ML20154J588
| ML20154J588 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/14/1988 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20154J579 | List: |
| References | |
| 5102K, NUDOCS 8809230077 | |
| Download: ML20154J588 (55) | |
Text
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l ATIACIMENT C ) .; i f, 1
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, LIST or AFFECTED SCHl(LCAL SPECIFI(*ATION PAGES i ,
Changed Paga_1 Comment l XIX l 2-4 Table 2.2.1-1 l
l B2-2 ;
l 82-3 l i B2-4 Deleted B2-5 Deleted B2-6 Deleted B2-7 Deleted ,
3/4.1-16 3/4.2-1 3/4.2-2b (New) Figure 3.2.3-lb ,
l 3/4.2-3 1 3/4.2-4 Insert "A" j t
3/4.2-5 Redrawn Figure 3.2.3-la ,
l 3/4.2-Sa (New) Figure 3.2.3-1b l
3/4.2-7 Insert "B" l
3/4.3-39 l l t l
3/4.3-53 Insert C1 & C2 to Table 3.3.6-2 l
3/4.3-54 Table 3.3.6-2 (Cont'd) .
3/4.4-1 f 3/4.4-2 Insert "D" i 3/4.4-6 3/4.6-1 l
3/4.7-34 Insert "E" 3/4.10-4 Deleted !
3/4.10-7 Deleted l
B3/4.2-1 Insert "F" l
B3/4.2-2 Deleted 83/4.2-3 Insert "G" (2 Pages)
I B3/4.2-4 B3/4.2-6 B3/4.3-3 Insert "H" B3/4.4-1 Insert "I" B3/4.5-2 B3/4.7-5 Insert "J" 5-4 Design Features, i Insert "K" l 0809230077 000914'~ i PDR ADOCK 05000374 :
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AUACIMENLD EAQEQSED TECHNICAL SPECITICATICti CIIM(GES i
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LIST OF FIGURES FIGURE PAGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /
CONCENTRATION REQUIREMENTS ........................ 3/4 1 21 3.1.5-2 SODIUM PENTABORATE (Na B 016 " 10 N20) 2 10 VOLUME / CONCENTRATION REQUIREMENTS ................. 3/4 1-22 3.2.1-1 MAXIMUM AVEAAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB176 8CRB219 an 8CRB071 ........................,.........,...d ....... 3/4 2 2
~, 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE FUEL TYP BPSCRB299L. .............................,.........E .
3/4 2-2(a) 3[2.3-lo. MINIMUM CRITICAi. POWER RATIO (MCPR) VERSUS t AT RATED FLOW .................................. 3/4 2-5 l
3.2.3-2 Kf FACTOR .........................................
3/4 2 6
' 3.4.1.1-1 CORE THERMAL POWE'R (% OF RATED) VERSUS TOTAL CORE
(.
~
' FLOW (% OF RATED) .................................. 3/4 4 2a 3.4.6.1-1 MINIMJM REACTOR ViSSEL HETAL TEMPERATURE VS. REACTOR VESSEL PRESSURE ....................... 3/4 4-19 4.7-1 SAMPLE PLAN 2) FOR SHUBEER FI,NCTIONAL TEST ........ 3/4 7 33 B 3/4 3-1 REAC',0R VESSEL WATER LEVEL ........................ B 3/4 3-)
8 3/4.4.6 1 CALCULATED FAST NEUTRON FLUENCE (E)1HeV at 1/4 T
..........)...........
l i
AS A FUY110N OF SERVICE LIFE 8 3/4 4-7 i
'5.1.11 EXCLUSION AR!A AND SITE BOUNDARY FOR GASEOUS '
l 1 i AND LIQUID EFFLUENTS .............................. 52 ,
' 5.1.2-1 LOW POPULATION ZONE ............................... 53 i
' 6.1 1 CORPORATE MANAGEMENT ..............................
l 6 11 l f. 1- 2 UNIT ORGANIZATION ................................. 6 12 6.1-3 i
3.A.5-tb MINIMUM SN1FT CREW COMPOSITION ........... =6.... 13 j m:rvsmurn e.gzTte.nu PowtA ft6To (ov PR) v ch15 4" I l AT uwD Ct.wJ Wat tNO cp c.ur.g ag4,incJ t.sLT44 i
(- Pum) VRIP LNb HAIN 'b'A.blN C hv P AM wst'tr tM I t
't N09 tth ik.E . . . . . . . . . .........~... .
. . . . . . . . bN 5 ' 0 **
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mAutmum MtMVE PtAN A8L LINGR W EAT 6CWCM ATII.N Elk W. L m k9t. % E VCR$u$
sunu % ves ANER% E Pt.AN AR 81 go3w . . . . . . . . . . . . . . . .'..C SUR E e. . .. pig 2.4 LA SALLE - UNIT 2 XIX Amnchent No ~5 2.
~. .m TABLE 2. 1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES g 1. Intermediate Range Monitor, Neutron Flux-High i 120 divisions of i 122 divisions -
- . full scale of full scale z.
p 2. Average Power Range Monitor: -
- 4. Neutron Flux-High, Setdown i 15% of RATED 1 20% of RATED THERMAL POWER THERMAL POWER i b N. Flow Biased Simulated Thermal Power - Upscale 0SoM*S9%
-* 1) Two Rectrculation Loop Operation o.50 c/ -* M%
1 t
a) Flow Blased -<maximum N with of a -< 0.'//.:
maximum of
. with a b) High F1w Clamped i 113.5% of RATED 1115.5% of RATED THERMAL POWER THERMAL POWER
- 2) Single Recirculatlan Loop Operation assa/ #SK.$% o.sa *-S v a %
a) Flow Blased -< 4.C'! + 5.7% with -< -0. T: . ^;.7% with a maximum of a maximum of b) High Flow Clamped i 113.5% of RATED 1 115.5% of RATED THERMAL POWER THERMAL POWER
- c. Fixed Neutron Flux-High < 118% of RATED -< 120% of RATED THERMAL POWER THERMAL POWER
- 3. Reactor Vessel Steam Dome Pressure - High i 1043 psig i 1363 psig
- 4. Reactor Vessel hter Level - Low, Level 3 > 12.5 inches above > 11' inches above instrument zero" instrument zero*
- 5. N in Steam Line Isolation Valve - Closure 1 8% closed i 12% closed
- 6. Main Steam Line Radiation - High 1 3 x full 1 powe.
3.6 x backor9und full power background
- 7. Primary Containsent Pressure - High 1 1.69 psig i 1.89 ps'3
- 8. Scram Discharge Volume Wter Level - High i 767' Sk" $ 767' 31s"
- 9. Turbine Stop Valve - Closure 1 5% closed i 7% closed g 10. Turbine Contro! Valve Fast Closure, C
g Trip 011 Pressure - Low . t 500 psig 1 414 psig C N.A. M.A. p k 11. Reactor Mode Switch Shutdown Position
,5 12. haual Scram N.A. N. A- C g 13. Control Rod Drive 1 113s M fg
& a. Charging W ter Header Pressure-Low '
3 1157 psig
- b. Delay Timer i 10 seconds i IC seconds
- See Bases Figure B 3/4 3-1.
f .
00 20 SAFETY LIMITS w
8ASES 2.1.2 THERMAL POWER. Hiah Pressure and Hioh Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters
, which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognize < that a departure from nucleate bolling would not necessarily result in damage tu BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for W ich more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit MCPR is determined using the General Electric Thermal 8
Analysis Basis, GETA8 , which is a statistical model that combines all of the uncertainties in operating pa'rameters and the procedures used to calculate crit,1 cal power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L),
GEXL correlation. , ,
Tb 0'.XL $ rbhth i: ;: lid ere- t' -egeaf :: dith : =;d h the t :te vi A d:,t: :::: t: 2;;h; th: ::T r; hth- "::t ce 9 tf :: Or:4
?-;;;ure. !^^ t; i t00 ;,;irr 6
g, q g . 796 73fh7 ,L .2
"::: "? r 0.' 10 hht 0- 2;;;1i74 O i. 1GG ^iv/is-L::: Te.,inv i.6i .6 . w. ,,er r:d t: - -
I.t7 ;0 : ir0;ri0r 7;d *
- a. "General Electric BWR Thermal Analysis Bases (GETA4) Data, Correlation ,
and Design Application," NE00 10958-A.
l LA SALLE - UNIT 2 8 2-2
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00 21 l l
l' SAFETY t.!MITS BASES i
THERMAL POWER. Hiah Pressure and Hich Flow (Continue 4)
, ,e -
Axial Peakina: Shape Max /Ava.
Unifom 1.0 Outlet Peaked 1.60 let Peaked 1.6 -
Double ak 1.46 and 1.38 Cosi 1.39 Rod Array 64 Rods in an 8 x ray The requir nput to the statistical model are the un rtainties listed !
in Bases Tab 2.1.2-1, the nominal values of the core parame rs listed in Bases Tab B2.1.2-2, and the relative assembly power distributio shown in Bases le 82.1.2-3. Bases Table 82.1.2-4 shows the R-factor distr utions tha are input to the statistical model which is used to establish the ety it MCPR. The R-factor distributions shown are taken near tae beginning h t i cycia l Thebgsesfortheuncertaintiesinthecoreparametersaregivenin NE00-20340 and a the basis for the uncertainty in the GEXL correlation is given in NE00-10958-A . The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe
., as the distribution used in the analysis.
l l
- a. "General Electric BWR Thermal Analysis Bases (GETA8) Data, Correlation and Design Application," NED0-10958-A.
- b. General Electric "Process Computer Perfomance Evaluation Accuracy"
! NED0-20340 and Admondment 1, NEDC-20340-1 dated June 1974 and 1
December 1974, respectively. .
LA 5ALLE - UNIT 2 ,
B 2-3 ,
l 00 22 )
(
BASES TABLE 82.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION ,
OF THE FUEL CLADDING SAFETY LIMIT
- STANDARD DEVIATION '
QUANTITY
(% of Point)
Feedwater Flo 1.76 Feedwater Tempe ture 0.76 r Reactor Pressure 0.5 t
'/ )-
i f ,
Core Inlet Temperature 0 W "
/
g s
Core Total Flow L '
Two Recirculation Loop peration 2.5 h'
Single Recirculation Loo Operation 6.0
~
, Channel Flow Area 3. 0 Friction factor Multiplier 10.0 Channel Friction Factor Multiplier 5. 0 TIP Readings Two Recirculation Lo Operation 0.7 i Single Rectreulati Loop Operation 6.8 R Factor 1. l Critical Power 3. 6
]
I r
- fe uncertainty analysis used to establish the core wide Safety Limi MCPR 5 based on the assumption of quadrant power symmetry for the reactor ore.
The values herein apply to both two recirculation loop operatien and si le
[ recirculation loop operation, except as noted.
LA SALLE - UNIT 2 8 2-4
- Amendment Ne 32 e
00 23 ;
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Bases Table 82.1.2_22 NOMINAL VALUES OF PARAMETDS IN .
THE STAT ICAL ANALYSIS OF FUEL CLAODF.C INTEGRITY SAFETY LIH!T THERMAL POWER 3293 MW l
Core Flow 102.5 M1b/hr l Dome Pressur 10.4 psig R-Facto 1.038 0 GWD/t 1.031 - /t 1.030 - 15 t 1.033 - to GWD/t f
DELE ~E PAGE i
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t LA SALLE - UNIT 2 B 2-5 -
Mendment No. 32 l
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Bases Table 82.1.2-3 RELATIVE BUNDLE POWER DISTRIBUTION USED IN THE GETA8 STATISTICAL ANA $
Perc el Bundles Within Range of Relative Bundle P e / Power Interval 1.375 to 1.425 5.1 1.325 to 1.375 1.275 to 1.325 1.225 to 1.275
\ 7.1 7.8
- 9. 8 1.175 to 1.225 7. 3 1.125 to 1.1 .8 1.075 to . 25 .
1.025 1.075 4.7
<1.
- 41.5 1E E L:~Th s
)
OF66 LA SALLE - UNIT 2 826 Amendeent Nc. 32
00 25 C
1
, Bases Table 82.1.2-4 R-FA O!$TRIBUTION USED IN GETA8 STATIST L ANALYSIS 8x8 Rod Array R-Factor Rod Sequence No.
1.038 3 1.038 2 1.037 1.035 /
1.030 / f i 1.030 / 8 through 64 l
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! LA SALLE UNIT ? 827 Arendment No. 32 1
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00 26 i
REACTIVITY CONTROL SYSTEM 3/4.1.4 CONTROL ROD PROGRAM CONTR0t$
k R00 WORTH MINIMIZER LIMITING CONDITION FOR OPERATION 3.1.4.1 The rod worth minimizer ( M ) shall be OPERA 8LE.
4 APPLICA81LITY: OPERATIONAL CON 0!TIONS 1# and 2"#, when THERMAL POWER is less l than or equal to 20% of RATED THERMAL POWER, the minimum allowable low power setpoint.
ACTION:
- a. With the M inoperable, verify control rod movement and compliance with the prescribed control rod pattern by a second licensed operator or other techni: ally qualified member of the unit technical staff ,
who is present at the reactor control console. Otherwise, control '
rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.
- b. The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.1.4.1 The M shall be demonstrated CPERABLE: [ $74 #[
- a. In OPERATIONAL CONDITION 2 prior to withdra 1 of c frodsfor the purpo', d in OPERATIONAL g CONDIT! h ,'prior of makingto the reactor critical, N'TEtst99w hen reducing THERMAL 8 POWER, by verifying proper annunciation of the selection error of at least one out-of-sequence control rod.
- b. In OP'5 RATIONAL CONDITION 2 prior to withdrawal of control rods for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of .
sequence control rod.
I
- c. In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after M autoe4 tic i initiation when reducing THERMAL POWER, by verifying the rod block l function by demonstrating inability to withdraw an out-of-sequence !
- d. By verifying the control rod patterns and sequence input to the M l i computer is correctly loaded following any loading of the program into the computer.
, "Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERulLITY of the M prior to I withdrawal of control rods for the purpose of bringing the reactor to j criticality.
l LA SALLE - UNIT 2 3/4 1-16 Amendment No. 30
00 27
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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION and 3.71-3 3.2.1 nil AVERAGE PLANAR LINEAR HEAT GE ERATION RATES (AFLHGRs) for each type of fuel as a function of AVERAGE PLANAR XPOSURE shall not exceed the limits
'h!*," ,I".bi"'!! _d'.!!}g3.2y 2g (( f.. M y.m L 2.,g,
;;:::.- '.~ n_,7 <:: i: .i:....,.
- 1:'...E:1,:J:
C' '"'""'"'""r APPLICABILITY: OPERATIONAL CON 0! TION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION o,nd 3.'2. l-3 With an APLHGR exceeding the limits of Figures 3. 2.1-1,4*+ 3. 2.1-2,^i ni ti a te corrective action within 15 minutes and restore APLHCR to within the required '[ i limits within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMA POWER within the next 4 hours. SURVEILLANCF REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limit; determined from Figures 3.2.1-Ig a 4 3.2.1-2 4 3 and 3. 2.1 -3:
- a. At least once per 24 hours, '
- b. Within 12 hours af ter completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and
- c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
l LA SALLE - UNIT 2 3/4 2-1 Anend.ent No, 3; _ _ _ ._ -- - +
~
g MAPLHGR Vs Average Planar Exposure ~ E Fuel Types BC300D and BC3208
, 14
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I 13
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/. -
's %
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/ %
12 / -
~
8C3000 - s
' BC320C
-~11 3 s 5 10 s 7
" { o
. Either the carves of thi,s figure or the latti,ce specific curves in GE g
NN a 9 decament liEDC-31510P shall be used~ x 'st
- =s
. for Incks'ical Specification 3.2.1. \
8 ,
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6 0 " ' ' ' ' ' ' '10 0 0 0 20000 30000 40000 50000 AveragePlanorExposure(mwd /St) figure 3.2.1-3 $ w
l , 00 29 POWER O!$TRIBUT!0N LIMITS 3/4.2.2 APRM SETPC!NTS C~ , LIMITING CON 0! TION FOR OPERATION i t i 3.2,1 The APRM flow biased simulated thermal powr-upscale scres trip setpoint l (5) and flow biased cfoulated thermal power-upscale control rod block trip 1 setpoint ($,g) shall be estabitshed according to the following relationships:
- a. l Two Recirculation Loop Operation 5 less than or equal to fWi:55P1 DNF (O.So W + 59 'f,) T S gg lesa than or equal to19;ssicamsg5'g (o.51W +"*fe) T
- b. $1ngle Recirculation Loop Operation
)
$ lesa than or equal to O ^^" s 6.O)T ( c. E's W + 54.3 *fe I T l
Rs less than or equal te (^.!*"a "E)T- ( 0. 5'8 w + 4 2. 3 *f. )T S where: $ and 3R8 are in percent of RATED THERMAL POWER, W = Loop meirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 108.5 million Ibs/hr, ( T = Lowst value of the ratio of FRACTION OF RATED THEPMAL POWER divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY. T 1s always less than or equal to 1. , o,,, %g g g ,o . ( APPL"CA8!LITY: OPERATIONAL CON 0! TION 1, when THERMAL POWER is greater than or equa to 15E of RATED THERMAL POWER. ACTION: With the APRM flow biased staulated thermal power-upscale scram trip setpoint and/or the flow biased steulated thermal powr-upscale control rod block trip sntpoint set less conservatively than $ or 3 as above determined, initiate cornctive action within 15 minutes and restNe, 5 and/or $ to within the requiredlimits"within2hourserreduceTHERMALPOWERto9essthan25%of RATED THERMAL POWER within the next 4 hours. ' SURVE!LLANCE REQUIREMENTS ___ t.2.2 The FRTP and the MFLPD for each class of fuel shall be deteisined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal powr-upscale scram and control rod block trip setpoint verified to be i within the above limita er adjusted, as required: , s. At least once per 24 hours, i
- b. Within 12 heurs after completion of a THERMAL POWER increase of at ;
least 155 of RATED THERMAL POWER, and i
- c. Initially and at least once per 12 hours when the reactor is operating
- with MFLP0 greater than or equal to FRTP.
"Wita NF' LPD greater than the FRTP up to 90% of RATED THERMAL POWER, rather than {
adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM read-
- ings are greater than or equal to 1005 times MFLPO, provided that the adjusted
- APRM reading does not exceed 1005 of RATED THERMAL POWER, the required gain t
adjustment increment does not exceed 105 of RATED THERMAL POWER, and a notice of the adjuntaent is posted on the reactor control panel. i LA SALLE - UNIT 2 3/4 2-3 Amendment No.17 f I
1 l 1 00 30 POWER DISTRIBUTION LIMITS i l 3/4.2.3 MINIMUM CRITICAL POWER RATIO
)
l i LIMITING CONDITION FOR OPERATION 3.2 A C "'":""" ca:T:C n te e ti it e ---<--e wa = :S:M ; e :n :: ; m te r ;r;. u r I
< :- ei .;; :.:.: ; e x t - x :; u r;ia;: o Ifr;: Fig..; 2.2.3-2 f:ru*** to: r::'r !:tI;a 1 vy .y;r: tie crd :I:!' 5 ;;.ei & '
I O re . 'haa the l'ai t- ;;;c-'n;; 7. .. T :Y.. ; 3. 2. 0- 1 ^ 0. 01 t' :: Q::;e..'er _4--g s q -- y 3.3-3 f;7 ;$3;;; 7;; ;;;;;;;7 ;;;; ;3..g43 APPLICABILITY: I OPERATIONAL CONOITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. X fV M ACTION
% 4;T "C " 1::: t' r the W i"k'e uC" rit 0;u rr'n; f r;- "';.r;;
b.2-3-1Ord??1-?,'-fti:t: I
- rr::ti;; ;;;':r. ;!t- 15 mi.;t:: ;;d "C"" te iin:a ;6 r;;.'r:d 'i-it 'ini- ? hau : :: r;t:: '""-
i :::t;r; Y X JE.O 3 ?;; " r 2 5% ; f "' T E 0 "E a"* L "OU " . l th i n i !,eweM N um . m SURVEILLAN*E REQtifRFMFNTS 4.2.3 MCPk, with: a. ' t" " = 0.86 prior to performance of the initial scram time measurerents
- for the cycle in accordance with Specification 4.1.3.2, or b.
t,y, determined within 72 hours of the conclusion of each scram time ! surveillance test required by Specification 4.1.3.2,
)
i shall be determined to L 6 sal to of greater than the applicatie MCPR limit determined from figures 3.2.3-1 and 3.2.3 2:
- a. At least once per 24 hours, i b. Within 12 hours after complet'..n of a THERMAL POWER increase of at least I n cf kATED THERMAL POWER, and 1 c.
Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL R00 PATTERN for MCPR. I l l LA SALLE UNIT 2 3/4 2-4 Amendment No. 32 I
., _ . . _ , _ _ . _ . _ _ _. ___ _ _.__._.___--,_-,m __ _ . _ .
=_.
4 4 00 31 , Insert A 3.2.3 - The MINIMUM CRITICAL TOWIR RATIO (MCPR) shall be equal' to or greater than the MCFR limit determined frons
- a. Single Recirculation Loop Operation Figure 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve 5 for a RBM setpoint of 110%) plus 0.01, times the kg determined from Figurs 3.2.3-2.
; b. Two Re:irculation Loop Operation Figure 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve 5 for a RBM setpoint of 110%) times the f(f determined from Figure 3.2.3-2.
- c. Two Recirculation Loop operation with Main Turbine Bypass Inoperable Figure 3.2.3-1b times the Xp determined from Figure 3.2.3-2, for two re-circulation loop operation, with the main turbine bypass system inoperable per specification 3.7.10 (any RBM setpoint determined per Specification Table 3.3.6-2 may be used).
- d. Two Recirculation Loop Operation with Ind-of-Cycle Recirculation Pump Trip system Inoperable .
Figure 3.2.3-1b times the K circulation loop operation, pwith determined from Figure the end-of-cycle 3.2.3-2, for recirculation twotrip pimp re-system inoperable as directed by Specification 3.3.4.2 (any RBM setpoint determined per Specification Table 3.3.6-2 may be used). t ACTIC4: ( l a. With MCPR less than the applicable MCPR limit as .' ... mined for one of the 1 above conditions:
- 1. Initiate corrective action within 15 minutes, and
- 2. Restore MCFR to within the required limit within 2 hours.
, 3. Othe rwise , reduce THIRMAL POWIR to less than 25% of RATID THIRMAL POWER within the next 4 hours.
- b. When operating in a cond1* ion not identified above, reduce THIRMAL POWER to less than 25% of RATED THERMAL POWER within 4 hours. l l
) I . ) - ' g 1
- - a .
. 00 32 POWER O!S7F 2 K' TION L2 M1TS
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u : t. REPLAW
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li LA SA'.LE N2
. 72 3 .4 .
[ A.9endment No. 32
~
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l-u, POWER IESTRI3UTION LIM::TS
* ~~
MCPR h 1.450
~,; .
m 1.400 1.350 - T 7 Curve B (RBf1 Setpoint = 110%) 1.300 1 30 L'. IL y Curve A (RDri 'leipoint = 106%) 1.27 1.260 1.200 o l
.687 .70 .72 .74 ..76 .78 .80 .82 .84 .86 e . EPR VERSUS T AT RATED FLOW S Figure 3.2.3-10
~
[ POWER DISTRIBljil0N 1.lM"S . i EPR e 1.45
~
1.40 E0C-RPIlnoperable ___ -
~
1*35 # q g= -
;*; se Main Turbine By) ass lno)erab,e _
1.30 1.25 1.20
.687 .70 .72 .74 .76 .78 .80 .82 .84 .85 WCPR VERSUS f AT RATED 10W l Tigure 3.2.3-1b s
00.35 (. POWER O!STRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CON 0! TION FOR OPERATION tnsa.H B 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed;N APPLLCA8IL TY: OPERATIONAL CON 0! TION 1, when THERMAL POWER is greater than or equa' to 2 M of RATED THERMAL POWER. ACTION: ) . With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or reduci THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. 4
$URVEILLANCE REQUIREMENTS 4.2.4 LHGR's shall be determined to be equal to or less than the limit:
- a. At least once per 24 hours,
- b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and 4 c. Intially and at least once per 12 hours when the reactor is operating i
on a LIMITING CONTRUL R00 PATTERN for LHGR. - 1 t l LA SALLE - UNIT 2 3/4 2-7
i i i 00 36 I . i. 1 6 Insert 3 l 1 t t i i
! a. 13.4 W/f t for fuel types !
l I *
- ' 1. SCRS176 i 1
s
- 2. SCR8219 .
o t i 3. BP8CRB299L i
?
I I b. 14.4 W/ft for fuel types ! 1. BC3000 )
- 2. 3C320C t
l
- i d i I t f
t d i i s , l i d I ( i f I ! i . i 3 , i , d i 4 ll a 1 i 1 2 l 1 . 1 1 . I 1
ou av C INSTRUMENTATION END 0F-CYCLE REC!RCULATION PUMP TRIP SYSTEM INSTRUMENTATION-LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (E0C-RPT system instrumenta-tionchannelsshowninTable3.3.4.2-1shallbeOPERA8LEw)iththeirtrip setpoints set consistert with the values shown in the Trip Setpoint column of i Table 3.3.4.2-2 and with the END-OF CYCLE RECIRCULATION PUMP TRIP SYSTEM RESP TIME as shown in Table 3.3.4.2-3. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 3rA of RATED THERMAL POWER. ACTION:
- a. With an end-of cycle recirculation purnp trip system instrumentstion channel trip setpoint less conservative than the value shown in the Allowable Values column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value,
- b. Witn the number of 0PERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within 1 hour. '
- c. With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement (s) for
, one trip systra and:
- 1. If the inoperable channels consist of one turbine control valve ;
channel ard one *urbine stop valve channel, place both inoperable l channels in the tripped condition within 1 hour. ' I
- 2. If the inoperable channels include two turbine control valve channels or two turbine stcp valve channels, declare the trip system inoperable.
- d. With one trip system inoperable, resture the inoperable trip systet Y g to OPERABLE status within 72 hours, reduce THERMAL POWER to less than 30% of RATED THERMAL POWER hin the next 6 hours.
>/
e I h e. With both trip systems inoperable, restore at least one trip syster to OPERABLE status within I hour, reduce THERMAL POWER to less than 30% RATED THERMAL POWER in the next 6 hours.
, ,wm emm voM (inc,W WN 6[$
o pratie n kC/A % h e, EC4.- M ino F # M S--
% g ,. sp.d eawo n 3. 2. . b w it:h i n ti x %" + N# 2Cf M f, Th,e,dneie..miSicns a p oc. sp.cificahc q M4 ce not LA SALLE - UNIT 2 3/4 3-39 A. mend. tent No. 33
- O 0
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LA SALLE - UNIT 2 , 3/4 3-53 Amendment No. 32
4
. 00 39 Insert C1 l a. When using the MCPR 6 0.66W + 37%**
- 0.66W + 40%**
l LCO frora Curve A of , Figure 3.2.3-la or : i the curves froen Figure 3.2.3-lb.
- b. When using the MCPR 6 0.66W + 41%** 40.46W + 44%**
LCO from Curve B of J Figure 3.2.3-la or the curves from j Figure 3.2.3-1b. i i Insert C2 1 4
- a. When using the nCPR tro.66W + 31.7%** 60.66W + 34.7**
LCO from Curve A of i Figure 3.2.3-la
- b. When using the MCPR 20.664 + 35.7%** 60.66W + 38.7%** !
LCO from Curve B of Figure 3.2.3-la f t .l 1 ! l , t t l l i l 1 J j 1 i . c- . -
, . , . - - . - , ~ , - - , m., - . . , . - , . , , , -_ ,,.--.n. - - - . - . - - - - - - . . _ . - - , - - ~ . . . - . , . . . - - . - . . . - - ,
r O l TAttE 3.3.6-2 (Continued) g CONTROL 300 WITHDRAC4L 340CK INSTRUMENTATION SETPOINTS E TS!P FUNCTION TRIP SETPo!NT ALLOWStE VALUE
- 5. Stun oIsCmacE vetuME
*) a. Water towel-High i 765' % " $ 765' %"
, b. Scram Olscharge Valene Switch in typass N.A. N.A.
- 6. REACTOR C00UWT SYSTEN RECISCULATION FLOW
- a. Upscale < 108/125 of full scale < 111/125 of full scale
- b. Ineperettwe R.A. R.A.
- c. Camperator i 10E flow deviation i 11% flew deviatten y aHoWo We val u Po,- *h b ed.m ct
= R
- 10 0 */8 fcca<culabioO IOo p hjo g ( w ) of 1
O C G. O e
00 41 3/4.4 REACTOR COOLANT SYSTER 3/4.4.1 RECIRCULATIDW SYSTEM RECIRCULATION LOOPS , LIMITING CONDITION FOR OPERATION in 3.4.1.1 Two Reactor cociant system recirculation loops shall be operation. , i APPLICABILITY t OPERATIONAL CONDITIONS 1 AND 2
> ACT!CM in
- a. Ytt ne (1) reactor coe,. ant system recirculation .cce.
operation , comply <t*.h Specification 3.4.1.5 ands .
- 1. Within feity 's 4 ) hours Place the recirculation flow control system in the e,
j~ Master Manual mode or lower. and b) Increase the MINIMUM CRITICAL POWER RATIO (MCPRI ] Safety Limit by 0.01 to 1.CS per Specification ' O. L. 2. and i c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) ' Limiting Condition for Operation by 0.01 per Specification 3.2.3, and, d) Reduce the Average Power Range Monitor (APRM) ; Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allevable Values to those applicable l 4
' to single rectreulation icop operation per ~~
3pecifications 2.0.1. 3. 2. 2. and 3.3. ] eC':t:: th: M A NIMVM-*VBAGE-PtAM AR-b 3 MEAA-ME AT-OHM *T!OM M ,T: ::tAFLF;GR i liwrt--to-o- vale- n r uS. I tt;;; the tre-rectrev4*ti+e=lec; spe o uca limet o+i 0;;;&itsau+e--Gr2rtr l t
- 2. The provisions of Specification 3.0.4 are not applicable.
l
- 3. Otherwise, to in at least HOT SHUTDOWN within the next twelve (10) hours.
b. With no reactor coolant recirculation 1 cops in operation: 'f Take the ACTION required by Specification 3.4.1.5 and 1.
- 2. Se in at least HOT SHUTDOWN within the next six (6) hours. .
. ?>. .' y /a q_q
l 00 4E3 SURVETLLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated CPERABLE at least once per 16 months bys
- a. Verifying that the control valve fails *as is' on ices of hydraulic pressure at the hydraulic power units, and i
- b. Verifying that the average rate of control valve movement ist
- 1. Less than or equal to 11% of stroke per second opening, and Less than or equal to 11% of stroke per second closing.
2. 1 l 1 't i I j %r b b] i l I i l l l l 1 I 1 . kJ kd Iff bM U g
l 1 00 43 ' REACTOR COOLANT SYSTEM [~ 3/4.4.2 SAFETY / RELIEF VALVis LIMITING CON 0! TION FOR OPERATION - g - -
/
v.1 m&!! nh= ne: M: ^*'
=1= f=iu riO a u;p:f
"; . ...-. f hd a=niv_.et-nfe-M =r:wr--
Mtt r nsytet u nt & ;;." 0- "C l J
- s. 4
- b. 4 safety /reitef valves 91205 psig + 12. -M i
- c. 4 safety /relfer valves 9 1195 psig + 15. - R -
j d. 4 safety /nlief valves 01185 peig + 15. -M e. safety / relief vr.ives 91175 psig + 1X, -R i 2 safety / relief valves 91350 pstg + 15. -M i APPLICA8!LITY: OPEAATIONAL CON 0!TIONS 1, 2, and 3. ! i .AE19$ , 1 i a. i With the safety valve function of one er more of the above required safety /reitef valves inopenble, be in at least NOT SHUTDOWN within i 12 hours and in COLD SHUTDOWN within the next 24 hours. b. 4 With one er mere safety / relief valves stock open, provided that suppression pool average water temperature is less than 110*F, close the stuck open rettef valve (s : i within 2 minutes er it suppres)s; ten pool average water temperatu! i 110*F er gnater, place the nectar mode swit,h in the Shutdown ' ) posittoa. & a e. reA uire. '
- c. WithSu&r se - yire ne l ve stes positten indicaters )
1 inoperable, rester, the inoperable stas positten indicaters te i } OPEAASLI status within 7 days er be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours, ) j SURVf!LLANCE REQUIREMENTS - l 4.4.2.1 ] valve shall be demonstrated OptRA8LE by performance of a:The safet I
- a. CMAMWL CHECK 4t least once per 31 days, and a
- b. CHANNEL CAL 18AAT!0N at least once per la months.**
- 4. 4. t. 2 The low low set function shall be demonstrated not to interfere with the OPERA 4!LITY of the safety / relief valves or the ADS by perference of a j
CHAMNEL CALIBRATION at least once per 18 months, d j 'The lift setting pressure shall cornspond to ambient condittens of the valves at neelnal operating temperatums and pressures. ! i i
#Up to two inoperable valves may be replaced with span OPERA 8LE valves with lower setpoints untti the next refueling outage.
j **The previstens of Specification 4.0.4 are not applicable provided the surveil-4 lance to perfere is performedtha test. within 12 hours after reacter stone pressure is adequate i l
- LA SALLE - UNIT t 3/4 4 6 Amendment No.15
,I l i
- _ , , . . - - . - . . - - - - . - - . . . - ~ - ..-n,---._,n.,_ _ , - . , - _ - , _ _ _ - - . - , . - - - - - - , ,-. n---~ _,--- - .-
' ' .,y l
INSERT _D 3.4.2 - %e safety valve function of 17 of the below listed 18 reactor coolant systern safety / relieve valves shall be OPERABLE with the specified code safety valve function lift setting *l s all installed valves shall be closed with OPERADLE position indication. l l l 1 5102K i
I 00 45 ( 3/4.6 CONTAI N NT SYSTEMS l 3/4.5.1 PRIMRY CONTAINMENT PRIMARY CONTAINM(NT INTEGRITY t!NITING CON 0! TION FOR OPERAYION t
- 3. 6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be maintained.
, APPLICA8!L!H: OPERATIONAL CON 0I'TIONS 1, 2,8 and 3 ACTION:
! Without PRIMARY CONTAINMENT INTEGRITY, restore PRIMARY CONTAINMENT INTEGRITY withi.' I hour or be in at least HOT SHUT 00W within the next 12 hours and in ; COLD SHUTDOW within the following 24 hours. (
SURVEILLANCE REQUIREMENTS 4 j 4. 6.1.1 PRIMARY CONTAINMENT INTEGRITY shall be demonstrated: ' 1 j a. 1 After each closing of each penetration subject to Type B' testing, except the primary containment air locks,1 f opened following Type A 4 by leak rate tasting the seal with gas at Pa. 39.5 psig 4 or8 and verify test,ingthatwhenthemeasuredleakagerateforthesesealsIs i added to the leakage rates determined pursuant to Surveillance l Requirement 4.6.1.2.d for all other Type 8 and C penetrations, the i coatined leakage rate is less than or equal to 0.60 La. '
- b. At least once per 31 days by verifying that all primary contairwent penetrations ** not capable of being closed by 0PERA8LE containment automatic isolation valves and required to be closed during accident ;
conditions are closed by valves, blind flanges, or deactivated i automatic valves secured in position, except as provided in Table 3.6.3-1 of Specification 3.6.3. l
- c. By verifying each primary containment air lock OPERA 8LE per
- Specification 3.6.1.3. !
l i
- d. By verifying the suppression chamber OPERA 8LE per Apecificatien J
3.6.2.1 , l ' i "5ee 5pecial Test Exception 3.10.1 i ) **Except valves, blind flanges, and deactivated automatic valves which are located inside the contaiwent, and are locked, saaled or othenvise secured j in the closed position. These penetrations shall be verified closed during sach COLD SHUTDOWN except such verificatica need not be performed when the i primary containment has not been deinerted since the last verification or 1 tore often than once per 92 days. 5!;::16T::t 5:ep44+e-MeKdr-l LA SALLE - UNIT 2 3/4 6-1 i
\
00 4G
, PLANT SYSTEMS 3/4.7.10 MAIN TUR$!NE SYPAS$ $YSTEM t.!NITING CON 0! TION FOR OPERATION 3.1.10 The sain turbine bypass systes shall be OPERA 4LE.
APPt.1 Call! .!TY: OPERATIONAL CONO!T[0N 1, when TMRMAL POWR is greater than or eque w 25X of RATED TM RMAL POWER. #
/ -
ACT ON1 2 3 i:.. = h L.; k. ;,- ::: ;,;t:: ' :;: c h , W 2'- O b:r: . tM ey-tr :: 2 "" l' :' r :r r:: x ; "'O r a . T : . i... e .. 2^4 Q.2:--* ^ T:" = ~ ~ we -h m. . .;
- t:ca s n su- + E/
$URVE!Lt.ANCE REQUIREMENTS 4.7.10 The sain tuttiine bypass systes shall be demonstrated CPERA8LE at least once per:
- a. 7 days by cycling each turbine bypass valve through at least one complete cycle of full travel.
- b. 18 months by:
- 1. Performing a systee functional test which includes simulated stuceatic actuation and verifying that each automatic valve actuates to its correct position.
- 2. Demonstrating TUR$1M 8YPA$$ SYSTEM RESPONSE TIME to be less than or equal to 200 milliseconds.
e LA SALLE - UNIT 2 3/4 7-34
i o . i .
. ,7 u
1 ImssmT_s i I L i r i l i ! I A. With the main turbine bypass system inoperable f q 1. If at least four bypass valves are capable of accepting steam flow { j per Surveillance 4.7.10.at j l a. Within 2 hours, either f 4 i ! 1) Restore the system to OPERABLE status, or l l j 2) Increase the MINIMUM CRITICAL POWER RATICH (MCPR) Limiting h i Condition for Operation (LCO) to the main turbine bypass [ inoperable value per Specification 3.2.3. ll l b. Otherwise, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. i 2. If less than four bypass valves are capable of accepting steam flow ( i per Surveillance 4.7.10.at ! l I I
- a. Within 2 ho*Jrs. increase the MCPR LCO to the main turbine bypass f
} inoperable value per Specification 3.2.3, and l 3 i
- b. Within the next 12 hours, restore the system to OPERABLE status,
- c. Otherwise, reduce THERMAL POWER to less than 25% of RATED f THERMAL POWER within the next 4 hours.
l ! I i B. The provisions of Specification 3.0.4 are not applicable. t l l i ! I 1 l j i f l I l ! l I I i 5102K : ) ! l I s 9
00 48 ( ECIAL TEST EXCEPTIONS l
. 3/ .4 RECTRCULATION t.00P5 DtACD i
I LIN! TING ON0! TION FOR OPERATION 3.10.4 The re frements of Specification 3.4.1.1 that recirculation loops be in operation awy suspended for up to 24 hours for the, performance oft ;
- a. PHYSICS TS, provided that THERMAL POWER does not exceed 5X of RATED THE L POWER, or
] b. The Startup i t Program, i l APPLICA8!LITY: OPERATIONAL. CON 0!TIONS I and 2, during first fuel cycle PHYS!C$ .
. TE5T5 and the Initial Startu Test Program. '
ACTION:
- a. With the above specified ime limit exceeded, insert all centrol rods.
! b. With the above specified THE L POWER limit exceeded, immediately ! place the reactor mode switch n the Shutdown position.
$URVE!LLANCE REQUIREMENTS . 1 4.10.4.1 The time durine which the above specift requirement has been suspended shall be verifled to be less than 24 hou at least once per hour during PHYSICS TESTS and the Startup Test Program.
- 4.10.4.2 THERMAL POWER shall be determined to be less an or equal to 5X of RATED THERMAL POWER at least once par hour during PHY C5 TESTS.
JEL=T$ f~ u - RGE LA SALLE - UNIT 2 3/4 10-4
. 00 49 l i
\$PECIALTESTEXCEPTIONS 3 10.7 CONFIRMATORY FLOW IN00CEO v!BRATION TEST LIM! CON 0! TION FOR OPERATION 3.10.7 provisions of Specifications 3.6.1.1 and 7.3 may be suspeaded to permit he drywell head to be removed and the RC systes to be incperable with a nit n supply line connected to the react vnssel at the RCIC i injection c nection in order to perfore the conf tory flow induced i vibration tes prior to first reactor criticali . In addition, the provisions of he following specifications whi ~ are applicable during j HOT SHUTDOWN sa be suspended so that the uni may be brought to HOT SHUTDOWN and sai tained in HOT SHUTDOWN for duration of the test by non-nuclear heat provided that initial ctor criticality has not l occurred. Upon s essful completion of test or initial reactor '
criticality, whiche r occurs first, thi specification is cancelled. b
- a. Specificatio 3.3.2, Table .3.2-1 for Trip Function A.1.c.1, Main team Line R diation - High Monitor.
- b. Specification 3.7.10. T le 3.3.7.10-1 for Instrument 1.a., Liquid Ra este Cf luont Line Monitor.
- c. Specification 3.3. 11 Table 3.3.7.11-1 for Instrument 1.a. Moble Gas Act i Monitor,
- d. Specification 3.4.3. for the primary containment atmosphere particulate and ga s radioactivity monitoring systems.
- e. Specification 3.5 fo the A05 valves and "B" 1,PCi toop.
- f. Specification 3. 1.1, 6.1.2, 3.6.1.3, and 3.6.1.4 ;
9 Specification 6. 2.1.
- h. Specification .6.3. Table .6.3-1 for valves in a.1, kain Steae ;
Isolation Va est a.3, Reac r Coolant System Samph Li/4 Valves; a.10, LPC3, CS, RCIC, and Injection Testa 6?n Ct,eck liyp4H ; Valves; a. , Drywell Pneumatt Valves; and a.14, T!P Guide Ta e Valve la Valve. ,
- i. Specif1 tion 3.4.3.2.d. isolatio valve leakage f6r "B" LPCI check Ive 1E12 F0418. ,
I APPLICA8!LITY OPERATIONAL CON 0! TION 3, during p fonsance of the confirmatory , ITE induc vibration test. ' ACT ON: ith the provisions of the above specification not satisfied, be in within 24 hours.
$URV6 LLANCE REQUIREMNTS s
4 0. 7 The reactor shall be verified not to have been critical w any wel assembly presently in the core within 24 hours prior to perfo nce of the test. - C)E LE T6 P%G ' LA SALLE - UNIT 2 3/4 10-7
1
\
00 50 i . (
- 3/4.2 POWER O!$TRIBL/ TION LIMITS BASES l .
The specifications of this section assure that the peak' cladding temperature following the postulated design basis loss of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.44. i 4
' i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE '
This specification assures that the peak cladding temperature following the postulated design basis loss of coolant accident will not exceed the limit specified in 10 CFR 50.44. TN. speeAce on awe assars.s t%t Ed red
). mechanical intsgrity is maintained during ree,vut and tynnskt oPertatten.s. l i
The peak cladding temperature (PCT) following a postulated loss-of-coolant i { accident is primarily a function of the average heat generation rate of all 1 i the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak l clad temperature is calculated assuming a LNGR for the highest powered rod which is equal to or less than the design LNGR corrected for densification. This LNGR times 1.02 is used in the heatup code along with the exposun dependent steady-state gap. conductance and rod-to rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) it this _MGR of the highest powered rod divided by its local peaking factor. ! i ; 'initin ^
=, .;... _ _ . . . : . _ _ Z.1
^
i 1::; :; c;^.g i ;;. ;;!;;,..
" r r. ..
. ;t.;i! 5: 9 t';ti;; ;, e f ^.
Of 0.'" f; :' ;!: i j
-- f -;;t :t' -- ' - ; :;; . ;;' .c..
; .!tig iac ie 4.;
m.. - ^ , r ;:-f: = ! I 3 N::;t' ' Mit! ; :t' ;; ^ :!pi; Mt_..c 1; c..;,;.;;t';c, ; -
@ ?; r;;'r:x_!^.t'--
values =rws Acet fee,tMa ru,Leod InitialdA The calculational procedure used to establish the APLHGR s on Figuns ! . y 3 1 1-1-4+ based on a loss-of coolant accident analysis. The analysis was i y P rTorned using General Electric (GE) calculational models which are con-i l I V sistent with the requirements of Appendix K to 10 CFR Part 50. A complete O g[ discussion of each code employed in the analysis is presented in Refonnce 1. Differences in this analysis compared to pnvious analyses performed with Reference 1 an (1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figure 3.2.1-1, (2) fission product decay is computed assuming an energy release rate of 200 MeV/ fission; (3) pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; and (4) the effects of core spray entrainment and counter-j current reflooding flow liettation as described in Reference 2. are included in the calculations. 3 i Mr.: r: et : " = t g =t :- -- m ur: e in : ... .: a .,,, 4 gp / { -t r: h:!: 9 ; r r e: '- := = m :. ; :.:.: 1 ) i i LA SALLE UNIT 2 8 3/4 2-1 . l
, ~ _ _ _ - . . -..- ,. - -.-- - ---_,_- - - _ _ _ _ _ - _- - _-__.. -_.- - - - _ _ - -
4 1 1 00 51 i Insert T The AFDIGR values for the reload fuel shown in Figure 3.2.1-3 are based on the fuel thermal mechanical design analysis. The improved SAFER /GESTR-LOCA analy-sis (Reference 3) performed for Cycle 3 used bounding MAPLHGR values of 13.0 and 14.0 kv/f t, independent of nodal exposure. These MAPLHGR values are higher than the expected "thermal-mechanical MAPLHOR" for both BP8x8R and GE8x813 fuel. Therefore, SAFER /GISTR established that for all SP8x8R and GI8x8E8 fuel designs . the MAPLHGR values are not expected to be limited by LOCA/ECCS eensiderations. ! However, MAPGGR values are still required to assure that the LHGR limits are i not componised and, consequently, fuel rod mechanical integrity is maintained, i I I l f 1 1 , f 4 l 1 i l t i ! i 1 I 1 1 1
00 52 i Bases Table 8 3.2.1-1 . j $!GN!FICANT INPtJT PARAMETERS 70 THE ; L053-0F-C00LANT ACCIDENT ANALYSIS gg Plant Paramete + j i Core THERMAL R .................... 3449
- which corresponds to 1 of rated steam flow i Vessel Steam outp .................'.. 14, 7 x 106 lba/hr which f 1 cprnsponds to 1055 of rated team flow 9
Vessel Steam Dome Pressu ............ 1055 psia Design Basis Recirculation (ne Break Area for j i Large I naks 3.1 ft 2, j a. j ' .j b. Small I naks 0.'10 ft . ' t l Fuel Parameters:
! EAK TECHNICAL INITIAL
$PECIFICATION DESIGN N!NIMUM LINEAR HEAT IAL CRITICAL FUEL BUM 0 GENERATION RATE P ING POWER j FUEL TYPE GE0METp (kw/ft) FACyR RATIO t to4t4ai Core e vi u.4 1. 4 \ 1.1. l l
A more detailed 11 ing of input of each'model and its sou is p nsented in Section !! of ofennce 1 and subsection 15.0-1 of the FS l *This power le 1 meets tt.e Appendix requirement of 103. The re i heatup cale ation assumes a bundle power consistent with operat n of
; the highe powered rod at 103 of its Technical Specification L! R T!ON RATE Itait, )
HEAT GEN
]
- P
- [
r C
. L,,
,1 i
==== 1* 4 1
i s LA $ALLE - UNIT 2 8 3/4 2-2 . 1 j 1
l 1 l . 00 sa ) ( POWER O!$TRIBLITION SYSTEMS I SASE5 ' i 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based
' on a power distribution which would yield the design LHGR at RATIO THERMAL POWER. The flow biased cieulated thermal power-upscale scraa setting and con- 1 tro1* rod blecit functions of the APM instruments for both two recirculation '
' loop operation and single recirculation loop operation must be adjusted to ensure that the MCPR does not become less than the fuel cladding safety limit or that ! ),11 plastic strain does not occur in the degraded situation. The scram settings and rod block settings are adjusted in accordance with the formula in this speci- l fication when the combination of THERMAL -POWER and MFLP0 indicates a higher 1 peaked power distribution to ensure that an LHG4 transient would not be 1 increased in the degraded condition. ' I , j 3/4.2.3 MINIML,M CRITICAL POWER RATIC 1 1he mquired operating limit MCPRs'at steady-state operating conditions i j as specified in Specification 3.2.3 are derfved from the established fuel 4 cladding integrity Safety Limit MCPR and an analysis of abnormal operational J transients. For any abnormal operating transient analysis evaluation with the i initial condition of the reactor-being at the steady-state operating limit, it I is required that the msulting MCPR does not decrease below the safety Limit ! MCPR at any time during the transient assuming instrument trip setting given '
. i in Specification 2.2. i To assure that the fuel cladding integrity Safety Lielt is not escoeded (
during any anticipated abnormal operational transient, the most limiting ; transients have been analyzed to deternine which result in the largest reduc- ; tion in CRITICAL POWER RATIO of flow, increase in pressure (CPR). The type and power, of transients positive reactivityevaluated insertion,were and loss coolant temperatun decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Lieft MCPR, the required einious operating l liett MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.3-1ct.. , pM0 ~ The evaluation of a given transient begins with the system initial parameters l shown in FSAR Table 15.0-1 that are input to a GE core dynamic be or transient i computer program. The code 8vs described I __ to evaluate in =, _=_ n n f St _ _ F.fm_ D F_-J_gl-H ,r, af,a _:_ f, h4. i.L ) events uS reg. 4 . _ , _ _ _
" ""^-1^"N The outputs of Dprogran'along with the initial MfB gy g .
fom the input for further analyses of the thermally lietting bundle ^ i _ 9; W.h t;' to mi r; m ! ';f-
- m ::2 A.; J... 'a C ".t M The principal msuit Jf this evaluation is the reduction in MCPR caused by the transient.
The need to adjust the MCPR operating limit as a function of scras time arises from the statistical approach used in the toplementation of tha 00YN computer code for analyzing rapid pressurization events. Generic statistical shalyses were perforsed for plant groupings of similar design which considered the statistical variation in several parameters, i.e. , initial power level, CR0 scram insertion time, and model uncertainty. These analyses, whie are LA SALLE - lJNIT 2 8 3/4 2 3 -
i j . 00 54 ) Insert 0 When the Rod Withdrawal Irror is the limiting transient event, two MCPR limits , may be provided. These limits are a function of the Rod Block. Monitor (RBM) ! setpoint. The appropriate limit will be chosen based on the current RSM setpoint. The flexibility of the variable RBM setpoint/MCPR limit allows of- , ficient use of the extended operating domain (ILLLA region), while maintaining ; j transient protection with the more restrictive MCPR limit, d Analyses have been performed to determine the effects on CRITICAL PCWIR RAT!0 (CPR) during a transient assuming that certain equipment is out of service. A detailed description of the analyses is provided in Reference 5. The analyses j performed assumed a single failure only and establised the licensing bases to
, allow continuous p? ant operation with the analysed equipment out of service. ! The following single equipment failures are included are part of the transient t i analyses input assumptions:
1
- 1. main turbine bypass system out of service,
- 2. recirculation pump trip system out of service.
! 3. safety / relief valve (5/RV) out of cervice, and ,
- 4. feedwater heater c t of service (corresponding to a 100 degree F re-duction in feedwater tegerature). ,
For the main turbine bype .a and recirculation pump trip systems, specific f j cycle independent MINIMUM CRITICAL PCWER RATIO (MCTR) Limiting Condition for Cperation (LCO) values are established to allow continuous plant operation with I ( these systems out of service. A bounding end-of-cycle exposure condition was l used to develop nuclear input to the transient analysis model. The bounding exposure condition assumes a more top peaked axial power distribution than the
- nominal power shape. thus yielding a bounding scram response with reasonable !
j conservatisms for the MCPR Leo values in future cycles. The cycle independent t i MCPR LCO values shown in Figure 3.2.3176for the main turbine bypass and re= circulation pump trip systema out of service are valid provided r
- 1. The cycle specific analysis for the Load Reject Without typass and Turbine 4 Trip without typass events yield MCPR LCo values less than or equal to 1.33 and 1.29 for options A and 3, respectively. l
- 2. The cycle specific analysis for the Feedwater Controller Tailure event (
- yields MCPR LCO values less than 1.25 and 1.21 for options A and 3, re- l
] spectively, when analysed with nomal feedwater temperature. j The analysis for main turbine bypass and reciculation pug trip systems inop-erable allows operation with either system inoperable, but not both at the same j time. 4 i Por operation with the feedwater heater out of service, a cycle specific anal-
. ysis will be perfomed. With reduced feedwater temperature, the Load Reject l Without typass event will be less severe because of the reduced core steaming
, rate and lower initial void fraction. Consequently, no further analyis is . 1 i
. l 00 55 needed for that event. However, the feedwater controller failure event becomes more severe with a feedwater heater out of service and could become the limiting transient for a specific cycle, consequently, the cycle specific analycis for the feedvator controller failure event will be performed with a 100 degree F feedwater togerature reduction. The calculated change in CPR for that event will then be used in determining the cycle specific McPR Leo value.
In the case of a single S/RV out of service, transient analysis results showed ' that there is no impact on the calculated McPR tco value. The change in CPR for this operating condition will be bounded by reload licensing calculations and no further analyses are required. The analysis for a single s/RV out of service is valid in conjunction with dual and single recirculation leep opera-I tien i I l i T i
)
i J i l l 4 4 f 4 4 J l 2 l J
00 56 POWER O!5TRIBUT!0N SYSTEMS
! BA$ts MIN! MUM CRITICAL PQWER RATIO (Continued) l l described further in Reference produced generic Statistical Adjustment '
; Factors which have been app 11 to plant and cycle specific 00VM results to yield operating limits which provide a 955 probability with 955 confidence i
' that the limiting pressurization event will not cause MCPR to fall below the fuel cladding integrity Safety Limit.
. As a result of this 95/95 approach, the average 205 insertion scram time
; sust be monitored to assure compliance with the assumed statistical distribu*
1 tion. If the mean value on a cycle cumulative, running average, basis were to
; exceed a 55 significance level compared to the distribution assumed in the 00YN statistical analyses, the MCPR limit must be increased linearly, as a
- function of the mean 205 scram time, to a more consarvative value which
; reflects an NRC determined uncertainty penalty of 4.45. This penalty is I
applied to the plant specific 00YN results, i.e. without statistical adjust-ment, .for the lietting single failure pressurization event occurring at the limiting point in the cycle. It is not applied in full untti the mean of all current cycle 20E scram times reaches the 0.M seconds value of Specifica-i tion 3.1.3.3. In practice, however, the requirements of 3.1.3.3 would most ' likely be reached, i.e. , individual data set average > 0.84 secs, and tne l required actions taken well before the running average exceeds Ot M secs. ] The 55 significance level is defined in Reference 4 as 3 I tg = p + 1.65 (Mg/ N) g 2, ! i1 where p = i nean value for statistical scras time distribution j to 205 inserted = g .tAA. ,o u j o = standard deviation of above distribution = g, i Ng = nuseer of rods tested at 50C, i.e., all operatie i rods l t n i IN ga total number of operatis nods tested in the l l tal current cycle OGM . The value for tg used in Specification 3.2.3 is g seconds which is conservative for the following reason l l For simplicity in formulating and impleuenting the LCO, a conservative i n I ] value for IN ofg .*94 was used. This raresentr. one full core data set J i=t at 50C plus one full core data set followiw a 120 day outage plus twelve l j 105 of core,19 rods, data sets. The 12 Cata sets are equivalent to l 1 24 opersting months of surveillance at ue increased surveillance i
! frequency of one set per 60 days ree,6'tred by the action statements of )
! Specifications 3.1.3.2 and 3.1.3.4. ) l ' I I LA 1*.' LE = UNIT 2 8 3/4 2 4 / ' t
1 1 00 57 ' i A0WER DISTRI8l1 TION SYSTEMS l l BASES MINIMUM CRITICAL POWER RATIO (Centinued)
References:
l l
- 1. General Electric Company Analytical Model for Loss-of-Coolant i Analysis in Accordanca with 1,0 CFR 50, Appendix K MEDE-20566, 1 November 1975. ,
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.ta. - r ,re.. .j =: c = = ).
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"*""I
. . .w . i.f .m.<.,Y 0'% _5 - '9"N. ,"*,._' .jf"..I'm""" ""
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5' *"' ""'d. m,,,, , , . u.. *.'. ,_Dd' - - .* .I"..*.!7* J c7 f. "Qualification of the One-Oisensional Core Transient Model for i Boiling Water Reactors" General Electric Co. Licensing Topical Report NEDO 24154 vols. I and II and NEDE-24154 Vol. III as sup-plemeated by letter dated Sestenber 5,1980, from R. H. Suchholz (GE) to P. 5. Check (NRC). 3/4.2.4 LINEAR HEAT GENERATION RATE The specification assures that the LINEAR HEAT GENERATION RATE (LNGR) in any rod is less than the design linear heat generation even if fuel pellet dansification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735 Suppleewnt 6, and assumes a linearly increasing variation in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power spikina- 1 j a, "La.klle. Coung Sfo% un;4 W J2 sr.pcR/Ges1R . w L44-oA Coolant Ac.4clen+ 4.n o h a n s * , Ge.n uut Repoc+ N E o c -31s10 P, D u.c m b u- 19 e v. trie.chW c c,, p 1 Uvul EDe<Aric. Grhandd hplicah o n b- Reg M > Mlii - 2.4oi1 -p- A , (. Ledced cLppro wJ re,vis ion) ,
- 5. " E dend d Opa.ecW q Dornain CLnd Equj pene.n t- '
Out -Of- 62.rvics $3 r- LaSalle. Coun GtwHon U"D I W C.) WDC. - 51%S blevonb.c , N Muhe 1961.
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f LA SALLE - UNIT 2 8 3/4 2-6 l l 1
i 00 58 INSTRUMENTATION l BASES i 3/4.3.4 RECIRCtlLATION PUMP TRIP ACTUATION INSTRUMENTATION The anticipated transient without scram (ATWS) recirculation pump trip systes provides a means of limiting the consequences of the unitkely occurrence l of a failure to scras during an anticipated transient. The response of the l plant to this postulated event falls within the envelope of study events in General Electric Company Topical Report NED0-10349,' dated March 1971 and NED0-24222, dated December, 1979, and Appendix G of the FSAR. ) The end-of-cycle recirculation pump trip (E0C-RPT) system is a part of the Reactor Protection System and is an essential safety supplement to the reactor trip. The purpose of the EOC-RPT is to recover the loss of thermal margin which occurs at the end-of-cycle. The physical phenomenon involved is that the void react.ivity feedback due to a pressurization transient can add ; positive reactivity to the reactor systes at a faster rate than the control i rods add negative scras reactivity. Each EQC-RPT system trips both recircula-tion pumps, reducing coolant flow in order to reduce the void collapse in the , core during two of the most limiting pressurization events. The two events I for which the EOC-RPT protective feature vill function are closure of the turbine stop valves and fast' closure of the turbine control valves. 9/
" A fact closure son:or from each of two turbine control valves provides ,
input to the EOC-RPT system; a fast closure sensor free each of the other two ' turbine control valves provides input to the second EOC-RPT system. Stailarly, : a position switch for each of two turbine stop valves provides input to one i EOC-RPT system; a position switch from each of the other two stop valves l provides input to the other EOC-RPT systas. For each E0C-rPT system, the j sensor relay contacts are arranged to fore a 2-out-of-2 logic for the fast , closure of turbine control valves and a 2-out-of-2 logic for the turbine stop i valves. The operation of either logie. will actuate the EOC-RPT system and trip both recirculation pumps. Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled. The manual bypasses and the automatic Operating , Bypass at less than 30% of RATED THERMAL POWER are annunciated in the control room. < The E00-RPT system response time is the time assumed in the analysis between initiation of valve rotion and complete suppression of the electric arc, i.e., 190 as, less the time allotted for sensor response, i.e., 10 as,
! and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test results, i.e., 83 as, and plant pre-operational test results.
3/4.3.5 REACTOR CCRE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION I The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater ficw to the reactor vessel without providing actuation of any of the emergency core cooling equipment, j
^
LA SALLE - UNIT 2 8 3/4 3-3 i
00 59 . 1 Insert H A generic analysis, which provides for continued operation with one or both trip systems of the ICC-RPT system inoperable, has been performed. . The analysis l determined bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Lim- I iting condition for Operation (LCO) values which must be used if the IOC-RPT l system is ir. operable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the 7 function inoperable. The analysis results are further discussed in the bases for specification 3.2.3. l l l
+
1 n - s
. ' (w e 3/4.4 REACTOR COOLANT SYSTEM \ BASES 3/4.4.1 RECIRCULATICN SYSTEM Operation with one reactor recirculation loop inoperable has been evaluated
~ and been found to be acceptable provided the unit is operated in accordance with i the single recirculation loop operation Technical Specifications herein. An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does present a hazard in case of a design-basis-accident by increasing the blowdown area and reducing the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring degradation. jet pump performance on a prescribed scheduled for significant Recirculation loop flow mismatch limits are in compliance with the ECCS LOCA analysis design criterion. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. Where the recir-culation loop flow mismatch limits can not be maintained curing the recir-culation loop operation, continued operation is permitted in the single recirculation loop operation mode, in order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop' temperatures shall be within 50*F of each other prior to startup of an idle loop. The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the water in the < upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145'F. ! l The possibility of thermal hydraulic instability in a BWR has been investi-gated since the startup of early BWRs. Based on tests and analytical models, j it has been identified that the high power-low flow corner of the power-to-flow map is the region of least stability margin. Tnis region may be encountered during startups, shutdowns, sequence exchanges, and as a result of a recircula-tion pump (s) trip event. To ensure stability, single loop operation is limited in a designated restricted region (Figure 3.4.1.1-1) of the power-to flow map. Single loop operation with a designated surveillance region (Figure 3.4.1.1-1) of the power-to-flow map requires moni ing of APRM and LPRM noise levels. ! I Ni 616 M h a.e, show n 3 /4. 4. 2 SAFETY / RELIEF VALVES ev A sr en % M+ soft as oim M-'l ef. g The safety valve function of the safety /r valves operate to prevent the reactor coolant system from being press zed above the Safety Limit of 1325 plig in accordance with the ASME Code. ;;;I :' 10 ":"r.: safety / relief valves ' :W: te i Mit reactor pressure to within ASME III mopvable allowable es for the worst case upset transient W e. 4 timQ J(L# ( 1 Demonstration of the safety / relief valve lift settings will occur only during shutdown Specification and will be performed in accordance with the provisions of 4.0.5. LA SALLE - UNIT 2 B 3/4 4-1 Amendment No. 32
00 61 Insert I Therefore, operation with any 17 SRV's capable of opening is allowable, although all installed SRV's must be closed and have position indication .to ensure that integrity of the primary coolant boundary is known to exist at all times. i r i f 3 a 't l g 4
- , , , .-w,-.. - - ..~,-n-,--- ---,_,_-.-,,n,--, , - - - , - -,,- . - - , - - - , - . , - . . - . - . , , - - - , - - ~ . - - , - , - - .
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, _ .._._...s
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, ENERGENCY Cott' COOLING SYSTEMS 9ASES gCS-OPERATINGandSHUT00WN(Continued) thw suppression pool into the reactor, bt no credit is taken in the hazards analyses for the gondensate storage tank star.
With the HPCS system inoperable, adequate core cooling is assured by the OPERASILITY of the redundant and diversified automatic depressurization system and both the LPCS and LPCI systems. In addition, the reactor core isolation cooling (RCIC) systaa, a systes for which no credit is taken in the hazards analysis, will automatically provide makeup at reactor operating pressures on a reactor low water level condition. The HPCS out of service period of 14 cays is based on the demonstrated OPERABILITY of redundant and diversified low pressure core cooling systems. Th6 surveillance requirements provide adequate assurance that the HPCS system will be OPERABL3 when required. Although all active cor, ",ents are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor vessel ) injectionrequiresreactorshutdown. The pump discharge piping is maintained full to prevent water hammer damage and to pregide cooling at the earliest soment. Upon failure of the HPCS systes to functic'n properly, if required, the automatic depressurizatibn system (ADS) automatically causes selected safety- l relief valves to open, depressurizing the reactor so that flow from the low pressure core cooling systems can enter the core in time to limit fuel cladding temperature to less than 2200'F. A05 is conservatively required to be OPERABLE whenever reactor vessel pressure exceeds 122 psig even though low pressure core cooling systems provide adegaate core cooling up to 350 psig. ,ac th' ADS automatically controls seven selected safety-relief valves. Six W valves ars required to be OPERA 8LE mt:r ": *">-^ *"d m:95
- 6*
C ..? A c-ft:: ::' . .;. o the vow 5 'd #N'# It is therefore appropriate to perm'it one#'f required valves to be out-of-service for up to 14 days without asterially /c' reducing systes reliability. V o a fd' n .M 4 3/4.5.3 SUPPRESSION P W BER "
~
The suppression chamber is also required to be OPERABLE as part of the ECCS to ensure that a sufficient supply of watar is available to the HPCS, LPCS and LPCI sys n.the event of a LOCA. This limit on suppression chamber minimum water.vol sures that sufficient water is available to parait recirculation cooling f1 to the core. The OPERABILITY of the suppression chamber in
- OPERATIONAL' CONDITIONS 1, 2 or 3 is required by Specitication 3.6.2.1.
Repair work sight require making the suppression chamber inoperable. This specification will permit those repairs to be made and at the same time give assurance that the irradiated fuel has an adequate cooling water supply when the suppression chambar must be made inoperable, including draining, in OPERATIONAL CONDITION 4 or 5. In OPERATIONAL CON 0! TION 4 and 5 the suppression chamber minimum required water volume is reduced because the reactor e.oolant is maintained at or below 200*F. Since pressure tureression is not required below 212*F, the minimum water volume is based on NPSH, recirculation volume, vortex preention plus a 3 2' 4" safety margin for conservatism. ~ LA SALLE - UNIT 2 B 3/4 5-2 Amendesnt No. 27
00 63 ( PLANT $YSTDts BASES SNUBLEk$ (Continued) Figure 4.7-1 was developed using "Wald's Sequential Probability Ratie Plan" as described in "Quality Control and Industrial Statistics" by . Acheson J. Duncan. Pemanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented end, if applicable, snut,ber life destructive testing was performed to qualify the snubber for the applicable design conditions.at either the com- ' pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the exemptions. The service life of a snubber is established via manufacturer input and information through considerettori of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, sea) replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. 3/4.7.10 MAIN TUR81NE SYPASS SYSTEM d "" I"
,__m}N_ ,..'N[I!'N. . .'. ..m.533j; W
rnsut T . i i l l a l ' LA SALLE - UNIT 2 4 3/4 7-5
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LERKET 2 1 i A generie analysis, which provides for continued operation with the main turbine bypass system inopetable, has boon performed. The analysis determined ' bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) values which must be used if the main turbine bypaJs system is inoperable. The MCPR LCO values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analysed I transient with the main turbine bypass system inoperable. Although analysis supports operation with all five turbine bypass valves inoperable, the specification provides for continued operation only if at least 4 bypass valves are capable of accepting steam flow. The analysis results are further discussed in the bases for Specification 3.2.3. 1
. I 1 1 f
~ 5102K l I d , r i J
l l 00 65 Of5fGN FEA.TURES 5.3 REACTOR CORE FUEL A55EM61.IES - 5.3.1 The resctor core shall contain 764 fuel assembliss with each fuel assembly containing 62 fuel rods and two water rods clad with Zircaloy -2. Each fuel rod shall have a nominal active feal length of 150 inches. The initial core loading shall have a maximum average enrichment of 1.89 weight percent U-235. Reload fuel shall be similar in r5ysical design to the initial core loading. ' INS 92T.K CONTROL R00 ASSEN8 LIES
, c+;"
5.3.2 M P. MThe ..; ereactor
- O r r ;.con err;_sbau
, ;. JQntain
...;..:...185 /.; contro'1 d ,____ __..- rod Eassemblies 7 m " - /_b ef be, er ;..-;.U , "J, ;n,tr :::::::td 5; : er;e:n :t:d Med-1=== a+--'
shettr. 5.4 RFACTOR COOLANT SYSTEM OESIGN PRESSURE AM0 TEMPERATURE , 5.4.1 The reactor coolant systes is designed and shall be maintained:
- a. In accordance with the code requirements specified in Section 5.2 of the FSAR. with allowance for normal degradation pursuant to the applicable Surveillance Requirements,
- b. For a pressure of:
- 1. 1250 psig on th's suction side of the ncirculation pumps.
- 2. 1650 psig from the ncirculation pump discharge to the outlet side of the discharge shutoff valve.
- 3. 1500 psig from the discharge shutoff valve to the jet pumps. -
- c. For a temperatun of 575'F.
VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation systee is
- 21,000 cubic feet at a nominal T,y, of 533*F.
- 5. 5 METEOROLOGICAt. TOWER LOCATION 5.5.1 The esteorological tower shall be located as shown on Figure 5.1.1-1.
n LA SALLE - UNIT 2 5-4
.n .- , , - , , _ . _ . ,, . ..,--_._n__ . . ._ _ _ . ______. _ , , ____.-___, , - . _. . -, _ - _ . , . _ - . _ . , _ , _ . _ . , . . _ . - .
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co6G l ! l i i I l l I l l l l INSERT K i l l l i 1 l l ! i t ! l l There are two possible types of control rods, one consisting of a creciform l array of stainless steel tubes containing 143 ir.ches of boron carbide, B 4C 1 i powder, surrounded by cruciform shaped stainless steel sheath, and the second
- typo contains 143 inches of absorber material of which the first 6 inches are i hafnium and the remainder is B 4C. l l
i 1 1 i 1 l l i l l [ I l i I I l l i l I 1 I 5102K II I ! l 1 I l i
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ATIAC3ENT E O9b7 1 . l SIGNIFICANT HAZARDS EVALUATION Commonwealth Edison proposes to amend Facility Operating License NPF-18 for LaSalle Unit 2 to support the Cycle 3 core reload. The proposed reload fuel and analyses including the previously approved SAFER /GESTR-LOCA Loss-of-Coolant Accident (LOCA) Analysis, changes resulting from analyses performed to expand the operating region and allow equipment out-of-service and changes that are administrative or provide clarification. The proposed ; changes for LaSalle Unit 2 are identical to those previously submitted and approved for use at LaSalle Unit 1, except for minor calculation differences in the results for transient analyses. DISCRIfl10N OF MiENDMENT REOUEST The Technical Specification changes for the LaSalle Unit 2 Cycle 3 (L2C3) reload include:
- 1. Provision for operation in the expanded operating domain including revised APRM and RBM satpoint changes incorporated using standard and previously approved methodology.
- 2. Use of extended burnup fuel (GE 8x8EB) with increased LHGR limit of 14.4 Kw/ft.
- 3. Use of improved transient and LOCA analysis methods which allow use of a lower tau-B value in determining the MCPR operating limit as a function of scram time, and deletion of the single loop MAPLHGR limit multiplier of 0.85.
- 4. Provision for operation with certain equipment inoperable or out of service. Specifically, one of the following systems or components may be out of service when the appropriate Technical Specification ACTIONS are satisfiedt
- a. Turbine Bypass System
- b. End-of-Cycle Recirculation Pump Trip (EOC-RPT)
- c. One Safety Relief Valve (SRV)
- d. Feedwater Heaters
- 5. Several changes for clarification or administrative purposes are proposed including: I
- a. Deletion of GEXL correlation and GETAB statistical model in the bases of the safety limit section.
- b. Revision to the Control Rod Program Controls Technical Specification to require the RWM to be demonstrated operable in Operational Condition 1, prior to reaching 20% power, when reducing thermal power.
d') 33 BASIS FOR PROEROSEQJQ_SIGHIf_ICAllLIIAZARDS_. CONSIDERATION DETEBli1Nld103 Commonwealth Edison has evaluated the proposed Technical Specifi-cations and determined that they do not represen; a significant hazards consideration. Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92(c), operetion of LaSalle Unit 2 Cycle 3 In accordance with the proposed changes will nots
- a. Involve a significant increase in the probability or consequences of an accident previously evaluated because:
(1) The use of the proposed operating limits are specifically analyzed to ensure the input assumptions of all existing transient and accident analyses remain valid. These analyses are performed 'asing a methodology which has received review and approval for other similar plants including LaSalle Unit 1. (2) The Technical Specification ACTIONS included in the proposed revisions do not significantly affect the probability of an accident previously analyzed because the required time intervals for corrective action are consistent with the existing specifications.
- b. Create the possibility of a new or different kind of accident from sny l accident previously evaluated because:
(1) The proposed MCPR, MAPLHGR, and LHGR limits represent limitations on reactor oper cing state which do not directly affect the operation, or function of any system or component. As a result, there is no I impact on or addition of any systems or equipment whose f ailure could initiate an sccident. (2) The proposed operating domain is evaluated to retain the originally I required design margins to system integrity during normal operation, transients and accidents and therefore do not cause significant new loads or stresses on mechanical systems or boundaries. (3) The proposed allowances for operation with prescribed equipment
]
inoperable or out-of-service do not cause physical changes to any i systems and therefore do not induce new failure modes.
- c. Involve a significant reduction in the margin of safety because:
(1) No change to Safety Limits are involved.
, , (J a b i (2) The analyses used to evaluate reactor and system performance are performed using standard methods and the calculated operating limits maintain conservative margin to safety limits to accommodate the l anticipated performance during transients and accidents. ;
1 (3) No changes to protective system logic or design are involved. (4) Changes which are administrative in nature do not affect the operating limits of the plant or the consequences of analysed transients. Guidance has been provided in 51 Fr 7744 for the application of q standards to license change requests for determination of the existence of l 1 significant hasards considerations. This document provides examples of 7 mnendments which are and are not likely considered to involve significant
- hazards considerations. This unendment request is similar to example (iii) of the examples that are not likely to involve significant hasards consideration.
Example (iii) "For a nuclear power reactor, a change resulting from e I nuclear reactor core reloading, if no fuel assemblies significantly dif ferent f roan those found previously acceptable to the NRC for a previous core at the f acility in question are involved. This assumes that no significant changes are made to the acceptance criteria for the technical specifications, that the analytical methods used to demonstrate conformance with the technical specifications and regulations are not significantly changed, and that NRC has previously found such methods acceptable." C This change clearly f alls within this example as the reload fuel for Unit 2 Cycle 3 is of the same design as reviewed and approved for LaSalle 1 Cycle 3. I This proposed snendment doea not involve a significant relaxation of i the criteria used to establish safety limits, a significant relaxation of the l bases for the limiting safety system settings or a significant relaxation of I the bases for the limiting conditions for operations. Therefore, based on the 4 guidance provided in the the rederal Register and the criteria established in 10 CFR 50.92(e), the proposed change does not constitute a significant hazards consideration. i i l l
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- 5102K l 1 ,
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4 _) ) 7) ATIACIG4DCt_t ) i 1 SUfELIEDLIAL InrORMAUCtl l l 1 i GE Document 23A5841, "Supplemental Reload Licensing Submittal for LaSalle County Station Unit 2 Reload 2 (Cycle 3)", dated July, 1988. l l l l l l 1 l l 5102K _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _}}