ML20091J788

Piping Design Review,Part 1

ML20091J788
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Site:	Perry
Issue date:	05/29/1984
From:	CLEVELAND ELECTRIC ILLUMINATING CO.
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ML20091J786	List: ML20091J783 ML20091J788 ML20091J789
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NUDOCS 8406060191
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{{#Wiki_filter:IRE CLEVELAl0 ELECIRlC lll U f5 lll Aill: 0 C-O ld ? A ll Y N"W Piping Design S" "i. ( ) FA O PAGE: Review REVISION: INTRODUCTION In early 1982, The Cleveland Electric Illuminating Company (CEI) conceptually defined the scope of a Design Verificaticn Program. This was done to further our confidence in the Gilbert Associates, Inc. (GAI) design program and our recognition of the regulatory environment and its changing direction. This report, Piping Design Review, represents one portion of that overall effort which includes design review programs addressing both Quality Assurance and Engineering objecti ::. v The Cygua Energy Consultants were selected to perform the piping design review on selected systems, as described in Part I Section 1-3. This review was completed as of February 28, 1984. The methodology of the review is described ~ in Part I Section 4.0 and Part II Volume I, Section 3. The results and all documentation from the review are contained in Part II Volumes 1 & 2 of this package. ( ) 8406060191 840529 PDR ADOCK 05000440 A PDR PERRY NUCLEAR POWER PLANT Servong The Best Location in the Nation PO box 97 e PERRY OHto 44081 e 7ELEPHONE (216l 269 3737 e ADOAESS 10 CEN7EA AOAO

THE C L E V E lls i: 0 ELECTRIC ! L L U L'I N A T il: 2 0 0 f4 F A l: Y WGE"W Piping Design S"" o 7 M PAGE: Review REVISION: INTRODUCTION (Cont'd) After completion of Cygna review, CEI initiated a follow-up review, to assure that potential ~ generic items identified are reviewed through the entire piping design control program. Cygna concentrated on the adequacy of the three systems within their review scope, whereas the CEI follow-up review addressed the potential for similar discrepancies to affect other safety-related systems. The CEI follow-up review and observation closure is described in Part I, Section 5 through 6 of this package. Observations are considered closed for the purpose of tilis report if action to insure it is addressed has been O' g developed. This will insure any generic problems are tracked to closure thereby receiving proper action. ,m ( PERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation 90 SON 97 e PERHY. OHIO 4404) e TELEPHONE #2166 259 3737 e ADOAESS to CENTE A ACAD

THE C L E \\' E L L I: D ELECTRIC ILLl!ffil,'ATIEC C O M F i. l: Y WGhW Piping Design S"" WNDM O r PAGE: Review ~ REVISION: TABLE OF CONTENTS Part I CEI Section-Description 1.0 Purpose 2.0 Scope 3.0 Participants Exhibit 3.1 Cygna Resumes Exhibit 3.2 CEI Engineer - Rasume 4.0 Methodology 5.0 Observation Follow-up 5.1 Mechanical 5.2 Piping Analysis 5.3 Pipe Support 6.0 CEI Conclusions GE Criteria Compliance Review Part II Cygna Final Report Volume 1 Section Description 1.0 Executive Sumraary 2.0 Program Review Scope 3.0 Mcthodology 4.0 Review Results and Conclusions Volume 2 Appendices A Definitions and Nomenclature B Documents Reviewed C Review Criteria D Checklists E Observations F Potential Finding Reports 9 PERRY hlICLEAR POV ER PLANT Serving The Best Location in the Nation PO DOX 97 o PERAY. OHIO 44081 e TELEPMONE d2161 259 3737 e ACDAESS 10 CENTEn nOAO L

r, e, - n i t y : :. 3, r, n r i ...r..a vuu.., a a. .......c W"W O ' f [X XJ % Piping Design S'" 1.0 \\./ lu PAGE: i gg g Review REVIsl0N: 0 t 1.0 PURPOSE The Piping Design Review was initiated to confirm the technical adequacy of Gilbert Associates' mechanical and piping design. While design control has been audited throughout the project, an additional technical review was initiated to evaluate the con-formance with design specifications, design criteria, licensing commitments and standard industry practices. This was accomplished byacompletedesignverificationandtechnicaldesigt$reviewbyan independent consultant. N I + PERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation PO BOX 97 e PEAAY. CHIO 44081 e TELEPHONE (2166 259 3737 o AcoAESS 'O CENTEA AOAO

IHE - CLEVELAP.D ELECiR!C I L L U M I N A ::::: 1" ? ;:: Y W"W Piping Design 5 " 2.0 m ) 7Fl!G-@- 4 PAGE. f2 Flestrienvy REVISION 0 2.0 SCOPE Three subsystams were chosen to give a good representation of CAI's piping analysis. This review approach has been called a'"vartical slice" because it started with the application of design requirements from CE, ASME, Federal Regulations, etc., and ended by reviewing the detail design drawings from GAI. The review followed the design process through system flow calculations, piping analysis, and pipe support design to the detailed drawings. The piping subsystems were selected to be consistent with the GAI piping and support analysis scope. The mechanical review scope was broader in order to review ry ( j the process calculations. The scope of the review is described in xs more detail within the Cygna final report. Following the independent consultant review, CEI reviewed all observations for generic implications to assess the effect of noted discrepancies on other safety-related systems. Any resulting findings were documented on an Engineering Design Deficiency Report to properly track the resolution and closure of these generic items. PERRY NUCLEAR POWER PLANT Serving The Best location in the Nation a p0 eOx 97 e. PERRY OHIO 44o81 s TELEPHONE (2166 255 3737 e ,ACOAESS. to CENTE A ACAO

THE CLEVELAND ELECTRIC ILLUMIN ATING COMPANY h Em m W Piping Design secu ~: 2.0 ESES'3% PAGE: 2 of 2 .,_ s ( l Rev,iew REVISION: 0 w/ The three systems chosen were: o 1-E22-G004 Class 1 High Pressure Core Spray Primary Reasons: 1. Important to Safe Shutdown or Cooldown 2. Strong Interaction with Other Systems ~ 3. Large Bore 4 Organizational Interface GE/GAI 5. Attached to RPV o 1-N22-G001 Class 1 Main Steam Drain g Primary Reasons: 1. Typical Standard Design ~

2. ' Small Bore 3.

Organizational Interface GE/GAI o 1-B21-G008 Class 3 SRV Discharge Line Primary Reasons: 1. To Review Class 2 or 3 Piping Analysis 2. Hydraulic Thrust Loads and Transients 3. Piping is Highly Stressed and Difficult Routing 4. Organization Interface 5. Important to Overpressure Protection of the Reactor h PERRY NUCLEAR POWER PLANT Serving The Best location in the Nation ' ADOAdss.to CQNfQn nOAO p(3 oor 91 e PEAAY OHIO 440st .e TELEPHONE (2169 259 3737 e

u; n -i r> rt: Ihw ts. c v L a t. a ,--er i !. i t '.. i a n. r e w- -. i. i w a . J hhN Piping Design 5'c" r-3.0 E O v PAGE. 1 of 2 Rev,iew REVISION. 0 3.0 PARTICIPANTS Cygna Energy Services provided most of the review organization which is shown in Figure #5. Their selection by CEI was bar;ed on their acceptability on all of the following items: 1. Design experience of personnel proposed for the review team. 2. Technical experts to supply backup for the review team if needed. 3. Independent of Gilbert Associates and the Perry Project design. Previous Cygna work on Perry represented less than 1% of their yearly revenue. In addition, this work did not involve design responsibility. 4. Cygna previously had performed two indepe.ndent design verifications. Cygna's previous reviews, as well as their proposal to CEI, included developing detailed acceptance cr.iteria and checklists prior to starting the review. CEI felt this well-organized approach would result in a meaningful review and a high confidence level in the outcome. The review board also included J. E. Meyer. Mr. Meyer is a recent addition to the CEI organization. llis expertise will increase,CEI's piping analysis capabilities, lie brought with him nine years of experience in piping at4d support analysis and han been active on l I variotas ANSI /ASME piping code committees. In addition, he was independent of prior Gilbert and CEI design decisions.

ERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation PO SON 97 e PERRY ohs 0 44081 o TELEPHONE #216) 2$9 3737 e ADOAESS to CEN7En nOAD

O[ O' O,_ FIGUftE +5 PRINCIPAL IN [HARGE J. A. Famiglietti l PROJECT MANAGER T.T. Wittig REVIEW BOARD 12-H0uSE CDRSULTANTS J.E. Meyer (CEI) E. van 5ttjgeren (Codes a standards) ~- T.T. Wittig J. Mtalchle11e J C.E. Wong k a 'Ws) PR(UECT SECRETART A5 A E I ER 5t s T. Acuna B. Good I i I PIPE SUPPORTS EROUP LEADER PROJECT ENGINEER AND MECHANICAL GA0uP LEADER PIPE STRESS GROUP LEADER ?.E. Wong L.J. Weingart R. Ness I 1 LEGEm0 i - PROJECT DIRECTION

=====COR$ ULT ATIO4 PIPING REVIEW ORGANIZATION i Cleveland Electric Illuminating Company i i Perry Nuclear Power Plant Design Review s 1 ,i

EXHIBIT 3.1 ?:::- C. 4 I 1 REVIEW BOARD i .i J. Meyer, CEI - SEE EXHBIT 3.2 T Wittig L. Weingart R. Hess C. Wong 4 N e i l 4 t J i I l . Op i L lO E,,,,,,, PAGE 1 OF 23

TED T. WITTIG O i EDUCATION: B.S., Civil / Structural Engineering, Michigan Technological University, Houghton, MI PROFESSIONAL REGISTRATION: Civil Engineer, California PROFESSIONAL EXPERIENCE: Mr. Wittig has over thirteen years of experience in structural engineering for nuclear power plants and is currently the Monoger of Projects. This experience includes criteria development, seismic analysis, high temperature effects, impact evaluations and soil-structure interaction. With Cygna, Mr. Wittig has acted as the Project Manager for the following projects: Independent Design Review for Mississippi Power & Light Independent Design Verification for Detroit Edison Company Third-Party Review for Cleveland Electric, Inc. Seismic Equipment Qualification for Washington Public Pnwer Supply System l The design reviews listed above covered a broad range of engineering design and design control activities, including structural, piping, pipe supports, cable tray supports, equipment qualification, electrical and mechanical. These reviews involved considerable interaction with the NRC in the form of developing a program plan and presenting the results. Prior to joining Cygna, Mr. Wittig was employed by a major architect / engineer. During this assignment he was responsible for the conceptual design and analysis of all structures on an LMFBR Study. He also acted as a liaison and technical reviewer for the LMFBR national team commissioned by the Department of Energy. His role as a technical reviewer covered the areas of structural, seismic, and planning / scheduling. Mr. Wittig also functioned as a structural engineer for a commerclol PWR plant. In this assignment he was responsible for the civil / structural design criteria, seismic analysis seismic specification for mechanical equipment and various special studies. The special studies included soil-structure interaction, tornado and turbine missile impact, and liquefaction. In addition, he was responsible for the design and analysis of the circulating water system Intake structures. I n v 2

TED T. WITTIG f (continued) ~ Mr. Wittig's previous experience has included design of roads, roilroads and seismic Category I structures for a major nuclear project. This experience included design and analysis of the containment building basemat and reactor cavity. It also included seismic analysis of the containment building and the design of major equipment supports. U O /[ 3 I

LEE J. WEINGART EDUCATION: B.S., Engineering. Son Francisco State University, Son Francisco, CA Undergraduate studies, Mechanical Engineering, Drexel University, Philadelphlo, PA Undergraduate studies, Cornmunicottons, Temple University, Philodelphia, PA PROFESSIONAL REGISTRATION: Registered Mechonical Engir.eer, California PROFESSIONAL AFFILIATIONS: Associate Member, American Society of Mechanical Engineers Member, American Nuclear Society PROFESSIONAL EXPERIENCE: Mr. Weingart has over ten years of experience with particular emphasis in the analysis of components, piping and support structures. He is presently assigned as a Senior Engineer ,Q of Cygna's Son Francisco office responsible for o brood range of engineering activities in V the Engineering Design Division, including: Assistant Project Engineer for the WPPSS-2 dynamic equipment quellfication. Lead Engineer for the Fermi-2 Independent Design Verification in the arcos of equipment qualification and piping. Project Engineer for the Grand Gulf Independent Design Review and Perry Independent Piping Analysis. Formerly, Mr. Weingart was employed as a Senior Engineer by Ouodrex Corporation, o West coast consulting engineering firm. Mr. Weingart was instrumental in computerizing standard calculations, modeling, and analysis. He created FORTRAN programs to facill-tote use of the SAGS program for computer modeling of pipe support structures, and per-formed static and nonlinear analysis of baseplates using STARDYNE. As a Structural Analyst for computer services support at Control Data Corporation, Mr. Weingort was actively involved in customer support services in structural osplications using ANSYS, EAC/ EASE 2, NASTRAN, SDRC/ SAGS, STARDYNE and STRUOL,f th and in piping opplicottons using DIS /ADLPIPE, NUPIPE cnd PIPESD. The capobilities o finite element programs include linear and nonlinear static, dynamic, and heat transfer 4 4

i LEE J. WEINGART O (c >> e d) onalyses of struct' res and piping systems. Mr. Weingart also served as the primary West Coast onalyst for : iping graphics opplications, in addition to organizing and participating (instructor) in training seminors for customers. Prior to t,e above, Mr. Weingart served as on Engineer at Bechtel Power Ccrporation where, es port of an overall Equipment Qualification effort, he located and sized the in4trumentation required to ver.ify dynamic transient onalyses which he performed (using avalloble computer programs such as STARDYNE and ANSYS) for both nuclear and fossil fuel power plant piping systems to determine restraint sizes and locations, and to assure system occeptability within code limits (ASME B&PV Section lll ond 031.1). He also per-formed thermal flexibility, weight and seismic calculations for both small and large pip-ing. He was also responsible for training new employees in analysis objectives and tech-niques, and coordinating their octivities. 9 U

ROBERT W. ESS G 1 EDUCATION: B.S., Engineering, University of Maryland, College Park, MD Graduate course work in Engineering Administration, George Washington University, Washington, DC Basic Project Monogement Course, Amerlean Management Association Air Conditioning and Refrigeration, Brevard Junior College, Cocoa, FL Cryogenics, Genesy's Extensten of University of Florida, Gainsville, FL PROFESSIONAL REGISTRATION: Professional Engineer, Mechanical, State of California PROFESSIONAL AFFILIATIONS: Member, American Nuclear Society Member, American Institute of Aeronautics and Astronautics PROFESSIONAL EXPERIENCE: Mr. Hess has more than eighteen years of experience in engineering and management. He is currently assigned as Engineering Monoger-Systems Engineering for the Western Region, in this capacity he is responsible for the supervision of multiple discipline groups including mechonicol, electrical, and instrumentation and control in the performance of systems onelysis and design, systems modification, computer opplications, and regulatory compliance projects. Formerly associated with NUS as General Monoger of its Western Engineering Office, he was responsible for the monogement, direction and staffing requirements of all engi-neerirg and design projects. in on earlier position as Monoger, Plant Engineering, his duties included technical direction and administrative activities associated with process development and system design of modifications to nuclear and fossil-fueled generating facilities. This included supervision of site investigations to determine system design requirements based on plant operations and site-specific constraints, technical approval of conceptual and detail design and monogement of assigned discipline ergineers and ' designers to meet schedule and budget requirements. Specific projects included NUREG 0612 compliance reports for Trojan and Crystal River Power Plants, ATWS modification requirements study for BWR's, preparation of emergency implementing procedures for a PWR, and modification of a pH control system for a fossil unit cooling tower. OU 6

1 l i 1 ROBERT W. lHESS (continued) As Project Engineer for the design of large waste treatment facilities for two fossil generating facilities, Mr. Hess was responsible for directing cnd sequencing project tasks to occomplish the work scope within budget and schedule, and maintaining formal com-munications with the client. This assignment required close coordination of design, procurement and construction efforts of process, mechanical, electrical, l&C, and civil / structural engineers. Other assignments with NUS included responsibilites for conceptual and detail design of make-up water and wastewater treatment systems for both nuclear and fossil power plants. These projects included specification of demineralizer systems, floating roof make-up water storage tanks, sand filters, pumps and tie-ins to existing systems. M r. Hess supervised engineers and designers in performance of discipline work scope within schedule and budget constraints; established system design criterio and coordinated inputs with other disciplines; prepared and supervised preparation of equipment specifications, construction bid packages, proposal bid evoluotions, P&lD's, equipment and piping layout drawings and engineering manhour estimates. Various other project experience includes engineering design and analysis of radioactive waste treatment systems for nuclear power plants, design and review of RCP oil enclosure systems, fossil plant fire water system modifications, and addition of fire suppression systems to the cable spreading rooms. While onigned to a core spray system modification project, he coordinated field engineering efforts and client inputs during the analysis and modifico-tion design, in addition to being responsible for the preparation of specifico+ ions, drawings and construction work packages for the installation of mechanical modifico-p) tions. Also, Mr. Hess prepared conceptual mechanical designs and weight analyses of t shippings casks for solid waste generated by nuclear fuel reprocessing plants (concepts included both rail and truck-mounted casks for high-and low-level wastes). Previously, Mr. Hess worked with Newport News Shipbuilding where he was responsible for the design and review of various fluid systems required for operation and support of a novol nuclear power plant. He participated the in formulation and composition of technical documents detailing and justifying system design characteristics, operating principles and maintenance requirements for primary shield water, reactor plant air and evacuation and nitrogen purge systems. As Lead Systems Engineer with Grummon Aerospace Corporation, Mr. Hess was responsi-ble for systems checkout and launch operations on the Lunar Module Propulsion Subsystems. His position required consideration of such items as test scheduling, manpower planning, review and approval of test procedures and direct supervision of engineers and technicians during pre-launch and launch operations. As Systems Engineer, he prepared and performed test procedures for fluid systems checkout, directed troubleshooting and repair of ground support and flight equipment, and participated in development and site start-up of high pressure gas and cryogenic k,oding equipment. O O lll111ll111111111111111lllllll 7 e

CHUN K. WONG o, EDUCATION: M.S., Structural Engineering, University of California, Berkeley, CA B.S., Civil Engineering, University of California, Berkeley, CA Ordinary Certificate Building Construction, Hong Kong Technical College, Hong Kong PROFESSIONAL REGISTRATION: Registered Professional Engineer (Civil), California Registered Civil Engineer, Ontario, Canodo PROFESSIONAL EXPERIENCE: Mr. Wong is currently on Engineering Supervisor in the Engineering Design Division at Cygna. He was assigned as Project Engineer for the design and analysis of the Control Rod Drive System for LaSalle Units I and 2. In this position, he was responsible for scheduling work and leading a group of ten engineers in the design of the support frames. His group used the ANSYS computer code to develop stiffnesses for the frames (for input to the pipe stress work) and to perform the final designs. O ere reesix. ur-wens werked en the 'imerick ceneretino Steeree areiect-se coordinated and supervised stress analysts in the performance of the analyses of piping systems in accordance with ASME Ill and B31.1 codes, and reviewed and approved stress calculations. For the Peach Bottom project, Mr. Wong coordinated and supervised analysts in the performance of NRC IE Bulletin 79-14, as-built analysis of nuclear piping systems. Mr. Wong also served as senior stress analyst, for the Surry Power Plant project and performed NRC 79-14 computer analysis of nuclear piping systems. Mr. Wong has also worked on such major projects as: Humboldt Bay Nuclear Power Plant, for which he performed dynamic seismic analysis of plant structures and soil-structure interaction analysis; Susquehanno Nuclear Power Plant, for which he performed pipe rupture time-history analysis of piping systems; Yankee Nuclear Power Station, for which he performed dynamic analysis of spent fuel pool; and Geyser Steam Gathering, for which he performed stress analysis of piping system. During his course of work at Cygna, Mr. Wong has gained extensive experience in structural dynamics and in the use of many commercial and Cygna proprietary programs such as ANSYS, PIPESD, PSA, SAPlV, NUPIPE, MEl01 (Bechtel Piping Program). ) i 1111111!!1111lll I

lO .i. t i 1 l PROJECT TEAM 4 i t i J. Minichiello R. Boliga S.Luo V. Phi i t h I I i U i 1 t e i I i I i i + 9

r i i RAVIPORANATH B. BALICA v EDUCATION: M. OcE., Ocean Engineering Structures, Oregon State University, OR M. Tech., Morine Structures, Mysore University, India B. E., Civil Engineering, Mysore University, India PROFESSIONAL REGISTRATION: Engineer-in-Trotning, Collfornia PROFESSIONAL EXPERIENCE: As a Senior Engineer in the Engineering Design Division of Cygno, Mr. 001190 is currently involved in the Independent Review of the Perry I plant, pipe support design group. Prior to this assignment rsn Perry, he was the Project Engineer in charge of the pl>e stress onalysis for Diablo Canyon Unit 1. As such, he is responsible for the scieduling, technical direction, and opproval of all work in this oreo of the project. He has been involved in a vorlety of tasks at Cygno, including: Midland Nuclear Stotlon O. eerformed nine 8,eow eeei<s s esieg PieEnuP oed designed feiie, resirerets. Susquehanna Nuclear Statlon Performed pipe stress analysis using MEl01 and ANSYS and a volve response study using ME632. Polo Verde Performed pipe stress analysis and evoluoted pipe suppori designs. Lo Solle Performed pipe stress onalysis on the CRD system using ADLPIPE. Peach Bottom / Limerick Performed pipe stress onelysis using MEl01. o 10 u

RAVIPORANATH B. BALIGA ] (continued) Vermont Yonkee Collected os-built piping dato and performed evoluotions in occordance with NRC lE Eulletins 79-01 and 79-14. Cygno Research and Development Developed a general purpose plotting package for in-house computer programs and performed dynamic analysis of beams on elastic foundations and two-way concrete slobs. Previously associated with Roy Desol Associated, he was responsible for the design of the foundations of stationary and bridge crones os well as retaining walls. Also performed dynamic onalysis of multi-story steel and concrete structures using finite-element computer methods. As o research assistont at Oregon State University, he performed a hydrodynamic study on sond waves in on estuary, dato collection in the field, computer programming and report writing. He also performed on experimental study in stochastic wave forces. Mr. Boliga also worked as on assistant lecturer at Mysore University, India, where he tought grophic statics, fluid mechanics and opplied mechanics, wC Prior to teaching, he worked as a consultant for planning and design of steel and concrete frame structures. PUBLICATlONS: " Influence of Hydrodynamics On Rote of Sediment Turnover Mechonics of Sond Wave Motion," National Science Foundation, Washington, D.C.,1976. " Estuarine Sediment Dispersion," report submitted to National Science Foundation, Washington D.C.,1976 " Stochastic Wave Tests on Test Cylinder - Dynamic Analysis of Hydrodynamic Force on Cylinder," report submitted to Continental Oil Company, Ponca City, Oklohoma, January 1978. "Evoluotion of Sand Waves in on Estuary," Journal of the Hydraulles Division, ASCE, February,1981. 11

r SIMON L UO b EDUCATION: M.S., Civil Engineering (structural), Texas Tech University, Lubbock, TX B.S., Civil Engineering, Tomkong University, Taipei, Tolwan, R.O.C. PROFESSIONAL REGISTRATION: Engineer-in-Trotning, Texas s PROFESSIONAL AFFILIATIONS: Member, American Concrete Institute Member, American Institute of Steel Construction PROFESSIONAL EXPERIENCE: Mr. Luo is a Staff Engineer currently assisting in program development for Cygna's CYTRAC computer program which tracks rodwoste in-plant. Other projects Mr. Luo has been involved in were the static and dynamic structural analysis and design evoluotions of the pipe sepport systems for Perry Unit 1, Comanche Peak Units I and 2, Diablo O Conyon Unit I and Lo Salle Unit 2. Previous assignments have included computer analysis for the Susquehanna Nuclear Power Plant pipe support system under seismic lood and documenting analysis results to meet ASME, ANS codes; computer pipe stress analysis for the Lo Solle Unit i Nuclear Power Plant CRD piping system under seismic, thermal and gravity loads. Formerly employed by the Hugh M. O'Neil Company, Mr. Luo was responsible for the design ond onolysis of a jib crane including the detailing of structure in steel. Other , design work required the opplication of finite element methods of dynamic analysis for o Lucky Stores' project. While working on his mosfer's at Texas Tech University, Mr. Luo was involved in the research of spall behavior for the U.S. Air Force. He develond a finite element computer program to simulate the stress wave propogotion due to 'mpact and by using a sultobie numerical integration scheme for the dynamic equation of motion involved in the stress wave oropogotion phenomeno. (/ M 12

SIMON LUO Iq (continued) g Additional industrial experience was acquired by Mr. Luo through his association with the Public Works Department, Taipei City. He was responsible for construction material quality and quantity control, sheer wall and basement construction design, schedule control. l PUBLICATIONS: 4 "A fracture spall finite element model in impact problems," Eleventh Southwestern Graduate Research in Applied Mechanics, Oklahoma State University, April i1, 1980. d 1 5

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VUONG PHI EDUCATION: B.S., Mechanicc.' Engineering, San Francisco State University, San Francisco, CA Bechtel Professional Training Program, Subjects: Piping Stress Analysis and Nuclear Power Plant Design PROFESSIONAL EXPERIENCE: As a Senior Engineer in the Engineering Design Division at Cygna, Ms. Phi is currently assigned as a pipe stress group leader for the Diablo Canyon Unit I reonalysis. With over six years in piping stress analysis, Ms. Phi is responsible for directing other engineers and solving technical problems as they crise. Just recently, Ms. Phi completed a; independent Design Review of the RHR piping analysis for the Grand Gulf Nuclear Station, in which she compared the octual analyses methods to that in the various ASME and NRC criteria and also checked that results met the stated criterio. Her previous work included assignment as Coordinator, Trojan Project, in which she evaluated the safety and operability of a variety of systems under the scope of I.E.B. 79-14 and of the auxiliary control modification area. Ms. Phi has served as Pipe Stress Analyst for o number of projects: Hope Creek, F.F.T.F., Susquehanna, Limerick, Peach Bottom, Yankee Rowe, Vermont Yankee, Vermont Main, Diablo Canyon, LaSalle and MP&L. In these, her duties included as-built s analysis and hydrodynamic loading analysis in accordance with the requirements of the appropriate codes. Prior to joining Cygna, Ms. Phi worked in the Power Division of Bechtel Power Corporation as a Stress Analyst. Her work involved analysis of piping subjected to thermal, gravity and seismic loadings. Ms. Phi is also experienced in using linear elastic finite element programs such as: ME 101, ME 632, PIPESD, ADLPIPE, PIPSYS, SUPERPIPE, and ANSYS. She also con-ducted a session on how to utilize ME 101 for doing piping stress analysis. ON.] 14

n JOI-N C. MINICHIELLO %) 1 EDUCATION: M.S., Applied Mechonics, Harvard University, Cambridge, MA B.S., Mechanical Engineering, Tuf ts University, Boston, MA PROFESSIONAL REGISTRATION: i Professional Engineer, Mechanical, Massachussetts and California PROFESSIONAL AFFILIATIONS: Member, Americon Society of Mechonical Engineers Member, Tou Beto Pi Engineering Soclety Member, American Nuclear Society PROFESSIONAL EXPERIENCE: Mr. 'Ainichiello is assigned as the Manager of the Engineering Design Division at Cygno. His responsibilities include technical direction of all projects within the Division, stof fing O "d '"ds ' ar aer# "a r a ' a " r " e"- a As part of his assignment, Mr. Minichiello served as the project engineer for the dynamic requalification of Mechonical Equipment for the Washington Public Power Supply System Unit 2 nuclear plant. This work involved upgrading the previous work to the new hydrodynamic loods and the new criterio (IEEE-344-1975). His division it currently also responsible for the stress analysis of the piping and the design of new pipe supports to meet the SEP requirements for the Yonkee Nuclear Station at Rowe, Mossochusetts, included in this evaluation is the analysis of the mechonical equipment (volves, steam generators, etc.) necessary to the operation of the plant. Other projects within his division included: the stress onalysis and support design for the control rod drive piping for the LaSalle stationi and reonalysis of piping and pipe supports for Diablo Canyon Unit 1. As Section Monoger for stress onolysis at Brown and Root Inc., Mr. Minichiello's responsibilities encompassed the overall direction of all mechon,ical onalysis and design activities for the company's nuclear and fossil projects. Activities included: a full range of piping design and analyses for the South Texas Nuclear Projecti computer.olded struc. tural analysis of on electric substation Insulating posts under 3-phase short circuit dynamic loodingi and development of stress destgo standards for Brown and Rnot. C' I is

i 1 JOHN C. MINICHIELLO

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(continued) As head of the component analysis section at NUS Corporation, he was responsible for proposal generation, direction and completion of the analysis (thermal, stress, and dynamic) of equipment in occordance with ASME, ANSI, and AISC codes. Projects included direction of the onolysis of a fuel pool skimmer tonk for dynamic looding dynamic analysis of vocuum relief volves, and the stress analysis of heat exchangers., th He was also responsible for technical direction for o team of 25 engineers performing the piping onalysis of 200 sub-systems for the Wm. H. Zimmer Nuclear Station. Mr. Minichiello generated proposals for linear and nonlinear (gopping) onalysis of heat exchanger component ports. For the Nine Mlle Island plant, he performed fracture onolyses of welds on the downcomers. This activity involved determining the stability of crock growth initiated by thermal cycling. His post work also included dynamic onalysis of high rodlotion sampling systems (ponels and piping), and analysis of various pressure vessels. As Lead Senior Engineer with EDS Nuclear, he was responsible for the design and onelysis for safety-related piping systems for the McCulte Nuclear Station. This effort involved the thermal transient and fatigue analysis required for ASME Class I systems and the identification of system modificottons, when required to alleviate thermal problems. Other projects included finite element analysis of penetration head fittings for thermal and structural loods and verification of the SUPERPIPE program per EDS OA standards. Mr. Minichiello's previous experience at NUS Corporotron includes fluid, thermal and structural analysis of nuclear systems and components using finite element codes such as ANSYS, STARDYNE ond PIPESD. These analyses included such evoluotions os the dynamic response of the auxiliary cooling piping for a reactor coolant pump test loop, the dynamic response of centrifugal chiller ossemblies, the dynamic response of high density spent fuel rocks and the high temperature respense of spent fuel shipping casks. He produced the hydraulle and thermal analysis report for the 57G reactor pressure vessel head and performed the flow calculations for the S7G purificotton filter. He has per-formed complete stress and thermal onalysis of the LOFT reactor vessel, including comparison of results to ASME code allowables and generation of the final stress report, a and was responsible for the computer code generoflon used to pre-ond post-process i finite element stress output to old in the evoluotion of ASME code requirements. As a stress engineer, Mr. Minichiello performed thermal and stress analysis of a purification filter using finite-difference and shell computer codes and performed the stress analysis of electrical plug plates per ASME Class lil criterio. Earlier, at Roytheon Co., Mr. Minichiello worked as o design engineer and was in charge of fabrication of a prototype onolog-digital computer Interfoce device. He also designed components of a control boord for missile trocking systems. AV it

1 l i l 4 i I I ) ~ CRITERIA VERIFICATKP4 l l I 1 l s i L. Kommerzell l D. Gordner j T. Nguyen i t r O i s i f f i I i 4 l l 1 c 4 ( I 1 ii l I i e 1 1 l I i l \\ 4 I i l i i I' I i 17 -,m-._..-. - -. _. -...., _,. _,,,. _ _... -,. ~. _. _. _ _ -.. - .m -.

r i LARRY L. KAMMERZELL ,3 V EDUCATION: M.B.A., National University (in progress), Son Diego, CA B.B.A., National University, Son Diego, CA i Third Year Inwstr101 Engineering, Drexel Institute of Technology, Philadelphlo, PA $PECIALTY COURSES: Business Management Seminors at General Atomic Company Naval T ilning: Navy Nuclear Power School Aovonced Submarine Engineering School Nuclear Deep Submersible Pilot and Power Plant Training PROFESSIONAL REGISTRATION: Professional Engineer (Nuclear), Collfornia l (O PROFES$lONAL AFFILIATIONS: l Member, American Nuclear Society (Post Chairman of Son Diego Seclion) Member, National Monogement Association (Post President, General Atomic Chapter) PROFES$10NAL EXPERIENCE: Mr.Kommerzell has twenty years of nucleor related experience covering a browf spectrum of Nuclear Power Plant rlA assessment, onolysis, testing, construction, and operoflons. He is presently serving as a Product Develcpment Monoger for Cygno. Previously, he acted as a discipline and project monoger for reliabl Illy, rlA assessment and rodwonte projects, and as monoger of Cygna's Son Diego of fice. Prior to joining Cygno, Mr.Kommerrell held responsible engineering and monogement positions with Stone & Webster Engineering Corp., United Engineers orxl Constructors, General Atomic Company and the U.S. Navy. The following summorizes his activilles over the post 20 years. At General Atomic Coneny Mr.kommerrell was Monoger of Systems Engineering, responsible for the coordination and technical Integration of the 18

LARRY L. KAMMERZELL ('; (continued) U various systems and component designs into on optimum plant design and to organize, direct and odminister overall systems engineering efforts on HTGR alonts including Sofety Analysis, Probabilistic Risk Assessment programs, and ' conomic Study Evoluotions, i in other positions held of General Atomic, Mr. Kommerzell was responsible for plant thermal performance evoluotions including the development of analytical techniques to determine the thermal performance risk associated with the specific plant design. As lead nuclear engineer at United Engineers and Constructors, he was respon-sible for the preparation of the safety analysis report for systems and facilities supporting the nuclear steam supply. These included the rodwoste, core cooling, and fuel storoge systems and the osso:loted building arrangements. At Stone and Webster, Mr. Kommerrell was responsible for evolootion of vendor l test and weld procedures. He was also responsible for the design, specification, ond field erection of nuclear power plant pumps, vessels and heat exchanges. Mr. Kommorrell held several positions in tee United States Navy. Representative of this period is his assignment os Nuclear power plant prototype instructor and assignment as M/A division of flcer on board the NR-l during 'he constructiori, testing, scotrials and inillol service. The NR-l is a (q huclear Powered Deep Submersible research submarine. Mr. Kommerrell had ) responsib'lity for: all phases of testing, trouble shooting, colibration ond molntenance of reactor, propulsion, and turbine generating equipmentt all power plont evolutknst and all underwater evolutions. He was the duty of ficer dur!ng power range testing and was responsible for testing during Inlllol crllicolity. I t .,m. 19 e

t DONALD A. GARDER, JR. EDUCATION: M.S., Nuclear Engineering, State University of New York at Buffalo, Buffalo, NY B.S., Aerospace Engineering, State University of New York at Buffalo, Buffalo, NY Associates Degree, Engineering, Auburn Community College, Auburn, NY ) ASME Radwoste Seminar, Georgia institute of Technology and Arizona State University PROFESSIONAL REGISTRATION: Professional Engineer, Nuclear, California PROFESSIONAL EXPERIENCE: Mr. Gardner is currently on Associate at Cygno, responsible for Rodwoste Engineering Services. Prior to joining Cygno, Mr. Gardner held supervisory and lead engineer positions at United Engineers and Constructors, General Atomic Company, Long Island Lighting Company, Virginia Electric and Power Company, and Niagara Mohawk Power Corpor-otion. His experience background is summarized as follows: OV Corporate Specialist for radweste systems working on the Seabrook I and 3 and Washington Public Power Supply Systems I and 4 radwoste system designs. In this capacity he acted as Lead Engineer on the conceptual design and cost evaluation of an Interim On-Site Rodwoste Storage Facility for the Seabrook I and 2 project. 4 -. Responsible for the performance of nuclear systems analyses, radioactivity release and dose ossessment analyses and radiation protection studies for the Brunswick I and 2, Indian Point 2, Seabrook I and 2, and Washington Public Power Supply System I and 4 nuclear plants. As Lead Nuclear Systems and Rodwaste Engineer on the SSoreham and James-port projects, Mr. Gardner was responsible for reviewing and evaluating NSSS and rodwoste system designs, opproving design changes related to these systems, and development of operating procedures for these systems. He also i supported plant licensing efforts for the two projects and routinely interfaced with NRC Staff in order to resolve open issues. Proposal Manager at_ United Engineers for the Advanced Packaging Facility Concepts Study Proposal to Battelle's Office of Nuclear Woste Isolation. h 20

DONALD A. GARDffR, JR. O (continued) b Program Coordinator for the Anticipated Transients Without Scram (ATWS) Program at General Atomic Company. In this capacity, he was responsible for ensuring the technical odequacy and timely completion of technical analysis and evoluotion of results. Coordinated major General Atomic Company design reports which described the resolution of the primary system parameter-related technological issues of the HTGR-GT Plant. Responsible for writing the bid specification and procurement activities for the Seabrook I and 2 Radweste Volume Reduction and Solidification System. Responsible for the design, analysis and system sizing of the Core Auxiliary Cooling System for the HTGR NSSS. He was oiso responsible for the design optimization of the HTCR-GT plant, using the CODER computer program. Lead Safety Engineer on the Ohio Edison Erie Nuclear project for PSAR Chapters 15 and 16. Lead Engineer responsible for core analysis and follow-up of operation for the two PWR's at Surry. Mr. Gardner was also a Lead Startup Engineer during the fm initial startup and criticality of Surry Power Station and performed nearly all d phases of the physics testing. In this capacity, Mr. Gardner co-authored the VEPCO submittals to the NRC, documenting physics tests results for the Surry I and 2 units. Mr. Gardner acquired extensive training and experience in the use of the major con puter codes used in the industry to evoluote and design nuclear fuel. Reactor physics and thermol/ hydraulics analyses were performed for the Nine Mile Point plant, relative to BWR fuel cycle optimization and reload assembly design. PUBLICATIONS: "Startup Physics Test Program at Surry Units I and 2," Transactions of the American Nuclear Society, June 1974. " Modular Interim Waste Storage Building for Low-Level Rodwoste," Waste Management '83, February,1983, t% k._ 111111111111111111111111111111 21

THitN DUC NGUYEN EDUCATION: Doctorate, Mechanical Engineering, University of Lyon, France Post Graduate Certificate, Applied Mechanics, University of Lyon, France M.S., Mechanical Engineering, Ecole Centrole de Lyon, France Certificates in Mechanics, Engineering Mothematics, Fluid Mechanics and Engineering Electrics, University of Lyon, France j PROFESSIONAL REGISTRATION: Registered Professional Engineer, California PROFESSIONAL EXPERIENCE: As a Senior Engineer at Cygna, Dr. Nguyen is currently assigned as the piping project engineer for the Yankee Nuclear Power Station at Rowe, Masochusetts. This work includes the stress analysis of the piping to the SEP requirements. Dr. Nguyen has personally performed the onolyses of those systems requiring special techniques such as displacement time history analyses or inclusion of the structural mass and stiffness in the piping model. g Dr. Nguyen was previously assigned as pipe stress group leader for the La Salle Unit 2 CRD piping analysis, in this function, he was responsible for issuing design criteria and work instruction, ccordinating work with the frame analysis group, cnd liaison with the client. Dr. Nguyen performed parametric studies which allowed the large number (370) of CRD lines to be qualified by the analysis of very few. In a similar position for the La Salle Unit I CRD piping, Dr. Nguyen's responsibilities included: sensitivity study of static, seismic, and hydrodynamic analyses of the CRD system composed of 370 similar lines. Analysis was principally performed through mode shape studies. evoluotion of seismic anchor movement, Annulus Pressurization displacement from time history data. generation of matching response spectro from time history and envelope spectro to use for each system. time history analysis for Annulus Pressurization displacements. study of c simplified model for the Hydraulic Control Unit. t i V 22 k

THlMi DUC NGUYEN O ~ (continued) U ^ establishing standards, such as charts related to maximum mass point spacing versus pipe sizes based on cut-off frequencies, and coding procedures conforming to ANSI B31.1 standards. writing procedures and final reports. Dr. Nguyen's other project work included static and dynamic analysis of class I and 2 piping systems in accordance with applicable codes and standards such as ASME lil, B31.1 for plants such as Vermont Yankee, Arkansas, Susquehanna, and Diablo Canyon. These analyses included the study of behavior of supports, finding the appropriate type of support through load, stress, and mode shape considerations; selection of spectro to be used according to eccentricity, elevation, location of attachment points, and envelope of spectro; evoluotion of the opplicability of previous thermal analysis to the suggested changes to the systems (cutting a relatively big system to small ones and using the overlapping techniques). In the performance of the work detailed above, Dr. Nguyen has acquired extensive experience in the use of computer programs such as PIPESD, INSPEC, ADLPIPE, NEWSPECTRA, and ANSYS. Dr. Nguyen's previous industry experience included serving as a senior engineer for an Q American architectural / engineering firm based in Saigon, Viet Hom. During this time he V concurrently provided private consulting engineering services foi a construction firm in Saigon, Viet Nam, which involved the study of unsteady flow in canal networks, hydraulic reduced scale models of outlets, gates, doms, and basins of dissipation of energy. Dr. Nguyen's academic experience includes holding the position of Professor and Dean of the School of Engineering, National Institute of Technology, Saigon, Viet Nam, for eight years. For five years, he was Assistant Professor at Ecole Centrole de Lyon, France. Dr. Nguyen concurrently performed research in the reduced scale compressor project for the Chatou Thermal Power Plant, France. TFESIS: " Study of the Secondary Effects of the F!ow of the Extremity of Blades in on Axial Com-pressor." The research was closely related to the rotating stoll phenomeno in oxial compressors. .O V 118111111llllll1llll1111111lll 23 L

EXHIBIT 3.2 l PO BOX 97 e PERRY. C H1'O 44081 m TELEPHONE (216) 259-3737 e ADDRESS-10 CENTER ROAD Serving The Best Location to the Nation JIMMY E. MEYER PERRY NUCLEAR POWER PLANT ENGINEER GENERAL BACKGROUND Mr. Meyer has experience in all design aspects of flexibility analysis and support of piping systems for chemical, petroleum and petrochemical processing facilities. He possesses a broad background in both manual and computerized techniques for solution of complex piping stress analyses. Since 1979, he has become active as a member of ANSI /ASME B31.3 Code Committee in the work group responsible for the design of Petroleum and Chemical Plant Piping. EXPERIENCE Cleveland Electric Illuminating Since joining the Cleveland Electric Illuminating Company in October, 1982, Mr. Meyer has been assigned to the nuclear plant being built outside of Cleveland. ("} Davy McKee v Summary of his job responsibilities at Davy McKee are listed below: 1961-October, 1982 - Senior Group Supervisor - Purchasing Responsible for purchasing and expediting of fab-ricated pipe and engineered pipe supports. Purchasing - Preparing inquiry packages. - Evaluating quotations and preparing bid tabu-lations. - Writing purchase orders based on lump sum or unit price agreements. - Miscellaneous purchasing of other engineered items. Expediting - Home Office expediting of Davy McKee isometric production and transmission to vendors. - Vendor expediting of: J.. Spool detailing from isometric drawings. B. Expediting of materials to vendor's shop. C. Expediting of vendor's fabrication and adherence to priority schedules. k f

Jimmy E. Meyer Page 2 -- t ) Davy McKee (continued) 1978-1981 Senior Group Supervisor - Piping Engineering (Approximately 25 engineers) Areas of Responsibility: A. Piping specifications. B. Piping line list development. C. Piping flexibility and stress analysis. D. Pipe support design. E. Review of isometric drawing to specify support and flexibility requirements. F. Pressure test circuit analysis. G. Field resolution of piping problems during plant start-up or operation. 1973-1978 - Engineer in group described above. Some of the major projects which he has participated in are. - Shell Oil Company, Marietta, Ohio - Thermoplastic Rubber Plant. - Carter / Exxon Co., Bayport, Texas - Coal Liquer... tion Pilot Plant. - Shell 011. Company, Geismar, Louisiana - Ethylene Oxide /Ethylen'e Glycol. (('N - Gulf 011 Chemicals, Cedar Bayou, Texas - Polypropylene. - Shell Oil Company, Geismar Louisiana - Chemical Processing r s facilities. - Diamond Shamrock Chemical Co., Greens Bayou, Texas - Fungicide Processing Facilities. - Esso FIOR, Puerto Ordaz, Venezuela - Iron Ore Direct Reduction. - Mobile Oil, Beaumont, Texas - Fuel Oil Desulfurization. - Texaco 011 Company, Nanticoke, Ontario, Crude & Fluid Catalytic Cracking Units. - Nipro, Inc., Augusta, Georgia - Caprolactam Plant. PROFESSIONAL DATA B.S., Mechanical Engineering, The University of Akron, Akron, Ohio. Registered Professional Engineer: Ohio Member: ANSI /ASME B31.3 Code Committee. (Until November, 1982) ANSI /ASME B31.1 (Appointed as a subgroup member 1983) DW119/G/2/sp-f3 -V s

iHE CLEt'ELEi 0 EC7~ i ! L ! :' " t ~ ~ ;"?A"Y hh2 Piping Design S " 4o ~ I TSU!GO@c O PAGE: 1 of 1 Review REVISiCN: 0 4.0 METHODOLOGY The review was conducted to predetermined checklists and acceptance criteria which are contained in the Project Manual for the piping design review. Separate checklists and acceptance criteria were prepared for each of the disciplines reviewed. 1. Mechanical 2. Piping Analysis 3. Pipe Supports Any discrepancies noted during this review were documented on the appropriate checklist. Significant items were resolved with CAI as [); observations. The identification documentation, and resolution e %/ cycle is described in Figure 6. CEI reviewed all discrepancies, as documented in the Cygna Final Report. GAI was responsible for determining reportability of all discrepancies per Appendix E of their Nuclear Quality Assurance Manual. CEI also reviE&ed all observations with respect to generic application to other systems. Any deficiencies found during this review resulted in the initiation of a CEI Engineering Design Deficiency Report (EDDR) per CEI Procedure 35-1501. p (v - PERRtNUCLEAR POWER PLANT - Serving The Best Location in the Nation PO BOX 97 e AEARY OHIO 44081 e TELE AMONE e216e 259-3737 e ADDRESS-to CENTER ROAD e

P7 PING DESIGN REVIEW FLoURE +6 METHODOLOGY I I CYGNA PROJECT TEAM. GILBERT REVIEWER REVIEW BOARD ASSOCIATES g TECHNICAL.EVIEW USING PREDETERMINED CHECK LISTS AND ACCEPTANCE CRITERIA, 4 ~ NOTE DISCREPANCY ON CHECK LIST 4 . REVIEW DETERMINES

lPROJECTTEAM' CONSULTED FOR GAI ENGINEERS r

POSSIBLE AGREES -e O [ L CLARIFICATION, SIGNIFICANCE p; YES IF REQUIRED NO OBSERVATION lDl-GAIRESPONSE-ISSUE RECORD l SUPPLIES JUSTIFICATION FOR DESIGN ? OR [ IDENTIFIES CORRECTIVE l' ) . ACTION BEING 'v' NO TAKEN REVIEW BOARD ACCEPTS RESPONSE-l ACCEPTANCE = YES l ' BASED ON ADEQUACY ' ($),(f) i OF DESIGN OR GAI 1 r OA PROGRAM GAI HAS i ADDRESSING RESPONSIBILITY THE PROBLEM TO EVALUATE FOR REPORTABILITY UNDER 10 CFR 21 7, DOCUMENTATION yp OF REVIEW REPORTABLE

l " FINAL REPORT" 1

1 P REFORT TO NRC PER QAD 600 1 r CEI FOLLOW-UP REVIEW OBSERVATIONS ,I FOR DESIGN l DEFICIENCY AND INITIATE 1 RESOLUTION ON AN EDDR 6 FORM l 4 e

u..-

vt.- -r r-- n ti i n i ..sa.. L.. s i.. : o 4 .w u. i.... hhN Piping Design. S" " 5.0 2#M O PAGE; Rev,iew t.EVISION 0 5.0 OBSERVATION FOLLOW-UP CEI was responsible for review of the resolution of all observations. Also, CEI completely reviewed all of the Cygna observations for any generic issues. A summary of this review is included as " Attach-ment B" to each observation. CEI considered the observation closed if any of the following conditions were met. 1. The Cygna observation did not represent a deficiency in the design, or documentation. 2. A generic review has been completed which verified the item has no effect on any system's ability to perform _ _, its intended safety function. 3c A generic review will be performed to approved procedures and tracked by Engineering Design Deficiency Report (EDDR). The procedures are developed specifically to address the generic issues. 7-(,) This section is organized in the following manner: 5.1 Mechanical - Observation Status Mechanical Observations - including: Cygna Observation Record Cygna Observation Record Review - Attachment A CEI Observation Closure - Attachment B 5,2 Piping Analysis - Observation Status Piping Analysis Observations - including: Cygna Observation Record Cygna Observation Record Review - Attachment A CEI Observation Closure - Attachment B 5.3 Pipe Support - Observation Status Pipe Support Observations - including: Cygna Observation Record Cygna Observation Record Review - Attachment A CEI Observati_on Closure - Attachment B c CERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation PO 80x 97 e PERRY QMiG 44081 e T E L E PHONE (216i 259 3737 o ADOAESS to CENTEA AOAo

I THE CLEVELAND ELECTRIC IL L UMlH A TIN G COMPANY MGEEN Piping Design S' " ". 5.1 ) O PAGE: Review ,,,s,o s ; o 5.1 MECHANICAL OBSERVATION STATUS REVISION O DATE 5-11-84 EDDR NO. F'OLLOW-UP SCHEDULED OBSERVATION DEFICIENCY OR ACTION COMPLETION DATE NO. YES/NO GC PRE NO. COMPLETE FOR FOLLOW-UP C0FDfENTS ME-01-01 YES 72 NO JUNE 30, 1984 ME-01-02 YES 72 NO JUNE 30, 1984 ME-02-01 YES 72/73 NO JUNE 30, 1984 ME-02-02 NO NA NA NA ME-02-03 YES 72 NO JUNE 30, 1984 ME-02-04 YES GC PRE-83 YES, 1/16/84 NA GC PRE-85 ME-02-05 NO NA NA NA ME-02-06 YES 72 NO JUNE 30, 1984 ME-02-07 NO NA NA NA ME-02-08 YES 72 NO JUNE 30, 1984 ME-02-09 YES 72 NO JUNE 30, 1984 ME-03-01 YES 72/73 NO JUNE 30, 1984 ME-03-02 YES 72 NO JUNE 30, 1984 ME-03-03 YES 72 NO JUNE 30, 1984 ME-03-04 YES 72 NO JUNE 30, 1984 ME-03-05 YES 72 NO JUNE 30, 1984 PERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation PO DOx 97 e PERAY. OHIO 44081 e TELEPHONE 1216) 259 3737 m ADDAESS 10 CENTER AOAD

Observation MDh' M Record 11111111111111111!I11111111111 Revision No. O Observation No. ME-01-01 sheet i of I Checkilet No. ME-01 MSSRVS Item #4 f 7 /j /g3 ) Date cristneted sy ge SQ, Date /2/9 /G gh [1,*_ 7 noviewed sy O O . 1.0 Description Safety relief valve discharge line sizing (flow and pressure drop) calculations could not be located by GAI. 2.0 Requirement Per the Perry FSAR Section 5.2.2.2.3.3, the discharge line is sized to prevent the backpressure on each safety / relief valve from exceeding 40 percent of the valve inlet pressure. The GE Process Diagram 105D5575 also states that the ASME relieving capacity of the S/RV's only applies when the back pressure at the discharge side of the S/RV's is,< 40% of the S/RV inlet pressure with a flow rate corresponding to nameplate. 3.0 Reference Documents O 3.1 Perry FSAR Amendment #7 (5-27-82), Section 5.2.2 3.2 Nuclear Boiler Specification, 22A4622, Rev. 5 3.3 Nuclear Boiler Data Sheet, 22A4622 AR, Rev. 2 3.4 Process Diagram Nuclear Boiler, 10505575, Rev. 0 3.5 Design Specification, DSP-B21-1-4549-00, Rev. 2 4.0 Potential Design Impact Due to the lack of verifiable and documented calculations, the adequacy. of the. S/RV discharge line size cannot be determined. However, per the Perry Supplemental Safety Evaluation Report #3 Table 6.4, the Perry S/RV discharge line size of 10" is the same as two similar nuclear power plants (Kuosheng and Grand Gulf). 5.0 Probable Cause l Document control. Attachments A. Observation Record Review Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review \\

i Observation [4 M'M Record Review ,g '""""""""""'"""" Attachment A Observation No. ME-01-01 Checklist No. ME-01 Revision No. O PFR No. Sheet 1 of } Yes No Closed X Extent 3 of 3 Systems with missing calculations Comments GAI submitted portions of piping engineering calculation P203, Rev. O, dated 1/20/83 as verification that the safety relief valve discharge piping was adequately sized. The original purpose of this calculation was to perform a thermal-hydraulic transient analysis on the MSSRV discharge piping and to generate a hydraulic transient force history for input to the TPIPE time history dynamic analysis. However, the submitted portions of calculation P203 do show that the discharge piping backpressure will be equal to or less than 40% of MSRV inlet pressure at a rated flow of 1.12 x 106 lb/hr. This meets the GE and FSAR requirement for this piping. Based on the above, this Observation does not have any impact on design or safety. l l 1 l Approvals ///6 /r y oe, n, w. h oa Project sne, 4 a 1 c om lhv/sq. Pro).ct Manag.r [L/,/g l[/4 Date Co R.p,. / yhN oa //ed/e4 Cleveland Elect,ric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review k

( 'J:El Observation Record Closure Attachment B j Observation No. Checklist No. Revision No. ME-01 -01 ME-01 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No CI: sed X is lated X P:tential Design impact X 1.0 Description See Cygna cbservation record and observation record review. 2.0 Discussion This was not a design deficiency, however, compliance with a GE requirement was not documented. 3.0 Action Taken A GE Criteria Compliance Review is being performed by gal. 7 O.0 Conclusion The review listed above will insure documentation of compliance with GE requirements. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compliance Review Procedure, Rev. O Approvals Originator g Date S:nior Proje Eh r Date CEi Supervisor Quality Aud t Unit Date Y/bl3'4 WA. 1P gal Project Minager 7 Date h !, n.l e. 5/W89' gal Manager Corpor TP/ogra ,Date y Tho Cleveland Electric illuminating Co~mpany: _Pcrry Nuclear Power Plant Piping Design Review DW138/9/Q/sp, l.

Observation h%i Record ilillllllilllilllilillllitilli j gU Rolsion No. g Observation No, ME-01-02 sheet of Checklist No. ME-01 MSSRVS. Item No. 3 y p Date /7ff/g3 Criginated By f,Q fM Date l2.[$ [83 { Reviewed By f \\ Q n,*_, j y Q 1.0 Description Vacuun. breaker valves F037 and F038 are 6 inch valves with a maximum resistance coefficient of K = 1.6 as specified in GAI Specification SP-639-4549-00 Rev. 1. Per information supplied by the vendor, Anderson, Greenwood and Co., the actual K = 1.408 and the flow area is 0.201 ft.2 This data results in an A/R factor equal to 0.17 ft.2, rather than the General Electric specified minimum of 0.30 ft.2 for each of these valves. In addition, no documented and verified calculations justifying the size of these valves could be located by GAI. 2.0 Requirement Genen1 Electric Specification 22A4622 Section 4.3.3.5 requires that two Idrallel vacuum relief valves be provided on each relief valve discharge line to m.nimize drawing water up into the line due to steam condensation following m termination of safety / relief valve operation. General Electric Specification Data Sheet 22A4622AR Section 3.1.20.1.2 states that the vacuum breaker A//T 2 ratio sht11 be equal to or greater than 0.30 ft. K is the effective loss coefficient of the vacuum breaker and its connecting pipe to the S/RVDL. 3.0 Reference Documents 3.1 Nuclear Boiler Specification, 22A4622, Rev. 5 3.2 Nuclear Boiler Data Sheet, 22A4622AR, Rev. 2 3.3 Specification for Vacuum Breaker, SP-63-4549-00, Rev.1 3.4 Anderson, Greenwood and Co. Assemt,1y, 6"-300 ANSI, CVIB SPCL Vacuum.. Breaker Valve N04-2217-530, Rev. D. O) \\_ Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record 4La i 111IIlllll1111llll11111111llll Revision No. 0 Observation No. ME-01-02 sheet 2 2 checklist No. ME-01-MSSRVS, Item No. 3 Date /2 /1)M Criginated By f, W, ph Date gf g /gg 0;wlewed By (M,

• _

y u o 4.0 Potential Design Impact Due to the lack of documented and verified calculations, the adequacy of the specified valves cannot be determined. Per the Perry Supplemental Safety Evaluation Report No. 3 Table 6.4, similar plants (Kuosheng and Grand Gulf) have two 10 inch vacuum breaker valves on each SRVDL instead of the 6 inch valves specified for Perry. The Perry SS~gR 3 Section 6.2.1.8.2 (Pg. 6.3) states, "This criterion (A/fK = 0.30 ft ) is met by the two 6 inch vacuum breakers at Perry." However, the General Electric Specification Data Sheet 22A4622AR indicates this criteria should be met by each valve and not by the sum of the two valves. 5.0 Probable Cause Design control. Attachments A. Observation Record Review i t l l Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation [.4 h' M Record Review O '* * * **"'" * " Attaehment A Cbservation No. ME-01-02 Checklist No. ME-01 Revision No-0 Sheet } of } PFR No. Yes No Closed X Extent 3 of 3 Systems with valve data inconsistent with GE Requirements Commentt Per the attached GE/GAI telecon of November 2,1983 (E. W50d, GE, to T. Daugherty, 2 criteria in the GE specification is to be interpreted as GAI), the A/fK = 0.30 ft the total ratio for both vacuum breaker valves. The vacuum breaker design provides 2 = 0.34 ft, which satisfies GE's requirements as explained in the 2 2 x 0.17 ft referenced telecon. Based upon this telecon, there is sufficient documentation to justify the sizing of l these valves. Accordingly, there is no impact on design or safety. O. i Approvals Date / /4/g3 I Originator , V, 7h "*~ "" ^* ~ O 'rj"c"t Manager V

1. f Date f

f/g Poe y /2,f/g/g3 bepresentative gg Date y Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review ~.

p7-yg, - 7 ~ / lEl Observation C' Record Closure Attachment B . Observation No. Checklist No. Revision No. ME-01-02 ME-01 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Closed X isslated X P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.6 Discussion This was listed as part of EDDR 72 to improve documentation of compliance with GE requirements. The design by gal did meet the GE requirements and the only documentation missing was an interpretation of tho GE requirement. 3.0 Action Taken p/ A GE Criteria Compliance Review is being performed by GAI. C 4.0 Conclusion The review listed above will insure documentation of compliance with GE requirements. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant-GE Criteria Compliance Review Procedure, Rev. O. Approvals Originator Date g S nior Projfe xAJ r/s/s+ Engineer Date l CEI Supervisor Quality ' Audit Unit Date b v t /u/ry gal Project Manager '~ Date 6~/ Way .,+ w gal Manager Corporat r' Date Tha Cleveland Electric illu'minating Company: Perry Nuclear Power Plant Piping Design Review [ W138/10/Q/sp

Observation [.4 1 %' M Record lilllllillllllittlilllllilllll Revision No. O Observation No. ME-02-01 of sheet 3 p Checklist No. ME-01 HPCS, Item No. I f g /j h Criginated By [ p, Date Reviewed By d* p Date gg /g 1.0 Description There are various inconsistencies between Table 1 of GAI Specification DSP-E22-1-4549-00 Rev. I and Rev. 2 and the General Electric Process Diagram 762E455. Specifically: a. GAI Table 1 defines both design conditions and operating conditions for the HPCS. In one region of the system, the operating conditions (234 psig 0104 F) exceed the design conditions (100 psig 0 212*F). Specifically, this occurs at locations 16,17 and 27 for operating mode B. 1 b. GAI Table 1 lists the pressure above the suppression pool as 15 psig in modes D through J. The GE diagram lists this pressure as 14.7 psia _. c. GAI Table 1 location No.1.5 pressure is stated to be 36 psig. This is O-higher than would be achieved by adding the static head of Weter in the tank to the General Electric stated atmospheric pressure of 14.7 psia in the tank. d. In GAI Table 1 for modes D through G, the difference in pressure between the source of suction and the reactor vessel does not match the General Electric requirements of 1550 gpm 01147 psid and 6110 gpm 0 200 psid. e. In GAI Table 1, mode H, the pressure at locations No.16, No.17, and No. 27 should be the same. Location No. 27 is given as 15 psig while No.16 and No. 17 are given as 25 psig. 2.0 Requirement General Electric Specification 22A3131, Data Sheet 22A3131AS and Process Diagram 762E455 are the design basis documents. They provide flow, pressure, and temperature data for which the system must be designed. l 0 J Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation A L% i Record tilllillllilitiiiiiiiiiiiiiiii Revision No. O Observation No. ME-02-01 sheet p of 2 Checklist No, M[.01 HPCS, Item No. 1 Criginated By f, W, h Date /1/j/f3 c.vi...e er Qu a,. W oata tale /en g 3.0 Reference Documents 3.1 Design Specification, 22A3131 Rev. 5 HPCS 3.2 HPCS Data Sheet, 22A3131 Rev. 2 3.3 Process Diagram, 762E455 Rev. 6 3.4 Design Specification HPCS, DSP-E22-1-4549-00 Rev.1 3.5 Design Specification HPCS, DSP-E22-1-4549-00 Rev. 2 4.0 Design Impact Since the GAI design specification is used for piping and pipe support design, inconsistencies in pressure, temperature, and flow data could cause O inaccuracies in this design effort. It is not clear what other design functions (valve sizing, I & C, etc.) use Table 1 data as design input information. 5.0 Probable Cause Failure to document the resolution of differences between corresponding General Electric and GAI specifications. Attachments A. Observation Record Review l l i i Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Obsorvation Record Review 4( < i Attachment A C, unaninninununu Cbservation No, ME-02-01 Checklist No ME-02 Revision No. O Sheet 1 of 1 pfrH No. Yes No Closed X Extent 2 of 3 Systems with inconsistencies between GE and GAI data Comments Based on the following GAI data and commitments, this Observation does not have any impact on the design or safety of components or systems within the scope of this review. GAI will revise the system design conditions portion of Table 1 in a. DSP-E22-1-4E49-00, Rev. 2, to reflect design conditions that envelop all system operating conditions. b. GAI will revise the Mode H operating pressure at locations #16 and #17 in Table 1 of DSP-E22-1-4549-00, Rev. 2, to be consistent with location #27, i.e., 1 ' O 15 Psis. c. GAI does not intend to correct any of the other inconsistencies and/or inaccuracies in Table 1 of DSP-E22-1-4549-00, Rev. 2. The GAI reason for not making addition.al revisions to this table is that the existing data is conservative for use in the design of system piping and pipe supports. As indicated by GAI in the title and Section 1:01 of Specification DSP-E22-1-4549-00, Rev. 2, Table 1 is intended to be used solely by the piping analysis and pipe support design groups. In addition, GSI has stated in various discussions that no other GAI procedures (other than piping procedures) specifically require the use of data in the E22 piping design specif

• cation as design input for other system / component design. Based on the fact that the systems review was limited to those items which may affect the piping analysis and that the existing Table 1 data is conservative for this purpose, Cygna concurs that a general revision to the Table is not required at this time.

Approvals o='* ////Jr4 l orson.<.c fl. -z.), f~u n oa ///s/sf rro ut une , J /A C x Or-er u R7=rr>r b/e u4-o '- // d / W 4(#W-c a'* l cui n r Cleveland Electric Illuminat ng; 83102 Perry Nuclear Power Plant Piping Design Review i

y- .i'z ~ : ;;;, f ' ~ (JEl Observation V Record Closure Attachment B Observation No. Checklist No. Revision No. ME-02-01 ME-01 0 EDDR No. QAD 600 No. Sheet of 72 and 73 N/A 1 1 Yes No Cl:: sed X is: fated X Potential Design impact X 1.0 Descriptior. See Cygna observation record and observation record review. 2.0 Discussion 2.1 Item 1) of this observation addresses design conditions and will be addressed by EDDR 72. 2.2 Items b, c, d, and e of this observation will be addressed by EDDR 73. 2.3 Subsequent review per EDDR 73 has resulted in a decision to review items b, c, d, and a under the GE Criteria Compliance Review. .0 Action Taken GE Criteria Compilance Review is being performed by gal. 4.0 Conclusion The Review listed as 3.0 above will insure that correct Design Conditions are listed. 5.0 References (9) EDDR 72 (10) EDDR 73 (25) Perry Nuclear Power Plant GE Criteria Compliance Review Procedure, Rev. O. Approvals Originator Data g ppg Senior Pro ngin r Date / MAM f8 8b / CEI Supervi A it Unit Date gal Project Manager Date ~~ b 5/ WOY -r-gal Manager Corpora Pfogr Data f (_Tha Cleveland Electric illuminating Company: / ~ ' rry Nuclear Power Plant Piping Design Review / \\_ DW138/23/Q/sp C

Observation [41 O Record llllllllllllllllfillllllllllll Revision No. 0 Observation No. ME-02-02 Sheet of Checklist No. ME-02 HPCS Item No. 1 3 ) Date jt /j / p Criginated By f_4, g gdg/gg Reviewed By %h Q,y Date V L) 1.0 Description In GAI Specification DSP-E22-1-4549-00 Table 1, the Hode A pressure drop across valve F010 is given as 522 ft., and the drop across valve F011 is given as 116 ft. These drops are well above the General Electric stated minimum of 62 ft., indicating that the valves are not fully open in mode A. Also, these pressure drops (throttled position) were not used in the flow and orifice sizing calculation for the system. 2.0 Requirement General Electric Process Diagram 762E455, Note 8, states that a 62 ft. pressure drop is the minimum drop for these valves and that they may be throttled to facilitate the piping arrangement. Note 16 of this process diagram recommends installing orifice R0-D004 to limit flow to 6110 gpm with valves F010 and F011 fully open. 3.0 Reference Documents 3.1 Process Diagram, 762E455, Rev. 6 HPCS 3.2 HPCS Design Specification, DSP-E22-1-4549-00, Rev. I and Rev. 2 3.3 Calculations HPCS Line losses, E22 A/J-CC Dated 2/8/79 4.0 Potential Design Impact The orifice, R0-D004, was sized based on (1) both F010 ad F011 being fully open and (2) dissipating an excess head of 945.3 ft. If valves F010 and F011 are throttled as indicated in Table 1 to absorb an additional 514 ft. of head [(522 - 62) + (116 - 62).), then the total system pressure drop at 6110 gpin will exceed the available head at this flow. 5.0 Probable Cause De.iign control. Attachments A. Observation Record Review lO i Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

b Observation L41 &'fd Record Review -Attachment A Observation No. ME-02-02 Checklist No. ME-02 Revis!on No. 0 PFR No. Sheet 1 of 1 Yes No X Closed Extent 1 of 3 Systems with inconsistent use of GE data Comments GAI has stated that they will revise Table 1 of Specification DSP-E22-1-4549-00 to indicate a pressure drop of 62 ft tr augh valve F010 and 62 ft through valve F011. In addition, the revised specificat: 1n will indicate that the remaining excess pump head is dissipated by orifice R0-D004 This is in accordance with the calculation of reference 3.3. GAI also verified in a telecon with Cygna on 11/16/83 that these changes to Table 1 will not affect any other design calculations, drawings or specifications. Based upon the above GAI staceiaents, this observation does not have any impact on design or safety. l l 4 r.. originator g,@ h D t- /e /4 /fa: gd/g Project Engineer Q h-Date O>e-M. i. -w>.4 ' i4/4/ss i Date fg/g/gy cai R r ntative h eJm U r r l Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review l ..--r,. e +,

w>=. m .-in wm-wm ( )El Observation '~' Record Closure Attachment B Observation No. Checklist No. Revision flo. ME-02-02 ME-02 0 EDDR No., QAD 600 No. Sheet of N/A N/A 1 1 Yes No CI:: sed X is: lated Not applicable P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion This observation resulted from Cygna using the data for something other than its intent. 3.0 Action Taken None required. (~'s Gj.0 Conclusion Conservative pressure drop values used by gal were taken by Cygna to indicate the valves are not fully open. This was an extrapolation of the data in the design specification beyond its intended purposo. This information has no effect on the piping design and, as Cygna noted, was not used in flow and orifice sizing. 5.0 References ~ (26) gal letter PY-gal /CEI 15478. Approvals ~ Originator Date g Senior ProjeFc. x x ~ -/ r /. l a ngineer Date a CEI Supervisor Quajity dit Unit Date / W1 Y/d l% 4 gal Project Manager & ' ' + N7/8Y Date gal Manager Corporap g Date he Cleveland Electric illuminating Company: [ erry Nuclear Power Plant Piping Design Review DW138/11/Q/sp

- - _ - - = - _ _ _ _.. Observation Record Mn i illlllilllllllllilllllllllllli gV Revision No. g Cbservation No. ME-02-03 sheet of Checklist No. ME-02 HPC3 Item No. 2 i p j7/j/gy Date Criginated By f, @, %p Date g/9/gg Reviewed By [dhf & \\J LJ 1.0 Description The location and arrangement of some equipment and piping is inconsistent with General Electric and NRC Criteria. Specifically: a. The HPCS suppression pool suction strainer is not located outside the safety relief valve discharge zone. b. Valve F023 is located approximately 14 ft. from the containment penetration. It should be located as close as practical to the penetration. Normally a distance of 5 ft. or less is achievable. c. The length of straight pipe after a valve and prior to flow orifice N007 does not meet the 43 ft. requirement. 2.0 Requirement a. General Electric Specification 22A3131, Section 4.2.4.6, states that the HPCS suction strainer shall be located away from safety relief valve ischarge zones. b. Both General Electric Specification 22A3131, Section 4.2.3.13 and 10CFR50 Appendix A Criterion 56 require that outside containment isolation valves, such as F023, be located as close to the containment penetration as practical. c. Per General Electric Specification 21A9505BV, Rev.1, 5 action 4.3.1.1 there should be 43 ft. of straight pipe between the outlet of a valve and the inlet of the flow measuring orifice. 3.0 Reference Documents 3.1 Design Specification HPCS, 22A3131, Rev. 5 3.2 General Design Criteria,10CFR50 Appendix A 3.3 Flow Orifice Assembly HPCS, 21A9505BV i Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review ~

Observation 9Ph i Record llllllllllllllllllllllllll1111 Revision No. O Observation No. EE-02-03 sheet 2 of 2 Checkitet No. ME-02 HPCS, Item No. 2 f dj jfy Crisinated By f, W, fh Date g gf g /g Reviewed By QQg ,g Date Q L) 3.4 Drawings 3.4.1 HPCS Plans and Sections D-304-701 3.4.2 HPCS Sections D-394-702 3.4.3 HPCS Reactor Building El. 620'-6" and 574'-10" D-304-703 3.4.4 MSSR Piping Inside Reactor Building El. 574 -10" and 599'-9" D-304-026 3.4.5 Discharge Quencher 767E676 I.C.D 3.4.6 Quencher Arrangement Design Envelope B-301-734, Rev. J O 4.0 rotentie: nesion I. Pact a. The location of the HPCS suction strainer within the quencher discharge zone could cause air or steam entrainment in the hPCS pump suction line. b. The location of F023 away from the containment penetration provides a greater length of nenisolatable piping which could lead to a breach of containment if it failed, c. The accuracy of flow orifice N007 could be affected by its proximity to the valve located upstream. 5.0 Probable Cause Design ove ight and lack of documentation of design variances. Attach ents A. Observation Record Review l Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review -~

Observation Record Review M i i amii;;;;;;;;iiiiii;;"" Attachment A Observation No. ME-02-03 Checklist No. ME-02 Revision 14o. O Sheet } of 1 PFR No, Yes No Closed X Extent 1 of 3 Systems with nonconformance to GE Equipment arrangement requiretrents Comments Based on the following GAI and GE data and documentation, this Observation coes not have any impact on design or safety. General Electric approved the location of the HPCS, LPCI, RCIC and RHR suction a. strainers within the SRV discharge quencher zones in Field Deviation Disposition Request No. KL1-301 approved on 6/6/83. This approval was based on I the pump vendor certification that the quantity cf ingested air (40% maximum in 1.5 seconds) is acceptable for pump operation. i I b. GAI has stated, based upon their review of the piping arrangment, that due to the proximity of other piping and the valve operator size, F023 cannot be located any closer to the containment penetration. GAI has stated that the current piping arrangement will trovide the 1% accuracy i c. specified for flow element E22-FE-N007 GE concurrence with the existing i piping arrangement was requested by GAI in letter PY-GAI/ GEN-2931, dated 12/30/83. i l l Approvals Date ///g/fsy originator 77, Q SA:;- Dete p/g3 /g4 Project Engineer d k _*___. % Date Q/g Project Manager h fyj can Represenistive gph& onte //yJ7/gp ClevelandEledricIlluminaIingk83102 Perry Nuclear Power Plant Piping Design Review

sw -c- ~.. = - -.= - -n ~ - ~ 'El Observation t Record Closure Attachment B Observation No. Checklist No. Revision No. ME 02-03 ME-02 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Closed X isolated X Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion These examples of incorporation of GE critoria have been accepted by GE as required, but documentation to that effect was not available in all cases. 3.0 Action Taken A GE Criteria Compilance Review is being performed by gal. U 0 Conclusion The review listed above will insure documentation of compliance with GE requirements. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compilance Review Procedure, Rev. O. Approvals Originator Date g, Senior Proje ngieer Date CEI Suparvisor Qualjty A dit Unit Date WA. W/d/T 'l au-gal Project Manager f Date th

• :n s/7/89 n-gal Manager Corpor

/_r Date y he Cleveland Electric illuminating Company: [ rry Nuclaar Power Plant Piping Design Review DW138/12/Q/sp h. k

Observation M M'TJ Record lillllllllllillllllllillllllli Revision No. O Observation No. ME-02-04 sheet i e' 2 Checklist No. ME-02-HPCS, Items No. 7 and 24 jz}j}f3 Otleinated By y, Q & Date g)q [p Reviewed By M Q, [ _ ,A Date \\ U G 1.0 Description The vendor print (Rockwell) for valve F005 indicates this valve is a lift check valve with no stem (i.e., no stem leak-off connection) or external operator for remote testing. In addition the pressure and temperatures indicated on the drawing approximately match a 600 lb. class valve. The General Electric data, CEI SAR and GAI P & ID all indicate this valve should be a remotely testable swing check valve with an air operator and stem leak-off connection. In addition, line specification D1-1 recommends valves of this size be 900 lb. class valves. 2.0 Requirement General Electric Specification 22A3131, Section 4.2.3.3 states that a testable check valve shall be provided in the HPCS discharge line inside the drywell. O The General Electric P & ID for the HPCS system, 795E873, indicates this valve has an air operator and stem leakoff connection. 10CFR50 Appendix A criterion 37 requires that the HPCS be designed to permit functional testing of the operability and performance of the active components of the system. 3.0 Reference Documents 3.1 HPCS Design Specification, 22A3131, Rev. 5 3.2 General Design Criteria,10CFR50 Appendix A 3.3 Amendment No. 3 Section 6.3.2.2.1, Perry FSAR 3.4 Drawings 3.4.1 HPCS P & ID, D-302-701, Rev. G l 3.4.2 Piping Design Specification HPCS, D-320-701, Rev. C 3.4.3 HPCS Reactor Building Elevation 620'-6" and 574'-10", D-304-703 Rev. G l 3.4.4 HPCS P & ID, 795E873, Rev. 1 3.4.5 Rockwell International Testable Piston Check Valve with indicator (GAI Tag No. RNU-237), D82-24401-18, Rev. C l \\q l / Cleveland Electric' Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

~. Observation ALni Record ll!!!!!!!!!!"""""""!!!! J Revision No. O Observation No, ME-02-04 An**t 2 of 2 Checklist No. ME-02, Items No. 7 and 24 f 7 /jfp Criginated my A W, 7h Date j[lqfg3 - Reviewed BY Q[ fj). I _ _ 71 Date G ~ u a 3.5 Letter PY-GAI/ GEN-1888 Dated 5/18/83, ECCS Testable Check Valves. 3.6 Letter PY-GEN /GAI-2656 Dated 4/25/83, ECCS Testable Check Valves. 4.0 Potential Design Impact The lift (piston) check valve has a higher flow resistance then the swing check valve and will affect the overall system pressure drop. The method of testing of this valve during normal plant operation is not given in any of the documents reviewed and therefore the design impact cannot be assessed. However, it appears that either a spare or new drywell penetration will be i required for the hydraulic test line. ALARA aspects of the testing of this valve should be reviewed, since, per discussion with GAI, personnel performing the test will now be located inside containment but outside the drywell, rather than outside containment. This location may expose test personnel to a. higher radiation field. The use of the Rockwell Valve was approved with comment by General Electric in Reference 3.5 but no NRC approval or FSAR amendment was found. 5.0 Probable Cause Inadequately documented design changes. Attachments A. Observation Record Review l Cleveland Electric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review . ~

Obt.ervation L4M' fd Record Review mmim!!!!!"!"""'!mi Attachment A Observation No. ME-02-04 Checkilst No. ME-02 Rnision No. O PFR Nc. Sheet } of } 1 Yes No Closed X Extent 3 of 3 Systems with inconsistencipt betwppn valvo data a nd r5 rarmi ramont e Comments Per GAI, valve E22-F005 is remotely testable by a fluid system which applies pressure to a test fitting on the valve and forces the piston to lift. The test fluid system is currently in preliminary design and is not yet reflected in design documents. The higher pressure drop through the piston type lift check valve was considered in the revised HPCS calculations (see Observation ME-02-09). 2 The GAI design condition for this valve was lowered from 1575 psig to 1475 psig at 140*F by ECN 12412-E22-001, Rev. O, dated 6/17/82. The manufacturer, Rockwell International, in a letter to GAI on 12/1/83, stated that the valve rating can be increased from Class 494 to Class 590 and that they will provide the new documentation by 1/27/84. Rockwell also stated in this letter that a motor operated version of this valve had previously been given a full 900 Class rating with the only exception being the corrosion allowance. Based on the above, this Observation does not have any impact on design or safety. l l Approvals Originator g,g g Date j /j3 /g af Project Engineer Q (_i_ T A - Date ,/g /g 4 Project Manager /h Date gf4 bl Representative (('gQg j' 7gg / Date l Cleveland Electric illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

=. .:. 4.g.. - .wg =. ~ PEI Observation V Record Closure Attachment B Observation No. Checklist No. Revision No, ME-02-04 ME-02 0 EDDR No. QAD 600 No. Sheet of N/A PRE-083, 085 1 2 Yes No CI:: sed X is: lated X . Pat:ntial Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The Cygna observations questioned 3 categories of discrepancy, i.e.; Valve Type / Testability, Pressure Drop, and Pressure / Temperature Rating. The first two items were resolved through confirming documentation. The Pressure / Temperature Rating concern resulted in evaluation via QAD 600 (Possible Reportable Event) (PRE) and review for generic applicability. O* O Action Taken Action taken regarding the Prossure/ Temperature Rating concern included two QAD-600 evaluations and a review of all Class 1 valves to preclude any potentially generic discrepancies. 4.0 Conclusion QAD 600, PRE-083 was closed based upon a vendor statement which was subsequently retracted, reopening the question in PRE-085. PRE-085 was closed based upon subsequent analysis. The review of all Perry Class 1 valves by gal concluded that the discrepancy was not generic in nature. Any remaining questions regarding availability of documentation will be resolved as a result of the GE Criteria Compliance Review. 5.0 References (11) PRE-083 (13) PRE-085 ) (25) Perry Nuclear Power Plant GE Criteria Compliance Review Procedures i O L.)

rf' " W " L.._ g...g. -- _ _; 3. ;;-;' ~? -; = ^, ^ '" * '=,_-[_-=^ W I El Observation Record Closure Attachment B Ob:ervation No. Checklist No. Revision No. ME-02-04 ME-02 0 EDDR No. QAD 600 No. Sheet of N/A PRE-083, 085 2 2 Approvals Originator Date S:nior Projecpt'gineer Date Y8 6Y

v. A kA&

CEI Supervis ty A t Unit Date . gal Project Manager s:<>' n-Date .T/ 7/8y 2 gal Manager Corpor Date (~ jha Cleveland Electric illuminating Company: / 'drry Nuclear Power Plant Piping Design Review W138/25/Q/sp

Observation Record l L*I FJt A-lillilillllillllilllillillilli Revision No. O Observation No, ME-02-05 sheet of Checkliet No. MF-02 HPCS, Items No. 10. 11 and 24 i 1 Criginated By [gh Date,7 L /j /g3 g/g[g R viewed By Q{ h, *__ Date (,) O 1.0 Description HPCS system check valve drawings for F002, F016, F024, and F007 do not show any provisions for checking free movement of the valve disc. 2.0 Requirement General Electric specification 22A313 Rev. 5 Section 4.5.1.4 requires that HPCS check valves be testable to verify free movement of the v'alve disc. 3.0 Reference Documents 3.1 HPCS design specification, 2A3131, Rev. 5 3.2 Drawings O 3.2.1 vaive assemsiy 1e incn, 900 ie. swing check (Borg Warner) GAI B/M RDQ 217, 81510, Rev. E 3.2.2 DUO. check valve (TRW Mission) GAI B/M ROQ 221, 21140, Rev. A, Sht. 12 4.0 Potential Design Impact Valve discs should be checked for free movement on a periodic basis to insure that valve is not binding or stuck in the closed position. If valves bind or stici closed, they will increase the overall system pressure drop or reduce the avai'.able NPSH to the HPCS pump. i 5.0 Probable Cause Design oversight. l Attachments A. Observation Record Review I l - n v l Cleveland Electric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review l l

Observation [ M &'fd Record Review nunninn.............no Atta hment A ~ O Observation No. ME-02-05 Checklist No. ME-02 Revision No. O PFR No. Sheet 1 of 1 Yes No Closed X Extent 3 of 3 Systems with inconsistencies between valve data and GE requirements Cornme nts Per telecon between T.S. Daugherty of GAI and D. Reich and S. Bellows of GE on 12/22/83, the GE requirement that HPCS check valves be testable to verify free movement of the valve disc can be met by system functional testing.- It is not GE's intent to require external manually or mechanically actuated operators to verify free movement. Based on the above, this Observation has no impact on design or safety. O l l Approvals orienetor 7/, w, h Date / //3// 9 Date g /g g,[g4 Project Engineet Q g- __4-Project Manager UTM ~ oste 6/g5 jl79ffg CEI Representative f.'& '//f &* Date y ClevelandEledricIlluminabg;83102 Perry Nuclear Power Plant Piping Design Review . - - - - ~

g:3.r ~=: m: gel Observation Record Closure Attachment B Observation No. Checklist No. Revision No. ME-02-05 ME-02 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No Closed X is: lated Not applicable Patential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required 3.0 Action Taken No additional action required. Nuclear Energy Services is developing a pump and valve testing program for CEl. O(",0 Conclusion Cygna einterpreted the GE requirement in a more stringent manner than GE intended. GE has confirmed that the gal interpretation is correct. 5.0 References None Approvals Originator gp Date Senior Projebc wJ .s/ sis + ngineer Date CEI Superv ty t Date gal Project Manager '~ Date ~ Y), 5l9lOY ,w gal Manager Corpory4 ff o Date 'fh3 Cleveland Electric illuminatirfg Company: / rry Nuclear Power Plant Piping Design Review DW138/13/Q/sp

4 Observation M L% i Record 1 p11111llltll1111lll111111111111 Q Revision No. O Observation No. ME-02-06 sheet 1 of 1 Checknet No. ME-02 HPCS Item #17 /g / / /Q Date Celeinsted my f, G, Q pqf q f g Ceviewed my gh* Q Date g o 1.0 Description The sizing calculation for pump C-003 minimum flow bypass orifice, R0-0003, is based on a minimum flow of 10 GPM and an assumed head loss of 96 feet. The specification for the pump and its attached " Design Requirement Summary Sheet" list two different minimum flows (i.e.,10 GPM and 15 GPM) for this pump. No sizing or pressure drop calculation could be located for this pump so the 96 feet of head available for orifice sizing could not be verified. 2.0 Requirement Specification SP-506-4549-00, Rev. VII Bill of Material Sheet 19 lists a minimum required flo; of 10 GPM and the attached design requirement summary sheet lists a minimum flow rate (continuous bypass) of 15 gpm. The Perry FSAR Amendment #3 Section 6.3.2.2.5 states that a low flow bypass is provided for 4 this pump to prevent overheating. 3.0 Reference Documents 3.1 Attachment #1 dated 2/8/79, Calculation E22 A-J, CC 3.2 Specification for Fabrication and Delivery of Water Leg Pumps, SP-506-4549, Rev. VII 3.3 Amendment #3 Section 6.3.2.2.5, Perry FSAR 3.4 HPCS Design Specification, 22A3131 Rev. 5 4.0 Potential Design Impact Dependent on the actual pump minimum flow requirement and available head, orifice R0-D003 may be incorrectly sized. 5.0 Probable Cause Incomplete and conflicting documentation. Attachments A. Observation Record Review O Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

1 Observation A(ni Record Review nunninununnunnui Attachment A O g,gy,g Checklist No* ME-02 Revision No. O Observation No. PFR No. Sheet of 3 } Yes No Closed X l Extent 2 of 3 Systems with calculation inconsistencies Comments GAI located preliminary pump design calculation SP-506-1. This calculation was verified and signed by GAI on 11/17/83 and issued as E22-7, Rev. O, on 12/27/83. The calculation contains some minor inaccuracies but verifies the capability of pump C003 to meet its design function. The vendor pump curve included with calculation E22-7 shows that the pump shutoff is 100' and not 106' as assumed in the sizing calculation for orifice R0-D003. This reduction in shutoff head will result in a reduced bypass / recirculation flow through orifice R0-0003 and could affect the heat dissipation capacity of the minimum flow bypass loop. GAI will ensure a minimum 10 gpm bypass flow during system performance testing and install a larger size i orifice, if required at that time. Based on the fact that pump C003 is adequate for its intended purpose and that the pump heat dissipation and orifice size adequacy will be verified by GAI in system tests, this Observation is closed. l ) I Approvals Originator %, W, M f//y/fg Date Protect Engineer dub ? M, Dete / /16,[8 4 O Ws/Yr%'L 'A /* a '- co Re.,e.e ieme N M04M-o' #24/u ClevelandElecdicIllumina$ing;83102 l Perry Nuclear Power Plant Piping Design Review

F """q_2&fr?;m..,Q ,~r- - C.;.. J .'g_ _4 ^ ~ El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. ME-02-06 ME-02 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Closed X isclated X Pctential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion As stated in the Cygna observation record review the pump is adequate and the orifice size will be verified by testing. 3.0 Action Taken The GE Criteria Compliance Review will be performed. .0 Conclusion The review listed above will provide additional assurance that all required calculations are available and verified. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compliance Review Procedure, Rev. O. 1 Approvals Originator Date S:nior ProJec Engineer / Date . M-N " + $ 6 fl3V CEI Supervisor Quali Au it Unit Date W Y/e/84 gal Project Manager $ ^ t '~ Date n ' r,n f/7/89 gal Manager Corpora P 6gr Date f {The Cleveland Electric liluminating' Company:[ 'ltrry Nuclear Power Plant Piping Design Review DW138/14/Q/sp

Observation 4h i Record Illlibilhilii (O Revision No. O Observation No. ME-02-07 sheet i of 1 Checklist No. ME-02-HPCS Item #24 Date /g /j/p Criginated my g, -1), 7M IRf9 N n: viewed my 4 (d._- Data V Q 1.0 Description I It is not apparent from the PalD or piping drawings how valves F001, F010, and F011 will be leak tasted. There do not appear to be any drain valves located such that meaningful test results can be obtained. 2.0 Requirement General Electric Specification 22A3131, Rev. 5 section 4.5.1.7 states that drains shall be provided which will permit leak testing valves F001, F004, F005, F010, and F011. 10CFR50 Appendix A Criterion 37 also requires that the HPCS system be designed to permit periodic pressure testing to assure the structural and leaktight integrity of its components. 3.0 Reference Documents 3.1 HPCS Design Specification, 22A3131, Rev. 5 3.2-General Design Criteria,10CFR50, Appendix A 3.3 Drawings l 3.3.1 HPCS P&ID, D-302-701, Rev. G 3.3.2 HPCS Piping, D-304-701, Rev. M l 3.3.3 HPCS Piping, D-304-702, Rev. L 3.3.4 HPCS Piping, D-304-703, Rev. G 4.0 Potential Design Impact Drains iney have to be added to the system piping in order to meet the leak test requirements for those valves. 5.0 Probable Cause Design oversight. Attachments l l A. Observation Record Review 1 O U 1 Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review ___ _ - _ _ _. ~ _.

i Observation dlni Record Review tiliiiiiiiii Attaehment A O Obrerwation No. ME-02-07 Checklist No. ME-02 Revision No. O Sheet I of 1 i PFR No. i Yes No X Closed Extent 2 of 3 Systems with inconsistencies between GE and GAI data Comments GAI has stated that the test method for the subject valves is currently being reviewed by the CEI/NTS (Nuclear Test Section) group. Additional drain valves may i be added as a result of this review. This review and any required design document changes will be completed in 1984 Based on the fact that this item is currently under review by GAI and CEl, this Observation is closed. , O i l Approvale Date jf)hllLj Orleineter }, Q, l fy.ogg, t Dele f/fg/gf Project engineer @ [h_ I _,Id- '/I'4-

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-c _g 1 .1El Observation Record Closure l Attachment B Observation No. Checklist No. Revision No. ME-02-07 ME-02 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No Closed X isolated Not applicable Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required. 3.0 Action Taken No additional action required - two testing programs are now being developed which will resolve all testing requirements. k.J.0 Conclusion The project is now reviewing all testing requirements. CEl/NTS (Nuclear Test Section) group is insuring all vents and drains needed for testing have been provided in addition NES (Nuclear Energy Services) is developing a pump and valve inservice testing program for CEl. 5.0 References None Required. Approvals Originator g Date Senior Project g r Date CEl Supervi ty A t U nit Date gal Project Manager Date 5/?lOY am gal Manager Corpor ro Date he Cleveland Electric illuminating Company: / rry Nuclear Power Plant Piping Design Review DW138/15/Q/sp

m Observation 4L n i Record lllllll11lllllllllllllllllllll Revision No. O Cbservation No, ME-02-08 sheet of 3 Checklist No. ME-02-HPCS i Originated By f, y fg Date j t / / /f3 i Date (2.f aj / g3 Ceviewed By QQ. ', y - g g 1.0 Description The following items either lack proper documentation or utilize inconsistent data. a. HPCS Fill Pump C003 sizing calculations could not be located by GAI. In addition, the specification for this pump (SP-506-4549, Rev. III) contains inconsistencies on pump minimum flow and discharge nozzle, size. The discharge nozzle size is also inconsistent between the vendor supplied pump curve and pump drawing. 4 b. The suppression pool suction strainer pressure drop utilized in all calculations is 1 PSI. Per the strainer specification this is the maximum drop at 8500 G.P.M. and would be lower at lower flow rates. Per the vendor pressure drop calculations, the actual drop thru the strainer at 8500 G.P.M. is 0.42 PSI in the clean condition and 0.60 PSI with the O straner 50% plugged. These pressure drops would then have to be adjusted 4 for the lower system flowrates of 7000 G.P.M., 6110 G.P.M. and 1550 G.P.M. c. Per the Perry FSAR section 6.3.2.2.1 reflief valve F014 has a capactiy of < 10 G.P.M. 10% accumulation with a set pressure of 100 PSIG. The valve data gives the capcity as 16.2 G.P.M. d. Per the Perry FSAR section 6.3.2.2.1, valve F039 is a thermal relief valve set at 15 P.S.I.D. The valve shown on the P&ID, physicals, and Bill of Material for Perry is a lift check valve with no specified opening pressure. e. The calculated size of orifice R0-D002 is 6.54" but the Perry Informatinn System (P7837151.5) lists the size as 6.51" The size of this orifice will be affected by inconsistencies in the flow pressure drop calculations with flow to the reactor vessel. f. The calculated size of orifice R0-D004 is 4.27" but the Perry Inforamtion System (P7837151.5) lists the size as 4.32". In addition, the calculation assumed valves F010 and F011 were fully open whereas specification DSP-E22-1-4549 Table 1 indicates the valves are in a throttled position. This would affect the size of R0-D004. AV Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation M L% i Record lpmtittitimittilllllllll Rniston No. O Observation No. ME-02-08 sheet of Checklist No. ME-02-HPCS p 3 Date f7/j/73 Crisinated By f,4,- h {M q / g3 C; viewed By Q( Q. _' ,,Q Date V O g. The calculated and specified size of orifice R0-D005 is 5.10".

However, this size may be affected by inconsistencies in the system pressure drop calculations i.e., strainer loss, valve losses, pump operating point, etc.

h. In calculation E22-1 on HPCS Pump C001 NPSH, an incorrect but conservative value is used for the loss thru the suction strainer and the pump runout flow. Also tha specific gravity of water at 212"F is approximately 0.96 not 1.0. 1. In calculation E22A/J-cc on page 13 it is indicated tha the RCIC is operating concurrently with the HPCS. No documentation was found of this operating condition, but the assumption leads to conservative suction losses. i j. Relief Valve F035 is a 900 lb. class valve. However Line Specification DI-2 calls for 150] lb. class valves in this size. O 2.0 Requirement Good engineering practice requires that design data be well documented and consistent through the design process. 3.0 Reference Documents 3.1 Water Leg Pumps, SP-506-4549-00, Rev. VII 3.2 Suction Line Stainers, SP-529-4549-00, Rev. III 3.3 Mac-Iron Pressure Drop Calculations dated 8/3/76, C.E.I. Job s.0. 52811-3 l 3.4 Amendment #3 dated 9/11/81, Perry FSAR 3.5 NPSH Calculations, Calculation E22-1 dated 12/10/81 l i 3.6 Line Losses, Calculation E22 A/J-cc dated 2/16/79 i 3,7 HPCS Restricting Orifices Attachment #1 to Calculation E22 A.J-cc dated 2/8/79 3.8 Byron Jackson Pump Curve dated 3/22/74 (GAI #4549-20-009-1), PC-741-S-1414 O Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review ~. -+

Observation Record 31 i i lilllllllllllillllllllilllllli Ruiston No. g Observation No, ME-02-08 sheet of Checkflat No. ME-02-HPCS 3 3 Date j g fj /g Criginated By [,y, fg/dj [g Date Reviewed By [ fd, _*_ ~ m _ l 3.9 Bingham Pump Curve (INQ #P-249-K) Water Leg Pumps, CA-3201-1 3.10 Perry Information System, P7837151.S dated 9/15/83 3.11 Design Specification HPCS, DSP-E22-4549-00, Rev. I and Rev. 2 3.12 Bingham-Willamette Pump Drawing (GAI #4549-21-034-3), E-17409X, dated 9/28/77 3.13 Check Valves Specifications, SP-531-01-4549-00 3.14 Relief Valves Specifications, SP-523-4549 4.0 Potential Design Impact The noted inconsistencies and lack of documentation could lead to design errors O and possibly incorrectly sized components.

c 5.0 Probable Cause Design control.

Attachments A. Observation Record Review I LO Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review.

Observation 4 f. n i Record Review minmimmmimmi Attachment A Cheervation No, ME-02-08 Checklist No. ME-02 Revision No. O PPR No, sheet 1 of 2 Yes No Closed X Extent 3 of 3 Systems with missing calculations and inconsistent data application Comments GAI has presented the following resolutions to the noted inconsistencies: "a. The HPCS fill pump calculation E22-7 was located and verified (see Observa-tionME-02-06). GAI has agreed to revise specification SP-506 to reflect the correct (2") nozzle size. b. GAI, in the independent HPCS calculations, has used a 1 psi drop at 6110 gpm for the suction strainer and adjusted this pressure drop at other flowrates. This is conservative and acceptable (see ME-02-09). O c. Per GAI memo from J.S. Smith to J. Hickson dated 1/13/84, FSAR pages 6.3-13 and 6.3-14 will'be changed to indicate that the capacity of relief valve F014 is less than 20 gpm. d. Per GAI memo from J.S. Smith to J. Hickson dated 1/13/84. FSAR pages 6.3-13 and 6.3-14 will be changed to indicate that valve F039 is a lift check vavle used for relieving thernally expanded fluid, ~ e. The HPCS independent calculations by GAI verify the adequacy of the 6.51" size of orifice R0-D002. f. The HPCS independent calculations by GAI verify the adequacy of the 4.32" size of orifice R0-D004. g. The HPCS independent calculations by GAI verify the adequacy of the 5.10" size l of orifice R0-D005. I i Approvals Orleinator [, W, g Date //1,o/fy f/gp/g Project Er.i,:r.::^._[M. _g Date I Project Manager j gf] Date g/gp/4 CEI Representative gj Date pj/g Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

h Obsorvation Record Review At t i immmmimmmmim n Attachment A L) i Observation No. ME-02-03 Checklist No. ME-02 Revision No-O PFR No. Sheet 2 of 2 Yes No Closed X Extent 3 of 3 Systems with missino calculations and inconsistent data application Comments h. Based on the fact that the pressure drop thru the strainer used in the calcula-tion is conservative and that the fluid specific gravity has no effect on the end result of the calculation, the calculated NPSH available is acceptable for system operation. 1. The HPCS independent calculations by GAI do not indicate that RCIC is operating 2 concurrently with HPCS. This matches other documentation and is acceptable from a system design standpoint. j. The 900 lb rating of relief valve F035 meets all system operating pressure and \\ temperature requirements. The line specifications only list recommended ratings for gate, globe and check valves, and do not apply to relief valves. Based on the above, this Observation has no impact on design or safety. Approv.le n D * //z o/# V orie net.e A 4s, h, O i.otunein,e >kLY cate i/ae/A4-Pre ma wox4-i.o. sa Tresem a '- si/ F F f cuiR.pr ne.tev. 4Lar 6 Data Cleveland Electric Illuminating; 83102 [ Perry Nuclear Power Plant Piping Design Review i ,.,,.__,n_.-__.

= -~z gm -- ~ TEI observation d Record Closure Attachment B Observation No. Checklist No. Revision No. ME-02-08 ME-02 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Closed X isolated X Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion Three of the ten inconsistencies were conservative assumptions (a, h, i) and one (J) was a misinterpretation of gal requirements. 3.0 Action Taken items a, c, d, e, f, L g represent questions of documentation which will be p addressed generically as a result of the GE Criteria Compliance Review. v 4.0 Conclusion The review listed above will insure that all calculations are consistent and documented. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compilanco Review. Approvals Originator g Date p g Senior Pro En in Date 9 CEI Supervi y it Unit Date gal Project Manager fA Data $/T/OY ^ gal Manager Corporap'Pfo --, Date ThJ Cleveland Electric lliuminatin~g Company: / grry Nuclear Power Plant Piping Design Review DW138/16/Q/sp

Observation Record dL 6i Oimummmmimmnlll l Revision No. O Observation No. ME-02-09 sheet i or 4 CheckHet No, ME-02-HPCS Item #20. 21. 22. & 23 n fj /Q Crielneted my f,Q h Date JQfn eviewed sy @). _ i oste 1.0 Description The following items sumarize the inconsistencies and inaccuracies noted in GAI ) Calculation E22-A/J-cc, HPCS Line Losses. The L/D used for valve F005 in all calculations is 135 (for a swing check a. valve). The valve is actually a lift check for which an L/D of 340 should have been used. Note: The vendor drawing for this valve indicates that C = 1993. y b. The static head used in Modes A & E is based on a condensate tank low water level of 633'-0". However, the worst case flow condition (max. AH ) would be just prior to switching to suppression pool suction. Thispoink is assumed to be at a tank level at the tank suction nozzle top and would add - 10ft. to the AH. In addition, Drawing D-302-102, Rev. G indicates 3 that the 150,000 gallon reserve in the condensate tank for HPCS is at level 630'-9". In Mode E a suction flow rate of 7800 G.P.M. is used fo calculating c. sucticn head loss, but the pump discharge head losses are based on a system discharge flow of 6110 G.P.M. This is inconsistent, but conservative. d. HPCS pump suction strainer D006 is not included as a head loss in the calculations. If this stainer is just used for startup and then has the element removed for normal operation, this should be stated in the calcu-lation. The physical drawing shows a large assembly for this strainer which may contribute some head loss even if the element is removed, On page 27 of the calcuation, a head loss of 0.4 ft. for valve F001 is e. added to the total head even though this valve was already included in total system equivalent length and head loss. This is a G.E. suppied valve and the 0.4 f t. drop is specified by G.E. This head loss should be used in lieu of, but not added to the previously calculated loss. I f. The head loss for valve F004 has been added to the total system head loss twice. Once as an equivalent length and once as 1.4 ft., the G.E. specified maximum. In addition, the loss of 1.4 ft. has been added to the 16 inch pipe segment on page 28 rather than the 12 inch segment in which the value is located. c Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review l-

Observation AMi Record (O l111111ll1ll111111111111111 V Revision No. 0 Observation No. ME-02-09 sheet 4 Checklist No. ME-02-HPCS Item #20, 21, 22. & 23 2 Date fg f /f 3 Criginated By , -{), 7h Date fMq/gg neviewed my VA Q.- ._J_a - g. In Mode B the suppression pool suction strainer head loss is given as 2.31. ft. on page 18. This is the maximum allowable drop with the strainer 50% plugged at 8500 G.P.M.. For 1550 G.P.M. and 50% blockage, this loss should not exceed 2.31(85 ) = 0.08 ft. h. Page 20 of the calculation lists the suppression pool low water elevation as 592'10", but the pump NPSH calcuation E22-1 lists the minimum level as 589'0". 1 In Mode C on page 22 of the calculation, the head loss of valve F015 has been added to the system loss twice. The stated loss for this G.E. supplied valve is 0.07 ft. at 6110 G.P.M. J. Page 23 of the calculation again adds the G.E. stated loss for valve F004 to the toal system loss which already includes valve F004. k. Page 32 of the calculation again adds the G.E. supplied drop for valve F015 to the total system loss which already includes valve F015. 1. The G.E. stated valve head loss was not used in the calculation of head losses for Mode F on page 23. The pump operating points used in the calculations for the various modes m. of operation do not appear to match the Byron Jackson Pump Curve Dwg. PC-741-S-1414. 2.0 Requirement Per the General Electric Process Diagram 762E455 and Specification Data Sheet 22A3131AS, the HPCS Piping System shcIl be designed to provide 1550 G.P.M. to the reactor vessel with the R.V. pressure 1147 PSI above source suction pressure and 6110 G.P.M. to the reactor vessel with the R.V. pressure 200 PSI above source suction pressure. The system should also limit the flow to the reactor vessel at 14.7 PSIA to 7800 G.P.M. or the tested runout flow of the pump, whichever is lower. kqJ l Cleveland Elet.tric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review u.

u Observation M%i Record llllllllllllllllllllllllllll11 Revision No. O Observellon No ME-02-09 s heet 3 of 4 Checklist No. ME-02-HPCS Item #20, 21, 22, a 23 f 4/j /g] Crielneted my [, M h Data l g( aj /g3 Deie noviewed sy g4 ' g - \\J N) 3.0 Reference Documents 3.1 HPCS Design Specification, 22A313, Rev. 5 3.2 HPCS Design Specification Data Sheet, 22A313, Rev. 5, 3.3 Process diagram HPCS, 762E455, Rev. 6, 3.4 Buryon Jackson Pump Curve (GAI #4549-20-009-1-0), PC-741-3-1414, dated i, 3/22/74 3.5 HPCS System NPSH, Calculation E22-1 (5/12/81) 3.6 HPCS - Line Losses, Calculation E22-A/J-cc (2/16/79) 3.7 HPCS Restricting Orifices, Calculation E22-A/J-cc Attachment #1 (2/8/79)

3.8 Drawings

3.8.1 HPCS Piping, D-304-701, Rev. M 3.8.2 HPCS Piping, D-304-702, Rev. L 3.8.3 HPCS Piping, D-304-703, Rev. G 3.8.4 Northeast Main Plant Area E-303-002, Rev. U 3.8.5 Sections & Details, E-303-016. Rev. H 3.8.6 Auxiliary Plans - Sections & Details. E-303-017 Rev. N 3.8.7 Plans and Details. E-303-002, Rev. F 3.8.8 Condensate Transfer and Storage, D-304-317. Rev. V 3.8.9 Condensate Transfer and Storage, D-304-315. Rev. E 3.8.10 Condensate Transfer and Storage, D-304-315 Rev. F 3.8.11 Condensate Transfer and Storage, D-302-102, Rev. G 3.8.12 HPCS, D-302-701, Rev. G 3.9 Testable Piston Check Valve w/ Indicator (HPCS System Valve F005), Rockwell International, Dwg. No. 082-24401-18, Rev. C 4.0 Potential Design Impact l The major design impact of the calculational inaccuracies will be their affect on the sizing of the system orifices. The result of improperly sized orifices may be off-nominal flow to the reactor vessel and/or inaccurate flow testing of the system. Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record Li l J L A 1111111llllll1lll!I1111lll1lll Revision No. O Observation No. ME-02-09 sheet 4 of 4 CheckHet No. ME-02-HPCS Item #20, 21, 22, & 23 ost. /2 / //f 3 Crisinated By g, y, y Date ( Af q f p Cowlewed By gl] _] g g 5.0 Probable cause Documentation inconsistencies and minor design oversights. Attachments A. Observation Record Review i ' O i i l O Cleveland Electric Illuminatingi 83102 Perry Nuclear Power Plant Piping Design Review

Obsorvation Record Review 4( t i mmmnimummn Attachment A O Cbservellon No. ME-02-09 Checklist No, ME-02 of g PFR No. 8 heet g Yes No Closed X i Extent 2 of 3 Systems with calculation inconsistencies and inaccuracios Commente GAI reanalyzed the HPCS system flow and head loss in calculations N22-3, N22-4, N22-5, N22-6 and N22-8 These new calcualtions utilized Tube Turns Piping 4 Engineering Chart 3 data for equivalent lengths of fittings and valves rather than the Crane Technical Paper 410 data which was used in the original calculations. This resulted in lower head losses for fittings and valves in the new calculation. Certain approximations are used in the revised calculations, but they have a negligible affect on the total system head loss. The new calculations indicate that eith the specified orifices installed, the system head exceeds requirements for all modes of operation. The adequacy of these calculations and orifice sizes will be ] confirmed by system performance and pre-operational testing. Based on the above, the system head losses are acceptable for design and this Observation has no impact on safety. j i Approvals Ortonneter f,W, %, D*'* ///f/$'/ ca'. /he/s4-M\\(A. a A O.Pr.>.t snen < s, A4-m rcAzia a '- .. t,., cei n.e, ni.*. M 43%, & o a'. //pWW ClevelandElec61cIlluminating;83102 Perry Nuclear Power Plant Piping Design Review

... g :m..q "" ~ - - - - " " = - - ^ - 'El Observation d Record Closure Attachment B Observation No. Checklist No. Revision No. ME-02-09 ME-02 0 EDDR No. QAD 600 No. Sheet of 72 1 1 Yes No Closed X isolated X Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The inconsistencies noted in the subject were considered to be minor since they did not affect the adequacy of the system. The GE Criteria Compliance Review will document the acceptability of such inconsistencies, which are primarily conservative assumptions to simplify the calculations. ,q0 Action Taken b' gal will complete the GE Criteria Compliance Review. 4.0 Conclusion The GE Criteria Compliance Review will insure verified calculations are available to document that the systems will perform their intended function. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compliance Review. Approvals Originator g Date g,g Senior Pr$ojeA AfA r/s ls + Engineef Date CEl Supervisor Quallt Audi Unit _Date W Y/c/94 -e _. gal Project Managerh ~ _. e

• ' w 5/ 9/ 84/

~ Date gal Manager CorporpA p(o Date The Cleveland Electric illuminating Company: / grry Nuclear Power Plant Piping Design Review DW138/17/Q/sp

Obscrvation Record 31 6 i Oinnauman a e... u.. o c.....u. u.,E.03 01 sheet i et p Checklist No. ME-03 MSDS Item #1 Date /t // /D orie netea ey A, -tj, g>& n o~ 11.I n I n u,i...a ev M

• Jr 1.0 Description l

The following inconsistencies within Table 1 of DSP-B21-1-1-4549 and between Table 1 and the General Electric system data are noted below: a. The indicated pressure drop in Table 1 from location 4 to 13 for a constant flow of 310 lb/hr varies from 47.7 PSI for mode A to 390 PSI for mode B and 100 PSI for mode E. b. In mode D of Table 1, the flow between locations 4 and 13 is given as 6,670 lb/hr and the pressure drop is listed as 100 PSI. This is the same pressure drop as given for mode E with a flow of only 310 lb/hr between these two locations. c. GAI Table 1 indicates a continuous drain flow of 310 lb/hr for modes A, 8 and E, i.e., drain valve F033 open. The General Electric Process Data 13I . O C7911C and Specification 22A4622 indicate that the drain valve F033 only 1 opens at power levels of 50% and below and that the flow rate through the ' orifice is 2,000 lb/hr. I d. Both Table 1 of the GAI Specification and the GE process data indicate that the drain flowrate between location 13 and 14 in mode C is 50 GPM at 125"F. This drain path consists of two 3/4" valves and approximately 125' of 3/4" pipe which will significantly restrict the actual drain rate. In addition. no pressure drop is indicated across the two drain valves with the 50 GPM flow through them, i.e., 100 PSIA indicated upstream and downstream of the valves. j 2.0 Requirement and Process Diagram 10505575 GE. Specification 22A4622, Process Data 131C7911C, hey provide flow, pressure, are the design basis documents for the system. T and temperature data for which the system should be designed. 1 3.0 Reference Documents 3.1 Nuclear Boiler Design Specification, 22A4622 Rev. 5 l 3.2 Nuclear Boller Design Specification Data Sheet, 22A4522AR, Rev. 2 l Lo Cleveland Electric 111uminatingt 83102 Perry Nuclear Power Plant Piping Design Review t

L Obscrvation Record AL 6 i O n..i... u.. o j c......... .. ME.03 01 ] s h.et 2 of 2 p ch hu.t 88.. ME-03 MSDS Item #1 ~ f o,se ..e av m -1.>, n.

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(- y g 3.3 Process Diagram Nuclear Boiler, 10605575. Rev. 0 i 3.4 Process' Data Nuclear boiler, 131C7911C, Rev. 5 i f' 3.5 Design Specification Nuclear Boiler System Piping and Pipe Supports, DSP-821P1P4549-00, Rev. 2 3 {# 3.ti Main, Rebeat, Extraction, and Miscellaneous Drains PAID, D-502-131 Rev. 0 t 4.0 Potentfal Design tapact I Since the gal design specification is used for piping and pipe support design, l incont.istencies in pressure, temperature, and flow data could cause inaccuracies in this dest y effort. It is not clear what other design functions (valve sizing, J AC, etc.) use Table 1 data as design input f information. i O 5.0 Probable Cause Failure to document the resoluttes of differences between the GAI design l l-specification and corresponding GE design data. Attachments A. Observation Record Review. l t il 1 l t - i l4 ) s i p , p' t Q y T Clevelend tjectrl( 111uminatings 83102 Perry thseleer Pen.cr Plant Piping Design Review ~ ^ ,a E u, .- h _ _, _.. _.. - ~.

i Observation [41*H M Record Review muimiumminimii Attachment A V Cbservation No. ME-03-01 Checklist No. ME-03 Revision No-0 PFR No. Sheet 1 of ~1 Yes No Closed X Extent 2 of 3 Systems with inconsistencies between GE and GAI data Comments GAI has stated that Table 1 of Design Specification DSP-B21-1-1-4549 will be updated to correct the inconsistencies noted in this Observation. Regarding items (c) and (d), GAI has obtained verbal concurrence from GE (reference 10/19/83 telecon between T. Daugherty and J. Hickson of GAI and E. Wood and D. Foster of GE) and has requested written agreement (reference letter PY-GAI/ GEN 2964 dated 1/3/84) on the following modes of system operation:

1) Continuous draining through the first MSIV before seat drain at all power levels. The resulting nominal drain rate will be approximately 310 lb/hr in lieu of GE-specified 2000 lb/hr at power levels below 50%.

2) A maintainance drain flowrate of less than 50 gpm to the clean radwaste system.

3) A maintainance drain rate of 50 gpm or greater to the main condenser, if condenser water quality require;nents are met.

Based on the above, this Observation has no impact on design or safety. l l l l l Approvals j //g/f sf originator M, W, h Date Me /s+ er.aci eneeer dx&Rm o i-0 r> e- ~ vs%e3m a '- 4 /* l hgg CEI Representative h hg, Date ClevelandEledicIlluminat$g;83102 Perry Nuclear. Power Plant Piping Design Review i l

y g- --_ ~I '*El Observation O Reco'rd Closure Attachment B Observation No. Checklist No. Revision No. ME-03-01 ME-02 0 EDDR No. QAD 600 No. Sheet of 72 and 73 N/A 1 1 Yes No Clased X ls lated X P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion No impact on system reviewed but generic effort to be addressed. 3.0 Action Taken Perform the GE Criteria Compliance Review and a review for any negative effects from operating data inconsistency. (~h v'.0 Conclusion The above review will insure inconsistencies which could affect system design will be resolved. 5.0 References (9) EDDR 72 (10) EDDR 73 (25) Perry Nuclear Power Plant GE Criteria Compliance Review Procedure. Approvals Originator Date S:nior Pr t Egin Date CEI Supervisor Q)uality Audj$ Unit V//a lTij Date W L gal Project Manager [<>R'h \\ I/ Date V6 'w 5/WW \\ gal Manager CorporpA"Pfogr Date Tha Cleveland Electric illuminating Company: / (~'7.rry Nuclear Power Plant Piping Design Review V DW138/22/Q/sp o i

Observation [4Li'TJ Record lilllillllllll!!!!!!llllllllll R*'38' "N-0 Observation No. ME-03-02 sheet of Checklist No. ME-01 MSDS Item #3 3 3 Date /g / j /p Criginated By f, q fhy Date ( 2, !g/$3 Reviewed By ((d,

• _ [ k -

1.0 Description No sizing calculation could be located for restricting orifice R0-D001. Therefore, no documented basis exists for the specified orifice size. 2.0 Requirement The General Electric Process Data 131C70911 gives the orifice R0-D001 flow conditions as 2000 lb/hr at greater than a 600 psi pressure drop. The G.E. Design Specification 22A4622, Rev. 5 states that a restricting orifice be provided for continuous draining of condensate during operation below 50 percent power level. 3.0 Reference Documents 3.1 Nuclear Boiler Design Specification, 22A4622, Rev. 5, (] 3.2 Process Data Sheet, 131C7911C, Rev. 5 3.3 Nuclear Boiler Process Diagram, 105D5575, Rev. 0 3.4 Nuclear Boiler Design Specification, DSP-B21-1-4549-00, Rev. 2, 3.5 Perry Information System, P1837151.S Dated 9/15/83, 4.0 Potential Design Impact Since no sizing calculation or documentation could be located for orifice R0-D001, its adequacy to perfonn the G.E. specified function could not be verified. 5.0 Probable Cause Design Control Attachments A. Observation Record Review \\O Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation (4D'n' fd Record Review ,nunmunnmnummu Attachment A Observation No. ME-03-02 checklist No. ME-03 Revision No. O Sheet 1 of 1 eFR No. Yes No Closed X Extent 3 of 3 Systems with missing calculations Comments GAI has generated a new calculation to verify the sizing of orifice R0-D001. Cygna's review of this calculation, N22-9 dated 11/15/83, verifies that the existing orifice size is adequate for all systen flow conditions. Based on the above, this Observation has no impact on design or safety. O . - -. ~ t-1 Approvals f////gy Date originator g, W,-SLu oa'* v"/s4-crowi ensa e 440. _ :_ N l 0 er.wtu e ffM ~ o =c* 1As/xs //za/h cuiR.pr. . gefrJl o=ia Cleveland Elec,tric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

y x3 .. annan ~~ v -- - = hEl Observation Record Closure Attachment B Observation No. Checklist No. - Revision No. ME-03-02 ME-03 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Cl=ed X istlated X Pr.tantial Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The case cited by Cygna was a documentation issue. 3.0 Action Taken 3.1 gal generated a calculation which verified that the orifice was adequate. 3.2 gal will perform the GE comp!!ance review, which includes a review that O all requirod calculations are available. D 4.0 Conclusions The review listed above will insure adequacy of documentation. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant, GE Criteria Compliance Review Procedure. Approvals Originator Date F, ,gj S:nior PfoJpet Engineer' Date j '2 - N-N- e-e-S/8 8Y CEl Superdvis r Quality Audit Unit Sh /sw Date A. W-a gal Project Manager n Date f) U gal Manager Corpor A P/og 5/7/M ~ Date h3 Cleveland Electric illuminating Company: / rry Nuclear Power Plant Piping Design Review DW138/18/Q/sp

Observation [9M'M Record !!illl@'!!!!!!!!!I!!!!!Ill Rnision No. g Observation No. ME-03-03 sheet e' Checklist No. ME-03 MSDS Item #8 and #9 i 2 Date g fj /p Criginated By g,W h Date lQ /p Reviewed By ([m* ,,y y ~ a 1.0 Description Calculation N22-3 page 13 is for sizing the 1st MSIV before seat drain line. This calculation does not match the physical piping arrangement and does not include all modes of operation. Specifically. The calculation is for a single 3" pipe from the 1st MSIV to the a. condesner. The actual piping arrangement consists of four 2" pipes (one from each MSIV) connected to a 3" drain header with a parallel orifice bypass line. The 3" pipe then ties into a 24" header which connects to the condenser. b. The calculation is based on a flow of 6670 lb/hr. However, the system t l design specification lists flows of 310 lb/hr and 50 gpm in addition to I the 6670 lb/hr. Also, G.E. lists a flow of 2000 lb/hr for low power operation, The calculation does not cover or show flow through valve F033 and R0-C001 c. or draining through valves F034 and F035. d. The calculation indicates no elevation difference between valve F016 and F021, whereas the physical piping drawing indicates a difference in elevation of approximately 15' feet. 2.0 Requirement 'GE Specification 22A4622 and process data 131C7911C provide the design requirements for the first MSIV before seat drain line. Section 4.6 of the specification states that the system should provide for draining the flooded. i main steam lines in a reasonable length of time and remove steam condensate l generated during heat-up and operation below percent power level. The process data lists a drain flowrate of 50 gpm and an operation below 50 percent power flowrate of 2000 lb/hr. 3.0 Reference Documents 3.1 Nuclear Boiler Design Specification, 22A4522, Rev 5 3.2 Nuclear Boiler Process Specification,105D5575, Rev 0 o Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation [. 4 M ' f J Record lillilitillllinilliiiimilii 3 Revision No. g Observation No. HE-03-03 sheet or Checklist No, ME-03 MS05 Item #8 and #9 p p f7///fJ Date Criginated By f, W, 7Q fy 9 [gg Reviewed By d ~_ ,p Date 9 v 3.3 Nuclear Boiler Process Data,131C7911C, Rev. 5 3.4 Design Specification Nuclear Boiler, DSP-B21-1-4539-00, Rev. 2 3.5 N22-Line Sizing, Calculation N22-3 (11/7/78) 3.6 Drawings 3.6.1 Piping N22, D-304-501, Rev. E 3.6.2 Piping N22, D-304-122, Rev. G 3.6.3 Piping N22, D-304-304, Rev. E 3.6.4 Piping N22, D-304-304, Rev. D 4.0 Potential Design Impact p The adequacy of the piping system to meat the design requirements cannot be determined based on the calculaticas presented. 5.0 Probable Cause Documentation control. Attachments A. Observation Record Review l l } s> Cleveland Electric Illuminating; 83102 ' Perry Nuclear Power Plant Piping Design Review

Observation [41t t f/ Record Review mmitmmimimmmil Attachment A Observation No. ME-03-03 Checklist No. ME-03 Revision No. O PFR No. Sheet } of } Yes No Closed X Extent 2 of 3 Systems with calculation inconsistencies and inaccuracies Comments GAI submitted revised calculation N22-3A, dated 1/6/84, to verify the adequacy of the size of the main steam drain piping from the first main steam isolation valve before seat drain to codenser connection 194 This calculation does not address flow through the 1" bypass line valve F033 and orifice R0-D001 which is the continuous drain path during no,rmal reactor operation. However, calculation N22-9 for verification of the adequacy of orifice R0-0001 indicates that sufficient margin exists in this flowpath to account for the 1" pipe and valve F033 losses. Based on the above, this observation has no impact on design or safety. O i l l Approvals Date ///g /f tj Originator f, Q_ g ffgfy Project Engineer [N.'-j Date Project Manager p {s{ V Date /g [7[g CEI Representative pggMy Date ClevelandElect[icIlluminding;83102 Perry Nuclear Power Plant Piping Design Review ,-%---+,.-%., ,w-m.3- .-,.---y. ,,,-,.m.--w, r_-,,,

,s .= (^';El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. ME-03-03 ME-03 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No CI: sed X is: lated X P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion Some inconsistencies noted resulted from documents being reviewed by Cygna beyond their intended purpose. None of these had any impact on the design of the system. 3.0 Action Taken (~) gal lias eliminated the inconsistencies in the cases cited by Cygna, la addition, U the GE Critoria Compliance Review will be performed on a generic basis. 4.0 Conclusion The GE criteria review will assure the adequacy of systems designed by gal. 5.0 References (9) EDDR 72 (25) Parry Nuclear Power Plant GE Criteria Compliance Review Approvals Originator g Date S;nior Proje.t Engineer ~ Date . A. A{^ ^ YB fR $ CEI Supervigua ity Au it Unit Date YMlN J-i - 4 4-gal Project Manager Date gal Manager Corporat Pf Date Tha Cleveland Electric illuminatin's Company: / gerry Nuclear Power Plant Piping Design Review \\ DW138/19/Q/sp '

Observation [.41 i'fA Record !!!!I11!!ll1!!!!!!11111111111 Revision No. O Observation No. ME-03-04 of Sheet } } Checklist No. M3-03 MSDS Item #10 f g/j /p Date Criginated By g,y 7g C viewed By [ ( ],,* _,,, [ Date 33 U G 1.0 Description Valves F034 and f035 are 3/4" Y pattern globe valves arranged in series withthe approximately 125 feet of 3/4" pipe attached to the outlet of valve F035. flowrate specified for this drain is 50 GPM of 125"F water with a pressure upstream of valve F034 of 100 PSIA. 2.0 Requirement Section 4.6.1 of G.E. Specification 22A4522 states that the main steam line drains shall drain the flooded steam lines in a reasonable length of time. The G.E. process data sheet 131C7911C states that the flowrate for this flowpath should be 50 G.P.M. 3.0 Reference Documents O 3.1 uuciear soiier cesisn specification, 22A4622, Rev. 5 3.2 Process Data Nuclear Boiler, 131C7911C, Rev. 5 3.3 Main. Reheat Extraction, and Miscel hnecus Drains, D-302-121, Rev. D 3.4 Piping N22, D-304-121, Rev. E 3.5 Piping N22, D-304-129, Rev. D 3.6 3/4" Series 1500 Y-Type Globe Valve, Kerotest Dwg. D-9955 4.0 Potential Design Impact The 3/4" drain size will restrict the drain flowrate to less than 50 GPM and increase the time required to drain the flooded main steam lines. 5.0 Probable Cause Design oversight. Attachments A. Observation Record Review A V Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review ~ r-n ++ - r -,w y -3y--

Observation = 4M'M Record Review imissimimimiimiliiii Attachment A O. Cbservation No. ME-03-04 Checklist No, ME-03 Revision No. O PFR No. Sheet } of } l Yes No Closed )( Extent 3 of 3 S_ystems with inconsistencies between valve data and GE requirements Comments GAI has discussed the drain flowrate requirement with GE (reference telecon dated 10/19/83 between T. Daugherty and J. Hickson of GAI and E. Wood and D. Foster of GE). The 50 gpm rate stated by GE is a nominal value and higher or. lower rates are acceptable. GE has stated that a faster rate can be achieved by draining to the condenser rather than the clean radwaste system as long as water chemistry limits a:e not exceeded. GAI has requested GE to confinn these discussions in writing (Ref. PY-GAI/ GEN-2964, dated 1/3/84). Based on the above, this Observation has no impact on design ur safety. O Approvals j /fy/f y enginator M, W, 7A.# Date f [g 4,[ g4 . Project Engine.r yh(d, _ ] g, Date Project Manager M f Date J fpgf_ f jj //zWe/ e o R.or.

t. m.

A,- P y ArJ o=' s 7 Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review 1 + ,-g.. .n.,

w - = = -. C EI observation V Record Closure Attachment B Observation No. Checklist No. Revision No. ME-03-04 ME-03 0 EDDR No. QAD 600 No. Sheet of 072 N/A 1 1 Yes No Clcsed X Is lated X Patential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion This item is also addressed in observation ME-03-01, item d. 3.0 Action Taken GE Criteria Compliance Review will be performed. i M.0 Conclusion There was no effect on the systems reviewed by Cygna, and the generic s cview listed above will insure documentation of reconciliation of variations from GE criteria. 5.0 References (9) EDDR 72 (25) Perry Nuclear Power Plant GE Criteria Compliance Review Procedure, Rev. O. i Aphrovals Originator g Date p S rdor Pr6jTEngineer 'x. Ar~ d s7's / e. + Date CEI Superv ty Unit Date gal Project Manager & J A Date $/7/Bf w L gal Manager CorporpK PF Date n,,,,nn 4 [,The Cleveland Electric illuminating Company: [ l rry Nuclear Power Plant Piping Design Review g v DW138/20/Q/sp. --s-.r --,w-.;.3- ,w w -,y-ey -,-wy-ygy,. g 9, -p'

i Observation l L41 6'fd Record lillillllllililllllllllllillli Revision No. O l . Observation No. ME-03-05 Sheet of Checklist No. hE-03 MSDS Item #4 & #5 3 3 fjj/p Date l Criginated By W t, yg f /g3 Raviewed sy y {ft Q j -- Date cy u 1.0 Description The closing speed specified for valves F016 and F019 in GAI Specification 521-02-4549-00 and bill of material RNU-202 is " Vendor Standard." The Borg-Warner vendor drawing 81180 states that the valve closing time is 20 seconds maximum. This closing time corresponds to a minimum closing speed of approximately 9 inches per minute for a 3 inch valve. 2.0 Requirement The GE Nuclear Boiler Design Specification Data Sheet 22A4522AR Section 3.1.17.1 states that valves F016 and F019 shall have a closing speed of at least 12 inches per minute. I 3.0 Reference Documents . O 3.1 Nuciear Bon er Desisn Sp - f utions 22u S22, a. 5 3.2 Nuclear Boiler Data Sheet 22A4522AR, Rev. 2 3.3 2-1/2 inch and Larger Valves SP-521-02-4549-00, Rev. 5 . 3.4 Valve Assembly, Ga%3 fnch,1,500 C S. Motor Operated Drawing 81180, Rev. H 4.0 Potential Design Impact Since the valve minimum closing speed of 12 inches per minute was not specified in the GAI purchase specification and the vendor drawing only indicates a maximum closing time of 20 seconds, it cannot be determined if the %7ve meets ~ ~- the GE criteria. 5.0 Probable Cause Design control. Attachments A. Observation Record Review l b Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

_..__.Z__. ~ ~* Observation AM' TJ Record Review nunnunninninininn Attachment A Observation No. ME-03-05 Checklist No. ME-03 Revision No. O PFR No. Sheet 1 of } Yes No Closed X Extent 3 of 3 Systems with inconsistencies between valve data and GE requirements Comments GAI has received verbal concurrance from GE (Ref. telecon PY-GAI/ GEN-2903T between T. Daugherty of GAI and E. Wood of GE, dated 11/4/83) and has requested written confirmation (Ref telecon PY-GAI/ GEN-2964, dated 1/3/84) of the acceptability of the closing speed of valves B21-F016 and B21-F019, which is slower than the GE requirement. Per memo T. Daugherty to M. Stewart dated 12/2/83, GAI is initiating an SAR change to Table 6.2-32 to reflect the 18.5 second closing time of these valves. Based on the above, this Observation has no impact on design or safety. O I Approvals D*'* //A//Y orien.ior 1{, tJ 7/m A G_* m n = o rhr./e4-Pr D, er >ct unen r onen u'****' %rT~dblL Vh /^4- //zd/#'/ Cai R r ai.iiv. 'j-g pah o*ie / Cleveland ETectric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

..my -n. -= n 7_ l (El Observation '~' Record Closure Attachment B Observation No. Checklist No. Revision No. ME-03-05 ME-03 0 EDDR No. QAD 600 No. Sheet of 72 N/A 1 1 Yes No Cl; sed X lstiated X P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required. 3.0 Action Taken GE acceptar.ce will be documented, and the GE Criteria Compliance Review will be performed. q (d.0 Conclusion The review listed above will insure that all GE criteria have been met, or deviations documented as acceptable. 5.0 References (3) EDDR 72 (25) Perry Nuclear Power Plant GE Criteria Compliance Review Procedure. Approvals Originator g Date p Senior Prof 6Nc 'ngineer 'fA r/ss /.s + Date A. A CEl Supervis ty it U Date gal Project Manager Date b_ ~ ^., $/WEY w gal Manager Corporap PF Date The Cleveland Electric illuminating Company: grry Nuclear Power Plant Piping Design Review [ DW138/21/Q/sp 1 w_.

Tile CLEVELAND ELECTP.1C ILLUMINATING COMPANY WGENN Piping Design S'"' ": 3.2 / O 3 PAGE: Rev,ew i ,,,,s,os; 0 5.2 PIPING ANALYSIS OBSERVATION STATUS EDDR NO. FOLLOW-UP SCHEDULED OBSERVATION 03FICIENCY OR ACTION COMPLETION DATE NO. YES/NO GC PRE NO. COMPLETE FOR FOLLOW-UP COMMENTS PI-00-01 NO NA NA PI-00-02 YES 64 YES, 1/13/84 NA PI-00-03 NO NA GA l I PI-00-04 YES 63 NO JUNE 30, 1984 FI-01-01 7ES 66 YES, 3/19/84 NA oPI-01-02 NO i.A NA

• IhVALID 03SERVATION PI-02-01 NO YES, 1/13/84 NA PI-03-01 NO YES, 1/30/84 NA I

I PERRY NUCLEAR POWER PLANT Serving The Best Location in the Nation PO box 97 e PEAAY OHIO 44081 e TELEPHONE #216e 25 4 -J 73 7 e ADDRESS-to CENTER AOAD L_

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g Observation MM' T3 Record lillilillillill!!" Hun'!lil l

Revision No. O Observation No. pl.00-01 Checkilet No. PI-01, -02, -03 General sheet 1 of I {%fAfg3 cricinated By @O' 2-Date /MMy Reviewed By { h,]g fg Date 1.0 Description Support flexibility is not considered in Class 2 or Class 3 piping analyses. Supports are input as rigid and then designed using a maximum deflection criterion of 0.1". 2.0 Requirement Cygna Review Criteria 83102-DC-1, Rev. O. Sect. 4.8.9. 3.0 Document Reference 3.1 gal Analysis Report No.1821G08A (MSRV) 3.2 GAI Class 1 Analysis Guide No. 04, Rev. C O 4.0 Design Impact Lars,e variations in as-built support stiffness as compared to the analyses could sis'lificantly change system mode shapes, load distribution, support loads and pipe stress. 5.0 Probable Cause Standard GAI practice. Attachments A. Observation Record Review f l l-O . Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review a_

m Observation &t Record Review i imimmmimmimiim Attachment A Observation No. pl.00-01 Checklist No. PI-01, -02, -03 Revision No. O PFR No. Sheet 1 of 1 Yes No Closed X Extent All 3 Systems Comments The use of rigid supports is acceptable provided that the GAI deflection criteria of 0.1 inches is sufficient to provide assurance that the flexibility of the supports will have no significant effect upon the piping analysis results (stresses and loads). An approximate evaluation of this issue can be made utilizing a cantilevered support (limiting case) with a pipe / support system f requency of 33 Hz (i.e., the " rigid" range of the seismic spectra). Under an applied load approximately equal to the tributary mass weight on the support, the deflection, 6, for this system is approximately l f= =+ 6 = 0.01" This is 1/10 of the value required by the GAI criteria. This shows that the supports, themselves, can be subjected to dynamic excitation due to loads well above the IPA level. Based on the above, Cygna performed a review of the pipe support deflections and stiffnesses for the Main Steam Relief Valve Discharge System 1821-G08. This review considered the GAI design calculations as well a's some approximate hand calculations by Cygna. The review indicated that the deflections of supports on this system were well below the 0.1 inch limit and that the corresponding stiffnesses were sufficient to provide confidence that there would not be any significant impact on the loads and stresses in this system. i Approvals l/(,,[g4 Originator Me -M Date s Project Engineer Q Yd Date ((4/g$ \\h Project Manager /,[h Date ffg4 g/y/gf CEI Representative fg'Mj Date Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review .1

CEI Observation (3 Record Closure U Attachment B Observation No. Checklist No. Revision No. PI-00-01 PI-01, 02, 03 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No CI: sed X ls: lated Not applicable Patential Design impact X 1.0 Description See Cygna cbservation record and observation record review. 2.0 Discussion Observation noted different techniques being used on Class 1 and Class 2 or 3, not a deficiency. 3.0 Action Taken l Nono required 4 h.0 Conclusion No design criteria has been exceeded or ignored in gal approach to this problem. gal has chosen different methods to mest the ASMd ree,uiremant. Both methods are adequate to insure the acceptability of the piping design for Perry N'sciear Power Plant. 5.0 References N/A Approvals ~ Originator Date Senior Pr(ec ngineer' Date b bh W' kY b CEI Supervisor Quali Audit Unit Date WA. 2' YH lTL/ gal Project Manager ~ () //. Date ~ 6)J,w 67 9/B9' -+ [ _ gal Manager Corporatey & Date The Cleveland Electric tiluminating Company: j/' _ Parry Nuclear Power Plant Piping Design Review DW138/4/Q/sp

Observation [4M' fd Record tilllll!!!Ill!!!!!Illllilllll! gU R nision No. O Observation No. PI-00-02 Sheet i of 5 Checklist No. PI-01, PI-02, PI-03 General Date

1. 2 /2 /s Crielnated By Q d. _*

cd (( -[ Q Reviewed By @R (&[ Date 0 1.0 Description The following items sununarize minor inconsistencies noted during the review of the MSRV, HPCS and MSD piping analyses: a. Deleted. b. In the functional capability check for the SRY discharge line (IB21-G08A, Rev. 2), the worst case was not examined for a reducing elbow. Specifically, the 12 inch end of a 12" x 10" 90" reducing elbow was examined, but not the 10" end. The was expressly omitted because a 10" 45" elbow, having higher stresses, had already been examined. In this case, and in general, such logic is not appropriate because the stress indice for a 45" elbow is nearly 30 percent lower than fcr a 90 elbow. c. In the calculation for codeling gate valves for the piping analysis, four O mass points are included: (1) operator, (2) stem and yoke, (3) bonnet and (4) body. There is ne mass point for the gate. Consequently, the mass moment of inertia is underestimated. For valve 1E22-F036, this technique results in the following calculated values: o mument arm w/o gate = 13.7 in. moment arm w/ gate = 14.4 in.* ratio = 1.05

• The actual moment arm shown on the vendor drawing is 14.90 in.

d. As shown on Fig.1. MSD piping is enclosed by a guard pipe from the l drywell to the shield wall. The guard pipe is connected to the drywell and is isolated from the shield wall and containment vessel by bellows. In performing the thermal modes analysis for MSD piping, thermal movement of the shield wall and containment vessel are expressly excluded due to the bellows at those points. Thernal movement of the drywell, on the other hand, is neither included nor addressed. f) l Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation [4M' M Record lillllllillllllllilllllllillli i Revision No. O Observation No. PI-00-02 5 sheet 2 Checklist No, pl.01, pl.02. PI-03 General Date gg/g Criginated By QQ

_ p kk T3 Date Reviewed By The weight of water was included in the deadweight and all dynamic e.

analyses for the MSD piping. This line is always filled with steam except during hydro testing. It should be noted that the thermal transient analysis was properly done considering the fluid properties of steam. 2.0 Requirement a. Deleted. b. Interim Technical Position " Functional Capability of Passive Piping Components," Mechanical Engineering Branch, Division of Systems Safety. c. Cygna Review Criteria, 8310?-DC-1, Section 4.7.6. Neights and centers of gravity shall be as specified on the applicable vendor supplied valve assembly drawings." d. All significant thermal anchor covements should be considered. e. N/A. 3.0 Document Reference 3.1 Deleted. 3. *c. Deleted. 3.3 GAI " Document Evaluation of Functional Capability of Piping Components", dated July 29, 1982. (b) 3.4 GAI Stress Analysis Report IB21G08A Rev. 2. (b) 3.5 Borg-Warner Drawing 81030. (c) 3.6 Borg-Warner Report No. 81030 (GAI No. 4549-94Q-386-1). (c) 3.7 GAI Calculation File No. 2.69.2, RNU 226. (c) 3.8 GAI Analysis Report IN2201C. (d) t Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation L4M'M Record lllllllillllilllllillllllllill Ruiston No. O Observation No. PI-00-02 sheet 3 5 Checklist No. PI-01, PI-02, PI-03 General 4[q/p Criginet.d By % h _. _* {~ Dat. ri-i-?s c..i...d er 6rg, g c;

o..

v 3.9 hutech Report Py-NTC-;AI-034, Rev. O. (d) 3.10 GAI Analysis 1N22G01C, Rev. 2. (e) 4.0 Potential Design Impact a. Deleted. b. This system still meets funcitonal requirements, however the margin is reduced from over 30 percent to 4 percent. If this same assumption was used for other similar lines, it could lead to the functional requirements not being met. c. It should be noted that the valve is appropriately modeled to simulate the fundamental frequency predicted by the vendor. O valve loads transferred into the piping are directly proportional to the moment arm. For Valve 1E22-F016, this corresponds to a 5 percent increase in loads, which is insignificant. However, this matter should be investigated for other g:te valves en PNPP. d. Thermal stresses in the guard pipe and at the piping / guard pipe iuncture may be incorrect. These predicted stresses will be unconservative only if the piping and drywell grow thermally in opposing directions. i f Cleveland Electric. Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation A L% i Record {'J] liiiiiiiiiiiiillililllllllilli s Revision No. O Observation No. pl.00-02 sheet 4 o' 5 Checklist No. PI-01, PI-02, PI-03 General Date ggfg3 f Criginated my Q Q." my ti-t p n.wi..eo ey QcK h o.te u e. The change in weight is summarized as follows: Input Actual Pipe Weight Weight Size (lbs./ft.) (lbs./ft.) Decrease 2" 11.91 10.94 8.1 3" 20.72 18.38 11.3 'l This 11.3% in mass could increase the frequencies by as much as 5.5%. This small shift in frequencies will not significantly affect the dynamic analysis due to the conservattsms of the response spectra analysis and the broadening of - ~ spectra peaks. 5.0 Probable Cause Minor oversights in the analysis and design. Attachments A. Observation Record Review l i Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record I a i t 11lll16111llll1111111111 e Revision No. O Observation No. pl.00-02 5 8h* 5 Checkitet No. pl.01, PI-02, PI-03 General Date jafs/g3 Criginated my LQ(d, ', _ j ~ [I h-Q Rwlewed my (Q, Date I FIGURE 1 M ') o* . #.........t# % .. /.7... h Process Pipe j Y-----------..-----------,_,---- k N Guard Pipe = s h. Y e Drywell Shield containment Wall Wall Vessel l 1 4 l O l Cleveland Electric Illuminating; 83102 Perry fluclear Power Plant Piping Design Review

Observation L4Dh' fd Record Review ninilinilliiiiiiiiiiiiiiiii Attachment A Observation No. p}.00-02 Checklist No. pl.01 -02, -03 Revision No. O PFR No. Sheet 1 of 1 Yes No i Closed X I Extent All 3 Systems Comments ) Based on evaluation of each of the noted items in this Observation, Cygna concludes that individually these items have no impact on design for the three systems reviewed. In addition, due to the small number of items per system, there are no cummulative effects. O l l l l 1 Approvals l/gf[g p Originator Q1d.__ .._ _ 4 Date fgfg Date Project Engin r Project Manager M Date /g i y,7g, Ce,R.,r t.tt,. p g u rj D.,. Cleveland ElectIc Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

CEI Observation {~T Record Closure w,/ Attachment B Observation No. Checklist No. Revision No. PI-00-02 PI-01, PI-02, PI-03 0 EDDR No. QAD 600 No. Sheet of 064 N/A 1 1 Yes No Cissed X isclated X Item B Other items not applicable Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion items A, C, D, E were conservative assumptions or insignificant as described below. 3.0 Action Taken item B initiated EDDR No. 64, and associated gal review for similar discrepancies. 4.0 Conclusion a. Deleted - Observation invalid b. Discrepancy was isolated as determined by EDDR 64 c. This item resulted in a 5% discrepancy in the mass moment of inertia of the valve. This is judged to be insignificant. In addition it should be noted that this value would very depending on the valve position. d. The thermal movement of a five foot concrete wall. from ambient temperature effect was correctly assumed insignificant. This was judged to be a conservative assumption which is acceptable e. regardless of the likelihood of the event. The effect of small frequency shifts will be offset by the additional mass of-the system. 5.0 References (1) EDDR 64 Approvals Originator Date g S:nior ProjVci4ngineer.a s' s % /44 Date P CEI Supervisor Quality Audit Unit Date W HW W// /TV gal Project Manager &) n F/W8'/ '~ Date gal Manager Corporat@7 Date The Cleveland Electric lliuminating Company: P;rry Nuclear Power Plant Piping Design Review [ DW138/5/Q/sp +- e-w-

.=:..-.... Observation M Uh i Record lililllillittiilililililllilli Ruiston No. O Observation No PI-00-03 Checklist No. pI-01. PI-02. PI-03 General sheet of i 3 Criginated By @ d.' yd-Date gfg/p l( h g3 Reviewed By hg( Date 0 1.0 Description The following items either lack documantation or utilize inconsistent data: a. GAI Specification B21 requires that SRV piping within the drywell be designed for a post-LOCA condition temperature of 250'F.195"F (185 + 10) was used. b. Deleted. c. Deleted. l d. There is no documentation within the calculation package justifying the thicknesses used in the thermal transient analysis for: Q

1) Reactor Nozzle (HPCS)

2) Sweepolet (HPCS)

3) Valycs (MSD and HPCS)

4) Penetration (MSD)

5) Tee (MSD)

.e. There is no documentation justifying the exclusion of the effects of bend or elbow ovalization for the HPCS. f. There is no documentation indicating that the movement of the Main Steam lines during turbine trip has been considered for its effect on the MSD l lines. I l 2.0 Requirement l a. GAI Project Design Specification, DSP-B21-1-4549, Rev. 1 Table 6. 1 b. Deleted. l c. Deleted. Cleveland Electric 111uminating; 83102 Perry Nuclear _ Power Plant Piping Design Review

Obsorvation Record A L.4 i 1 m !ummuumqilllllllllll 7 observation No. PI-00-03 Revision No. 0 ' Checklist No. PI-01, PI-02, PI-03 General sheet p o' 3 Criginated By QA, *

j Date dpy 1-[ k 73 Reviewed By h [)( [gg Date V

W d. Standard industry practice. e. ASME B & PV Code Section III 1974 with addendum through Winter 1975, Subsection NB, Paragraph NB-4223.2. f. N/A. t 3.0 (4teference Documents r: m 3.1 GAI Analysis Report No. IB21G08A, Rev. 2. (a) 3.2 Deleted. 3.3 Deleted. 3.4 GAI Analysis Report Nos. IN22G01C, Rev. 2 and IE22G04C, Rev. 2. (d)

3. f. GAI Analysis ' Report No.1E22G04C, Rev. 2.

(e) 3. GAI Analysi_s Report No. IN22G01C, Rev. 2. (f) j.4.0 Potential Design impact a. The following table shows the temperature considered in designing a portion of the SRV piping. TH1 TH2 TH3 TH4 SECTION (UPSET) (UPSET) (NORMAL) (POST-LOCA) \\ 1 450"F 450"F 145"F 195"F .b e, + 1'95"F TO h50"F is a significant temperature rise, which could impact design stresses. However, taking into account the other design conditions (upset temperature = 450"F) and the higher allowable normally associated with post 40CA event, the cversight in design will have no impact. k'. .$. },' L. p ' ' !> f I Cleveland Elects,1c Illuminating; 83'107 hrry Nuclear Power Plant Piping Design Review

u

^ Observation L4 L% i Record lilillililll!!!nunmi!!!ir Revision No. O Observation No. pl.00-03 Checklist No. PI-01, PI-02, PI-03 General sheet 3 of 3 Date (S/pg Cricinated sy QQ_*yA qQd q C Q @,, Date noviewed my 6 b. Deleted. c. Deleted. d. Individual loose sheets indicate that the values are appropriate. These sheets should be incorporated into the analysis package. The following calculation shows that the pressure stress indice may e. increase by as much as 3 times. Per NB-4223.2 ovality is limited to.08 x Do as a maximum (could be less) s .. F, = 1 +.08 ({) 3 O = l +.08 (12.75) (1.5) 1050 .687 (1+.455(12.75)3 27X10) 6 .687 F =3 la kl ..X{=Fla

• K1=3x1=3 l'

e This would be a maximum. For ANSI B16.9 elbows, the out-of-round may be less. f. Additional stresses may occur in the drain lines due to the movement of the Main Stream lines to which they are attached. 5.0 Probable Cause I l Document and design control. Attachments {' _ A. Observation Record Review L j' - \\ L. i. r & lit-$N t ~ Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation dim i Record Review nmannmumunnnin Atiachment A Observation No-PI-00-03 Checklist No. PI-01, -02, -03 Revision No. O PFR No. Sheet I cf 1 Yes No Closed X Extent All 3 Systems Comments Based on evaluation of each of the noted items in this Observation, Cygna concludes that individually these items have no impact on design for the three systems reviewed. In addition, due to the small number of items per system, there are no cummulative effects. . O Approvale - Ihs/e+ oren. tor skh & w c=' f[y[g Project Engineer T._Q Date ifgfg Project Manager j jj Date p/y'/pg ll Date CEI Representative g [ Q_g jy Cleveland Electric Illuminating; 83102 Perry Nuclear' Power Plant Piping Design Review

CEI Cbservation Record Closure d Attachment B Observation No. Checklist No. Revision No. PI-00-03 PI-01, 02, 03 0 EDDR No. QAD 600 No. Sheet of NJA N/A 1 2 Yes No Closed X lsolated Not applicable Pstential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required. 3.0 Action Taken None required. 4.0 Conclusion a. No impact because the upset temperatures bound the post loca ambient g- ~ temperatures. In addition, the GAI final piping analysis calculation review (79-14) includes an ambient temperature review. This will insure this kind of oversight will not affect the design of any system. b. Delete - Observation invalid. c. Delete - Observation invalid. d. Descriptions of all models used in the thermal transient analysis are provided in the analysis package. Discontinuity thicknesses used in the one-dimensional heat transfer analysis are based on the maximum thickness ~ occurring dt distance from the location being analyzed. In cases where maximum thickness is used, no additional justification is given or considered necessary. For those cases or models for which maximum -thickness cannot be used, sketches and descriptio.ns are provided in th'e calculation, e. .The gal assumption was that this term has -a negligible effect. gal has demonstrated that this is a correct assumption for the pipe sizes and wall thickness used on the Perry project. j f. The movement of the main steam line at this point was correctly assumed j to be negligible. The maximum movement from the GE stress report was 0.02 inches. l O. V' +-~o nr - e-,< r v-m-ww +s-,-,

---

v-

• > -A+

nn v

% 4y39 7 [~^ ~}gfG; " L( El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. 'PI-00-03 PI-01, 02, 03 0 EDDR No. QAD 600 No. Sheet of N/A N/A 2 2 5.0 References (15) gal memo dated March 19, 1984 from J. T. Zalewski to C. W. Whitehead O Approvals Originator Date S:nior Pr inee Date CEI Sup Qu Audit Unit Date gal Project Manager Date gal Manager CorporapA~ Date Tha Cleveland Electric liluminating Company: P rry Nuclear Power Plant Piping Design Review [ DW138/27/Q/sp P1 . Q,) o e4

Observation L4h i Record lilllllllllllllllillllllllill! Ruiston No. O Cbservation No. PI-00-04 Sheet of Checklist No. PI-01, PI-02. PI-03 General 1 p l-} _hh Date Criginated By C, K, b)ggCy f/3/g Reviewed By (((k[__ Date (I C 1.0 Description The following analysis oversights are noted for Jet Impingement load calculations: Main Steam Safety Relief system,1821 G08(A), Rev. 2, Shts.17.5 thru e 17.10. al. Case 6.b in Table 7 of specification. The jet load input at node point 11 should be -F instead of F* (597.3#), since local coordinates are used for thatnodepoTnt. High Pressure Core Spray, 1E22G04(C), Rev. 3. e bl. Item IC of Table 7 in specification (break LPB2LL). The total load computed is 6902.6#. The total load specified in the design specification is 7488#. b2. Item 2 of Table 7. Break SD3A. F component should be included in the input. z b3. Item 3 of Table 7. Break SB3A. F component should be included in the input. 2 b4. Item 7J of Table 7. Force input at node A18 should be at node B18 (difference of 0.566' in i elevation). b5. Item 8 of Table 7. B33 Break RD7 (header side) Loop "B". Jet loads on piping and valve E22-F036 are not included in the calculation. This is listed as an analysis exception in the Class 1 Stress Report, P-1001, Rev. O. b6. The load input for nodes 18 and A18 (Jet 6D) are interchanged. b7. At node 13, a negative load of -1122.0# was input as a positive load. O V i I Cleveland Electric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review

Observation LM% i Record lll1111111111111lllll11111ll11 Revision No. O Observation No. PI-00-04 Checklist No. PI-01, PI-02, PI-03 General sheet 2 2 {-3-% ) Criginated By (, h, gp Date 1/3/84 Reviewed By g), [ _ g,7_ k Date b b 2.0 Requirement 1. GAI specification - DSP-B21-1-4549-00, Rev. 2. 2. GAI specification - DSP-E22-1-4549-00, Rev. 2. 3.0 Reference Documents 3.1 GAI analysis IB21G08(A), Rev. 2. (a) 3.2 Computer output for 1821G08(A), Run #JOHNVXW (1/12/83) (a) 3.3 GAI Analysis 1E22G04C, Rev. 2, Run #3, E22G4J Run 10=A0XZGCL (3/28/83) (b) 3.4 GAI Class 1 Stress Calculation IE22G04C, Rev. 3. (b) 4.0 Potential Design Impact 1. Individually, no significant impact. 2. The combined effect could impact the accuracy of the analysis. 5.0 Probable Cause Analysis oversights. Attachments A. Observation Record Review 1 l l l

O i

Cleveland' Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review l

Observation [4M' TJ Record Review m!!m!!!mm!mmiimi Attachment A Observation No. PI-00-04 Checklist No. PI-01, -02, -03 Revision No. O PFR No. S h.e t } of 1 Yes No Closed X Extent 2 of 2 Systems with Jet Loading Comments Further review indicates the following: a. This item had been noted by the GAI verifier in Calculation 1821G08A and was determined not to be significant enough to warrant reanalysis for the MSRV system. Cygna concurs with this conclusion. b. As a result of this Observation, GAI has performed a reanalysis for the HPCS system incorporating all the specified corrections. Cygna has reviewed the input calculation for this reanalysis. GAI has stated that there was not any significant change in the results (it should be noted that per GAI, the piping -is now shielded from B33 Break RD7 which closes item b5). Based on the above, this Observation has no impact on design or safety. Approvals D * I-M'O orleinator C, % @(v\\St. i pro moi a - A n. 2. 2 th484-o.'. O'e,one,u..e.r % g pr R ' iA4pa. o.s. rXy/sv CaiR.,r. ... /scr m o.i. Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

CEI Observation c'S Record Closure V' Attachment B Observation No. Checklist No. Revision No. P I-00-04 PI-01 PI-2, PI-03 0 EDDR No. QAD 600 No. Sheet of 065 N/A 1 1 Yes No ClI; sed X !s:; lated X Pctential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion This observation resulted from CYGNA reviewing the calculation before the Jet work was completed. The Jet maps have been updated since the CYGNA review and will be again, before they are final. ~ 3.0 Action Taken The generic jet calculation (File Code X176) is being finalized. This will include a matrix of all jets, shields, and piping or supports impacted. it will document that (-) all jets are accounted for and all piping impacted is properly analyzed and all pipe supports impacted are either shielded or designed to withstand the impact. 4.0 Conclusion Based on the jet calculation X176, this item will not impact plant safety. 5.0 References (2) EDDR 65 (16) Generic Jet Calculation - File Code X176 Approvals Originator Date Senior Pr Et-in ,Date CEI Su ervisor Quality Audit Unit Date

M_ J. R s u T/9/sW GA1 Froject Managf

~~ Date x f//d/8V ^ gal Manager Corporp Pf gg g Date ' The Cleveland Electric lliuminati6g Company: / P;rry Nuclear Power Plant Piping Design Review uW138/Q/37/es

i Observation 41 n i Record 111llll11!!!!!!n""""'"" Revision No. O Observation No. PI-01-01 sheet i of 2 CheckHet No. PI-01 MSRV ghf Q Caleinsted By M._ * -M Date l[._ 2.Q ] Reviewed By hg Qg Date 1.0 Description The stress intensification factors (SIF's) at points 2, F1, and F2 are not input properly. POINT ACTUAL SIF INPUT SIF ANALYSIS (PIPE / FLANGE) 2 2.1 2.1 1.083 F1 1.9 1.9 2.889/1.766 F2 1.9 1.9 2.889/1.766

Where, actual SIF = ASME value t

input SIF = value input to the TPIPE analysis l analysis SIF = value utilized by TPIPE 2.0 Requirement ASME B & PV Code, Section III,1974 with Addenda to Winter 1975 Subsection ND, Fig. 3673.2 (b)-1. 3.0 Document Reference 3.1 GAI computer analysis 1821G08 Rev. 2 l 3.2 TPIPE Manual. O \\ Cleveland Electric 111uminating; 83102 \\ Perry Nuclear Power Plant Piping Design Review \\ __,m._

Observation ADhi. Record O ll111111111ll11lllll1lll1 11111 Revision No. O .l Observation No. PI-01-01 8h"' Checklist No. PI-01 MSRV 2 2 Date ggfg/ g3 Criginated By M Q (I k.@3 Date Reviewed By K, (y 4.0 Potential Design Impact Using the actual intended SIFs at these points results in the following ratio of maximum to allowable stress: POINT MAX. STRESS / ALLOWABLE 2 0.16 F1 0.26 F2 0.28 These revised stresses are clearly well within the allowable limits. 5.0 Probable Cause O This observation resulted from the analyst's attempt to override an internally computed SIF. This is specifically cautioned against in the TPIPE manual. In addition, the analyst did not review the program's interpretation of the SIF input. Attachments A. Observation Record Review I Cleseland Electric Illuminating; 83102 i Perr/ tluclear Power Plant Piping Design Review

Observation t4Ph i Record Review Q""""""""""""""" Attachment A observation No. PI-01-01 Checklist No. pI-01 Revision No. O PFR No. Sheet 1 of 1 Yes No Closed X Extent 1 of I Class 3 Systems Comments As shown in Section 4.0, the increased stresses using the correct SIFs, are still within the Code limits. Therefore, there is no design impact on these three systems. Even though there is no design impact on this system, GAI plans to correct the SIFs and include the corrected stresses in the analysis package. J t Section 4.0 also shows that stresses at the points of concern on the SRV discharge j increased to up to 28% of the Code allowable when the correct SIFs are applied. Cygna did not evaluate the impact of this issue on systems where the design margin may be less than that found in the SRV discharge. O l Approvals I/es/s( orwn. tor Lf_\\. qd, ? - A.- o ='a s f /jg[gy Project Engineer Q,M, Q _ _ _ {} 7 [~ Date Project Manager ((gfj Date

l. jg.3 4

//20/8v cEin.pr. i.uv. p,r ' h o =i. ClevelandEMricIlluminaTing;83102 Perry Nuclear Power Plant Piping Design Review

~- El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. PI-01-01 PI-01 0 EDDR No. QAD 600 No. Sheet of 065 N/A 1 1 Yes No Cli; sed X isolated X P;tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required 3.0 Action Taken Initiated EDDR No. 66 and associated gal review to determine that it was an isolated case. [').O Conclusion v As indicated on the EDDR and the gal memo, there was no effect, and this was an isolated case. 5.0 References (3) EDDR No. 66 jfq(15) gal memo dated March 19, 1984 from J. T. Zalewski to C. W. Whitehead. Approvals Originator Date g g,g S3 nice Projecf ineer / Date 5///8/ 8h A. N"W CEI Supervisor&uall y A T//a/ Ty it Unit Date .sr gal Project Manager // Date 2Q]M 5/7/or v gal Manager Corpora @K pf o Date The Cleveland Electric illuminatin's Company: / crry Nuclear Power Plant Piping Design Review DW138/6/Q/sp e g +

Observation A f.% i Record 11lll1111111!!!!!!!!!'nn'111 Revision No. O Observation No. PI-01-02 sheet of Checklist No. pl.01 MSRV 3 3 g/gg Criginated ByQgf M Mo#g g gpg* Date [f gg Reviewed By 'Ql], * * ]/ Date s 1.0 Description MSRV seismic anchor movements.(SAM) in the z-direction are applied in the x-direction at point J1. 2.0 Requirement Standard Industry practice. 3.0 Reference Documents GAI TPIPE Computer Output 1821G08, Rev. 2. 4.0 Potential Design Impact

n Inputting SAM in the wrong direction will result in an incorrect stress distribution that may impact design of the MSRV piping supports.

v 5.0 Probable Cause j Analysis oversight. This occurs at one out of two points where movements are input in the analysis for subsystem IB21-G008. Attachments A. Observation Record Review l !o l Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record Review 4 f. t i t~imunnunmmimu Attachment A Observation No-PI-01-02 Checklist No. PI-01 Revision No. O I ,FR No. Sheet 1 of 1 Yes No Closed X Extent 1 of 3 Systems Comments The seismic anchor movements were correctly input by GAI in a local coordinate system corresponding to the direction of the restraint at point J1. Therefore this observation is invalid. l l i i i l ! O i t l \\ l Approvals orie neter 1/un A.W / n Date 1 8'f O, Pre >ct eneneer4@Qic A ~ l o * ilaled ,..c. e u Ts,;/,ff,,

1. 4jo o.te Date }/jo/B 4/

Can Repreeeeeeuwe M.40* M*-a - Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

1~s

c:. m w=,-m m m -

n s lEl Observation Record Closure Attachment B IJbservation No. Checklist No. Revision No. PI-01-02 pl-01 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No Closed 'X isolated Not applicable Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None ;equired 3.0 Action Taken None required (l.0 Conclusion V Observation was invalid. 5.0 References None Approvals Originator Date g Senior Proja perg n Date CEI Supervisor _ Quajjty Au it Unit Cate T, ' /T V &. A. _n /s. gal Project Manager Date Qh O~/ 9/gs) n gal Manager Corporap P7 Data he Cleveland Electric illuminating Company: / erry Nuclear Power Plant Piping Design Review DW138/7/Q/sp J

Observation Record L*Il L i lllll111111111ll1111111ll11111 Rnision No. O Cbservation No. PI-02-01 of CheckNet No. PI-02 HPCS sheet i 2 Celeinsted my Q], jg Date ggfG /2/,2/g3 Ceviewed my (),gj, C Q,,JLoff p oste p'~ L ' / / 1.0 Description The fatigue analysis did not consider the different thermal gradients (AT1 and AT2) for the sweepolet and socket welded boss. The piping thermal gradients were input as the default values and these were not overridden for the sweepolet and socket welded boss. The thermal transient analysis indicates that the only instances for which this happens to be non-conservative is for the sweepolet (Point C24) during the up transients. In addition, the thermal transient analyses considered the flow to be zero at these same points. While this may be conservative when determining the A - T ), it is non-conservative in the calculation of discontinuity stresses (T B the thermal gradients through the thickness (aT1 and AT2). 2.0 Requirement a. ASME B&PV Code Section III 1974 with Addendum through Winter 1975, Subsection NB, Paragraph NB-3653. 3.0 Document Reference 3.1 GAI Analysis Report and TPIPE computer output 1E22G04C, Rev. 3. 4.0 Potential Design Isapact The stress increases at the sweepolet (based upon the original thermal transient analyses) are listed below: Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record A L.n i I;liiiiiiiiiiiiiiiiiiiiiiiilli Revision No. O Observation No. PI-02-01 Checklist No. PI-02 HPCS sheet 2 of 2 4ftlg crieinsted my @ M Date ajj'/n Mt_, _ cq,M o.t. c.,s...d my Event Sweepolet Piping Temp. Increase Stress Increase E =aT E=aT K E=aT 3 2 3 3 AT1 AT2 AT1 AT2 AT1 AT2 z(l-v) 1-v Z(1-v) ("F ) ("F) ( *F ) ("F) ("F) ("F) (PSI) (PSI) (PSI) 12D 133 23 53 8.5 110 14.5 13306 3509 22620 20A 144.5 25.0 54.0 8.5 90.5 16.5 10947 3991 18610 20A 120.5 20.5 45.5 7.0 75. 13.5 9072 3266 15422 O It should be noted that the magnitude of the increase will go up when flow is considered. Tne sweepolet is already overstressed. Usage factor requirements are also exceeded for the sweepolet (2.7481) and the socket welded boss (0.2744 - No Break Zone). Both of these components will require more refined analyses as noted in the Class 1 stress report. The reanalysis should incorporate the impact of thic observation. In addition, these concerns should t,e addressed with regard to all Class 1 analyses due to the fact the impact may not be insignificant as shown by the above table. 5.0 Probable Cause Analyst oversight. Attachments I l A. Observation Record Review O l l Cleveland Electric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review

Observation Record Review 4L t i muunnsmum Attachment A l Observation No. PI-02-01 Checklist No. PI-02 Revision No. 0 ,FR No. 8heet 1 ef 1 Yes No Closed X Extent 1 of 1 Systems with Branch compor.ent where Branch piping is not modeled. Comments GAI has reanalyzed these components using a 20 finite analysis method (P-267 Rev. 1). Cygna has not reviewed this analysis and does not intend to do so within the scope of this review. Per GAI, in this analysis flow was considered in the crotch area and the results show that the components in question now meet ASME Code requirements. Based upon the above, this Observation is considered not to have any impact on the design or safety of the HPCS system. j

O i

j i i Apotevole I l Orieinster YM _ [d ii3/84 D*te l O,,ebet Easin*** b MOtA-s1 g j g g fgp. Date e t ee, mf7.y;2m A3/ 4 on,. cas nepreeeneetko M'My Syd/fa) Date ClevelandElechic111uminatTng;83102 Perry Nuclear Power Plant Piping Design Review

- j ~e e y= ~-= <~+ s

-%* w a= - 1

, _ ~ _ _ .. =. - ~ )El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. PI-02-01 PI-02 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No CI: sed X ls: lated Not applicable P;tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The Cygna review noted this observation in the middle of the gal design process. The component was identified by gal as being over stressed based on simplified analysis and was scheduled for a more detailed analysis. 3.0 Action Taken p() gal continued with their design process and reanalized the component using a 2D finite element method. 4.0 Conclusion No descrepancies of the type were noted on completed calculations. 5.0 References (17) calculation P-267, Rev. 3 s Approvals Originator g Date Senior Projiic ee Date 4 CEI Supervisor QuJal] y Au6t Unit W/.rv Data s v m. gal Project Manager 'g Date / VA A c L* w N7/94 gal Manager CorporategPrg@r Date The Cleveland Electric illuminating Company: / grry Nuclear Power Plant Piping Design Review DW138/28/Q/sp I_ J

_...~._...._._. _.. _ _ _ _ _ _.-_..___ _ _-- L L Observation 4 Ld i i Record ,,.m. mm...... m. Revision No. O Observation No. PI-03-01 sheet 3 of 3 checklist No. PI-03 MSD Crielnated By dk1_ *__,M Date 4fg/ g Date g h Ocylewed By ] ).g < t 7 g 1.0 Description The review of the thermal transient reanalysis (P-256, Rev. 0) did not consider the following discontinuities for evaluation of T -TB A 1. Valve coupling to 2" pipe. 2. 3" x 3" x 2" tee to 3" pipe. 3. 3" x 3" x_2" tee to 2" pipe. 4. 3" pipe to 3" valve. 5. 3" pipe to penetration. This analysis was rerun due to errors in fluid properties. It should be noted that the original analysis did consider these discontinuities. In addition there is no documentation to indicate that the fatigue analysis is to be rerun t using the later transient analysis data. Furthermore, the tee sections did not consider any additional thickness in the crotch area of the component. 2.0 Requirement ASME B & PV Code Section III 1974 with addendum through Winter 1975, Subsection NB-3653. 3.0 Refereace Documents 3.1 GAI Analysis 1N22G01C, Rev. 3. 3.2 GAI Analysis P-256, Rev. O. 4.0 Potential Design Impact i effects at these discontinuities, as well as the thermal gradient The Tg-Tg effects at the tee crotch areas, may be underestimated which may lead to failure in meeting ASME Code Requirements. 5.0 Probable Cause Analyst oversight. Attachments A. Observation Record Review i Cleveland Electric Illuminating; 83102 l Perry Nuclear Power 91 ant Piping Design Review

Observation Ai n i Record Review ur""'"'"'"ullis WhM A Observation No. PI-03-01 Checklist No. PI-03 Revision No. O of } Sheet } PFR No. l Yes No j Closed X Extent 1 of 2 Class 1 Systems Comments GAI has performed a 2D thermal discontinuity analysis (P-258, Rev. 0), for items 1, 4 and 5, and plans to incorporate this information in their next revision of the fatigue analysis. Regarding the tee components, GAI has performed a study using a thickness increase of 50% in a 1D thermal analysis. Based on vendor drawings, this is a reasonable value to assume at the crotch region for the purpose of this study. This analysis showed a maximum increase of 295% in the thermal stresses (from 1900 P51 to 5600 PSI). However, due to the very high margin to both Code allowable stress (15900 PSI = 30%) and break exclusion allowables (43%) at these components, this increase does not impact the design or safety of the Main Steam Drain system. O i I Approvei. 2ho/84 orieinator WL_ _M. 0*'* g,f(,fh6 O,Pr net uneneerMS.f 3]' Date ./,,f,4

m.,= 2 o...

,.....aoer 2/8/8f Ces Repreeenseewe ff D * '* Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

_w- --=--%, .c x'El Observation Record Closure Attachment B Observation No. Checklist No. Revision No. PI-03-01 PI-03 0 EDDR No. QAD 600 No. Sheet of N/A N/A 1 1 Yes No CI: sed X is: lated Not applicable P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Disc.ussion The purpose of calculations is to insure that the design requirements have been met, not that the exact stress is known. 3.0 Action Taken gal performed a study to insure that the conservatism in the simplified analysis p would compensate for discontinuities not considered. U 4.0 Conclusion gal has adequately demonstrated that a more detailed analysis will reduce the calculated stresses enough to offset the increases from discontinuities which were not considered. This approach insures all ASME design requirements have been met. 5.0 References (18) gal letter to Mr. J. M. Lastovka dated January 30, 1984 PY-G Al/CEl-15279. Approvals Originator g Date p g, g 4 Senior Pr6 ee Date CEI Supervisor Quality Audit Unit Date Pl. Yh % /#4 gal Project Manager Date gal Manager Corpora WP'ro Date Tha Cleveland Electric illuminatinh Company: / Oerry Nuclear Power Plant Piping Design Review DW138/8/Q/sp

THE CLEVELAND ELECTRIC ILLUMINATING COMPANY W En m W Piping Design "'" " 5 3 O PAGE. Review ,,,,s,os. o 5.3 PIPE SUPPORT OBSc,F'.nTION STATUS EDDR NO. FOLLOW-UP SCPEDULED OBSERVATION DEFICIENCY OR ACTION COMPLLTION DATE NO. YES/NO GC PRE NO. COMPLETE FOR FOLLOW-UP COMMENTS PS-00-01 YES 67 NO 9-28-84 PS-00-02 YES 68 YES, 5/10/84 NA PS-00-03 YES PRE-084 YES, 1/16/84 NA PS-00-04 was changed ( ) to PI-00-04 PS-00-05 NO NA NO 9-28-84 PS-00-06 YES 69/139 NO 6-15-84 PS-00-07 YES 70 NO 9-28-84 PS-01-01 NO NA YES, 1/24/84 NA PS-02-01 YES 71 YES, 3/31/84 NA PS-02-02 YES 65 NO 6-30-84 l I i rEn.Y NUCLEAR POWEn PLANT Sctvang The Best location on the Nation 20 UOx 97 e PE 8t n v OHIO 44081 e TELEPHONE #2t61 259 3737 e AOr)nt g3. 'O Ct Nf tn 840 A D C

Observation 4L n i Record I flillililllllillilitellililill t Cbservation No. PS-00-01 Russion No. O Checklist No. pS-01. PS-02. PS-03 General sheet 3 of 6 orle nsted my 46 a Date 1/3/gg D. i. t %

c. i..ed av G g (h 0

1.0 Description I The following items either lack documentation or utilize inconsistent data: e Main Steam Safety Relief System a. Support * -1821-H061 Drawing S-322-605, Sht. 061.2, Rev. B. Location plan dimensions require revision per ECN 9152-44-1111. The dimensions shown on drawing do not incorporate all of the specified changes. j } b. Support MK-BP1-H062 Drawing S-322-605, Sht. 062.2. Rev. E. A 10" x 10" x 1/2" base plate was utilized in the design. This was not properly specified on the drawing. The 1/4" all around fillet weld to the embedment plate was not specified. c. Support MK-1821-H063 The design calculation and verification calculation (pg.1.9 thru l 1.15) were based on Rev. C of the drawings, whereas the current drawing revision is "D". Effects from support 1G61-H033 are not evaluated (Ref. ECN 9627-44-1291). d. Support MK-1B21-H064 4 RAP No. OR-SV-231 has not been incorporated in the drawing (Dwg. S-322-605, Sht. 064.2, Rev. C). e. Support MK-1821-H066 4 The support rear bracket angle used in the design (Dwg. Rev. C) does not match the angle calculated from the dimensions shown on the drawings (Rev. D). Horizontal angle (48") shown on Sht. 066.1 of the drawing is in conflict with the angle computed using dimensions shown on Sht. 066.2 of the drawing (S-322-605, tev. D). The design loads shown on the drawing for the emergency condition i (+18400 'bs -19500 lbs) are incorrect. The correct design loads qQ are: upset = 118400 lbs and emergency " t19500 lbs. l Cleveland tiectric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation L4 L n i Record 'llllllllllllll!!"H8!!'H!!! Observation No. PS-00-01 Revision No. O Checklist NJ. PS-01, PS-02, PS-03 General sheet 2 of 6 Date l/3/g Originated By f,h Date l_3 _K Cowlewed By C ' Q, (g o f. Support MK-1821-H067 The restraint direction used in the design calculation differs by about 9" from the direction calculated based on the current support configuration (Dwg. S-322-605, Sht. 067.2, Rev. D). g. Support MK-1821-H068 The verification calculation references Rev. "0" of Dwg. S-322-605, Sht. 068.1 and 2. Letters A, B, C and D were actually used for revision number (pg. 1.34). The snubber size (catalog number P/N 1801172) and the pin-to-pin dimension shown on the drawing do not match the size specified in the design verification calculation (pg. 1.37). h. Support MK-1821-H112 The snubber size (catalog number P/N 1801172) and the pin-to-pin dimension shown on the drawing do not match the size specified in the design verification calculation (pg. 1.45). Nates on Sht.1.40 of verification calculation refer to Shts. 3 and 4 for sketches. The sheet numbers are incorrect and should be Shts. 1.41 and 1.42. 1. No calculation was provided for Support MK-1821-H436.

j. For the Main Steam Safety Relief System (1821-G08) pipe support design, the assumption that no jet impingment load was acting on the supports requires verification. No such verification was provided in the design calculation.

High Pressure Core Spray System k. General Design verification record pg.1.1,1.2 and 1.3 is not properly filled out. Specifically, the pertinent items are not checked off. O Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation dLni Record lll""'"""analtlilillil Rnision No. O Observation No. PS-00-01 sheet 6 Checklist No. PS-01. PS-02. PS-03 General 3 Date s/.3/g g Criginated my f3 [g o.i. t _g % c.vi.. 4 sy C. R Way 0 1. Support MK-1E22-H004 The dimensions of Item "L" on Dwg. Sht. 2 do not match the dimensions shown on Sht. 3, Detail "L" in the design calculation. The restraint direction shown on Dwg. S-322-701, Sht. 1, Rev. D, is incorrect. c m. General There is insufficient information on the design verification sheet. The supporting documents section references " latest analysis." l e Main Steam Drain System n. Support MK-1N22-H017R The elevation shown on the isometric differs from the support O drawing. There is a total elevation difference of 2.04 feet which considerably exceeds the standard criteria of one pipe diameter. o. Support MK-1N22-H018 A 45" bracing member was used in design calculation (P.10.31)}. brace was specified in drawing (S-322-12f, Rev. A whereas a 30 Also, the plan view of Items "E" and "F" is not consister.t with Section A-A on Sht. 018.3 of the drawing. p. Supports MK-1N22-H126; -H127; -H128; -H129; -H130 and -H131 In each of the calculations, an LCD sheet for a special piping clamp (Power Piping Co.) was included, i>ut was not referenced or used in the calculation. Furthermore, the clamps specified in the corresponding support drawings are BE-419N series (National Valve and Manufacturing j Co). Clarification of tha purpose of the LCD sheets is required. l i L O Cleveland Electric 111 aninating; 83102 Perry Nuclear Power Plant Piping Design Review l

Observation ALus Record Hlllllliiiiiiiiiiiiiiiiiiiiii Revision No. 0 Cbservation No. PS-00-01 sheet o' Checklist No. pS-01. PS-02. PS-03 General 4 6 Crisinated my 4, f y / p, gg(y Date 3/.3,fg g Date \\-3-%4 c.vi...e my C.s. M owev I q. Many of the supports of this MSD system have revised or changed support stiffnesses. (Examples are H016, H017,H018 H130, H132...etc.) The aggregate effect of these changes have not been confirmed by analysis. r. Deleted. s.

Deleted, t.

Support No. H003 Calculation does not show the detailed design of snubber and attach-ment. Cold setting and offset are not shown on the support drawing. i u. Support No. H004 Calculation is not shown for the support attachment. I v. Support No. H007 There is no calculation of stiffness presented. w. Support No. H014 Design calculation gives loads for X and Y directions. X-direction load is for Support H014 Y direction load is for support H148, t x. Support No. H148 I A separate calculation is not provided for this support or its connec-1 tion. Only snubber sizing is done as a partial calculation on support l H014 calculation sheet. l l 2.0 Requirement Standard practice and proper documentation. 3 \\ O Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review i

'3-a = + 3 o 4.. Obscrvation 0 /V Record simummme 'o Revlolon No. O Cheervation No, p3-00-01 sheet 5 e' 6 Cheepet No.' PS-01, PS-02, PS-03 General Date I/.3,/gG oryr.ated my 6, ;gPm Dete I-3-b4-Ceviewed av /C K, W(g 3.0 Reference'lbcuments i e Mair Steam '4M af System 3.1 GAI Sup;et Design Calculatien 1821G08(B), Rev. 0 (a thru j) I 3.h DrawhgS322-605,Sht.061.2,Rev.8(a) '3.3 ECN 9152-44-1111 (a) 3.4 Drawing 5-322-605, Sht.' 062.2, Rev. E (b) 3.5' ECN %27-44-1291 (c) ( / 3.6 RAPNo.OR-SV-231l(d) p 3.7 Drawings-322-60V,Sht.064.2,Rev.C(d) 3.8 Drawing S-322-605', Sht. 066.1, Rev. D (e) 3.9 Drawing 5-322-605, Sht. 066.2, Rev. D (e) l 3.10 Drawing S-322-605, Snt. 067.2, Rev. D (f) 3.11 Drawing.5-322-605, Shts. 068.1 and 068.2, Rev. D (g) ( High Pressure Core Spray System e 3.12 GAI Support Design Calculation IE22-G04(B), Rev. 1 (k thru m) 3.1,3 Drawing S-322-701, Shts. 2 and 3. Rev. D (1) Mat'n Steam Drain System e. 3'.14 GAI. Support Design Calculation 1N22-G01(B), Rev. 1 (n thru x) j 3.15 GAI Load Capacitar Data Sheets of Class 1 Component Supports, P-2010, Rev. O. (n thru x) 3.16 GAI program M093 Rev. 1. Load Combination Computer output, J71 A i j,; (dated 5/10/83) for N22G01 (n thru x) 7 3.17 Power Piping Co., Pipe Hanger Catalog and Load Capacity Data Sheets (n thru x)

l l(

3.18 Pacific Scientific Co., Mechanical Arrestor Catalog and Load CapacityDataSheets(nthgx) l y \\. t. L h e. Cleveland Electric 111uminatingt 831 N iF Perry Nuclear Power Plant Piping Design Review 3 f k4 y

4 Observation = L41 &'fd Record lliiiiiiiiiiiiiiilllllllllill! Observation No. PS-00-01 Revision No. O Checklist No. PS-01, PS-02, PS-03 General sheet 6

• f 6

Date gh/g4 Criginated By f,[g Date \\-344 n.wi...d my

6. R

@ ms -~ 0 Capacity Data Sheets (n thru x)g Co., Bas c Engineering Load 3.19 National Valve and Manufacturin 3.20 Drawing S-322-121. Sht. 018.3, Rev. A (o) .g 4.0 Potential Design Impact 1. Individually these itens have no significant impact on design based upon: ei ' A spot check of the above listed items. I, ' b The design margin used in the Perry Project. e k* E. The cumulative effect of the noted documentation problems could lead to a design deficiency. 5.0 Probable Cause Design control. Attachments f) A. Observation Record Review \\ -, ll t, ! u i . Cleveland Electric Illuminating; 83102 ~ Perry Nuclear. Power Plant Piping Design Review .. - ~.

1 h Observation [.4M' M Record Review O * * * "" * " ** *

• Attachment A Cbservation No.

p3 00-01 CheckIlst No. PS-01, 02, 03 Revision No. O PFR No. Sheet 1 of 4 Yes No Closed X Extent All 3 Systems Comments Further review and discussions with GAI reveal the following: a. The referenced ECN was written 7/30/82. In this change, an interference was noted which required relocation of the P.A. by 2" north and 4" east. This was incorporated in drawing Rev. B issued 8/31/82. The change block should have noted this. b. ECN 10130-44-1485, Rev. A, issued 9/19/83 deleted the baseplate from the design. Rev. F of the support drawing notes this but is not issued pending incorporation of ECNs after Phase II inspection. Elimination of the baseplate and welding directly to the embedded plate did not require back-up ( calculations. c. Back-up calculations for Rev. D of the design, which include the effects of 1G61-H033, are contained in the "pending revision" book for subsystem 1821-GOS(3), Rev. 1. d. ECN 9781-44-1341, Rev. A, issued 7/26/83 against Rev. C of the support drawing makes the necessary changes. e. Based on the dimensions shown on Drawing S-322-605, Sht. 066.2, Rev. D, and taking into consideration the length of the rear bracket, the computed angle is 36.1". This closely matches the 35.6" angle used in the stress analysis. Thus, only the coordinate system shown on pg.1 of the drawing would require l revision to be correct. Per GAI, this will be corrected in their upcoming cosmetic update program prior to fuel load. Per GAI, load sumary sheets are not updated for a revised analysis if no hardware changes are necessary due to the revised loads. Their current program provides for updating miscellaneous items on the support cover sheet (cosmetic revisions) after Phase 11 tagging by field QA. This will occur prior to fuel load. l Approvals Originator h, ( ]g, ( Q""[- g Date Project Engineer d1 Date ) %-l[g4 Project Manager ((f ff.7/g Data ggg CEI Representative g L Cleveland Elec /tric Illuminating; 83102 Perry Nuclear Power Plant Piping Desigr. Revicw

Observation [m

M Record Review mimmmmmmmmii Attachment A Observation No-PS-00-01 Checklist No. PS-01, 02, 03 Revision No.

0 PFR No. Sheet 2 CI 4 Yes No Closed X Extent All 3 Systems Comments f. Calculations supporting Rev. D of the drawing are contained in pending Rev. I file to IB21-G08(B), Rev. 1. This was in response to RI #477. g. This is a minor documentation error. Sheet ib of iii gives the correct drawing revision number for H068.

g. & The snubber size and pin-to-pin dimension was changed on the Power Piping h.

drawing when they re-detailed the sheet. In accordance with GAI Fabrication Specification SP-527, fabrication drawings are submitted to the engineer (GAI) for approval prior to use for fabrication. O h. The incorrect sheet number reference is a minor documentation error which was overlooked when renumbering the sheets. 1. Calculation is contained in pending revision file for 1B21-G08B, Rev.1. j. Jet impingement work is still in progress. Per GAI, upon completion of this work, the assumption will be removed. k. Page 1.1 is a superceded form. The current Design Control Procedure (DCP) utilizes GAI form 468 which is contained in the referenced package. Per GAI, pages 1.2 and 1.3 are not an official part of the design control program. l 1. This piece was changed per RI #865 from PPC. When revising the GAI drawing, l the bill of material was changed but not detail "L". The correct dimensions are shown on the PPC drawing. Regarding the restraint direction, the support location plan is correct and consistent with the analysis. The discrepancy exists in the cartesian coordinate system sketch on the support cover sheet. Approvals l -2~"[- % Originator (.-[ C G W 3,, Date D at t. ] [Q,'y [Q Project Engineer f __ { d l[g7[g, ject Manager g Date i/5/sv CEi R.pr. ni.tw. pg h. o = Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

m a s-M e-5 11-- Observation L4M' fd Record Review O * ***** * *' Attachment A Observation No. PS-00-01 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 3 of 4 Yes No Closed X Extent All 3 Systems Comments m. Although the design verification sheet does not reference any specific analy-sis revision in the supporting documentation section, Section 7 of the package provides all the analysis data used as reference or supporting documents. n. Per GAI, the piping was re-routed and the analyst considered a new support location but the relocation was not picked up. This has since been corrected. Support relocations of this nature would have been picked up by the as-built program. o. ECN 9631-44-1294, Rev. A, shows the proper orientation of the brace. Per GAI, the ECN is the governing document and the calculation will be updated to incorporate any specified changes prior to fuel load. p. Per GAI, there was a transition period during which the PPC clamp was replacing the equivalent clamp from National Valve. In accordance with GAI Fabrication Specification SP-527, these changes are submitted to the engineer (GAI) for approval prior to fabrication. q. Per GAI, there is a design loop to confirm final stiffness of the design with that in the analysis. This will also be accomplished when as-built dimensions are confirmed. r. Deleted. s. Deleted. t. The designer referenced the snubber size required and this was verified. LCD sheets provide the capacities. No offset was intended and a lack of cold set would require PPC to set the snubber at mid-stroke. This would accommodate the movement of 0.16". 1 1 \\ l Approvala l 2_T-4 Originator C.k (dh, Date g f2,g % Project Engineer Md Date /hM4 Projec.t Mrmager [M' Date gy[p CEI Represbatative gg# Date Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

===- Observation [41d;M Record Review mmmmmmumunm Attachment A l Observation No. 95-00-01 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 4 of 4 Yes No Closed X Extent All 3 Systems Comments Per GAI, on the previous calculation, 3" of the 1/4" weld was determined by u. inspection to be adequate for a load of 1,300 lbs. For Rev. A of the support, the load decreased to 1,000 lbs. The existing weld was again determined adequate by inspection. Cygna agrees with this assessment. v. Per GAI, calculation book 1N22-G01(B), Rev. 2, contains .e stiffness calculations for this support. Rev. 2 of this calculation was not in the Cygna review scope.

w. & The orf ginally specified x and y restraint was designed as two individual x.

support marks, H014 and H148. This was requested per RAP #6701. O As stated in Section 4.0, individually these items do not have impact upon design. In addition, based on the explanations above, Cygna does not consider the cumulative effect of these items to be a potential problem for the three systems reviewed. Approvals ( 1~) -h j Originator

k. N. N@

Date Project Engineer { lfghSf Date l ) Project Manager h~ /[77 /g, Date Mggg CEI Representative Date e Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

CEI Observation Record Closure .O_ Attachment B ' I bservation No. Checklist No. Revision No. O -PS-00-01 PS-01, 02, 03 0 EDDR No. QAD 600 No. Sheet of 67 N/A 1 2 Yes No Closed X isolated X Pctential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The items cited are examples of incomplete details in an in-process design, for which l explanatory information was available in other files in most cases. None of the-items were found to compromise design adequacy. To facilitate future reviews, GA t is preparing a. summary procedure (road map) clarifying availability of additional information augmenting documentation in support design changes. q.0 Action Taken ,, ) gal Engineering has initiated a series of programs to ensure completeness and availability of the documentation: 1. A data base is being maintained which lists all change documents and PPC shop sheets against a particular support. A copy of each of these change documents and shop sheets will be included in each support calculation package. '2. An effort is underway to close out all items identified as requiring later confirmation. 3. A snubber and spring' can " cold set form" will be completed for every snubber and spring can, and issued to the field on a subsystem basis. This form will confirm the snubber /sprin's size, confirm the movements, and assure that the proper settings are issued for construction. ~4. 'A ' Generic' Jet calculation (File Code X176) is being finalized to include a matrix of all Jets, shields, and piping or supports impacted. It will document that all jets are accounted for, that piping impacted is properly analyzed, and that pipe supports impacted are either shielded or designated to withstand the impact. 5. ~ Pipe inspection reports (submitted at 75% 'and 100Y, Construction) will be l . reviewed by gal Engineering to ensure that the as-built configuration at the piping system and the support type, direction, and location match the as-analyzed condition. Any discrepancies will be resolved in a letter of. l reconciliation. _._._,,-.a. .L.

l CEI Observation rw Record Closure b) Attachment B ~ Observation No. Checklist No. Revision No. PS-00-01 PS-01, 02, 03 0 EDDR No. QAD 600 No. Sheet of ~ 67 N/A 2 2 4.0 Conclusion The items cited by CYGNA are addressed by the five action above as follows: 1. Procedure clarifying availability of supporting documentation: Items g, k, and m. 2. Data Base: Items e, b, c, i, I, o, u, y, w, and x. 3. Closeout of items requiring later confirmation: Items o and p. 4. Cold Set Program: Items h and t. 5. Generic Jet File: Item J. I 6. As-built. review: Items e, f, n and q. 7. Not addressed (deleted by CYGNA): ltems r and s. .0 References 1 (20) PPM Appendix AA - General Procedure for IE Bulletin 79-14. l (4) EDDR 67 1 Approvals Originator Date Senior Prof' et Er3gineer / Date - 7 M-NW 9f8 % CEI Suparvisor Quali Au Unit Date 1, ' ' _d f/Ty gal Project Manper w, m,_ Date i gal Manager Corpor A Date .Tha Cleveland Electric illurdinating Company: / Pcrry Nuclear Power Plant Piping Design Revidw ' gW138/Q/31/es i D e

u t= = = = = a m Observation [4M'M Record 111ll11111ll111111111111111111 R "I*' a N - Observation No. PS-00-02 n sheet of Checkilet No. PS-01, PS-02, PS-03 General 3 9 l-3-h Date Criginated By b.K MCy { f3 /gd neviewed sy QC _ h - Date M d 1.0 Description The following items are not consistent with design commitments, requirements or criteria: a. The GAI method for combining dynamic inertial loads and dynamic displacement loads differs from the General Electric specification. The difference is shown below: g )1/2 + (OBE 2 + SRVD) 2 + SRV GAI method: (0 bey D General Electric method: [(0 beg + OBE ) + (SRV; + SRV )* D D where OBE = Operating Basis Earthquake SRV = Safety Relief Valve P I = Inertial Load D = Displacement Load b. GAI Design Specifications B21 and E22 do not include Faulted Load Case No. 8 as specified in Table 3.9-21 of the PNPP FSAR. 2.0. Requirement a. General Electric Design Specifications 22A5454, Rev.1 and 22A6547, Rev. O. b. PNPP FSAR, Amendment No. 3, dated 9/11/81, Table 3.9-21. l 3.0 Reference Documents ' ~ ~ ~ - ~ ~~' l l 3.1 General Electric Specification for ECCS Piping Systems No. 22A6547 Rev. 0 (Table 5, Sht. No. 21) (a) 3.2 General Electric Specification for Main Steam Piping l No. 22A5454 Rev. 1 (Table 8, Sbt. No. 28) (a) i 3.3 GAI Support Design Calculations for HPCS Calculation E22G04B (a) 3.4 Computer Load Combination Output E22G04C (4/18/83) (a) j p3 3.5 Program M093 LOC 1 (a) V Cleveland Electric illuminating; B3102 -Perry Nuclear Power Plant Piping Design Review

Observation Record Al t i lillllllllllllllllill!!lllllli Revision No. 0 Observation No. PS-00-02 Sheet p of g Checklist No. PS-01, PS-02, PS-03 General t-3 44 Criginated By C, g, @gy cate l[3[gh Keviewed By k( Date M Q \\ 3.6 Load Capacity Data Sheets of Class 1 Component Support P-2001, Rev. 0 (a) 3.7 GAI Design Specification DSP-B21-1-4549-00, Rev. I and 2 (b) 3.8 GAI Design Specification DSP-E22-1-4549-00, Rev. I and 2 (b) 3.9 GAI Support Design Calculation 1E22-G04(B), Rev.1 (b) 3.10 GAI Support Design Calculation 1821-G08(B), Rev. 0 (b) 3.11 GAI Support Design Calculation IN22-G01(B), Rev.1 (b) 4.0 Potential Design Impact a. By inspection, the GAI method for combining loads is more conservative than the General Electric recomended approach. This conclusion is supported by the following sensitivity calculations: O GAI GENERAL ELECTRIC g SRV; OBE SRV COMBINATION COMBINATION % DIFFERENCE CASE OBE D D 1 100 100 100 100 283 283 0 2 100 1 100 1 200 200 0 3 100 1 1 100 200 143 -40 4 100 100 1 1 143 143 0 5 4397 390 2313 5478 10361 8914 -16 Where Case 5 is an actual loadir.g case for Support 1E22-H005. Consequently, the GAI method is conservative and may be up to 40% conservative. b. More severe design loads may result due to the excluded load combination. 5.0 Probable Cause Standard GAI practice. Attachments A. Observation Record Review q v Cleveland Electric Illuminating; 83102 Perry Nuclear' Power Plant Piping Design Review

e Observation L41 &'fd Record Review mmmimilimilmimii Attachment A Observation No. PS-00-02 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 1 of 1 Yes No Closed X Extent All 3 Systems Comments Further review indicates the following: a. As stated in Section 4.0, the GAI method for combining inertial' loads and dynamic displacement loads is conservative. l b. Consideration of FSAR Load Case No. 8, for the three systems reviewed, does not i result in any significant increase in support design loads. Based on the above, this Observation does not have any impact on design or safety. Approvals g_g-% Originator C K, Qg9 Date Project Engineer QQJ, '_ j[g[g Date 'f[g/j4 7 [jjj d Project Manager Q Date [fj/gg CEI Representative g gf Date Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

. r-m__. m _ _. m _a., _ ('El Observation 's/ Record Closure Attachment B cbservation No. Checklist No. Revision No. PS-00-02 PS-01 PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of OG8 N/A 1 l' Yes No CI: sed X is: lated Not Applicable Pct ntial Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion item A is conservative in all cases and item B is controlled by other load cases and allowables for all systems. 3.0 Action Taken The items were reviewed under EDDR 68 for any generic effect. .0 Conclusion Their is no effect on plant safety from these items. 5.0 References (5) EDDR 68 Approvals Originator Date g p S nior Pro t Engineer Date ftol6lO CEI Supervisor ~M. k& b ualit Audit Unit Date W/a / 84 W 'M gal Project Manager $ ^^ + ~ Date 5//O/ M v d)Ral%2H"M*1m L ~ 'hoht Efi5 Clevela%l@rg/ Illuminating Compan'y: Pcrry Nuclear PoWr Plant Piping Design Review DW138/Q/38/mm L ~, -, - -

Observation [. 9 M ' f d Record lillllilllilllll!!Illilillllli C Revlalon No. O Observation No. PS-00-03 Checklist No. PS-01, PS-02, PS General Sheet 1 of } l-3 % Date Criginated By C g Qgy j h /gd Reviewed By {h, [ 0- p Date C Q 1.0 Description The signs of Jet Impingement load ingut for support load combinations in utilizing the computer program "M093 were not properly considered (e.g., HPCS, E22G04(C), Run No. J484, dated 4/18/83, the dynamic Jet Impingement input loads are all positive). 2.0 Requirement 1. GAI Design Specifications DSP-B21-1-4549, Rev. I and 2 and DSP-E22-1-4549-00, Rev. 1 and 2. 2. Perry FSAR Amendement No. 3, dated 9/11/83. 3.0 Reference Documents 3.1 GAI Support design calculation 1E22-G04(B), Rev. 1. 3.2 GAI Support design calculation IB21-G08(B), Rev. O. 4.0 Potential Design Impact Incorrect signs will give incorrect design load combinations and may lead to underdesign of some supports. 5.0 Probable Cause Design oversight. Attachments A. Observation Record Review NOTE: Jet Impingement load is not applicable to the Main Steam Drain Line, IN22-G01, per GAI memo from D. H. Hunt to J. Chang, dated 9/27/83. O Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Record Review Al t i mninmununiminmi Attachment A Observ.tlon No. PS-00-03 Checklist No. p3 01, 02, 03 Revision No. O sheet 1 of 1 PFR No. 01 Yes No Closed X Extent 2 of 2 Systems with jet loading Comments Standard GAI practice for input to the "M093" combination program is to use the same sign for the support loads as that found in the TPIPE output. In general, it is critical that the signs are properly input, however, any inaccuracies in sign input drd of minor consequence for the HPCS system due to the small magnitude of the weight loads. In addition, during the course of performing further review to explain inconsistencies between input loads and output combination values for the MSRV system, GAI has discovered a bug in the "M093" program. The problem occurs when j considering the negative jet impingement loads in the emergeacy load combinations. A value of zero is always used in this situation due to taking the maximum (instead of the minimum) between the negative load and zero. This could result in situations where support stresses exceed Code allowables due to the loads being underestimated. Due to the potential design and safety impact associated with.this problem, a PFR has been written. I i Approv.ls 1-R-80 orie.. G y W m o ='* ili8/s. proi.ci so r 4td. ? ~ 1 oat

• O ro-i e.,

Mia4 o rmlym r //ro/Bf Cai R.pr . w. M pj M' o = '* y y v r Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

p gp% = z -m >l[ El Observation Record Closure Attachment B E. Obcervation No. Checklist No. Revision No. PS-00-03 PS-01, PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of N/A GC PRE-084 1 1 Yes No Clused X is: lated X (Yes with respect to programming error) Patential Design impact X 4 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required 3.0 Action Taken Evaluate pipe supports which could have been affected by this program error, and correct program. .0 Conclusion The programming error did not affect the design of any of the 100 supports sampled. From this it was concluded that a substantial safety hazard was not created on the

l Perry Project.

5.0 References (12) GC Pre-084 (24) gal memo dated January 16, 1984 from J. B. Muldoon to F. L. Moreadith Approvals - Originator g Date S nior Project 4*.g2n x. - 4 s /s is+ i r 7 Date .CEI Supervisor Qual ty Au ' Unit Date W-Y/o E4 4" gal Project Manager Date gal Manager Corpora F Date ~Tha Cleveland Electric liluminating Company: brry Nuclear Power Plant Piping Design Review / DW138/1/Q/sp . =

Observation d L% i Record 111111111llllll111111111111111 Revision No. O Observation No. PS-00-05 Sheet of Checklist No. PS-01, PS-02, PS-03 General } 4 Date 1/3 /f4 Crleinsted By ((y Date [ ~3 4 4 R: viewed By G, k, ( d hy, 0 1.0 Description The following design oversights were noted: e Main Steam Safety Relief System a. Support MK-1821-H062 The wrong eye nut allowable load was used. (Pg. 1.5)' b. Support MK-1821-H163 The design is based on the calculation for support IB21-H179 (Rev. 0) with enveloped design loads. O b.1 The calculation and assumptions shown on Pg. 10.30 are not applicable since they do not represent the actual condition of the support. b.2 Allowable stress used is 1.2 S - Sh was mistakenly stated as Sy h (Pg. 10.31). b.3 The width of the ring is 5-1/2", but 12" was used in the calcula-tion. Consequently the section properties were incorrect (Dwg. S-322-605, Sht. 163.2, Rev. E). b.4 Penetration sleeve was specified as schedule 40. It should bt: schedule 30 based on the thickness of 0.375" (Pg. 10.32). b.5 The thickness of the ring is 0.875", but 1.1875" was used in computing the section modulus. (Pg. 10.36) GAI is currently redesigning this support due to the overstress caused by this item and item b.3. b.6 The Lug size Li used ir the computer analysis did not match the actual size of the Lug. b.7 The design was based on a very simplified analysis. There are other load conditions which were not considered (e.g., friction loads etc.) A more detailed analysis model is recommended to p/ reduce the stress level and obtain more accurate results. s. Cleveland Electric Illuminating; 83102 J Perry Nuclear Power Plant Piping Design Review

Observation L4M'M Record llllllilllllilllillllllillllll Revision No. O Observation No. pS-00-05 sheet 7 of 4 Checklist No. PS-01, PS-02. PS-03 General %/g & Cricinated By 4{ Date ( __3 % Reviewed By C,k (yg Date 4 High Pressure Core Spray System e c. Support MK-1E22-H003 k The cold load (10.87 ) used by the verifier (Pg.1.5) was based on an incorrect calculation in Section 11 (Pg. 5). The correct cold load calculation procedure in Section 10 (Pg. 10.29) should be used to update the loads and to perform verification. d. Support MK-1E22-H004 The property of a solid circular section instead of a hollow tube section was used in the stiffness calculation (Pg. 7.2). e. Support MK-1E22-H006 k k L = 25.9 ) were TheproperloadingsfromH005(Pg.1.9;F{=13.9,FSectionlifSupport not used in the design calculation (Pg. 5 H006). The support frame weight was not included in the design. Main Steam Drain System f. Support MK-1N22-H017 The moment arm used in checking the existing W12x40 should be calculated as the distance from the point of load application to the center of W12x40 beam. The distance used in the calculation was measured only to the top of the flange. l l g. Deleted. I h. For most of the supports (MK-1N22-H126; -H127; -H128; -H129, etc.), Youngs Modulus was not adjusted for temperature effects. 1. Deleted. Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation Alni Record n ll111lll!!I1111111111111111111 U Observation No. PS-00-05 Revision No. O checklist No. PS-01, PS-02, PS-03 General sheet 3 of 4 Criginated By 4 [w [ k /$ db O I* A ' ~ ~ ~ Reviewed By C. K ( UCMU 4

j. Deleted.

k. The following supports do not meet the GAI stiffness criteria: 1. H004 5. H011 2. H006 6. H012 3. H008 7. H014 4. H009 8. H015 1. Deleted. 2.0 Requirement 2.1 Standard Practice -Q 2.2 ME B&PV Code Section III, 1974 with Addenda to Winter, 1975, Subsection 2.3 GAI Design Specification DSP-B21-1-4549-00, Rev. I and 2 (MSRV and MSD) 2.4 GAI Design Specification DSP-E22-1-4549-00, Rev.1 and 2 (HPCS) 3.0 Reference Documents e Main Steam Relief System 3.1 GAI Support Design Calculation 1821G08(B), Rev. 0 (a thru b) l 3.2 Drawing S-322-605, Sht.163.2, Rev. E (b) High Pressure Core Spray System e l 3.3 GAI Support Design Calculation IE22-G04(B), Rev. 1 (c thru e) 3.4 ECN-8857-44-1004, Rev. C (e) I Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation M%i Record liiiiiiiiiiiiiiiilHililllill Observation No. PS-00-05 Revision No. O sheet 4 of 4 Checklist No. PS-01, PS-02, PS-03 General Date %fy y Criginated By f, [w l Rowlewed By C. K ( yCwf. Date {-3_R4 i e Main Steam Drain System 3.5 GAI Support Design Calculation 1N22-G01(B), Rev. 1 (f thru 1) 3.6 GAI Load Capacity Data Sheets of Class 1 Component Supports, P-2010, Rev. 0 (f thru 1) I 3.7 GAI program M093, Rev. 1, Load Combination Computer. output, J71 A l (dated 5/10/83) for N22G01 (f thru 1) i e General - All Systems i l 3.8 Power Piping Co., Pipe Hanger Catalog and Load Capacity Data Sheets. 3.9 Pacific Sciantific Co., Mechanical Arrestor Catalog and Load O canec$ts at sheets-o 3.10 National Valve and Manufacturing Co., Basic Engineering Load Capacity Data Sheets. 4.0 Potential Design Impact 1. Individually these items have no significant impact on design based upon: i e A spot check of the above listed items. o The design margin used in the Perry Project. 2. The cumulative effect of the noted oversights could lead to a design _ deficiency. i 5.0 Probable Cause l Design control. l Attachments A. Observation Record Review O Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

T-Observation [4M' fd Record Review ammmumimumn Attachment A PS-00-05 Checklist No. PS-01, 02, 03 Revision No. O 0bservation No. PFR No. Sheet } of 2 Yes No Closed )( Extent All 3 Systems Comments Further review and discussions with GAI reveal the following: a. The reference calculation compared the applied load of 4312 lbs to 8900 lbs. the actual allowable should have been 8000 lbs. b. Per GAI, the original configuration of this support was 12" long with a thickness of 1.1875" and no lugs. Due to constructability concerns, this was revised to the 5-1/2" configuration with lugs. Welded attachment calculations are contained in the A Calculation, which in this case references Calculation P-584, a finite element analysis of the configuration shown on Rev. E of the support drawing. Cygna has not reviewed this analysis due to O GAI's detailed attention to this support. c. For this subsystem, the line is normally cold in the operating mode but the vessel is hot. This creates the maximum differential condition. One analysis (thermal case with hot vessel and hot line) showed a 0.8" displacement at the spring. A second analysis (hot vessel and cold line case) showed a 1.5" displacement at the spring. The true condition during normal operation is somewhere in between. The corresponding spring cold setting for this more realistic condition should be 10.8K + (0.8 + ) x spring constant = 13.1K. 2 Since spring settings are verified as part of GAI's Phase III program, the 11.8K setting on Rev. C of the drawing does nnt create a safety concern. d. The calculation reviewad was a preliminary calculation used initially for estimating. Per GAI, the noted discrepancy was picked up by the designer when i reviewing final stiffnesses with the analyst in a later revision of the calculation. l Approvals om b r7-M c.x. wow o, e,onou e, Liv A _ m iIn/e4 om O ,0>.ot , e., %mLL iai/s o '- Cen.,,. p p ;p m Jg, d/5/s/ om ClevelandElectricIlluminating$83102 Perry Nuclear Power Plant Piping Design Review

..~ Observation [4h' TJ Record Review 11111111111111111111111llI111 g Attachment A V Observation No. PS-00-05 Checklist No. p3 01, 02, 03 Revision No. O PFR No. Sheet 2 of 2 Yes No Closed X Extent All 3 Systems Comments E K e. The loads of 13.9 and 25.9 are 150% of the actual loads. Per GAI this was done to provide margin in long lead time hardware at the time supports were designed. This would allow for a substantial variation in load when spring stiffness was included in the later analysis. The actual loadings were used for the design of the support structural steel (consideration of frame weight is addressed in PS-00-06). f. Per GAI, the W12X40 is a structural nember checked in the load confirmation effort by structural engineers. g. Deleted. h. Per Table I-6.0 of ASME Subsection NA, Young's modulus varies with temperature frcm 27.9 ksi at ambient to 27.3 ksi at 330", which is the accident temperature inside drywell. Since this property is only used for the calculation of support deflection and support stiffness, there is potentially a 2% maximum variation in calculated values. This would have a negligible impact on design. 1. Deleted.

j. Deleted.

k. Per GAI, their stiffness criteria was a guideline established for Class I work to aid designers in new designs and minimize iterative cycles between analysis and design. Final stiffnesses are included in the "C" calculation and have been addressed by the analyst. I 1. Deleted. As stated in Section 4.0, individually these items do not have impact upon design. In addition, based on the explanations above, Cygna does not consider the cumulative g((g of these items to be a potential problem for the three systems reviewed. l 1-r1-84 Orwaior cy mm o=t* Project Engineer O. *} Date

[2.*7 / k

1. ((4 7

Project Manager Date l ffgg g CEI Representative gg Date Cleveland Elec(ric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

CEl_ Observation Record Closure Attachment B Observation No. Checklist No. Revision No. PS-00-05 PS-01, PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of 67 N/A 1 2 Yes No Closed X isclated - X Pctential Design impact X 1.0 Description I See Cygna observation record and observation record review. 2.0 Discussion The majority of these observations resulted because the gal packages were stiil being developed at the time of the review. It is-difficult to freeze design documentation for a review of this type. A " Road Map" is being prepared to assure that all necessary documentation is identified to facilitate future reviews, h.0 Action Taken 1. A data base is being maintained which lists all change documents and PPC shop sheets against a particular support. A copy of each of these change documents and shop sheets will be included in each support. calculation package. 2. A snubber and spring can " cold set form" will be completed for every snubber' and spring can, and issued to the_ field on a subsystem basis. This form will confirm the snubber / spring size, confirm the movements, and assure that the proper settings are issued for construction. 4.0 Conclusion ~ ~ ~ All discrepancies noted during the~ Cygna-Yeview will'be addressed by the programs ~ describe,d in Section 3.0. The items cited in this observation have been addressed as follows: d 1. Deleted by CYGNA: Items g,' i, ~ a nd J. -- 2. Data Base: Items b, d, e, f, h, k, l. 3.- Cold Set Program: Item c. ~ ~ 4. Minor documentation error of no consequence: Item a. G w v* g m- ,y. .y -,g -,--ym,. e .-ww,,yy,.w,-..-e,.y w.e..w.w.,-.e=w--.---.sii.meewi,1- .e.-.-, m--.------

CEI-Observation es-Record Closure h Attachment B Observation No. Checklist No. Revision No. PS-00-05 PS-01, PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of 67 N/A 2 2 5.0 References (4) EDDR 67 (19) Spring and snubber setting forms. 't } O Approvals Originator Data S:nior Pro 4*A* E giAn-d rivl sr r/ Date '4""PX2"'"*YJ"C"i

mar, gal Project,ManagW

.Date ,' e 3//o/PV gal Manager CorporapK Date The Cleveland Electric illumi'nating Company: / Perry Nuclear Power Plant Piping Design Revie6 [ ' DW138/Q/34/es l20 l b e , ~,. ---+-,,,% ..m-.- [--- _ - - - ~, _,,,

l Observation A f.% i Record 11llll!I1111111ll1111111111111 Observation No. PS-00-06 Revision No. O Checklist No. PS-01, PS-02 & PS-03 General sheet of 3 1 Date /3/g44 Crloinated my f, b Reviewed my C.M@g Date l_3 % 0 1.0 Description The design of the supports does not consider the following items: a. Dead weight of the support itself. b. Inertial loads due to support self-weight excitation. 2.0 Requirement Standard industry practice. 3.0 Document Reference 3.1 GAI support design calc. IN22G01 (B), Rev. 1. 3.2 GAI support design calc.1821G08 (B), Rev. O. C 3.3 GAI support design calc. 1E22G04 (B), Rev. 1. 4.0 Potential Design Impact a. This is critical only for frame-type supports which have a small margin with respect to allowables. b. This is most critical in the unrestrained direction for frame-type supports where high accelerations must be considered. 1 In the restrained direction this is only critical when the margin with respect to allowable is small. Note: " Restrained direction" is defined as the line of action of the support. 5.0 Probable Cause GAI standard practice. Attachments A. Observation Record Review O Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Pipinn Design Review

Observation [. 4 M ' f d Record Review imumimmumummi Attachment A 0bservation No. p3 00-06 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 1 of 2 Yes No Closed X Extent All 3 Systems Comments The GAI standard practice is that the consideration of dead weight and inertial loads due to support self-weight excitation is made at the designer's discretion. During the design process, the designer makes a judgment as to whether these factors are critical to the design and integrates them in his calculations as appropriate. GAI has performed an c/aluation of the three systems within the scope of this review to determine the most critical support (s) for the loads of concern. Their determination was that a frame comprised of supports H004 and H009 in subsystem IN22G01(B) was most critical. This judgement was based on the following three factors: 1. The support frame appears to be flexible in the out-of-plane direction. O 2. The frame is attechee at two structerei Points (derweii waii and bio-shieie platform steel) which are highly excited. 3. The support frame is located in containment building where the most severe transient loadings are found. GAI's evaluation was made by analytically determining the natural frequencies in the three orthogonal directions. Once the frequencies were found, the corresponding accelerations were read from the response spectrum curves. The accelerations were applied to the frame mass, resulting in the self-weight inertial loads. GAI then performed a static analysis combining the out-of-plane inertial loads (in two directions) with in-plane piping loads, in-plane inertial loads,.and supp. ort. dead weight. The resulting stresses for loadings in different directions were added direr.tly. This is conservative since it is unlikely that the maximum inertial loadings would occur simul-taneously in three orthogonal directions. Per GAI, the results showed that for this l conservatively combined loading case, the stresses were within code allowables. l Aporovals l Originator C, K. W my Date 1.- ] - @(4 Qygd Project Engineer %Q[ _ [- Date Project Mar:ager Nh Date g,/7/4 Q~8-gq l Cns Representatsve ff f&r Date Cleveland Electric Illuminatingk 83102 Perry Nuclear Power Plant Piping Design Review ~

~... Observation Record Review AL t i inimmmm!!aaa!!!! Attachment A Observation No. PS-00-06 Checklist No. PS 01, 02, 03 Revision No. O PFR No. Sheet 2 of 2 Yes No Closed X Extent All 3 Systems Comments Cygna has not reviewed this analysis, and based upon an independent assessment of the supports for the three systems within the scope of this review, Cygna requested GAI to perform a similar evaluation as that described above for support IN22-H132. The results of this analysis showed that the stress levels are acceptable. However, GAI has decided to install bracing for this support in order to provide additional out-of-plane stability. Based on the above, this Observation does not have any impact on the design or safety of the MSRV, HPCS, or MSD systems. O l l l A,provals Date 2, SQ Orleinetw C. K, OcQCp p,[7/g y Project Engine.r C.' Date ~ Pros.ct wanneer W hWM' hg/g4, Data l Cnl R.or at.m. f/WMe d-h'- M o * ** Cleveland detric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

l CEI Observation p Record Closure ( Attachment B Observation No. Checklist No. Revision No. PS-00-06 PS-01, PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of 069 N/A 1 1 Yes No Clased X is: lated X Patential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion GAI Engineering includes support weight and support self-weight excitation as part of the calculation when they could have a potential design impact. Therefore, this consideration is only noted when necessary in the design. '3.0 Action Taken Self-weight excitation will be addressed as the systems are walked down at 75% and 100% completion. Supports identified by the walkdown as being flexible in the out g of plane direction will be investigated and braced if necessary. 4.0 Conclusion The effects of support weight and support weight excitation are negligible for the majority of supports. Those supports either requiring further calculation to document their adequacy or needing bracing will be identified by the walkdown. 5.0 References (6) EDDR 69 (14) EDDR 139 l Approvals 1 h$WW ~~ Date Originator 5~ 9 - 89 Senior Proje'cAtgineer # Date 7.# Ne G 9 8h CE uperyi_so ty A it Date GAI Project Many' Date , gal Manager Corporap Date '~~ E Cleveland Electric llidminating Company: / _Pcrry Nuclear Power Plant Piping Design RevTew -vW138/Q/32/es D

Observation t4 (.% i Record !!!I ) O l11111111111111lll11111111 Rniston No. O Observation No. PS-00-07 shut i 7 of Checkitet No. pS-01. PS-02 & PS-03 General Crfginated Er 6hw / D. A A t I C A-D*'* V3/ 6 (J-i-3-94 Reviewed By C,M, h p Data U 1.0 Description The following items were noted in relation to the setting for springs and snubbers. Mainstream Relief Valve System a. Deleted. f High Pressure Core Spray System e b. Supports MK-1E22-H003 and MK-1E22-H006 Neither a cold setting calculation nor an indication of the proper normal thermal mode for design was given in the verification O ' '" ' ' " ' ' ' " 9 '"' ' " '" a a d * * * ( d ' '* ' " a "' ' c ) - e Main Steam Drain System c. Deleted. d. Support MK-1N22-H019 Incorrect thermal movement was used in calculating the cold load (See Sht. 019.2 of Dwg. S-322-121, Rev. A). Also the normal thermal mode (THN2) displacement was not used. e. Supports MK-1N22-H008, H131, H127, H013, H011 i Snubber setting was computed in calculation, but was not specified on the drawing. The drawing indicates "N/A" for setting. Per GAI this instructs installer to set the snubber at midstroke. The actual settings should be: H008 2.875" H011 1.25" H127 2.82" H131 Max thermal = 2.0156", H013 0.325" but no setting was calculated. i O I > Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

3 x c. '="- Observation L4M' TJ Record llllllllilllllllllllillllllll! Observation No. PS-00-07 Revision No. O Checkilst No. PS-01, PS-02 & PS-03 General Sheet p of p lh)g4 v Crieinsted av f & l Q. jf A L f C A Date c.vi...a er C. K (Aowq. o=t* 1-3-M i e General f. There,is no indication that bottoming or topping out of springs is checked for combined thermal and dynamic movements. There are no calculations performed combining the displacements due to dynamic s loading. 2.0 Requirement Staridard industry practice. 3.0 pocument Reference 3.1 GAI support \\ design calc. 1H22G01 (B), Rev. 1. O 3.2 GAI support design calc.11521G08 (B),,Rev. O. 3.3 GAI support design calc. 1E22G04 (B), Rev. 1. C.0. Potential Design Impact Improper settings may result in a spring or snubber bottoming or topping out. This would result in the support not performing its intended function. s 5.0 Probable Cause -Minor design / analysis oversights. Attachments A. Observation Record Review I l O Cleveland Electric Illuminating; 83102 l Perry Nuclear Power Plant Piping Design Review

Observation Record Review M i i mmunniiiiiiiiiiiiiiini Attachment A Observation No. PS-00-07 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 1 of 2 Yes No Closed X Extent 2 of 3 Systems Comments Further review and discussions with GAI indicate the following: a. Deleted. b. Appropriate settings are shown on the drawings but not document"ed in the calcula-tions. c. Deleted. d. Per GAI, a value of 0.549" down (f rom a previous analysis) was used versus the current actual value of 0.387". Fo:- the spring rate of 200 lbs/ inch, this would Q change the setting from 390 lbs to 423 lbs. Cygna agrees that this deviation is not sufficient to warrant a drawing revision at this time, pending as-built infor-mation. The settings specified en the drawing bill of material are correct for H011, H013 e. and H127. Per GAI, for H008, the PPC drawing has this snubber set at mid-stroke. GAI has committed to update the drawing to reflect this setting during the upcoming " cosmetic revision" cycle prior ta fuel load. Regarding H131, the therrtel i'csement exceeds the specified mid-stroke setting by 0.015". However, per G?J., 'l settings will be reviewed as part of the as-built program prior to fua! T 'aq. l f Approvale lm-M orwn.ior C M.1 Ocre o = tlu /e+ Pr w en e e.r %A - EA o='- (( 7 /g Date Project yanager ((f f/3p/g( can & resoneetko M Date Cleveland Electric Illuminating; 83102 Perry Nuclear' Power Plant Piping Design Review l

Observation e Record Review 4L t i nminn!!!9ammmn!! Attachment A Observation No. PS-00-07 Checklist No. PS-01, 02, 03 Revision No. O PFR No. Sheet 2 .f 2 Yes No Closed X Extent 2 of 3 Systems Comments f. In general, inertial movements are small compared to thermal movements. Spring cans are selected to achieve a center set as much as possible. Per GAI, travel is then restricted to the recommended load range which permits a minimum 1/2" margin on each end to pre 'nt bottoming out. It is also important to note that it is standard design practice to locate springs either adjacent to equipment or near large concentrated masses where they provide constant dead weight support. GAI states that dynamic displacements of 1/2" do not occur at these locations since they could not be tolerated by the piping or supporting equipment. Based on the above, this Observation does not have any impact on design or safety. O l I Approvals t

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nei. , [v.7 /4 Project EnWneer dh), _ ;,30 oete O ,,o-t vms;Ar wa. o.t. ppg CEI Representative g' oate v Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design keview

CEl Observation p Record Closure v Attachment B Observation No. Checklist No. Revision No. PS-00-07 PS-01, PS-02, PS-03 0 EDDR No. QAD 600 No. Sheet of 070 N/A 1 1 Yes No CI: sed X isclated X P:tential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The correct thermal movement, size and load for all springs / snubbers will be issued before construction. '3.0 Action Taken A snubber and spring can cold set form will be completed for every snubber and spring can, and issued to the field by subsystem. The documentation confirms the snubber / spring size and the thermal movement used, and assures that the proper cold xC set has been issued for construction. 4.0 Con'clusion All spring and snubbers will be at the correct settings prior to hot functional -testing, and the actual movements will be verified during the testing of the systems. 5.0 References ,.u. mu (7) EDDR No. 70 (19) Spring / snubber forms Approvals l E'S'"** r f g 4, f_ <7. g f Senior Proje e Date - "' M' '22"*'19L"'*. Uwrv ~ gal Project Managf /J Date f/Yd//V Y n ~ Date 1 gal Manager CorporpA Pfo The Cleveland Electric illuminating Company: / ~ _ P rry Nuclear Power Plant Piping Design Review / - DW138/Q/35/es. + 4 g

Observation L41*$2fd Record llllillllllllllllllllillllllli R "8 8'*" N *- Observation No. PS-01-01 n sheet of Checkliet No. PS-01 MSRV 3 3 l-3 d$ Criginated By C.{ M Date g/3/ g Reviewed By ' d, [__ Date D C 1.0 Description For the design of Main Steam Safety Relief system pipe supports, there is no indication that the hydro test load is considered in the design. 2.0 Requirement All pertinent loading conditions should be considered. 3.0 Reference Documents GAI support design calculation IB21G08(B), Rev. O. 4.0 Potential Design Impact . O so e suvaarts > be "aeerdesis"ed $< a dro test ioed es #ot co#s4dered-r 5.0 Probable Cause Improper assumption that the discharge line does not require hydro test. Attachments A. Observation Record Review P O l Cleveland Electric Illuminating; 83102 . Perry Nuclear Power Plant Piping Design Review

Observation Record Review Al i i Attachment A limmimmummmmi Observation No. PS-01-01 Checklist No. PS-01 Rulsion No. O PFR No. Sheet 1 of } Yes No ~ Closed X Extent 1 of 1 Steam Systems Comments Further review indicates the following: 'a. The rigid supports for this system are designed for an upset load which is larger than 1.9 x deadweight load (hydro-test load). b. Per GAI, Power Piping Company designs springs and variable supports in accordance with the " Manufacturers Standardization Society" (MSS) Standard Practice SP-58. This practice requires that elements designed for use with hydrostatic test stops be capable of supporting up to two times the normal operating load. i Q c. The structural support steel associated with variable spring support H062 is sufficient to withstand the additional loading due to hydro-test. Cygna has not reviewed the support detail for variable spring support H468 due to the fact that this is a recently added support which was not part of the Rev. O calculation. This support is included in the Rev. I calculation which was not within the scope of this review. j Based on the above, this Observation does not have any impact on design or safety, l l Approvals orwn. r C. K. uh 1-2.4 -M g erecino r t./1rt W oa i/ W 84 Project Meneger h [M/g Date t/3/s/ Cai R.pr. pr #Sa;hc o = '. ~ y E Cleveland Electric Illuminating; 83102 Perry Nuclear Power Plant Piping Design Review

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'.D Record Closure Attachment B Observation No. Checklist No. Revision No. PS-01 -01 PS-01 1 EDDR No. QAD 600 No. Sheet of _ N/A' N/A 1 1 Yes No Cicsed X lsolated X Pctential Design impact i X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion The inclusion of hydrotest load should not be considered safety related because it is not a_ safety function of the system. 3.0 Action Taken None required O.o Ce ciosie-The system was designed adequately for the hydrotest loads, in addition, this case was isolated based on the following. 1. gal normal procedure is to include the hydrotest load in their analysis. This system was not considered for this load case because it is an open ended piping system. 2. All 19 lines which are open ended still required hydrotesting are adequate for the hydrotest load. 3. Most of this hydrotesting has been completed with no adverse-effect. 5.0 References None Approvals Originator Date g S:nior ProJ6 t-Engineer Date 3 5 A $' 8fb Y CEI Supervis ua Audit Unit Date 70 m T/u/ry gal Project Manager Date yjpp ' fAl Manager Corporatey r r Date 1 1 Tha Cleveland Elactric illuminating Company: / P rry Nuclear Power Plant Piping Design Review-DW138/2/Q/sp -

Observation d (*h' i , Record lilllllllill!!nimerirIIII Rnision No. O Observation No. PS-02-01 sheet of Checklist No. PS-02-H001 HPCS 3 3 Date l-3-% Criginated By C.M, (IL)Oy(cy g Jgf gd Reviewed By '{ Date 1.0 Description The following design oversights were noted for support 1E22-H001: a. Wrong section properties were used in shear and deflection calculations (Pg. 10.4). b. Young's modulus "E" has not been adjusted for temperature effect in the stiffness calculation (Pgs.10.1 and 10.2). c. Welding between items D and F is overstressed. d. Dimensions of some items on the support drawings are not clearly defined (e.g. length of item D, and length of weld between F and D). 2.0 Requirement 2.1 ASME B&PV Code, Section III, 1974 with addenda to Winter 1975 Subsection NF. 2.2 Standard Industry Practice. 3.0 Document Reference i 3.1 GAI Support Design Calculation IE22-G04(B), Rev. 1. 3.2 Support drawings for MK-1E22-H001, S-322-701, Sht.1 and 2, Rev. E. 4.0 Design Impact Support is not adequate. 5.0 Probable Cause Design oversight. l Attachments I 1 A. Observation Record Review llO l l Cicycland Electric 111uninsting; S3102 l Perry Nuclear Power Plant Piping Design Review t

Observation A (. n i Record Review g mmmmmmmmumi Attachment A Cbservation No. PS-02-01 Checklist No. pS-02 Revision No. O sheet i of 1 PFR No. 02 Yes No Closed X Extent i of 3 Systems Comments a. Per GAI, in April 1983, due to reanalysis, the design loads increased by approximately 50%. This required substitution of a 6" schedule 160 pipe for a previous 5 x 5 x 1/2" tube section. For the shear and deflection calculation, the higher section properties were not used to update the calculation. This is conservative, b. Per Table I-6.0 of ASME Subsection NA, Young's modulus varies with temperature from 27.9 ksi at ambient to 27.3 ksi at 330, which is accident temperature inside drywell. Since this property is only used for the calculation of support ,O deflection and support stiffness, there is potentially a 2% maximum variation in calculated values. This would have a negligible impact on design, c. Due to the potential impact on design and safety associated with the overstressed weld, PFR 02 has been written. d. Per GAI, dimensions of a minor nature are not always provided on the GAI drawing. The GAI drawing is an engineering drawing which is re-detailed by the fabricator (PPC) for use as a fabrication / installation drawing. In accordance with the GAI fabrication specification SP-527, fabrication drawings are submitted to the engineer (GAI) for approval prior to use for fabrication. Adherence to this specification that the hardware will be properly dimensioned and that there will be no impact on design or safety. 1 Approvals orwn.ior c. K. ( ho, o=i. I-14--84 Proioci enen r L&n.1,d, (W/st o=i. g-Proi.ct uen oerWMjh Date ([24[$4' M 'gg m 4/PS'/ Ceiner. ni.iiv. o =i. Cleveland Elect.ric illwninating; B3102 Perry Nuclear Power Plant Piping Design Review i

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~ - - i l 1 1 ^( :El V Observation Record Closure Attachment B Observation No. Checklist No. Revision No. PS-02-01 PS-02-H001 0 EDDR No. QAD 600 No. Sheet of 071 N/A 1 1 Yes No Cl: sed X is lated X Potential Design impact X 1.0 Description See Cygna observation record and observation record review. 2.0 Discussion None required 3.0 Action Taken Review of other work by the designer to insure it was an isolated design error, and revise support so all ASME allowables have been met. A ().O Conclusion No substantial safety hazard would have been created if this had gone uncorrected. Follow-up documented on EDDR 071 verifies this was an isolated design error. 5.0 References (8) EDDR 071 (23) gal memo from B. M. Stevens to C. W. Whitehead dated March 31, 1984 Approvals Origindor g Date g, Senior Profd' no Date CEI Supervisor Quality udit Unit Date TJ. V//s/g4 o gal Project Manager 'Y Date '~ $/W8Y ~ -L gal Manager Corporap 'Pr Date ThJ Cleveland Electric illuminatirig Company: garry Nuclear Power Plant Piping Design Review / DW138/3/Q/sp

Observation 4(L i Record O' lillllillillllllilill!Illilll! ~ Revision No. g Observation No. PS-02-02 Sheet cf Checklist No. PS-02-H001 & H002 HPCS } 1 Date {-3 -% Criginated my d, K, @g t/p/gg coviewed my ig i case 1.0 Description The Jet loads on supports H001 and H002 are specified in the design specification, but were not included in the support design calculations. 2.0 Requirement 2.1 GAI Design Specification DSP-E22-1-4549-00, Rev. I and 2., 2.2 Perry FSAR Amendement No. 3, dated 9/11/83. 3.0 Reference Documents GAI Support design calculation IE22-G04(B), Rev. 1. 4.0 Potential Design Impact Design loads will be increased and.may necessitate redesign of the supports. 5.0 Probable Cause Design oversight. Attachments A. Observation Record Review t i l l l O i Cleveland Electric 111uminating; 83102 Perry Nuclear Power Plant Piping Design Review

Observation 4 f.% i Record Review a""""""""""'"" Attaehment A Cbservation No. PS-02-02 Checklist No. PS-02 Revision No. O PFR No. Sheet 1 of 1 Yes No Closed X Extent 1 of 2 Systems with Jet Loading comments Further review indicates that the jet map drawings are used in conjunction with the design specification to determine which jets strike particular supports. These drawings are continually updated as source shields are added. Per GAI, as a result of this process, supports 1E22-H001 and H002 are now. shielded from all breaks. i Based on the above, this Observation does not have any impact on design or safety. O l l. Apot vais \\-2 4 ; orwa.ior C..M. (a m9, D ML th' 7 o @ *l84 O,er winava, a4/u ,.-,...ee vwjsja o... $///f3/ f D * '* ff Av ces Repreeentattve Cleveler:3 Elect 71c illuminating; 83102 Perry Nuclear Power Plant Piping Design Review i

CEI observation Record Closure t Attachment B Observation No. Checklist No. Revision No. PS-02-02 PS H1101 L H002 0 EDDR No. QAD 600 No. Sheet of 065 N/A 1 1 Yes No Cl:: sed X is: lated X Potential Design impact X 1.0 Description See Cygna observation record and observation record review. I 2.0 Discussion This observation resulted from CYGNA reviewing the calculation before the Jet work was completed. The jet maps have been updated since the CYGNA review and will be again, before they are final. '3.0 Action Taken The generic Jet calculation (File Code X176) is being finalized. This will include i O a matrix of all Jets, shields, and piping or supports impacted.all Jets are accounte It will document that supports impacted are either shielded or designed to withstand the impact. 4.0 Conclusion Based on the Jet calculation X176, this item will not impact plant safety. 5.0 References (2) EDDR 65. a : u - u. (16) Generic Ja't Calculation - File Code X176 Approvals Originator Date g s.nior Projppin j .Date L CEl r ty i Unit Date gal P'roject Ma.no e Di.te 4 g ---*} gyg gal Manager _ Corpora K Date g 'The Cleveland Electric illuminefing Company: Perry Nuclear Power Plant Piping Design Review [ W138/Q/36/es

THE-C L E V E L A!: 0 E L E C i f. ! C I L L U E !!: Ali n G C O M P A l' Y B:#"W Piping Design 5 ' '" " - 6-I d 9 ??e O PAGE: Rev,iew REVISION. 0 6.0 CEI CONCLUSIONS Based on a thorough review of the Cygna report on the three systems, CEI concluded that the selected systems are adequately designed and will perform the intended safety functions. In addition, CEI believes the Cygna review was broad enough to uncover any generic design problems, which could affect the ability of other systems to perform their intended safety functions. As identified earlier in this report, CEI reviewed all observations and identified items of generic applicability to insure they will not affect any other safety systems. As a result of the piping design review, follow-up programs have been initiated in two areas. In the first area, three activities have been initiated to address { ) observations in the pipe support design program. These include finalizing a generic jet calculation, review of pipe inspection reports and the snubber cold set. program. In the second area, programs have been initiated to address documentation of the mechanical design. This will involve a thorough update of the mechanical process calculations, and will insure all GE criteria have been designed into the systems and the design properly documented. This CE Criteria Compliance Review program has been expanded to include all other disciplines (see Attachment 1). All of these follow-up activities will be tracked to closure (per Procedure 35-1501 EDDR, ur. der CEI's Quality Assurance program). l I i PET.RY NUCLEAR PCnNER PLANT Serving The Best Location in the Nation

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~. ~. ~~.#. se t.- ~ v I I '7.LI-EI,TT ' 5 5 ~ ;'- 60 Piping Design j;" Pd Me O at.GE Review mEv:S C. 0 6.0 CEI CONCLUSIONS (continued) With the activities underway to ensure complete closure of all EDDRs including issues of a generic nature, CEI concludes that the Perry Nuclear Power Plant mechanical and piping design will assure the safety function capabilities are maintained and the design is in compliance with all applicable codes, standards and regulations. Reviewed 4-f Senior Project Engine 4r Approved Ytc.w b c Manager - Nuclear Engineering Departnent I (G CEPRY NUCLEAR POWER PLANT Serving The Best location en the Nation PO Box 27 e PEARY OwiO 44081 e TELEPHONE e216e 2$9 3737 e A 00 A E S S.10 CENTER ROAO t PERRY NUCLEAR POWER PLANT l GE Criteria Compliance Review Procedure l 1.0 PURPOSE The purpose of the GE Criteria Compliance Review is to provide additional l assurance that the appropriate GE Criteria requirements have been properly incorporated and documented for the GAI scope of design. 2.0 SCOPE GE is the designer for the Perry reactor and emergency core cooling systeurs. 'Most major equipment for these systems are supplied by GE under the NSSS contract. GAI is the designer for the plant facility into which the NSSS will be installed as well as for a significant portion of these GE systems. This includes interconnecting piping, instrumentation, and most valves and other equipment. In addition, GAI is the designer for all support systems (cooling water, HVAC, compressed air, etc.) (-'s GE design requirements are contained in text-type documents (design (,) specifications, application engineering information, data sheets, ect.) and drawings (P&ID's, Process Diagrams, FCD's, IED's, etc.). It is GAI's responsibility to comply with these requirements or obtain GE and/or CEI approval of any deviations from them. _ The primary source document for review of system and plant requirements is the the PNPP 1 & 2 Master Parts List (MPL), GE Document No. 18NS06803, Revision 11. The baseline revision level of each of the GE documents used in the review will be that reported in GE CDCS Report RPT01, dated March 9,1984 Review results will be documented per this procedure, and any questions arising from the review will be resolved with GE as ex-peditiously as possible. As necessary design changes be initiated, which will be proposed to the Project Manager via Change Request (CR). Once the baseline review has been established, subsequent revisions of GE criteria docuacnts received by GAI will be reviewed and incorporated as appropriate. / v}}