ML18116A263

Revised Rept on Reanalysis of Safety-Related Piping Sys

ML18116A263
Person / Time
Site:	Surry
Issue date:	08/31/1979
From:	STONE & WEBSTER ENGINEERING CORP.
To:
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ML18116A262	List: ML18116A261 ML18116A263
References
NUDOCS 7908020481
Download: ML18116A263 (82)

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{{#Wiki_filter:SED REPORT ON THE IS OF SAFETY-RELA!ED PIPING SYSTEMS SURRY FO~ER STATION, UNIT l ELECTRIC AND POWER COMPANY ---**--*----p

I I I I I I I I I I I I I I I I

• 1 I

I SURRY POWER STATION, UNIT 1 REVISED REPORT ON THE REANALYSIS OF SAFETY-RELATED PIPING SYSTEttS SURRY POWER STATION, UN'IT 1 VIRGINIA ELECTRIC AND POWER COMPANY AUGUST 1979 s:,.az,ao l;k-8-F1C\\ t't0862...C>4-t8 '::r"'e-... ~ v-c::::c- -,-*... **c::::--*c: _..., __._ __, --C.--c.-.. --i::---

I I I I* I. I I I I I I I (' I I I., I I I I SURRY POWER STATION, UNIT l TABLE OF CONTENTS Section Title

l.

SUMMARY

AND CONCLUSIONS SCOPE OF REANALYSIS

3.

PIPE STRESS RESULTS **

4.

PIPE SUPPORTS RESULTS *

5.

DECISION MATRIX ****

6.

BRANCH LINE

SUMMARY

7.

s.

FIELD VERIFICATION

SUMMARY

APPLICATION OF BUILDING DISPLACEMENTS AND ANCHOR MOVE~.EN! STRESSES. *. * * * * * *. * * * * * * * * * * * *. APPENDIX A. B.

c.

SYSTEMS AFFECTED *****.*** FLOW DIAGRAMS - IDENTIFICATION OF PROBLEMS REANALYZED. CORRESPONDENCE WITH NRC ***********..*** i . l-l

• 2-l 3-l
• * * *
• 4-1
• 5-l 6-l 7-l
  • • 8-l
• *
• A-l
• B-l
• C-l

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 1

SUMMARY

AND CONCLUSIONS In response to the Nuclear Regulatory Commission's Order to Show Cause, dated March 13, 1979, a reanalysis is being conducted of safety-related piping systems which were originally dynamically analyzed using the SHOCK 2 computer program. This program, which used an earlier load combination methodology, is no longer considered acceptable by the NRG. This report updates information previously presented in the report submitted June 5, 1979, CVE?CO Serial No. 453), and subsequent letters from VEPCO to the NRC. The reanalysis work performed to date is consistent with the methods discussed in the earlier report; a detailed presentation of these methods will not be repeated herein. The present report summarizes progress to date, and based on this program, it seems appropriate to permit operation of the Surry Unit l plant on an interim basis. At this point, all the piping systems affected by the Order to Show Cause, both inside and outside the contain~ent, have been reanalyzed using the ~~ogram NU?I?! ~hich is acceptable to the NRC. So=e ~~cii!icaticns to these ~~ :ao:e 3-3. So~e 1-:

I I I I I I I I I I t I I I I I I I I SURRY POWER STATION, UNIT 1 analysis method, but, by far, the majority are attributable to differences in as-built conditions and other reasons. All of these modifications have been made or will be made on or about August 10, 1979, and all Surry l piping meets

SAR allovables for the DBE condition.

All pipe supports inside the containment affected by the Order to Show Cause have been evaluated for the revised support loads from the pipe stress reanalysis. Modificatiqns resulting from the support reanalysis are shown in Table 4-4; again, in most cases, the modifications are caused by differences between the original design conditions and the actual field as-built conditions. All of these hardware modifications have been made or will be made on or about August 10, 1979. The remaining pipe supports outside containment will be analyzed and any modifications identified within 60 days of interim startup. Any over-stressed conditions identified during this reanalysis for which the loads exceed one-half the ultimate capacity of the support will be reported to the NRC and appropriate actions taken as required by the Technical Specifications, as describedin Section 5 of this report. Based on the results of the analysis of supports inside containment, it is expected that very few, if any, supports outside containment have a safety factor of less than 2 with respect to ultimate capacity. Table 4-4 indicates 1-2

--.. -~ SURRY POWER STATION, UNIT 1 that almost all of the supports inside containment meet this criteria. The remainder have not been shown to meet this safety factor primarily due to limitations of the analysis. !he analytical techniques used are intended to produce conservative results for design purposes, and are not intended to be used for the calculation of ultimate capacities, for which different techniques are needed. These techniques include large deflection theory, increased damping effects on elastic analysis, and finit~ element analysis for such details as welded attachments. When such factors are taken into

• consideration, it is likely that few of any supports outside containment will have a safety factor of less than 2 with respect to ultimate capac*ity.

Moreover, the systems outside containment are less critical to safe shutdown than those inside containment:

Obviously there is no potential for a loss-of-coolant accident as the primary coolant pressure boundary is inside containment. Most of the supports outside containment are en component cooling lines, which are not required for safe shutdown. Systems which are required for safe shutdown are redundant. There are two low head safety injection pumps, and the auxiliary feedwater and supply to the high head safety injection systems are cross-connected to Unit 2.

Thus, even if an overstressed condition in a support is identified after interim startup, it is unlikely to affect an essential system, and almost all modifications can be made while the unit is in operation.

Finally, all support evaluation will be 1-3

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION; UNIT l completed within 60 days of startup and the likelihood of an earthquake approaching the DBE in this period is extremely small. Analysis of the pipe and supports foi the OBE conditions will proceed as soon as possible after the analysis for the DBE. Pending completion of this

analysis, VEPCO agrees to voluntarily suspend operations of Surry Unit l if a site ground acceleration exceeding O.Ol g is measured by the onsite accelerometer.

This temporary suspension would continue until all those piping systems, for which the OBE reanalysis are net yet been completed, were inspected, and found to be ~n operable condition. l-4

I., I ,I, I \\I, I* I -' I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 2 SCOPE OF REANALYSIS As described in the NRC Order to Show Cause, March 13, 1979, some piping systems in the Surry Power Station, Unit 1, were dynamically analyzed with a computer program that is not currently acceptable to the NRC. All systems or portions of systems that were analyzed by the SHOCK 2 computer program have been identified Csee Appendix A). These piping systems have been reanalyzed using the NUPIPE computer code~ The results* of the reanalysis are compared with code allowable pipe stresses,, allowable loads for nozzles and penetrations, and with the results of original design loads for pipe supports. 2-1

I I I ,I I \\l I I 1: I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 3 PIPE STRESS RESULTS A total of ?STRESS/SHOCK 2 63 pipe stress problems were originally analyzed by the computer program therefore specifically addressed problems are being analyzed by that used algebraic by the Show Cause two groups: Stone summation and are Order. These stress iiebster Engineering Corporation CS&W) in Boston, Massachusetts, and Nuclear Services Corporation CNSC) in Campbell, California, as indicated in the following table: PIPE STRESS PROBLEMS TOTAL 42 21 63 Field-verified piping isometric drawings provide the basis for program inputs for the pipe stress reanalysis of the SHOCK 2 problems. The reanalysis is concucted using the NU?!?~ computer program. NUPIPE calculates intra-modal seismic forces using a modified square root of the sum of the squares CSRSS) 3-l

I II I ,_-_ I II I I I I I I I I I I

1 I

I I /~- SURRY POWER STATION, UNIT l technique which is always mo~e conservative than the approved SRSS method, and an SRSS technique for inter-medal combination. Additionally, in some cases, piping is analyzed utilizing amplified response spectra CARS) that are developed using soil structure interaction techniques CSSI-ARS). The resultant stresses and loads are used to evaluate piping, supports, nozzles, and penetrations. In accordance with the NRC letter of May 25, 1979 to Virginia Electric and Power tompany CVEPCO), the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a facto~ of l.5 for the DBE condition. All 63 SHOCK 2 problems have been reanalyzed and approved by Engineering Assurance or Quality Assurance. Table 3-1 shows the results for these 63 stress problems. In Table 3-1, the figures for Original Total Stress, at the point cf maximum total stress in the pipe, and Original Seismic Stress, at the same

point, are extracted from Table 4.1 of the "Seismic Design Review Equipment and Piping, Surry Power Station," dated September 15, 1971.

The original calculations for the seismic design review report are no longer available; therefore, correlations were

iade to the

• original stresses in Table 4.1 on the basis of the MSK's.

In !able ~ -, the colu.::ns fer New Total Stress, at the point of ma:dmur:1 total stress in the pipe, and ~ew Seismic stress, at the same point, were taken from 3-2

I I I .I .I I I I I I I I I I I I I I SURRY POWER STATION, UNII l the NUP!PE computer runs vith the seismic inertial stress magnified by a factor of 1.5 for runs using the SS!-ARS, per the NRC letter to VEPCO of ?-!ay 25, 1979. 0£ the 63 problems in Table 3-1, 56 used the SSI-ARS and 7 used the original ARS. The original total and original seismic stresses shewn in Table 3-l were computed using the SHOCK 2 programs fer the original design conditions. !he nev total and new seismic stresses were computed by the NUP!PE program using different mass models and i~ some cases different ARS's than the original calculations. ~ora importantly, the reanalyses were based on field-verified, as-built conditions in 1979, vhich in some cases* differ from the original design conditions. Fer this

reason, the nev stresses and the original stresses in Iable 3-l are not comparable, as they do not necessarily represent the same physical conditions:

!able 3-2 su~marizes the nozzles and penetrations evaluated un~er the reanalysis program. All of t~e a~ nozzles en problems originally analyzed by the S~OCK 2 program have been found acceptable by S&W. Fer these problems in 11cc~ o: 1.5

.e

.;,.,c-*-i _,

---

* '--~-

!'lOZZle loads been the N:tC le4:ter

I I I I I I I I I I I I I I I I I I I SURRY POTJER STATION, UNIT l Table 3-3 lists the hardware modifications necessary to bring the pipe stress analysis to within code allcwables. Of the 63 SHOCK 2 problems reanalyzed, hardware modifications were made to 19 problems due to pipe stress. Th~se modifications consisted of 63

added, modified, or deleted supports.

No physical ~edifications to piping were required. !able 3-4 lists all modifications to reduce nozzle and penetration loads. 3-.'..

---

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

SURRY POHER STATION, UNIT l Prob-lem lJ_Q_._ System .J!fil!l.!L 555 Low Head Safety Injection 1555 Low Head Safety Injection 706A Low Head Safety Injection 707A Low Head Safety Injection 708 Low Head Safety Injection 7."llA Low Head Safety Injection 731B Low llead Safety Injection 743 Lou Head Safety Injection 727 Low Head Safety Injection 735 High Head Safety Injection

• 525A/

Containment & 1525A Recirculation Spray 546/ Containment & 560 Recirculation Spray 5461 Containment & 5600 Recirculation Spray 546/ Containment & 5620 Recirculation Spray TABLE 3-1 PIPE STRESS REEVALUATION SUHMARY Pipe Stress <psi> Isa. fuh_ Line Size CNPSl Original Total* Original New New Seismic* Total Seismic SHOCK2 Problems 122 Dl 10 11 12 11 122 Ll 12 11 122 Hl 6 11 122 Jl 611 122 Kl 611 127 El 811 127 E2 811 127 Fl 1011 127 Cl 611 127 C2 10 11 127 Gl 4", 611 127 G2 811, 10 11 123 Al 811 10 11 123 Dl 811 123 El 10 11 123 F3 811 10 11 123 F2 811 10 11 29290 NA 25290 NA 18451 10707 22436MM l9577MM 20552MM NA 22671MM NA 22671 ** NA 24649MM NA 28909MM NA 23011 13220 11999 10866 28209 24753 28209 24753 28209 24753 1 of 5 10733 5311 10036 3855 15953 1002,. 19229 12556 17489 10941 21890

• 12238 22484 3781 14789 5496 29157 26678

.19296 15708 5978 4986 10299 8402 21279 . 2658 21003 2499 Allow-able 30882 30882 30769 28980 28980 24750 24750 33750 29970 33750 33750 32616 32616 32616

- - - - - - - i-.... - - - -... - - - -.. SURRY POWER STATION, UNIT l TABLE 3-1 <Cont> Prob-Line PiHe Stress <Hsi~ Allow-lem System Isa. Size Original Original New Hew [Q_,__ NamL fuh.._ CHfS) TotalM Seismic* Total Seismic able 548C Containment & 123 H2 10 11 15785~0E ll241MM 5610 2685 29970 Recirculation Spray 547 Containment & 123 Cl 8" 20953 5688 7451 5634 31482 Recirculation Spray 10 11 744/ Containment & 123 Jl 8" 26721** NA 23820 16773 33750 754 Recirculation Spray 548A Containment & 123 Bl 8" 11955 11256 14927 11598 31428 Recirculation Spray 1011 1

• 548B Containment &

123 Hl 10 11 28660 26790 17028 15505 32616 .1 Recirculation Spray 544 Containment & 123 Gl 10 11 13402 6986 9211 6133 28485 Recirculation Spray 123 G2 544A Containment & 123 R2 10 11 Recirculation Spray 12853 11256 6430 3556 29970 544B Containment & 123 Rl 1011 Recirculation Spray 12853 11256 6448 4541 28485 751 Containment & 123 Nl 10 11 6010 5169 7085 5206 28485 Recirculation Spray 123 N2 5621 Containment & 123 Fl 10 11 12610MM HA 19919 6618 30000 546 Recirculation Spray 123 E2 745 Containment & 123 Kl 8" 25702 24579 26294 2382 33750 Recirculation Spray 323A Hain Steam 100 Dl 30 11 13824 6343 12988

• 354 27000 322A Main Steam 101 Dl 30 11 13031 5548 11546 852 27000 334A Hain Steam 102 Dl 30 11 18635 11082 15319 463 27000 346 Hain Steam 103 Al 30 11 19970 12563 23988 7305.

33750 323B Feedwater 100 Gl 1411 15829 590 12899 8061 27000 322B Feedwater 101 Gl 14". 17927 13521 15432 878 27000 2 of 5

SURRY POWER STATION, UNIT l TABLE 3-1 <Cont) Prob-Line fiEe Stress (Esi> Allow-lom System

Isa, Size Original Origi,1al New New ll!L_

Name tl.2.:._ <HPS > Total* Seismic* Total Seismic able 334B Feedwater 102 Gl 1411 16025 12281 17169 10791 27000 '*17 Auxiliary Feedwater 118 Al 311 8568 HA 21+530 ll+036 27000 118 A2 607 Auxiliary Feeduater 118 Gl 411 18681 NA 18126 5347 27000 118 G2 611 636 Pressurizer & Relief Spray 125 Al 411 17671 9527 20304 13833 28800 630 Pressurizer Spray 124 Al 311' 411 205000 NA 29421 24839 33372 & Relief 124 A2 6 II' 12 11 540 Residual Heat 117 Bl 311' 411 14746** 3374MH 19144 9020 25785

• Removal 6"' 12 11 508 Residual Heat 117 Al 10", 12", 16627 12375 13203 8757 25785 Removal 117 A2 1411 465 Service Water 119 Al 24 11 19101 18285 6935 5826 21600 119 A2 119 A3 119 A4

'*BBi Component Cooling 112 C 18" 26723** NA 16263 13680 27000 480 112 Al 5071 Component Cooling 112 Fl 8" 26695** NA 21499 19895 27000 .I 481 112 Bl 18 11 614 Component Cooling 112 AEl 12" 25448** NA 20601 17 57 5 27000 112 AE2 1811 112 AE2 10", 611 512 Component Cop ling 112 ANl l 811 26 792MM . NA 17748 15621 27000 603A Component Cooling 112 Sl 18 11 18710 13461 5396 3554 27000 766 Component Cooling 112 AR 8" 10449** NA 19041 17263 27000 112 T 605A Component Cooling 112 AAl 311' 6 II I 12420 6727 20269 17990 27000 112 AA2 18" 3 of 5

Prob-lem Hfu_ 605B System Name Component Cooling 509A/D Component Cooling 612 Component Cooling 1512 Component Cooling 2529/ Component Cooling 2527 2526 5~7 603B 526C Component Cooling Component Cooling* Component Cooling Component Cooling 527A/C Component Cooling 509B Component Cooling . 509C Component Cooling CVl Containment Vacuum 746 3 11 HP Steam CFl Fire Protection SURRY POWER STATION, UNIT 1 TABLE 3-l (Cont) Iso. li.Q..:_ 112 112 Line Size <NPS) AAl 311, 611 1 AA2 18 11 Original Total* 12420 112 Gl 811, 12 11, 26566** 18 11, 24 11 112 AKl 1811 17958** 112 J 18 11 26792** 112 AH 311, 611 1 26701** 8 11, 10 11, 112 AJ 112 Ml 112 M2 112 M3 112 Sl 112 Ll 112 L2 112 L3 14", 18 11. 2'2" 611, NA 8 11, 10 11.' 411,6 11, 8 11, 10 11, 14", 18 11 18" 6 11 ' 8 11 17958** 18710 HA 112 Tl 411, 611, l0449MM 112 Tl 8 11, 10 11, 1411 112 G2 811, 12 11, 26566** 18", 24 11 112 G3 811 1 12 11 1 26566MM 18", 24 11 137 Al 811 25750MM 131 Al 311 26766MM 131 B2 4 11 131 C3 144 Al 211 1 611, 22255IOE 12" 4 of 5 Pipe Stress (psi) Original New New Seismic* Total Seismic 6727 NA NA NA NA NA NA 13461 NA NA NA NA llA NA NA 14775 20374 14034 21455 18405 24840 14686 5794 22573 11004 16992 9374 19998 16809 16269 13239 2624 22258 22803 . 21266 22705 19319 18862 16007 16039 1824 18674

• 11200 9911 8353 Allow-able 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000 27000

SURRY POHER STATION, UNIT l TABLE 3-1 (Cont) I' rob.,. Lino Pipe Stress Cpsi) .lom Syr; Lam !so. Size Original Original Uew New 1111... __ __II,1mq__ ~ <UPil IotalK Seismic* Total Seismic C:fl I Firu l'rutoction 144 Bl l l/2 11, 211, 16" 10966KM 3733MM 5323 1190 l (llt O I Dini;n.l l111fflor Exhaust 143 Al 24 11 11857KM NA 7340 6756 J,t:(;l;;l![!I H T;1hlo 4-l of soismic design reviev equipment and pipirig Surry Power Station, S11pl:11mbor 1971, or H H :,pbsuq11unt ro.'.lnalys is. l.ug11nd: II/\\ = llot /\\vailabla /\\ I.I. m,.;i h lo S l ro s s = L. 8 Sh llmt Tot:d. Struss = SLP + SDW + 1.5 SDBEI + SDBEA <for SSI/ARS> llu,1 Su imnic ::: L. 5 S l)BlU < for SSI/ ARS) Ihm Tot:aJ. Stress = SLP l SDW + SDBEI + SDBEA <original ARS> llm, Soismi.c :: S ]>BK[ (original ARS> Allow-able 27000 225.00 Tl10 original totaJ.H and original soismicK stresses shown in Table 3-1 were computed using SIIOCK2 for t:hu original design conditions. The new total and new seismic stresses were computod by tho IIUPIPE program using different mass models and, in some cases, different ARS's than tho original calculations. Hore importantly, the reanalyses were based on field-verifiedl a~-built co11ditions in

1979, which, in some cases, differ si,nificantly from the origina dnrd gn conditions.

For this reason, the new stresses and the original stresses in Table 3-1 aro not co1nparablo, as they do not necessarily represent the same physical conditions, 5 of 5

I SURRY POT,lER STATION, UNIT l I

• TABLE 3-2 I

NOZZLE AND PENETRATION

SUMMARY

(Status as of July 27, 1979) SHOCK 2 Problems I No. Accep-Modifi-Vendor I System Total No. table cations or Confirmation and of Nozzles/ Evaluation No. Under Additions Being Prob. No. Penetrations CCom12lete) Evaluation Reouired Obtained I Low Head Safety Injection I 555 l l 0 0 l I 1555 l l 0 0 l 706A N/A 0 N/A N/A N/A I 707A 0 N/A N/A N/A N/A 708 0 N/A N/A N/A N/A I 731A 0 N/A N/A N/A N/A I 73lB 0 N/A tUA N/A N/A 743 0 N/A N/A N/A N/A I 727 2 -') 0 0 2 I High Head Safety Injection I 735 3 3 0 0 3 Containment I Recirculation So!'ay I 525A/1525A 0 N/A N/A N/A N/A 546/560 l l 0 0 0 I 546/5600 0 N/A N/A N/A N/A 546/5620 0 N/A N/A N/A N/A I I l of 4

I . SURRY POWER STA:'ION, UNIT 1 I TABLE 3-2 CCont) I No. Accep-Modifi-Vendor System Total No. table cations or Confirmation I and of Nozzles/ Evaluation No. Under Additions Being Prob. No. Penetratio~s CComolete) Evaluation Reauired Obtained I 548C 1 1 0 0 0 547 0 N/A N/A N/A N/A I 744/754 1 1 0 0 1 548A 0 N/A N/A N/A N/A I 548B 1 1 0 0 0 I 544 2/2 2/2 0/0 0/0 0/0 344A 2 0 0 Confirmed I 544B 2 2 0 0 Confirmed 751 2 2 0 0 0 I 562 1 l 0 0 0 I 745 1 1 0 0 1

1-!ain Steam I

323A l/l l/1 0/0 0/0 0/0 322A l/1 l/1 0/0 0/0 0/0 I 334A l/1 1/1 0/0 0/0 0/0 I 346 0/0 N'/A N/A N/A N/A Feeci'l:ater I .32.33 l/1 l/1 0/0 0/0 0/0 3223 l/1 1/1 0/0 0/0 0/0 I 334:3 l/1 l/l 0/0 0/0 0/0 I Au:-:.

" e e ~*r:2.-: ~ '!"

.:..17 0 N'/A tUA N/A N/A I - o:.... I

I SURRY POWER STATION, UNIT l I TABLE 3-2 (Cont) I No *. Accep-Modifi-Vendor System Total No. table cations or Confirmation I and of Nozzles/ Evaluation No, Under Additions Being Prob. No. Penetrations CCcmolete) Evaluation Reouired Obtained I 60i 3 3 0 0 3 Pressurizer Soray & Relief I 636 l 1 0 0 1 I 630 5 5 0 0 5 Residual Heat I Removal 540 1 l 0 0 l I 508 8 8 0 0 4 Service Wate~ I 465 4 4 0 0 0 I Comoonent Cooling 488/480 4 4 0 0 0 I 507/481 4 4 0 0 0 614 0 N/A N'/A N/A N'/A I 512 0 N/A N/A lf/A N'/A I 603A 1 1 0 0 1 i66 2 2 0 0 0 I 605A 3 3 0 0 2 I 605B 3 3 0 0 2 509A/D 3 3 0 0 0 I 612 2 0 0 1512 0 1:UA ~/A N/A l:i/A I I 3 of 4

I SURRY POWER STATION, UNIT 1 I TABLE 3-2 CCcnt) I No. Accep,.. Modifi-

• Vendor System Total No.

table cations or Confirmation I and of Nozzles/ Evaluation No. Under Additions Being Prob. No. Penetrations CComolete.) Evaluation Reouired Obtained I 252712529 2 2 0 0 2 2526 0 NIA NIA NIA NIA I 527A/C 2 2 0 0 517 4 4 0 0 2 I 603:B 1 1 0 0 1 I 526C 0 NIA N/A N/A N/A 509B 0 NIA NIA l:UA NIA I 509C 0 NIA NIA N/A NIA I Containment Vacuum CVl 0 NIA N/A N/A NIA I 3" HP Steam I 746 1 1 0 0 1 Fire Protect!on I CF-1 0 NIA NIA NIA NIA CF-2 0 NIA NIA N/A NIA I Diesel Muffler Exhaust I 1040 3 3 0 0 0 I N/A = not applicable I I I 4 of 4

~URRY f9W~R ~TA.TlPtt. µttJ+ l Probl,em HKS tio. : l!Q.i.. 508 ' ll 7Al 540 1 ll 7Bl 548A 562 727 735 743 745 746 l23Bl l23E2 127Cl 127Gl 127Fl 123Jl 123Kl 131Al B2 C3 System RHR RHR C&RS C&RS LIISI HHS I LHSI C&RS C&RS

• HP Steam 122Al,Cl CCH TABLE 3-3 PIPE STRESS HARDWARE HODIFICATION

SUMMARY

Overstressed Condition DBET overstressed RHR pump Discharge elbows and overloaded

nozzles, 1, Thermal overstress due to support which was not included in original analysis.

2. Overstress in 211 line which was not included in original analysis.

DBEI overstress condition DBE! overstress DBE! overstress DBE! overstress DBEI overstress DBEI overstress DBEI overstress DBE! overstress DBE! overstress i of 2 Attributable To: Seismic Reanalysis As-Built As-Built As-Built Seismic Reanalysis As-Built As-Built As-Built As-Built As-Built As-built Resolution Six supports added to alleviate overstress condition.

1. Delete NS & RH'S.

2, Add Anchor to 2" line. Addition of a horizontal snubber. Addition of EW, Replace RH with SH, Replace SH with VC. Addition of an LSS. Addition of two guides. Shim existing constraint to VC/LC, Modification and re~ placement of constraints to reflect original analysis. Modification and re-placement of constraints to reflect original analysis. Modification and addition of constraints to reflect original original analysis. Add two guides to Problem 480 and two lateral restraints to .I

-~----------------- SURRY POWER STATION, UNIT 1 Problem MKS No,' No. System 507/4(31 509A/D 112Bl, Fl CCW 517 526C 1 i 527A/C 605A 766 112G 112M 1121 112T 112AA 112AR 2527/.2529 112 tWTE :, ccw ccw ccw ccw ccw ccw ccw TABLE 3-3 (Cont> Overstressed Condition DBEI overstress DBEI overstress DBEI overstress DBEI overstress DBEI overstress DBE! overstress DBE! overstress and thermal overstress DBE! overstress Attributable To: As-built As-built As-built Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built Resolution Problem 488. Add two guides to Problem 481. Added one vertical restraint. Seven restraint changes are required. Added a lateral restraint and modify an SH to a rigid vertical constraint. Delete anchor A36 and - shim exchanger supports. Add four lateral restraints. Add horizontal guides and delete an axial restraint. Same as Problem 527C.

• The use of ARS curves causes higher acceleration levels than the procedure for below-grade structures formerly used as described in the Seismic Design Review revised September 15, 1971.

2 of 2

I I I I I I I I I I I I I I I I I I I Problem No. 746 746 746 746 746 746 322B 323B 334B SURRY POwER STATION, UNIT 1 TABLE 3-4 .HARDWARE MODIFICATION SUMHARY DUE TO NOZZLE AND PENETRATION OVERLOADING Overstress No. l-FW-P-2 inlet nozzle l-FW-P-2 inlet nozzle 1-FW-P-2 inlet nozzle 1-FW-P-2 inlet nozzle 1-FW-P-2 inlet nozzle 1-FW-P-2 inlet nozzle Feedwater pene.tration Feedwater penetration Feedwater penetration Attributable To: As-Built/ thermal As-Built/ thermal As-Built/ thermal As-Built/ thermal As-Built/ thermal As-Built/ thermal Seismic reanalysis Seismic reanalysis Seismic reanalysis 1 of 1 Resol1-1tion Add vertical shock suppressor Add vertical shock suppressor Add spring hanger Add north/south shock suppressor Replace LC by spring hanger Add spring hanger Add guide Add guide Add quide

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 4 PIPE SUPPORT RESULTS Tables 4-lA and 4-lB summarize the pipe supports evaluated in the reanalysis program. There are 1005 supports (487 inside containment, 518 outside containment) on lines originally analyzed by SHOCK 2. All 487 pipe supports within the reactor containment have been reanalyzed, and 66 hardware modifications have been identified. All pipe support modifications within containment will be completed prior to startup. The modifications identified to date <inside containment) due to pipe support analysis are listed in Table 4-2. A support is considered acceptable if all the reaction components are lower in magnitude than the reactions for which the support was originally designed. !f any reaction component is greater than the original design reaction, the support is reanalyzed using the new reactions to determine if it is acceptable. In cases where SS!/ARS was supports are multiplied by 1.5. calculation of support stresses

used, the DB~

seismic inertial reactions on Anchor movements are not considered in the for the DBE con:itions.

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 Pipe support analysis outside the reactor containment will be completed within 60 days after startup.* Any modifications resulting from this analysis shall be processed per the Decision Matrix Csee Section 5). - 2-- -

I SURRY POWER STATION, UNIT 1 I TABLE 4-lA I PIPE SUPPORT STATUS

SUMMARY

INSIDE CONTAINMENT - SHOCK 2 I No. Accep-Modifica-System table tions or I and Total No. of <Evaluation No. Under Additions Prob. No~ SUP'OCrts Comolete) Evaluation Reouired Comment I Low Head Safety Injection I 555 14 14 4 706A 10 10 I 707A 9 9 I 708 22 22 3 1555 9 9 I Containment and Recircu-lation Sora:z I 525A/1525A 35 35 11 I 544 12 12 2 544A 6 6 I 544B 6 6 1 546/560 18 18 7 I 546/5600 11 11 I 546/5620 11 11 2 547 13 13 I 548A 2 546B 22 22 5 I 548C 25 -~ 2 562 19 19 6 I l of 3 --~-

I I I System I and Prob. No. I ttain Steam 322A I 323A I 334A Feedw-ater I 322B 323B I 3343 I Auxiliary Feed*.:ater I 417 Pressurizer I Spray and Relief 630 I 636 I Residual Heat

temaval I,

508 540 I Se!""J'~ i:e :.late:- t..65 I I I SURRY POWER STATION, UNIT l !ABLE 4-lA (Cont) Na. Accep-table ttodifica-tions or Total No. of (Evaluation No. Under Additions Suooorts Ccmolete) Evaluation Reauired Comment 5 17 3 8 14 8 26 27 33 50 10 s 5 17 3 8 14 8 26 27 33 so 10 8 l l 3 1

I I I I I I I I I I I' I* I I,, 1* I I I System and Prob. No. Component Cooling 603A 603B 605A 605B SURRY POWER STATION, UNIT 1 !ABLE 4-lA (Cont) Total No. of SU'C'OOrts 5 5 14 10 No. Accep-table (Evaluation No. Under Ccmolete) Evaluation 5 5 14 10 3 of 3 ttodifica-tions or Additions Reauired 5 2 5

• 3 Comment

I SURRY POWER STATION9 UNIT 1 I TABLE 4-lB I PIPE SUPPORT STATUS

SUMMARY

OUTSIDE CONTAINMENT* - SHOCK 2 I No. Accep-Modifica-System table tions or I and Total No. of <Evaluation No. Under Additions Prob. No. Supports Comolete) Evaluation Reouired Comment I Low Head Safety Injection I 727 5 1 4 731A 4 1 3 1 I 731B 4 1 3 1 I 743 5 2 3 High Head I Safety Injection I 735 30 2 28 Containment and Recircu- ,1 lation Sorav 744/754 10 8 2 I 745 8 8 I 751 0 0 Main Steam I 346 54 33 21 Auxiliary I Feedwater 607 12 9 3 I I I 1 of 3

I I I I I*

• I I

I I

• 1 I

I I I I 1* I I I System and P-rob. No. Component Cooling 488/480 507/481 509A/D 509B 509C 512 517 526C 527A/C 612 614 766 1512 2526 2527/2529 ContaiI".ment Va::'.!um CVl / '-0 SURRY POWER STATION, UNIT l !ABLE 4-lB (Cont)

• No. Accep-table Total No *. of CEvaluition No. Under Supoorts Comolete)

Evaluation 30 27 32 9 13 9 13 17 8 20 24 18 8 20 7l 13 12 9 13 17 8 20 24 18 8 zo 7l 13 Modifica-

• tions or Additions Reauired Comment

I I I I I I. I I I I I I I I System and Prob. No. Fire Protec-tion CFl CF2 Diesel Muffler 1040 NOTES: SURRY POWER STATION, UNIT 1 TABLE 4-lB (Cont) No. Accep-table Total No. of (Evaluation No. Under Suooorts Comolete) Evaluation 17

7.

18 3 15 Modifica-tions or Additions Reouired 3 Comment The numbers in the "Modifications" or "Additions" column are already included in the status shown in the "Acceptable (Complete)" column. The number in the 11Total Supports" column equals the number in the, 11Acceptable" plus the number in the 11Under Evaluation" column. 3 c: 3

---~--~--~----~--~- Problem No. 334B - 465 508 508 508 525A 525A 525A 544 544 544B 548B 548B 548B 548B Support No. PT 89 H-21 H-42A PT 26 H-20 Hl4 PT 65 H70 H34A H37A H46 H87 H72A H72B H84 SURRY POWER STATION, UNIT 1 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION SUMUARY Description of Modification or Overstressed Condition Compar-ison to Ultimate Attributable Capacity To: Snubber modification INSIDE REACTOR CONTAINMENT Note 6 Support function Support member overstressed Snubbers over allowable Snubbers over allowable Weld over allowable Prevents growth in one direction Add shock suppressor due to H-1 overstressed Support overloaded Overstressed with new loads Overstressed with new loads Weld over stress allowable No lateral restraint capability 6~way restraint not properly functioning 6-way restraint not properly functioning No lateral movement due to thermal and seismic Note 5 As-built Not~ 7 -Seismic Reanalysis Note 7 Seismic Reanalysis

• Note 9 Note 9 Note 3 NA Note 6 Note 6 Note 3 Note 6 Note 9 Note 9 Note 6 1 of 6 As-built As-built Seismic Reanalysis As-built As-built As-built As-built As-built As-built As-built As-b.uilt Resolution Add horizontal shock suppressor.

Removed Support H-21, Add horizontal snubber. Add horizontal and vertical snubbers. Modify support. Modify support. Add horizontal shock suppressor. Replace spring hanger. Redesigned and modified to VC-LC. Redesigned and modified to VC-LC. Support modified. Redesigned and modified support to receive lateral load. Redesigned (modified to 3-way restraint>. Redesigned (modified to 3-way restraint>. Restore to original configuration.

- - -.. - -', -* -i _\\.. --- - - - Problem No. 548B 548C 548C 555 555 5:;5 555 5461560 5461560 5461560 5461560 5461560 5461560 5461560 562 Support No. Hl8A Hl8B Hl7A Hl7B Hl7C H75 HlO H28 H29 H30 H39A PT-185 PT-208 PT-130 PT-216 PT-310 H50 H26 SURRY POWER STATION, UNIT 1 TABLE 4-2 (Cont) Description of Modification or Overstressed Condition U-bolts overstressed Lugs overstressed on gang hanger - anchor U-bolt subjected to lateral loads over allowable loads Integral attachment overstressed Horizontal and vertical structural member restraining pipe movement Vertical member restraining pipe motion Verticai member restraining pipe motion Integral attachment over-stressed for support H39A Added due to H39A Added due to H39A Added due to H39A Added due to H39A Added due to H39A Integral attachment over-stressed for support HSO Modified due to H39B. Inte-gral attachment overstressed for H39B. Compar'"'. ison to Ultimate Attributable Capacity To: Resolution Note 9 As-built Redesigned and modified support by adding bracing member. Note 8 As-built Redesigned and modified by removing attachments to relieve pipe overstress conditions. Note 9 As-built Replace U-bolt with framed member. Note 1 Seismic Add lateral shock Reanalysis suppressor. Note 6 Thermal Restore to original intent. movement Note 6 Thermal Restore to original intent. movement Note 6 Thermal Restore to original intent. movement Note 8 As-built Replace anchor by a 3-way restraint. NA As-built Add vertical shock suppressor to reduce loads. NA As-built Add lateral constraint. NA As-built Add lateral constraint and vertical constraint. NA As-buiit Add lateral constraint, NA As-built Add lateral constraint. Note 9 As-built Add North South East West restraint. NA As-built Modify support by adding U-bolt. 2 of 6

Problem No. Support No. 562 H29A H29B 562 PT-265 562 PT-340 562 H-39B 562 PT-183 546/5600 Hl5B 546/5600 PT-155 546/5620 H-15A 546/5620 PT-230 636 H36 708 PT-1 708 Hl27 708 PT-380 SURRY POWER STATION, UNIT l TABLE 4-2 (Cont) Description of Modification or Overstressed Condition Modified due to H39B. Inte-gral attachment overstressed for H39B. Added due to H39B. Integral attachment overstressed for H39B. Modified due to H39B. Inte-gral attachment overstressed for H39B. Integral attachment over-stressed Added due to H39B. To reduce load on hanger H39B. U-bolt modified to reduce loads on Hl7B Added to reduce anchor loads U-bolt modified to reduce loads on Hl7A Added to reduce anchor loads Plate overstressed Lateral load on U-bolt overstressed Member overstressed Lateral restraint. over-stressed Compar-ison to Ultimate Attributable Capacity To: NA NA NA Note 8 NA NA NA NA NA Note 3 Note 2 Note 3 Note 3 3 of 6 As-built As-built As-built As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built Resolution Modify 2-way support to 3-way support. Add lateral constraint. Reduce loads at anchor point. Modify vertical constraint. Replace anchor by a 3-way restraint. Add 2-way support. Redesigned and modified ganghanger from 1-way rLC>. Add LC-VC to reduce loads at anchor. Redesigned and modified ganghanger from 1-way CLC> to 2-way CLC-VC). Add LC-VC to reduce loads. Modify anchor by adding gusset. Add lateral restraint. Modify support to vertical and lateral constraint. Modify support to vertical and lateral constraint.

---

~~~~~---~-

SURRY POWER STATION, UNIT 1 Problem No. 1525A 1525A 1525A 1525A 1525A 1525A 1525A 1525A 603A 603A 603B 603B 605B Support No. PT-1000 PT-5000 PT-8000 PT-l.34 PT-232 PT-314 PT-2510 PT-6300 3, 4, 7 5 13 10 4 TABLE 4-2 (Cont) Description of Modification or Overstressed Condition Integral attachment CH-2) overstressed Delete due to H-2 U-bolt and frame over-stressed Added due to HlOA Added due to H2 Added due to HlOA Add due to H2 Hanger overloaded Pipe straps overstressed Vertical restraint only re-quired, but support had zero lateral clearance. I Uplift restraint is re-quired, but support had negative vertical restraint only. Vertical support only re-quired, but support is a vertical restraint and lateral restraint. Support members overstressed Compar-ison to Ultimate Attributable Capacity To: Note 9 NA Note 1 NA NA NA NA NA Note 3 Note 6 Note 6 Note 6 Note 9 4 of 6 Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built. As-built As-built Resolution Modify anchor to VC to reduce anchor load. Delete axial support and 1-way LC to reduce anchor load. Modify frame. Add LC to reduce anchor loads. Add LC to reduce loads. Add LC to reduce anchor loads. Add LC to reduce loads~ Upgrade existing support H71. Add gussets to all and increase weld size. Modify support to provide lateral clearance. Modify support to provide uplift restraint, Remove top and side members of support. Replace two W4xl3 members with TS4x4xl/2.

_/ ____ _ r SURRY POWER STATION, UNIT 1 TABLE 4-2 (Cont) Compar-Description of Problem Support Hodification or ison to Ultimate Attributable No. No. Overstressed Condition Ca2acity To: Resolution 605B 6 Pipe strap weld stress over Note 9 As-built Add two gussets to reduce load allowable at connection. 603A H-2 Lateral clearance insufficient NA Thermal Modify restraint. movement 605B H-2 Lateral clearance insufficient NA Thermal Modify restraint. movement 605A PT-13 Need vertical restraint Note 6 As-built Modify vertical support. 605A PT-23 Need vertical and lateral restraint Note 6 As-built Modify vertical support. 605A PT-28 Need vertical and lateral restraint Note 6 As-built l'iodify vertical support. 605A PT-4295 Additional support NA As-built Add new support. 605A PT-4300 Additional support NA As-built Add new support. GENERAL NOTES: NA= Not applicable COMPARISON TO ULTIMATE CAPACITY NOTES:

1.

Support load is less than 50 percent of ultimate capacity based on linear elastic analysis.

2.

Combined lateral and vertical load on U-bolt is less than 50 percent of ultimate capacity based on ultimate load analysis.

3.

Support load is less than 50 percent of ultimate capacity based on ultimate load analysis.

4.

Snubber load is less than 50 percent of ultimate capacity based on ultimate load analysis.

5.

Problem is acceptable with support removed entirely,

6.

Support modified to restore support to originally intended function. Support not overstressed. Comparison to ultimate capacity not applicable. 5 of 6

\\. ---*---------~------ SURRY POWER STATION, UNIT 1 TABLE 4-2 (Cont)

7.

Load may be greater than 50 percent of ultimate capacity based on preliminary simplified analysis and conservative loadings as of July 31, 1979. More realistic analysis is continuing, and further consideration may indicate loads less than 50 percent of ultimate capacity. 8, Ultimate capacity not computed for as-built conditions, which differ from original design conditions. As of July 31, 1979, preliminary calculations do not confirm that the load is less than 50 percent of ultimate capacity. Analysis continues which will affect the conclusions *

9.

Calculations not available as of July 31, 1979. Lengthy hand calculations are continuing. 6 of 6

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION UNIT 1 SECTION 5 DECISION MATRIX 5.1 DECISION MATRIX FOR SIGNIFICANT.ITEMS IDE~TIFIED AFTER INTERIM STARTUP The following is the procedure that will be used in the continuing reanalysis program to determine the significance of, and to define the reporting requirements for, possible equipment or support modifications identified after start of interim plant operation. This procedure addresses: Equipment nozzles for which loads exceed listed values. Pipe hangers or supports for vhich loads exceed code allowable capacities. 5.1.1 'Equipment Nozzles and Penetrations !f calculated loads exceed original loads after an analyst has reviewed all model parameters Cfor

example, the translating of information on piping isometric drawings into a computer program), the supervisor will review all computer input, program controls, data sources, and all output printouts; this review will include an overview of all the input/output work and a review to determine whether a

more detailed mathematical model will reduce the conservatism, or whether any of the input assumptions used to simplify problem solutions should be revised to create a longer, but more correct, calculation. 5-1 ~'

I I I I I I I I I I I I I 1* I I I I I SURRY POWER STATION UNIT l If the supervisor's review does not result in a reduction of calculated loads to listed values, he will contact a system engineer familiar with the actual operating modes and conditions to determine if any of the input conservatisms can be reduced, based on the way operating conditions are actually combined as opposed to the way they are normally combined to produce worst possible case inputs to the problem and computer runs. !his may require multiple computer

reruns, but produces less overconservatism than combining inputs for a worst case run.

.1.- the load is still in excess of listed values after the above is completed, a system engineer and an experienced stress engineer will immediately review the problem together to determine if the answer is comparable to what past experience would indicate is to be expected for this system arrangement. If it appears unreasonable based on experience, they will immediately review with the supervisor all the conditions or input parameters that contribute to the largest component load. !hey will continue their review for all inputs until they deter~ine that an incorrect input exists, or conclude that the load is correctly calculated and is over listed values. If so, the engineers wiil perform an expeditious review of the nozzle capacity to determine whether total nozzle actual capability is exceeded. If total nozzle actual capacity is exceeded, the prime causes of the overload condition will be deter~ined and the S&w stress review project engineer will be notified. The S&~ stress review project engineer will notify the Virginia Electric and Power Co~pa~y CVE?CO) Station ttanager =>-~ overcapacity lead has been

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION UNIT 1 calculated, and will define the location in the pipe system and possible corrections. He will then initiate a study of the overall impact of this stress level and provide additional information as it develops to the Station Manager. The reporting process to the

NRC, including determination of system nonoperability status, has been previously established and will compliance with the Surry Technical Specification Section T.s. 6.6.

be in 5.1.2 Pipe Hangers and Supports The following flow diagram represents typical steps used in making decisions in the evaluation of pipe hangers.

1.

Calculate actual allowable hanger capacity from sketches '1t Compare load from pipe stress run with actual allowable hanger capacity~ ~ Hanger overloaded Hanger OK t i Would SSI-ARS reduce Complete lea/ Yes ~ ~No

3.

Initiate SSI computer run. Go to 2. Is new load of ultimate greater than 50 percent capaci\\

4.

Yes Stress engineer review

5.

~ If pipe stress appears OK with support inef:ec-tive, go to 5. Pipe requires active support to maintain allowable stress levels. t 5-3 No List hangers to be reworked during steam generator repair outage.

I I I I I I I I I I I I I I I I I I I

6.

9.

SURRY POWER STATION UNIT l / \\ S&W Stress Review Project Engineer will inform S~ition Manager that support exceeds "Notification of NRC Stress Criteria" and is in final review to determine any other mitigating conditions.

i.

s.

S&W Engineers Review.

a.

Determine what por-tion of the load is due to self-limit-ing forces, e.g., anchor displace-ments which have only a finite tra-vel possibility, dead load which would be transfer-red to adjacent supports after dis-placement of this hanger, and reduce total by this amount for tempor-ary use.

b.

c.

Review system physical equipment arrangement to determine if ~tresa would increase in equipmen~ or valves or only in piping if hanger was ineffective. Review loads and hanger design to determine if model-ing and computer analysis CSTRUDL) will yield more accurate and lower stress. Advise Station Manager of hanger evaluation results and information to and in determining if shutdown is or is not required. Station Manager determines impact on system operability and technical specifications and notifies NRC, !&E, Region !I within 24 hours in accordance with technical specification procedures. 5-4

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 6 BRANCH LINE SUIB'iARY 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 b~ evaluated separately in each of the three translational coordinates and nust be sufficient to prevent overstresses in the branch due to thermal and seismic displacements imposed on the branch pipe at the connection to the run pipe. This procedure is intended to provide a secondary stress check based on new run pipe displacements resulting from the current Task Force effort. !he evaluation cf branch lines utilizes sinplified beara formula and USAS B31.l material allowable stress values to establish the existence of sufficient ~ 1

• '
• 1.._

=-eXl.Ol.-l... y, Based upon the computer analysis results of the stress evaluation which ir:.cluc:es the ~un

' \\J -** -

dis p !. a:: er:er:.ts

pipe, the resulting to fc:::::i

~-' c!isplacenents the centerline !1:ese a::.

a!" - ea.ct

I I I I I I I I I I I I I I I I I I I ~~-~ SURRY POWER STATION, UNIT 1 Secondary stresses in the branch pipe which exist independent of the run pipe displacements are represented by computing the unrestrained thermal growth at the run/branch connection and superimposing those with the run pipe displacements. !he minimum equivalent length of branch pipe is determined by utilizing the cantilever beam analogy, based upon geometric properties of the branch pipe. Input displacement (deflection) is the sum of intersection displacements. Allowable stress to be used is determined by dividing the B31.l code allowable S~ by the appropriate stress intensification factor. The minimum equivalent required length calculated must be less than the actual effective length existing in the as-built condition. To determine the actual effective length, it is necessary to consider the additional flexibility of elbows as well as the flexibility of straight piping l~ngths. For purposes of this evaluation, the actual effective length may be considered to be the actual length of straight pipe segments which provide bending flexibility Cthose which are perpendicular to the direction of displacement) Plus the effective length of bends and elbows which can be approximated as the factor k times the length of the curve, measured on the centerline. The flexibility factor k may be obtained from Appendix D of USAS B3l.l. 6-2

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 The evaluation of branch line flexibility is accomplished by preparing a form and incorporating it into the pipe stress problem package. 6-3

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 7 FIELD VERIFICATION

SUMMARY

Io ensure that the pipe stress reanalysis is performed as accurately as possible) field verification of as-built conditions has been performed. The field verification produced detailed piping isometric drawings upon which reanalysis is based. All field-verified piping isometrics are independently verified by Surry Power Station quality control per~onnel. Pipe support as-built drawings are prepared in similar fashion when complete* design details or support sketches are not available. Approximately 60 percent of all seismic pipe supports within the scope of the reanalysis effort have had as-builts prepared to data. These as-built drawings are used in pipe support analysis. In all cases, the as-built drawing is* the second support inspection since initially the support is inspected for de$ign type functions, locations, orientation, clearance, completeness of installation, and physical damage. 7-1

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 SECTION 8 APPLICATION OF BUILDING DISPLACEMENTS AND ANCHOR MOVEMENT STRESSES The maximum building displacements and accelerations are computed from the time history response of the dynamic model of the building structure, using the computer model.FRIDAY. These results are based on the FSAR earthquake spectrum and the average strain-compatible free-field soil properties from the final iteration of the SHAKE computer program. The building accelerations are used to generate the Amplified Response Spectra CARS) used in the piping dynamic analysis; the building displacements are used to compute the effects of anchor movements on the pipe. !he NRG has required that dynamic DBE pipe stresses computed using the Soil-Structure Interaction Amplified

Response

Spectra CSSI-ARS) be increased SO percent to account for a soil shear modulus of elasticity 50 percent larger than the average value from the final iteration of SHAKE. A 50 percent increase in the soil shear modulus vould cause a significant decrease in the building displacements. Fer consistency, then, the piping stresses due to anchor movements should be decreased in correspondence with the increase in dynamic ~nert~al stresses due to the SSI-A~S. !n actuality, in the pipe a-1

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT l stress reanalysis, the anchor movements are based upon the average strain-compatible shear modulus, and no decrease or pipe stresses is made,, al though the increase inertial of the stresses corresponding are increased 50 percent. !his approach results in very conservative values for seismic pipe stresses. a-:

I I I I I I I I I I I I I I I I I I I hl284622-lh 06/05/79 SURRY POWER STATION, UNIT 1 APPENDIX A SYSTEMS AFFECTED 042 __ ____J

I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT 1 APPENDIX A The reanalysis included those safety-related lines originally computer-analyzed with the SHOCK 2 program. The systems line numbers, the associated computer problem numbers, and the flow diagram numbers are listed below. The flow diagram numbers refer to the Surry, Unit 1, FM and FB drawings included in Appendix B. System Pressurizer Spray & Relief Low Head Safety Injection Line No. 4"-RC-15-1502 4"-RC-14-1502 12"-RC-36-602 611-RC-39-1502 6"-RC-42-602 6"-RC-38-1502 6"-RC-41-602 6"-RC-37-1502 6"-RC-40-602 4"-RC-34-1502 3"-RC-35-1502 6"-RC-20-602 3"-RC-61-1502 6"-RC-62-602 12"-RC-23-1502 12"-SI-46-1502 10"-RH-16-1502 12"-RC-22-1502 12"-SI-45-1502 6 "-RC-.16-1502 6"-SI-49-1502 6"-RC-21-1502 6"-SI-50-1502 6"-RC-18-1502 6"-SI-48-1502 6"-SI-143-1502 6"-SI-49-1502 611-SI-50-1502 8"-SI-92-153 a:'-SI-14-153 A-1 Problem No. 636 630 555 1555 i06A 707A 708 73!A,B Flow Diagram No. FM-103 FM-103 FM-103 & 106 FM-103 & 106 FM-103 & 106 FM-103 & 106

M-103 & 106 F~-106

I I SURRY POWER STATION,

• UNI! l FloY I

Problem Diagram System Line No. No. No. I 10"-SI"."'6-153 743 FM-106 6"-SI-48-1502 727 FM-106 6"-SI-49-1502 I 611-SI-153-1502 10"-SI-148-153 10"-SI-149-153 I 10"-SI-150-153 10"-SI-151-153 10"-SI-16-153 10"-SI-13-153 I 10"-SI-84-152 10"-SI-152-1502 811-SI-14-153 I 18 11-SI-83-152 8"-SI-92-153 I High Head Safety 811-SI-7-152 735 FM-105 & 106 Injection 8"-SI-102-152 10"-SI-6-153 8"-SI-17-152 I 6"-SI-18-152 611-SI-19-152 6"-SI-78-152 I 6"-CH-201-152 6"-CH-202-152 6"-CH-203-152 8"-CH-204-152 I 8"-CH-206-152 8"-CH-17-152 6"-CH-72-152 I 6"-CH-18-152 6"-CH-19-152 4"-CH-112-152 I

• 8"-S!-92-153 8"-SI-14-153 8"-SI-170-153 8"-SI-171-153 II 8"-SI-170-152 Contain."llent and 811-CS-23-153 525A/1525A FM-101 I

~ecirculation 8"-CS-22-153 Spray 10"-CS-3-153 10"-CS-4-153 I .10"-CS-4-153 547 FM-101 011-cs-J.:..-153 I A-2 I

I I SURRY POWER STATION., UNIT l Flov I Problem Diagram System Line No. No. No. I 8"-RS-21-153 546/560 FM-101 10"-RS-4-153 8"-CS-34-153 744/754 FM-101 I 4"-CS-36-153 10"-CS-3-153 548A,B FM-101 I 8"-CS-33-153 10"-RS-3-153 10"-RS-10-153 544, 544A,B

• FM-101 I

10"-RS-9-153 10"-RS-1-153 10"-RS-2-153 I 10"-RS-10-153 751 FM-101 10"-RS-9-153 li."-RS-14-153 I 4"-RS-15-153 10"-RS-12-153 562 FM-101 t 8"-RS-23-153 8"-CS-33-153 745 FM-101 I 10"-RS-3-153 546/5600 FM-101 an-Rs-20-153 4"-CS-35-15 I 10"-RS-11-153 546/5620 FM-101 8"-RS-22-153 I, 10"-RS-11-153 548C FM-101 I Residual li."-RC-197-153 540 FM-103 & lOli. Heat 4"-RH-15-152 Re:::oval 3"-CH-66-153 3"-CH-65-152 I 14"-RH-1-1502 508 FM-103 & lOli. 14!'-R:i-lS-602 I l~"-RH-Z-601 lO"-R?!-.:..-50: 10"-R~-S-602 12"-R:i-6-602 I 2"-~:!-1:-602 C"-~~-5-602 0"-RH-9-602 -1

\\- ~

'I

I

• 1 I

I I

I t

• I II I

I I I I I I System Component Cooling Water SURRY POWER Line No. 12"-RH-19-602 10"-RH-7-602 10"-RH-10-602 18 11-CC-227-121 18"-CC-228-121 18"-CC-229-121 18"-CC-230-121 18"-CC-5-121 18"-CC-220-121 18"-CC-6-121 18"-CC-224-121 18"-CC-237-121 18"-CC-236-121 18"-CC-235-121 18 11-CC-225-121 18 11-CC-232-121 18"-CC-233-121 18 11-CC-234-121 18"-CC-231-121 18"-CC-7-121 6"-CC-177-151 6"-CC-169-151 12 11-CC-24-121 10 11-CC-167-121 24 11-CC-226-121 12 11-CC-27-121 18 11-CC-16-121 18"-CC-15-121 18"-CC-16-121 1a 11-cc-s-121 8"-CC-69-151 811-CC-67-151 8"-CC-67-121 10"-CC-!.Z.3-121

O"-CC-6i-l::

on-cc-~73-151 S!AIION, UNIT l Flow Problem Diagram !fo. No. 488/480 FM-22 507/481 FM-22 614 FM-22 1512 FM-22 512 F!i-22 603A,3 FM-22 766 FM-22

,.-4

I I SURRY POYER STATION, UNIT 1 Flow

• 1 Problem Diagram System Line No.

No. No. I 8"-CC-61-151 527A/C FM-22 4"-CC-61-151 6"-CC-62-151 8"-CC-75-151 6"-CC-81-121 10"-CC-81-121 14"-CC-67-121

1 18"-CC-10-121 18"-CC-10-121 605A,B FM-22 I

611-CC-105-151 3"-CC-105-151 3"-CC-107-151 3"-CC-112-151 t 6"-CC-207-151 18"-CC-17-121 4"-CC-108-151 II 4"-CC-113-151 18"-CC-8-121 509B,C,A/D FM-22 18"-CC-237-121 II 18"-CC-236-121 18"-CC-235-121 8"-CC-311-151 1: 8"-CC-32-151 6"-CC-32-151 4"-CC-32-151 a 6 "-CC-28.6-15 l 24"-CC-235-121 18"-CC-226-121 18"-CC-10-121 I 12"-CC-27-121 24 11-CC-226-121 18"-CC-9-121 10"-CC-89-151 611-CC-89-151 6"-CC-97-151 I\\ 18"-CC-14-121 612 FM-22 8"-CC-148-151 ~I 14"-CC-72-121 10"-CC-72-121 4"-CC-148-151 6"-CC-146-151 I 14"-CC-70-121 517 FM-22 10"-CC-70-151 4"-CC-66-151 I A-5 I ~-- ---~~

I., SURRY POWER S!ATION, UNI! l Flow I Problem Diagram

• System Line.No.

____NQ., ____N_Q. I 8"-CC-66-151 611-CC-64-151 l 0"-CC-72-121 I 18"-CC-17-151 811-CC-70-151 8"-CC-71-151 1: 8"-CC-78-151 526C FM-22 18"-CC-17-121 611-CC-85-151 I 611-CC-93-151 611-cc-101-151 10"-CC-101-121 2 l/2"-CC-104-151 I 4"-CC-144-151 2527/2529 rn:-22 6"-CC-145-151 l 1 8"-CC-144-151 10"-CC-143-151 lli. 11-CC-143-121 I 18"-CC-7-121 10"-CC-161-121 6"-CC-161-151 I 8"-CC-153-151 611-CC-181-151 2526 FM-22 I. 2 l/2"-CC-184-151 6"-CC-173-151 10"-CC-181-151 6"-CC-165-151 I 811-CC-156-151 18"-CC-14-121 t II Zli."-WS-26-10 li.65 FM-21A 2 411-1-iS-2 8-10 2 4"-\\-iS-30-10 2 4"-\\-.'S-32-10 I Hain S tea.rn 30"-SHP-1-601 323A

'M-14.A 30"-SHP-2-601

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I SURRY POWER STATION, UNIT l I I I I I I APPENDIX C CORRESPONDENCE WITH NRC I I I I I I I I I I I I

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION, UNIT l .APPENDIX C CORRESPONDENCE WITH NRC The following is a listing of correspondence with the NRC related to the reanalysis effort. Signature 3/13/79 Denton 4/2/79 Stello 4/13/79 Stello 5/18/79 Stello 5/25/79 Eisenhut 3/30/79 Spencer 4/19/79 Stallings 4/23/79 Spencer 4/24/79 Spencer l./27/i9 Spencer 5/2/79 Stallings S/2/79 Spencer 5/22/79 Ragone 5/24/79 Spencer Addressee NRC to VEPCO Proffitt Proffitt Proffitt Proffitt Proffitt Denton/ Stello O'Reilly O'Reilly O'Reilly Denten/ Stell.a Denton Stello Hendrie - Stello C-1 Letter No./Subject Show Cause Order Addendum to Show Cause Order Use of Soil Structure Interaction Techniques Request for Furth~r SSI Information Factor Adjustment to SSI Calculated Stresses 198/Initial Response to Show Cause Order Z70/LER 79-0l0/03L-0 289/Response to I.E. Bulletin No. 79-07 288/Response to I.E. Bulletirt No. 79-07 3ll/Transmittal of Ivo Sample Problems to EG&G Observations on Reanalysis Effort 260/Submittal of SSI In-formation Comments on Moratorium/Surry Reanalysis Response to NRC Letter cf 4/2/79

I I I I I I I I I I I I I I I I I I I 6/12/79 8/1/79 SURRY PO~ER STATION, UNIT 1 V.EPCO to NRC Addressee Letter No./Subject Stello pencer Response to NRC Letter of Denton Denton 5/18/79 Submittal of Report on Reanalysis Additional Information, Report on Reanalysis of Piping Systems Denton pencer Modification Information, Reanalysis of Piping Systems Denton pencer Schedule and Support Informa-tion Denton Support Modifications Denton pencer Support Information, Re-analysis of Piping Systems Denton Spencer Submittal of Revised Report on Reanalysis

I I I I I I I I I I I I I I I I I I Date 3/22/79 3/30/i9 4/3/79 l./6/79 4/6/79 4/ll/79 4/13/79 .:../18/79 4/27/79 4/27/79 5/8/79 6/4/79 6/12/79 Signature Kennedy Jacobs Jacobs Kennedy Jacobs Jacobs Jacobs Jacobs Jacobs Jacobs Rossier Jacobs Jacobs SURRY POWER STATION, S&'li-1 to NRC Addrec:see Denton Herring Bezler Denton Stello Bezler Stello Hartman Bezler Stello Neighbors 3ezler 3ezler

-3 UNIT 1 Letter No./Sub iect Transmittal of S&W Comput.er Programs Submittal of Computer Outputs Submittal of Benchmark Problem to Brookhaven National Laboratory Transmittal of S&W Computer Programs Plan for Verification of Dynamic Analysis Codes Submittal of Computer Outputs Update and Status of Veri-fication Plan for Dynamic Analysis Codes Submittal of Computer Outputs Submittal of Benchmark Problems Status of Verification Plan

or Dynamic Analysis Codes Draft Outline of SSI-ARS Report Submittal of Benchmark

?rcblems Submittal of Benchoark Probler:is}}