ML20206P956
| ML20206P956 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 08/08/1986 |
| From: | Olshan L Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8608270394 | |
| Download: ML20206P956 (44) | |
Text
- . . . . . . -
. . (, 0 8 gg gg i
Bocket*No.: STN 50-454 .
1 LICENSEE: Commonwealth Edison Company FACILITY: Byron Station, Unit 1
SUBJECT:
MEETING
SUMMARY
- BYRON 1 T-P0T LEG REDUCTION On July 25, 1986, the NRC staff met with Commonwealth Edison Company and i Westinghouse Electric Corporati,on to discuss a program to reduce the hot leg
- temperature at Byron Station, Unit 1. A list of attendees is given in l Enclosure 1. The slides presented during the meeting are included in Enclosure 2.
On May 17, 1986, Commonwealth Edison voluntarily reduced power at Byron 1 to 1 75%. This action was taken to reduce the potential for steam generator tube
} primary water stress corrossion cracking. At 75% power, hot leg temperature t is about 600"F (at 100%, hot leg temperature was about 618"F), and the rate of
- stress corrossion cracking decreases as temperature decreases. To further i reduce the potential for stress corrossion i
- racking, the Byron 1 steam generator tubes will be shot peened during the next refueling outage. "
- The Byron T-hot leg reduction program is to be dcne in two phases. Phase I, as proposed by the licensee, will be done in accordance 10CFR Section.50.59; i
i.e., no changes to the Technical Specifications are required and no prior approval by NRC is required. The plant will be operated at 94% power with T-hot leg of 600"F. The licensee intends to begin implementing changes for this mode of operation on August 2,1986. The NRC questioned the appropriateness of implementing this new mode of operation under 50.59 and thought that a license amendment may be needed. The NRC committed to inform 3
the licensee within a short time after the meeting if a license amendment would be needed prior to implementing the new mode of operation.
~
Phase II of the T-hot leg reduction program involves a T-average program that ,
is based on actual steam generator tube plugging data. A license amendment -
will be required and the licensee intends to submit the amendment in Tebruary 1987.
4 The licensee has not yet determired if a similar program will be implemented
- on-Byron 2, Braidwood 1 or Braidwood 2.
i lenny Olshan, Project Manager ge27oa9486o808 PWR Project Directorate #5 P ADocK 05000454 Division of PWR I.icensing-A PDR *
- 0FC PD#5
- : : : : : :
' NAM . Olshan:ss : : : : : :
jDATE :8/ 6/86 : : : : : :
0FFICIAI. RECORD COPY
, i ,
Mr'. Dennis L. Farrar Byron Station Commonwealth Edison Company Units 1 and 2 CC' -
Mr. William Kortier Ms. Diane Chavez Atomic Power Distribution 508 Gregory Street Westinghouse Electric Corporation Rockford, Illinois 61108 Post.0ffice Box 355 Pittsburgh, Pennsylvania 15230 Regional Administrator, Region III U. S. Nuclear Regulatory Commission Michael Miller 799 Roosevelt Road Isham, Lincoln & Beale Glen Ellyn, Illinois 60137 One First National Plaza 42nd Floor Joseph Gallo, Esq.
Chicago, Illinois 60603 Isham, Lincoln & Beale
, Suite 1100 Mrs. Phillip 8. Johnson 1150 Connecticut Avenue, N.W.
1907 Stratford lane Washington, D. C. 20036 Rockford, Illinois 61107 Douglass Cassel, Esq.
Dr. Bruce von Zellen 109 N. Dearborn Street Department of Biological Sciences Suite 1300 Northern Illinois University Chicago, Illinois 60602 DeKalb, Illinois 61107 Ms. Pat Morrison Mr. Edward R. Crass 5568 Thunderidge Drive .
Nuclear Safeguards & licensing Rockford, Illinob 61107 Sargent & Lundy Engineers 55 East Monroe Street Ms. Lorraine Creek -
Chicago, Illinois 60603 Rt. 1 Box 182 Hanteno Illinois 60950 Mr. Julian Hinds U. S. Nuclear Regulatory Commission Byron / Resident Inspectors Offices 4448 German Church Road ,
Byron, Illinois 61010 l Mr. Michael C. Parker, Chief Division of Engineering )
Illinois Department of Nuclear Safety l 1035 Outer Park Drive i Springfield Illinois 62704 - l I
1
Meetino Summary Distribution Docket or Central File NRC Participants NRC PDR Local POR C. Berlinger PD+5 Reading File M. Dunenfeld J. Partlow R. lobe'l V. Noonan I. 01shan Project Manager 0GC-Bethesda E. Jordan B. Grimes ACRS (10)
M. Rushbrook cc: 1.icensee and Plant Service 1.1 s t I
1 I
l i I r
I
,( p " %' (k t.'NITED STATES
{- P( g NUCLEAR REGULATORY COMMISSION L /"j WASHINGTON, D. C. 20555
- f 0 8 pyg Docket No.: STN 50-454 -
I.ICENSEE: Commonwealth Edison Company -
FACII.ITY: Byron Station, Unit 1
SUBJECT:
MEETING
SUMMARY
- BYRON 1 T-POT LEG REDUCTION On July 25, 1986, the NRC staff met with Commonwealth Edison Company and Westinghouse Electric Corporation to discuss a program to redu:e the hot leg l temperature at Byron Station, Unit 1. A list of attendees is given in Enclosure 1. The slides presented during the meeting are included in Enclosure 2.
On May 17, 1986, Commonwealth Edison voluntarily reduced power at Byron I to 75%. This action was taken to reduce the potential for steam generator tube primary water stress corrossion cracking. At 75% power,- hot leg temperature is about 600"F (at 100%, hot leg temperature was about 618"F), and the rate of
~-
stress corrossion crading decreases as temperature decreases. To further reduce the potential for stress corrossion cracking, the Byron I steam '
generator tubes will be shot peened during the next refueling outage.
The Byron T-hot leg reduction program is to be done in two phases. Phase I, as proposed by the licensee, will be done in accordance 10CFR Section 50.59; 1.e., no changes to the Technical Specifications are required and no prior approval by NRC is required. The plant will be operated at 94% power with
,- T-hot leg of 600"F. The licensee intends to begin implementing changes for this mode of operation on August 2, 1986. The NRC questioned the appropriateness of implementing this new node of operation under 50.59 and thought that a Ifcense amendment may be needed. The NRC comitted to inform the licensee within a short time after the meeting if a license amendment would be needed prior to implementing the new mode of operation.
Phase II of the T-hot leg reduction program involves a T-average program that 3
is based on actual steam generator tube plugging date. A license amendment I
will be required and the licensee intends to submit the amendment in February 1987.
The licensee has not yet determined if a similar program will be implemented on Byron 2, Braidwood 1 or Braidwood 2.
- j $ . !h l.enny Olshan, Project Manager PWR Preject Directorate #5
. Division of PWR I.icensing-A i
Enclosure 1
, ., MEETING ATTENDEES: BYRON 1 T-HOT LEG REDUCTION JULY 25, 1986 NRC COMMONWEAI.TH EDISON CT Berlinger W. Ainger ~ -
M. Dunnenfeld A. Miosi R. lobel C. Moerke -
L. 01shan p. Reister T. Tramm ,
WESTINGHOUSE OTHER D. Augustine L. Connor, Doc Search J. Bass Association E. Manz J. Bell, AEP3C T. Sinback B. Rederstorff, AEPSC G. Smith V. Vanderburg, AEPSC S. Ditommaso 1 .
l l ._
- . . =
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July 25,1986 -
Byron 1T-HotReduction ~l
' ~
Agenda 1.IntroductionAndObjectives KenAinger 2.DescriptionofChanges TomTramm 3.PhaseIOverview DebbieAugustine 4.SafetyAnalyses ,
SteveDitommaso 5.ImplementationAtByron1 PennyReister
. _ . _ 6.SummaryAndConclusion KenAinger e e a
e i
I
AMENDMENT 42 STEAM OUTI.ET j }l}l}l TO TUR8INE p SECONDAR Y r MOISTURE SEPAR ATOR -
PRIMARY '
MOISTURE F MANWAY SEPARATOR T7 _
Ng ~
1
" -l< ,
1 - - -
~ ANTI VIBRATION 5A 4,.y BARS
._ m ,
l -TUBE BUNDLE C' TUBE SUPPORT PLATE (S)
/
0l PRE HEAT AREA
[
FEEDWATER INLET
' 'I
_-x TU8E SHEET COOLANT kl ll 4 INLET
- MANWAY BYRON /BRAIDWOOD STATIONS ,
FIN AL S AFETY ' AN ALYSIS REPORT '
fir,URE 5.4-11 STEAli 03ENERATOR I
J
- s i .
e F Longitudinal Cracks At Roll Transition i
l Cracks
]
\
//)(I'l'{'['(Ihg
\
!- Tube Sheet ' .
Roll Transition Zone
)
i i '
1 -
i l
)
i -
h .
O O .
$U NO
$N 3,-
""" O N'- N XM '
oE o SO
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, f s ,
Tube Sheet Roll Transitions, Initial Eddy Current Indications Temperature, (DEG. F) 630 620 610 600 590 1000 g i i 580 570 i i i i M
O 100 :::
3 * **
.N -
g T *
+
! H S - ~
10 -
Legend e = Full Depth Hard Roll
- b
~
A= Kiss Roll 4
+ = Partial Depth Hard Roll .r.-
1 '
1.66 1.67 1.68 1.69 1.70 1.71 1.72 1.73 1.74 1.75 1/T,(DEG. K 1)
(10"3)
~ . .. .
. BYRON STATION. . = . .
618.4 oF .
618.4 F TNot
- /
l . .
l h
i i
e.
l 588.4 .F ,- 588.4 F Tave O
6 ,'
v s
/
e -
U a /
b
/
/
u [ /
I c
h - / s' ,
5570f"j',' /
558.4 F Tcold i
I j 100 Reactor Power (%)
. . u.-
1 I
- 1092 -
1 a
= .
.e4 900 i'
e u
- s '
l M M .
- e u
E, e .
, a en i
100 Reactor Fower (%)
- i
594CF That
~
6 0
,- 559.60F Tave e -
w "
c ,-
a -
m u
s'- .
[E 547
. G h
e 5300F Tcold 100 Reactor Power (%)
e a
m
.,e ,
m 0- 1005 m
h w
a m
M 0
w Oe E a m
O .
a M
720 100 Reactor Power (%)- . ,
o
.M *
^
May17,1986 f
Byron 1 T-Hot Limited To 600 E' UntilShotpeeningIsCompleted
- 9 6
B
.* O t
BYRON STATION 618.4 618.4 F F
Thot
/ 6000F Tho't
/ . o
/
/
'/ .
+ m g
588.4'!F . 588.4 F Tave
/
O /
~
/
8 '
3 l D / .
w
/
,, 570.90F Tave a / ,#
E / -
J f / #'
/
5570F ,
558.4 F
.- s s Te s
s N.
I ' -
541.8 F Tcold 94 100 Reactor Power (%) .
8 i *
- 1092 s e s ee4 s n s, i a s 990 's 990
. G a
's s n s n s e s 64 s
s E
867 's 867 .
e W
- u W
94 100 Reactor Power (%) ,
.T-HotReduction -i PhaseIWork -
Objectives: - Justify Power Increase At 600 F Complete ASAP No Tech Spec. Changes . .
Schedule: AnalyticalWork: May28,1986ToJuly23,1986 Report: July 31,l'986 InplementChanges: August 2,1986 Y
e .
O p' .
T-HotReduction . , .
PhaseIIWork I Objectives: - Revise Documentation (FSAR, Manuals, Drawings)
Obtain NRC Approval Of " Operating Window"
' Schedule: RequestNRCApproval: February 1987 InplementOnByron1:
Spring 1987
.p. O h e e e 9
^
l z
ObjectiveOfOperatingWindow -
OperateReactorOnT-AveProgram WhichIsBasedUponActual
. SteamGeneratorTubePluggingData e 6
-= .a . -
e
- O t
a t
1
e . .
~ :.
OperatingWindowRange
. T-Hot: 600F-618F
~
% Plugged: 0%-5pprox.10%
--a.-
O l
OperatingWindowfeatures -
- 1. Analyze in advance to justify T-hot for the range of plugging, 2.ReviseT-AverageProgramwhenevertubesareplugged.
3.Notaday-to-daychange.
~
4.Licensechangeneeded. ,
5.Minimizesemergencylicensechanges.
6.MinimizescoststoCommonwealthEdison 7.Inplacebyendoffirstrefueling Y
- e
- t
- 3YRO\ \:~'i s ,
u,.. Ra o.. u \. 3 h_ RA ,
3-ASE OVERV .EW 8
-o e
- J.E30RA- AAS NE u _Y 25, ' 985 i
S t
l 4
. . , . . . - , _ --- __-,__,__m-,,,,--
PlPOSES 07 PnASE :
-i 8 PERFORM CONSERVATIVE EVALUATIONS TO SUPP OPERATION OF BYRON UNIT 1 AT 94% POWER FO REMAINDER OF CYCLE 1
- SAFETY
- OPERATIONAL 8 IDENTIFY EFFORT FOR PHASE II:
i
- MORE RIAUS"lC
~
DETAILED ANALYSIS / DOCUMENTATION
- ASSESS NEW MARGIN
, . COWPARISON OF CURRENT AND REDUCED-TEMPERATURE DESGN POWER CAPABillTY PARAMETERS
. REDUCED' TEMPERATURE CURIENT 100% 94%
PARAWETERS:
_DESGN THERWAL POWER THERWAL POWER (1) (2) (3) !
NSSS POWER (X) 100 100 94 (WWt) 3425 3425 3220 l THERWAL DESIGN R.0W 94400 94400 94400 RCS PRESSURE (PSIA) 2250 2250 2250 RCS TEMPERATURES fF):
CORE OUTIIT 621.7 603.5 603.3 VESSEL OUTLET 618.4 600.0 600.0 CORE AVERAGE 591.8 572.2 573.8 l VESSEL AVERAGE 588.4 569.1 570.9 VESSEL / CORE El.ET 558.4 538.2 541.8 i SG OUTIET 558.1 537.9 541.5 SIEAW GENERATOR:
l STEAW TEMPERATURE (*F) 543.3 522.1 527.6 STEAU PRESSURE (PSIA) 990 827 867 STEAM FLOW (10' LB/HR TOTAL) 15.13 15.03 14.05 i FEED BPERATURE PF) 440 440 435
~
APPROX. FDUUNG FACTOR .00005 .00005 .00005 1
ZERO LDAD TEMPERATURE 557 557 557 l
l
l GEN ERA _ AR AS Or sVES DA"Os 8 FINAL SAFETY ANALYSIS REPORT (FSAR) SAFETY REANALYSIS / EVALUATIONS 0 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) SYSTE EVALUATIONS
- ^ ,. =
e .
. 9 NS.SS C.0MPONENTS EVALUATIONS 8 NSSS/ BALANCE-0F-PLANT (B0P) INTERFACE EVALUATIONS
FSAY SATBY EANA_YS S/NA_UA"'0.sS 1
8 LOSS OF C0OLANT ACCIDENT (LOCA) 1
. - UMITING LARGE BREAK REANALYZED AT REDUCED T COLD 8 NON-LOCA ACCIDENT EVALUATIONS
- REDUCED TEMPERATURES GENERATE DNB MARGI
- NON-DNB ACCIDENTS ACCEPTABLE 0 MASS AND ENERGY RELEASES l - STEAMUNE BREAK (INSIDE AND OUTSIDE i
CONTAINMENT)' B0UNDED BY CURRENT FSAR -
- LOCA/ CONTAINMENT SUBCOMPARTMENT B0Ul EXISTING FSAR ANALYSES l ,
l
NSSS SYSTEMS EVALUATIONS
., [
t DEMONSTRATED FEASIBluTY OF GENERATING STEAM DUM SETPOINTS WHILE PRESERVING 50% LOAD REJECTION CAPABluTY O REVISED CONTROL SYSTEM SETPOINTS PROV!DED 8 PERFORMANCE OF FLUID SYSTEMS UNAFFECTED 8 CONFIRMED THE FOLLOWING VALVES WILL FUNCTION
. ... AS INTENDED AT NEW TEMPERATURES:
- PRESSURIZER SAFETY VALVES
- PRESSURIZER PORVS
- PRESSURIZER SPRAY VALVES
- S/G SAFETY VALVES 8 NSSS DESIGN TRANSIENTS MODIFIED TO REFLECT NEW TEMPERATURES 8 CONFlRMED CURRENT LOCA HYDRAUUC FORCING FUNCTIO B0UNDING
NSSS COMPONENTS EVA_UA"l0NS S VERIFIED STRUCTURAL NTEGRITY AT REVISED CONDIT
- STEAM GENERATORS .
- REACTOR VESSEL
- REACTOR INTERNALS
- CONTROL R0D DRIVE WECHANISMS
-=-
- PRESSURIZER
- RCS PlPING AND SUPPORTS
- REACTOR COOLANT PUMP MOTOR i
i
l NSSS/ BOP INTERFACE EVALUATl0NS 1
0 TURBINE - GENERATOR
, - INVESTIGATED HP TURBINE M0DS FOR WWe BENEFIT
- ADDRESSED-lMPACT OF LOW TEMPERATURES ON LP -
TURBINE AND MSR'S 0 MASS AND ENERGY RELEASES B0UNDED BY EXISilNG ANA 0 CONRRMED EXISTING INSTRUMENT RANGES STILL APPU AT REDUCED. TEMPERATURES i
1 0 CONRRMED NO CHANGE TO MAIN FEEDWATER CONTROL VALVE POSm0N, FUNCTION l
t
~1 CONCLUSIONS 0 CONSERVATIVE SAFETY /0PERATIONAL EVALUA110NS DEM FEASIBluTY OF OPERATION AT 94% RATED NSSS POWER, REDUCED TEMPERATURES FOR:
- FSAR SAFETY ANALYSES
- NSSS SYSTEMS / COMPONENTS
- NSSS/B0P INTERFACE CONSIDERATIONS l
)
l
1 BYRON PHASE 1 T-HOT REDUCTION -
SAFETY EVALUATION
~f
- RCS COMPONENTS EVALUATED: i STEAM GENERATOR PRESSURIZER 1
REACTOR COOLANT PUMP / MOTOR VESSEL AND INTERNALS PIPING AND SUPPORTS
'.,._ CONTROL ROD DRIVE WECHANISMS ap -
- AUXIUARY SYSTEMS AND ASSOCIATED COMPONENTS EVALUATED:
FEEDWATER WAIN STEAM CHEMICAL AND VOLUME CONTROL RESIDUAL HEAT REW0 VAL
- N0' ADVERSE AFFECT ON SAFE OPERATION OF TH5 PLANT
e a BYRON PHASE ~ 1 T-HOT REDUCTION SAFETY EVALUATION ,[
e LARGE BREAK LOCA ANALYSIS POTENTIAL FOR HIGHER PCT FROW PAST SENSITMTIES USE OF 1981 EVALUATION WODEL -
1978 WODEL UMITING BREAK:
~~
DECLG 0.6 DISCHARGE COEFFICIENT
- WAXIWUW SAFETY INJECTION RUN 1981 WODEL FOR WINIWUW AND WAXIWUW SAFEGUARDS o RESULTS (WAXIWUW SAFEGUARDS UWITING)
PCT OF 1988.7* F I -
MAXlWUW LOCAL WETAL-WATER REACTION 3.97%
TOTAL CORE WETAL-WATER REACTION * .34 CORE GE0WETRY AMENABLE TO C00UNG
BYRON PHASE 1 T-HOT REDUCTION SAFETY EVALUATION .
- CONTAINWENT WASS/ ENERGY RELEASE
- SHORT TERW RELEASE RATES INCREASE DUE TO:-
INCREASED SUBC00 LED FLOW REDUCED BREAK SIZE OFFSETS INCREASED RELEASE -
- CASES EVALUATED:
REACTOR VESSEL INLET BREAK (REACTOR CAVITY)
,, SPRAY UNE RUPTURE (UPPER PRESSURIZER CUBICLE)
~
DECL AND DEHL (LOOP COMPARTWENTS) 51AIN STEAMUNE RUPTURE (STEAMPIPE CHASE) e RESULTS REACTOR CAVITY: 49% OF FSAR VALUE PRESSURIZER COBICLE: 78% OF FSAR VALUE l DEHL: 66% OF FSAR VALUE ,
DECL: 59% OF FSAR VALUE
- STEAWUNE CHASE: FSAR VALUE BOUNDING (LOWER INITIAL TEWP. AND VOLUME) l
_ - - - - - _ _____ _ -. _ - _ _ . _ - _ . - _ -- i
BYRON PHASE 1 T-HOT REDUCTION l
SAFETY EVALUATION .
. ,. l e LOCA HYDRAUUC FORCES 1
- INCREASED HYDRAUUC FORCES:
LOWER TEWPERATURES HIGHER WATER DENSITY
._ - 0FFSET BY UTIUZING REDUCED BREAK SIZE ;
- UWITING BREAK LOCATION: REACTOR VESSEL INLET N0ZZLE !
144 SQ.IN.
- THREE CASES ANALYZED: '
144 SQ. IN./ ORIGINAL TEWPERATURE 144 SQ. IN./ REDUCED TEMPERATURE ;
REDUCED BREAK / REDUCED TEMPERATURE l l
- RESULTS
- l i !
- FIRST SENSITMTY: 14% INCREASE IN PEAK FORCES !
- SECOND SENSITMTY:
- 14% DECREASE IN PEAK FORCES I
- WARGIN AT OTHER LOCATIONS TO ACCOUNT FOR INCREASED FORCES !
- ORIGINAL DYNAWlC ANALYSIS REMAINS VAUD !
l
BYRON PHASE 1 T-HOT REDUCTION SAFETY EVALUATION .
~1 e NON-LOCA ACCIDENT EVALUATIONS PARAWETERS WITH IMPACT ON NON-LOCA TRANSIENTS:
'. 17.5 DEGREE-F REDUCED Tavg .
16.6 DEGREE-F REDUCED Tin 123 PSI REDUCED STEAW PRESSURE 6% REDUCED FULL POWER 5% REDUCED NARROW RANGE SG WATER LEVEL OVERPOWER AND OVERTEMPERATURE SETPOINTS STILL VALID TO PROTECT CORE LIMITS CASES ANALYZED AT NO LOAD CONDITIONS ARE UNAFFECTED OPENING OF SG SAFETY OR PORY STEAM SYSTEM PIPE FAILURE BANK WITHDRAWAL FROM SUBCRITICAL RCCA ROD EJECTION
'FEEDWATER SYSTEM WALFUNCTION STARTUP OF RCP AT INCORRECT TEMPERATURE CVCS (BORON DILUTION) i
._ . _- _ _ ._. . . _ _ _ - - _ - . . __ - - . . .. L
BYRON PHASE 1 T-HOT REDUCTION SAFETY EVALUATION -
l 1
l e NON-LOCA ACCIDENT EVALUATIONS l l
DNB TRANSIENTS BOUNDED BY FSAR DUE TO:
l REDUCTION IN TEMPERATURE
~-
REDUCTION IN POWER INCREASED DNBR TH$SE INCLUDE:
[
FEEDWATER SYSTEM MALFUNCTION l INCREASE IN SECONDARY STEAM FLOW l TURBINE TRlP RCCA WISOPERATION !
, INCREASE IN REACTOR COOLANT INVENTORY.
OPENING OF PRESSURIZER SAFETY OR PORY i
DECREASE IN RCS FLOW RATE i
BYRON M(ASE 1 T-NOT REDUCTION SAFETY EVALUATION -
- NON-LOCA ACCIDENT EVALUATIONS -
1
- FEEDWATER SYSTEM PIPE BREAK KNEFITS RESULT FR0W:
LOWER Tovg LOWER SG WATER LEVEL LOWER STEAM PRESSURE GREATER SUBC00UNG MARGIN EXISTS TO NOT LEC 80 lung FSAR ANALYSIS REMAINS VAUD
-RCCA EJECTION AT FULL POWER
_ BENEFITS RESULTS FROW:
REDUCED POWER REDUCED TEMPERATURE FUEL R00 THERWAL TRANSIENT LESS SEVERE FSAR ANALYSIS REMAINS VAUD 1
CVCS WALFUNCTION (B0RON DILUTION) .
EVALUATED AT BOL (NIGNEST B0RON CONC.)
4-5 ADDITIONAL PPW TO OFFSET POSITIVE REACTMTY 4
4
~
BYRON PHASE 1 T-HOT REDUCTION SAFETY EVALUATION -
'1 e NON-LOCA ACCIDENT EVALVATIONS STEAMUNE BREAK MASS / ENERGY RELEASE ;
INSIDE CONTAINMENT:
~~
LOWER STEAM TEMPERATURE AND PRESSURE REDUCED SG LEVEL / REDUCED TOTAL MASS RELEASED l- '
REDUCED POWER (LESS DECAY HEAT) !
FSAR BLOWDOWN DATA IS BOUNDING
\
OUTSIDE CONTAINWENT (SUPERHEAT):
LOWER STEAM PRESSURE RESULTANT LOWER PEAK ENTHALPY EARUER LOW-LOW SG TRIP
~
s SI AND STEAMUNE ISOLATION ACTUATED EARUER REDUCED POWER (LESS DECAY HEAT)
CURRENT WASS/ ENERGY RELEASES BOUNDING l
i BYRON PHASE 1 T-H0T REDUCTION
' ~
SAFETY EVALUATION
- NON-LOCA ACCIDENT EVALUATIONS CONDITION 11 EVENTS P0TENTIAL FOR PRESSURIZER OVERFILL LIMITING TRANSIENTS ARE:
LOSS OF NORMAL FEEDWATER LOSS OF 0FFSITE POWER UNCONTROLLED RCCA WITHDRAWAL AT POWER l : THESE TRANSIENTS WERE REANALYZED
/
. .--,---,.,-.,.--n , , . . - - - , - - . , . . _ . - , -
. - - , , . _.,,-.,,,_,...n.,.,,-... n-,. -. . _ .,
- BYRON PHASE I T-HOT REDUCTION SAFETY EVALUATION ,
. , -. l 1
e LOSS OF 0FFSITE~ POWER (WITH' REACTOR TRIP)~
- LOSS OF NORWAL FEEDWATER UNCERTAINTIES WODIFIED TO WAXIWlZE PRESSURIZER INSURGE ANALYSIS'ASSUWPTIONS:
102% OF 94% RATED THERWAL POWER (3220 WWT)
Tavg WINUS UNCERTAINTIES
~~~
NOWINAL PRESSURIZER LEVEL PLUS UNCERTAINTIES PORV'S OPERABLE i WAXIMUM PRESSURIZER SPRAY i
SG LEVEL AT 61%
l e RESULTS WAX. PRESSURIZER VOLUWE FOR BOTH CASES REWAINS BELOW PRESSURIZER FULL VOLUWE OF 1865,CU. FT.
FSAR' ANALYSIS REWAINS VAllD
1 .
- BYRON PHASE 1 T-HOT REDUCTION SAFETY EVALUATION .
~1 e RCCA BANK WITHDRAWAL AT POWER FULL POWER CASE DNB LIMITING NEGLIGIBLE EFFECT ON 10% POWER CASE 60% POWER INITIAL CONDITIONS CHANGE ANALYSIS ASSUMPTIONS:
RCS PRESSURE AND Tavg AT 60% OF FULL POWER e RESULTS PRESSURIZER DOES NOT OVERFILL
, DNB DESIGN BASIS WET e
FSAR ANALYSIS REMAINS VALID l
L__.___.- - - . _ .
~b
~
BYRON PHASE I T-HOT REDUCTION SAFETY EVALUATION
- BASIS FOR 10 CFR 50.59 APPUCABluTY
.__ FSAR DESIGN BASES MET FOR ALL CASES ANALYZED
, NO PROTECTION SYSTEM SETPOINT CHANGES NECESSARY
- t NO TECHNICAL SPECIFICATION CHANGES REQUIRED PLANT WILL NOT OPERATE OUTSIDE ANALYZED CONDITIONS NO UNRESOLVED SAFETY ISSUES IDENTIFIED l
,--,,----,,.-----------w-- - ' ' ' ' " ' - ' ' ' ' ' ' ' " ~ ~ ' ~ ~ " * ^ ' ' ' '
g4_,_ w . _ _ , - - - - - - - - - - - - - - - -
O b O e
TYPES OF CHANGES TO BE IMPLEENTED
. INSTRUENT CHANGES
. PROCEDURE CHANGES TRAINING 0F LICENSED SHIFT OPERATORS
- a.-
se .
t O
I i
- 01/0013
T
~;
INSTRUMENT CHANGES
. STEAM 6ENERATOR LEVEL
. TAVE PROGRAM NIS ADJUSTMENTS
. NOMINAL TAVE TERM IN OP AT, DTAT SETPOINT EQUATIONS
. STEAM DUMP CONTROLLER
.n
= *
- 01/013
^
PROCEDURE CHANGES i
. ANNUNCIATOR RESPONSE
. INSTRUMENT CALIBRATION ALL REFERENCES TO REVISED SETPOINTS
. STANDING ORDER ON SHIFT
. 914 % POWER LIMIT
- 6 8
e
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