ML20041C407

Tech Spec 15.4.2 Re Inservice Insp of Primary Sys Components

ML20041C407
Person / Time
Site:	Point Beach
Issue date:	11/10/1981
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20041C405	List: ML20041C404 ML20041C407 ML20041C430 ML20041C432
References
NUDOCS 8203010288
Download: ML20041C407 (25)

Text

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15.4.2 IN-SERVICE INSPECTION Cr PRIMARY SYSTEM CCMPCNENTS

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• Applicability f Applies to in-service inspection of Reactor Coolant System Components.

Objectives To provide assurance of the continuing integrity of the Reactor Coolant System.

Specifications A. Steam Generator Tube Inspection Requirements

1. Tube Inspection Entry from the hot-leg side with examination frem the point of entry completely around the U-bend to the top support of the cold-leg is 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 requirements:

1. The inservice inspection may be limited to one steam generator on an alternating sequence basis. This ewamination 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 condition of the tubes in one generator is found to be more severe 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.

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

Unit 1 Amendment No. 10 15.4.2-1 July 12, 1976 Unit 2 Amendment No. 12 0203010288 820223 PDR ADOCK 05000266 G PLR

Catenary C-1: loco than 5% of tho total number of tubes examined are degraded but none are defective. ,

Catenorv C-2: Between 5% and 10% of the total number of -

tubec examined are degraded, but none are defective or one tabe to not more than li of the sample is defective.

Category C-2: Mere than 10% of the tet:1 n--her of tubcs examined are degraded, but none are defective o_r more than 14 of the sample is defective.

In the first sample of a given steam generator during any inservice inspection, degraded tubes not beyond the pluggina limit detected by the prior examinations in that steam generator 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) Tubes shall be selected for examination primarily from those i areas of the tube bundle where service experience has shown the most severe tube degradation. I (d) In addition to the sample size specified in Table 15.4.2-1, the tubes examined in a given steam generator during the first exandna-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 ereminations of any inservice inspection, the tubes inspection nay ce lLnited to those sections of the tube lengths where imperfections were detected during the ,

prior examination.

3. Examination Method and P.ecuirements (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 vessel Code - Section V " Nondestructive Examination". '

(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,Section 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. 12

4. Incpectien Int ^rvnin (a) Inservice inspections shall not be more than 24 calendar months apart. ,

I (b) The inservice inspections may be scheduled to be coincident 8

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 inservice inspections covering a 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.

(d) If the results of the inservice inspection of steam generator ,

tubing conducted in accordance with Table 15.4.2-1 requires that  ;

a third sample examination must be performed, and the results of this fall in category C-3, the inspection frequency shall be reduced to not a re than 20 months intervals. The reduction shall apply until a subsequent inspection demonstrates that a third sample examination is not required.

(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.

5. Acceptance Limits (a) Definitions:

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

Degradation means a service induced cracking, wastage, wear, or general corrosion occurring on either inside or outside of a tube.

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

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

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Dafcet in cn imparfection of such stvarity thst it excerds 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 te return to power from a refueling or inservice inspection conditi:n.*

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 description 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

• 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 i Amendment No. 56 15 . 4. 2-Ic November 10, 1981

techniqu;a will be comm rcially availabls and that nec02cary acc soibility can be gnin:d to compon:nts to ellow intptction. At th2 cnd of tha firct five years of the inspection period, a review of the inservice inspection program will be conducted. This review will evaluate the results obtained to date in view of possible modifications to the inspection program.

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

IN-SERVICE INSPECTION PROGRAM (NOTE 1)

By 1/3 of inspection eeriod - 40 menths RV flange and head flange welds Volumetric of 25% of each weld RV nozzle 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%

Closure head cladding Visual and surface of 2 patches Pressurizer cladding Visual (Note '3)

Reactor vessel nozzles to piper Visual, surface, and volumetric of pressurizer surge nozzle to 25% of welds (Note 4) pipes steam generator primary no nozzles to pipe welds Unit 1 Amendment No. 10 15.4.2-ld July 12, 1976 Unit 2 Amendment No. 12

.Cire aferential pipe welds Visual and volumetric of 6% of welds Surveillance samples Tensile, Charpy, wedge-opening-load f teste (Note 5) ,/

Reactor coolant pump flywheels Visual, as accessible without removing flywheel By 2/3 of inspection period - 80 months RV flange and head flange welds Volumetric of additional (over previous inspection) 25% of each weld RV 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%

Closure 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 pressurizer surge nozzle to (over previous inspection) 25% (Note 4) '

pipe; steam generator primary nozzles to pipe welds Circumferential pipe velds Visual and volumetric of additional (over previous inspection) 6% of welds i

Reactor coolant pump flywheels Volumetric, as accessible without removing flywheel l

! End of inspection period - 120 months 1 RV shell welds Volumetric of 10% of longitudinal and 5% of

{ circumferential welds l

Reactor head welds Volumetric of 10% of longitudinal and 5% of l

l circumferential welds l

RV flange and head flange Volumetric of remainder (left from previous welds ,

inspections) of each weld l

l RV nozzle to vessel welds Volumetric of 2 inlet nozzles

,cnd inside radii l

RV nuts and studs Volumetric and visual of remainder (left from previous inspections) (Note 2) 15.4.2-2 l

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I RV elegura vach2ra tnd Vicuni cf remaind2r (1sft from previtun I bushings inspections)

Closure head cladding Visual and surface of additional (over previous inspection) 2 patches Reactor vessel cladding Visual (Note 3)

Reactor vessel internals Visual (Note 6) and supports Pressurizer shell and Visual and volumetric 10% of longitudinal head welds and 5% of circumferential Pressurizer cladding Visual (Note 3)

Steam generator primary head Visual and volumetric 5% of circumferential to tube sheet weld Reactor vessel nozzles to Visual, surface and volumetric of remainder pipe; pressurizer surge nozzle (lef t from previous inspection) (Note 4) to pipe; steam generator primary nozzles to pipe welds Circumferential welds Visual and volumetric of additional (over previous inspections) 13% of welds Reactor coolant pumps casing Visual welds 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 i

Reactor coolant pump flywheels Volumetric, 100%

As accessible for other reasons Control rod drive penetration Visual welds Pressurizer bolting, steam Visual generator bolting, valve l

l bolting, RC pump bolting Steam generator studs Volumetric, when removed l

Steam generator primary head Visual (Note 3) cladding i Valve hangers Visual and surface NOTE (1): The inspection period contemplated is 10 years with the cycle then repeating itself.

15.4.2-3

o NOTE (2): Threads in v;ccal not included: thsta cre relatively Icw ctrees creau end cbility to do a meaningful examination is

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• Stud strctching will be dona aftsr cach refueling doubtful .

and is considered a better test. s NOTE (3): Nn particular patches prepared general inspection will be made.

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

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

NOTE (6) : Internals 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 remo'ved prior to end of period (120 months), inspection for that period will be made when barrel is removed otherwise, barrel will be removed at 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, ,

Summer 1971 Addenda. It must be recognized, however, that equipment and l \

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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 are believed practical at this time for this reactor. In most areas scheduled for test, a detailed pre-service mapping 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 reactors, the recommended time schedule and lo~ tion of inspection might be altered, or should new technicues be developed, consider-ation will be given to incorporate these new techniques into this inspection program. -

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

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.' The use of conventional non-destructive, direct visual and remote visual test techniques can be applied to the inspection of all primary loop components except 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 tes t 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 pe riods .

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

At this time it is judged that such a flaw is substantially larger than 1/2 inch by 1 inch which is the degree of detectability. The inspection method is designed to detect flaws of this magnitude.

(1)FSAR - Section 4.4 l

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i 15.4.2-5 i

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STEAM GENERATOR TUBE INSPECTION PER UNIT

• POINT BEACH UNITS 1 4 2 b

i IST SAMPLE EXAMINATION 2ND SAMPLE EXAMINATION 3RD SAMPLE EXAMINATION Sample Size Result Action Required Result Action Required Result Action Required A minimum of C-1 Acceptable for N/A N/A N/A N/A S tubes per Continued Service bI**'

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

'* ~ ceed with 2nd sample

_ examination of 2S tubes C-2 Plug tubes exceeding C-1 Acceptable for 6"# INI"U in same steam generator the plugging limit Continued Service and proceed with 3rd Plug tubes exc. plug sample examination of C-2 linit. Acceptable fc where: 4S tubes in same contirued service steam generator Perfora action requii

! N is the C-3 under C-3 of 1st number of sanple examination

-stean genera- -

Perform action requir-tors in the C-3 ed under C-3 of 1st N/A N/A Pl ant - 2 sample examination C-3 Inspect essentially all C-1 in Acceptabic for N/A N/A n is the tubes in this S.G., plug other Continued Service number 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 ed during an examination of 2S tubes other ed under C-2 of 2nd Gxamination in the other steam S.G. sample examination generator. above Report results to NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in Unit 1 Amendment No. 10 accordance with Technical Unit 2 Amendment no.12 July 12, 1976 Spec.ification 15.6.5.2.A. 3.

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• ENCLOSURE 4 DEVIATIONS FROM ASME SECTION XI ISI REQUIREMENTS

1. Reactor Vessel Nozzle 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 WE intends to perform only UT exams of these welds in the future.

2. Reactor Vessel Closure Studs Category BGl; Item No. Bl.8 An MT exam was performed on 33-1/3% 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 33-1/3% 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 Vessel Interior Category BN1; Item No. B l . l _5_

A visual examination of the reactor vessel interior was not performed each refueling outage since only a 10-inch band around the vessel wall is accessible during a normal re-fueling outage as shown in Figure 1. The onPf time the

O Enclosure 4 Page 2 vessel surfaces are accessible is when the core barrel is removed for the 10-year reactor vessel examination. The accessible area of the reactor vessel during normal re-fueling outages is observed by Maintenance and operations personnel, but their examinations are not performed in accordance with ASME Section XI.

The reactor vessel interior examination will be performed during the next scheduled 10-year reactor vessel examination.

This examination is not practical once each period.

5. 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 pump casing. A VT exam was performed on the support lugs vice a UT exam.

6. Regenerative Heat 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. As shown in Figure 2, 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.

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