ML17263A479

Summary of 930629 Meeting W/Util Re S/G Replacement for 1996 & TS Conversions.List of Attendees Encl

ML17263A479
Person / Time
Site:	Ginna
Issue date:	12/02/1993
From:	Andrea Johnson Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 9312100051
Download: ML17263A479 (49)

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Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COIVllVllSSlON WASHINGTON, D.C. 20555 0001 December 2, i993 LICENSEE:

Rochester Gas and Electric Corporation FACILITY:

R.

E. Ginna Nuclear Power Plant

SUMMARY

OF MEETING WITH ROCHESTER GAS AND ELECTRIC CORPORATION JUNE 29, 1993 - STEAN GENERATOR REPLACEHENT AND FUEL RELOAD CHANGES FOR 1996 AND IMPROVED STANDARD TECHNICAL SPECIFICATIONS, R.

E.

GINNA NUCLEAR POWER PLANT Provided below are the summaries for meetings with the licensee to discuss two subjects:

1) steam generator replacement for 1996, and 2) Technical Specifications (TS) conversions.

Rochester Gas and Electric Corporation (RGKE) by letter of July 21, 1993, provided a summary of the discussion on the above subject meeting regarding steam generator replacement at Ginna station.

The staff concurs with the RGKE summary and is issuing this meeting summary for record purposes.

The steam generator replacement at Ginna Station is scheduled for the spring of 1996, and the conversion of the Ginna Station TS to a Standard Technical Specification (STS) format, will be accomplished concurrently.

RG&E presented to the NRC staff their appr oach and schedule in preparing safety analyses to support changeout of steam generators pursuant to 10 CFR 50.59.

RGKE indicated that the steam generator replacement will not constitute any unreviewed safety question consistent with 10 CFR 50.59.

RG8E described the methodology used to support the 10 CFR 50.59 evaluations regarding design basis accident analyses and the structural design basis.

Descriptions of the evaluations, modeling and analyses

process, as well as preliminary conclusions, were presented and discussed.

The staff recommended that RGKE closely examine the computer codes/models that will be used to ensure their applicability to Ginna.

The NRC cautioned RGLE to ensure that all design considerations for interfacing systems (primary and secondary safety valve sizing and leak-before-break technology were specifically called out) be adequately addressed by their review.

RGKE's presentation at the meeting also provided preliminary information concerning RGEE's int'ention for transition to an 18-month fuel cycle, starting with the spring 1996 reload.

Preliminary schedules for submittals to support the reload were discussed by RG8E.

The staff believed that adequate resources for NRC review could be provided.

,93121000S1 931202 PDR ADOCK 05000244 PDR III~ HLKC~ MV I

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December 2,

1993 The staff has had the experience of a TS conversion at Crystal River Unit 3 and Watts Bar.

Through the utility owners groups, the staff is working to improve the STS regarding the resolution of generic action items (e.g.,

the steam generator testing program, pressure/temperature limits, low temperature overpressure, battery and inverter limiting condition for operation and new 10 CFR Part 20 restructure).

Any future TS conversion would take advantage of these resolved items.

The NRC recommended additional meetings and further discussions as new information develops.

Enclosures:

1.

List of Attendees 2.

Horning Agenda and Discussion Material Original signed by:

Allen R. Johnson, Project Manager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next page

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!NAME'::;::';:::::"::::i PDI LA

) tie PDI-3: P KCotton:dt PDI-3:PH i AJohnson PDI-3:D WButler

'DATE"

:

93 ck cR 93 h~ <93 2 93 OFFICIAL R CORD COPY DOCUMENT NAME: A:iGIMTG2.SUH

December 2,

1993 The staff has had the experience of a TS conversion at Crystal River Unit 3 and Watts Bar.

Through the utility owners groups, the staff is working to improve the STS regarding the resolution of generic action items (e.g.,

the steam generator testing program, pressure/temperature limits, low temperature overpressure, battery and inverter limiting condition for operation and new 10 CFR Part 20 restructure).

Any future TS conversion would take advantage of these resolved items.

The NRC recommended additional meetings and further discussions as new information develops.

Enclosures:

1.

List of Attendees 2.

Horning Agenda and Discussion Haterial Allen

nson, Project Hanager Project Directorate I-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation cc w/enclosures:

See next page

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 Ms.

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.

Reynolds Winston

& Strawn 1400 L St.

N.W.

Washington, DC 20005-3502 Hs.

Thelma Wideman

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

Lyons, New York 14489 Ms. Mary Louise Heisenzahl Administrator, Monroe County Office of Emergency Preparedness 111 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

0 f

P r'

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

E.

GINNA NUCLEAR POWER PLANT STEAM GENERATOR REPLACEMENT - ACCIDENT ANALYSIS APPROACH JUNE 29, 1993 Allen Johnson Project Hana er Bob Eliasz Eugene J.

Domaleski Hartin Parece John F. Smith Brian J.

Flynn William D.

Maxham Bob Eckert Chris Schieck Mark Beaumont Bernard Carrick Hark D. Smith Geor e Wrobel Steven R. Jones Mark A. Caruso Frank Orr J.

Rajan USNRC NRR PDI-3 Rochester Gas and Electric RG&E Babcock

& Wilcox Nuclear Service BWNS BWNS RG&f RG&E BWNS Babcock

& Wilcox International BWI BWNS Westin house RG&E Bechtel RG&f USNRC NRR SPLB USNRC NRR SRXB USNRC NRR SRXB USNRC NRR EHEB

-,Enc1osure 2

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NRC-RG&E Meeting Steam Generator Replacement Analysis Approach Morning Agenda

==

Introduction:==

Overview:

Accident Analysis:

1996 Fuel. Reload:

G. Wrobel J. Smith B. Flynn R. Kliasz 29 June 93 Slide l

E ui ment Overview Steam Generators Ordered From B&W Detailed Engineering in Progress Major Components Ordered Forgings, Japan Steel Works Shell Plate, Creusot Loire Tubing, Sumitomo

. Licensing Support-Babcock dk Wilcox Nuclear Services First Tubesheet Arrives in Canada mid-July 1993 SiG Shipment mid-February 1996 29 June 93 Slide 2

'0 Installation Overview Contract Signed Week of June 21, 1993 Bechtel Construction to do Detailed Design and Installation Conceptual Design Completed by Bechtel in 1990 and Updated in 1992 Concept Involves Cutting One or More Openings in Containment Removal/Reinstallation Thru Openings Detailed Design to be Done in 1994 and 1995 Pre-Outage Modifications 1995 Installation Spring 1996 29 June 93 Slide 3

Steam Generator Designed to Minimize Analysis Impact

1. Primary Side PP < Existing Generator at 0% Plugging

2. Primary Side T,, Remains Unchanged 3.

PrimarylSecondary Volume Changes Minimized 4.

Level Changes Minimized 5.

Physical Envelope Unchanged 29 June 93 Slide 4

p; R.E. Ginna Steam Generator Replacement Comparison of Existing vs. Proposed Manuf./Model Existing W/44 Replacement (Preliminary)

BWI Primary Side Flow for above dp's 33 E06 lbm/hr Primary Side Pressure Drops (0% plug)

Nozzle inlet to Nozzle outlet 32.3 psi 32.3 psl 33 E06 Ibm/h Heat Transfer Areas 0% Plugging 15% Plugging 20% Plugging Tubing Outside'Diameter 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 131 cu. ft.

654.5 cu. ft.

131 cu. ft.

54000 sq. ft.

43200 sq. ft.

0.750 in 0.0431 in 4765 Alloy 690, TT 131.5 cu. ft.

710 cu. ft.

131.5 cu. ft.

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

Secondary Mass Flow, 100%

Steam Line Orifice Size Initial Steam Pressure, 100%

4580 cu. ft.

81200 Ibm 118300 lbm 3.3 E06 lbm/hr 4.37 sq. ft.

800 psia 4480 cu. ft.

84100 Ibm 114200 lbm 3.3 E06 lbm/hr 1.4 sq. ft..

875 psia 29 June 93 Slide 5

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Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RSG Required 15.1.1 Decrease in Feedwater Temperature MDNBR (Bounded)

H.T. Area

+22%

No 15.1.2 Increase in Feedwater Flow MDNBR (Bounded)

Flow Rate H.T. Area NC

+22%

No 15.1.3 Load Increase MDNBR (Bounded)

H.T. Area

+22%

No 15.1.5 Steam Line Break Core Response Break Area H.T, Area

-70%

+22%

Containment Response Break Area SG Inventory

- 70%

Yes 15.1.6 SG ARV Ec FW CV Failures Bounded by LOF and SLB No 29 June 93 Slide 7

Preliminary Review of UI'SAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RS6 Required 15.2.2 Loss of External Electrical Load PRI Press SEC Press H.T. Area MSSV Cap

+22%

NC Yes 15.2.5 Loss of Offsite AC Power PRI Press MDNBR (Bounded)

H.T. Area MSSV Cap.

+22%

NC No 15.2.6 Loss of Normal Feedwater PRI Press Sec Mass @ Trip D.H. Rem Cap Setpoint No 15.2.7 Feedwater Line Breaks MIN RCS Mass Sec Mass @ Trip Setpoint No 29 June 93, Slide 8

Preliminary Review of UI'SAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RSG Required 15.3.1 Single RCP Coastdown RCS Flow (SG DP)

No 15.3.1 Complete Loss of Forced RC Flow MDNBR RCS Flow (SG DP)

No 15.3.2 Locked Rotor Pins in DNB RCS Flow (SG DP)

No 29 June 93 Slide 9

Event Preliminary Review of UFSAR Safety Analyses Acceptance Important Criteria Parameters Analysis RSG Required 15.4.1 RCCA Withdrawal &om Subcritical MDNBR 15.4.2 RCCA Bank Withdrawal at Power MDNBR Rod Worth Kinetics COEF Rod Worth Kinetics COEF NC NC NC NC No No 15.4.3 Startup of an Inactive RC Loop Kinetics COEF Rod Position NC NC No 15.4.4 CVCS Malfunction-Boron Dilution Time to Critically CVCS Flow RCS Vol NC No 15.4.5 RCCA Ejection 15.4.6 RCCA Drop Fuel Enthalpy RCCA Worth Kinetics COEF RCCA Worth Kinetics COEF NC NC NC NC No No 29 June 93 Slide 10.

Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RSG Required 15.5 Inadvertent ECCS Operation at Power N/A (Low Head)

No 15.5 CVCS Malfunction-RC Inv. Increase PRI Press CVCS Flow PZR Level NC NC No 29 June 93 Slide 11

Event Preliminary Review of UFSAR Safety Analyses Acceptance Important Criteria Parameters Analysis RSG Required 15.6.1 Inadvertent Opening of a PSV/PORV PSV Area CVCS Flow NC NC No 15.6.2 Rad. Consequences of Small Lines Carrying RC Outside Containment Offsite Dose RCS Activity Leak Rate NC NC No 15.6.3 Steam Generator Tube Rupture

/

Offsite Dose RCS Activity Tube I.D.

SEC Volume NC

-14%

-2%

Yes 15.6.4.1 SBLOCA

- PCT (Bounded)

H.T. Area PRI Volume

+22%

+.5%

No 15.6.4.2 LBLOCA PCT RP Volume NC

+.5%

No 29 June 93 Slide 12

Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RSG Required 15.7 Rad. Releases

&om Subsystems Offsite Dose NC No 5.2.2 LTOP-RC Pump Restart LTOP-ECCS Actuation PRI Press PRI Press H.T. Area PORV Cap ECCS Flow

+22%

NC NC Yes No Station Blackout No Core Uncovery Seal Leak PRI Mass NC NC No 29 June 93 Slide 13

Analysis Summary Accidents that willbe Analyzed Loss of External Electrical LoadlTurbine Trip Main Steam Line Break Core Response Containment Profile Steam Generator Tube Rupture Low Temperature Overpressurization 29 Sune 93 Slide 14

Anal sis/Evaluation Schedule Overview Activit Model Preparation Analyze Containment Analyze Accidents Prepare UFSAR/TS Changes ifRequired Licensing Report (10CFR50.59 Evaluation)

Com letion 10l1/93 4l1/94 8/1/94 9/1/94 12/1/94 29 June 93 Slide 15

1996 Fuel Reload Current Contract Ends 1995 Reload Evaluating Proposals for 1996 Reload Proposals Received Westinghouse Siemens 29 June 93 Slide 16

Westin house Pro osal Performance +

0.400 in. OD Maximum Assembly Burnup:

Up to 55000 MWD/MTU ZIRLO Cladding Coated Cladding Through Bottom Grid DFBN Burnable Absorber-Zirconium Diboride Siemens Pro osal Custom Design 0.424 in. OD Maximum Assembly Burnup:

Up to 57,000 MWD/MTU Zircaloy Cladding DFBN Burnable Absorber-Gadolinia 29 June 93 Slide 17

Change Starting with 1996 Reload F~ increase from 1.66 to 1.70 F< increase from 2.32 to 2.50 Cycle Length Increase from Annual to 18 Months Possible Tavg Decrease up to 15'F 29 June 93 Slide 18

Ginna Accident Analysis 15.1 15.2 15.3 15.4 15.5 15.6 15.7,

, 15.8 6.2.1 5.2.2 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 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 Decrease in RCS Flowrate 15.3.1 Flow Coastdown Accidents 15.3.2 Locked Rotor Accident 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 Increase in RCS Inventory 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 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 Anticipated Transients Without Scram Chapter 6, Chapter 5 2 Containment Integrity Low Temperature Overpressurization 29 June 93 Slide 19

Analyses that willbe Updated with Reload 15.1.1 15.1.2 15.1.3 15.1.4 15.1.5 15.1.6 15.2.3 15.2.5 15.2.6 15.2.7 15.3.1 15.6.3 15.6.4.1 15.6.4.2 15.7.3 5.2.2 Decrease in Feedwater Temperature Increase in Feedwater Flow Excessive Load Increase Incident Inadvertent Opening of a SG RV SLB (Both core and MkE)

SG RV 8~, FW Control Valve Failure Loss of External Load/Turbine Trip Loss of Offsite Power to Station Aux Loss of Normal Feedwater Flow Feedwater System Pipe Breaks Flow Coastdown Accidents SG Tube Rupture SBLOCA LBLOCA Fuel Handling Accidents Low Temp. Overpressurization 29 June 93 Slide 20

Schedule Award Contract Preliminary Design Packages Safety Analysis Report Submit Analysis to NRC Scheduled Startup January 1994 January 1995 June 1995 September 1995 June 1, 1996 29 June 93 Slide 21

NRC-RGBK Meeting Steam Generator Replacement Analysis Approach Afternoon Agenda

==

Introduction:==

Overview:

G. Wrobel J.

Smith Component Structural Analysis:

B. Carrick Overview Analysis Methodology Hydraulic Analysis Structural/Component Loadings 29 June 93 Slide l

E ui ment Overview Steam Generators Ordered From BdkW Detailed Engineering in Progress Major Components Ordered Forgings, Japan Steel Works Shell Plate, Creusot Loire Tubing, Sumitomo Licensing Support-Babcock dk: Wilcox Nuclear Services First Tubesheet Arrives in Canada mid-July 1993 S/G Shipment mid-February 1996 29 June 93 Slide 2

Installation Overview Contract Signed Week of June 21, 1993 Bechtel Construction to do Detailed Design and Installation Conceptual Design Completed by Bechtel in 1990 and Updated in 1992 Concept Involves Cutting One or More Openings in Containment Removal/Reinstallation Thru Openings Detailed Design to be Done in 1994 and 1995 Pre-Outage Modifications 1995 Installation Spring 1996 29 June 93 Slide 3

Structural Evaluation of Effected Components A Systems 29 June 93 Slide 4

OBJECTIVE Demonstrate Acceptable Structural Response Following S/G Replacement 29 June 93 Slide 5

GOALS Demonstrate As-Built Supports Are Acceptable Minimize Changes to Design Basis Calcs 29 June 93 Slide 6

39 MAIN STEAM LINE STEAM GENERATOR IA 38 MAIN STEAM LINE STEAM GENERATOR I

REACTOR COOLANT PUMP 18 II FEEOWATER LINE

)4 FEEOWATER LINE UPPER SUPPORTS ANO SNUBBERS {TYP.)

COLO LEG HOT LEG HOT LEG INTERMEOIATE SUPPORTS (TYP.)

CROSSOVER LEG COLO LEG LOWER SUPPORTS ITYP.)

REACTOR COOLANT PUMP IA REACTOR VESSEL VIEW OF NSSS SYSTEM FOR GINNA NUCLEAR STATION

METHODS Preferred:

Confirm Current Analysis Bounds Expected

Response

i.e. Loadings/Response no Greater than.

at Present Alternate:

Perform Analysis to Show Expected Response is within Current Acceptance Criteria 29 June 93 Slide 8

Anal sis Methodolo Evaluate Hydraulic Transients Develop StructurallComponent Loadings Evaluate Resulting Stresses 29 June 93 Slide 9

Develop Structural Model (BW Span)

Analysis Methodology Develop Hydraulic Model (Craft)

Benchmark Model Benchmark Model Evaluate Hydraulic Transient Develop Loadings (Old/New)

Old Yes Bound

'7 Done No Evaluate Stresses (New Loads)

Modify Models as Required Stresses Acceptable

?

No Propose Modification 29 June 93 Slide 10 Yes Done

H draulic Anal sis Primar Pi e Breaks-Leak-Before-Break Methodology Surge Line RHR SI Secondar Pi e Breaks High Energy Line Break Criteria Main Steam Feedwater 29 June 93 Slide 11

H draulic Anal sis H draulic Transients C

Com uter Code - CRAFT Results - Internal Forces

- External Jet Loads

- Mass-Energy (M/E) Release 29 June 93 Slide 12

H draulic Anal sis Com artment Anal sis Based on M/K Release Com uter Code - COMPAR2 Results-Asymmetric Cavity Pressure/Loads 29 June 93 Slide 13

Hydraulics Loop

Model, Craft Building Model Com are Generator Model Craft Internal Force Time Histories Energy Asymmetric Cavity Delta P Mass 8 Energy Generator Force Time Histories Surge Line RHR Safety Inj.

Jet Imp.

To Structural Anal sis Steam Line Feedwater Jet Imp.

Structural/Com onent Loadin s

Models BWSPAN Old S/G New S/G Basemat Up BWSPAN Structural/Piping Beam Elements Linear Static/Dynamic 29 June 93 Slide 15

Structural/Com onent Loadin s

In ut Loads Deadweight Thermal Seismic Hydraulic Internal Forces Jet Impingement Asymmetric Cavity 29 June 93 Slide 16

Structural/Com onent Loadin s

Anal sis Resulting Stresses Stress Combinations Controlling Stresses Compare to:

1-Existing Analysis Results 2-Current Allowables 29 June 93 Slide 17

Old Gen.

Orig. Meth.

Structural Model Loop BWSPAN New Gen.

Orig. Meth.

New Gen.

improved Meth.

65 E0 LL D.W.

Thermal Expansion Seismic High Energy line Break (LOCA)

Compare Normal Operating Seismic LOCA FaUlted ModifyStress Re orts Report

EVALUATIONS ULE Assembly of Inputs Structural Model Analy. Sehmic DWTherm old -

old new new Gen. new en.

High Energy Break Loads Comparison/Write off Hydraulic Model Loop Break Steam/Full Break Gen. F(t)

Bulldlng Assy. Delta P's Support Embedment Review Margin Evaluabon Thermal (Tavg.) Evaluabon 1993 1994 JulAug SeptOct Nov Dec Jan Feb Mar Apr May Jun JulAug Sept Oct Nov

, DISTRIBUTION:

Docket= File=

'- NRC

& Local PDRs PDI-3 Read'.ng THurley/FHiraglia LJCallan SVarga JCalvo WButler AJohnson KCotton SLittle OGC EJordon SJones

HCaruso, 8E23

FOrr, 8E23

JRajan, 7E23 RLo, llE22

LHarsh, 13D18

CGrimes, llE22

AHansen, 13E21

RLaufer, 10D22 MGamberoni, 13D18 CHoon, llE22 ACRS (10)

VMcCree, EDO JLinville, RI

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