ML20212N546

Comprehensive Vibration Assessment Program for Reactor Internals During Preoperational Testing

ML20212N546
Person / Time
Site:	Seabrook
Issue date:	08/31/1986
From:	PUBLIC SERVICE CO. OF NEW HAMPSHIRE
To:
Shared Package
ML20212N542	List: ML20212N539 ML20212N546
References
PROC-860831, NUDOCS 8608280277
Download: ML20212N546 (10)

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! Seabrook Station Unit No.1 .

l Docket Number 50-443 ..

COMPREHENSIVE VIBRATION i ASSESSMENT PROGRAM FOR

! REACTOR INTERNALS DURING

PREOPERATIONAL TESTING l

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COMPREHENSIVE VIBRATION . ASSESSMENT ..

PROGRAM FOR REACTOR INTERNALS DURING PREOPERATIONAL TESTING PUBLIC SERVICE COMPANY OF NEW HAMPSHIRE NEW HAMPSHIRE YANKEE DIVISION SEABROOK, NEW HAMPSHIRE SEABROOK STATION UNIT 1 DOCKET ! UMBER S0-443 AUGUST 1986

TABLE OF CONTENTS

.G REFERENCES 111 INTRODUCTION 1

SUMMARY

2 FEATURES EXA!4INED 3 CONCLUSION 5 11 8

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REFERENCES

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1. U.S.N.R.C. Regulatory Guide 1.20, Comprehensive Vibration Assess-ment Program for Reactor Internr.ls During Preoperational and Initial Startup Testing, Revision 2, May 1976.

2. Seabrook Station FSAR Subsection 3 9(N).2 3, Dynamic Response Analysis of Reactor Internals Under Operational Flow Transients and Steady State Conc'itions.

3 Seabrook Station FSr.R Subsection 3 9(N).2.4, Preoperational Flow-Induced Vibration Ttst.ng of Reactor Internals.

4. Seabrook Station FSAR Table 14.2-3, Item 43, Reactor Post-Hot Functional Inspection.

5. Seabrook Station Preoperational Test,. 1-PT(I)-43, Revision 1, Reactor Post-Hot Functional Inspection.

6. Seabrook Station Detailed Cleaning Package, RC-F-5.

7. Westinghouse Drawing 6125?73, Vibrational Check-Out Functional l Test Inspection Data.

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8. Westinghouse Assembly Specification 2463A68.

4 9 ANSI N45.2.1-1973, Cleaning of Fluid Systems and Associated Components Durind Construction Phase of Nuclear Power Plants.

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INTRODUCTION This report describes the inspection program of reactor internals and is submitted in accordance with USNRC RG 1.20, Rev. 2.

The Indian Point No. 2 plant has been established as the prototype for a four-loop plant internals verification program, and was fully I

instrumented and test:sd during Hot Functional Testing.

I Seabrook Station is similar to Indian Point No. 2. The only signifi-

cant differences are the modifications resulting from the use of I 17 x 17 fuel, replacement of the annular thermal shield with neutron shielding panels, and the change to the UHI-style inverted top hat i support structure configuration. These differences are addressed in detail in FSAR Subsection 3 9(N).2 3 Since the Seabrook reactor internals design configuration is well characterized, the vibration measurement program was omitted and the inspection program was imple-mented, as discussed in USNRC RG 1.20.

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SUMMARY

The overall inspections performed on the. reactor internals were governed by Westinghouse Assembly Specification 2463A68. This specifi-cation documents the verification of critical areas on the internals package from initial installation prior to Cold Hydre Testing through Hot Functional Testing up to the final Westinghouse Quality Release prior to core loading.

Incorporated as part of this series of internals inspections is Westinghouse drawing 6125E73, vibrational Check-Out Functional Test Inspection Data, which delineates 28 items which were verified both prior to Cold Hydro Testing and after Hot Functional Testing.,

Documentation of the completion of these inspections and any findings are included as part of prenperational test 1-PT(I)-43, Reactor Post-Hot Functional Inspection.

Examinations included visuni inspection of welds, surface conditions ,

and locking devices and feeler gauge verification of critical contact aaeas. Welds were examined using 5-10x ma.nification. Indications of cracks located on the vessel clevis pins were discovered prior to testing and removed after Hot Eunctional Testing. No other indica-tions vere evident. Interface sur f ac'es were inspected for any evidence of damage. None was found. Bearing surfaces were examined using 5-10x magnificatioa for evidence af (amage. No damage was found. Locking devices were verified to 'oe crimped and undamaged.

Inserts were exanined for the required maximum gap using the

, appropriate feeler gacger. Two upper core plate inserts were found to have excessive gap and t'ere subsequently restored to meet the design gap and contact area. No apparent defects were found at the other insert locations.

During the performance of the vessel and internals inspections, and any work resulting thereof, the refueling cavity access was controlled by appropriate houcekeering requirements. Tool control was maintained to assure that equipment could not be left in the system. The inter-nals were maintained wrapped in plastic and the vessel covered when-ever work on these components was not in progress. After Hot Functional Testing, personnel working on the internals wore clean overalls and protective foot covering.

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4 FEATURES EXAMINED l

A. All major load-bearirag elecents of the reactor in to'rnals relied upon to retain the core structure in place.

l - Hold down spring interface surface condition. ,

j - Accessible welde on suppcet column lower nozzles.

- Upper support skirt to plate girth weld. 'i

- Upper support skirt to flange girth weld.

] - Upper barrel to flange girth weld.

- Upper barrel to lower barrel girth weld.

j - Lower barrel to core support girth weld.

j - Core support columns and their screw locking devices.

. - Secondary core support housing to base plate weld.

! l i B. The lateral, vertical and torsional reatraints provided within the t j vessel.

i. - Upper core plate inserts.

- Upper core plate aligning pin welds and bearing surfaces.

- Radial support key welds, a

1 C. Those locking and bolting devines whose failure could adversely l affect the structural integrity of the internals.

j - Accessible support column and core plate insert screw locking  !

devices.  !

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- Lower core plate to core barrel screw locking devices accessible at the O*, 90', 180' and 270* axes.

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- Locking devices and contact of the cruciform columns where

attached to the' lower core support forging and tie plates.

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- Locking devices and contact of the secondary core support offset and butt columns at the lower core suppert forging and at the tie plates.

- Radial support key locking arrangements and bearing surfaces.

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- Head and vessel aligning pins screw locking deviceand bearing surfaces.

- Vessel clevis locking arrangements and bearing surfaces.

D. Those other locations on the reactor internal components which are similar to those which were examined on the prototype designs.

- Upper support column nut to extension welds.

- Accessible guide tube welds.

- Outlet nozzle interface surface condition.

- Neutron shield panci screw locking devices.

- Interface surfaces at the spacer pads along the top and bottom ends of the neutron panels.

- Baffle assembly screw locking arra~ngements at the twc top and the two bottom former elevations.

- Irradiation specimen guide screw looking devices and dowel pins.

- Vessel nozzle interface surface condition.

- Guide tube screw locking devices.

E. The inside of the vessel will be inspected before and after the Hot Functional Test, with all the internals removed, to verify that no loose parts of foreign material are in evidence.

Prior to Hot Functional Testing, the vessel, closure head, upper internals and lower internals were inspected to a cleanliness level that meets and exceeds Class B requirements as specified 'n .

ANSI N45.2.1-1973 After Hot Functional Testir.g, the vessel, head and internals were inspected for foreign maturial. The full flow filters were removed and inspected for any debris. No loose parts were found. The inspection prior to testing is documented by pro- ,

cedure RC-F-5. Inspections performed after testing are documented by preoperational test 1-PT(I)-43, Section 6 3 4

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to CONCLUSION During Hot Functional Testing the internals were subjected to greater than 240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br /> of operation at greater than normal full-flow con-i ditions This subjected the main structural internals elements to 1 over 10I cycles.

i three pumps operating Additionally, in thefour-loopthere was operation system. with one, welds, Important two, or bearing surfaces, and alignment and locking devices in the internals i have been visually inspected.

} The combination of tests, predictive analysis, and post-test inspec-tion provide adequate assurance that the reactor internals will, l during their service lifetime, withstand the flow-induced vibration of reactor without loss of structural integrity.

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