Text

Rept on Reanalysis of Safety-Related Sys,Surry Power Station,Unit 2, Revision 1,resubmitted Due to Errors in Original Submittal
	ML18139A328
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Site:	Surry
Issue date:	04/11/1980
From:	; EBASCO SERVICES, INC.
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REPORT ON THE I I I

I

REANALYSIS OF I

I SAFETY-RELATED 1

PIPING SYSTEMS I

I SURRY POWER STATION-UNIT 2 I

I VIRGINIA ELECTRIC I

AND POV\\IER COMPANY I

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REVISION 1 EBASCO SERVICES INCORPORATED----

JERICHo, NEW YORK I

so o a 1 s,0;2~J

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EBASCO SERVICES INCORPORATED ENGINEERS -

CONSTRUCTORS CONSULTANTS ONE JERICHO PLAZA JERICHO, N. Y. 11753 CABLE AOORESS -eBAScoe-Mr Harold R Denton, Director Office of Nuclear Reactor Regulation.

US Nuclear Regulatory Corrmission Washington, OC 20555

Dear Mr Denton:

Subject:

References:

SHOW CAUSE ORDER REANALYSIS REPORI' REVISION NO. 1 SURRY Pav.ER STATION UNIT *2 Serial No.:

Docket No.:

License No. :

138B 50-281 DPR-37 June 12, 1980 NRC-l Ebasco Sel'.Vices, Inc is*resubrnitting forty-five (45) copies of Revision 1 of the "Report on the Reanalysis of Safety-Related Piping Systems, Surry Power Station Unit 2, Virginia Electric and Pc:mer Company," originally hand delivered April 11, 1980.

Please replace the forty-five (45) copies sul:rnitted on April 11, 1980 (white cover) with those enclosed (beige cover).

The resutrnittal is for two (2)* reasons:

1)

General illegibility of Appendix B flow diagrams

2)

Errata on Appendix B, Page B-2,

\\

Main Steam Drawing No. 11548-FM-14.A.

Feedwater Drawing No. 11548-FM-18A *-.

are Revision 1 but were not indicated as such.-*

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I EBASCO SERVICES INCORPORATED Mr Harold R Denton June 12, I980 We apologize for any inconvenience this may have caused.

HWN/JNR/bfl Enclosures cc:

Mr Victor Stello, Director Very truly yours, 91~~

H W Nelson Project M3nager Office of rns:i:ection and Enforcerrent Mr James P O'Reilly, Director Office of Ins:i:ection and Enforcerrent, Region II

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I VIRGINIA ELEC:rRIC AND POWER COMPANY RIC:HHOND,VUiiG:INIA 23261 April 11, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission

Washington, DC 20555

Dear Mr. Denton:

SHOW CAUSE ORDER REANALYSIS REPORT REVISION NO. 1 SURRY POWER STATION UNIT 2 Serial No. 138B PSE&C/GLS:mac:wang Docket No. 50-281 License No. DPR-37 In the letter of February 22, 1980 (Serial No. 138), Vepco requested start-up of Surry Power Station Unit 2 based on the "Report on the Reanalysis of Safety Related Piping Systems, Surry Power Station, Unit 2" of the same date.

The report reflected the results of the pipe stress and pipe support analyses subject to final verification and modification installation.

The purpose of this submittal is to update the original -report to reflect revisions necessitated during modification installation and to make minor typographical corrections to the text to enhance clarity and consistency.

The subject changes, 1,Jhich are noted in the margins, in no way alter our original conclusion that the analytical work completed and the modifications inst~lled at the time of start-up provides a high degree of confidence that the integrity of safety systems for Unit 2 can be assured during the DBE or OBE events.

This revised report also incorporates our recent corrmitment to the NRC Staff (Vepco letter of March 21, 1980, Serial No. 138A) to complete the installation of all modifications associated with the Order to Show Cause prior to start-up of the unit.

The additional commitm~nts to complete certain portions of the work associated with I.E.Bulletins 79-02 and 79-14 are firm as*

outlined in the February 22 letter.

If you have any questions with regard to this submittal, please contact us.

Very truly jours,

/,r~-~**

~

/_,/

~...-,r.A.

, t'/t,;./

{'.. i ~/ \\,;,*(

W. C.

pencer Vice President - Power Station Engineering and Construction Services Attachment cc: Mr. Victor Stello, Director Office of Inspection & Enforcement Mr. James P. O'Reilly, Director Office of Inspection & Enforcement, Region II

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I REPORT ON THE REANALYSIS OF.

SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION-UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY REVISION 1 EBASCO SERVICES INCORPORATED---

JER1cHo, NEW YORK

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SURRY POWER STATION -

UNIT 2 I

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I REPORT ON THE I

REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION -

UNIT 2 VIRGINIA ELECTRIC AND POWER COHPANY I

FEBRUARY 22, 1980 I

REVISION 1 APRIL 11, 1980 I

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EBASCO SERVICES INCORPORATED JERICHO, NEW YORK I

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I Section I*

1.0 2.0 3.0 4.0 I

5.0 6.0 7.0 I

7.1 7.2 7.3 1,

7.4 8.0 I

9.0 10.0 I

10.1 10.2 10.3 I

10.4 I

Appendix A

B I

C D

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I TABLE OF CONTENTS Title

SUMMARY

AND CONCLUSIONS.........................

SCOPE OF REANALYSIS.............................

PIPE STRESS RESULTS.............................

PIPE SUPPORT RESULTS............................

SCHEDULE FOR COMPLETION........*................

HIGH ENERGY LINE BREAKS.........................

CONSERVATISMS...................................

Field Verification of As-Built Conditions.....

Quality Assurance and Engineering Assurance...

Use of Amplified Response Spectra.............

Conservatisms Applied to Inertial Stress......

SYSTEM OPERABILITY EVALUATION...................

BRANCH LINE

SUMMARY

RESPONSE TO THE NUCLEAR REGULATORY COMMISSION'S CONCERNS.........*.................

Support Stiffness...... ****~************......

NUPIPE Computer Code..........................

Problem 2538 - Support H-15...................

Benchmark Problems.....*.**...................

Systems Affected *****, *.*, **.*,.*.***..**.....**.

Flow Diagrams - Identification of Problems Analyzed Response to IE Bulletin 79-04......... ~.........

Correspondence with the NRC *.********..**.*.***.

1-i 1-1 2-1 3-1 4-1 5-1 6-1 7-1 7-1 7-1 7-1 7-2 8-1 9-1 10-1 10-1 10-1 10-1 10-2 A-1 B-1 C-1

D-1

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I Table 3-1 3-2 3-3 3-4 4-1 4-2 5-1 LIST OF TABLES Title PIPE STRESS RE-EVALUATION

SUMMARY

NOZZLE AND PENETRATION

SUMMARY

PIPE STRESS HARDWARE MODIFICATION

SUMMARY

HARDWARE MODIFICATION

SUMMARY

DUE TO NOZZLE AND PENETRATION OVERLOADING PIPE SUPPORT ANALYSIS

SUMMARY

PIPE *SUPPORT HARDWARE MODIFICATION

SUMMARY

SCHEDULE FOR COMPLETION 1-ii

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I SURRY POWER STATION -

UNIT 2 SECTION 1

SUMMARY

AND CONCLUSIONS In response to the Nuclear Regulatory Commission's Order to Show Cause, dated March 13, 1979, a reanalysis was conducted of safety related piping systems for Surry Power Station Unit 2 which were originally dynamically analyzed using the SHOCK 2 computer program.

The SHOCK 2 program, which used an earlier load combination methodology, is no longer considered ac-ceptable by the NRC.

This report discusses the details of the analysis work and results of the pipe and support analyses within the scope of the reanalysis for Surry Unit 2.

Further, this reanalysis is consistent with the methods used on Surry Power Station Unit 1, which were discussed in earlier reports submitted on June 5, 1979 (Vepco Serial No. 453) and on August 1, 1979 (Vepco Serial No. 453A) and on January 15, 1980 (Vepco Serial No, 048).

This report summarizes the total reanalysis effort £or all aspects of the March 13, 1979 Order to Show Cause for Surry Power Station Unit 2.

All piping systems affected by the Order to Show Cause, both inside and outside the containment, have been reanalyzed using the NUPIPE

program, which is acceptable to the NRC.

Table 3-3 (Pipe Stress Hardware Modi-fication Summary) and Table 3-4 (Hardware Modification Summary Due to Nozzle and Penetration Overloading) identifies all modifications to the piping systems which have resulted from this reanalysis.

While some of these modifications are attributable to the seismic analysis

method, the majority of modifications result from differences in the as-built conditions and other miscellaneous reasons.

All of these modifications have been or will be made prior to startup of the unit following. the steam generator replacement outage.

With the installation of these modifications, all Surry Unit 2 piping within the scope of this report will meet the Final Safety Analysis Report (FSAR) allowables for both the Operating Basis Earthquake and Design Basis Earthquakes (OBE and DBE) conditions.

All pipe supports both inside and outside containment affected by the Order r 1 to Show Cause have been evaluated for the revised support loads from the pipe stress reanalysis.

All of these hardware modifications have been or will be installed prior to start up of the Unit following the steam gen-erator replacement outage.

Table 4-2 (Pipe Support Hardware Modification Summary) reports all the modifications resulting from the support re-analysis.

As was the case in the piping system reanalysis, most of the modifications are the result of differences between the original design conditions and the actual field as-built condition.

During the reanalysis of Surry Unit 2, 79 stress modifications (22 due to pipe stress, 57 due to nozzle overload) and 258 pipe support modifications were identified; while 63 pipe stress and 66 pipe support modifications 1

were identified on Surry Unit 1.

The differences in the number of modif-ications is not considered significant, due to the conduct of the re-analysis applied to Surry Unit 2.

Modifications on Surry Unit 1 were 1-1 1

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I SURRY POWER STATION -

UNIT 2 identified after many analytical iterations;. whereas, on Surry Unit 2

,nodifications were designed based on fewer iterations.

The conduct of the reanalysis in this manner served to identify modifications faster so that systems could be upgraded more quickly in order not to substantially inter-fere with the completion of the Steam Generator Replacement Project.

In addition, the Surry Unit 2 reanalysis included the OBE condition.

Further, all pipe supports were as-built in the field and QC verified prior to re-analysis.

Lastly, to limit the interface problems between the Show Cause scope and other piping, supports were added to facilitate the NUPIPE re-analysis.

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SURRY POWER STATION -

UNIT 2 SECTION 2 SCOPE OF REANALYSIS As described in systems in the with a SHOCK 2

the NRG.

the NRG Order to Show Cause, March 13, 1979, som*e p1.p1.ng Surry Power Station, Unit 2

were dynamically analyzed computer program that 1.s not currently acceptable to

Al 1 systems or portions of systems that were analyzed by the SHOCK 2

computer program have been identified in. Appendix A.

These systems were reanalyzed by Stone

& Webster Engineering Corporation (Stone Webster) and Ebasco Services Incorporated (EBASCO) using a

NUPIPE com-puter code.

Responsibility for the reanalysis 1.s also identified 1.n Appendix A by system and problem number.

The results of the reanalysis are compared with code allowable

stresses, allowable loads for nozzles and penetrations, and are in the evaluation of pipe supports.

2-1 pipe used

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SECTION 3 PIPE STRESS RESULTS A total of 62 pipe stress problems were originally analyzed by the PSTRESS/

SHOCK 2 computer, program that used algebraic summation and are therefore specifically addressed by the Show Cause Order.

These stress problems are being analyzed by two groups:

Stone

& Webster Engineering Corporation (Stone & Webster) in Boston, Massachusetts, and Ebasco Services Incorpora-tion (EBASCO) in Jericho, New York, as indicated in the following table:

Stone & Webster 13 PIPE STRESS PROBLEMS EBASCO 49 Total 62 Responsibility for the reanalysis 1s identified by system and problem number in Appendix A of this report.

Field-verified piping isometric drawings provide the basis for program in-puts for the pipe stress reanalysis.

The reanalysis is conducted using the NUPIPE computer program.

NUPIPE calculates intra-modal seismic forces using a modified square root of the sum of the squares (SRSS) technique which is always more conservative than the approved SRSS method, and an SRSS technique for inter-modal combination.

Piping is analyzed in most cases utilizing amplified response spectra (ARS) that are developed using soil structure interaction techniques (SSI-ARS).

The resultant stresses and loads are used to evaluate piping, supports, nozzles, and penetrations.

In accordance with the NRC letters of May 25, 1979 and November 15, 1979 to Virginia Electric and Power Company (VEPCO),

the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a factor of 1.5 for the DBE and 1.25 for OBE conditions.

All 62 problems have been reanalyzed.

Table 3-1, Pipe Stress Re-Evaluation Summary, presents the results for these 62 stress problems.

In Table 3-1, the figures for Original Total Stress, at the point of maximum total stress in the pipe, and Original Seismic Stress, at the same point, are extracted from original design stress isometrics (MSK's).

In Table 3-1, the columns for New Total Stress, at the point of maximum total stress in the pipe, and New. Seismic Stress, at the same point, were taken from the NUPIPE computer runs with the seismic inertial stress multi-plied by a factor of 1. 5 and then added to the Seismic Anchor Movement (SAM) Stress for runs using the SSI-ARS.

Even though Table 3-1 reports DBE results, stress analysis is performed for OBE also and modifications de-signed wherever necessary.

The Original Total and Original Seismic Stresses shown in Table 3-1 were computed using the SHOCK 2 programs for the original design conditions.

The New Total and New Seismic stresses were computed by the NUPIPE pro-3-1

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SURRY POWER STATION -

UNIT 2 gram using different mass models and in most cases different ARS's th~n the original calculations.

More importantly, the reanalyses were based on as-built conditions, field verified in 1979, which in some cases differ from the original design conditions.

For these reasons, the new stresses and the original stresses in Table 3-1 are not comparable, as they do not necessarily represent the same physical conditions.

Table 3-2, Nozzle and Penetration Summary, summarizes the nozzles and pene-trations evaluated under the reanalysis program.

For all the problems in which the SSI-ARS are used, the seismic inertial nozzle loads have been increased by a factor of 1. 5 for DBE per the NRC letter.of May 25, 1979, and by a factor of 1. 25 for OBE per the NRC letter of November 15, 1979.

Table 3-3, Pipe Stress Hardware Modification Summary, lists the hardware modifications necessary to bring the pipe stress analysis to within code allowables.

Of the 62 problems reanalyzed, hardware modifications were made to 17 problems due to pipe stress.

These modifications consisted of I 1 22 added modified, or deleted supports.

The modifications include those necessary to the flexibility analysis of the branch lines.

A branch line (Problem No. 2508B) was rerouted as a result of thermal reanalysis, not as I 1 a result of seismic reanalysis.

Table 3-4, Hardware Modification Summary due to Nozzle and Penetra_tion Overloading, lists all modifications to reduce nozzle and penetration loads.

Of the 62 problems reanalyzed, hardware modifications were made to 17 problems due to nozzle overload.

These modifications consisted of 57 added, modified, or deleted.supports.

Those modifications which result from the p1.p1.ng reanalysis are identified in Section 3.

Only the modifications which result from the pipe support reanalysis are reported in Table 4-2, Pipe Support Hardware Modification Summary.

Final verification of piping and support stresses and Engineering Assurance review has yet to be completed for a few problems.

It 1.s ex-pected, however, that the number and type of modifications due to the stress and support analysis are correct and final.

3-2

SURkY POWER STATION -

UNIT 2 TAl!LE 3-1 ShPP.t I of 5

!.!!~TR~~-~E-EVALUATION SUUMARY SystP.m NamP.

LinP. Size Pipe StrP.ss (psi) and RP.analysis MKS NPS Original O~iginal New NP.w

+

Problem Number Responsibility Number (Inches)

Total SP.ismic Total SP.ismic Allowable Low HP.ad Sa fie,ty

.!!!Jection 2555 E

122Dl 10, 12 12043 NA 7974 2449 30690 2709 E

l22LI 12 NA NA 19173 Il427 33750 2537/2540/2540B E

122AI, 4,6, 12350 NA 2739 883 33750 ll 7Bl 10, 12 2539 E

l22JI 6

30368 NA 14 771 7330 32985 l22KI 11 2727 S&W 127Cl 6

21179 NA 24352 17453 33750 127C2 8, 10 2681 E

l27Kl 8

1677 307 1220 185 281185 2682 E

127K2 8

1677 307 1174 164 28485 2695 E

l27Dl 8

21179 NA 2094 1103 28485 2697 E

127D2 6

21179 NA 1981 1022 281185 High Head Safety Injection 2689 E

127Fl 10 246119 NA 11773 9571 33750 2735 E

127GI 3,4,6, NA NA 26660 17772 33750 127G2 8, 10 Containment and Recirculatinn Spray 2521 S&W 123AI 8, 10 14904 NA 7790 4276 33561 2523 S&W I23A2 8, 10 14904 NA 7977 6013 33561 254 7 S&W l23Cl 8, 10 127 lJ NA 23.532 19739 33561 2546

f.

l23DI 8, 10 3.528 1576 8892 7328 28800 251,1 E

123D2 8, IU 3528 15 76 16338 16636 28800 2542 E

123D3 8, IU J521i 1576 1119 31 17252 28800

-.... -'-............. --\\..

SlJHl~Y l'U\\;1,.1{ SlA'l'IUU -

llk 11' 2 TAUI.F. 3-1 Sheet 2 o[ 5 l'll'E STllESS RE-EVALUATlUH

SUMMARY

SyN te111 Name Linl! Size

'~Stress (_p!J.L ____ -----------

anJ lleanalysie MKS NI'S Original Original New New Prob I cm Hunobe r llc~~~ib_!l i!x_

Number

(_Inches)

Total Seismic Total Seismic Allowable Con la i 111,,ent and Ile-circu!_ation S£_!!!1'_ (Cont '<I) 254)

E 12JD4 8, 10 352B 157b 6li98 6988 29970 25bll E

I 2J~;I 111 7JJl1 NA 12775

!0995 29970 25bl E

IZJt:2 Ill 7JJl1 NA 6'.:.87 3576 29970 25114 E

12JG1 10 11605 7",22 2912 IIBI -

211485 2533 E

l23G2 10 11605 7922 5874 31113 211485 25411 E

123111 10 15785 11241 3904 2437 29970 2';45 E

123H2 10 15785 11241 2397 676 29970

2741, E

123JI 8

7966 51111 16143 15107

)5820 2745 E

12JKI 6

24114) 22577 15621 12559 33750 27';3 E

1231,l 12 6136 2818 1344 394 JJ750 2754 E

123HI 12 5649 NA 1999 949 JJ750 2751 K

123NI IO 6010 NA 21234 13842 281,85 2752 E

123112 10 6010 NA 18613 13934 28485 2549 S&W 12JC2 8

11955 10125 10061 8397 33561 2755 E

1231' 1 4,8 10369 6324 14)11 7340 33750 2756 E

l23Ql 10 NA NA 5705 21,66 28485 27'H E

12JQ2 IU 5811)

NA 3t;t, 7 2381 26485 fa in Steam 1577 S&W IOOIJI

)(J l:lll24 NA 1076]

288) 33750

!';BB S&w l01DI JO 11163'.>

NA 125 IJ 3041

))750

!579 S&W I021J2 JO IJOJI NA 11434 4120 33750 D46*

S&W IOJAI 30 19970 NA 12568 2 54 77 337SO lll3A2

SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 3 of 5 PIPE STRESS RE-EVALUATION

SUMMARY

Line Size Pipe Stress ( psi)

System Name Reanalysis MKS NPS Original Original New New and (Inches)

Total Seismic Total Seiemic Allowable Responsibility Number Problem Number Feedwater 14 14499 NA 13681 9360 27000 2569 S&W IOOGI 14 16025 NA 12970 8376 27000 2573 S&W IOIGI 14 17927 NA 14230 8443 27000 2571 S&W I02Gl Auxiliary Feedwater ll8AI 3,6 8568 2407 20963 17736 27000 2473 E

l 18A2 118GI 4,6 21230 NA 12988 9188 27000 2683 E

ll8G2 Pressurizer Spray 4

18560 NA 8013 3088 30690 277i E

125Al Pressurizer Safety and Relief 3,4 9093 NA 8824 542L 30636 2000 E

124Al 124A2 6, 12 Residual Heat Removal ___

2540/2540B E

Listed Under Low Head Safety Injection System l0,12,14 NA NA 13112 8461 24570 2508A/2508B E

II 7Al 6

NA NA 12008 8824 29970 2554 E

ll 7CI Service Water 24 NA NA 6529 6212 21600 2465 E

119AI 24 NA NA

'6561 6214 21600 2467 E

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SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 4 of 5 PIPE STRESS RE-EVALUATION

SUMMARY

System Name Line Size Piee Stress (esi) and Reanalysis MKS NPS Original Original New New Problem Number Reseonsibilitr Number (Inches)

Total Seismic Total Seismic Allowable Service Water (Cont'd) 2469 E

l l 9A3 24 NA NA 14687 13730 21600 2471 E

119A4 24 NA NA 12459 11614 21600 Comeonent Cooling 2601/2603 E

ll2Sl 18 9696 NA 7043 5246 21600 ll2S2 2604/2605 E

112AA1 18 9696 NA 7074 5204 21600 112AB1 Containment Vacuum 2650 S&W I37Al 8

25,750 NA 13659 13037 21600 HP Steam to Auxiliary Feedwater Pume 2869/2862/2864 E

131Al 3,4 22609 NA 24229 19559 27000 l31Bl 131Cl

Legend:

E EllASCO S&W Stone & Webster NA Not Available Allowable Stress = 1. B sh New Total Stress (for SSl/ARS)

New Seismic (for SSI/ARS)

New Total Stress (original ARS)

New Seismic (original ARS)

SURRY POWER STATION -

UNIT 2 TABLE 3-1 PIPE STRESS RE-EVALUATION

SUMMARY

SLP + SDW + l.S SDBEI + SDBEA l.S SDBEI + SDBEA SLP + SDW + SDBEI + SDBEA = Original Total Stress

= SDBEI + SDBEA = Original Seismic Stress Where SLP Longitudinal Pressure Stress Dead Load Stress Seismic Inertial Stress, Design Basis Earthquake SDBEA = Seismic Stress due to Anchor Movements, Design Basis Earthquake Sh

= Allowable stress at maximum (hot) temperature Note:

Sheet 5 of 5 The original total and original seismic stresses shown in Table 3-1 were computed using SHOCK 2 for the original design conditions.

The new total and new seismic stresses were computed by the NUPlPE program using different mass models and, in most cases, different ARS's than the original calculations.

More importantly, the reanalyses were based on field-verified, as-built conditions in 1979, which, in some cases, differ significantly from the original design conditions.

For *this reason, the new stresses and the original stresses in Table 3-1 are not comparable, as they do not necessarily represent the same physical conditions

Soil Structure Interaction (SSI) Amplified Response Spectras (ARS) were used in the new analysis for ell problems except Problem 23.46 which utilizes a combination of SSI-ARS and the original ARS.

+ Problems having / are counted as separate problems example:

2869/2862/2864 are counted as three problems.

Using this method of counting the total number of pipe stress problems equal 62.

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TABLE 3-2 Sheet 1 of 4 NOZZLE AND PENETRATION

SUMMARY

I SHOCK 2 Problems Responsi-Vendor I

System bility Total No.

No. Acceptable Nozzle Confirmation and For of Nozzles/

(Evaluation Modification Being Problem No.

Analisis Penetrations Com:elete)

Re9uired Obtained I

Low Head Safety Injection I

2555 E

1/0 1

0 1

I 2709 E

1/0 1

0 0

2537/2540 E

1/0 1

0 1

I 2539 E

0/0 NA NA NA 2727 S&W 2/0 2

0 2

2681 E

0/0 NA NA NA I

2682 E

0/0 NA NA NA 2695 E

0/0 NA NA NA I

2697 E

0/0 NA NA NA High Head Safety Injection 2689 E

0/0 NA NA NA I

2735 E

3/0 3

0 3

I Containment and Recirculation Sprar I

2521 S&W 0/0 NA NA NA 2523 S&W 0/0 NA NA NA I

2547 S&W 0/0 NA NA NA I

2546 E

0/0 NA NA NA I

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I TABLE 3-2 Sheet 2 of 4 I

NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems I

Responsi-Vendor System bility Total No.

No. Acceptable Nozzle Confirmation and For of Nozzles/

(Evaluation Modification Being Problem No.

Analx:sis Penetrations Complete)

Required Obtained I

Containment and Recirculation I

Spray (Cont'd) 2541 E

0/0 NA NA NA I

2542 E

0/0 NA NA NA 2543 E

0/0 NA NA NA I

2560 E

1/0 1

0 0

I 2561 E

1/0 1

0 0

2544 E

1/0 1

0 0

I 2533 E

1/0 1

0 0

2548 E

1/0 1

0 0

I 2545 E

1/0 1

0 0

2744 E

0/0 NA NA NA 2745 E

0/0 NA NA NA I

2753 E

1/0 1

0 0

2754 E

1/0 1

0 1

I 2751 E

2/0 2

0 0

2752 E

2/0 2

0 0

I 2549 S&W 0/0 NA NA NA I

2755 E

2/0 2

0 2

2756 E

2/0 2

0 0

I 2757 E

2/0 2

0 0

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SURRY POWER STATION -

UNIT 2 I

TABLE 3-2 Sheet 3 of 4 NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems I

Responsi-.

Vendor System bility Total No.

No. Acceptable Nozzle Confirmation I

and For of Nozzles/

(Evaluation Modification Being Problem No.

Analysis Penetrations Complete)

Required Obtained Main Steam I

2577 S&W 1/1 1/1 0/0 0/0 I

2588 S&W 1/1 1/1 0/0 0/0 2579 S&W 1/1 1/ l 0/0 0/0 I

2346 S&W 0/0 NA NA NA Feed water I

2569 S&W 1/1 1/1 0/0 0/0 I

2573 S&W 1/1 1/1 0/0 0/0 2571 S&W 1/1 1/1 0/0 0/0 I

Auxiliary Feedwater 2473 E

0/0 NA NA NA I

2683 E

3/0 3

0 3

I Pressurizer Spray 2771 E

1/0 l

0 l

I Pressurizer Safety and Relief I

2000 E

5/0 5

0 l

/1 Residual Heat I

Removal 2540 E

(Listed under Low Head Safety Injection System)

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I System and Problem No.

Responsi-bility For Analysis Residual Heat Removal (Cont'd) 2540B E

2508A/2508B E

2554 E

Service Water 2465 2467 2469 2471

omponent Cooling 2601/2603

~604/2605

ontairunent 7acuum

'.650 E

E E

E S&W IP Steam to Auxiliary ieedwater Pump

'.862/2864/2869 E

SURRY POWER STATION -

UNIT 2 TABLE 3-2 NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems Total No.

of Nozzles/

Penetrations 0/0 8/0 0/0 1/0 1/0 1/0 1/0 2/0 2/0 0

1/0 No. Acceptable (Evaluation Complete)*

NA 8

NA 1

1 1

1 2

2 NA 1

'OTES:

NA = Not Applicable E = EBASCO S&W = Stone & Webster Nozzle Modification Required NA 0

NA 0

0 0

0 NA 0

.Sheet 4 of 4 Vendor Confirmation Being Obtained NA 8

NA 0

0 0

0 1

1 NA 0

System Name and Problem No.

Low Head Safety Inj1:_1:_t i~~

2709 25)9 Containment and Rec ircu lat ion Spr~------

2549 2544 2745 2752 Pressurizer Sp~---

2771 Residual Heat Removal 2508B Reanalysis Resp~nsibi!_ity E

E E

S&W E

E E

MKS No.

l22Ll 122Al 117Bl l22J l l22Kl l2JC2 12JGI 12JK1 123N2 125Al ll 7Al SURRY POWJ,:R STATION -

UNIT 2 TABLE 3-3 P ll'li STRl,SS 111\\Rmll\\R:l MOD IF [CAT [ON SUr*IHARY Over st t*essed Condit ion Seismic over:;tr~s9 Thermal overstress Thermal overstress (Branch Line)

Pipe contacts crane wall during seismic condition.

Thermal overstress Seismic anchor move-ment overstress TI1ermal overstress

( Branch Line)

Seismic overstress Thermal overstress Branch line Attributed To:

Seismic Reanalysis As-built As-built Seismic Reanalysis As-built As-built As-built As-built As-built 2540 (Listed under Low Head Safety Injection System)

Resolution Spring hanger rep laced by rigid restraint.

Removed a restraint, anchor rep laced by restraints and a snubber.

Removed a restraint Lateral support added.

Removed a restraint.

Vertical restraints replaced by spring hangers at two loca-tions Removed a restraint.

Two restraints added.

Rerouting of 1-1/2 in.

pipe Sheet I of 2 No. of Modifications 2

2 l

2

System Name and Problem No.

Residual Heat Removal (Cont'd) 2540B Component Cooling 2604/

2605 HP Steam to Auxiliary Feedwater Pump 2862/

2864/

2869 Feedwater 2569 Notes:

E EBASCO Reanalysis Responsibility E

E E

S&W S&W Stone & Webster MKS No.

Il 7B Il2AA1 112AB1 I31Al 131Bl 13IC1 lOOGl SURRY POWER STATION -

UNIT 2 TABLE)-)

PIPE STRESS HARDWARE MODIFICATION

SUMMARY

Overstressed Condition Thermal and seismic overstress Seismic overstress Thermal and seismic overstress Insufficient branch line flexibility Attributed To:

As-built Seismic Reanalysis As-built/

Seismic Reanalysis As-built Resolution Anchor replaced by vertical restraint, horizontal restraint removed.

Two restraints added.

Two anchors replaced by restraints, two snubbers and a spring added.

Remove existing U-bolt on 3/4 in. line Sheet 2 of 2 No. of Modifications 2

2 5

System Name and Problem No.

Low Head Safety Injection 2555 High Head Safety Injection 2735 Containment and Recirculation

~

2544 2533 2753 2754 2751 2752 2755 2756 Reanalysis Responsibility E

E E

E SURRY POWER STATION -

UNIT 2 TABLE 3-4 HARDWARE MODIFICATION SUM~IARY DUE TO NOZZLE AND PENETRATION OVERLOADING Sheet l of 2 Equipment No.

Attributed To:

Resolution No. of Nodifications 2:._Sl-TK-lB 2-CII-P-IA 2-CH-P-18 2-CH-P-lC 2-RS-E-lD 2-RS-E-IC 2-CS-P-18 2-CS-P-lA 2-11S-P-2A 2-RS-P-2ll 2-CS-P-18 2-CS-P-lA 2-RS-E-lA 2-RS-P-lA As-built/

Seismic Reanalysis Seismic Reanalysis As-built As-built As-built/

Seismic Reanalysis As-built/

Seismic Reanalysis Seismic Reanalysis Seismic RPanalysis As-built/

Seismic Reanalysis As-built One restraint added, One spring hanger replaced by a two direction restraint Ten restraints added, one vertical restraint replaced by spring hanger, two anchors added One restraint removed Two restraints added One vertical restraint replaced by spring hanger, two restraints added One vertical restraint replaced by spring one horizontal restraint added.

One restraint added One restraint added One restraint added two vertical restraints removed.

One anchor and one restraint removed 2

13 2

3 2

1

System Name and Problem No.

Containment and Recirculation Spray (Cont'd) 2757 Auxiliary Feedwater 2683 Pressurizer Safety & Relief 2000 Residual lleat*

Removal 25081i Service Water 2471 Component Cooling 260 I /2603 EBASCO Reanalysi11 Responsibility E

E E

E SURRY POWER STATION -

UNIT 2 TABLE 3-4 HARDWARE MOIHFICATlON

SUMMARY

DUE TO NOZZLE AND PENETRATION OVERLOADING Equipment No.

2-RS-E-IB 2-IIS-P-IB 2-rn-r-2 2-FW-P-3B 2-rn-P-3A 2-RC-TK-2 2-RII-P-IA 2-RII-P-IB 2-RS-E-ID 2-RII-E-18 Attributed To As-bu i 1t Seismic Reanaly11is As-built Seismic Reanalysis As-bui It As-built Sei11mic Reanalysis Sheet 2 of 2 Resolution No, of Modifications One anchor removed TI1ree horizontal and two vertical restraints added, A spring replaced by a rigid hanger and a horizontal snubber, two restraints replaced by snubhers, lateral restraint deleted, Six snubbers added, two springs replaced by restraintY, one spring hanger added, one vertiral restraint added.

One* restraint removed Four vertical restraints replaced by springs and 1mubbers, two restraints added, 5

4 JO 6

I I

I SURRY POWER STATION -

UNIT 2 SECTION 4 PIPE SUPPORT RESULTS Table 4-1, Pipe Support Analysis

Summary, summarizes the pipe support reanalysis program.

Six hundred ninety four (694) supports (467 inside the containment, 218 outside the containment) on lines originally analyzed using Shock 2, were reanalyzed as part of this Show Cause effort.

Two hundred fifty eight (258) hardware modifications (17 5 inside the containment, 83 outside the containment) have been identified.

The modifications identified due to. the pipe support reanalysis are listed in Table 4-2, Pipe Support Hardware Modification Summary.

Those modifications which result from the piping reanalysis are identified' in Section 3.

Only the modifications which result from the pipe support reanalysis are reported in Table 4-2.

Of the modifications identified, only 109 were the result of seismic reanalysis of the piping systems iden-tified in the Show Cause Order, while 149 were the result of differences identified between the as-built conditions and the original design.

These conditions are identified in the table for each problem.

For all the problems in which the SSI-ARS are used, the seismic inertial loads have been increased by a factor of 1.5 for DBE per the NRG letter of May 25, 1979, and by a factor of 1. 25 for OBE per the NRG letter of November 15, 1979.

4-1 I

I i1 I

SURRY POWER STATION -

LilHT 2 TAbLE 4-1 Sheet 1 of 4 PIPE SUPPORT ANALYSIS SUNNARY System Name Total Hodifications and Analysis Number of Evaluation or Additions Problem Number

!_lesponsibilit_l'_

Location

_Supports Complete Required Low llead Safety ln jection System 2537/2540 E

IC

)3 33 17 1

2.'>55 E

IC 17 17 9

2~39 E

IC 7

7 5

2681 E

oc 4

4 3

2682 E

O(;

4 4

]

1 2695 E

QC 10 Hi 4

2697 E

oc 9

9 6

au9 E

IC 9

9 3

2727 S&W oc 17 17 8

I 1 lligh Head Safety Injection System 2689 E

QC 5

5 4

2735 E

oc 52 52 23 I 1 Containment and Recirculation Spray 2521 S&W IC 15 15 4

252]

S&W IC 16 16 5

2':J47 S&IJ IC 13 13 6

2549 S&IJ IC 4

4 1

254b E

IC 12 12 6

I 1

SURRY POI/ER STATION -

UNIT 2 TAIILE 4-1 Sheet 2 of 4 PIPE SUPPORT ANALYSIS SU!IHARY System UaiJe Total Modifications and Analysis Number of Evaluation or Additions Problem Number Responsibility Location Suppor~

_Comple~

Required Containment and Recirculati~

~~

(Cont'd)

Z541 E

IC 11 11 6

Z'.>4Z E

IC 11 11 4

Z'.>43 E

IC 12 12 6

11 E

I(;

4 4

1 Z'.>60 Z561 E

IC 5

5 3

2544 E

IC 6

6 2

I 1

2533 E

IC 5

5 1

Z'.>48 E

IC 15 15 6

11 2545 E

IC 17 17 9

2744 E

oc 4

4 3

2745 E

oc 4

4 2

I 1 2753 E

oc 4

4 l

E oc 3

3 l

Z754 2751 E

oc 5

5 1

2752 E

oc 4

4 0

E oc 8

8 4

1 27'.>5 2756 E

IC 7

7 2

2757 E

IC 9

9 3

SURRY POWER STATION -

UNIT 2 TAllLE 4-1 Sheet 3 of 4 PIPE SUPPORT ANALYSIS

SUMMARY

System Name Total Modifications and Analysis Number of Evaluation or Additions Problem Number Responsibility Location Supports Complete

~~~

llain Steam 2577 S&W IC 9

9 2

2588 S&W IC.:

2 2

0 2579 S&W IC 5

5 3

2346 S&W oc 41 41 6

Feedwater 2%9 S&W IC.:

9 9

4 2573 S&W IC 3

3 0

2571 S&W IC 6

6 l

Auxiliary Feedwater 2473 E

IC 34 34 12 1

2683 E

oc 22 22 8

Pressurizer Spray and Relief 2771 E

IC 32 32 7

20UU E

lC 29 29 15 j 1 Residual Heat Removal

25081\\/B E

IC 45 45 5

5 4

1 254Ull E

IC 2554 E

oc l

l l

SURRY PO\\IER STAT IOU -

UNIT 2 TAIJLE 4-1 PIPE SUPPORT ANALYSIS SUHHARY

---~~~~~--~--~~~

System tfame and Problem Number Analysis Responsibility Location Total Number of

?upports Service Water

1465

1467 2469 2471 Component Cooling 26Ul 26U3 2604 2605 Containment Vacuum 2650 High Pressure Steam to Aux Feedwater Pum~

2862 2864 2869 Hotes:

E EBASCO S&W Stone~ Webster IC Inside Containment OC Outside Containment E

E E

E S&\\I E

E E

IC IC IC

IC IC IC IC IC oc oc oc oc 2

2

,16 13 17 17 3

8 3

7 Evaluation Complete 2

2 16 13 17 17 3

8 3

7 Sheet 4 of 4 Modifications or Additions Required 1

0 5

5 7

4 3

1 u

1 1

1

SY STEI 1 NAI-U, AIIIJ PRIJllLEN llUHBER Low Head Safety Injection Systmn ZS37/2S4U ZS'.:17 ZS4U 2S5S

'l.539 ANALYSIS RESPONSIBILITY E

E E

SURRY POI/ER STATION -

utUT Z TABLE 4-2 PIPE SUPPORT llt\\RDWARE MODIFICATION SUNHARY LOCATION IC I(;

IC IC HKS SUPPORT HUMBER 2,4 6

9 s

7,8,9,lU,13,26 12,28 17,19 1

llt\\

8 3

4,S 9,12 12,.

13, 14 2

REASON FOR tlODlFICATION Local pipe wall stress over allowable Support member over allowable ATTRillUTAilLE TO Seismic Reanalysis l\\s-built Support does not allow lateral As-built movement Insufficient clearance Support member over allowable Pipe clamp over allowable U-bolt over allowable Weld over allowable Upward vertical restraint required Support member over allowable Weld over allowable Local pipe wall stress over allowable Support member over allowable Upward vertical restraint required Seismic Reanalysis As-built As-built As-built As-built Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Insufficient lateral clearance l\\s-built Upward vertical restraint required Upward vurtical restraint requtred Seismic Reanalysis Seismic j{!Janalysis Sheet L ot Lt.

RESOLUTION Hodify SUplJOrt Hodify support Hodi fy support llodify support Modify support Modify support Hodi fy support 1,dd weld Modify restraint for uplift load

~1odi (y support Add weld 1-lodi fy support Hodi fy support NoJHy restraint for uplift load Noctify support Hodify restraint for uplift loa<.j llodity restra!11t

!,l)f upl Ht load 11

SYSTEM NAME AND PROBLEM NUMBER Low Head Safety inTection ~x_st~ (Cont'd) 2539 (Cont'd) 2709 2727 2695 ANALYSIS RESPONSIBILITY E

E S&W E

SURRY POWER STATION -

UIIIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC QC oc MKS SUPPORT NUMBER 11-1**

3,4 7

2,3 8

13 11 14 15 16 18 19 11-50 A-16 C-53 A-17 REASON FOR MODIFICATION Support member over allowable Support member over allowable Support member over allowable Local pipe wall stresses over al lowab_le Support over allowable Support and weld over allowable Loads out of range of spring Loads out of spring range Loads out of spring range Loads out of spring range Loads out of spring range Supports restraint lateral movem<;int Support member over allowable Support member over allowable Support member over allowable Support member over allowable ATTRIBUTABLE TO As-built As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built As-built Sheet 2 of 12 RESOLUTION Modify support Modify support Modify support Modify support Modify Support Hod i. fy Support Replace spring Replace spring Rep lace spri ni,;

Ile place spring lleplace spring Modify Support Modify support Modify support Mnd ify support Modify support I i

SURRY POWER STATION -

UNIT 2 TABLE 4-2 Sheet 3 of 12 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBlLITY LOCATION NUMBER MODIFICATION TO RESOLUTION

_!.~w llead Safety Injection System (Cont'd) 2697 E

QC 11-49 Support member over allowable As-built Modify support C-7 Support not acting As-built Removed support C-17,C-18 Support member over allowable As-built Modify support A-14 Support member over allowable As-built Modify support A-15 Support member over allowable As-built Modify support 2681 E

oc 2,4 Support member over allowable As-built Modify support 5

U-bolt over allowable As-built Modify support 2682 E

oc

1.

Support member over allowable As-built Modify support 2

Support member over allowable As-built Modify support 5

U-bo lt over allowable As-built Modify support ligh Head Safety

nject ion System 2689 E

QC C-38,C-39 Support member over allowable As-built Modify support C-40,C-41.

Support member over allowable As-built Modify support 2735 E

QC 19,34,42 Support member over allowable As-built Modify support I 1 22,37,45,28 36,21,23,44 1,3,16,31,39,24 Weld over allowable As-built Add weld 4,6,18,33,41 U-bolt over allowable Seismic Modify support I 1 reanalysis 26 Local pipe will stress Seis,uic Modify support over allowable reanalysis I 1 ontainment and ecirculation Spray 2521 S&W IC 2

U-bolt capacity for side-Seismic Add members to load in insu f fie ient Reanalysis r~sist side lo!Jd

SYSTEM NAIIE MU PRUliLEM NUMllL:R Containment and Recirculation Spray {Cont'd) 2521 (Cont'd) 2523 2547 AN,\\LYS1S RESPONSIBILITY S&W S&I~

S&W SURRY POWER ST/,TIUN -

UNIT 2 TABLE 4-2 PIPE SUPPORT IIARUWARE MOUIFICATION SUHMARY LOCATION IC IC IC HKS SUPPORT NUMllER 3

4 5

1 2

3 4

5 2

4 6

I!

REASON FOR MOUIFICATION Frame overstressed with new loads out of springs range U-strap has insufficient capacity Local stress U-bolt capacity for side-load is insufficient Frame overstressed Capacity of springs in-sufficient U-strap has insufficient capacity Local stress Lateral load fails U-bolt Rod hanger cannot resist upward load Insufficient clearance for thermal movement Insufficient lateral clearance for thermal mo 11ement ATTRIIIUTAIJLE TO Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis As-buiJ t As-built Sheet 4 of 12 RESOLUTlUN Hodify existing frame and replace springs Replace existing stra!J.with new framing lsliminale anchor and add. vert/lat restraint Add members to resist sideload Replace existing frame Replace springs Replace existing strap with new framing Eliminate anchor and add vert/ lat restraint Add lateral restraint lleplace with sway st rut Remove lateral stop Remove lateral stops.(angle)

Ii

SYSTl::N Nl\\llli

.Jill PROIILEM NUHIJER Containment and Recirculation~ (Cont'd) 2547 (Cont'd) l54'J 254b 2541 2542 11NALYSIS RESPONSIBILITY S&I~

S&W E

E E

SURRY POWER STATION -

UNIT 2 TABLE 4-2 l'lPE SUPPORT HARDWARE H0DlFICA1'l0N SUHHARY LOCATION IC IC IC IC IC tlKS SUPPORT NUHIJER 10 11 2

6 9

7,8 2

22A 6,21 7,9 2,8 15,23 13, 17 REASON FOR HODIHCATION Insufficient lateral clearance for thermal movement Local stress and support frame overstressed U-bolt failure Support member over allowable Insufficient lateral &

vertical clearance lnsu f ficient lateral clearance U-bolt restricts lateral movement Anchor stress over allowable Support member over allowable lnsu f ficient lateral clearance U-bolt restricts lateral movemE!nts Support member over allowable Insufficient lateral clearance ATTRIIJUTABLE TO 11s-built As-built As-built As'--built Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis ns-built Seismic Reanalysis As-built As-built Seismic Reanalysis bheet 5 of ll RESOLUTION Remove lateral I

stops (angle) llodify structure Add lateral restraint Modify support Modify restraint Modify restraint I 1 Hodify support Relocate adjacent restraint Modify sup1,ort Modify support liodify support 1

Nodify support l

Modify support

SURRY POWER STATION -

lJNIT 2 TAllLE 4-2 Sheet 6 of 12 PIPE SUPPORT HARDWARE MOD1HCATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTAllU:

PR06LEN NU~1llER RESPONSIUILITY LOCATION NUNllER MODIF'ICATION TO RESOLUTION Containment and Recirculation Spray 2543 E

IC 15,23 Support member As-bu ii t Nod i fy Rupport over allowable 12,13,14 lnRufficient lateral SeiRmic Modify reRtraint clearance ReanalyAis 24A Support member SeiAmic Relocate adja-over allowable ReanalyRis cent restraint 11 2560 E

IC 11-50 U-bolt over SeiRmic Nodi fy Aupport allowable ReanalyAis 2561 E

IC 11-91 Upward vertical SeiRmic Modify for

~eAtraint required ReanalyAis up lift load 11-50 U-bol t over Seismic Modify Aupport 11 allowable Reanalysis ll-98A Support member AA-built Remove vertical over allowable reAtraint 2544 E

IC 11-67 Upward vertical Seismic Modify Aupport reRtraint required

. ReanalyRiA for uplift load 11-68 Support member AR-bu ii t Nodify Rupport over allowable 2533 E

IC 11-3 Support member As-built

~1odify support over allowable 2548 E

IC 7

Support member As-built Modify Rupport over allowable 10 Support member As-built Modify Rupport over allowable 12 U-bolt over As-built Modify Aupport allowable 9A Upward ve rt ica 1 SeiRmic

~1od i fy support restraint required ReanalysiA for uplift load 11 Support member As-built Modify support over all9\\./able

SY STEN llnl-lE 1,l(l) l'llUULEli 1-lUMllER AN11LYSIS REbl'ONSllllLlTY l.outainment and Recirculation~ (Cont'd)

Z~4tl tCont'd)

E

'l.744 E

'L74'.J E

E E

SURllY l'OI/ER STATIOll -

UNlT 2 T!.llLC 4-2 Pll'C SUPPORT llARDUARC NOIJ!FlCATlON SUHMARY MKS SUPPORT LOCATIOll NUIWER IC IC oc Ol.

oc or:

tl 3

14 4

7 lU,15 6,8 2,3 1

4 3

2 REASOll FOR llO!HFICATlOll Support member over allowable Support member o vcr allowable Upward vertical restraint required U-bolt over allowable Support member over allowable Support member over allowable 1/eld over allowable Support member o\\/er allowable Loads out of spring range Local pipe wall stress over allow-able Upward vertical restraint required Support does not al-low lateral movi,ment Local pipe wall stress O\\'er allowable Support member over allowable ATTRI llUTAllL G TO l\\s-built As-built Seismic Reanalysis As-built As-built As-built As-built As-built Seismic Reanalysis Seis1uic Reanalysis As-built Seismic Reanalysis Sheet 7 of 12 RESOLUTION Nodity sup1,ort I 1 Modify support Hodify restraint for uplift loau llodify Sllp(JOrt Modity support Modify support t><ld 11eld Hodify support Two rii;id re-straints rt:!-

i,laced by sprin1c;s l*lodify support 1

Hodiiy support Modify support I 1 Modify support Hodily support

SURRY POWER STATION -

UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MOVUICATION

SUMMARY

SYSTEN AND PROllLEM NUMBER ANALYSIS RESPONSIBILITY Containment and Recirculation Spray (Cont'd) 2754 E

2755 E

2756 E

2757 E

Main Steam 2577 S&W 2579 S&W 2346 S&w MKS SUPPORT LOCATION.

NUMBER oc l

oc 7,8 9

11 IC 54,55 IC H-90 11-63, 11-88 IC 9

IC l

4 5

oc 1,2,3 REASON FOR MODIFICATION Support member over allowable Upward vertical restraint required Upward vertical re11traint required Weld over allowable Support member over a 11 owab le Upward ve rt ica 1 restraint required Support member over allowable Local stre11s ex-ceedR allowable Local stress ex-ceeds allowable Spring variability ratio exceeded Loads outside spring range Spring variability ratio exceeded.

Local stress over allowable.

Loads outside spring range, local over-stress in lug ATTRIBUTABLE.

TO

. As-built Seismic Reanalysis Sei11mic Reanalysis As-built Af1-buil t Seismic Reanalysis AR-built As-built AF1-bu i It Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Sheet 8 of 12 RESOLUTION Modify support

~1od i fy support for uplift load Redesign support for uplift load Add weld Modify support Modify Flupport for.

uplift load Modify,mp port Modify lug Replace lug with clamp Replace spring Replace spring Rep lace 1,"pr ing Replace pipe lug with clamp.

Replace springs, modify lug,

SY::.TE!l I\\NIJ l'RlJl.iLEM llUHLEl{

111.tin !>team (Cont'd)

'.l346 (Cont'd)

/,llt\\L Y SIS RESPOlJSlHILITY S&\\J SURRY POI./ER STATIOt; -

Ul'<lT 2 TABLE 4-2 I' lPE SUPPORT llARlJ~/ARE tlOIHFICATlON SlJHMARY

~1KS SUPPORT LOCATIOU llUlli.iER oc

'.>,7,9 RE'1S0ll FOR ll0D1FICATION S11ubbcrs, local stress, and sup-port hlembers are overstressed ATTRlliUTABLE TU Seismic Reanalysis Sheet I!;, of 12 REWLli'l'l!Jli

,*iodify snubber,;

and lug,;

SYSTi'.M NAME AND PROliLEN NUMl!ER Feedwater 25b'J 2571 Auxiliary Feedwater 2473 2683 ANALYSIS RESPONSllllLITY S&W S&W E

E SURRY POWER STATION -

UNIT 2 TABLE 4-2

(.......

PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC MKS SUPPORT NUMBER 2,3 6

7 5

8, 18 14 16 REASON F'OR MODIFICATION LoadR outRide spring range Thermal movement lnRufficient clearance for lateral movement LoadR outside spring range Lateral clearance in-RU f fi_c ient Upward vertical restraint required Upward vertical restraint required Weld over allowable

... <1111-,

ATTRIBUTABLE TO Seismic Reanalysis Seismic Reann lys is Sei11mic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis 11,13,26,29,32 22 Support member over allowable AR-built As-built As-built oc 28, 31 It-BA 11-11 H-9 11-1( l 18Gl),

l1-5(118Gl) 11-4 11-5( 118G2)

Support member over allowable U-bolt over allowable U-bolt over allowable Support member over allowable Local pipe wall *'stress over al Jow_able U-bolt over allowable Upward vertical restraint required As-built As-built As-built Seismic ReanalyRis As-built Seismic

_Reana I yn is Sheet 'J of 12 RESOLUTION Replace springs Reduce pin-to pin dimension Modify'support Replace springs Modify reRtraint Modify restraint for uvlilt load Modify restraint for uplift load Modify support Modify Rupport Modify support Modify support Modify support

~,od i fy support Modify support Modify support Modify support I

SYSTEM NAME AND PROBLEH NUMBER Auxiliary feedwater (Cont'd) 2683 {Cont'd)

Pressurizer Spray and Relief 2771 2000 Residual Heat Removal 2508A/2508B 2540B ANALYSIS RESPONSIBILITY E

E E

SURRY POWER STATION -

UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE HODIFICATIOll SUHMARY LOCATION oc IC lC IC IC HKS SUPPORT NUMBER 11-l( ll8G2) 33 5,23 6,24 26,27 4,12,H-5,B,10,17 15 13 4A,7,H-1A, ll,21A,113 11-2 ll-36,H-15,11-17 11-12 20,21 REASON FOR IIODHICATION Support does not allow lateral movement Upward vertical restraint required l~eld over allowablj!

Support member over allowable Upward vertical restraint required Local pipe wall stress over allowables Insufficient vertical clearance Support restraints lateral movement Support member over allowable Weld over allowable Support member over allowable Vertical support not required Upward vertical restraint required ATTRIBUTABLE TO As-built Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built Seismic Reanalysis Seismic Reanalysis Sheet 10 of 12 RESOLUTIOII Modify support Modify support for uplift load Add weld Hodify support Modify for uplift load Modify support Modify* support Modify support Modify support AdJ weld Modify support Remove support 11 I

11 1

HoJify for uplift 11 load

SY8TEH NAME AND PROllf,Ell NUHBER Residual Heat

~tl (Cont'd) 2540!1 (Cont'd) 25'>4 Service Water 2465 2467 2471 Component Cooling 2601 2603 2604 ANALYSIS RESPONSIBILITY E

E E

SURRY POWER STATIO!J -

U!HT 2 TABLE 4-2 PIPE SUPPORT HARDWARE !IODIFICATION SUHHARY

)

LOCATION MKS SUPPORT NUHBER REASON FOR MODIFICATION ATTRIBUTABLE TO IC 23 22 oc l!-31 IC IC IC 2

IC ll-32A, ll-28,H-44 ll-30 11-31 IC 4

6,8 9

10 IC ll-38A Insufficient lateral clearance Seismic Reanalysis Support member over allowable Insufficient lateral clearance Support member over allowable Support member over allowable Local pipe wall stress over allowable Uplift vertical restraint required Insufficient lateral clearance Weld over allowable Insufficient lateral clear-ance Support member over allowable Upward vertical restraint re4uired Upward vertical restraint required Upward vertical restraint required As-built Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Sheet 11 of ll RESOLUTION Modify support Hodify support Modify support Modify support Modify support 1-lodify support Modify support uplift Modify support Add weld Modify support Hodlfy support Modify support uplift load-Modify support uplift load Modify support uplift load for for for for 1

1

SYSTLm NAHE Alli>

PROBLEtl NUHBER

£omponent Cooling (Cont'd) 2604 (Cont'd) 2605 Containment Vacuum 2650 High Pressure Steam To Aux, Feedwater Pump 2862 2869 Notes:

ANALYSIS RESPONSIBILITY E

E S&W E

E

- Originally Problem No, 2708 j..

SURRY POI/ER STATION -

UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC oc oc.

oc NKS SUPPORT NUHUER U-J8B H~J5,H-36,H-38C H-38 H-38D U-25 U-25B U-25A 11-23 2

3 3

5 RMSON FOR MODIFICATION Up>1ard vertical restraint required Weld over allowable Support member over allowable Local pipe wall stress over allowable Support member over allowable Upward vertical restraint required Upward vertical restr~int required Weld over allowable Support restraint lateral movement Local stress at at-tachment on pipe exceeds allowable Local stress Sup1>0rt member over allowable Support member over allowable ATTRIBUTABLE TO Seismic Reanalysis As-built As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis As-built As-built As-built Seismic Reanalysis As-built As-built Sheet 12 of 12 RESOLUTION Modify support for 11 uplift load Add >1eld 11 Modify support Modify support 11 Modify support Modify support for uplift load Modify support for uplift load Add >1eld 11 Hodify support Move support above l

elbo>1 and use trun-I nion Modify support 11 Modify support Modify support

).....,....

Location of rroblem lnslde Containment Systems Outside t:ontainment Systems I.ow liead Safety Injection lligh llead Sal ety injection Containment Recirculation Spray

,,ulliliary Feedwater Balance of Systems NOTES:

E SC.W EIIASCO Stone & Webster SURRY l'OWlrn STATION -

UtllT 2 TAULE 5-l SCIIEOUl,E FOR COMl'l,I\\TlON Status Reau.1Iysls Respwsibi lity

- Stress--Support/Restni.1.nt -*--* l-lodlflcation

E/S&W E/S&W E

E E

E/S&W_

Reanaly~is Reanalysis..

____ _l!.!.'!~al_!._a_t_!.~---

Complete Complete Complete Complete Complete t;omplete Complete

.Complete Complete Complete Comvlete Complete l'rlor to start-up following SGR outar,e l'rior to start-up fol.lowJ.ng SGR outage Prior to start-up following S~R outage Prior to start-up followinr, SGR outage rrior to start-up fo llowlng S(;R outage Prior ~o start-up following SGR outage S~R Steam Generator Replacement

- -.(.. -

I,,

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I SECTION 5 SCHEDULE FOR COMPLETION The status of the reanalysis of those systems subj.ect the installation of modifications identified as being reanalysis is shown in Table 5-1, Schedule for Completion.

to Show Cause required by and the Reanalysis on all systems is complete pending final review.

modifications on all lines will be installed prior to start Unit following the Steam Generator Replacement Outage.

5-1 Required up of the 1

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. 1-I I

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SURRY POWER STATION -

UNIT 2 SECTION 6 HIGH ENERGY LINE BREAKS For the high energy lines outside the containment addressed in Appendix D of the Final Safety Analysis Report (FSAR), only the main steam lines are included in this stress reanalysis.

Each of the main steam lines has two terminal break locations,. one at the containment penetration and the other at the main steam manifold.

Each of the risers to the main steam relief valve headers has two terminal break locations, one at the main steam lines, the other at the tee into the main steam header.

These terminal breakpoints are predetermined and are not changed as a result of the stress reanalysis.

Two intermediate break locations were originally determined based upon maximum primary plus secondary stresses.

Upon reanalysis, two additional breakpoints on each of the steam lines were located.

One of these points is located immediately upstream of* the check valve (TV-MS201A, TV-MS201B, TV-MS201C) and the other point is at the elbow just downstream of the check valve.

All of these points will be included in the augmented inservice inspection program.

6-1

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I SURRY POWER STATION -

UNIT 2 SECTION 7 CONSERVATISMS The conservatisms applied to the design of the p1p1ng systems for Surry Power Station Units 1 and 2 were extensively delineated in Section 7 of the VEPCO June 5, 1979 submittal (Serial Number 453).

The seismic capability of nuclear piping and the seismic event probability at the Surry Power Station were discussed in that submittal.

The design of Unit 2 closely follows the design of Unit 1, applying the same conservative criteria with respect to safety systems and system redundancies.

Similiarly, the reanalysis efforts on Surry Power Station Unit 2 close*ly follows that of Unit 1, applying the same stress limits and soil struc-ture interaction amplified response spectra (SSI-ARS).

Paragraphs 7.1, 7.2, 7.3 and 7.4 describe the differences in the conser-vatisms applied to the Unit 2 reanalysis.

7.1 FIELD VERIFICATION OF AS-BUILT CONDITIONS To ensure that the pipe stress and pipe support reanalysis is performed as accurately as possible, field verification of as-built conditions has been performed.

The field verification produced detailed piping isometric drawings and pipe support sketches for each support upon which reanalysis is based.

All field-verified piping isometrics and pipe support sketches are independently verified by Surry Power Station quality control personnel.

7.2 7.2.1 QUALITY ASSURANCE/ENGINEERING ASSURANCE EBASCO QUALITY ASSURANCE The EBASCO QA Topical Report ETR-1001, Revision 7, as approved Nuclear Regulatory Coilllllission on December 15,

1978, is being to the Surry Unit 2 reanalysis activities.

7.2.2 STONE & WEBSTER QUALITY ASSURANCE/ENGINEERING ASSURANCE by the applied The Stone

& Webster Quality Assurance program described in the VEPCO June 5, 1979 submittal to NRC, is being applied to the Surry Unit 2 reanal-ysis activities.

7.3 USE OF AMPLIFIED RESPONSE SPECTRA The use of amplified response spectra was extensively discussed in the June 5, 1979 submittal.

The soil structure interaction amplified response Spectra (SSI-ARS) are being used in the reanalysis in most cases.

For pipe runs extending over a range of elevations S&W and EBASCO utilized an amplified response spectra enveloping the acceleration of the mass points spanning the elevation of the piping run.

7-1

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I.,

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I SURRY POWER STATION -

UNiT 2 7.4 CONSERVATISMS APPLIED TO INERTIAL STRESSES In accordance with the NRC letters of May 25, 1979 and November 15, 1979 to VEPCO, the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a factor of.1. 5 for the DBE and l. 25 for OBE conditions, 7-2

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I SURRY POWER STATION -

UNIT 2 SECTION 8 SYSTEM OPERABILITY EVALUATION This section has been deleted.

Since all modifications will be installed prior to startup following the Steam Generator Replacement Outage, a system operability evaluation is no longer necessary.

~1 1

I I

I SECTION 9 BRANCH LINE

SUMMARY

Branch lines are evaluated to assure that sufficient flexibility exists between the run pipe and the first few restraints on the branch piping.

The flexibility of the branch pipe must be evaluated separately in each of the three translational directions and must be sufficient to prevent overstresses in the branch/run pipe interface due to thermal and seismic displacements imposed on the branch pipe.

The procedure is intended to provide a secondary stress check based on run pipe displacements result-ing from the current analysis.

If a branch line is part of the scope of work under IE Bulletin 79-14, a detailed evaluation is performed as part of the IE 79-14 ~ffort.

S&W has performed evaluation of branch lines in accordance with Section 6 of the August 1, 1979 report for Unit 1 (Vepco Serial No. 453A).

EBASCO has performed evaluation of some of the branch lines by coding for the NUPIPE program and analyzing it for seismic anchor movement and thermal analysis.

Engineering judgement is used in qualifying the branch lines with small displacements in the remaining cases.

Thermal analysis is conducted by applying the thermal displacements from the run pipe and the operating temperature of the branch line.

The seismic anchor movement analysis is performed by applying seismic inertia displacements.

The applicable stress intensification factor (SIF) at the branch connection is included in the analysis.

The stresses from both analyses are combined by absolute sum.

Allowable stress is considered to be SA, 9-1

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SURRY POWER STATION -

UNIT 2 SECTION 10 RESPONSE TO NRC STAFF CONCERNS A meeting was held with the Nuclear Regulatory Commission at Ebasco Services, Jericho Offices on October 24,

1979, to review pipe stress analyses within EBASCO' s scope of work.

As a result of the discus-

sions, four starf concerns were identified as delineated in the NRC Summary of Meeting Notes dated November 13, 1979.

These concerns were:

1)

2)

3)

4)

The validity of support stiffness used in the piping reanalysis when, for example, a vertical trunion is welded onto a horizontal wide flange.

The pertinence of the version of B31. l Code implemented in Contro1Data Corporations' NUPIPE program, which was used in the EBASCO reanalysis program.

The identification of the original loads on support H-15 in problem 2538.

The verification of the NUPIPE computer program (benchmark problems).

These concerns are addressed i.~ the following sections.

10. 1 SUPPORT STIFFNESS The original piping analysis of Surry Unit 2 did not consider the actual stiffness of the supports.

Representative support stiffness was considered during the current reanalysis.

During the pipe support reanalysis effort it has been observed that cer-tain anchor type supports expose wide flange members to torsional moments.

This type of loading condition results in a very flexible support.

As a part of the pipe support reanalysis effort, anchors have been reviewed for this type of loading and members modified to resist torsion as required.

10.2 NUPIPE COMPUTER CODE At EBASCO's

request, Nuclear Services Corporation (NSC) conducted a

thorough review of the NUPIPE program against the source codes, NSC has determin'!d that all values utilized by the program, but not spec-ified by the user as input, are pertinent to the 1967 and earlier versions of the B31. l Power Piping Code.

The code of record for Surry Unit 2 is B31.l 1955 with code class N-7.

10.3 PROBLEM 2538 -

SUPPORT H-15 Problem 2538, a portion of the Low Head Safety Injection System (LHSIS) was originally analyzed as a hand calculation by S&W; SHOCK 2 was not used, therefore, problem 2538 is not within the scope of the Show Cause Order.

In the original analysis, decal loads were applied to the re-straints in this problem.

These decal loads.did not include moments.

10-1

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I SURRY POWER STATION -

UNIT 2 In the EBASCO NUPIPE analysis of the portion of the LHSIS within pro-blem 2538, the system was not overstressed.

However, the loads iden-tified by the NUPIPE analysis as existing at hanger 15 caused local pipe wall and support anchor stresses to exceed allowables by an order of magnitude.

Support H-15 has been modified so that it will relieve the local overstress conditions.

10.4 BENCHMARK PROBLEMS EBASCO has performed four pipe stress problems supplied by the NRG to verify the NUPIPE computer program.

The results were submitted to the staff in the EBASCO letter to Dr M Hartzman dated January 3, 1980 (Letter Number VEP/NRC/002).

I I

SURRY POWER STATION - UNIT 2 I

I I

I APPENDIX A I

SYSTEMS AFFECTED I

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I.

A-1

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I*

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SURRY POWER STATION - UNIT 2 The reanalysis included those safety related lines originally computer-analyzed with the SHOCK2 program.

The systems line numbers, the associated computer problem numbers, and the flow diagram numbers are listed below.

The.

following table includes all seismically analyzed lines.

The figure numbers re-fer to the FSAR drawings, and the Surry Unit 2, FM and FP drawings included in

/ 1 Appendix B.

System Low Head Safety Injection High Head Safety Injection Responsi-bilities Line No.

for Analysis 8-SI-214-153 E

8-SI-292-153 E

8-SI-214-153 E

8-SI-292-153 E

10-SI-284-152 S&W 10-SI-216-153 S&W 8-SI-292-153 S&W 10-SI-351-153 S&W 6-SI-249-1502 S&W 10-SI-349-153 S&W 8-SI-214-152 S&W 10-SI-283-152 S&W 10-SI-213-153 S&W 6-SI-248-1502 S&W 10-SI-352-1502 S&W 10-SI-350-153 S&W 10-SI-349-153 S&W 10-SI-348-153 S&W 8-SI-214-153 S&W 12-SI-247-602 E

12-SI-247-1502 E

12-RC-324-1502 E

10-RH-117-1502 E

12-SI-246-602 E

12-SI-246-1502 E

12-RC-323-1502 E

10-RH-116-1502 E

6-SI-248-1502 E

6-SI-249-1502 E

6-SI-250-1502 E

6-RC-321-1502 E

6-SI-343-1502 E

12-SI-245-602 E

12-SI-245-1502 E

12-RC-322-1502 E

10-SI-206-153 E

6-CH-372-152 E

4-CH-412-152 E

3-CH-373-152 E

8-SI-214-153 E

A-2 Problem No.

2695 2697 2681 2682 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2537 2537 2537 2537 2555 2555 2555 2555 2539 2539 2539 2539 2539 2709 2709 2709 2689 2735 2735 2735 2735

tv.lKS No.

127Dl 127D2 127Kl 127K2 127Cl 127Cl 127Cl 127Cl 127Cl 127Cl 127Cl 127C2 127C2 127C2 127C2 127C2 127C2 127C2 127C2 122Al 122Al 122Al

122Al, 117Bl 122Dl 122Dl 122Dl 122Dl 122Kl 122Kl 122Jl 122Jl 122Kl 12211 l22Ll 12211 127Fl 127Gl 127Gl 127Gl 127G2 Flow Diagram No.

FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-lGoA FM-10~1' FM-106A FM-106B FM-106B FM-106B FM-106B, 104A F:tvl-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106A FM-105B FM-105B FM-105B 106A

I I

SURRY POWER STATION -

UNIT 2 Responsi-Flow I

bilities Problem MKS Diagram System Line No.

for Analysis No.

No.

I High Head 8-CH-504-152 E

2735 127Gl, G2 105B Safety 8-CH-317-152 E

2735 127Gl, G2 105B Injection 8-SI-217-152 E

2735 127Gl, G2 106A I

(Cont'd) 3-SI-292-153 E

2735 127Gl 106A 6-SI-218-152 E

2735 127Gl, G2 105B, 106.A:

6-SI-219-152 E

2735 127Gl, G2 105B, 106A 6-SI-278-152 E

2735 127Gl, G2 105B, 106A I

6-CH-501-152 E

2735 127Gl 105B 6-CH-502-152 E

2735 127Gl 105B 6-CH-503-152 E

2735 127Gl 105B I

8-CH-505-152 E

2735 127G2 105B 8-CH-506-152 E

2735 127Gl, G2 105B 8-SI-207-152 E

2735 127G2 106A 8-SI-302-152 E

2735 127G2 106A I

8-SI-170-153 E

2735 127G2 106A 8-SI-172-153 E

2735 127G2 106A 10-SI-206-153 E

2735 127G2 106A I

6-CH-318-152 E

2735 127Gl 105B 6-CH-319-152 E

2735 127Gl 105B I

Residual 14-RH-101-1502 E

2508B 117Al FM-104A Heat Removal 14-RH-102-602 E

2508B 117Al FM-104A L

10-RH-104-602 E

2508B 117Al FM-104A 10-RH-105-602 E

2508B 117Al FM-104A I

12-RH-106-602 E

2508B 117Al FM-104A 10-RH-107-602 E

2508B 117Al FM-104A 10-RH-108-602 E

2508B 117Al FM-104A I

10-RH-109-602 E

2508A 117Al FM-104A 10-RH-110-602 E

2508A 117Al FM-104A 12-RH-112-602 E

2508B 117Al FM-104A 14-RH-118-602 E

2508B 117Al FM-104A 1::

1.

12-RH-119-602 E

2508A 117Al FM-104A 12-RH-112-602 E

2540 117Bl FM-104A 3-RH-113-602 2540B E

117Bl FM-104A I 1.

I 4-RH-115-152 E

2540B 117Bl FM-104A 10-RH-116-1502 E

2540 117Bl FM-104A 10-RH-117-1502 E

2540 117Bl FM-104A I

6-RH-120-152 E

2540 117Bl FM-104A 10-RH-137-602 E

2540 117Bl FM-104A 6-RH-120-152 E

2554 117Cl FM-104A, 101A I

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I A-3

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_SURRY POWER STATION -

UNIT 2 Responsi-Flow I

bilities Problem MKS Diagram System Line No.

for Analysis No.

No.

I Main.Steam 30-SHP-101-601 S&W 2577 1000 FM-14A 11 30-SHP-102-601 S&W 2588 101D FM-l4A 30-SHP-103-601 S&W 2579 1020 FM-14A I

30-SHP-101-601 S&W 2346 103A FM-14A 30-SHP-102-601 S&W 2346 103A FM-14A 30-SHP-103-601 S&W 2346 103A FM-14A 30-SHP-124-601 S&W 2346 103A FM-14A I

30-SHP-123-601 S&W 2346 103A FM-14A 30-SHP-122-601 S&W 2346 103A FM-14A 1-Feed water 14-WFPD-117-601 S&W 2569 100G FM-18A 14-WFPD-113-601 S&W 2573 101G FM-lBA 14-WFPD-109-601 S&W 2571 102G FM-18A I

Auxiliary 6-WAPD-101-601 E

2473 llBAl, A2 FM-IBA, 18B Feed water 6-WAPD-102-602 E

2473 118A2 FM-18A, 18B 3-WAPD-109-601 E

2473 118Al, A2

-FM-18A I

3-WAPD-110-601 E

2473 118Al, A2 FM-18A 3-WAPD-111-601 E

2473 118Al, A2 FM-1SA 3-WAPD-112-601 E

2473 118Al, A2 FM-18A I

3-WAPD-113-601 E

2473 118Al, A2 FM-18A 3-WAPD-114-601 E

2473 118Al, A2 FM-lBA 6-WAPD-150-601 E

2473 118Al, A2 FM-18A, 18B 6-WAPD-151-601 E

2473 118Al, A2 FM-18A, 18B I

6-WAPD-101-601 E

2683 11SG2 FM-18A, 18B 6-WAPD-I02-601 E

2683 118GI FM-ISA, 18B 6-WAPD-I03-601 E

2683 118Gl, G2 FM-IBA I

6-WAPD-I04-60I E

2683 118Gl, G2 FM-18A 4-WAPD-105-60I E

2683 118GI, G2 FM-18A 4-WAPD-I06-601 E

2683 118Gl, G2 FM-l8A 4-WAPD-I07-601 E

2683 118GI, G2 FM-18A I

4-WAPD-IOS-601 E

2683 118Gl, G2 FM-18A 6-WAP0-50-601 E

2683 118GI, G2 FM-18A,18B 6-WAPD-52-60I E

2683 118Gl, G2 FM-18A, I8B I

Service Water 24~WS-126-10 E

2465 119Al FM-21A 24-WS-I28-10 E

2467 ll9A2 FM-21A I

24-WS-I30-10 E

2469 119A3 FM-21A 24-WS-132-10 E

2471 119A4 FM-21A I

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I A-4

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SURRY POWER STATION -

UNIT 2 I

Responsi-Flow bilities Problem HKS Diagram System Line No.

for Analysis No.

No.

I Pressurizer 4-RC-334-1502 E

2000 124Al FM-103B Safety and 3-RC-335-1502 E

2000 124Al FM-103B Relief 3-RC-361-1502 E

2000 124Al FM-103B I

6-RC-320-602 E

2000 124Al, A2 FM-103B 6-RC-362-602 E

2000 124Al, A2 FM-103B 12-RC-336-602 E

2000 124Al, A2 FM-103B

1 6-RC-337-1502 E

2000 124Al, A2 FM-103B 6-RC-338-1502 E

2000 124Al FM-103B 6-RC-339-1502 E

2000 124Al FM-103B 6-RC-340-602 E

2000 124Al, A2 FM-103B I

6-RC-341-602 E

2000 124Al, A2 FM-103B 6.;..RC-342-602 E

2000 124Al, A2 FM-103B I

Pressurizer 4-RC-314-1502 E

2771 125Al FM-103B Spray 4-RC-315-1502 E

2771 125Al FM-103B 2-CH-368-1502 E

2771 125Al FM-103B I

HP Steam to 4-SHP-125-601 E

2862 131Al FM-14A Auxiliary 3-SHP-132-601 E

2862 131Al FM-14A Feed water 3-SHP-128-601 E

2862 131Al FM-14A I

Pump 3-SHP-131-601 E

2862 131Al FM-14A 3-SHP-157-601 E

2862 131Al FM-14A 4-SHP-126-601 E

2864 131Bl*

FM-14A I

3-SHP-129-601 E

2864 131Bl FM-14A 4-SHP-127-601 E

2869 131Cl FM-14A 3-SHP-130-601 E

2869 131Cl FM-14A 3-SHP-135-601 E

2869 131Cl FM-14A*

I Containment 10-CS-104-153 S&W 2521 123Al FM-lOlA and Recir-8-CS-123-153 S&W 2521 123Al FM-lOlA culation Spray 10-CS-103-153 2523 123Al FM-lOlA I

I S&W 11 8-CS-122-153 S&W 2523 123Al FM-lOlA 10-CS-103-153 S&W 2547 123Cl FM-lOlA 8-CS-133-153 S&W 2547 123Cl FM-lOlA I

10-CS-104-153 S&W 2549 123C2 FM-lOlA 8-CS-134-153 S&W 2549 123C2 FM-lOlA I

10-RS-112-153 E

2546 123Dl FM-lOlA 8-RS-123-153 E

2546 123Dl FM-lOlA 10-RS-104-153 E

2541 123D2 FM-lOlA 8-RS-121-153 E

2541 123D2 FM-lOlA I

10-RS-103-153 E

2542 123D3 FM-lOlA 8-RS-120-153 E

2542 123D3 FM-101A I

l 0-RS-111-15 3 E

2543 123D4 FM-lOlA 8-RS-122-153 E

2543 123D4 FM-lOlA I

I A-5

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I SURRY POWER STATION _- UNIT 2 Responsi-bilities Problem System Line No.

for Analysis No.

Containment 10-RS-112-153 E

2560 and Recir-10-RS-104-153 E

2561 culation Spray 1 O-RS-110-153 E

2544 (Cont'd) 10-RS-109-153 E

2533 10-RS-103-153 E

2548 10-RS-111-153 E

2545 8-CS-134-153 E

2744 8-CS-133-153 E

2745 12-CS-102-153 E

2753 12-CS-101-153 E

2754 10-RS-109-153 E

2751 4-RS-114-153 E

2751 10-RS-110-153 E

2752 4-RS-115-153 E

2752 8-CS-133-153 E

2755 8-CS-134-153 E

2755 4-CS-135-153 E

2755 4-CS-136-153 E

2755 4-CS-105-152 E

2755 1/2-CS-108-153 E

2755 4-CS-106-152 E

2755 10-RS-101-153 E

2756 10-RS-102-153 E

2757 Component 18-CC-15-121 E

2604 Cooling 18-CC-9-121 E

2605 18-CC-7-121 E

2601 18-CC-14-121 E

2603 Containment 8-CV-108-151 S&W 2650 Vacuum Note:

E = EBASCO S&W = Stone & Webster A-6 Flow MKS Diagram No..

No.

123El FM-lOlA 123E2 FM-lOlA 123Gl FM-101A 123G2 FM-101A 123Hl FM-101A 123H2 FM-101A 123Jl

.FM-lOlA 123Kl FM-101A 12311 FM-101A 123Ml FM-lOlA 123Nl FM-lOlA 123Nl FM-101A 123N2 FM-lOlA 123N2 FM-101A 123Pl FM-lOlA 123Pl FM-101A 123Pl FM-lOlA 123Pl FM-lOlA 123Pl FM-lOlA 123Pl FM-lOlA 123Pl FM-lOlA 123Ql FM-101A 123Q2 FM-lOlA 112AA1 FM-22A.

112AB1 FM-22A 112Sl FM-22A 112S2 FM-22A 137A FM-102A

I I

SURRY POWER STATION -

UNIT 2 I

I I

I APPENDIX B I

FLOW DIAGRAMS -

IDENTIFICATION OF PROBLEMS REANALYZED I

'I I

I I

I B-1

I I

I SURRY POWER STATION - UNIT 2 APPENDIX B FLOW DIAGRAMS - IDENTIFICATION OF PROBLEMS REANALYZED Title Main Steam Feed water Cross-Connects for Auxiliary Feed Circulating and Service Water Component Cooling Containment and Recirculating Spray

~ontainment Vacuum and Leakage Monitor Reactor Coolant Sheet 1 Reactor Coolant Sheet 2 Residual Heat Removal Chemical and Volume Control Sheet 2 Safety Injection Sheet 1 Safety Injection Sheet 2 Refueling Water Storage Tank Crosstie B-2 Drawing No.

11548-FM-14A 11548-FM-lBA 11448-FM-lBB 11548-FM-21A ll 548-FM-22A 11548-FM-lOlA 11548-FM-102A 11548-FM-lOJA 11548-FM-lOJB 11548-FM-104A 11548-FM-l 05B 11548-FM-l 06A 11548-FM-106B 11448-FM-106C 1

I I

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I FL.DN DIAGRAM

.MAIN STEAM

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1S7Z U'TfNSIOH-SURRY l'OWf" STATION ILECTIIIC AND POWEii COiii-AN 11s4a-n,,1-14A~*

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. -r*~v.:1s-5oa.a i..FVL=L Tli.'AN5MITTlf.~S SEE ACCIJMVl.4TV.~ TA1'o'r ~~~REF DW::i5. - FLll*IO<;.:,A l,. AOOJTlc>>JAL. Ci.tECK VALVE'S.~!:ITALLE.~~~

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ORIGfNAL ISSUE t' 11*9-76 J" f L
FLOW a VALVE NUMBERS R.WS.T. CROSSTIE SURRY POWER STATION V.
E.
P.
C.
0.
11448-FM-IOSC l
a 4
II T
I I
SURRY POWER STATION - UNIT 2 I
I I
I I
I I
I APPENDIX C RESPONSE TO I
IE BULLETIN 79-04 I
I I
I I
I I
I C-1
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I I
I I
I SURRY POWER STATION - UNIT 2 APPENDIX C RESPONSE TO IE BULLETIN 79-04 Velan swing check valves, sized 3 and 6 inches, following seismic Category I piping systems:
a)
Chemical and volume control system b)
Safety injection systems A detailed listing by line number is contained in are installed in the the following tabl.e.
Lines with 6 inch check valves were originally seismically analyzed by computer program or hand calculations.
The re-evaluation of these systems using the correct valve weight is currently being done under the NUPIPE program.
The results have shown that the pipe stress ts within the allow-able for all lines.
Lines with 3 inch check valves were tions.
An estimated weight, overly actual valve weights:
The incorrect calculations and re-evaluation is not the related pipe lines are included C-2 analyzed originally by hand calcula-conservative, was used instead of valve weight has no effect on these required, however, these valves and in the scope o*f IE Bulletin 79-14.
I I
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I I
I I
1.
I I
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I I
SURRY POWER STATION -
UNIT 2 LISTING OF VELAN SWING CHECK VALVES COVERED BY IE BULLETIN NO. 79-04 SAFETY INJECTION SYSTEMS - UNIT 2 6 Inch 3 Inch 2-SI-79 2-SI-82 2-SI-85 2-SI-88.
2-SI-91 2-SI-94 2-SI-228 2-SI-229 2-SI-238 2-SI-239 2-SI-240 2-SI-241 2-SI-242 2-SI-243 2-SI-224 2-SI-225 2-SI-226 2-SI-227 CHEMI.CAL AND VOLUME CONTROL SYSTEM - UNIT 2 3 Inch 2-CH-196 2-CH-258 2-CH-267 2-CH-276 2-CH-309 2-CH-312 C-3 6-RC-317-1502 6-RC-319-1502 6-RC-320-1502 6-RC-318-1502 6-RC-316-1502 6-RC-321-1502 6-SI-249-1502 6-SI-249-1502 6-SI-248-1502 6-SI-249-1502 6-SI-250-1502 6-SI-345-1502
.6-SI-344-1502 6-SI-353-1502 3-SI-346-1503 3-SI-270-1503 3-SI-347-1503 3-SI-272-1503 3-CH-500-1502
~~3-CH.;..381-1503 3-CH-3 0 2-1503 3-CH-303-1503 3-CH-3 7 9-1503 3-CH-301-1502~~
I I
SURRY POWER STATION - UNIT 2 I
I I
I I
I I
APPENDIX D I
CORRESPONDENCE WITH THE I
NRC I
I I
I I
I I
I D-1
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I I
I SURRY POWER STATION - UNIT 2 APPENDIX D CORRESPONDENCE WITH NRG The following is a listing of correspondence with the NRC related to the reanalysis effort.
Item No.
1 2
3 4
5 6
7 8
9 10 11 12 13 Date 3/13/79 4/2/79 4/13/79 5/18/79 5/25/79 7 /18/79 8/15/79 8/27/79 10/5/79 10/23/79 10/24/79 10/25/79 11/15/79 Signature Denton Stello Stello Stello Eisenhut O'Reilly O'Reilly Denton 0' Reilly Murphy Murphy Murphy Eisenhut Addressee NRC TO VEPCO D-2 Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Letter No./Subject Show Cause Order Addendum to Show Cause Order Use of Soil Structure Interaction Techniques Request for Further SSI Information Factor Adjustment to SSI Calculated Stresses Information Pertaining to IE Bulletin No.
79~14, Revision 1 Letter of Guidance on IE Bulletin No. 79-14 Lifting of Suspension Required by the Order to Show Cause Confirmation of Concur-rence Refers to NRG Inspection of Sept" 10-13 and Sept.
19-21, 1979 Refers to NRG Inspection of Sept. 13-14, 1979 Refers to NRG Inspection of Sept. 26-28, 1979 Refers to Soil Structure Interaction
I I
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I Item No.
14 15 16 17 18 19 20 21 22 23 24 25 26 27 Date 3/30/79 4/19/79 4/23/79 4/24/79 4/27 /79 5/2/79 5/2/79 5/22/79 5/24/79 5/24/79 6/5/79 6/8/79 6/8/79 6/12/79 SURRY POWER STATION - UNIT 2 APPENDIX D (Cont'd)
CORRESPONDENCE WITH NRC Signature Addressee VEPCO*TO NRC Spencer Denton/
Stello Stallings O'Reilly Spencer Spencer Spencer Stallings Spencer Ragone Spencer Spencer Spencer Spencer Spencer O'Reilly O'Reilly Denton/
Stello Denton Stello Hendrie Stello Stello Denton Denton Stello Spencer Denton D-3 Letter No./Subject 198/Initial Response to Show Cause Order 270/LER 79-010/0131-0 289/Response to IE Bulletin No. 79-07 288/Response to IE Bulletin No. 79-07 311/Transmittal to Two Sample Problems to EG&G Observ.ations on Reanalysis Effort 260/Submittal of SSI Information Comments on Moratorium/
Surry Reanalysis Response to NRC Letter of 4/2/79 Response to NRC Letter of 5/18/79 Submittal of Report on Reanalysis Additional Information, Report on Reanalysis of Piping Soil Structure Interac-tion Report Modification Informa-tion, Reanalysis of Piping Systems
I I
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I I
I I
I I
I Item No.
28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 Date 6/15/79 6/19/79 6/25/79 8/1/79 8/21/79 8/31/79 8/31/79 9/13/79 10/3/79 10/4/79 10/ 4/79 10/15/79 10/23/79 10/30/79 11/28/79 SURRY POWER STATION -
UNIT 2 APPENDIX D (Cont'd)
COR..~SPONDENCE WITH NRC Signature Spencer Spencer Spencer Spencer Spencer Spencer Spencer Spencer Proffitt Spencer Spencer Spencer Spencer Spencer Spencer D-4 Addressee Denton Denton Denton Denton Denton Denton O'Reilly Eisenhut Denton O'Reilly O'Reilly O'Reilly O'Reilly O'Reilly Denton Letter No./Subject Schedule and Support Information Support Modifications Support Information, Reanalysis of Piping Systems Submittal of Revised Report on Analysis Analysis Completion of Designated Supports -
Outside Containment Reanalysis of Piping Systems 60-Day Response for IE 79-14 Response to NRC Letter of 5/25/79 Seismic Analysis of Piping Systems Response to NRC Letter of 7 /2/79 Response to IE Letter Dated 9/7/79 Extension of IE Bulletin 79-14 Deadline Extension of IE Bulletin 79-14 Deadline, Unit 2 120-Day Response to IE 79-14 Seismic Analysis of Piping Systems 1
I I
SURRY POWER STATION -
UNIT 2 APPENDIX D (Cont'd)
I CORRESPONDENCE WITH NRC I
Item No.
Date Signature Addressee Letter No./Subject 43 12/7 /79 Spencer O'Reilly Response to NRC Letter I
of 11/8/79 44 12/13/79 Spencer Denton Show Cause Order I
Reanalysis 45 12/21/79 Spencer O'Reilly Show Cause 60 Days I
Analysis Completion 46 2/1/80 Spencer Denton Show Cause Modification Schedule Revision I
47 2/22/80 Spencer.
Denton Start-up Request for Surry Unit 2 I
48 3/21/80 Spencer Denton Amended Start-up Request for Surry Unit 2 1
I 49 3/28/80 Spencer Denton Show Cause Report Errata for Surry Unit 1 I
S&W to NRC so 3/22/79 Kennedy Denton Transmittal of S&W I
Computer Programs 51 3/30/79 Jacobs Herring Submittal of Computer Outputs I
52 4/3/79 Jacobs Bezler Submittal of Benchmark Problem to Brookhaven I
National Laboratory 53 4/6/79 Kennedy Denton Transmittal of S&W Computer Programs.
I 54 4/6/79 Jacobs Stello Plan for Verification of Dynamic Analysis I
Codes 55 4/11/79 Jacobs Bezler Submittal of Computer I
Outputs 56 4/13/79 Jacobs Stello Update and Status of Verification Plan for I
Dynamic Analysis Codes I
D-5
I I
SURRY POWER STATION -
UNIT 2 APPENDIX D (Cont'd)
I CORRESPONDENCE WITH NRC I
Item No.
Date Signature Addressee Letter No./Subject 57 4/18/79 Jacobs Hartman Submittal of Computer I
Outputs 58 4/27/79 Jacobs Bezler Submittal of Benchmark I
Problems 59 4/27 /79 Jacobs Stello Status of Verification Plan for Dynamic I
Analysis Codes 60 5/8/79 Rossier Neighbors Draft Outline of SSI-I ARS Report 61 5/9/79 Kennedy Stello Reference SHOCK 0 I
Program 62 5/11/79 Kennedy Stello Reference SHOCK 0 Program 1
I 63 5/14/79 Kennedy Denton Proprietary Computer Codes I
64 6/4/79 Jacobs Bezler Submittal of Benchmark Problems I
65 6/12/79 Jacobs Bezler Submittal of Benchmark Problems I
66 9/6/79 Allen Stello Re~ponse to NRC Letter of 8/10/79 I
Ebasco to NRC 67 9/7 /79 Nelson Hartzman Benchmark Problem VEP/NRC/001 I
68 1/3/90 Nelson Hartzman Benchmark Problem VEP/NRC/002 I
I I
I D-6

ML18139A328

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Dear Mr Denton:

Subject:

References:

Dear Mr. Denton:

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