ML18139A144
| ML18139A144 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/11/1980 |
| From: | Spencer W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML18139A145 | List: |
| References | |
| NUDOCS 8004140219 | |
| Download: ML18139A144 (84) | |
Text
{{#Wiki_filter:, e VIRGINIA ELEC:I'RIC.!\\.ND POWER COMPANY RICHMOND,VIRGINIA 23~61 April 11, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Denton:
SHOW CAUSE ORDER REANALYSIS REPORT REVISION NO. 1 SURRY POWER STATION UNIT 2 Serial No. l38B PSE&C/GLS:mac:wang Docket No. 50-281 License No. DPR-37 In the letter of February 22, 1980 (Serial No. 138), Vepco requested start-up of Surry Power Station Unit 2 based on the 11Report on the Reanalysis of Safety Related Piping Systems, Surry Power Station, Unit 2 11 of the same date. The report reflected the results of the pipe stress and pipe support analyses subject to final verification and modification installation. The purpose of this submittal is to update the original report to reflect revisions necessitated during modification installation and to make minor typographical corrections to the text to enhance clarity and consistency. The subject changes, which are noted in the margins, in no way alter our original conclusion 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. This revised report also incorporates our recent commitment to the NRC Staff (Vepco letter of March 21, 1980, Serial No. 138A) to complete the installation of all modifications associated with the Order to Show Cause prior to start-up of the unit. The additional commitments to complete certain portions of the work associated with I.E. Bulletins 79-02 and 79-14 are firm as outlined in the February 22 letter. If you have any questions with regard to this.submittal, please contact us. Attachment W. C./ pencer Vice President - P6 er Station Engineering and Construction Services Ao7.:> er s 1/lf o A~ry. t:1,t ~ H. 0&/To;y I cc: Mr. Victor Stello, Director Office of Inspection & Enforcement Mr. James P. 0 1Reilly, Director Office of Inspection & Enforcement, Region II 80 041402/. C/
I I. I I I I I I I I I I I I I I I I I RES?iULflTnij'J rnioc"ET fill E K:fi1T}\\\\J . ~.. n HJ.tu u a\\ . e,l,i blh a REPORT ON THE REANALYSIS OF SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION-UNIT 2 VIRGINIA ELECTRIC AND POWER COMPANY REVISION 1 EBASCO SERVICES INCORPORATED--- .. REGULATOR'{ DOCKET Fu.LE COPl JER1cHo, NEW YORK Docket#, '>.50- 2-8 I Control # s*,:,04-, 'I-o :z I 'I Date !_ /, I /80 of Documeiif: ll!§§U~fiQ3X P~:cgr fU.~ ' ., 80 04140 2..2'-f
I I ii I I I I I I I I I I I I I I I I VIRGINIA ELEC:rRIC AND POWER COMPANY RXCHMOND,VI:e:GXNIA 23261 April 11, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission
- Washington, DC 20555
Dear Mr. Denton:
SHOW CAUSE ORDER REANALYSIS REPORT REVISION NO. 1 SURRY POWER STATION UNIT 2 Serial No. 1388 PSE&C/GLS:mac:wang Docket No. 50-281 License No. DPR-37 In the letter of February 22, 1980 (Serial No. 138), Vepco requested start-up of Surry Power Station Unit 2 based on the "Report on the Reanalysis of Safety Related Piping Systems, Surry Power Station, Unit 2 11 of the same date. The report reflected the results of the pipe stress and pipe support analyses subject to final verification and modification installation. The purpose of this submittal is to update the original report to reflect revisions necessitated during modification installation and to make minor typographical corrections to the text to enhance clarity and consistency. The subject changes, which are noted in the margins, in no way alter our original conclusion 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. This revised report also incorporates our recent commitment to the NRC Staff (Vepco letter of March 21, 1980, Serial No. 138A) to complete the installation of all modifications associated with the Order to Show Cause prior to start-up of the unit. The additional commitm~nts to complete certain portions of the work associated with I.E. Bulletins 79-02 and 79-14 are firm as outlined in the February 22 letter. If you have any questions with regard to this submittal, please contact us. . ~ '\\;~(!,(.~ Very t~rly. *ours, W. C. pencer Vice President - Po~er 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
I I I* I I I I I I I I I I I I I I I I be: Mr. Mr. Mr. Mr. Mr. Mr. Mr. Mr. Mr. Mr. J. H. Sam C. W. C. C. M. E. A. J. w. D. W. W. C. B. R. J. C. Ferguson Mr. Brown, Jr. Mr. Spencer Mr. Sta 11 i ngs Mr. Baum Mr. Waddil 1 Mr. Speidell, Jr. Mr. Daley Mr. Sylvia Mr. Harris, Jr. G. A. A. L. R. H. F. M. R. M. C. M.
- w. T.
J. L. B. R. Helm Parrish, III Wood a 11, II I Alligood, Jr. Berryman Robinson, Jr. Davidson - Surry Wilson - Surry Crowe - S&W Mr. Mr. Mr. Mr. Mr. Mr. Mr. Mr. Mr. D. G. M. H. J. R. J. T. s. F. Cochran L. Strickler
- w. Maupin
- w. Nelson - Ebasco M. Daly K. MacManus L. Perkins A. Peebles - Surry C. Rossier - S&W
I I,.*.. I 1* .I I I* I I* I I I: II I* I, I I I REPORT ON THE REANALYSIS OF . SAFETY-RELATED PIPING SYSTEMS SURRY POWER STATION-UNIT 2 VIRGINIA ELECTRIC . AND POWER COMPANY. REVISION 1.** EBASCO SERVICES INCORPORATED--- .. --~Mid JERICHO, NEW YORK
ii I 1, 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 22, 1980 REVISION 1 APRIL 11, 1980 EBASCO SERVICES INCORPORATED JERICHO, NEW YORK
I I I Section 1.0 I 2.0 3.0 4.0 I 5.0 6.0 7.0 I 7.1 7.2 7.3 I 7.4 8.0
- 9.0
- 1 10.0 I
10.1 10.2* 10.3 I 10.4 Appendix 1* A B I C D I 1, I I,, II 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................................... 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 Flow Diagrams - Identification of Problems Analyzed B-1 Response to IE Bulletin 79-04................... C-1 Correspondence with. the NRG..................... D-1 1-i
I I I I 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 both inside and outside containment affected by the Order f 1 to Show Cause have been evaluated for the revised support loads from the pipe stress reanalysis. All of these hardware modifications have been or will be installed prior to start up of the Unit following the steam gen-erator replacement outage. Table 4-2 (Pipe Support Hardware Modification Summary) reports all the modifications resulting from the support re-analysis. As was the case in the piping system reanalysis, most of the modifications are the result of differences between the original design conditions and the actual field as-built condition. During the reanalysis of Surry Unit 2, 79 stress modifications (22 due to pipe stress, 57 due to nozzle overload) and 258 pipe support modifications were identified; while 63 pipe stress and 66 pipe support modifications were identified on Surry Unit 1, The differences in the number of modif-ications is not considered significant, due to the conduct of the re-analys*is applied to Surry Unit 2. Modifications on Surry Unit 1 were 1-1 1
I I I I 1* I I I I I I I I SURRY POWER STATION - UNIT 2 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 inter-fere with the completion of the Steam Generator Replacement Project. In addition, the Surry Unit 2 reanalysis included the QBE condition.
- Further, all pipe supports were as-built in the field and QC verified prior to re-analysis.
Lastly, to limit the interface problems between the Show Cause scope and other piping, supports were added to facilitate the NUPIPE re-analysis. 1-2
I I I I I
- I I
I I I I I,. I I SURRY POWER STATION - UNIT 2 SECTION 2 SCOPE OF REANALYSIS As described 1.n systems in the with a SHOCK 2 the NRC. the NRC Order to Show Cause, March 13, 1979, some p1.p1.ng Surry Power Station, Unit 2 were dynamically analyzed computer program that is not currently acceptable to 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 is also identified 1.n Appendix A by system and problem number. The results of the stresses, allowable in the evaluation of reanalysis are compared with code loads for nozzles and penetrations, pipe supports. 2-1 allowable and are pipe used
I I I I I I I I I I I 'I SURRY POWER STATION - UNIT 2 SECTION 3 PIPE STRESS RESULTS A total of 62 pipe stress problems were originally analyzed by the PSTRESS/ SHOCK 2 computer I 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-ARS). The resultant stresses and loads are used to evaluate piping, supports, nozzles, and penetrations. In accordance with the NRC letters of May 25, 1979 and November 15, 1979 to Virginia Electric and Power Company (VEPCO), the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a factor of 1.5 for the DBE and 1.25 for OBE conditions. All 62 problems have been reanalyzed. Table 3-1, Pipe Stress Re-Evaluation Summary, presents the results for these 62 stress problems. In Table 3-1, the figures for Original Total Stress, at the point of maximum total stress in the pipe, and Original Seismic Stress, at the same point, are extracted from original design stress {sometrics (MSK' s). In Table 3-1, the columns for New Total Stress, at the point of maximum total stress in the pipe, and New Seismic Stress, at the same point, were taken from the NUPIPE computer runs with the seismic inertial stress multi-plied by a factor of 1.5 and then added to the Seismic Anchor Movement (SAM) Stress for runs using the SSI-ARS. Even though Table 3-1 reports DBE results, stress analysis is performed for OBE also and modifications de-signed wherever necessary. The Original Total and Original Seismic Stresses shown in Table 3-1 were computed using the SHOCK 2 programs for the original design conditions, The New Total and New Seismic stresses were computed by the NUPIPE pro-3-1
I I I I I I I I,. I I I I 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 under the reanalysis program. For all the problems in which the SSI-ARS are used, the seismic inertial nozzle loads have been increased by a factor of 1. 5 for DBE per the NRC letter of May 25, 1979, and by a factor of 1.25 for OBE per the NRC letter of November 15, 1979. Table 3-3, Pipe Stress Hardware Modification Summary, lists the hardware modifications necessary to bring the pipe stress analysis to within code allowables. Of the 62 problems reanalyzed, hardware modifications were made to 17 problems due to pipe stress. These modifications consisted of 22 added, modified, or deleted supports. The modifications include those necessary to the flexibility analysis of the branch lines. A branch line (Problem No. 2508B) was rerouted as a result of thermal reanalysis, not as 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 Engineering Assurance review has yet to be completed for a few problems. It is ex-pected, however, that the number and type of modifications due to the stress and support analysis are correct and final. 3-2
'I SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet l of 5 PIPE STRESS RE-EVALUATION
SUMMARY
System Name Line Size Pi12e Stress ( 12si) and Reanalysis MKS NPS Original Original New New Problem Number + Res12onsibilit:t: Number (Inches) Total Seismic Total Seismic Allowable Low Head Safety Injection 2555 E 122Dl 10,12 12043 NA 7974 2449 30690 2709 E 12211 12 NA NA 19173 11427 33750 2537/2540/2540B E
- 122Al, 4,6, 12350 NA 2739 883 33750 117Bl 10,12 2539 E
122Jl 6 30368 NA 14771 7330 32985 11 122Kl 2727 S&W 127Cl 6 21179 NA 24352 17453 33750 12JC2 8,10 2681 E I27Kl 8 1677 307 1220 185 28485 2682 E 127K2 8 i677 307 1174 164 28485 2695 E 127Dl 8 21179 NA 2094 1103 28485 2697 E 127D2 8 21179 NA 1981 1022 28485 High Head Safety Injection 2689 E 127Fl 10 24649 NA 11773 9571 33750 2735 E 127Gl 3,4,6, NA NA 26660 17772 33750 127G2 8,10 Containment and Recirculation S12rai 2521 S&W 123Al. 8, 10 14904 NA 7790 4276 33561 2523 S&W 123A2 8,10 14904 NA 7977 6013 33561 2547 S&W 123Cl 8,10 12713 NA 23532 19 739 33561 2546 E 123Dl 8,io 3528 1576 8892 7328 28800 2541 E 123D2 8,10 3528 1576 18338 16636 28800 2542 E 123D3 8,10 3528 1576 18931 17252 28800
I SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 2 of 5 PIPE STRESS RE-EVALUATION
SUMMARY
System Name Line Size Pi:ee Stress (:esi) and Reanalysis MKS NPS Original. Original New New Problem Number Res:eonsibiliti Number (Inches) Total Seismic Total Seismic Allowable Containment and Re-circulation s:erai (Cont'd) 2543 E l23D4 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 7922 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 123Kl 8 24843 22577 15621 12559 33750 2753 E 123Ll 12 6136 2818 1344 394 33750 2754 E l23Ml 12 5649 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 11955 10125 10081 8397 33561 2755 E 123Pl 4,8 10369 6324 14311 7340 33750 2756 E 123Ql 10 NA NA 5705 2686 28485 2757 E 123Q2 10 5810 NA 3847 2381 28485* Main Steam 2577 S&W 100Dl 30 13824 NA 10763 2883 33750 2588 S&W 101Dl 30 18635 NA 12513 3041 33750 2579 S&W 102D2 30 13031 NA 11434 4120 33750 2346* S&W 103Al 30 19970 NA 32568 25477 33750 103A2 L_ ___ -
SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 3 of 5 PIPE STRESS RE-EVALUATION
SUMMARY
System Name. Line Size Pipe Stress (psi) and Reanalysis MKS-NPS Original Original New New Problem Number Responsibility 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 27000 2571 S&W 102Gl 14 17927 NA 14230 8443 27000 Auxiliary Feedwater 2473 E 118Al 3,6 8568 2407 20963 17736 27000 118A2 2683 E 118Gl 4,6 21230 NA 12988 9188 27000 118G2 Pressurizer Spray 2771 E 125Al 4 18560
- NA 8013 3088 30690 Pressurizer Safety and Relief 2000 E
124Al 3,4 9093 NA 8824 5421 30636 124A2 6, 12. Residual Heat Removal 2540/2540B E Listed Under Low Head Safety Injection System 2508A/2508B E ll 7Al 10,12,14 NA NA 13112 8461 24570 2554 E 117Cl 6 NA NA 12008 8824 29970 Service Water 2465 E 119Al 24 NA NA 6529 6212 21600 2467 E 119A2 24 NA NA 6561 6214 21600
SURRY POWER STATION - UNIT. 2 TABLE 3-1 Sheet 4 of 5 PIPE STRESS RE-EVALUATION
SUMMARY
System Name Line Size PiEe Stress (Esi) and Reanalysis MKS NPS Original Original New New Problem Number ResEonsibility Number (Inches) Total Seismic Total Seismic Allowable Service Water (Cont'd) 2469 E 119A3 24 NA NA 14687 13730 21600 2471 E 119A4 24 NA NA 12459 11:614 21600 ComEonent Cooling 2601/2603 E 112Sl 18 9696 NA 7043 5246 21600 112S2 2604/2605 E 112AA1 18 9696 NA 7074 5204 21600 112AB1 Containment Vacuum 2650 S&W 137Al 8 25,750 NA 13659 13037 21600 HP Steam to Auxiliary Feedwater Pum:e 2869/2862/2864 E 131Al 3,4. 22609 NA 24229 19559 27000 131Bl 131Cl
SURRY POWER STATION - UNIT 2 TABLE 3-1 Sheet 5 of 5 Legend: E = EBASCO S&W = Stone & Webster* NA = Not Available 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 + l.S SDBEI + SDBEA l.S SDBEI + SDBEA SLP + SDW + SDBEI + SDBEA = Original Total Stress SDBEI + SDBEA = Original Seismic Stress Where SLP SDW Longitudinal P~essure Stress Dead Load Stress SDBEI SDBEA sh Note: Seismic Inertial Stress, Design Basis Earthquake Seismic Stress due to Anchor Movements, Design Basis Earthquake = Allowable stress at maximum (hot) temperature The original total and original seismic stresses shown in Table 3-1 were computed using SHOCK 2 for the original design conditions. The new total and new seismic stresses were computed by the NUPIPE program using different mass models and, in most cases, different ARS's than the original calculations. More importantly, the reanalyses were based on field-verified, as-built conditions in 1979, which, in some cases, differ significantly from the original design conditions. For this reason, the new stresses and the original stresses in Table 3-1 are not comparable, as they do not necessarily represent the same physical conditions
- Soil Structure Interaction (SSI) Amplified Response Spectras (ARS) were used in the new analysis for all problems except Problem 2346 which utilizes a combination of SSI-ARS and the original ARS.
+ Problems having/ are counted as separate problems example: 2869/2862/2864 are counted as three problems. Using this method of counting the total number of pipe stress problems equal 62. 1 1
I I SURRY POWER STATION - UNIT 2 TABLE 3-2 Sheet 1 of 4 I NOZZLE AND PENETRATION
SUMMARY
SHOCK 2 Problems I Responsi-Vendor System bility Total No. No. Acceptable Nozzle Confirmation and For o~ Nozzles/ (Evaluation Modification Being I Problem No. Analysis Penetrations Complete) Required Obtained Low Head I Safety Injection 2555 E 1/0 1 0 1 I 2709 E 1/0 1 0 0 I 2537/2540 E 1/0 1 0 1 2539 E 0/0 NA NA NA I 2727 S&W 2/0 2 0 2 2681 E 0/0 NA NA NA I 2682 E 0/0 NA NA NA I 2695 E 0/0 NA NA NA 2697 E 0/0 NA NA NA I High Head Safety Injection I 2689 E 0/0 NA NA NA I 2735 E 3/0 3 0 3 Containment and Recirculation I Spray 2521 S&W .0/0 NA NA NA I 2523 S&W 0/0 NA NA NA I 2547 S&W 0/0 NA NA NA 2546 E 0/0 NA NA NA I I
I I SURRY POWER STATION - UNIT 2 TABLE 3-7 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 Penetratioqs Compl~te) Required Obtained Containment and I Recirculation 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 1 0 0 I 2561 E 1/0 1 0 0 2544 E 1/0 1 0 0 I 2533 E 1/0 1 0 0 2548 E 1/0 1 0 0 I 2545 E 1/0 1 0 0 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 1 0 1 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 2683 Pressurizer Spray 2771 Pressurizer Safety and Relief E E E 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 Sheet 3 of 4 Vendor Nozzle
- Confirmation Modification Being Required Obtained 0/0 0/0 0/0 0/0 0/0 0/0 NA NA 0/0 0/0 0/0 0/0 0/0 NA 0
0 0 0/0 NA 3 1 1 (Listed under Low Head Safety Injection System)
I I SURRY POWER STATION - UNIT 2 TABLE 3-2 .Sheet 4 of 4 I NOZZLE AND PENETRATION
SUMMARY
SHOCK 2 Problems I Respon1;i-Vendor System bility Total No. No. Acceptable Nozzle Confirmation and For of Nozzles/. (Evaluation Modification Being I Problem No. Analysis Penetrations Complete) Required Obtained Residual Heat I Removal (Cont'd) 2540B E 0/0 NA NA NA I 2508A/2508B E 8/0 8 0 8 2554 E 0/0 NA NA NA I Service Water I 2465 E 1/0 1 0 0 2467 E 1/0 1 0 0 I 2469 E 1/0 1 0 0 2471 E 1/0 1 0 0 I Component Cooling 2601/2603 E 2/0 2 0 1 I 2604/2605 E 2/0 2 0 1 I Containment Vacuum 2650 S&W 0 NA NA NA ~...,.. I HP Steam to Auxiliary I Feedwater Pump 2862/2864/2869 E 1/0 1 0 0 I NOTES: NA = Not Applicable I E = EBASCO S&W = Stone & Webster I I
I l System Name and Problem No. Low Head Safety Injection 2709 2537/2540 2539 Containment and Recirculation Spray 2549 2544 2745 2752 / / Pressur;i;r Spray/ 2771 Residual Heat Removal /' 2508B Reanalysis Responsibility E E E S&W E E E E E MKS No. I22Ll 122Al 117:Bl I22Jl 122Kl I23C2 I23Gl I23Kl I23N2 125Al 117Al SURRY POWER STATION - UNIT 2 TABLE 3-3 PIPE STRESS HARDWARE MODIFfCATION
SUMMARY
Overstressed Condition
- Seismic overstress Thermal overstress Thermal overstress (Branch Line)
Pipe contacts crane wall *during. *seismic condition. Thermal overstress Seismic anchor move-ment overstress Thermal overstress (Branch Line) Seismic overstress Thermal overstress Branch* line Attributed To: Seismic Reanalysis As-built As-built Seismic Reanalysis As-built As-built As-built As-built As-built 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 added. Rerouting of 1-1/2 in. pipe 2540 (Listed under Low Head Safety Injection System) Sheet 1 of 2 No. of Modifications 1 2 l l 1 2 l 2 l
System Name and Problem No. Residual Heat Removal (Cont'd) 2540B Reanalysis Responsibility E Component Cooling 2604/ 2605 HP Steam to Auxiliary Feedwater Pump 2862/ 2864/ 2869 Feedwater 2569 Notes: E EBASCO S&W Stone & Webster E E S&W MKS No. 117B 112AA1 112AB1 131Al 131Bl 131Cl lOOGl SURRY POWER STATION - UNIT 2 TABLE 3-3 PIPE STRESS HARDWARE MODIFICATfON
SUMMARY
Overstressed Condition Thermal and seismic overstress Seismic overstress Thermal and seismic overstress Insufficient branch line flexibility Attributed To: As-built Seismic Reanalysis As-built/ Seismic Reanalysis As-built Resolution Anchor replaced by vertical restraint, horizontal restraint removed. Two restraints added. Two anchors replaced by restraints, two snubbers and a spring added. Remove existing U-bolt on 3/4 in. line Sheet 2 of 2 No. of Modifications 2 2 5 1
System Name and Problem No. Low Head Safety Injection 2555 High Head Safety Injection 2735 Containment and Recirculation ~ 2544 2533 2753 2754 2751 . 2752 2755 2756 Reanalysis Responsibility E E E E E E E E E E SURRY POWER STATION - UNIT 2 TABLE.3 HARDWARE MODIPICATION
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-lB 2-CS-P-lA 2-RS-P-2A 2-RS-P-2B 2:-CS-P-lB 2-CS-P-lA 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 Sei~mic Reanalysis Seismic Reanalysis As-built/ Seismic Reanalysis As-built Resolution No. of Modifications One restraint added*, One spring hanger replaced by a two direction restraint Ten restraints added, one ve~tical 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 1 2 3 2 1 1 3 2 1
System Name and Problem No. Containment and Recirculation Spray (Cont'd) 2757 Auxiliary Feedwater 2683 Pressurizer Safety & Relief 2000 Residual Heat Removal 2508B Service Water 2471 Component Coolin!!; 260 I/ 2603 Note: E EBASCO SURRY POWER STATION - UNIT 2 TABLE 3 HARDWARE MODIFICATION
SUMMARY
- DUE TO NOZZLE AND PENETRATION-"*OVERLOADING Reanalysis Responsibility E E E E E E 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-lD 2-RH-E-lB Attributed To As-built Seismic Reanalysis As-built Seismic Reanalysis 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 vertical restraint added. One restraint removed Four vertical restraints replaced by springs and snubbers, two restraints added. 5 4 10 l 6
I I I I I I I I I I I I I I I I I I I SURRY POWER S'rA.TION - UNIT 2 SECTION 4 PIPE SUPPORT RESULTS Table 4-1, Pipe Support An~lysis Summary, summarizes the pipe support reanalysis program. Six hundred ninety four (694) supports (467 inside the ~ontainment, 218 outside the containment) on lines originally analyzed using Shock 2, were rean~lyzed as part of this Show Cause effort. Two hundred fifty eight (258) hardware modifications (175 iqside the containment, 83 outside:the conta:tnment) hav~ been identified. The modifications identified due to i the pipe support reanalysis are l;i.sted in Table 4-2, P:i,pe Support Hardware Modification Summary. Those modifications which result from the piping reanalysis are identified in Section 3. Only the modifications which result*. from the pipe support reanalysis are reported in Table 4-2. Of the modifications identified, only 109 were* the result* of seismic reanalys;i.s of the piping systems iden-tified in the
- Show Cause Order, while 149 were the result of differences identified between the as-built cot1ditions and the original design.
These conditions are identified in the table for each problem. For all the problems in which the SSI-ARS are used, the seismic inertial loads have been increased by a factor of. 1. 5 for DBE per the NRC letter
- of May 25, 1979, and by a factor of 1. 25 for OBE per the NRC letter of November 15, 1979 *.
1 .:.~
SURRY POWER STATION - U!HT 2 TABLE 4-1 Sheet 1 of 4 PIPE SUPPORT ANALYSIS
SUMMARY
System Name Total Modifications and Analysis Number of Evaluation or Additions Problem Number Responsibility Location Supports Complete
- Required Low Head Safety Inject:ion System 2537/2540 E
IC
- 33.
33 17 f 2555 1 E IC 17 17 9 2539 E IC 7 7 5 2681 E OC 4 4 3 2682 E OC 4 4 3 2695 E OC 10 HJ 4 1 2697 E oc 9 9 6 27u9 E IC 9 9 3 2727 S&W OC 17 17 8 I 1 High Head Safety Injection System 261:19 E oc 5 5 4 2735 E OC 52 52 23 I 1 Containment and Recirculation Spray 2521 S&W IC 15 15 4 2523 S&W IC 16 16 5 2547 S&W IC 13 13 6 2549 S&W IC 4 4 1 2546 E IC 12 12 6 I 1
SURRY POWER STATION - UNIT 2 TABLE 4-1 Sheet 2 of 4 ~. PIPE SUPPORT ANALYSIS
SUMMARY
System Name Total Modifications and Analysis Number of Evaluation or Additions Problem Number Res12onsibility Location Su1212orts Com12lete Required Containment and Recirculation ~ (Cont'd) 2541 E IC 11 11 6 2542 E IC 11 11 4 2543 E IC 12 12 6 11 2560 E IC 4 4 1 2561 E IC 5 5 3 2544 E IC 6 6 2 11 2533 E IC 5 5 1 2548 E IC 15 15 6 Ji 2545 E IC 17 17 9 2744 E OC 4 4 3 -*-~-- 2745 E oc 4 4 2 11 2753 E QC 4 4 1 2754 E oc 3 3 1 27 51 E oc 5 5 l 2752 E oc 4 4 0 27.'.>5 E oc 8 8 4 1 ~756 E IC 7 7 2 2757 E IC 9 9 3
SURRY POWER STATION - UNIT 2 TABLE 4-1 Sheet 3 of 4 ~ PIPE SUPPORT ANALYSIS
SUMMARY
System Name Total Modifications and Analysis Number of Evaluation or Additions Problem Number ResEonsibility Location su12Eorts ComElete Required Main Steam 2577 S&W IC 9 9 2 2588 S&W IC 2 2 0 2579 S&W IC 5 5 3 2346 S&W QC 41 41 6 Feedwater 2569 S&W IC 9 9 4 2573 S&W IC 3 3 0 2571 S&W IC 6 6 1 Auxiliarx Feedwater 2473 E IC 34 34 12 2683 E QC 22 1 22 8 Pressurizer SEray
- and Relief 2771 E
IC 32 32 7 2000 E IC 29 29 15 11 Residual Heat Removal 2508A/B E IC 45 45 4 2540B E IC 5 5 4 1 2554 E QC l l l
System Name and Problem Number Service Water 2465 2467 2469 2471 Component Cooling 2601 2603 2604 2605.. Containment Vacuum 2650 High Pressure Steam to Aux Feedwater Pump 2862 2864 2869 Notes: E = EBASCO S&W = Stone & Webster IC= Inside Containment OC = Outside Containment SURRY POWER STATION - UNIT 2 TABLE 4-1 PIPE SUPPORT ANALYSIS
SUMMARY
Analysis Responsibility E E E E E E E S&W E E E Location IC IC IC IC IC IC IC IC oc oc oc oc Total Number of Supports 2 2 1 1 16 13 i7 17 3 6 3 7 Evaluation Complete 2 2 1 l 16 13 17 17 3 8 3 7 Sheet 4 of 4 Modifications or Additions Required l l 0 1 5 5 7 4 3 l (J 1 1 1
- 1.
11
SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 1 of lL. PIPE SUPPORT HARDWARE MODIFICATION
SUMMARY
SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE I PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION I Low Head Safeti I Injection Sistem I 2537/2540 E IC 2537 E IC 2,4 Local pipe wall. stress Seismic Hodify support over allowable Reanalysis 11 6 Support member over allowable As-built Modify support 9 Support does not allow.lateral As-built Hodify support /1 movement 2540 E IC 5 Insufficient clearance Seismic llodify support Reanalysis 7,8,9,10,13,26 Support member over allowable As-built Modify support t1 12,28 Pipe clamp over allowable As-built Modify support 11 17,19 U-bolt over allowable As-built Modify support 1 Weld over allowable As-built Add weld 11 8A Upward vertical restraint Seismic Modify restraint required Reanalysis for uplift lo_ad 2555 E re; 8 Support member over allowable As-built Modify support I 1 3 Weld over allowable As-built Add weld 4,5 Local pipe wall stress over Seismic Modify support 11 allowable Rea_nalysis 9,12 Support member over allowable Seismic Modify support / 1 Reanalysis 12A Upward vertical restraint Seismic Modify restraint required Reanalysis for uplift load 13,14 Insufficient lateral clearance As-built Modify support 11 2539 E IC 2 Upward vertical restraint Seismic Modify restraint required Reanalysis for uplift load 1 Upward 'vertical restraint. Seismic Hodify restraint required Reanalysis for uplift load
SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 2 of 12 PIPE SUPPORT HARDWARE MOOIFICATION
SUMMARY
SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION Low Head.Safety Injection sistem (Cont'd) 2539 (Cont'd) E IC H-1** Support member over allowable As-built Modify support 11 3,4 Support member over allowable As-built Modify support 2709 E IC 7 Support member over allowable As-built Modify support 2,3 Local pipe wall stresses over Seismic Modify support allowable Reanalysis 2727 S&W oc 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. 14 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 11 18 Loads out of spring range Seismic Replace spring Reanalysis 19 Supports restraint lateral As-built Modify Support movement 2695 E oc H-50 Support member allowable As-built Modify 11 over support A-16 Support member over allowable As-built Modify support C-53 Support member over allowable As-built Modify support A-17 Support member over allowable As-built Modify support
SYSTEM NAME AND PROBLEM NUMBER Low Head Safety Injection System (Cont'd) 2697 2681 2682 High Head Safety _Injection System 2689 2735 Containment and Recirculation Spray 2521 ANALYSIS RESPONSIBILITY E E E E E S&W SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION
SUMMARY
LOCATION QC QC oc oc oc IC MKS SUPPORT NUMBER H-49 C-7 C-17,C-18 A-14 A-15 2,4 5 4 2 5 C-38,C-39 C-40,C-41 19,34,42 22,37,45,28 36,21,23,44 1,3,16,31,39,24 4,6,18,33,41 26 2 REASON FOR MODIFICATION Support member over allowable Support not acting Support member over allowable Support member over allowable Support member over allowable Support member over allowable U-bolt over allowable Support member over allowable Support member over allowable U-bolt over allowable Support member over allowable Support* member*over allowable Support member over allowable Weld over allowable U-bolt cive:r allowable Local pipe* *will stress over allowable U-bolt capacity for side-load in insufficient ATTRIBUTABLE TO As-built As-built As-built As-built As-built As-built As-built As-built As-built As-built As-built As-built As-built As-built Seismic reanalysis Seismic reanalysis Seismic Reanalysis Sheet 3 of 12 RESOLUTION Modify support Removed support Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify support Modify support Add weld Modify support Modify support Add members to resist side load I 1 I 1
SYSTEM NAME MID PROBLEM NUMBER Containment and Recirculation Spray (Cont'd) 2521 (Cont'd) 2523 2547 ANALYSIS RESPONSIBILITY S&W S&W S&W SURRY POWER STATION - UNIT 2 TABLE 4 - PIPE SUPPORT HARDWARE MODIFICATION
SUMMARY
LOCATION IC IC IC MKS SUPPORT NUMBER* 3 4 5 l 2 3 4 5 2 4 6 8 REASON FOR MODIFICATION Frame overstressed with new loads out-of springs range U-strap has insufficient capacity Local stress U-bolt capacity for side-load is insufficient Frame overstressed Capacity of springs in-sufficient U-strap has insufficient capacity Local stress Lateral load fails U-bolt Rod hanger cannot resist upward load Insufficient clearance for thermal movement Insufficient latera-1 clearance for thermal movement ATTRIBUTABLE TO Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis As-built As-built 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)
- *- - - - -*.... ll'II - - - - - -* - SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 5 of 12 PIPE SUPPORT HARDWARE MODIFICATION
SUMMARY
SYSTEM NAME ~JKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION Containment and Recirculation SErai (Cont'd) 2547 (Cont'd) S&W IC 10 Insufficient lateral As-built Remove lateral 11 clearance for thermal stops (angle) movement 11 Local stress and support As-built Modify structure frame overstressed 2549 S&W IC 2 U-bolt failure As-built Add lateral restraint 2546 E IC 6 Support member over allowable As-built Modify support 9 Insufficient lateral & Seismic Modify restraint vertical.clearance Reanalysis 7,8 Insufficient lateral Seismic Modify restraint clearance Reanalysis 2 U-bolt restricts lateral As-built Modify support I 1 movement 22A Anchor stress over allowable Seismic Relocate adjacent Reanalysis restraint 2541 E IC 6,21 Support member over allowable As-built Modify support 7,9 Insufficient lateral Seismic Modify support clearance Reanalysis 2,8 U-bolt restricts lateral As-built Hodify. support 1 movements 2542 E IC 15,23 Support member over allowable As-built Modify support ! 1 .r-- 13,17 Insufficient lateral clearance Seismic Modify support Reanalysis
-* - - - - -) -.. - - - - SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 6 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 Spray 2543 E IC 15,23 Support member As-built .Modify support over allowable 12,13,14 Insufficient lateral Seismic Modify restraint clearance Reanalysis 24A Support member Seismic Relocate adja-over allowable Reanalysis cent restraint 11 2560 E IC 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-50 U-bolt over Seismic Modify support 11 allowable Reanalysis 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 Rupport 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 over allowable 10 Support member As-built Modify support over allowable 12 U-bolt over As-built Modify support allowable 9A Upward vertical Seismic Modify support restraint required Reanalysis for uplift load* 11 Support member As-built Modify support over allowable
SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 7 of 12 PIPE SUPPORT HARDWAP.E MODUICATION SUHMARY SYSTEM NAME MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTA!lLE PROBLEM NUMBER RESPONSIBILITY LOCATIOU NUMBER MODIFICATION TO RESOLUTION Containment and Recirculation SErai (Cont'd)
- .!548 {Cont'd)
E IC 8 Support member As-built Hodify support I 1 over. allowable 2545 E IC 3 Support member As-built Modify support over allowable 14 Upward vertical Seismic Modify restraint restraint required Reanalysis for uplift load 2 U-bolt over
- As-built Hodify support allowable 4
- iupport 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 Hodify support over allowable
- .!744 E
QC 2,3 Loads out of spring As~built Two-rigid re-range straints re-placed by springs 1 Local pipe wall tieismic Modify support stress over allow-Reanalysis 1 able 2745 E oc 4 Upward vertical Seisruic Modify support restraint required Reanalysis 3 Support does not al-As-built Modify support I low lateral movement 1 .:751 K QC 2 Local pipe wall Seismic Modify support stress over Reanalysis allowable L753 E oc 1 Support member As-built Hodify support over allowable
SURRY POWER STATION - UNIT 2 TABLE 4-2 Sheet 8 of 12 PIPE SUPPORT HARDWARE MODIF!CATION
SUMMARY
SYSTEM MKS AND ANALYSIS SUPPORT REASON FOR ATTRIBUTABLE PROBLEM NUMBER RESPONSIBILITY LOCATION NUMBER MODIFICATION TO RESOLUTION Containment and Recirculation SEra::t: (Cont'd) 2754 E OC 1 Support member As-built Modify support over allowable 2755 E oc 7,8 Upward vertical Seismic Modify support restraint required Reanalysis for uplift load 9 Upward vertical
- Seismic Redesign support restraint required Reanalysis for uplift load 11 Weld over allowable As-built Add weld 2756 E
IC 54,55 Support member A,;-built Modify support over allowable 2757 E IC H-90 Upward vertical Seismic Modify support for restraint required Reanalysis uplift load H-63, Support member As-built Modify support H-88 over allowable Main. Steam 2577 S&W IC 1 Local stress ex-As-built Modify. lug ceeds allowable 9 Local-stress ex-As-built Replace lug with ceeds allowable clamp 25}9 S&W IC 1 Spring variability SeiRmic. Replace spring ratio exceeded Reanalysis 4 Loads outside Seismic Replace spring spring range Reanalysis 5 Spring variability Seismic Replace spring ratio exceeded. Reanalysis Replace pipe lug Local stress over with clamp. allowable. 2346 S&W oc 1,2,3 Loads outside spring Seismic Replace springs, range, local over-Reanalysis modify lug stress in lug
SYSTEM AND PROBLEM NUMBER Hain Steam (Cont'd) 2346 (Cont'd) ANALYSIS RESPONSlllILITY S&W SURRY POWER STATION - UNIT 2 TABLE 4-2 ~ PlPE SUPPORT HARDWARE HODIFICATION
SUMMARY
!!KS SUPPORT LOCATION NUMBER oc 5,7,9 REASON FOR MODIFICATION Snub be rs, local stress,. and sup-port members are overstressed ATTRIBUTABLE TO Seismic Reanalysis Sheet !la of 12 RESOLUTION
- ,1odify snubbers and lugs
SYSTEM NAME AND PROBLEM NUMBER Feedwater 2569 2571 Auxiliary Feedwater 2473 2683 ANALYSIS RESPONSIBILITY S&W S&W E E SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MOD1FICATION
SUMMARY
LOCATION IC. IC IC OC MKS SUPPORT NUMBER 2,3 6 7 5 8,18 14 16 11,13,26,29,32 22 28,31 H-8A H-11 H-9 H-1(118Gl), H-5(118Gl) H-4 H-5(118G2) . REASON FOR 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 Upward vertical restraint required ATTRIBUTABLE TO Seismic Reanalysis As-built Seismic Reanslysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built As-built AR-built Seismic Reanalysis As-built Seismic Reanalysis 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 Modify support l-1 11 I 'I
SYSTEM NAME AND PROBLEM NUMBER Auxiliary Feedwater (Cont'd) 2683 (Cont'd) P*ressurizer Spray and Relief 2771 2000 Residual Heat Removal 2508A/2508B 2540B ANALYSIS RESPONSIBILITY E E E E E SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MCWIFICATION
SUMMARY
MKS LOCATION QC IC IC IC IC SUPPORT NUMBER H-1(118G2) 33 5,23 6,24 26,27 4,12,H-5,8,10,17 15 13 4A,7,H-lA, ll,ZlA,113 H-2 H-36,H-15,H-17 H-12 20,21 REASON FOR MODIFICATION Support does not allow lateral movement Upward vertical restraint required Weld over allowablf! Support member over allowable Upward vertical restraint required Local pipe wall stress over allowables Insufficient vertical clearance Support restraints lateral movement Support member over allowable Weld over allowable Support member over allowable Vertical support not required Upward vertical restraint required ATTRIBUTABLE TO As-built Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built As-built As-built As-built Seismic Reanalysis Seismic Reanalysis Sheet 10 of 12 RESOLUTION Modify support Modify support for uplift load Add weld Modify support 11 Modify for uplift load Modify support 11 Modify support Modify support - Modify support Add weld 1 Modify support Remove support Modify for uplift 11 load
SYSTEM NAME AND PROBLEM NUMBER Residual Heat Removal (Cont'd) 2540B (Cont'd) 2554 Service Water 2465 2467 2471 Component Cooling 2601 2603 2604 ANALYSIS RESPONSIBILITY E E E E E E E E SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE MODIFICATION
SUMMARY
LOCATION IC ex: IC IC IC IC IC IC MKS SUPPORT NUMBER 23 22 H-31 1 1 2 H-:-32A,H-28,H-44 H-30 H-31 4 6,8 9 10 H-38A REASON FOR MODIFICATION ATTRIBUTABLE TO Insufficient lateral clearance Seismic Reanalysis Support member over allowable Insufficient lateral clearance Support member over allowable Support member over allowable Local pipe wall stress over allowable Uplift vertical restraint required Insufficient lateral clearance Weld over allowable Insufficient lateral clear-ance Support member over allowable Upward vertical restraint requir_ed Upward vertical restraint required Upward vertical restraint required As-built Seismic Reanalysis As-built As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis Seismic Reanalysis Sheet 11 of 12 RESOLUTION Modify support Modify support Modify support Modify support Modify support Modify support Modify support for uplift Modify support Add weld Modify support Modify support Modify support for uplift load Modify support for uplift load Modify support for uplift load 1 .i 1
SYSTEH NANE AND PROBLEM NUHBER Component Cooling (Cont'd) 2604 (Cont'd) 2605 Containment Vacuum 2650 Uigh Pressure Steam To rtux. Feedwater Pump 2862 2369
- otes:
ANALYSIS RESPONSIBILITY E E S&W E E
- ~ Originally Problem No. 27ue*
SURRY POWER STATION - UNIT 2 TABLE 4-2 PIPE SUPPORT HARDWARE M01'IFICATION SUilllARY LOCATION IC IC oc oc oc MKS SUPPORT NUMBER H-38B H-35,H-36,H-38C H-38 H-38D H-25 H-25B H-25A H-23 1 2 3 3 5 REASON FOR MODIFICATION Upward vertical restraint required Weld over allowable Support member over allowable Local p,ipe wall stress over allowable Support member over allowable Upward vertical restraint required Upward vertical restraint required Weld over allowable Support restraint lateral movement Local stress at at-tachment on pipe exceeds allowable Local stress Support member over allowable Support member.over allowable ATTRIBUTABLE TO Seismic Reanalysis As-built As-built Seismic Reanalysis As-built Seismic Reanalysis Seismic Reanalysis As-built As-built As-built Seismic Reanalysis As-built As-built. Sheet 12 of 12 RESOLUTION Modify support for 11 uplift load Add weld 11 Modify support Modify support 11 Modify support Modify support for uplift load Modify support for uplift load Add weld 11 !1odify support Move support above 1 elbow and use trun- [ nion Modify support J1 Modify support Modify support
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 5 SCHEDULE FOR COMPLETION The.status of the reanalysis of those systems subject the installation of modifications identified as being reanalysis is shown in Table 5-1, Schedule for Completion. to Show Cause required by and the Reanalysis on all systems is complete pending final review. modifications on all lines will be installed prior to start Unit following the Steqlll Generator Replacement Outage. 5-1 Required up of the 1
Location of Problem Inside Containment Systems Outside Containment Systems Low Head Safety Injection High head Safety Injection Containment Recirculation Spray fiUXiliary Feedwater Balance of Systems NOTI:S: E S&W EBASCO Stone & Webster SURRY POWER STATION - UNIT 2 TABLE 5-1 SCHEDU~OR COHPLETION Reanalysis Responsibility E/S&W E/S&W E E E E/S&W Stress Reanalysis Complete Complete Complete Complete Complete Complete S~R Steam Generator Replacement Status Support/Restraint Reanalysis Complete Complete Complete Complete Complete Complete Modification Installation Prior to start-up following SGR outage Prior to start-up following SGR outage Prior to start-up following St;R outage Prior to start-up . following SGR outage Prior to start-up following SGR outage Prior to start-up following SGR outage 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 6 HIGH ENERGY LINE BREAKS For the high energy lines outside the containment addressed in Appendix D of the Final Safety Analysis Report (FSAR), only the main steam lines are included in this stress reanalysis. Each of the main steam lines has two terminal break locations, one at the containment penetration and the other at the main steam manifold. Each of the risers to the main steam relief valve headers has two terminal break locations, one at the main steam lines, the other at the tee into the main steam header. These terminal breakpoints are predetermined and are not changed as a result of the stress reanalysis. Two intermediate break locations were originally determined based upon maximum primary plus secondary stresses. Upon. reanalysis, two additional breakpoints on each of the steam lines were located. One of these points is lol;!ated immediately upstream of the check valve_ (TV-MS201A, TV-MS201B, TV-MS201C) and the other point is at the elbow just down:;;tream 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 SURRY POWER STATION - UNIT 2 SECTION 7 CONSERVATISMS The conservatisms applied to the design of the p1p1ng systems for Surry Power Station Units 1 and 2 were extensively delineated in Section 7 of the VEPCO June 5, 1979 submittal (Serial Number 453). The seismic capability of nuclear piping and the seismic event probability at the Surry Power Station were discussed in that submittal. The design of Unit 2 closely follows the design of Unit 1, applying the same conservative criteria with respect to safety systems and system redundancies. Similiarly, the reanalysis efforts on Surry Power Station Unit 2 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 vatisms applied to the Unit 2 reanalysis. in the conser-7.1 FIELD VERIFICATION OF AS-BUILT CONDITIONS To ensure that the pipe stress and pipe support reanalysis is performed as accurately as possible, field verification of as-built conditions has been performed. The field verification produced detailed piping isometric drawings and pipe support sketches for each support upon* which reanalysis is based. All field-verified piping isometrics and pipe support sketches are independently verified by Surry Power Station quality control personnel. 7.2 7.2.1 QUALITY ASSURANCE/ENGINEERING ASSURANCE EBASCO QUALITY ASSURANCE The EBASCO QA Topical Report ETR-1001, Revision 7, as approved Nuclear Regulatory 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 in the VEPCO... June 5, 1979 submittal to NRC, is being applied to the Surry Unit 2 reanal-ysis activities. 7.3 USE OF AMPLIFIED RESPONSE SPECTRA The use of amplified response spectra was extensively discussed in the June 5, 1979 submittal. The soil structure interaction amplified response Spectra (SSI-ARS) are being used in the reanalysis in most cases. For pipe runs extending over a range of elevations S&W and EBASCO utilized an amplified response spectra enveloping the acceleration of the mass points spanning the elevation of the piping run. 7-1
I I I I I I I I I I I I I I I I I I SURRY POWER STATION - UNIT 2 7.4 CONSERVATISMS APPLIED TO INERTIAL STRESSES In accordance with the NRC letters of May 25, 1979 and November 15, 1979 to VEPCO, the seismic inertial stresses and loads computed using the SSI-ARS have been increased by a factor of 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 This section has been deleted. Since all modifications will be installed prior to startup following the Steam Generator Replacement Outage, a system operability evaluation is no longer necessary. ~1 1
I I I 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 analysis. Engineering judgement is used in qualifying the branch lines with small displacements in the remaining cases. Thermal analysis is conducted by applying the thermal displacements from the run pipe and the operating temperature of the branch line. The seismic anchor movement analysis is performed by applying seismic inertia displacements. The applicable stress intensification factor (SIF) at the branch connection is included in the analysis. The stresses from both analyses are combined by absolute sum. Allowable stress is considered to be SA. 9-1
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 10 RESPONSE TO NRC STAFF CONCERNS A meeting was held with the* Nuclear Regulatory Commission at Ebasco Services, Jericho Offices on October 24,
- 1979, to review pipe stress analyses within EBASCO' s scope of work.
As a result of the discus-
- sions, four 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 pi-oblem 2538.
- 4)
The verification of the NUPIPE computer program (benchmark problems). These concerns are addressed in the following sections. 10~1 SUPPORT STIFFNESS Th~ original piping analysis of Surry Unit 2 did not consider the actual stiffness.of the supports. Representative support stiffness was considered during the current reanalysis. During the pipe support reanalysis effort it has been observed that cer-tain anchor type supports expose wide flange members to torsional moments. This type of loading condition results in a very flexible support. As a part of the pipe support reanalysis effort, anchors have been reviewed for this type of loading and members modified to resist torsion as required. 10.2 NUPIPE COMPUTER CODE At EBASCO' s
- request,. Nuclear Services Corporation (NSC) conducted a
thorough review of the NUPIPE program against the source codes, NSC has determinl!d that all values utilized by the program, but not spec-' ified by the user as input, are pertinent to the 1967 and earlier versions of the B31. l Power Piping Code. The code of record for Surry Unit 2 is B31.l 1955 with code class N-7. 10.~ 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 withfn the scope of the Show Cause OrdE!r. 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 SURRY POWER STATION - UNIT 2 I,n the EBASCO NUPIPE analysis of the portion of the LHSIS within pro-blem 2538, the system was not overstressed.
- However, the loads iden-ti~ied 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 b~en modified so that it will relieve the local overstress conditions. 10.4 BENCHMARK PROBLEMS EBASCO has performed four pipe stress to verity the NUPI?E 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 SURRY POWER STATION - UNIT 2 I I I .1 I I I I APPENDIX A SYSTEMS AFFECTED I I I I I I I I I A-1
I I I I I I I I I I I I I I I I I --- -----~------ ~~- ----------~- - - ~-~- ---- ---- --- ~-- SURRY POWER STATION - UNIT 2 lhe reanalysis included those safety related lines originally computer-an~lrzed with the SHOCK2 program. The systems line numbers, the associated cqmputer problem numbers, and the flow diagram numbers are listed below. The following table includes all seismically analyzed lines. The figure numbers re-fe+ to the FSAR drawings, and the Surry Unit 2, FM and FP drawings included in
- ,Appendix,B, s;rstem Low Head Safety Injection High Head S~fety Injection Line No.
8-SI-214-153 8-SI-292-153 .8-SI-214-153 8-SI-292-153 10-SI-284-152 10-SI-216-153 8-SI-292-153 10-SI-351-153 6-SI-249-1502 10-SI-349-153 8-SI-214-152 lO-SI-283-152 10-SI-213-153 6-SI-248-1502 10-SI-352-1502 10:-SI-350-153 10-SI-349-153 10-SI-348-153 8-SI-214-153 l i-SI-24 7-602 12-SI..,.247-1502 12-RC-324-1502 10-RH-117-1502 12-SI-246-602 12-SI-246-1502 12-RC-323-1502 10-RH-116-1502 6-SI-248.-1502 . 6-SI-249-1502 6-SI-250-1502 6-RC-321-1502 6-SI-343-1502 12-SI-245-602 12-SI-245-1502 12-RC-322-1502 10-SI-206-153 6-CH-372-152 4-CH-412-152 3-CH-373-152 8-SI-214-153 Responsi-bilities for Analysis E E E E S&W S&W S&W S&W S&W S&W S&W S&W S&W
- S&W S&W S&W S&W S&W S&W E
E E E E E E E E E E E E E E E Problem No. 2695 2697 2681 2682 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2727 2537 2537 2537 2537 2555 2555 2555 2555 2539 2539 2539 2539 2539 2709 2709 2709 E 2689 E 2735 E 2735 E 2735 E 2735 A-2 MKS No. 127Dl 127D2
- 127Kl 127K2 127Cl 127Cl 127Cl
.127Cl 127Cl 127Cl 127Cl 127C2 127C2 127C2 127C2 127C2 127C2 127C2 127C2 .122Al 122Al 122Al
- 122Al, 117Bl 122.Dl 122Dl 122Dl 122Dl 122Kl 122Kl 122Jl 122Jl 122Kl 122Ll 122Ll 122Ll 127Fl 127Gl 127Gl 127Gl l27G2 Flow Diagram No.
FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM:-106A FM-106A FM-106A FM-106A FM-106A FM-106A FM-lGoA FM-!Ot;J\\ FM-106A FM-106B FM-106B FM-106B FM-106B, 104A F:tv1-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106B FM-106A FM-105B FM-105B FM-105B 106A I i
-~~-- --------~-- I I SURRY POWER STATION - UNIT 2 Responsi-Flow bilities Problem MKS Diagram I System Line No. for Analysis No. No. No. High Head 8-CH-504-152 E 2735 127Gl, G2 105B I Safety 8-CH-317-152 E 2735 127Gl, G2 105B Injection 8-SI-217-152 E 2735 127Gl, G2 106A (Cont'd) 8-SI-292-153 E 2735 127Gl 106A I 6-SI-218-152 E 2735 127Gl, G2 105B, 106A 6-SI-219-152 E 2735 127Gl, G2 105B, 106A 6-SI-278-152 E 2735 127Gl, G2 105B, 106A 6-CH-501-152 E 2735 127Gl 105B I 6-CH-502-152 E 2735 127Gl 105B 6-CH-503-152 E 2735 127Gl 105B 8-CH-505-152 E 2735 127G2 105B I 8-CH-506-152 E 2735 127Gl, G2 105B 8-SI-207-152 E 2735 127G2 106A 8-SI;...302-152 E 2735 127G2 106A 8-SI-170-153 E 2735 127G2 106A I 8-SI-172-153 E 2735 127G2 106A 10-SI-206-153 E 2735 127G2 106A 6-CH-318-152 E 2735 127Gl 105B I 6-CH-319-152 E 2735 127Gl 105B Residual 14-RH-101-1502 E 2508B 117Al FM-104A I Heat Removal 14-RH-102-602 E 2508B 117Al FM-104A 1 10-RH-104-602 E 2508B 117Al FM-104A 10-RH-105-602 E 2508B 117Al FM-104A 12-RH-106-602 E 2508B 117Al FM-104A I 10-RH-107-602 E 2508B 117Al FM-104A 10-RH-108-602 E 2508B 117Al FM-104A 10-RH-109-602 E 2508A 117Al FM-104A I 10-RH-110-602 E 2508A 117Al FM-104A 12-RH-112-602 E 2508B 117A.l FM-104A 14-RH-118-602 E 2508B 117Al FM-104A 12-RH-119-602 E 2508A 117Al FM-104A I 12-RH-112-602 E 2540 117Bl FM-104A 3-RH-113-6 02 E 2540B 117Bl FM-104A I 1 4-RH-115-152 E 2540B 117Bl FM-104A *" I 10-RH-116-1502 E 2540 117Bl FM-104A 10-RH-117-1502 E 2540 117Bl FM-104A 6-RH-120-152 E 2540 117Bl FM-104A I 10-RH-137-602 E 2540 117Bl FM-I04A 6-RH-120-152 E 2554 117Cl FM-104A, 101A 1* I I I A-3
I I
- suRRY POWER STATION -
UNIT 2 Responsi-Flow I bi lit ies Problem MKS Diagram System Line No. for Analysis No. No. No. Main Steam 30-SHP-101-601 S&W 2577 1000 FM-14A 11 I 30-SHP-102-601 S&W 2588 1010 FM-14A 30-SHP-103-601 S&W 2579 1020 FM-14A .30-SHP-101-601 S&W 2346 103A FM-14A I 30-SHP-102-:-601 S&W 2346 103A FM-14A 30-SHP-103-601 S&W 2346 103A: FM-14A 30-SHP-124-601 S&W 2346 103A FM-14A 30-SHP-123-601 s&w 2346 103A FM-14A I 30-SHP-122-601 S&W
- 2346, 103A FM-14A Feed water 14-WFPD-117-601.
S&W 2569 100G FM-18A I . 14-WFPD-113-601 S&W 2573 101G FM-18A 14-WFPD-109-601 S&W 2571 102G FM-18A I . ~1.1,xiHary 6-WAPD-101-601 E 2473 118Al, A2 FM-18A, 18B Feed water 6-WAPD-102-602 E 2473 118A2 FM-18A, 18B 3-WAPD-109-601 E 2473 118Al, A2 FM-18A 3-WAPD-110-601 E 2473 118Al, A2 FM-18A I 3-WAPD-111-601 E 2473 118Al, A2 FM-18A 3-WAPD-112-601 E 2473 118Al, A2 FM-18A 3-WAPD-°l 13-601 E 2473 118Al, A2 FM-18A I 3-WAPD-114-601 E 2473 118Al, A2 FM-18A
- 6-WAPD-150-601 E
2473 118Al, -A2 FM-18A, 18B 6-WAPD-151,-601. E 2473 ll8Al, A2 FM-18A, 18B I 6-WAPD-101-601 E 2683 118G2 FM-18A, 18B 6-WAPD-102-601 E 2683*. 118Gl FM-18A, 18B 6-WAPD-103-601 E 2683 118Gl, G2 .FM-18A 6-WAPD-104-601 E 2683 118Gl, G2 FM-18A I 4-WAPD-105-601 E 2683 118Gl, G2 FM-18A 4-WAPD-106-601 E 2683 118Gl, G2 FM-18A 4-WAPD-107-601 E 2683 118Gl, G2 FM-18A I 4-WAPD-108-601 E 2683. 118Gl*, G2 FM.:..18A 6-WAPD-50-601 E 2683 118Gl, G2** FM-18A, 18B 6-WAPD-52-601 E 2683 118Gl, G2 FM-18A,18B I Service Water 24-WS-126-10 E 2465 119Al FM-21A 24-WS-128-10 E 2467 119A2 FM-21A .24-WS-130-10 E 2469 119A3 FM-21A I 24-WS-132-10 E 2471 119A4 FM-21A I I I I A-4
I I SURRY POWER STATION - UNIT 2 Responsi-Flow bi lities ProblE;!m MKS Diagram I ~ystem Line No. for Analysis Ne;>. No. No. Pressurizer 4-RC-334-1502 E 2000 124Al FM-103B I Safety and 3-RC-335-1502 E . 2000 124Al FM-103B Relief 3-RC..:.361-,-1502 E 2000 124Al FM-103B 6-RC-320-602 E 2000 124Al, A2 FM-103B -1 6-RC-362-602 E 2000 124Al, A2 FM-103B 12-RC-336-602 E 2000 124Al, A2 FM-103B 6-RC-337-1502 E 2000. . 124Al, A2 FM-103B 6-RC-338-1502 E 2000 124Al FM-103B I 6-RC-339-1502 E 2000 124Al FM-103B 6-RC-340-602 E 2000 124Al, A2 FM-103B 6-RC~341-602 E . 2000 124Al, A2 FM-103B I 6-RC-342-602 E 2000 124Al, A2 FM*:)03B ~ressurizer 4-RC-314-1502 E 2771 125Al FM-103B Spray 4-RC-315-1502 E 2771 125Al FM-103B I 2-CH-368-1502 E 2771 125Al FM7103B HP- ~team to 4..,.SHP-125-601 E 2862 131Al FM-14A I Auxiliary 3-SHP-132-601 E 2862 131Al FM-14A Feed water 3-SHP-128-601 E 2862 131Al FM-14A Pump 3-SHP-131-601 E 2862 131Al FM-14A I 3-SHP-157-601 E 2862 -*. 131Al FM-14A 4-SHP-126-601 E 2864 l~lBl FM-14A . 3-SHP-129-601 E 2864 131Bl FM-14A 4-SHP-127-601 E 2869. 131Cl FM-14A I 3-SHP-130-601 E 2869 131Cl FM-14A 3-SHP-135-601 E 2869
- 131Cl FM-14A I
Containment 10-CS-104-153 S&W 2521 123Al FM-lOlA and Recir-8-CS-123-153 S&W 2521 123Al FM-lOlA culation Spray I 10-CS-103-153 S&W . 2523 123Al FM-lOlA 11 8-cs...:122-153 S&W 2523 123Al FM-lO~A I 10-CS-103-153 S&W 2547 123Cl FM-lOlA 8-CS-133-153 S&W 2547 123Cl FM-lOlA 10-CS-104-153 S&W 2549 123C2 FM-lOlA-~ I 8-CS-134-153 S&W 2549 123C2 FM-lOlA 10-RS-112-153 E 2546 123Dl FM-lOlA I 8-RS-123-153 E 2546 123Dl FM-lOlA 10-RS-104-153 E 2541 123D2 FM-lOlA 8-RS-121-153 E 2541 123D2 FM-lOlA I 10-RS-103-153 E 2542 123D3 FM-lOlA 8-RS--120""" 153 E 2542 123D3 FM-101A I 10-RS-111-153 E 2543 123D4 FM-lOlA 8-RS-122-153 E 2543 123D4 FM-lOlA I I A-5
I I
- SURRY POWER STATION - UNIT 2 Responsi-Flow I
biHties Problem MKS Diagram System Line No. for Analys;i.s No. No. No. Containment 10-RS-112-153 E 2560 123El FM-101A I and Recir-10-RS-104-153 E 2561 123E2 FM-101A culation Spray 10-RS-110-153 E 2544 123Gl FM-lOlA (Cont'd) 10-RS-109-153 E 2533 123G2 FM-lOlA I 10-RS-103-153 E 2548 123Hl FM-101A 10-R$-lll-153 E 2545 123H2 FM-lOlA 8-CS-134-153 E 2744 123Jl FM-lOlA s.... cs-133-153 E 2745 123Kl FM-lOlA I 12-CS-102-153 E 2753 12311. FM-lOlA 12-CS-101-153 E 2754 123Ml FM-lOlA 10-RS-109-153 E 2751 123Nl FM-lOlA I 4-RS-114-153 E 2751 123Nl FM-lOlA 10-RS-110..-153 E 2752 123N2 FM-lOlA I 4-RS-115-153 E 2752 123N2 FM-101A 8-CS-133-153 E 2755 123Pl FM-lOlA 8-CS-134-153 E 2755 123Pl FM-lOlA I 4-CS-135-153 E 2755 123Pl FM-lOlA 4-CS-136-153 E 2755 123Pl FM-lOlA 4-CS-105-152 E 2755 123Pl FM-lOlA I 1/2-CS-108-153 E 2755 123Pl FM-lOlA 4-CS-106-152 E 2755 123Pl FM-101A 10-RS-101-153 E 2756 l23Ql FM-lOlA I 10..-R~-102-153 E 2757 123Q2 FM-lOlA Component 18-CC-15-121 E 2604 112AA1 FM-22A I Cooling 18-CC-9-121 E 2605 112AB1 FM-22A 18-CC-7-121 E 2601 112Sl FM-22A 18-CC-14-121 E 2603 112S2 FM-22A I Containment 8-CV..-108-151 S&W 2650 137A FM-102A Vacmnn .~ I I I I Note: E = EBASCO 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 APPENDIX B 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~ UNIT 2 APPENDIX B FLOW DIAGRAMS - IDENTIFICATION OF PROBLEMS REANALYZED Title Main Steam Feed water Cross-Connects for Auxiliary Feed Girculating and Service Water Component Cooling Containment and Recirculating Spray Containment Vacuum and Leakage Monitor Reactor Coolant Sheet 1 Reactor Coolant Sheet 2 Residual Heat Removal Chemical and Volume Control Sheet 2 Safety Injection Sheet 1 Safety Injection Sheet 2 Refueling Water S~qrage Tank Crosstie B-2 Drawing No. 11548-FM-14A 11548-FM-lSA 11448-FM-lSB ll 548-FM-21A
- 11548-FM-22A 11548-FM-lOlA 11548-FM-10 2A 11548-FM-l 03A
- 11548-FM-103B 11548-FM-~04A ll548-FM-105B 11548-FM-106A ll548-FM-106B 11448-FM-106C
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I I SURRY POWER STATION - UNIT 2 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 I C-1
11 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 swing check valves, sized 3 and 6 inches, are inst al led in the following seismic Category I piping syst*ems: a) Chemical and volume control system b) Safety injection systems A detailed listing by line number is contained in the following table. Iiines with 6 inch check valves were originally seismically analyzed by computer prograill-or hand calculations. The re-evaluation. of these systems using the correct valve weight is currently being done t,tnder the NOPIPE program. The results have shown that the pipe stress is within the allow-abJ.e for all lines. Lines with 3 inch check valves were analyzed originally by hand calcula-tions. An estimated weight, overly conservative, was used instead of actual valve weights. The inc~rrect valve weight has no effect on th_ese calculations and re-evaluation is not required, however, these valves and the related pipe lines are included in the scope of IE Bulletin 79-14. C-2
I I I I I I I I I 1* I I I I I I I I I SURRY POWER STATION - UNIT 2 LISTING OF VELAN SWING CHECK VALVES COVERED BY IE BULLETIN NO. 79-04 SAFETY INJECTION SYSTEMS - UNIT 2 6 Inch 3 Inch 2-SI-79 2-SI-82 2-SI-85 2-SI-88 2-SI-91 2-SI-94 . 2-SI-228 2-SI-229 2-SI-238 2-.SI-239 2-SI-240 2-SI-241 2-SI-242 2-SI-243 2-SI-224 2-SI-225 2-SI-226 2-SI-227 CHEMICAL AND VOLUME CONTROL SYSTEM - UNIT 2 3 Inch 2-CH-196 2-CH-258 2-CH-267 2-CH-276 2-CH-309 2-CH-312 C-3 6-RC-317-1502 6-RC-319-1502 6-RC-320-1502 6....;RC-318-1502 6-RC-3l6-1502 6-RC-321-1502 6-SI-249-1502 6-SI-249-1502 6-SI-248-1502 6-SI-249-1502 6-SI-250-1502 6-SI-345-1502 6-SI-344-1502 6-SI-353-1502 3-SI-346-1503 3-SI-270-1503 3-SI-347-1503 3-SI-272-1503 3-CH-500-1502 3-CH-381-1503 3-CH-302-1503 3-CH-303-1503 3-CH-379-1503 3-CH-301-1502
,I I SURRY POWER STATION - UNIT 2 , I I 'I I I I I APPENDIX D I CORRESPONDENCE WITH THE I NRG 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 I I I SURRY POWER STATION - UNIT 2 APPENDIX D CORRESPONDENCE WITH NRC 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 Addressee NRC TO VEPCO Denton Proffitt Stello Proffitt Stello
- Proffitt Stello Proffitt Eisenhut Proffitt O'Reilly Proffitt O'Reilly Proffitt Denton Proffitt O'Reilly Proffitt Murphy Proffitt Murphy Proffitt Murphy Proffitt Eisenhut Proffitt D-2 Letter No./Subject Show Cause Order Addendum to Show Cause Order Use of Soil Structure Interaction Techniques Request for Further SSI Information Factor Adjustment to SSI Calculated Stresses Information Pertaining to IE Bulletin No.
79-14, Revision 1 Letter of Guidance on IE Bulletin No. 79-14 Lifting of Suspension Required by the Order to Show Cause Confirmation of Concur-rence Refers to NRC Inspection. of Sept. 10-13 and Sept. 19-21, 1979 Refers to NRC Inspection of Sept. 13-14, 1979 Refers 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 1~ 19 20 21 22 23 24 25 26 27 Date 3/30/79 4/19/79 4/23/79 4/24/79 4/27/79 5/2/79 5/2/79 5/22/79 5/24/79 5/24/79. 6/5/79 6/8/79 6/8/79 6/12/79 SURRY POWER STATION - UNIT 2 APPENDIX D (Cont'd) CORRESPONDENCE WITH NRC Signature Addressee VEPCO TO NRC Spencer Denton/ Stallings Spencer Spencer Spencer Stallings Spencer Ragone Spencer Spencer Spencer Spencer Spencer Spencer D-3 Stello O'Reilly O'Reilly O'Reilly Denton/ Stello Denton Stello Hendrie Stello Stello Denton Denton Stello Denton Letter No./Subject 198/Initial Response to Show Cause Order 270/LER 79-010/013L-O 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 Reanalysis Effort 260/Submittal of SSI Information Comments on Moratorium/ Surry Reanalysis Response to NRC Letter of 4/2/79 Response to NRC Letter of 5/18/79 Submittal of Report on Reanalysis Additional Information, Report on Reanalysis of Piping Soil Structure Interac-tion Report Modification Informa-tion, Reanalysis of Piping Systems
I I SURRY POWER STATION - UNIT 2 APPENDIX D (Cont'd) I CORRESPONDENCE WITH NRC I Item No. Date Signature Addressee Letter No./Subject 28 6/15/79 Spencer Denton Schedule and Support I Information 29 6/19/79 Spencer Denton Support Modifications I 30 6/25/79 Spencer Denton Support Information, Reanalysis of Piping Systems I 31 8/1/79 Spencer Denton Submittal of Revised Report on Analysis I 32 8/21/79 Spencer Denton Analysis Completion of Designated Supports - I Outside Containment 33 8/31/79 Spencer Denton Reanalysis of Piping Systems I 34 8/31/79 Spencer O'Reilly 60-Day Response for IE 79-14 1 I 35 9/13/79 Spencer Eisenhut Response to NRC Letter of 5/25/79 I 36 10/3/79 Proffitt Denton Seismic Analysis of Piping Systems I 37 10/4/79 Spencer O'Reilly Response to NRC Letter of 7 /2/79 I 38 10/4/79 Spencer O'Reilly Response to IE Letter Dated 9/7 /79 39 I 10/15/79 Spencer O'Reilly Extension of IE Bulletin 79-14 Deadline 40 10/23/79 Spencer O'Reilly Extension of IE Bulletin I 79-14 Deadline, Unit 2 41 10/30/79 Spencer O'Reilly 120-Day Response to I IE 79-14 42 11/28/79 Spencer Denton Seismic Analysis of Piping Systems I I D-4
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I I . SURRY POWER STATION - UNIT 2 APPENDIX D (Cont'd) I CORRESPONDENCE WITH NRC I Item No, Date Signature Addressee Letter No,/Subject 43 12/7 /79 Spencer O'Reilly Response to NRC Letter I of 11/8/79 44 12/13/79 Spencer Denton Show Cause Order I Reanalysis 45 12/21/79 Spencer O'Reilly Show Cause 60 Days I Analysis Completion 46 2/1/80 Spencer Denton Show Cause Modification Schedule Revision I 47 2/22/80 Spencer Denton Start-up Request for Surry Unit 2 I 48 3/21/80 Spencer Denton Amended Start-up Request for Surry Unit 2 I 1 49 3/28/80 Spencer Denton Show Cause Report Errata for Surry Unit 1 I S&W to NRC so 3/22/79 Kennedy Denton Transmittal of S&W I Computer Programs 51 3/30/79 Jacobs Herring Submittal of Computer I Outputs 52 4/3/79 Jacobs* Bezler Submittal of Benchmark Problem to Brookhaven I National Laboratory 53 4/6/79 Kennedy Denton Transmittal of S&W I Computer Programs 54 4/6/79 Jacobs Stello Plan for Verification I of Dynamic Analysis Codes 55 4/11/79 Jacobs Bezler Submittal of Computer I Outputs 56 4/13/79
- Status of Jacobs Stello Update and I
Verification Plan for Dynamic Analysis Codes I D-5
I I SURRY POWER STATION - UNIT 2 I APPENDIX D (Cont'd) CORRESPONDENCE WITH NRC I Item No, Date Signature Addressee Lette~ No./Subject I 57 4/18/79 Jacobs Hartman Submittal of Computer Outputs 58 4/27 /79 Jacobs Bezler Submittal of Benchmark I Problems 59 4/27 /79 Jacobs Stello Status of Verification I Plan for Dynamic Analysis Codes I 60 5/8/79 Rossier Neighbors Draft Outline of SSI-ARS Report 61 5/9/79 Kennedy Stello Reference SHOCK 0 I Program 62 5/11/79 Kennedy Stello Reference SHOGK 0 I Program 1 63 5/14/79 Kennedy Denton Proprietary Computer Codes I 64 6/4/79 Jacobs Bezler Submittal of Benchmark Problems I 65 6/12/79 Jacobs Bezler Submittal of Benchmark Problems I 66 9/6/79 Allen Stello Response to NRC Letter of 8/10/79 I Ebasco to NRC 67 9/7 /79 Nelson Hartzman Benchmark Problem I VEP/NRC/001 68 1/3/90 Nelson Hartzman Benchmark Problem I VEP/NRC/002 I I I D-6}}