ML20149K754
| ML20149K754 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/10/1988 |
| From: | Wagner D Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8802230440 | |
| Download: ML20149K754 (30) | |
Text
d' February 10, 1988 Docket Nos. 50-266 _ DISTRIBUTION:
and 50-301 S' pocket.Filesa JPartlow NRC~& l.6c~al PDRs ACRS(10)
PDIII-3 r/f KPerkins LICENSEE: WISCONSIN ELECTRIC POWER COMPANY DWagner OGC-WF1 FACILITY: P0 INT BEACH NUCLEAR PLANT, UNITS 1 AND 2 EJordan
SUBJECT:
SUMMARY
OF FEBRUARY 4, 1988 MEETING At the request of the licensee, a meeting was held on February 4,1988 in the Rockville, Maryland offices of the NRC to discuss the licensee's accident reanalysis to support increased allowable core peaking factors at Point Beach 1 and 2. A list of meeting attendees is included in Enclosure 1. The slides used during the presentation are included in Enclosure 2. These slides discuss the goals of the reanalysis effort, the strategy involved in achieving these goals and the potential Final Safety Analysis Report / Technical Specification changes resulting from the reanalysis.
~
DU5TiaTsigTei! gg David H. Wagner, Project Manager Project Directorate III-3 Division of Reactor Projects
Enclosures:
As stated cc: See next page Office: LA PD 11-3 PFF/FDIII-3 PD/PDIII-3 N
Surname: P r DWagner/tg KPerkins Date: 0 j/88 02/1o/88 02//o/88 8802230440 G80210 PDR ADOCK 05000266 P PDR
Mr. C. W. Fay Point Beach Nuclear Plant Wisconsin Electric Power Company Units 1 and 2 cc:
Mr. Rruce Churchill, Esq. .
Shaw, Pittman, Potts and Trowbridge 2300 N Street, N.W.
Washington, DC 20037 Mr. James J. Zach, Manager Point Beach Nuclear Plant Wisconsin Electric Power Company 6610 Nuclear Road Two Rivers, Wisconsin 54241 Town Chairman Tcwn of Two Creeks Route 3 Two Rivers, Wisconsin 54241 Chairman Public Service Comission of Wisconsin Hills Farms State Office Building Madison, Wisconsin 53702 Regional Administrator, Region III U.S. Nuclear Regulatory Commist ton Office of Executive Director for Operations 799 Roosevelt Road Glen Ellyn, Illinois 60137 Resident Inspector's Office U.S. Nuclear Regulatory Comission 6612 Nuclear Road Two Rivers, Wisconsin 54241
ENCLOSURE 1 ACCIDENT REANALYSIS TO SUPPORT INCREASED ALLOWABLE CORE PEAKING FACTORS AT POINT BEACH 1 AND 2
~
LIST OF ATTENDEES NAME AFFILIATION David Wagner NRC Summer Sun NRC Lambros Lois NRC Y. Gene Hsii NRC Wayne Hodges NRC Randall Fieldhack WisconsinElectric(WE)
Harve Hanneman WE Bob LaMuro Westinghouse (W)
Janelle Ivey W Beth Hall E n
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O
I ENCLOSURE 2 )
l 1
1
- l NRC/WE/ WESTINGHOUSE MEETING '
I 9"
ACCIDENT REANALYSIS TO SUPPORT INCREASED ALLOWABLE CORE PEAKING FACTORS AT_
POINT BEACH NUCLEAR PLANT UNITS 1 AND 2 l
l l
l February 4, 1988 Harv Hanneman (WE)
Bob LaMuro (W) g Randy Fieldhack (WE) ~
I n
v
MEETING AGENDA INCREASED ALLOWABLE CORE PEAKING FACTORS POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 ,
February 4, 1988 Introduction / Schedule WE I.
II. Non-LOCA Transients A. Scope of Analyses & Evaluations W B. Status of Generic WCAP Approval NRC
- 1. WCAP-ll397 Revised Thermal Design Procedure (RTDP)
- 2. WCAP-ll394 WOG Dropped Rod Generic Methodology III. Small-Break LOCA A. NOTRUMP Analysis WE B. Reanalysis Strategy WE IV. Possible Technical Specification /FSAR Changes WE A. Increased Peaking Factors B. Part-length Hafnium Absorber Assemblies 3
Thimble Plug Removal f C.
D. Increased Steam Generator Tube Plugging E. Enhanced OFA Fuel Features ,
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PURPOSE OF MEETING o Present WE's Reasons, Strategy, and Schedule for Requesting a License Amendment to Increase Allowable Core Peaking Factors (F g and F H) at Point Beach o Cbtain Feedback from the NRC Staff on Strategy and Schedule so WE Can Proceed With Desired Fuel Cycle Denign and Implementation (Unit 1, Cycle 17 - Spring 1983 Refueling)
. o Resolve Break-Spectrum Issue for Small-Break LOCA Analysis c ,.
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ACCIDENT REANALYSIS AT INCREASED PEAKING FACTORS GOALS: o Implement Low-Low Leakage Loading Pattern (L4P) Fuel Management Scheme with Part-Length Non-Burnable Absorbers in Peripheral Assemblies to Reduce Reactor Vessel Neutron Fluence
- Enhance Ability for Additional Twenty (20)
Years of Vessel Life
- Address Current Reactor Vessel Safety Issues (PTS, Low Upper Shelf Material Toughness, Heatup and Cooldown Restrictions) and Uncertainty in Future NRC Requirements (e.g., Proposed Rev. 2 to R.G. 1.99) o Implement L4P and Enhanced OFA Fuel Features (e.g., Axial Blanketsj High Burnup) to Reduce Fuel Cycle Costs o Improve Operational Flexibility
- Removal of Thimble Plug Devices
- Increased Steam Generator Tube Plugging SCHEDULE: Implement L4P Reload Designs As Follows:
.4 o Unit 1 - Spring 1989 (Order Fuel - April 1988) o Unit 2 - Fall 1989 (Order Fuel - September 1988) .
A L=
l 1
l INCREASED PEAKING FACTOR STRATEGY .
l o Reanalyze Limiting Transients First (2/15/88) l
- Dropped RCCA (Ph yg)
- Rod Ejection (Fg) o Complete Reanalysis / Evaluation Effort for Remaining Accidents Including SBLOCA (6/30/88)
I i
o Coordinate with Upper Plenum Injection /Best-Estimate l LOCA Analysis Project o Request License Amendment (i.e., Tech Spec Revision) to Allow Increased Peaking Factors (8/31/88):
- 1. r on - 1.70
- 2. Fq = 2.50
- 3. Certain RPS Settings l
- Overtemperature edt Overpower AT l
- 4. Control Rod and Power Distribution Limits I
- 5. DNB Safety Limit Curves
- 6. Non-Burnable Absorber Assemblies h
o Reload Core 10CFR50.59 Safety Evaluation (April 1989) .
- Removal of Thimble Plug Devices .
- Insertion of Part-Length Non-Burnable Absorbers in 3 c,
Peripheral Fuel Assemblies I.
l
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REQUIRED ANALYSES / EVALUATIONS FOR LICENSE AMENDMENT, l
l o LOCA Reanalysis at Higher F g of 2.50 f
- Large Break - Best-Estimate WCOBRA/ TRAC
- Small Break - NOTRUMP l
o FSAR Chapter 14 Accident / Transient l
Reanalysis / Evaluations at Higher F,g of 1.70 (Rod Ejection at F g of 2.50)
- Revised Thermal Design Procedure (RTDP)
(WCAP-ll397 Submitted 3/16/87) ;
1
- WOG Dropped Rod Generic Methodology j (WCAP-11394 Submitted 5/22/87) l
- Operational Considerations - Maintain 2000 psia, Up l to 14% SG Tube Plugging, Eliminate Thimble Plugs
- OFA Fuel Enhancements - Higher Burnup, Axial Blanket, IFBAs, DFBN 7..
1 1
1 POINT BEACH INCREASED PEAKING FACTOR EFFORT -
NON-LOCA CONSIDERATIONS l
l AREAS OF IMPACT INCREASE IN ALLOWABLE F dH DNB - CRITERIA EVENTS 0 CORE LIMITS /SETPOINTS 0 FLOW TRANSIENTS 0 DROPPED ROD (WCAP 11394)
INCREASE IN ALLOWABLE F g FUEL / CLAD TEMPERATURE CRITERIA EVENTS THIMBLE PLUG REMOVAL INCREASE CORE BYPASS FLOW RTDP: REVISED THERMAL DESIGN PROCEDURE (WCAP 11397)
CORE LIMITS /SETPOINTS a*.n s
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s EVENTS REQUIRING NEW ANALYSIS FOR INCREASED PEAKING FACTORS EVEt4T Core Limits ESAR Edh Eg Bvoass OTDT/0PDT Setpoint Calculation X X Rod Withdrawal at Power 14.1.2 X Rod Withdrawal from Subcritical 14.1.1 X Dropped Rod 14.1.3 X Boron Dilution at Power 14.1.4 X at Startup at Shutdown at Refueling Startup of an Inactive Loop 14.1.5 X Reduction in Feedwater Enthalpy 14.1.6 Excessive Load Increase 14.1.7 X Loss of Flow 14.1.8 X Locked Rotor 14.1.8 X X Loss of Load 14.1.9 X Loss of Normal Feedwater 14.1.10 X Station Blackout 14.1.11 X Steamline Break, Core Response 14.2.5
, Containment Response 14.2.5 Rod Ejection 14.2.6 X a
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NOTRUMP ANALYSIS .
l o TMI Requirements
- NOTRUMP Description l WCAP - 10079-P-A, "NOTRUMP: A Nodal Transient J l
Small Break and General Network Code" l
WCAP - 30054-P-A, "Westinghouse Small Break ECCS l i
Evaluation Model Using the NOTRUMP Code" l
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- Generic Study WCAP - lll45-P-A, "Westinghouse Small Break LOCA ECCS Evaluation Mode.1 Generic Etudy With The NOTRUMP Code" o Wisconsin Electric Status
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GENERIC STUDY - WCAP-lll45-P-A -
Assurptions: 2-Loop Plant Downflow Barrel-Baffle Configuration 0% Steam Generator Tube Plugging 1709 MWt Core Power 2.32 F g Results: Break Size NOTRUMP WFLASH i
(inches) (PCT) (PCT) 3 NCU* 1132 4 796 1713 6 757 1547 l
- No Core Uncovery
,i 4
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9
s WFLASH 74 EM e Earlier Point Beach Analysis:
- Assumptions: Standard Fuel 1518 MWt Core Power 2.32 F q Downflow Barrel-Baffle Configuration
- Results: Break Size WFLASH (Inches) (FCT) 3 1469.
4 1688.
6 1167.
e Earlier Prairie Island Analysis:
- Assumptions: Standard Fuel 1709 MWt Core Power 2.32 F q Downflow Barrel-Baffle Configuration Break Size WFLASH
- Results:
(Inches) (PCT) 3 1152.
4 1713.
6 1547.
e
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3
s WFLASH - OCT 75 EM o Current Point Beach Analysis
- Assumptions: OFA Fuel 1518 MWt Core Power 2.32 F g 44F Steam Generators Downflow Barrel-Baffle Configuration
- Results: Break Size WFLASH (Inches) (PCT) 4 890.
6 992.
8 750.5 o Kori-l Analysis
- Assumptions: OFA Fuel 1723.5 MWt Core Power 2.32 F 9
Upflow Barrel-Baffle Configuration
- Results Break Size WFLASH (Inches) (PCT) 4 1163.
6 1375.
8 1286.
o R.E. Ginna Analysis
- Assumptions: OFA Fuel 1520 MWt Core Power 2.32 F g Downflow Barrel-Baffle Configuratio3
- Results: Break Size WFLASH (Inches) (PCT) [
4 976 6 1092 8 758 .,
s NOTRUMP o Generic Analysis:
- Assumptions: 1709 MWt Core Power 2.32 F g 0% Steam Generator Tube Plugging Downflow Barrel-Baffle Configuration
- Results Break Size NOTRUMP (Inches) (PCT) 3 NCU*
4 796.
6 757.
o Prairie Island Analysis:
- Assumptions: 1650 MWt Core Power ,
l 2.5 F g 10% Steam Generator Tube Plugging Downflow Barrel-Baffle Configuration K(Z) Third Line Segment Removal i 1
- Results: Break Size NOTRUMP (Inches) (PCT) l 3 NCU* , [
4 1000, 6 NCU*
- No Core Uncovery 4
-A
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s COMPARISON OF POINT LEACH AND PRAIRIE ISLAND
- NOTRUMP INPUT ASSUMPTIONS POINT BEACH PRAIRIE ISLAND 1518 MWt Core Power 1650 MWt Core Power 2.50 F g 2.50 F g 1.70 SAH 0 $aH l K(Z) Third Line Removal K(Z) Third Line Removal 25% SGTP 10% SGTP Upflow configuration Downflow Configuration l
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SBLOCA Reanalysis Strategy o Select Expected Limiting Break Size 4" Cold Leg Break ,
o Analyze This Break Using NOTRUMP o Evaluate For Acceptable Results Criteria: 1. PCT I 1600 F
- 2. Core Uncovery 1
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BASIS OF PCT CRITERION A 1600 F PCT Cutoff Will:
- Ensure PCT for All Break Sizes Fall Below the 2200 F Limit:
Two-Loop Plant SBLOCA Analyses Results Max. A PCT Between Model Break Sizes WFLA511 74 EM 600 F WFLASH 75 EM 340 F NOTRUMP 40 F
- Prevent Any Significant Zirc-Water Reaction
- Ensure Significant Margin to the 2200 F Limit )
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POTENTIAL FSAR/ TECH SPEC CilA'GES o Increased Peaking Factors
- Technical Specification Changes
- 1. Core Safety Limits
- 2. OPAT, OTAT
- 3. Rod Insertion Limits
- 4. F O' AH
- 5. Third Line Segments, K(Z) Curve
- 6. RAOC Band
- Accident Analyses - Chapter 14 FSAR o Part-Length Absorber Assemblies - Tech Spec Design Features o Thimble Plugs - FSAR Change I
o Increased Steam Generator Tube Plugging - Future License Amendment o Enhanced OFA Fuel Features
- IFBA/ Axial Blankets - Add to Tech Specs, FSAR
- Debris Filter Bottom Nozzles - 10CFR50.59 Review
- Removable Top Nozzles / Higher Burnup - 10CFR50.59 Review u 0
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FIGURE 15.3.10-1 l CONTROL BANK INSERTION LIMITS POINT BEACH UNITS 1 AND 2 U60 1
(74g) 100 :. -
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0 20 (34%)40 60 80 100 l POWER LEVEL (t OF RATED POWER) o
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Unit 1 Amendment No. 86 May 22, 1985 Unit 2 Amendment No. 90 October 5, 1984 g.- Q .' 3' M ' . , . +
% MV a&'
B. Power Distribution Limits
- 1. a. Except during low power physics tests, the hot channel factors defined in the basis must meet,the following limits":
F9 (Z)1(2.21) x K(Z) for P> 0.5 P
F9 (Z)14.42 x K(Z) for P5 0.5 N
F <l.58 x [1 + 0.3 (1-P))
AH Where P is the fraction of full power at which the core is operating, K(Z) is the function in Figure 15.3.10-3 and Z is the core height location of F .
q b.
Following a refueling shutdown prior to exceeding 90 percent of rated power and at effective full power monthly intervals there-after, power distribution maps using the moveable incore detector system shall be made to confirm that the hot channel factor limits l are satisfied.
The measured hot channel factors shall be increased in the following way:
l (1) The measurement of total peaking factor, F eas , shall be increased by three percent to account for manufacturing tolerances and further increased by five percent to account for measurement error.
(2) The measurement of enthalpy rise hot channel factor, F q,
shall be increased by four percent to account for measurement error.
- c. If a measured hot channel factor exceeds the full power limit of Specification 15.3.10.B.1.a. the reactor power and power range high setpoints shall be reduced until those limits are met. If
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subsequent flux mapping cannot, within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, demonstrate that the full power hot channel factor limits are met, the overpower Unit 1 Amendment No. 86 May 22, 1985 '
Unit 2 Amendment No. 90 15.3.10-2 October 5, 1984 '
FIGURE 15.3.10-3 POINT BEACll UNITS 1 AND 2 i
\
HOT CHANNEL FACTOR NORMALIZED OPERATING ENVELOPE ,
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0 2 4 6 8 10 12 Core Height (Ft) v4 Unit 1 Amendownt No. 86 May 22, 1985 Unit 2 Amendment No. 90 October 5, 1984 -
4
P FIGURE 15.3.10-4 FLUX DIFFERENCE OPERATING ENVELOPE POINT BEACH UNITS 1 AND 2 -
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