ML18138A044

Requests Facility Startup Following Installation of Piping Sys Mods Identified in Encl Rept on Reanalysis of Safety- Related Piping Sys. Remaining Analytical & Mod Work Outside Containment Will Be Expedited During Operation

ML18138A044
Person / Time
Site:	Surry
Issue date:	02/22/1980
From:	Spencer W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML18138A045	List: ML18138A044 ML19296C290
References
138, NUDOCS 8002250534
Download: ML18138A044 (84)

Text

{{#Wiki_filter:e Vepco VIRGINIA ELECTRIC AND POWER COMPANY, RICHMOND, VIRGINIA 23261 February 22, 1980 Mr. Harold R. Denton Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Denton:

START UP REQUEST SURRY POWER STATION UNIT 2 Serial No. 138 PSE&C/CMRjr:L/wang Docket No. 50-281 License No. DPR-37 The purpose of this letter is to request start-up of Surry Power Station Unit 2 after installation of certain piping system modifications identified in the attached 11Report on the Reanalysis of Safety Related Piping Systems, Surry Power Station, Unit 2 11 , dated February 22, 1980. The basis of this start-up request is the confidence of system operability during the seismic events associated with the Design Basis Earthquake (DBE) and the Operating Basis Earthquake (OBE). The attached report documents the final results of all aspects of the analysis associated with the Nuclear Regulatory Commission Order to Show Cause of March 13, 1979 for Surry Power Station Unit 2. Modifications are identified in the report relating to the stress analysis of piping systems and pipe support evaluations. Our request for start-up is based on the following:

1.

Completion of pipe stress reanalysis and resulting modifications installed for all stress problems originally run on the Shock II computer program.

2.

Completion of detailed support analyses and resulting modifications installed for Shock II problems inside the containment structure.

3.

Completion of detailed support analyses and resulting modifications installed on the SHOCK II portions of the following four systems outside the containment structure:

a.

Low Head Safety Injection

b.

High Head Safety Injection ~

c.

Containment Spray d)_\\

d.

Auxiliary Feed Water System '? \\tl{} Completion of the above work was the basis for interim operation of Unit 1 \\\\ while analysis continued on systems outside the containment other than the four safe shutdown systems noted above. The Unit 2 work includes analysis to both the DBE and OBE load cases, whereas only the DBE load case was completed for Unit 1 at the time of interim operation. ~* SO 022 50 S3'-/

I e e In addition to that stated above, a substantial amount of work has been or will be completed beyond that required for interim operation of Unit 1. The additional work that will be completed prior to start up of Unit 2 is as follows:

1.

Completion of pipe stress analysis and resulting modifications installed for all I.E. Bulletin 79-14 pipe stress problems inside the containment.

2.

Completion of detailed support analyses and resulting modifications installed for all I.E. Bulletin 79-14 lines inside the containment.

3.

An on going effort for I.E. Bulletin 79-02 involving inspecting and testing one accessible anchor bolt per base plate and analyzing base plates and anchor bolts for flexibility. All work inside the containment will be completed and 80% of the work outside the containment will be completed prior to start-up. Completion of the above items represents a substantial milestone in the overall analysis effort. All piping and supports analyzed and modified as a result of the Show Cause and IE Bulletin 79-14 analyses are capable of performing their intended function without experiencing an overstressed condition. All modifications associated with the above items will be completed prior to start up of the Unit following the steam generator replacement outage. At the present time, it appears that these modifications can be completed in May, 1980 depending on the modification installation rate our field forces are to able to attain. The remaining work to be completed after start-up will be completion of the support analysis and any modifications that might be required on the remaining systems outside the containment for both Show Cause and IE Bulletin 79-14. The reporting process for significant modifications will follow the procedure outlined in our letter of November 28, 1979 (Serial No. 972). We would appreciate your prompt review and approval of our request for returning Unit 2 to service. We believe that the analytical work completed and the modifications installed at the time of start-up provides a high degree of confidence that the integrity of safety systems for Unit 2 can be assured during the DBE or QBE events. The remaining analytical and modification work outside the containment will be expedited during the operation of the Unit. Very truly yours, W. C. Spencer Vice President - Power Station Engineering and Construction Services Attachment cc: Mr. Victor Stello, Director Office of Inspection & Enforcement Mr. James P. 0 1Reilly, Director Office of Inspection & Enforcement, Region II

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REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION-UNIT 2 VIRGINIA ELECTRIC AND.POWER COMPANY FEBRUARY 22, 1980 EBASCO SERVICES INCORPORATED--- JERICHo, NEW YORK 80 022 50 53'1 - _j

I I I I I I I I I I I I I I I I Vepco VIRGINIA ELECTRIC AND POWER COMPANY, RICHMOND, VIRGINIA 23261 February 22, 1980 Mr. Harold R. Denton Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Denton:

START UP REQUEST SURRY POWER STATION UNIT 2 Serial No. 138 PSE&C/CMRjr:L/wang Docket No. 50-281 License No. DPR-37 The purpose of this letter is to request start-up of Surry Power Station Unit 2 after installation of certain piping system modifications identified in the attached "Report on the Reanalysis of Safety Related Piping Systems, Surry Power Station, Unit 2 11 , dated February 22, 1980. The basis of this start-up request is the confidence of system operability during the seismic events associated with the Design Basis Earthquake (DBE) and the Operating Basis Earthquake (DBE). The attached report documents the final results of all aspects of the analysis associated with the Nuclear Regulatory Commission Order to Show Cause of March 13, 1979 for Surry Power Station Unit 2. Modifications are identified in the report relating to the stress analysis of piping systems and pipe support evaluations. Our request for start-up is based on the following:

1.

Completion of pipe stress reanalysis and resulting modifications installed for all stress problems originally run on the Shock II computer program.

2.

Completion of detailed support analyses and resulting modifications installed for Shock II problems inside the containment structure.

3.

Completion of detailed support analyses and resulting modifications installed on the SHOCK II portions of the following four systems outside the containment structure:

a.

Low Head Safety Injection

b.

High Head Safety Injection

c.

Containment Spray

d.

Auxiliary Feed Water System Completion of the above work was the basis for interim operation of Unit 1 while analysis continued on systems outside the containment other than the four safe shutdown systems noted above. The Unit 2 work includes analysis to both the DBE and QBE load cases, whereas only the DBE load case was completed for Unit 1 at the time of interim operation.

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I I I I In addition to that stated above, a substantial amount of work has been or will be completed beyond that required for interim operation of Unit 1. The additional work that will be completed prior to start up of Unit 2 is as follows:

1.

Completion of pipe stress analysis and resulting modifications installed for all I.E. Bulletin 79-14 pipe stress problems inside the containment.

2.

Completion of detailed support analyses and resulting modifications installed for all I.E. Bulletin 79-14 lines inside the containment.

3.

An on going effort for I.E. Bulletin 79-02 involving inspecting and testing one accessible anchor bolt per base plate and analyzing base plates and anchor bolts for flexibility. All work inside the containment will be completed and 80% of the work outside the containment will be completed prior to start-up. Completion of the above items represents a substantial milestone in the overall analysis effort. All piping and supports analyzed and modified as a result of the Show Cause and IE Bulletin 79-14 analyses are capable of performing their intended function without experiencing an overstressed condition. All modifications associated with the above items will be completed prior to start up of the Unit following the steam generator replacement outage. At the present time, it appears that these modifications can be completed in May, 1980 depending on the modification installation rate our field forces are to able to attain. The remaining work to be completed after start-up will be completion of the support analysis and any modifications that might be required on the remaining systems outside the containment for both Show Cause and IE Bulletin 79-14. The reporting process for significant modifications will follow the procedure outlined in our letter of November 28, 1979 {Serial No. 972). We would appreciate your prompt review and approval of our request for returning Unit 2 to service. We believe that the analytical work completed and the modifications installed at the time of start-up provides a high degree of confidence that the integrity of safety systems for Unit 2 can be assured during the DBE or OBE events. The remaining analytical and modification work outside the containment will be expedited during the operation of the Unit. Attachment cc: Mr. Victor Stello, Director Office of Inspection & Enforcement Mr. James P. O'Reilly, Director Very truly/yours, I .**~~o,h~*,,--- W. C. Spencer Vice President - Power Station Engineering and Construction Services Office of Inspection & Enforcement, Region II

I I I I I I I I,, I I,, I I\\ I I,, I SURRY POWER STATION - UNIT 2 REPORT ON THE REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION - UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY FEBRUARY 19 80 EBASCO SERVICES INCORPORATED JERICHO, NEW YORK

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~- ~ ---=----- --=- -- ~- -- --- --=---- --=-- _- - --- ~ -- --- ~~-, I Section I 1.0 2.0 I 3.0 4.0 5.0 6.0 I* 7.0 7.1 I 7.2 7.3 7.4 I, 8.0 9.0 10.0 I 10.1 I 10.2 10.3 10.4 t Appendix I A B C I D I t* I,, I\\ I SURRY POWER STATION - UNIT 2 TABLE OF CONTENTS Title

SUMMARY

AND CONCLUSIONS......................... SCOPE OF REANALYSIS............................. PIPE STRESS RESULTS............................. PIPE SUPPORT RESULTS...........*.............. *.. SCHEDULE FOR COMPLETION......................... HIGH ENERGY LINE BREAKS......................... CONSERVATISMS................................... Field Verification of As-Built Conditions..... Quality Assurance and Engineering Assurance... Use of Amplified Response Spectra............. Conservatisms Applied to Inertial Stress...... SYSTEM OPERABILITY EVALUATION................... BRANCH LINE

SUMMARY

RESPONSE TO THE NUCLEAR REGULATORY COMMISSION'S CONCERNS........................... Support Stiffness............................. NUPIPE Computer Code.......................... Problem 2538 - Support H-15................... Benchmark Problems.... -........................ Systems Affected *..*.*.....*.......*............. Flow Diagrams - Identification of Problems Analyzed Response to IE Bulletin 79-04.........*...**.... Correspondence with the NRG *.......*........*... 1-i 1-1 2-1 3-1 4-1 5-1 6-1 7-1 7-1 7-1 7-1 7-2 8-1 9-1 10-1 10-1 10-1 10-1 10-2 A-1 B-1 C-1 D-1

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'I I I* I I I I I I I I I I Table 3-1 3-2 3-3 3-4 4-1 4-2 5-1 ---:=.--- -- --~-~---~ SURRY POWER STATION - UNIT 2 LIST OF TABLES Title PIPE STRESS RE-EVALUATION

SUMMARY

NOZZLE AND PENETRATION

SUMMARY

PIPE STRESS HARDWARE MODIFICATION

SUMMARY

HARDWARE MODIFICATION

SUMMARY

DUE TO NOZZLE AND PENETRATION OVERLOADING PIPE SUPPORT ANALYSIS

SUMMARY

PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

SCHEDULE FOR COMPLETION 1-ii

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I I I,. I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 1

SUMMARY

AND CONCLUSIONS In response to the Nuclear Regulatory Commission's Order to Show Cause, dated March 13, 1979, a reanalysis was conducted of safety related piping systems for Surry Power Station Unit 2 which were originally dynamically analyzed using the SHOCK 2 computer program. The SHOCK 2 program, which used an earlier load combination methodology, is no longer considered ac-ceptable by the NRC. This report discusses the details of the analysis work and results of the pipe and support analyses within the scope of the reanalysis for Surry Unit 2. Further, this reanalysis is consistent with the methods used on Surry Power Station Unit 1, which were discussed in earlier reports submitted on June 5, 1979 (Vepco Serial No. 453) and on August 1, 1979 (Vepco Serial No. 453A) and on January 15, 1980 (Vepco Serial No. 048). This report summarizes the total reanalysis effort for all aspects of the March 13, 1979 Order to Show Cause for Surry Power Station Unit 2. All piping systems affected by the Order to Show Cause, both inside and outside the containment, have been reanalyzed using the NUPIPE program, which is acceptable to the NRC. Table 3-3 (Pipe Stress Hardware Modi-fication Summary) and Table 3-4 (Hardware Modification Summary Due to Nozzle and Penetration Overloading) identifies all modifications to the piping systems which have resulted from this reanalysis. While some of these modifications are attributable to the seismic analysis

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

All of these modifications have been or will be made prior to startup of the unit following the steam generator replacement outage. With the installation of these modifications, all Surry Unit 2 piping within the scope of this report will meet the Final Safety Analysis Report (FSAR) allowables for both the Operating Basis Earthquake and Design Basis Earthquakes (OBE and DBE) conditions. All pipe supports inside the containment and all supports outside the containment for the High Head Safety Injection, Low Head Safety Injection

System, Containment Recirculation Spray and Auxiliary Feedwater Systems affected by the Order to Show Cause have been evaluated for the revised support loads from the pipe stress reanalysis.

All of these hardware modifications have been or will be installed prior to start up of the Unit following the steam generator replacement outage. Table 4-2 (Pipe Support Hardware Modification Summary) reports all the modifications resulting from the support reanalysis. As was the case in the piping system reanalysis, most of the modifications are the result of differ-ences between the original design conditions and the actual field as-built condition. For the remaining supports outside the containment, other than those four systems listed above, modifications are expected to be completed in June 1980. 1-1

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I I I SURRY POWER STATION - UNIT 2 During the reanalysis of Surry Unit 2, 79 pipe stress modifications and 256 pipe support modifications were identified; while 63 pipe stress and 66 pipe support modifications were identified on Surry Unit

1.

The differences in the number of modifications is not considered significant, due to the conduct of the reanalysis applied to Surry Unit 2. Modifica-tions on Surry Unit 1 were identified after many analytical iterations;

whereas, on Surry Unit 2 modifications were designed based on fewer iterations.

The conduct of the reanalysis in this manner served to identify modifications faster so that systems could be upgraded more quickly in order not to substantially interfere with the completion of the Steam Generator Replacement Project. In addition, the Surry Unit 2 reanalysis included the OBE condition. Further, all pipe supports were as-built in the field and QC verified prior to reanalysis. Lastly, to limit the interface problems between the Show Cause scope and other piping, supports were added to facilitate the NUPIPE reanalysis. Any overstressed condition identified during the reanalysis remaining supports outside the containment, for which the loads one-half the ultimate capacity of the support, will be reported NRC and appropriate actions taken as required by the Technical fication, as described in Section 8 of this report. of the exceed to the Speci-Based on the results of the analysis of supports done to date, it is not expected that any significant number of the remaining supports should have a safety factor of less than two with respect to ultimate capacity. 1-2

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I I I I I I I I I I I I ---~---~-------------~-- SURRY POWER STATION - UNIT 2 During the reanalysis of Surry Unit 2, 79 pipe stress modifications and 256 pipe support modifications were identified; while 63 pipe stress and 66 pipe support modifications were identified on Surry Unit

1.

The differences in the number of modifications is not considered significant, due to* the conduct of the r~an~lysis applied to Surry Unit 2. Modifica-tions on Surry Unit 1 were identified after many analytical iterations;

whereas, on Surry Unit 2 modifications were designed based on fewer iterations.

The conduct of the reanalysis in this manner served to identify modifications faster so that systems could be upgraded more quickly in order not to substantially interfere with the completion of the Steam Generator Replacement Project. In addition, the Surry Unit 2 reanalysis included the OBE condition. Further, all pipe supports were as-built in the field and QC verified prior to reanalysis. Lastly, to limit the interface problems between the Show Cause scope and other piping, supports were added to facilitate the NUPIPE reanalysis. Any overstressed condition identified during the reanalysis of the remaining supports outside the containment, for which the loads exceed one-half the ultimate capacity of the support, will be reported to the NRC and appropriate actions taken as required by the Technical Speci-fication, as described in Section 8 of this report. Based on the results of the analysis of supports done to date, it is not expected that any significant number of the remaining supports should have a safety factor of less than two with respect to ultimate capacity. 1-2 /

I I I I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 2 SCOPE OF REANALYSIS As described in the NRC Order to Show Cause, March 13, 1979, some p1.p1.ng systems in the Surry Power Station, Unit 2 were dynamically analyzed wi_th a SHOCK 2 computer program that is not currently acceptable to the NRC. All systems or portions of systems that were analyzed by the SHOCK 2 computer program have been identified in Appendix A. These systems were reanalyzed by Stone Webster Engineering Corporation (Stone Webster) and Ebasco Services Incorporated (EBASCO) using a NUPIPE com-puter code. Responsibility for the reanalysis 1.s also identified in Appendix A by system and problem number. The results of the reanalysis are compared with code allowable

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

2-1 pipe used

I .I I I I I 'I I I I I I I I I I I I SURRY POWER STATION SECTION 3 UNIT 2 PIPE STRESS RESULTS A total of 62 pipe stress problems were originally analyzed by the PSTRESS/ SHOCK 2 computer, program that used algebraic summation and are therefore specifically addressed by the Show Cause Order. These stress problems are being analyzed by two groups: Stone & Webster Engineering Corporation (Stone & Webster) in Boston, Massachusetts, and Ebasco Services Incorpora-tion (EBASCO) in Jericho, New York, as indicated in the following table: Stone & Webster 13 PIPE STRESS PROBLEMS EBASCO 49 Total 62 Responsibility for the reanalysis is identified by system and problem number in Appendix A of this report. Field-verified piping isometric drawings provide the basis for program in-puts for the pipe stress reanalysis. The reanalysis is conducted using the NUPIPE computer program. NUPIPE calculates intra-modal seismic forces using a modified square root of the sum of the squares (SRSS) technique which is always more conservative than the approved SRSS method, and an SRSS technique for inter-modal combination. Piping is analyzed in most cases utilizing amplified response spectra (ARS) that are developed using soil structure interaction techniques (SSI-ARSf. The resultant stresses and loads are used to evaluate piping, supports, nozzles, and penetrations. In accordance with the NRC letters of May 25, 1979 and November 15, 1979 to Virginia Electric and Power Company (VEPCO), the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a factor of 1.5 for the DBE and 1.25 for DBE conditions. All 62 problems have been reanalyzed. Table 3-1, Pipe Stress Re-Evaluation Summary, presents the results for these 62 stress problems. In Table 3-1, the figures for Original Total Stress, at the point of maximum total stress in the pipe, and Original Seismic Stress, at the same point, are extracted from original design stress isometrics (MSK's). In Table 3-1, the columns for New Total Stress, at the point of maximum total stress in the pipe, and New Seismic Stress, at the same point, were taken from the NUPIPE computer runs with the seismic inertial stress multi-plied by a factor of 1. 5 and then added to the Seismic Anchor Movement (SAM) Stress for runs using the SSI-ARS. Even though Table 3-1 reports DBE results, stress analysis is performed for DBE also and modifications de-signed wherever necessary. The Original Total and Original Seismic Stresses shown in Table 3-1 were computed using the SHOCK 2 programs for the original design conditions. The New Total and New Seismic stresses were computed by the NUPIPE pro-3-1

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SURRY POWER STATION - UNIT 2 gram using different mass models and in most cases different ARS's than the original calculations. More importantly, the reanalyses were based on as-built conditions, field verified in 1979, which in some cases differ from the original design conditions. For these reasons, the new stresses and the original stresses in Table 3-1 are not comparable, as they do not necessarily represent the same physical conditions, Table 3-2, Nozzle and Penetration Summary, summarizes the nozzles and pene-trations evaluated u'nder the reanalysis program. For all the problems in which the SSI-ARS are used, the seismic inertial nozzle loads have been increased by a factor of 1. 5 for DBE per the NRC letter of May 25, 1979, and by a factor of 1.25 for QBE per the NRC letter of November 15, 1979. Table 3-3, Pipe Stress Hardware Modification Summary, lists the hardware modifications necessary to bring the pipe stress analysis to within code al lowab les. Of the 62 problems reanalyzed, hardware modifications were made to 16 problems due to pipe stress. These modifications consisted of 22 added, modified, or deleted supports. The modifications include those necessary to the flexibility analysis of the branch lines. A branch line (Problem No. 2508) was rerouted as a result of thermal reanalysis, not as a result of seismic reanalysis. Table 3-4, Hardware Modification Summary due to Nozzle and Penetration Overloading, lists all modifications to reduce nozzle and penetration loads. Of the 62 problems reanalyzed, hardware modifications were made to 17 problems due to nozzle overload. These modifications consisted of 57 added, modified, or deleted supports. Those modifications which result from the piping reanalysis are identified in Section 3. Only the modifications which result from the pipe support reanalysis are reported in Table 4-2, Pipe Support Hardware Modification Summary. Final verification of piping and support stresses, and Engine-ering Assurance review has yet to be completed for a few problems. It is expected, however, that the number and type of modifications due to the stress and support analysis are correct and final. 3-2

System Name and Problem Number+ Low Head Safety Injection 2555 2709 2537/2540/2540B 2539 2727 2681 2682 2695 2697 High Head Safety Injection 2689 2735 Containment and Recirculation Spray 2521 2523 2547 2546 2541 2542 SURRY POWER STATION - UNIT 2 TABLE 3-1 PIPE STRESS RE-EVALUATION

SUMMARY

Reanalysis Responsibility E E E E S&W E E E E E E S&W S&W S&W E E E MKS Number 122Dl 12211

122Al, 117Bl 122Jl 127Cl 127C2 127Kl 127K2 127Dl 127D2 127Fl 127Gl 127G2 123Al 123A2 123Cl 123Dl 123DZ 123D3 Line Size NPS (Inches)-

10,12 12 4,6, 10,12 6 6 8,10 8 8 8 8 10 3,4,6, 8,10 8, 10-8, 10 8, 10 8,10 8, 10 Original Total 12043 NA 12350 30368 21179 1677 1677 21179 21179 24649 NA 14904 14904 12713 3528 3528 3528 Pipe Stress (psi) Original New New Seismic Total Seismic NA NA NA NA NA 307 307 NA NA NA NA NA NA NA 1576 1576 1576 7974 19173 2739 14771 24352 1220 1174 2094 1981 11773 26660 7790 7977 23532 8892 18338 18931 2449 11427 883 7330 17453 185 164 1103 1022 9571 17772 4276 6013 19 739 7328 16636 17252 Sheet 1 of 5 Allowable 30690 33750 33750 32985 33750 28485 28485 28485 28485 33750 33750 33561 33561 .13561 28800 28800 28800 \\' *- \\ I

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

SUMMARY

System Name Line Size PiEe Stress (Esi) and Reanalysis MKS. NPS Original-Original New New Problem Number Res Eons ihilit~ Number (Inches} Total Seismic Total Seismic Allowable Containment and Re-circulation SErai: (Cont-'d) 25-43 E 123D4 8,10 3528 1576 8498 6988 29970 2560 E 123El 10 7334 NA 12775 10995 29970 2561 E 123E2 10 7334 NA 6587 3576 29970 2544 E 123Gl. 10 11605 7922 2912 1181 28485 2533 E 123G2 10 11605 1n2 5874 3813 28485 2548 E 123Hl 10 15785 11241 3904 2437 29970 2545 E 123H2 10 15785 11241 2397 676 29970 2744 E 123Jl 8 7966 5118 16143 15187 35820 2745 E l23Kl s 24843 22577 15621 12559 33750 2753 E 123Ll 12 6136 2818 1344 394 33750 2754 E 123Ml 12 564~ NA 1999 949 33750 2751 E 123Nl 10 6010 NA 21234 13842 28485 2752 E 123N2 10 6010 NA 18613 13934 28485 2549 S&W 123C2 8 fl955 10125 10081 839T 33561 2755 E 123Pl 4,8 10369 6324 14311 7340 33750 2756 E l23Ql 10 NA NA 5705 2686 28485 2757 E 123Q2 10 5810 NA 3847 2381 28485 Main Steam 2577 S&W lOODl 30 13824 NA 10763 2883 33750 2588 S&W 101Dl 30 18635 NA 12513 3041 33750 257CJ S&W 102D2 30 13031 NA 11434 4120 33750 2346* S&W 103Al 30 19970 NA 32-568 25477 33750 103A2

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

SUMMARY

System Name Line Size PiEe Stress- ( :es i) and Reanalysis MKS NPS Original Original New New Problem Number Res:eonsibilitl Number (Inches) Total Seismic Total Seismic Allowable Feedwater 2569 S&W lOOGl 14 14499-NA 13681 9360 27000 2573 S&W 101Gl 14 16025 NA 12970 8376-21000 25-71 S&W 10-2Gl 14 17927 NA 14230 8443 27000 Auxiliari Feedwateic-2473 E 118Al 3,6 8568 2407-20963 17736 27000 118A2 2683 E 118Gl 4,6 21230 NA 12988 9188 27000 1I8G2 Pressurizer s:erai 2771 E 125Al 4 18560 NA 8013 3088 30690 I I Pressurizer I Safeti and Relief 2000 E 124Al 3,4 9093 NA 8824 5421 30616 124A2 6,12 I I Res--iaual !feat 1

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2540/2540B E Listed-Under Low Head Safety fojectfon System 2508A/2508B E 117Al lG,12,14 NA NA 13112 8461 24570 2554 E 117Cl 6 NA NA 12_008 8824 29970 Service Water 2465 E Il9Al 24 NA NA 6529 6212 21600 2467 E Il9A2 24-NA NA 6561 6214 21600

System-Name and Problem Numb"er Service Water (Cont'd) 2469 2471 Component Cooling Z6CH/2603 260412605 Containment Vacuum 2650 HP Steam to_ Auxiliary Feedwater Pump 2869/2862/2864 Reanalysis Responsibility E E E E SliiW SURRY POWER STATION - UNIT 2 TABLE 3- - - - PIPE STRESS RE-EVALUATION

SUMMARY

MKS Number l19A3 lI9A4-112Sl 112S2-Tl2AA1 112AB1 137Al 131Al .131Bl 131Cl Line Size NPS (Inches) 24 24 18 18 8 3,4 Original Total NA NA 9696 9696 22609 Pipe Stress (psi) Original New New Seismic Total-- Seismic NA-14o87 13730 NA 12459 11614 NA 7043 5246 NA 7074 5204 NA 13659. 13037 NA 242-29 Sheet 4 of 5 Allowable 21600 216-00 I I 21600 21600 21600 27000

SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 5 of 5 Legend: E EBASCO S&W Stone&* Webster NA Not Applicable Allowable Stress = 1.8 sh New Total Stress (for SSI/ARS) New Seismic (for SSI/ARS) New Total Stress (original ARS) New Seismic (original ARS) PIPE STRESS RE-EVALUATION

SUMMARY

SLP + SDW- + 1* 5 SDBEI + SDBEA 1* 5 SDBEI + SDBEA SLP + SDW + SDBEI + SDBEA SDBEI Where SLP Longitudinal Pressure Stress Dead Load Stress Seismic Inertial Stress, Design Basis Earthquake SDBEA = Seismic Stress due to Anchor Movements, Design Basis Earthquake Sh Allowable stress at maximum (hot) temperature Note: The original total and origina1 seism:i:c stresses shown in Table 3-1 were computed using SHOCK 2 for the original des~gn conditions. The new total and new seismic stresses were computed by the NUPI~ program using different mass models and, in most cases,. different. ARS' s than the original calculations. Mo Fe importantly, the reanalyses were based on field-ve:rified, as-built conditions in 1979, which, in some cases-, differ significantly-from the original-desig_n conditions. For this reason, th~ new stresses and the original stresses in Table 3-1 are not comparable, as they do not necessarily represent the same physical. conditions.

• Soil Structure Interaction (SSr) Amplified Response Spectras (ARS) were used in the new ana.lysis_ for all problems except Problem 2346 which utilizes a combination of SSI-ARS and the origjnal ARS.

+ Problems having. / are counted as separate probiems example: 2869/2862/2864-are counted as i::hree problems. Using th.is method of counting the total number of pipe stress probl-ems. equal 62. I . I I I* _l

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SUMMARY

SHOCK 2 Problems I Responsi-Vendor System bility Total No. No. Acceptable Nozzle Confirmation and For of Nozzles/ (Evaluation Modification Being I Problem No. Analysis Penetrations Complete) Required Obtained Low Head Safety I Injection 2555 E 1/0 1 0 1 I 2709 E 1/0 1 0 0 I 2537/2540 E 1/0 l* 0 1 2539 E 0/0 NA NA NA I 2727 S&W 2/0 2 0 2 0/0 2681 E NA NA NA I 2682 E 0/0 NA NA NA 2695 E 0/0 NA NA NA I 2697 E 0/0 NA NA NA I High Head Safety Injection I 2689 E 0/0 NA NA NA 2735 E 3/0 3 0 3 I Containment and Recirculation I Spray 2521 S&W 0/0 NA NA NA I 2523 S&W 0/0 NA NA NA 2547 S&W 0/0 NA NA NA I 2546 E 0/0 NA NA NA I I

~ ---- I I SURRY POWER STATION - UNIT 2 TABLE 3-2 Sheet 2 of 4 I NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems I Responsi-Vendor System bility Total No. No. Acceptable Nozzle Confirmation and For of Nozzles/ (Evaluation Modification Being I Problem No. Analysis Penetrations Complete) Required Obtained Containment and Recirculation I Spray (Cont'd) 2541 E 0/0 NA NA NA I 2542 E 0/0 NA NA NA I 2543 E 0/0 NA NA NA 2560 E 1/0 l 0 0 I 2561 E 1/0 l 0 0 2544 E 1/0 l 0 0 I 2533 E 1/0 l 0 0 I 2548 E 1/0 l 0 0 2545 E 1/0 l 0 0 I 2744 E 0/0 NA NA NA 2745 E 0/0 NA NA NA I 2753 E 1/0 1 0 0 2754 E 1/0 l 0 l I 2751 E 2/0 2 0 0 I 2752 E 2/0 2 0 0 2549 S&W 0/0 NA NA NA I 2755 E 2/0 2 0 2 2756 E 2/0 2 0 0 I 2757 E 2/0 2 0 0 I I

I I I I I I I I I I I I I I I I I I I -= - ~-- -- -- - System and Problem No. Main Steam 2577 2588 2579 2346 Feedwater 2569 2573 2571 Responsi-bility For Analysis S&W S&W S&W S&W S&W S&W S&W Auxiliary Feedwater. 2473 E 2683 E Pressurizer Spray 2771 E Pressurizer ~afety and Relief 2000 E Residual Heat Removal 2540 E SURRY POWER STATION - UNIT 2 TABLE 3-2 NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems Total No. of Nozzles/ Penetrations 1/1 1/1 1/1 0/0 1/1 1/1 1/1 0/0 3/0 1/0 5/0 No. Acceptable (Evaluation Complete) 1/1 1/1 1/1 NA 1/ 1 1/1 1/1 NA 3 1 5 Nozzle Modification Required 0/0 0/0 0/0 NA 0/0 0/0 0/0 NA 0 0 0 (Listed under Low Head Safety Injection System) Sheet 3 -of 4 Vendor Confirmation Being Obtained 0/0 0/0 0/0 NA 0/0 0/0 0/0 NA 3 1 5

I I I I I I I I I I I I I I I I I I I -=-- -- ~--- -~ --~~ SURRY POWER STATION - UNIT 2 TABLE 3-2 NOZZLE AND PENETRATION

SUMMARY

SHOCK 2 Problems Responsi-System bility Total No. No. Acceptable and For of Nozzles/ Problem No. Analysis Penetrations Residual Heat Removal (Cont'd) 2540B E 2508A/2508B E 2554 E Service Water 2465 E 2467 E 2469 E 2471 E Component Cooling 2601/2603 E 2604/2605 E Containment Vacuum 2650 S&W HP Steam to Auxiliary Feedwater Pump 2862/2864/2869 E NOTES: NA= Not Applicable E = EBASCO S&W = Stone & Webster 0/0 8/0 0/0 1/0 1/0 1/0 1/0 2/0 2/0 0 1/0 (Evaluation Complete) NA 8 NA 1 1 1 1 2 2 NA 1 Sheet 4 of 4 Vendor Nozzle Confirmation Modification Being Required Obtained NA NA 0 8 NA NA 0 0 0 0 0 0 0 0 0 1 0 1 NA NA 0 0

System Name and Problem No. Low Head Safety _Injection 2769 2537/2540 2539 Containment and Recirculation Spray 2549 2544 27~5 2752 Pressurizer Spray 277-1 Residual Heat Removal 2508 Reanalysis Responsibility E E S&W E E E E E SURRY POWER STATION - UNIT 2 TABLE 3 - - - - - Sheet l of 2 PIPE STRESS HARDWARE MODIFICATION

SUMMARY

MKS No, 122Ll l 22AL 117Bl 122Jl 122Kl l 23C2-l23Gl 123Kl 123N2. 125AJ 117Al Overstressed Condition ..;..:...;..:...;..:...;;.....;..:...;.--=-- Seismic overstress Thermal overstre&s Thermal overstress (Branch Line) ~ipe contacts crane wall during seismic condition. Thermal overstress Seismic anchor move-ment overstress Thermal overs-tress (Branch Line) Seismic overstress Thermal overstres.s Branch line Attributed To: Seismic Reanalysis As-built As-built Seismic Reanalysis As-built As-built As-built As-built As-built Resolution Spring hanger replaced by rigid restraint. Removed a restraint, anchor replaced by restraints and a snubber. Removed a restraint Lateral support added. Removed a restraint, Vertical restraints replaced by spring hangers at two loca-tions Removed a restraint. Two restraints a~ded. Rerouting of 1-1/2 1n. pipe No, of Modifications l 2 2 l 2 2540 (Listed under Low Head Safety Injection System)

System Name and Problem No. Residual Heat Removal (Cont'd) 2540B Component Cooling 2604/ 2605 HP Steam to Auxiliary Feedwater Pump 2862/ 2864/ 2869 Feedwater 2569 Notes: E EBASCO Reanalysis Responsioility E E E S&W S&W Stone & Webster MKS No. 117B 112AA1 112AB1 131Al 131Bl 131CI lOOGl SURRY POWER STATION - UNIT 2 TABLE 3-3 PIPE STRESS HARDWARE MODIFICATION

SUMMARY

Overstressed Condition Thermal and seismic overstress Seismic overstress Thermal and seismic overstress Insufficient branch line flexibility Attributed To: As-built Seismic Reanalysis As-bui-lt/ Seismic Reanalysis As-built Resolution Anchor replaced by vertical restraint, horizontal restraint removed. Two restraints added. Two anchors replaced by restraints, two snubbers and a spring added. Remove existing U-bolt on 3/4 in. line Sheet 2 of 2 No. of Modifications 2 2 5 1

System Name and Problem No. Low Head Safety Injection 2555 High Head Safety Injection 2735 Containment and Recirculation Spray 2544 2533 27-;53 2754 275-1 2752 2755 2756 Reanalysis Responsibility E E E E E E E E E E

• SURRY POWER STATION -

UNIT 2 TABLE 3-4 HARDWARE MODIFICATION

SUMMARY

Sheet 1 of 2 DUE TO NOZZLE AND PENETRATION OVERLOADING Equipment No. 2-SI-TK-lB 2-CH-P-lA 2-CH-P-lB 2-CH-P-lC 2-RS-E-lD 2-RS-E-lC 2-cs-P-IB 2-CS-P-JA 2-RS-P-2A 2-RS-P-2B 2-CS-P-lB 2-CS-P-JA 2-RS-E-lA 2-RS-P-lA Attributed To: As-built/ Seismic Reanalysis Seismic Reanalysis As-built As-built As-built/ Seismic Reanalysis As-built/ Seismic Reanalysis Seismic Reanalysis-Seismic Reanalysis As-built/ Seismic Reanalysis As-built Resolution No. of Modifications Two restraints added. Ten restraints added, one vertical restraint replaced by spring hanger, two anchors added One restraint removed Two restraints added One vertical restraint replaced by spring hanger, two restraints added One vertical restraint replaced by spring one horizontal restraint added. One restraint added One restraint added One restraint added two vertical restraints removed. One anchor and one restraint removed 2 13 l 2 3 2 3 2

System Name and Problem Na. Containment and Recirculation Spray (Cont'd-) 2757 Auxiliary Eeedwater 2683 Pressurizer Safety & Relief 2000 Residual Heat Removal 2508B Service Water 2471 Component Cooling 2601/ 2603 Note: E EBASCO Reanalysis Responsibilit)" E E E E E E SURRY POWER STATION - UNIT 2 TABLE 3-_4 HARDWARE MODIFICATION SUM!".ARY DUE TO NOZZLE AND PENETRATION OVERLOADING Equipment No. 2-RS-E-lB 2-RS-P-lB 2-FW-P-2 2-FW-P-3B 2-FW-P-3A 2-RC-TK-2 2-RH-P-lA 2-RH-P-lB 2-RS-E-JD 2-RH-E-lB Attributed To As-fiuilt Seismic Reanalysis As-built Seismic Re-analysis As-built As-built Seismic Reanalysis Sheet 2 of 2 Resolution No. of Modifications One anchor removed Three horizontal and two vertical restraints added. A spring replaced by a rigid hanger and a horizontal snubber, two restraints replaced By*snubbers, lateral restraint deleted. Six snubbers added, two springs replaced by restraints, one spring hanger added, one verticar reoiraint added. One restraint removed Four vertical restraints replaced by springs and snubbers*, two restraints added. l 5 4 10 6 I' -I I

I I I I I I I I I I I I I I I I I I I ~URRY POWER STATION - UNIT 2 SECTION 4 PIPE SUPPORT RESULTS Table 4-1, Pipe Support Analysis

Sunnnary, sunnµarizes the pipe support re.;ir,.alysis program.

Six hundred four (604) supports, (427 inside the con-tainmep.t, 177 outside the coatainment) on lines originally analyzed using Shock 2, were reanalyzed as part of this Show Cause effort. Two hundred fifty six (256) hardwar~ modifications have been identified. The podificationl:i identified due to the pipe support reanalysis are listed in Table 4-2, Pipe Support Hardware Modification Summary. Those modifications which result from the piping reanalysis are identified in Section 3. . Only the modifications which result from the pipe support reanalysis are reported ~n Table 4-2. Of the modifications ~dentifi~d, only 104 were the result of seismic reanalysis of the piping systems identi-f~ed in the Show Cause Order, while 152 11ere the resµlt 9f differences

i.denttfied between the as-built condittons and tqe original design.

These condit~ons are iqentified in the table for each problem. For aU the problems in which the SSI-!RS are used, the seismic inertial loads havr been increased by a factor of l.5 for DBE per the NRG letter of May 25, 19 79, and by a factor of 1. 25 for OBE per th~ NRG letter of November 15, 1979. 4-1

System Name and Problem Number Low Head Safety Injection System 2537/2540 2555 2539 2681 2682 2695 2697 2709 2727 High Head Safety Injection System 2689 2735 Containment and Recirculation Spny 2521 2523 2547 2549 2546 SURRY POWER STATION - UNI.T 2 TABLE 4-1 PIPE SUPPORT ANALYSIS

SUMMARY

Analysis Responsibility E E E E E E E E S&W E E S&W S&W S&W S&W E Location IC IC IC oc oc QC QC IC QC oc QC IC IC IC IC IC Total Number of Supports 33 17 7 4 5 9 9 9 17 5 48 15 16 13 4 11 Evaluation Complete 33 17 7 4 5 9 9 9 17 5 48 15 16 13 4 11 Sheet 1 of 4 Modifications or Additions Required 16 8 5 3 3 4 6 3 10 4 21 4 5

6.

1 8 I I I I.

System Name and Problem Number Containment and Recirculation ~ (ConL'd-) 2541 2542 2543 2560 2561 2544 2533 2S-48 2545 2744 2745 2753 2754 2751 2752 2755 2756 2757 SURRY POWER STATION - UNIT 2 TABLE 4-1 PIPE SUPPORT ANALYSIS

SUMMARY

Analysis Responsibility E E E-E E E E E E E E E E E E E, E E Location IC IC IC IC IC IC IC IC IC oc oc oc oc oc oc oc IC IC Total Number of Supports 11 11 11 3 5 8 5 15 17 4 4 4 3 5 4 10 8 10 Evaluation Complete 11 11 I1 3 5 8 .5 15 17 4 4 4 3 5 4 10 8 10 Sheet.2 of 4 MQdifications oE Additions Required 6 4 6 2 3 2 1 7 9 3 1 l 1 1 0 4 2 3 r I.. I \\' I I I I I I I I I I I

System Name and Problem Number Main Steam 2577 2588 2579 2346 Feedwater 2569 2573 2571 Auxiliary Feedwater 2473 2683 Pressurizer Spray and Relief 2771 2000 Residual Heat Removal 2508A/B_ 2540B 2554 SURRY POWER STATION - UNIT 2 TABLE 4-1 PIPE SUPPORT ANALYSIS

SUMMARY

Analysis Responsibility S&W S&W S&W S&W S&W - S&W S&W E E E* E E E E Location IC IC IC oc IC IC IC IC oc IC IC IC IC oc Total Number of Supports 9 2 5 41 9 3 6 34 22 32 31 47 5 1 Evaluation Complete 9 2 5 41 9 3 6 34 22 32 31 47 5 1 Sheet 3 of 4 Modifications or Additions Required 2 0 3 6 4 0 1 10 7 7 15 4 5 0

---

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

. I I,.. i' I . I I I I I I I I I I I I l I I I I

SURRY POWER STATION - UN-IT 2 TABLE 4-1 PIPE SUPPORT ANALYSIS

SUMMARY

Total Analysis Number of System Name and Problem Number Responsibility Location Supports Service Water 2465 2467 2469 2471 Component Cooling 2601 2603 2604 2605 Containment Vacuum 2650 High Pressure Steam. to Aux Feedwater Pump 2862 2864 28.69 Notes: E EBASCO S&W Stone & Webster IC Inside Containment OC: Outside Containment E* E E E E E E E S&W E E E IC IC IC IC IC IC IC IC oc oc oc oc 2 2 2 2 16 13 16 17 3 12 3 7 Evaluation Complete 2 2 2 2 16 13 16 17 3 12 3 7 Sheet 4 of 4 Modifications. or Additions Required 1 1 0 1 5 4 8 3 2 1 0-1 I I -

SYSTEM NAME AND PROBLEM NUMBER Low Head Safety Injection System 2537/2540 2537 2540 2555 2539 ANALYSIS RESPONSIBILITY E E E E SURRY POWER STATION - UNIT 2 TABLE 4 - PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC MKS-SUPPORT NUMBER 2,4 REASON FOR MODIFICATION Local pipe wall stress exceeds allowable ATTRIBUTABLE TO: Seismic reana-lysis 6 5 Support member over allowable As-built IC IC IC 7,8,9,10,1-2, 13,26 28 17,19 1 BA 8,16 3 4,5 9, 12 12A 2 1 l** 3,4 Insufficient clearance Support member over allowable Pipe clamp over allowable U-bolt over allowable Support member over allowable Upward vertical restraint required Support member over allowable Weld over allowable Local pipe wall stress exceed~ allowable Support member over allowable Upward vertical restraint required Upward vertical restraint required Upward vertical restraint required Support member over allowable Support member over allowable Seismic Reanalysis As-built As-built As-ouilt As-built Seismic reanalysis As-built-As-built Seismic reanalysis Seismic Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built Sheet 1 of 12 RESOLUTION Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify restraint for uplift load-Modify support Add weld Modify support Modify support Modify restraint for uplift load Modify restraint for uplift load Modify restraint for uplift load Modify support Modify support

111D - -.. - - - - I.. I SURRY POWER STATION - UNIT 2 TAllLE-4-2 Sheet 2 of 12 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE-PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO: RESOLUTION Low Head Safety Iniection System (Cont'd) 2.1og-E IC 7 Support member over al-lowabte-As-built Modify-support-2,3 Local pipe wall stresses-over Seismic Modify support allowable Reanalysis 2727 S&W ex:- 8 Support over allowable Seismic-Modify. Support Reanalysis 13 Support_ and weld over Seismic Modify Support allowable Reanalysis 11 Loads-out of range of spring As-built Replace spring 1-4 Loads out of spring range Seismic Replace spring. Reanalysis 15 Loads out of spring range Seismic Replace spring Reanalysis 16-Loads out of spring range Seismic Replace spring Reanalysis 18 Loads out of spring range Seismic Replace spring Reanalysis 19 Support restrains lateral As-built Modify support mo..v.ement E oc H-l,H-17 Weld over allowable As-built Add weld 2695 E* QC H-50 Support member over allowable As-built Modify support A-16 Support member over allowable As-built Modify support C-53 Support membe-r over allowable As-built Modify-support A-17 Support member over allowable As-built Modify support

- - - - - -.. - - - -...... - - - -I SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 3 of 12 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO: RESOLUTION Low Head Safeti Injection System (Cont'd} 2697 E QC R-49 Support member over allowable As-built Modify support c-1 Support not acting As-built Removed support. C-17,C-18 Support member over allowable As-built Modify support A-14 Support member over allowable As-bui-lt Modify support A-15 Support member over allowabre As-built Modify suppor_t 2681 E QC 2,4 Support member over allowable As-built Modify support 5 U-bolt over allowable As-built Modify support 2682 E QC 4 Support member over allowable-As-bu-ilt* Modify support 2 Support member over allowable As-built Modify support 5 U-bolt over allowable As-built Modify support High Head Safety Injection System 2689 E oc C-38,C,-39 Support member over allowable As-built Modify support C-40,C-41 Support member over allowable As-built Modify support 2735 E QC 22,___37,45,28 Support member over al-low ab le-As-built Modify support 36,21,23,44 1,3,16,31,39,24 Weld-over allowable As-built Adel weld 4,6, 18,33,41,56 U-bolt over allowable Seismic Modify support

26.

Local pipe wall stress Seismic Modify support exceeds allowable reanalysis Containment and Recirculation Sprai 2521 S&W-IC 2 U-bolt capacity for side-Seismic Add members to load is insufficient Reanalysis resist sideload

SYSTEM NAME AND PROBLEM NUMBER Containment and Recirculation Spray (Cont'd) 2521 2523 2547 ANALYSIS RESPONSIBILITY S&W S&W S&W SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC MKS SUPPORT NUMBER 3 4 5 1 2 3 4 5 2 4 6 s* REASON FOR MODIFICATION Frame ov:erstressed with new loads out of springs range U-strap ha~ insufficient capacity Local stress U-bo.lt capacity for side-load is insufficient. Frame overstressed Capacity of springs in-sufficient U-strap has insufficient-capacity Local stress Lateral load fails U-bolt Rod hanger cannot resist upward: load Insu~ficient clearance for thermal movement Insufficient lateral clearance for thermal movement ATTRIBUTABLE TO: Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis. As-built As-built 11111* Sheet 4 of 12 RESOLUTION Modify existing frame and replace springs Replace existing strap with new framing Eliminate anchor and add vert/lat restraint Add members to resist sideload Replace.. existing frame Replace springs Replace existing strap with new framing Eliminate anchor and add vert/lat restraint Add lateral restraint Replace with sway strut Remove lateral" stop Remove lateral stops (angle)

SYSTEM NAME AND PROBLEM NUMBER Containment and Recirculation Spray (Cont'd) 2547 2549 2546 2541 2542 ANALYSIS RESPONSIBILITY S&W S&W E E E SURRY POWER STATION - UNIT 2 TABLE 4-2. -.... *-.. PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC IC IC MKS .SUPPORT NUMBER

10.

11 6 9 Z., 8 22:, 26" 2 22A 6,21 7,9 2 22 15,23 13,17 REASON FOK MODIFICATION Insufficient lateral clearance for the:r:mal movement: Local *.stress and-support* frame overstressed U-bolt failure

Support member over allowable ATTRIBUTABLE TO:

As-built As::bui1t As-built. As-built Insufficient lateral & Seismic vertical clearance Reanalysis Insufficient lateral Seismic clearance Reanaly.sis Support member over allowable* As-built U-bolt restricts lateral As-built movement Anchor stress over al-lowable Supp*ort member over allowable Insufficient lateral clearance U-bolt restricts lateral movements Support member over allowable Support member over allowable S-eismic Reanalysis As-built Seismic Reanalysis As-built As-built As-built. Insufficient lateral clearance Seismic Reanalysis I I Sheet 5 of 12 RESOLUTION* Remove lateral stop.a (angle)* Modify structure Add lateral restraint Modify -support Modify re.straint Modify restraint Modify suppor.t Modify support Relocate-adjacent restraint Modify support Modify support Modify support Modify support Modify support Modify support I I I I i

i...... - ll!!!l 1lill...... - ,., 'illla, SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 6 of 12 PIPE S~PPORT HARDWARE MODIFICATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION Containment and Recirculation Spray 2543 E IG 15, 2'.J Support member As-built Modify support over allowable 12,13,14 Insufficient lateral Seismic Modify restraint clearance Reanalysis 24A Support member Seismic Re locate adj a-over allowable Reanalysis cent restraint 2560 E IC H-49A Support member As-built Modify support over allowable. H-50 U-bolt over Seismic Modify support allowable Reanalysis 2561 E IC H-91 Upward vertical Seismic Modify for restraint required Reanalysis uplift load H-49B Supp_ort member As-built Modify support over allowable H-98A Support member As-built Remove vertical over allowable restraint 2544 E IC H-67 Upward vertical Seismic Modify support restraint required Reanalysis for uplift load H-68 Support member As-built Modify support over allowable 2533 E IC H-3 Support member As-built Modify support over allowable 2548 E IC 7 Support member As-built Modify support I over allowable I 10 Support member As-built Modify support I over allowable I 12 U-bolt over As-built Modify support allowable 9A

• Upward vertical Seismic Modify support restraint required Reanalysis for uplift load

I 1 I SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 7 of 12 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

SYSTEM NAME MKS AND ANALYSIS-SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION Containment and Recirculation Spra:t (Cont'd) 2548 g-- IC 11 Support member As-built Modify support I over allowable I 6,8 Support member As-built. Modify support over allowable 2545 E IC 3 Support member As-built Modify support over al low ab le 14 Upward vertical Seismic Modify restraint restraint required Reanalysis for uplift load 2 U-bolt over As-built Modify support allowable 4 Support member As-built Modify support over allowable 7 Support member As-built Modify support over allowable 10,15 Weld over allowable As-built Add weld 6,8 Support member As-built Modify support over allowable 2744 E QC 2,3 Loads out of spring As-built Two rigid re-range straints re-placed by springs l Weld. over allowable As-built Add weld 2745 E QC 4 Upward vertical Seismic Modify support restraint required Reanalysis 2751 E QC 2 Local pipe wan Seismic Modify support stress over Reanalysis allowable 2753 E QC 1 SupporJ: member As-built Modify support over allowable 2754 E oc Support member As-built Modify support over allowable

SYSTEM AND PROBLEM NUMBER I.. SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

ANALYSIS RESPONSIBILITY LOCATION MKS SUPPORT NUMBER REASON FOR MODIFICATION Containment and Recirculation Spray (Cont'd) 2755 E oc 7--,8 Upward vertical restraint re<l!lired 9 Upward vertical restraint required 11 Weld over allowable 2756 E IC 54,55 Support member over allowable 2757 E IC H-90 Upward vertical restraint required H-63, Support member H-88 over allowable Main Steam 2577 S&W IC Local stress ex-ceeds allowable 9 Local stress ex-ceeds allowable 2579 S&W IC-1 Spring variability ratio exceeded Loads* outside. spring range 5 Spring variability ratio exceeded. Local stress over allowable. 2346 S&W QC 1,2,3 Loads outside spring range, local over-stress in lug 5,7,9 Snubbers, local stress, and sup-port members are overstressed ).,,, - ATTRIBUTABLE TO Seismic Reanalysis Seismic Reanalysis As-built As-built Seismic Reanalysis* As-built ~s-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Re-analysis Seismic Reanalysis-Seismic Reanalysis Sheet 8 of 12 RESOLUTION Mod-ify support for uplift load Redesign support for uplift load Add weld Modify support Modify support for uplift load Modify support Modify lug Replace lug with clamp Replace spring Replace spring Rep.lace spring

• Replace pipe lug with clamp.

Replace springs, modify lug Modify snubbers and lugs I I -I I i ]. i I 1\\ J

SYS !EM NAME AND PROBLEM NUMBER Feedwater 2569 2571 Auxiliary Fe*edwater 2473 2683 ANALYSIS RESPONSIBILITY S&W S&W E E SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC QC MKS SUPPORT-NUMBER 2,_3 6 7 5 8 14 16 11, 13, 26, 29, 32 22 28 H-2,H-8A H-11 . H-9 H-1,H-5 H-4 REASON* FUR MODIFICATION Loads outside spring range Thermal movement Insufficient clearance for lateral movement Loads outside spring range Lateral clearance in-sufficient Upward vertical restraint required Upward vertical restraint required Weld over allowable Support member over allowable Support member over allowable U-bolt over allowable U-bolt over allowable Support member over allowable Local pipe wall stress over allowable U-bolt over allowable ATTRIBUTABLE TO Seismic Reanalysis As-built Thermal., Movement Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built As-built As-built Seismic Reanalysis As-built Sheet 9 of 12 RESOLUTION Replace springs Reduce pin-to pin dimension Modify support Replace springs Modify restraint Modify restraint for uplift load Modify restraint for uplift load Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify support by adding bracing members I...

SYSTEM NAME AND P-ROBLEM NUMBER Pressurizer Spray and Relief. 2771 2000 2508 2540(B) ANALYSIS RESPONSIBILITY E E E E SURRY POWER STATION - UNIT 2 -TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC IC MKS SUPPORT NUMBER 33 5,23 6 26,27 32 4,12,H-5 15 13 4A,7,8,H-1A, 10,U,21A,113 17 2 H-36,H-15,H-17 H-12 20,21 REASON FOR MODIFICATION Upward-vertical restraint required Weld over allowable Support member over allowable Upward vertical restraint required Insufficient lateral clearance Local pipe wall stress over allowables Insufficient vertical clearance Support restraints lateral movement Support member over allowable Support member over allowable Weld over allowable Support member over allowable Vertical support not required Uplift vertical restraint required ATTRIBUTABLE TO Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built_ As-built Seismic Reanalysis Seismic Reanalysis 1*.. I Sheet 10 of 12 RESOLUTION Modify-support for uplift load Add weld Modify support Modify for uplift load Modify support Modify support Modify support Modify support Modify support Modify support Add weld Modify support Remove support Modify for uplift load

SYSTEM NAME AND PROBLEM NUMBER Pressurizer Spray and Relief (Cont'd) 2540(B) Service Water 2465 2467 2471 Component Cooling 2601 2603 2604 ANALYSIS RESPONSIBILITY E E E E E E E SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC IC IC re* IC IC MKS SUPPORT NUMBER 22 3,25 1 1 2 H-32A,H-28,H-44 H-30 H-31 4 6 9 10 H-38A . H-38B REASON-FOR MODIFICATION Support member over allowable Local pipe wall stress over allowable Support member over allowable Support member over al-lowab-le Local pipe wall stress over allowable Uplift vertical restraint required Insufficient lateral clearance Weld over allowable Insufficient lateral clear-ance Support* member over allowable Upward vertical restraint required Upward vertical restraint required Upward vertical restraint required Upward vertical restraint required ATTRIBUTABLE TO As-built-Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic* Reanalysis Seismic Reanalysis Seismic Reanalysis - - ~ I I I Sheet 11 of 12 RESOLUTION Modify support Modify support Modify support Modify support Modify support Modify support uplift Modify support Add weld Modify support Modify support Modify support uplift load Modify support uplift load Modify support uplift load Modify support uplift load for for for for for I I .\\ I I I I I \\ ' -

SYSTEM-NAME AND PROBLEM NUMBER ANALYSIS RESPONSIBILITY Component Cooling (Cont'd) 260.4 E 2605 E Containment Vacuum 2650 S&W High fressure Steam TQ Aux. Feedwater Pume 2862 E 2o69 E Notes:

• • Originally Problem No. 2708 SURRY POWER STATION -

UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION

SUMMARY

LOCATION IC IC oc oc oc MKS SUPPORT NUMBER H-35,H-36,H-23, H-38C H-38 H-38D H-25 H-25B H-25A l 2 3 5 REASON FOR MODIFICATION Weld over allowable Support member over allowable Local pipe wall stress exce.eds allowable Support member over allowable Upward vertical restraint** required Upward vertical restraint required Support restraint lateral movement Local stress at at-tachm.ent on pipe exceeds allowable Support member over allowable Support member over allowable ATTRIBUTABLE TO As-built As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built I' I I Sheet 12 of 12 RESOLUTION Add weld Modify support Modify support Modify support Modify support for uplift load Modify support for uplift load Modify support Move support above elbow and use double trunnion Modify support Modify support I, I I I I i i

I I I I, l I. l I

, I.,.

I I I I 1* I I I SURRY POWER STATION - UNIT 2 SECTION 5 SCHEDULE FOR COMPLETION 'fhe status of the reanalysis of those systems subject to Show Cause and* the installation of modifications identified as being required by the reanalysis is shown in Table 5-1, Schedule for Completion. Reanalysis on all systems is complete pepding final review. Apy required modifications on lines inside containment and on the High Head Safety In-

jection, Low Head Safety Injection, Containment Recirculation Spray, and Auxiliary Feedwater Sy~tems will be installed prior to startup of the Unit following the stea~ generator replacement outage.

Installation of modifica-tion on the balance of systems outside containment is expected to be 'Com-plet~ in Ju~e 1980. 5-1

Location of Problem Inside Containment Systems Outside Containment Systems Low Head Safety Injection High Head Safety Injection SURRY POWER STATION - UNIT 2 TABLE 5-1 SCHEDULE FOR COMPLETION Reanalysis Responsibility E/S&W E/S&W E Stress Reanalysis Complete Complete Complete Containment Recirculation Spray E Complete Auxiliary Feedwater E Complete Balance of Sys-terns E/S&W Complete NOTES: E = EBASCO S&W = Stone & Webster SGR = Steam Generator Replacement Status Support/Restraint Reanalysis Complete Complete Complete Complete Complete Complete Sheet 1 of 1 Modification Installation Prior to start-up following SGR outage Prior to start-up following SGR outage Prior to start-up following SGR outage Prior to start-up following SGR outage Prior to star_t-up following SGR outage Complete by June, 1980 ~ I

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 6 HIGH ENERGY LINE BREAKS For the high energy lines outside the containment addressed in Appendix D of the Final Safety Analysis Report (FSAR), only the main steam lines are included in this stress reanalysis~ Each of the main steam lines has two terminal break locations, one at the containment penetration and the other at the main steam manifold. Each of the risers to the main steam relief valve headers has two terminal break locations, one at the ma~n steam lines, the other at the tee into the main steam header. These terminal breakpoints are predetermined and are not changed as a result of the stress reanalysis. Two intermediate break locations were originally determined based upon maximum primary plus secondary stresses, Upon reanalysis, two additional breakpoints on each of the steam lines were located. One of these points is located immediately upstream of the check valve and the other point is at the elbow just downstre~ of the check valve. All of these points will be included in the augmented inservice inspection program. 6-1

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 7 CONSERVATISMS The conservatisms applied to the design of the piping systems for Surry Power Station Units 1 and 2 were extensively delineated in Section 7 of the VEPCO June 5, 1979 submittal (Serial Number 453). The seismic capability of nuclear piping and the seismic event probability at the Surry Power Station were discussed in that submittal. The design of Unit 2 closely follows the design of Unit 1, applying the same conservative criteria with respect to safety systems and system redundancies. Similiarly, the reanalysis efforts on Surry Power Station Unit 2 closely follows that of Unit 1, applying the same stress limits and soil struc-ture interaction amplified response spectra (SSI-ARS). Paragraphs 7.1, 7.2, 7.3 and 7.4 describe the differences in the conser-vatisms applied to the Unit 2 reanalysis. 7.1 FIELD VERIFICATION OF AS-BUILT CONDITIONS To ensure that the pipe stress and pipe support reanalysis is performed as accurately as possible, field verification of as-built conditions has been performed. The field verification produced detailed piping isometric drawings and pipe support sketches for each support upon which reanalysis is based. All field-verified piping isometrics and pipe support sketches are* independently verified by Surry Power Station quality control personnel. 7.2 7.2.1 QUALITY ASSURANCE/ENGINEE¥ING ASSURANCE EBASCO QUALITY ASSURANCE The EBASCO QA Topical Report ETR-1001, Revision 7, as approved Nuclear Regulatory Commission on December 15,

1978, 1s being to the Surry Unit 2 reanalysis activities.

7.2.2 STONE & WEBSTER QUALITY ASSURANCE/ENGINEERING ASSURANCE by the applied The Stone & Webster Quality Assurance program described 1n the VEPCO June 5, 1979 submittal to NRC, is being applied to the Surry Unit 2 reanal-ysis activities. 7.3 USE OF AMPLIFIED RESPONSE SPECTRA 'l;he use June 5, Spectra of amplified response spectra was extensively discussed 1979 submittal, The soil structure interaction amplified (SSI-ARS) are being used in the reanalysis in most cases. in the response For pipe runs ex-tending over a range of elevations S&W and EBASCO utilized an amplified response spectra enveloping the acceleration of the mass points spanning the elevation of the piping run. 7-1

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

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 8 SYSTEM OPERABILITY EVALUATION When an analysis is complete, the following procedure will be used in the decision making process as to the reportability to the NRG, including de-termination of system non-operability status of over-stressed supports/ restraints. If a potential support modification is identified, a check is made to de-termine if the new load is greater than 50 percent of the ultimate capacity of that support. A parallel effort is initiated whereby a preliminary support modification is designed and the preliminary support modification 1s sent to the site. If the new load is greater than 50 percent of ultimate, the station manager 1s notified of a potential nonconformance. If the station manager deter-mines a nonconformance exists and if the preliminary modification has not been installed, then the Technical Specification governs. If the new load is less than 50 percent of ultimate, the support is expedi-tiously installed after any nonconformances have been corrected. 8-1

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 SECTION 9 BRANCH LINE

SUMMARY

Branch lines are evaluated to assure that sufficient flexibility exists between the run pipe and the first few restraints on the branch piping. The flexibility of the branch pipe must be evaluated separately in each of the three translational directions and must be sufficient to prevent overstresses in the branch/run pipe interface due to thermal and seismic displacements imposed on the branch pipe. The procedure is intended to provide a secondary stress check based on run pipe displacements result-ing from the current analysis. If a branch line is part of the scope of work under IE Bulletin 79-14, a detailed evaluation is performed as part of the IE 79-14 effort. S&W has performed evaluation of branch lines in accordance with Section 6 of the August 1, 1979 report for Unit 1 (Vepco Serial No. 453A). EBASCO has performed evaluation of some of the branch lines by coding for the NUPIPE program and analyzing it for seismic anchor movement and thermal ~nalysis. Engineering judgement is used in qualifying the branch lines with smali displacements in the remaining cases. Thermal analysis is conducted by applying the thermal displacements from the run pipe and the operating temperature of the branch line. The seismic anchor movement analysis is performed by applying seismic inertia displacements. The applicable stress intensification factor (SIF) at the branch connection is included in the analysis. The stresses from both analyses are combined by absolute sum. Allowable stress is considered to be SA. 9-1

I I I I I I I I I I I I I I I I,, I I SURRY POWER STATION - UNIT 2 SECTION 10 RESPONSE TO NRG STAFF CONCERNS A meeting was held with the Nuclear* Regulatory Commission at Ebasco Services, Jericho Offices on October 24, 1979, to review pipe stress analyses within EBASCO' s scope of work. As a result of the discus-

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

These concerns were:

1)

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

2)

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

3)

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

4)

The verification of the NUPIPE computer program (benchmark problems). These concerns are addressed in the following sections. 10.1 SUPPORT STIFFNESS The original piping analysis of Surry Unit 2 did not consider the actual stiffness of the supports. Representative support stiffness was considered during the current reanalysis. During the pipe support reanalysis effort it has been observed that cer-tain anchor type supports expose wide flange members to torsional moments. This type of loading condition results in a very flexible support. As a part of the pipe support reanalysis effort, anchors have been reviewed for this type of loading and members modified to resist torsion as requ1recf. 10.2 NUPIPE COMPUTER CODE At EBASCO's

request, Nuclear Services Corporation (NSC) conducted a

thorough review of the NUPIPE program against the source codes, NSC has determined that all values utilized by the program, but not spec-ified by the user as input, are pertinent to the 1967 and earlier versions of the B3 l. 1 Power Piping Code. The code of record for Surry Unit 2 is B31.l 1955 with code class N-7. 10.3 PROBLEM 2538 - SUPPORT H-15 Problem 2538, a portion of the Low Head Safety Injection System (LHSIS) was originally analyzed as a hand calculation by S&W; SHOCK 2 was not used, therefore, problem 2538 is not within the scope of the Show Cause Order. In the original analysis, decal loads were applied to the re-straints in this problem. These decal loads did not include moments. 10-1

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

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

Support H-15 has been modified so that it will relieve the local overstress conditions. 10.4 BENCHMARK PROBLEMS EBASCO has performed four pipe stress to verify the NUPIPE

• computer program.

the staff in the EBASCO letter to Dr M (Letter Number VEP/NRC/002). 10-2 problems supplied by the NRC The results were submitted to Hartzman dated January 3, 1980

--i---i-- 1 I I I I I I I I I I I I I I I I I - ------------------------~--------=-:=- - -- - - -- -----1 SURRY POWER STATION - UNIT 2 APPENDIX A SYSTEMS AFFECTED A-1

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-~,---~~----- SURRY POWER STATION - UNIT 2 I The re.analysis included those safety related lines originally computer-analyzed with the SHOCK2 program. The systems line numbers, the associated computer problem numbers, and the flow diagram numbers are listed below. The I following table includes all seismically analyzed lines. The figure numbers re-

• fer to the FSAR drawings, and the Surry Unit 2, FM and FB drawings included in Appendix B.

I Responsi-Flow bilities Problem MKS Diagram I System Line No. for Analysis No. No. No. Low Head 8-SI-214-153 E 2695 127Dl FM-106A Safety 8-SI-292-153 E 2697 127D2 FM-106A I Injection 8-SI-214-153 E 2681 127Kl FM-106A 8-SI-292-153 E 2682 127K2 FM-106A 10-SI-284-152 S&W 2727 127Cl FM-106A I 10-SI-216-153 S&W 2727 127Cl FM-106A 8-SI-292-153 S&W 2727 127Cl FM-106A 10-SI-351-153 S&W 2727 127Cl FM-106A 6-SI-249-1502 S&W 2727 127Cl FM-106A I 10-SI-,349-153 S&W 2727 127Cl FM-106A 8-SI-214-152 S&W 2727 127Cl FM-106A 10-SI-283-152 S&W 2727 127C2 FM-106A I 10-SI-213-153 S&W 2727 127C2 FM-106A 6-SI-248-1502 S&W 2727 127C2 FM-106A 10-SI:-352-1502 S&W 2727 127C2 FM-106A I 10-SI-350-153 S&W 2727 127C2 FM-106A 10-SI-349-153 S&W 2727 127C2 FM-106A 10-SI-348-153 S&W 2727 127C2 FM-106A 8-SI-214-153 S&W 2727 127C2 FM-106A I 12-SI-247-602 E 2537 122Al FM-106B 12-SI-247-1502 E 2537 122Al FM...:106B 12-RC-324-1502 E 2537 122Al FM-106B I 10-RH-117-1502 E 2537

122Al, FM-106B, ll 7Bl 104A 12-SI-246-602 E

2555 122Dl FM-106B 12-SI-246-1502 E 2555 122Dl FM-106B I 12-RC-323-1502 E 2555 122Dl FM-106B 10-RH-116-1502 E 2555 122Dl FM-106B 6-SI-248-1502 E 2539 122Kl FM-106B I 6-SI-249-1502 E 2539 122Kl FM-106B 6-SI-250-1502 E 2539 122Jl FM-106B 6-RC-321-1502 E 2539 122Jl FM-106B I 6-S:I-343-1502 E 2539 122Kl FM-106B 12-SI-245-602 E 2709 12211 FM-106B 12-SI-245-1502 E 2709 12211 FM-106B 12-RC-322-1502 E 2709 12211 FM-106B I High Head 10-SI-206-153 E 2689 127Fl FM-106A Safety 6-CH-372-152 E 2735 127Gl FM-105B ,I Injection 4-CH-412-152 E 2735 127Gl FM-105B 3-CH-373-152 E 2735 127Gl FM-105B 8-sr..:214-153 E 2735 127G2 106A I A-2 I

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- -~----

I ~ -~ - I SURRY POWER STATION - UNIT 2 I Responsi-Flow bili ties Problem MKS Diagram System Line No. for Analysis No. No. No. I High Head 8-CH-504-152 E 2735 127Gl, G2 105B Safety 8-CH-317-152 E 2735 127Gl, G2 105B I Iqjection 8-SI-217-152 E 2735 127Gl, G2 106A (Cont'd) 8-SI-292-153 E 2735 127Gl 106A 6-SI-218-152 E 2735 127Gl, G2 105B, 106A, 6-SI-219-152 E 2735 127Gl, G2 105B, 106A! I 6-SI-278-152 E 2735 127Gl, G2 105B, 106A, 6-CH-501-152 E 2735 127Gl 105B 6-CH-502-152 E 2735 127Gl 105B I 6-CH-503-152 E 2735 127Gl 105B 8-CH-505-152 E 2735 127G2 105B 8-CH-506-152 E 2735 127Gl, G2 105B 8-SI-207-152 E 2735 127G2 106A I 8-SI-302-152 E 2735 127G2 106A 8-SI-170-153 E 2735 127G2 106A 8-SI-172-153 E 2735 127G2 106A I 10-SI-206-153 E 2735 127G2 106A 6-CH-318-152 E 2735 127Gl 105B 6-CH-319-152 E 2735 127Gl 105B I Residual 14-RH-101-1502 E 2508 117Al FM-104A Heat Removal 14-RH-102-602 E 2508 117Al FM-104A 10-RH-104-602 E 2508 117Al FM-104A I 10-RH-105-602 E 2508 117Al FM-104A 12-RH-106-602 E 2508 117Al FM-104A 10-RH-107-602 E 2508 117Al FM-104A I 10-RH-108-602 E 2508 117Al FM-104A 10-RH-109-602 E 2508 117Al FM-104A 10-RH-110-602 E 2508 117Al FM-104A I 12-RH-112-602 E 2508 118Al FM-104A 14-RH-118-602 E 2508 ll 7Al FM-104A 12-RH-119-602 E 2508 117Al FM-104A 12-RH-112-602 E 2540 ll 7Bl FM-104A I 3-RH-113-602 E 2540 ll 7Bl FM-104A 4-RH-115-152 E 2540 117Bl FM-104A 10-RH-116-1502 E 2540 ll 7Bl FM-104A I 10-RH-117-1502 E 2540 117Bl FM-104A 6-RH-120-152 E 2540 ll 7Bl FM-104A 10-RH-137-602 E 2540 117Bl FM-104A 6-RH-120-152 E 2554 117Cl FM-104A, I 101A I I I A-3 I

1~--------- ------------------------------ ---------------------------- ------------~= __: ______ --- 1 I I I I I .I I I I I I I I I I I I I I System Main Steam Feedwater Auxiliary Feed water Service Water SURRY POWER STATION - UNIT 2 Responsi-bilities Prob le,~. Line No. for Analysis No. 30-SHP-102-601 S&W 2588 30-SHP-102-601 S&W 2588 30-SHP-103-601 S&W 2579 30-SHP-101-601 S&W 2346 30-SJ{P-102-601 S&W 2346 30-SHP-103-601 S&W 2346 30-SHP-124-601 S&W 2346 30-SHP-123-601 S&W 2346 30-SHP-122-601 S&W 2346 14-WFPD-117-601 S&W 2569 14-WFPD-113-601 S&W. 2573 14-WFPD-109-601 S&W 2571 6-WAPD-101-601 E 2473 6-WAPD-102-602 E 2473 3-WAPD-109-601 E 2473 3-WAPD-110-601 E 2473 3-WAPD-111-601 E 2473 3-WAPD-112-601 E 2473 3-WAPD-113-601 E 2473 3-WAPD-114-601 E 2473 6-WAPD-150-601 E 2473 6-WAPD-151-601 E 2473 6-WAPD-101-601 E 2683 6-WAPD-102-601 E 2683 6-WAPD-103-601 E 2683 6-WAPD-104-601 E 2683 4-WAPD-105-601 E 2683 4-WAPD-106-601 E 2683 4-WAPD-107-601 E 2683 4-WAPD-108-601 E 2683 6-WAPD-50-601 E 2683 6-WAPD-52-601 E 2683 24-WS-126-10 E 2465 24-WS-128-10 E 2467 24-WS-130-10 E 2469 24-WS-132-10 E 2471 A-4 MKS No. 101D 101D l02D 103A 103A 103A 103A 103A 103A lOOG 101G 102G 118Al, A2 118A2 118Al, A2 118Al, A2 118Al, A2 118Al, A2 118Al, A2 118Al, A2 118Al, A2 118Al, A2 118G2 118Gl 118Gl, G2 118Gl, G2 118Gl, G2 118Gl, G2 118Gl, G2 118Gl, G2 118Gl, G2 118Gl, G2 119Al 119A2 119A3 119A4 Flow Diagram No. FM-14A FM-14A FM-14A FM-14A FM-14A FM-14A FM-14A FM-14A FM-14A FM-18A FM-18A FM-18A FM-18A, 18B FM-18A, 18B FM-18A FM-18A FM-18A FM-18A FM-18A FM-18A FM-18A,18B FM-18A,18B FM-18A, 18B FM-18A,18B FM-18A FM-18A FM-18A FM-18A FM-18A FM-18A FM-18A,18B FM-18A,18B FM-21A FM-21A FM-21A FM-21A

i-~,------~--- ------- - ---- -- SURRY POWER STATION - UNIT 2 I Responsi-Flow bilities Problem MKS Diagram System Line No. for Analysis No. No. No. I Pressurizer 4-RC-334-1502 E 2000 124Al FM-103B Safety and 3-RC-335-1502 E 2000 124Al FM-103B I Relief 3-RC-361-1502 E 2000 124Al FM-103B 6-RC-320-602 E 2000 124Al, A2 FM-103B 6-RC-362-602 E 2000 124Al, A2 FM-103B 12-RC-336-602 E 2000 124Al, A2 FM-103B I 6-RC-337-1502 E 2000 124Al, A2 FM-103B 6-RC-338-1502 E 2000 124Al FM-103B 6-RC-339-1502 E 2000 124Al FM-103B I 6-RC-340-602 E 2000 124Al, A2 FM-103B 6-RC-341-602 E 2000 124Al, A2 FM-103B 6-RC-342-602 E 2000 124Al, A2 FM-103B I Pressurizer 4-RC-,314-1502 E 2771 125Al FM-103B Spray 4-RC-315-1502 E 2771 125Al FM-103B 2-CH-368-1502 E 2771 125Al FM-103B I HP Steam to 4-SHP-125-601 E 2862 131Al FM-14A Auxiliary 3-SHP-132-601 E 2862 131Al FM-14A I Feedwat~r 3-SHP-128-601 E 2862 131Al FM-14A Pump 3-SHP-131-601 E 2862 131Al FM-14A 3-SHP-157-601 E 2862 131Al FM-14A 4-SHP-126-601 E 2864 131Bl FM-14A I 3-SHP-129-601 E 2864 131Bl FM-14A 4-SHP-127-601 E 2869 131Cl FM-14A 3-SHP-130-601 E 2869 131Cl FM-14A I ~-SHP-135-601 E 2869 131Cl FM-14A Containment FM-lOlA 10-CS-104-153 S&W 2521 123Al I and Recir-8-CS-123-153 S&W 2521 123Al FM-lOlA culation Spray 10-CS-103-153 S&W 2533 123Al FM-lOlA 8-CS-122-153 S&W 2533 123Al FM-lOlA 10-CS-103-153 S&W 2547 123Cl FM-lOlA I 8-CS-133-153 S&W 2547 123Cl FM-lOlA 10-cs-104.:...153 S&W 2549 123C2 FM-lOlA 8-CS-134-153 S&W 2549 123C2 FM-lOlA I 10-RS-112-153 E 2546 123Dl FM-lOlA 8-RS-123-153 E 2546 123Dl FM-lOlA 10-RS-104-153 E 2541 123D2 FM-lOlA I 8-RS-121-153 E 2541 123D2 FM-lOlA 10-RS-103-153 E 2542 123D3 FM-lOlA I 8-RS-120-153 E 2542 123D3 FM-lOlA 10-RS-Ul-153 E 2543 123D4 FM-lOlA I 8-RS-122-153 E 2543 123D4 FM-lOlA I I A-5

I SURRY POWER STATION - UNIT 2 I Responsi-Flow bilities Problem MKS Diagram System Line No. for An~lysis No. No. No. I Containment 10-RS-112-153 E 2560 123El FM-lOlA ~nd Recir-10-RS-104-153 E 2561 123E2 FM-lOlA I culation Spray 10-RS-110-153 E 2544 123Gl FM-lOlA (Cont'd) 10-RS-109-153 E 2533 123G2 FM-lOlA 10-RS-103-153 E 2548 123Hl FM-lOlA I 10-RS-111-153 E 2545 123H2 FM-lOlA 8'-CS-134-153 .E 2744 123Jl FM-lOlA 8-CS-133-153 E 2745 123Kl F~-lOlA 12-CS-102-153 E \\ 2753 123Ll FM-lOlA I 12-CS-101-153 E 2754 123Ml FM-lOlA 10-RS-109-153 E 2751 123N~ FM-lOlA 4-RS-i 14r-l 53 E 2751 123Nl FM-lOlA I 10-RS-110-153 E 2752 123N2 FM-lOli\\ 4-RS-115-153 E 2752 123N2 FM-101A I 8-CS-133-153 E 2755 123Pl FM-lOlA 8-CS-134-153 E 2755 123Pl FM-lOlA 4-CS-135-153 E 2755 123Pl FM-lOlA I 4-CS-136-153 E 2755 123Pl FM-101A 4-CS-105-152 E 2755 123Pl FM-lOlA l/2-CS-108-153 E 2755 123Pl FM-lOlA I 4-CS-106-152 E 2755 123Pl FM-lOlA 10-RS-101-153 E 2756 123Ql FM-lOlA 10-RS-102-153 E 2757 123Q2 FM-lOlA I Cpmponent; 18-CC-15-121 ~ 2604 112AA1 FM-22A Cpoling 18-CC-9-121 E 2605 112AB1 FM-22A I 18-CC-7-121 E 2601 11281 FM-22A 18-CC-14-121 E 2603 11282 FM-22A I Containment 8-CV-108-151 S&W 2650 137A FM-102A Vacu~ I I I Note: E = EBASCO I S&W = Stone & Webster I I A-6

I I I I I I I I I I I I I I I I I I I =---- ----- -- SURRY POWER STATION - UNIT 2 APPENDI~ B I FLOW DIAGRAMS - IDENTIFICATION OF PROBLEMS REANALYZED B-1

---

=--- ----- ------ --- ----------- - ----------------- --- - -----~~--

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNtT 2 APPENDIX B FLOW DIAGRAMS - IDENTIFICATIQN Of PROBLEMS REANALYZED Title Main Steam Feedwater Cross-Connects for Auxiliary Feed Circulating and Service Water Component Cooling Containment and Recirculatipg Spray Containment Vacuum and Leakage Monitor Reactor Coolaµt Sheet 1 Reactor Coolant Sheet 2 Residual Heat Removal Chemic~l anp Volume Control Sheet 2 Safety Injection Sheet 1 Safety Injec~ion Sheet 2 Refueling Water Storage Tank Crosstie / B-2 Drawing No. 11548-FM-14A 11548-FM-lBA l1448-FM-18B ll548-FM-2.1A H 548-FM-22A 11548-FM,-101.t\\. 115.48-FM-102A 1154~-FM-103A 11?48-FM-103B 11548-FM-104A 11,548-FM-lOSB 11548.,.FM-106A 11548-FM-106B 11448-FM-106C

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rVOS-60C-2

WCMU-161-151 LO Pl 6~WCIIU-152-151

,VOS-60C (TYP) 2-FW-P-4A EIIERGE~Y MAKE-UP PUMPS Z*WCIIU-161-151

-VGS-608 (TYP) Lvc;w-1SA (TYP) '-G!w<:IIU-152-151 VC\\11-l!IA (TYP) 6'-WCMU-156-1!'. EMERGENCY MAKE-UP SYSTEM FOR AUXILiARY FEED UNIT NO. 2 I.ODED: "-LVE.,...D DClE1T D'tAIN'-"LV!: I-FW-P-38 (FM-IBA, G-8) 6*-wcuu-111-1s1 TO SUCTION 2-FW-P-2 (II 548-FN-ISA,0-&) r6~WCIAU-153-l~I 4~ WCIIU-110-151 6"-WCMU -154-151 TO SUCTION 2-FW-P-3A.. (11548-FM-ISA, E-8) 4"-WCIIU-109-151 TO SUCTION 2-FW-P-311 (11548-FM-IBA, G-8) MOV FW151A INSIDE REACTOR CONTAINMENT fvos-60c

• -6~WAPO-l-601 (FII-IBA, C-6)

V 6"-WAPD-150-601 ISOuH ION VALVE - LEAK TEST r~S-60C. 110,000 GAL ,---;:::====13~~~K CROSS-CONNECTS FOR UNIT NO. 2 AUXILIARY FEED FROM l.!NIT NO. I UOY FWl5111 I INSIDE REACTOR CONTAINMENT (TYP) VCW-60A I V ISOI.A TIOIII VALVE LEAK.. TEST 3* 4-YOS-60C CROSS-CONNECTS FOR UNIT NO I AUXILIARY FEED FROM UNIT NQ 2 -Tm: n.,t)ISl.'l'IDI QI! !B'!.S Dl{ll!II'. lllT IOT U CCIPIBD 01 UDD l'OI: OTHD TUI YD co1suoct1aa. D.DrDDCS cm BPAI& GP nm 1'Uft *&eWn lSICllDD U ~. 1111.a ll.OCI:.

• NOTES:

-NEW 110,000 GAL CONDENSATE .-------*---IS*IORAGE TA .-------i 2-CN-TK-IA U54llflMIIJ!il --EXISTING R~FERl;NCE DRAWINGS: FLOW DIAGRA11°:.. FEEDWATER FM-IBA FLOW *01AGRAM-FEEDWATER 11~&-FII-IIA VALVE OPERATING NUMBERS-CROSS" CONNECTS FOP. AUXILIARY FEED Fll-611B

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FLOW DIAGRAM CROSS-CONNECTS FOR AUXILIARY FEED ,..; *&UIIRY ?OWE~ STATll)lf VIRGINL"\\ ELECTRIC AND POWDI COMP-"". 11448-FM~IBB -* J r

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• FURNISHED BY C:QUIPMENT MA...,FACTURER MCO-1-'aAIN CON.Tli:.'J:... B0>.20

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f. PROVIDE ROOT VALVES AT ALL FLOW ORIFICES DET-

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~ \\.. NOTE II ',..~,-i II\\ '. ~ (~Mf'*-1~0~1 ! ~1-Ts~ 1~-~.* L'lOTE II - t~**W,2.-1502. 4~Rc- ;m,-1soz. e.--., _rli:*TSe>z s{*DG 273*15"<-i'\\ ""4'1'>--<>' (FM-10011>)(9*7)./ \\-,t~c-4-0S ~IS 0~ (1'

• i

'* '-*

• l *TS&

-~ * --**-' '-l-il.c-.:ot.-1}"02. [_c:ot,JNE TION FOR AT"rl>.CHING. PLA'=a,1 H0~6 !COR DETERMINING WATER. L VEL. PR..ICR TO REFUELING OR LOOP IN,E:NANCE, TO BE.LCCI\\TED AT HIGH POI T IK PIPE, f* CH-l23* 150Z (FM*IOSC~ NO,ES: GENERA.L NOTE'a> ~ REFER.ENCE DHG FM-10?> e, REACTO~',""' \\ :-\\*,R::~::* L VE~='.EL ~-'....oet----t1V (NOTE 10) \\ f-------.......,>---------------- ONNECTION FOR A"TTAC.l.!ING MK~

• I 1 ';,A.I PLASTIC HO&E TO e,LIND FL.-'NGE

..___z:'-RC-:IS6*IS02 ON VENT 111:ADER (i"-YA*IOB-l!:14) (fM*IOOB) ~ 1°Rc.-1502.\\ l"RC.-1501'.---._~* ,yr-** --~ ~. n.oo 7 RC'a> i* "OT LE6

• PIP6" fRC-407*1-C-N9 TH£ lililFO:UOTION or, THIS DRAl!i11'1G !ll.Y MOT EE CortED OR USED FOR OTHER THAN THE COJriSfJtUCTIGll, IWNTDl-HC£ OR REPAIR Of TMI PLANT fAaUrf DISQUIJ) SN 1k[ mu aocK.

ORIGINAL 1551.1;: DESCRIPTION FLOW DIAGRAM REACTOR COOLANT SYST~M SHEET I 1972 EXTENSION-SURRY POWER STATION VIRGINIA ELECTRIC AND POWER COMPANY STONE & wkiisTEH *::O.GINEKRING CORroRA.TION II05TON. MASS. 11548-FM-103 A 4 7 I* " t.i i-:,*.

• ~,: *.'.

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l47!-,

{*RC-501-ISI:- PP.I MARY GR,-Df- -,-"'4-----__._-,-,--...,r'i <<~~FM-290) / ,.e. ':>*ICC*'!'S,-:52 *. ( A-7i /*-3'-PG*,O-IS~ Z.-*PG*S5*15"t. ~1448-FM-290) (,Fi) -:_: 0'":i-21'2.-t=c..-, V' ~ '::, --~'---j ~~~*'C'.:P-.) (c-7) L DUCIIPTIO!I 2.'-R-.:-3"1=,*'52 ~-vC.*323 -1s2. (FM *ICC"' (J-<\\) 'Z."*DG-217-15'2. (F>.1-100 B) (c-3) 1-:.*-c.::-:!~-:.::: v---l l*T58 2.4C.7 G.'-.:C-342-~2. ,;.oc.-323-IS,2. (*"1*'C='8) (J*+) -.;,:. e:- 3,.f.1-..:..;.: A".J>'. ~Ay*~-----.,_ ______._-t MKS-IZSAI ci.ia~, it*v.::x_ G.":>1 JTC"!:*!,. ::,,Y'S,_.,..,_........... (FM-1(--S.::.) P<:V 2~:,:,A '*. 00.-220*152. ffM-IOOBJ(IH PCV

• 'HHB ftx.*225*15 (Fl'l*IOOE))

(H*.. ) TE: z*,,

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1. n J/P,.a,a.. n*.,.., m:m ~

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• CIID8 bl.ta...U.f DnAUD 11?!11 ROS -

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• l'Sl9IlD.

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• 1111mm... ftS'1'WBnDU!f~

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a. m.,_..

A 1CI I 'ID. & llaall DP 4'-6 !/Ill*. ALL BffA!ll IDOP PIPDI Alm m aaum.a aAI. u.a..,...,. llall,dml'

• m ya, ar Y&U,

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2'-<ec- " 51*1S02

• f ss**::>0'2.-I* C-N<;>

/ 1--- ~MPUNC. ~'I'$ (fM-,l"I.A)t1'-1) 58-CL 1,;02...J-CL*I-C.-N9 ~IF'fCES IF REQUIAEO. Sl.ANK ORI'~ PlAT[ wtLL Y: DRflLED IN FIELD AS REQUIRH.

30. OR.A.IN UN[S TO 8E SLOPED DOWN n:tOM LOOP 5£1L5 TO COIOttCT WITH COMMON RO.lEF \\..\\Ut TO PR£S,S1JR'.2EI\\ RW£F 'T"'<<*

LEGENI> A.C. - INDICA.TEO MAJJUA.L ~LYE POSITlml MAINTA.INED BY.IDHINISTit.'1"'1£'..!" CONTROL MUCH IS CONSIDERED EQUl'ALENT TOALOC.KED W.YE. J~[::i~~E) M1,s1LE BARRIER AJQ'OIII EDNlWIY HILD F.O.* f'Ali. OPEN f.C.

• FAIL CLOSED D *. LOC.,l tJSUtt.lN V * \\.Oc..AL VEMT

'

• TRFPED ev ~RDS INITIATlCIN fflTEM C'l'l - NELDL£. VALVE A.q,. - HEAT TAAClHG cll:i -OIAPHRAGIII \\IALVE wnK Al.. OPE.RATOR LMC *LfAKAGE MONllORING CONNl!IION REFLRCNCE owm, COMPON£.hT coa..lN. svsn. "'

SAMPLING SYSTI:M FM-22AfB FM*3ZA FM-1000 F'M-10?.f.' FM*l04A FM-I058(C F'M-10!11!

• VENTe'DRAIN S't'STEM REACTOR COOLANT SYSTEM RESIDUAL HCAT REM:NAL CHEM ( VOLUME CONTRO..

SAFETY INJE.CTl6N SYSTl'.M BORON RtS:0\\/ERY SYSTE'M 114<118-FM*'29AIO 7 OtmllO? REA!.TDR CONTAINMENT D£5ClttfTION FLDN DIAGRAtv1 REACTOR COOL/\\NT SYSTEM

• SHEET 2 teT2 EXTENSl<'N-SURRY POWEii S'MTION

VIAGlfl'IIA ELEC1"RIC AND POWl[R COMPANY 11548-FM-1638 . L. l

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• 1a*-11.11*JO'l*

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• • COHTAINr.ofE.NT I

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• • -"*-cs-1!1-112.

(FM-101,._) (<-3)

---

~

4"*R*<-U5-15Z -t~ Rt-,OHM>tlt I.GOP fMO'TLI.Q (Fl!l*IOU) (H-J) DISCalntON RER1'ENC£0WH, Fl.OW DIAGRAM COWOtl£NT COOLING Ft... Z."2.A l'lOW DIAGRAM S~INI> f'Y*>ZA 1'1.0W CfAwlllAII VENT & CllAIN FM*I008 f"LOW DlAi&ill.AA\\ COll?IM.IRT i,t!CIII.C.SIUN IY6. FM*IOIA FLOW OIMtAAM 1\\1:AC.TfM COO\\.A*T

• ra1-1o3A4B l'LOW OIAGIIAM CHh\\lC.._
• YD*UMI CM7AD.:
• 11-IOSC*

f\\.ow*O-AM SAF(TY III.IECTION nt--8 ~Vl OPlfl. NO. R!.51DUAL HEAT REMOVAi. SYS -.FM--97A LC<<NO: D OIDl~ATES.1,.,X.AL IJltA\\NCOMlrilEC.TION Y oYDICATE 5 LOU<\\. VENT CONIIUTIOII Qq OUT&ll>L Ml!'>ilLE. EIARRl&lt IM. 1~,11>6 MIHl~66.. IUU3"- 'l' PLIJ6-&£D CONNE.CTIDM NQTH:*:. FLOW* DIAGRAM ICUAL !-£AT REM)VAL SYSTE~~ 197Z EX'l£N5IOII - SURRY POWER 5fA1ION VIRGINIA ELECTFllC AND l"OWER COMPANY aftHd

• wtiisf& IS'diN&idito ooitrou.'IION 11548.:FM-104A 7

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• I I I' j' I I I ' I: I 0

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• B; I' *; -.:..:.

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• -<:i--~~~...

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• f-'>-4t !)'I...
• X4ZD

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• NOT!: o.1 2~gcP3//

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2. Yz."*VOS*ISV 2, ~--'<<~1'5W 1,,....._,OS*ISY FLOW DIAGRAM HEM & VOLUME CONTROL SYS SHEEi-2 i97Z EXTEN510N

• sukiW POWER STATION

'IIRQlnlA ELl!:CTRle A.:D POWU COMflAN'i' mn:

• w&Hlii iN&A&i&d OOJIPl111.l'l'IOII

!1548-FM-1058 7 I ~ J* J-u

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• 1-1 '1f-U II Dl:5C.klPTIOM DECCIHll'TION*

-=-'". I .. _.., ~. ffli f,iflWCll,natl GIi l1US DfWlneG IIAY 80I' el COPIED oa usm FOR OTHU TlWI TM[ tOAr.JNCTl(IL

• • NTO.AIIICI r* a£P.Ul. 01 TKl Pl.Alff fAQUTf

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4.

LOCATE CONMl:CTION CLOSE TO ~ON IMJ(CTIO.. TA~lf e F1V - z.:i,;,3 TO BE ~AUC.£0 ~URIN& "'IYD*J TE'STU.r-c LOCATE CONNECTION CLOSE TO~VE! HCW*l8c.7C ii.~ D lHIS IS A SPECIA.L VALVE* FUMC.n(.'l!IIS lOT.H ~ A CCNTROL ANO A RELIEF VALVC TAPS MUST IS£ Z"*O'"JPART SAM?i.E CONNECTION SliAU. BE BRDIJGHT IIITO THE ANtJUL.L' 5 MGIOII B[Hl>II A COlUMk 1'r;ilCH ~HIE'LDS 't&~i£ A!OM ACtl:~IJLATOR. 1.'..i. y.1.,.,"':. :.~.* 1 ':*.!': ~ !"'P~S '!'"C -*.,,;.... _...,.,e..... !. t'Jil' N *.~ IA.L,i. ~YM'QOL PC.. DENOTES f','.\\\\. 4=.L.OS.E~ V~VE. S'fH.M>L A..C.. t)Et,JOTE5 ADM\\Nl~Thlrl.T:'11'~ COkTRoJ\\.. V"""--.*L SYl"~OL - Qi.NOTE.:, r\\'Z:llr,.T TRM..I. VN.V"C 'w'i.U> '-NO KEPA.U-.TIOW fOtl SC"t.lt lO~ INloT"LL ID!. TO 'iOS TA>.lil~1TIDII PiUt..

• .*.""=u:=::, -.,.,... :"*-:--:...:~

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• D'!A.U..llil.6.E IIIJXIL\\A.t."C" et.O~

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• WD8TD Zl'ICHIQIDlll'IQ CXJm'OIIAno!f Je 1972 EXTENSIL)N-SURRY P,'WER STAT :'N VIRGINIA ELECTRIC AND POWER COMPANY

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• ..11..i.

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• I I

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• ""151DE. ""S'::ilL.E
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• rc.*~t~So *ISOZ

"";:) LOCAL SAMPLE SEE NOTE"F" l I I I

• I

....JT!:S:

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---

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• g ':'~ E.XTEN ~l\\"'!"-J-S.. *P.RY F.:)WER STA.T1:t.i

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0.

114 48-FM-106C ll II 7 I

I SURRY POWER STATION - UNIT 2 I I I I I I I I APPENDIX C I RESPONSE TO IE BULLETIN 79-04 I I I I I I I I I C-1

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 APPENDIX C RESPONSE TO IE BULLETIN 79-04 Velan ~wing check valves, sizeli 3 and 6 inches, following seismic Category I piping sy~tems: are inst al led in the a) Chemical and volume control syst~m b) Safety injection systems A detailed listing by line number 1.s contained in the following table. l.,ines with 6 inch check valves were origin~Uy. seismically analyzed by computer program or hand calculations. The re-evaluation of tl).ese systems using the correct v.!:llve weight is currently being done under the NUPIPE program. The results have shown. that the pipe. stress is within the allow-able for all lines. Lines w;ith 3 inch check valves were tions. .An estimated weight, overly actual valve weights. The incorrect calculations and re-ev~luation is n9t the related pipe tines are included C-2 analyzed originally by /hand calcula-conservative, was used instead of valve ¥eight has no effect on these required' however' these valves and in the scope of l;E Bulletin 79-14.

I I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 LISTING OF VELAN SWING CHECK VALVES CQVERED BY IE BULLETIN NO~ 79-04 SAFETY INJECTION SYSTEMS - UNIT 2 6 Inch 3 Inch 2-SI-79

• 2-sr-s2 2-SI-85 2-SI-88 2-SI-91 2-SI-94 2-SI-228 2-SI-229 2-SI-238 2-S1-239 2-131-240 2-SI-241 2-Sir242 2-SI-243 2-SI-224 2-SI-225 2-SI-246 2-SI-227 CHEMICAL AND VOLUME CONTRoi SYSTEM - UNIT 2 3 Incµ 2-CH-196 2-CH:-258 2-CH-267 2-CH-276 2-CH-309 2-CH-312 C-3 6-RC-317-1502 6-RC-319-1502 6-RC-320-1502 6-RC-318-1502 6-RC-316-1502 6-l{C-321-1502 6-SI-249-1502

~-SI-Z49-1502 6-SI-248-1502 6-SI-249-1502 6-SI-250-1502 6.-SI-345-1502 q-SI-344-1502 6-SI-353-1502 3-SI-346-1503 3-SI-270-1503 3-SI-34 7-1503 3-SI-272-1~03 3-CH-500-1502 3-CH-381-1503 3-CH-302-1503 3-CH-303..-1503 3:-CH-379-1503 3-CH-301-1502

I SURRY POWER STATION - µNIT 2 I I I I I I I I APfENDIXD I CORRESPONDENCE WlTH THE NRC I I I ,I I I I I I D-1

I I I I I I I I I I I I I I I I .,... 1 I I SURRY POWER STATION - UNIT 2 APPENDIX D CORRESPONDENCE WITH NRG The following is a listing of correspondence with the NRC related to the reanalysis effort. Item No. 1 .2 3 4 5 6 7 8 9 10 11 . 12 13 Date 3/ 13/79 4/2/79 4/ 13/79 5/18/79 5/25/79 7 /18/79 8/15/79 8/27/79 10/5/79 10/23/79 10/24/79 10/25/79 11/15/79 Signature Denton Stello Stello Stello Eisenhut O'Reilly 0 1 Reilly Denton O'Reilly Murphy Murphy Addressee NRG TO VEPCO Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt Proffitt* Proffitt Proffitt Proffitt Murphy Proffitt Eisenhut Proffitt D-2 Letter No./Subject Show Cause Order Ad~endum to Show Cause Order Use of Soil Structure Interaction Techniques Request for Further SSI Information Factor Adjustment to SSI Calculated Stresses Information Pertaining to IE Bulletin No. 79-14, Revision 1 Letter of Guidance on IE Bulletin No. 79-14 Lifting of Suspension Required by the Order to Show Cause Confirmation of Concur-rence Refers to NRC Inspection of Sept. 10-13 and Sept. 19-21, 1979 Refers to NRC Inspection of Sept. 13-14, 1979 Refer~ to NRC Inspection of Sept, 26-28, 1979 Refers to Soil Structure Interaction

I I I I I I I I I I I I I I I I I I I Item No. 14 15 16 17 18 19 20 21 22 23 24 25 26 . 27 Date 3/30/79 4/19/79 4/23/79 4/24/79 4/27/79 5/Z/79 5/2/79 5/22/79 5/24/79 5/24/79 6/5/79 6/ 8/79 6/ 8/ 79 6/12/79 SURRY POWER STATION ** UNIT 2 APPENDIX D (Cont'd) CORRESPONDENCE WITH NRG Signature Addressee VEPCO TO NBC Spencer D~nton/ Stallings Spencer Spencer Spencer

• Stallings Spencer Spencer Spencer Spencer Spencer Spencer Spencer D-3 Stello O'Reilly O'Reilly (I)' Reilly Denton/

Stello Denton Stello Hendrie Stello Stello Denton Denton Stello Penton Letter No./Subject 198/Initial Response to Show Cause Order Z70/LER 79-0l0/013L-0 289/Response to IE Bulletin No. 79-07 288/Response to IE Bulletin.No. 79-07 311/Transmittal to Two Sample Problems to EG&G Observations on Reanaly~is Effort 460/Submittal of SSI Information Comments on Moratorium/ Sµrry Reanalysis Response to NRG Letter of 4/2/79 Response to NRG Letter of 5/18/79 Sub~ittal of Report on Reanalysis Additional Information, Report on Re~nalysis of Piping Soil Structure Interac-tion.Report Modification Informa-tion, Reanalysis of Piping Systems

I I I I I I I I I I I I I I I I I I I Item No. 28 29

rn 31 32 33 34 35 36 37 38 39 40 41 42 Date 6/15/79 6/19/79 6/25/79 8/1/79 8/21/79 10/4/79 10/3/79 10/4/79 10/15/79 11/28/79 12/7 /79 12/13/79 12/21/79 3/22/79 3/30/79 SURRY PQWER STATION - UNIT 2 APP~NDIX D (Cont'd)

CORRESPONDENC~ WITH NRG Signature Addressee Spencer Denton Spencer Denton Spencer Denton Spencer Denton Spencer Denton Spenc~r O'Reilly Proffitt Denton Spencer O'Reilly Spencer O'Reilly Spencer Denton Spencer O'Reilly Spencer Denton Spencer O'Reilly S&W to NRG Kennedy Dent op Jacobs Herring D-4 Letter No./Subject Scheduie and Suppqrt information Support Modifications Support Information, Reanalysis of Piping Systems Submittal of Revised Report on Analysis Analysis Completion ot. Desig~ateq Supports - Outside Containment Response to NRG Letter of 7/2/79 Seismic Analysi~ of Piping Systems Response to IE Letier Dated 9/7/79 Ex tens i<;>n of IE Bulletin 79~14 Deadline Seis~tc Analysis of Piping Systems Response to NRG Letter of 11/8/79 Show Cause Order Reanc;llysis Show Cau~e 60 Days An~lysis Compl~tion Transmittal of S&W 9omputer Pr?grams Submittal of Comput~r Outputs

I SURRY POWER STA~ION - UNIT 2 I APPENDIX D (Cont'd) CORRESPONDE~CE WITH NRC I Item No. Date Signature Adqressee Letter No./Subject I 43 4/3/79 Jacol;>s Bezler Submittal of Benchmark Proble~ to Brookhaven I National Laboratory 44 4/6/79 l{.ennedy

Penton Transmittal of S&W Computer Programs I

45 4/6/79 Jacobs Stello Plan for Veritication of Dynamic Analysis I Codes 46 4/ 11/79 Jacobs Bezler Submittal of Corqputer I Outputs 47 4/ 13/79 Jacobs Steqo Update and Status of Verification Plan for I Dynamic Aq.alysis Codes 48 4/18/79 Jacobs Submitt?l of Computer Hartman I Out;puts 49 4/27 /79 Jacobs Bezler Submittal of Benchmark Problems I 50 4/27/79 Jacobs Stello Status of Verification Plan for Dynamic I Analysis Co<:).es 51 5/8/79 Rossier Neighbors Draft Out line of SSI-I ARS Repqrt 52 .5/9/79 Kennedy Stello Reference SHOCK 0 Program I 53 5/ 11/79 Kennedy . Stello Reference SHQCK 0 Program I 54 5/ 14/79 Kennedy Denton Proprietary <;:omputer Cod.es I 55 6/4/79 Jacobs Bezler Submittal of ;Benchmark Prob~ems I 56 6/12/79 J?CObS Bezler Submittal of Benchmark Problems I I D-5

I I I I I I I I I I I I I I I I I I I Item No. 57 58 59 Date 9/6/79 9/7/79 1/3/90 SURRY POWER STATION - UNIT 2 APPENDIX D (Cont'd) CORRESPONDENCE WITH NRC Signature Addressee Allen Stello Ebasco to NRC Nelson Hartzman Nelson Hartzman D-6 Le~ter No./Subject Resp?nse to NRC Letter of 8/10/79 Benchmark Problem VEP/NRC/001 B~nch~ark Problem VEP/NRC/002}}