ML19093A140

Final Results of the Inservice Inspection Program, Refueling Outage No. 1, Surry Power Station, Report No. ISI 75-3

ML19093A140
Person / Time
Site:	Surry
Issue date:	04/01/1975
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
ISI IR 1975003
Download: ML19093A140 (18)

Text

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I I FINAL RESULTS OF THE INSERVICE INSPECTION PROGRAM I REFUELING OUTAGE NO. 1 I SURRY POWER STATION I APRIL 1, 1975 I REPORT NO. ISI 75-3 I

DOCKET NO. 50-280 I LICENSE NO. DPR-32 I

I I THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE OFFICE OF REGULATION:* THEY HAVE BEEN .CHARGED TO YOU FOR A LIMITED TIME I PERIOD ANS MUST BE RETURNED TO THE CENTRAL RECORDS STATION 008. ANY PAGE(S)

REMOVED FORREPRODUCTION MUST BE RETURNED I To ITS/THEIR 0R1G,~AL ORDER ..

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1 I MARY JINKS, CHIEF I CENTRAL RECORDS STATION yCf? I

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I FINAL RESULTS OF THE INSERVICE INSPECTION PROGRAM I REFUELING OUTAGE NO. 1 I SURRY POWER STATION I APRIL 1, 1975 I

REPORT NO. ISI 75-3 I

DOCKET NO. 50-280 I LICENSE NO. DPR-32 I

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I Vepco I VIRGINIA ELECTRIC AND POWER COMPANY I

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I TABLE OF CONTENTS I Page No.

I I. INTRODUCTION . . * * . . *... 1 II. INSPECTION S~Y ** 4 I III. INSPECTION RESULTS

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2.

Reactor Coolant System Piping Systems Containing Sensitized Stainless Steel ...*

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I 3. High Energy Line Piping

• 8 4 * . Low Pressure Turbine Rotor Blading 9 I 5. Material Irradiation Surveillance Capsule 9 I 6.

7.

Steam Generator Tube Examinations

• Reactor Internals Inspection ... .

9 12 I IV. CONCLUSIONS 14,

v. REFERENCES . . . . . .. 15 I

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I I I. INTRODUCTION I In accordance with the requirements of Technical Specification 6.6.C, this report contains a summary of the results of the inservice inspection I activities performed during the first refueling outage of Unit No. 1 of I the Surry Power Station during the period November 1, 1974 through December 31, 1974.

I The document entitled, Inservice Inspection Program, Refueling Outage No * ..!., Surry Power Station, Unit No._!_, Report No. ISI 74-1, I dated July 18, 1974, provides the specific details concerning the in-I spections which were scheduled to be performed. Pages 3 through 9 of the aforementioned document summarizes the specific areas to be inspected.

I Referring to this listing, the following items were omitted or modified during the inspection period:

I ISI-261 Tech Spec I Component Ref Ref Remarks Reactor Vessel 1.3 1.3 Will be done by remote UT at I a later refueling 1.4 1.4 Will be done by remote UT at I 1. 9 1.9 a later refueling Will be done by remote UT at a later refueling I Associated Aux. 4.9 4.5 A surface examination (PT) was Piping done instead of volumetric (UT)

I because of configuration The above deviations comply with the requirements of Technical Specification I 4.2.

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I I Some individual welds inspected were different from the ones I designated in Report No. ISI 74-1 due to configuration or accessibility.

Details of the inspections performed are on file at the Surry Power I Station and the Richmond General Office.

The inservice inspections were conducted by Vepco and Westinghouse I personnel. The items which the Virginia Electric and Power Company accomplished during the inspection are detailed below:

I Component Tech Spec Ref Area Inspected Method of Inspection I Reactor Vessel 1. 7 Primary nozzel to safe end weld VT,PT I Misc. Inspections 1.8 7.2 Closure Studs Low head SIS piping PT VT I 7.3 in valve pit LP Turbine rotor VT,MT,PT blades I 8.1.2 Cir. welds and VT branch connections I 4" dia. and smaller I The items which the Westinghouse Electric Corporation accomplished during the inspection are detailed below:

I Tech Spec Area Method of Component Ref Inspected Inspection I Reactor Vessel 1.3 1.8 Closure Head to Flange Closure studs and nuts VT/UT VT/UT 1.10 Closure washers VT I 1.13 1.14 Closure head cladding Vessel cladding VT/PT VT 1.15 Vessel internals VT I

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I I Component Tech Spec Ref Area Inspected Method of Inspection I Pressurizer 2.1 2.1 Circumferential welds Longitudinal welds VT/UT VT/UT 2.6 Manway bolts VT I Nozzle to safe end welds Skirt weld VT/PT/UT VT/PT I Steam Generator Loop A 3.1 3.3 Channel Head to Tube Sheet Weld Nozzle to safe end weld VT/UT VT/PT/UT 3.5 Manway bolts VT I Steam Generator 3.5 Manway bolts VT Loop B I Steam Generator 3.3 Nozzle to safe end weld VT/PT/UT Loop C I Reactor Coolant 3.5 4.1 Manway bolts Pipe to safe end welds VT VT/PT/UT Piping I 4.2 Circumferential Butt Welds VT/UT Auxiliary Piping 4.4 Pressure retaining bolting VT I 4.2 Circumferential Butt Welds Nozzle root connections VT/UT VT/UT Socket Welds VT/PT Nozzle root connections VT/PT I 4.5 4.6 Integrally welded supports Piping supports and hangers VT/PT VT I Reactor Coolant Pump (Loops A, 5.5 5.7 Seal Housing Bolting Support structures VT VT B, C) 5.8 Flywheels VT/UT I Valves 6.7 Supports and hangers VT Miscellaneous 8 .1.1 Circumferential welds and VT/UT I Inspections branch connections larger than 4" diameter 8.1.3 Socket welds and branch VT/PT I connection welds 4" diameter and smaller 8.2.1 Containment and recircu- VT/PT I 8.2.2 lation piping Remaining sensitized stain-less steel piping and cold VT/PT bends I High Energy Designated high energy line VT/UT Line Piping welds in TS Figure 4.15 I All welds exceeding 4" OD TS Figure 4.15 other than VT/UT designated welds I

I I II. INSPECTION

SUMMARY

I Inservice examination of Class I components and piping systems, sensitized stainless steel piping, and main steam and feed water piping I was accomplished from November 11, 1974 to December 12, 1974. The in-spections performed utilized visual, surface, and volumetric non-I destructive testing methods. The extent of which S)'Stems and components are subject to inspection was established in accordance with ASME Section I XI, Rules for Inservice. Inspection of Nuclear Reactor Coolant Systems, I with the Summer 1972 addenda. Procedures, UT instruments, UT transducers, UT Calibration Block Certifications, Couplant Certifi~ation, *Liquid I Penetrant Material Certifications, and Personnel Qualifications were reviewed and approved. The arrangement and detail of the Unit No. 1 I piping systems and associated components were designed. and fabricated before I any of the examination requirements of Section XI of the Code were formalized.

Consequen.tly, the performance of the examinations has been limited to the I extent practical due to accessibility and geometric configuration.

The piping systems of Unit No. 1 contain welds which are- inaccessible I for examination or examinations are limited to less than 100 per cent I of the weld and adjacent base material. Elbow, valve, and tee configuration restricted angle beam examination of the weld and lT on each side as I required by the Code. These welds were generally examined by the following techniques: (1) 100 per cent angle beam of the weld and from the pipe I side; (2) longitudinal wave inspection of the pipe side, weld metal, I and component areas where search unit contact is possible within the one weld thickness zone; and (3) partial angle beam examination from I the component side, search unit contact permitting. This technique I

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I satisfies code requirements for inspection of the weld, but does not inspect base metal for lT on component side of the weld.

I The calibration standards utilized for weld inspections were I machined from pieces of the same grade material as the systems being inspected. Transfer functions were performed in all material exceeding I one inch thickness where required. Transfer corrected gain adjustments were made when required.

I The auxiliary piping examination results record notes examination restrictions due to geometry and obstructions. Removal of existing hangers I for access was not considered because of the high risk of stressing adjacent I piping weldments in the system.

A volumetric examination of integrally welded pipe supports was not I accomplished since meaningful volumetric examination of these supports cannot be made with present techniques. As a substitute, a liquid penetrant I and a visual examination was performed on those supports which would have I been examined by ultrasonics.

Liquid penetrant examination of the sensitized piping cold bends was I accomplished by examining a minimum area of one inch wide by one foot long on each bend and recording the location of the area examined.

I Ultrasonic examination of the main steam designated pipe welds, Loop I 1 Weld Number 1, Loop 2 Weld Number 155, Loop. 3 Weld Number 275, could not be accomplished as these welds are well inside the containment wall I penetrations and are not accessible. As a substitute for these welds, the next available weld in the main steam valve house was ultrasonically I examined. These welds are designated Westinghouse Weld Number 13 for I Loops 1, 2 and 3.

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I I III. INSPECTION RESULTS The inservice inspections performed during the period covered by I this report included the basic areas listed below:

I 1. Reactor coolant system, including reactor vessel, pressurizer, steam generator welding and bolting, I auxiliary piping, reactor coolant piping and valves.

2. Piping systems containing sensitized stainless steel, I including safety injection system, charging system, I reactor coolant system (lines less than 4 inches in diameter), containment spray system, recirculation.

I spray system, and other miscellaneous piping containing sensitized stainless steel.

I 3. Designated high energy line welds described in Technical I 4.

Specification 4.15.

Low pressure turbine rotor blades.

I 5. Materials irradiation surveillance capsule.

6. Steam generator tube inspections.

I 7. Reactor Internals Inspection.

I Results of each of the above inspections are sununarized below.

1. Reactor Coolant System I Inservice examination of components and piping systems within this area were performed by the Westinghouse Electric Corporation. The in-I spections performed utilized visual, surface and volumetric non-destructive I testing methods. The results of examinations performed by the three methods and disposition of any indication are listed below.

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I a. Volumetric Examinations Volumetric examinations performed did not reveal I any rejectable flaw indications.

I b. Surface Examinations Surface examinations revealed the following I linear indications:

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Loop 1 - 2" Fill Header ISOMETRIC DRAWING NO.

RC-198-1502 WELD NO.

2 I Loop 2 - 2" Drain Header RC-57-1502 3 Loop 3 - 2" Fill Header RC-200-1502 2 I

All indications were located in the base material I of the adj~cent casting. The indications were I ground out, repaired and re-inspected by liquid penetrant. The re-inspections showed no rejectable I indications.

c. Visual Examination I Visual examination performed revealed the following I conditions:

SYSTEM ITEM: CONDITIONS I Loop 3 - Drain Header Welded support Arc Strike 2" Drain Header All welded channels I and U-bolts Support bracket T Carbon steel Broken plate I Loop 3 - Seal Injection Loop 3 - Pressurizer Pipe strap rod A Spring hangers R Broken rod Readings off scale Spray maximum I Loop 3 - Accumulator Spring hangers D Readings off scale Discharge maximum I

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I SYSTEM ITEM CONDITIONS Loop 3 - Accumulator Spring hangers C Readings off scale I Discharge Loop 3 - Pressurizer Spring hangers C maximum Readings off scale I Surge maximum The indications on the welded support on the Loop 3 drain I header were removed and re-inspected by liquid penetrant I and visual examination. Re-inspection showed no indications.

The remaining items were repaired and/or adjusted to proper I setting.

I Vepco personnel performed the examination of the reactor I vessel closure studs and the primary nozzles* to safe end welds by visual and/or liquid penetrant non-destructive I test methods. The results of the examinations were accept-able, with only small rounded indications on the primary I nozzles to safe end welds indicated by liquid penetrant I 2.

examinations .

Piping Systems Containing Sensitized Stainless Steel I Piping systems containing sensitized stainless steel were visually examined. A number of arc strikes were noted in the valve pit area. The I indicated arc strikes are not significant; however, they will be removed I for future inspections.

3. High Energy Line Piping I The high energy line welds designated in Technical Specification 4.15 were volumetrically examined by ultrasonic non-destructive testing I methods. No discrepancies were noted.

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I I 4. Low Pressure Turbine Rotor Blading The low pressure turbine blading was examined by visual and surface I non-destructive testing methods. The results of the inspection and I disposition are sununarized below:

AREAS METHOD OF I EXAMINED EXAMINATION INDICATION DISPOSITION Blading Visual Arc Strikes Ground out and re-I Magnetic 5 cracked last inspected satisfactorily Blades were replaced particle stage blades I Stellite erosion shields Visual One shield missing Replaced shield I Lashing lugs Liquid Penetrant Cracked lashing lugs Ground out and repair welded. Re-inspected satisfactorily.

I Undershroud welds Liquid Cracking in Ground out and repair Penetrant under shroud welded. Re-inspected I welds satisfactorily.

5. Material Irradiation Surveillance Capsule I Battelle Columbus Laboratories is presently examining the first I surveillance capsule. The Technical Specification requirements were satisfied.

I 6. Steam Generator Tube Examinations Dur.ing the refueling outage, Eddy current inspections were performed I on all three steam generators. In conj.unction with the conversion to all I volatile treatment of the feedwater, sludge lacing was accomplished in parallel with the inspection program.

I The inspection program required examinations at 400 KHZ to detect and measure potential tube defects with supplemental examinations at I

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I I 100 KHZ to provide an assessment of low level wall thinning and at 25 KHZ to measure sludge deposits on top of the tubesheet. A nominal 100 I per cent inspection was also performed on the inlet side of steam generator A up to the first support. This was accomplished along with U-bend in-I spections of peripheral tubes. Based on the results of the planned in-I spection on steam generator A,* the inspection program for steam generators Band C was expanded to include those areas of the tubesheet array where_

I defects were noted in steam generator A.

.1 A total of 3852 tube inspections, including inlet and outlet, were performed at.40() KHZ in steam generator A~ Additionally, 1243 tube in-I spections were performed in steam generator C* . As a -result of this in-spection program, 55 tubes were explosiveiy plugged in steam generator A, I 2 in Band 84 tubes in C steam generator. All those tubes plugged in steam* generator A exhi_bited defects of 50 per cent or greater. The two I tubes plugged in B steam generator having defects of 30 per cent and 38 I per cent were plugged. Two of the 84 tubes plugged in steam generator Chad defects of 25 per cent and 47 per cent. All others exhibited I defects of -50 per cent or greater.

A summary of the inspection results for each steam generator is I given below:

I Steam Generator A A total of 3262 tubes were examined at 400 KHZ from the inlet side.

I Of these tubes, 271 exhibited detectable wall :Penetration (i.e. >20 per cent), 55 defects were equal to or greater than 50 per cent, 81 were I between 40 and 49 per cent, 92 were between 30 and 39 per cent and 43 I fell in the range of 21 to 29 per cent. An additional 80.tubes were .

. noted to have probable *defects less than or equal to- 20 per' cent.

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I I Outlet inspections of 400 KHZ were performed in 580 tubes. Only 4 tubes were noted to have defects in excess of 20 per cent and these I were confined to the 21 and 29 per cent range. Three others exhibited I defects of 20 per cent or less.

Crud level measurements were made from both inlet and outlet sides I at 25 KHZ prior to sludge lancing. Following the lancing, 25 KHZ sludge measurements were performed from the inlet side only.

I Steam Generator B I Inlet side inspections were made at 400 KHZ to a total of 1204 tubes.

Three of these tubes exhibited detectable wall penetration, two were in I the range of 21 to 29 per cent and one between 30 and 39 per cent. Another 39 tubes were noted to have defects less than or equal to 20 per cent.

I No unacceptable defects were noted.

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Thirty-nine tubes were inspected from the outlet side with no defects I Crud level measurements were made from both inlet and outlet sides following the sludge lancing effort.

I Steam Generator C I Of the 1514 tube inspections performed at 400 KHZ from the inlet side, 206 tubes exhibited detectable wall penetration (i.e. >20 per cent).

I Eighty-two tubes had defects of 50 per cent or greater, 66 tube defects fell in the range of 40 to 49 per cent, 39 defects in the range of 30 to I 39 per cent and 19 tubes had defects in the range of 21 to 29 per cent.

I Another 57 tubes were noted to have defects less than or equal to 20 per cent.

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I A total of 3 detectable defects were noted in the 84 tubes examined from the outlet side. All of these were in the 21 to 29 per cent range.

I Another 4 defects were detected and assessed to be. less than or equal to I 20 per cent.

Crud level measurements were made from both inlet and outlet sides I following the sludge lancing program.

7. Reactor Internals Inspection I During the refueling all of the fuel assemblies were removed from I the reactor vessel. As a result, the opportunity. was taken to perform an inspection of the reactor internals and core components. However, the I reactor lower internals were not removed; therefore, the outside of the lower internals and certain other parts of the reactor vessel were not.

I inspected. The Technical Specifications requirements were satisfied.

I The emphasis of the inspection was on the upper internals, accessible areas of the lower internals, vessel clad, and selected control rod drive I

• shafts and rod cluster control assemblies (RCCA's).

All inspections were carried out by certified inspectors using I closed circuit underwater television systems and recorded on video tape.

I The reactor internals were found to be in good condition and only one defect was noted. The locking cups for fasteners of the upper guide I

• tube on the upper internals at location D...;.4 had not been crimped during construction. These locking cups were crimped in*the field. utilizing I approved procedures.

I Information of particular importance in the evaluation of each of the reactor components inspected included the following:

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I a. Condition of .wear and alignment surfaces.

b. Integrity of critical structural joints.

I C, Condition of mechanical fasteners, alignment pins I d.

and locking devices.

Free movement of movable components.

I e. Corrosion product formations.

f. Presence of debris, I g. Mechanical distortion I Components included in the evaluation were the following:

a. Reactor upper internals I b.

• Reactor lower internals c, Drive line components I d. Reactor vessel I The program included the inspection and evaluation of. certain important areas of the internals which are good indicators of performance. The I overall effects of conditions found were evaluated on the basis of design information, experimental results and previous experience with I other reactors.

I A small diameter television camera was lowered through the irradiation specimen access holes at the 65 degree and 285 degree locations for the I purpose of in~pecting the vessel clad at these locations.

Two six (6) by six (6) inch patches were inspected at each location. One patch at I each location was adjacent to and above the top of the irradiation specimen basket at each location. The second patch was approximately I six (6) feet higher *. All four patches of vessel clad appeared to be in I excellent condition.

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• I I IV. CONCLUSIONS The results of the inservice inspections performed verified the I integrity of the systems and components examined and satisfied the I requirements of Technical Specification 4.2.

were corrected.

The discrepancies noted I Based on the results of the inservice inspection program, as sunnnarized herein, the safety systems and components inspected have I not experienced degradation and there is reasonable assurance that they I will continue to perform their design function in a safe and satisfactory manner.

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I I v. REFERENCES

1. Reactor Internals and Core Components Evaluation, Surry Unit 1, November 1974, Refueling Shutdown, March 1975, Westinghouse I Electric Corporation.

2. Report of .Inservice Inspection Conducted EI. Vepco Personnel, I Refueling Outage No. 1, Unit No. 1, Surry Power Station, March 1, 197 5, Virginia Electric and Power Company.

I 3. Inservice Inspection Report of the Surry Unit No . .!. Nuclear Power Station, January 1975, Westinghouse Electric Corporation.

Inservice Inspection Program, Refueling Outage No * .!., Surry Power I 4.

Station, Unit No. 1, ISI 74-1, July 18, 1974, Virginia Electric and Power Company.

I 5. Section 4.2, Technical Specifications, Surry Power Station, Unit Nos. 1 and 2, Virginia Electric and Power Company.

I 6. Field Service Report MRS 4.4 (VPA-2), Surry No. 1-VPA, Westinghouse Electric Corporation.

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• 7. Battelle Columbus Laboratories letter of April 22, 19.75.

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