ML20010A293

to Comprehensive Vibration Assessment Program Final Rept.

ML20010A293
Person / Time
Site:	San Onofre
Issue date:	07/31/1981
From:	ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY
To:
Shared Package
ML13323A163	List: ML13323A162 ML20010A293 ML20010A297 ML20010A303
References
CEN-168-(S), CEN-168-(S)-R01, CEN-168-(S)-R1, NUDOCS 8108110286
Download: ML20010A293 (10)

Text

\

- r  :

R E C EIV E D Jt)L 1 0 1981 H.B. RAY San Onofre Nuclear Generating Station Unit 2 Docket No. 50-361 CEN-168 (s)

Revision 01 COMPREHENSIVE VIBRATION ASSESSffENT PROGR#1 FINAL REPORT ]

[ July 1981 Combustion Engineering, Inc.

Nuclear Power Systems

Power Systems Group Windsor, Connecticut 06095 l

l

,i .

%'9Q **

ED

- 8108110286 810807 PDR ADOCK 05000361 A PDR

1 r  %

i LEGAL NOTICE l

l THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COM3tisTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSOrJ ACTING ON ITS BEHALF:

A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN TH17 HEPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIWLITIES WITH RESPECT TO THE USE OF,OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.

9 4

l

~

l 4 __ ._ ._ _ _ _. ; z ! '_ __ __ .. . ___. - . 12 L . ._ _. . , 2 =- - ....__.-.-._._u____ ..___.n__ --

r ;__._ _ __-

. g }

Table of Contents

1. INTRODUCTION

2. SUffMRY AND CONCLUSION 3 2

3. VIBRATION ANALYSIS PROGRAM 3

4. VISUAL INSPECTION PROGRAM 4 REFERENCES 7 AlTACHitENTS:

1. " Precritical Vibration Monitoring Program Standard Procedure for Visual In-spection of Reactor Vessel Internals for 3410 Type Plants" Specification No.

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

2. 'Pracritical Vibration Monitoring Program Project Procedure for Visual In-susction of Reactor Vessel Internals for Southern California Edison Company San Onofre Nuclear Generating Station Units 2 and 3", Specification No.

4 1370-RCE-413. Revisi n 00, dated August 13, 1980.

h i

u -

San Onofre Nuclear Mnerating Station Unit 2 i Comprehensive Vibration Assessment Program 1, INTRODUCTION The Comprehensive Vibration Assessment Program (CVAP; fonnerly referred to as the Precritical Vibration Monitoring Program, PVMP) reported herein satisfies the NRC Regulatory Guide 1.20, Revision 1 (Reference 1), requirements for verifying the structural integrity of the reactor internals for flow induced vibrations prior to comercial operation. The CVAP provided confirmation, based upon prototype PYMP programs, an analytical program and an inspection program, that the hydraulic excitations and structural responses of the Sen Onofre Nuclear Generating Station Unit 2 (SONGS 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 Maine Yankee and Fort Calhoun reactors are designated jointly as the Valid Prototype for the SONGS CVAP, with SONGS 2 and 3 designated as Non-Prototype Category 1 reactors. Reference 3 (Question 112.7) and Reference 4 (Section 3.9.2.3) state that the NRC staff has conditionally accepted the prototype designation; with the acceptance con-l tingent upon the results of NRC's review of the Arkansas Nuclear One .- Unit 2 (ANO-2) augmented intarnals inspection, reported in Reference 5.

Reference 1 requires that an analysis program and a measurement or inspection program be performed for the CVAP for Non-Prototype Category 1 reactors. The analysis program for the SONGS 2 and 3 CVAP was reported in Reference 2. Section

. 3.9.2.6. A visual inspection program with photographic documentation was per-formed for SONGS 2 in lieu of a measurement program.

This report summarizes the results of the SONGS 2 CVAP and provides an evaluation l

.i of those results.

1 -

2. SUtttARY AND CONCLUSIONS The SONGS 2 CVAP was successfully completed in accordance with the requirements of NRC Regulatory Guide 1.20 Revision 1 (Reference 1).

)

The vibration analysis program, performed in accordance with regulatory posi-tion C.3.1.1 of Reference 1, provided sufficient evidence to support the classification of SONGS 2 as Non-Prototype, Category 1, with the Valid Proto-type designated jointly as Maine Yankee and Fort Calhoun. In support of the NRC's acceptance of the Valid Prototype designation, the results from the Arkansas Nuclear One - Unit 2 PVMP were provided in Reference 5; no corrective action was required and no indications were observed that would necessitate reactor internals modifications on SONGS 2 and 3.

The SONGS 2 vibration inspection progrem, performed in accordance with the guidelines of regulatory position C.3.1.3 and C.2.3 of Reference 1, included inspections of the SONGS 2 reactor internals both prior to and following pre-core hot functional testing. The pre-core hot functional tests'ng in-cluded all steady-state and transient modes of reactor coolant purg operation.

Neither real nor dunny fuel assenblies were in position for the testing.

It was shown by analysis that the absence of fuel assemblies would yield conservative results for the CVAP of reactor internals. The critical reactor internals component with the lowest natural frequency is the Core Support Barrel (CSB). Based upon the minimum significant response frequency of-the CSB, the critical reactor internals components were subjected to greater than 107 cycles of vibration during the pre-core hot functional testing.

The inspection program was performed without deviation from the specified operating conditions. No unanticipated observations or inspection anomolies were encountered. The inspections of the SONGS 2 reactor internals revealed no defects, evidence of unacceptable motion, or excessive or undue wear.

l 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.

2

I 1

f In sumary, the SONGS 2 CVAP inspection program was entirely consistent with -

the PVMP of the Maine Yankee and Fort Calhoun reactors (and with that of ANO-2) and with the SONGS 2 CVAP analysis program.

l Evaluation of the results of the SONGS 2 CVAP concludes that a significant

( margin of safety for the structural integrity of the SONGS 2 reactor internals will be maintained during all nomal steady-state and transient conditions of l

l reactor coolant pump operation.

I i

l 3. VIBRATION ANALYSIS PROGRAM The Maine Yankee and Fort Calhoun reactors together constitute a Valid Prototype for the purpose of the SONC3 2 and 3 CVAP. The San Onofre Units' reactor in-ternals configuration have subs tantially the same arrangement, design, size and operating conditions as the Val 1d Prototype. Nominal differences in arrangement, I design, size and operating conditions have been shown by test or analysis to have no significant effect on the vibratory response and excitation of those reactor internals important to safety; for these reasons, the SONGS 2 and 3 reactors are designated Non-Pretotype. Categcry 1, for the CVAP.

l As mentioned in Reference 2, Saction 3.9.2.4, theoretical prediction analyses l

l were performed for Maine Yankee (Reference 8) and Fort Calhoun (Reference 9) to estimate the amplitude, time, and spatial dependency of the steady state and transient hydraulic and structural responses to be encountered during pre-critical testing. The PVMP for Naine Yankee and Fort Calhoun were completed '

successfully and reported in Rcferences 10 and 11, respectively. Comparisons of the measured and predicted responses for Maine Yankee and Fort Calhoun demonstrate that the theoretical prediction methods used provided accurate i estimates of the steady-state response of the core support barrel system, when

! reasonable best estimate values for the umgnitude of the inlet pressure fluctuations are used. It was concluded from these programs that flow induced

' vibrations of the Maine Yankee and Fort Calhoun reactor internals are wel?

within design allowables and are acceptable for all normal steady-state, and transient flow modes of reactor coolant pump operation.

I 3

_ __. _ _ ___ _. __._ _._~ __ _ _ .._ _ _ - __

~

, i Reference 2. Table 3.9-7, presents a suunary of the significant hydraulic and structural design parameters for the SONGS 2/3, Maine Yankee and Fort Calhoun reactor designs. The effects of these structural and hydraulic parameters on the flow-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 levels. In general, the analysis of the San Onofre Units demonstrates that:

A. The predicted structural response of the San Onofre reactor internals are well within design allowables and are acceptable for all normal steady-state, and transient flow modes of primary coolant pump operation.

B. The prototype precritical vibration monitoring programs for Maine Yankee and Fort Calhoun adequately account for the specific design features of the San Onofre Units which are shared by the Valid Prototype reactor designs.

4. VISUAL INSPECTION PROGRN1 The SONGS 2 CVAP inspection program was perfonned per the procedure of References 6 and 7 (Attachments 1 and 2), which meet the intent of regulatory positions C.3.1.3 and C.2.3 of Reference 1. The inspection program included photographic documentation of the condition of the SONGS 2 reactor internals, both prior to and after pre-core hot functional testing.

The inspection was conducted in two phases. The first phase (baseline inspection) was completed on September 7,1980. The second pha% (post-hot functional, pre-core, inspection) was completed on April 20, 1981.

Reference 1 requires that the reactor intemals critical components be subjected to at least 10 7cycles of vibration prior to the CVAP final inspection, based upon the component's computed minimum significant response frequency. The SONGS 2 Core Support Barrel (CSB) was calculated to have the lowest natural frequency of the critical reactor internals components. During pre-core hc+.

functional testing, the SONGS 2 reactor internals were subjected to 11.5 days 0

of cold flow (below 3600F) and 43.3 days of hot flow (above 360 F). Based

/

4

uran the minimum significant response frequency of the CSB, the internals' 7

critical components were subjected to greater than 10 cycles of vibration.

Neither real nor dumy fuel assemblies were 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 report, prepared in accordance with References 6 and 7, includes photographic documentation and descriptions of conditions observed during both phases, in addition to commentary on changes from the baseline inspection. The inspections were performed and quality assured by qualified inspectors. The inspection procedures provide the tabulation of all reactor internals components and local areas inspected, which includes:

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

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

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

D. Those surfaces that are known to be or may become contact surfaces during operation.

E. Those critical locations on the reactor internal components es identified by the vibration analysis.

F. The inter-ior of the reactor vessel for evidence of loose parts or foreign matter.

' The analysis program (Reference 2. Section 3.9.2.6) identified .the core support barrel upper flange region to han the maximum stress intensity. This region 5

s was included in the SONGS 2 inspections tr verify the results of the vibration analysis, that the maximum stress intensities are below allowable stress criteria.

A comparison of the baseline surface conditions with those of the post-hot functional inspection indicated that no abnormal flow-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 indica-tions 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 r.o wear was indicated, but close fits'were evident by discoloration and some surface burnishing. Contact between the reactor vessel., upper guide structure flange, CSB flange, and closure head appeared unifom 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 any indication of abnormal movement and calculational results l

l based on post-hot functional dimensions, it is concluded that the internals I were provided with adequate lateral and axial support. In general, all in-ternals components were found to be in very good condition, their contact areas all appeared norraal and as expected following hot functional testing and compared favorably with the prototype inspections.

i 4 ,

6

. - , - - , . - ----n-.n. ,. , __ ._,.,,. .,,,n. , _ _ , , . , , , . , . _ _ _ _ , , . , , , , _ , , - .- , - -

REFERENCES:

1. " Comprehensive Vibration Assessment Program for Reactor Vessel Internals During Preoperational and Initial Startup Testing", NRC Regulatory Guide 1.20, Revision 1, dated June 1975.

2. " Final Safety Analysis Report San Onofre Nuclear Generating Station Units 2 and 3", Docket Nos. 50-361 and 50-362.

. 3. " Responses to NRC Questions, San Onofre Nuclear Generating Station Units 2 and 3", Docket Nos. 50-361 and 50-362, (Supplanent to Refers:nce 2).

4. " Safety Evaluation Report by the Office of Nuclear Kaactor Regulation U.S.

Nuclear Regulatt ry Commission, Related to the Operation of San Onofre Nuclear Generating Station Units 2 and 3", Docket Nos. 50-361 and 50-362 NUREG-0712.

5. "ANO-2 Precritical Vibration Monitoring Program Visual Inspection of the Reactor Internals with Photographic Documentation", Arkansas Power and Light, CEN-91 (A), Combustion Engineering, Inc., April 1978.

6. " Precritical Vibration Monitoring Program Standard Procedure for Visual In-spection of Reactor Vessel Internals for 3410 Type Plants", Specification No. 00000-RCE-413, Revision 00, dated 6/12/80 (Attachment 1 herein).

7. " Precritical Vibration Monitoring Program Project Procedure for Visual In-

spection of Reactor Vessel Internals for Southern California Edison Company San Onofre Nuclear Generating Station Units 2 and 3", S >ecification No.

1370-RCE-413. Revision 00, dated August 13,1980 (Attac zgent 2 herein).

8. " Analysis of Flow-Induced Vibrations: Maine Yankee Precritical Vibration Monitoring Program Predictions", Combustion Engineering, Inc., CENPD-55, 1 May 30,1972. ,

9. " Analysis of Flow-Induced Vibrations: Fort Calhoun Precritical Vibration Monitoring Program", Combustion Engineering Inc., CENPD-85, January 1973.

10. " Maine Yankee Precritical Vibration Monitoring Program Final Report",

Combustion Engineering Inc., CENPD-93. February 1973.

11. " Omaha Precritical Vibration Monitoring Program, Final Report", Combustion Engineering, Inc., CEN-7(0), May 1974. _

4 I

7

. - - - - - - _ _ _ . _ . _ _ , _ . , _ _ _ _ _ _ _ . _ , , . . . . , _ _