ML20027C601

Independent Seismic Design Verification:Turbine Driven Portion Emergency Feedwater Sys, Final Rept.One Oversize Drawing Encl.Aperture Card Is Available in PDR

ML20027C601
Person / Time
Site:	Summer
Issue date:	10/15/1982
From:	STONE & WEBSTER ENGINEERING CORP.
To:
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ML20027C600	List: ML20027C599 ML20027C601
References
NUDOCS 8210190500
Download: ML20027C601 (105)

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i... INDEFENDETT SEISMIC DESIGN VERIFICATION -? TURBINE DRIVEN PORTION EMERGENCY FEEDilATER SYSTEh e j V.C. SUMMER NUCLEAR STATION F FINAL REPORT: OCTOBER 15, 1982 3 .a t_

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? s* 3 ,.q S J.O. 14236 October 15, 1982 e .3 goton osoo - - +. .~ y n----- -.,._-_-.y --r y.

...a...- ...- ~.. b. TABLE OF CONTENTS

1.0 INTRODUCTION

e 1 1.1 General Scope 3 1.2 Stone & Webster Engineering Corporation Qualificatio'ns and Independence ,. 4 1.3 Evaluation Process J 2.0

SUMMARY

AND CONCLUSIONS

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2.1 Conclusions ~ 2.2 Field Verific'ation Summary 2.3 Analysis and Evaluation Summary --'j 2.4 Audit of Design Control Summary 3.0 FIELD WALKDOWN 3.1 Scope 3.2 Walkdown Procedures and Criteria il 3.3 Walkdown Results gj 3.4 Conclusions 4.0 STRESS ANALYSIS AND EVALUATION e.3 -] s d 4.1 Scope 4.2 As-Built Data 7f ll3 4.3 Stress Analysis Procedures i .2 4.4 Evaluation Criteria 4.5 Pipe Stress Review 9 4.6 Support Load Review ,3 4.7 Equipment Nozzle Load Review 4.8 Reactor Building Penetration Load Review n 4.9 Thermal Movement Review }] 4.10 Open Item Reports 4.11 Potential Discrepancie's 4.12 Final Analysis (} 4.13 Resolution of Open Item Reports and Potential Discrepancies l 4.14 Conclusions Q 5.0 AUDIT REPORT, DESIGN CONTROL PROGRAM ( ' ;;: 5.1 Purpose 5.2 Scope ~) 5.3 Approach 5.4 Evidence Examined l 5.5 Results ,. ) 5.6 Conclusions ..1 .e APPENDIX A: Status Report July 9, 1982 "A \\ !.3 APPENDIX B: SCE&G Responses to Draft Final Report 1 i

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n I 'd ~! LIST OF TABLES ^j 7, TABLE 1-1 Project Personnel 1-2 Project Procedures ..] 4-1 Maximum Piping Stresses from the Comparison Analysis 4-2 Support Load Review of Package 101 - Comparison Review ~ 4-3 Support Load Review of Package 102 - Comparison Review ,,] 4-4 Support Load Review of Package 103 - Comparison Review ..1 m LIST OF ATTACHMENTS Att. 1-1 Statement Regarding Potential or Apparent t Conflicts of Interest ^ 5-1 Audit Participants "I LIST OF FIGURES Fig. 3-1 Simplified Emergency Feedwater Flow Schematic (with SWEC Notations) 4-1 Worksketch for Package 101 4-2 Worksketch for Package 102 4-3 Worksketch for Package 103 4-4 Worksketch for Package 104 5-1 Simplified Emergency Feedwater Flow Schemhtic .,J 5-2 Design' Input Flow Chart 5-3 Quality Management Program Document Hierarchy I e. O .4 wm IT .ka E , 3 ( ii '1 J I ! s;.a ,~ww-.. _w ,,-v,, c

q SECTION 1.0 W INTRODUCTION 7 1.1 GENERAL SCOPE 1 Stone & Webster ' Engineering Corporation (SWEC) was engaged by South j Carolina Electric & Gas Company (SCE&G) to perform an independent review of the seismic design for the flow path of the Turbine Driven Pump of the Emergency Feedwater System to Steam Generator C at the V.C. Summer Nuclear Station, Unit 1. The scope was outlined in SCE&G-Procurement Specification DSP-544C and consisted of three major tasks, specifically: ..,c: Field Walkdown: Verification of the as-built piping configuration .1d Stress Analysis and Evaluation: Analysis of the as-built piping system, review of stresses, and comparison of support ? loads with results obtained by Teledyn:: Engineering Services

j (TES), which performed piping stress analysis of this system for Gilbert Associates Incorporated (GAI), the designer of y

V.C. Summer Nuclear Station, Unit 1 ~ Design Control Audit: Review .o f the design control procedures and implementation thereof by GAI. .'.2 J, This report presents SWEC's findings, conclusions and recommendations. An earlier status report dated July 9,' 1982 is enclosed as Appendix A.

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] 1.2 STONE & WEBSTER ENGINEERING CORPORATION QUALIFICATIONS AND INDEPENDENCE ) SWEC has extensive experience in the engineering, design, construction, and startup operations for nuclear power plant projects as well as special expertise involving seismic design analysis, field verification i efforts, and pipe stress and support reanalysis required by recent NRC I&E Bulletins. SWEC also has extensive experience in the quality + assurance aspects of the nuclear power industry and in auditing of g large, highly technical and complex projects. SWEC is justifiably proud of its record and large staff of capable and experienced personnel. h.

SWEC, its parent company Stone & Webster, Inc.,

its affiliated companies, and all personnel assigned to this evaluation are independent of South Carolina Electric & Gas Co. Work performed by Stone & Webster Engineering Corporation and its affiliated companies J for SCE&G represents only a minuscule portion of SWEC's business. Stone & Webster, Inc. and its subsidiaries have no holdings of SCE&G ] securities. The Employee Savings Plan of Stone & Webster, Incorporated i

j and participating subsidiaries is administered by the Chase Manhattan Bank, N.A. as trustee.

Funds may be invested in the Enployee Benefit 1-1 h.: ~ 17 .[ ~~ ' ~[

L. ..: ^ L q l.l 7 -Investment Funds, Equity Fund of the Chase Manhattan Bank which is a j comingled fund. SWEC exercises no direct control over the investment of such funds. El Table 1-1 lists personnel assigned to the various tasks. Dr. P. 4 Dunlop, Project Manager, had overall responsibility for the project. Dr. K. Y. Chu was Project Engineer responsible for the technical ~] evaluation (Tasks 1 and 2) and was independent of Mr. J. H. MacKinnon d who was respersible for auditing the GAI design control program (Task 3). All key technical personnel assigned to the proj ect signed ~! disclosure statements (Attachment 1-1). A ~ 1.3 EVALUATION PROCESS .q f All work was performed in accordance with project procedures (Table 1-2). Whenever a reviewer noticed anything outside the criteria, or had any question about the information or data, the reviewer identified 3 this. Specific procedures for identifying questions were different for di each of the three major tasks and are explained in the task-specific project procedures (Table 1-2). 33 1.3.1 Field Walkdown (as-built verification) All field measurements were recorded directly on the piping isometrics. "] Whenever the measured values differed from the isometric values by more than the criteria presented in VCS-1, Field Walkdown Procedure, the iscmetric values were circled on the isometrics and also recorded on .l Difference List-(DL) Forms. Copies of the marked-up isometrics and DL Forms were provided to SCE&G at the. end of the Field Verification Effort. Section 3.0 presents complete details of this effort. 4 1.3.2 Stress Analysis and Evaluation All analyses were performed in conformance with VCS-3, Analysis and 'i Evaluation Procedure, and VCS-4, Analysis and Evaluation Criteria. These provided the procedures and criteria 'for performing the piping reanalysis. Procedures for highlighting differences are defined in ~. Procedure VCS-3. Questions raised by the stress analyst were formally J recorded and resolved. A two-step procedure was used. An Open Item Report (OIR) was initiated for all items requiring clarification or ~' confirmation. The OIRs were formally transmitted to SCE&G for their ,.j review and evaluation. If a satisfactory resolution was received, the OIR was formally closed out. If a possible error or incons2.stency was g confinned, a Potential Discrepancy (PD) was written. Section 4.0 ,j presents complete details of this effort. 1.3.3 Design Control Audit i :3 ' a Of the three tasks, the procedures and resolution of items for this task were more subjective. The personnel assigned to this effort were ] experienced certified auditors who performed the audit in conformance .J with general SWEC standards for such audits. Section 5.0 presents the complete audit report for this effort. ! 1 ) 1-2 l -. ~

..--..-.~ ~.....,...... TABLE 1-1 .a PROJECT PERSONNEL a Project Manager: Peter Dunlop y 4 Project Ecgineer: K. Y. Chu ..i. Assistant Project Engineer: J. F. Pam '.1 TASK 1 - FIELD WALKDOWN N. Roth (Lead Field Verification Engineer) x K. Anderson J. Y. Chen "4 D. Loffa A. Moss L. Peterson V. Saleta e i TASK 2 - STRESS ANALYSIS AND EVALUATION 4 T. Wei (Lead Engineering Mechanics Engineer) J. Y. Chen 1 J. Chiang _a Y. Chin J. Chu .j D. Loffa ..) Design Control Audit Manager: J. H. MacKinnon 6 TASK 3 - DESIGN CONTROL AUDIT l' D. Malone R. Twigg . 'da E. i i .a n .:n 1-3 7

n 4 .3 I TABLE 1-2 6 PROJECT PROCEDURES .] G TASK-SPECIFIC PROCEDURES Field Walkdown Effort ~[ VCS-1 Field Walkdown Procedure L. 4 Stress Analysis and Evaluation e-J VCS-3 Analysis and Evaluation Procedure ~ VCS-4 Analysis and Evaluation Criteria Design Control Audit Design Control Verification Plan i: PROJECT GE.YERIC PLANS /PROCEDLTES Quality Assurance Plan ~ Document Control Procedure - VCS-2 Quality Assurance Records Procedure - VCS-5 Engineering Assurance Audit Program E. A. Review Plan 1720 - Independent Seismic-Design Verification - Field Walkdown Effort w.S [ 2 a l 5 m e 6.$ k 7 =e* 9 .i 1-4 i .2

s "J + i ATTACHMENT l-1 J.O. 14236 . )4 INDEPENDENT SEISMIC DESIGN VERIFICATION 3 V.C. SUMMER NUCLEAR STATION, UNIT NO. 1 i SOUTH CAROLINA ELECTRIC & GAS CO. f J _J Statement Regarding Potential or Apparent,. Conflicts of In erest To: Stone & Webster Engineering Corporation 0*' Whereas, the undersigne4 employes (" Employee") understands that he or she is assigned as a participant to provide services to South Carolina Electric & Gas Company with respect to the Design Verification Program .eJ for the V.C. Summer, Nuclear Station; and Whereas Employee understands that it is necessary that the participants 4 1 r gj be screened for any potential,or apparent conflicts of interest with respect to this assignment; '9i Therefore, for the above stated purposes Employee makes the following representations 4 o Stone & Webster Engineering Corporation: t m i 1. Employee has not engaged in any work or business involved with or related to the engineering or design of the V.C. Summer Nuclear e Station other than this Design Verification Program; 71 ] 2'. 'Neither Employee, nor Guy members of his or her immediate fusily, own any beneficial interest in the South Carolina Electric'& Gas Company, including but not limited to common or preferred stock. 1 bonds or other securities issued on behalf of the South Carolina Electric & Gas Company; and ~ M'j 3. None of the members of Employee's immediate family are enployed b by South Carolina Electric & Gas Company i This statement is based upon the Employee's best information and belief ~' ..J and any exceptions to the rupresentations contained herein have been i described on the reverse side of this document. n

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.:) 7 SECTION 2.0 J

SUMMARY

AND CONCLUSIONS i

2.1 CONCLUSION

S 2.1.1 General Cocclusions j..J . The. obj ective of this project was to draw a conclusion regarding

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the adequacy of the seismic design of the V.C. Summer Nuclear Station .i Unit.I piping systems. The technique agreed upon was to perform an indepth evaluation of 'one representative subsystem which incorporated 4 , ~; major analyses techniques, vendor and subcontractor interfaces, and practices upon whien seismic design of the plant was based. The ,? l subsystem selected by SCE&G, with the concurrence of the NRC, was the flw path of the turbine driven portion of the Emergency Feedwater g ..J Syrten to Steam Generator C. Based on results of the specified technical review, the audit of design !1 conirol procedures, and actior.s taken by SCE&G, it is SWEC's opinion I J that the piping seismic design of the V.C. Summer Nuclear Station meets the d'esign criteria and is adequate to withstand the specified seismic event's. 2.1.2 Specific Conclusions The 'following are the specific conclusions 'for each of the three tasks in this independent review of the seismic design of the turbine driven portion of the Emergency Feedwater System to Steani Generator C at the 1. V.C. Summer Nuclear Station Unit 1. Areas of generic concern are j identified in each task conclusion along with reference to SCE&G action l regarding these concerns .a ,.j Field Walkdown - In general, the field walkdown verified that the as-auilt condition of the piping subsystem reflected the design layout as presented on the isonitric drawings. Where differences exceeded the i stringent criteria they did not affect the stress results 2 significantly. Stress Analysis and Evaluation - The as-built piping was found to be d within code-allowable stresses throughout.

However, numerous j

• differences in support loads were discovered.

These were due to three 7 causes: failure to include Diesel Generator Building se' a'-C-response j' spectra and movements in one portion of the pip m

Asystem, misorientation and mislocation of impingement jets, atQ aodeling 3

differences. Errors were subsequently corrected and are reported a

herein, i.a i

Because of the significance of the omission of the seismic effects of ] the Diesel Generator Building and the finding in the design control J audit related to response spectra, it is recommended that piping systems be reviewed to ensure that all appropriate response spectra and j] seismic anchor movements had been incorporated into the analysis of the ! j 2-1 lJ .7....

...... a : 1 1 ] as-built piping systems. SCE&G had undertaken action on this generic

.J concern and has completed the review.(Appendix B).

This generic concern is therefore resolved.

• m Ll Because of the inconsistencies in the jets, their orientation, 2

location, and combination with other loads, it is recommended that these items be carefully reviewed. It is further recommended that GAI l specification 1902 be updated to clearly reflect the design criteria .d applicable to jet impingement. SCE&G had undertaken action on this generic concern and has completed the review and specification updating ] (Appendix B). This generic concern is therefore resolved. Several differences in

• piping modeling were identified'. These related to stiffnesses of skewed supports which were less than stiffnesses of 4

global supports, location of mass points, and flexibility of an elbow. SWEC performed additional review of the TES computer input and piping i modeling to evaluate the impact of these differences. It was found that the modeling assumptions were reasonable and that the results were. J acceptable. Therefore, there is no generic concern regarding modeling. 7 Design Control Audit - This task had three parts: review of the GAI J design control program, verification of program application, and confirmation of consistent utilization of respon:;e spectra. SWEC's conclusions are: l GAI had an adequate design control program meeting the requirements of 10CFR50 Appendix B relative to the specific, areas investigated in this. seismic design verification i program. Most of the control methods rcviewed were generic and apply to all phases of design. .y The implementatiori of the program was adequate, except there y were cases of inconsistencies in the utilization of design inputs. These appear to be due to documentation problems. The complete audit results are presented in Section 5.0, with a recommendation to determine the extent' of incomplete documentation and provide compilation of all appropriate design criteria so that a clear and traceable record is developed and maintained. SCE&G is undertaking actions to ensure complete documentation and to provide clear 'a} identification of design criteria (Appendix B). These actions should resolve the problems and this generic concern is therefore considered resolved. The audit showed that response spectra were consistently labeled throughout the design process. 2.2 FIEI.D VERIFICATION

SUMMARY

In general, the field walkdown verified that the as-built condition of ] this piping system reflected the design layout as presented on the isometric drawings. The following is a brief d'escription of all the differences identified: 7, J 2-2 .h ,e r,

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.x 7 Gaps between piping-and support steel larger than criteria - two j occurrences. The largest of these was 9/32 inch whereas the criteria allowed only 5/32 inch. No major effects from these were ebserved in the subsequent stress analysis. SCE&G subsequently perfot$med a review m { of all similar supports (Appendix B). j Clearances between piping and structural components - three occur-c, 3 rences. Two of these were small clearances between pipe and structural component (0 and 7/64 inch) The 0-inch clearance was found at N penetration P-IB-1-041 which had not yet been reviewed under SCE&G's 3 sleeve clearance program when the field work was performed. This .d penetration was subsequently addressed by SCE&G's penetration program and is now acceptable '(Appendix B). The 7/64-inch clearance was found a to be more than adequate based on the results of SWEC's subsequent stress analysis. The third occurrence was a sleeve through a wall 1 which was found to be partially grouted. This was subsequently determined to have been identified by SCE&G (ECN 2316) and the grout ], had been removed when SWEC field personnel again visited the site on June 7, 1982. P Struts at angles other than identified on the isometrics - three occurrences of struts more. than 3 degrees from the values on the isometrics. The maximum difference was 11 degrees. No effects from .q these were observed in the stress analysis. SCE&G subsequently l gt performed a review of all similar skewed supports (Appendix B). Dimensional data outside the criteria specified for SWEC's field l p walkdown effort - 15 occurrences. The maximum difference.was 5.3 d inches for a span 6f 11.6 feet. .All dimensional differences were within SWEC's standard criteria. No. effects from these were observed .Q in the stress analysis. tJ Drafting errors - five occurrences. These were confirmed by reviewing q the support or piping drawings. All differences were noted and included (except the difference at penetration P-IB-1-041 which as indicated above is now acceptable) in ,j the subsequent stress analyses. No significant impacts of these differences were observed in the stress analysis. ~ J 2.3 ANALYSIS AND EVALUATION

SUMMARY

All piping stresses were found to be within code allowables and all ] thermal movements were within the criteria. Review of support, anchor, j penetration, and nozzle loads showed a number of cases where SWEC loads exceeded GAI loads. These were reviewed with GAI and TES and found to be due to three main causes: 2 l (1) Seismic effects from the Diesel Generator Building were not included.in the original GAI design analysis for subsystem fi-01 ] (SWEC stress package 101). J (2) Several jets were misoriented or mialocated in the original ~ j analyses. 2-3 -?) di _-------,g y wm -p ,p r

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~1, W (3) Differences in the capabilities of the two programs used and .,j differences in modeling of stiffnesses, lumped mass locations, geometrical differences, 'and engineering judgments. n + Causes 1 and 2 as well as some geometrical differences in cause 3 were subsequently corrected in the design analyses. SCE&G also undertook action on generic concerns (causes 1 and 2) and has identified their i findings and actions (Appendix B). SWEC confirmed that the other differences were insignificant. r 7 Overall, the analysis of the piping systems is judged adequate based on 'j our review and the actions undertaken by SCE&G. 2.4 AjiDIT OF DESIGN CONTROL SLMARY 7'.1 The three parts to this task were: 7 Review of the GAI design control program a Verification of design control program application ]j Confirmation that the structural dynamic analysis output was consistent with response spectra provided to TES for analysis y of the turbine driven portion of the Emergency Feedwater System. The following are SWEC's conclusions based on the design control audit. .IJ Procedural Program I"! An adequate Design Control Program, meeting the requirements of 10CFR50 j Appendix B, was in place for the transmittal and utilization of input data for pipe stress analyses of subsystems EF-01, 02, 03, and 22 of y the Emergency Feedwater Piping System (GAI and TES subsystem j numbering). Only one instance was observed in the existing program where there was i no formally approved procedure for the maintenance and distribution .d of a mechanical specification index. This was performed using an updated, uncontrolled instruction with no evidence that the instruction f had been approved. Although unapproved, the procedure was adequate and .j was being implemented. Program Implementation The procedures associated with the activities reviewed during the audit were adequately implemented except that the utilization of inpute to ,j pipe stress analysis in some cases was not consistent with program d requirements. The instances found in the audit are apparently documentation problems that would not-affect the design. One case ?'_'s) affecting the design was subsequently found during the Stress Analysis and Evaluation Task (Sect' ion 2.3(1)). The following were found during the audit: m 1 ~ 2-4 1 a

^ -^- ' ~ .r.-- -- -~ 9U The pipe stress analysis

• package for subsystem EF-01 did not utilize Figure 64 response spectra as_ specified on the isometric.

Although GAI had approved the deletion of Figure 64 in a request for information (RFI), there was no evidence i ~ j that the isometric had been marked-up to indicate that Figure 64 should be deleted nor was there documentation in the pipe 4 n stress analysis package that justified the deletion of Figure 3 64 (such as by reference to the GAI-approved RFI). There was no documentation in the pipe stress analysis- '] package for EF-22 that the differences between the thermal i movements utilized in the analysis and the movements on the isometric ha'd been evaluated. A letter to GAI from TES initiated, as a result of this audit indicated that.the ,5 differences had been evaluated when the analysis was performed and that reanalysis was not necessary. '1 (The project scope was expanded to include SI-09 because of e the difference noted in EF-22 above). The pipe stress 4 j analysis package for subsystem SI-09 apparently utilized ~. anchor movement information from a Westinghouse letter rather than the movements identified on the isometric. There was no evidence that GAI had approved or transmitted this m information for use. In addition, the pipe stress analysis package did not indicate that the movements utilized were different than the isometric and the reasons for the differences. A letter submitted by TES to GAI as a result of 3 the audit indicated that the Westinghouse anchor movement a information had.been used in the analysis. ~~ The nozzle loadings in pipe - stress analysis packages were ? noted as acceptable by " trade-off." There was no documentation in the pipe stress analysis packages that n identified the method or the acceptability of the method. j .j There were approved RFIs in GAI files that addressed load ~ I trade-offs, but they were not referred to in the packages. SCE&G is undertakirg corrective action regarding documentation (Appendix B.) These actions should resolve the problems and SWEC's ~ Concerus. - i Another area that was not clearly documented was the application of damping factors. Although the application of damping factors complied 7 with the FSAR, this could not ha discerned unless reference was made 4 collectively to the FSAR, Specification 702, pipe stress analysis packages, a GAI study, and minutes of a meeting. SCE&G is undertaking actions to ensure that the appropriate design criteria are clearly defined (Appendix B). These actions should resolve SWEC's concerns. g Response Spectra Consistency

A

' J The response spectra utilized in the pipe stress analysis were found to be consistent with the dynamic (structural) analysis output. In some E ! :A 2-5 t 1 2 l

.L cases additional spectra were utilized when it did not appear ,3 necessary. The stress analysis and evaluation portion of the seismic design verification effort subsequently identified one case in which ~ response spectra had been omitted from the analysis. SCE&G has hadertaken a complete review of response spectra to ensure correct i" application (Appendix B). Recommendations Procedures ~2 A procedure governing the preparation and distribution e7 a specification index for mechanical specifications (and for other discipline specifications if necessary) should be formalized as part of the project program. SCE&G has addressed this finding (Appendix B). SWEC considers this resolved. O Implementation J The extent of incomplete documentation in pipe stress analysis packages ~) should be determined and appropriate corrective action implemented. .1 J To preclude future misunderstanding and provide clear traceability regarding application of damping

factors, corrective action, in f.'.

the form of either a revision to Specification

702, or a

memorandum of explanation in the pipe stress analysis packages, or other appropriate equivalent, should be performed. .gM Both of these items appear resolved based on SCE&G response (Appendix B). -n c.; 7 .,.j 2J ~

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i ~ SECTION 3.0 FIELD WALKDOWN 4 ~ 3.1 SCOPE The as-built piping geometry of the Emergency Feedwater (EF) System for U the flow path of the turbine driven EF pump to Steam Generator C, shown on Figure 3-1, was determined. The piping walkdown included ~~I identification of valve locations and orientation, support location, orientation, function,,and other dimensions as necessary for the stress analysis. The walkdown continued beyond the identified flow path to equipment nozzles, terminal anchors, or a series of constraints remote from the flow path for the purpose of terminating the mathematical model of the subsystem at a point where the boundary condition would have no practical effect on the structural response of this subsystem. 3.2 WALKDOWN PROCEDURES AND CRITERIA ] 3.2.1 Walkdown Procedures Prior to commencing any field work, a project procedure, VCS-1, Field r. Walkdown Procedure, was prepared. A copy was submitted to SCE&G. This L procedure provided all necessary steps, documentation, and criteria required to proceed with the work in an orderly, consistent, and efficient manner. J 3.2.1.1 Measuring Devices l ~ The following devices were used for field measurements: i i 12-foot Engineers measurement tape, Lufkin, Ultralok, W312D ~i 6-foot folding ruler, Lufkin Rugged Red End Engineers 6-inch stainless-steel rulers, General Hardware Manufacturing Co., Inc., Nos. 300 and 616 Protractor, General Hardware Manufacturing Co., Inc., No. 18 ,,d Universal Protractor, by Sears Craftsman l 3 Feeler gauge, Starrett EDP 51170, Engineers gauge No. 245. l ..a: 3.2.1.2 Document Provided by SCE&G All field measurements and observations were recorded on the following 1 six GAI piping isometric drawings. l j_l C-314-081, Sheet 27, Rev. 3 I C-314-085, Sheet 1, Rev. 2 l 1 c. l 3-1 .. ~...... . ~.. .~

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• • 8 C-314-085, Sheet 2, Rev. 2 q

y C-314-085, Sheet 3, Rev. 3-i r l C-314-085, Sheet 4, Rev. 2 C-314-085, Sheet 5, Rev. 2 Other drawings, such as concrete outline drawings, turbine driven EF pump drawing, Reactor Building penetration No. 213 drawing, and EF piping drawings, were also provided to SWEC-for the purpose of 4 y providing orientation and dimensions which could not be measured. .n 3.2.1.3 Survey Teams

1. 2 A Lead Field Verification Engineer was assigned responsibility for this task.

He supervised three two-man survey teams. Each team was 7 assigned a portion of the subsystem to survey. 3.2.1.4 Reference Points t ] There were basically three reference points used for this survey; the turbine driven EF pump, the Reactor Building penetration No. 213, and i the steam generator no::le. The locations and elevations of these t ') three reference points were taken from the construction and fabrication drawings. fl 3.2.1.5 Survey and Documentation i d i The piping geometry was measured in segments to identify all locations of pipe supports, valves, flanges, tees, elbows, reducers, branch connections, penetrations, and orifices. These measurements were ~ I recorded directly on the isometric drawings. l Pipe clearances at penetrations and pipe supports were also measured i 2 and recorded. Pipe support types (functions) and orientations were verified and noted on the drawings. Orientations and lengths of valve operators were also measured. Also, all valve numbers were checked. 4 J Any dimensions - found outside the tolerance criteria in Section 3.2.2 g were circled on the isometric drawings and recorded on a Difference - 3 List (DL forms). All dimensions verified as being within the tolerance; 1 criteria were noted with a check mark (V ) on the isometric drawings. 3 The DL forms and the -isometric drawings were reviewed and approved by the Lead Field Verification Engineer after they had been completed, signed, and dated by the two survey team members. ] 3.2.2 Tolerance Criteria m ~ In order to compare the accuracy of the dimensions on the isometric ,7 drawings with SWEC's measurement, a set of tolerance criteria was id established based on SCE&G's !fF-14 Walkdown Procedure. All of SWEC's dimensions deviating from the dimensions on the isometric drawings by 7 .s ~U 3-2 .y 2..-- z = .~

m. .a. .T.~.. ...s...._...... =_ ~ ~.. -_ b ] more than the values listed below were entered onto a Difference List n (DL Form). 0.50 inch between an anchor or nozzle and the closest support a 2.0 inches between two adjacent supports m

J.

2.0 inches for segment length up to 6 inches a 3.0 inches for segment length greater than 6 inches and up to ~"" 24 inches 4 V 6.0 inches fbr segment length greater than 24 inches and up g to 60 inches 1: ~ 10.0 percent for segment length greater than 60 inches 3 degrees for angle measurements u 5/32 inch for total gaps between piping and support steel P. 3.3 WALKDOWN RESULTS r4 Approximately 800 feet of piping and 116 supports were field-walked. h The geometrical data, orientations, and functions of pipe supports on the six isometric drawings were generally accurate. There were some minor differences found waich exceeded-the stringent tolerance h criteria. 'These differences are summarized as,follows: 2.- 3.3.1 Difference in Span Lengths Between an Anchor / Nozzle and the f ',' Closest Support l Y l There were eight occurrences of differences exceeding the 0.5-inch e criterion. The maximum difference was 2.5 inches in a length of 5.5 ]"q feet, which represents a difference of only 3.8 percent. The complete list of these differences is as follows: 1 Difference (in.) Span length (ft) Percentage (%) .) i 2.50 5.45 3.80 l 9 1.39 3.93 2.95 l d 0.82 2.60 2.62 1.45 5.37 2.25 m 0.65 4.40 1.23 A 1.80 16.40 0.90 0.84 9.20 0.76 0.91 11.80 0.64 q Li J 3.3.2 Differences in Span Lengths Between Two Adjacent Supports 7 Thets were six. occurrences of differences exceeding the 2.0-inch 'i m ..,I 3-3 \\ 2

... -.~ - ~. criterion as listed below: Difference (in.) Span Length (ft) Percentage (%) 4.43 5.12 7.21 3.94 6.22 5.27 5.29 11.60 3.80 3.43 9.62 2.97 2.95 8.34 2.95 2.65 18.95 1.17 3.3.3 Differences in_ Segment lengths 'O 4 Only one difference exceeded the criterion. This difference was 3.4 inches, which was measured from the center line of a support to the 'd nearest elbow. This difference was a portion of the total cumulative difference between two adjacent supports, which was listed in Section "i 3.3.2. .s 3.3.4 Support Orientations W.y rf Three occurrences had angles different from the values on the isometric ~~ drawings by more than 3 degrees. These were 11, 11, and 9 degrees for sup orts EFH-4024, 4028, and 103, respectively. '2 3.3.5 Gaps Setween Piping and Support Steel Larger Than Criteria fl, Two lateral supports (EFH-080 and - EFH-099) had gaps between piping and support steel exceeding 5/32 inch. They were 9/32 and 1/4 inch. l-3.3.6 Clearance to Allow Piping Movements Clearances were observed and measured to ensure that piping movements eq as designed were not obstructed by rigid elements, such as other pipes, support steel, penetration slee?as, or sealants. Three instances were "3 . identified as follows: The lateral clearance of the vertical support EFH-4020 was only 7/64 inch. This is rather small. However, SWEC's pipe c.- stress analysis verified that the total pipe movement due to 3 thermal and seismic effects would amount to only 1/32 inch. Thus, this clearance is acceptable. Although penetration P-IB-1-041 was at least 2 inches larger g, in diameter than the pipe, there was hardly any clearance at one point. As indicated by SCE&G, the penetration program to identify sleeve clearance and QC surveillance requirements had not been completed at the time of this field measurement. This was subsequently addressed and resolved by SCE&G (Appendix B). 'M j Fenetration P-AB-4-049 was partially grcuted for a depth of about 1 1/2 inches. This was subsequently determined to have 3 r c 3-4 m --wm...

9 been identified by SCE&G (ECN 2316) and the grout had been removed when SWEC fieldwalk personnel again visited the site on June 7, 1982. ~ 3.3.7 Drafting Errors On Isometric Drawings ~' The. functions of four supports . ere labeled incorrectly. This w conclusion was confirmed by the agreement between the as-built supports J and the original pipe support drawings. Also, one elbow radius was written as SR, which should have read 0.5R, a standard long radius elbow for the 4-inch pipe. ~ 3.4 CONCIUSIONS 9 All data verified by the field walkdown were recorded on the isometric ~' drawings and used for the independent seismic design verification analysis. All differences exceeding the tolerance criteria were addressed. Although the tolerance criteria were very stringent, only a few differences were found and most of these were minor. No significant impacts of these were observed in the stress analysis. 9 Overall, the fieldwalk verified that the as-built condition of this piping subsystem reflected the design layout as presented on the isometric drawings. .4 D*% e. 4 L. ~. 8 4 3 $5d d2 3 4 .a 3-5 m u

l Icz..:, L.J L.El LJ E.53 G.M L. /.J !!. 3 E.i3 L I.' EJ l'.J. J L.) h.'.2 j Li.t) L..', J L.J L _..] (2J i t i 1 I e gf F bl* l~s t [ (xt-l*1- -ix1- - - r Do -.- r - M ^"c" a C A'*tfTM/.t a g g - -['7)- -{>()' - 3 l ( T8 siAtout soH p a-.- (l r g FROM CONDENSA1E g % AGE 183 I 3 f SloRM~>C 1AHK l' - f EFH nt3 8 ANCHOR i I .i ~g ORE 55E R g [_ g COOPLlHG l -t>1 D4-f (~ ) t-f EFH-It5 g rygt Asg gog

s. r AucuoR ;

I, MOIO" D#NEN Pus Act le2 8 l PUMP A 3 -+ u 3.- g,' FAGME lOI (hell (DDE D trH-sie l - (trH in b ~~ ~ AHCHOR n q $3g Au .gy)- _{x).. abN I -b.i 0 O' i 68 tat R AloR C ruusu j- - * = O FMOM D1-l----D4" -IXI-l/I d l'!-D4 (g"3$g j -ruaAGE sol rus AGE ses pg rg .E_5fst ses g)4 g_._ MOTOR DalVEN { g-g 4_g )_ h _.3 Pump a Gu aT AtuMENI E s'g ut.in An tou ~ u a Q ~ y -EFH-ilt " r:3 gy}- pJ ).-{>() ' AHctioR 1..,,, D4 s s .m "I i t ti etz 3 AnciioR runst.,som t -id-D<1 ( FIGilHE 3-1 l'A(#A6e 808 S1HPI.IFIEI) EHEHGENCY FICEIMATER FIJW SCllEHATIC TunusuE Daivtu (with SWEC Notations) Puue } Flowpatie of Turbinie-1) riven EF Puenp g._g[ a to Stease (;enerator C t

.<.a. 9 i c, SECTION 4.0 STRESS ANALYSIS AND EVALUATION ~ 1 4 4.1 SCOPE l The piping along the flow path of the Turbine Driven Emergency J Feedwater Pump to Steam Generator C was independently analyzed and results evaluated. This verificatica analysis was based on SWEC's field walkdown data and design criteria provided to SWEC by SCE&G and GAI. This task did not include a review of original licensing commitments nor of construction quality assurance. The scope of the evaluation included comparison of pipe stress with allowables, load comparison of pipe supports and anchors with design loads, and load comparison of equipment nozzles and Reactor Building penetration with allowable loads as provided by various design documents. Individual load cases were dead load, design pressure, thermal, seismic, and jet impingement loads. s p 4.2 AS-BUILT DATA J SWEC's field walkdown data as recorded on the GAI isometric drawings were reviewed by the stress analysts. This review identified a need for additional information and clarification. The stress analysts issued a Field Information Request (FIR) for each item to be verified. A field verification team was then assigned to make an additional ,1i survey in. order to respond to the FIRS. The FIR responses were incorporated in the stress analysis. 4.3 STRESS ANALYSIS FROCEDURES {2 i Two project procedures, VCS-3, Analysis and Evaluation Procedure, and VCS-4, Analysis and Evaluation Criteria, were developed to provide 1 design input information, load combinations, reference documents, guidelines for calculation preparation, evaluation criteria, and other documentation and procedural requirements in order to ensure a uniform analysis approach. Initially, all requests for design criteria from SWEC were addressed to T SCE&G. SCE&G either responded directly or requested GAI to providr b information to SWEC. Occasionally, telecopy or phone calls were l utilized to expedite the effort. All telecopies were filed on the I g) project and all phone calls transmitting data or decisions were g recorded and filed. On July 7, 1982, a meeting among GAI, TES, and SWEC was held at the TES office in Waltham, Ma. The purpose of the meeting was to ensure that 2 SWEC understood the design criteria and their application to the analysis. All items discussed were documented officially as meeting G notes, copies of which were distributed to SCE&G, GAI, and TES. Based d on the information provided in this meeting and other criteria provided previously, the " Initial Analysis" was performed. Results of this Q J 4-1 .5 0 ~

~^ ~ T ^. ~: - ~ .$d 7, analysis were reviewed and compared with the GAI design loads. 3 Differences between SWEC's and GAI's support loads, penetration and equipment nozzle allowables were identified as Open Item Reports q (OIRs). Copies were sent to SCE&G and GAI for review and ti clarification. at SCE&G called a meeting with GAI, TES, and SWEC on July 28, 1982 at GAI ] offices in Reading, Pa. The purpose of the meeting was to review the - W OIRs issued as a result of the Initial Analysis. GAI advised that some of the design criteria provided to SWEC were erroneous. Therefore, all ] the OIRs were voided and a new analysis was required. The new dj corrected data and criteria were documented in a G,AI letter addressed to SCE&G. The new? analysis, called " Comparison Analysis," was performed in order to incorporate the new information. Section 4.4, 9i Evaluation Criteria, identifies all applicable criteria for this Comparison Analysis. -j' Review of pipe stress, support load, equipment nozzle load, penetration 4 load, and thermal movement is presented in Sections 4.5 to 4.9 based on the result of the " Comparison Analysis." Open Item Reports were 3 written to document differences between SWEC's results and GAI's design M loads or allowables. Copies of OIRs were sent to SCE&G, GAI, and TES for review and clarification. e 4 m I! On August 13, 1982, a meeting among SCE&G, GAI, TES, and SWEC was held. at the TES office in Waltham, Ma. for the purpose of reviewing the OIRs . issued as a' result of the Comparison Analysis. GAI advised in the h. meeting that some of the. jet - forces.were redefined.and TES was in the {/4 process of revising their analysis to include these revised jet forces and the effects of seismic response spectra and movement from the N Diesel Generator Building which were omitted originally. After receipt ! b of the official transmittals from GAI documenting the revised jet forces and new support and nozzle load summary sheets, SWEC performed a m, " Final Analysis," which is presented in Section 4.12. 4.3.1 Desian Documents Provided by SCE&G and GAI Reference documents from SCE&G or GAI are listed as follows: 1. Design Specification, DSP-544C-044461-000, " Emergency Feedwater 'r'i System Piping and Pipe Supports," Rev. 5, 4-30-82, V.C. Summer d Nuclear Station, Unit No. 1. y1 2. Pipe Line Specifications for Nuclear Safety Class

Piping, y

SP-545-044461-000, Rev. 7, 11-25-80, V.C. Summer Nuclear Station, Unit No. 1. ~j 3. Pipe Line Specifications for Conventional Piping, SP-337-4461-00, l Rev. 8, 9-29-77, V.C. Summer Nuclear Station, Unit No. 1. 7il l3 4-2 =.::- = - = - - -. z

=

4 3 Design Specification for Reactc,r Building Piping Penetrations, g ASME B&PV

Code, Section

III, Division 1,

Class 2, DSP-606-044461-000, Rev. 9, 2-1-82, V.C. Summer Nuclear Station, 3 Unit'No. 1. ,J 5. Specification, Seismic

Analysis, Testing and Documentation, SP-702-4461-00, Rev. 4, 2-11-77, V.C. Summer Nuclear Station, Unit

.j No. 1. J 6. Design Specification, Motor Driven Emergency Feedwater Pumps, ASME ',j III, Class 3, DSP-508A-4461-00, Rev. 2, 7-8-77, V.C. Summer ~ J Nuclear Station, Unit No. 1. 3 7. Design Specification, Turbine Driven Emergency Feedwater Pumps, 'j ASME III, Class 3, DSP-508B-4461-00, Rev. 2, 4-2-76, V.C. Summer Nuclear Station, Unit No. 1. "4 8. Steam Generator Design Loads, Auxiliary Feedwater Nozzle, Model D u (51-D) Steam Generator, Design Specification 679060, Rev. 6,- 11-3-80, Westinghouse Electric Corporation. m d 9. Letter from G.J. Braddick, Gilbert / Commonwealth, to C.A.

Price, SCE&G, CGGS-27683, dated May 27, 1982.

10. " Jet Loadings on ASME Section III Piping," Gilbert Associates, Inc., Report No. 1902. II. SWEC Letter.to C.A. Price, SCE&G, dated June 1, 1982. a 12. Memorandum from K.R. Gabel, GAI, to K.Y. Chu, SWEC, dated June 4, p 1982. u 13. Letter from G.J. Braddick, GAI, to C.A. Price, SCE&G, CGGS-27890, n dated June 15, 1982, with

Attachment:

Memorandum from K.R. Gabel ] to J.R. Helwig, dated June 11, 1982. 14. Record of telephone conversation, from K.Y. Chu, SWEC, to K.R. '4 Gabel, GAI, dated June 29, 1982. A 15. Letter from SWEC to C.A. Price, SCE&G, dated June 11, 1982. r Ej 16. Memorandum from K.R. Gabel to J.R. Helwig, both GAI, dated June 28, 1982. 2 j 17. Letter from Teledyne Engineering Services (TES) to GAI, 4813-9, dated Fov. 24, 1980, with

Attachment:

Minutes of Meeting. -j 18. Record of telephone conversation among GAI, TES, and SWEC, July 16, 1982. ~ ] 19. Letter from G. J.

Braddick, GAI, to C.

A.

Price, SCE&G, J

CGGS-28392, dated July 29, 1982, with Attachments. JJ 4-3 -9 .:t a

F, 1 Q D 20. Letter from G.J. Braddick, GAI, to C. A. Price, SCE&G, CGGS-28528, gj dated August 16, 1982, with Attachments. r2 + 21. Letter from G.J. Braddick, GAI, to C. A. Price, SCE&G, CGGS-28587,

ij dated August.24, 1982, with Attachments.

22. Letter from G.J. Braddick, GAI, to C. A. Price, SCE&G, CGGS-28697, dated September 13, 1982, with Attachments. 23. Letter from G.J. Braddick, GAI, to C. A. Price, SCE&G, CGGS-28744, ] dated September 20, 1982, with Attachments. 2 24. Letter from G.J. braddick, GAI, to C. A. Price, SCE&G, CGGS-28800, l p dated September 30, 1982, with Attachments. .d ~ 25. Letter from G.J. Draddick, GAI, to C.A. Price, SCE&G, CGGS-28807, dated October 4, 1982, with Attachments. U 26. GAI Telephone and Conference Memorandum, File I-1.23, 4.8-SW, dated 10/8/82, by K.R. Gabel, GAI, (

Subject:

EFH-101). 11 27. Letter from G.J. Braddick, GAI, to C. A. Price, SCE&G, CGGS-28857, dated October 12, 1982. (Subj ect: EFH-083). 1 28. Letter from P.D.

Harrison, TES, to C.N.

Rentschler, GAI. 4813-184,- October 8, 1982. (

Subject:

EFH-4028). i 4.3.2 Stress Packages-Upon receipt of the field verified isomet'ric drawings,- the Lead }L, Engineering Mechanics Engineer reviewed, logged in, and divided the subsystem into four stress packages for mathematical modeling. Each of these stress packages was terminated at six-wry restraints (anchors, .n equipment nozzles, Reactor Building penetration), except for Package j 102, in which the subsystem was extended and terminated after several restraints at a point where the boundary conditions would not affect the flow path piping being analyzed. These four packages are as shown 'U on Figure 3-1.

• J Package 101:

Supply line from Dresser coupling to turbine-driven EF l Ti Pump XPP-8-EF ga [ Package 102: Discharge line from turbine-driven pump XPP-8-EF to '.3-inline anchor EFH-112 g

• 41 Package 103:

From in-line anchor EFH-112 to Reactor Building penetration No. 213 '1 ,d Packa p 104: From Reactor Building penetration No. 213 to Steam Generator C 3' '.d Appendage vent, drain, and instrument piping up to 1 1/2 inch were not included in the mathematical models because the moments of inertia of r, 4-4 i ? N

7 these are much smaller than the moments of inertia for the main runs, .] and their coupling effects are.therefore negligible. These small pipe lines were not.a part of the scope. However, when considered necessary, a concentrated weight was added at the branch point to ~4 -account for the contributing weight. The Lead Engineering Mechanics Engineer assigned stress analysts to work on these packages. Concurrently, he assigned an engineer to J develop for each package a set of digitized response spectra to envelop the floor response spectra for the locations and elevations where the pipe supports are attached. Highest Elevation of Package Building Support Attachment Figures To Be Envelooed ' D' 101 Intermediate Bldg. 433'-6" 62 (El. 436') 61 (El. 412') 'I J Diesel Generator Bldg. 425'-0" 30 (El. 427') R 102 Intermediate Bldg. 432'-0" 62 (El. 436') 16 61 (El. 412') n 103 Intermediate Eldg. 436'-0" 62 (El. 436') y 61 (El. 412') Reactor Building 443'-1" 8 (El. 462')

r.,

'.{ 7 (El. 435') ~ 104 Reactor Building 441'-1" 8 (El. 462') l' 7 (El. 435') Interior Concrete 477'-3" 21 (El. 475') ^ 20 (El. 462') jE} 19.(E1. 445') 4.3.3 Field Information Requests b Additional field information or clarification was requested by completing a Field Information Request (FIR). Twenty-two FIRS were y submitted to the Project Engineer, who ensured that they were logged in J and indexed. Responses to the FIRS were documented on FIR Response i .) Forms by the field verification team. These FIR responses were ~ J. provided to the stress analysts and incorporated into the analysis. Copies of all FIRS and FIR responses were transmitted to SCE&G. j '* .It r 1 1 a b i d 4-5 m-l

.i n .J 4.3.4 Analysis Input Criteria ^~ 4.3.4.3 Deadweight In ad,ition to the weight of the run pipe, water, valve, flanges, and other fittings, the weights of pipe support attachments and most of the vent, drain, and instrument lines were included in the analysis as concentrated weights. In a few cases the weight of support attachments was represented by a distributed weight along the pipe length with a limitation of not being longer than one pipe diameter on each side of the support. ~ 4.3.4.2 Thermal Condttions Two thermal conditions were considered which included the maximum and minimum temperatures specified for various derign, operating, and environmental conditions. They were as follows: Thermal Condition 1 Entire subsystem 32*F Thermal Condition 2 Line from check valve 1038C-EF 600*F 3 to steam generator nozzle iU Line from Reactor Building Penetration 120*F to the check valve The rest of the line 110*F p Thermal d'isplacements at the steam generator nozzle were also j considered. 4.3.4.3 Internal Pressure i 9, The values of pressure used for the analysis were selected from the maximum of various plant and system operating conditions as follows: Stress Package Internal Pressure (psig) q 101 27 - Supply lines from condensate -y storage tank 60 - Supply lines from service water q system i 102 2250 - Discharge line from turbine driven EF pump to stop-check valves ,a 1020A-EF, 1020B-EF, and 1020C-EF. l 3 Recirculation line up to orifice 55 - Recirculation line' downstream of ] orifice a 1360 - All other portions 3 4-6

2

..e

n. _.

.i F-.1 103 1715 - Discharge lines from motor driven EF pumps to stop-check valves 1019A-EF, 1019B-EF, ' and 1019C-EF. Recirculation line up to orifice 55 - Recirculation line downstream of a orifice 1360 - All other portions. !.] 104 1360 4.3.4.4 Seismic Response Spectra Both Operating Basis Earthquake (OBE) and Design Basis Earthquake (DBE) were analyzed by means of the response spectrum approach. The a contribution of closely spaced modes was considered by the grouping method as addressed in NRC Regulatory Guide 1.92. ?'4 The dynamic analysis considered. all significant modes up to and & _i including the first frequency exceeding 33 Hz. 7 The individual spectra used for enveloping were taken directly from the ?'. GAI Seismic Analysis Specification for OBE with a 1 percent damping factor. The DBE response spectra were obtained by scaling the OBE , g response spectra for a 2 percent damping factor with the following ,i factors as defined in the GAI specification: Reactor Building 1.50 tm t'O Intermediate Building 1.55 i:;1 Diesel Generator Building 1.62 U In addition, depending on th location of pipe supports, GAI applied a n different type of scaling fc.: tor (Gamma) to the vertical components of -1 the response spectra. This factor throughout all four piping packages was 1.0. l 4.3.4.5 Seismic Anchor Movement All components of the seismic anchor movements were taken from the GAI ? isometric drawings and referenced GAI correspondence. Another specific ,j-criterion provided to SWEC was that if all three directions of relative seismic anchor movement between two adjacent supports were equal to or a less than 1/8 inch, the differential movement was not considered in the y analysis. ., w 4.3.4.6 Jet Impingement Load u a Break point, jet orientation, and jet impingement forces were provided in GAI Report No. 1902, Jet Loading on ASME Section III Piping. There l}2 f. were seven break points to be considered, five from the 4-inch steam line' to the turbine and two from the discharge side of this subsystem. ] u 4-7 n + .d .~. ... - ~.. -.... -

l 9M W I "1 The jet from one of these seven break points was not analyzed per GAI's Q direction that a shield installation negated this jet force. q The jet impingement forces given in the report did not include the g dynamic load factor and shape factor. SWEC assumed a shape factor of 0.60.and a dynamic load factor of 2.0 for the initial stage of jet impingement. (See Section 4.12 for subsequent revision of these .,ll criteria for specific jets.) During the initial stage of jet !J impingement, all shock suppressors (snubbers). were considered effective. After the initial stage, the jet load becomes a stationary O force. Therefore, a second analysis was performed for this condition, d, in which no dynamic load factor was included and the shock suppressors were considered deactitated. 1: 4.3.5 Calculation Preparation 9 Stress analyses were performed using the NUPIPE-SW (ME 110) computer fl program. u { Wcck sketches representing the mathematical models of the stress }].a packages were prepared (Figures 4-1, 4-2, 4-3, and 4-4). The data in the work sketches included dimensions, pipe support types and orientations, node and mass points, valves and operators, elbows, and other fittings. All work sketches were checked by a stress analyst y other than the preparer for completeness and accuracy. During the process of preparing work sketches, the stress analysts C identified seven items requiring clarification. Each item was d documented in an Open Item Report (OIR) and resolved as discussed in Section 4.10.1. b The stress calculations were prepared and reviewed in accordance with SWEC Engineering Assurance Procedure (EAP) 5.3, Preparation and Control of Manual and Computerized Calculations (Nuclear Proj ects). In ,1 addition to the normal standard presentation of a calculation, the ~ stress package included comparison of pipe support loads, anchor loads, and thermal movements with the data received from GAI, and comparison i of penetration and equipment nozzle loads with the given allowables. 4.4 E?ALUATION CRITERIA .d 4.4.1 Piping Allowable Stresses I l 7 The piping is to meet the requirements of 1971 ASME Boiler and Pressure I,j Vessel Code, Section III (ASME III), Division 1, Classes 2 and 3, with addenda up to and including the Summer 1973 issue and Code Case N-240. Loading combinations together with their design criteria are as ) j follows: 1 s .] 4-8 ~~ i s.J ........... ~ y .___.-,m __--,e".,. ,y- ---N-

T.: v .s 1) System Normal / Upset. I Operating Condition 'J NC 3600 Allowable Equations Combination Stress q aU 8 DL + LP Sh 9 DL + LP + OBEI 1.2 S 0 11 DL + LP + TH + OBEA Sg+Sh 2 2) System Upset II Operating Condition (Plant Emergency) NC 3600 Allowable (.; Equations Combination Stress .. t 9 DL + LP + JI 1.5 Sh .y .0 3) System Emergency Conditica (Plant Faulted) q NC 3600 Allowable l Equations Combination Stress 3 9 DL + LP + DBEI 1.8 S q h 1 u where Sh= all wable stress at maximum (hot) temperature } S = all wable stress at minimum.(cold) e temperature SA= f(1.25 Sc + 0.25 S ), f=1.0 the s

q stress range reduction factor ij DL =

Deadweight LP = Longitudinal pressure stress OBEI/DBEI = Inertia effects of OBE/DBE. ,i OBEA = OBE anchor movements d TH = Thermal load JI = Jet impingement load _I 4.4.2 Pipe Support and Anchor Load Combinations "7 The pipe support and anchor loads from the following load combinations j were compared with the loadings from GAI pipe support drawings and TES documents. If the loads exceeded the original design values by 15 { --, percent or more, and if they also exceeded them by 100 lb or 100 ft-lb, an Open Item Report (OIR) was generated and submitted to the Lead Engineering Mechanics Engineer, Proj ect Engineer, and Project Manager for review and resolution. .3

• • S, 4-9 l

.2 - -.. ~.

3 ) System Operation Loading Combination Normal. DL + TH OBEA )\\ 2 Upset I DL + TH + (OBEI + Upset.II DL + JI Emergency DL + DBEI ~ 4.4.3 Equipment Nozzle Loads 4.4.3.1 Steam Generator Nozzle ~ ~ Forces and moments from individual load cases were compared with the allowables given in I)esign Specification 679060, Rev. 6, 11-3-80, Westinghouse Electric Corporation. 4.4.3.2 Pump Nozzles r Forces and moments derived from the same load combinations as for pipe supports were compared with the allowables given in the design specifications for motor and turbine driven EF pumps. o 4.4.4 Reactor Building Penetration No. 213 The forces and moments from each individual load case were first .~ transformed to axial and shear forces, and torsion and bending moments, t which were then compared with the allowables given in the GAI Design Specification, DSP-606-044461-000, Rev. 9, 2-1-82. Load comparisons were made at both ends of the penetration. 4.4.5 Comnarison of Thermal Movements !l Thermal movements from the two thermal conditions at pipe supports were compared with those presented in GAI pipe support drawings. If the movements exceeded the. original values by 15 percent or more, and if they also exceeded them by 0.02 inch, an OIR was generated and sent to the Lead Engineering Mechanics Engineer, Project Engineer, and Project Manager for review and resolution. i 4.5 PIPE STRESS REVIEW 7 All piping stresses were found to be within allowables for all analyses ) performed. Maximum stresses from each stress packages are presented in Table 4-1. 4.6 SUPPORT LOAD REVIEW .a All four load combinations for all supports were tabulated and compared with the design loads from GAI pipe support drawings and supplements provided by GAI. For terminal anchors, the loadings from SWEC's 2 analysis were combined with the loadings from the interfacing side which war not analyzed by SWEC. These loadings were provided to SWEC by GAI and TES. 4-10 ...p..

,U. l The load comparisons proceeded in two steps. The -first step was to i calculate the difference between SWEC's values and the original design + values. The second step was to calculate the ratio of the difference to y the original design value. J Based on the information provided in the pipe support load comparison tables, a summary table for each package was prepared to indicate differences in values and in percentages. Unless noted otherwise, 3 J SWEC's load in the summary tables is for an Upset Condition, generally the controlling design case. The GAI load is for the same load 9 combination. The value in the column "Differeneg" is SkIC's load minus / the corresponding GAI loac.. This value divided by the corresponding GAI load is recorded ?in the column " Percentage." The last column, " Dominant Factor," indicates the load case contributing most to this g difference. There are probably four major factors that contributed to /- differences. One is the effect of seismic response spectra (seismic inertial). The second is the effect of differential seismic support ,4 movement, noted in the tables as " Seismic Movement." The third is the .J jet impingement

effect, which could have been caused by misinterpretation of the impingement target area.

The fourth is 7 modeling differences due to geometrical differences, program h differences, and engineering judgment, in assigning mass points and support stiffnesses. 4.6.1 Comparison Review - Package 101 This package contains 32 supports and two anchors. The load comparison [- indicated that 20 supports and both anchors had load differences and 1 4_ ratios exceeding 100 (1b, ft-lb) and 15 percent. The primary contributor for all these differences appeared to be the seismic l y] response spectra and movement of the Diesel Generator Building, which ~ were not considered in the original analysis. Table 4-2 identifies all differences. 4.6.2 , Comparison Review - Package 102 This package contains 23 supports and 4 anchors; The load comparison 7j indicated that 5 supports and 3 anchors had load d-fferences and ratios i exceeding 100 (lb, ft-lb) and 15 percent. Four of these were probably caused by the difference in the effects of the seismic response ] spectra. One anchor (EFH-113) had a large discrepancy in My. This was l J caused by the difference in the mathematical models. SWEC's model l represents the physical location of the anchor, i.e., I foot away from lq a vertical riser, while the GAI's model assumed the anchor located at the intersection where the vertical riser joins the horizontal run. ~ One support, ETH-048, is near the jet impingement target from break number 32. The GAI load did not include this effect due to o, , alj misorientation of this jet. For anchor EFH-111 all force and moment components except Mx were within the comparison criteria. The difference of Mx appeared to be caused by deadweight. SWEC's analysis

1 showed that Mx due to deadweight was almost entirely caused by a valve 2

located only few inches away and having its center of gravity 8.5 m 3 4-11 ~) J

m ~ inches off the pipe axis. The difference for the remaining support ,j ETH-057 seemed to be from thermal effects. Table 4-3 identifies all differences. ~, j 4.6.3 Comparison Review - Package 103 This package contains 32 supports and 4 anchors. The load comparison indicated that 6 supports and one anchor had load differences and ./ ratios exceeding 100 lb and 15 percent. Three of these were caused by the effect of jet impingement from break No. 33. This jet was '~ ! misoriented in the original GAI data. The higher total loads in f, the anchor and two other supports were probably caused by higher thermal loads. The di'fference for the remaining support EFH-4029 load was primarily due to relative seismic movement. This was the vertical and lateral support closest to the Reactor Building penetration. Table 4-4 identifies all differences. 9 4.6.4 Comparison Review - Package 104 This package contains 12 supports. None of SWEC's support loads 'l exceeded the comparison criteria. 4.7 EQUIPMENT N0ZZLE LOAD REVIEW 4.7.1 Steam Generator Nozzle All forces and moments from each. individual load case' were substantially smaller.than the allowables ' specified for this nozzle. 4.7.2 Pump Nozzles 1-. L.; The forces and moments from all required load combinations' at the motor driven pump nozzles and at the discharge nozzle of the turbine driven

a pump were smaller than the specified allowables.

However, at the ~, suction nozzle of the turbine driven pump the X-force component and the force resultant from Upset II Condition were greater than the allowables. Also, the moment resultant from Upset I Condition was greater than the specified allowable. l. -. ' 4.8 REACTOR BUILDING PENETRATION LOAD REVIEW t [j The shear forces of SWEC's analysis for deadweight and seismic load cases at the outside interface exceeded the specified allowables, but i.; fell within the values from TES's analysis. Since GAI had concluded previously that the penetration was acceptable for the TES forces, it must be acceptable for SWEC's forces. At the inside interface of penetration, SWEC's analysis indicated that all forces and momenta were within the specified allowables. 7

• m)

.y 4 4-12 t .b . ~ _..., _

• 1 N

~1 4.9 THERMAL MOVEMENT REVIEW l ol ~ 4.9.1 Thermal Movement at Support Locations Thermal displacements in unrestrained directions .at. all support d locations were reviewed and compared with the values on the GAI pipe support drawings. No significant difference was found between these two analyses. In two instances the difference exceeded 0.01 inch, but cd did not exceed 0.02 inch. ] 4.9.2 Thermal Movement at Supports with Excessive Gaps -i The field verification' effort identified two supports with excessive 7 gaps. These were EFH-099 and ETH-080. .7 EFH-099 (Stress Package 101) was originally designed as a north-south restraint. During field walkdown the clearance was found to exceed the criteria; therefore, this support was assumed ineffective in the stress d analysis. The result of the analysis verified that this assumption was correct. The maximum thermal displacement is 0.019 inch at this point. q The total displacement including the effects of deadweight, thermal, Q and seismic will amount to 0.030 inch, much less than the existing clearance. EFH-080 (Stress Package 103) was originally designed as a vertical and 't east-west restraint. Since the hori:: ental gap exceeded the criteria, the east-west restraint was considered ineffective in the analysis. h, The result of the analysis indicated that thermal movement would be G 0.019 inch, and total displacement including the effects of thermal, seismic, and deadweight would be 0.032 inch, much less than the f* existing clearance. Therefore, the assumption was verified to be ,y correct. .n 4.9.3 Thermal Movement at Supports with Small Clearance 1 EFH-4020 is a box-type vertical support with very little lateral clearance, 7/o4 inch. Normally, a clearance of at least 1 inch would ?j be expected. However, the analysis verified that the pipe lateral U movement is expected to be very small, 0.019 inch for thermal case, and 0.022 inch for maximum load combination. Therefore, this clearance is acceptable. l 4.10 OPEN ITEM REPORTS i j 4.10.1 Open Item Reports for Interpreting Field Walkdown Data During the stage of preparing work sketches, seven OIRs requesting ] clarification of field walkdown data were filed by the stress analysts. u These were reviewed by the Lead Engineering Mechanics Engineer, Project Engineer, and Project Manager. These OIRs were resolved based o t the l3 responses to FIRS. j 2 I

)

i 4-13 ~ r dud a .e>g. . g. e, s es== g es>.ee -*4

• • • • e 9

g . Pe e-- J

.. : ~ ,~ ~ 5. - :..:E.J.: ~ 73 [ 4.10.2 Open Ites Reports for Analysis Review u 4.10.2.1 Initial Analvsis 9] The initial analysis was completed based on the design criteria provided by SCE&G and GAI. Review o f. the results indicated many ly supports, some equipment nozzles, and the Reactor Building penetration-

i exceeded the original design values.or allowables by. the amount specified in the comparison criteria.

Each of these_ items was documented on an OIR and submitted to. the Lead Engineering Mechanics @*J l

Engineer, Project Engineer, and Project Manager for review and resolution.

i These OIRs were forwarded to SCE&G and GAI for their review to be 1 1 :. certain that the. input criteria provided to SWEC were complete and ~ correct. A meeting was held in GAI's office and attended by j - representatives of SCE&G, GAI, TES, and SWEC. During review of each

}

OIR, it became evident that some of the data transmitted to SWEC were inconsistent and required corrections. These were: 4 G Seismic anchor movement for Diesel Generator Building il Jet impingement effect should not be considered positive and l negative (t) L]s j Seismic effects due to response spectra and anchor movement should be. combined as square-root-of-the-sum-of-the-squares (SRSS) instead of absolute summation As a result of these corrections a new analysis was performed. This was called " Comparison Analysis" and was the basis for SWEC's i .d evaluations. l m 4.10.2.2 Comparison Analysis i; rq I A comparison analysis was made to incoporate the changes of criteria plus other minor adjustments, such as distribution of deadweight and a j consideration of pipe support attachment points offset from the pipe axis. The following OIRs were written and submitted for further review and resolution as a result of this analysis. Copies of OIRs were ] forwarded to SCE&G, GAI, and TES for their review and action. c.2 Packase No. Total Number of OIR Review Categorv 7y 101 20 Support

Load, Section

'4.6 2 Anchor Load, Section 4.6 4D q 1 Nozzle Load, Section 4.7 '1 s 102 5 Support

Load, Section 4.6 3

9 Anchor Load, J:c.et, ion 4.6 3. 103 6 Sup' port

Load, Secti3n 4.6

^ 1 Anchor Load, Section 4.6

l. 3 1

Penetration Lead, Section" 4.8 ..w* 104-None ~, 4-14 ~\\ ^ x

~I 4.11 POTENTIAL DISCREPANCIES e 'The OIRs were scrutinized and those reaching the following conditions classified as Potential Discrepancy (PD) items

• for which further were g

,,1 evaluation and corrective action might be required. 1. Difference between SWEC's and GAI's design loads is 1 substantial .4 2. SWEC's maximum load is significant in respect to support ~ capacity

.)

3. Adequacy can' only be justified with additional calculation ' :n j'j The majority of the OIRs stemmed from three F.eneric Potential Discrepancies which were: 1. Diesel Generator Building Seismic effects, including response spectra and support F3

movement, fro,e the Diesel Generator Building were not

d included in the original GAI design data for Subsystem EF-01 (SWEC Stress Package 101).

-.j 2. Jet Impingement ( Two jets were misoriented. In one instance the target area hl of a jet impingement in the design document (1902) appeared ~2 to be inappropriate. Subsequent communication indicated that the jet need not be included in the analyses because shield 'I installation negated this break load. 1 3. Mathematical Modeling Techniques 7 These could have included inconsistent pipe support a stiffnesses, lumped mass locations, geometrical differences, and differences in engineering j udgments. Several c., differences identified' which could have contributed to frequency shift and load differences were the following. SWEC's analyses used a consistent stiffness value of 1 x 1012 (lb/in, in-lb/ rad) for all supports to simulate the TES criteria which basically used rigid supports. TES's ? analyses however actually used several stiffness values, i.e., infinite stiffness for supports oriented in global 8 axes, 1 x 10, 1 x 10s or 1 x 1012 lbs/in for supports not oriented in global axes, and 3.5 x 104 lb/in for the horizontal direction of support ETH-4029 to represent the actual stiffness for analyzing the effect of seismic 's anchor movement. a s s- '.z-s 4-15 ~ ~. .,... ~. y..

. :: a.

..:.g :. ~+.

-?. .= / ~; .\\ u ,9 SWEC's model did not have any mass points at support y locations, while TES's did have mass points at support locations. Also, the number of mass points between j supports was different in these models. Since' lumped mass dynamic analysis is an approximation, selection of "~ mass points is based mainlyJ on engineering judgment and different program limitation.. Depending on the piping t support.. system and number = of, mass - points - between i supports, these' differences may sor may: not, have any effect'on support loads. m .s + Several minor geometrical differences 1were identified. ^ Some o'f these geometricaP 7 differences had been g identified ' earlier and indicated in GAI isometric. -) drawings as analytically acceptable. Other differences were the following: l One of the elbows in EF-01 was not c:odeled as an elbow in-TES's analysis. This elbow has a a lug welded to it to facilitat e the connection to the hanger rod EFH-105. An elbow is more j flexible than TES's model with two straight pieces of pipe ~ joined at 90 degrees. However, because cf .m the lug acting.as stiffener, the standard elbow flexibility may not represent the true flexibility of this elbow. Therefore, 2S ' s nodel is' considered acceptable. i. + TES's analysis for EF-02 assumed the anchor EFH-113 i located at the intersection where a vertical riser joined the horizontal run. SWEC's model represented the physical location o!-' the anchor, i.e., I foot away from the vertical riser. ~) . 'l 7i TES's model for EF-03 assumed that the horizontally ~ skewed support ETH-083 was offset to the east of the pipe axis. The functional orientation was

4 modeled correctly; however, the support is

actually

-d offset to the west of the pipe axis. ) 4.11.1 Package 101 / ,.s All 23 OIRs (Section 4.10.2.2) were primarily due jto the. effects of Diesel Generator Building seismic response-spectra and-support 3 movement (.,which were

j not included in the original design (Potential Discrepancy No. 12 3

- - -.j A jet impingement force caused the suction' nozzle of the turbine driven pump to exceed the specified allowables - (Potential Discrepancy No. 2). / w A s fm ! jj j 4-16 1 I \\

m

,+

n.. 4 Comparison of the first mode frequency indicated a minor difference y between these two analyses. SWEC's mathematical model had a natural frequency of 11.04 Hz in the first mode versus 11. 9 ' Hz for TES's analysis. (This was subsequently reduced to 11.7 Hz in the TES revised analysis.) This lower frequency caused greater seismic inertial response in SWEC's analysis of the piping subsystem. There could have been numerous reasons contributing to this difference, such as number an6'. location of lumped mass points, magnitude of masses, geometrical d difference, and stiffness ef pipe supports (Potential Discrepancy No.

3) '.

• )

4.11.2 Package 102 a ? -g Three of the. eight OIRs (Section - 4.10.2.2) were classified as PDs. ]' These are for support EFH-048 (Potential Discrepancy No. 2), and for anchors EFR-113 and 114 (Potential Discrepancy No. 3). The other five OIRs were judged to be satisfactory without further evaluatica or r M corrective action. The bases for the judgments are: d' The difference does not occur in the controlling design ~~{ condition (supports EFH-051 and 4005, anchor EFH-111) L! Difference appears to be insignificant for the as-built g support structure (support EFH-057) I) Maximum design load appears to be substantially smaller than . any commercially available support. component (support j p} EFH-182), 4.11.3 p -Package 103 L.; Five of the seven OIRs. (Section 4.10.2.2) were classified as PDs. . Three of them for supports EFH-060, 61, and 62 were caused by an r: erroneous jet impingement target (Potential Discrepancy No. 2). Two 'j others are linear type supports not oriented in global axes of the model. These two supports (EFH-083 and 4029) were represented in TES's analysis by a stiffness value smaller than-others (Potential Discrepancy No. 3). W Two OIRs for supports EFH-082 and 115.were judged to be satisfactory 9, without furt.ier evaluation or corrective action. The basis for this '.j ' judgment is that the difference exists in the Normal Condition only, which is not the controlling design condition. The OIR for Reactor i Building penetration No. 213 was resolved satisfactorily based on 3: ' additional information provided by GAI and TES. According to GAI, during the original design stage they recognized that TES's loads exceeded the specified allowables. Further evaluation was made and it f was concluded that the penetration was designed adequately for those 'J loads. SWEC's loads are similar to or smaller than TES's loads, and therefore the same conclusion is valid. 9L O

u 4-17 m

t 4.12 FINAL ANALYSIS

)

On August 13, 1982, a meeting among SCE&G, GAI, TES, and SkIC was held at the 'TES office in Waltham, Ma. The purpose of the meeting was to e review the OIRs. issued as a result of the Comparison Analysis. In the meeting, GAI advised that TES was to rerun the computer analysis for EF-01 (SWEC Stress Package 101) to include the seismic input from the Diesel Generator Building. Also, three jet forces would be redefined. GAI confirmed this information by a copy of a letter to SCE&G, dated August 16, 1982 (Ref. 20, Section 4.3.1), f u which three jet forces, 7 one each for subsystems EF-01, EF-02, and EF-03, were reduced. In ? addition, these forces were to be multiplied by a factor of 0.75, representing the combined effect of the dynamic and shape factors. qcuQ SWEC performed a Final Analysis using these new jet forces and factors. Results of this analysis were reviewed and compared with the new m support load summary sheets which were transmitted to SkIC by GAI as attachments to letters to SCI &G dated August 24 and September 13, 1982 (Refs. 21 and 22, Section 4.3.1). The support load summary sheets ~ included all supports except spring hangers for Subsystem EF-01, one C support (EFH-113) for Subsystem EF-02, and one support (EFH-062) for '.U Subsystem EF-03. q

,j 4.13 RESOLUTION OF OPEN ITEM REPORTS AND POTENTIAL DISCREPANCIES p

The Open Item Reports and Potential Discrepancies issued as a result of ,i the Comparison Analysis were reviewed again with the latest ' knowledge of the new load summary sheets. Section 4.11 identified three Potential Discrepancies for which corrective actions might be required. ,,,7 Potential Discrepancies Nos. I and 2.were subsequently corrected in

• ES's reanalysis.

] The Potential Discrepancy No. 3, Mathematical Modeling Techniques, a could not be readily reso_ved. SWEC, therefore, reviewed TES's models and input data for seismic inertial analysis for Subsystems EF-01, EF-02, and EF-03. The results of this review indicated some ,-j differences between TES's and SVEC's models as addressed in Section l 4.11. The frequencies of the first modes in TES's models are 11.7, 14.11, and 'W 4.513 Hz for EF-01, EF-02, and EF-03, respectively, versus 11.04, 13.36, and 4.361 Hz in SWEC's models. These differences of 5.6, 5.3, '3 and 3.4 percent are considered insignificant for the overall structural - J.j response of the piping system. The lumped mass approach for_ vibration l analysis it. only an approximation; minor differences in numerical l .e j results are bound to occur. 'U Localized effects on EFH-113 resulting from the correction of SWEC's identified differences were included in TES's new support load summary /1 sheets. Therefore, this Potential Discrepancy No. 3 was resolved i M satisfactorily. E! u 4-18 D S l

_./__ ..~ -..s - _1 t c. ?

• Ub

~~ 4.13.1 Package 101 .s. Comparison between the result of SWEC's Final. Analysis 'and TES's new - 3 ' load summary sheets indicates the following for the 23 - OIRs issued earlier: 1. Eight OIRs for Supports EFH-094, 095, 096, 098, 4019, 4022, q 4031 and 4034 are closed, since TES loads from the reanalysis -4 s with Diesel Generator Building seismic response spectra and anchor movement are now within the evaluation criteria. i .s !d 2. OIR for Support EFH-4023: SWEC's loads 'n Upset' and i Emergency Conditions are within the evaluation criteria. q'j SWEC's load in Normal Condition is 141-lb larger than TES's load; however, this load.is not a controlling design case; it amounts to only 64 percent of the load in Upset Condition. q Therefore, this OIR is closed. r 3 w 3. OIR for Support EFH-4026: SWEC's loads in the Upset Condition are 2487 lb compression and 5251 lb tension. The I tensile force is within the evaluation criteria, while the compressive force exceeds the TES's load by 344 lb and 16 percent. Since the support is a short rigid strut, pin.to pin 13 5/8 inches, the controlling design load must be the -r gh larger tensile force. This OIR is considered closed. i 4. OIR for Support EFH-4046: SWEC's load in the Emergency 3 Condition is.881 lb,.19 percent over TES's load. However,. this is not a controlling design case. Both SWEC's and TES's loads in the Upset Condition are approximately 8400 lb. N Ther.efore, this OIR is considered closed. .'y 5. Anchor at Tiode No. 441: This is an anchor embedded in 9 concrete. Only-the negative Mx moment in the Normal pj, Condition exceeds the evaluation criteria. However, this value is only about one-half of the positive Mx - value. Therefore, this OIR is considered closed. i n.-

h. i.

d 6. Suction Nozzle of the Turbine Drive EF Pump: The moment resultant is still higher than the specified allowable, however, it is lower than TES's value. Since GAI considered TES's value acceptable, SWEC's value must also be acceptable. This OIR is therefore closed. G~ 3 7. OIR for Support EFH-4035: Except the positive load in the Upset Condition, all other loads from SWEC's analysis are within the evaluation criteria. However, this positive load E-j of 903 lb is much smaller than the actual controlling design load. Therefore, this OIR is closed. ] 8. OIR for Support ETH-097: SWEC's model did not include the .c support EFH-099, which was designed as a north-south m I lj 4 - y

{

~

---

,,_a

a w- ..e w _ r m. c.;. +_ ~ restraint, but was assumed ineffective due to the excessive a gap (Section 4.9.2)..SCE&G concurred with SWEC's finding and informed SkIC that the intended design function was restored. ~~ SWEC performed a computer analysis to include EFH-099. The 1 .i result of this analysis indicated that the maximum support load for EFH-097 r' educed significantly to fall within the evaluation criteria. Therefore, this OIR is closed. 9. Six OIR's for Supports EFH-102, 103, 104, 106, and 4021, and one concrete anchor (SWEC's node No. 1, GAI No. 158): The i higher support loads from SWEC's analysis were primarily due a to the first, mode contribution from the seismic inertial load case. The first. mode frequency from TES's analysis is F-j considered to be sufficiently accurate as discussed before, and since the amplified acceleration response spectra had already been peak spread to account for minor difference in m frequency calculation, the acceleration resulting from TES's frequency excitation can therefore be considered acceptable. An analysis was then performed by SWEC to apply the acceleration value corresponding to TES's first mode ] frequency to calculate the seismic inertial loads. The a result of this analysis indicated that the loads reduced substantially to fall within the acceptance criteria. Therefore, these OIRs are considered closed. 10. Two OIRs for Support EFH-101 and '4028: The analysis described above also indicated a substantial. reduction of loads for these two supports. These. loads still exceeded ~ the design loads by more than the criteria; however, no generic reason could be identified. The maximum loads for -,2 the Upset and Emergency Conditions are 1545 lb and 2341 lb M for EFH-101, and 2852 lb and 2655 lb for EFH-4028. These loads were therefore transmitted to GAI and TES for 7 evaluation. It was confirmed that the supports as designed J are acceptable for these loads. Therefore, these OIRs are closed. .~ 3 ...f,, 4.13.2 Package 102 i i l 3 Five of the original eight OIRs were resolved satisfactorily in Section j 4.11.2. The remaining three OIRs are discussed in the following: e 1. OIR for Support ETH-048: SWEC's Final Analysis with the 71 revised jet force indicates' that the support loads fall gi within the evaluation criteria. This OIR is therefore closed. .-1 .d 2. OIR for Anchor EFH-113: The new load summary sheet from TES ~ i indicates a My moment much larger than SkIC's value. Therefore, the concern expressed before is resolved

d satisfactorily.

This OIR is therefore closed. _v .0 4-20 y 2 s.1

n:) ] ] 3. OIR for Anchor EFH-114: The major contribution to the a maximum design load was the seiamic inertial load, which was calculated based on the absolete summation o'f the seismic n inertial loads from the two analytically independent piping .1 subsystems terminated at this anchor. This approach is very conservative. SkTC would normally combine these inertial loads with the square-root-of-the-sum-of-the-squares n ,1 spproach. When this approach. is used, SWEC's maximum load 4 reduces substantially to fall within the evaluation criteria. Therefore, this OIR is closed. .+ i; 4.13.3 Package 103 ,7l Two of the original seven OIRs were resolved satisfactorily in Section 4.11.3. The remaining five OIRs are discussed in the following: 1. Three OIRs for Supports ETH-060, 61 and 62: SkIC's Final m l Analysis with the revised jet force indicated that the support loads are within the evaluation criteria as a result of correcting Potential Discrepancy No. ,2, Jet Impingement. ] Therefore, these OIRs are considered closed. 2. OIR for Support EFH-4029: SWEC's Final Analysis indicated a much higher horizontal load than TES's. SWEC made an analysis using a horizontal support stiffness value approximately equal to the actual stiffness of this support. This had been done by TES to more accurately model the distribution.o f load between EFH-4029 and the containment penetration which is only 1.7 feet away. The result of this analysis indicated that the support load was reduced to less "I than TES's value when the representative stiffness was used. Therefore, TES's' design value is considered acceptable. This OIR is considered closed. Cj 3. OIR for Support EFH-083: This horizontally skewed support was modeled offset to the east of the pipe axis _ by TES (Section 4.11). SWEC's loads were transmitted to GAI for p,i evaluation. GAI confirmed that this support as designed was acceptable for thece loads. Therefore, this OIR is closed. 4.13.4 Package 104 .i No Open Item Reports (OIRs) and no Potential Discrepancies (PDs). ~. ~) 'O ..i 1 Li 4-21 "i ).'.: =~2.-. ' ~ ~ ~ "

.1 b 'l 4.14 CONCLUSIONS .J The independent seismic design verification analysis confirmed that all ] piping stresses in' the flow path of the Turbine Driven EF Pump of the y Emergency Feedwater System to Steam Generator C were within the ASME Code allowables. Review of the piping thermal movement, Reactor ~ Building penetration No. 213, and all pump nozzles but the suction 'd nozzle of the turbine driven pump led to the conclusion that adequate design data were properly used by GAI. However, SWEC's Comparison Analysis showed that loads for 26 supports, four anchors, and the suction nozzle of the Turbine Driven EF Pump were substantially higher ?J than design loads. These differences appeared to be caused by three Potential Discrepancies. Two of these discrepancies, i.e., Diesel 7 Generator Building seismic input and jet impingement locations and jj orientations, were subsequently corrected by GAI and TES. The other potential discrepancy, Mathematical Modeling Techniques, was studied extensively by reviewing TES's mathematical models and computer input n } data. Some minor geometrical differences were reconciled. Other differences were considered insignificant. This potential discrepancy was, therefore, resolved satifactorily. d Overall, the analysis of this piping subsystem was performed adquately. The design loads of the supports were considered sufficient to maintain the structural integrity of this piping subsystem with the required safety margin. 1 e C N ~3 2 "? .l ..] h1 J 4-22

L:,,_.- C,2 [J (} kd CN} [ 53 L:A3 i. :: 3 U CU O Edi UM b* O b bdI li l! t i-TABI.E 4-1 r HAXINIIH PIPING STHESSES FROH Tile COMPARISON ANALYSIS Package.No. System Conilition Egg tion Nocle No. Max. Stress (psi _) Allowable Stress (psi) i 101 Normal 8 323 2,097 15,000 llpset I 9 197 4,991 18,000 .II I 20,188 37,500 it Ilpset iI 9 227 2,251 22,500 Emergency 9 197 5,833 27,000 102 Normal 8 950 8,374 15,000 '4 Upset 1 9 950 10,593 18,000 i i1 524 30,356 37,500 'l Upset iI 9 950 8,378 22,500 Emergency 9 950 10,648 27,000 103 Noreal 8 1008 9,428 15.000 l llpset I 9 1008 12,801 18,000 II 14 28,104 37,500 Upset 1I 9 1008 9,429 22,500 Emergency 9 1008 13,331 27,000 l-204 Normal 8 94 1,336 15,000 Upset I 9 123 17,391 18,000 Ii 123 30,069 37.500 Ilpset 11 9 No Jet Impingement Emergency 9 123 17,923 27,000 'I i I 3 I t t II 1 l .i e

L. _. L._] k.13 L_ J l.ij t3 L.II EU l] E: "O (22 'OJ [ 31 CI 2 J C.I23 L%) i.'d:] t t l 1 TAllt.E 4-2 SUpp0RT 1.0AD REVIEW OF PACKAGE 101 - COMPARISDN REVIEW 1i I Snpport No. SWEC's SWEC's Correep. GAI Percentage Dominant Support EFil-Node No. I.oad I.oa d Ili f ference (%) Factor _ Type 094 95 1352 783 569 73 Seis^mic Movement U-Holt t 095 85 1424 1230 194 16 Seismic Novement U-Ilolt 096 83 321 210 til 53 Seismic Novement Clamp 097 71 891 577 314 54 Seismic Movement Clamp 098 63 527 399 128 32 Seismic Incertial U-Bolt lut 145 2727 1034 1693 166 Seismic Inertial Snubber 102 153 1342 1039 303 29 Seismic Inertial Framing 103 165 3410 2307 1103 48 Seismic Inertial Strut 104 175 2173 1637 536 33 Seismic Inertial Rod 106 197 2478 1153 1325 115 Seismic Inertial Framing 4019 235 549 22 527 2396 Seismic Novement Sunbber 4021 213 1277 672 605 90 Seismic inertial Snubber l 4022 245 2292 1083 1209 112 Seismic Novement Strut 4023 343 1282 966 316 33 Seismic Inertial Strut 4026 257 5251 2692 2559 95 Seismic Novement Strut 4028 163 4171 2307 1864 81 Seismic Inertial Strut 4031 213 5050 2302 2748 119 Seismic Novement Framing 4034 275 1613 110 1503 1366 Seismic Novement Strut 4035 273 2110 1216 894 74 Seismic Hovement Strui A 4046 4! 881* 683 198 29 Seismic Hovement Strut Anchor 1 1426 1057 369 35 Seismic Inertial Concrete Anchor 441 922 785 131 18 Seismic Inertial Concrete NOTE: I.oads are forces in Ib.

• Emergency Condition

La ( La L,. :. ] L..t.i L. i3 L. 1 i....] LI ' 25 C JJ bd 2 lI Ld k*d I-' #' 2 I i i TAlli.E 4-3 SUPPORT LOAD REVIEW OF PACKAGE 102,,00HPARISON REVIEW Support No. SWEC's SWEC's Corresp. GAI Percentage Dominant Support EFH-5' ode No. Load I.nad Difference (%) Factor Type i 048 6 3') 1227 280 947 338 Jet Impingement Strut l 051 754 '38 59 129 219 Seismic Inertial Snubber + 057 Y' 239 102 43 Thermal Framing li st-tio 1416 200 1216 608 Dead Weight Trunnion 113'- 824 III76 2732 8444 309 HodeI Trunnion 3W" 970 477 336 141 42 Seismic Inertial Trunnion l 182 544 162*** 19 143 753 Sci.smic Inertial Snubber l 430', 532 355M* 117 238 203 Seismic Inertial Strap i r l' I! NOTE: linless otherwise noted, loads are fo rce.s in Ih.

• EFil-ll3 is an anchor; the indicated loads are Homent.Hy in i n-lie.
  • EFil-114 is an anchor; the indicated loads are For ces Fx.
    • l.oads indicated are from Emergency Condition.
    • fEFil-lit is an anchor; the indicated loads are Homent Hx (in-lb) in Normal Condition; all others are within the criteria.

g i I ( J

L',, Q tg ; U.J I_;.* 4 kidi hh g ' 'a) 4.2 bM' bu5 b b-

• E
  • 5' lI i

'l TABl.E 4-4 SilPPORT 1.0All REVIEW OF PACKAGE 103 - COHPARISON REVIEW 3' l l' Support No. SWEC's SWEC's Corresp. GAI Percentage Dominant Support EFil-Noile No. Load Load Difference (%) Factor Tyge__ i 060 1019 995 757 238 31 Jet Impingement Si r..t 061 1020 925 412 513 125 Jet Impingement Strut. 062 1018 678 166 512 308 Jet Impingement Strap 082 1029 365*** 189 176 93 Thermal Framing 083* 1500 2012 1160 852 73 Thermal Skewed Strut i15** 1400 4420 3509 911 26 Thermal Trunnion 4029* 1001 2148 914 1234 135 Seismic Movement Framing I i l' l. f f NOTE: *0rientation of support not in global coordinate system.

• • Anchor; the indicated loads are Forces Fz in Normal Condition.
  • Al.oads in Normal Condition.

I.oads are forces in Ib. I h I i

c_; ! !..J L.3 !.1.d E iS L1 LJ i.- 3 l' " ' ^

• ~

i sio. .s..auc n co oa. o r*< s a<ss s nca n ,,,,,,,,,o. c,,,o,,,,,,,, c,.. c,3 u,,,,,,,,,,,,,.,.,..so .,,.,,,,, s,.,,,,, o,.,, t,, ,o ,,,,,,, g, y,,,,,3 ;; 3,,,,,,,,,,,,,,,,,, r7...,.,,,,_, y,,,,,,,,,, w,m. /,.,. j.g ..... isx:urM-espios aivisioa i._ <ie<><ie/ rat u s... m.

y =,,.= y,.=,7, y,g..,7,,.,.m,.,7,.,

(tru aositt ru-not tee e. 02.)(ten to2r) (41 161 IRI. IRI c,,, is. . iss-

iss, 3st ng at at ne s,

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J nt J 5.0 AUDIT REPORT, DESIGN CONTROL PROGRAM 7 2 m 55 4 l.: / / PREPARED BY: Dr L./M& lone 1 Audit Team Leader

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s n . REVIEWED BY: J H. MacKinnon , Design Control Audit Manager S. L3 R j APPROVED BY: D. C. Shelton Division Chief, Engineering Assurance -~ ..J D*, J I .3 7, 9 'J ] .b c. d 1 .J

s CONTE.TS .td Section Pase ] 5.1 Purpose 5-3 4 5.2 Scope 5-4 J 5.3 Approach 5-6 .4 5.4 Evidence Examined 5-15 7 ) 5.5 Results 5-21 1 5.6 Conclusions 5-28 m iW Figures ~] 5-1 Simplified Emergency Feedwater 5-5 a Flow Schematic y 5-2 Design Input Flow Chart n1 Sheet 1 5-8 .7 Sheet 2 5-9 5-3 Quality Management Program 5-10 {, Document Hierarchy Attachment q 1. Audit Participants l .t .;,e 9 ~' 5-2 I J

• • 1 i

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lm J ,j 5.1 PURPOSE The purpose of this audit was to independently verify that an adequate design control program, meeting the requirements of 10CFR50 Appendix B, was in place and implemented for transmittal.'.nd utilization of input data for activities associated with the seismic analysis of the Emergency Feedwater ] Piping System for the flow path of the turbine driven Emergency Feedwate Pump of the V.C. Summer Nuclear Station, Unit No. 1. O s 7d -4 1 4 '3 u

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T'; 3.) 5.2 SCOPE 3 The audit included the. inputs used to perform the seismic analysis (pipe stress analysis) of that portion of the Emergency Feedwater System identified as subsystems EF-01, EF-02, EF-03 and EF-22 on Figure 5-1. It also included a review of the procedures for controlling design inputs ~ generated by Gilbert Associates, Incorporated (GAI) or provided to GAI by manufacturers, through transmittal to the input user, Teledyne Engineering Services (TES). Control of inputs such as the following were included: o Response Spectra [ o Design Specifications / Requirements / Conditions 4 o Manufacturer's Data f Control of inputs to pipe stress analyses for other piping systems was examined when necessary to provide sufficient basis to justify conclusions. J Ld P w. ,~ s

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L - 5.3 APPROACH 1 5.3.1 General ..at The first stage of the audit was a pre-audit conference with GAI. (GAI is the Architect / Engineer (A/E) for the V.C. Summer Nuclear Station). The pre-audit conference was held May 19, 20 and 21. In addition to presenting _.3 the purpose and. scope ' of the. audit, the conference was used to gain ~- unt erstanding of the GAI organization and procedural program. The flow. of design information relating to pipe ~ stress analysis was g discussed in detail. Based on this discussion and a review of applicable + procedures, a preliminary flow chart depicting the basic flow of pipe stress analysis input was prepared during this period. This flow chart was used as a reference document during the course of the audit and was ~~j modified to reflect observed information flow. (See Figure 5-2, Sheet 1). ,o Sheet 2 of Figure 5-2, which is based on discussions with GAI and procedure review, is included for information only. The scope of the audit did not directly include pipe support design for, or field walkdown of, piping systems. Figure 5-2, Sheets 1 and 2, are simplified for clarity. Not all documents, procedures and feedback loops are shown. 1 l d The GAI quality assurance program document hierarchy is ' shown on Figure - 5-3.. The procedures most directly applicable to the audit were contained in the Project Management Manual (PMM) and Design Control Procedures (DCPs). The Reference PMM and the DCPs underwent major restructuring in 1977. The restructured program was invoked. on the V.C. Summer Project in November d 1981. As explained by GAI, many of the changes in the program dealt,mainly with format. The significant difference in the program was the increased requirements for controlling design verification (e.g., design verification l status reports). The present program requires that designs be verified - prior to installation except piping design (pipe stress analysis and pipe support design). Verification of these designs may be performed af ter ., _i installation, but prior to fuel load. This exc.eption is previded for in the PMM. Discussion at the pre-audit conference included clarification of the interface between GAI and TES. South Carolina Electric and Gas Company contracted with TES to perform pipe stress analyses using inputs supplied by GAI. The interface between GAI and TES is controlled by an' interface ..y procedure (an appendix to the Project Management Manual for the V.C. Summer Proj ect). a 3 Following the pre-audit conference, an audit checklist was developed. The ,3 checklist questions developed were basically of two categories: 1. Questions that related to tracking a specific pipe stress analysis

l. j input from its source (e.g., a GAI calculation or a vendor drawing) to a

its use in pipe stress analysis. El S 5-6 j 9 i .+ e ~. e -w-

\\,, y N with a specific control aspect (e;g.i. ~ 2. General questions that. dealt M control of specifications or control of vendor drawings). A-Initial avamination of evidence began at GAI on May 24, 1982 and continued s until May 28, 1982. On May 28, 1982, a status meetir.g was held with GAI. The purpose of the meeting was to.addse GAI that the major part of the ~ audit was complace but the audit would resume in approximately one,veek 4 after SWEC had time to evaluate audit results to date. After 'this evaluation was complaced it was determined that additional documentation should be examined. This was accomplished on June 9, 1982. On June 10, 1982 a post audit conference was held to present the audit results. ? s Audit Participants, including attendees at the pre-audit conference, status meeting and post-audit conference, are identified on Attachment 1. a on W .%) r.1 9 s 4y .*si e. g

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6d FIGURE 5-3 QUALITY MANAGEMENT PROGRAt: u DOCUMENT HIERARCHY 9 U CORPORATE PHILCSOPHY ,i I ~ ,l CUALITY g MANAGENENT 4 MANUAL '.1 l l l ~~ h NUCLEAR ACn!NISTRATIVE OUALITY NANUAL NON-NUCLEAR CCR* CRATE ASSURANCE (IF RECUIREDj OUALITY

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MANUAL ASSURANCE NANUALS y, NANUAL 1 .a CUALITY ASSURANCE i I TOPICAL REPORT d DESIGN PROCUREMENT CONTROL CONTRCL y , PROCEDURES PROCEDURES .) I REFERENCE I w PROJECT CIVISION l .'.3: . MANAGEMENT J LEVEL Q MANUAL PROCE'UP.ES m ,) i 4J INDIVIOUAL l 7 PROJECT PRCJECT !. f-MANAGENENT LEVEL ' d PRCCEDURES nANUAL 1 "J 4, l ENGINEERING PROC'REMENT J T AND ORAFTING SERVICES DEPARTNENT .5 STRUCTURAL PURCHASING . INSTRUCTIONS ELECTRICAL EXPEDITING SI AN ANUA,RC3 MECH /NVC SPECIFICATICNS l sU.Cc: 3" IEC EILLS OF BUILDING SERVICE MATERIAL a ETC. i '} I o i

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'.J: 'l 5.3.2 ' Examination of Evidence 7 J 5.3.2.1 Inputs to Pipe Stress Analvsis The initial planned approach and sampling are discussed below. In some n) cases it was necessary to select additional items; these cases are discussed in Section 5.5, Results. 3 Response Spectra (OBE) The piping isometric drawings (isemetrics) .:3 identify the response spectra to be utilized by reference to a Figure Number in GAI Specification SP-702-4461-00 (Specification 702). The figures contained in Specification 702 were compared to the retults of the dynamic analysis (computer printout section of GAI calculation). This was i accomplished by comparing one or more points (at a peak or peaks) on tha curve in the specification against the dynamic analysis results. d. It was then determined if the spectra (Figure Nos.) called out on the isometrics were appropriate, considering piping location (building, q elevation). a The TES pipe stress M ysis packages (EF-01,02,03 and 22) were then reviewed to determint a.i the spectra identified on e the isometries were utilized in the ana'. sis.- This was determined by comparing the spectra listed on the input section of the stress analysis packages against the isometrics. 9 y All response spectra identified as applicable to the turbine driven pump portion of the Emergency Feed Water System (subsystems EF-01, 01, 03 and

22) were compared in this manner, mq 0

DBE Factors - DBE response spectra are obtained by. f actoring the OBE spectra by a value that is dependent on the pipihg location (building). The DB2 factors are contained in Specification 702. The DBE factors in Specification 702 (for the buildings of interest) were q compared to the factors in the FSAR for the same buildings. The factors '1 utilized in the pipe stress analysis (as evidenced in the input section of the analysis) were then reviewed to determine if they were appropriate for the location of the piping analyzed. y.j All DBE factors associated with subsystems EF-01,02,03 and 22 were tracked in this manner. c-, ,j Dampina Factors The response spectrum for a particular building elevation, and direction is represented by several ' curves, each corresponding to a damping factor. d r.2 The damping factors utilized in the pipe stress analysis packages (as evidenced in the input section of the analyses) were compared to those ,q provided by GAI. The damping iictors provided were also compared to the '3 FSAR. This methodology was applied to all subsystems within the scope of the 3 audit. d 5-11 m

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.J Design Conditions - A sample of " design conditions (one temperature and ~~. three pressures) was selected from GAI Design Specification DSP-544C-

j 044461-000 (Specification 544C) and compared against the source document (e.g., GAI calculation or k 'tinghouse Design Specification).

The input !q section of the two pipe stresa analysis packages that would utilize these ,g inputs were reviewed to deter =ine if these design conditions were evaluated. O Anchor Movements - Anchor movements (thermal and seismic) are identified on d the isometrics. A sample of three sets of seis=ic movements and one set of thermal movements were selected from the isometrics and compared to the results of GAI calculations. These movements were also compared to the M values utilized in the pipe stress analysis. Since the input section of the pipe stress analysis package did not list anchor movements, it was necessary to review the echo print from the computer run to determine values used. .) Jet Loadings - A sample of jet loadings (five load cases) transmitted by a GAI to TES, were selected and compared to the results of GAI calculations. ..i The echo print of the computer run associated with the pipe stress analysis was reviewed to deter =ine if the transmitted loadings were utilized. ] Piee Meterials A sample of pipe materials was selected from the 1sometrics consisting of material sizes and schedules in the run of piping between the turbine driven E=ergency Feedwater Pump to the Steam Generator. G Comparisons were =ade between the isometric, design specification and j piping drawing to determine if the materials and sizes were consistent. Since the pipe stress analysis did not define materials but material properties, the modulus and allowables in the pipe stress analysis packages id were compared to those specified in ASME Boiler and. Pressure Vessel Code. Valve Weights - A sample of ten valves was selected from flow diagrams. ] The weights of the valves, as deter =ined from the suppliers' drawings, was L compared against the valve weights listed in the system design specification. The weights were also compared to the values used in the pipe stress analysis 'oy reviewing the echo print of the computer printout j associated with the pipe stress analysis. Valve Center of Gravity (cgs) - Valve cgs are shown on the isometries. The l same sample of valves used to compare valve weights was used in comparing i l a valve cgs. j e; The CG locating dimensions shown on the isometric were compared against the 1 "'! supplier drawing for each valve. By reviewing the echo print of the computer printout associated with the pipe stress analysis, the locating dimenss.ons shown on the isometric were compared to the dimensions utilized j in the pipe stress analysis. a Nozzle Loadings - A sample of components was selected (steam generator, ' y] containment penetration, and two pumps). The allowable loads on a nozzle or nozzles of the components were determined from the procurement specifications for the component. These values were connared against the calculations contained in the nozzle load summary section of the pipe W stress analysis packages. = 5-12 3 ~ m, ~

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• I J

5.3.2.2 Control Methods t r J Certain aspects associated with control of pipe stress analysis inputs through transmittal and utilization were evaluated. The first step was to O determine if procedures were available to cover these aspects.

Secondly, d

the procedures were reviewed to determine if basic elements were addressed. In addition, documentation was examined to determine if specific areas of the procedures were consistently implemenced. ,- i . \\i Vendor Drawina Control - Applicable procedures were reviewed to determine if the following were addressed: q

3 o

Receipt control (logging / indexing) Review and approval by appropriate personnel ,g o o Distribution to appropriate personnel 9 o Retention / filing .J The index of vendor drawings was reviewed to determine if it is maintained g* ) up-to-date and that the listed drawings were clearly identified. Specification Centrol - Applicable procedures were reviewed to determine if the following were addressed and evidence was examined to determine U.f implementation: o Distribution of specifications (including revisions) to appropriate !?. personnel 33 o Maintenance and distribution of' indexes 8.T M In addition, the specification issues identified on the index were compared d against those issues identified on isometrics and against those issues transmitted for use in the pipe stress analysis. d Drawina Control - Procedures were reviewed to deter =ine if the following were addressed and evidence was examined to determine implementation: .5 o Distribution of drawings (including revisions) to appropriate personnel 3 o Maintenance and distribution of indexes .3 l In addition, the isomeeric' revisions, identified on the index, were l .g compared against the 1.sometric revisions referred to in the pipe stress .( analysis packages. Change Mechanians GAI utilizes several advance change mechanirms. ,.j However, only one, the Engineering Change Notice (ECN) system was evaluated as since only ECNs w(re used to make changes to isometric 5, p The procedure was reviewed to determine if the following were addressed and 3 evidence was examined to deter =ine implementation: o Identification on the ECN of affected documents. ) l m 5-13 13 ~ - - -

• .a

..J - o a method of identifying EC3 status and tracking the change through ~ ,j incorporation into the affected documents. Interface Control (GAI/TES) The procedure establishing the. interface ..j requirements between GAI and TES was reviewed to determine if methods for N-transmitting information were provided and implemented, including subsequent changes to previously transmitted information. G .i One of the interface con:munication mechanisms is the use of a for= called " Request for Information" (RFI). Methods for identifying and logging RFIs were examined. ~3 Computer Programs - Procedures were reviewed to deter =ine if the following were addressed and evidence was examined to determine i=plementation: m h o Requirements for utilization of verified / certified computer programs. o Methods for identifying / tracking the use of computer programs that are j not verified / certified to ensure later verification / certification. Design Verification Procedures were reviewed to determine methods 'l utilized in performing and documenting design verification. Evidence was examined to determine implementation. 3 o .k 6 4 e 'b _3 I ( =J b 1 9 I 5-14 m \\ )U e em .g - - - - ~ + - = = = = - * * * *

• N-5.4 EVIDENCE EXAMINED The following is a list of the major documents examined during the audit:

PROCEDURES q Project Manasement Manual (October 1971 - November 1981): J 6.6 Manufacturer's Drawings 7.3 Design Review and 7erification 8.1 Drafting Interface Information 1 10 Design Changes 13 Schedules Appendix 5A As-built Piping Verification - GAI and Subcentractor Interface Control Docu=ent Project Management Manual (Effective November 1981): 6.07 Vendor's Drawings 7.19 Design Verification y 7.20 Vendor's Drawings and other Documents 8.0 Document and Record Control 9.0 Design Changes Appendix 7A As-built Piping Verification - GAI and Subcontractor ',- j Interface Control Document L: Design Control Manual (DCRs},(Old): ~ 1.5 Design Ccarrol Progra: 3.2.1 Identification of GAI Drawings 3.4.1 Vender's Drawing Control i-3.6.1 Design Verification L. 3.12.1 Computer Program Development and Maintenance 3.12.2 Computer Program Verificatien/ Certification '~i 4.2.1 Design Analyses / Calculations 4.3.2 Design Specifications q. Design Control Manual (DCPs) (Effective November 1981) (New): 2.05 Design Verification 3.05 Vendor Drawings 4.15 Procurement Documents l

• h Office Procedures:

7j, 10.1 Correspondence Action Control 10.2 Distribution of Project Documents Specifications and Bills of Materials Department Instrue:1ons: l Instruction 1.4 - Setting Up the Specification Program '/- Instruction 1.5 - Developing and Maintaining the Engineering and Purchasing .l. Schedule n Other: O Piping Engineering Standard DS-8, General Procedure for Design Verification Computer Applications Manual (CAM) Ji SCE&G Quality Control Procedura MF-14 .._t u 5-15 '4, .:.) ~...... --q-e

---

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o; ANALYSES / CALCULATIONS / REPORTS .? TES Technical Reports (Pipe Stress Analvsts Packages): Technical P. aport TR-4813-8, Rev.1, Strass Analysis and Support Load Summary of i Emergency Feedwater Subsystem EF-01 Piping for Virgil C. Summer Nuclear Power a Plant April 22, 1982. Technical Report TR-4813-9, Rev. 2, Stress Analysis and Support Load Summary of Emergency Feedwater Subsystem EF-02 Piping for Virgil C. Summer Nuclear Power Plant, April 23, 1982. Technical Report TR-4813-10, Rev. 1, Stress Analysis and Support Load Summary of 4 Emergency Feedwater Subsystem EF-03 Piping for Virgil C. Summer Nucient Power Plant, April 23, 1982.

• j Technical Report TR-4813-15, Rev.1, Stress Analysis and Support Load Summary of Emergency Feedwater Subsyste.n EF-22 Piping for Virgil C.

Summer Nuclear Power e5 Plant, April 22, 1982. Yt GAI Calculations: File Code 2.9.2 File Code 2.4.3.13 File Code S-14:05 t File Code S-14:01 Q File Code S-14:06 File Code EF-01 File Code EF-02 9 File Code EF-03 .IJ File Code EF-21 File Code EF-22 I; File Code IF-02 and 03 a GAI Reports: [8, Report No.1902, Jet Loadings on ASME Section III Piping, dated 1/77 ~d1 Report W.O. Number 04-4461-000 dated 9/10/81 3

• 1*

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9 VENDOR DRAk'INGS ~$ IMS-50181-3 Fischer $4A 7509-D ..J DfS-50176-4 Fischer $4A 7513-D ~ Dis-25-242-4 Adefior Darling 3342-3-D IMS-25-602-1 Anchor Darling 93-14530-A IMS-25-222-3 Anchor Darling 3316-3-C -s ~~ IMS-25-276-7 Anchor Darling 3379-3-C IMS-25-273-3 Anchor Darling 3317-3 IMS-25-224-3 Anchor Darling 3318-3 IMS-25-695-1 Anchor Darling 93-15061-A IMS-06-083-0 Pathways Bellows D-50-1776 I' r~. la' e5 J R

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7 EL SPECIFICATIONS O d DSP-544C-044461-000 Rev. 5, Design Specification Emergency Feedwater System Piping and Pipe Supports (ASE III Division 1 Class 2 & 3) O

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DSP-508A-4461-00 Rev. 2, 7/8/77, Design Specification Motor Drive E=argency Feedwater Pumps (ASE III Class 3) m 'j DSP-508B-4461-00 Rev. 2, 4/2/76, Turbine Driven E=ergency Feedwater Pumps a.s DSP-606-044461-000 Rev. 9, 2/1/82, Design Specification for Reactor Building Piping l} Penetrations (ASE III Div. 1, Class 2) SP-545-044461-000 Rev. 17 11/25/80 - Pipe Line Specification for Nuclear Safety Class Piping

.)

3 SP-702-4461-00 Rev. 4 2/11/77 - Seismic Analysis, Testing and Documentation. 0.] Westinghouse Steam Generator Specification, dated 11/3/80, Revision 6 3 9d a el

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I DRAWINGS / DIAGRAMS Isometric Drawings: C-314-085 Sheet i Revision 2 e, C-314-085 3 2,3 C-314-085 2 2 C-314-085 4 2 C-314-085 5 2 C-314-085 27 3 Flow Diagrams: ~ D-302-083 Revision 17 D-302-085 Revision 13 ~. Piping Drawings: E-304-087 Revision 6 E-304-090 Revision 4 .n. ~~ r*. v ~~t p.4 .a .m r 22 r-5 r n e- ~: e- , L*, .s.i e~ '. 's 5-19 M. j

.. ? MISCELLANEOUS Engineering Change Notices (ECNs): ECNs - 1891, 2134, 2143, 2202, 2206, 2219, 2230 Recuest for Information (RFIs): RFIs - 1240, 1241, 1242, 1243. 1244, 0131, 0180, 0152, 0140, 0017, 0195, 0082, 0285 GAI Letters of Soecification Transmittal: CGGS - 10760 (2/23/77), 4815 (6/30/75), 1819 (1/21/74), Number not Recorded (5/17/72), 24079 (7/23/81), 22117 (12/8/80), 19996 (2/26/80), 16207 (8/2/78), 6936 (3/15/76), 5196 (8/15/78), 23248 (4/13/82, 20886 (7/3/80), 4 18654 (6/7/79), 13095 (8/23/77), 6209 (2/5/76) 7 Other: \\ ~~ ASME III 3o11er and Pressure Vessel Code, Section III (1971 Edition, Summer 1973 Addendum) U FSAR, V.C. Summer Nuclear Station -] Certification of Compliance for Turbine Driven Pu=p L: Manufacturers Print Index Activity List dated 4/14/82 ,,j Engineering Change Status Report dated 5/13/82 GAI letter CGGS-2341'1/CGGT-0048 of 4/27/81, transmittal of infor=ation to TES O j GAI letter of ' October 10, 1980, transmit +.a1 of documents to TES 1 GAI letter CGGS-22092/CGGT-0014of 12/1/80, transmittal of information to TIS ,i.1 Request For Infor=ation Log (.) TES letter 4813-159 of 5/26/82, Confirmation of Specification revisions Specification and Drawing Distribution Control Forms for specifications '] DSP-544C, SP220, and SP702 .3 ej d 9 8) 5 Ela Gd

,~: 1 a 5.5 RESI!I,TS ij 5.5.1 Inputs to Pipe Stress Analvsis Resconse Soectra - For all response spectra reviewed, the spectra figures N i contained in GAI Specification 702 agreed with the results of the GA1 dynamic (structural) analysis. The response spectra called out on the isometrics were appropriate for the s piping location except that it did not appear necusary to utilize Figure 64 (elevation 463 of inter =ediate building) as specified on the isometric for subsystem EF-01 since the piping did not appear te extend to the elevation represented by this figure. This was confirmed as indicated below. ,f The response spectra utilized in the analyses were consistent with those i d specified on the isometrics with the following exceptions / comments: i ] 1. The isometrics covering subsystem EF-01 specifies the enveloping of Figures 61, 62 and 64. The pipe stress analysis package indicates only Figures 61, and 62 were enveloped. There was ne docu=entation in the package to indicate why Figure 64 had not been used. An RFI - (TES-0082) from TES which addressed the ideletion of Figure 64 was located in the GAI files. This RFI was approved by GAI. However, the package did not reference the RFI nor was there evidence that TES had =arked up the ISO (as required by the interface procedure) to show the deletion of Figure 64 2. The pipe st'ress analysis package for subsyste: EF-02 indicates that Figures 7,8,61,62 and 64 are used in. the analysis. Figures 7 and 8 are for the Reactor. Building and Figure 64 is for elevation 463 of the Inter =ediate Building. Subsysten! EF-02 terminates in the Intermediate j Buil. ding and does not extend to elevation 463 of the Intermediate Building. Therefore, it appears that the use of Figures 7, 8 and 64 l was not necessary. 3. The pipe stress analysis package for subsystem EF-03 indicates that l Figure 64 (elevation 463 of the Intermediate; Building) was used in the ~* analysis (along with Figures 7, 8, 61 and 62). EF-03 piping does not j extend to elevation 463. Therefore, it appears that the use of Figure 64 was not necessary. ? NOTE: When an isometric depict: more than one subsystem all l.j applicable spectra figures are listed. The appropriate spectra for each subsystem is then selected from that list. m ~ ] DBE Factors - The DBE factors contained in Specification 702 agree with the U factors contained in the FSAR. The DBE factors utilized in the pipe stress analysis packages for subsystems EF-01, 02, 03 and 22 were appropriate for G the locations of the piping. ta d 5-21 i eeuf .;.3 -.3 .. ~....

u Damoing Factors - Amendment 26 to the FSAR specifies the following damping j factors: a CBE DBE j 12 inch or Smaller Piping 1.0% 2.0% j Specification 702 presents damping factors as: J Working Stress No More At or Just Below l Than About 1/2 Yield Point Yield Point Vital Piping Systemd 0.5% 1.0% ")'j The FSAR damping factors were used as the basis for pipe stress analysis. An additional factor, " gamma factor", is defined in Specification 702. The ~1 The gamma factor accounts for vertical flexure in certain slabs under a seismic conditions. To obviate application of the gamma factor, GAI performed a study (W.O. No. 04-4461-000, dated 9/11/81) that de=enstrated that the direct use of 0.5% vertical damping would account for the gamma q factor and meet FSAR requirements of 1.0% damping. This information was supplied to TES and the other subcontractor performing pipe stress analysis. Pipe Stress analysis packages for subsystems EF-01, 02, 03 and 22 all used damping factors of 0.5% vertical and 1.0% horizontal, which meets or exceeds FSAR requirements. Design Conditions The sample of design conditions selected from 'the design specification agreed with or were more conservative than the scurce documents and the pipe stress analysis packages used the design conditions. .s Anchor Movements - Of the four sets of anchor movements initially selected, l two sets as depicted on the isometries did not exactly agree with the GA2 j calculation. The movements are very small and the differences were negligible (e.g., 0.0722 versus 0.07064). In one of the GAI calculations, the verifier had noted the differences as negligible. Since the movements a were so small, TES (in the pipe stress analysis packages for EF-01 and 02) V documented that anchor movemerts were not considered in the analysis. O In the other two cases (one thermal, one seismic) the movements specified

• j on the isometric agreed with the GAI calculations.

The seismic movements utilized in the pipe stress analysis package agreed with the isometric. However, the thermal movements utilized in the pipe stress analysis package q*] (for EF-22) did not appear to agree with the isometric for x-direction u (-0.718 on isometric,-0.9645 in analysis) and slightly different for z-direction (-1.679 on isometric,-1.698 in analysis). 9 b Later correspondence initiated by the audit (TES letter dated June 23, 1982) indicates that TES had observed that the movements had been changed 3 on Revision 1 of the isometric, had considered the changes insignificant i1 and that remnalysis was not required. The fact that there wars differences d and that the differences were evaluated was not documented in the pipe stress analysis package. Is W 5-22 _d

~~) 1} Due to the above differences, an additional pipe stress analysis package from a different system was selectec for review; subsystem SI-09. The anchor movements utilized in the pipe stress analysis package did not agree with the anchor movements called out on the isometric. u.i The above sentioned TES letter of 6/23/82 states, 7 "The SAM displace =ents used in the analysis f or the SI-09 subsystem were obtained via a copy of Westinghouse letter, nu=ber CGWG-2290, dated February 18, 1981 from Mr. Ja es B. Cookinha= of Westinghouse to Mr. H.E. Yocom of GAI. This let.:er defined,the CBE Seis=ic Movements =! for a number of subsystems including SI-09. The copy was transmitted informally at the V.C. Sum =er Station during an infor=al =eeting for which no record could be located. It was Teledyne's understanding, at 7 the ti=e, that the C-314 isometrics for SI-09 would be revised to include these movements. This was not done and the discrepancy still exists between the drawings and the analysis. It is Teledyne's understanding that the SAMs used in the analysis are correct and, therefore, the C-314 iso =etrics should be revised to incorporate them.". ~j There was no docu=entation in the pipe stress analysis package for SI-09 to d indicate why the anchor =ove=ents utilized were different than the isometric values. In addition, there was no evidence that GAI had approved or trans=1ttad the =ove=ents utili:ed. I I d In each of the five load cases selected, the loadings Jet I.oadings transmitted to TES agreed with the results of the GAI calculations and, the loadings utilized in the pipe stress analysis package.s were consistent with ...) those trans=itted. (The values utilized in the analysis were twice.the valdes transmitted since a dyna =ic factor of 2'.0 was utilized). P'. Pipe Materials and Sizes - The co=parison of pipe materials, sizes and schedules between piping drawings, flow diagra=s and design specifications revealed correlation between input docu=ents. The allowable stresses and I modulus identified within the pipe stress analysis were in agreement with those identified within ASME Section III 1971 Editica, including Surmer 1973 Addenda for Class 2 piping. Valve Weights - The comparison of valve weights contained on the latest issues of vendor drawings, design specifications and pipe stress analysis were in agreement. .) The valve weights used in pipe stress analysis of subsystem EF-01 were 10: greater than certified weights since certified valve weights were not ~) available when the analysis was originally performed. This 10% margin was j consistent with the system design specification. l Valve Centers of Gravity (CG)s - The valve cgs shown on the latest issue of l "l vendor drawings agreed with the piping isometric. The valve cgs utilized in tha pipe stress analysis packages were consistent with the cgs shown on the isometrics, or if different were justified by GAI approved RFIs which l 3 were referenced within the pipe stress analysis packages. t 1 i 5-23 t ) .i

[ -s) Nozzle Loadings - Loadings on seven of the nine nozzles audited were less 7 } than the allowables loads established within the cocponent design specifications. .F Pipe stress cnalysis packages for subsystems IF-03 and EF-22 indicate that nozzle loads exceeded che established allowable loads.for Reactor Containment Penetration No. 213 (inside and outside ends) and for Motor ~ q Driven Pump XPP-21A-EF. Notes on the pipe stress analysis packages 1 indicate that the exceeded allowables are "ok by trade-off". However, the packages do not identify or reference what trade-off methods were used. An RFI (TES-0285) was located during the audit that discussed allowable load .i trade-offs. However, this RFI did not apply to subsystems EF-03 or EF-22.

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The allowable loadd evaluated for motor driven pump XPP-21A-EF were 7 compared for the DBE event rather than OBE event as required by t{ Specification 508A. When the allowable loads are compared to DBE load

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combinations the allowable loads are not exceeded. Therefore utilization of " trade-off" methods is not required for this case. O n ~

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~1J 5.5.2 Control Methods .:5 Vendor Drawings - PMM 6.6 (old), PMMs 6.07 and 7.20 (new), DCP 3.4.1 (old) and DCP 3.05 (new) establish the methods for control of vendor drawings. 9 These procedures provide for receipt control, review and approval, j distribution and retention. The Manufacturers Print Index Activity List of 4/14/82 was reviewed. This i index contained the GAI number, namber of sheets, revision, descrip tion, W vendor drawing number, date received, (purchase order nu=ber or system bill of material number). The index was consistent with all vender drawings examined in conjunction with valve weight and center of gravity input ]. comparisons. PMM 13.0 established the requirements for Specification Control -.j maintenance of an Engineering-Purchase Schedule. This schedule functions as an index for procurement specifications. Specifications and Bills of r8 Material Department (SBMD) Instructions 1.4 and 1.5 and Office Procedures ] (OP) 10.2 provide amplification such as distribution requirements. A .J procedure for maintaining a mechanical design specification index was in use but had not been formally promulgated. ]m Distribution of specifications and revisions was performed in accordance with procedures for a sample of three selected specifications, DSP $44C, SP ~ 702 and SP 220. 1! i L1 Comparison of revisions from the specification indexes to that called out on the isometrics and that transmitted to TES was conducted. It was noted N that Rev. 5 to DSP $44C dated 4/30/82 had not yet been formally transmitted to TES. The information contained in Rev. 5 wcs a reformatting of ~p previously provided data which would not affect the analysis. i C 'Drawina Control - PMM 8.1 (old), PMM 8.0.(new), DCP 3.2.1 (old), DCPs 1.30 I C and 3.20 (new), and OP 10.2 established the requirements for drawings and index distribution and maintenance. Indexes were updated and distributed in accordance with procedures. j The revisions to the isometrics used in the pipe stress analysis packages j were consistent with the isometric index with the exception that one sheet of an isometric series did agree with the index. The index had not been "d updated to reflect recent revisions of this sheet. The latest isometric 7 revision had been used in the analysis. An additional sample of five d controlled tracings was compared to the index. The issue numbers agreed. Interface - The interface between GAI and the subcontractors was formally = .i established by an interface procedure, PMM Appendix 5A (old) and 7A (new), which was contractually invoked by SCI &G. ] The input information was formally transmitted to TES by GAI. However, the '2 first transmittal of the input information did not clearly identify the revisica of all documents forwarded to TES. Subsequent correspondence and discussions with TIS confirmed that ;he latest revisions had been received.

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?.3 The Request for Information (FII) Log contained information such as RFI number, applicable subsystem, date reviewed, date answered, etc. The log ] was maintained in accordance with the interface procedure. i .3 5-25 C [.j

I 4 Computer Programs - The requirements for using verified / certified programs are addressed in the existing procedural program as are methods for ,J tracking the use of programs that have not been verified / certified. Computer program verification / certification requirements are established in the Computer Applications Manual (CAM). A " Design Verification Record" form is required by DCP 2.05, for each analysis. If an unverified / ~ ~ ' uncertified compu;er program is utilized, the Design Verification Record is annotated to indicate an assumption requiring later confirmation. This information is also reflected on the Design verification Status Report (DVER). The DVSR is a listing of open design verification ite=s and their current status. GAI calculation, file code 2.9.2 (verified 8/11/81), was evidence of implementation of this tracking method. The verifier recorded the use of a program requiring verification / certification on the Design Verification Record form. The DVSR appropriately reflected that the calculation used an W unverified / uncertified computer program. As required by the CAM, a list of certified computer programs is issued semi-annually. The latest listing was dated 4/17/82. Distribution of the list includes all holders of the CAM. r-, During the course of tracking pipe stress analysis inputs, the use of one J program for which there was no evidence of verification / certification and no direct evidence of tracking the prc' gram use was observed. (Three other computer programs used in analyses perfor ed in 1974 and 1980 had been verified / certified.) The c =puter progra= was identified as 50f1 (OA number) and was used in GA! Calculation S-14:01 for developing the response spectra for the Reactor Building. This calculation was perfor:ed in 1972 prior to any for=al require =ents for ccmputer progra= verification / certification. DCP 4. 2.1', issued October 1972' addressed the use of l verified / certified computer programs in analyses and DCP 3.12.1, issued October 1973, addressed computer program development and maintenance. According to GAI, this program (5051) had been tested but the material had i not been compiled into a for=al certification package. GAI was apparently i aware that formal verification / certification was required as evidenced by a GAI memo dated 8/6/80. Verification / certification of this program was completed during the audit. ~ As a result of this one instance, additional investigation was performed. Fourteen additional computer programs were selected. All had been verified / certified. However, due to the difficulty in reconstructing the historical usage of computer programs, especially usage circa 1971, GAI ~ performed a survey of all Departments to determine if any unverified / uncertified computer programs had been used in finalized safety related analyses for the V.C. Summer Plant. The results of the survey indicated no l such usage. Design Verification - PMM 7.3 (old), PMM 7.19 (new), DCP 3.6.1 ( old), DCF 2.05 (new), and Piping Engineering Standard DS-8 provided direction for controlling design verification activities. l l The detailed implementing procedure for piping design (includin.g pipe I " stress analysis) was DS-8. The stated purpose of the procedure was, "To 5-26 .... ~.

review, conform or substantiate a design by one or more methods in order to provide assurance that system design meets the specified design inputs and that these inputs were - selected in accordance with, appropriate design 4 criteria". - The implementation and documentation of the verification is by use of various forms and checklists. For example, form 2.3, Review of Analysis, includes: 2.3.1 Applicable Drawings-1. Have the latest revisicus been used as a source of , input for che analysis? z 2.3.2 Modeling 1. Is the system configuration as analyzed representative of the layout depicted by the latest information? .2 u 2. Has acceptable modeling theorf been utilized? The form continues and asks similar questions regarding: design }i conditions; static analysis; dynamic analysis; output; supports and restraints. C Implementation of design verification was evident in.all GAI calculations 1 and TES pipe stress analysis packages reviewed during the audit. l A major tool in controlling design verification is the Design Verification Status Report ~ (DVSR).. The DVSR is ' a listing compiled from information ~~ supplied by all disciplines that identifies all items (e.g., calculations) requiring verification. The DVSR identifies, for example, if a particular l,* item has been verified and if assumptions have been confirmed. The DVSR is a computer based information system that is up-dated on a l continuing basis. There is no specified frequency for issuing the DVSR but .j recent DVSRs were issued approximately quarterly. $-s =$8 E

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5.6 CONCLUSION

S i 5.6.1 Procedural Program An adequate Design Control Program, meeting the requirements of 10CFR50 Appendix B, was in place for the transnittal and utilization of input data for pipe stress analyses of subsystems EF-01, 02, 03 and 22 of the Emergency Feedwater Piping System. t Only one instance was observed in the existing program where there was no formally approved procedure. Although formal procedures were available for indexing of design and procurement specifi~ations,the maintenance and c ,I distribution of a mechanical specification index was performed using an undated, uncontrolled instruction with no evidence that the instruction had y been approved. Although unapproved, the procedure was adequate and was ] being implemented. In the early stages of the proj ect there were no formal procedures 'l governing the verification / certification of computer programs and their During the course of audit the use, in 1972, of one program for which use. there was no evidence of verification / certification was observed. (Three p other programs used in analyses performed in 1974 and 1980 had been i. verified / certified). This led to additional investigation. A review of additional program usage, procedures and tracking mechanisms indicates that the existing program does address this area and controls the use of computer programs. In addition, GAI conducted a survey to determine if any 1 other unidentified uses of unverified / uncertified programs had occurred; no other instances were revealed by this GAI survey. I'!.a e t ? '* ~ .. f I ~.1 ]

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.~ ..] 5.6.2 Program Implementation 7j The procedures associated with the activities reviewed during the audit were adequately implemented except that the utilization of inputs to pipe scrass analysis in some cases was not consistent with program .y requirements. The instances are apparently documentaticn problems that would not affect the design adequacy. o The pipe stress analysis package for subsystem EF-01 did not utilize Figure 64 response spectra as specified on the isometric. Although GAI had approved the deletion of Figure 64 in an RFI there was no evidence that the isometric had been marked-up to indicate that Figure 64 should be deleted nor was there documentation in the pipe strese analysis package that justified the deletion of Figure 64 (such as by reference to the GAI approved RFI). ,l o There was no documentation in the pipe stress analysis package for EF-22 tha.t the differences between the thermal movements utilized in .q the analysis and the movements on the isometric had been evaluated. A letter to GAI from TES initiated as a result of this audit indicated that the differences had been evaluated when the analysis was perfor=ed and that reanalysis was not necessary. i' o The pipe stress analysis package for subsystem SI-09 apparently utilized anchor =ove=ent infor=atics from a Westinghouse letter rather than the =ove=ents ida.tified on the isometric. There was no evidence that GAI had approved or trans=itted this infor=atit for use. In ~ addition, the pipe stress analysis package did not identify that the movements utilized were different than the isometric and the. reasons i for the differences. A letter submitted by TES to GAI after the audic indicated that the Westinghouse ' anchor movement information had been used in the analysis, a.1 o The nozzle loadings in pipe. stress analysis packages were noted as acceptable by " trade-off". There was no docu=entation in the pipe stress analysis packages that identified the method or the acceptability of the method. There were approve.d RFIs in GAI files that addressed' load trade-offs, but they were not referred to in the packages. [ Another area that was not clearly documented was the application of damping factors. Although the application of damping factors complied with the FSAR, this could not be discerned unless reference was made collectively to [~ the FSAR, Specification 702, pipe stress analysis packages, a GAI study, U and minutes of a meeting. The underlying cause of this condition was apparently due to not updating Specification 702 to reflect the issuance of l q Amendment 26 to the FSAR. .i The response spectra utilized in the pipe stress analysis were consistent i with the dynamic (structural) analysis output. In some cases additional l 'y spectra were utilized when it did not appear necessary. Utilization of these additional spectra adds to the conservatism of the design. ^1 l ..T 5-29 b ~ d z .; - -~

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. - ~.. Ci .) 5.6.3 Recommendations 7 ..{ Procedures A procedure governing the preparation and distribution of a specification ',j index for mechanical specifications (and for other discipline e specifications if necessary) should be formalized as part of the project program. Oj_ Implementation The extent of incomplete documentation in pipe stress analysis packages 'l should be determined and appropriate corrective action implemented.

• J To preclude future misunderstanding and provide clear traceability R

regarding application of damping factors, currective action, in the form of j either a revision to Specification 702, or a.nemorandum of explanation in the pipe stress analysis packages, or other appropriate equivalent, should m be performed. 34 d' bd 77 p 'f 7 .w .g q a 'E L ~ Se q ij ie

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, *O .2 ATTACHMENT 1 TO SECTION 5.0 AUDIT PARTICIPANTS _s PRE-AUDIT CONFERENCE ACENDEES (MAY 19, 1982) CAI ~ G.J. Braddick Project Manager K.R. Gabel Project Engineer J.R. Helwig Project Control Engineer D.R. Kershner Piping Engineer (and Primary Contact during Audit) H.A. Manning Quality Assurance Progra: Manager F.L. Moreadith , Manager of Engineering J.B. Muldoon Department Manager, Specialty Engineering C.C. Paschall Manager, Design Control

• C.N. Rentschler IJ Piping As-Built Verification Task Manager K.W. Sandman Project Piping Support Designer SWEC J. MacKinnon Design Control Audit Manager D.L. Malone Audit Tean Leader R.W. Twigg Auditor
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. a :........ L -_ ... ~..... - ~ ~.. -~.- - i s t GAI PERSONNEL CONTACTED DURING EXAM!"ATION OF EVIDENCE j R.S. Chang R.F. Ely J.R. Helwig D.R. Kershner j G. Khurshudyan J.E. Lisney q H.A. Manning J.B. Muldoon C.C. Pascha11 J.W. Reitnauer 5: R.J. Sheldon .e

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~ STATUS MEETING ATTENDEES (MAY 28. 1982) GAI s..: N.R. Larker Vice President & General Manager QA Division F.G. Boutros Manager, Nuclear Section G.J. Braddick Project Manager C. Chen Manager, Structural Department E.C. Goodling Section Manager, Piping Stress Analysis J.R. Helwig Project Control Engineer .s D.R. Kershner Piping Engineer J.E. Lisney Structural Project Engineer E.A. Manning QA Project Manager W.E. Mack Vice President, Projects F.L. Moreadith Manager of Engineering J.3. Muldoon Manager, Specialty Engineering ?3 C.C. Paschall ' Manager of Design Control C C.N. Rentschler Section Manager, Pipe Support Design R.J. Sheldon Mechanical Engineer I ? Jt SWEC e .g P. Dunlop Project Manager

i. 3 J.H. MacyJ nnon Design Centrol Audit Manager d

D.L. Malone Audit Team Leader R.W. Twigg Auditor r i o lC

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c 4 a FOST-AUDIT CONFERENCE ATTENDEES (JUNE 10, 1982) 1 GAI ) F.G. Boutros Manager, Nuclear Section 1! G.J. Braddick Project Manager J K.R. Gabel Piping Project Engineer J.R. Helwig Project Control Engineer Vice President E.K. Hess D.R. Kershner Piping Engineer '~ J.E. Lisney Structural Project Engineer H.A. Manning QA Project Manager J.B. Muldoon Manager, Specialty Engineering C.C. Paschall }pnager of Derign Control W.F. Sailer Manager, Program Management QAD Division 4.:. Sheehan, Sr. Manager of Prcjects R.J. Sheldon Mechanical Engineer S'n*EC 3 J.H. MacKinnon Design Control Audit Manager D.L. Malone Audit Team Leader 7 .a ~~ a i l e i t J

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. -l t s e J w ..i ,,,, v '- .g s. s.. 4 s. ..k I J l ..u APPENDIX A: STATUS REPORT JULY 9, 1982 ~1 J e M a 1 F.*me* M f.) t1 . k id 'mt e C s.5u .=g we 6 'd.

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..c . n.. ~ ... :, :.rc. g e n .s 4 s _, 5 g A. 1 1 \\ . I s r L.e N 3 s, w q. s s 4 . ~, ( j c e y 5] .) -t J, ss e 7 g\\ INDEPENDENTJDSMIC DESIGN VERIFICATION \\ 3 l' URBINE DRIVEN SECTION 1 ? w 7 N "s ;, EMERGENCY FEEDWATER SYSTEM d-h;% ' _V.C. SUMMER NUCLEAR STATION .s a .= 5' STATUS 'AEPORT: JULY 9, 1982 A 4 e,. - t r.g i u l prepared for -'i 1 g. ELECTRIC & GAS COMPANY SOUTH CAROLINA j 0, .t 2 m A ed

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== ~~ ~ .3 i 1 t 1. INTRCDUCTION: i 1.1 GENERAL SCOPE 1 Stone & Webster Engineering Corporation (SWEC) was engaged by South Carolina Electric & Gas Company (SCE&G) to perfcrm an a independent review of the seismic design for the Turbine Driven portion of the Emergency Feedwater System at V.C. Summer Nuclear Station, Unit No. 1. The review consisted of three major tasks, specifically; 8 1) Field Walkdown:Verific'ation of the as-built piping i L configuration 2) Stress Analysis and Evaluation: reanalysis of the as-built T-piping system, review of stresses and support loads, and 3) Design Control Audit: review of the design control pro-cedures and implementation thereof by Gilbert Associates 'Incorperated (GAI), the designer of V.C. Summer Nuclear Station, Unit 1. a 1.2 STONE & WEBSTER QUALIFICATIONS AND INDEPENDENCE i j SWEC has extensive experience in the engineering,. design, con-struction and startup operations for nuclear power plant projects as well as special expertise involving seismic design analysis, a field verification efforts, and pipe stress and support reanalysis required by recent NRC I&E Bulletins. SWEC also has extensive experience in Quality Assurance aspects of the nuclear power in- 'l dustry and in auditing of large highly technical and complex pro-jects. Stone & Webster is justifiably.p'roud of its record and large staff of capable and experienced personnel. a Jt SWEC, its parent company Stone & Webster, Inc., its affiliated companies and all personnel assigned to this evaluation are in-dependent of South Carolina Electric & Gas Co. .4 Work performed by Stone & Webster and its affiliated companies for SCE&G represents only a miniscule portion of Stone & Webster's a business. All key technical personnel assigned to the project signed disclosures (Attachment.1-1). Table 1-1 lists personnel + ~ assigned to the various tasks. Dr. P. Dunlop, Project Manager, has overall responsibility for the project. Dr. K. Y. Chu is 1,, Project Engineer responsible for the technical evaluation (Tasks 'J l and 2) and is independent of Mr. J. H. MacKinnon who is respon-sible for auditing the GAI design control program (Task 3).

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1.3 EVALUATION PROCESS All work was performed in accordance with project procedures m (Table 1-2). Whenever a reviewer noticed anything outside the ' criteria, or had any question about the information or data,-the ~' reviewer highlighted this. Specific procedures for highlighting 1 questions were different for each of the three major tasks and are explained in the task specific project procedures (Table 1-2). 7 -1.3.1 Field Walkdown (as-bu'ilt verification) All field measurements were recorded directly on piping isometrics. Whenever the measured values differed from the isometric values by more than the criteria, presented in VCS-1 Field Walkdown Procedure,- u the recorded values were circled on the isometrics and also recorded ,~ on Difference List (DL) Forms. Copies of the marked-up isometrics 9 and DL forms were provided to SCE&G at the end of the Field Veri-t fication Effort. 1.3.2 Stress Analysis and Evaluation g7 All analyses are to be performed in conformance with VCS-3, Analysis and Evaluation Procedure and VCS-4, Analysis and Evaluation Criteria. These provide the procedures and criteria for perfer=ing the piping reanalysis. Procedures for highlighting differences are fefined in Procedure VCS-3. Questions raised by the stress analysts are for-mally recorded and resolved. A two step procedure is used. An Open Item Report (OIR) is initiated fer all items requiring clarification or confirmation.. If a satisfaciary resolution is received, the'OIR is formally closed out. If a possible erro~r or inconsistency is confirmed a Potential Discrepancy (PD) is written. These PD's will be fermally transmitted to SCE&G for their review and evaluation. 1.3.3 Desien control Audit ~ Of the three tasks the procedures and resolution of items for this task are more subjective. The personnel assigned to this effort were certified auditors and performed the audit in conformance with gen-7 l L-eral Stone & Webster standards for such audits. i 7 2.0 PROGRAM STATUS e t ~' As of July 9, 1982, SWEC has completed Tasks 1 and 3. Task 2 is currently in progress. To date nothing has been found which would ~ require the initiation of'a 10CFR21 review. The detailed status of each task is given below. e ~; Task 1: Field Walkdown - verification of the as-built piping jj geometry. This task has been completed and all Difference List (DL) items have been forwarded to SCE&G for their review and informatio'n. The following is a brief description of the differences identified. y ~, A* l 2. J.6 p. g-- --y 7 y w y -, ..e w-sw e-m-ew -w = w-- w-v- r

(1) Gaps between piping and support steel larger than criteria - two occurrences. The largest of these was 9/32 inch whereas the criteria allowed only 5/32 inch. (2) Clearances between piping and structural components - three ~ occurrences. Two instances of small clearance between pipe and structural component (0 and 7/64 inch) to be reviewed ~ ~ during stress analysis. A sleeve through a wall was also found to be partially grouted. This was subsequently deter-mined to have been identified by SCE&G (ECN 2316) and the grout had been removed when SWEC field personnel again visited the + site on June 7,,1982. 7 (3) Struts at angles other than identified on the isometrics - three occurrences of struts more than 3 degrees from the values on the isometrics. The umximum difference was 11 degrees. ~ (4) Dimensional data outside the criteria specified for SWIC's J-~ field walkdown effort - 15 occurrences. The maximum difference was 5.3 inches for a span of 11.6 feet. All dimensional y differences were within SWEC's standard criteria. ~ (5) Drafting Errora - five occurrences. These were confirmed by reviewing the support or piping drawings. i All field measured dimensions will be input to the stress analysis in Task 2. Any impact.of the above o,n the stress analysis will therefore be obtained.. Ci Task 2: Stress Analysis and Evaluation - reanalysis cf the piping m system with as-built geometry, comparison of pipe stress with allow-ables and support loads. Thi's task which consists of codin'g piping / support geometry and design criteria.into the NUPIPE program is ^ currently in progress. No detailed results have yet been obtained to compare with the piping allowable stress or with the original .2 design loads for supports. Three inconsistencies were identified during correspondence with GAI relative to design criteria. These are: (1) During the field walkdown and subsequent data review it was found that several supports on subsystem EF-01 were in the i,J Diesel Generator Building. This subsystem therefore should be analyzed considering seismic response spectra from the Diesel Generator Building. The piping isometric does not indicate ~; this requirement. (2) During reivew of data received an inconsistency in jet orientation and jet location was identified. 4.,

M (3)

In one instance the target area of a jet impingement in t the design document (1902) appeared to be inappropriate. Subsequent communication indicates that the jet need not be included in the analysis because shield installation negates 4 this break load. ? G 3. d . - 2: - ~".-- ,. ~ ~~"

It is not known what impact these inconsistencies'might have on the detailed stresses and support loads in the piping reanalysis. This ~~ ,,j task is expected to be complete by July 27,, 1982. Task 3: Design Control Audit - This task consists of three parts. These are: J (1) Review of the GAI design control program (2) Verification of program application >^ 4 y (3) Confirmation that the structural dynamic analysis output was consistent with response spectra provided.to Teledyne Engineering Service (TES) for analysis of the turbine driven portion of the Emergency Feedwater System. The above three parts of this task have been completed. The following are SWEC's conclusions based on the design-control audit. Procedural Program An adequate Design Control Program, meeting the requirements of 10CFR50 Appendix B, was in place for the transmittal and utili-zation of input data for pipe stress analyses of subsystems EF-01, 02, 03 and 22 of the Emergency Feedwater Piping System. Only one instance was observed in the existing program where there was no formally approved procedure. Although formal procedures were available for indexing of design and procure-ment specifications, the maintenance and distribution of a mechanical specification index was performed using an undated, uncontrolled instruction with no evidence that the instruction had been approved. Although unapproved, the procedure was adequate and was being implemented. Program Implementation The procedures associated with the activities reviewed during the audit were adequately implemented except that the utilization of inputs to pipe stress analysis in some cases was not con-sistent-with program requirements. The instances are apparently documentation problems that would not affect the design. The pipe stress analysis package for subsystem EF-01 did not utilize Figure 64 response spectra as specified on the isometric. Although GAI had approved the deletion of Figure ..j 64 in a request for information (RFI) there was no evidence -h 9 4. '.3 1,

• SMS 4

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...._....1 1 that the isometric had been marked-up to indicate that ^ Figure 64 should be deleted nor was there documentation in. the pipe stress analysis package that justified the I deletion of Figure 64 '(such as by reference to the GAI W approved RFI). There was no documentation in the pipe stress analysis o 2 package for EF-22 that the differences between the thermal movements utilized in the analysis and the movements on the isometric had been evaluated. A letter to GAI from TES 't

• initiated as a result of this audit indicated that the differences had,been evaluated when the analysis was per-formed and that reanalysis was not necessary.

o (The project scope was expanded to include SI-09 because of the difference noted in EF-22 above). The pipe stress analysis package for subsystem SI-09 apparently utilized anchor movement m i information from aWestinghouse letter rather than the movements identified on the isometric. There was no evidence that GAI had approved or transmitted this information for use. In addi- .J tion, the pipe stress analysis package did not identify that 3 the movements utilized were different than the isometric and the reasons for the differences. A letter submitted by TES to GAI as a result of the audit indicated that the Westinghouse anchor movement information had been used in the analysis. .3 The nozzle loadings in pipe stress analysis packages were noted o as acceptable by " trade-off". There was no documentation in 'a the pipe stress analysis packages.that identified the method or the acceptability of the method. There were approved RFI's in GAI files that addressed load trade-offs, but they were not referred to in the packages. Another area that was not clearly documented was the application of damping factors. Although the application of damping factors complied with the FSAR, this could not be discerned unless reference was made collectively to the FSAR, Specification 702, pipe stress analysis packages, a GAI study, and minutes of a meeting. The underlying cause of this condition was apparently due to not updating Specification 702 ,a to reflect the issuance of Amendment 26 to the FSAR. 7 Response Spectra Consistency The response spectra utilized in the pipe stress analysis was consistent with the dynamic (structural) analysis output. In some 's cases additional spectra were utilized wh'en it did not appear. necessary. Utilization of these additional spectra adds to the q conservatism of the design. 4 .J 5. n .) J

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$4 ATTACHMENT l-1 J.O. 14236 INDEPENDENT SEISMIC DESIGN VERIyICATION V.C. SUMMER NUCLEAR STATION,-UNIT NO. 1 SOUTE CAROLINA EI.ECTRIC & GAS CO. ia F7 Statement Regarding Pcten:ial or Apparent Conflicts of Interest I To: S:ene & Webs:er Engineering Corporation Whereas,, :he undersigned e=ployee (" Employee") understands that he or she '2 is assigned as a participan: :o provide services to Sou:5 Carolina Electric & Gas Company vich respect to :he Design Verification Progra= for the V.C. Summer Nuclear Station; and ~. ' ') Whereas E=ployee understands : hat it is necessary that :he participants se screenec for any potential or apparen: conflicts of in:eres: vi:n respec: :o this assignment; [.1 Therefore, for :he above stated purposes E=ployee =akes the following _) representations to 5:ene & Webster Engineering Corpora: ion: m i 1. E=playee has not engaged in any work or business involved vi:h or related to the engineering or design of the V.C. Su=mer Nuclear Station other than this Design Verifica: ion Program; .s.. 2 ' 2. Neither Employee, nor any members of,his or her i= mediate family, own any beneficial in:erest in the South Carolina Elec:ric & Gas Company, including but not limited :o common or preferred s:ock, '~ bonds or.other securities issued on behalf of the Sou:h Carolina Elec:ric & Gas. Company; and a 3. None of the me=bers of E=ployee's L==edia:e fa=11y are e= ployed l by South Carolina Elec:ric & Gas Cocpany. J This statement is based upon the E=ployee's best infor=ation and belief f. and any exceptions to the representations contained herein have been described on the reverse side of this document. '.T.. eih Da:ed

,a Signature

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w Print Name e '.t .f .3 ~4 . -4 J i t I. 3

i i i 1 TABLE l-1 PROJECT SERSONNEL ..-4 Proie t Manacer: P*.ter Dunlop Pre +ect Eneineer: K. Y, Chu . Desizn Centrol Audit Manarer: Assistant Preiede Encineer: J. F. Pa= J. P.. MacKinnen TASK 1 ?! ELD *a*ALKDC'*7; TASK 3 DESIGN CONTROL AI.TT 4 N. Roch D. Malone K. Anderson R. Twigg /~ J. Y. Chen D. Loffa "~ A. hoss L. Peterson V. Saleta s. TASK 2 STRESS ANALYSIS AND EVALUATION T. k'ei D. Loffa J. Y. Chen J. Chiang ~ Y. Chin' J. Chu = ve%0 a 4 p. 4 e 1 ~} .a k e , ~. - -. _ - - L* ~.." * * * * * ' * ' ' ' * ' ~ ~ * * * ~ ~

• TABLE l-2 PROJECT PROCEDURES i

'(A) TASK SPECIFIC PROCEDURES FIELD WALKDOWN EFFORT VCS-1 Field Walkdown Procedure STRESS ANALYSIS AND EVALUATION i 4 VCS-3 Analysis and Evaluation Procedure VCS-4 Analysis and Evaluation Criteria DE' SIGN CCNTROL AUDIT Design Control Verification Plan (3) PRO.?EC'" GEERIC ?LANS/PROCDURES Qualt:y Assurance Plan Docunent Control Procedure - VCS-2 Quality Assurance Records Procedure - VCS-5 i Engineering Assurance Audit Progran m. gb O I f ...w 6 i ? ~ M i

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1 ?7 ' 48 1 e J. APPENDIX B: SCE&G RESPONSES TO DRAFT FINAL REPORT r_ r s. n i;rw l ~ b.: s.. .a a ?"4 ?: . 3 0 eimi .7 a

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^ 92 ~ SOUTH CAROLINA ELECTRIC & GAS COMPANY man orries sex vu COLUMBI A. 5. C. 29218 ] October 12, 1982 sos-74a. 3722 t.J Mr. Peter Dunlop Stone & Webster Engineering Corporation ~' .i One Penn Plaza 250 West 34th Street-t g New York, New York.10ll6 i

Subject:

Virgil C. Summer Nuclear Station Independent Seismic Design il Verification a NE File No. 1.1801 p

Dear Mr. Dunlop:

South Carolina Electric and Gas Company (SCE&G) has reviewed, in detail, the draft final report submitted by your letter dated September 30, 1982. There are several recommendations from Stone and Webster Engineering Corporation (SWEC) and also open issues which must be addressed for the final report. Accordingly, we wish to provide you with information such that these areas may be closed. ~ SWEC recommended that seismic response spectra and seismic anchor movements be reviewed for other piping systens for the Virgil C. a Summer Nuclear Station. The following is our response to that recommendation. 5 Prior to the beginning of the Independent Design Verification Program, a problem was discovered with analysis code CC08. It was determined that not all of the seismic response-spectra curves had been identified on the isometric for input to the j,' analysis. Subsequently, a 100% review of computer analyzed probless was undertaken except for those problems analyzed by Westinghouse. q. Westinghouse had been provided the entire' Specification 702 from which to extract response spectra curves. All other 'i organizations relied on the figure numbers indicated on each as isometric. This effort also included a review to ensure that the latest response curve revisions were actually used in the 7 analyses and to ensure that changes in buildings were picked up. .l As a result of these reviews, several other analysis codes, including EF-01, were f ound to have this same problem. However, most were acceptable without reanalysis because the additional 1 curves were enveloped'by existing spectra. Analysis codes EF-05, U CC-05A, and SW-04 as well as CC-0.8 were reanalyzed. Their supports were subsequently reverified. The additional resp 6nse

!9 spectra for analysis code EF-01 was reviewed by Teledyne and

' ij determined not to adversely impact the analysis. 9 4 ev- - -.--.-ew-,e-- g-e _y ,-y e -iw_a-.,. --,ng y s-M .9g y-w-- uy--, ~--*-9

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.m U Peter Dunlop. October 12, 1983 ,l Page $2 71 d In late June and early July of 1982, another 100% review was under-taken for Seismic Anchor Movements (SAM). Calculations generating the seismic Anchor Movements were checked. A review of the SAM's indicated on each isometric was also made. The results of these reviews were that one other case, piping to the Reactor Coolant Pump upper bearing oil cooler analysis codes CC-09 through CC-13, did ,.2" not utilize proper 3AM's. SCE&G has accepted analysis codes without hardware modifications. n SWEC also recommended that the GAI specification 1902 be updated ~ to clearly reflect the design criteria applicable to jet impinge-ment. The following is cur response to the recommendation. A typographical error in the Jet Impingement GAI Report 1902 resulted in the incorrect acclication of jet cases IF-2 and EF-3. Il Additionally, it was found that through lack of clarity in the J report, Teledyne applied the je: forces to the Emergency case instead of the Upset case. This was identified during the SWEC p-program.- As a result an extensive program was undertaken to address this report and its application. Inputs.to the report were checked resulting in the following: 1. Elimination of 22 cases due to the addition of jet shields. 2. Reduction of nine jet case loads. l 3. One jet case had an increase ic load. 3 4. Revision of Figures 37 and 58 due to the typographical error. The application of the Report 1902 was then checked for safety i related piping. The results of this effort were the following: I 1. Jet impingement design input had not been considered for Westinghouse analysis. v . r. 2. Two other analysis problems were found not to have utilized proper jet impingement design input. 3. Ten jet cases were found not to have been considered in the application of the seismic spacing criteria, GAI Report 1923. ] 4. Lead factors had not been chosen properly. _J 5. Forces for support design purposes had not been included l7 in the proper design loading conditions. g ..m.. ,y w e""'. u

= ~t u i Peter Dunlop October.12, 1982 y Page #3 ."a n -1 The corrective action taken was to revise the GAI Report 1902 to ~ resolve these problems. Each of the affected analytical problems was reviewed for the correct design input. Although a few sup-y ports have had minor load increases, there have not been any support hardware mod 4fications as a result of items 2, 3, 4, and 5. Westinghouse has completed their review of the applicable jet cases d and has concluded that there are no pipe stress problems. GAI has completed their review of 144 of the 146 supports involved. The l 144 supports have been accepted without hardware modifications. [q The remaining two supports are still under review. 1 s SWEC recommended that the extent of incomplete documentation in Fi pipe stress analysis packages be aetermined and corrected. Also it d was recommended that either a revision to Specification 702 be performed, or its equivalent, to clarify the application of da= ping }] factors. The following is our response to these recommendations. As uncovered by the design control audit portion of the indepen-dent design verification, there was some confusion in the appli-r, by cation.of damping factors. Amendment 26 of the FSAR revised Table.3.7-1 to be. consistent with Reg.. Guide 1.61.

However, Table 3.7-9 of the FSAR was not revised causing a disagreement l

within the FSAR document. FSAR Change Notice 368 has been-U issued to delete damping factors from Table 3.7-9. This change notice will be incorporated in Amendment 34. Table 1 of GAI 9 Specification 702 presents the damping factors to be used for l buildings, assemblies and structures, components, etc. It does not agree with the information in the FSAR; however, the information in Specification 702 is more conservative. It has been decided that, i in order to avoid confusion and unnecessary revisions to many qualification documents for the Virgil C. Summer Nuclear Station, l a this specification will not be revised. Instead, SCE&G will include l 2 in an appropriate design criteria the danping factors used for 7. piping analysis. It is our position that this design criteria is more conservative, will resolve the question, and will also provide documentation for future application. l The design control audit portion also identified a concern involving documentation problems in the pipe stress analysis packages. This problem was recognized by SCE&G in our audit

a program.

We have been working'with GAI to correct this situation. At this tim'e we are in the beginning phases of a program to again review every analytical package. It is our position that these j documentation problems would not affect the design. In addition to the recommendations made by SWEC, the following t i information is provided to address the problems discovered during the field walkdown. 9

Peter Dunlop October 12, 1982 Page #4 As a result of the SWEC walkdown, two (2) supports were found with excessive clearances between the pipe and box type retraints I' requirements on box type restraints. A significant deficiency under the provision of 10CFR50.55(e) was filed on this problem. Corrective action resulted in a 100% review of computer analyzed gap requirements of box type supports. Several other supports ~ were found with excessive as well as not enough gap and subsequently corrected. Three conditio'ns were found in which the direction of a skewed strut was misidentified. A 100% r.eview of computer analyzed skewed supports was undertaken. As a result several other supports on other systems were found. Review by the Analysts (Teledyne, EDS and Gilbert) resulted in acceptance as is of not only the three (3) found by SWEC, but also those f ound by SCE&G. One of the last phases in the piping design v'erification process was SCE&G's two hot functional tests and the final Thermal Expan-sion Test, TE-3. The TE-3 program is in its last phase of completion at this time. Two very important parts of these prograns are to check / set all spring cans and inspect in detail for non-supporting structures. Additionally, a separate penetra-tion program was undertaken for safety related piping to ensure proper clearances. Penetration P-IB-1-041 was addressed by this l program and accepted. l l If you require additional information, or have comments to the l responses which we have provided, please advise. (/ % F Gary Moffatt, Mechanical Engineer Nuclear Engineering

b. O.

M N C. A. Price, Manager 8 Nuclear Engineering i /fje I + +er==**.===. m&}}