ML17263A332
| ML17263A332 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/21/1993 |
| From: | Mecredy R ROCHESTER GAS & ELECTRIC CORP. |
| To: | Andrea Johnson NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9307290199 | |
| Download: ML17263A332 (58) | |
Text
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A.CCEjLERAT. D DOCGNIKNTDISTRIBUTIONSYSTEM REGULA'I INFORMATION DISTRIBUT101YSTEM (RIDE)
ACCESSION NBR:9307290199 DOC.DATE: 93/07/21 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 6 Electric Corp.
RECIP.NAME RECIPIENT AFFILIATION JOHNSON,A.R.
Project Directorate I-3
SUBJECT:
Summarizes 930629 public meeting w/NRC re effort to evaluate SG replacement at plant,per 10CFR50.59 w/regard to util design basis accident analysis a structural design basis.
Matl presented at meeting encl.
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ROCHESTER GAS AND ELECTRIC CORPORATION o
89 EAST AVENUE, ROCHESTER N.K 14649-0001 ROBERT C MECREDY Vice Prerident Cinna Nudear Production July 21, 1993 TELEPHONE AREA CODE 71B 546 2700 U.S. Nuclear Regulatory Commission Document Control Desk Attn:
Allen R. Johnson Project Directorate I-3 Washington, D.C.
20555
Subject:
Steam Generator Replacement and Fuel Reload Changes for 1996 R.E.
Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Johnson:
On June 29, 1993 Rochester Gas
& Electric and members of the Reactor Systems Branch (RSB), Mechanical Engineering Branch (MEB),
and Project Directorate I held a public meeting to discuss our efforts in evaluating the Steam Generator Replacement at Ginna
- Station, scheduled for the spring of 1996.
This letter provides a
summary of the discussion held at the meeting.
As previously discussed, RG&E believes that the replacement will not constitute an unreviewed safety question.
Consequently, consistent with 10CFR50.59, the replacement is planned to be performed without prior NRC approval.
The purpose of the meeting was to describe the methodology we are using to support the 10CFR50.59 evaluation as it pertains to our Design Basis Accident Analysis and our structural design basis.
Descriptions of the evaluation, modeling and analysis process as well as preliminary conclusions were presented and discussed.
Attachment A to this letter provides copies of,the material presented.
While neither branch expressed concerns with the methodology or preliminary conclusions, both recommended that we closely examine the computer codes/models that will be used to ensure their applicability to Ginna.
We were also cautioned 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 our review.
It was recommended that the Materials Branch be included in future discussions.
The meeting also provided preliminary information concerning our intention to transition to eighteen month fuel
- cycles, starting with the spring 1996 reload.
Preliminary schedules for submittals to support the reload were discussed.
The RSB felt that adequate time for NRC review was provided by the proposed schedules, provided no new analysis methodology review was 280l..";.";
9'307290i99'3072i PDR ADOCK 05000244 ark~ p,-g~ J.y
required. If the latter were the case more time would be required.
Both parties agreed that additional meetings/teleconferences would be beneficial as details become more firm.
RGGE appreciates the valuable input and cooperation that was provided at this meeting.
Very truly yours, Robert C. Mecre BJFi289 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 Ginna Senior Resident Inspector
Attachment A
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. Eliasz 29 June 93 Slide 1
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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 A Wilcox Nuclear Services First Tubesheet Arrives in Canada mid-July 1993 S/G Shipment mid-February 1996 29 June 93 Slide 2
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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
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Steam Generator Designed to Minimize Analysis Impact
- 1. Primary Side AP < Existing Generator at 0'/0 Plugging
- 2. Primary Side T,, Remains Unchanged 3.
Primary/Secondary Volume Changes Minimized 4.
Level Changes Minimized 5.
Physical Envelope Unchanged 29 June 93 Slide 4
L I
A
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 ibm/hr Primary Side Pressure Drops (0% plug)
Nozzle inlet to Nozzle outlet 32.3 psi 32.3 psi 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. A.
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 ibm 3.3 E06 ibm/hr 4.37 sq. ft.
800 psia 4480 cu. ft.
84100 ibm 114200 ibm 3.3 E06 ibm/hr 1.4 sq. ft.
875 psia 29 June 93 Slide 5
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Accident Evaluation Methodology Define Critical Parameters for each Accident Analysis Develop Plant Model With OSG Compare RSG to OSG Validate Plant Model No Could Change Affect Approach
, to Acceptance Criteria Yes Develop Plant Model With RSG Perform Accident Analysis with OSG; with RSG Containment Analysis No Does State Point Move in an Unfavorable Direction?
Perform DNB Calculation Develop Contempt Model Calculate Containment Pressure Calculate Mass 8c Energy Document Evaluation
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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%
No Containment Response Break Area SG Inventory
- -70%
Yes 15.1.6 SG ARV &, FW CV Failures Bounded by LQF and SLB No 29 June 93 Slide 7
Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RSG Required 15.2.2 Loss of External Electrical Load 15.2.5 Loss of Offsite AC Power PRI Press SEC Press PRI Press MDNBR (Bounded)
H.T. Area MSSV Cap H.T. Area MSSV Cap.
+22%
NC
+22%
NC Yes 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
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Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RS0 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 RS6 Required 15.4.1 RCCA Withdrawal &om Subcritical MDNBR Rod Worth Kinetics COEF NC NC No 15.4.2 RCCA Bank Withdrawal at Power MDNBR 15.4.3 Startup of an Inactive RC Loop Rod Worth Kinetics COEF Kinetics COEF Rod Position NC NC NC NC No No 15.4.4 CVCS Malfunction-Boron Dilution Time to Critically CVCS Flow RCS Vol 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.
I I4
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 r 15.6.4.1 SBLOCA
- PCT (Bounded)
H.T. Area PRI Volume
+22%
+.5%
No 15.6.4.2 LBLOCA PCT AP Volume NC
+.5%
No 29 June 93 Slide 12
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Preliminary Review of UFSAR Safety Analyses Event Acceptance Important Criteria Parameters Analysis RS6 Required 15.7 Rad. Releases from Subsystems Offsite Dose 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
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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 June 93 Slide 14
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Anal sis/Evaluation Schedule Overview Activit Model Preparation Analyze Containment Analyze Accidents Prepare UFSAR/TS Changes ifRequired Licensing Report (10CFR50.59 Evaluation)
Com letion
~
10/1/93 4/1/94 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
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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 ofNormal 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 29 June 93 Slide 19
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I, 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 MATE)
SG RV A 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-RGAE 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 1
E ui ment Overview Steam Generators Ordered From BAW 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 8echtel 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 Coinponents A Systems 29 June 93 Slide 4
OBJECTIVE Demonstrate Acceptable Structural Response Following S/6 Replacement 29 June 93 Slide 5
GOALS Demonstrate As-Built Supports Are Acceptable Minimize Changes to Design Basis Calcs 29 June 93 Slide 6
38 MAIN STEAM LINE STEAM GENERATOR IA 38 MAIN STEAM LINE STEAM GENERATOR I
REAC'IOR COOLANT PUMP IB 14 FEEOWATER LINE l4 FEEOWATER LINE UPPER SUPPORTS ANO SNUBBERS
{TYP ~
)
COLO LEG HOT LEG HOT LEG INTERMEOIATE SUPPORTS
{TYP.)
CROSSOVER LEG COLO LEG LOWER SUPPORTS (TYP.)
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
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Anal sis Methodolo Evaluate Hydraulic Transients Develop StructuraVComponent 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 Loads Yes Bound Done No Evaluate Stresses (New Loads)
Modify Models as Required Stresses Acceptable 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 Com uter Code - CRAFT Results - Internal Forces
- External Set Loads
- Mass-Energy (M/E) Release 29 June 93 Slide 12
H draulic Anal sis Com artment Anal sis Based on M/E 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 Mass 8 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 Feed water Jet Imp.
Structural/Com onent Loadin s
Models BWSPAN 01d 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
Deadweight Thermal Seismic Hydraulic In ut Loads 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 BNfSPAN New Gen.
Orig. Meth.
New Gen.
Improved Meth.
(Q U
E0 LL D.W.
Thermal Expansion Seismic High Energy Line Break (LOCA)
Compare Normal Operating Seismic LOCA Faulted ModifyStress Re orts
~ Report
STRU EVALUATIONS Assembly of inputs Structural Model Analy. Seismic DWTherm old -
old new new Gen. new Gen.
High Energy Break Loads Comparison/Write off Hydraulic Model Loop Break Steam/Full Break Gen. Fg)
Building Assy. Delta P's Support Embedment Review Margin Evaluation Thermal (Tavg.) Evaluation 1993 1994 JulAug SeptOct Nov Dec JanFeb Mar Apr May Jun JulAug Sept Oct Nov