ML20133M523

Submits Info Re Other Facilities Affected by 790313 Order to Show Cause Concerning Use of Pipe Supports on 2 1/2 to 6 Inch Piping Designed for Loads Derived from Shock 2 Computer Calculations.Supporting Documentation Encl

ML20133M523
Person / Time
Site:	Beaver Valley, Surry, Maine Yankee, FitzPatrick, 05000000
Issue date:	01/03/1980
From:	William Kennedy STONE & WEBSTER ENGINEERING CORP.
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20133M133	List: IA-85-301, Forwards Draft Temporary Instruction Providing Guidance for Inspecting & Evaluating Licensee & CP Holder Responses to IE Bulletin 79-14, Seismic Analysis for As-Built.... Comments Requested by 790723.Supporting Documentation Encl ML20133M130 ML20133M162 ML20133M174 ML20133M190 ML20133M211 ML20133M223 ML20133M232 ML20133M247 ML20133M261 ML20133M293 ML20133M428 ML20133M451 ML20133M483 ML20133M505 ML20133M523 ML20133M555 ML20133M575 ML20133M607 ML20133M621 ML20133M635 ML20133M646 ML20133M662 ML20133M685 ML20133M699 ML20133M722 ML20133M737 ML20133M757 ML20133M771 ML20133M786 ML20133M842 ML20133M853 ML20133M891 ML20133M908 ML20133M919 ML20133M959 ML20133M996 ML20133N121 ML20133N130 ML20133N148 ML20133N166 ML20133N203 ML20133N381 ML20133N390 ML20133N407 ML20133N418 ML20133N421 ML20133N434 ML20133N440
References
FOIA-85-301 IEB-79-02, IEB-79-2, NUDOCS 8508130156
Download: ML20133M523 (31)

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STONE 6 WEBSTER ENGINEERING CORPORATION 245 Sumur.n SinscT. tiesvoN. M a ssacHuticTrs

, . AoOncss ALL COMHCbPONDENCE TO P.O. PO4 3329. notte'N. MASE c2107

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P0nf t as*D. C AE GON Mr. Harold Denton January 3, 1980 -

Director Office of Nuc1 car Reactor Regulation .

U. S. Nuclear Regulatory Commission -

i Washington, D. C. 20555

Dear Sir:

During the process of collecting pipe support loads for use in the IE Bulletin 79-02 Baseplate Evaluations, it came to our attention that some of the pipe supports on 2 1/2 to 6 inch piping on the Beaver Valley 1 Power Station are designed for loads that were derived from Shock 2 Computer calculations.

It had previously been thought that the calculations of record for these supports were simplified manual calculations.

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In response to the telephone request by your Messrs. Wigginton and Wychman, we have obtained the following information with respect to the other projects that were af fected by the March 13, 1979 Order to Show cause:

Surry 1 and 2 - No Shock 2 computer runs were identifled in connection with 6 inch and under piping.

Maine Yankee. - There were a number of Shock 2 runs identified in connection with 6 inch and under piping and all were reanalyzed as a part of the Show Cause reanalysis.

J. A. FitzPatrick- There were no Shock 2 runs identified in connection with 6 inch and under piping.

We do not believe that there are any safety implications from this finding since previous reanalysis of Shock 2 calculations did not result in any significant changes.

Very rul* yo s,

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-(%.. D idaho Falls, Idaho 83401 July 13, 1979 Mr. R. E. Tiller Director Reactor Operations & Program Division ,

Idaho Operations Office - DOE .  !

Idaho Falls. ID 83401

SEISMIC REEVALUATION OF PIPING ASSOCIATED WITH THE NRC SHOW CAUSE ORDER OF MARCH 13, 1979 BEAVER VALLEY POWER PLANT (A6156) - JAD-146-79

Dear Mr. Tiller:

Attachment 1 documents the technical assistance EG&G Idaho, Inc. provided to NRC-NRR in their re-review of certain Beaver Valley Power Plant piping systems.

Attachment 2 is the input provided by the NRC to perform the audit calculation described in Attachment 1 and is returned by this transmittal to R. G. LaGrange, NRC-DOR. This transmittal completes Node PA-16 page 21 of the A6156 PERT chart dated June 8,1979.

Very truly yours.

Q _ .E-J. A.

Dearien,

Manager Code Assessment and Applications Program BFS:snt Attachments:

As Stated cc: LR. G. LaGrange, NRC-DOR s R. W. Kiehn, EG&G Idaho w/o attachments l

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BEAVER VALLEY POWER STATION UNIT 1 PIPING REEVALUATION D. K. Morton G. L. Thinnes R. W. Macek

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INTRODUCTION EG&G Idaho personnel are, at the request of NRC-DOR Engineering Branch, participating in the seismic review of safety related piping systems at the five plants affected by the NRC Show Cause Order of March 13, 1979. The scope of this effort includes participating in meetings of the plant specific NRC review teams, performing audit calculations, and performing miscellaneous technical assistance.

In response to a March 21, 1979 request from the Nuclear Regulatory Commission's Division of Operating Reactors, a visit was made by EG&G Idaho and NRC personnel to the Beaver Valley Power Station (BVPS) Unit 1 on March 23, 1979. The purpose of the visit was to acquaint the personnel with BVPS and prepare for the review. Three meetings were held in Boston, Massachusetts (March 29 & 30,~ April 11 &

12, and June 5, 6, & 7.) at Stone & Webster to initiate the review of ,

the various piping system analyses.

Five safety related systems (Stone and Webster Problems 203C, 783A, 204, 255A, and 1200B) were chosen for independent audit review.

Figures 1 through 7 are computer model plots of these piping systems.

EG&G Idaho received the final data necessary to perform the review on April 23. Preliminary results were supplied over the telephone to the NRC on May 4 and selected EG&G Idaho results, compared to Stone and Webster results, were transmitted to the NRC on May 8.

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., ANALYSIS Tiiis audit review entailed performing response spectrum seismic analyses using independently formulated finite element models.

Several parameters were varied in the audit review to assess areas of concern. The Stone and Webster results are believed to be based on i rigid (=10 16 lb/in.) rather than an estimate support stiffness value 0 7

(=10 -10 lb/in.) and not to have considered the current requirements for closely spaced modes as outlined in Regulatory Guide 1.92. Some uncert'ainty exis'ted as to what change in magnitude would occur when using SRSS instead of algebraic summation methods of

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intramodal spatial combination. Table I defines the parameters varied for performing case by case comparisons. No valve eccentricities were supplied and, therefore, were not reflected in the audit calculations.

4 Two sets of earthquake spectra were used in each analysis. The first was for an operational basis earthquake (OBE) and the second was for a design basis earthquake (DBE). Figures 8 through 19 are the response spectrum used in the audit calculations. Table II lists which types of seismic displacements were supplied with each problem. .

Seismic displacements were applied exactly as Stone and Webster data indicated. No attempt was made to obtain a worst case combination.

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The EG&G Idaho audit calculations were performed using the Nuclear Services Corporation (Campbell, California) computer code NUPIPE-1I. This version of NUPIFE-II uses normal SRSS techniques for 2

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TABLE I CASES COMPLETED OBE DBE Best Estimate Restraints Yes Yes Rigid Restraints Yes Yes Without Closely Spaced Yes Yes ,

J Modes Algebraic Sum Yes Yes SRSS Yes Yes TABLE II SEISMIC DISPLACEMENTS OBE DBE OBE DBE Intermediate Intermediate S&W Problem Anchor Anchor Constraint Constraint 203C No Yes -

No No 204 No No No No No Yes No No 12008 255A Yes Yes Yes Yes -

783A Yes Yes Yes Yes 3

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combining seismic response. The " algebraic sum" cases were performed using a version of the NUPIPE-II computer code modified specifically for this effort.

RESULTS A direct comparison between EG&G Idaho results and Stone and Webster SHOCK II results was made (Tables III-VIII). The selected results compared reasonably sl1 except for Problems 783A and 1200B.

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results using algebraic sum versus SRSS on the seismic response combination. Problem 783A also indicates a significant difference in results (factor of 2.3) due to the seismic response combination method used. In addition, Problem 783A highlights how stress results can change by using an estimated support stiffness rather than assuming rigid supports. In this one case, stresses due only to seismic response increased by a factor of 2.2.

CONCLUSION The results of the BVPS audit calculations indicate that seismic stresses may be significantly altered depending on support stiffnesses used and which method of seismic response combination (algebraic sum vs SRSS) is employed. However, all calculated stress levels were below code allowables. .

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N: BEAVER VALLEY LINE: 16"-NFPD-22-601 PRO 8 NO.: 783A Snubber & Hanger Snubber & Hanger Load Case Support Stiffness Support Stiffness Ratio S&W Allowable Estimate Rigid (1)O G)q)

SRSS Q) Algb Sun @ SRSS @ Algb Sun @

Dead Load + P 6022 X X 6035 7299 15,000 X X Dead Load + P + OBE X X X X 12.172 3854 15,677 18,000

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9824 22,313 27,000 X X 7617 3265 4091 3399 X X 2.33 1.20 8517 4267 3794 4369

• 2.00 .87 No anchor movements included in audit results.

No closely spaced modes.

Stress suspary data are maximum values irregardless of location.

Anchor movement loads calculated consist of all constraint displacements supplied. It is believed that S&W Included their anchor displacements in their primary stress calculations. Due to the fact that they used only the end anchor displacements, this contribution was small. If the intermediate constraint displacements are also included, the ODE combined stress for Case I would be about 25% over allowable while the DDE stress would be about 10% ovc . If,the _ _

seismic anchor stresses are included in the secondary stress calculations instead of the primary, as pers .hted in current code criteria, all-stresses would be under allowable.

Hanger H-6 which was not shown on the work sketch was modelen at the location shown on the hanger drawings.

Modes 1 and 2 were the primary contributors to the seismic stress.

The effect of two modal changing as N guencies dynamic support from 3.836 stiffnesses hz and from 6.594 hz maggitudes for 101 in the to 3.807 10165.463 hz and rangehzto for 10610gange shifted the first

. This shifts the secon't ed:: I:.to a r- of the spectrum.

TABLE IV Pressurizer Spray, PUWT: BEAVER VALLEY UNIT NO. 1 LINE: Safety & Relief System '

PROS NO.- 12006 Snubber & Hanger Snubber & Hanger load Case Support Stiffness Support Stiffness Ratio S&W Allowable Estimate Rigid (DQ Qq)

SRSS Q) Algb Sun O SRSS Q) Algb Sun @

Dead Load + P 5846 psi X X 5346 psi 7976 16000 psi X X X

13 084 psi 7422 psi 16873 19200 psi X X 13728 psi 7584 psi 15930 28800 psi I

• OBE 8051 psi 2349 psi 8050 psi 2347 psi X X 3.43 3.43 DBE 8695 psi 2509 psi 8695 psi 2507 psi X X 3.47 3.47 No anchor movements included in tabulated audit results.

ho closely spaced modes.

! tress summary data are maximum values irregardless of location.

Anchor movement loads calculated consist of all constraint displacements supplied using only the direction indicated by S&W. No OBE anchor movements were suppifed. No intermediate earthquake constraint movements were supplied. It appears-that 51W included earthquake anchor movements in their primary stress calculation. Our results indicate as S&W's do that the primary stresses with anchor movements included are within allowable for the D8E.

Modes 1 and 2 are the primary contributors to the selsmic stress.

The effect of changing constraint and dynamic support stiffnesses from 105 to 1016 on the first two modal frequencies are small for this problem. -

No valve eccentricities were given, therefore, eer model does not reflect this effect.

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1 TABLE V W: BEAVER VALLEY UNIT 1 LINE: RHR System PROS NO.* 255A 1 of 2 i

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Load Case Support Stiffness Support Stiffness S&W Allowable Rati*

l Estinate Rigid GO G@

SRSS @ Algb Sun @ SRSS @ Algb Sun @

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Dead Load + P 5885 psi X X

• 5885 psi 6111 16300 psi X X had M + P + M 6444 psi 6378 psi 7719 19560 psi X X

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Dead Load + P + N E 6791 psi X X 6663 psi 8898 29340 psi X X i E 959 psi 1241 psi 956 psi 1241 psi I I

.77 .77 1382 psi X X 1793 psi 1377 psi 1792 psi ,77 ,j7 No anchor movements included in tabulated audit results.

No closely spaced modes. -

Stress summary data are maximum values irregardless of location.

Anchor anvenent calculated loads consist of all constraint displacements supplied. OBE & D8E anchor mer ments and

, intermediate constraint movements were given. -

All modes were at the zero pet sod acceleration level. Three to four modes provided the primary contribution to the

response.

The effect of changing constraint stiffnesses on system frequency was small.

S&W may have included earthquake anchor movements in their primary stress calculation. Since the earthquake anchor movement stresses are low (3000 psi) it doesn't make much differen:e where they are included.

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PLANT: BEAVER VALLEY L M

• RHR System PROB NO.* 255A 2 of 2 Souh6er & Manger Souther & Hanger Lead Case Support Stiffness Sapport Stiffness Ratio S&W A11omable Estimate Rigid GO G@

SRSS @ Algb Sun @ SRSS @ Algb Sun @

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Dead Lead + P 6649 X X 1 6630 ---

16300 X X Dead Load + P + OBE 8257 8125 i ---

19563 Dead Lead + P + ISE I I u 8497 X X

8382 ---

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• 721 745 531 639 I I

.97 .91 DOE 993 1034 862 909 I X .96 .95 he == char movements included la tabulated audit results.

No closely spaced modes.

Stress summary deta' art maximum values irregardless of location. -

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• TABLE VII PLANT: BEAVER VALLEY LINE: S.G. Aux. Feedwater PR08 NO.: 203C Snubber & Henger Snubber & Hanger Load Case Support Stiffness Support Stiffness S&W Allowable Ratt8 Estimate Rigid (LO Q@

SRSS @ Algb Sun CD SRSS @ Algb Sun @

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• ' I I 3110 3106 363*. 15000 Dead Load + P + OBE 47j9 X X X X 4202 5909 18000 Dead Load + P + DBE 5029 I I 5170 6751 27000 X X 08E 3498 3550 3500 3536 X X 1.00 .99 4360 4424 I I 4459 4505 1.00 .99 No anchor movements. .

No closely spaced modes.

Stress sunnary data are maximum values irregardless of location. k The first three modes are:

Best Es'timate Stiffness Rigid Stiffness

1. 12.509 Cps 1. 12.559

2. 13.618 - 2. 13.684

3. 15.039 3. 15.053 Three =c es . o *d , r -trary contribution to the response. -

. t' TABLE VIII PLANT: BEAVER VALLEY LINE: S.G. Aux. Feedwater (Discharge) PROS NO.: 204 Snubber & Hanger Snubber & Hanger Load Ease Support Stiffness Support Stiffness Ratio S&W Allowable Estimate Rigid GO G@

SRSS @ Algb Sun @ SRSS @ A19b Sun @

Dead Load + P 5371 5440 5138 15000 X X had Load + P + ME 8225 8380 7323 18000 Dead Load + P + DBE 8860 X X 8931 10842 27000 X X ME 4098 4762 3697 4263 .86

,. .87 08E 4734 5347 4318 4814 X X .89 .90 No anchor movements included in tabulated audit results.

No closely spaced modes.

Stress suonary data are ma,ximum values irregardless of lecation.

No seismic anchor movements specified.

For them the inEGAG Idaho analysis, their seisulc analysis.rod hangers were input in dynamic support configuration. S&W apparently did not include

The first two modes provided the primary contributton to the response.

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