ML17309A540
| ML17309A540 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/13/1994 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9404190372 | |
| Download: ML17309A540 (56) | |
Text
ACCELERATED DI TRIBUTION DEMONS TION SYSTEM I
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Cy REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:9404190372 DOC.DATE: 94/04/13 NOTARIZED: NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G 05000244
,AUTH. NAME AUTHOR AFFILIATION MECREDY,R.C. Rochester Gas & Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R. Project Directorate I-3 R
SUBJECT:
Summary of 940324 public meeting w/util to discuss Steam generator replacement 6 fuel reload changes for 1996.Agenda encl. D DISTRIBUTION CODE: A001D COPIES RECEIVED:LTR ENCL SIZE:
TITLE: OR Submittal: General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72). 05000244
/
A RECIPIENT COPIES RECIPIENT COPIES D ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD1-3 LA 1 1 PD1-3 PD 1 1 JOHNSON,A 2 2 D INTERNAL: NRR/DE/EELB 1 1 NRR/DORS/OTS B 1 1 NRR/DRCH/HICB 1 1 NRR/DRPW 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 1 NUDOCS-ABSTRACT 1 1 OC/~LCB 1 0 OGC/HDS1 1 0 ILE 01 1 1 EXTERNAL: NRC PDR 1 1 NSIC 1 1 D
D D
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t ROCHESTER GAS AND ELECTRIC CORPORATION o e9 EAST AVENUE, ROCHESTER N.Y. 14649-0001 ROBERT C MECREDY TELEPHONE Vice President AstEA cooE T le 546-2700 Glnna Nuclear Production April 13, 1994 U.S. Nuclear Regulatory Commission Document Control Desk Attn: Allen R. Johnson Project Directorate I-3 Washington, D.C. 20555
Subject:
Public Meeting on March 24, 1994 to discuss Steam Generator Replacement and Fuel Reload Changes for 1996 R.E. Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson:
On March 24, 1994 a public meeting was held with Rochester Gas and Electric and members of the NRC Staff to update the staff on the evaluation of the steam generator replacement and fuel reload changes at Ginna Station, currently scheduled for the spring of 1996. This letter provides a summary of the meeting.
J. Cutchin (NRC/OGC) began the meeting with a discussion of the use of 10CFR50.59 for steam generator replacement. It was noted that the NRC Staff determined that 10CFR50.59 could not be used for replacement at V.C. Summer because several changes to the Technical Specifications were intimately connected with the installation of the new larger steam generators. Mr. Cutchin stated that this did not signal a change in staff philosophy. Under the terms of 10CFR50.59, even if a replacement does not involve an unreviewed safety question, prior NRC approval is required if it involves a change in the Technical Specifications. Mr. Cutchin indicated that if the plant cannot be operated following a change without certain Technical Specification amendments, then prior approval would be required. However, a replacement project could be broken down into several discrete plant changes (as has been done at other plants),
with the result that prior approval via the Technical Specification amendment process would only be required for certain specific changes, not the entire project. The staff observed that Ginna appears to be different from the V.C. Summer case in that the new steam generators will be physically similar to the existing ones, and that no necessary Technical Specification amendments have been identified.
RG&E presented the current status of the Steam Generator Replacement Project. RG&E noted that the project would be broken down into two distinct parts: the component design; and the 9404i90372 Tl404i3 oo PDR',i ADOCJC 05000244'=.PDR t )
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installation activities. There are currently no intentions to further break down the component design aspect of the project. The installation activity aspect, however, would be further divided into discrete work activities, notably the containment openings, rigging and handling of heavy loads, pipe cutting and restoration, insulation of the generator, and storage of the old steam generators. Current evaluations indicate that all changes associated with the steam generator replacement can be accomplished without prior staff approval. A schedule was presented indicating when final safety evaluations for these items would be available.
These will be provided to the staff for information only.
RG&E presented a comparison of the existing and replacement steam generators. The staff noted that the generators appear very similar. RG&E discussed the preliminary safety evaluation of the replacement steam generator design, and the detailed thermal hydraulic models of the primary system and containment being developed to reanalyze some of the UFSAR transients. The staff questioned the need for detailed models, given the similarities of the generators.
RG&E presented an overview of changes associated with the 1996 Fuel Reload. Ginna will transition from annual fuel cycles to eighteen month fuel cycles with the 1996 reload, and therefore there would be some Technical Specification amendments associated with the reload. A schedule for these submittals was presented.
RG&E presented an overview of some of the design features incorporated in the replacement steam generator to correct deficiencies identified in the current generators. The staff observed that these improvements seemed prudent, however they cautioned that close attention should be paid to the materials used for weld transitions.
RG&E outlined the inten'ded methods to be used for the structural evaluation of the restored reactor coolant system. Specifically, RG&E intends to include explicit modeling of walls and supports when analyzing the restored system. The staff cautioned that while this appears acceptable, close attention should be given to this to ensure that consistent and conservative modeling of the wall-RCS interfaces are made, particularly with respect to damping factors.
RG&E also presented information on a displacement model of the reactor coolant system developed to help predict pipe movement during cutting activities. The staff was questioned as to the appropriateness of using temporary supports or restraints, but was agreed that further detail would be needed before a response it could be provided.
RG&E presented the efforts to date to model the containment to analyze the effects of the temporary construction openings on the structure both during construction and after restoration of the containment. The staff noted that the model appeared to be of
sufficient detail to analyze all effects of the openings.
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Copies of the overheads presented by RG&E at this meeting are attached. RG&E appreciates the valuable input and cooperation that was provided by the staff at this meeting.
Very truly yours, Robert C. Mec edy Attachment BJF/327 xc: Mr. Allen R. Johnson (Mail Stop 14D1)
Project Directorate I-3 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King of Prussia, PA 19406 USNRC Ginna Senior Resident Inspector
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Rochester Gas and Electric GINNA STATION STE<AM GL<NERATOR IMPLACEMENT NRC STATUS UPDATE MARCH 24 1994 AGENDA
1.0 INTRODUCTION
GEORGE< WROBE<L 2.0 SAFETY AND LICENSING BRIAN FLYNN BOB ELIASZ 2.1 PRELIMINARY SAFETY EVALUATION 2.2 RELAP MODEL 2.3 CONTEMPT MODEL 24 FUEL CONTRACT 3.0 STATUS / STRESS ANALYSIS UPDATE JOHN SMITH BE<RNIE CARRICK 3.1 EQUIPMENT / INSTALLATIONSTATUS 3.2 METHODOLOGY AND APPROACI-I 3.3 ACCEPTANCE CRITERIA 3.4 STATUS AND SCI-IEDULE 3.5 DISPLACEMENT MODEL 4.0 S/G DE<SIGN FEATURE<S JOHN SMITH 5.0 CONTAINMENT OPE<,NING MODEL JOHN SMITH 6.0 SUBMITTALSCHEDULE GEORGE WROBEL 7.0 RE<PLACEML<'NT VIDE<0 JOHN SMITH 94p41 9p372 24 March 94
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Rochester Gas and E/ectric R.E. Ginna Stcam Gcncrator Rcplaccmcnt Comparison of Existing vs. Proposed Existing Replacement Manuf./Model W/44 (fccdring) BWI (fccdring)
Primary Side Prcssure Drops (0% plug)
Nozzle inlet to Nozzle outlet 33.5 psi 31.1 psi Primary Side Flow for above dp's 34.6 E<06 lbm/hr 34.9 E<06 lbm/h IIeat Transfer Areas 0% Plugging 44430 sq. ft. 54000 sq. ft.
15% Plugging 37765 sq. ft.
20% Plugging 43200 sq. ft.
Tubing OutsideDiameter 0.875 in '.750 in Avg. Wall Thickness
~ 0.050 in 0.0431 in Number of Tubes 3260 4765
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Material Inconcl 600, MA Alloy 690, TT Volumes, primary side Inlet Plenum 133 cu. ft. 132.5 cu.ft.
Tubes 654.5 cu. ft. 710 cu.ft.
Outlet Plenum 133 cu. ft. 132.5 cu. ft.
Secondary Volume, Total 45SO cu. ft. 4513 cu. ft.
Secondary Water Mass, nominal 100% (1520 MWt) S4,500 Ibm S6,200 lbm 0% (HZP) 11S300 Ibm 115,100 Ibm Secondary Mass Flow, 100% 3.3 E06 lbm/hr 3.3 E06 lbm/hr Stcam Linc Orifice Size 4.37 sq. ft. 1.4 sq. ft.
Initial Stcam Pressure, 100% 800 psia 875 psia 24 March 94
STEAM OUTLET TO TURBINE GENERATOR DEMISTERS SECONDARY MOISTURE SEPARATOR SECONDARY MANWAY ORFICE RINGS UPPER SHELL SWIRL VANE PRIMARY MOISTURE SEPARATOR FEEDWATER RING FEEDWATER INLET ANTIVIBRATION BARS TUBE BUNDLE DOWNCOMER FLOW RESISTANCE PLATE LOWER SHELL WRAPPER TUBE SUPPORT PIATES SECQMDARY MAMOMOLE BLOWOOWN LINE TUBE SHEET TUBE LANE BLOCK PRIMARY MANWAY PRIMARY COOIANT OUTLET PRIMARY COOIANT INLET QK SERIES 44 STEAM GENERATOR NOT 70 SCALE 922$
Babcock 8 Witch Canada Advanced Series
'PNR,Rtep/acement Steam Generator Steam nozzle High elliciency moisture separators Bundle inspection access duct Feedwalef inlel Feedwater header Stainless steel lnttice4sar lube support gird PVF&).
Tube bundle Handholes
, I Integral vertical suppon Primary nozzle Primary.side manway Design Objectives
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sna oversn per 4nnsnce consisleni >>ith exntrng sleem generelon snd reecror vertu ernenrs oesrgn lor magnum retr ebony in opersbon Act>>eve high circulation ratios
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lstntstln presswe boundsry Integri;y dunng seismic and bunt pipe events Stnrmtxe lube resiausl soess Avoid no>>endured varrsnon Assure high steam rturrrty (rbavo 99.7$ r r) under
~ Orrvnere crevices ond poiernisl rlovv tegnsuon ~ a operating conCNttons
~ prevent loose ports
~ use norvngra hare suppons spectton ond nleinlenslrce
Ginna Station Bechtel Lsmpson Rochester Gas and E/ectric Ginna Accident Analysis EVALUATE ANALYZE 15.1 Increase in I-Ieat Removal by the Secondary System X 15.1.1 Decrease in Feedwater Temperature X 15.1.2 Increase in Fcedwater Flow X 15.1.3 Excessive Load Increase Incident X 15.1.4 Inadvertent Opening of a SG Relief/Safety Valve X 15.1.5 Steam Line Breaks Inside and Outside Containment X 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 X 15.4.1 Uncontrolled RCCA Withdrawal from Subcritical X 15.4.2 Uncontrolled RCCA Withdrawal at Power X 15.4.3 Startup of an Inactive Reactor Coolant Loop X 15.4.4 CVCS Malfunction X 15.4.5 RCCA Ejection X 15.4.6 RCCA Drop X 15.5 Increase in RCS Inventory X 15.6 Decrease in RCS Inventory X 15.6.1 Inadvertent Opening of a Pressurizer Safety or Relief Valve 15.6.2 Radiological Consequences of Small Lines Carrying RC Outside Containment 15.6.3 Steam Generator Tube Rupture 15.6.4 Primary System Pipe Ruptures 15.6.4.1 SBLOCA X 15.6.4.2 LBLOCA X 15.7 Radiological Release From a Subsystem or Component 15.7.1 Radiological Gas Waste System Failure X 15.7.2 Radiological Liquid Waste System Failure X 15.7.3 Fuel Handling Accidents X 15.8 Anticipated Transients Without Scram Chapter 6, Chapter 5 6.2.1.2 Containment Integrity 5.2.2 Low Temperature Overpressurization 24 March 94
Ginna Station SW Bechtel Rochester Gas and Electric STEAM GENERATOR REPLACEMENT
- TRANSIENT ANALYSIS STATUS ACTIVITY COMPLETION MODELS RELAP COMPLETE CONTEMPT COMPLETE JALYZE CONTAINMENT 5/94 ANALYZEACCIDENTS 9/94 10CFR50.59 EVALUATION 12/94 24 March 94
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Sl f AM $ )>.f S)EAM LINE 955 960 ~965 ~9>0 0 ~ 1b OI0 490 I>b I>0 645 ~ 60 455 I50
~ >4 ~N I) >40 490 CC 0
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>44 >40 ~ 70 660
>64 >60 ~ F 5
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~ 00 1)0 120 61 220 2)0 l)d C3-I )4 ~)) 1 110 Iod IOO 700 2nd 210 21) 214 7)d IN 164 1>4 I IO 240 2>6 t>o 246 2)I 160 )40 pQM> >)>M > ten 260 I~ L 164 lbd tdI 264 140 f~)64 tdo 24 lvlARCH 94
Ginna Slatlort I
Bechtel ljldlp$ 0h ANO Rochester Gas and Electric CORE MODEL GINNA REACTOR VESSEL AND 364 hot leg hot leg 360 nozzle leakage path nozzle D upper plenum oC9 351 107 103 102 350 354 106 368 302 upper plenum 349 348 326 338 1
101 100 336 104 324 cold leg cold leg nozzle nozzle 322 334 320 332 318 316 328 312 375 310 380 24 MAI<CII 94
Ginna Station BIVI Bechtel 1
828mOSOn CD CD Rochester Gas and Electric MAIN FEEDWATER A WSIOE CONTAlNMENT AFW 820 SGA 801 MFW Pump A 02 806 808 810 18121814 81 6 i --424!
818 822i-4~824 ~83D 625 SGB 888 ~~82 906 908 910 918
+-~~~22L~924 ~30 931; 933 934 725 901 MFW 920 Pump B - -iw ==-
AFW MAIN FEEDWATER 8 Ginna MFW RELAP5 Node Diagram 24 ivlAl<Cfl 94
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1996 FUEL RELOAD
~ NEW WESTINGHOUSE FUEL FABRICATION CONTRACT
~ 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
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Rochester Gas and Electric FVKI ASSKMBI Y CHARACTERISTICS COMPARISON OF CURRENT VS. 1996 CHARACTERISTIC CURRENT 1996 TYPE 14 X 14 OFA, VANTAGE 5 FUEL ROD O.D. 0.40 IN 0.40 IN FUEL CLADDING MATERIAL Z;4 Z,-4 ACTIVE FUEL LENGTH 141.4 IN 141.4 IN BLANKET REGION/
ENRICHMENT 6 IN/NAT. 6 IN/2.6 W/0 CENTER REGION UP TO 4.25 UP TO 5.0 ENRICI-IMENT W/0 W/0 BOTTOM NOZZLES DFBN DFBN GRIDS 2 TOP AND BOTTOM INCONEL-718 'NCONEL-718 7 MID Z,-4 Z,-4 INTERMEDIATE FLOW MIXING NONE NONE DISCHARGE BURNUPS LOW 40s MID 50s GWD/MTU GWD/MTU 24 MARCIA 94
Ginna Station Bechtel Rochester Gas and Electric 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 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 Valve 15.6.2 Radiological Consequences of Small Lines Carrying RC 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 6.2.1.2 Containment Integrity 5.2.2 Low Temperature Overpressurization 24 March 94
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+npsan ANALYSES THAT WILL BK 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 M&E)
I 15.1.6 SG RV Ec FW CONTROL VALVE FAILURE 15.2.7 LOSS OF EXTERNAL LOAD/TURBINE TRIP 15.3.1'LOW COASTDOWN ACCIDENTS 15.6.3 SG TUBE RUPTURE 15.6.4.1 SBLOCA 15.6.4.2 LBLOCA 15.7.3 FUEL HANDLING ACCIDENTS 5.2.2 I.OW TEMP. OVERPRESSURIZATION (BWNT) 24 MARCH 94
Ginna Station Bechtel Rochester Gas and Electric SCHEDULE DATA PREPARATION 4/I/94 FINALIZEINPUT DATA 6/1/94 START ANALYSIS 6/1/94 DRAFT REPORT 6/1/95 FINAL REPORT 7/1/95 SUBMIT REPORT TO NRC 8/1/95 CYCLE 26 STARTVP 5/1/96 24 MARCH 94
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E UIPMENT STATUS
~ FABRICATION BY BOW INTERNATIONAL
~ ALL MAJOR COMPONENTS ORDERED MAJOR FORGINGS, JAPAN STEEL WORKS TUBING, VALINOX SHELL PLATE, CREUSOT-LOIRE
~ PRIMARY HEADS CLADDING COMPLETE NOZZLE DAM RINGS BEING INSTALLED PRIMARY NOZZLE BUTTERING UNDERWAY
~ TUBESHEETS CLADDING COMPLETE READY FOR GUNDRILLING OF TUBEHOLES
~ SECONDARY SHELLS LOWER SHELL CONES WELDED HANDHOLES AND INSPECTION PORTS BEING INSTALLED
~ TRANSITION CONK FORGINGS HANDHOLE OPENINGS CUT 24 March 94
Ginna Station st t swI Bechtel E UIPNIENT STATUS CONT'0 TUmNG PRE PRODUCTION UNDERWAY AT VALINOX PREPRODUCTION COMPLETE JUNE 1994 PRODUCTION MATERIALBEING MELTED AT INCO PRODUCTION COMPLETE DECEMBER 1994 24 March 94
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INSTALLATIONSTATUS
~ INSTALLATIONCONTRACTOR BKCHTKL POWER DETAILED DESIGN INSTALLATION
~ DETAILED ENGINEERING 1994
~ PROCEDURE PREPARATION 1995
~ ACTIVITIES TO DATE PROJECT INTERFACE PROCEDURES VIDEO PREPARATION CONTAINMENT OPENING STUDY QA PROCEDURE MANUAL PROJECT ENGINEERING PROCEDURES MANUAL INSUI.ATION STUDY DRAFT DESIGN CRITERIA FOR CONTAINMENT STRUCTURAL WORK
~ MAJOR SUBCONTRACTORS POWER CUTTING LAMPSON PSI 24 March 94
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Rochester Gas and Electric R. K. GINNA STEAM GENERATOR REPI ACKMKNT
\
STRVCTVRAL EVALUATION OF EFFECTED COMPONENTS & SYSTEMS=
24 March 94
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l REACTOR COOLANT SYSTEM LATEtST ANALYSIS
~ 1988 SNUBBER REPLACEMENT
~ IMPLEMENT LBB/HELB CRITERIA
~ RIGID STRUTS S/G COLD SPRING ACCEPTANCE CRITERIA
~ PIPING USASFANSI B31.1 Q EQUIPMENT ASMK SECTION III O SUPPORTS ASME SECTION III 24 March 94
Ginna Station Bechtel Ijlfhp50h AHD CD Rochester Gas and Electric 38 HAIN S1EAM LINE 38 HAiN StEAM LINE STEAM GENERnlOR )A S)EAM CENERA1OR REACIOR COOLAN1 PUMP )B 14 FEEOWATER LINE 14 FEEOWATER LINE UPPER SUPPORTS ANO SNUBBERS l'TYP.)
COLO LEG HOT LEG NOT LEG INTERHEOIATE SUPPORTS lTYP.)
LOWER SUPPORTS lTYP.)
CROSSOYER LEG COLO 'LEG REACTOR COOLANT PUHP lA REACtOR vESSEL VIEW OF NSSS iTEM FOR GINNA NUCLEAR STATION
Ginna Stalion Bechtel STRUCTURAL MODELS BWSPAN STRUCTURAL CODE 3 MODELS BENCHMARK ENHANCED W/OLD S/G
~ ENHANCED W/NKW S/G DEMONSTRATE UNDERSTANDING OF CURRENT BASIS AND LOOP BEHAVIOR GENERATE DETAILED LOADING/STRESS INFO FOR CURRENT S/G DISTINGUISH EFFECTS OF MODEL ENHANCEMENTS AND S/G DIFFKRKNCK CONFIRM/CALCULATELOADS, STRESSES, AND THERMAL MOTIONS FOR NKW S/G 24 March 94
Ginna Station Bill t Bechtel Rochester Gas and Electric LOADING CONDITIONS BENCHMARK DEADWEIGHT 0 OBK ENHANCED MODEL W/OLD S/G
~ DEADWEIGHT 0 OBE ENHANCED MODEL W/NEW S/G Oe DEADWErCHT
~ THERMAL OBK/SSK
~ LOCA/HKLB Q COLD SHUTDOWN EARTHQUAKE 24 March 94
Ginna Station Bechtel I
ACCEPTANCE CRITERIA
- 1. COMPARE TO CURRENT ANALYSIS NKW LOADS < OLD LOADS = OK
- 2. COMPARE TO ALLOWABLKS 0 pnnwG I
- B31.1 ALLOWABLKS CURRENT LBB CRITERIA EQVIPNtKNT - NOZZLKS LOADS
- SUPPORT LOADS AUX LINKS - < '/,6 " ADDED DEFLECTION 24 March 94
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NKW OLD CONSISTENT MASS LUMPED MASS MODELING EXPLICIT MODELING OF STIFFNESS MATRIX SUPPORTS FREQUENCY CUTOFF 100 Hz 30Hz N-411 DAMPING 2%/4% DAMPING SINGLE ANCHOR Pt./SINGLE MULTIPLE ANCHOR SPECTRA Pts./ENVELOPE SPECTRA CLOSELY SPACED MODES EPSILON RULE VIA 10 Z. RULE EXPLICIT ACP ANALYSIS FACTOR ON DISCHARGE FOR HKLB COEFFICIENT 24 March 94
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lampsan LOADING METHODS SEISMIC DWhTHKRMAL STATIC OBEYS SE RESPONSE SPECTRA
~ 3 AXIS EXCITATION
~ CURRENT DESIGN SEISMIC SPECTRA MULTIPLE CASKS FOR SUPPORTS 0 MODES COMBINED SRSS a cr.osKz,v spAcKD moDEs vw. io z RUr.E TIME HISTORIES FOR ARSs ON S/G SHELL 24 March 94
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LINEARTIME HISTORY ANAI,YSIS PIPINGLINTKRNALS TRANSIENT LOADS 0 M/E RELEASE FOR ACP ANALYSIS
~ ACP ON COMPONENTS
~ CONFIRM SUPPORTS ACTIVE RCS: LEAK-BEFORE-BREAK RHR ZZNE SURGE LINK SI LINK IIELB: TERMINALLOCATIONS ONLY MAIN STEAM FEED WATER BLOWDOWN RECIRC NOZZLE 24 March 94
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Rochester Gas and Electric CVRRKNT EXPECTATION DKADWKIGIIT INCREASE (<5%)
THERMAL - SAME SEISMIC - DECREASE '
LOCA INCREASE (< 15%) '
HKLB DECREASE
- 1. MODEL ENHANCEMENTS = MORE MARGIN INCREASED WKIGIIT= LESS MARGIN OVERALL EFFECT - EXPECT MORE MARGIN
- 2. BLOWDOWN INITIALCONDITIONS WILL USK 15'F REDUCED TAvE 24 March 94
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t Rochester Gas and Electric STATUS dk SCHEDULE MODELS NKAmWG COMPLETION STRUCTURAL BENCHMAKKINGSTARTED SEISMIC ANAIVSIS: COMPLETE 7/94 BLOWDOWN ANALYSES: COMPLETE 10/94 LOAD COMBINATION dt.
COMPARISONS COMPLETE 1/95 24 March 94
DISPLACEMENT MODEL PURPOSE RG&K TECHNICAL OVERSIGHT OF RCS PIPING CUT &
WELD CONCERNS COLD SPRING (BEFORE 7 AFTER)
WELD FIT-UP RCS TEMPORARY SUPPORT DESIGN VIODEL ANSYS FINITE ELEMENT MODEL OF RCS PLATE & BEAM ELEMENTS MODEL RCS PIPE AS SHELLS 3400 ELEMENTS / 3400 NODES ONE LOOP MODEL STATUS MODEL NEAR COMPLETION 24 March 94
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S/G DESIGN FEATURES PROBI KM DESIGN FEATURES TUBESHEET DEFECTS CLOSED CREVICE HYDRAULIC EXPANSION INCONEL 690 TUBING SLUDGE HIGH CIRCULATION ACCUMULATIONON RATIO TUBESHEET INSPECTION/MAINTEN-ANCE PORTS ACCESSIBLE FOR SLUDGE LANCING DEFECTS AT 7UBE LATTICE GRIDS SUPPORT PLATES STAINLESS STEEL CONSTRUCTION INCONEL 690 MATERIAL HIGH CYCLE FATIGUE FAN BAR SUPPORT SYSTEM WATER HAMMER GOOSE NECK AT FEEDWArER RING INLET I-TUBES J-TUBE FAILURES INCONEL 690 FOR EROSION RESISTANCE 24 March 94
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S/G DESIGN FKATVRKS PROBI KM DESIGN FKATVRKS MOISTURE CARRYOVER HIGH EFFICIENCY SEPARATORS 0.10%
GUARANTEE PRESSURE BOUNDARY FORGED AND PLATE WELD FAILURES COMPONENTS NO CORNER WELDS STRICT PRE AND POST HEAT REQVIREMENTS PWSCC OF U-BFNDS LARGE MINIMUM RADIUS BENDS STRESS RELIEF OF FIRST 8 ROWS SECONDARY LOOSE NO FASTENERS, 100%
PARTS WELDED STRUCTURE PRIMARY SIDE ACCESS 18" DIAMETER MANWAYS SECONDARY SIDE 6-8" HANDHOLES ACCESS 14-2" INSPECTION PORTS 1-18" MANWAY PRIMARY NOZZLE 316 LN SAFE ENDS WELDING NARROW GAP WELDING SPARE ELBOWS 24 March 94
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CONTAINMENT OPENING MODEL CONTAINMENT OPENING DESIGN INCLUDED IN BKCHTKL WORKSCOPK RGK MODEL FOR OVKRCHKCK OF BKCHTKL DESIGN ANSYS FINITE ELEMENT MODEL INCLUDES ENTIRE CONTAXNMKNTSTRUCTURE ROCK ANCHORS BASE rVrATS WALLS AND TENDONS SPRjNG LINE AND TRANSITION DOME STEEL, CONCRETE AND LINER PLATE MODEL DEVELOPMENT COMPLETE VERIFIED AGAINST CLASSICAL SOLUTIONS
~ WILL BE AVAILABLETO VERIFY BECHTEL DESIGN AND FOR CONSTRUCTION.
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Ginna Station Bechtel SCHEDULE FOR INk Ol&'IATIOWAI SUBMITTAIS COMPONENT ACTIVITIES PRELIMINARY SAFETY EVALUATION MAY 1994 FINAL REPORT/50.59 EVALUATION MAY 1995 nS rAI,r.ETIO~ WCTVmIKS SAFETY EVALUATIONOF CONTAINMENT OPENING AUGUST 1994 SAFETY EVALUATIONOF RIGGING AND HANDLING OCTOBER 1994 SAFETY EVALUATIONOF STEAM GENERATOR PIPING DECEMBER 1994 SAFETY EVALUATIONOF STEAM GENERATOR INSULATION DECEMBER 1994 TESTING AND INSPECTION PLAN MARCH 1995 24 March 94
Ginna Station Bechtel Ipmp0an Rochester Gas and Electric SCHEDULE< FOR SUBMITTAI,S FOR REIVIKW I-690 RELIEF REQUEST MAY 1994 CURRENT STFAM GENERATOR TUBE RUPTURE ANALYSIS JULY 1994 FUEl RELOAD REPORT AUGUST 1995 24 March 94 I
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