ML17213B127

Comprehensive Vibration Assessment Program,Final Summary Rept.

ML17213B127
Person / Time
Site:	Saint Lucie
Issue date:	02/28/1983
From:	Robert E. Uhrig FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17213B125	List: ML17213B124 ML17213B127
References
RTR-NUREG-0843, RTR-NUREG-843 CEN-244(L), NUDOCS 8303150538
Download: ML17213B127 (15)

Text

Florida Pov!er 5 Ligh>> Co., St. Lucie Plant Unit -.":2 Docket Hn. 50-389 CVi-2ee(L }

Comprehensive Vibration Assessment Prnoram Final Summar Re ort February 1983 Combustion "-ngineering, Inc.

Nuclear Pou,'er Systems Power Systems Group

!!indsor, Connecticut 06095 8303i50538 830310 PDR ADOCK 05000389 E PDR

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TAP>LE OF CO<'lTEl!TS

1. I tlTRODUCT I 0,'l SUtUQRY AND CO."iCLUSIOl'lS

3. VI BRAT IOil AiNLYSIS PROGRAll

4. VISUAL Il(SPECTIOil PROGPAW 2of 9

Florida Power and Light Cn., St. Lucie Plant Uni t;-2 Comprehensive Vibration Assessment Program IHTROOUCT IOll The Comprehensive Vibration Assessment Program (CVAP; occasionally referred to as the Precritical Vibration!lonitoring Program, PV!1P) reported herein satisfies the HPC Regulatory Guide 1.20 (Reference 1), requirements for verifying the structural integrity of the reactor int mals for flow induced vibrations prior to commercial operation. The CVAP prnvideh confirmation, based upon prototype PVhP programs, an analytical progran and an inspection program, that the hydraulic excitations and structural responses oZ the Florida Power and Light, 1

St. Lucie Plant Unit 2 reactor internals are within design estimates and are acceptable for all normal steady state and transient modes of reactor coolant pump operation.

As stated in Reference 2, Section 3.9.2.4, the ilaine Yank e and Fort Calhoun reactors are designated jointly as the Valid Prototype for the St. Lucie Plant Uni . 2 CVAP, with St. Lucie Plant Unit 2 designated as a !ion-Prototype Category 1 reactor. Reference 3 {Section 3.9.2.3) states that the .'lRC staff has accept-ed the St. Lucie 2 program provided that "the applicant submits a correlation of the St. Lucie 2 observed vibrational characteristics with the results from the prototype reactors".

Reference 1 requires that an aralysis program and a measurement or inspection program be performed for the CVAP for !!on-Prototype Cateqory 1 reactors. The analysis program for St. Lucie Plant Unit 2 CVAP was reported in Reference 2, Section 3.9.2.6. A visual inspection program with photographic documentation was performed for St. Lucie Plant Unit 2 in lieu of a measurement program.

This report summarizes the results of the St. Lucie Plant Unit 2 CVAP and provides an evaluation of those results.

2. SUi!l'!ARY Ai'l0 CO!lCLUSIOl!S The St. Lucie 2 CVAP was successfully completed in accordance with the requirements of .'lRC Regulatory Guile 1.20, Revision 2 {Reference'1).

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The vibration analysis program, performed in accordance with regulatory position C.3.1.1 of Reference 1, provided sufficient evidence to support the classification of St. Lucie 2 as Non-Prototype, Category 1, with tho Valid Prototype designated jointly as Viaine Yankee and Fort Calhoun. Comparison of the St. Lucie 2 results with thos. from the prototype reactors was favorable, no corrective action was required and no indications were observed that would necessitate reactor int mals modifications on St. Lucie 2.

The St. Lucie 2 vibration inspection program, perfor.",ed in accordance with the guidelines nf reoulatory position C.3.1.3 and C.2.3 of Reference 1, included inspections of the St. Lucie 2 reactor internals both prior to and following pre-core hot functional testing. The pre-core hot functional testing included all steady-state and transient modes of reactor coolant pump operation. Heither,real nor durity fuel assemblies were in position for the testing. It was shown by analysis that tho absence of fuel assemblies would yield conservative results for the CVAP. of reactor internals. The critical reactor internals component wi th the lowest natural frequency is the Core Support Barrel (CSB). Based upon the minimum significant response frequency of the CSB, the critical reactor internals conponents wer subjected to approximately 8.6 X 106 cycles. of vibration during the pre-core hot functional testing. t(OTE: Regulatory Guide 1.20 Revision 2 recommends,a minimum of 1 X 196 cycles.

The inspection program was performed without deviation from the specified operating conditions. Ho unanticipated observations or inspection anomolies were encountered. The insp ctions of t"..e St. Lucie 2 reactor internals revealed no defects; evidence of unacceptable motion, or excessive or undue wear. The interior of the reactor vessel was visually inspected after the pre-core hot functional testing and found to be absent of any loose parts or foreign material.

In summary, the St. Lucie 2 CVAP inspection program was entirely consistent with the P'/llP of the Haine Yanf:ee.and Fort Calhoun reactors and v)ith the St. Lucie 2 CVAP analysis program.

Evaluation of the results of the St. Lucie 2 CVAP concludes that a significart margin of. safety for the structulal integrity of the St. Lucie 2 reactor 4of 9

V internals will be maintained durino all normal steady-state and transient conditions of reactor coolant pumo operation.

VIBRATIO!I AHALYSIS PROGR,"Jl The f1aine Yankee and Fort Calhoun reactors together constitute a Valid Prototype for the purpose of the St. Lucie 2 CVAP. The St. Lucie 2 Plant reactor in-ternals configuration has substantially the same arrangement, design, size operation conditions as the Valid Prototype. Hominal differences in

'nd arrangement, design, size and operating conditions have been shnv!n by test or analysis to have no significant effect on the vibratory response and ex-citation of those reactor internals important to,safety; for these reasons, the St. Lucie 2 reactor is designated Won-Prototype, Category 1, for the CVAP.

As mentioned in Reference 2, Section 3.9.2.4, theoretical prediction analyses were performed for,ilaine Yankeo (Reference 5) and Fort Calhoun (Reference 6) to estimate the amplitude, time, and spatial dependency of the steady-state and transient hydraulic and structural responses to be encountered during precritical testing. The PVl/P for f )aine Yankee and Fort Calhoun demonstrate that the theoretical predication methods used provided accurate estimates 1

of the steady-state response of the core support barrel system, when r asonable best estimate values for the magnitude of the inlet pressure fluctuations are used. It was concluded from these programs that flov!, induced vibrations of the!laine Yankee and Fort Calhoun reactor internals are well within design allowables and are acceptable for all normal steady-state, and transient flow modes of reactor coolant pump operation.

Reference 2, Table 3.9-4, presents a sugary of the significant hydraulic and structural design parameters for the St. Lucie 2, Haine Yankee and Fort Calhoun reactor designs. The effects of these structural and hydraulic parameters on the flov!-induced vibratory response of the reactor internals are presented in Reference 2, Section 3.9.2.6, where it is shown that the nominal differences have no significant effects on the stress 1 vels. In general, the analysis of the St. Luc i e 2 Unit demon stra t s tha t:

A. The predicted structural responses of the St. Lucie 2 reactor internals are well within design allowables and are acceptable for all normal steaRy-state and transient flov! modes of pr imary coolant pump operation.

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B. The prototype precritical vibration monitoring programs .or )/aine Yankee and Fort Calhoun adequately account for the specific design features of the St. Lucie 2 Unit which are shared by the Valid Prototype reactor designs.

VISUAL Il(SPECTIOll PROGRAM The St. Lucie 2 CVAP inspection program was performed per the procedure of Reference 4, which meets the intent of regulatory positions C.3.1.3 and C.2.3 of Reference 1. The inspection program includes photographic document-ation of the condition of the St. Lucie 2 reactor in.ernals, both prior to and after pre-core hot functional testing.

The inspection was conducted in two phases. The first phase (baseline inspect-ion) was completed on April 22, 1982. The second phase (post-hot functional, pre-core, inspection) was completed on January 10, 1983.

Peference 1 requires that the reactor internals critical components be sub-jected to at least 10 cycles of vibration prior to the CVAP final inspec,ion, based upon the component's computed minimum significant response frequency.

The St. Lucie 2 Core Support Barrel (CSB) was calculated to have the lowest natural frequency of the critical reactor internals components. During pre-core hot functional testing, the St. Lucie 2 reactor internals were subjected to 1.19 X 10 sec. of col'd flow (below 350'F) and 1.61 X 106 sec, of hot flow (above 350'F). Based upon the minimum signi icant response frequency of the CSB, the internals critical components were subject to approximately 8.6 X 10 cycles of vibration. .")either real nor Rummy fuel assemblies !vere included in the.hot functional testing. The lack of fuel serves to provide greater flow velocities and forces on reactor internals components and, therefore, yields conservative results for the CVAP.

The detailed inspection proceduro prepared in accordance with 'Reference 4 requires photographic documentation and 'descriptions of conditions observed during both phases, in addition to commentary on changes from the baseline inspection. The inspections were per ormed and quality assured by qualified inspectors. The inspection procedures provide the tabulation af all reactor internals components and local areas inspected, which includes:

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A. All major load-bearing elements of the reactor internals relied upon tn retain the core support structure in position.

B. The lateral, vertical, and torsional restraints provi ded within the vessel.

C. Those locking and bolting components whose failure could adversely affect the s.tructural integrity of the reactor internals.

0. Those surfaces that are known to be nr may become contact surfaces during operation.

E. Those critical locations nn the reactor internal components as identified by the vibration analysis.

F. The interior of the reactor vessel for evidence of loose parts or foreign matter.

The analysis program (Peference 2, Section 3.9.2.5) identified the core support barrel upper flange region to have the maximun stress intensity. This region was,included in the St. Lucie 2 inspections to v rify the results of the vib-ra ion analysis, that the maximum stress intensities are below allowable stress cri teri a.

A comparison of the baseline sur ace conditions with those of the post-hot functional inspec .ion indicated that no abnormal low-induced vibration had occurred and that no reduction in the structural integrity of the internals components, closure head or reactor vessel had occurred. There were indicat-ions of normal amounts of relative thermal growth between the stainless steel internals and the carbon steel vessel. At areas where contact occurred between core support barrel (CSB) snubbers, guide lugs, and alignment keys, little or no wear was indicated, but close fi;s were evident by discoloration and some surface burnishing. Contact between the reactor vessel, upper guide structure flange, CSB flange, and closure head appeared unifnm with no wear. All structural threaded fasteners and lockbars appeared secure and showed no indications of loading. The girth welds on the CSB all appeared sound as did the core shroud welds.

Due to lack of indication of abnormal movement and calculational results based on post-hot functional dimensions, it is concluderl that the internals were provided with adequate lateral an4 axial support. In oeneral, all in-ternals components were found to be in very good condition, their contact'of 9

II areas all appeared normal and as expected 4

following hot fUnctional testing and compared favorably with the prototype inspections.

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QEFEREHCES

1. "Comprehensive Vibration Assessment Prograr. for Reactor Vessel Internals During Preoperational and Initial Startup Testing.", IIRC Reoulatory Guide 1.20, Revision 2, dated hay 1976.'I

2. "Final Safety Analysis Report, St. Lucie Plant Unit 2", Docket I'!o. 50-389.

3. "Safety Evaluation Report by the Office of !uclear Reactor Regulation, U.S.

Huclear Regulatory Cornission, Related to the Operation of St. Lucie Plant Unit 2", Docket Ho, 50-389 and HUREG-0843.

4. "Precritical Vibration t'ionitoring Program Standard Procedure for Visual In-spection of Reactor Vessel Internals for 3410 iype Plants", Specification Hn.

00000-RCE-413, Revision 00, dated 6/12/80

5. "Analysis of Flow-Induced Vibrations: ilaine Yanke Precritical Vibration 11onitoring Program Predictions", Combustion Engineering, Inc. CEHPD-55, .

iiay 30, 1972.

6. "Analysis of Flow-Induced Vibrations: Fort Calhoun Precritical Vibration I/onitoring Program", Combustion Engineering, Inc., CE~!PD-85, January 1973.

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7. "'!aine Yankee Precritical Vibration ilonitoring Program, Final Report",

Combustion Engine ring, Inc., CE.'IPD-93, February 1973.

8. "Omaha Precritical Vibration lionitoring Program, Final Report", Combustion Engineering,, Inc.,"0 ~CEil-7 0, Iiey 1974.

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