ML19093A139

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

ML19093A139
Person / Time
Site:	Surry
Issue date:	01/20/1975
From:	Virginia Electric & Power Co (VEPCO)
To:	Office of Nuclear Reactor Regulation
References
ISI IR 1975001
Download: ML19093A139 (7)

Text

I I REGtJLATORy DOCKET FILE COJ>y, I

I PRELIMINARY RESULTS OF THE I INSERVICE INSPECTION PROGRAM I REFUELING OUTAGE NO. 1 SURRY POWER STATION I UNIT NO. l I

I JANUARY 20, 197 5 i'I REPORT NO. ISI 75-1 I 50-280 DOCKET Ng, DPR-32 LICENSE N.

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Vepco I

• I Virginia Electric an d Power Company

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I I I. INTRODUCTION I In accordance with the requirements of Technical Specification 6.6.C, this report contains a suIIUllary of the preliminary results of I the inservice inspection activities performed during the first re-I fueling outage *of Unit No. 1 at the Surry Power Station. The document entitled, Inservice Inspection Program, Refueling Outage No._!_," Surry I Power Station, Unit No._!_, Report No. ISI 74-1 dated July 18, 1974 provides the specific details concerning the inspections performed.

I This report is intended to be a preliminary report of the results I of the inservice inspections to be submitted prior to returning the unit to service following the first refueling outage, pursuant to I Technical Specification 6.6.C. A more detailed report will be submitted within 90 days.

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SUMMARY

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

Reactor coolant system, including reactor vessel, 1.

I pressurizer, steam generator welding and bolting, auxiliary piping, reactor coolant piping and I valves.

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

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I 3. Designated high energy line welds described in I Technical Specification 4.15.

4. 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 summarized below.

Reactor Coolant System I Inservice examination of components and piping systems within this area were performed during the period of November 11, 1974 to December 12, I 1974 by the Westinghouse Electric Corporation. The inspections performed I utilized visual, surface and volumetric non-destructive testing methods.

The results of examinations performed by the three methods and disposition I of any indication are listed below.

a. Volumetric Examinations I Volumetric examinations performed did not reveal any I b.

rejectable flaw indications.

Surface Examinations I Surface examinations revealed the following linear indications:

I SYSTEM ISOMETRIC DRAWING NO. WELD NO.

I Loop 1 - 2" Fill Header Loop 2 - 2" Drain Header RC-198-1502 RC-57-1502 2

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

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

indications.

Visual Examination I Visual examination performed revealed the following conditions:

I SYSTEM ITEM CONDITIONS I Loop 3 - Drain Header 2" Drain Header Welded support All welded channels and U-bolts Arc Strike Carbon steel Support bracket T Broken plate I Loop 3 - Seal Injection Pipe strap rod A Broken rod I Loop 3 - Pressurizer Spray Spring hangers R Readings off scale maximum Loop 3 - Accumulator Spring hangers D Readings off scale I D:i,.sc;:harge maximum Loop 3 - Accumulator Spring hangers C Readings off scale I Discharge maximum The indications on the welded support on the Loop 3 drain I header were removed and re-inspected by liquid penetrant and visual examination. Re-inspection showed no indications.

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

I Vepco personnel performed the examination of the reactor vessel closure studs and the primary nozzles to safe end I welds by visual and/or liquid penetrant non-destructive I

I I test methods. The results of the examinations were I acceptable, with only small rounded indications on the primary nozzles to safe end welds indicated by liquid I penetrant examinations.

I Piping Systems *Containing Sensitized Stainless Steel Piping systems containing sensitized stainless steel were visually I examined. A number of arc strikes were noted in the valve pit area.

The indicated arc strikes are not significant; however, they will be I removed for future inspections.

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

I 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 II disposition are sununarized below:

AREAS METHOD OF I EXAMINED Blading EXAMINATION Visual INDICATION Arc Strikes DISPOSITION Ground out and re-inspected I Magnetic 5 cracked last 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 I Liquid Cracking in satisfactorily Ground out and Undershroud welds I Penetrant under shroud welds repair welded.

Re-inspected satisfactorily.

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t Material Irradiation Surveillance Capsule t The materials irradiation surveillance capsule was removed from the reactor vessel. Battelle Columbus Laboratories is presently t performing the analysfs of the capsule. No preliminary results are available at this time.

Steam Generator Tube Examinations The steam generator tubes of all three steam generators were examined by eddy current non-destructive testing. The inspections were performed in accordance with Regulatory Guide 1.83. As a result of the inspection, approximately 137 tubes which had wastage greater than 50 per cent were plugged.

Reactor Internals Inspection An inspection of the reactor internals was performed during the third week of November 1974. All reactor internal items were found to be in normal condition with the exception of the locking cups on the guide tube upper flange bolts at position D-4 which were found not to be crimped. The uncrimped locking cups were crimped prior to completing refueling.

III. CONCLUSIONS The results of the inservice inspections performed verified the integrity of the systems and components examined. The discrepancies noted were corrected.

Based on the results of the inservice inspection program, as surrnnarized herein, the safety systems and components inspected have not experienced degradation and there is reasonable assurance that I

I they will continue to perform their design function.

As required by Technical Specification 6.6.C, a detailed report I of the inspections performed will be submitted with 90 days.

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