ML17309A540

Summary of 940324 Public Meeting W/Util to Discuss Steam Generator Replacement & Fuel Reload Changes for 1996.Agenda Encl

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 I

'tt" i~

Cy TION SYSTEM REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9404190372 DOC.DATE: 94/04/13 NOTARIZED: NO FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G

,AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.

Rochester Gas

& Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.

Project Directorate I-3

SUBJECT:

Summary of 940324 public meeting w/util to discuss Steam generator replacement 6 fuel reload changes for 1996.Agenda encl.

DISTRIBUTION CODE:

A001D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: OR Submittal: General Distribution DOCKET 05000244 R

D NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

/

05000244 A

RECIPIENT ID CODE/NAME PD1-3 LA JOHNSON,A INTERNAL: NRR/DE/EELB NRR/DRCH/HICB NRR/DSSA/SPLB NUDOCS-ABSTRACT OGC/HDS1 EXTERNAL: NRC PDR COPIES LTTR ENCL 1

1 2

2 1

1 1

1 1

1 1

1 1

0 1

1 RECIPIENT ID CODE/NAME PD1-3 PD NRR/DORS/OTS B NRR/DRPW NRR/DSSA/SRXB OC/~LCB ILE 01 NSIC COPIES LTTR ENCL 1

1 1

1 1

1 1

1 1

0 1

1 1

1 D

D D

NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

TOTAL NUMBER OF COPIES REQUIRED:

LTTR 16 ENCL 14 D

D S,

t lllilSllill llISIZ IIIIII III l>V~@

ROCHESTER GAS AND ELECTRIC CORPORATION o

e9 EAST AVENUE, ROCHESTER N.Y. 14649-0001 ROBERT C MECREDY Vice President Glnna Nuclear Production TELEPHONE AstEA cooE T le 546-2700 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 PDR',i ADOCJC 05000244'=.PDR oo t

)

4 t

I l

I(

k CJu~~~ 0,

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 it was agreed that further detail would be needed before a response 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.

/

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, Attachment BJF/327 Robert C.

Mec edy 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

0 0

~

e f

I

Ginna Station BWI+

Bwl t Beep(

Rochester Gas and Electric GINNA STATION STE<AM GL<NERATOR IMPLACEMENT 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 FUEL CONTRACT GEORGE< WROBE<L BRIAN FLYNN BOB ELIASZ 3.0 STATUS / STRESS ANALYSISUPDATE JOHN SMITH BE<RNIE CARRICK 3.1 3.2 3.3 3.4 3.5 EQUIPMENT / INSTALLATIONSTATUS METHODOLOGY AND APPROACI-I ACCEPTANCE CRITERIA STATUS AND SCI-IEDULE DISPLACEMENT MODEL 4.0 S/G DE<SIGN FEATURE<S 5.0 CONTAINMENTOPE<,NING MODEL 6.0 SUBMITTALSCHEDULE 7.0 RE<PLACEML<'NT VIDE<0 JOHN SMITH JOHN SMITH GEORGE WROBEL JOHN SMITH 94p41 9p372 24 March 94

C

R.E. Ginna Stcam Gcncrator Rcplaccmcnt Comparison of Existing vs. Proposed Ginna Station stvl I recit te(

R Rochester Gas and E/ectric Manuf./Model Primary Side Prcssure Drops (0% plug)

Nozzle inlet to Nozzle outlet Existing W/44 (fccdring) 33.5 psi Replacement BWI (fccdring) 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 15% Plugging 20% Plugging Tubing OutsideDiameter Avg. Wall Thickness

~

~

Number of Tubes Material 44430 sq. ft.

37765 sq. ft.

0.875 in 0.050 in 3260 Inconcl 600, MA 54000 sq. ft.

43200 sq. ft.

'.750 in 0.0431 in 4765 Alloy 690, TT Volumes, primary side Inlet Plenum Tubes Outlet Plenum 133 cu. ft.

654.5 cu. ft.

133 cu. ft.

132.5 cu.ft.

710 cu.ft.

132.5 cu. ft.

Secondary Volume, Total 45SO cu. ft.

4513 cu. ft.

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

Secondary Mass Flow, 100%

Stcam Linc Orifice Size Initial Stcam Pressure, 100%

S4,500 Ibm 11S300 Ibm 3.3 E06 lbm/hr 4.37 sq. ft.

800 psia S6,200 lbm 115,100 Ibm 3.3 E06 lbm/hr 1.4 sq. ft.

875 psia 24 March 94

STEAM OUTLET TO TURBINE GENERATOR DEMISTERS SECONDARY MOISTURE SEPARATOR SECONDARY MANWAY ORFICE RINGS UPPER SHELL FEEDWATER RING SWIRL VANE PRIMARY MOISTURE SEPARATOR FEEDWATER INLET TUBE BUNDLE ANTIVIBRATIONBARS DOWNCOMER FLOW RESISTANCE PLATE LOWER SHELL WRAPPER SECQMDARY MAMOMOLE TUBE SUPPORT PIATES 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 I

Handholes Integral vertical suppon Primary nozzle Primary.side manway Design Objectives

~ Aetrsn ee ierlnrnrtpoints. outslde~

sna oversn per 4nnsnce consisleni >>ith exntrng sleem generelon snd reecror vertu ernenrs

~ oesrgn lor magnum retrebony in opersbon

~ Act>>eve high circulation ratios

~ Orrvnere crevices ond poiernisl rlovvtegnsuon

~ use norvngra hare suppons

~ lstntstln presswe boundsry Integri;y dunng seismic and bunt pipe events

~ Stnrmtxe lube resiausl soess

~ Avoidno>>endured varrsnon

~ Assure high steam rturrrty(rbavo 99.7$ rr) under

~a operating conCNttons

~ prevent loose ports spectton ond nleinlenslrce

Ginna Station Bechtel Lsmpson Ginna Accident Analysis Rochester Gas and E/ectric EVALUATE ANALYZE X

X X

X X

X X

X X

X X

X X

X X

X X

X X

X 15.1 Increase in I-Ieat Removal by the Secondary System 15.1.1 Decrease in Feedwater Temperature 15.1.2 Increase in Fcedwater 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 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 ANALYSISSTATUS ACTIVITY COMPLETION MODELS RELAP CONTEMPT COMPLETE COMPLETE JALYZE CONTAINMENT ANALYZEACCIDENTS 10CFR50.59 EVALUATION 5/94 9/94 12/94 24 March 94

A>>0 BVII+

Ginna Station B VII 1

Bech tel 2)

Rochester Gas and Eiectric GINNA RELAP5 MODEL Sl fAM $ )>.f M)LV 4)I)

S)EAM LINE I)

CC0 I

955 960

~965 ~9>0 0

~ 1b OI0

~ >4

>40 1>0 490 I>b I>0 645

~ N

~ 60 455 I50 490 4>0 N4 No eeb

~40

>44

>40

~ 70 660

>64

>60

~ F 5 MlA I" M>5 MI 924 I)O hkC'74 0'I 02 01 1)) 01 1>) 04

11) OL

>1) 06 IS) 0) 1)) 07

>1) 01 10ll 12 I)

I~

Ib 14 0>

OI OL 06 01 0>

-01

11) OI OI, 06

>)7 01

>i) oe

>11 04

>>) 06 1>>'>

>)1 0>

>))+MI MI))

~1 I>>5 MI IA I:4 ~

~ IM 674 I)0 424 01 Ot

.01 e)i o>

4)).06 ON.06

~)) 06

~)) 07 4)).at IS) 01 I)) 01 01

-oe

-04

.06 07 02 276-01

• 10

~ 11

~ltl)

.I~

~ Id

~ I~

-OI I

Oe 4)S 01 4)) 04 I)) 06

~7) 06 I)) 07

~)) 07 4)S OI 1 79 171

~ 00 2tl tte 1)0 120 61 220 2)0 l)d I )4

~)) 1 C3-110 Iod IOO 700 2nd 210 21) 214 7)d IN 164 1>4 IIO 240 2>6 t>o 246 2)I 160

)40 pQM>

>)>M >

ten 260 I~ L 164 lbd 140 f~)64 tdI 264 tdo 24 lvlARCH 94

ANO I

Ginna Slatlort Bechtel ljldlp$0h Rochester Gas and Electric GINNAREACTOR VESSEL AND CORE MODEL hot leg nozzle 102 350 106 Do C9 302 364 360 leakage path upper plenum 354 upper plenum 368 hot leg nozzle 351 103 107 348 326 338 349 100 104 cold leg nozzle 322 334 324 336 1

101 cold leg nozzle 320 332 318 316 328 312 375 24 MAI<CII 94 310 380

Ginna Station BIVI 1

Bechtel 828mOSOn CD CD Rochester Gas and Electric MAINFEEDWATER A AFW WSIOE CONTAlNMENT 801 MFW Pump A 02 806 808 810 820

--424!

822i-4~824

~83D i

818 18121814 81 6

SGA 625 888 ~~82 901 MFW Pump B 906 908 910 918

+-~~~22L~924

~30 931; 933 934 920

- -iw==-

SGB 725 AFW MAINFEEDWATER 8 24 ivlAl<Cfl94 Ginna MFW RELAP5 Node Diagram

Ginna Station Blvll Bectt t 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

kl

Ginna SIalion BIVII BeChtel Rochester Gas and Electric FVKI ASSKMBIY CHARACTERISTICS COMPARISON OF CURRENT VS. 1996 CHARACTERISTIC TYPE FUEL ROD O.D.

FUEL CLADDING MATERIAL ACTIVE FUEL LENGTH BLANKETREGION/

ENRICHMENT CENTER REGION ENRICI-IMENT BOTTOM NOZZLES CURRENT 14 X 14 OFA, 0.40 IN Z;4 141.4 IN 6 IN/NAT.

UP TO 4.25 W/0 DFBN 1996 VANTAGE 5 0.40 IN Z,-4 141.4 IN 6 IN/2.6 W/0 UP TO 5.0 W/0 DFBN GRIDS 2 TOP AND BOTTOM 7 MID INTERMEDIATE FLOW MIXING DISCHARGE BURNUPS INCONEL-718 Z,-4 NONE LOW 40s GWD/MTU

'NCONEL-718 Z,-4 NONE MID 50s GWD/MTU 24 MARCIA 94

Ginna Station Bechtel Rochester Gas and Electric Ginna Accident Analysis Valve Outside Containment 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 ofSmall Lines Carrying RC 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

I 1

Ginna Stalion BWI (

ge~ht

+npsan ANALYSES THATWILLBK UPDATED WITHRELOAD 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 VALVEFAILURE 15.2.7 LOSS OF EXTERNAL LOAD/TURBINETRIP 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 HANDLINGACCIDENTS 5.2.2 I.OW TEMP. OVERPRESSURIZATION (BWNT) 24 MARCH 94

Ginna Station Bechtel Rochester Gas and Electric SCHEDULE DATAPREPARATION 4/I/94 FINALIZEINPUT DATA 6/1/94 START ANALYSIS DRAFT REPORT FINALREPORT 6/1/94 6/1/95 7/1/95 SUBMIT REPORT TO NRC 8/1/95 CYCLE 26 STARTVP 5/1/96 24 MARCH 94

BIYI)

Beef)te/

E UIPMENT STATUS

~

FABRICATIONBY BOW INTERNATIONAL

~

ALLMAJOR COMPONENTS ORDERED MAJOR FORGINGS, JAPAN STEEL WORKS TUBING, VALINOX SHELL PLATE, CREUSOT-LOIRE

~

PRIMARYHEADS CLADDINGCOMPLETE NOZZLE DAMRINGS BEING INSTALLED PRIMARYNOZZLE BUTTERING UNDERWAY

~

TUBESHEETS CLADDING COMPLETE READY FOR GUNDRILLINGOF TUBEHOLES

~

SECONDARY SHELLS LOWER SHELL CONES WELDED HANDHOLES AND INSPECTION PORTS BEING INSTALLED

~

TRANSITION CONK FORGINGS

HANDHOLE OPENINGS CUT 24 March 94

Ginna Station swI st t Bechtel E UIPNIENT STATUS CONT'0 TUmNG PRE PRODUCTION UNDERWAYAT VALINOX PREPRODUCTION COMPLETE JUNE 1994 PRODUCTION MATERIALBEING MELTED AT INCO PRODUCTION COMPLETE DECEMBER 1994 24 March 94

Ginna Station BltII+

BlYI)

Bechtel Lpmpson l INSTALLATIONSTATUS

~

INSTALLATIONCONTRACTOR BKCHTKL POWER DETAILED DESIGN INSTALLATION

~

DETAILEDENGINEERING 1994

~

PROCEDURE PREPARATION 1995

~

ACTIVITIESTO DATE PROJECT INTERFACE PROCEDURES VIDEO PREPARATION CONTAINMENTOPENING 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

Ginna Station Blvl t Bechtgl Rochester Gas and Electric R. K. GINNA STEAM GENERATOR REPI ACKMKNT

\\

STRVCTVRALEVALUATION OF EFFECTED COMPONENTS & SYSTEMS=

24 March 94

Ginna Station BlYIl Beep(

REACTOR COOLANT SYSTEM LATEtST ANALYSIS

~

1988 SNUBBER REPLACEMENT

~

IMPLEMENTLBB/HELB CRITERIA

~

RIGID STRUTS S/G COLD SPRING ACCEPTANCE CRITERIA

~

PIPING Q

EQUIPMENT O

SUPPORTS USASFANSI B31.1 ASMK SECTION III ASME SECTION III 24 March 94

AHD Ginna Station Bechtel Ijlfhp50h CD Rochester Gas and Electric 38 HAIN S1EAM LINE STEAM GENERnlOR

)A 38 HAiN StEAM LINE 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.)

CROSSOYER LEG COLO 'LEG LOWER SUPPORTS lTYP.)

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 DETAILEDLOADING/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 LOADINGCONDITIONS 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

EQVIPNtKNT AUXLINKS

- B31.1 ALLOWABLKS CURRENT LBB CRITERIA

- NOZZLKS LOADS

- SUPPORT LOADS

- < '/,6 " ADDED DEFLECTION 24 March 94

c!'

'1

~

4 w~

J

~ ~

Ginna Stalion Bechtel La npsen Rochester Gas end Electric MODEL ENHANCEMENTS i

NKW CONSISTENT MASS MODELING EXPLICIT MODELING OF SUPPORTS FREQUENCY CUTOFF 30Hz N-411 DAMPING OLD LUMPED MASS STIFFNESS MATRIX 100 Hz 2%/4% DAMPING SINGLE ANCHOR Pt./SINGLE MULTIPLEANCHOR SPECTRA Pts./ENVELOPE SPECTRA CLOSELY SPACED MODES VIA10 Z. RULE EXPLICIT ACP ANALYSIS FOR HKLB EPSILON RULE FACTOR ON DISCHARGE COEFFICIENT 24 March 94

i

)

s Ginna Station Bwl t Bechtel lampsan LOADINGMETHODS SEISMIC DWhTHKRMAL STATIC OBEYSSE RESPONSE SPECTRA

~

3 AXIS EXCITATION

~

CURRENT DESIGN SEISMIC SPECTRA MULTIPLECASKS 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

1

'0

~+~

~ 4

~e

& 8 ~

~

~.

Ginna Stalion BlVI+

Bechtel Rochester Gns and E/eclric LOADINGMETHODS LOCALHK'LB CRAFT dk, COMPAR2 CODES I

LINEARTIME HISTORY ANAI,YSIS PIPINGLINTKRNALSTRANSIENT 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 MAINSTEAM FEEDWATER BLOWDOWN RECIRC NOZZLE 24 March 94

"f

~ RL

Ginna Station ewt I Bechtel Rochester Gas and Electric DKADWKIGIIT THERMAL SEISMIC LOCA HKLB CVRRKNT EXPECTATION INCREASE (<5%)

- SAME

- DECREASE '

INCREASE (< 15%) '

DECREASE 1.

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

BLOWDOWNINITIALCONDITIONS WILLUSK 15'F REDUCED TAvE 24 March 94

t I

4

Ginna Station Bwl t gech~e/

Rochester Gas and Electric STATUS dk SCHEDULE MODELS NKAmWG COMPLETION STRUCTURAL BENCHMAKKINGSTARTED SEISMIC ANAIVSIS:

COMPLETE 7/94 BLOWDOWN ANALYSES:

COMPLETE 10/94 LOAD COMBINATIONdt.

COMPARISONS COMPLETE 1/95 24 March 94

DISPLACEMENT MODEL PURPOSE RG&K TECHNICALOVERSIGHT 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

Ginna Station BlVI)

Etet f)t~]

Lpneson Rochester Gas and Electric

t

~CJi.

II

Ginna Sialion SWI -,

Bechtel lpmpean a S/G DESIGN FEATURES PROBI KM TUBESHEET DEFECTS SLUDGE ACCUMULATIONON TUBESHEET DEFECTS AT 7UBE 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 HAMMER J-TUBE FAILURES GOOSE NECK AT FEEDWArER RING INLET I-TUBES INCONEL 690 FOR EROSION RESISTANCE 24 March 94

'l I

'iM

,It~

F %~

Ginna Station Bwl )

gecgfg(

S/G DESIGN FKATVRKS PROBI KM MOISTURE CARRYOVER PRESSURE BOUNDARY WELD FAILURES PWSCC OF U-BFNDS SECONDARY LOOSE PARTS PRIMARY SIDE ACCESS SECONDARY SIDE ACCESS PRIMARYNOZZLE WELDING DESIGN FKATVRKS HIGH EFFICIENCY SEPARATORS 0.10%

GUARANTEE FORGED AND PLATE COMPONENTS NO CORNER WELDS STRICT PRE AND POST HEAT REQVIREMENTS LARGE MINIMUM RADIUS BENDS STRESS RELIEF OF FIRST 8 ROWS NO FASTENERS, 100%

WELDED STRUCTURE 18" DIAMETER MANWAYS 6-8" HANDHOLES 14-2" INSPECTION PORTS 1-18" MANWAY 316 LN SAFE ENDS NARROW GAP WELDING SPARE ELBOWS 24 March 94

"C

,VX'A)

Ginna Slation Bwl

)

Bechtgl CONTAINMENTOPENING MODEL CONTAINMENTOPENING DESIGN INCLUDEDIN BKCHTKLWORKSCOPK RGK MODEL FOR OVKRCHKCKOF 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

~

WILLBE AVAILABLETO VERIFY BECHTEL DESIGN AND FOR CONSTRUCTION.

24 March 94

V' C

n

Ginna Station Bechtel SCHEDULE FOR INkOl&'IATIOWAISUBMITTAIS COMPONENT ACTIVITIES PRELIMINARYSAFETY EVALUATION MAY 1994 FINAL REPORT/50.59 EVALUATION MAY 1995 nS rAI,r.ETIO~ WCTVmIKS SAFETY EVALUATIONOF CONTAINMENTOPENING AUGUST 1994 SAFETY EVALUATIONOF RIGGING AND HANDLING SAFETY EVALUATIONOF STEAM GENERATOR PIPING SAFETY EVALUATIONOF STEAM GENERATOR INSULATION TESTING AND INSPECTION PLAN OCTOBER 1994 DECEMBER 1994 DECEMBER 1994 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

E 0