ML20138J844

Summary of 851205 Meeting W/Util & Westinghouse Re 851115 Reload Submittal for Cycle 2.Non-LOCA Transient Analysis, LOCA Analysis,Tech Spec Changes & Radiological Consequences Discussed.Viewgraphs & List of Attendees Encl

ML20138J844
Person / Time
Site:	Callaway
Issue date:	12/12/1985
From:	Alexion T Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8512180077
Download: ML20138J844 (39)

Text

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Docket No.: 50-483 DEC 12 1985 LICENSEE: Union Electric Company FACILITY: Callaway Plant, Unit 1

SUBJECT:

MINUTES OF MEETING WITH UNION ELECTRIC AND WESTINGHOUSE REGARDING THE NOVEMBER 15, 1985 RELOAD SUBMITTAL FOR CYCLE 2 AT CALLAWAY On December 5, 1985, representatives of the Union Electric Company and. Westing-house met with the staff to discuss the November ~ 15, 1985 reload submittal for Cycle.2 at Callaway. Meeting attendees are listed in Enclosure 1. Slides used during the presentation are in Enclosure-2.

A summary of the reload submittal was presented. There were discussions on the non-LOCA transient analysis, the LOCA analysis, technical specification changes and radiological consequences. Regarding the licensee's use of the im-proved thermal design procedure, the staff noted that the licensee should document that the uncertainties used-in the Westinghouse procedure are appli-cable to Callaway. Regarding the optimized fuel assembly (OFA) design and methodology as described in WCAP-9500, the staff noted that the licensee should either address or provide references for the plant-specific open items resulting-

.from the staff's' generic review of WCAP-9500. -Regarding the W-3 R-Grid Correla-

-tion which was used in.the analysis of the Cycle 1 design, the staff noted that the technical specification bases should continue to address this correlation until the' core consists of 100% OFA fuel. Regarding the revised large break LOCA analysis, which was performed using the Westinghouse' BASH model, the staff noted that the generic review of this model has not yet been completed. If the staff identifies a significant open item in its generic review, it could delay approval of the reload submittal.

GRIGINAL SIGNED BT2 Thomas W. Alexion, Project Manager PWR Project Directorate #4 Division of PWR Licensing-A

Enclosures:

- As stated cc:- see next page h

h1 -

.PWR#4/ L Ph R-A TAlex on f B; i g 1 od 12/ j(' /85 12/ 8 8512180077 851212 PDR ADOCK 05000483 P_ PDR J

I -

Mr. D. F. Schnell Callaway Plant Union Electric Company Unit No. 1 i

)

cc

[

Mr. Nicholas A. Petrick Executive Director - SNUPPS Mayor Howard Steffen Chamois, Missouri 65024 i 5 Choke Cherry Road

[ Rockville, Maryland- 20850 Professor William H. Miller i Missouri Kansas Section American l Gerald Charnoff, Esq. Nuclear Society i Thomas A. Baxter, Esq. Department of Nuclear Engineering 1026 Engineering Building

[

Shaw, Pittman, Potts & Trowbridge 1800 M Street, N. W. University of Missouri Washington, D. C. 20036 Columbia, Missouri 65211 Mr. J. E. Birk Mr. Robert G. Wright Assistant to the General Counsel Assoc. Judge, Eastern District Union Electric Company County Court, Callaway County,

. Post Office Box 149 Missouri St. Louis, Missouri 63166 Route il F- Fulton, Missouri 65251

' Lewis C. Green, Esq.

U.- S. Nuclear Regulatory Comission Green, Hennings & Henry i Resident Inspectors Office Attorney for Joint Intervenors

RRf1 314 N. Broadway, Suite 1830 F

Steedman, Missouri 65077 St. Louis, Missouri 63102 Mr. Donald W. Capone, Manager Mr. Earl Brown Nuclear Engineering School District Superintendent Union Electric Company Post Office Box 9 Post Office Box 149 Kingdom City, Missouri 65262 St. Louis, Missouri 63166 Mr. Harold Lottman L A. Scott Cauger, Esq. Presiding Judge, Dasconade County Assistant General Counsel for the Route 1 Missouri Public Service Comm. Owensville, Missouri 65066 Post Office Box 360 Jefferson City, Missouri 65101 Mr. John G. Reed j Route #1

Ms. Marjorie Reilly Kingdom City, Missouri 65262 Energy Chairman of the League of Women Voters of Univ. City, M0 Mr. Dan I. Bolef, President 7065~Pershing Avenue Kay Drey, Representative University City, Missouri 63130 Board of Directors Coalition for the Environment j Mr. Donald Bollinger, Member St. Louis Region _

t Missourians for Safe Energy 6267 Delmar Boulevard

, 6267 Delmar Boulevard University City, Missouri 63130

! University City, Missouri 63130 1

I e

, - . - _- - . - - . - _ _ . . _ _ . _ . . _ , _ . . ...m __ _ _

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DEC 121985 l Callaway Plant 0 Unit No. 1 8

i

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cc: -

Regional Administrator U. S. NRC, Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 I Mr. Ronald A. :'ucera, Deputy Director LI Department of Natural Resources

. P. O. Box 176'

.l Jefferson City, Missouri 65102 o Mr.'Glenn L. Koester i Vice President - Nuclear Kansas Gas and Electric Company L

201 North Market Street Post Office Box 208

' Wichita, Kansas 67201

1

' Eric A. Eisen, Esq.

Birch, Horton, Bittner and Moore

' ' Suite 1200 1155 Connecticut Avenue, N. W.

Washington, D. C. 20036 i

f.

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[.. ..

ENCLOSURE 1

' CYCLE 2 RELOAD SUBMITTAL FOR CALLAWAY g

j DECEMBER 5, 1985 I:

NAME ORGANIZATION i.

} Richard Lobel NRC Tom Alexion NRC

f. Walter Brooks NRC

'C Carl- Berlinger NRC Shih-Liang Wu NRC 4 Brian McIntyre Westinghouse-

[ Jayashree Iyengam Westinghouse Phil McHale Westinghouse Carole Leach Westinghouse .

Keith Forchl Westinghouse Barbara Tuttle Westinghouse Randy Irwin Union Electric Alan Passwater Union Electric Dave Shafer Union Electric Mike Fletcher SNUPPS L

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,.A.. ~._m. ___. ._ _ . . . . - - . ~ . . -

l ENCLOSURE 2 I.*

F UE-NRC MEETING AGENDA

{* DECEMBER 5, 1985 CALLAWAY PLANT, CYCLE-2 LICENSE SUBMITTAL l

i-

[' o Introduction

- Summary of contents of package UE

- Schedule for Refuel-1 and subsequent startup

- Effect of NRR reorganization on the Callaway review NRC I

I W f

o Non-LOCA transient analysis - description and results o LOCA Analysis description and results W_

- Status of NRC BASH review NRC l

L j o Summary of Technical Specification changes UE F

o Radiological consequences UE

~t o Summary UE I

j o open Discussion 6

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. . - - - . _ . . ~ . - - . ~ - - . . . . . ~ . . . _ . _ - - . .

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5 r

I CALLAWAY CYCLE 2' LICENSING SUBMITTAL l:- -REFERENCE ULNRC-1207, 11/15/85 p

i-F CYCLE 2 BEGINS A TRANSITION TO WESTINGHOUSE 17x17 0FA 1

40 0FA AT 3.4 W/0 U-235 44 0FA AT 3.8'W/0 U-235 109 LOPAR ASSEMBLIES t

CORE SAFETY. ANALYSIS HAS BEEN PERFORMED AT 3565 MWys DESIGN THERMAL POWER = 3565 MWys RATED THERMAL POWER = 3411 MWrs NOTE: WE ARE NOT REQUESTING T0 OPERATE AT DESIGN THERMAL.

POWER AT THIS TIME.

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} CONTENTS OF SUBMITTAL t

TRANSMITTAL LETTER f, ATTACHMENT A - CHANGED TECHNICAL SPECIFICATION PAGES.

L 3

ATTACHMENT B - SAFETY EVALUATION FOR TRANSITION TO 0FA 1 ATTACHMEl4T C - NON-LOCA ACCIDENT ANALYSIS ATTACHMEi4T D - LOCA ACCIDENT ANALYSIS LARGE BREAK - BASH l

4- SMALL BREAK - NOTRUMP ATTACHMENT E - SIGNIFICANT SAFETY HAZARD EVALUATION ATTACHMENT F - RADIOLOGICAL CONSEQUENCES l

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.SIGNIFICANT CHANGES FROM PREVIOUS CORE l

t DESIGN CHANGES I 17x17 0FA y_a 17x17 LOPAR -

WCAP-9500-A WCAP-9272-P-A I WET ANi4ULAR BURNABLE ABSORBER (WABA) RODS WCAP-10021-P-A i

a ANALYTICAL CHANGES g .

i USE OF IMPROVED THERMAL DESIGN PROCEDURE WCAP-8567 (ITDP)

USE OF WRB-1 CORRELATION WCAP-8762-P-A l

! USE OF PAD CODE WCAP-8720 USE OF WOTRUMP SMALL BREAK LOCA MODEL WCAP-100514-P-A USE 0F BASH LARGE BREAK LOCA MODEL WCAP-10266 REAdALYSIS PERFORMED AT 3565 MWi s 10% OF STEAM GENERATOR TUBES ASSUMED PLUGGED em i

B i

- . - - - - - - , - , - , . , , - - - - , - - . , - . . - , - , ~ , ,, ,- - - - - - - - - - - - - - - - - - - - - - - - - -

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I CURRENT PLANT STATUS-i DECEMBER 1, 1985 g

F i

PLANT 0PERATING AT 100% POWER LEVEL (19 DAYS)

CUMULATIVE CAPACITY FACTOR (CF) SINCE COMMERCIAL I

I 82%

OPERATION HAS BEEN

)f

- CONTINUED OPERATION AT 80% CF - 3/10/86 END OF CORE LI CONTINUED OPERATION AT 90% CF - 2/27/86 END OF CORE LIFE h

REFUEL 1 WILL BEGIN ON 2/28/86

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M REFUEL-1-MAJOR MILESTONE SCHEDULE y

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[ 2/28/86 OPEN BREAKERS

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3/7/86 ENTER MODE 6

!, 3/d/86 REMOVE RV HEAD 3/10-3/14 REMOVE FUEL 3/14-3/18 RE-FUEL L

3/20/d6 C. LOSE RCS

3/24/86 ENTER MODE 5 r

4/4/86 ENTER MODE.4 ,

4/5/86 . .

ENTER MODE 3' 4/7/86 ENTER MODE 2 t 4/10/86 ENTER MODE 1

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]- 4/11/d6 CLOSE BREAKERS

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REFUEL'I - MAJOR ITEMS t

REFUELING - (CRITICAL PATH)

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INSPECT MAIN GENERATOR ROTOR AND ALL TURBINE VALVES h

REPLACE SEALS ON RCP CSD - INSPECT ALL 4 MOTORS la MONTH INSPECTION ON BOTH DIESEL GENERATORS STEAM GENERATOR WORK l

- SLUDGE LANCING

.20% EDDY CURRENT INSPECTION ON ALL 4 SG's L- .

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l' LICENSING SUBMITTAL DESIGN BASES i f, NON-LOCA SAFETY ANALYSES I

FUEL DESIGN CHANGE 17 X 17 LOPAR TO 17 X 17 OFA t

NON-CHAMFERED TO CHAMFERED PELLET DESIGN ANALYTICAL ASSUMPTIONS: PLANT CONFIGURATION 10 % STEAM GENERATOR TUBE PLUGGING ALLOCATION ENGINEERED SAFEGUARDS FEATURE-DESIGN RATING I

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I , i INPUT CHANGES L'

NON-LOCA SAFETY ANALYSES s

FUEL TEMPERATURES (OFA/W REVISED THERMAL SAFETY tlODEL)

REACTIVITY PARAMETERS (DFA)

CONTROL ROD DROP TIME (OFA)

CORE /NSSS POWER (ESFDR)

REACTOR COOLANT TEMPERATURES (ESFDR/SGTPALLOCATIbN)

STEAM PARAMETERS (ESFDR/SGTP ALLOCATION) f-l

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i METHODOLOGY NON-LOCA SAFETY ANALYSES I

CONSISTENT WITH W RELOAD SAFETY EVALUATION METHODOLOGY: WCAP-9272-P-J ANALYSES ARE APPLICABLE TO TRANSITION AND FULL OFA CORES l

l ITDP AND WRB-1 DNB CORRELATION USED: - WCAP-8762-P-A AND WCAP-8567 CODES AND ANALYSIS METHODOLOGY CONSISTENT WITH MOST RECENT NRC ACCEPTED FACTRAN LOFTRAN TWINKLE THINC 3 I

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5 CALLAWAY NON-LOCA SAFETY ANALYSES IN SUPPORT OF OFA

) REASON FOR REANALYSIS i ACCIDENT ESFDR/NSSS DEO QQNE1QUS911gN I

FSAR SECTiON 15.1 X X FEEDWATER MALFUNCTION X X EXCESSIVE LOAD INCREASE X

g STEAMLINE BREAK

FSAR SECTION 15.2 X X LOSS OF LOAD ,

X LOSS OF NON-EMERGENCY AC X

. LOSS OF NORMAL FEEDWATER X

FEEDLINE BREAK FSAR SECTION 15.3 X X PARTIAL LOSS OF FLOW X X COMPLETE LOSS.0F FLOW X X' LOCKED ROTOR FSAR SECTION 15.4 X

l ROD WITHDRAWAL FROM SUBCRITICAL X X

, ROD WITHDRAWAL AT POWER X X l DROPPED ROD X X STARTUP OF INACTIVE LOOP X X

-BORON DILUTION '

X X r ROD. EJECTION

. FSAR SECTION 15.5 X X INADVERTENT ECCS OPERATION AT POWER X

CVCS MALFUNCTION FSAR SECTION 15.6 X X l RCS DEPRESSURIZATION i

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CONCLUSION L

THE NON-LOCA SAFETY ANALYSES PERFORMED FOR THIS LICENSING SUBMITTAL :

I APPLICABLE FOR TRANSITION AND FULL OFA CORES PENDING TECHNICAL SPECIFICATION REVISIONS, MEET ACCEPTANCE CRITERIA OF STANDARD REVIEW PLAN CHAPTER 15, DEMONSTRATE NO ADVERSE IMPACT UPON PLANT SAFETY.

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' CALIAWAY (SNUPPS) UNIT i

LOSS - OF - COOIANT ACCIDENTS .

ANALYSIS BRIAN A. MCINTYRE WESTINGHOUSE ELECTRIC CORPORATION e

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L INTRODUCTION i

- BACKGROUND l

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- SCOPE ,

-- IARGE BREAK LOCA ANALYSIS

- SMALL BREAK LOCA ANALYSIS

' - RESULTS t

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BACKGROUND r -

REASONS FOR LOCA ANALYSIS r

- LICENSE CONDITION FOR CYCLE 2 START UP j - UPRATING THE PIANT TO 3565 MWT (CORE DESIGN THERMAL POWER)

- 10 PER CENT STEAM GENERATOR TUBE PLUGGING

- OPTIMIZED FUEL ASSEMBLY IMPLEMENTATION i

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LARGE BREAK ANALYSIS

- 3 DOUBLE-ENDED COLD-LEG GUILLOTINE BREAKS WITH DISCHARGE COEFFICIENTS OF 0.4, O.6 AND 0.8 r - MINIMUM AND MAXIMUM SAFEGUARDS CONDITIONS i

- BASH EVALUATION MODEL SMALL BREAK ANALYSIS E - 3 COLD LEG BREAKS OF 3, 4, AND 6 INCHES

- MINIMUM SAFEGUARDS CONDITION t

- NOTRUMP EVALUATION MODEL I

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COMPUTER CODES USED w

1981 EVALUATION MODEL VERSIONS

- SATAN-VI

- WREFIDOD

- COCO

- LOCTA-IV BASH

- COMBINES BART WITH IMPROVED THERMAL-HYDRAULIC MODELS l

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CALCULATES HOT ASSEWLY I AVERADE n00 COPOffl0NS M AT TRANSFER COEFFICIENT I M AT TRANSFER COEFFICIENT I 9r l l l I BAnT g jg MOT ASSDelv TERMONYDRAULIC g

COPOffl0NS OURING REFLOOO I LNIT ASSOMit.Y l l y

MASS VELOCITY I 1 L----_---_ _____ ____________,_s Cane INLET FLOW.ENTNALPY.fHESStpE r___ - ____ __ _.__ ______ ___ __,

SATAN l BASH l l I RCS. CONE AT EOB dl RCS CONDITIONS DURING REFLOOD I l

RCS. CORE T}ERMONYORALE.lC CONDITIONS DIRING 6 LOWDOWN l l l

i RCS AT EOS l ORE RET N . N M M CORE M ET W .EN M W l 1 1r l 1 I I BART g

' CORE TERMOHYDRALE.IC COPOITIOPS g g y

OURING REFLOOD g

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! RErILI. flw AmATOR & S.I. NOWS

,j CONTAINMENT PRESSunE MASS EFFRGY K LEASE F-"""--"-"""-""""*""""""""""-"""""-1 I wnEFLOOD/ COCO I I I I wnErLOOo I N CALCIA ATES BnEAM uASS.ENtnov nELEASE 1

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JL g C LCULATES CONTAINENT PRESStpE ,,

., AT EDS I I t l COCO I I CALCULATES CONTAlbaENT FHES9URE l

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Figure 15.6-4 Evaluation Model -

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CODE DESCRIPTIONS (LARGE BREAK ANALYSIS) l

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, - SATAN-IV ANALYZES THE THERMAL-HYDRAULIC TRANSIENT IN THE RCS DURING BLOWDOWN e -

- WREFLOOD

- CALCULATES THE SYSTEM THERMAL HYDRAULIC TRANSIENT DURING THE REFILL AND REFLOOD PHASES OF THE TRANSIENT

- COCO

- CALCULATES CONTAINMENT PRESSURE DURING ALL 3 PHASES OF THE TRANSIENT ,

- LOCTA-IV

- CALCULATES CORE WIDE AVERAGE ROD INITIAL CONDITIONS AT BOTTOM OF CORE RECOVERY

- BART

- CORE THERMAL HYDRAULIC CONDITIONS DURING REFIDOD

- BASH

- THERMAL-HYDRAULIC SIMULATION OF THE REACTOR CORE

'AND RCS DURING REFLOOD PHASE OF THE TRANSIENT

- IDCTA/BART

- DETAILED FUEL ROD MODEL

- COMPUTE THERMAL TRANSIENT OF THE HOTTEST FUEL ROD i

DURING THE THREE PHASES OF THE ACCIDENT t

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I - RCS RAPIDLY DEPRESSURIZES TO SATURATION i

- SIDW GRAVITY DRIVEN DRAIN OF THE RCS

- LOOP SEALS HAVE IMPORTANT INFLUENCE-t l

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CODES USED (SMALL BREAK ANALYSIS)

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-CONTROL VOLUME APPROACH

-MASS STORAGE

-TRANSIENT EQUATIONS

-DRIFT FLUX, SEPERATED FLOW MODELS

-THERMODYNAMIC NON-EQUILIBRIUM

-LOCTA-IV

-CLADDING THERMAL ANALYSIS e

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? O O T C R T CORE PRESSURE, CORE U A FLOW, MIXTURE LEVEL M AND FUEL' ROD POWER P HISTORY O< TIME < CORE COVERED j '

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[ TABLE 15.6-2 r -

InDut Parameters Used in the ECCS Analyses Parameter Laroe Break Small Break Reactor Core Design Thermal Power * (Mwt) 3565 3565 Peak Linear Power (kw/ft) 13.47 .12.77 (includes 102% factor)* at 6.0 ft at 10.0 f t Total Peaking Factor (Fg) at peak 2.32 2.20 i

Power Shape Chopped See Figure P

Cosine 15.6-45 l Fue'l Assembly Array 17 X 17 17 X 17 Optimized Optimized I

Nominal Cold Leg Accumulator 850 850 Water Volume (ft 3/ accumulator)

Nominal Cold Leg Accumulator 1364 1364 Tank Volume (ft 3/ accumulator)

Minimum Cold Leg Accumulator 600 600

, Gas Pressure (psia)

Pumped Safety Injection Flow See Table See Table r

15.6-3 15.6-9 Steam Generator Initial Pressure (psia) 935.0 935.0 Steam Generator Tube 10 10 Plugging Level (%)

Containment Parameters (See FSAR Section 6.2)

Initial Flow In Each Loop (1b/sec)

, -9710.7 Vessel Inlet Temperature (*F) 554.93 Vessel Outlet Temperature (*F) 619.47 t Reactor Coolant Pressure (psia) 2280.

t

• i j Two percent is added to this power to account for calorimetric error.

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T. 4 TABLE 15.6-5 (Sheet 1)

Laroe Break LOCA Results Fuel Claddina Data

'! Minimum Safeauards I

CD = 0.8 Co = 0.6 CD = 0.4 DECLG DECLG DECLG

[

g RESULTS I Peak Clad Temperature (*F) 1939 2009 1750 Peak Clad Temperature Location (ft) 7.0 7.00 7.25 Local Ir/H O2 Reaction (maximum %) 4.20 6.32 2.55 Local Zr/H O Location for maximum 6.00 6.00 6.75 reaction (2ft) r Total Ir/H O 2 Reaction, (%) <0.3 <0.3 <0.3 Hot Rod Burst Location, (ft) 6.00 6.00 6.75 Maximum Safeauards CD = 0.6 DECLG

. RESULTS Peak Clad Temperature (*F) 2026

~

.g Peak Clad Temperature Location (ft) 7.0

{. Local 2r/H O2 Reaction (maximum %) 5.94 Local 2r/H2O Location for maximum reaction (ft) 6.0 i

Total Ir/H O 2 Reaction, (%) <0.3 f; Hot Rod 8urst Location, (ft) 6.00 4

1.

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TIME AFTER START OF REFLOOD (S)

Figure 15.6 35 Reflood Mixture Levels, CD = 0.6 DECLG, MAXSI i

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TABLE 15.6-8 Small Break LOCA Results

j. Fuel Claddina Data i

3 in 4 in 6 in RESULTS Peal- Clad Temperature (*F) 1299 1184 969 Peak Clad Location (ft) 12 12 11.75 Local Ir/H O2 Reaction (maximum %) 0.23 0.084 0.071 i Local Ir/H 2O Reaction Location for 12 12 11.25 I maximum reaction (ft)

Total .2r/H 2O Reaction, (%) <0.3 <0.3 <0 3 Hot Rod Burst Time, (sec) N/A N/A N/A n Hot Rod Burst Location, (ft) N/A N/A N/A t

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SUMMARY

OF TECHNICAL SPECIFICATION CHANGES REFER TO ATTACHMENT B,. TABLE 2 OF SUBMITTAL

[

t Introduce Design Thermal Power and Rated Thermal Power Use'of ITDP at Design Thermal Power and WRB-1 Fig. 2.1 revised core safety limits T-2.2-1 -

revise min. measured flow revise various OT A T a'nd OPliT definitions, allowances and sensor errors ,

3/4.1.3 - Revised rod drop time to 2.4 seconds F 3.1-1 -

Revised to indicate rod insertion limits based on Rated Thermal Power 3/4.2.3 - Removed RCS flow rate on nuclear enthalpy

-A rise hot channel factor specification and replaced with new FNA H specification f 3/4.2.5 - Flow rate now covered under this specific ~ation T-3.2-1 --

Revised Tave and added RCS-flow

- (We have revised the Bases accordingly) 6

-i i

4

- - . . _ ,r _. _,

r

SUMMARY

OF i RADIOLOGICAL CONSEQUENCES I

+

EVALUATION I

l - FISSIm PRODUCT INVEIRORY

- NEGLIGIBE IffACT ON SOURCE TERMS ,

i f

- STEAM SYSTEM PIPIfr FAILURE

[

! -SmLLDECREASEINTHYROIDlbSE

- NO CHANGE IN $ 0E BODY DOSE

- LOSS OF NON-EMERGENCY AC PMER TO THE PLANT AUXILIARIES

- NO SIGNIFICANr OKANGE IN THYROID m

! $ 0 E BODY DDSE

- REACTM C00Uwr PUMP SHAFT SEIZURE

- REDUCTIM IN THYROID AND WHOE BODY DOSE BYAPPROXIMATELY25%

, -SGTR

[i -LATER

~

- ROD EJECTIM ACCIDENT; BREAK IN INSTRlNENT LINE; LOCA; PADIMCTIVE WASTE GAS DECAY IANK FAILURE; RADIMCTIVE LIQUID WASTE SYSTEM FAILURE; FusL HANDLINGACCIDEIR

- NO CHANGE IN IHYROID m WHOE BODY DOSE

+ + . - . _ - - - . -. -- -.. ~ ..--. ......_. -- ,- -- - . - - - .

(

i

. 2-,

~

MEETING

SUMMARY

DISTRIBUTION DEC 12 E

#$ME$8th ._ NRC Participants NRC PDR R.-Lobel '

L'PDR-T. Alexion NSIC. - W. Brooks LPRC System. C. Berlinger

'PWR#4 Reading File' .

' Project Manager Tom Alexion ~

Mc Duncan Attorney,~OELD

J. ' Partlow E. Jordan

'B Grimes

-ACRS (10)

.OTHERS

, bcc: Licensee & Service: List 6

9

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