ML17278A320
| ML17278A320 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 07/26/1985 |
| From: | Bradfute J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-60804, NUDOCS 8508060425 | |
| Download: ML17278A320 (57) | |
Text
Docket No.
50-397 JUL p 6 1985 8508060425 850726 PDR ADOCK 05000397 P
PDR APPLICANT:
Washington Public Power Supply System (WPPSS)
FACILITY:
WPPSS Nuclear Project No.
2 (HNP-2)
SUBJECT:
WNP-2 REFUELING OVERVIEW - l1EETING MINUTES The Supply System requested an opportunity to present and discuss the initial plans and preparations for its forthcoming Cycle 2 Core Reload currently scheduled for spring l986.
The meeting was held in Bethesda on June 27, and included representatives from the Supply System, Exxon Nuclear, the Supply System's attorney and the NRC.
The attendees are listed in Enclosure l.
The Supply System presented the general scope of their reload program.
A copy of their viewgraphs is presented as Enclosure 2.
Technical and schedular issues were discussed in some detail and no significant difficulties were evident.
The Supply System's proposed schedule was believed to be tight but achievable.
The technical issues appear to be properly understand.
DISTRIBUTION L~il NRC PDR LPDR PRC System NSIC LB82 Reading
- Attorney, OELD WButler JBradfute EHylton DL:M DL:L852 JBradfute:tm WButler 07/zW/85 07/~(/85 brieteal sKened byt John 0. Bradfute, Project t1anager Licensing Branch No.
2 Division of Licensing
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o~ R<av P0 Cy A
O
/p +**y4 Docket No.
50-397 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555
~JUL 26 $85 APPLICANT:
Washington Public Power Supply System (WPPSS)
FACILITY:
WPPSS Nuclear Project No.
2 (WNP-2)
SUBJECT:
WNP-2 REFUELING OVERVIEW - MEETING MINUTES The Supply System requested an opportunity to present and discuss the initial plans and preparations for its forthcoming Cycle,2 Core Reload currently scheduled for spring 1986.
The meeting was held in Bethesda on June 27, and included representatives from the Supply System, Exxon Nuclear, the Supply System's attorney and the NRC.
The attendees are listed in Enclosure l.
The Supply System presented the general scope of their reload program.
A copy of their viewgraphs is presented as Enclosure 2.
Technical and schedular issues were discussed in some detail and no significant difficulties were evident.
The'Supply System's proposed schedule was believed to be tioht but achievable.
The technical issues appear to be properly understand.
Enclosures:
As stated John 0. Bradfute, Project Manager Licensing Branch No.
2 Division of Licensing cc:
See next page
e t.
Hr. G. C. Sorensen, Hanager Washington Public Power Supply System WPPSS Nuclear Project No. "
(h'NP-2)
CC:
Nicholas Reynold, Esquire
- Bishop, Cook, Liberman, Purcell 5 Reynolds 1200 Seventeenth
- Street, N.
W.
Washington, D. C.
20036 Hr. G.
E. Doupe, Esquire Washington Publ'ic Power Supply System P. 0.
Box 968 3000 George Washington Way Richlard, Washington 99532 hr. Curtis Eschels, Chai> man Energy F'acility Site Evaluation Council tiail Stop PY-ll Olympia, Washinoton 98504-P. L. Powell, Licensing Hanager Washingtcr, Public Power Supply System P. 0.
Box 968, HD 956B Richland, Washington 99352 Hr.
W. G.
Conn Burns and Roe, Incorporated c/o Washington Public Power Supply System P. 0. Box 968, HD 994E Richland, Washington 99352 R. B. Glasscocl', Director Licensing and Assurance Washington Public Power Supply System P. 0.
Box 968, HD 280
- Richland, Wa hin9xon 99352 Hr. C. H. Powers WNP-2 Plant Hanager Washington Public Power Supply System P. 0.
Box HD 927H
~
Richland, Washington 99352 Regional Administrator, Region V
U.S. Nuclear Regulatory Commission 1450 Haria Lane, Suite 210 Walnut Creek, California 94596 Hr. James N. Horgan, Hanager Customer Service Engineering EXXON Nuclear Company, Inc.
600-108th
- Avenue, N.W. P.O.
90777
- Bellevue, Washington 98009
WNP 2 REFUELING f1EETING 27 JUNE 1985 ATTENDANCE RECORD Enclosure l
Wayne Hodges Jim l'lorgan Daniel Fieno Howard Riching Dick Collingham George Schwenk Walter L. Brooks Dave L. Larkin Sanfrod Hartman H. L. Aeschlman H. J.
Campagnone S. L.
WU John 0. Bradfute ORGANIZATION NRC/RSB Exxon Nuclear NRC/CPB HRC/CPB Exxon Huclear NRC/CPB HRC/CPB WPPSS BLCPSR WPPSS NRC/DL NRC/CPB HRC/DL TELEPHONE (301 492-9410 (206) 453-4399 (301) 492-9474 (301) 492-9418 (509) 375-8392 (301) 492-9421 (301) 492-9498 (509) 372-5574 (202) 857-9823 (509) 372-5389 (301) 492-9476 (301) 492-9476 (301) 492-9779
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Enclosute 2
WNP-2 CYCLE 2 RELOAD OVERVIEW JONE 27, 1985 LARRY AEsCHLIMANg WPPSS DAVE LARKINg WPPSS BILL WOLKENHAUERi WPPSS JIM NORGAN ENC DICK COLLINGHAMi ENC SANFORD HARTMANi 'BLCPRR LICENSING NUCLEAR FUEL NUCLEAR FUEL
II
il c
4 I
0
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0 BESCRIBE PLANS AND SCHEDULE FOR RELOAD FUEL
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IDENTIFY EXPECTED I'INP-2 CYCLE 2 LICENSE MODI F ICATIONS
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."lUTUAL AGREEMENT ON LICENSING SUBMITTAL CONTENT ESTABLISH LICENSING APPROACH AND SCHEDULE
Y
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INTRODUCTION
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RELOAD SCOPE
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SAFETY ANALYSIS
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PROPOSED TECH SPEC CHANGES LICENSING AMENDMENT PACKAGE
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SIGNIFICANT HAZARDS BETERMINATION
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BHR/5 REACTOR
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COMMERCIAL OPERATION BEGINNING 12/13/84
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INITIAL CORE:
764 6E 8X8 RP FUEL ASSEMBLIES
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RELOAD CORE:
- 196 ENC 8X8 FUEL ASSEMBLIES
(CONTINGENT ON CYCLE 2 STARTUP DATE)
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SUBMITTAL FOR LICENSING AMENDMENT'ARcH 1, 1986 END OF CYCLE 1 APRIL 1, 1986
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BEG INNING OF CYCLE 2 JUNE 1, 1986
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END OF CYCLE 2 APRIL 1, 1987
1
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/4
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'I 4
PLANT CLASS CONTAINMENT RATED THERMAL POWER NO>
OF FUEL BUNDLES NO OF CONTROL RODS BMR/5 NARK II 3323 NWT 760 185 RATED CORE FLOW AVERAGE POWER DENSITY AVERAGE LINEAR HEAT RATE DOME PRESSURE 108,5 NLB/HR 09,1 Ktl/L 5,39 KM/FT 1020 Ps I A
ill I
FUEL BUNDLE DESIGN 0
f'lECHANICAL 8X8 ROD ARRAY 62 FUEL RODS/2 WATER RODS 2IRC-2 FUEL ROD CLADDING 7
SPACERS
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NUCLEAR 138" ENRICHED> 12 NATURAL 3 BUNDLE TYPES HIGHS MEDIUMS AND NATURAL ENR I CHMENTS 4
GD RODS (SX MAX GD)g AXIALLYDISTRIBUTED UP TO 6 PIN ENRICHMENTS/BUNDLE CORE DESIGN 0
764 BUNDLES IN A 3 RADIAL ZONE LOADING PATTERN 0
CORE AVERAGE ENRICHMENT OF 1>881 U"235
0 f1ECHANI CAL 8X8 ROD ARRAY 62 FUEL RODS/2 WATER RODS ZIRC-2 FUEL ROD CLADDING 7 SPACERS
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NUCLEAR 138" (2,89K 0-235); 12" NATURA 5
GD RODS (2X GD)p AXIALLYUNIFORM 5 PIN ENRICHMENTS/BUNDLE THERMAL-HYDRAULIC COMPATIBLE WITH EXISTING RE 8X8RP FUEL 0
568 CYCLE 1 6E + 196 EXXON RELOAD ASSEMBLIES
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12-MONTH CYCLE OF 809 GWD ENERGY
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SCATTER LOADING'IGHTH-CORE SYMMETRY
e "EXXON NUCLEAR METHODOLOGY FOR BOILING HATER REACTORS-NEUTRONIC i"IETHODS FOR DESIGN AND ANALYSIS'p XN-NF-80-19 (P) (A), VOL, 1 AND YOLK 1 SUPPLEMENTS 1 AND 2, FUEL ASSEMBLY ANALYSIS METHODOLOGY (XFYRE)
CORE SIMULATION METHODOLOGY (XTGBNR)
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RELOAD BATCH SIZE SELECTED TO OBTAIN DESIRED DISCHARGE EXPOSURES BASED ON CYCLE ENERGY FUEL BUNDLE DESIGN TO MEET CYCLE ENERGY REQUIREMENTS AND POWER DISTRIBUTION CRITERIA
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SAFETY RELATED DESIGN CRITERIA SHUTDOWN MARGINS 1X ~K/K FUEL THERMAL MARGINS
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DEPENDENCY ON CYCLE 1 OPERATION RELOAD DESIGN AND ANALYSIS BASED ON MOST CURRENT CYCLE 1 OPERATING HISTORY PLUS PROJECTION TO EOC"1 RELOAD ANALYSIS PERFORMED ASSUMING AN EOC-1 EXPOSURE WINDOW
CO
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PRIME 9950 COMPUTER INSTALLED IN MAY 1985, r
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INSTALLATION OF POWERPLEX AT SITE SCHEDULED FOR OCTOBER 1985
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OPERATIONAL FOR SIX MONTHS PRIOR TO RELOAD CYCLE g
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FULLY CONSISTENT WITH EXXON NUCLEAR ANALYSIS METHODOLOGY,
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POWER DISTRIBUTION UNCERTAINTIES INCORPORATED INTO MCPR SAFETY LIMITI
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CORE LOADING VERIFICATION PURPOSELY ASSURE CORE IS CORRECTLY LOADED 0
CONTROL ROD FUNCTIONAL (INSERT AND WITHDRAWAL CHECKS)
PURPOSELY ASSURE PROPER CONTROL ROD FUNCTION
". VERIFICATION OF SUBCRITICALITY 0
SUBCRITICAL SHUTDOWN MARGIN DEMONSTRATION PURPOSE' ASSURE MINIMUM REQUIRED SHUTDOWN MARGIN EXISTS WITH THIS STRONGEST WORTH ROD FULLY WITHDRAWN TO DETERMINE ACTUAL AMOUNT OF SHUTDOWN MARGIN TO COMPARE PREDICTED VS>
MEASURED CRITICAL ROD POSITION
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TIP ASYMMETRY PURPOSELY TO ASSURE PROPER OPERATION OF TIP SYSTEM TO CHECK CORE SYMMETRY
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I'1CPR SAFETY LIMITS CALCULATED FOR ENC AND GE FUEL (XN-NF-524 (P) (A)),
0 TRANSIENT ANALYSIS PERFORMED TO ESTABLISH LIMITING TRANsIENT DEI TA-CPR (XN NF 79 71 (P) p AND XN-NF-80-19 (P) VoL, 5, AND XN-NF-81-22 (P) (A)),
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OPERATING LIMIT BASED ON SUM OF SAFETY LIMIT AND DELTA-CPR, 0
PREVIOUSLY LICENSED FOR CYCLE 1 OPERATIONs
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QUALITATIVEJUSTIFICATION THAT CYCLE 1 NAPLHGR REDUCTION FACTORS AND NCPR SAFETY LIMIT DELTA REMAINS CONSERVATIVELY APPLICABLE TO CYCLE 2e Sf I
NO CHANGES ANTICIPATED(
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CONTROL ROD WITHDRAWAL ERROR IS EVALUATED (XN-NF-80"19 (A)g VOL) 1)
AND ROD BLOCK MONITOR SETTINGS DEVELOPED<
0 FUEL LOADING ERRORS INCLUDING A MISLOADED FUEL BUNDLE AND A MISORIENTED FUEL BUNDLE ARE EVALUATED (XN-NF"80-19 (A)p VoL, 1),
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FUEL SEISMIC AND HYDRODYNAMIC LOADS ARE EVALUATED BYl COMPARING GE AND ENC FUEL RESPONSE CHARACTERISTICS VERIFYING THE APPLICABILITY OF EXISTING HNP"2 ANALYSIS TO ENC FUEL>
CONTROL ROD DROP ACCIDENT ANALYSIS'ONE ON A GENERIC BASIS (XN-NF-80"19 (A) g VOL> 1) p IS VERIFIED TO APPLY TO I'lNP-2,
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LIMITING BREAK CALCULATIONS ARE PERFORMED IN ACCORDANCE WITH APPENDIX K REQUIREMENT USING APPROVED ECCS MODELS TO DETERMINE APLHGR LIMITS (XN-NF-80-19 (P)
(A) VOt UMES 2, 2A, 28, 2C),
0 LIMITS FOR GE FUEL REMAIN UNCHANGED'N CYCLE 2,
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TRANSIENT AND LOCA ANALYSIS IS PERFORMED TO EVALUATE OPERATION AT GREATER THAN RATED FLOW (105K FLOW AT 100K POWER),
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OPERATION AT RATED POWER AND FLOWS AS LOW AS 90K WAS PREVIOUSLY LICENSED FOR CYCLE 1 OPERATION AND WILL BE CONTINUED IN CYCLE 2>
E 0
NUCLEAR DESIGN ANALYSIS THERMAL HYDRAULIC DESIGN ANALYSIS HECHANICAL DESIGN ANALYSIS PLANT TRANSIENT ANALYSIS STATISTICAL ANALYSIS CRITICAL POWER HETHODOLOGY CRITICAL POWER CORRELATION L0CA/ECCS HETHDDQLDGV (EXEfh/BMR)
FUEL ROD HEATUP ANALYSIS (HUXY)
FUEL THERMAL-NECHANICALANALYSIS CLAD SHELLING AND RUPTURE NODEL (GENERIC RELOAD APPLICATION XN-NF-80-19 (A), VoL. 1 XN-NF-80-19, VOL., 3 XN-NF-81-21 (A)
XN-NF-79-71 I
XN-NF-81-22 (A)
XN-NF-520 (A)
XN-NF-512 (A)
XN-NF-80-19 (A), VoL, 2 XN-CC-33 (A)
XN-NF-81-58 (A)
XN-NF-82-07 (A)
XN-NF-80-19 (A), VoL, 0 APPROVED UNDER REVIEW APPROVED UNDER REVIEW APPROVED APPROVED APPROVED APPROVED APPROVED APPROVED APPROVED APPROVED
tl
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PRIOR TO CYCLE 2 STARTUP TO PROVIDE NRC REVISED ANALYSES SHOWING. AFFECTS OF HIGH-BURNUP FISSION GAS RELEASE ON LOSS-OF-COOLANT ACCIDENTs e
PRIOR TO CYCLE 2 STARTUP TO PROVIDE NRC A REVISED STABILITY ANALYSIS s
~
iREQUEST CLOSURE OF LICENSE CONDITION<
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JNLPP.
PROPOSED 03/01/86 04/01/86 04/01/86 06/01/86 04/01/87 LICENSE SUBMITTAL END OF CYCLE 1 SHOLLY NOTICE IN F/R EARLIEST BOC 2 END OF CYCLE 2
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HPPJKED 4
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CHANGES MOTIVATED BY FUEL CHANGE
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CHANGES NOTIYATED BY OPERATION AT 105K FLOW
I
0 OVERVIEW OF TECH SPEC CHANGES
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IDENTIFICATION OF TECH SPECS TO BE CHANGED
0 1
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(HANGES OR ADDITIONS TO DEFINITIONS 0
CHANGES TO SAFETY LIMITS 0
CHANGES TO LIMITING CONDITIONS FOR OPERATION
l
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4
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ADD DEFINITION OF AVERAGE BUNDLE EXPOSURE
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NODIFY CRITICAL POWER RATIO DEFINITION (1,8)
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HODIFY FRACTION OF LIMITING POWER DENSITY DEFINITION.(1>13)
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NODIFY LINEAR HEAT GENERATION RATE DEFINITION (li20)
l
- 1
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Etc e>ICE
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NODIFY BASEST NODIFY THERMAL POWERS LOW PRESSURE OR LOW FLow" (2,1,1)
NODIFY THERMAL POWERS HIGH PRESSURE AND HIGH FLow" (2,1,2)
NODIFY REACTOR PROTECTION SYSTEM INSTRUMENTATION sETPOINTsj AvERAGE PowER RANGE MoNIToR (2 2
~ 1 s2)
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0 REACTIVITY ANOMALIES (3/ f e 1 e 2)
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AVERAGE PLANAR HEAT GENERATION RATE (3/4,2,1)
APRN SETTINGS (3/ f y 2 y 2)
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MINIMUM CRITICAL POWER RATIO (3/4,2,3)
I LINEAR HEAT GENERATION RATE (3/4,2,4)
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END OF CYCLE RECIRCULATION PUMP TRIP INSTRUMENTATION (3/4,3,4,2)
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HAIN TURBINE BYPASS SYSTEM (3/4e? o8)
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NOD IFY BASES l REACTIVITY ANOMALIES (3/4 ~ 1 y 2)
CONTROL RODS (3/4sly3)
CONTROL ROD PROGRAM CONTROLS (3/4il 4)
AVERAGE PLANAR LINEAR HEAT GENERATION RATE (3/412 1)
APRN sETTING (3/4,2,2)
MINIMUM CRITICAL POWER RATIO '3/ fg2o3)
LINEAR HEAT GENERATION RATE (3/4,2,4)
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COVER LETTER/NO SIGNIFICANT HAZARDS DETERMINATION 0
RELOAD
SUMMARY
REPORT
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PROPOSED TECH SPEC CHANGES
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ENC RELOAD ANALYSIS REPORT
~
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PHYSICS TEST PROGRAM
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C BASIS FOR PROPOSED NO SIGNIFICANT 0
ENSURE THAT THE FOLLOWING GENERAL STATEMENT CAN BE MADE'HE PROPOSED TECHNICAL SPECIFICATION CHANGES DO NOTl 1)
INVOLVE A SIGNIFICANT INCREASE IN THE PROBABILITY OR CONSEQUENCES OF AN ACCIDENT PREVIOUSLY EVALUATED>
OR 2)
CREATE THE POSSIBILITY OF A NEW OR DIFFERENT KIND OF ACCIDENT FROM ANY ACCIDENT PREVIOUSLY EVALUATED; OR 3)
INVOLVE A SIGNIFICANT REDUCTION IN A MARGIN OF SAFETY<
~
FOR EACH PROPOSED CHANGE (OR GROUP OF SIMILAR CHANGES)p PROVIDE A BRIEF DESCRIPTION OF THE REASON FOR THE CHANGE AND THE RELATIONSHIP TO THE APPROPRIATE EXAMPLE OF AMEND" MENTS NOT LIKELY TO INVOLVE SIGNIFICANT HAZARDS CONSIDERA" TIONS FROM 48 FR 14870,
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