ML18114A655

Rept on Reanalysis of Safety-Related Piping Sys,Unit 1.

ML18114A655
Person / Time
Site:	Surry
Issue date:	06/05/1979
From:	VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:
Shared Package
ML18114A653	List: ML18114A652 ML18114A655
References
H1284622-1A, NUDOCS 7906070298
Download: ML18114A655 (162)

Text

REPORT 0~ THE REANALYSIS or SAFETY-RELATED PIPING SYSTEMS NOTICE THE ATTACHED FILES ARE OFFICIAL RECORDS OF THE DIVISION OF DOCUMENT CONTROL. THEY HAVE BEEN CHARGED TO YOU FOR A LIMITED TIME PERIOD AND MUST BE RETURN.ED TO THE RECORDS FACILITY BRANCH 016. PLEASE DO NOT SEND DOCUMENTS CHARGED OUT THROUGH THE MAIL. REMOVAL OF ANY PAGE[S) FROM DOCUMENT. FOR REPRODUCTION Ml!_Sf ~

BE REFERRED TO FILE PERSONNEL. \ \\\la~

.. . tl\\\\t\.. .

OEADLINERE\~~,~i'\ ~*

0

"~~\\ \~ Docfmtfl:5J- ~W

t. ~ \ n\\) ico~~./ if *1q o to .o* lD ill~

\du\~*n

,~rt,' ~ n ~ \ \ c .., n~r*1ment:

1.1~*~ 4J "S 1) L- * , ::.*1'~

• R,~it ~~. . I !,..,/,.,ii.. ....... ll .. ' ,.;;:~ .

RECORDS FACILITY BRANCH

h1284622-la 06/05179 . 44 I SURP.Y PO'.-l'ER ST A!IOH 1 UNIT 1

',~,

I r

REPORT ON THE 1.12 -

~I REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS 1 ;14

• 1 SURRY POWER STATION, UNIT 1 VIRGINIA ELECTRIC AND PO~ER COMPANY 1.18

'_I I

RETURN TO REACTOR DOCKET

.I fllES I .JUNE 5 , 19 7 9 1. 20 I _I 1

_:I Stone & webster Engineer~ng Corporation

• -1 I Boston, Massachus~t~s

I

.1 Vepco VIRGINIA ELECTRIC AND POWER COMPANY, RICHMOND, VIRGINIA 23261 I

June 5, 1979 Mr. Harold R. Denton, Director Serial No. 453 Office of Nuclear Reactor Regulation PSE&C/CMRjr:mc U.S. Nuclear Regulatory Commission I Washington, DC 20555* Docket Nos.: 50-280 50-281 License Nos.: DPR-32 I DPR-37

Dear Mr. Denton:

I REPORT ON THE REANALYSIS OF SAFETY RELATED PIPING SYSTEMS.

SURRY POWER STATION UNIT 1 The Nuclear Regulatory Commission Order to Show Cause of I March 13, 1979 required that certain piping systems associated with Surry Power Station Units 1 and 2 be reanalyzed using an appropriate piping code to account for seismic loads. We complied with the Order requiring shut-I down of the Units within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Since that time, an intense effort has been under way to analyze all affected piping systems in a manner acceptable to the NRG staff and I commensurate with our commitment to provide a safe and reliable source of power for our customers. We have had the benefit of numerous discussions with the NRG staff to clarify and amplify their specific concerns with I regard to the details of our reanalysis effort. We have been and are totally committed to provide the staff, on an expedited basis, with any information they require for their.review of the Surry units.

I We believe the culmination of the pipe stress analysis effort is at hand. The analysis to date, while continuing, has shown that the piping systems are impacted only slightly even after a thoroughly rigorous

• I* reanalysis. It has been unequivocally demonstrated that the impact on the piping systems is wholly incompatible with the severity of the Commission's Order. It is on this basis that we submit the attached Report and request I immediate start up of Surry Power Station Unit 1.

Correspondence with the staff has transmitted a vast amount of information between the parties. A compilation of the transmitted I information is tabulated in the attached Report in Appendix G for your convenience and reference.

I I

I I VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton SHEET NO. 2 I We would not feel justified in requesting an immediate lifting of the Order if the reanalysis had shown demonstrable and persistent modifications to piping systems. Such has not been the case. For example, I of the approximately 74 piping problems to be reanalyzed, 29 have been completed as of June 2, 1979. Results show that no piping of any size will have to be replaced or repaired. Of the approximately 873 total supports

• 1 to be reanalyzed, 138 analyses have been completed. None of those supports will require modification. In a cursory look at the balance of the supports to be completely reanalyzed, we have so. far identified four supports which will require some modification. These modifications include I addition of one snubber, shimming of one support and lateral braces for two supports. These modifications are not only minor, they do not even occur.

because of seismic stress conditions. Modifications are discussed in some I detail in Section 5 of the Report. On the basis of these analyses and conservatisms contained in our analysis techniques as explained in our attached Report, we believe we have substantial justification for start up of Surry Unit 1.

I As we continue our reanalysis effort, it is possible that other

• 1 potential support modifications may surface. We will evaluate each of these potential modifications on a case by case basis in accordance with the guidelines delineated in Section 5 of the attached Report. We have several methods available to evaluate the necessity of a potential modifi-I cation. For those modifications which we deem to be major in nature, we will contact you and solicit your involvement. Such modifications, once identified, will be expedited.

I Modifications for the design basis earthquake (DBE) case which are considered to be less than major in nature in accordance with the guidelines in Section 5 of the Report will be made at advantageous* times I in the operating schedule of the unit.

The following two paragraphs specifically address the two items of your May 25, 1979 letter.

I_ Your letter of May 25 requested information regarding operating basis earthquake (OBE) design requirements. For those supports which meet I DBE requirements but do not meet the FSAR QBE design requirements, we have not as yet identified a requirement to reduce the present FSAR QBE design value. We will evaluate the QBE requirement as stipulated in Item 2 of

• I your May 25 letter on a continuing basis for those piping systems which meet DBE design requirements but do not meet OBE design requirements. The basis for evaluation will be amplified response spectra (ARS) compatibility between the DBE and QBE cases. That is, if soil structure interaction is I used in a piping system evaluation for the DBE case, it will also be used in the OBE case.

I I

I

I I VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton SHEET NO. 3 Your May 25 letter also requested information on the capability I of piping systems to safely withstand all earthquakes up to and including the DBE. An investigation of the effects of earthquakes smaller than the DBE leads to the conclusion that the effects of the DBE are not exceeded by smaller earthquakes. This investigation will be covered in Section 7 of a detailed report on SSI-ARS to be submitted on or before June 8, 1979.

Capability of piping systems can also be addressed in terms of the numerous

• 1 conservatisms involved in the overall analysis. These are addressed in detail in the attached Report in Section 7.

Enclosure three of your April 2 letter addressed verification of I certain computer codes, including the NUPIPE code being used on Surry Units 1 and 2, with standard benchmark problems developed by the staff and Brookhaven National Laboratory. These have all been previously forwarded I except for one benchmark problem involving the analysis of a two loop NSSS, the results of which will be forwarded to the staff on or before June 8, 1979 by Stone & Webster Engineering Corporation.

I Prolonged discussions have been held with the staff regarding the methodology and use of soil structure interaction in the development of amplified response spectra. A*detailed report is presently being prepared I to fully describe its use on Surry Units 1 and 2. The report will be submitted on or before June 8, 1979 and will be entitled "Soil Structure Interaction in the Development of Amplified Response Spectra for Surry I Power Station, Units 1 and 2. 11 With the submittal of the SSI-ARS report (on or before June 8, 1979), the submittal of the two loop benchmark problem (on or before I June 8, 1979), the submittal of information regarding the status and schedule of IE Bulletin 79~02 (letter dated June 4, 1979, Serial No.

146/030879A), and the information contained in the Report attached to this I letter, we believe we have complied with all of the staff's outstanding requests for information. We plan no further submittals, except*the final report on the piping analysis, unless subsequent evaluation of the above information by the staff leads to further inquiries. Because of the severe I economic consequences of the present shutdown status of the plant, we plan to respond as quickly as possible to any questions the staff may have.

However, we believe there is sufficiently deta.iled information available to I the staff from this and past submittals, meetings, and telephone conversations to evaluate quickly and with confidence our request to lift the Order and resume operation of Unit 1.

• I We believe it to be in the best interests of our customers and the citizens of the Commonwealth of Virginia to minimize this country's dependence on oil. For Surry to be allowed to restart and to function I during the coming hot months is commensurate with that goal. To be allowed I

I I

I I VIRGINIA ELECTRIC AND POWER COMPANY TO Mr. Harold R. Denton SHEET NO. 4 I to do this requires a commitment to address safety concerns to the satisfaction of both ourselves and the NRC. We believe we have gone the I extra mile in the case of the Surry pipe stress reanalysis effort and our findings fully justify our position to start up.

As stated in Section 4 of the Report, all reanalysis of Unit 1 systems will be completed and fully reviewed by Engineering Assurance personnel by October 1, 1979.

I The staff's accessibility during our reanalysis effort is gratefully acknowledged and appreciated.

Prompt consideration and affirmation of our proposal would be I appreciated.

I W. C. Spencer I Vice President - Power Station Engineering and Construction Services I Attachment I

I I

I

• I I

I I

I

I I h-1284622-79 06/05/79 SURRY POWER STATION, UNIT 1 042 I TABLE OF CONTENTS 1.12 -

I Section

1.

SUMMARY

AND CONCLUSIONS 1-1 1.15 1.17 I 2. SCOPE OF REANALYSIS

• 2-1 1.19

3. PIPE STRESS RESULTS 3-1 1. 21 I 4. PIPE SUPPORTS RESULTS . 4-1 1.23

5. SCHEDULE FOR COMPLETION. 5-1 1.25 I 5.1 Pipe Stress and Support Reanalysis 5-1 1. 27 5.2 Hardware Modifications . . 5-1 1. 28 I 5.3 6.

Review of SHOCKl Program HIGH ENERGY LINE BREAKS 5-4 6-1

1. 29 1.31 I 7. CONSERVATISMS

• 7-1 1.33 7.1 Stress Limits. 7-2 1.35 I 7.2 7.3 7.4 System Redundancies .

• Safety Systems . * .

Field Verification of As-Built Conditions.

7-3 7-4 7-4

1. 36 1.37

1. 38 7.5 Quality Assurance/Engineering Assurance . *

• 7-5 1. 39 I 7.6 7.7 Use of Amplified Response Spectra . * * .

• Development of Soil Structure Interaction Amplified 7-6 1.40

1. 41 Response Spectra CSSI-ARS)

• 7-8 1. 42 7.8 Seismic Event Probability * . . * . *

• 7-1.0 1. 52

8. LICENSEE EVENT REPORTS AND RESOLUTIONS * . 8-1 1.54 APPENDIX 1. 56 A. SYSTEMS AFFECTED . . * * . * * *

• A-1 1. 58 B. FLOW.DIAGRAMS - IDENTIFICATION OF PROBLEMS REANALYZED B-1 2.2 I C.

D.

SNUBBER CAPACITIES . * .

RESPONSE TO IE BULLETIN 79-04.

C-1 D-1 2.4 2.6 I E.

F.

COMPARISON OF SHOCKO RESULTS WITH NUPIPE.

SEISMIC CAPABILITY OF NUCLEAR PIPING.

E-1 F-1 2.8 2.10 I G. CORRESPONDENCE WITH NRC . . . G-1 2.12 I i

• I

I h1284622-lc 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

.,I SECTION 1 1.10 SUMH:ARY AND CONCLUSIONS 1.12 I In response to the Nuclear Regulatory Commission's Order to Show Cause, gated 1.16 Harch 13, 1979, a reanalysis is being conducted of safety-related Riping 1.17 I systems which were originally dynamically analyzed using the SHOCK2 ~omputer 1.18

.,I program. Ihis program, which used an earlier load combination methodology, is no longer ~onsidered acceptable by the NRC .

1.19 1.20 Ihis report addresses details of the analysis work, results of pipe and 1.21 I support analyses to date,~ discourse on conservatisms, and other topics 1.22 I within the scope of the reanalysis task. It is in support of our effort to restart Unit 1 as discussed in the transmittal letter with this report. Ihe 1.23 1.24 I report is regarded as a culmination of all work to date and is in addition to other submittals previously forwarded Eince the Order to Show Cause. A 1.26 I listing of correspondence with the NRC, through the date of this report, is included in Appendix G for reference.

I

-I Ihe seismic reanalysis uses methodology currently is based on a piping analysis program, NUPIPE, that acceptable to the NRC. Ihe results to date 1.27 1.28 I indicate that the subject systems will be able to perform their intended 1.29 I

I 1-1 I

I hl284622-lc 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I safety functions yncler the maximum seismic conditions specified in the Final 2.1 Safety Analysis Report. Ihe reanalysis effort has clemonstratecl the 2.2 conservative nature of the original seismic analysis. Ihe piping systems have 2.3 been found to be *impacted only slightly after a thoroughly rigorous I ~eanalysis. Results also show that no piping of any size will have to be 2.5 replaced or repaired. A few systems may require addition of minor pipe 2.6 I support hardware to limit stresses to code allowable yalues; however, these 2.7 I changes are clue to reasons other than the algebraic summation process.

I In aclclition to the systems formerly analyzed with SHOCK2, systems which were originally analyzed ~ith SHOCKO and SHOCKl (predecessors of SHOCK2) are being 2.8 2.9 I reevaluated to demonstrate existing seismic adequacy.

I I

I I

• I I

I I 1-2 I

I 06/05/79 46 hl284622-lj I SURRY POWER STATION, UNIT 1 I

'I SECTION 2 . 1. 7 SCOPE OF REANALYSIS 1. 9 I As described in the NRC Order to Show Cause,.Harch 13, 1979, some piping 1.15 systems in the ~urry Power Station, Unit 1, were dynamically analyzed with a 1.16 I computer program that is not currently acceptable to the NRC. 1.17 I In.response to the Order to Show Cause, the following actions were taken: 1.18 I

1. Safety systems or portions thereof that were dynamically analyzed 1. 20 I using the £Omputer program SHOCK2 were identified. Ihese are listed 1. 22 I in Appendix A. Ihe specific piping reanalyzed is shown on the flow diagrams in Appendix B.

1. 23 I

Z, These systems are being reanalyzed using computer programs hased on methodology currently acceptable to discussed in Sections 3 and 4.

the NRC. Ihese programs are

1. 25

1. 26 I

• I J. Results of the reanalysis are compared with code allowable pipe stresses, with £llowable loads for nozzles/penetrations, and with the 1.27 1.28 1, results of original design loads for pipe supports.

I  !+/-. In those cases where the reanalysis indicated that stresses or loads 1.29 may be in excess of £llowable values, using the newer methodology, 2.1 I

2-1 I

I hl284622-lj 06/05/79 46 I SURRY POWER STATION, UNIT 1 I

I further examination is modifications are identified.

initiated and, where required, gquipment 2.2 In addition to the* analyses addressed in the Order to Show Cause, those 2.4 1* problems originally analyzed ~ith the SHOCKO and SHOCKl programs were 2.5 identified and are being reanalyzed to demonstrate existing seismic adequacy.

I Ihis investigation is discussed in Section 5.3. 2.6 I Information regarding s~fety-related piping that was not originally subjected 2.7 I to computer ~eismic analysis (for example, small diameter pipe done by calculation) is included in the VEPCO letter of May 24, 1979 Iesponding to the hand 2.8 2.9 I NRC letter dated April 2, 1979.

I!

I

'_,I I\

1*

2-2 I

I'

I hl284622-lw 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I SECTION 3 1.10

-1 PIPE STRESS RESULTS 1.12 I'

A total of 87 pipe stress problems have been identified for reanalysis. 1.15 I

I' Qf these, 69 problems were originally analyzed by the PSTRESS/SHOCK2 computer 1.16

-,, ~rogram that used algebraic summation and are therefore specifically addressed by the Show Cause Order, ~ were hand calculations, and 12 problems were 1.17 1.18 originally analyzed by various versions of PSTRESS/SHOCKO. Ihis latter 1.19 t program is not specifically addressed by the Show Cause Order but is now considered not gquivalent to currently .accepted practice. Ihese stress 1. 21 I' problems are being analyzed by two groups: Stone & Webster tngineering 1. 22 1* Corporation in Boston, Massachusetts, and Nuclear Campbell, California, as indicated in the following table:

Services Co~poration in

1. 23 I PIPE STRESS PROBLEMS 1. 25 I

• I S&W NSC TOTAL 1. 28 I SHOCK2 42 27 69 2.1 1*

3-1 I

I

I h1284622-lw 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

SHOCKl 5 7 12 2.2

-\1 HAND CALCULATIONS TOTAL 6

53 0

34 6

87 2.3 2.4 I Status as of June 2, 1979 2.6 I [ield verified piping isometric drawings provide the basis for program inputs 2.9 I' for the pipe stress reanalysis of the ~HOCK2 problems. Ihe reanalysis is 2 .11

-, conducted using. the NUPIPE computer program. NUPIPE calculates intra-modal seismic forces using a modified square root of the sum of the squares CSRSS) technique which is always more conservative than the approved SRSS method, and 2.12 2 .13 I an SRSS technique for inter-modal combination.

I Additionally, in some cases, piping is analyzed utilizing amplified response 2.14 I spectra CARS) that are developed ysing soil structure CSSI-ARS). Ihe resultant stresses and interaction techniques loads are used to evaluate piping, 2.15 2.16

't supports, nozzles, and penetrations. Ihese techniques are discussed in 2.17 Section 7.7. In accordance with the NRC letter of Hay 25, 1979 to Virginia 2.18 Electric & Power Company CVEPCO), the seismic inertial stresses computed using

.f the SSI-ARS have been increased by a factor of 1.5 for the DBE condition. 2.19 1*

1*

I' 3-2 I

I hl284622-lw 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I Qf the 69 SHOCK2 problems, 24 have been reanalyzed and approved by the Stone & 2.20 Webster Engineering Assurance Division up to this time. Ihis constitutes a 2.22 sample of 24/69 = 35% of the total SHOCK2 stress problems at Surry 1. Iable 2.23 3-1 shows the list of.problems to be reanalyzed including the results for I' these 24 SHOCK2 problems ~lus 5 of the total of 6 hand calculations rerun on 2.24 t liUPIPE and 4 of the 12 SHOCKl problems, for a total of 33 completed, stress problems.

accepted In Table 3-1, the figures for Original Total Stress, at the 2.25 2.26 I point of maximum total stress in the pipe, and Qriginal Seismic Stress, at the same point, . are extracted from Table 4.1 of the "Seismic Design Review 2.27

• 1 .t_qu_i.pment and Piping, Surry Power Station," dated September 15, 1971. Ihe 2.29 original calculations for the seismic design review are no longer available,

• I and £Orrelations were made to the original stresses in Table 4.1 on the basis 3.1 I of the MSK's. In some cases, particularly where hand calculations were used (problems 1020A, 1020B, 1020C, 1030, and 1010A), ~he original stresses are not 3.2 3.3 I available.

') In Table 3-1, the columns for*New Total Stress, at the point of maximum total stress in the pipe, and New Seismic stress, ~t the same point, were taken from 3.4 3.5 t the NUPIPE computer runs with the seismic inertial stress magnified by a

-I factor of 1.5 for ~uns using the SSI-ARS, per May 25, 1979.

the NRC letter to VEPCO Qf the 33 completed problems in Table 3-1, 24 used the SSI-ARS of 3.6 3.7 I

I' 3-3 I

'I

-1 hl284622-lw 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

a and 8 used the original ARS. Ihe stresses after the 1.5 magnification for 3.8

-, SSI-ARS are below the allowable stress for all 33 completed problems.

Ihe original total -and original seismic stresses shown in Table 3-1 were 3.9 I computed using the SHOCK2, iHOCKl, or SHOCKO programs or hand calculated for 3.10 t the original design conditions. Ihe new total and new seismic stresses were computed by the NUPIPE program using different mass models and in some £_ases 3.11 3.12 I different ARS's than the original calculations. tlore importantly, reanalyses were based on field-verified, as-built conditions in 1979, which in the 3.13 I some c~ses Qiffer significantly from the original design conditions. [or this 3.15 a reason, the new stresses and the original stresses in T~ble 3-1 comparable, as they do not necessarily Iepresent the same physical conditions.

are not 3.16 I Iable 3-2 summarizes the nozzles and penetrations evaluated under the 3.17 I reanalysis program.

by the SHOCK2 Qf a total of 67 nozzles on problems originally program, 14 analyzed have been evaluated and found to be acceptable, 9 3.18 3.19 t are under evaluation, and 34 Are problems for which the pipe is not complete and nozzle loads are not yet available.

stress analysis None of the nozzles 3.20 3.21 t

., for which evaluation is complete has been found to be unacceptable.

problems in which the SSI-ARS are used, the seismic inertial nozzle loads have been increased by a factor of 1.5 per the NRC letter of 25 May 1979.

[or those Ihere 3,22 3.24 1* are an additional 30 nozzles in problems which were originally computed by the 3-4

I hl284622-lw 06/05/79 042

'I SURRY POWER STATION, UNIT 1 I SHOCKO program or by hand calculations; of these, 5 have been evaluated to 3.25

-, date and all are acceptable.

Ihe SHOCK2 stress problems include 8 penetrations, Qf which 2 have been 3.27 I'V evaluated and found to be acceptable. Ihe remaining six are in problems for 3.28 which analysis is not complete, consequently the loads on the penetrations are I not yet ~vailable. 4.1 I

I

'I II

1

• t

• I' I

3-5

,.. .... -- ,. -* ... .. ... .. -; .. .. .. *- (-~ ... - ... .. -

,c hl284622-lx

(

06/05/79 041

/

1. 20 SURRY POWER STATION, UNIT 1 1. 22 TABLE 3-1 1. 24 PIPE STRESS REEVALUATION

SUMMARY

1. 26 NA - Not Available 1. 29

• Table 4-1 of Seismic Design Review 1. 30 Equipment and Piping Surry Power 1. 31 Station, Sept. 15, 1971 or 1. 32 Preliminary Criteria Subsequent Reanalysis** 1.33 Prob- Line Pioe Stress (]2Si) 1. 35 lem System Iso. Size Original Original New New Allow- 1. 36 li9..!....__ Name fuL_ CNPS) Total* Seismic* Total Seismic able 1.37 SHOCK2 Problems 1. 40 555 Low Head Safety 122 Dl 10 11 29290 NA 11180 5307 30882* 1.43 Injection 12 11 1.44 1555 Low Head Safety 122 Ll 12" 25290 NA 10392 3855 30882 1. 46 Injection 1.47 706.1\ Low Head Safety 122 Hl 6" 18451 10707 19830 8439 30769 1.49 Injection 1. 50 707A Low Head Safety 122 Jl 6" 1. 52 Injection 1. 53 708 Low Head Safety 122 Kl 6" 1.55 Injection 1.56 731A Low Head Safety 127 El 8" 22671** 984 21503 13940 24750 1.58 Injection 2 .1 731B Low Head Safety 127 E2 8" 22671** 984 21800 16004 24750 2.3 Injection 2.4 743 Low Head Safety 127 Fl 10 11 24649lOE 3738 16119 5496 33750 2.6 Injection 2.7 727 Low Head Safety 127 Cl 6" 2.9 Injection 127- C2 10" 2.10 735 High Head Safety 127 Gl 411' 6" 2.12 Injection 127 G2 8"' 10 11 2 .13 525A/ Containment & 123 Al 8" 11999 10866 9409 3846 33750 2.15 1525A Recirculation Spray 10 11 2.16 546/ Containment & 123 Dl 8" 2.19 560 Recirculation Spray 123 El 10 11 28209 24753 31976 16024 32616 2.20 1 of 7

-/ - *- .. - .. .. - - ..... .. .. -

... '11111 hl284622-lx

... r . .,

06/05/79 SURRY POWER STATION, UNIT 1 041 TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION SU:tiliARY Preliminary Prob- Line Pi2e Stress C2siJ lem System Iso. Size Original Original New New Allow-

~ Name fuLe_ CNPS) Total* Seismic* Total Seismic able 5461 Containment & 123 F3 8 II 2.23 5600 Recirculation Spray 10" 2.24 546/ Containment & 123 F2 8 ti 2.27 5620 Recirculation Spray 10 11 2.28 548C Containment & 123 HZ 10" 2.31 Recirculation Spray 2.32 547 Containment & 123 Cl 811 20953 5688 21960 19284 31482 2.35 Recirculation Spray . 10" 2.36 744/ Containment & 123 Jl 811 2.39 754 Recirculation Spray 2.40 i,

548A Containment & 123 Bl 811 2.43 Recirculation Spray 10 11 2.44 548B Containment & 123 J:{l 10 11 28660 26790 23251 18529 32616 2.47 Recirculation Spray 2.48 544 Containment & 123 Gl 10 11 13402 6986 6386 3766 28485 2.51 Recirculation Spray 123 G2 2.52 544A Containment & 123 R2 10 11 12853 11256 6814 3556 29970 2.55 Recirculation Spray 2.56 544B Containment & 123 Rl 10 11 12853 11256 6628 4541 28485 3.1 Recirculation Spray 3.2 751 Containment & 123 Nl 10 11 6010 5169 7085 5206 28485 3.5 Recirculation Spray 123 N2 3.6 562/ Containment & 123 Fl 10 11 3.9 546 Recirculation Spray 123 E2 3.10 745 Containment & 123 Kl 811 3.13 Recirculation Spray 3.14 323A Main Steam 100 Dl 30 11 13824 6343 13064 354 27000 3.17 2 of 7

(- .. . .. .. .. .. -' - .. .. - - -* -* .. - .. -

hl284622-lx 06/05/79 SURRY POWER STATION, UNIT 1 041 1: ,*

TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION

SUMMARY

Preliminary Prob- Line PiPe Stress (osi) lem System Iso. Size Original Original New New Allow-

~ Name No. CNPS) Total* Seismic* Total Seismic able 322A Hain Steam 101 Dl 30" 13031 5548 11532 400 27000 3.19 334A Hain Steam 102 Dl 30 11 18635 11082 15407 463 27000 :3. 21 346 Hain Steam 103 Al 30" 3.23 323B Feedwater 100 Gl 14" 15829 590 12923 8061 27000, 3.25 322B Feedwater 101 Gl 14" 17927 13521 15965 1796 27000. 3.27 334B Feedwater 102 Gl 14" 16025 12281 16145 9828 27000 3.29 417 Auxiliary Feedwater 118 Al 311 8568 NA 26769 14036 27000 3.32 118 AZ 3.33 607 Auxiliary Feedwater 118 Gl 411 18681 NA 18331 5467 27000 3.36 118 G2 6 II 3.37 636 Pressurizer Spray 125 Al 411 3.40

& Relief 3.41 630 Pressurizer Spray 124 Al 3"' 411 3.44

& Relief 124 A2 611' 12 II 3.45 540 Residual Heat 117 Bl 311' 411 3.48 Removal 6"' 12" 3.49 508 Residual Heat 117 Al 10", 12 11 I 3.52 Removal 117 AZ 14" 3.53 465 Service 1-fater 119 Al z4n 19101 18285 7778 5826 21600 3.56 119 A2 3.57 119 A3 3.58 119 A4 4.1 488/ Component Cooling 112 C 18" 4.4 480 112 Al 4.5 5071 Component Cooling 112 Fl 8 II 4.8 481 112 Bl 18" 4.9 3 of 7 wn

hl284622-lx 06/05/79

\

SURRY POWER STATION, UNIT 1 041 TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION

SUMMARY

Preliminary Prob- Line Pioe Stress Cosi) lem System Iso. Size Original Original New New Allow-

~ Name ~ CNPS) Total* Seismic* Total Seismic able 614 Component Cooling 112 AEl 12 II 4.12 112 AE2 18" 4.13 512 Component Cooling 112 ANl 18" 4.16 603A Component Cooling 112 Sl 18 11 4.18 766 Component Cooling 112 AR 8 II 4.21 112 T 4.22 605A Component Cooling . 112 AAl 311' 6 I I ' 4.25 112 AA2 18" 4.26 605B Component Cooling 112 AAl 3 II I 611' 4.29 112 AA2 18 11 4.30 509A Component Cooling 112 Gl 8"' 12 11 , 4.33 18 II I 24 11 4.34 I

612 Component Cooling 112 AKl l 8 11 4.37 1512 Component Cooling 112 J 18 II 4.39 2529 Component Cooling 112 AH 311' 6 I I ' 4.42 8"'11 10 11 11, 4.43 14 , 18 4.44 2526 Component Cooling 112 AJ 2 **2 11 ' 6 t i ' 4.47 811' 10" 4.48 2527 Component Cooling 112 i\L 411' 6 t i ' . 4.51 811' 10" 4.52 527A Component Cooling 112 Tl 411' 6 t i ' 4.55 8"' 10 11 ' 4.56 14" 4.57 517 Component Cooling 112 Ml 4rr' 6 t i ' 5.2 112 M2 8 t i ' 10 11 ' 5.3 112 M3 14 11 ' 18" 5.4 4 of 7

hl284622-lx

\-

06/05/79 SURRY POWER STATION I UNIT 1

... 041

... ...... ., *I TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION

SUMMARY

Preliminary Prob- Line Pioe Stress (]2Si) lem System Iso. Size Original Original New New Allow-1!Q_,_ Name No. CNPS) Total* Seismic* Total Seismic able 603B Component Cooling 112 Sl 18" 5.7 526A Component Cooling 112 13 611 I 8" 5.9 526B Component Cooling 112 L1 6" I 8" 5.11 526C Component Cooling 112 L1 6 II I 8 II 5.14 112 12 5.15 112 13 5.16 527B Component Cooling 112 T2 411 I 6 l l I 5.19 8 II I 10 11 1 5.20 14" 5.21 527D Component Cooling 112 T3 4" I 6ll I 5.24 8 II I 10 11 I 5.25 14" 5.26 509B Component Cooling 112 G2 8 11 I 12", 5.29 18 II I 24 11 5.30 509C Component Cooling 112 G3 811, 12 11 1 5.33 18" 1 24" 5.34 509D Component Cooling .112 G4 8 II I 12", 5.37 18 11 I 24" 5.38 CVl Containment Vacuum 137 A1 8" 25750** 1029 17554 1209 27000 5.41 746 3" HP Steam 131 Al 311 5.44 131 B2 411 5.45 131 C3 5.46 CFl Fire Protection 144 Al 2" I 611 I 5.49 12 11 5.50 CF2 Fire Protection 144 Bl 1 l/2 11 , 5.53

? II

- I l 6 11 5.54 1040 Diesel Muffler Exhaust 143 Al 24" 5.57 5 of 7

/- - ... - - .. - .. -* /- - - - -** - *- - - -

hl284622-lx 06/05/79 SURRY POWER STATION, UNIT 1 041 TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION

SUMMARY

Prelimin2.ry Prob- Line Pioe Stress (Psi) lem System Isa. Size Original Original New New Allow-1iQ_,__ Name ~ CNPS) Total* Seismic* Total Seismic able Other Problems CHand Calculations and SHOCK0/1) 6.3 1000A Low Head Safety 127 Jl 2", 6TT 6.6 Injection 6.7 1010A Low Head Safety 127 J2 211 NA NA 24709 5423 33750 6.10 Injection 6" 6.11 1020A Low Head Safety 127 J3 2" NA NA 28401 6168 337 50 . 6 .14 Injection 6" 6.15 1020B Low Head Safety 127 J4 6" NA NA 12305 5587 33750 6.18 Injection 6.19 1020C Low Head Safety 127 JS 6" NA NA 22453 1270 33750 6.22 Injection 6.23 1030 Service Water 1119 Al 24" NA NA 3092 1421 21600 6.26 537 Low Head Safety 122 Al 411 6"' 19247 13944 22928 17042 25789 6.29 Injection 122 A2 10 11' , 12" 6.30 755 .Containment & 123 Pl 12" 3950 2235 2400 867 33750 6.33 Recirculation Spray 6.34 756 Containment & 123 Ql 12" 2077 1230 2638 1205 33750 6.37 Recirculation Spray 6.38 611 Auxilary Feedwater 118 Ll 4"' 6" 6.41 554 Residual Heat 117 Cl 6" 16627 12375 25955 21992 32238 6.44 Removal 6.45 606 Component Cooling 112 ABl 12" 6.48 613 Component Cooling 112 ADl 3"' 4" , 6.51 8", 6" 6.52 502 Component Cooling 112 Dl 18" 6.55 6 of 7

/- .. .. - - -' - ... _,.. - - - ... - - - -

hl284622-lx 06/05/79 SURRY POWER STATION, UNIT 1 041

.,.i,"tr, TABLE 3-1 (Cont)

PIPE STRESS REEVALUATION

SUMMARY

Preliminary Prob- Line Pioe Stress (psi) lem System Iso. Size Original Original New New Allow-tl.Q_;__ Name ~ CNPS) Total:>E Seismic" Total Seismic able 506 Component Cooling 112 El l 8 11 6.58

. 0 112 E2 7 .1 112 E3 7.2 747 Spent Fuel Cooling 128 Al 12 11 7.5 748 Spent Fuel Cooling 128 Cl 12 ' 16" 11 7.7 749 Spent Fuel Cooling 128 Bl 12 11 7.9 Legend: 7.12 Allowable Stress = 1.8 Sh. __ 7.14

/. 7 Total Stres,;... = \p + SOW + SDBEI + s 7.16 s e1.sm1.c

• * - 1,aDBE I s_

+ '""!TBEA DBEA 7.18 7 of 7

I hl284622-lu 06/05/79 041 1.18 I SURRY POWER STATION, UNIT 1 1.19 Page 1 of 7 1. 21 I TABLE 3-2 Status: 6/2/79 1. 22 1.25 I 1. 27 SURRY POWER STATION, UNIT 1 NOZZLE AND PENETRATION

SUMMARY

1.29 No. Accep- Modifi- 1. 32 System Total No. table cations or 1.33 I and Prob. No.

of-Nozzles/

Penetrations Evaluation CComolete)

No. Under Evaluation Additions Required Comment 1.34 1.35 SHOCK2 PROBLEMS 1.37 I

Low Head 1.39 I Safety Injection 1.40

1. 41 1 1 0 0 SSI-ARS 1.44 I 555 1555 '1 1 0 0 SSI-ARS 1.46 I 706A 707A 0

0 N/A N/A N/A N/A N/A N/A SSI-ARS 1.48 1.50 I 708 0 N/A N/A N/A 1. 52 731A 0 N/A N/A N/A Origi.nal ARS 1.54 I 731B 0 N/A N/A N/A Original ARS 1. 56 I 743 727 0 N/A N/A N/A 1.58 2.2 I High Head Safety Iniection 2.5 2.6 2.7

• -1 735 3 *

• Incomplete 2.9 Containment 2.12 I Recirculation SEray 2 .13 2.14 I 525A/1525A 546/560 0

1 N/A 1

N/A 0

N/A 0

SSI-ARS SSI-ARS 2.16 2.19 I

I

I hl284622-lu 06/05/79 041 I SURRY POWER" STATION, UNIT 1 Page 2 of 7 I TABLE 3-2 (Cont)

I System Total No.

No. Accep-table Modifi-cations or

• 1 and Prob. No.

of Nozzles/

Penetrations Evaluation (Complete)

No. Under Evaluation Additions Reauired Comment 546/5600 0 N/A N/A N/A 2.21 I 546/5620 0 N/A N/A N/A 2.23 548C 1 0 0 2.25 I 547 0 0

IUA N/A N/A Original ARS 2.27 I 744/754 548A 1

0 0

N/A 1

N/A N/A Incomplete SSI-ARS 2.29 2.31 I 548B 1 1 0 0 SSI-ARS 2.33 544 2/2 2/2 0/0 0/0 SSI-ARS 2.35 I 544A 2 2 0 0 SSI-ARS 2.37 I 544B 751 2

2 1

2 1

0 0

0 Original ARS Original ARS 2.39 2.41 I 562/546 745 0

1 N/A 0

N/A 1

N/A Incqmplete 2.43 2.45 I Hain Stearn 2.48 323A l/1 l/* 0/* 0/* SSI-ARS 2.50 I 322A l/1 l/* 0/* 0/* SSI-ARS 2.53

-I 334A 346 1/l 0/0 11*

N/A 0/*

N/A 0/*

N/A SSI-ARS Inc.ornplete 2.55 2.57 I Feedwater l/1 0/* SSI-ARS 3.2 3.4 323B l/* 0/*

I 322B l/1 l/* 0/* OnE Original ARS 3.7 I

I

I hl284622-lu 06/05/79 041 I SURRY POWER STATION, UNIT 1 Page 3 of 7 I TABLE 3-2 (Cont)

I System Total No, No. Accep-table Hodifi-cations or

• 1 and Prob. No.

of Nozzles/

Penetrations Evaluation (Comolete)

No. Under Evaluation Additions Reouired Comment

-.*. 334B l/1 l/* 0/* 0/* SSI/ARS 3.9 I Au:x:. Feedwater 3.12 417 0 N/A N/A N/A SSI-ARS 3.14 I 607 3 2 1  ::o SSI-ARS 3.17 I Pressurizer Spray & Relief 3.20 3.21 636 1 0 1 0 Incomplete 3.23 I 630 5

• Incomplete 3.25 I Residual Heat Removal 3.28 3.29 N/A N/A N/A 3.31 I 540 508 0

4 0 4 0 Incomplete 3.33 I Service Water 465 4 4 0 0 SSI-ARS 3.36 3.38 I Comoonent Cooling 3.41 488/480 4 Incomplete 3.43 I 507/481 4 *

• Incomplete 3.46 NIA 3.48

• I 614 512 0

0 NIA N/A NIA 3.50 I 603A 766 1

2 0 0 0 Incomplete Incomplete 3.52 3.54 I

605A 2

• Incomplete 3.56 I

I

I hl284622-lu 06/05/79 041 I SURRY POWER STATION, UNIT 1 Page 4 of 7 I TABLE 3-2 (Cont)

I System Total No.

No. Accep-table Hodifi-cations or

• 1 and Prob. No.

of Nozzles/

Penetrations Evaluation (Complete)

No. Under Evaluation Additions Reouired Comment 605B 0 NIA NIA NIA 3.58 I 509A 3

• Incomplete 4.2 Incomplete 4.4 I 612 1512 2

0 NIA NIA N/A 4.6 I 2529 2526 0

0 N/A N/A N/A N/A N/A N/A 4.8 4.10 I 2527 0 N/A N/A N'/A 4.12 527A 2 . *

• Incomplete 4.14 I 517 4 *

• Incomplete 4.16 I 603B 526A 0

0 N/A N/A N/A N/A N/A N'/A 4.18 4.20 I 526B N'/A N'/A N/A 4.22 4.24 526C 0 I 527B 0 N/A N'/A NIA 4.26 527D 0 N'/A N/A N'/A 4.28 I 509B 0 N/A N'/A N/A 4.30

-I 509C 509D 0

0 N'/A N/A N/A N/A N'/A N/A 4.32 4.34 I Containment Vacuum 4.37 4.38 I

CVl 1 1 0 0 SSI-ARS 4.40 I

I

I 06/05/79 041 hl284622-lu I SURRY POWER STATION, UNIT 1 Page 5 of 7 I TABLE 3-2 (Cont)

I System Total No.

No. Accep-table Hodifi-cations or

• 1 and Prob. No.

of Nozzles/

Penetrations Evaluation (Complete)

No. Under Evaluation Additions Required Comment 3" HP Steam 4.43 I 746 1

• Incomplete 4.45 Fire Protection 4.48 I CF-1 0 N/A N/A N/A 4.50 I CF-2 Diesel Muffler 0 N/A N/A N/A 4.52 4.55 Exhaust 4.56 I 1040 0 N/A  !'f/A l'f/A 4.58 I

I I

I I

• I I

I I

I

I hl284622-lu 06/05/79

• 041 I SURRY POWER STATION, UNIT 1 Page 6 of 7 I TABLE 3-2 (Cont)

I No. Accep- Modifi-

-, System and Prob. No.

Total No.

of Nozzles/

Penetrations table Evaluation CComolete)

No. Under Evaluation cations or Additions Reouired OTHER PROBLEMS CHAND CALCULATIONS AND SHOCK0/1)

Comment 5.4 I Low Head Safety 5.7 5.8 Injection 5.9 I 1000A 0 N/A N/A NIA 5 .11 I 1010A 1020A 0

0 NIA N/A N/A N/A N/A SSI-ARS SSI-ARS 5.13 5.15

• 1 1020B 0 N/A N/A N/A SSI-ARS 5.17 1020C 0 - N/A N/A N/A SSI-ARS 5.19 I Service Water 5.22 SSI-ARS 5.24 I 1030 Low Head 4 4 0 0 5.27 Safety 5.28 I Injection 537 1 Incomplete 5.29 5.31 I Containment &

Recirculation &

5.41 5.42 Spray 5.43 I 755 1 0 1 0 Incomplete 5.45

-1 756 Aux. Feedwater 1 1 0 0 Original ARS 5.48 5.51 I 611 3 0 3 0 Incomplete 5.53 5.56 Residual Heat I Removal 554 0 N/A N/A N/A SSI-ARS 5.57 6.1 I .. ,*

I

I 06/05/79 041 hl284622-lu I SURRY POWER STATION, UNIT 1 Page 7 of 7 I TABLE 3-2 (Cont)

I System Total No.

No. Accep-table Hodifi-cations or

• 1 and Prob. No.

of Nozzles/

Penetrations Evaluation (Complete)

No. Under Evaluation Additions Required- Comment Component 6.4 I Cooling 606 0 N/A N/A N/A 6.5 6.7 I 613 3

• Incomplete 6.10 502 4 Incomplete 6.12 I 506 5

• Incomplete 6.14 6.17 I Spent Fuel Cooling 6.18 747 4 Incomplete 6.20 I 748 2

• Incomplete 6.23 I 749 2 *

• Incomplete 6.25 I

I I

• I I

I *Stress analysis not complete; loads not available N/A not applicable I

I

I hl284622-lv 44 I SURRY POWER STATION, UNIT 1 I

I SECTION 4 1. 9 i -*1 PIPE SUPPORT RESULTS 1.11 I

Iable 4-1 summarizes the pipe supports evaluated in the reanalysis program. 1.14 I Ihere are 846 supports on lines originally analyzed by SHOCK2; of these, 117, 1.15 I about l/7 1 have been evaluated and found acceptable.

acceptable if all the reaction components are lower A support is considered in magnitude than the 1.17 I reactions for which the support was originally designed. If some reaction 1.19 component is greater than the original design reaction, the £upport is 1.20 I reanalyzed using the new reactions. Qf the total SHOCK2 supports, 300 are 1.21 I being reevaluated at this time.

for which stress analysis is

~n additional 429 supports are in problems not yet complete and hence for which support 1.22 1.23 I reactions are not available.

I In cases where _SSI/ARS was used, the DBE seismic inertial reactions on 1.24 supports are multiplied by 1.5.

I

• I I

I 4-1 I

I

I h1284622-lt 06/05/79 042 1.18

1. 20 I SURRY POWER STATION, UNIT 1 Page 1 of 6 1. 22 I TABLE 4-1 Status 6/2/79 1.23 1.26 I PIPE SUPPORTS STATUS

SUMMARY

1. 28

• 1 System and Total No.

of No_. Acceptable Evaluation No. Under Modifications or Additions

1. 31

1. 32

1. 33

-.*. Prob. No. SuP2orts CCom2lete) Evaluation Reguired Comment 1.34 I SHOCK2 PROBLEMS 1.36 Low Head 1. 38 I Safety Injection 1.39

1. 40 I 555 1555 14 9

5 5

9 4

1. 42 1.44 I 706A 707A 14 2 12 8

1.46 1.48 8

I 708 25 2 23 1. 50 731A 4 1 3 1.52 I 731B 4 1 3 1.54 I 743 727 5

5 5 1.56 1.58 I High Head Safety 2.2 2.3 Injection 2.4 I 735 30 2.6

-I Containment &

Recirc. SPray 2.8 2.9 1525A 16 11 5 2 .11 I 525A 15 13 2 2.13 I 546/56'0 546/5600 11 11 11 11 2.15 2.17 I

I

I hl284622-lt 06105179 042 SURRY POWER STATION, UNIT 1 I

Page 2 of 6 I TABLE 4-1 (Cont)

I' System Total No. No. Acceptable Modifications or and of Evaluation No. Under Additions

• 1 Prob. No. Supports (Complete) Evaluation Required Comment 54615620 11 11 2.19 I 548C 11 11 2.21 547 12 3 9 2.23 I 74417 54 5 5 2.25 2.27 I 548A 548B 1

22 1

1 0

21 0

2.29 I 544 12 6 6 1

2.31 2.33 544A 5 4 I 544B 5 3 2 2.35 751 0 0 0 NIA 2.37 I 5621546 3 2.39 I 745 Main 8 1 7 2.41 2.43 Steam 2.44 I 323A 15 15 0 0 2.46 I 322A 334A 3

2 3

2 0

0 0

0 2.48 2.50

• I 346 50 2.52 Feedwater 2.54 I 323B 10 10 0 0 2.56 2.58 I

322B 334B 5

3 3

3 2

0 0 3.2 I

I

I h1284622-lt 06/05/79 042 I SURRY POWER STATION', UNIT 1 Page 3 of 6 I_ TABLE 4-1 (Cont)

I System Total N'o. N'o. Acceptable Modifications or

-1 and Prob. N'o.

of SuPPorts Evaluation CComPlete)

N'o. Under Evaluation Additions Reauired Comment Aux. Feedwater 3.5 I 417 26 12 14 3.7 607 12 4 8 3.9 I Pressurizer 3.12 Spray & Relief 3.13 I 636 29 29 3.15 630 21 3.17 I Residual Heat 3.19 Removal 3.20 I 540 30" 30 3.22 22 3.24 I 508 Service Water 22 3.26 I 465 4 Component Cooling 4 3.28 3.30 I 488/480 24 3.32 507/481 20 3.34

.I 614 14 3.36

• I 512 603A 5

4 3.38 3.40 I 766 605A 5

8 3.42 3.44 I 605B 7 3.46 I

I

I hl284622-lt 06/05/79 042 I SURRY POWER STATION, UNIT 1 Page 4 of 6 I TABLE 4-1 (Cont)

I System Total No. No. Acceptable Modifications or and of Evaluatiqn No. Under Additions

• 1* Prob. No. Supports (Complete) Evaluation Reouired Comment 509A 21 3.48 I 612 3 3.50 1512 7 3.52 I 2529 42 3.54 I 2526 2527 13 29 3.56 3.58 I 527A 8 4.2 517 13 4.4 I 603B 4 4.6 4.8 I 526A 526B 10 9 4.10 I 526C 527B 17 11 4.12 4.14 I 527D 16 4.16 509B 9 4.18 I 509C 8 4.20

-I 509D Containment 4.22 4.24 Vacuum 4. 25 I tvl 13 13 4.27 I 3" HP Steam 746 7 4.30 4.32 I

I

I hl284622-lt 06/05179 042 I SURRY POWER STATION, UNIT 1 Page 5 of 6 I TABLE 4-1 (Cont)

I System Total No. No. Acceptable Modifications or

.*1 and Prob. No.

of Suooorts Evaluation (Complete)

No. Under Evaluation Additions Reouired Comment Fire Protection 4.36 I CF.l 5 4.38 CF.2 1 4.40 I Diesel Muffl'er 4.43 4.44 Exhaust I 1040 15 6 9 4.46 OTHER PROBLEMS CHAND CALCULATIONS AND SHOCK0/1) 4.49 I Low Head 4.52 Safety 4.53 I Injection lOOOA . 4 2 2 4.54 4.56 I 1010A 2 2 4.58 1020A 8 7 1 5.2 I 1020B 8 6 2 5.4 5.6 I 1020C Service Water 11 6 5 5.9 I 1030 4 Low Head Safety Injection 4 0 0 5.11 5.15 I 537 28 28 5.17 Containment & 5.21 I Recirculation Soray 755 3 3 5.22 5.24 I 756 3 3 5.26 I

I

I h1284622-lt 06/05/79 042 I SURRY POWER STATION, UNIT 1 Page 6 of 6 I TABLE 4-1 (Cont)

I System Total No. No. Acceptable Modifications or and Prob. No.

of Supports Evaluation CComPlete)

No. Under Evaluation Additions

---'R=e=g~u=i=r~e=d..._,._ Comment Aux. Feedwater 5.30 I 611 8 8 5.32 Residual Heat 5.36 I Removal 5.37 554 7 7 5.39 I Component Cooling 5.42 606 5.44 I 613 3

10 5.46 I 502 506 16 17 5.48 5.50 I Spent Fuel Cooling -

5.52 5.53 I 747 748 4

6 5.55 5.57 I 749 6 6.1 I

• I I

I I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I SECTION 5 1.10

• 1 SCHEDULE FOR COMPLETION 1.12 I

1,1 PIPE STRESS AND SUPPORT REANALYSIS 1.15 I

I All reanalysis of Unit 1 systems will be completed and reviewed by Engineering Assurance personnel by October 1, 1979. Ihis includes all pipe stress, 1.16 1.18 I equipment nozzle, penetration, and pipe support evaluations. All required 1.19 modifications will be identified in advance of this date.

I I d,2 HARDWARE MODIFICATIONS 1. 20 I ~tress analysis of a sample of safety-related piping has progressed to the 1. 21 point where some ~tresses which would exceed code allowable levels have been 1. 22 I identified. ~ome have been resolved by the use of more detailed or refined 1.23 modeling techniques or by use of ~oil structure interaction amplified response 1.24 I spectra (SSI-ARS). Qthers have been designated to be corrected by physical 1.25

• I hardware additions rather than pursue further done where analysis. ~enerally this the addition of restraints or damping devices (snubbers) would be is 1.26 I easier and less time-consuming than it would be to continue calculational 1.27 I

5-1 I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I procedures. Host of these as-built and original design.

modifications In addition are to due the to differences between basic verification for 1.28

'1.29

• 1 SHOCK2, new information is being incorporated in calculations to Qpgrade the 2.1 analyses where important changes to the input have occurred or where I additional data have been generated since the original analysis. Ihis 2.3 produces some stresses that exceed code allowables. Ihese cases are also 2.4 I being corrected at this time even though they are not part of the Show Cause I Order.

I Ihose hardware modifications now identified inside the containment in areas 2.5

~hich are not accessible because of radiation levels during plant operation 2.6 I ~ill be performed prior to startup of Unit 1. 2.7 I Ihose hardware modifications outside the containment which are presently 2.8 I identified will be £erformed within 30 outside the days. Any additional. modifications containment which may be determined in the process of completion 2.10 I Qf the stress analysis will be completed within 30 days of the decision to 2.11 modify hardware.

I

• I If any further modifications are determined containment, a detailed gvaluation of the severity of the to be needed condition inside will the be 2.12 2.13 I

I 5-2 I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I prepared and a proposal made to the NRC in order to getermine whether shutdown is warranted or whether modification at ihe next outage is reasonable.

2.14

• 1 Ihis evaluation would consist for example of comparing calculated stresses 2.15 I with Ieal yield stress; determining whether a hanger would deflect within 2.16 acceptable limits Qr actually be damaged; and review for redundancy and/or 2.17 I isolability of a line. Qther aspects of an evaluation program will be 2.18 I determined on a case-by-case basis.

I feveral modifications are currently identified even though the analysis and 2.19 evaluation ~tage is not yet complete. Ihese modifications are described in 2.21 I the following paragraphs for the problem identified. ~ecause final design of 2.22 each of these modifications is incomplete, they are currently included under I ,!;_he heading of "No. Under Evaluation" in Table 4-1. 2.23 I Problem No. 743 - Low Head Safety Injection. In order to meet the allowable 2.25 I stress for the DBE condition, a box-type !estraint on this line requires shimming to close £n existing gap between the restraint and the pipe. Ihis 2.27 I shim is necessary for the pipe support to function as a lateral restraint as I well as a vertical restraint. Ihe addition of this shim is considered to be a very minor modification £nd its addition reflects the intent of the original 2.28 2.29 I design. Ihe analysis of the support itself is incomplete. 3.1 I

5-3 I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT l I

I Problem 548A Containment Recirculation Soray System.

as-built condition for this line differs from the Ihe field-verified, condition used in the 3.3

-,,, original analysis. Ihis problem was within code allowable stresses using the 3.4 SSI-ARS but exceeds the DBE allowable when the 1.5 magnification factor is 3.5 I applied to the seismic inertial stress, per the NRC letter of May 25, 1979.

Ihis condition will require the addition of a seismic snubber to the piping 3.6 I system. Ihis modification is considered minor and is attributable to the 3.7 I difference between the as-built design of the snubber is i?complete.

and the Qriginal design conditions. I.he 3.9 I

Problems 731A and 731B Low Head Safety Injection. Ihese two problems each 3.11

,11 contain a box type support. ~ased on preliminary analysis using the original 3.12 FSAR ARS, these pipe supports will deflect excessively gnder DBE loading 3.13 I conditions. Although these supports have not been reanalyzed using the SSI- 3.14 I ARS loads, to be added.

it is evident that these supports ~ill require a structural brace I.he design of the required modification is incomplete.

3.15 3.16 I

}.3 REVIEW OF SHOCKl PROGRAM 3.19 I

• I J.welve Surry* 1 pipe stress problems were computed by versions of the program fSTRESS/SHOCKl. I.his program performed intermodal combination by the so-3.20 3.23 I

called "Navy method," :lihich consists of the absolute sum of the largest 3.24 5-4 I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I magnitude modal response and the square root of the sum of the squares of the 3.25 remaining modes. Intra-modal responses due to multi-directional earthquake 3.26 excitation were not calculated 12.ecause SHOCKl only produced responses parallel 3.27 to a given earthqUak~ component excitation ~i.e., the responses were 3.28 I considered uncoupled). [or this reason, the SHOCKl code is not considered 3.29 consistent with current analysis techniques. Ihe Navy method, being more 4.1 I conservative than a straightforward square root of the ~um of the squares of 4.2 the modal responses, generally Erovides more than adequate conservatism. 4.3 1**

~omparative results calculated for Maine Yankee Atomic Power Station< 1 > have 4.4 I led to the conclusion that iHOCKl is suitably conservative to ensure that the 4.5 piping systems meet the allowable stress levels.

1*

A listing of the latest version of the SHOCKl program was sent to the NRC 4.7 I (letter from S&W to Mr. Denton, NRC, dated April 6, 1979). However, no safety 4.8 I systems at Surry 1 are known 'to have been analyzed using ihis version of the program, and so no verification of this £rogram for Surry 1 has been done.

4.9 4.10 1* Ihe 12 problems cited were analyzed using earlier versions of the program, now 4.12 called SHOCKO, for which no listings are now available. ~omparative analyses 4.13 I given in Appendix E, using the NUPIPE program together with similar ~tudies 4.14

-1 made for the Maine Yankee Atomic Power Station, show that SHOCKO stress results comparable to £ccepted programs and provides assurance that the produced 4.15 4.16 I FSAR criteria are met. Ihe studies and reanalyses performed to date reaffirm 4.17 I

5-5 I

I

I h-1284622-77 06/05/79 041 I SURRY POWER STATION, UNIT l I

I that the Seismic Category I ~iping is conservatively designed to withstand the effects of the design Qasis earthquake, 4.18 4.19

1 tour of the 12 pipe stress problems originally calculated by SHOCKO/SHOCKl 4.20

-1 - have been* recompui:ed ysing NUPIPE - and- field-verified, as-built conditions. _ 4:. 21 I Ihese four problems, Table 3-1 given earlier.

numbers 755, 756, 537, and 554, are summarized in Ihe SSI-ARS was used only on Problem No. 554, and 4.22 4.24 I' the seismic inertial stress was multipled by 1.5. In all four cases the 4.25

-, computed maximum total stress is less than* the allowable cases yery much less than the allowable stress.

stress Also, and in for comparison two 4.27 1* purposes, two additional SHOCKI/SHOCKO stress problems were without the yse rerun on of field~verified data; consequently these do not represent NUPIPE 4.28 I the results from the Surry 1 reanalysis effort.

studies were the same as those Ihe mass models used in these used originally. Ieveral default program 5.1 5.2 I values, particularly assumed ~upport stiffnesses, were different in the NUPIPE 5.3 reruns ~o that the mathematical models were not identical, although they are. 5.4 I as similar as would be ~xpected in normal production work. It is believed 5.6 that these comparative examples fairly illustrate the results that would be I obtained by ~reduction stress analysts using the two programs. In addition to 5.8

-I the two Surry- 1 examples given here, the Maine Yankee Atomic Power Station<

verified the lHOCKl program against NUPIPE and compared 10 SHOCKO problems 1 >

5.9 I ~ith SHOCKl runs, and 3 SHOCKO problems with NUPIPE runs, with essentially 5.10 I

1* 5-6 I

I

,. h-1284622-77 06/05/79 SURRY POWER STATION, UNIT 1 041 I

I similar results.

problems, the*

Appendix E gives a description of respective isometric drawings, each stress of the plots, two and Surry

~tress 5.11 5.12

-,,) comparisons by node point.

• 1 ~ecause the SHOCKO program is not equivalent to current practice, the Virginia 5.13 I Electric and fower Company has decided that all Surry 1 pipe originaliy analyzed stress problems by PSTRESS/SHOCKO or SHOCKl Eill be reanalyzed using the 5.14 5.15 I benchmarked accepted program NUPIPE, using field-verified data.

the conservative nature of the SHOCKO stresses as In determined view* of by the 5.16

• 1

,. comparative analyses, and in view matter of Haine Yankee Atomic Qf the determination by the NRC Staff in the Power Station<2>, Docket No, 50-309, that 5.17 5.18

,. SHOCKO/SHOCKl did not use the algebraic summation method, 12 SHOCKO/SHOCKl problems of the SHOCK2 problems.

reanalysis will be given a lower priority than the reanalysis Ihe Virginia Electric and Power Company of commits the to 5.19 5.20 I reanalyze all the remaining SHOCKO problems and to provide the Iesults to the NRC Staff by October 1, 1979.

5.21 1*

t <1 > S&W Report, "Verification of SHOCKl Program for Haine Yankee Atomic Power 5.24

.1 Station," April 19, 1979.

I I

• 1 5-7 I

I

,. h-1284622-77 06/05179 SURRY POWER STATION, UNIT 1 041 I

I <.2> NRC "Safety Evaluation by the Office of Nuclear Reactor Regulation 11 5.25

-, May 24, 1979 I

I I

t' I

I I

I I

1* 5-8 I

I h-1284622-63 06105179 042 I SURRY POWER STATION, UNIT 1 I

I SECTION 6 1.10 I HIGH ENERGY LINE BREAKS 1.12 I

I Qf the high energy lines addressed in Appendix D of the FSAR, only the main steam lines outside ihe containment are included in this stress reanalysis.

1.15 1.16 I tach of the main steam lines has two terminal break locations, one at the 1.17

• 1 containment ~enetration and the other at the main steam manifold. Ihese 1.19 terminal breakpoints are predetermined and are not changed as a result of the

'I stress reanalysis.

Iwo intermediate break locations were originally determined based upon maximum 1.20 I primary plus secondary stresses. ypon initial inspection, it is felt that the reanalysis will not significantly affect their location.

1.21 I

I [or piping required.

downstream of the main steam manifold; no aeismic analysis is 1.23

-1 In summary, the reanalysis will not create any appreciable change to 1.24 I Appendix D of the FSAR.

I

. 6-1 I

I

I 06/05/79 042 h-1284622-63 I SURRY POWER STATION, UNIT 1 I

I Ihis will be verified with main steam line calculations after the stress reanalysis is completed ~nd the highest stress points reviewed by the same

1. 25 1.26 I procedure used in the original break analysis.

are noted, they will be incorporated into the If any break location changes existing station inspection

1. 27

1. 28 I program within 90 days.

I I

I I

I' I

I I

• I I

1*

1* 6-2 I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I SECTION 7 1.10

a CONSERVATISMS 1.12 I Ihere are many conservatisms inherent in the seismic design of the Surry Power 1.15 Station, Unit 1. iome of these are: 1.16 I

'I

• Elastic dynamic damping values.

analyses Ihis point are is performed discussed using in conservatively low greater detail in 1.18 1.19 I Appendix F.

t

,, . Multiple-directional seismic input, having equal intensity, is considered in design with each horizontal ,component earthquakes are typically stronger in one direction.

of plants. Actual

1. 20

1. 22 I ,. In the design of structures and equipment, it is convenient to assure 1.23 I' that all elements Qf the structur,e or equipment are designed .t.o 1.25 stress levels well below the actual strength of the materials, so I

., that any permanent delormation'is very small.

the need for complex behavior would significantly and costly inelastic reduce. structural Ihis approach obviates analyses.

response Inelastic prior to

1. 26

1. 27 I failure. rrom the standpoint of functionability, piping systems and 1.28 7-1 I

1*

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I components can withstand the Qeformation and usually even the failure of some supports due to redundancy, i.e., the existence Qf multiple 1.29 2.1

-a _load paths.

a -* Stress limits, whether elastic or inelastic, are based upon material 2 .2 I behavior contain a under static limited loading conditions. iince dynamic amount of energy, the margin (between the stress loads 2.3 I limits and failure) under dynamic loads is greater than under loads if elastically calculated peak static response is compared to the 2.4 I stress limits with strain rate effects neglected, 2.5 I

.,,

• Pipe and structural support members are selected

§.Vailable sections, and consequentl-:,r have gene-rally greater than the minimum Iequirements by the analysis.

from standard strength Hence, the computed 2.6 2.7 2.9 I stresses are often lower than the allowable stresses.

difference constitutes an additional margin in the actual plant.

Ihis 2.10 I

• Computed pipe stresses are magnified by code stress intensification 2 .11 I factors. As discussed in Appendix F, these stress intensification 2.12 factors are primarily fatigue factors, and are not strictly 2.13 applicable to the seismic DBE condition.

I 7-2

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I

• Piping seismic in general

£nalysis is is conservatively designed, even when no dynamic performed. [ossil-fueled power plants, 2.14 2.16 I refineries, and process plants have survived major earthquakes in _.2.17 California, Alaika, Guatemala, and other loc~tions with little or no I' piping damage. Ihis experience includes earthquakes considerably 2.18 larger than the DBE for Surry Unit 1. Ihe experience with piping 2.19 I performance in earthquakes is reviewed in detail in a Ieport included 2.20 I here as Appendix F.

• 1 In addition to the conservatisms listed above, which are inherent in any 2,22

~esign of nuclear facilities, there are additional conservatisms specific to 2.23 I the Surry units. Ihese conservatisms are not theoretical concepts, but indeed 2.24 I are real ~nd existing margins of safety.

difficult, Io quantify these conservatisms but this in no way negates the sound £onservative premise on which is 2.26 2.27 I the Surry reanalysis effort is discussed below.

based. Ihese additional conse~vatisms are 2.28 I'

1,1 STRESS LIMITS 3.9 I

.*1 Ihe analyses. and reanalyses of Seismic Category I piping systems are based upon the £Onservative stress limit of l.8Sh under the limiting faulted or DBE 3.10 3.11

• I loading conditions. Ihe present ASHE Section I I I Code specifies the piping 3.13 I

7-3 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I stress limit to be 2.4Sh under the [aulted DBE Condition.

NUREG/CR-0261 In July 1978, report* used the limit moment theory to address the Code rules, the 3.15

• 1 and it was established that gross plastic deformation may occur when primary 3.16 stress exceeds 1.5 *to 2.0 times the yield strength CSY) of piping material, 3.17 I. QUt for stresses below these values, functional capability was maintained. 3.18

[or Surry Power Station, Unit 1, the majority of carbon steel piping material 3.19 I is of SA-101 ~rade B steel. ~sing the lower limit of 1.5Sy from NUREG/CR- 3.21 I 0261 and representative properties of SA-106 ~rade B steel, the added margin of conservatism is the ratio Cl.5Sy/l.8S~), which ranges from 1.4 at 650°F to 3.22 I 1. 94 at 100°F.

1* Ihe Surry 1 reanalysis calculations have included the seismic stress due to 3.27 I snchor displacements in the DBE condition.

stresses was not Inclusion of the anchor movement required by ANSI Code B31.1, used for the original design, 3.29 4.1 I and is not required by current 1979 codes, for the faulted Addition of this stress component is~ significant conservatism.

DBE condition.

4.2

• E.C. Rodabaugh and S.E. :t-foore, "Evaluation of the Plastic Characteristics of Piping Products in Relation to Code Criteria," NUREG/CR-0261, July 1978 I 7-4 I

I

I h-1284622-60 06105/79 042 I SURRY POWER STATION, UNIT 1 I

I l.2 SYSTEM REDUNDANCIES 4.4

-1 All systems essential for shutdown or for mitigation of a design basis 4.5

-, accident CDBA) contain Iedundant flow redundant paths and driving equipment.

paths and equipment are analyzed as separate problems with complete Ihese 4.8 I input yerification, engineering analysis, and engineering review.

a common mode problem is precluded.

Iherefore, tven identically designed redundant 4.10 4.11 I systems may not always experience similar seismic excitation due to mounting locations Hith structural filtering effects.

different Ihus, even a postulated 4.13 I loss of a redundant component will not mean a loss of function for the system.

,- l.3 SAFETY SYSTEMS 4.15 1

All safety systems, not just a sampling or a portion of redundant systems, 4.16 which were originally seismically analyzed with the ~HOCK2 program have been 4.17 included in this reanalysis. Ihe systems involving the check valves whose 4.18 1 weight has been revised (see Appendix D) have also been completed. Iherefore 4.19 pipe stress reanalysis has been completed for the systems in this reanalysis I program that interface with the Ieactor coolant system. Ihe auxiliary feed 4.21

-I system and the steam supply to the steam driven auxilary feed pump have also been reanalyzed.

I, I

I 7-5 I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I Ihe new weights for safety injection accumulator discharge motor-operated valves supplied QY Velan have ben reviewed. Ihe problems were rerun with 4.23 4.25 I higher weights and no additional changes to stresses resulted.

I l,4 FIELD VERIFICATION OF AS-BUILT CONDITIONS 4.27 I Io ensure that the pipe stress reanalysis is performed as accurately as 4.28 I ROssible, field verification of as-built conditions has been field Verification produced detailed piping performed.

isometric drawings upon which Ihe 5.2 I reanalysis is based. Ihis confirmation of input data provides assurance that 5.3 analytical results are correct. All field-verified piping isometrics are 5.4 I independently verified by Surry Power Station quality £Ontrol personnel. 5.5 I 1icensee Event Reports were discrepancies a~e being corrected.

filed when discrepancies were found and these iee Section 8.

5.6 5.7 I l,5 QUALITY ASSURANCE/ENGINEERING ASSURANCE 5.9 I

A comprehensive and extensive Quality Assurance program has been developed ~nd 5.11 I applied to the reanalysis activities. Ihe unusual nature of the evaluation 5.13

.1 and analysis program required the development of new ~echnical .procedures for use in performing evaluation and analysis work. A detailed project procedure 5.14 5.15 I was developed that includes provisions for design control, document £Ontrol, 5.16 I

7-6 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I and interface controls. tach new procedure developed received a full review 5.17 by the S&W Engineering Assurance CEA) staff; ~roject procedures required 5.18 approval by EA.

-, Ihe normal Quality Assurance auditing activity has been extensively expanded 5.19 to provide the highest possible degree of confidence in the reanalysis work. 5.20 I Instead of auditing on a sampling basis, each pipe stress/pipe support problem 5.21 I package is ~ubjected to a detailed EA inspection according inspection plan. Ihe EA to a written inspection plan is sufficiently broad in scope to EA 5.22 5.23 include all significant technical attributes in addition to the usual 5.24 programmatic attributes inspected in audit plans. tach problem package is 5.26 inspected at the completion of the stress evaluation and again At the 5.27 completion of support/nozzle/equipment evaluations. Ihis type of inspection 5.28 I activity, by increasing the confidence which can be placed on task group I QUtput, is considered to confirm the high degree of conservatism inherent in the engineering work.

5.29

'I

• ..../

In addition to the stress and support package inspections, EA has confirmed 6.1 I the quality of tey input information used in the stress/support evaluation. 6.2

-I. Ihe as-built. documentation effort has been extensively audited by S&W EA and VEPCO QC. tach as-built document created was subjected to this joint S&W 6.3 6.4 I EAIVEPCO QC inspection, £Onducted at the p~ant site, to confirm the actual as- 6.5 I

7-7 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I built configuration.

subjected to a

[inally, comprehensive, the development technical audit.

of the SSI-ARS ~ill Ihe effort to confirm the be 6.7 6.8 I accuracy of the as-built configuration and the amplified response spectra gdds 6.9 additional confidence to the quality of the reanalysis effort.

I I l,6 USE OF AMPLIFIED RESPONSE SPECTRA 6 .11 I Ihe peaks variation in in amplified material floor response spectra are broadened to account for properties and app roxim.ations in modeling. .E,eak 6.13 6.15 I broadening is intended to reflect a range of uncertainty in the precise location Qf the resonant peak of the response curve and not to indicate that 6.16 I multiple peak resonant response is likely within the broadened range. Rhat, 6.18 I in fact, exists is a "family" of resonant curves, each having only of maximum Iesonant response.

one point 6.19 I It would be more precise to analyze systems and components for a number of 6.20

'I unbroadened spectra which sre members of the broadened family of possible 6.21 amplified response spectra. iince there can only be one single peak frequency 6.22 I for a given system, the use of peak broadened floor Iesponse spectra as 6.23

,I practiced in* seismic design is a conservative analytical gxpediency that 6.24 results in an additional margin of safety for systems, components, and I supports.

I I 7-8 I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I In addition, when a piping system has multiple supports, the system is 6.25 analyzed with a response spectrum that envelops the spectrum at each support 6.26 that results in an additional margin of safety for systems, components, and 6.27

-, supports.

fince pipe runs generally extend over a range of elevations from beginning to 6.28 I end of run, and since the magnitude Qf acceleration associated with each ARS 6.29 I increases with elevations in a structure, the ARS applied in the analysis of each run is ~elected coincident with the higher elevations along run.

each piping 7.1 I 7.7 DEVELOPMENT OF SOIL-STRUCTURE INTERACTION AMPLIFIED RESPONSE 7.4 I SPECTRA CSSI-ARS) 7.5 I Reevaluation of piping systems for induced earthquake loading has employed conservatisms in the development of ARS based on iSI and the application of 7.8 7.11 I resulting loads to the pipe stress analysis. Ihese conservatisms involve the 7.13

• methodology of developing ARS, the range in soil £roperty variations in 7.14 I developing amplified spectra, the application of particular ARS to each piping 7.15

• I run, the effects of three dimensional input, and an increase (bumping) in inertia forces applied to each pipe run after computer calculation of stress the 7.16 I and support loads.

I 7-9 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I Ihe methodology used in ARS based on SSI is based upon a layered elastic media model for soil and a lumped mass model for the structure. Analysis using 7.18 7.19 I these models involves Ca) the calculation of frequency-dependent ~tiffness at 7.20 the surface of a layered medium using the program REFUND, lb) modification of 7.21 I a specified surface motion to account for embeclment of the structure, Cc) the 7.22 I application of kinematic interaction principles to specified at the modify translation surface to both a translation and Iotational motion at the input 7.23 7.24 I base of the rigid structure foundation analysis_ of the structural model using the program KINACT, and supported on frequency-dependent springs (cl) 7.25 7.26 I using the program FRIDAY. Ihe resulting ARS developed fro~ this methodology 7.27 were compared ~ith ARS developed using a detailed finite element 7.28 I representation of the gnderlying soil medium with a lumped mass representation 7.29 I of the ~ontainment structure using the program PLAXLY.

of acceleration computed using the REFUND/KINACT/FRIDAY method Ihe amplified values

£re generally 8.2 8.3 I 30 to approach.

100 percent larger than values computed using the more rigorous PLAXLY 8.4 I

~ariations in soil properties have generally been accounted for by developing 8.5 I ARS Qsing mean values of soil moduli and damping ratio values adjusted for 8.6

• 1 strain ARS.

levels £ssociatecl with earthquakes, ancl peak spreading the resulting 8.7 I

I 7-10 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I iubstantial conservatism is also gained by the application of a 20 percent increase in the* seismic inertia forces acting on each piping §.ystem above 8.9 8 .10

• 1 those forces computed using the ARS developed from the REFUND/KINACT/FRIDAY 8.11 approach. Ihe 50 percent increase in inertial seismic forces is a requirement 8.12 I of the NRC letter of May 25, 1979. Ihe 50 percent increase accounts for a 8.13 I wide range of possible input parameters and, in fact, adds again ~onservatisms already_ accounted for by using peak spreading techniques.

8.14 I iubstantial elaboration ~f the techniques described in this section on the use 8.15

• 1 of soil structure interaction_ techniques will be gocumented in a subsequent 8.16 submittal to the NRC staff on or about June 8, 1979.

I I 1,8 SEISMIC EVENT PROBABILITY 8.18 I Ihe seismic hazard at the Surry Power Station site is small. ~alculations incorporating the effects of potential seismic zones far from ihe site as well 8.20 8.22 I as random local seismicity indicate that the annual risk of equaling or exceeding ihe DBE of 0.15 g is about 1.1 x 10-s, which corresponds to a hazard 8.23 I of 4.5 x 10-~ over the 40 year life of the plant. 8.24

-I ~uring any one month, there is a hazard of 4.5 x 10-~ of equaling or exceeding 8.25 I an garthquake with a peak acceleration of 0.04 g. Ihus, the chances are very 8.27 I

7-11 I

I

I h-1284622-60 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I slight that the plant will experience any significant shaking due 10 an 8.28 earthquake during the period that systems are being checked and any corrective

• 1 action taken. Ihe chances of experiencing the DBE are extremely small. 8. 29 I Ihe calculations leading to these conclusions use methods developed by 9.1 I Prof. C. Allin Geological Cornell Survey. Ihe Qf M.I.T. and Dr. Robin K. McGuire of the seismicity data are derived from historical records U.S. 9.2 9.3 I and are developed into seismic zones QY the methods of Dr. Edward Chiburis of 9.4 Weston Observatory. Ihe analytical techniques - explicitly account for 9,5

• 1 randomness of occurrence of earthquakes in space and time and for uncertainty 9.6*

in attenuation relations.

I I

I I

I

-I I

I I 7-12 I

I h-1284622-66 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I SECTION 8 1.10

• 1 LICENSEE EVENT REPORTS AND RESOLUTIONS 1.12 I Ihe pipe supports addressed in Licensee Event Report LER79-0l0/03L-0 will be 1.15 I shimmed or restored to meet the requirements of the stress analyses on the yarious systems.

performed final as-built verifications will be performed to 1.16 1.18 I demonstrate correspondence Qf the stress analysis models with hardware and 1.19 system arrangements prior to startup.

I Licensee Event Report LER79-004/01T-O describes the £Omputer code 1.21 I nonconservatism which is being restored by this report. Ihe variance in valve 1.22 I weights from those originally given by the valve vendor are being included in this effort also. £These valve weights are also the subject of IE Bulleti~

1.23 1.24 I 79-04.)

I I

• I I

I I 8-1 I

I hl284622-lh 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I

• 1 \

I I APPENDIX A 1. 9 I SYSTEMS AFFECTED 1.11 I

I I

I I

I

--1 I

I I

I

I h1284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Ihe reanalysis included those safety-related lines originally computer- 1.12 I analyzed with the SHOCK2 program. Ihe systems line numbers, ~he associated computer problem numbers, and the flow diagram numbers are listed below.

1.14 I Ihe figure numbers refer to the FSAR drawings, and the Surry, Unit 1, FM ~nd FB drawings included in Appendix B.

1.16

• 1 System Line No.

Problem No. F.igure No.

1. 26

1. 27 Pressurizer 4".:.RC-15-1502 636 4.2.1-1 1.29 I Spray & Relief 4!!-RC-14-1502 12"-RC-36-602 630 4.2.1-2 4.2.1-2 1.30

1. 32 6"-RC-39-1502 1.33 I 6 11 -RC-42-602 6"-RC-38-1502 1.34 1.35 6"-RC-41-602 1.36 I 6"-RC-37-1502 6"-RC-40-602 4"-RC-34-1502 1.37 1.38 1.39

• 1 3 11 -RC-35-1502 6"-RC-20-602 3"-RC-61-1502 1.40 1.41 1.42 6 11 -RC-62-602 1. 4.3 I Low Head Safety Injection

. 12"-RC-23-1502 12"-SI-46-1502 555 4.2.1-1 6.2.2.1-2 1.45 1.46 I 12"-RC-22-1502 12"-SI-45-1502 1555 4.2.1-1 6.2.2.1-2 1.48 1.49 I 6"-RC-16-1502 6"-SI-49-1502 706A 4.2.1-1 6.2.2.1-2 1.51 1.52 6"-RC-21-1502 707A 4.2.1-1 1. 54 1* 6 11 -SI-50-1502 6.2.2.1-2 1. 55 6"-RC--18-1502 708 4.2.1-1 1. 57 I 6"-SI-48-1502 6"-SI-143-1502 6"-SI-49-1502 6.2.2.1-2 1. 58 2.1 2.2

• I 6"-SI-50-1502 8 11 -SI-92-153 731A,B 6.2.2.1-1 2.3 2.5 8"-SI-14-153 2.6 I 10"-SI-6-153 743 6.2.2.1-1 2.8 I

I A-1 I

I hl284622-lb 06/05179 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No. No.

1000A Figure No, 4.2.1-1 2.11 6 11 -RC-17-1502 I 2"-SI-74-1502 6"-SI-145-1502 6.2.2.1-2 2.12 2.13 2.16

• 1 6"-SI-48-1502 6"-SI-49-1502 6"-SI-153-1502 727 6.2.2.1-1 2.17 2.18

-.-. 1Oi'-SI-148-153 2.19 I 10"-SI-149-153 10"-SI-150-153 lP"-SI-151-153 2.20 2.21 2.22 10"-SI-16-153 2.23 I 10"-SI-13-153 2.24

• 2. 26 6"-RC-19-1502 1010A 4.2.1-1 I 2"-SI-85-1502 6"-SI-144-1502 6.2.2.1-2 2.27 2.28 6"-RC-20-1502 1020A,B,C 4.2.1-1 2.30 I 2"-SI-75-1502 6"-SI-153-1502 6.2.2.1-2 2.31 2.32 6"-SI-15-1502 2.33 I High Head Safety 6 11 -SI-145-1502 8"-SI-7-152 735 6.2.2.1-1 2.34 2.36 Injection 8"-SI-102-152 2.37 I 10"-SI-6-153 8 11 -sr..:.17-152 2.38 2.39 6"-SI-18-152 2.40 I 6"-SI-19-152 6"-SI-78-152 6 11 -CH-201-152 2.41 2.42 2.43 2.44 I 6 11 -CH-202-152 6"-CH-203-152 8 11 -CH-204-152 2.45 2.46 8"-CH-206-152 2.47 I 8"-CH-17-152 6 11 -CH-72-152 6"-CH-18-152 2.48 2.49 2.50

• I 6"-CH-19-152 4"-CH-112-152 2.51 2.52 Containment and 8"-CS-23-153 525A FM-101A 2.54 I Recirculation Spray 8"-CS-22-153 10"-CS-3-153 2.55 2.56 2.57 10"-CS-4-153 I 10"-CS-4-153 547 FM-101A 3.1 I A-2 I

I hl284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No.

8"-CS-34-153 No. Figure No.

3.2 I 8"-RS-21-153 10"-RS-4-153 546/560 FM-101A 3.4 3.5

• 1 8 11 -CS-34-153 744(7 54) FM-101A . 3. 7 3.8 3.9 I 10"-CS-3-153 8"-CS-33-153 548A,B FM-101A 3.11 3.12 10"-RS-3-153 3.13 I 10"-RS-10-153 544, 544A,B FM-101A 3.15 10"-RS-9-153 3.16 I 10"-RS-1-153 10"-RS-2-153 3.17 3 .18 10"-RS-10-153 751 FM-lOlA 3.20 I 10 11 -RS-9-153 3.21 10"-RS-12-153 562 FM-101A 3.23 I 8 11 -RS-23-153 8"-CS-33-153 745 FM-101A 3.24 3.26 10"-RS-3-153 546/5600 FM-101A 3.27 I 8"-RS-20-153 3.28 10"-RS-11-153 546/5620 FM-lOlA 3.30 I 8"-RS-22-153 10"-RS-11-153 548C FM-101A 3.31 3.33

'I Residual 4"-RC-197-153 540 4.2.1-2 3.37 Heat 4"-RH-15-152 9.3-1 3.38 I Removal 12"-RH-19-602 10"-RH-16-1502 3"-RH-13-602 3.39 3.41 3.42

-I 6"-RH-14-602 10"-RH-'23-602 10"-RH-17-1502 3.43 3.44 3.45 I 14"-RH-1-1502 14"-RH-18-602 14"-RH-2-602 508 9.3-1 4.2.1-1 3.47 3.48 3.49 I 10"-RH-4-602 10"-RH-5-602 3.50 3.51 I A-3 I

I hl284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No.

12"-RH-6-602 No. Figure No.

3.52 I 12"-RH-12-602 10"-RH-8-602 10"-RH-9-602 3.53 3.54 3.55

• 1 12 11 -RH-19-602 10"-RH-7-602 10 11 -RH-10-602 3.56 3.57 3.58 I Component Cooling Water 18 11 -cc-227-121 18"-CC-228-121 18 11 -cc-229-121 488/480 9.4-4 4.3 4.4 4.5 18 11 -CC-230-121 4.6 I 18"-CC-5-121 18 11 -cc-220-121 4.7 4.8 4.9 18 1 '-CC-6-121 I 18"-CC-237-121 18"-CC-236-121 507/481 9.4-4 4.12 4.13 18 11 -cc-235-121 4.14 I 8 11 -CC-311-151 18 11 -CC-225-121 4.15 4.16 18 11 -CC-232-121 4.17 I 18 11 -CC-233-121 18"-CC-234-121 18"-CC-231-121 4.18 4.19 4.20 I 18"-CC-7-121 614 9.4-4 9.4-3 4.22 4.23 12 11 -cc-27-121 9.4-6 4.24 I ..

• 18 -CC-16-121 11 1512 9.4-1 9.4-4 4.26 4.27 I 18"-CC-15-121 512 9.4-1 9.4-4 4.29

4. 30 .

9.4-6 4.31 I 18 11 -CC-16-121 18 11 -cc-8-121 603A,B 9.4-1 4.33 4.34

-1 e11 -cc-69-151 8 11 -cc-67-151 766 9.4-3 4.36 4.37 I

1*-

I A-4 I

I h1284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No.

8 11 -CC-61-151 No.

527A,B,D Figure No.

9.4-3 4.40 I 411 -cc-61-1s1 6 11 -cc-62-1s1 8 11 -CC-75-131 4.41 4.42 4.43

-,- 6 11 -cc-81-121 10 11 -cc-81-121 14 11 -cc-67-121 4.44 4.45 4.46 18 11 -CC-10-121 4.47 I 18 11 -cc-10-121 6 11 -cc-105-151 605A,B 9.4-1 4.50 4.51 3* 1 -cc-107-151 4.52 I 3 11 -cc-112-1s1 6"-CC-207-151_

4.53 4.54 18"-CC-17-121 4.55 I 4"-CC-108-151 4tt-CC-113-151 4.56 4.57 18 11 -CC-8-121 509A,B,C,D 9.4-1 5.1 I 18 11 -CC-237-121 18"-CC-236-121 9.4-2 9.4-4 5.2 5.3 18 11 -CC-235-121 5.4 I 8 "-CC-311-151 8"-CC-32-151 6"-CC-32-151 5.5 5.6 5.7 4"-CC-32-151 5.8 I 1 l/2"-CC-30-151 6"-CC-286-151 5.9 5 .10 24 11 -cc-235-121 5.11 I 18"-CC-226-121 18 11 -cc-10-121 12"-CC-27-121 5.12 5.13 5.14 24"-CC-226-121 5.15 I 18"-CC-9-121 10 11 -cc-89-151 5.16 5.17 6 11 -CC-89-151 5.18 I 6"-CC-97-151 18"-CC-14-121 612 9.4-4 5.19 5.21

• I 14 1*-cc-n-121 10"-CC-72-121 9.4-3 5.22 5.23 5.24 4 11 -cc-148-151 5.25 I 6tt-CC-146-151 18 11 -CC-14-121 9.4-3 5.26 5.28 517 I

14 11 -cc-n-121 14"-CC-70-121 9.4-5 9.4-6 5.29 5.30 I A-5 I

I hl284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No.

• No. Figure No.

5.31 18"-CC-70-151 I 4"-CC-66-151 8"-CC-66-151 6"-CC-64-151 5.32 5.33 5.34 5.35

• 1 10"-CC-72-121 18"-CC-17-151 8"-CC-70-151 5.36 5.37 a11 .:.cc-71-1s1 5.38 I . :*:.:.:..~.-- ,.. . 8"-CC-78-151 18"-CC-17-121 526A,B,C 9.4-2 5.41 5.42 6"-CC-85-151 5.43 I 6 11 -cc:..93-151 10 11 -CC-104-121 5.44 5.45 6"-CC-101-151 5.46 I 4"-CC-144-151 6"-CC-145-151 2527 9.4-5 5.49 5.50 8"-CC-144-151 5.51 I 10"-CC-143-151 5.52 611 -CC-181-151 2526 9.4-6 5.54 I 2 l/2 11 -CC-184-151 6 11 -CC-173-151 10"-CC-181-151 5.55 5.56 5.57 6"-CC-165-151 5.58 I 8 11 -CC-156-151 18"-CC-14-121 6.1 6.2 I 3"-CC-127-151 14"-CC-143.:..121 18"-CC-7-121 2529 9.4-3 9.4-5 6.4 6.5 6.6 6.7 I 10"-CC-161-121 6"-CC-161-151 8"-CC-153-151 6.8 6.9 I Service Water 24"-WS-33-10 24"-WS-34-10 24 11 -WS-35-10 1030 9.9-1 6.12 6.13 6 .14

• I 24"-WS-36-10 24"-WS-26-10 465 9.9-1 6.15 6.18 24"-WS-28-10 6.19 1-- 24"-WS-30-10 24"-WS-32-10 6.20 6.21 I

Main Steam 30"-SHP-1-601 323A 10.3-1 6.23 I A-6 I

I hl284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Problem I System Line No. No. Figure No.

30"-SHP-2-601 322A 10.3-1 6.25 I 30"-SHP-3-601 334A 10.3-1 6.27

• 1* 30 11 -SHP-1-601 30"-SHP-2-601 346 10.3-1 6.29 6.30 30'.'-SHP-3-601 6.31 I High Pressure Steam 411 -SHP-25-601 4"-SHP-26-601 4"-SHP-27-601 746 10.3-1 6.34 6.35 6.36 3"-SHP-28-601 6.37 I 3"-SHP-29-601 3"-SHP-30-601 6.38 6.39 3"-SHP-32-601 6.40 I 3"-SHP-33-601 3"-SHP-34-601 3"-SHP-35-601 6.41 6.42 6.43 I Feedwater 14"-WFPD-17-601 323B 10.3.5-2 6. 47 14"-WFPD-13-601 322B 10.3.5-2 6.49 I 14i'-WFPD-9-601 334B 10.3.5-2 6.51 I Auxiliary Feedwater 3"-WAPD-9-601 3"-WAPD-11-601 3"-WAPD-13-601 417 10.3.5-2 6.54 6.55 6.56 3"-WAPD-10-601 6.57 I 3"-WAPD-12-601 3"-WAPD-14-601 6.58 7,1 6"-WAPD-1-601 7.2 I 6"-WAPD-2-601 6"-WAPD-1-601, 607 10.3.5-2 7.3 7.10 6"-WAPD-2-601 7 .11 I 6"-WAPD-3-601 6"-WAPD-4-601 7.12 7.13 4"-WAPD-5-601 7.14

• I 4"-WAPD-6-601 4-WAPD-7-601 4"-WAPD-8-601 7.15 7.16 7.17 I Containment---

Vacuum 8"-CV-8-151 CV-1 FM-102A 7.19 7.20 I

Fire Protection See Figure See Figure CF-1 CF-2 FB-3D 7.22 7.23 I A-7 I

I h1284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 Diesel Muffler See Figure 1040 FB-25L 7.25 I Exhaust In addition to the SHOCK2 computer calculations, the above list includes 7.26 7.30 I five hand calculations which were reanalyzed because of incorrect VELAN check valve weights ~IE Bulletin No. 79-04).

7.31 7.32

• 1 H6nsafety-related SHOCK2 problems were protection and diesel muffler exhaust.

not reanalyzed, gxcept for the fire 7.34 Ihe following is a listing of safety-related lines that were analyzed with the 7.35 I ~HOCKO/SHOCKl programs. Ihese problems are also identified on the flow diagrams in Appendix B.

7.37 Problem 7.40 I System Low Head Safety Line No.

12"-SI-47-1502 537 No. Figure No~

4.2.1-1 7.41 7.44 I Injection Containment and 12"-RC-24-1502 12"-CS-1-153 755.

6.2.2.1-2 FM-101A 7.45 7.49 Recirculation 12n-cs-2-153 756 FM-101A 7.50 I Spray 7.51 Residual 6"-RH-20-152 554 9.3-1 7.55 I Heat Removal 7.56 7.57 Component 12TT-CC-33-121 606 9.4-3 8.3 I Cooling Water 12"-CC-34-121 8.4 8.5 I 4"-CC-37-151 311 -CC-38-151 6"-CC-37-151 613 9.4-4 8.8 8.9 8.10 6"-CC-287-151 8.11 I 24 & 18 11 -CC-225-121 481 9.4-4 8.13 18"-CC-232-121 8.14 I 24 & 18 11 -CC-233-121 18"-CC-234-121 8.15

.8 .16

-I 18"-CC-1-121 18"-CC-2-121 18"-CC-3-121 502 9.4-4 8.18 8.19 8.20 18"-CC-4-121 8.21 I 18 11 -CC-223-121 l 8 11 :...cc-227-12 C 18"-CC-228-121 8.22 8.23 8.24 I *18"-CC-2 29-121 8 11 -CC-314-151 8.25 8.26 I A-8 I

I hl284622-lb 06/05/79 042 I SURRY POWER STATION, UNIT 1 8.27 I 8"-CC-287-151 18"-CC-14-121 506 9.4-4 8.37 18"-CC-15-121 8.38 I 18"-CC-16-121 18"-CC-17-121 18"-CC-19-121 8.39

8. 40

• 8.41

• 1 6 11 -CC-20-151 6"-:-CC-:22-151 539 9.4-4 8.43 8.44

• 1 6"-:-CC-222-151 6 11 -CC-:21-151 8.45 8.46 Fuel Pit 12"-FP-4-152 747 9.5-1 8.49 8.50 I Cooling 12 11 -FP-5-152 12"-FP-3-152 8.51 12"-FP-32-152 748 9.5-1 8.54 I 12"-FP-33-152 16"-FP-18-152 8.55 8.56 I 12"-FP-1-152 12"-FP-2-152 749 9.5-l 8.58 9.1 Auxiliary 6"-WCMU-5-151 611 10.3.5-2 9.4 I Feedwater 6 11 -WCMU-6-151 6 11 -WCMU-7-151 9.5 9.6 411 -WCH:U-9-151 9.7 I 4"-WCMU-10-151 6 11 -WCH:U-11-151 6 11 -WCMU-8-151 9.8 9.9 9.10 I

I I

• I I

I I A-9 I

I hl284622-li 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I

-1:

I*

I I APPENDIX B 1.10 FLOH DIAGRAMS - 1.12 I IDENTIFICATION OF PROBLEMS REANALYZED 1.13 1.14 I

I I

I I

• I I*

I

• 1 I

I hl284622-ll 06/05/79 041 I SURRY POWER STATION, UNIT 1 FLOW DIAGRAMS 1.10 I 1.13 Figure No.

I Reactor Coolant - Sheet 1 Reactor Coolant - Sheet 2 4.2.1-1 4.2.1-2 1.16 1.17

• 1 Safety Injection - Sheet 1 Safety Injection - Sheet 2 Chemical and Volume Control 6.2.2.1-1 6.2.2.1-2 9.1-2 1.18 1.19

1. 20 Containment & Recirculation Spray 11448-FM-lOlA 1. 21 I Residual Heat Removal Component Cooling - Sheet 1 Component Cooling - Sheet 2 9.3-1 9.4-1 9.4-2

1. 22

1. 23 1.24 Component Cooling - Sheet 3 9.4-3 1. 25 I Component Cooling - Sheet 4 Component Cooling - Sheet 5 9.4-4 9.4-5

1. 26

1. 27 Component Cooling Sheet 6 9.4-6 1. 28 I Fuel Pit Cooling Service Water Main Steam 9.5-1 9.9-1 10.3-1

1. 29 1.30

1. 31

• 1 Feedwater Containment Vacuum Fire Protection 10.3.5-2 11448-FM-102A 11448-FB-3D

1. 32 1.33

1. 34 Diesel Muffler Exh~ust 11448-FB-251 1.35 I

I*

I I

I I

I

• -1 I

I I B-1 I

I FIG. 4.2.1-1 OCT.15,1970

'v'tOl-~.:1-8171711 FT FT I H~C-P*IA

\

g

• S!>*IO'?>*I-C*"19

(!'M-'3?.'S) 1'~B

-~*TS&

CL*r*C.*119 IC.L-1'&0?.

PT 1403 PT

/

./

t*SS*IOH*C*ll9 (fM-~2111

,2~RC-S8*l50Z DET. C (FM*I0.5&)

1*11.C*P* IC l>EACT02 PR.ES..S ~ E ~ 11.E-"C.oR.

COCLAIJT PUMP

roe _J ,E (F~~1ii1s) f *T55- COOL.-.NT PUMP 1401

'3'=-ec~104-15o?..

NOTE 2i ~- ,.-* ©

.I

'-'11.EO TO Fr'r \ltfLL F.O l!E/0'011.w.'lSEL f*RC-32-I*C*N9 FLANGI: L~i,.K OETf.CTlON LIIIE.

I FROM OVTE FROM IN>a.11.

CHAHBl!2 C:HA!1SER.

I I

I f*RC *l!I0* 15:>2,_/ Lf*D&-65*1Sl

(~M-1009)

I z*-r.. .-~ -1111t

a. FT

!"*(Pllf*R8J I

CL 1502., ,t.L 152.

CM, U* IIUl I l*RC.-P*IB 14z.+

12.EACTOli!.

VE55E.L (l'olOTE IO) 1-Tsa...!

p21c,; 1"0 ~.JEUNG- OR.

LOOP MA.l=ANCE.,

NOTES:

GENERAL NCTE.9

• REFEIUNCI. DWGe. FM-\OJII R.f.i,.CT02 CON>JECTIO>J FOi! ,O.TTi,.CHII<<', PLAf>TIC I

COOLANT PUMP MOSE TO 111.INO FLi,.NGE Im VE.NT HEi,.DER. (.1~-VA*8*1S4-)(1'1"1*I008)

%°RC*I07*1:*t.-N9 I 1*-11.c-1'!102.- 't"*li!.C-15D'l.

I I

I ~--c.~-,-1so2 12.0* _./

1*-2c-1soz_l 1 I A-A I

C.t-lAll&IUC, LINE

\

'---..:.......1<:'>'----;*- (Fl-\*106C) f oG-C.*152 REACTOR COOLANT (F'M*IOOe) 2*-1<e-&!Hso2 ~ * -Cn~:,tc.,e.)

SYSTEM SHEET I.

L------------.-------'---------*---------------------------' \_l2."*Sl*46*IS02 SURRY POWER STATION

• -Sfk)CKf I

I FIG. 4.2.1-2 OCT.15, 1970 I 9E'OHI\J.:l-9Vt'II MAJl~.J~ £LEV~iTION 15 rT A.IOVt I

PUl!P CONNECT10M, (TYP.)

1*-ttc--.-reL-l\.7FTdltJ),*CWE C.Of<N£C'.Tf0f1 TO

~

--.._y I-RC,9-15"2

~

l~ZU.

1"-0IC'.* llll*flil EL 7 FT NIN ABOVE V ..!!£L_

,,zzc.

f:R.c'":'se

1. 7' 2.

1/za, 2'1- -s n* ana tc mm ~ua:Tlll.

.U. c.&SS 2,0U PIPIJG lm'ft.llD n lll.'ltmala& IOI m:!All.1 Ill.: us:. Gb76J.U, a6'76)42 &Ill G6'76)4J. AU. OIUI. CMS11 ..

PUDO ID'IC. lrm-20.

J. .i.i. 11m FaaSau 1'D13 10 u IUIUD on 11m 1,,ao LI IP PVOWP RO J*YSIIN JlW &m ~ JXJaDQ s,aw. onu,m1.

AR~*._,:J' ""'11!.t..!_!'::'!.A'r_\~1.L\I'[!, Will IIA\'t A wr.*tff'Mli.l.

mc~ *

s. ca.ou Y.&Lns 1t111W.Lt mr.w.a vmr n.ov tall Sid'

6. ~;~~**:~m~*.,~~~~~~~*Pv~~~~t~J.j ~:R*'4c, I 2"*RC*65*1Se -

STANDPIPE O!!IFltE

- j.

<o'*R.C-l'l0-152"'}- .*"

7. SCUOOLI 140 PD'I, I, l'K>llm IITIISlDW HIN l'CI .ID1tlSIKlft ZJU1llm SIIIIIL OJIIDl&Dla, 9, BDQl:ll.S - Pll'DG an:cIJ'ICltlDIS,

10. ACCl.1.$ l'CR DSPICU)I or Ule!OI aJD1.lft PIIS m ~ talS, Sll.ZU WILDS '!O U ftllYIDID, 11, n.tt a>IUCTit:IS JO& UJOr S'IDP YAl,ff ANAL'"* 15

12. ILlCI tialC'!IJl:.9 il !l1flDN OJ' rn.

' *~ *

• SWllmPE j COMPONE.NT l). samu.& 120 PUS, u.

I COOLING o*.as* KJU fflK)g_ HLH JII.UO, f"-R.C-1'2-152----._;:

(F'M*Z'i'A) 1,. IDUn B1l£1W IZ.nillDI or U&C!OI lD.1& . . . . . .

1*-P.c-ea.ise ---< ::._-

1'1. wean (l)DICfJDI °

16. U>Cil. a>U&"rJDI JODT Kl& DnliMDDQ L&AUlS ldL IIU>V llll.lml'UL cana J.lll.

11, SlT Lai B!Pj,S.j LDI SQ)OJIS - &LL 1DC&!ID

• 1&111 ~

2"-RC*71l*ISe 2'-~c-e2-,s2 """' 1UD (SIi !ICl'lDI .A-i),

19. lftl N.IIDOLD &ID MD'* 8tlffl.DD &S P&mA 9,V 'o",f',Tl[JeNOUSE(IMNWOI.Q Ul'WllDllm.t 217 LC!C).

2-l'<C-73-152 *20,* ~c:.&!I Y&l.TE &Pl'llllDIATILI 9'I , _ N&IDaLII I

21. LOcan CIJIIIDl'lDI 01 ma 1IDD or rm cncmnaaca

.,~.

22. lDCd'I TAI.ft AIDYI CIITII Lm IUT.Ulll a, IIACD Tai&

23, 1Q!IL ~ OP mt ~ 2" Sa:Tlll anrQIJI OP lftl 1111Dm.11 2-Jl'4'20 i '!O I 'ID U i Ml!MtM OP 4'-6 1/ZI",

BD NAIDOU. SiU,U. IAYI UCIAIU JDIUfl>I mo m Ml:IP IDDT 24, 1DT'1. I.DOiii OP CDLD LS, SEC'f'Tnl UPSTMAK OP IITD IIAIIPOU, C 10 D 10 a i Jl&lnrml OP 6 1 -6",

ALL 9TPAS6 LOOP PDDIO AID I

t'"*DG.*14

• 15? 31'-RC*Z *Z'501R 2*-oc;..11-1s2. 2.'-D<.*16-152. 2', 1!11 DftW.I!> II m.L (FM-100'1) (FM*IOllA) ~-bG-l'Z.~*162. (F.. *1008) (l'M*IOO*) 26. P.&l&I.Ul. PIP& PATB:S smu a AJll'mllll&!&t cmaar UIQ!II.

uv ,m LaJTlil or uca 1* PD'DD l&'flll au IDI' llCIID (F'"*IOO!,) '\~o*.

f*RC*i;l!l-t52.

27. PROVIOC WAT[R SEAL fOR RELIEF' ,V.O.LVE~

ZR J.DCAT[ CRUICE IN BYPASS liETJJAN LINI:: WITH MINIHUM 0, 20 PIPE D/,IIUTEP.~ Of' UPSTP.EA.M STRAIGHT RUN.

29. 'fUNG.U 1~1.UI F:OR INSERTION DP' F.LOW LIMITING OP.l~ICES I~ IW2111P.ED. BLANK ORIFICE PLI\TE WILL IIE I '-IZ'*RC-36*602 l*T'S&

V-i (TYP) 30. : ~ ' t ~ : S f ' ~ B ~ \ ~ R ~ FROM LOOP ;£.t.l'i TO LEGEi-iD CONNECT WITH COMl.tOH AtLltF UNE TO P911E!tSURl2.11l IU\.IUTAN<<.. .

A..C.- INDICAT!DWJCJAL \IU.IIJl'OSJTIIDll"IArNTi\JN!D IV ADMIWISTMTIW! .

CONTROL. wHfgl I! cca.smtRl!!J [OUIVAL.ENT TO A LDC.lllD VAl,.\.'I I W&T[R SEAL ARllAN6[NEfrlT

~: I~~} MISSILE. 9"RRIER AND/Oil SfCDIIII...V fO 0

V j

.,,.K._

FC - ~IL CLOS!' l

- LOC.A.L D, \Ill

• LOC.A.\... v* IT

- TRIPPl:D ffl $A.F£TY INJE.c.TION ~~

SHIELD

~ - DJ.t.!l'HRA.li,M VALVE WITH Ate OPlltATOft I*

AUX SPRAY l)j

foCH5*1!1Z (Fl"4*100B)

I *~

Un I L2--DG-12-152.

TE 1452 I 4'*RC*l5*1502 <f-RC-14*1502. - IZ'*RC *IO*Z'SOIR.(NOT£ 1)

I CFM-IO!A) (F1,t~03A)

I REACTOR COOLANT SYSTEM - SHEET 2.

I SURRY POWER STATION I

I FIG. 6.2.2.1-1 FEB. 11, 1972 I

, 51-';Co=~---iv? *TS8 l*T58

~------<Fs-Q/1)

TO FLOOI\ DRAIN

,-:__ cr-f*91*SZ*IU

, -~- f-....-1~-lliZ. ..0

,,... 2,11}

PT 1-IJ.+ " ' ,

f *T5!1 F"ROM BOR1t Atll) TR.*N! . ( ~ IIUFICE PUMPS (rl',4-1054) \_l~Clt-iff-4R I ,~cH*:rzo-,sz 3*-S!-70* 1503 FROU URTY INJECTION SloNIIL---G---1

~

-~6_'-!~

(FM*IOSl5)

I f*CLIS2 TO F:.~R. C:-P.AINS 1 (Fl!H~~

I PUMIP Dr9CHA.Rt.E HE.A.DER {FM-IOS~

r-4;;N*4*l-C*NII (FM-10711)

!1'*Sl*147*1S03 2"*CH*Zll?-IS%

CL 1$Z * , fL r-c-1111 s---~-'---< (FM*IOS~

f CL&

I ir-----:

. 'IO fl.COi! llRA1'6

,., ,.,.... BU)G. ~ .. ( ~

FJICM AErU£UIIO W>TlA ffllAAG£

,.....__.___ .,,. :a..,....,,...., F1ilCAil CONTA.U,,1),0,IT PROB 11JU\(FM*101-'l.

-, . ,SO.:.,TION s~*n::,.,. - -;z5.s-

~

I S!IOO

/':.GH*I""°'

D I f'*U*IO.l*1'2 I I

I MYDRO. TEST CON" TO n.ooR.

(~~~i

_rtf*SI*8*1SZ 1l*TS6 1i"-SI-lOl!*ISZ raoM 1154e-f" I

I I

I 2'*SI*l31 *ISZ IN!o1Dt 2filCIUARD!I, r-SI-209*1503 I .t.RtA

-(l548-n4-ID6AI 11 4',i:2 R£D

'I TO !>AF'ElaU.IJlOII

- A R O , SUMP F"M*IOOA llOTn;

+: SUPPUED WTTH BOTTLES.

.t.. LOCATE' CO~WECTION CLOSI TO PQM l..,ICT'TOll w.NK B RV*S1*189' TO l!E 61\GG[D D""ING ""11110 T&ITI~

C LOU.TE CONNECTION CLOSE TI) v*L~E9 MOV*SX* INTCI D D , THIS IS A. SPECIAL VAL\/!* FUNCTION' !OTH AS A CONTP.OL ANO A. RELIEF WALY£ TAPS MUST B,[ Z.'*O*,t,pA.RT .

SAMPL£ CONNECTION SHALL I [ IIAO(£HT INTO TH£ ,._MtlUI.US

,i I TO f"LOOR DRAINS(FP-~Ai' G

M RE.IOIIII REM\Nt, Ji.. CDL \.l~IOI QI\.~*~ VAJ..'VI. ~ ~U~

V"\.VE STU'\ L t - Pl~E TD MA'\IY. 30001.II SS UII\ON AT V"-\.VI<

tl.EC~ROMN.NETIC "\,\E"T..R IN VUl'Ttei.l. 'PIPE IIUN SYMl!OL AC llENOTES ADMINISTRATIVE CONTROL V1'LYE

,~&OL - - DENOTES HE-'T TR.*CE i*TS&" * ~AJi'fLf1~ 0, ~0./6~PlsR~~'mrf."J. f : i ~ L \O.S I Plf:.Ot5 r31

!!VMBOC l'C OIHJO"T<.~

IIEFEREH,t OR...WI NC.:!,

1'0.IL CLO'!,[t> I/ALI/£ LM'

• LE1'KAC.E t,f,o!l1TOl11NG CONN.

l'LOOR OR&ll,l~C.l ....,~ILl"RV eLO&.

FLOW DIAG VEN'T ! DR-'IN SVS

'Fa* .....

S"M* 11101'1.0 FLOW OIAG CONTAINM~NT ( ~ttlllC S~RAV SY,; ~tot- IOI "

lzo DI-" FM--to, ... fe I

., FLOW [)IJtJI RE'-CTOI! COOL4NT SYS 1 41 1*

- ~ *s.t*1:0,*l~*N9 rLOW Ol-'Q R!:S'IOUAL !<£Al CU:"°"'"'- S't~

FLOW OIAQ CH[>,t ! 1/0l.UO,ll C'ONTAOL SYS FM- I04 A.

"'1* IM.1.fe

---£-st-141-1-c.-..-, rLOW OIAG' sa.rtTY INJ(.CTIO .. !'fS !>MI "'M* 106&

FLOW DIAS SAF[TY iNJECTION SYS 11~*8 ..J"M*IQ(>A

,.__~1,'T f*YO!-~"°' FLOW DIAC. BORON RE,01/Erl'{ !'{~ FNI *ftD I '-COOLIN(;, COIL BY )!

I 4 D

°-T!>e SAFETY INJECTION SYSTEM I

, "14 TI-< 1* H)IOOT*O~ CN THIS DIU.WUIG IU' co1>:c~ CIC US(D J::lR OTH[~ nu* TH[ cc NAl ..TO.Al'iCL 0 .. 11[,AIII Off DUC1t1N,D .. TNI TiTU: ILIXL OtlCIOtNAL ISSUC D(ICtt1P'll;;,,I fMl PIMf I\

UNIT I SHEET SURRY POWER STATION I

I FIG. 6.2.2.1-2 I

FEB II, 1972 i*S[E NO~~;. (TVP)

I l j 3 (50 \ 21 3,*-s*-,4£:-1~c'-:.. , " (FM - 106A)

'I 47

~-(~8 I

r-SI *70*1503 ~ _.l. ..,.-T (FM-JDC.A)

I I

I I lll!IM HIS$1L[ )

8AARII* I ~

OVT1IDl HISIII.C b. . .

-:-*TSI , ---i

!,I D <. _..

• _-. : - 2

, ez .......

.~*

.........- - ..__ e=:*-.:~-:scz~ 6-Sr-143-15C-2-trr*-

( ~

I I

'I I IIOTES:

,_.Nf:RAl "'1TE!i t REF. DWG,!, l'M--

,6. ADDITIONAL CHECK VALVES INSTALLED il)

I PROVIDE 2 VALVE ISOl.lll'JON OF S£N51TIZC&

PIPING FROM AEACTOII COOLAIIT IOUNIIAJl'I'.

¢ MOIJI' IB~A THRU F TO BE HAND OPEUTC~

ELECTRICAL CONNECTION Wl~L NOT If: Rf'1mAJl:D LOOP e CO~CI LE l"CS}

(l'M*IO

--- -SHOCkf I

0 1"-sx-co-1101~-

Z. -Sl-61-602 cf-st-'ZT-602 - r-----------"'::-.°-6 --(FH-ID(o.t,)

I I ,&.9

.---------~=--<.:""-1oui)

I -

"'"-TV 1N,l'1'10II Sl(oNAL l II I

I SAFETY INJECTION I

, A(ACTOIICOlfTAI-NT SIJHP (PN*QA) - - - - - ' .__ _..J_ _ _ _. __

• ' ...10 I

/0. ,,;

• TIii ..........TIOll Oil THl5 DU.lllllt. ...., ml"I COPllD Pit USID ro* OTHt* TkAlt TNI . . . . , _

IWlff'Dl&IU CIII &ala OIi ,,_ k - ...,,..

~ - , ... flJU * , u SYSTEM UNIT I SHEET 2 SURRY POWER STATION

(;[ .. fRAL Rl**!>ION ,,.. ,ca*

• - ..-* ~*/ "\\

I

--'ll aar,ie:,a,l-*c-11*...,,. DtlC*IP1'10N OtsCflJP'TIOJII

I FIG. 9.1-2 OCT. 15, 1970 I 8<;01-11\f.:l-'aVVII H't'l>~'l',E~J

,1Jr1 ~,*

l\f..'.."1'F!J:;

TO OE.~AlH..lr;, CATION A.wci

• tiAl~Et BED OF't41HE~\Z.ER'S I t"*CH-25~-!!01*

,*-vos-,sot='

~ ! ' l . ,...TlllJCII:

FIL'-

1"-VCS-ISOC I

I I

I 4-;

FC.V l"C o----...;--.. . ____

11'.0'~'

,rCJ.. ISZ r~ CM-8-:503 T0'1.0CMlDMIIS* f'.1'511--=.:::, **

/

,..EL 11"0

-s=~=='I"'""----,

3--o..i92.-1s2.

,*X4Z.D- ....._

(F6-9Al I 15

• j I

3"*S1*7'l-150l!

(FM-I~

3'-CH-~-1so:1~

fc.usz ,

-!"<H-17-tS2

~

FGV ro

,t!.'c~-5'-,Sl- Fl?OM BoRIC ACII> FIL.T!.R.

(Fi,1-IOSA)

I I I

! C)JTSIOE R£AC'TOR tOHTAINMlNT ~*-tH-89-IS09 MOV 3'CH-191-152. ~:?-ce;,O '.?-"4ZD 1.. Z~CtM9'-15,2 FILL

,-.,... , ~ '.$2.

V 1373-I 3"-0Ml3*1SOS 5 CLIS05 --I--CLIS2 Cl-lEIVIICAL MIXIIJG TANK 1-CH-'TIC-3 I i........

z.* HYOl10 ='S,j

-4'!.CM-89-1150~

---

'---------.------------------.-;r.,

CON .. -........__

4~CM-S0-1503 <':B i,o,

,-moo FAI D MDV

• z*cw-*2-1~2 I ~*-CH- Z-19:)3 3!cH-113-l':,03 3'-0l-71-1503

.~/

1?67C

/'1ze"c

\'

..CM-S7-l'S2.

NOTE.':>:

1110 (MOTE"')

DE.VELOPE.O FROM Vt~B+J903 I 37 GENERAL t-DTE.S ~ REFE:RHICE.ORAWINGS------FM-105 ...

VALVES LISTED &LOW SUPPLIED BY ,TONE.¢ WEBSTE.R.

ALLOTHE.RVALVf.5 !5UPPLIE.0 BY WE.5T1Nr.HOUSE..

z Yi-vos-l&Y 2 YZ*-vcS-ISX 1 1*-vcs-1soc I 1*-vos.-150F

..ee-<*!<i--

m""".-4

S t" - vDS-1£.Y I 2 I"* VCS-l!IX Pt.V/CH 100 PCV/CHIOI PCV/CHl01t FCV/BRl49

... I SOV/ BA.149 I _J

~9'!:)-0f:IK!"-......:i-Fl RV /C.H 100 I --~-...,.p I .:. ~4-0G*ll... *152

.CHARGING PUMPS I CHEMICAL aVOLUME CONTROL SYSTEM UNIT I SHEET 2 I SURRY POWER STATION I

I

• * - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -**.;.,i:,~_!.!~'.:T_t;

---

. ":.A.-... ~,:.£_

-- *cacT:.

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

I .;:teas ~4-1

,-,.1,.,,,.,,11,:*,T

~

1~c*5**;... **,.'T" ,..,..~":,.*,.-

T3~-~',.....

_.***1~**!i.

1*, ~'.  :-..... ,i.;.,.j

• 1,.':.*ISA ,.Iii( ..., :._Ill( ~ **

I *.- ,. *.

l*-?'-0,.* .~.

t

. I CG . . . ,sz I

I ti-~'  ;;

I I

.: *. ~ 11'\*t!IZ I

• I ..

~

I COHTAINMDIT 1 c:~ .. s.,u3 I !PRAV

~UM~

Tnf'l.()(I .... ~ ..

RiFUU.IMG 1111111 0

Rf.C:lllt.UI.ATI~

"""" 1 R&PU!llN~ -.TU IU.1'81<iotll...TIQOII I

• n:~ IINiT.

LMC*LtMAW: -ITOIU"IC, CONN[CT,OM!I IIUllll:"'TE:':> P~l"4=D 1W ~l'OellT ... R.

I  ::.~.i:= ~,\f-r::.~~~

lt-.L~ *Tf:Ml.~\L 'TO~C-C>.

l'LL JU)W MID P l l l l ~ -TRI.MUT ~TIOtft TC:*R\'4",

VALVi llllol "'"'~C<T ~IP(S WIie.

~o HIW( SIIO(Htss .......

l'T VAi.Vt I

aEFt:.f.113 DR,..W,NG'S:

l'LOW IXII09- MAIN 91'£>>1 _,,. l'M-MA l'LOW I l l - - f.&WYCli. ~

Ill.OW 1!11,QIIAM ~ N T CIIOLIW. *Tilt PL0N OIAClll"'M a&DIIAL OC,.T llllfJMI. $'I!,

,-..c,w DIM*"'" c,c,i. t '<<ll. C:CNft(II. eYf,

'M*llo\

,i'I..UD I ~l,QW DIMau.M &AFlT'I INff.CTI°"' !'~ F>-*IOM r1.00111 CIIAI- llt.U:TOR *oo,TAINUCIIT p ......

1 '1N (.tA,rwH,H.' FI t( t,'1- :. * ~. l '."

Y/it,1 VI ,,...,..kA.1 ltf.!", t,,a , .... r~- .. Tt.*-1vc*-.,

l,c.[i...1\llULA\ l",11 ",.t-kA.'1. ~' ** T UJI now Dll'GR"'.'l VENT &.DR...IN l'l'STltM -

I *- * =~ \A

....... ._.__N ,_ _

1-

, T05'.M? ...t** 161--tSA D - ,

I FLON DIAGRAM CONTAINMENT & RECIRCUL.ATIC~

SPRAY SYSTEM I aui.111v l'OW(i. aTATION Vlll,GINIA ILECTIUC AND ,:,owEfl COMIJIIIAN I -

........ --mGMtll aaazua hiirPtWA"""

11<<8-FM-IOIA

I L FIG. 9.3-1 OCT. 15, 1970 I

I

_ _!NSIDE-l.....~.D.E_RtllCTOfl I

I COHTAIN,..E.HT I

TO Ll:.TDOWH HOR (FM-IOSC) 2°*CH*2Jl*602 I

---r (l'M*IO!le) li"*C6*31*1.SZ.

(l'M*IOI"')

.  ! DC:*Sl-11&

(F"M*IOOI!.)

PRIMARY ORAII< TR1'HSf'ER

!!! - * -* - ** TK. RELIEF VALVE DISCH.

I 0 ,-Z"-RH*3*'4Z CF"M*IOOBl

~UMULA.'TO~

,'l' ~TLltll!.Rl!.l.ll:,,:.

V"'LVt.(l't'l*I~)

RESIDU"- MEAT REMDV"'I.. £ ICIWl<i,PlS Z."DCi*2*1!1Z I l"Al'-

.OPl:N 1:51-!ll*ISt

• I IIOTES

~VAl.l/E'5, ARE t.JORMM.LY 11.!ST,t\LLEO WITH fLOW UNDE.P. 'f>E:A, Z.RJ;.O.CTOR COOWl<l,IT PRE.',,';,UP.1!. ltl'TERLOCK PR!:1/E~

ISOL/>,TION VALVE OPEtllNt/1 ,._BOVI!: st.'t PP.£:SSUP.fa I

!LOCATE 1'50LATIOl,J VALVE CLOSE: 'TO MNt.J KU..01!.R 4.LO<.ATE C.ONNEC.TION ltl 90T'TC>t1 MA\.I' 01" REK.TOII.

C.OOL"-KT Pl'PINO- ON 4J 0 ,....C.l..t. TI) ~Tl'-'L.

l'*CC*IOSH51 -~ fl,._NG,1!.f, ,t\11.t. :500 LIi, ST liiT\. TYPR :116

~tllANICAL SU.I.

l'Aa.~IUMC. REFERENCE. DWGs:

COOI.IWQ (P:M* ttN now DIAGRM'l I

COM-NT COOLING FM*U,.UI

( TY'e) r'LOW 01-'GRAM SAMPLIN~ FNl*:n"'

FLOW DIAGRAM VENT( DRAIM f'M*IDOt.

fLQW DIAGRAM GOIITAINMEWT4RECIRC5PRAV IIVS FWI-IOl ,t\

FLOW D1AGR~'1 RU.tTOII COOLANT FM* ID*A now DIAGRM,1 CHEMICAL e VOLLJN1£ COKTROL FM-IOS-I FLOW DWl"""'4 SAF.. TY INJE CTIOK FM-1

'Rt!>IDUl>L. MEAT Rt M!IVAL PI/IIIPS LE~ND 0 LOCAL DRl<III CD\NECTION V LOCAL I/ENT COMKECTION 0MB OUTS!~ MISSILE -P.Rll!.11 I IMEi INSIDe. Ml~I~ l!ol,RRll!:Q T PLUC:C:ED CONNECTION I

, ~- .51/()C ~I RESIDUAL HEAT REMOVAL SYSTEM SURRY POWER STATION I

I FIG.9.4-1 OCT.15, 1970 L

I ., 1a*-cc-e-121 fvo~-,o.:~

1

~t:,**~-..,c I ,~:c.:-,0-121 P5 PA

'00\ C.LISI DET,.IL A I ,:cc-zo1-151 YO>-,ocm-'7

,1.*

I TDu,m, MG LUii[

Oi. COOL[~ (TYP.l- - CCIZS8 PA

,1.* ..

'i li!E5ID1.JAL HEAT _.,-

REMOVAL EXCHANGER:>

~

iii I

CJ i  !!'

u u u

1*. u TO LOWU 811G. LUBl I OIL Ccou.t (TY,)-- -

TO THCRMIIL IIAARIUI {T'1P.)-

I I VOS-Goc~.f 1*-1

~-

10 Y"--FUNNI!\.

~ T O FLOoR D2AIN I 6 LOCA.L S,._MPLE DETAIL

!ii NOT!:S:

".:e,L !:~':;~11.E C:ONU$ 'To ALL TC"'PERATUR.[ COAINECTICNS TO e£. l"wmi er VOS*loOC E ~

I" UCEPT i'IS NOTED I ~

v NUMBER SHDWN A'T R[AC:l"OR CONTAINI\AENT LIN[ 1ND1Ci'1Tl5 PIPING PENETRATION Nl)Ml![II.

ALL LOCA'. SAMP~E. POIH1'31t> BE Fk.)'f"ID[O WITH OPEN FlJNN[L5 TO FLOOR PRAIN

• INDICATS PEN~TIU.TION NO ON F>'*IA MCB-MAIN CONTROL 6~A.O .

I &-COMPUTER PRoVIDE ROOT VAL"E~ (

r\05 *"C)C) A.'T A.LL R.OW CONN'S .

INDICATES FUltNISHU wrTJI EOUll'MI.NT REFEllE.,t.E DWG5. :

PLUHf>ltJC. SEIM~ i!.I..OG.. Fr, - 271!,

I FLOW DIAQRAM-COM-LNT COO(.ING ~TER -SH:2 - FM* 17 l!o FLOW DIAliRAM*CCMPO"ENT COOi.iNG WA'l'Ell-SH.~ -FM-22C n!)W DIAG'Cl:AM-COMPOWEIJT CC,ot....lN6W.A,TE'- *Stl4 -FH-220 ROW ::llAGRAM *COMFlotP.cN-,, COCUNG .....TER -511.5 - rl'I

• 22E f'I.DW Ol,t.~RAM-C0'1POl'ENT COOUN~ W..TER*SH.6 - ' "

• llF Jl(AC TOR C.O"T"IWME>Jl" PCI.IE TII.A.'TIONS - F'I

• I I<

Fl.OW OIACRAM*AUX 5TEAH tAIR IW'IOVAL - H*-1G ..

FLOW C1A.5RAM- C.Ot:OEN~AT~ - F'M- Ilk FLOW [)IA.,:;RI.M-1!,0l!DN Rl!:C.DYER'f

• FM* 29 ..

1* FLOW t.lAGRAM *WASCE Dl'SPO"'-'L -Fll-=00.

FLOW 01.. G£1o.H - SAMPLING S'fS'T£M -FM- ?al<

FLOW OIAGRP.H-YEMT tD'<"'"' 5YST£M -FH-100&.

Fl.OW Cl-'<il! ...... -CHEH ~ VG~OME CONTROL -FH

• 1056 I 1*

l*Ct-E-2 F\DW C(._GR .. M-COMPENT CODLIN6 W'-1"Ut(115~8)-F .... -l.1 B FLOW Dl'6UM -"l!CUl,,(l'll'fl. 'SE~VIC!: WA'Tl.R (IIS4!)* Fl'o* ZI,..

4 FLOW 01'61W-\ - ..UX S'T jtJll l!EMOVA: /1154&', -FM- ll,,..

PIPE I PENETRATION COOUNG COILS (INNER)

I (FM-2f

• CC*l'MZI

'l?eo8 60.s-COMPONENT COOLING SYSTEM I UNIT I SHEET I I l ----------------------------------------------------------------------------1 SURRY POWER STATION

I "FIG. 9~-2 OCT.15, 1970 I 2 2 - r,i.:1- 8t,t,I I _::r.a..~t ON VEIJ"T. C'IJJU..

.... ...-...... .... ~ Q t : L ~N\

OUTSIDE RtACTOR IN.SIDE ltlACTOlt CONTAINMtNT QI

- 5EU:cr.9 ,-TCH f"ROM CONTAll<MENT

~-: ISOL"'-TION.Sl'3T!.III I

-P'-t<T*'ll\.'f: M"D APING ll!CIV1' Tll~Ll'IE sov

.-T 1

"JNT

-t!"':.==~-~

~11>"

• GC\11 S(JPP\.:ED 8Y GQ;L MF~.

1 ~o/j S'(} I PRIMARY .SHIUD PE"LTRATla'I COOLING COILS

FE

=~ . - sov* _,,,--. ' /~* ~-

I

-,:_CL 151  : ~----------t--~u:----r.5ii; /

• HC-£-!G CC-£-3C :

• I

--- ",--~~-~-.

• : - I il!!t-1 ~o.1-I I 1, Vll5*r.OC I f FE mn-:

rl.P.~ ~P£ !l'-oL '!. S.

ftE 'SK>P tlr.lSUrll-ED UJ EU'l'P""'\I ~M\EL.t) 'TAIO~

1* ~ TY'PE i31G> L &.~

DET-"A*.,/ I PIPE ~

!kV F1£LI' IN9TALL~D I f (FM*22A.)

t~~ri.~~,'~iLC.

~FA&RICA."r-.D l"*~TYPE!llfoLS.'6.

~INST.\LLED ve;ae YCIS"'°' IIYl'U:1.D I ITYfl I

ffYP.)

I R£""1::R =:1'-"'IIILNT AIR RECIRCULATION COOLEPS PRIM ..R'I' ~HIELD WATER WALi. C.OOLER PAMl:L ~EC."TIO~

I I H I 1*

.. 1...t... _*--=--:. *. :....::.___.

1-

,I I

I REACTOR I

l-.£uTIIOH SHIELD TANK I .. ~-.......-~'-ISim i!

1*

I 1-RC:-t:5*1 DETAILB I r-1-CC*l*Z PIPE PDCTRATION

/

COOLING COIU I ,*-cc-m,*.,,

(OUT[Rj "TO RU.CTOII' COi.r,"-'J ..IA'!:MT

!S<MP (F-,..\tt>B) COMPONENT COOLING SYSTEM

-CC* 17*111 I OET.A (P"M*r2A) UNITI SHEET2 SURRY POWER STATION I

I _,

FIG.9.4.-3 OCT. 15, 1970 J22-V*L:l-8t7171 I TO ":..O~PCUl[UT C.Ol)LllrJG I*-* PRoB 61~

I P.JMP:. ~TAR.lltt... i C.lRCUtl "!'.~- 1a* r.::.-1 1& t:.C*IO*l?I lt'.1-I --VG :i OUB Ctl"i1 Tl

, - 1'4-CC*Z,"1'1>1

.-*; -cc.-s.;.-,s, -Vls-c.oc

,,=cc-51 .,,

I Z"*C.C.*ZM-1'51

._T_l_

M/.1._J C.OOLAUf MOT LL' SEN el.OW!r."1.111 C.C.lll.A I VGH,<>ll- r*c.:. 50.,51 "-!:,...:-,.1oo11(T<Pl ,,*-cc-!IIH!II Pl f\V - I" tcii,"

z:UN.t4S*l51 CCl'.lt.r ,,( ..

(FM*llO!I)

~ -~OOPUO, l"CC-JOt-151 rvo:;-*oc l*~S ~TE.i..M f-cc-~-1111

_I E*?a6 !i.C:t..i Pl I RV c~:

I"

,,.cc-!i~

111.V ccir,,.~ '-.I-

_ e:ov.:::,:l\!,'1,1 fl*CC-~9*151 l"-CC-303-151 eti;o

-r*CC-30&-151 IN / ---r

~ L:>OP ~Cl':

run-I r GAS STIUl'l'U r-YGW-15A I I*!>~*

[-~

LIQUID

~rfLUl:,.T STC.:.'°'

ws-'°c-L---

1\~cc-Z9?-151 WASTE GA~

D1Al'HRA5M

-1114 COMl'PESICIH

-1f-cc-z91-ei ~- I"'*(-*!. .r.* I !:;.1 I 1*-cc-z""-1s1

-~ll.--

cc,o54 a:-*

Pl.['5~URIZtl<

LIQUII) l'M*IO!il!I

~PAC[

I

1, 2

I LIQUID wa=in 1*cc-2.(o5-1~1 1f-cc-e;,-1s1 I ....

'i I ~H~A  :~OC(TY~

c,.

HE

~

'--vG=i-*Oa [YAPOIIATOR I 1"-CC*Z70-III

~,-

~J m:v tel00, CCI04-6 CCIOS A CCIOS 11 (H!IM>-=iA (1-1111-P*lUI Cl*Bll*P. 4,_

(l*lllt-P-48 l!ODOM L\aP DIST l'IUMP)

.. /IUJI'(

lllt,\lj "TANK PUM")

I CCIOlo" Cl-llll-P.C." BOIIDII - P CIIIC P-P)

f>;foB CC/01,8 ( 1-'Ba-P.r.8 CCIOU, CI-BR-P.7,_ ~"TAIPKlt C.lllC PU-)

CX.I07.B C1-a11-P-7B SI~ CCIOeA (1-LW-P-i, M1C... 1.Lttl.-'!ST£DQ1,1"Tk

<,~w-p-~a

--D a:t0a a CX.IOIII fl-Lll*P-4 ...s'T1: D5P 'f.V"P OlaT* PUMP)

I  !!

CCIIO H-LW*P-e (1-811-C*IA (l*IIIK-111 DE"TA.IL l!I

-...S"TC D1aP'f.V"-P CIQC l'UlolP) 11 a,,'~ COl&Plltno*)

ol II )

I I ti"*cc-~-1!i1 I

I ll!)"CC-17-IZ "tt-1&-IZI I COMPONENT COOLING SYSTEM UNIT I - SHEET 3 I SURRY POWER STATION I

I .

FIG. 9.4-4 OCT. 15, 1970

?-,Y I p,eo8 FS L I r-- *;a,;...,.,--:.- r18'"-L::.-7-12.l(flh22~

Z.4"-cc-zz,-12.1 I+~. ~ JH ** 'L 1 lu I \

'-' F-12.A

• -,e:.-cc.-Z'Z6,-IZ/

1ct:'C100*

\

'-----------------i-""'~* '-'-*-*

fl. /"-- , ..

r 18"-CC.-8-12.1 (FM*22"l!I I  ! '11 (r~~blcc.,ootf 111(

[Filla! 7-TO

\

UNITIIO I -~

r18""*CC- ICrlZICFU*22C )

• 1.-'w*c:;. -"I ]

I ~

RA V rc-,~ - U- -iow.

I r1 I ~ C/91 I "'~

~ cu, I

• u u

f I ' I .

I I ll'-;<l *S7-IS1 I

(

iii

~

TO DlliCoG

_PL 106Cp CHILLED CC!\.IIPONENT COOLERS A

L.

~W-1'::-

PI

/i C.Cl301..

8 ~. I (

• YPJ-_,. .J-r-/"' g CCl68A

• I  ;;,

~ ---

E.'-cc-11a-1s, ~ "

6"-tt-157*151 ~

WL '

--~

lvr;w-15,,. -, . _41

('TT")

1-C.O*P*,,.-'

0 PI

~ci:i'io"i,

,.,~.._ '!><le--""""

TS*ISA (TYi')

,f:'_-_cc_._-_11,_-_1t_*__,1, V5W-15A

'4 t131C>-151 cc,30& _ ~- {F'M*22C)

~

.i

~

I 6**tt*H9*\1SI~ 1

, "\

t"*CC--ISA (F11ift,,\ .

I

..,i u

~-*CC *199*151-..

l*CD*P*II!

1ol 'o PA.l'L ~ Pl

.:: ~-

~ - *--~--r-

.----- ~C.U!O:

.;;),- rr f Pl cc1&ec

-t-~'----~. .---'t.__"*"-*-ll_._._ --t I

6 111 1

'-*b-CC-117-151- . . .- .... - _

I I I l*C.D*P-IC.

CHILLED COMPONENT COOLING PUMP!:-

--=--I- - - ! ' - - - - --------- 4--*1_--,,-u-Jo-,51, I 1. *1} *-cc-2c.*:i-1S'1 AUX. SLDli.

I

• 1i*-c.c-2s,-1s,

./hea_tS SCf

~~

V e~~~11 l{i,N~9o\l I RV ffi208\ . \ ,*

~

Pl cciiii'""'

'r .

Tl

~Jrr.l== __

CCIB7 ~-

l_.4 .:J*

cciiil

"!I Lr

~i TCV iiiiti

~~,*=e

\'G,'N-11jA.

- .. ,M~CI--------'-~--~ i--.c. 1 c:-cc-z~-,~,

6"*t\J-l~-*51, 1,\" 1

~- **111i!*E*12 I - P/e()8*

L\....e.e..._ \ --......1.--.....1 tll-"'Pl!--""'---...*T*R-_IP::;;-

I *-~

~. C.00._LE.R_..,.l,_6..,..._

' - . : . _ V&'w- 15-,;:J. "TQIM Pl c111ou~L~

__ , "Tl ) ,,..6~*

I

~------------'

,-cD-£-2.8 )

~

PI CC.l5~ ~ i l *

~_!,'ll8"*CC*314*15i

• ~5*C:.O* 7. ~,.,*.cc.-3S1-1~*

--~- **-

l*ll**P.4A<l!I,

~

j I

.-, .,, 1-lllo'-P'*tACa l l?EFUE.UNGWATER STORAGE l"ANK ::,10LERS CLl51-

. ....- CLl21-> ~

1__ '-------------""':~-,:~~~""':~~*~-ll~::_*f,;:\rci -l8*CC.-1..S-1?:* (""1*.t'--7' I - 1 COMPONENT COOLING SYSTEM l *l&"*CC.*lb* " 'F .

• z.t ... t UNIT 1.- SHEET 4.

I I SURRY POWER STATION I

I L FIG. 9A-5 OCT. 15, 1970 3 22-Vt:l-8vt>II I

I ~Llll!,,1 3*:u.-1~1-151 I"

I r**-cc-zez-151 2f*CC*l'll*1!1I Y6S*i;OII 1fcc-n9-151 r-CC-141-1~1 I -**-=-~*-51 RCSlllUAL MEAT R[MCJW'AL LOOP ND.1 l~CC*1*D-1S1 I LOOP NIU!

STCAM QCN

~---11-4~1-~ TE*2144 I r-cc-1,-i-151 V0S*60C(TYP)

(11541H"M*I05 I - - RCl:ENtR,.TIIIE

!£AT EXCHANC:tR I S ....PLC CDO\.ERi C.Llf,/

l'CC*.S.~I 1f-cc-11!HSI I

CLILI PRUSUIIIZIEA LIQUIII I

SPACE I r-cc-~

I I

I I

I COMPONENT COOLING SYSTEM UNIT I. - SHEET 5.

I SURRY POWER STATION I

I "

FIG. 9.4-6 OCT.15,1970 *

.£2-VH-Bvvll I

nlOU CONUl><W{NT I ISOLATION 5YS TCN

• l

• 11~&.6

'* t*-- -;

na-~A. e*-ws-1.&11-1!1:.

,, -115411-FM-22A 1

1 =Fl j f{1154&*FM-21A1 FUJA.T Vl!l.l..v'L(TYP)

,-e--w,;-149-l!to

' t

'S-*151(1154P.**FM- IS.A)

I S"'-f,iA.-10~-301 -  : ,

54 6 21

-~---..,*--"__ _-_"_0o\_-_ _"_)__,,..,a.__ _,

~1546-FM-iw")

T I ___ -v!T~P,15<.

=1 1

.:,W~*.101-151 II) i,..,....c....~

L--~~2<D-~-,~s*~~ -

0

,p

~ :l I .. ,. ,.. * .*1)

I ~L I pq

• cw11.LED W~R CONDEN:.'Ee

/

1Ltl<,P ,c.._

(*"IC,*1~1

• VGM-15[-*

,_ ,~; l~

V6S*--:-,.i_

1*ws-2a.,... ,,.)

.~

V~ ~

It

.~':>A- 151

• J:WC;,~J.~

YGM-1st-

'~-I.Iii V

• 1 V65*MIII r"*CL-151-..,_ f_,,,-*YGS~ y .,-l"*CL'-151 I TO FLOOR - - - . ,

1 2*CO*EJ*2A

'1 *'TO Fl.DOIi DltMI 2*CD*E.1*29 CH1L:..£D WATER CONDCNSER AIR E.lfCTOll5 I

l!r-c:IC*l!HIIII I

I I

I ,o~c:w-,.a-151 - . . . __

I I

I I

I COMPONENT COOLING SYSTEM UNIT I -SHEET 6 I SURRY POWER STATION I

I FIG.9.5-1 OCT.15,1970 I

I I FUEL. .,,-T COOLEf.S V8S*l5Y Pl I

• re 105 I 12.*-F .... )) "'lli2 2f FP-'26 *152-\

A.C.

HOS£ CONNE'CTION WITH BLANK F'LAHO!

l'!.°*fP*3*152.

I ~VOS*l~Y

~ llo'*F,*tB*IIZ l-l'C*P*3" SPENT FUEL PtT PVP.ll'ICATICN. PUMPS

.t'*PG-o?9*1ii.1\

(FM*29C) \

3 *

'I * * *2'*F~-U-ISZ.

-e"*<P*Z.1-152 RcrtR.:NCE ow.:s C'OMPOH[NT COvLIM; "H~TCP. 'iVSTtM B<l<',C., "-ECO\ltll.Y STHl..

'W'A1Tt 01'120'3.lrl. 'SYSTEM VEIIT.t DRAtlll ~YSTt ..

_..,ne S'l11*29C:

"5'-,0C

.....'°°"

I rY11"1*15Y

\

NOTES, I

FUEL l'tT ~YM80L A.C. OENOTCS AOMINISTRATJVE <;:ONTROl V,.LVt FILTER I* fC*FL*Z I SPENT FVn. PIT VIIS*ISY 2FP*2.B-152.

I Pl C:IOl.t.

---

**-- r-.1.-,.4-,....;,_,..--n~c".OO)

}'llaS*ljY fls.<:"-E>O) r.*csc~.~ii.~-1s2.

I PI

,c:103.t.

S~NT J'IJ£L. PIT SPENT F~L i'IT ~,ur,.O~ER ,NMPS SKl~~t~

I FILTER

,*-Dl'l*!l!l*lll

(<M*IOOA)

FUEL PIT l~ DI\*"* IS.?

(FM*IOOA) COOLING SYSTEM SURRY POWER STATION I

I SF FIG 9.9-1 F'~-1'.?~*

OCT. 15, 1970 I

~LOWC'CWN TAUi"': ~RAIM OISCHAPS~ TUNt..lEL 5* FROM Al;f COt~

UtllT (i'B*-41A) 2 2 ~ "*TO SLU~E iRE).:.H I

.;, ~ (FP*71A) 1 3

u 3 '

3

'.iII' -VVF -ti.lT1P)

~ ~ w:N MOV c.~. °' II' CWl~OBL\

I l*CW*P-'ZC SLUICl k'A'TER PUMP I

1°TO RAD CAP 850 C.PN ~

~MON!TOR -.c WON1iOR1wG SYS. ZZO FT TOH lflol*l?~A) lTn) Pl VQT*ZO<, (l'l'Pp,

......~~CC-E-IA I

I I tlllt\J.ATINI; WATER PUMPS l'.APAtlTY. zro,aoa GPM IIUI>

• 28'TDII crsin:J

'I UNIT lt'z.

I -24.MH 0

COMP<lMNT COOL I NG H1.0 4Z *'16-12-IO HE...T* e-.cHANGERS

+2:ws-13-10 l*CW*P*IO V~F-l?D I r-----

1 L. _____

CMILLED WATtR.

UNIT C.ONDlN~t.Q 1-CD*SC.*I I VAC PRIM Pp NOTE':,,-

&, DENCTES COlotPUTEII' POINT'!,

-;;j~r

'"~

PIPE ~IZE FOR 30" TO 96" PIPING IS IN'ol!)[ Dl"METta.

PIPING TO BE IN ACCORDANCE WITH SPECIFICATIOfr\l ll'OR WELDED CIRCULATING WATER PIPING.(EXClP"T 1>,S NOTED)

I

s

2. CTC.StaWN AT EQUIP CONNS INDICATES MFR'S CONN

• !iJfeU r'lfli. 1 (TVP) Nl'MBER-CO,lD CWG N4*22SD-~R[Bl'SOIXI

• REACTOR CCNTAINMENT Pf:NtTltATIOHS U411-fV*IA. *

-VCT*20*i ALL f'AE3SURE CONNS TO BE f_ EXCEPT 116 NOTED.

(TVP: . ALL TE..,PERATURE CONNS TO lie 1:EXCEPT ~ NOTED.

• I 4**\\SW-4*136 -

,*** *~  ;,,. *: ..sve:,r.1i:o

. -j- .. J.L

• FUl<Ml'!,HtD l!>Y EQUIPMENT MPII' I I LDB, CM I

I *LOS.. 1..:.1~

C.WIOSC C.WI~ CWIOSA*

J.!-os* (TVO) c.w,oso 4 *WSW-7*136

'VVf*l2A.

C.P (lYP) su:v:cE zsoo WATEt ,WPS l*S't-P-4,*48 GP11@"15fT TOM (ruu. ~IZt) z*

3*.r.,..!.!,*IK, EII\ER.11,ENC.Y WA.'t?.R PUM\116 IIERVIC.11.

!.N.&IM£ OIZl\l'!.M CAP 15000 G... EhF'T ,:Dit TlllP)

~:;~~\~~~;~~Efa!~l~::o I

L._~lDW)

~~010 U. MlC.\I) ON \.OW !IE.0 'T'tUCK

'" SERVICE WATER SYSTEM

~.-

, ** -;:1-LA SURRY POWER STATION I

I FIG. 10.3-1 OCT.15, 1970 V~I-V'J..:l-9Vvl I I O>IE CJJ' OIi E ... rn I-IE. ...DER TO e,.; 'ill-l1PPEO ... OOSE FOR FIELD WEU)IIIG PlllOII To C>4E"IICA\. CLEAM*IIG. *

\

J. §t&J

,*.t

>" ...04

'-,4 ,-'TO GLUrilD ~'TM R~ULA.1'11111

. (l='P1- ll!A) r- \4'"-SMP-15-WOi

'-'--'------t::>1-.....,,0---..-,- . ~r.w.*oA I 't1Clf*V.,a.*

I I

I ...s

r:z:

NlfN -~ "'

',i, I t.,.-

TO ATYOS

>ISICli!&

8-SOHY*S-121 s

.!e

% .,.......t.-CAP ,o K LOOla I'~ Pl&L.D

'Mol,~D - l"DkAl'I--

-*-DECAY HE"T

-= / WI.lo.II Arfl,. C.loll.M C.L~A.Nl~

FM*ii:&A.

I RELEASE ~

I ij-t A. v I

~a l*NS*E-18

,H---+--+--"""""""'lr*lll ~- t '

I' I LOO";E r,a F1u.D Wf1DUr1'J AHU (Ml CLPNIM<.

1--11--...L.-'-T.;.;;.;.-'-.;;.;.;.----+.....L.'---,

---

lCON~ 110 I Pl?oB 34, I

l.,------......1.<<l-\'!'lt!J~-!J,C>+N.., ~

I l[

t I OUT'510t 3._St-!PD*S-G.OI lll'llt\£ TUll9.IILD6.

I

[WAU.

Y<,W I o-....._...,........______L__. ,. ."...,. -----15--.-,---l I DIWft FM-leAl'TYP)

.. II

"'OTi!,:

I FLO'tV' OIAGRA.M ~Y'MBOLS. S.TO HM*IOO 'TO HM*to8 INCL E SK**C *7i,3, RE\f S.61.

.ALL PRE'S~URE CO.. NS TO SE .-. *

-,.* *SHP*a9*601 ALL TEMPl:RATURE ~ONNS TO BE ll REAMED HOLE, EXCEPT. .-\.!., NOTH

• :N01C.A1 ES r;,.,;;:,11~HED BY EQU1PhNT MFA.

lETC ~HOWN AT E.QI..IP C.ONN5 INOIC.ATE$ MF'Rb <.i:>N ..

-c..:,,..r.,.

I

• • ** t*J\B[rt py;:.. N-t-22,0-~*ta"l'*SOIXI.

c!---*-

~

--~:M CONTAINMENT

~ATJON &YaTEM

• TURB owe, 680..:'!IB7.

- REACTOR ':.ONTAINMENT PEP'.IETRA.TION'a lt448*FV*IA.

• INOJ(U~ l'OWl~ RtU[f V"1.V[ .-itATIJll"tl'(

4 !H°"A'T£~ COUPln'Ellil PCtNT MAIN STEAM SYSTEM I *-~~~

TO ST. Ul~N. SLOWOFF ,.ANKFNli-lZ.2.A AUX :- EEO P......, DRIVE FOfl AUX :,T \,,EN FEED PUMP c,-,.w-P-ZJ

=~~;:K~~:~~--:.Trow.TOM"U811 UNIT I.

SURRY POWER STATION I

I "

FIG.10.3.,"'"2 OCT.15,1970

\'f9HU-9v vii

~,'-'

~-,,

LI I

STEAM !IENEIIATO"S

-Wl*IOa .. -

/ \

T

...h!...*.

14T& 1..I LI

'41&

--~

~

,~u* *--...U-LI

.,c l ..7G, FEED WATEJt H£AT£11'&

• 167C.

I ' I

,~**wFl'D*n*60I V6W*90B  :

FCV

,'iii* l.

i,0',-150<

DfTYP)

- - ,sou,T10N """vr I

CLGclC.L901 LEAlt TEST vos-c.oc;(.l'vP)

I 1:

I 0 H *WFPD-t3*601 I .I I

i Ii I LI iiiii V&S-i;Olh J:il__

LIi LT '.

Y011-1SOF

(

-- \

It ~'

.-vcw-soA Q"TEMP.COtll'l l4" IIV flELD (fM*Uw,)

CONDENSEII 1-Cll*SC-IA

,)

I _,.,

I I

I STEAM GENEIIATOII fvo .. -c. - FEED PUMPS I ,---

GAL CONOli NS,.T!

I-C:.. *Tlt-1

'STOIIAGt 'TAMI

_j 13,800GPM@l l700FT TOH HAL.FSIZE NOTES:

I I.

I I

Ii! ETC SHOWN AT EQUIP CONII INDICATES MFR CONN NO -C:CND OWG N4*i!Z5D*RREBT*501lU

• REACTOR CONTAINMENT PENE,RATIONS ll+<f8*FV*IA ALL PRESSU~E CONN TO BE ALL TEMPERATURE C.CH>i TO 8[

f.

• ll REAMED HOt.~.

~)

• INCICATES FUR~IS;l[D BY EQUIP MFR.

I I &, INDIC,.Tt~ COMPUUII PDINT lRoB r Gtr*--....---

L___ ~_

ro* r.w-11-1$1

___J I

,oacPM9c?T!OFT TDH ~so .. ? ... *s Z730 FT TDH FEEDWATER SYSTEM

• " - .sHOCKI FULL SIZE AUX TURBINE DAIi/i MALF SIZE MOTOR DRIVES SURRY POWER STATION I --~----------------------~...----~-----'"-"------------------...-.J

I OUTSIOli f'-CTOtll IN.\l:>W. 'IIII-C,-0111.

I CON'TAINNIS:IJT C.ON'T,..l>U.1\W.\S'T I ,-o CONTAIIJME>JT ISOLATION AND SAFl!!C.UI.RDS l~ITll<TICIJ I ~'f:'>TENl'i>

I u r-LM*lALL OPE~ PRE.SG.JRE TAPS I

CPE,-.TAl':.)*117~

I ..i:.;_

/LM\02

.. rV0~*401l ***I.M*ll*N'P

.,,~ I

_:~ ~l ~-

I J*

~:~* =~

i I .

, T\*

LMIOOG

.::.!._~

\ L.M"?~:,

50Y UIIO!A. \

VO:, *,;o,t.(T'I' P)-

J( . ~

$('/

,j

. ~ '\._!L_ P:ROM C:lMTAltJM~tJT-E-~---*"--;

ISOLATION s*,*s,*EM i::-~~ {

I LI.UDOH~

i , L-.*o:c.

-LM-\IS-N.,

I \

\

I il

.'-v0S*40X~ _ti_

1'01 Lr,1103'

~ i

-'I0!,-4C')(.-

CCN.TAitlMENI -*-

I LEAKA:.E Vt..:UU!w1 F..."£.C.TOR YCW*ISA MONlT:)~\t-',*i l*CV:l'.J*I

l. SE"-E t:> S't'ETS.111 r-\".::.-..i..:-)( J VA:L:u,, PU'IIP

--,-! *1 .. LM*J'*I LMC BAltOM£'Tl!.I\ MAN01,i1E'Tl

R ..___-.------...... 1- ..1.st..-2.CJ-!~I UO'TE'S:

I (ATMOl;.P,..E.R;C, * !~

,/LMioo LM*l8*Nio 5V

'...._ /

~~t-A--1oe,)Cc**to) f'TD*LMI0<>-1 THRI.' Zl) "r.> MEA~Rl!!.. REAc:TOR C~NT... INMEIJT AIR "l l!MPERA"TUR'I!. AT Dll"FLRE.IJT P01Ni5 .

M'T*LM!00-1 &1. TO MEAS\Jf<E REA::TOR COlffAl\,lME>n Al'Ri: h'\O\STUR~ co,.rre.\.:,..

'l>'"To-lH100-1,ua :;e.uo s1<o>J"'- ,-:, -n-uuoo-1, z. or.1 il I Pt.I LMIOI' I MAll,I COl,ITotOI.. i!>OAP.t>.

RT'.>*U4\00- I Tl\~IJ '"* S\'!t.lD 51Co~""-TQ t>A.,.A 1..oi;cr.e M'T*LMIOO *I 51!.>>IC>S l!'.ir.'-IAI.. T'O MJ:*LIIIIOO *I Ql,I 11\Alt-:> C:!.,TRCI.. 111:*Af't>.

MT-LMI004. .!,1<,.n:, .. f,ICl,IJAI.. To l:A'TA. Loa.i!.e.,;i *

• INDIC ...TE.S FU><*J151\\'!':> B'( S:QIJI PM EN"T M.FR*-

I M,\.t,,IC1METER 2"<¥-4-t!i4~ r2"*C\l-3*1S4" ir-:.w-210 -154

!l'M-10~1,7) .---

1 LI.IC

• L::Al<AGE t.f.Nll'JRING CONtlS REFEREtoJ::.E DRl,WING'!.

FLOW DtllGW,t-.Wll:llN\Y STEAM*l'NWltl' PLANT SIil FL,lW DIAGRHO CO"""-&!.SU> AIR. SYSTEM -~tl.1 FM-1,a*:

FM*ZSG (Dlf'FEkEt1'r1A.;) V6!:o-lSZ ( FLON D1"61U.11 r....'2:JUS WI.STElel'CSA\.S'ISTDA Slll(i FM-Im'(&

I .- I* !.A* J'L*.::

l*CV-P*IA)

I VALVE OPEl\l'oTl:~G t.lUM&EII.~ CONT"-'NMENT VA.CUUM

• LE~l\l,Gt MOIIITOlllNCH,YUt.M

_FM*8S~

~~<u<AC.t Ti::,l C:CNTA!NME.1-.T I ..

"G1

-~

z.

01 METE k Fh.. Ti:.P H

e,;..y

• -2*-cv-2-1054 I

A,;._ cv""*.~.C'E.R

• ,i,.15! J"'NM, *T,J* 0£.... DQU,T A1~J

~**G'N*.W*l54 (FIA*ICTl!.)(H>)

I I I I I :F;:i FLOW DIAGRAM CONTAINMENT VAC.& LEA~<AGE MONITOR SYSTEM

~-

SURRY ..OWER STATION I *I VIRGINIA ELECTRIC AND POWER CO'!'PANY RTOXB

• WBBIITKII KSUIXKl<RlNO <.'ORPOR4'110:'li

• '(< a_

• :.- :.;...J...:*

DtSCR!PTIO~i I ,[SCftlPTIO~

IIIIAWOHG, 11448-FM-102A I

I Ct:- ~.:I - 9t'VII I

a"Tf.i,,f M."**,FC..\.t. 19ull ,4.li' NO'i,(:JM"l.'t-~hl.' fr~....,IP~

I ~

!J t'*n1r-rn,,11

) nDr ,nn.11 i

I I I I

,. . .,14~ ~l.Ollf

/ Llill Al.6.M

~u rt-rt'

'---"'----==---11'-

I ~

I 0 fmL flll!UM" TO tll'!>lL EtU,INE.

1N,DMDU1t I

I I DIAGRAM DIAGRAM

• HE.AT E.XC.HANG.ER LOOP AUTO PRESS SWITCH CONN I

\NH

(*_~ElldtlSD)

I I F'-F KALl.*l'*l'*r DIAGRAM.-

ORI FLC.E FLOW METER. CONN NU I

• lUtTU.~

I f KIILI *t"../1' E.ltf PT Al wanD td.*H:U:. MT£~ ~ :.U,OlD * - ~ "

• lt1DICATE5 FITTltlC.b 3UP~!.f£D eY ~ " " ,urc.

SU Nffl.'lo t,1A.~\IU.,.IOU6 Ql,I Drlt6, .... K I r .._.._ ~

• • a.

t,~trlf~

.* <\~ - ., .. - .. .!-4" OI.) M,:, C.J-.11,(fo["'(!'f ;j

' \. '.' ... , .. *:* lot~ "\ ... : ., i-.

I "~~_u_"_*:~-~-----l '-:~~-~~-~-~---- ~ - c:: - ---- - .-- :1 ,**~'!""" - ; i

... -.~ . ,.

I L. -- - - c - *- '"--~----= --=-==---'--=-=--*--c _1*-----=----:..::= _:;_-:_:c-::.:c,.::--_:__:__:__-=*.::;::.:c"--=:=::c::-: _:::._--::: ..._ _ ,

,_.._,n. *Mo*te*i*~

1-r'Y 'FM(FA*M,~

PART PLAN C.

t ll~JQII\.DllT£TC~

n~-0-7 II

w /

I I

• !'I:'""*~*-

I b ..., ' : ' : , t

-p~,.:,~~

• _-\~~::::; _.-

--V -,~q-:--

1 L,.,,.l'S.Pl'Ol.~*--*--------....:a-----,----"""".'-t I A*A

......:;:.*r. ... .=.;,.a*

)006,1'11 {TYP) YARD - WATER & FIRE PRO_TECTION LINES SH -4 I

I DtlO:.L*'1*0"" 1 4\/l Olltl1n10;,.

- 11448 - ra I

I I 1814

?.5'- 2"'

I ,: II~

I C. :I*

1** E

... / +

(

f

,-*." 1'2S ~c,::, C.i:'M (M>>j...

i-1'2.~.:..-:.c C.FM (MA>C) -- \25 . .:,::,,:, C.~M ( ~

A .. L.:.uvE!ZQ','?)

"--"'""-i>":

• • ~*11,**,

'. *4*

I

+F1 +1111

,, G'OOO f.FM(M1l,,Q: 500C. CFM (1'11 ..) -* 5000 CAw\(Ma..1) "-:"' s:-"  !~TAKE". 0 *A,

... , -1 I

(re-~5 .r..'.) _ ~ --_ ~ __ _

L ** -

I k.

I /

T - --:-£AA SA.m(Ut

• (T'l'P)

I

,---- --!OILEII

.... .* (.

\

. . . 82JiECMING 1>\ *~. ~ 1->,---,_ M* --I -r

____\____ 'Z.tt~':!~------*-- --

' *.;, C

-:; .; ~...----1+--+llcc=~-'"T'h I IXH F"U \

1*-IS F ~,,.,\

500C\ r.,-M~~.J-~:..\!:, ...:_

/.------~+i-+----f-1

5XN ,,,,,.,

I I-iL J NOTa:

W[A.TING 501&.EA. kl.1 Vl NTI ._ .. TH>>J ('ll*Jil'A.)

-r ___-_, .-----*.l -- - , Cx I

I .*

~:

I

~

C.

rI I I -'I - .. I.

C I,._ 32'-c

• JU'-0" J~'-0"' 3 '-*0" 1

I 14 L.

A 16 PLAN EL 2.7'-0' 17 !B 19 l*VS*F*iZt

-VEk'TICAL ";tsT"RA.INT (Dr . .3) b:====c,----r;,P;;--;-LE:;-NU~~-,i :/*.

I E I( a 1'-0"

• • t.1. 53,'. ,1i'" . ' _____,-1 _ _ __.___

.  :-:,,,.:.c _J__ ___ .~,,..-------- .

8-B F1~E::> _::i ,*;t_;;. '.FA--,4. C"TYP)

..~'-:_ ~ "'C.6 '-c* Ot=:J.,t1t..&Cc s.o:ALE y.-. r-o" I NOTE~ .

6tN£RAL wons - L~HND" ~a-26A I 1-CLT S':..!.' r t,,'S. ~~TED

,. , ":.TEM lrJLlMOE:.2~-~P£.C.\F,C.A.T\ON L~.G..END

~AD SELF l\~TUJ~ DA.Mf'£2 I  ; .: :-.r::

0

. '* !I. JI',[ ,

/.~:: -:- ............~ ......

..

• _ _ _ _ _ ___:,:_.=_.:;..-=.=.*.* - * - * * * *

.:7~-

.. C.H.Jt i;>ATCQ

.\,;, CISC:HA~G,t I -*!

DETAIL A

>J:3-::AL(

AfT"':.'.;::.

I

,ETA.I~ C, I St.A.E *~*.r-r>* \ EtHILATIOl'J & AIR. CCNC.HCNING SERVI CE cUI LDI NG-SI-!. I I I * ~ :1 ;**:;,,;;,*:,* ...,.~ ......,. :1,

'.t*--* - *:

C[~CAIPflO', DlSC.11.1,TIOH P[SCRIPTIO~ Ml..'t DUCR1,r1c~

I

I h1284622-lm 06105179 042 I SURRY POWER STATION, UNIT 1 I

I

'1 APPENDIX C 1. 9 I SNUBBER CAPACITIES 1.11

)I

• 1 I

'I I

I I

• -1 I

I I

Pipe Hanger Division I 260 West Exchange Street Providence, Rhode Island 02901 Telephone (401) 831-7000 April 11, 1979 Stone & Webster Engineering Corp.

245 Summer Street

'I Boston, MA 02210 Attention: Mr. Mel Pedell I

Subject:

ITT Grinnell Snubber Load Ratings

Dear Mel:

I As we discussed on April 10, 1979, ITT Grinnell Load Ratings shown in our 1969 and 1972 catalogs for Hydraulic

• I(

Snubbers were the normal condition Load Ratings. Not published in those catalogs, was a one-time Load Rating,;.

which was available to our Engineering Staff for use on contracts whose specifications required this Loading Condition.

I The one-time Load Ratina is described as an event whose Loading on the Snubber is expected only once, after which event the Snubber will be inspected and replaced. Examples I of such events are Pipe Whip Restraint and SSE or DBE Seismic.

The design criteria used for this vintage of Snubber was a maximum design stress in the Snubber of approximately 0.45Sy

• for normal condition and 0.9Sy for the one-time rating.

I As I understand the situation, you have an OBE Seismic condition that would correspond to an occassional Loading ii Condition as defined by ANSI B31.l. Based upon this, we can

. , increase our allowable stresses for the Snubber by 20% according to paragraph 121.1.2 (A.l) of ANSI B31.l 1967 and 1973 editions .

This, in turn, will permit us to allow an increase of 20% for our Snubber catalog Load Ratings, prior to ASME III Subsection NF, if your OBE Seismic is categorized as an Occassional Load.

These increases will apply to our Hydraulic Snubbers manufactured I during the period of time in question for any of the five plants for which you are providing piping re-analysis. Our cylinder vendors*during this time were Lindco, Lynair, Tompkins-Johnson and possibly Miller Fluid Power.

I C-l I

I -,

. , Fl358 Printed in U.S.A .

PAGE 1 r' TO Stone & Webster Engineering Corp.

.245 Summer Street April 11, 1979 Boston, MA 02210 If you have any further requirements that would be of

• 1 service to this endeavor, please feel free to contact me.

Very truly yours,

• 1 ITT GRINNELL CORPORATION

/.*

R. J. Masterson, Manager Research and Development Engineering Pipe Hanger Division I RJM: jp cc: D. Brown I( R. Lundgren I

I I

-l I

I 1*- C-2 1

I 06/05/79 46 hl284622-lz I SURRY POWER STATION, UNIT 1

'I

• 1 "I.

I APPENDIX D 1.10 RESPONSE TO IE BULLETIN 79-04 1.12 I*

• 1

• 1.:

'I

.I I

I

• -1 I

I

• 1 I

I h-1284622-65a 06/05/79 041 I SURRY POWER STATION, UNIT 1 I APPENDIX D 1.12 I RESPONSE TO IE BULLETIN 79-04 1.14

• 1

-, ~elan swing check valves, sized 3 and 6 inches, are installed in the following Seismic Category I piping sy~tems:

1.17 I £* Chemical and volume control system 1.19 I Q, Safety injection system 1. 20

• 1 A detailed l1isting by line number follows. 1. 22 tines with 6 inch check valves were seismically analyzed by computer £rogram. 1. 24 I- Ihe re-evaluation of these systems using the correct valve weight is currently 1. 25 being done under the NUPIPE program. Ihe results have .shown that the pipe I stress is Rithin the allowable for all lines.

1. 26

1. 27 I

,, tines with 3 inch estimated weight, overly weights. Ihe check valves were conservative, analyzed was used by hand calculations.

instead of actual An valve incorrect valve weight has no effect on these calculations and

1. 29 1.30

'I re-evaluation is not required .

.*1 I

'I D-1 I

h-1284622-65a 06/05/79 041 SURRY POWER STATION, UNIT 1 LISTING OF VELAN SWING CHECK VALVES 1. 33 COVERED BY IE BULLETIN NO. 79-04 1.34 SAFETY INJECTION SYSTEM - UNIT l 1.36 6 Inch l-SI-79 6-RC-17-1502 1.40 l-SI-241 6-SI-145-1502 1. 41 l-SI-82 6-RC-19-1502 1. 42 I l-SI-85 l-SI-88 l-SI-91 6-RC-20-1502 6-RC-18-1502 6-RC-16-1502 1.43 1.44 1.45 l-SI-94 6-RC-21-1502 1. 46 I l-SI-242 l-SI-243 l-SI-239 6-SI-144-1502 6-SI-153-1502 6-SI-49-1502

1. 47 1.48 1.49

• 1 l-SI-238 l-SI-240 l-SI-228 6-SI-48-1502 6-SI-50-1502 6-SI-48-1502 1.50

1. 51

1. 52 l-SI-229 6-SI-49-1502 1.53 I 3 Inch l-SI-224 l-SI-225 3-SI-146-1503 3-SI-70-1503 1.56 1.57

• 1 l-SI-226 l-SI-227 3-SI-147-1503 3-SI-72-1503.

1.58 2.1

• t 3 Inch l-CH-258 CHEMICAL AND VOLUME CONTROL SYSTEM - UNIT 1 3-CH-81-1503 2.5*

2.8 l-CH-267 3-CH-2-1503 2.9 I l-CH-276 l-CH-196 l-CH-430 3-CH-3-1503 3-CH-200-152 3-CH-1-1502 2.10 2 .11 2.12 I l-CH-312 l-CH-309 3-CH-1-1502 3-CH-79-1503 2.13 2.14

-1 I

• 1 I D-2 I

I

,, hl284622-laa 06/05/79 SURRY POWER STATION, UNIT 1 46

'I I

I I APPENDIX E 1.10 COMPARISON OF SHOCKO RESULTS WITH NUPIPE 1.12 I

1*

1,

1 I

..I a

I I

I

I h-1284622-65 06/05/79 041 I SURRY POWER STATION, UNIT l I

I, APPENDIX E 1.10

.,, COMPARISON OF SHOCKO RESULTS WITH NUPIPE 1.12 I

1.15

• 1: Problem No. 747 - Fuel Po61 Cooling

,. tor this example the natural frequencies computed by NUPIPE differ from those originally computed by SHOCKO, but the computed pipe stresses do not Qiffer 1.18

• 1.19 appreciably.

substantially Ihe SHOCKO and NUPIPE stresses show similar patterns, with peak stresses at the same locations, and the maximum stress computed by more conservative than the maximum NUPIPE stress.

SHOCKO is Ihe SHOCKO

1. 20

1. 21 1.23

/I support reactions are also more conservative than original licensed amplified response spectra the NUPIPE results. Ihe were used for both analyses.

1.24

'I iupporting data follow. 1. 25 I

Problem No. 606 - Comoonent Cooling Water 1. 29

.1

.*1 Ihis example compares the results obtained from the original SHOCKO IUn with a reanalysis using the latest version of SHOCKl (which was yerified for 2.3 2.5 I the Maine Yankee Atomic Power Station). Although the natural frequencies 2.6 l

E-1 I

I

I

,. h-1284622~65 06/05179 SURRY POWER STATION, UNIT 1 041 I

I. computed by SHOCKl differ from those computed Qy SHOCKO, the resulting pipe 2.7

-,. stresses are very similar. Hhere differences between the two runs occur, the SHOCKO results are the more ~onservative. Ihe resultant forces and moments at 2.8 2.10 supports are also consistent between the two programs. ~upporting data 2.11 I follow.

I

• 1 I

• 1 I

I

• a

.*1 I

-1 I E-2 II

I ----

.,,I~=-:-::-------~----

., r4 7 R.:E;..,_ ~ COO~\ NG,

/2 rf 11../ 11 Nf'j

~

- - NuP1PE.

I. II _ _ _ S i-to::..f:. 0 /i

• 1 I\

I I\

,, / 1

.I

/I

• 1 I\

I J-II I. II l I I I

.I I I

I ~ i

• 1 I l 5 9 IS ~ 40 4:, ss 20 ii Bl qq cs I01 11q 130 140 ISO ,ss-

• 1

,-~*

I - ;E-3 ..

)

J

~

1 I * ,,

t'f\"k,,Y .. it£!...li! I I I j

i I I

i I . !,111 h

U*U o .* ,4.o o.,.

i .

o.re o**h6 MAT, s.s.

flt,1f J?o* f

• r.us. *,

I ro I'>*\

(11.f',t. 0, 000 q,, ( 'l

[C i1. 4 .. ,o' r?o1

.0 ll,75'

• re. 7(.i 1- .211., /0.

D, 8, 2.. /4,0' .-HB /).1.. .

Z.tJ, ~

I I .2.7,(j .0//2- /0,

I ------------

STONE & WEBSTER ENGINEERING eo~P('ll=IATimJ CALCULATION SHEET J. O. /W.O. / CALCULATION NO. REVISION PAGE

.f..50 REVIEWER /CHECKER /DATE INDEPENDENT REVIEWER/DI.TE

. f>R~;AREF: 'YTE SUBJECT TITLE QA CATEGORY/ CODE CLASS s v;<?,e I

., i'

! /v~O.::

.5/7"t:? C,t:;" a/"

.STREr.S P""1rA (C?&'~)

t-JlrP l PE CCLD ,-...R.<;)

I i PtMV'7' P.-9 rtrc, B-Z.8-~ 9 Rll4o05~ s/~f 1 e; I I'

! I /ZB~ tb.23 I I s /~~B .2082-

! 9 /?G B\'2..

I _u

/,5 / t:; oc:; 7D5 29b

/08/

• 1 4,-C; Z9.:7 441 .

"1- ~ 6 t::) J:,, 965

't;":,S- ./.S-.73 /524 z,,, ,4b,4--l

• 1 7/

7 ?/~

d.~? 95-3 E9 J .s- .:.!" lb::, I I I 99 /Z/Z ~69 I '='3E::,

1. /OS /~-,:(}

/tt, 9 /7.:!0 0r~

)/ 9 7.s, 202- I I 130 ZS9 2!S 19~

• 1 14,0

/S-0 40/

'?/G 943, 406

• .I /S~ . .,. -~.S- I I 4.14)( rlff'T~L sr.e~r  ;::: /?.t., ~ 7//, r P&'e" -=- // ~4,i!)

-I ..e-/c'c:>- S'#e) CK ~ V/1"' A~t-aG,t.JA~ Le z. 7~ Cl &Jo PS/

--1 I '

I

• I CALCULATION SHEET STONE & WEBSTER ENGINEERING CORl'>nnn,,..,.,

J.0./W.O. /CALCULATION NO. REVISION PAIIE I REVIEWER/ CHECkER / DATE INDEPENDENT 111£VIEWER/DATE OA CATEGORY/ CODE CLASS I

I:

l>>o~ .$;,/(')Ck 0/1 ~lV Pl P=. (OL-D f.:..~S)

• 1 I ,Pt:7/N.,. .ol'9reo a-zt9-e9

,Cy R.\l40boo s/9/79 r:y .

2.::, (-,

9 .'5°A~

pz_

I i

+~

Fx I ...;-6 ~

-i?'

Fz.

,S,,e:,~* :zee, l=x PY

250

/"9 .278 I I /~O I

,=JC 4 '7 ~ ~ ~-~

Fz

,27 F~ ~~

12 I  ;

• 1 I

I I

• 1

-1 -

I I-I I

I .... *--*** _ . ***-**- -- **-** --~.""'"'C.~~-=:!:.-.... **-*-*--** _ ...-.C~--- *****'------**--**, E-6_* ~,..,.._______ ~ -----*"*--*----*-- *- -~- .. - .-.-~-~-------*-

I CALCULATION SHEET STONE & WEBSTER ENGINEERltJEi CORl"QRATtOtJ J.O. /W.O. / CAl.CUl.ATION NO. REVISION PAGE REVIEWER/ CHECKER/ DATE INDEPENDENT REVIEWER/DATE I QA CATEGORY/ CODE CLASS I

I. I C?Y/v/th?/ c - ,,e-~.t:? l~OP\Pi::. (OL.D ¥S)

-1*

II /tl'ooc S,;,,'oc. z 9'/..2...3 69' R114obee B/9 /79 I 2./, 9.5 6, 76t, I 2 3

!Ir 9-f-2G,!3 7,638

/0,7~

I 5

~ 2~-05 2 9,.0 7

/3,S/

151-G ..

I 6 7*

2--;,:; B ~

32..£/

/6,46,

/7,20 I 0 9

= ..3,/.:Sc, 3:3,B7

/8, bZ.

/8,82.

t /0 II 36'.8

<ft'. 67' 20.9/

.22.24 I 12

/3

"'77-57

~ 3,72-23.4G'-

2f,.5~

I /~

/~

5?.. b,;;,,

6 $. eJ:}

.2 30,45 BO, t,z_

I lb

/7 59: .13- 32,81 b5,?o 3?.,, "11 I 18 b5,22, 4/, /2-( ~ 7::5/ /-f- +1,?G

.*1 ?o 73,47 4-5,85

.t I R3.s/ . 4B,61 I 22 ~O. 68 52,0'?

l? 85,.::>v 53,l<'i l ).' 4 /0, 5'} ob,37 2";/

I ~'? G P7 -

I "'

I I

r SUBJECT/ TITLE PROB

., Sh'oc.c Swoci: I o

/("r::

,,1s.£t:.t r£ c,-:t-GV

~e. ,.-;

,R!/rl :/I Go,

ti; zo I: 0,-....------1(:)

0~ l-"'IAY 7'7

"-'"~""'"'l;... *~. . -'"'""-:::::.""'s)"'

s\...._\,"'

I I """'. t,ooo I ~

~

I V)

~

V\

?DbO

-t bJ

~

"'I I 'folili I '2;,000

('

I I I

I I

I 'J.DOO

• I 1000 I

I " ., " I& ti 2.3 5"1 5'1 SG 91 i,I 5 1 ID 12. IS IB 22. SO ,;2, S5 51 (,O I )JtJ Pe /' TS .

I E-8. --

I & so,o.tD STONE & WEBSTER ENGINEERING CORPORATION Page No.

Preliminary _ _ _ _ _ _ __

CALCULATION SHEET Item----------

I 1 LocationSu~:r 1... Est. No.

5 8 (oO b 1

Cj J.O. l\o. /2-8'7{, , Z '2..

2 Subject f:;7-'C ./-Cf:.-£- A 7t!, Date / By ~ C O - ~

I 3 4 Based on

-/

;;; - / {.

SHcc. Y-:;:

r:"

CC)"'1t,-..lit:/?. fl)J.l f< I 74-'.>(;,CV 4/; 7 7Cf Checked Revised By By 5

I 6 7

8 I

I 14 I 15 16 .*- __ .-*--.. X. ..l)

• ____, ______ .., ....-- -*---0--------------*--*-**"*-- ...----*** ---* *--- - --

17 '=- IL  :..,_ C? :____ *_ . ___ *-* *-- N. ________. ____________. ---- .----*--**-------..

I 18 19  :>-

... ----- d:)*- -- -~-::.

.oo C'"

-- ---------*- ~ *\J"}----..c-*--:-'*---

---'--o*

-,e---.....l_ ................ -.... _ " ..... -- -----------.,.-::....i~8-+-+""---"-----..- .*,-- . *----. ---* .. __

~- V,

. * .... <:t" o* .. '

20 I 21 22 -: ~~~; :~O~N4-_" __ ~----~ _______ . . . . . . . . .

..,.;_ : ~:_ :; .--J~~ ;_,_:-: :_ :=-- ~=-: *-* .-: :-

23 24 25 ___ ; ._____ :_ .-* . ____,, _______ _

I 26 *-* :

27 28

-Cl J.q

--- "c 29

- - *-----***-*-* - -- ~- - ;__ , --- , - ,_ :__ '___ ---- - , . .

I __

' I * '

=a 30

---

... . . ---*r!'- -*-* --- *-  : . __ . --------. -------- . -----*--------- ----------- -----* ..

31 tv J..ci ....__ ' ..-,.-" : *-. ----------- .__ .,,_________________........... ,--*- ......._,, ______ _

I :: ~ ia{ :*)- - ~

1_ _ _ _ _ ----.----- * - * - - - - - - * - - * - - - - - --- **- - - - - - - * - - - ------

34  : O'.-:

35 --- :_ __ \./) __ ~ E-Llli * *&s **- -- *------ .  : --- ' .  : . . . . : ' . --*--****-**---: -**

I :~ --: !~-~.-~: 2:. -;.. _._.; _ _ : . -'--' --;-  :---*--------: -- . --- -*----***. **- '-------  :-- .------- *---*-**--***------- -- . ***-*

• I ::--! :-;_ _ :_~,l~------ ~~

; \1,j ,.,j . .I;

-- :-~--..-*-- ._ ...r.,. :,-- ............____ . ____ .. __

---

* .----~---- :*- - '_ : : - ' *-

--* *__ ' - ------------ -~ ----- *----

40  : , . 2 v') ~ .

41 - - --- ----8. 3..-b--------*-* .  : . .  : . . . -*- .. --** *-* - - -------. -

I 42 43

---

. ----------*---*--- ---------. *s;* . . ~-: .... *---: , _,. "" . . ------.. --

i

• '{\!*-- _... *- _----. - - ------------*- ----- :.......... *-. ---*-----*-----*,....---* -------------------- .

I 44 45 46*

-c:i .. *-*.

I 47 48 .C?

49

---

*-**** ---* " --- \i, . t ..... -.**--**

I 50

I CALCULATION SHEET STONE & WEBSTER ENGINEERING CORPORATION J.O.i'fl.0./CALCULATIOH HO. REVISION PAGE I

PREPARER/DATE REVIEWER/ CHECKER/ DI.TE INDEPENDENT REVIEWER/DATE

~-Q O }....V~ ~\°'.'n'\

SUB~ECT / TITLE OA CATEGORY/ CODE CLASS I I j "'-.><:::.~'f-;:.

~ ~'10~~\1,,"\\

- Rv~-i.. ~c\o <;\1..~q ""\>~ "!..~~~"<:\."'

~\ 5\°'\,""-

• 1  !

/

'?~\\$.

\

S\_\tt._<...,

\°'\\\

a S'M.~c.~ \

'20.t.\,

, ..... C)~.:>\o

"'-.) \.J ?\~E

\. ~ \

I I

'Z.

~

\~'2..."2.

~t:::_C\'2_

\~~'

.L\~ \ \

\~\

~\\

1* II t

I

'::::, ~~s~ '"'\,C\'2- \ \1..o~

I I

t C-,

~~s

,.g, vi

'6~~

\,7 9-, ,~, Co \o

~ss I °'

\()

A._'?:,.°'

'2.S~ 2..2~

\ \.o \

.,")

\\ s._C\I)_ ~\4. \ '2..

I- \1..

\"\

\~\'l.

\'"2_~

\1..:~C\

\1.'2..'6 I

\SO

\SO I \( 0_L\0 ~'s. L. " l

\\o \\u, \,'LC\ \\i I \~ \ \.,,C') \ \~°' \ \~~

\C\ \\....~ \ \. L.\ 4...1..o NO l;)c,\~'\

I 2..'=::, ,R2:)

(_<;!$,-"}

i*"-

\<:::;°'~ \ l....t.o

<:a \~ L-\:, \ '\_°\~ \ \c,(')

I ~\ \ ; l.o~ \'),_ 2..,\ \ ~

c:: 2.. "\ .L\. \).l..\ 1.."6~, '"?--. \ ..q

• I c;~ ~a,"-~ '<_\;°' '204

(~ C\~n C\~\...., "-\~

a* ~b sJo\ ~1._, \()~

~l -"\ '8.3. 41.~ ~a, 1*

-- c;q 60

'3:.~2

\.lo<;.

~~l:,

\~u

• 7..._Cfi,,

\\ Q 1 C:, \ 262 2S.~ \\.<;

b2. C\O, ~c;:a '241 I * * **-"--*-

... =. --~-=-*=- - - ~ ~ * - - - - - - - - E-10 f

I CALCULATION SHEET STONE & WEBSTER ENGINEERING CORPORATION J.0. /W.O. / CALCULATION NO. REVISION PAGE INDEPENDENT REVIEWER/DATE I

REVIEWER/ CHECKER/ DATE QA CATEGORY/ COOE CLASS 1  !

N~)I:__ I ~""-'-aC.,\~\..~ Q.'JN ~\..~ ~\\:.\""\°-. ~ . ,....,f--,J?.."'n ~

-1 i Ve:)\~ ~~(:)(__~

C\ 4-g_

() S::,~ .::x_ ~ \

°'-,..,.

~"'""~ lo'l:l\:, <,\~\-...o...

\-..)'-,) r' '~ ~

! b"\ ':,. ':::,. \\\

I I  !

I b( C:, 4~ 6\\ \Q'2 i

II

! ~6 \~-~°" \~'\.L.. \\\

I I 2:) i i b~ \\\\

\~~"'::,

'.~'"'-;)

-,_~

\ '""A(.

,o~

2-'"\\

2q.:::::

-I ' I RO '-'-\\:>\. '2.2-\ \ ~'uc:)

I I-I *~:-

I I

I

• I I

I,,-..

I I ,,,..,.,............,,._,..._~-=--~-~~~~-~.~~~~~~~- E-ll

I CALCULATION SHEET STONE & WEBSTER ENGINEERING CORPORATION J.O./l!i.0./CALCULATION NO. REVISION PAGE I

PREPARER/ DATE REVIEW ER/ CHECKER/ DI.TE INDEPENDENT REVIEWER/DATE

~~ s.\q_\,p\

SUBJECT/ TITLE QA CATEGORY/ COOE CL ASS

~~Ob. C:.O'° I

I 'c,1._;:,c.._,\..\1e.~  ;\1o\*,1=, ~~ ,~..,.,~

i iI t-...)0\,~

~

q,~uc::..~ a,- I ~I,)\...) ~01.o S:,~~c..~ \ "'-.:l'-J°i"'\.~'S:..

I I ~<;),.s,;:

*------ q__'""\A..0~ ";"\A....\""\I:\

I i

i

\ ~'I-.-=- ~"\"2.

~'\ =~\)\

~ c;o~

'2...\C\

°'~

C\1 j

i ~ ~= '2:.2,C) °'.4..~ ~~

I I I

~\?....== \c:A'l.

'r'\.Y- = 2S,01.o~~

'r~= \-::::i°'\

'2..1.,,.A..S

'f ~'::. \S:~

3;"L\

• -- r-- -**-

I h\l-= 6 \6~ h°'.~::, '"3:,Q~

'i\~= '°""' ~lo\ "3:,"3, '?

I I

,~ s;o I

'r\~=

~i~

Y'-\,-=

T"'z= 3.o.._~

12\'l

\Cl I.\

'2. li,

\\ (.(___-:. \ <;.""\ \o

\ \ \,..

lo~

..a..~:z.

I

\"\~~ '5;~1-C\'2..

~-4.

~iR I ~~=- ..0..01 ~~ '= L.\\ '2.. r~ ~,SC\

't"\.~ -:: '\\)\ \~\J1 '2.<;C)

I 't"\. \\ :: ol!.\""\S\o 4°'.~~ .L\o<;.

I 'r\~: '2.Jo \4-'::i 7.2f\

I 'i\.~:: A;~, A.'- n.R.-= c;:~~ i\Q:: S~..l\

~\J I ~'/...-=- 201 ,01.:, °'-.x I F'/ = \1..~ '2.'2-'2. \. 1.--.0...

r-z = \"-\0\ \ ~<;°' '"SL\. C)

• I "t-R:= ,,,i. ~ ~ -:. \ I.e. O.l\ w_-::: ':)°'(}__

"'"'-= 1mi b.0...1.~ \'\~.L\

I* r\. 'I = s.001.o '"°\~ ~ \ ~o,

'r\~:: \ 4-.~"'\ L.~\ L\:~'\

I,,,.,,,.. \\.~~ ~'l~"l 'r'\Q.:= 'Q ""S\\... \\~= \~1..""""\

I I <E-12...-

I STONE & WEBSTER ENGINEERING CORPORATION CALCULATION SHEET J.O. /W.O. / CALCULATION NO. REVISION PASE PREPARER/ DATE REVIEWER /CHECKER/ DATE INDEPENDENT REVIEWER/DATE I >Q ~ . c,. \~"""\~

,- SUBJECT/ TITLE

'* 0 ~ ~,~~ C,'f QA CATEGORY/ CODE CLASS 1*  !

I

~~~'°"'-at,\~\6'\) '2.v\...J-.. ~OC~\\,\,~ °':;'>p_,: \.-...)! ~ " ~

~n.r...:::.<...:01,;, c,,\e...,,-..,;.,.

I I 'n. c:i~"c.

• 1 I'*

I I \

S""'-ac.~

.'2.~~

0 "::,.~~~---*'

1...1..,°'-.\o

.. ~-...) l'"Vti:.

\ ~ ,1-. lo<;'.

I I I i

I z

~

"?:.\:.A '2.

4..0t\0 J,.lo, Q\)

~4-'l'-\

\<..,-\'~,\

7-oA~

I!

I I I ~

A ~<;-~\

~o;:;\

~<;.°'

"S;Ch '2...l.\.

'2..~,'2.~

"\o ,S,'l I r:

b\ ,"'\15 ~' . .;1 31. A S.l.o "l \v,\~ 51, ~'::.. .L\1.,A..°'\

I ~ ~'2..:,a ~C\-0'2 ~~' 6.l.\\o C\ ~~~,

° \""\ (, l-,, 0~ \

• a-l I .

I I

• I 1*

'I~

I.

I h-1284622-74 06/05/79 041 I SURRY POWER STATION, UNIT 1 I

I

• 1 I

I APPENDIX F 1.10 1.12 SEISMIC CAPABILITY OF NUCLEAR PIPING I by 1.14 Robert L. Cloud 1.16 I Robert L, Cloud Associates Inc.

Menlo Park, Calif, 1.17 1.18 May 1979 1.19 I

I I

I I

• I I

I I

I

I

• h-1284622-7 5 06/05/79 041 I SURRY POWER STATION, UNIT l 1.12 I APPENDIX F TABLE OF CONTENTS 1.14 I Section 1,

2.

INTRODUCTION **

SEISMIC ANALYSIS OF NUCLEAR PLANTS.

Page F-1

. . F-4 1.17 1.19

1. 21 I 3.

4.

P!PING ANALYSIS

• ANSI B31,l CODE, .

F-6 F-8 1.23

1. 25 I 5. B31.l AND LATER CODES , . . F-15 1.27

6. SEISMIC PERFORMANCE OF POWER PIPING F-18 1.29 I 6.1 6.2 Long Beach Steam Station **

Kern County Steam Station.

F-19 F-20

1. 31 1.32 I 6.3 6.4 6.5 Alaska Earthquake of 1964 **

San Fernando, California, 1971

• Manaiua, Nicaragua, 1972 * * * *

• F-22 F-25 F~21 1.33 1.34 1.35 I 7.

8.

BASIS FOR SEISMIC CAPABILITY OF POWER PIPING.

CONCLUSIONS AND IMPLICATIONS FOR MODERN NUCLEAR PLANTS.

• F-29

• F-33 1.37 1.39 I 9. REFERENCES. F-35 1.41 I LIST OF TABLES 1.45 Table Title 1.48 I F-1 Seismic Analysis of Nuclear Plants 1.50 F-2 Seismic Analysis of Piping Systems 1.52 I F-3 Damaged Equipment at the Enaluf Power Plant 1.55

-I I

I I F-i I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I 1, INTRODUCTION 1.9 I ~efore the development of the ANSI B31.7 Code for Nuclear Piping in the-late 1.10

~I 1960's and £ubsequent inclusion of piping under the provisions of Section III, all nuclear safety class piping Ras designed ASME to Code, meet the 1.11 1.12 requirements of the ANSI (formerly USAS) B31.l Code for Power Piping. as a 1.13 I result, many of the :operating nuclear power plants in the United States today I were gesigned and built to meet the provisions of the B31.l code. 1.14 I A general review of the methods applied to the seismic analysis of B31.l 1.15 safety class piping is given including reference to the historical evolution 1.16 I of these methods. Ihe B31.l code itself is discussed and it i~ demonstrated 1.17 I that, contrary to the belief of many, ihis code rests on an advanced technical base; £ufficiently advanced in fact that very few changes had to be made, 1.18 1.19 I other than notation, io upgrade it to the B31.7 nuclear code and then to Section III, Ihe older ASME piping code, unlike that for vessels, contained all 1.20 1.21 I the main features of current codes. fower plant, chemical plant, and refinery 1.22 piping designed to B31.l is not outdated and Qehaves very well in earthquakes. 1.23 I

• I Ihe available data on performance of piping in seismic events are reviewed, and it is shown ihat ~ngineered piping systems have performed extremely well 1.24 1.25 I in power plants that have ~xperienced substantial earthquake-induced ground 1.26 I

F-1 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER ~TA!ION, "

UNIT 1 I

I motion.

which Ihis outstanding performance has been exhibited the piping even at systems Here designed for seismic lo-dings far less severe plants in 1.27 1.28

-,. than current criteria would associate ~ith the ground motions actually 1.29 experienced. It appears that piping systems engineered for the pressure and 2 .1 I temperature conditions typical Qf power plants are inherently resistant to the 2.2 effects of seismic-induced motions their supports, whether or not such 2.3 I

Qf effects are specifically addressed in the design process.

I fower plant piping always has been designed to demanding standards. Hith the 2.5 introduction of nuclear power, these standards have been maintained and strengthened to some degree. Ihu9, it is reasonable to expect that piping 2.6 I systems in nuclear power plants that may gxperience earthquake motions will 2.7 I perform as well as have the piping systems in non-nuclear power plants.

I tarly in the introduction of nuclear power, a major development effort began 2.8 in methods of dynamic analysis of structural response to earthquake ground 2.9 I motion. Ihis development was focused, almost exclusively, on systems Cof 2.10 piping and 2 upporting structures) conservatively assumed to respond in a 2.11 I linear elastic mode. Qn the other hand, criteria of allowable piping stress, 2.12

.*I which had been developed before remained relatively unchanged 2 eismic by* the loadings s~bstantial were of major interest, gvolution of methods of 2.13 2.14 I dynamic analysis. ,S_tress criteria which had been rationally and 2.15 I

F-2 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I conservatively developed for relatively liell-defined pressure-temperature-time conditions were* applied with minor modifications 10 the less well-defined and 2.16 2.17

• 1 .very different earthquake conditions.

I In consequence, the inherent seismic resistance of nuclear power plant piping 2.18 systems designed garlier in the evolution of dynamic analysis may go 2.19 I unrecognized. fince systems for which earthquake effects ~ere not even 2.20 I considered in design clearly have ~ubstantial resistance to such is effects, unwarranted to only judge the seismic safety of a ~articular piping system it 2.21 2.22 I by the particular ground motion specified and analytical methods ~sed to 2.23 predict response at. the time of its design. Io illustrate, there was a period 2.24 I when algebraic combination of intramodal seismic effects iof earthquake ground 2.25 I motions in differing directions) was common practice throughout the industry.

Ihis technique could either overestimate or underestimate earthquake loads on 2.26 I a piping system. In the limit, it might indicate essentially zero earthquake loading; i.e., equivalent to Qmitting earthquake from the design conditions.

2.27 2.28 I Qbviously this is in no way the same as designing with zero resistance to 2.29 earthquake. Qn the contrary, the system, designed to conservative criteria 3.1 I for pressure-temperature-time £onditions, would certainly be resistant to 3.2

• I substantial earthquake ground motion as may be ~een power plants discussed herein.

by comparison with the 3.3 I

I F-3 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I* £* SEISMIC ANALYSIS OF NUCLEAR PLANTS 3.5

• 1 Iable F-1 shows a rough chronological development of some of the main features 3.7 of seismic design and.analysis methods for nuclear plants. Ihe first plants 3.9 I were designed with static methods using lateral force coefficients as static loads in the manner of various building codes. Ihese plants were, in the 3.11 I main:, built in regions of low seismicity.

I ~ynamic considerations were introdiced at about the time plants were built in 3.12 I regions Qf higher seismicity. In recognition of the amplified response 3.14 possible when shaking motions have frequencies gt or near the natural 3.15 I frequencies of buildings and equipment, design ground response ePectra were 3.16 I introduced for design.

application of response spectra

~everal papers methods gre that describe contained in the derivation and the section on 3.17 3.18 I Seismic Analysis of Reference 1. Ihis reference wa~* compiled to provide technical background for the advances and changes of yarious codes for design 3 .19 3.20 I and construction of pressure vessels and piping, especially for nuclear applications. As such, the key papers that influenced the development of 3.21 I nuclear

• seismic technology by sei,smic ePecialists such as Newmark, Hall, 3.22

• I Clough, Cornell, and others are reprinted conveniently in one place. 3.23 I

I F-4 I*

I

I 06/05/79 042 h-1284622-76 I SURRY POWER STATION, UNIT 1 I

I Io obtain the seismic response of piping systems, it is necessary to study the passage Qf ground motion through the soil, buildings, and equipment, all of 3.24 3.25

• 1 which cause modifications Qf the motion before it reaches the piping. 3.26 Qriginally, design response spectra were applied to piping in the simplest way 3.27 I considering j;_he first mode of each span and taking the response directly from 3.28 I the ground spectrum. I.his approximation was an improvement over purely static methods, but is quite simplified ~ompared to later methods.

3.29 4.1 I fubsequently in the 1960's, the effect of building motion on piping systems 4.2 I was incorporated into the design process on an industry-wide basis although 4.3 the concept had Qeen developed much earlier.<2> ~onceptually, this is done by 4.5 I analyzing the building for the effect of ground motion and geveloping new 4.6 I spectra at the floors and walls of the building where piping is supported.

practice this was done at first using records of actual earthquakes, Taft, In El 4.7 I Centro, etc, normalized to the design acceleration level chosen for the site. 4.8 I.he accelerations were applied to lumped mass building models in a time- 4.9 I history fashion. At first, very few masses would be used to represent the 4.10 building, say less than 10. Also approximate methods were devised to obtain 4.11 I the effect of building amplification on j;_he design spectrat 3 > directly without 4.12

-I a time-history analysis of the building. Qesign floor spectra were by these means and used for several plant designs.

developed 4.13 I

I F-5 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I In the 1970's, analysis were made.

several major changes in methods of nuclear plant seismic Ihe key changes were a standardization of design ground 4.14 4.15 I spectra, a requirement for three-directional ~nalysis, and use of increased 4.16 damping values. Ihe net effect was a more rational approach. to seismic 4.17 I analysis, but in any given case, ~omputed seismic stresses tended to be 4.18 comparable to those obtained by the more approximate methods since the higher 4.19 I damping compensated for the additional imposed motion. In any event, this 4.20 I appendix is addressed more to B31.l plants and not be discussed further.

subsequent developments ~ill 4.21 I

J. PIPING ANALYSIS 4. 23 .

I

~eismic analysis of piping systems in nuclear plants has also undergone an 4.24 I ~volution, outlined in Table F-2, consistent with the growth and gevelopment 4.27 I of seismic static analysis methods using for a

the plant as a whole.

constant lateral force tarly methods were based on coefficient that was a 4.28 I ~pecified fraction of the total mass of that part of the piping under 4.29 consideration. As mentioned previously, when spectra were first used, 5.1 I spectral accelerations consistent with span frequency were ~pplied to the 5.2

• I piping system.* Ihese would be applied normal to the plane of the pipe, in the worst 'direction' and combined with a vertical component.

i.e., 5.3 I

I F-6 I

I

I h-1284622-76 06105179 042 I SURRY POWER STATitiN, UNIT 1 I

I 1ater, amplified modal floor response spectra analysis was applied to safety class piping as

~pectra became available and were specified.c~> Ihe 5.4 5.6

• 1 application of this approach varied between different organizations and with the times. Although the fundamental steps and basic mathematics were 5.7 I generally common to all, certain choices had to be made in combining responses 5.B for each direction and each mode. Ihese combinations are in a sense arbitrary 5.9 I since the modal response spectra analysis method relinquishes time as a 5.10 I parameter and relationships with respect to time, including phasing, are lost.

I ~ome analyses have been done by evaluating each of three directions separately 5.11 and combining ~ontributions from each direction. In many cases the horizontal 5.13

.I direction that causes the highest stress is combined with the vertical and this 2-D Rlanar response becomes the basis for evaluation. 5.14 I

I Ihe directional combinations have also been made in other ways.

various response quantities are signed, algebraic summation of responses

~ince the from 5.16 I each direction liithin each mode .has been done. Analyses have been completed 5.18 using other options, SRSS and absolute sum. Ihe latter is probably overly 5.19 I conservative.

-1 After combinations have been made so "that the response for each mode is 5.20 I complete, the sum Qf all the modal responses must be obtained. Analyses have 5.22 I

r-7 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT l I

I been completed using several different methods of combining these responses.

Ihe methods include a straightforward "square-root-of-the-sum-of-the-squares" 5.23 I or SRSS, the absolute value of the ~ingle maximum modal response plus the SRSS 5.24 of the remaining modes, and other combinations including absolute values of I ~esponse of closely spaced modes plus the SRSS of the remaining modes. 5.25 I Ihe general impetus for the advance of seismic analysis and evaluation methods 5.26 I came from a widely felt need both Rithin the industry and regulatory to better understand seismic behavior of piping and equipment.

agencies As results 5.27 5.28 I became available from development activities, they would be used for specific plant analysis. Ihe impetus for doing so would as often come from the utility 5.29 I or the manufacturer as from the regulatory agency. It was a period of rapid 6.2 technical growth in which all groups concerned with the issue participated.

I I ~. ANSI B31.l CODE 6.4 I friar to the appearance of the ANSI B31.7 and the ASME Section III Code, all 6.5

~afety class piping was evaluated according to the ANSI iformerly USAS) B31.l 6.8 I Code for Power Piping. [or the present discussion, the 1955 and 1967 versions 6.9 I of this code are the issues of interest.

in B31.1 Ihere was little or no basic between the 1955 and 1967 versions.

change Ihe 1955 version however was a 6.10 6.11 I major departure from the_prev~ous issue of 1942 and supplements. In fact it 6.12 I

F-8 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I was in philosophy the 1955 were version established of B31.l for the in which the basic rules and technical design of power piping that are 6.13 I essentially in existence today.

I Ihe advanced features and underlying technical sophistication of the B31.l 6.14 Code have gone xelatively unnoticed in this era of rapid technical change and 6.15 I innovation. Ihe B31;1 approach, first established in 1955, contained 6.16 I provisions for limiting the thermal strain range; recognized ihe self-limiting nature of thermal stress; contained design rules for low cycle fatigue; 6.17 I incorporated ihe maximum shear stress theory; and contained other 6.18 improvements. Ihe ASME Boiler and Pressure Vessel Code contained none of 6.19 I these features at that time. In fact it was not until the ASME III Nuclear 6.20 Vessel Code came out nine years later (1964) that these technical improvements I were ~pplied to pressure vessels. 6.21 I Ihe fundamental intent of piping design lies in developing a system that has 6.22 I sufficient flexibility but is sufficiently well controlled as discussed 6.23 further below. Ihe concept of controlled flexibility is the key to successful 6.24 I piping design. Ihe Code recognizes this with an entire section devoted to 6.25 I piping flexibility. Ihe approach can be seen from the following, quoted from paragraph 119.5 of the Code:

6.26 I

I F-9 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I .'..:_Power piping systems shall be designed to have sufficient flexibility to prevent pipe movements from causing failure from overstress of the pipe 6.28 6.29 I material or anchors, leakage at joints, or detrimental gistortion of 7 .1 connected equipment resulting from excessive thrusts and moments.

I [lexibility shall be provided by changes of direction in the piping 7.2 through the use of bends, loops, or offsets; or Rrovisions shall be made 7.3 I to absorb thermal movements by utilizing expansion, swivel or ball joints, I or f.Orrugated pipe." 7. 4 I ~xplicit guidance is given to obtain balanced systems and to avoid problems of 7.6 strain concentration caused by non-uniform flexibility. In this connection 7.8 I the concept of elastic follow-up is discussed. ~esign configurations 7.9 I vulnerable to strain concentration are explained and cautioned against.

I Ihe phenomena of low cycle fatigue are accounted for in the design of B31.l 7.10 piping systems also. Ihe basic allowable value of expansion stress is 7 .11 I multiplied by a factor which is related t-0 the number of stress f.YCles. Ihe 7.13 factor functions as an allowable stress reduction factor due to fatigue I service. Ihe values off are given below, where N is the number of stress 7,14 I cycles.

I I

F-10 I

I

I h-128~6Z2-76 I

I I tl 7.17 I 7,000 and less 1. 0 7.19 7,000 to 14,000 0.9 7.20 I 14,000 to 22,000 0.8 7. 21 22,000 to 45,000 0.7 7.22 I 45,000 to 100,000 0.6 7.23 I 100,000 and over 0.5 7. 24 I Ihese stress range reduction factors are based upon tests of full size pipes 7.28 made by Harkl.< 5 > tlot only is the basic fatigue process considered, but also 7.29 I the deleterious effect on fatigue strength of various fittings, glbows, tees, 8.1 etc. Ihis is accomplished by a requirement to multiply the basic components 8.2 I of the expansion stress by ~stress intensification factors" denoted by i. Ihe 8.4 I numerical values of i were also derived from full scale tests and are given in the Code. Ihe stress intensification factor bears only a nominal relation to 8.5 I the stress concentration factors of glasticity; rather, i for a given fitting 8.6 is related to the ratio of the fatigue gtrength for the fitting to that of 8.7 I straight pipe. It is in fact a fatigue strength reduction factor. 8.8 I Ihese various fatigue considerations have been condensed and codified in 8.9 I apparently simple terms; Qut it is important to keep in mind that the approach 8.10 I

F-11 I )

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I has a basis in full scale testing and, where they are conservative. It is also true that 2 implifications even today have been made, with apparently 8.11 8.12 I inexhaustible computer resources available, extraordinarily complex structure and in a single a single nuclear piping plant 2 ystem the is an safety 8.13 I class piping might Iesolve down to as many as 90 to 100 piping problems. It 8.15 can be seen the simplifications are not only desirable, they are necessary.

I I Although an evidently straightforward consideration, the use of the maximum shear stress instead of the maximum normal stress Casa limit of strength) is 8.16 8.18 I wortµ mentioning. Ihe advanced technical nature of B31.l can be better 8.19 understood when it is realized that the widely accepted ]oiler and Pressure 8.20 I Vessel Code used the less accurate maximum principal stress theory up until 1964.

I I Ihe Code has a brief paragraph applicable, must be considered; however, no that states explicit that earthquake loads, when guidance is provided.

8.21 8.22 I Ihis matter would ordinarily be left to the designer. However, in nuclear 8.24 practice, the magnitude of design basis earthquakes is established as part of I the licensing process. [urther, the methods used to seismically qualify a 8.25 plant are subject to regulatory body approval, so this combination of I Iequirements governed seismic design of B31.l piping on nuclear plants. 8.26 I

I F-12 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I As discussed previously, in all except the very early plants, a seismic ground motion in the form of zround spectra and appropriate acceleration levels would 8.27 8.28 I be specified. Ihis motion would be. applied to the buildings and 8.29 amplifications of the*ground motion at various levels throughout the buildings 9.1 I would be computed in the form of floor response spectra. It is the latter 9.2 that were used as design bases for nuclear piping.

I I Ihe qualification of large piping systems of safety class categories is nearly always done by means Qf a computer analysis. A dynamic analytical model of 9.3 9.5 I the piping system is derived in which the mass of the system is concentrated at a finite number of mass points and the flexibility of the system is 9.6 I represented by springs connecting the masses. Iystem damping is included as 9.7 viscous damping, normally with highly conservative numerical values of 0.5 or I l percent of* critical damping. Ihe completed model is then analyzed for the 9.9 appropriate seismic spectral motion on the computer.

I ~sually, one amplified floor response spectrum is used as an input 9.10 acceleration at every ~oint of support or connection to the building. Ihis 9.12 I simplification can be an important conservatism especially for piping systems I traversing different vertical levels or different buildings.

piping Ihe model of the system is passed through the computer several times to account for all 9.14 I directions of motion and both the operating and design basis earthquakes. 9.15 I

F-13 I

I

I 06/05/79 042 h-1284622-76 I SURRY POWER STATION, UNIT 1 I

I Inertia forces are developed first vibration, then- the contributions of each mode for all directions within each mode of sre combined to obtain the 9.16 9.17 I total force. A current controversy lies in the fact that force combinations 9.18 within each mode were in some £ases combined algebraically so that some loads 9.19 I would subtract from the total. Ihe alternative would be to combine forces in 9.20 such a way that subtraction could not occur, which is the case if an SRSS I sPProach is used. 9.21 I tffects of the inertial forces are combined with effects from relative 9.22 I building displacements, gravity <weight> effects, and internal/external 9.23 pressure loadings on the pipe.

I 9.24 I Hhen load combinations are complete, bending piping system are computed according to B31.l equations.

moments and stresses in the

~asically, twice the 9.26 I maximum shearing stress in the pipe due to bending and tension is computed and limited to 1.2 \ for the OBE and 1.8 ~ for the DBE in a manner very 9.27 I comparable to ASME III today. \ is the tabulated value .of allowable stress 9.28 as provided by the Code, in the hot condition. In B31.l, Sh is based on the 9.29 I

,. lower of 5/8 Yield Strength or l/4 Ultimate Strength at operating temperature, AXcept certain austenitic materials temperatures up to 90 percent of are yield permitted strength Sh values Qecause at operating of the greater 10.1 10.2 I toughness and ductility of these materials. Ihese values of allowable stress 10.3 I

F-14 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1.

I I are the lowest in use for any piping in the United States.

nuclear piping has higher allowables, as does B31.3 Refinery ASME III Class 1 and Chemical 10.4 I Plant Piping. ~31.4 respectively permit allo~able and B31.8 stresses for up Gas to and Oil 72 percent Transmission Qf the piping ultimate 10.5 10.6

• 1 strength. Ehen nuclear plant piping was moved under the aegis of ASME 10.7 Section III, Safety Class 3 .snd 2 continued to be designed by B31.l; however 10.8 I the allowable stress for the faulted plant condition ~as raised to 2.4 Sh. 10.9 I Hention is conservative made nature of certain of these facts as an observation of of the B31.l ~ode even when compared to other codes that .10.11 the 10.10 I

,, use the same calculational basis.

Ihe method of stress evaluation just described is a simplified overview of the 10.12 I: acutal process. Qne of the more troublesome aspects of the work is accounting for elbows, tees, .sttachments, and other stress raisers. Ihis is accomplished 10.13 10.15 I by a mandatory multiplication of the stress at tabulated "stress j,ntensification factors" or i factors.

points of concentration by 10.16 I

2* B31.l AND LATER CODES 10.18 I

I Ihe first version of the B31.l second edition ~as published in 1942.

Code was published in 1935, and a revised Ihen a third edition was issued in 10.19 10.22 I 1951. Ihis was a period of rapid development in piping design methods and it 10.23 I'

e F-15 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT l I

I was found 2esirable to publish another revised edition of the Code in 1955.

brief history is given in the foreword to the 1955 edition of B31.l.

b Ehat is 10.25 10.26 not mentioned there, however, is that the 1955 edition of the piping code had several far reaching engineering improvements, which have been mentioned 10.27 earlier herein.

Ihe development of the 1955 edition and some of the changes therein are 10.28 I discussed in References 6, 7.

1967,

~ubsequently, a new edition was published and although there were a number of changes ~nd minor revisions, no new in 11.1

11. 2 I concepts were introduced.

I

,. In 1969 the ANSI B31.7 Code for nuclear piping was first published.

philosophy of this code was to have nuclear primary system piping designed similar criteria ~s nuclear primary system vessels, Ihe basic Ihis required B31.7 to to

11. 4

11. 6 I adopt similar approaches to

~rovide comparable margins the different possible types of failure with Section III of the ASME Code * .Ihe modes of and

11. 8

• I failure for which protection is provided explicitly by the stress evaluation

~rocedures of Section III are bursting, excessive plastic deformation, 11. 9 I

,, progressive distortion, ~nd thermal and mechanical fatigue failure.

other possible types of Qf course failure are considered in other areas of the Code, specifically in materials ~election and fabrication guidelines.

11.11 11.12 I}

I' F-16

'I

I h-1284622-76 042 I SURRY POWER STATION, UNIT 1 I

I Ihe obvious approach vessels was ~o attempt to to develop a piping code comparable to S~ction III for adapt the existing B31.l Code, which was the 11.13 11.14 I approach taken. However, as it turned out, the B31.1 Code already contained almost every provision of Section III, in a different format perhaps, but all 11.15 11.16 l the basic concepts were in place. Ihe divelopment of B31.7 then was a matter 11.17 I of recasting the original provisions of B31.1 into ~ection I I I one technical addition format.

was required that could be considered a new concept, Qnly 11.19 I and that was the sddition of consideration £or through pipe walls. In certain situations radial temperature gradients or processes this could be an 11.20 11.21 I important consideration, but in nuclear plants it Iarely determines the 11.22 acceptability of piping systems. Ihe net result is that B31.7, even though 11.23 I different in appearance and permitting slightly thinner ~ipe walls due .to 11.24 I higher Section III B31.l Code.

S values, was not fundamentally different Ihis was especially true in the most important aspects of from the piping 11.25 t design, the limitation on the main expansion strain Iange and thermal fatigue considerations. Ihe stress indices, C2 and K2 of B31.7 Cand Section III), are 11.26 11.27 I even generally related to the old i indice~ of B31.l. 11.28 t-

., 12.1 I

I I~

F-17 I

I h-1284622-76 06/05/79 042 I SURRY POwER STATION, UNIT 1 I

I Ihis relationship and other background on the development of the current ASME Section III Piping Code are in a forthcoming edition of the ASME Criteria 12.4 12.5 I Background Booklet,<e>

f Ihe essential point of the preceding discussion has been to make clear that 12.6 safety class piping 2esigned to meet the requirements of the older ASA B31.l 12.7 I Code would almost without exception also meet the Iequirernents of the latest 12.8 I version of the ASHE Code.

however. Ihe B31.l Code A little more needs to be said about seismic design of 1967 and 1955 clearly spells out that seismic 12.9 12.10 I stresses are to be considered but does not say how. [or nuclear plants built 12.12 to those codes, however, this is not significant for present purposes ~ince 12.13 I rigorous seismic analysis was completed for these plants to satisfy licensing I Iequirements. 12.14 I f, SEISMIC PERFORMANCE OF POWER PIPING 12.16 I Although Qf there appear to be no controlled experiments of seismic performance actual piping systems, there is, nevertheless, a surprising amount of very 12.17 12.18 t

interesting gata on the response of power piping to actual earthquakes.

the following, power plant behavior in 1972, San several recent earthquakes, In Managua Fernando 1971, Alaska 1964, Kern County 1952, and Long Beach 1933, 12.21 12.22 1*

I t F-18 I

I h-1284622-76 06/05/79 042 1: SURRY POWER STATION, UNIT 1 1,

I is discussed.

rather No attempt has been made to sort or classify the observations, all significant data that could be found in a ghort time are reported.

12.23 12.24 I

-1 fossibly the most interesting of the observations are those pertaining to the Kern Steam Station in the Kern County earthquake, .and the Enaluf Steam Plant 12.25

12. 26

• in the Managua earthquake. ioth these plants were designed by conventional 12.27 I procedures, both underwent severe ground shaking, *nd neither suffered any 12.28 I failures of the piping systems. Ihe maximum ground accelerations were estimated to be as high as possibly 0.6 g at Enaluf, which was *djacent to the 12.29 13.1 r main Plant.

fault causing the quake, and about 0.25 g for the Kern County Steam Iime and again it is seen that piping systems correctly designed for 13.2 I normal service are relatively impervious to earthquake damage. Ihe basic 13.4

'I: concept of controlled flexibility built into power piping renders sy~tems more resilient ihan the buildings from which they are supported.

these 13.5 t:

Q,l. Long Beach Steam Station 13.7 1,

1 Ihis station was located on Terminal Island in Long Beach, California, about 13.8 I ~ miles from the fault that caused the Long Beach earthquake on March 10, 13.9 I 1933. Ihis earthquake was of magnitude 6.3 and caused accelerations at the site of the steam plant eitimated to be about 0.25 g. Damage in Long Beach 13.11 13.13 I

I' F-19 I'

,I

I h-1284622-76 06/05/79 042

1. SURRY POWER STATION, UNIT 1

• I I itself was very extensive, but there were no actual accelerometer records of the garthquake,

• 13.14 I

At the steam station site there were actually three independent plants. 13.15

1 E.lant 1 consisted of one unit and was built in 1911. It was either out of 13 .17 I service or in intermittent damaged in ,the earthquake.

service in 1933 and the building was severely Elant 2 consisted of two units and was built in 13.19 I 1922. Elant 3 consisted of three units and was built in 1928.

subsequent information was obtained from W.F. Swiger of the Ihis and Stone & Webster 13.21 I

Engineering Corporation, gesigners and builders of the plant. [or other 13.23 reasons it was necessary to re-examine the design of the plant at a later time and it was getermined the plant structures were designed for lateral static 13. 24 I forces of 0.2 g.

mats supported

[oundations of both plants were heavily reinforced by wooden piles 50 to 60 feet long driven to hard sands.

concrete lio 13.25 13.27 information is available on seismic design of the piping and equipment, but considering the state of the srt it is probable that either the 0.2 g static 13.28 design was used, or else seismic design was not £Onsidered. 13.29 lieither plant, that is to say, none of the five units, suffered any 14.1 I significant damage.

reported;

~ome minor damage such as to lighting fixtures however, the steam plants either Qperated through the earthquake or was 14.2 14.3 I were shut down due to loss of load and were back ~n Qperation the same day. 14.4 I

I F-20

I 06/05/79 h-1284622-76 042 I SURRY POWER STATION, UNIT 1 I

I Ihe important point is that five steam units designed with at most static methods to a g level i0,2) probably lower than actually experienced C0.25) 14.5 14.6 I were undamaged and, in particular, no piping was damaged.

I Q,2 Kern County Steam Station 14.8 t Ihis oil-fired 60 MW steam plant was designed and built in 1947-8. It is 14.10 I located on the Kern River near Bakersfield, California, the gpicenter of the July 21, 1952 Kern County earthquake.

about 25 miles from 14.12 I'

i Ihis earthquake, sometimes referred to as the Taft, the Tehachapi, or the 14.13 Arvin-Tehachapi, was Qf magnitude 7.7. It was the most severe earthquake 14.15 I recorded in the continental United States since that of 1906 in San Francisco.

It occurred along the White Wolf fault south and east of Bakersfield. Qamage 14.16 14 .18 I was extensive in Bakersfield and to oil production facilities in the area and to the Southern Pacific Railroad. Ihe railroad tunnel near Bealville crossed 14.20 i the fault and was destroyed,c1c>

I Ihe structures of the plant were designed for 0.2 lateral load on a ~tatic 14.22 I basis with stress limits increased earthquake loadings.

by 0.33 ,for combined dead, live, foundations are soil bearing footings at shallow depth.

~nd 14.23 14.24 I

I I F-21 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 t

I Anchorage systems of all major equipment including switchgear Leviewed for resistance to lateral loads.

were carefully 14.25 14.26 I Ihis is one of the first electric power plants to have piping gesigned by 14.28 I dynamic analysis. Ihe Biot<2> smoothed response spectrum was used for the 14.29 I design of the main steam ~nd boiler feedwater piping.

was normalized to 0.1 g at ground level and 0.3 g at the Ihe response spectrum top floor of the 15.2 15.3 I buildings, with linear interpolation at other levels. In this amplified.response spectra was available at every floor, even though it ~as of way an 15.4 15.5 I narrow band and heavily damped compared to spectra used for nuclear plants.

Ihe spectra were applied for the steam and feed lines by calculating the first 15.6 I natural frequency of each span of pipe considered as a simply supported beam, 15.7 I then applying the appropriate lateral g force.

of the main piping, psuedo-static

~ased on the dynamic analysis g loads were developed for Qther piping 15.9 15.10 I systems. Ihese loads were also used to design guides and stops loads acting on the ~upporting structure.

and to It is of interest to note that some find 15.11 15.13 t guides and stops on the main steam line had gaps or Iattle space of as much as 15.14 I 2 inches,

I An acceleration record obtained epicenter than the Kern County Plant.

at Taft, California, was farther from the tlaximum acceleration recorded at Taft 15.16 15.17 I was 0.17 g and it was estimated that ground acceleration at the plant site was 15.18 I

t F-22 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I a very substantial 0.25 g, significant damage.

Ihe plant operated through the earthquake with It was shut clown after the earthquake clue to loss of load no 15.19 15.20 I but was returned to service in a few hours.

oil Ihere was some minor damage tank seals and a small house turbine thrust bearing, QUt no damage at all to 15.21 15.22

'I to piping systems.

complete seismic Ihis is a very clear and graphic protection that is example of the provided by even the most rudimentary seismic design procedures (by today's standards). Qf course, there was almost even 15.23 15.24 15.25 l~

greater inherent reserve in the piping systems due to their natural £Ontro~led flexibility.

15.26 I 2,3 Alaska Earthquake of 1964 15.28 I

I Ihis earthquake of 8.4 magnitude was the largest recorded earthquake times. It was centered east of the city of Anchorage, near the Qf modern town of 16.1 16.3 I Valdez. Ihere was widespread destruction throughout the area, not only from garth vibration, but from the tsunami, the failure of poor soils, and fire.

16.4 16.5

• I iome observations by knowledgeable engineers of power piping are available, QUt there is more detailed information that is yet to be obtained. In a panel 16.6 16.8 I discussion on the Nuclear Piping Code, observations were noted of Rower piping behavior by an experienced piping engineer with a leading 16.9 I Architect/Engineer 111 >, tlr, Fred Vinson reported that he reviewed the damage 16.10 I

I F-23 I

h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

1 at two power stations immediately following the earthquake.

at an air base Ihe power station in the earthquake zone had no damaged piping although there 16.12 16.13 I were some bent hanger rods," damaged lighting

~ontrol panel due to absence of anchor bolts.

fixtures, and an overturned 16.14 i

I A second power plant in the earthquake zone incurred more damage to the plant,

~lthough there was no failure of power piping. Ihere were failures of some 16.15 16.17 I equipment supports made of *malleable iron, and an ash handling line connected 16.18

• 1 with patented couplings is reported to ~ave failed due to improper support.

• 1 Ihe significant finding Of the observations of Reference 11 is that two £Ower plants rode out the Alaska earthquake with no failures of the power piping, 16.20

• a even though the exact g levels at the sites were not reported and the design basis was not given Qther than to say "very little was done in the way of 16.21 16.22 I seismic design for the 12,rotection of anything." 16.23 I A brief mention is made in Reference 10 of* the Chugach Electric Company plant 16.24 in Anchorage. Ihis fossil-fueled plant of about 50 MW was built between 1949 16.26 I and 1957. Ihe plant was designed to 0.1 g by the Uniform Building Code. Ihe 16.28 I buildings were* of steel frame construction with corrugated panel walls.

was no damage in the turbine room nor to piping and critical equipment.

Ihere Ihere 16.29 17.1 I was minor damage in the boiler room consisting of bending of some Qracing 17.2 I

I F-24 I

~, h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

1 members and appreciable damage to framing supporting the coal bunkers. tlany 17.3 piping hangers on the main steam lines were broken, but the piping itself ~as 17 .4 I u*ndamaged. Ihe plant was returned to service at full power in less than 17.5

• I 10 days.

Ihe consulting firm of Ayres and Hayakawa of Los Angeles was asked ~o review 17.7 all nonstructural damage to buildings due to the Alaska earthquake as £art of 17.8 I

the investigation performed by the National Academy of Sciences at the xequest of President Lyndon Johnson. In their report< 1 2> power plants were not 17.9 17.10 I discussed separately, rather observations Qf piping systems of all types were 17 .11 discussed on a generic basis. Ihe discussion is based on a study of large 17.12 I' modern structures located, with few exceptions, in Anchorage. 17.13 I Ihe reference report addresses general piping systems of all type*s, but mainly 17.15

f that required in modern buildings. Hith the exception of certain fire 17.16 protection piping, none was seismically designed. iecause of the broad basis 17.17 I of the report, the following paragraph is quoted girectly from the section 17.18 I entitled "Piping Systems."

I ~The overall damage to piping systems was surprisingly instances were reported where piping systems remained intact, despite low. tlany

~he 17.21 17.22 l significant structural and nonstructural damage suffered by the building.

I I F-25 I

I h-1284622-76 06/05/79 042

-I SURRY POWER STATION, UNIT 1 I

• 1*

Ior example, the plumbing pipes in the Enlisted Men's Service Club at Iort 17.24 Richardson* remained standing after the earthquake although the walls I around them ~ollapsed. ~ontractors also reported that most put back into service when pressure-testing ~evealed no leaks."

systems were 17.26 17.27 I

I Ihe general resistant.

conclusion was that piping systems are basically earthquake Iailures occur if at all at threaded fittings. Eelded steel pipe 17.29 18.2 I does not fail. Qne instance of power piping failure was noted.

pipe drain lines anchored to building walls were torn from the steam imall steam line as 18.4 18.5 I it responded to the earth~uake at the Fort Ri~hardson power plant. Ihis is 18.6 the type of unbalanced design warned against in the piping code. froperly 18.7 I detailed systems had no problems.

I Q,4 San Fernando, California, 1971 18.9 I Ihe San Fernando Earthquake of 1971 was centered in the northern part of the 18.11 I San Fernando Valley. ~round accelerations of 0.1 to 0.19 g were recorded in 18,13 Los Angeles at ~istances of 35 km and 0.37 g at Lake Hughes, 25 km from the 18.14 I epicenter. [igure F-1, shows recorded g levels for the 1971 earthquake at 18.15 I various locations near structures in the valley.

Los Angeles. Ihere was severe damage to a number of 18.17 I

I I F-26 I

I.

h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 Ii II. Ihe Valley Power Plant is a fossil fuel plant with three units on the ~ite located 2.8 miles from the line of surface rupture (Lakeview Segment) of the 18.19 I primary fault break. Accelerations at the site were estimated to be in excess 18.20 of 0.25 g based upon the location of various recordings. Ihe station was 18.22 I

,, designed to 0,2 or 0.25 g although actual details are not known.

In any event there was no damage to the plant. It was tripped off the line by 18.24 I action of sudden pressure relays and loss of load, QUt was back inside of 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s< 1 3>.

on Ihere was significant motion of the piping and seismic the line 18.25 18.26 I holddown bars came into play< 1 ~>, but Qther than insulation, the piping itself 18.27 was undamaged. Ihis is a graphic eY.ample of the basic point that well 18.28 I designed piping to ,regular commercial practice is highly resistant to 18.29 I earthquake resistant.

damage. £iping designed to nuclear standards is that much more 19.1 I Ihere were other power plants in the area at Playa del Rey, San Pedro, and 19.2 I Seal Beach ihat were not as close to the epicenter as the Valley Plant and 19.3 none of these were damaged. Ihe San Fernando Power Plant is an old hydro 19.4 I plant built in 1921 and there ~as a structural failure of the building which 19.5 I led to a penstock failure. Ihere were numerous failures transmission facilities due to cracking Qf porcelain bushings .and movement of electric of 19.6 19.7

.I I

,, F-27 I

I.

h-1284622-76 06/05/79 042 I

SURRY.POWER STATION, UNIT 1

,I poorly anchored equipment. Ihere were no power piping failures in the San 19.8 Fernando earthquake.

I

~~5 Managua, Nicarag~a, 1972 19.10 I

An earthquake of magnitude 7.5 struck Managua on December 25, 1972. Ihere was 19.12 I much damage and great loss of life. Ihe loss of life was largely unrelated to 19.14 I damage of industrial near midnight.

buildings and facilities since the earthquake occurred A report on the damage was. sponsored by £he National Science 19.15 19.16

.1 Foundation and ~everal professional societies together with the Ministry of 19.17 Public Works Qf Nicaragua< 15 >. 19.18 I

I [igure F-2 taken from Reference 15 shows the fault lines along which movement occurred running through the city of Managua. Ihe location of two industrial 19.20 19.21 I facilities, the ESSO refinery and the ENALUF Power Plant, ~re also noted.

earthquake response of these two facilities will be discussed since Ihe they 19.23 I £Ontain industrial piping systems of interest for present purposes. 19.24 I A complete accelerograph record was obtained at the ESSO refinery. Ihe peak 19.26 I. measured acceleration was 0.39 g E-W and 0.34 g N-S. Ihe design of refinery met provisions of the Uniform Building Code for 0.2 g, including tall the 19.27 19.28

.( fractionating towers, some of which exceeded several hundred feet. Ihere was 19.29 I

F-28 I

I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I almost no damage at the refinery and none to the piping systems.

jumped out of s~ddle supports and was pushed back into ~lace.

~ome piping Ihe facility 20.1 20.2 I was shut down for an inspection but was operating at full £apacity within 20.3 20.4 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> even though there was a loss of offsite power. Ihe refinery provides I a clear example of the seismic capacity of welded steel ~ipe that has been 20.5 I designed for seismic conditions, albeit statically.

I ~ased on the earthquake magnitude, acceleration record at the Iefinery, and the location of the ENALUF Plant immediately adjacent to the ca~sative fault, 20.7 I it is probable this plant experienced accelerations on the order of 0.6 g. 20.8 Ihe power plant consists of three oil-fired units, one of 50 MW and two of 20.9 I 20 MW. All three units were taken off-line by protective relays. Ihe plant 20.11 I suffered some damage but none to the piping systems.

industrial It was one of the facilities restored to service after the earthquake.

first Qne unit was 20.12 20.13 I operating in two weeks, the second in three weeks.

delayed due to turbine problems.

Qperation 0£ Unit 3 was 20.14 I

Ihe specific damage to the three units is listed in Table F-3. Note that no 20.16 I damage occurred to the piping, and that many of the problems Iesulted from 20.17 I absent or inadequate anchors. [or example, turbine bearings were lost because emergency de oil pumps were inoperative £Ue to the batteries tumbling out of 20.18 20.19 I their racks.

I I F-29 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I Ihe basic facts about the power piping are that, with unknown seismic gesign applied, but ceitainly less rigorous than used for nuclear plants, the Riping 20.21 20.22 I survived site accelerations on the order of 0.6 g with no failure. tlodern highly 20.23 welded steel piping with built-in controlled flexibility is inherently I ~esistant to earthquake damage. 20.24 I l* BASIS FOR SEISMIC CAPABILITY OF POWER PIPING 20.26 I In the previous section the performance of piping systems in power plants and 20.27 I a ~efinery during actual earthquakes was reviewed. It was shown that there 21.l were no piping failures even though ground accelerations up to 0.6 g were I gxperienced and seismic design was usually based on static analysis to the 21.2 I lower value of 0.2 g. Ihis approach to seismic design would be considered rudimentary by nuclear standards.

21.3 I In the following paragraphs, the probable reasons for the excellent 21.4 I performance of piping ~ystems in earthquakes is explored. Ihe fundamental 21.6 seismic capability of piping systems apparently derives from three sources:

I I 1. The power conservative.

piping design and construction code, ANSI B31.l, is quite 21.8 I

I I F-30 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I £, Designs that are successful for thermal expansion also provide good

~eismic capability.

21.9 21.10 I J. The large damping factors that become operative in severe shaking sre 21.12 I neglected in normal design practice.

I Iaking the above factors one at a time, it is first noted that in Sections 4 21.14 I and 5 of this xeport, the B31.l code for power piping was discussed and it was shown that the nuclear power piping codes derived from B31.l have much in 21.15 21.16 I common with the parent code. However the basic conservatism was not covered 21.17 in detail, Ihere is substantial margin provided by the design rules of B31.l. 21.18 I Ihe average stress in the pipe wall due to the design pressure is limited to 21.19 I l/4 of the tensile strength of the steel. Ihermal expansion of the pipe may cause stresses due to restraint of expansion, Qut these are displacement or 21.21 21.22 I strain controlled. Ihat is, the strains will not become larger than indicated by the associated temperature ~nd will always be stable, unlike a dead load or 21.23 21.24 I pressure stress. Ihe strain range duet~ thermal expansion is limited to a 21.25 very small fraction of the strain £apability of the pipe, considering the 21.26 I repetitive nature of the thermal expansion ioading. Ihe code attempts to 21.27 I consider all the categories of loading that a piping and maintain the pipe system will experience in a state of small strain, including the effects of 21.28 I stress intensification at elbows and tees.

I I F-31 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I However, the significance of the rules for fabrication and construction given by the code tena io be overlooked in discussions of design capability. Ihe 21.29 22.2 I provisions for sound weld design, weld qualification tests, heat treatment, inspection, and tests all combine to pr9duce piping systems as sound in the 22.3 I field as they appear on the grawing board. Ihe significance of the 22.5 I requirements for construction becomes even references that describe the results more of visible field as one inspections reads ihe following 22.6 I earthquakes. Qccasional references to failures of piping in plumbing are made, e.g., Reference 12. In these cases the problems invariably occur at systems 22.7 22.8 I threaded joints and occasionally at flanged joints. Rrought iron and cast 22.9 iron pipe also perform poorly in earthquakes. However, properly designed and 22.10 I hung welded steel power piping did not fail in even very severe earthquakes.

'I tvidently made for the pipe controlled flexibility built into well designed piping systems imparts substantial seismic capability also.

displacement due to If, in the design, provision thermal growth, the pipe is then later is 22.11 22.12 I untroubled by forced seismic displacements. Ihe provision for flexibility may 22.14 be the most important aspect of seismi~ design and is an especially important I consideration in selecting and sizing pipe hangers. It is significant that 22.16 I piping hangers were reported on one occasion to have failed 1 2, QUt the piping itself did not. iound piping material can undergo cyclic strain of several 22.17 22.18 I

I I F-32 I

I h-1284622-76 06/05/79 042 1: SURRY POWER STATION, UNIT 1 I

I percent earthquake.

for the limited number of cycles that would Qe imposed by an 22.19 I

Ihe damping associated with severe shaking is one of the most important 22.20 I conservatisms in existing approaches to nuclear £iping design. liormally 22.22 I viscous damping critical damping.

is assumed with damping factors In a large earthquake however, several of l/2 or 1 percent of energy dissipating 22.23 I mechanisms will become operative; Qrdinary material damping, impact damping, friction or coulomb damping, and plastic deformation when there sre large pipe 22.24 22.25 I motions. Iaken together, it is clear that damping ratios much greater than 22.26 design values can be expected.

I I fohm 1 b systems.

has presented a reasonably comprehensive survey of damping in reactor Ynfortunately, the data available were all for relatively small 22.27 22.28 I deflections.

amplitude of However, vibration.

there is a clear correlation of damping values with

[igure F-3, taken from Reference 16, shows the 22.29 23.1 I increase in damping with deflection for the data obtained from iests of full 23.2 scale nuclear plants. Ihere are also some data from the San Onofre Nuclear 23.3 1* Plant in the El Cajon and ~an Fernando earthquakes. 23.4 I- It is interesting to note that the San Fernando earthquake produced ground 23.5 I accelerations Qf 0.018 g maximum at the San Onofre site and damping of between 23.6 I

I F-33 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I 2 and 4 percent for deflections of ~bout 0.03 inches were measured by plant instrumentation- on the primary equipment. ~amping of 3 to 8 percent was 23.7 23.8 I reported to have been measured in pluck tests at the Tsuruga Nuclear Plant.

In general, damping that is much higher than the design value was measured in 23.9 I several tests at very small deflections ~nd it increases with amplitude of 23.10 I deflection. txtrapolating the curve yields 10 percent gamping.

of Figure F-3 to 0.5 inch deflection 23.11 23.12 I As plasticity develops in the piping even in small amounts, damping ratios of 23.13 I 10 percent and higher are definitely to be expected. In fact, there is a 23.15 major project underway at the present 17 to develop seismic restraints based on I cyclic 2lasticity of the supports. Ihe essential quality of the relationship 23.17 I between damping, acceleration level, and damage is that damage to £iping does not increase proportionately with input acceleration levels and this is due in 23.18 I large part ~o increases in damping levels as deflections increase. 23.19 I ~. CONCLUSIONS AND IHPLICATIONS FOR MODERN NUCLEAR PLANTS 23.21 I Ihe evolution of seismic design methods in nuclear power plants has been 23.22 I ~eviewed together with the development of the piping codes.

nuclear plants that meet the older It was shown that B31.1 code will more than likely also 23.25 I ~atisfy the new nuclear codes that have better quantified conservatism. 23.26 I

I F-34 I

I h-1284622-76 06/05/79 042 I SURRY PO~ER STATION, UNIT l I

I Available reviewed.

data on the actual seismic performance of power piping systems were It was shown that operating power plants do indeed have very high 23.27 23.28 I levels of seismic capability. Qf the several plants that sustained severe 23.29 ground motion from 0.2 to 0.6 g, there were no failures of welded steel power 24.1 I piping. ~onsidering the magnitudes of the earthquakes and the variability of 24.2 I the design practices, ~his is an excellent record and can only have been possible by the natural resiliency of ~ower piping.

made 24.3 24.4 I Ihe probable reasons for this natural resiliency were discussed next. It is 24.6 I believed that the main reasons are: first, the substantial conservatism of the

~ode for Power Piping, B31.l, including the provisions for materials, 24.7 I fabrication, and ~onstruction; second, that design of piping for thermal 24.8 I expansion provides inherent seismic increases very rapidly with deflection capability; levels. Ihe

~nd third, large that damping damping factors 24.9 24.10 I prevent these buildup reasons of seismic disturbances in resonant systems.

explain the remarkable performance of piping It is believed systems in 24.11 I earthquakes.

I ~ased upon the foregoing observations, it is very improbable that piping- 24.12 I related safety-problems would occur States due to seismic disturbances.

in nuclear plants in the eastern United Ihese plants have maximum ground motions 24.13 24.14 I of 0.15 g; they have been designed by dynamic analysis; ~nd all safety piping 24.15 I

I F-35 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I systems have _been specifically scrutinized.

the Kern County-plant where 0.25 g was

~ontrast this situation with say actually experienced and ~xplicit 24.16 24.17 I analysis was performed only on the steam and feed lines; or the ENALUF plant which was ~robably designed statically and experienced perhaps 0.6 g. Ihe 24.19 I contrast is simply too great; piping failures of nuclear safety systems should I not result from earthquakes in the United States. 24.20 I ~- REFERENCES 24.22 I Cloud, R.L. et al. tditors, Pressure Vessels and Piping, Design and 24.25.

Analysis, Amer. Soc. of Mech. Engrs. H,Y., N~Y., 1972. 24.26 I

I Biot, M.A., Analytical and Experimental Methods in Engineering Seismology, Irans ASCE 108 p 365-408, 1942, 24.28 24.29 I Biggs, 1964.

J.M., Introduction to Structural Dynamics, McGraw Hill Book Co., 25.1 I ~- Berkowitz, L., Seismic Analysis of Primary Piping Systems for Nuclear 25.2 I Generating* Systems, Reactor and Fuel Processing Technology, Lab, Fall, 1969.

• Argonne Natl 25.3 I

I I F-36 I

I 06/05/79 042 h-1284622-76 I SURRY POWER STATION, UNIT 1 I

I 2* Markl, A.R.C., Fatigue Tests of Piping Components, Trans ASME V 74 1952. 25.4 I Brock, J.E., Expansion and Flexibility, Chap. 4, Piping Handbook, 5th Ed., 25.5 25.6 King and ~rocker (Eds.) McGraw Hill Publishing Co., 1967.

I 1.

I Markl, A.C.R., Piping Flexibility Analysis, Trans ASME, 1955, p 419. 25.7 I ~- Criteria of the ASME Boiler and Pressure Vessel Analysis, Amer. Soc. of Mech. Engrs., 1979 Cto be published).

Code for Design by 25.8 25.9 I

~. Swiger, W.F., personal communication, May 1979. 25.10 I

I 10. Swiger,. W.F., Notes on Plants Designed Experienced 1arge Earthquakes, 1979, Unpublished.

by Stone & Webster Which Have 25.11 25.12 I

11, How Nuclear Piping Code Rules Will Influence Piping Design Today and 25.13 I Tomorrow, Heating, Piping, and Air Conditioning, June, 1970, p 69. 25.14 I 12, The Great Alaska Earthquake .of 1964, Engineering, National Academy of 25.15 I Sciences, Eashington, D.C., 1973. 25.16 I

I I F-37 I

I h-1284622-76 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I 13. San Fernando, California, Earthquake of February 9, 1971, Leonard Murphy,

~ci. Coord,* U.S. Dept. of Comm., NOAA, Washington, D.C., 1973.

25.17 25.18 I

• 1 14. Snyder, Arthur I., Damage to Mechanical Equipment as a Result of the Feb. 9, 1971 tarthquake in San Fernando, California, Seismic Design ~nd 25.19 25.20 I Analysis, Amer. Soc. of tlech. Engrs., 1971. 25.21 I 15. Managua, Nicaragua Earthquake Research Inst., Nov., 1973.

of Dec. 23, 1972, Earthquake Engineering 25.22 25.23 I

16. Bohm, George J. Damping for Dynamic Analysis of Reactor Coolant Loop 25.24 I Systems, Qptical Meeting on Reactor Safety, Salt Lake City, Utah, March 25.25 I 1973, Conf-730304 Available NTIS.

I 17. Bush, Spencer, Battelle Northwest Laboratories, Personal Communication. *25. 26 I

I I

I I

I F-38 I

hnHn11111*nt:ill\' 1Tc11nl1*cl maxlnu1111 .:rut11

@ at'l:t*li*r:111 .. 11 In 111*a,*IL1l11111:il lun:i, l-:11rlh11uaku lnit*nall)' al rq11,rtlnlJ' 1-:illuun fnl'lllllt,~. ~lndlll,*<1 ~kn*nlll !icalu (Llnnh ,1*l11*rc nu <l:11111

'U~.1-.J' Scale In mllt*a CD ~luoq'1ark f'EIIN/\NDO QGoulil Dove rs El Sc1:undo

<icncmlln~ Slnllon ANA

\__,, ..

L.~

'/'

PACIPIC OCEAN

/

(------~

J ., **

s~~ ~p:~~ 1i:N/r.:;.~

.Fig. F-1 r**/

1-

• -- -----~=-~------=-----

~--*-**-

ef/ALU~ 5TE/-),0 0(4)

LAK( MANAGUA POIA!E/2 /?LIJ;,/r

// ----

~

~ ....

RAGUA~ACIFIC RAILROAD 1 Ml, PAN~:2!:_ ft I CAN 0

r;:~~~;r

__J . If I G IIW,\ '(

0 OLO rt.AN

. ... .. ..... G) t_ esso ...

.... ... .. 2 Ml.

E LAKE ASOSOSCA 1

= ... .

,,,,,,/  :

.r

()

-t . .

~

<f (J

lt .... .

~

<f Nonn1 0 1.

Scnlc I11 (vlilcs ~ 1972 PRINCIPAL F/\lJL'TS MANAGUA, NICARAGUA EARTHQUAKE 1111111111111 1972 SECONDAflY FAULTS 23 DEC. 1972 (fil) SURVEY SI TE NO, Fig. F-2 l\l:ap ,>i" l\fau:igua, Ni<:aragua ~lu,wiug llil' lu.:al11>11~ *>i' i":1L*1li1i,:, . a11d huildi11g~ rd°l.!rr,*d !D i111 I

• 11:, rep 11 11. Scc T ;a lJ I*~ l for a li\l or Ilic i"acilitics.

10 9

8 7

6 5

4

. _J

<(

L.)

3 I-a:

L.)

I-2

,, A .

- - * - - - .. --'--j--C.if-.:.,-.

N.: ,1......: : . :., ~.

+~--b,-

. _l..,,-'

z LLJ L.)

a: .....

LLJ 1 CL

(.!J z

0.9 0.8 0.7 0.6

~

t-CL 2!

<(

0,5 ...

D f-OA A IPP Shaker T. 2.9 Mi SL Ge11 Rartt6t K EGCR Shaker T. 5.92 Hz & 5.82 Hz SL Ge11. EW & NSll91 0.3 ._ B IPP Shaker T. 6.5 Hz S*L Ge11. Ratl 161 L EGCR Blasl T. 5.8 Hi SL Gen. Ns121 .

C IPP Shaker T. 2.48 Hz SI Gen. Rar1 1s1 M SONGS Shaker T. 5.87 Hz & 3.18 Hz Pump NS & EW 131 D IPP Shaker T. 3.15 Hz SL Gen. Radt6t N SONGS Shaker T. 2.87 Hz Pressurizer NSt 41 E IPP Shaker T. 4.5 Hz Pump Ta11t61 0 SONGS Shaker T. 1.87 Hz SI. Gen. EW141 0.2 i- F IPP Shaker T. 5.3 H1 Pump Rari'"' P SONGS Lille Creek Earlh. 2.0 Hz Sr. Gc11. EWt 41 G IPP Shakl?r T. 2.6 H1 SI Ge11. Tan161 Q SONGS S. Fernando Earlh. 1.9 Hz & 2.85 Hz SI. Gen. Tan & Rad 110 '

fl IPP Shaker T. 2.43 H1 SL Gen Ta11°r.1 R IPP Preop. T. 2.4 Hz & 3.25 Hz St. Gen Rad (1.40)161 I IPP Shaker T. 3. 15 H1 S1 Gr.11 Ta11161 S IPP Preop. T. 2.4 Hz & 3.25 Hz St. Gen. Tan (1.4oJl6t J IPP Shaker T. 6.9 Hi s-1 Geri. Ta11t6t T

• IPP Preo11 T. 4.0 Hz & 4.9 Hz Pump 11.411) 161 0.1 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

.001 0.01 0.1 DEFLECTION (IN)

Fig. F*-3 Percent of Critical Damping for Reactor Coolant Loop Components Deflection (in)

I h-1284622-67 06/05/79 042 I SURRY POWER STATION, UNIT 1 TABLE F-1 1. 8 I CHRONOLOGY FOR 1.10 1.12 SEISMIC ANALYSIS OF NUCLEAR PLANTS I 1955 Static Methods 1.15 I 1960 1965 Introduction of Ground Spectra; Buildings Considered Rigid Buildin~ Motion and Amplification of .Spectra Considered 1.17 1.19 I Dynamic Analysis and Amplified Response Spectra First Applied to iiping

1. 21

1. 22 I Ground Spectra Change 1.24 1970 Soil Structure Interaction Considered; Ground Spectra 1. 26 I Change 3 Directional Earthquakes Regulatory Guides 1.92,

1. 27

1. 29 1.61, 1.60 Damping Changed 1. 30 I 1975 Higher Site g Levels Considered; Systematic Reevaluation 1. 32 Program; Seismic Safety Research 1. 33 I

I I

I I

I I

I I 1 of 1 I

I 06/05/79 041 h-1284622-68 I SURRY POWER STATION, UNIT 1 TABLE F-2 1.10 I CHRONOLOGY FOR SEISMIC ANALYSIS OF PIPING SYSTEMS 1.12 I 1955 Static Methods 1.15 1960 Static Application of Spectral Accelerations 1.16 I 1965 Response Spectra Dynamic Analysis; Consideration of Broadened 1.17 Amplified Spectra; ~31.7 Code - Evaluation Criteria 1.18

• 1 1970 ASME Code Section III Applied 3 Directional Earthquakes; Damping 1.19 Changed; Regulatory Guides 1.92, 1,61, l.60 1.20 I 1976 Occasional Time History Analysis; Occasional Plastic Analysis 1. 21 I

I I

I I

I I

.1 I

I I 1 of 1 I

I h-1284622-69 06/05/79 041 1.12 I SURRY POWER STATION, UNIT 1 1.13 TABLE F-3

• 1.16 I DAMAGED EQUIPMENT AT THE ENALUF POWER PLANT 1.18 I Unit 1

1. Forced~draft fan was out of alignment,

1. 21

1. 23

2. Induced-draft fan was out of alignment. 1. 25 3 *. Bearings of the condensate pump burned out. 1. 27

4. 440 V ac panel fell, 1. 29

,I 5.

6.

Condensate pump intake valve was broken.

Some tubing and refractory walls of the boiler were broken.

1. 31 1.33 I 7.

8.

Deaerator number 1 fell from its base.

Stack suffered broken splice bolts at mid-elevation.

1.35 1.37 I Unit 2 *1.39

1. Forced-draft fan was out of alignment, 1. 41 I 2. Induced-draft fan was out of alignment. 1.43 Refractory walls of the boiler were damaged. 1.45 I 3.

4.* Deaerator number 2 fell from its base. 1. 47 I 5.

Unit 3 The condensate pump intake valve was broken. 1.49

1. 52 I 1. One 440 V ac control center fell. 1.54

2. Main transformer bushings were broken. 1. 57 I 3. Starting transformer bushings were broken. 2.1 I 4.

5.

Some preheater seals were damaged.

Four turbine bearings burned out when the de-powered emergency 2.3 2.5 lube oil pump batteries broke. 2.6 I -*

6. A 69 kV switch bushing was broken. 2.8

7. Boiler support tubes over the preheater were broken. 2.10 1 1 of 2 1-

I h-1284622-69 06/05/79 041 I SURRY POWER STATION, UNIT 1 TABLE F-3 (Cont)

I

8. Forced-draft-fan control linkage was damaged. 2.12 I 9. Miscellaneous air tubes and other tubing were broken. 2.14

10. Evaporator drip valve was broken. 2.16 I 11. Three recirculating valve bodies were broken. 2.18

• 1 12. Batteries in the battery room fell from their supports and broke. 2.20 Miscellaneous Damage 2.23

.1. 1. Turbine bay crane rails were bent and electrical supply conductors were broken. Crane remained in place.

2.25 2.26 I 2.

3.

One 138 kV substation fell, Several transformer bushings were broken.

2.29 2.31 I 4. Five lightning rods (69 to 138 kV) were broken or damaged. 2.33

5. One capacitor transformer was broken. 2.35 I 6. Miscellaneous insulators were broken. 2.37 I 7.

8.

Water softener units fell from their supports and were damaged, One end of the bridge crane in the building that housed the diesel~

2.39 2.41 electric generators fell from the crane girder. 2.42 I 9. Other miscellaneous minor damage. 2.44 I

I I

I I

I 2 of 2 I

I hl284622-lab 06/05/79 042 I SURRY POWER STATION, UNIT 1 I

I I

-I I

I APPENDIX G CORRESPONDENCE WITH NRC

1. 9 1.11 I

I I

I I

I I

I I

I I

I 1.12 h-1284622-58b 06/05/79 041 I SURRY POWER S!A!ION, UKIT 1 1.13 1.16 I APPENDIX G CORRESPONDENCE ~ITH NRG 1.18 I Ihe following is reanalysis ~£fort.

a listing cf correspondence with the NRC related to the 1. 21

l. 22 I Signature Addressee Letter No./Subject 1.24

-, 3/13/79 Denton NRC to VEPCO Proffitt Show Cause Order 1.27 1.30 4/2/79 Stello Proffitt Addendum to Show Cause 1.32 Order 1.33 4/13/79 Stello Proffitt Use of Soil Structure 1.35 I 5/18/79 Stello Proffitt Interaction Techniques Request for Further SSI 1.36 1.38 I 5/25/79 Eisenhut Proffitt Information Factor Adjustment to SSI 1.39

1. 41 Calculated Stresses 1.42 I VEPCO to NRC 1. 45 I 3/30/79 Spencer Denton/

Stello 198/Initial Response to Show Cause Order 1.48 1.49 I 4/23/79 Spencer O'Reilly 289/Response to I.E. Btilletin No. 79-07

l. 51 1.52 I 4/24/79 Spencer O'Reilly 288/Response to I.E. Bulletin No. 79-07

l. 54 1.55 4/27/79 Spencer Denton/ 311/Transmittal of Two Sample 1.57 I Stello Problems to EG&G 1.58 5/2/79 Spencer Stello 260/Submittal of SSI In- 2.2 I 5/22/79 Ragone Hendrie

-formation Comments on Moratorium/Surry 2.3 2.5 Reanalysis 2.6 I 5/24/79 Spencer Stello Response to NRC Letter of 2.8 4/2/79 2.9

-1 6/5/79 Spencer Denton Submittal of Report on 2.11 I G-1 I

I 1- h-1284622-58b 06105/79 SURRY POWER STATION, UNIT 1 041 I Date Signature Addressee Letter No./Subject Reanalysis 2.12 I S&W to NRC 2.16 I 3122179 Kennedy Denton Transmittal of S&W Computer Programs 2.19 2.20 3130179 Jacobs Herring Submittal of Computer Outputs 2;24 I 4/3/79 Jacobs Bezler Submittal of Benchmark 2.28 Problem to Brookhaven 2,29

,I* National Laboratory 2.30 2 ,34-_

S&W to NRC (Cont)

I* 4/6/79 Kennedy Denton Transmittal of S&W Computer Programs -~,._.......

2.37 2.38 I 4/6/79 Jacobs Stello Plan for Verification of Dynamic Analysis Codes 2.40 2.41 I 4/11/79 4113/79

-* Jacobs Jacobs Bezler*

Stello Submittal of Computer Outputs Update and Status of Veri-2.43 2.45 fication Plan for Dynamic 2 .46 I 4118/79 Jacobs Hartman Analysis Codes Submittal of Computer Outputs

2. 47 2.49 I 4/27/79 Jacobs Bezler Submittal of Benchmark Problems 2.51 2.52 I 4127/79 Jacobs Stello Status of Verification Plan for Dynamic Analysis Codes 2.54 2.55 5114/79 Kennedy Denton Submittal of SHOCKl Program 2.57 I* Listing 2.58 I

I I

I. G-2 I