each unit shall be inspected during inservice inspection in accordance with the following i requirements:

1 l

l 1. The inservice inspection may be 14=dted to one steam generator en an alternating sequence basis. This avamination shall include at least 6% of the tubes if the results of the first or a prior inspection indicate that both generators are performing in a comparable manner.

2. When both steam generators are required to be examined by Table 15.4.2.1 and if the cendition of the tubes in one generator is found to be more severo than in the other steam generator of a unit, the steam generator sampling sequence

- at the subsequent inservice inspection shall be modified to examine the steam generator with the more severe condition. I s

(b) The minimum sample sire, inspection result classification and the associated required action shall be in conformance with the requirements specified in Table 15.4.2-1. *he results of each sampling examination of a steam generator shall be classified into the following three categories:

i 1

i Unit 1 Amendment No. 10 15.4.2-1 July 12, 1976 Unit 2 Amendment No.12 l

1 4

t 1 . .

l

- $ Catecarv c-1:

~

locs than 5% of the total numbar of tubos examined are dearaded but none are defective. <

Catecorv C-2: Between 54 and 10% of the total number of tubes examined are degraded, but none are defective E one tube to not nere than 1% of the sample is defective.

Catecerf C-2: Mere than 1% of the tetal n-kcr of tubes examined are degraded, but none are defeetive g more than 14 of the sample is defective.

l

1 In ths first sample of a civen steam cenerator during any inservice inspection, decraded tubes not beyond the pluggine limit detected l- by the prior examinations in that steam cenerator shall be included in the above percentage calculations, only if these tubes are demonstrated to have a further wall penetration of greater than 10% of the nominal tube wall thickness. '.

1 (c) "ubes shall be selected for examination primarily from those i areas of the tube bundle where service experience has shown the ,

most severe tube degradation.

(d) In addition to the sample sire specified in Table 15.4.2-1, the tubes examined in a given steam generator during the first examina-tion of any inservice inspection shall include all non-plugged tubes in that steam generator that from prior examination were degraded.

(e) During the second and third sample exami..ations of any inservice inspection, the tubes inspection my be limited to those sections of the tube lengths where imperfections were detected during he I

prior examination.

I j 3. Examinatien Method and P.eeuiremests (a) Steam generator tubes shall be examined in accordance with the method prescribed in Article 8 " Eddy Current Examination of Tubular Products", as contained in ASME Boiler and Pressure

{

i Vessel Code - Section V " Nondestructive Examination".

I l

(b) The examination method of 15.4.2.A3 (a) shall apply until Appendix IV,

, " Eddy Current Examination Method of Non-Ferromagnetic Steam '

Generator Heat Exchanger Tubine" is incorporated and become

. effective rules of the ASME Boiler and Pressure vessel Code,Section XI - Inservice Inspection of Nuclear Power Plant Components.

At that time, the rules of ASME Code, Secticn XI shall be used

. in lieu of 15.4.2.A3 (a) .

I Unit 1 Amendment No.10 15.4.2-la July 12, 1976 Unit 2 Amendment No. "2 .

4. Incoection Int"-rv'ltt

. (a) Incorvico inspections shall not be moro than 24 calendar montha apart.

, (b) The inservice inspections may be scheduled to be coincident '

with refueling outages or any plant shutdown, provided the inspection intervals of 15.4.2.A.4 (a) are not exceeded.

(c) If two consecutive insei'vice inspections covering a e time span of at least 12 months yield results that fall in C-1 category, the inspection frequency may be extended to 40 month intervals. ,

1 (d) If the results of the inservice inspection of steam generator j tubing conducted in accordance with Table 15.4.2-1 requires that ,

j

a third sample examination must be performed, and the results of !

l this fall in category C-3, the inspection frequency shall be l l reduced to not more than 20 months intervals. The reduction !

shall apply until a subsequent inspection demonstrates that a third sample ev==4 ution is not required.

3 (e) Unscheduled inspections shall be conducted in accordance with Specifications 15.4.2.A.2 on any steam generator with primary-to-secondary tube leakage exceeding Specification 15.3.1.D.4.

All steam generators shall be inspected in the event of a seismic occurrence greater than an operating basis earthquake, a LOCA requiring actuation of engineered safeguards, or a main steam line or feedwater line break.

e s 5. Acceptance Limits (a) Definitions:

l Imoerfection is an exception to the dimension, finish, or contour l of a tube from that required by fabrication drawings or speci-fications. Eddy-current testing indications below 20% of the nr= 4 a= 1 tube wall thickness, if detectable, may be considered as imperfections.

I Degradation means a service induced cracking, wastage, wear, or general corrosion occurring on either inside or eutside I

of a tube.

Deeraded Tube is a tube that contains imperfections caused by degradation greater than 20% of the nominal tube wall thickness.

t Unit 1 Amendment No. 10 15.4.2-lb July 12, 1976 Unit 2 Acendment No. 12

. . ~ ._,,,.~.,.___.---,,w - - , - ~ . , - - _ , . . - _ . . , . . . _ . , . _ , , , . , _ - _ _ , , _ _ _ . . .

.=. . . - . - -

6'

.. Defact is en imparfecticn of such ssvarity that it exceeds the minimum acceptable tube wall thickness of 50%. A tube containing a defect is defective.

Plugging Limit is the imperfection depth beyond which the tube must be removed from service, because the tube may become defective prior to the next scheduled inspection.

The plugging limit is 40% of the nominal tube wall thickness.

B. Corrective Measures All tubes that leak or have degradation exceeding ^the plugging limit shall be plugged prior to return to power from a refueling 1or inservice inspection condition.*

C. Reports

1. After each inservice examination, the number of tubes plugged in each steam generator shall be reported to the Commission as soon as practicable.

2. The complete results of the steam generator tube inservice inspection shall be included in the Operating Report for the period in which the inspection was completed. In addition, all results in Category C-3 of Table 15.4.2-1 shall be reported to the Commission prior to resumption of plant operation.

3. Reports shall include:

\

(a) Number and extent of tubes inspected (b) Location and percent of all thickness penetration for each indication (c) Identification of tubes plugged

4. Reports required by Table 15.4.2 Steam Generator Tube Inspection shall' provide the information required by Specifi-cation 15.4.2.C.2 and a descriptica of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

B. In-Service Inspection of Reactor Coolant System Components Other Than Steam Generator Tubes ,

The in-service inspection program is generally based on the recommendations of ASME Boiler and Pressure Vessel Code,Section XI, Summer 1971 Addenda as practical for a plant whose design and construction preceded issuance of the recommendations. The commitments herein are made assuming that

> the necessary inspection I

Point Beach Nuclear Plant Unit 1 may be operated at power with up to six tubes in one steam generator having degradation exceeding the plugging limit provided these tubes have been repaired by insertion of sleeves into the tubes to bridge the degraded or defective portion of the tube. The plugging limit is 35% of the nominal sleeve wall thickness for tubes that have been repaired by sleeving.

Unit 1. Amendment No. 56 15 . 4. 2-I c November 10, 1981

.techniquac will be comm:rciall'i availablo end that necoscary cecassibility

. ,can b3 gain d to compon:nts to allew insp;ction. At th3 cnd of tha first ss five years of the inspection period, a review of the inservice inspection program will be conducted. Titis review will evaluate the .results ebtained to date in view of possible modifications to the inspection program.

These modifications may increase or decrease surveillance regairements as experience dictates.

IN-SERVICE INSPEC"JION PPCGRAM (NOTE 1) i.

Bv 1/3 of inspection period - 40 menths RV flange and head flange welds Volumetric of 25% of each weld l

l ,

RV noz=le to vessel welds and Volumetric of 2 outlet nozzles inside radii RV nuts and studs Volumetric and visual on 25% (Note 2)

RV closure washers and bushings visual of 25%

l Closure head cladding Visual and surface of 2 patches i

Pressurizer cladding Visual (Note '3) l Reactor vessel nozzles to piper Visual, surface, and volunetric of

( pressurizer surge noz=le to 25% of welds (Note 4)

! pipes steam generator primary no nozzles to pipe welds l

l l .

Unit 1 Amendment No. 10 15.4.2-Id July 12, 1976 Unit 2 Amendment No. 12

- . ,,,_....m.. . .-__ _ ,,,_ c _.____,_._...__,...,.m. _ _ _ - - . . . . . , . , _ . . , - . , , . _ . . . -

i c$inferancial pips weldo Vicuni and volumatric of 6% of walda -

Surveillance samples Tensile, Charpy, wedge-opening-load ,

tests (Note 5) ,d Reactor coolant pump flywheels Visual, as accessible without removing flywheel ;

By 2/3 of insnection eeriod - 80 months RV flange and head flange velds Volumetric of additional (over previous inapection) 25% of each weld R7 nozzle to vessel welds and Volumetric of 2 SIS nozzles inside' radii RV nuts and studs Volumetric and visual on additional (over previous inspection) 25% (Note 2)

RV closure washers and bushings Visual of additional (over previous

- inspection) 25 l

Clo'sure head cladding Visual and surf ace of additional (over previous inspection) 2 patches

Pressurizer cladding Visual (Note 3)

Reactor vessel nozzles to pipe; Visual, surface and volumetric of additional ~

i pressurizer surge nozzle to (over previous inspection) 25: (Note 4)

~

pipe; steam generator primary ,

nozzles to pipe welds Circumferential pipe welds Visual and volumetric of additional (over previous inspection) 6% of welds t

Reactor coolant pump flywheels Volumetric, as accessible without removing

! flywheel End of inspection eeriod - 120 months I

i RV shell velds Volumetric of 10% of longitudinal and 5% of

! circumferential welds i

Ret .ar head welds Volumetric of 10% of longitudinal and 5% of I

circumferencial welds RV flange and head flange Volumetric of remainder (left from previous welds inspections) of each veld l RV nozzle to vessel velds Volumetric of 2 inlet nuzzles and inside rsdii RV nuts and studs Volumetric and visual of remainder (lef t from i

previous inspections) (Note 2) l

! 15.4.2-2 l

, ' , . RV ciecuro wachcra and Vicual of remaindsr (1sf t f rom previtua

.* bushings inspections) .

Closure head cladding Visual and surface of additional (over previous inspection) 2 patches Reactor vessel cladding Visual (Note 3) l Rasetor vessel internals Visual (Note 6) l and suppercs

! 1 Pressurizer shall and Visual and volumetric 10% of longitudinal head welds and 5% of circumferential Pressurizer cisdding Visual (Note 3) i stesst generator primary head Visual and volumetric 5% of circumferential

, to tube sheet veld l

i Reactor vessel nozzles to Visual, surf ace and volumetric of remainder pipe; pressurizar surge nozzle (lef t from previous inspection) (Note 4) to pipe; steam generator primary nozzles to pipe welds Circumferential velds Visual and volumetric of additional (over previous inspections) 13% of welds $

Reactor coolant pumps casing Visual welds l

Valve body welds Visual and volumetric of one large gate valve and one large check valve Valve internals Visual of one large gate valve and one large

! check valve Reactor coolant pump flywheels Volumetric, 100%

{ As accessible for other reasons

, Control rod drive penetration Visual wMs Pressurizer bolting, steam Visual generator bolting, valve bolting, RC pump bolting Steam generator studs Volumetric, when removed Steam generator primary head Visual (Note 3) cladding Valve hangers Visual and surface NOTE (1): The inspection period contemplaced is 10 years with the cycle then repeating itself.

15.4.2-3 en, , , _ _ , , , , , , , _ , . - - - , - - _ . _ .

r.

NOTE (2): Thrtsdn in vtecal not included: tharo are relatively icw strota creas end ability to do a metningful examination is

. ... doubtful. Stud etrotchiny will be dons after each refualing

- and is considered a better test. s NOTE (3): No particular patches prepared; general inspection will be made. <

NOTE (4) : Visual and surface of RV nozzle to pipe welds will be top surface only.

L NOTE (5): Subsequent tests will be scheduled based on results of examina-tions made at first 40 month interval.

NOTE (6): Intsrnals will be inspected as accessible during normal refueling.

> Removal of core barrel to allow additional inspection of reactor vessel internal areas- shall be done once during inspection interval. If core barrel is reme'ved prior to end of period (120 months), inspection for that period will be made when barrel is ' removed; otherwise, barrel will be removed z.t end of period specifically to allow inspection.

Bases The proposed inspection program is, where practical, in compliance with the recommendations of ASME Boiler and Pressure Vessel Code,Section XI, s i

Susumer 1971 Addendz. It must be recognized, however, that eculpment and l

. techniques to perform the inspections are still in development. It is recognized, however, that examinations in certain areas are necessary and therefore a schedule is proposed that includes areas and frequencies that l

are believed practical at this time for this reactor. In most areas scheduled for test, a detailed pre-service maoping will be conducted using techniques which can be used for post-operation inspections. The areas indicated for inspection represent those of relatively high stress and therefore will serve to indicate potential problems before significant flaws develop there or at other areas. As more experience is gained in operation of pressurized-water reactor's, the recommended time schedule ard location of inspe'etion might be altered, or should new technicues be developed, consider-ation will bc given to incorporate these new techniques into this inspection i

program.

Unit 1 Amendment No.10 15.4.2-4 July 12, 1976 Unit 2 Amendmont No.12

. ... , ___. _ _ - - , _ , - _ . . - - _ _ . _ _ ~ _ - . - . - . - _ . - _ - _ _ _ _ _ _ _ _ _

n t '.

The use of conventional non-destructive, direct visual and remote visual test techniques can be applied to the inspection of all primary loop components e'xcepc for the reactor vessel. The reactor vessel presents special problems because of the radiation levels and remote underwater accessibility to this component. Because of these limitations on access to the reactor vessel, several steps have been incorporated into the design and manufacturing procedures in preparation for non-destructive test techniques which may be available in the future.

The techniques for in-service inspection include the use of visual in-spections, volumetric (ultrasonic or radiographic) and surface (dye penetrant or magnetic particle) testing of selected parts during refueling periods.

i -

The intent of the inspection is the detection of flaws large enough to initiate fast fracture and gross leakage prier to subsequent inspection.

At this time it is judged that such a flaw is substantially larger than l

1/2 inch by 1 inchwhich is the degree cf detectability. The inspection method is designed so detect flaws of this magnitude.

l (1)FSAR - Section 4.4 15.4.2-5

STEAM GENERATOR TUBE INSPECTION PER tlNIT POINT BEAC 'IITS 1 4 2 ..

.~ e .

IST SAMPLE EXAMINATION 2ND SAMPLE EXAMINATION 3RD SAMPLE EXAMINATION

! Saarple Size Result Action Required Result . Action Required Result Action Required i

A ninimum of C-1 Acceptable for N/A N/A N/A N/A S tubes per Continued Service

' C-2 Plug tubes exceeding the C-1 Acceptable for ', N/A N/A plugging limit and pro- continued Service

ceed with 2nd sample exam na n u es - ug tubes exceedng C-1 AcceptaMe for i S=3(N/n)% in same steam generator the plugging limit Continued Service and proceed with 3rd Plug tubes exc. plus sample examination of C-2 linit. Acceptable j

' ** 4S tubes in same contirued service j N is the steam generator Morg action req 1 C-3 under .-a of ist ntnber of sanple examination

~

stean genera- Perform action requir-tors in the ed under C-3 of 1st J C-3 N/A N/A

] Pl ant = 2 sample examli6ation C-3 Inspect essentially all C-1 in Acceptable for N/A N/A other Continued Service

. n is the tubes in this S.G., plug

nic5er of tubes exceeding the S.G. __

steam genera- plugging limit and tors inspect- proceed with 2nd sample C-2 in Perform action requir- N/A '

N/A a ed during an examination of 25 tubes other ed under C-2 of 2nd i exanination in the other steam S.G. sample examination generator. above __

ltcport results to NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in C-3 in Inspect essentially all N/A N/A accordance with Techni- other tubes in S.G. and plug

. cal Specification S.G. tubes exceeding the plug-

! 15.6.5.2.A.3. ging limit. Report to NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in Unit 1 Amendment No. 10 oni n o u h Technical Unit 2 Amendment flo.12 July 12, 1976 Spec.i fication 15.6.5.2. A. i.

L-o F t 4 ENCLOSURE 4 DEVIATIONS FROM ASME SECTION XI ISI REQUIREMENTS

1. Reactor Vessel Safety Injection Nozzles Safe End Welds Category BF; Item No. Bl.6 A visual, dye penetrant and ultrasonic exam was performed on all the reactor vessel nozzle buttered connections with the exception of both safety injection nozzle buttered connections. A VT and PT could not be performed on these welds since they are enclosed by a concrete sleeve.

A UT exam of these welds was performed from the inside diameter of the nozzles-during the second period with the aid of a remote controlled examination device (par device) . The VT and PT examinations are not practical and relief was granted per NRC letter dated August 31, 1982, amended _by NRC letter dated February 17, 1983, to include Unit 2 welds.

2. Reactor Vessel Closure Studs Category BGl; Item No. Bl.G An MT exam was performed on 331/3% of the reactor vessel closure studs for the first interval. As stated in our inservice inspection plan, WE. intended to perform only the third period requirements for this exam (33 1/3%) since these requirements were adopted during the third period.

Therefore, only 331/3% of the studs were examined vice 100% for the interval. A VT and UT exam was performed on 100% of the studs during the first 10-year interval.

3. Reactor Vessel Closure Nuts Category BGl; Item No. Bl.8 An MT exam was performed on 33 1/3% of the reactor vessel closure nuts for the first 10-year-interval. The reason is the same as for the reactor vessel closure studs. A VT and UT exam was performed on 100%

of the nuts during the first 10-year interval.

4. Reactor Coolant Pump Integrally Welded Support Category BKl; Item No. B5.4 A UT or PT exam of the reactor coolant pump support lugs was not practical due to the surface roughness of the stainless steel pump casing. A VT exam was performed on the support lugs vice a UT exam. Relief was granted per NRC letter dated August 31, 1982.

5. . Regenerative Feat Exchanger Integrally Welded Support Category BH; Item No. B3.7 A PT or UT Exam of the integrally welded support attachments to the heat exchanger shell was not practical due to the weld configuration.

The integral support welds are partial-length fillet welds and are not suitable for UT or PT examination. A VT examination was performed vice a UT examination. Relief was granted per NRC letter dated August 31, 1982.

o --

,o; ,:=-

6. Reactor Coolant Pump Casing Welds

. Category'BL1, Item No. B5.6

'An RT Exam of the casing welds was not practical due to the high radiation exposure, cost, and time. The casing is cast stainless steel which provides too much attenuation for a UT exam. In place of the RT exam, a 33 1/3% PT exam and 100% VT exam of the casing welds were accomplished as granted by NRC letter dated June 1,1983.

7. Reactor Coolant Pump Casing Internals Category BL2; Item No. B5.7 Since the pump was not opened for the RT of the casing welds, the visual examination of the casing internals could not be accomplished. The 100%

VT and 331/3% PT of the casing welds exterior serve as alternate exams for the casing internals VT exam per NRC letter dated June 1,1983.

8. Reactor Vessel Nozzle to Vessel Welds and Inside Radiused Sections Category BD; Item No. Bl.4 A visual, dye penetrant, and ultrasonic exam was performed on all the

- reactor vessel nozzle-to-vessel welds and inside radiused sections during the second period when the core barrel was removed. Due to the significant

. amount of exposure and costs associated with reactor vessel nozzle-to-vessel welds and inside radiused sections, a third period exam was determined.to be impractical as granted by NPC letter dated August 31, 1982, amended by NRC letter dated February 17, 1983, to include Unit 2 nozzle welds and inside radiused sections.

_. 9. Regenerative Heat Exchanger Shell and Nozzle-to-Shell Welds Category BB; Item No. B3.1 and B3.2 The twelve 'shell welds and twelve nozzle-to-shell welds described in Wisconsin Electric's third period inspection plan included both Class 1 and Class 2 welds. There are actually six Class 1 shell welds and six Class 1 nozzle-to-shell welds. Enclosure 1 only reflects the six Class 1 shell and nozzle-to-shell welds.

10. Circumferential Pipe Welds-Category BJ; Item No. B4.5 Per Wisconsin Electric's third period inspection plan, 24 welds would be examined. Actually 27 welds were examined giving 9.25% and meeting the interval requirements by completing 28% of the welds. The more restrictive Technical Specification amount of 13% was not met, but an even period-by-period distribution was achieved and interval requirements were exceeded.

p-a c _; ..o-,

f.

,11. Piping Socket Welds Category -BJ; - Item No. B4.8

' Per Wisconsin Electric's third period-inspection plan, approximately 31 welds were to be examined. ' Actually 24 welds were examined giving L7.50% which exceeded interval requirements with 27.50% and satisfied ASME Section-XI third period requirements of "the remainder of 25%."

k

__

v t e Letter Sequence Request
TAC:48762, (Open)
Initiation Request, Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other:
MONTHYEAR1982-08-20 [Table View]

ML20080R854

Text

SUMMARY

SUMMARY

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