ML17263A640

Summary of 940324 Meeting W/Util Re Update of Plans for SG Replacement at Plant

ML17263A640
Person / Time
Site:	Ginna
Issue date:	05/10/1994
From:	Andrea Johnson Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
TAC-M86500, NUDOCS 9405170052
Download: ML17263A640 (59)

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Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205554001 May 10, 1994 LICENSEE:

Rochester Gas and Electric Corporation FACILITY:

R.E.

Ginna Nuclear Power Plant

SUMMARY

OF MEETING WITH ROCHESTER GAS AND ELECTRIC CORPORATION ON MARCH 24, 1994 STEAN GENERATOR REPLACEMENT (TAC NO. M86500)

Rochester Gas and Electric Corporation (RG&E), in a meeting with the NRC staff on Narch 25, 1994, at NRC Headquarters, presented an update of their plans for steam generator (SG) replacement at the Ginna plant.

RG&E plans to replace SGs at Ginna during the 1996 refueling outage.

RG&E intends to replace like-for-like SGs under the provisions of 10 CFR 50.59.

Under the provisions of 10 CFR 50.59, RG&E intends to replace the SGs without changes to the plant Technical Specifications (TSs),

and make available to the NRC its safety analyses to support its determination and finding that "no unreviewed safety question exists."

RG&E stated that no TS changes are necessary to replace the SGs because the new SGs are not larger than those currently installed and involve:

(1) no alterations to core operating limits, (2) no changes to reactor trip setpoints, (3) no modifications to departure-from-'nucleate bpiling parameters, (4) no changes to shutdown limits, and (5) no changes to he engineered safety features actuation system setpoints for SG water levels.,

RC&E's preliminary analyses is scheduled to be subtiIitted to NRC in Hay 1994.

The final analyses is scheduled for submittal to NRC with the final 10 CFR 50.59 report in Hay 1995.

If RG&E's position changes from that described

above, and requires changes:

(1) to plant TSs, or (2) to the facility involving an unreviewed safety question pursuant to 10 CFR 50.59, then RG&E will be required to submit an application to amend the operating license using the procedures in 10 CFR 50.91.

The 10 CFR 50.91 procedures call for RG&E to provide its analysis about the issue of no significant hazards consideration using the standards in 10 CFR 50.92.

The Commission may then publish a notice for public comment (an individual or periodic) in the Federal Receister for which it may make a

proposed determination that no significant hazards consideration is involved, thus informing the public and providing an opportunity for hearing pursuant to the 10 CFR 50.91 procedures.

RG&E presented to the NRC a breakdown and status of their Steam Generator Replacement Project (SGRP).

RG&E's presentation divided the SGRP into:

(1) component design and (2) installation activities.

The RG&E safety

'nalyses schedule for installation activities was proposed and included such work activities as:

(1) containment

openings, (2) rigging and handling of heavy loads, (3)

SG pipe cutting and welding, (4)

SG insulation, (5) storage of old SGs, and (6) the testing and inspection plan.

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I Hay 10, 1994 RG&E presented to the NRC a comparison of the new and old SGs in terms of their similarity.

The NRC staff noted that they were very similar.

RG&E presented to the NRC a discussion on their preliminary safety analyses on the SG design and detailed thermal hydraulic models of the primary system and containment with regard to reanalyzing several postulated accident transients contained in the updated final safety analysis report.

RG&E also presented an overview of the improvements in the design features of the new SGs.

'The staff cautioned that close attention should be given to materials used for weld transitions.

RG&E presented to the NRC their intended methods to be used for the structural evaluations of the restored reactor coolant system (RCS).

RG&E's intent is to explicitly include modeling of walls and supports when analyzing the restored RCS.

The staff cautioned that close attention should be given to this modeling to ensure that consistent and conservative modeling of wall-to-RCS interfaces are made, particularly with regard to damping factors.

RG&E also presented to the NRC information on a displacement model of the RCS developed to predict pipe movement during pipe cutting.

The staff questioned the appropriateness and use of temporary supports and restraints, and requested RG&E to pursue the engineering activity in more detail.

RG&E presented their efforts to date in modeling the containment to analyze the effects of the planned temporary construction openings during and after restoration of the containment dome.

The staff noted that the model appeared to include, sufficient detail to analyze all effects of the openings;, however the staff would'.review the model and results when completed.

P RG&E also presented an overview of several changes that would accompany the 1996 new fuel reload for the Ginna plant which will undergo a transition from an annual to 18-month fuel cycle.

RG&E described several TS amendments associated with the 1996 fuel reload.

A copy of a list of meeting attendees is included in Enclosure I.

Enclosure 2

is a copy of the meeting agenda and Enclosure 3 is the discussion material.

Enclosures:

1.

List of Attendees 2.

Heeting Agenda 3.

Discussion Haterial Allen R. Job son, Project Hanager Project.Di'rectorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next page

May 10, 1994 RG&E presented to the NRC a comparison of the new and old SGs in terms of their similarity.

The NRC staff noted that they were very similar.

RG&E presented to the NRC a discussion on their preliminary safety analyses on the SG design and detailed thermal hydraulic models of the primary system and containment with regard to reanalyzing several postulated accident transients contained in the updated final safety analysis report.

RG&E also presented an overview of the improvements in the design features of the new SGs.

The staff cautioned that close attention should be given to materials used for weld transitions.

RG&E presented to the NRC their intended methods to be used for the structural evaluations of the restored reactor coolant system (RCS).

RG&E's intent is to explic'itly include modeling of walls and supports when analyzing the restored RCS.

The staff cautioned that close attention should be given to this modeling to ensure that consistent and conservative modeling of wall-to-RCS interfaces are made, particularly with r'egard to damping'factors.

RG&E also presented to the NRC information on a displacement model of the RCS developed to predict pipe movement during pipe cutting.

The staff questioned the appropriateness and use of temporary supports and restraints,

'and requested RG&E to pursue the engineering, activity in more detail:

I RG&E presented their efforts to date in modeling the'containment to analyze the effects of the planned temporary construction openings during and after restoration of the containment dome.'he staff noted that the model appeared to include sufficient detail to analyze all effects of the openings; however the staff would review 'the model" and results when completed.

I RG&E also presented an overview of several changes. that would accompany the 1996 new fuel reload for the Ginna plant 'which will undergo a."transition from an annual to 18-month fuel cycle.i RG&E described"several TS amendments associated with the 1996 fuel reload.

A copy of a list of meeting attendees is, included in Enclosure 1.

Enclosure 2

is a copy of the meeting agenda and Enclosure 3 is the discussion material.

Enclosures:

1.

List of Attendees 2.

Meeting Agenda 3.

Discussion Material Or.ig'inal signed by Allen R. Johnson,'roject Manager Project Directorate I-3 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next a

e OFF1CE LA:P PM:PDI-3 4 D: PDI-3 HAHE DATE SLittle 94 AJohnson:mw WButler 5 /IQ/94 4

/94 OFFICIAL RECORD COPY FILENAME: A:iMTSUM.G I

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DISTRIBUTION w enclosures:

,. Docket File

'RC 5 Local PDRs PDI-3 Reading J. Linville, RI A. Johnson DISTRIBUTION w enclosures 1

8 2:

W. Russell/f. Hiraglia L. Reyes S.

Varga J.

Calvo W. Butler S. Little OGC E. Jordan C.

P.

Tan F. Orr H. Caruso G. Lainas J.

Rajan K. Hanoly R.

Hermann A. Lee ACRS (10)

A. Thadani B. Sheron P. Patnik, RI H. Hodes, RI G.

Wunder J. Cutchin G. Bagchi R. Jones J.

Norberg W. Lazarus, RI

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R.E. Ginna Nuclear Power Plant CC:

Thomas A. Hoslak, Senior Resident Inspector R.E.

Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road

Ontario, New York 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs.

Donna Ross Division of Policy Analysis

& Planning New York State Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Charlie Donaldson, Esq.

Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Nicholas S. Re)holds Winston Strawn'400 L St.

N;W.

Washington, DC 20005-3502 Ms. Thelma Wideman

Director, Wayne County Emergency Management Office Wayne County Emergency Operations Center 7370 Route 31

Lyons, New York 14489 Hs.

Mary Louise Heisenzahl Administrator, Monroe County Office of Emergency Preparedness ill West Fall

Road, Room ll Rochester, New York 14620 Dr. Robert C. Mecredy

.Vice President, Nuclear Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649

I

Enclosure 1

LIST OF MEETING ATTENDEES NRC MEETING WITH ROCHESTER GAS AND ELECTRIC CORPORATION R.E.

GINNA NUCLEAR POWER PLANT STEAM GENERATOR REPLACEMENT MARCH 24, 1994 NAME TITLE A. Johnson, Project Mana er Bob Eliasz John F. Smith Bernard J. Carrick Brian J.

Flynn Dan Sten er Lynn Connor Bob Borsum Mark Smith Chen P.

Tan'eor e Wrobel Frank Orr Mark Caruso Gus Lainas Walter Butler J.

Rajan Kamal Manoly Robert Hermann Arnold Lee ORGANIZATION NRC DRPE PDI-3 RG&E RG&E RG&E RG&E Winston

& Strawn STS BWNT Bechtel NRC NRR ECGB RG&E NRR SRXB NRR SRXB NRR ADRI NRR PDI-3 EMEB DE NRR EMEB DE NRR EMEB DE NRR EMEB/DE/NRR

nclos re 2

Ginna Station Rochester Gas and Electric GINNA STATION 4

STEAM GENERATOR REPLACEMENT NRC STATUS UPDATE MARCH 24 1994 AGENDA

1.0 INTRODUCTION

2.0 SAFETY AND LICENSING 2.1 PRELIMINARYSAFETY EVALUATION 2.2 RELAP MODEL 2.3 CONTEMPT MODEL 2.4

'FUgL CONTRACT 3.0 STATUS'/ STRESS ANALYSIS UPDATE 3.1 EQUIPMENT / INSTALLATIONSTATUS 3.2

~ METHODOLOGYAND APPROACH 3.3 ACCEPTANCE CRITERIA 3.4 STATUS AND SCHEDULE 3.5 DISPLACEMENT MODEL GEORGE WROBEL BRIAN FLYNN BOB ELIASZ JOHN SMITH BERNIE CARRICK 4.0 S/G DESIGN FEATURES 5.0 CONTAINMENTOPENING MODEL 6.0 SUBMITTALSCHEDULE 7.0 REPLACEMENT VIDEO JOHN SMITH JOHN SMITH GEORGEWROBEL JOHN SMITH 24 March 94

Enclosure 3

BWI Ginna Station Bechtel R.E. Ginna Steam Generator Replacement Comparison of Existing vs. Proposed Rochester Gesand Electric Manuf./Model Primary Side Pressure Drops (0% plug)

Nozzle inlet to Nozzle outlet Existing W/44 (feedring) 33.5 psi Replacement BWI (feedring) 31.1 psi Primary Side Flow for above dp's 34.6 E06 lbm/hr 34.9 E06 Ibm/h Heat Transfer Areas 0% Plugging 15% Plugging 20% Plugging Tubing OutsideDiameter Avg. Wall Thickness Number of Tubes Material Volumes, primary side Inlet Plenum Tubes Outlet Plenum 44430 sq. ft.

37765 sq. ft.

0.875 in 0.050 in 3260 Inconel 600, MA 133 cu. ft.

654.5 cu. ft.

133 cu. ft.

54000 sq. ft.

43200 sq. ft.

0.750 in 0.0431 in 4765 Alloy 690, TT 132.5 cu.ft.

710 cu.ft.

132.5 cu. ft.

Secondary Volume, Total 4580 cu. ft.

4513 cu. ft.

-Secondary Water Mass, nominal 100% (1520 MWt) 0% (HZP)

Secondary Mass Flow, 100%

Steam Line Orifice Size Initial Steam Pressure, 100%

84,500 lbm 118300 lbm 3.3 E06 lbm/hr 4.37 sq. ft.

800 psia 86,200 Ibm 115,100 lbm 3.3 E06 lbm/hr 1.4 sq. ft.

875 psia 24 March'94

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'I ANO Glnna Station Beehgtel Ginna Accident Analysis Rochester Gas and Electrfc EVALUATE ANALYZE X

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X rl Valve RC Chapter 6, Chapter 5 6.2.1.2 Containment Integrity 5.2.2 Low Temperature Overpressurization 15.1 Increase in Heat Removal by the Secondary System 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Feedwater Flow 15.1.3 Excessive Load Increase Incident 15.1.4 Inadvertent Opening of a SG Relief/Safety Valve 15.1.5 Steam Line Breaks Inside and Outside Containment 15.1.6 SG Relief Valve and Feedwater Control Valve Failure 15.2 Decrease in Heat Removal by the Secondary System 15.2.1 Steam Pressure Regulator Malfunction 15.2.2 Loss of External Electrical Load 15.2.3 Turbine Trip 15.2.4 Loss of Condenser Vacuum 15.2.5 Loss of Offsite Power to the Station Auxiliaries 15.2.6 Loss of Normal Feedwater Flow 15.2.7 Feedwater System Pipe Breaks 15.3 Decrease in RCS Flowrate 15.3.1 Flow Coastdown Accidents 15.3.2 Locked Rotor Accident 15.4 Reactivity and Power Distribution Anomaities 15.4.1 Uncontrolled RCCA Withdrawal from Subcritical 15.4.2 Uncontrolled RCCA Withdrawal at Power 15.4.3 Startup of an Inactive Reactor Coolant Loop 15.4.4 CVCS Malfunction 15.4.5 RCCA Ejection 15.4.6 RCCA Drop 15.5 Increase in RCS Inventory 15.6 Decrease in RCS Inventory 15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief 15.6.2 Radiological Consequences of Small Lines Carrying Outside Containment 15.6.3 Steam Generator Tube Rupture 15.6.4 Primary System Pipe Ruptures 15.6.4.1 SBLOCA 15.6.4.2 LBLOCA 15.7 Radiological Release From a Subsystem or Component 15.7.1 Radiological Gas Waste System Failure 15.7.2 Radiological Liquid Waste System Failure 15.7.3 Fuel Handling Accidents 15.8 Anticipated Transients Without Scram

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Rochester Gas and E/ectrfc STEAM GENERATOR REPLACEMENT TRANSIENT ANALYSISSTATUS ACTIVITY COMPLETION MODELS RELAP CONTEMPT ANAI.YZECONTAINMENT ANALYZEACCIDENTS 10CFR50.59 EVALVATION COMPLETE COMPLETE 5/94 9/94 12/94 24 March 94

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Glnna Station Bdchtei Rochester Gas and E/ectrtc GINNA REACTOR VESSEL AND CORE MODEL hot leg nozzle 102 350 106 302 364 360 leakage path upper plenum 354 352 upper plenum 368 hot leg nozzle 351 107 103 348 100 104 cold leg nozzle 326 324 322 338 336 334 349 105 101 cold leg nozzle 320

318, 330 316 328 312 375 24 WAR('ll 94 310 380

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Ginna Station BWt 'I Bet:hgg f A29D Rochester Gas and Electric MAINFEEDWATER A INSOE CONTANMENT d

801 MFW Pump A 02 806 808 810 I812 814 815 816 820 818 822 824 830 SGA 625 900 MFW Pump B 906 908 910 HIA

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CC Rochester Gas and E/ectrfc 1996 FUEL RELOAD

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NKW WESTINGHOUSE FUEL FABRICATION CONTRACT

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INSERT VANTAGE-5 FUEL VS. OFA FUEL F. INCREASE FROM 1.66 TO 1.75 Fq INCREASE FROM 2.32 TO 2.50 CYCLE LENGTH INCREASE FROM ANNUALTO 18 MONTHS INCREASE SFP ENRICHMENT FROM 4.25 W/0 TO 5.0 W/0 POSSIBLE T,, DECREASE UP TO 15'F 24 MARCH 94

Ginna Stalion BWIt gegQtg J lornpean Rochester Gas and Electric FUEL ASSEMBLY CHARACTERISTICS COMPARISON OF CURRENT VS. 1996 CHARACTERISTIC CURRENT 1996 TYPE FUEL ROD O.D.

FUEI. 'CLADDING MATERIAL ACTIVE FUEL LENGTH 14 X 14 OFA 0.40 IN Z,-4 141.4 IN VANTAGE 5 0.40 IN Z;4 141.4 IN BLANKETREGION/

ENRICHMENT CENTER REglgN ENRICHMENT BOTTOM NOZZLES 6 IN/NAT.

UP TO 4.25 W/0 DFBN 6 IN/2.6 W/0 UP TO 5.0 W/0 DFBN GRIDS 2 TOP AND BOTTOM 7 MID INTERMEDIATE FLOW MIXING INCONEL-718 Z;4 NONE INCONEL-718 Z;4 NONE DISCHARGE BURIES LOW 40s GWD/MTU MID 50s GWD/MTU 24 hciARCH 94

BWI Ginna Station Bach+tel Rochester Gas and Electrfc Ginna Accident Analysis 15.1 Increase in Heat Removal by the Secondary System 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Feedwater Flow 15.1.3 Excessive Load Increase Incident 15.1.4 Inadvertent Opening, of a SG Relief/Safety Valve 15.1.5 Steam Line Breaks Inside and Outside Containment 15.1.6 SG Relief Valve and Feedwater Control Valve Failure 15.2 Decrease in Heat Removal by the Secondary System 15.2.1 Steam Pressure Regulator Malfunction 15.2.2 Loss of External Electrical Load 15.2.3 Turbine Trip 15.2.4 Loss of Condenser Vacuum 15.2.5 Loss of Offsite Power to the Station Auxiliaries 15.2.6 Loss of Normal Feedwater Flow 15.2.7 Feedwater System Pipe Breaks 15.3 Decrease in RCS Flowrate 15.3.1 Flow Coastdown Accidents 15.3.2 Locked Rotor Accident 15.4 Reactivity and Power Distribution Anomaities 15.4.1 Uncontrolled RCCA Withdrawal from Subcritical 15.4.2 Uncontrolled RCCA Withdrawal at Power 15.4.3 Startop of an Inactive Reactor Coolant Loop 15.4.4 CVCS Malfunction

,15.4.5 RCCA Ejection 15.4.6 RCCA Drop 15.5 Increase in RCS Inventory 15.6 Decrease in RCS Inventory 15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve 15,6.2 Radiological Consequences ofSmall Lines Carrying RC Outside Containment 15.6.3 Steam Generator Tube Rupture 15.6.4 Primary System Pipe Ruptures e

15.6.4.1 SBLOCA 15.6.4.2 LBLOCA 15.7 Radiological Release From a Subsystem or Component 15.7.1 Radiological Gas Waste System Failure 15.7.2 Radiological Liquid Waste System Failure 15.7.3 Fuel Handling Accidents 15.8 Anticipated Transients Without Scram 6.2.1.2 Containment Integrity 5.2.2 Low Temperature Overpressurization 24 March 94

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Psoll Rochester Gas and Electric ANALYSES THAT WILLBE UPDATED WITH RELOAD 15.1.1 DECREASE IN FEEDWATER TEMPERATURE 15.1.2 INCREASE IN FEEDWATER FLOW 15.1.3 EXCESSIVE LOAD INCREASE INCIDENT 15.1.4 INADVERTENT OPENING OF A SG RV 15.1.5 SLB (BOTH CORE AND MkE) 15.1.6

.SG RV 0 FW CONTROL VALVEFAILURE p

15.2.7 LOSS OF EXTERNAL LOAD/TURBINETRIP 15.3.1 FLOW COASTDOWN ACCIDENTS 15.6.3 SG TUBE RUPTURE n

15.6.4.1 SBLOCA 15.6.4.2 LBLOCA 15.7.3 FUEL HANDLINGACCIDENTS 5.2.2 LOW TEMP. OVERPRESSURIZATION (BWNT)

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Ginna Station BWI I BBCtltel Rochester Gas and Electric SCHEDULE DATAPREPARATION 4/1/94 "FINALIZEINPUT DATA 6/1/94 START ANALYSIS DRAFT REPORT 6/1/94 6/1/95

'FINALREPORT 7/1/95 jl SUBMIT REPORT TO NRC 8/1/95 Y

CYCLE 26 STARTUP 5/1/96 24 MARCH 94

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FABRICATIONBY BdkW INTERNATIONAL

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ALL1VlASOR COMPONENTS ORDERED MAJOR FORGINGS, JAPAN STEEL WORXS TUBING, VALINOX SHELL PLATE, CREUSOT-LOIRE

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PRIMARY HEADS CLADDINGCOMPLETE NOZZLE DAM RINGS BEING INSTALLED PRIMARYNOZZLE BUTTEKlNG UNDERWAY pi TUBE SHEETS CLADDINGCOMPLETE READY FOR GUNDRILLINGOF TUBEHOLES SECONDARY SHELLS LOWER SHELL CONES WELDED ED&lDHOLES AND INSPECTION PORTS BEING INSTALLED TWWSITION CONE FORGINGS HANDHOLE OPENINGS CUT 24 March 94

Ginna Station Rochester Gas and Electric K UIPMENT STATUS CONT'D

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Ginna Station

'" j Bechteli Rochester Gas and Electric INSTALLATIONSTATUS

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INSTALLATIONCONTRACTOR BECHTEL POWER DETAILED DESIGN INSTALLATION DETAILEDKNGINKKRII.'ttG 1994 PROCEDURE PREPARATION 1995 ACTIVITIESTO DATE PROJECT 1NTERFACE PROCEDURES VIDEO PREPAIMTION CONTAINMENTOPENING STUDY QA PROCEDURE MANUAL PROJECT ENGINEERING PROCEDURES MANUAL INSULATIONSTUDY DK'WT DESIGN CRITERIA FOR CONTAINMENT STRUCTURAL WORK

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gee Rochester Gas and Electric R. K. GINNA STEAM GENERATOR REPLACEMENT STRVCTVRALKVALVATION OF KF ECTED COMPONENTS dk, SYSTEMS

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24 March 94

Ginna Station BWI Bechgtel Rochester Gas and E/ectrfc REACTOR COOLANT SYSTEM LATEST ANALYSIS

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19SS SNUBBER REPLACEMENT

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IMPLEMENTLBB/HELBCRITERIA

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RIGID STRUTS

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S/G COLD SPRING ACCEPTANCE CRITERIA

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PIPING

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EQUIPMENT

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SUPPORTS USASV.NSI B31.1 ASMK SECTION III ASME SECTION III 24 March 94

8WI Ginna Station Bechtet Rochester Gas and Electric 38 MAIN STEAM t.INE STEAM GENERATOR IA 38-Malu STEAM LtNE SIEAM GENERAIOR I

REAC'TOR COOLANT PUMP I8 II FEEOWATER LINE I4-FEEOWATER LINE UPPER SUPPORTS ANO SNU88ERS ITYP.)

COLO LEG HOI LEG HOT LEG INTERMEOIATE SUPPORTS ITYP.)

CROSSOVER LEG COLO LEG I

~ r LOWER SUPPORTS (TYP.I REACTOR COOLANT PUMP I A REACTOR VESSEL

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VIEW OF NSSS SYSTEM F'R GINNA NUCLEAR STATION

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Ginna Station h+~ i Behhte/I Rochester Gas and Electric STRUCTURAL MODELS BWSPAN - STRUCTURAL CODE 3 MODELS

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BENCHMARK

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ENHANCED W/OLD S/G ENHANCED W/NKW S/G

.DEMONSTRATE UNDERSTANDING OF CURRENT BASIS AND LOOP BEHAVIOR GENERATE.DETAILEDLOADING/STRESS INFO FOR CURRENT S/G DISTINGUISH EFFECTS OF MODEL ENHANCEMENTS AND S/G DIFFERENCES CONFIRM/CALCULATELOADS, STRESSES, AND THERMAL MOTIONS FOR NKW S/G 24 March 94

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DEADWEIGHT

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DEADWEIGHT

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OBE ENHANCED MODEL W/NKW S/G

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Rochester Gas and Electnc ACCEPTANCE CRITERIA 1.

COMPARE TO CURM<"NT ANALYSIS NKW LOADS < OLD LOADS

= OK-2., COMPARE TO 'ALLOWABLKS PIPING.

pi EQUIPMENT

- 831.1 ALLOWABL'KS

- CURRENT LBB CRITERIA

- NOZZLKS LOADS

- SUPPORT LOADS AUXLINKS

/" ADDED DEFLECTION 24 March 94

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Ginna Station SW Btachtel Rochester Gas and E/ectrfc MODEL ENHANCEMENTS NEW CONSISTENT MASS MODELING EXPLICIT MODELING OF SUPPORTS FREQUENCY CUTOFF,.,

30Hz OLD LUMPED MASS STIFFNESS MATRjlX 100 Hz N-411 DAMPING 2%/4% DAMPING SINGLE ANC OR Pt./SINGLE MULTIPLEANCHOR SPECTRA ',,

Pts./EI.'&&LOPE SPECTRA CLOSELY SPACED MODES VIA10% RULE EXPLICIT ACP ANALYSIS FOR HELB EPSILON RULE FACTOR ON DISCHARGE COEFFICIENT 24 March 94

I

Ginna Station sec~g~ei Rochester Gas and E!ectric LOADINGMETHODS SEISMIC DW<THKRMAL O

STATIC OBE>8SE

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RESPONSE SPECTRA

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3 AXIS EXCITATION

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CURRENT DESIGN SEISMIC SPECTRA

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MULTIPLECASES FOR SUPPORTS

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MODES COMBINED SRSS CLOSELY SPACED MODES VIA 10% RULE

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TIME HISTORIES FOR ARSs ON S/G SHELL 24 March 94

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Ginna Station Bechgtel Rochester Gas and Electric LOADINGMETHODS LOCASIELB CRAFT dk COMPAR2 CODES LINEARTIME HISTORY ANALYSIS

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PIPINGUNTERNALS TRANSIENT LOADS

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M/K RELEASE FOR ACP ANALYSIS

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ACP ON COMPONENTS

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CONFIRM SUPPORTS ACTIVE RCS: LEAK-BEFORE-BREAK 3MR LINK

. SPRGK LINK SI 'LINK HELB: TERMINALLOCATIONS ONLY MAINSTEAM FEKDVVATER BLOWDOWN RECIRC NOZZLE 24 March 94

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Bechtel Rochester Gas and Electric CURIMNT EXPECTATION DEADWEIGHT THERMAL SEISMIC LOCA HKLB

- INCREASE (<5%)

- SANK

- DECREASE '

INCREASE (< 15%) '

DECREASE p

1.

MODEL ENHANCEMENTS = MORE MARGIN INCREASED WEIGHT = LESS MARGIN OVERALLEFFECT - EXPECT MORE MARGIN 2.

BLOWDOWNINITIALCONDITIONS WILLUSE 15'F REDUCED T~~

24 March 94

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qP Ginna Station Rochester Gas and Electric STATUS dk SCHEDULE C

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MODELS NEARING COMPLETION STRUCTURAL BKNCKNARKINGSTARTED SEISMIC ANALYSIS:

BLOWDOWN ANALYSES:

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LOAD COMBINATIONdk COMPARISONS:

COMPL'ETK 7/94 COMPLETE 10/94 COMPLETE 1/95 24 March 94

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c Ginna Station SWt Bechtet Rochester Gas and Electric DISPLACEMENT MODEL PURPOSE RG8rE TECHNICALOVERSIGHT OF RCS PIPING CUT dk WELD CONCERNS COLD SPRING (BEFORE i AFTER)

WELD FIT-UP RCS TEMPORARY SUPPORT DESIGN MODEL ANSYS Fg(ITE ELEMENT MODEL OF RCS PLATE dk BEAMELEMENTS MODEL RCS PIPE AS SHELLS 3400 ELEMENTS / 3400 NODES ONK LOOP MODEL STATUS MODEL NEAR COMPLETION 24 March 94

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Ginna Station SWl Bechtel Rochester Gas and Electric

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Rochester Gas and Electric S/G DESIGN FEATURES PROBLEM TUBESHEET DEFECTS SLUDGE ACCUMULATIONON TUBESHEET DEFECT 'AT TUBE SUPPORT PLATES DESIGN FEATURES CLOSED CREVICE HYDRAULIC EXPANSION INCONEL 690 TUBING HIGH CIRCULATION RATIO INSPECTION/MAINTEN-ANCE PORTS ACCESSIBLE FOR SLUDGE LANCING LATTICE GRIDS STAINLESS STEEL CONSTRUCTION INCONEL 690 MATERIAL HIGH CYCLE FATIGUE FAN BAR SUPPORT SYSTEM WATER HA2vtMER J-TUBE FAILURES GOOSE NECK AT FEEDWATER RjNG INLET J-TUBES INCONEL 690 FOR EROSION RESISTANCE 24 March 94

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S/G DESIGN FEATURES Ginna Station swt t Bechtel PIOll Rochester Gas and Etectrtc PROBLEM DESIGN FEATURES MOISTURE C&KYOVER HIGH EFFICIENCY SEPAL.TORS 0.10%

GUARANTEE PRESSURE BOA'43ARY WELD FAILURES PWSCC OF U-BENDS SECONDARY LOOSE PARTS PRIMARY SIDE ACCESS SECONDARY SIDE ACCESS PRIMARYNOZZLE WELDING FORGED Aj.'K) PLATE COMPONENTS NO CORNER WELDS STRICT PRE AND POST HEAT REQUIREMENTS LARGE MINIMUM IUD)IUS BENDS STRESS RELIEF OF FIRST 8 ROWS NO FASTENERS, 100%'ELDED STRUCTURE 18" DIAMETER MANWAYS 6-8" HANDHOLES 14-2" INSPECTION PORTS 1'-18" MANWAY 316 LN SAFE ENDS NMK.OWGAP WELDING SPARE ELBOWS 24 March 94

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f Ginna Slalion Bechtel Rochester Gas and Electric CONTAINMKNTOPENING MODEL CONTAINMENTOPENING DESIGN INCLUDEDIN BKCHTKLWORKSCOPK RGK'ODEL FOR OVKRCHECK OF BECHTEL DESIGN

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ANSYS FINITE ELEMENT MODEL

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INCLUDES ENTIRE CONTAINMKNTSTRUCTURE ROCK)ANCHORS BASE'ATS WALLS AND TENDONS SPRING LINE AND TRANSITION DOME STEEL, CONCRETE AND LINER PLATE MODEL DEVELOPMENT COMPLETE

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VERIFIED AGAINST CLASSICAL SOLUTIONS

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WILLBE AVAILABLKTO VERIFY BKCHTKL DESIGN AND FOR CONSTRUCTION.

24 March 94

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IHO BWI Ginna Station Be eh+tel Rochester Gas and Electrtc SCHEDULE FOR INFORlV1ATIONALSVBMITTALS COMPONENT ACTIVITIES PRELIMINARYSAFETY EVALUATION MAY 1994 FINAL REPORT/50.59 EVALUATION MAY 1995 INSTALLATIONACTIVITIES SAFETY EVALUATIONOF CONTAINMEN OPENING SAFETY EVALUATIONOF RIGGING AND HANDLING SAFETY EVALUATIONOF STEAM GENERATOR PIPING SAFETY EVALUATIONOF STEAM GENERATOR INSULATION TESTING AND INSPECTION PLAN AUGUST 1994 OCTOBER 1994 DECEMBER 1994 DECEMBER 1994 hhMCH 1995 24 March 94

%k Ginna Station sw Bechtel Rochester Gas and Electric SCHKDULK FOR SUBMITTALSFOR REVIEW I-690 RELIEF REQUEST MAY 1994 CURI&NT STEAM GENERATOR TUBE RUPTUI& ANALYSIS

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FUEL RELOAD REPORT KJLY 1994 AUGUST 1995 24 March 94

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