Text

Proposed Tech Specs,Supporting Second Reload for Plant
	ML20148U192
Person / Time
Site:	LaSalle
Issue date:	01/19/1988
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML19341D859	List: ML20148U186; ML20148U192;
References
	NUDOCS 8802030494
	Download: ML20148U192 (59)
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1 PROPOSED TECHNICAL SPECIPICATION CHANGES 3

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C INDEX LIsi 0F FIGURES FIGURE

[ AGE 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

CONCENTRATION REQUIREMENTS.............................

3/4 1-21 3.1.5-2 SODIUM PENTABORATE (Na:B o0 s. 10 H O) i t

2 VOLUME / CONCENTRATION REQUIREMENTS......................

3/4 1-22 l

3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION l

RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB176, 8CRB219, AND SCRB071................................................

3/4 2-2 3.1.1-2 MAXIMUM A.R P NAR LINEAR HEAT GENERATION RATE (M LHTA)

. SUS AVERAGE PLANAR EXPOSURE, FUEL TY BP 299L...................................

3/4 2-2a 3.2.3-14 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS t AT RATED FLOW........................................

3/4 2-5

-+

3.2.3-2 K FACTOR..............................................

3/4 2-6 f

3.4.1.1-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED).................................

3/4 4-lb 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE............................

3/4 4-18 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST.............

3/4 7-32 B 3/4 3-1 R EACTO R VE S SE L WAT ER LEVEL.............................

B 3/4 3-7 s

B 3/4.4.5-1 CALCULATED FAST NEUTRON FLUENCE (EJ1MeV) at 1/4 i AS A FUNCTION OF SERVICE LIFE..........................

B 3/4 4-7 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASECUS AND LIQUID EFFLUENTS...................................

5-2 5.1.2-1 LOW P O P U LAT I O N Z O N E....................................

5-3 6.1-1 CORPORATE MANAGEMENT.................................

6-11 6.1-2 UNIT ORGANIZATION......................................

6-12 6.1-3 MINIMUM SHIFT CREW COMPOSITION.........................

6-13 p.a. ) -if--

pl#Namum cntTnni. power AAT80 ( Mcm) Ef35ui t AT RMTEo Ft.oiaJ foA F4o of" '.rt'E. AEcintuuarted Am#

7tv Awo mas 4 Tursu#g 6Yf A55 S YSTims INo#tAAst.E.....

3/'l A-5~4 La.t - 3 Max amum Avannot pt.nN4( ssNdaA usar Gr#Jeamwou A*fd (m4#6Msa) Vensus AvtAAs.E ft.nwast. fues uaa 3/4A4h funt. T w ss se }orn sng sc n o s..............t........

LA 5ALLE - UNI [ 1 XIX Amencment No. 40

i 1

TA8LE 2.2.1-1 5

REACIOR PROIECIl0N SYSTEM INSTRUMENIATION SEIP0lNIS h

ALLOWABLE l

TRIP SETPOINT VALUES l;;

FUNCTIONAL UNIT d

1.

Intermediate Range Monitor, Neutron Flux-High 1 120 divisions of

$ 122 divisions full scale of full scale atU 2.

Average Power Range Monitor:

a.

Neutron Flux-High, Seldown i 15% of RATEG THERMAL POWER 1 20% of RATED THERMAL POWER b.

Flow Blased Simulator Thermal Power - Upscal 1)

Two Recirculation Loop Operation o.suWt n%

o.stw + p%

a) Flow Blased 1 -G. =. ",1% wi th a

$ G. =. "M with maximum of maximum of b) High flow Clamped i 11J.M oi san 1 115.",% ui TED THERMAL POWER 1HERMAL POWER i8 2)

Single Recirculation Loop Operation o.5tw 4 fM 3 %

o.rtW t n.3%

rp a) Flow Biased

$-G. =.

4',. M with 1 G.M ; 4G. 3 withl maximum of a maximum of b) High Flow Clamped

$ 113. n of D

1

..,% ef on THERMAL POWER THERMAL POWER c.

Fixed Neutron Flux-High i 118% of RATED THERMAL POWER 1 120% of RATED THERMAL POWER 3.

Reactor Vessel Steam Dome Pressure - High 1 1943 psig 5 1963 psig 4.

Reactor Vessel Water Level - Low, Level 3

-> 12.5 inches above instrimient

~> 11.0 inches zero" above instrument zero*

~5.

Main Steam Line Isolation Valve - Closure

$ 8% closed 1 12% closed l1 g

6.

Main Steam Line Radiation - High 1 3.0 x full power background i 3.6 x full power background E

y 7.

Primary Containment Pressure - High 1 1.69 psig i 1.89 psig t;;

8.

Scram Discharge Volume Water Level - High

$ !&7' 54"

$ 767' SV A

c-- n., ca.. n 1/4 3-1,

SAFETY LIMITS m

BASES 4

i 2.1.2 THERMAL POWER. Hich Pressure and Hich Flow r

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 parametars which result in fuel damage are not directly observable during reactor aceration, the thermal and hydraulic conditions resulting in a departure frem nucleate boiling have been used to mark the beginning of the region where fuel damage could occur.

Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to SWR fuel r:cs, the critical power at which boiling transition is calculated to oc:ur nas oeen accoted as a convenient However, the uncertainties in monitoring tne ::rs coerating state and limit.

in the procedures used to calculate the critical power result in an uncertainty in the wiue of the critical power.

Therefore, the fuel cladding integrity Safety Limit it defined as the CPR in the limiting fuel assemely for wnich l

sort taan 99.9% of the fuel rods in the core are exoected ta avoid boiling transitie,n considering the power distribution within the core and all uncertainties.

Analysis Basis, GETA8,MCPR is determined using the General E l

The Safety Limit i

3 uncertainties in operating parameters and the procedures used to calculate 4

The probability of the occurrence of boiling transition is critical power.

determined using the General Electric Critical Quality (X) Boiling Length (L),

GEXL correlation.

The 0:XL w,,

eietiea i; alid ;ar e; cs;; ;f saittis:.;; '- e; t::t: :f e: d:n 2:d e :=:!:; c: ::rr:!:ti=,

Thn: =:iti=: ;r;;

1 Ei wa am w.

GO t; 1000 sie

";;; fle.

O, i ;; 106 ;;, 3,,;6,

7;2

- alet 0,tseltas.

O te 100 L /10 i

Ln ;' ?n';ia;.

i.0; it ; s. a ? r;d 0; j

-?.07 ;; en laterier ce i

i, "General Oectric 6VR Thermal Analysis Bases (GETA8) Data, correlation a.

and Design Application," NE00-10958-A.

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LA SALLE - UNIT 1 82-2

.. /

1

SAFETY LIM!TS BASES

~

THERMAL power, Hich Pressure and Hich Flew (Continued)

rf;' *:: ::

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u The bgsas for the uncertainties in the core parameters are given in in NE00-10958-A,the basis for the uncertainty in the GEXL co NE00-20340 and The power distribution is based on a typical 764 assemoly core in wnich the red pattern was arbitrarily chosen to produce a skewed ocwer distribution having the greatest numoer of assemolies at the nignest power i

levels. The worst distribution durjng any fuel cycle.would not be as severs as the distribution used in the analysis.

i I

i

- General Electric SWR Thermal Analysis Bases (GETAB) Data, Correlation A

a.

and Design Application," NE00-10958 A.

General Electric "Process Computar Performance Evaluation Ac:uracy" b.

NE00-20340 and Admondment 1, NE 0-20340 1 dated June 1974 and f

Oecameer 1974, respectively.

j t

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LA SALLE UNIT 1 8 2-3 I

O F

Bases Table B2.1.2-1 UNCERTAINTIES USE0 IN THE DETERMINATION OF THE FUEL CLAODING SAFETY LIMIT

Standard Deviation

(% of Point)

Quantity 1.76 Feedwater Flow 0.76 Feedwater Temperature 0.5 Reactor Pressure

0. 2 Core Inlet Temperature 2.5 Cors Total Flow Two recirculation Loop Operati on 6.0 Single recirculation Loop Opera 3.0 Channel Flow Area h

10.0 Friction Factor Multiplier 4

Channel Friction Factor h

5. 0 Multiplier 8.7 l

TIP Readings Two Recirculation loop Operation 6.8 Single Recirculation Loop Operation

.6 l

R Factor 3.

Critical Power w

" The uncertainty analysis used to establish the core wide fety Limit MCPR is based on the assumption of quadrant power symmetry for t e reactor core.

The values herein appply to both two recirculation loop oper tion and single recirculation loop operation, except as noted.

LA SALLE - UNIT 1 B 2-4 Amen @ent No. 40

OELETE Bases Table 82.1.2-2 NAL VALUES OF PARAMETERS USED IN THE STATISTICAL AMA 0F FUEL CLADDING INTEGRITY SAFETY LIMIT THERKAL POWER 93 MW l

Core Flow 102.

b/hr l

1010.4 psig Dome Pressure 1.038 - 0 GWD/t R-Factor 1.031 - 7 GWD/t 1.030 - 15 GWO/t 1.033 - 20 GWD/t I

f I

l I

i i

f Amendment No. 40 B 2-5 LA SALLE - UNIT 1

DE LET'E Bases Table B2.1.2-3 RE VE BUNOLE POWER DISTRIBUTION USED IN ETAB STATISTICAL ANALYSIS Percent of Fuel Bundles Within Rance of Relative Bundle Power Power Interval N

5.1 1.375 to 1.425 s

1.325 to 1.375 7.3 1.275 to 1.325 8

1.225 to 1.275 9.

1.175 to 1.225 7.3 1.125 to 1.175 11.8 1.075 to 1.125 4.7 l

1.025 to 1.075 4.7 41.5

<1.025 IUU~U 9

,s LA SALLE - UNIT 1 B 2-6 meendment No. 40 I

l b

Bases Table 82.1.2-4 R-FACTOR OISTRIBUTI USED IN GETAB STATISTICAL. ANALYSIS 8

Array Rod Sequence No.

R-Factor 1

1.038 1.038 3

1.037 4

1.037 5

y 1.035 1

6 1.035 7

1.030

<1.030 8 through 64 s

r E

E -

ke a

REACTIVITY CONTROL SYSTEM CONTROL R0D MAXIMUM SCRAN INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maxis scram insertion time of each control rod from the fully withdrawn positio to notch position 05, based on de-energization of the scram l

pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a.

With the maximum scram insertion time of one or more control rods exceeding 7.0 seconds:

1.

Declare the control rod (s) with the slow insertion time inoper-able, and 2.

Perform the Surveillance Requirements of Specification 4.1.3.2.c at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , b.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scraa insertion time of the control rods shall be demonstrated through measurement with reactor coolant pressure greater than or equal to 950 psig and, during single control red scram time tests, the control rod drive pumps isolated from the accumulators:

a.

For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER following CORE ALTERATIONS"" or after a reactor shutdown that is greater than 120 days, b.

For specifically affected individual control rods following maintenance on or modification to the control rod or control red drive system which could affect the scram insertion time of those specific control rods, and For at least 10% of the control rods, on a rotating basis, at least c.

once per 120 days of operation.

genp w'.ve 10 47 -., ;; pseitien t0 ;; th; fully withdre-a p;;iti:r, ";r wiwi...

- Eecept movement of SRM, IRM or special movable detectors or normal control rod movement.

LA SALLE - UNIT 1 3/4 1-6 Amendment No. 48

REACTIVITY CONTROL SYSTEM CONTROL R00 AVERAGE SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.3 The average scram in rtion time of all 0PERABLE control rods from the fully withdrawn position,W based on de-energization of the scram pilot l

valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.43 39 0.86 25 1.93 05 3.49 APPLICABILITY:

OPERATIONAL CON 0!TIONS 1 and 2.

ACTION:

With the average scram insertion time exceeding any of the above limits, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.

4.s t

"Centrel ie; 10 47 ee, -ee peeitien ?C ee the falli ithe i-o Peiitish fei l

-Cy-le 2.

l l l

LA SALLE - UNIT 1 3/4 1-7 Amendment No. 48 lu

REACTIVITY CONTROL SYSTEM FOUR CONTROL R00 GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, l

for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 45 0.45 39 0.92 25 2.05 05 3.70 APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

With the average scram insertion times of control rods exceeding the a.

above limits:

1.

Declare the control rods with the slower than average scram insertion times inoperable until an analysis is performed to determine that required scram reactivity remains for the slow four control rod group, and 2.

Perform the Surveillance Requirements of Specificaticn 4.1.3.2.c at least once per 60 days whe.1 okeration is continued with an average scras insertion time (s) in excess of the average scram insertion time limit.

Otherwise, be in at least HOT SHUTDOW within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

The prosisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement 4.1.3.2.

"Cer;re' rea 10-t7 ae, oee peeitier. tC ;; the fully ithe e r. positier, for l

-C elu 2.

i LA SALLE - UNIT 1 3/4 1-8 Amendment No. 48

REACTIVITY CONTROL SYSTEM i

CONTROL R00 CRIVE COUPLING LIMITING CONDITION FOR OPERATION 3.1.3.6 All control rod [ hall be coupled to their drive mechanisms.

l APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, and 5*.

ACTION:

In OPERATIONAL CONOITION 1 and 2 with one control rod not coupled to a.

its associated drive mechanism:

1.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either:

a)

If permitted by the RWH and RSCS, insert the control rod drive mechanism to accomplish recoupling and verify recoupling by withdrawing the control rod, and:

1)

Observing any indicated response of tne nuclear instrumentation, and 2)

Demonstrating that the control rod will not go to the overtravel position.

b)

If recoupling is not accomplished on the first attempt or, if not permitted by the RWM or RSCS then until permitted by the RWM and RSCS, declare the control rod inoperable and insert the control rod and disarm the associated directional control valves ** either:

1)

Electrically, or 2)

Hydraulically by closing the drive water and exhaust water iso'ation valves.

2.

Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

In OPERATIONAL CONDITION 5" with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either:

1.

Insert the control rod to accomplish recoupling and verify recoupling by withdrawing the control rod and demonstrating that the control rod will not go to the overtravel position, or 2.

If recoupffng is not accomplished, insert the contr?) rod and disarm the associated directional control vains'* either:

(

a)

Electrically, or b)

Hydraulically by closing the drive water and exhaust water isolation valves.

l The provisions of Specification 3.0.4 are not applicable.

c.

t "At least each withdrawn control rod.

Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.

Atiel,ed 10-4T i. --+; fer Cy;ie 2 pr;vid;d the red i; felij insert;d h;n 1;;; then er e @ el te 2 3 ef "ATE 0 THE Z L FCWER.nu n m ran i n u ru-

nt;tien r;;pene; is v;rified dur'n; r;d withdr e1.

LA SALLE - UNIT 1 3/4 1-11 Amendment No. 48

~T.

REACTIVITY

CONTROL SYSTEM 3/4.1.4 CONTROL R00 PROGRAM CONTROLS

\\

R00 WORTH MIN!MIZER I.!MITING CON 0! TION FOR OPERATION 3.1.4.1 The rod worth minimizer (R'aN) shall be OPERABLE.

APPLICABILITY:

OPERATIONAL CON 0!TIONS 1 and 2",

when THERMAL PCWER is less enan or equal to {0% of RATED THERMAL POWER, the minimum allowable icw power setpoint.

ACTION:

With the RWM inoperable, verify control red movement and compliance with a.

the prescribed control red pattern by a second licensed operator or other technically qualified member of the unit technical staff who is present at the reactor control console. Othemise, control rod movement may be only by actuating the manual scram or placing the reactor mode switch in the Shutdown position.

t b.

The provisional of Specification 3.0.4 are not applicable.

[

SURVEILLANCE REQUIRENENTS 4.1.4.1 The RWM shall be demonstrated CPERABLE r*eacN'n3 A07o of N

(

In OPERATIONAL CONDITION 2 prior to withdrawal of control rods for a.

the purpose of making the reactor critical, a in OPERATIONAL CCN0! TION 1 prior to L" ee.; ti; i-iti; tin when reducing THERMAL POWER, by verifying proper annunciation of the selection error of at least one out-of-sequence control rod.

j b.

In CPERATIONAL CONDITION 2 prior to withdrawal of control rods for the punose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-I j

sequence control rod.

i In OPERATICNAL CONDITION 1 within one hour af ter hN automatic l

i c.

initiation when recucing THERMAL PCWER, oy verifying the red block l

function by demonstrating inact11ty to withdraw an out-of sequence control rod.

By verifying the control toe patterns and sequence input to the LM d.

computer is correctly leaded following any loading of the program into the computer.

c.: 4 14 "Entry inta GER4;CNAL ;;N0*TICN

-4 '

-1

' u I t's: tau

.1.N per: sittee for the purpose of determining tne WEAAsi'.ih of tr.e Ra.4 prior to s

withdrawal of control rods for the purpose of bringing the reactor to cri t,1 cali ty.

l LA SALLE - UNIT 1 3/4 1 16 9

l e

i 1

3/4.2 POWER OISTRIBUTION LIMITS _

3/4.2.1 AVERAGE PLANAR t,INEAR NEAT GENERATION RATE LIMITING CONDITION FOR OPERATION and L a.t-3 3.2.1 All AVERAGE PLANAR LINEAR HEA GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLA4AR EXPOSURE shall not exceed the limits shcwn i,n Fipres 3.2.1-1 em> 3.2.li p.. ij@ of fjp....y".y.,~

l 3

.C '. _! J_.. ' "' C. J. ' - l "_. '7 ' ' J. '..'..J '..' ".7. _. _..'. C.. 7. ' _. ' C.. _

I J

APPLICABILL :

0 T ONA CONDITI 1,

POWER is greater than or equal to Zuz of RATED THERMAL POWER.

and 3.A.1-5 ACTION:

With an APLHGR exceeding the limits of Figures 3.2.

Igne.3.2.1-initiate corrective action within 15 minutes and restore AP to within Ge required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to les than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to less than the limits determined from Figures 3.2 -1gne3.2.1-g l

ena LA.)-3' a.

At least once per

ours, I

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is l

c.

operating with a LIMITING CONTROL R00 PATTERN for APLHGR.

l 1

LA SALLE UNIT 1 3/4 2-1 Amendment No. 40 i

b MAPLHGR YS. Average Planar Exposure Fue Types B0301 A and B0320B 14.0 r

s s 13.0 g:

U?

BC301A

' s

~

/

s 12.0 g,

' - ~

s s

s s E ll y0

--- BC3208

' s

~

g10.0 k

- s y

s ys 3

9.0 4

' =

A

' ' 9" '

80 the lattice specific curves in GE g

document 23A5836 shall be used sg 7.0 for Technical S)ecification 3.2.1

\\\\

\\

6.0 0

10,000 20,000 30,000 40,000 50,000 Average Planar Exposure (mwd /St) figure 3.2.1-3

-__--=.

l l

4 l

l 3/4.2.2 APRM SETPCINTS i

I I

LIN! TING CON 0! TION FOR OPERATION The APM flow biased simulated thermal power-upscale seres trip setpoint 3.2.2 (5) and flow biased simulated thermal power-upscale centrol red block trip setpoint (5,g) shall be established acesrding to the following relationships:

Two Recirculation Lee [0peration (o.5 W + U N T s.

5 less than ee equal (to

'0.""" : "ZT less than er eqqa1 to (0 """

")T-(o.#8W + O N T S

l gg b.

Single Recirculati n Loop Operation (0 8 W + #437* T 5 less than er eque te (0 M

'". '")T-S less than er equ 1 to '0.""" ^ 00.'" )" (o.rsw + %3 7. T gg 5 and 5.g are in pe nent of RATH M MAL M,

where:

W e Loop recirculatten flow as a percentage of the 1eep recirculatten i

flow which produces a rated core flow of 108.5 millien

/%

T e Lowest value of the ratis of FRACTION OF RATED THE R

divided by the MAXIMUM FRACTION OF LIMITING POWER ITY. T is always less than er equal to 1.0.

A, m 4 g APPLPCA8!'."TY: OPERATIONAL CON 0! TION 1, when THERMAL POWER is or equal to a of RATED THEMAL POWER.

M:

With the APM flow biased simulated themal power-upscale scres trip setpoint l

and/or the flow biased simulated themal power-upscale centrol red block trip corrective action within 15 minutes and restNe, as aheve detemined l

setpoint set less conservatively than 5 or 5 5 and/or 5 te within the required limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months er reduce THERMAL POWER taSess than 255 of RATED THERn4L POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE!LLANCE pf00!REMENTS The FATP and the WLPD for each class of fuel shall be detamined, the 4.2.2 value of T calculated, and the most recent actual APRM flow biased simulated thermal power-gscale scras and control red block trip setpoint ve.rtfied to be within the above limits er adjusted, as required:

a.

At least once per 24 hoves, Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completten of a THERMAL POWER increase of at b.

least 155 of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.-

with W LPD greater than er equa h rate.

"With MFLPD greater than the FRTP w to 905 of RATED R, rather than l

adjusting the APM setpoints, the APRM gain may be adjusted such\\ that APM readings are greater than er equal to 1005 times MFLPO. provided'that the adjusted APRM reading does not ascoed 1005 of RATED THERMAL POWit, the required gain adjustment increment does not exceed 10E of RATED THEWEL POWER, and a notice of the adjustment is posted on the reactor control panel.

LA SALLE UNIT 1 3/4 2-3 Amendment No. 30 I

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION

-.li k -

.1 A.._ _

_A PL _

L A f L19 t A1 RA

,M999AA1 mat 'PM MATTM

/ L AA N M T k

3. 2. 3 s 3......

. m... o.. zu r,. a... s <,.

,e--m

~,...v~.

u 5.re )..)

$$ee he $ de ei siisud II;I Ih; I

liGii det;.~1in;d fi;C 5

'r;: "i;;r-2.2.0-2 for t,; r;;ir;;l;ti;n 1;;; ;;;r;ti;n ;nd ;h;11 b; ;;;;l t; ;r gr;;;;r th;r th; "C?1 'imit d;t. rain;d fr;; rigore 0.2.01

0. 01 time, -

l th; K - catermined f r;a rig re 0.2.0 2 fer eingle recii;wl.Lvu ivvg vg..a vn. 7 I

.APr ICABILITY:

c 1 M52R7 A OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION I

Wifh MCT" le; then th esplicedle "CI" limit determined frea rigores 2.2.0-1

nd 0.2.: :, initist; ;;rr;;;i.; ;;;i;n ithin 15 min tes end rester; "CPR te n;;r; ;r r;d.;; T":R"",L POU:n te lee; th;n

.ithin the.,re.a irac limit it,hin._

ou.

e, _,,..

or-o.,

.m

,e,,

.m..m

...-.s

...o...

o.

ov o...

SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

0.86 prior to performance of the initial scram time measurements a.

t

=

for the cycle in accordance with Specification 4.1.3.2, or ave t,y, determined within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time b,

surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined frem Figures 3.2.3-1 and 3.2.3-2:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , a.

b.

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

with a LIMITING CONTROL R00 PATTERN for MCPR.

l LA SALLE UNIT 1 3/4 2-4 Amendment No. 40

INSERT A (T.S. 3/4.2.3. Page 3/4 2-4) 3.2.3 - The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit determined from:

a.

single Recirculation Loop Operation Figure 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve B for a RBM setpoint of 110%) plus 0.01, times the kg determined from Figure 3.2.3-2.

b. Two Recirculation Loop Operation Figure 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve B for a RBM setpoint of 110%) times the kg determined from Figure 3.2.3-2.

Two Recirculation Loop Operation with Main Turbine BYDass InoDerable c.

Figure 3.2.3-lb times the kg determined from Figure 3.2.3-2, for two recirculation 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 End-of-Cycle Recirculation Pump Trip System Inoperable Figure 3.2.3-lb times the kg determined from Figure 3.2.3-2, for two recirculation loop operation, with the end-of-cycle recirculation pump trip system inoperable as directed by Specification 3.3.4.2 (any RBM setpoint determined per specification Table 3.3.6-2 may be used).

ACTION:

a. With MCPR less than the applicable MCPR limit as determined for one of the above conditions:

1.

Initiate corrective action within 15 minutes, and 2.

Restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

3.

Otherwise, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

When operating in a condition not identified above, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4081K

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, 6 8

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-r

POWER DISTR BUT ON L M TS a

'f MCPR 1.45 R.

1.40 1.35 w

~

Curve B 'RBM setpoint = 110%)

1.31

~&

1.30 Curve A {RBM setpoint = 106%)

g 1.25 I

1.20

.687

.70

.72

.74

.76

.78

.80

.82

.84

.86 MC3R VERSUS T AT RK~ED 10W l

Figure 3.2.3-la

POWER D STR BUT ON.M"S f

M:PR i

' L5 c-1.40

-))n ge%

E0C-RP 1 35 C'By3 ass no) era 3 e a=

Main Turaine C

M 1.30 1.25 1.20

.687

.70

.72

.74

.76

.78

.80

.82

.84

.86 MC3R V RSUS f AT RK ED FLOW

~

igure 3.2.3-Ib

t r

L PowfR OISTRI80 TION LIMITS 3/4.2.4 LINEARHEATUENERATICNRATE 7%

LIMITING CONDITION FOR OPERATION

/

I 3.2.4 The LINEAR HEAT GENERATION RATE '(LHGR) shall not e sec "1. t ', /"'B'.

9ssar APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER s greater than or equal to zu.J. of RATED THERMAL PC%ER.

ACT!ON:

With the LHGR of any hel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the nex 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ..

i SURVEILLANCE REOUIREMENTS f

4.2.4 LHGR's shall be determined to be equal to or less than the limit:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after comoletion of a THERMAL POWER increase of i

b.

at least 15% of RATED THERMAL PCVER, and I

Intially and at least once car 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating c.

on a LIMITING CONTROL R00 PATTERN for LHGR.

h I

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I I

LA sal.LE - UNIT 1 3/4 2-7

INSERT B (T.S. 3/4.2.4. Page 3/4 2-7)

a. 13.4 kw/ft for fuel types:

1.

8CRB176 2.

8CRB219 3.

BPBCRB299L

b. 14.4 kw/ft for fuel types:

1. BC301A

2. BC320B 4081K

INSTRUMENTATION ENO-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumenta-tion channels shown in Table 3.3.4.2-1 shall be OPERA 8LE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.

ACTION:

With an end-of cycle recirculation pump trip system instrumentation a.

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.

With the number of OPERABLE 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 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , With the number of OPERABLE channels two or more less than required c.

by tne Minimum OPERABLE Channels per Trip System requirement (s) for one trip system and:

1.

If the inoperable channels consist of one turbine control valve channel and one turbine stop valve channel, place both inoperable channels in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , i

2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system inoperable.

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .or, reduce THERMAL POWER to less than 30% of RATED THERMAL POWER wJttfin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

s' A

With both trip systems inoperable, restore at least one trip system e.

to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , g, reduce THERMAL POWER to less s

than 30% of RATED THERMAL POWE sthin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

N

/k

~

Okew;se, e Ac:

PMEK MTIO (MCfA) LWfiq CO'M I. sw era se t he ensum ctZrtcAL O

4Le Ece.-t PT ;u pe r4f e vsla e p r {prfwk M m

m.pr&n (L e s') s o

\\

for m na a u.a o r>

~ -.

aff calle U

Tke. gre v.u ;m of Spe c:0es} :e 103 are nef LA SALLE - UNIT 1 3/4 3-39 Amendment No. 40

o l

IABLE 3.3.6-2 E

CONIROL R0ll WilllDRAWAL BLOCK INSIRUMINIAIION Sell %INIS U

All0WABLE VALUE 1 RIP $[IPOINI

(

1 RIP fuMC110N 1.

ROD BLOCK MONIIDR E

4.

scale U

1)

Two Recirculation Loop Operation 1 9. 55 " ^ ?W -

a.55 u. 3y -

ggg4)

Single Recirculation ce 35, y 2

< n cgu 1

x a. Ef" ^ 32. M -

loop Operation

%4sarC ?. 4 M.A.

N.A.

sq> erat we

> 5% of RAIED IEERMAL POWLR

> 3% of RAIED llERMAL POWER c.

Downscale 2.

APilM Flow Biased Simulated a.

Thermal Power-Upscale 1)

Two Recirculation n

v7 p

sd D

toop Operatinn

< 0.Mr W + ef1*

< 0.etr W

4S%*

<a

~

2)

Single Recirculatiori A

4g.3 p

y43 y

loop Operation 1 0.46W + 36-1%

$ 0.66W + 34-11" g

M.A.

b.

Inoperative M.A.

> 5% of RAlfD 1HEllMAL POWER

> 3% of RAlfD ll4RMAL POWER c.

Downscale d.

Neutron flux-liigh

[12%ofRAl[D1H[HMALPOWER

[14%ofRAll0IIERMALPOW[R 3.

SOURCE RANG [ MONI10R$

N.A.

Detector not full in M.A.

< 2 x 10$ cps

< 5 x 10$ cps a.

b.

Upscale R.A.

R.A.

c.

Inoperative

> 0.5 cps d.

Downscale

> 0.7 cps 4.

INIIRMLDIA11 RANGE MONIIORS y

N.A.

Detector not full is N.A.

a.

< 108/125 of full scale

< 110/125 of fusi scaie S

b.

upscale R.A.

Inoperative R.A.

S c.

> 5/125 of full scale

> 3/125 of full scale d.

Downscale 5

h

t INSERT C1 (T.S. TABLE 3.3.6-2. Pace 3/4 3-53)

a. When using the MCPR s 0.66 W + 37%**

1 0.66 W + 40%**

LCO from Curve A of Figure 3.2.3-la or the curves from Figure 3.2.3-lb.

b. When using the MCPR 1 0.66 V + 41%**

$ 0.66 W + 44%**

LCO from Curve B of r

Figure 3.2.3-la or the curves from Figure 3.2.3-lb.

L l

INSERT C2 (T.S. TABLE 3.3.6-2, Pace 3/4 3-53)

P

a. When using the MCPR s 0.66 V + 31.7%**

$ 0.66 W + 34.7%**

r LCO from Curve A of Figure 3.2.3-la.

b. When using the MCPR

$ 0.66 W + 35.7%**

s 0.66 W + 38.7%**

LCO from Curve B of Figure 3.2.3-la.

i l

I r

i i

l L

i 4081K L

f i

[

I

TABLE 3.3.6-2 (Continued)

E CONTROL N00 WlIHDRAWAL BLOCK INSTh2NIAT10N SEIPOINIS C

Att0WA8tE VALK TRIP SETPOINI k

TRIP FINICTION 5.

SCRAM oiSCNAaGE WOLLBEE Water Level-High 1 765' % "

1 765' % "

h*

a.

-4 b.

Scram Discharge Volume l

M.A.

F Switch in Bypass M.A.

6.

REACTOR COOLAIIT SYSTEM RECIRCULATION FLOW a.

Upscale i 108/125 of full scale i 111/125 of full scale l

N.A.

M.A.

b.

InsperaLive c.

Comparator i 105 flow deviation i 11% flow deviation e,

"The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow o

tion 3.2.2.

The trip setting of this function must be maintained in accordance with y

(W).

e_ allo s I Valut for a rwcimie6on pC) 3 0

loop Gl Y ( W ) of SO*lo.

9 i

i i

e I

a

.W 03

e 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION

[

Two reactor coolant systes dcirculation oops shall be in operation.

3.4.1.1 idand2k APPLICA8ILITY: OPERATIONAL CONDITIO ACTION:

With one reactor coolant system recirculation loop not in operation:

a.

1.

Within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s:

a)

Place the recirculation flow control system in the Master Manual mode, and b)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and, c)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting I

Condition for Operation by 0.01 per Specification 3.2.3,

and, U

5t$ N N"'.. N M,bl Y'bi9 @I E

"'I M...'S. '.2..".'".. ',.11", '.!_ !'.'.L..I ".. ' O...'f!. Y ' ",, ?,.! ".. L ? "L..

y,

.m.

d)

Reduce the Average Power Range Monitor (APRM) Scram and Rod i

+

Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single loop recirculation loop operation per Specifications 2.2.1, 3.2.2, and 3.3.6.

2.

When operating within the surveillance region specified in Figure 3.4.1.1-1:

a)

With core flow less than 39% of rated core flow, initiate action within 15 minutes to either:

1)

Leave the surveillance region within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or w

2)

Increase core flow to greater than or equal to 39% of rated' flow within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b)

With the APRM and LPRM# neutron flux noise level greater than three (3) times their established baseline noise levels:

1)

Initiate corrective action within 15 minutes to re-store the noise levels to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , otherwise 2)

Leave the surveillance region specified in Fig-ute 3.4.1.1-1 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

!:: !;;; h' T;;; Oces, tie, '.10.t.--

0 evels A an C of one LP string per core octant plus detector levels

etector i

A and C of one LPRM string in the center region of the core should be monitored.

l LA SALLE - UNIT 1 3/4 4-1 Amendment No.4U

I REACTOR COOLANT $YSTEM 3/.

2 SAFETY /RE IEF VALVES LIMITING CON 0! TION FOR OPERATION l

3 A.? % ufety-veh: '=:thr. Of eight= nrur :=1=t :y:^- n":ty!

InseYt M }ef-ve h:j 3:11 M ^^:r" : d tt. O.; :; uifi u ;; i eefety velve f.r.; tie.,

p 44.

21%"JdDlR:M unt$"W, 5 a.

4 safety / relief valves 91205 psig +3 d.

4 safety / relief valves 01175 psig +N, -5 e.

2 safety / relief valves 01150 psig +3, -5 APPLICA81LITY: OPERATIONAL [:0N0!TIONS1,2and3.

ACTION:

With the safety valve function of one er more of the above required

'a.

safety / relief valves inoperable, be in at least NOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i b.

With one or more safety / relief valves stuck open, provided that suppression 1 overa water temperature is less than 110*F, close

'nflu v Ivets

if unable te close the open valve (s) tar er if suppre)ssi pool average water temperature is i

the s k

l withi a

er greater, place the reac mode switch in the Shutdown si",iontop Ae deve yetufred th one er serv 6aTety/re11ef valve stas position indicators c.

i erable, restore the inoperableJ positten indicators to OP LE status within 7 days or be in at least HOT SHUTDOWN within 122 hours0.00141 days 0.0339 hours 2.017196e-4 weeks 4.6421e-5 months and in C WM within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 1

acDs

$URVEILLANCE REQUIREMENTS t

I l

4.4.2.1 The safety / relief valve stas position indicators of each safety / relief i

valve shall be demonstrated OPERA 8LE by performance of a:

CHANNEL CHECK at least once per 31 days, and a I

a.

CHANNEL CALIBRATION at least once per 18 months.**

b.

The low low set function shall be demonstrated not to interfere with 4.4.2.2 the OPERA 41LITY of the safety n11ef valves or the ADS by performance of a CHANNEL CALISAATION at least once per la months.

"The lif t setting pressum shall correspond to ambient conditions of the valves at nominal operating temperatuns and pressums.

Up to two inoperable valves may be replaced with span CPEAA8LE valves with lower setpoints until the next refueling outage.

- The provisions of specification 4.0.4 are not applicable provided the survett-Iance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate to perfore the test.

LA SALLE-UNIT 1 3/4 4-5 Amendment No. 28

i r

l I

INSERT D (T.S. 3/4.4.2. Pace 3/4 4-5) 1 3.4.2 - The safety valve function of 17 of the below listed 18 reactor coolant

[

system safety / relief valves shall be OPERABLE with the specified code safetyvalvefunctionliftsettin/Iallinstalledvalvesshallbe i

closed with OPERABLE position indication.

I i'

f r

f j.

3 I

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f I

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l I

t e

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4081K i

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- - - - " - ~ ~ ~

'N 3/4.6 CONTAINMENT $YSTEMS v

~

- - ~ ~

3/4.6.1 PRIMARY CONTAINMENT

~

PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATICN O

3.6.1.1 PRIMARY CCNTAINMENT INTEGRITY shg be saintained.

AP8LICA!!LITY: OPERATICNAL CCNDITIONS 1

and 3 ACTION:

Withcut PRIMARY CONTAINMENT INTEGRITY, restere PR. RY CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTCC'aN within the next 12 hcurs and in COLD SHUTCCkN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE REOUIREw!NTS

4. 6.1.1 PRIMARY CCNTAINMENT INTEGRITY shall be demonstrated:

a.

After each closing of each penetration subject to Type B tasting, or 5 test, by leak rate testing the seal with gas at Pa 39.g Type A except the primary containment air locks, if opened followino psig, v

and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Surveillance Requirement 4.6.1.2.d for all other Type 3 and C penetrations, the i

comnined leakage rata is less than or equal ts 0.60 La.

b.

At least once per 31 days by verifying tnat all crimar/ c:ntainment l

penetratiens** not caeanle of being closed by OPERABLE c:ntainment aut:matic isolation valves and.ecuired to be closed during accident c:ncitions are closed by valves, blind flanges, or deactivated aut:matic valves secured in ;csition, except as provided in Table 3.6.3-1 of Specification 3.6.3.

c.

By verifying each prist.ry c:ntainment air lock CPERAELE per Specification 3.6.1.3.

d.

By verifying the suppression chamcer CPEEABLE per Specification 3.6.2.1.

"See 5pecial rest F.xception 3.10.1

- Except valves, blind flanges, and deactivated aut:matic valves which are located insica the c:ntainment, and are locked, sealed or otherwise secured i

in the c' sed ':csiti:3.

These ;t strat"s shall te nMf tec :lesad hM.;

d naen CL3 SHJT::'.N e.:::: suca veM 'i:.: :n neet r;;t : 4 er :- ec '..nen

.e y c nt....,:......

.. - i. -...: sin:

n,e '..4 : 'arif1:a i n :r more of ten than once per 92 days.

s N ; ';;;i.,....

ro. v n.......

LA SALLE - UNIT 1 3/4 6-l' 4.

PLANT SYSTEMS 3/4.7.10 MAIN T1)RSINE SYPAS$ SYSTEM LIMITING CON 0! TION FOR OPERATION The main turbine bypass systes shall be OPERA 8LE.

3.7.10 OPERATIONAL CONDITION 1, when THERMAL POWER is greater than l

APPt.ICAll!LITY:

or equa. to l

ACTION: With th: ni-t;-tin; bp ::: ;y:t= i n; ;retl e,.i ;,',',. 2 ;,....

'Y

"U ;; I""

0;;

[iiOi52.ibi.;d32.00 5 57 5.n '.'.;!? 5.mm~.;323; i

5 x

n...

m..

.n. n. -. r.... n..

F Zasen Q VEILLANCE RECUIREMENTS The main turbine bypass systes shall be demonstrated OPERA 8LE at least 4.7.10 once per:

7 days by cycling each turbine bypass valve through at least one a.

complete cycle of full travel, b.

18 months by:

Perfonsing a systes functional test wnich includes simulated 1.

automatic actuation and verifying tnat eacn automatic valve l

actuates to its correct position.

t Demonstrating TijRSINE 8YPASS SYSTEM RESPONSE TIME to be less 2.

than or equal to 200 milliseconds r

I i

I I

e I

3/4 7-33 Assndment No.18 LA SALLE - UNIT 1 l

9 INSERT E (T.S. 3/4.7.10. Paes 3/4 7-33)

A.

With the main turbine bypass system inoperable :

[

t 1.

If at least four bypass valves are capable of accepting steam flow per Surveillance 4.7.10.a:

4 a) Witisin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , either:

l

1) Restore the system to OPERABLE status, or 2)

Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) to the main turbine bypass inoperable value per Specification 3.2.3.

i b) Otherwise, reduce THERMAL POWER to less than 25% oi RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

2.

If less than four bypass valves are capable of accepting steam flow per Surveillance 4.7.10.a:

a) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months increase the MCPR LCO to the main turbine bypass inoperable value per Specification 3.2.3, and b) Within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months restore the system to OPERABLE status.

i c) Otherwise, reduce THERMAL POWER to less than 25% of RATED THERMAL l

i POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

B.

The provisions of Specification 3.0.4 are not applicable.

I i

i

\\

l t

4081K 3

i

PECIAL TEST EXCEPTIONS 3

10.4 RECIRCULATION LOOPS hQ 6N, j

LIMITI CONDITION FOR OPERATION i

N i

3.10.4 The bequirments of Specification 3.4.1.1 that recirculation loops be in operation be suspended for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for the performance of:

i l

a.

PHYSIC TESTS, provided that THERMAL POWER does not exceed 5% of RATED TH L POWER, or

{

b.

The Startup est Program.

i APPLICABILITY:

OPERATIO

,L CONDITIONS 1 and 2, during first fuel cycle PHYSICS TESTS and the Initial Sta up Test Program.

ACTION:

a.

With the above specif d time limit exceeded, insert all control rods.

b.

With the above specified RMAL POWER limit exceeded, immediately s

place the reactor mode swi in the Shutdown position.

O SURVEILLANCE REQUIREMENTS 4.10.4.1 The time during which the above speci ed requirement has been suspended shall 'ce verified to be less than 24 ho s at least once per hour i

during PHYSICS TESTS and the Startup Test Program, i

4.10.4.2 THERMAL POWER shall be determined to be less than or equal to 5%

of RATED THERMAL POWER at least once per hour during PH ICS TESTS.

l 1

l LA,SALLE - UNIT 1 3/4 10-4 l

I

~

SPECI TEST EXCEPTIONS 3/4.10.

CONFIRMATORY FLOW INDUCEO VIBRATION TEST LIMITING DITION FOR OPERATION 3.10.7 The p visions of Specifications 3.6.1.1 and 3.7.3 may be suspendec to permit the 11 head to be removed and the RCIC system to be inoperaele with a nitrogen upply line connected to the reactor vessel at the RCIC injection connect on in order to perform the confirmatory flow induced In addition, the' vibration test pri r to first reactor criticality.

provisions of the f 110 wing specifications which are applicable during HOT SHUTDOWN may be spended so that the unit may be brought to HOT SHUTOOWN and maintain in HOT SHUTOOWN for the duration of the test by non-nuclear heatup pro dad that initial reactor criticality has not occurred. Upon success 1 completion of the test or initial reactor criticality, whichever oc urs first, this specification is cancelled.

Specification 3.3.. Table 3.3.2-1 for Trip Function a.

A.1.c.1, Main Ste Line Radiation - High Monitor, b.

Specification 3.3.7. 0. Table 3.3.7.10-1 for Instrument 1.a., Liquid Radwaste Effluent Line Monitor, c.

Specification 3.3.7.11 Table 3.3.7.11-1 for Instrument 1.a Noble Gas Activity nitor.

d.

Specification 3.4.3.1 fo the primary containment atmosphere particulate and gaseous r ioactivity monitoring systems.

Specification 3.5.1 for the ADS valves and "B" LPCI loop.

l.

e.

f.

Specification 3.6.1.1, 3.6.1 2, S,6.1.3, and 3.6.1.4.

Specification 3.6.2.1.

g.

h.

Specification 3.6.3, Table 3.6. -1 for valves in a.1, Main Steam Isolation Valves; a.3, Reactor olant System Sample Line Valves; a.10, LPCS, HPCS, RCIC, and RHR, jection Testable Check Bypass Valves; a.12, Drywell Pnei.ssatic Va ves; and a.14, TIP Guide Tube Valve Ball Valve.

3::n:q,'a>c '-'-"

'- ' ~ <

APPLICABILITY:

OPERATIONAL CONDITION 3, during per ormance of the confirmatory flow induced vibration test.

ACTION: With the provisions of the above specificatio not satisfied, be in COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLINCE REQUIREMENTS The reactor shall be verified not to Mve been critica with any 4.10.7 fuel assembly presently in the core within 24 h' burs prior to pe formance.of the test.

I LA SALLE - UNIT 1 3/4 10-7

3/4.2 POWER OISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature the postulated design basis loss-of-coolant accident will not exceed followingF limit specified in 10 CFR 50.46.

the 2200 AVERAGE PLANARllNEARSEAT GENERATION RA 3/4.2s1v v

v v

v This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exc ed the limit d GA* Won aho$55"'*5 NF M mi cA*^/"I

.specified.s maint C,FR 50.4,6. Tb. sfand +r ns M ora +'on5.

in 10.nd 4Le 5 me=*

ngrity The oak cladding temperature PCT) followin a post lated

-o-rate of all en er fuy at any axial location and is dependent only w

1 i

the rods of a fuel ass secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specific ion AVERAG P LI RH T GENE TI RAT y i s local paa ng of s

er to y

(APLH i

li;ith; =h; f;r A L"0" h :h= h Ff;;= 2.2.H, f:r te fer r.

.h:

al.;; en:.11 = L.iti;11;d ;, e feetor ef 0.0 h:; ;; m tha.

";=:h:; ;; = tha. PS : ltipihr S :t:-h:d fx: ;;r ch;h n:i:uhth inith; en.ly;te beta t ; rui.;eletien lee; end ;h '-

gigew

=i = cf th

n

,ecirceleti...Jovy v;.. u v...

R shown on Figures he cil ational procedure used to establish the A 3.2.1-k W6 s on a loss-of-coolant accident analysis, nalysis was i

are o,J 3.21-a r performed usi g General Electric (GE) calculational models wh c A complete th the requirements of Appendix K to 10 CFR 50.

on of each code employed in the analysis is presented in Reference 1.

consist n fferences in this analysitcompared to previous analyses performJd with di -

(1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figure 3.2.1-1, (2) fission Reference 1 are:

product decay is computed assuming an energy' release rate of 200 MEV/ fi (3) pool boiling is assumed after nucleate boiling is lost during the flow I

stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitation as described in Reference 2, are included in the ooding calculations.

- w F -+

A u tt e sn: e ga m = t p unt e;;t ; = = = = = tn: = * = : = t

=h=t =1y:S S 3==ud b :== T= : :.2.1-:.

Amendment No.18 LA SALLE - UNIT 1 B 3/4 2-1

INSERT F (Bases 3/4.2.1, Page B3/4 2-1)

The APLHGR 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 analysis (Reference 3) performed for Cycle 3, used bounding MAPLHGR values of 13.0 and 14.0 kw/ft, independent of nodal exposure. These MAPLHGR values are higher than the expected "thermal-mechanical MAPLHGR" for both BP8x8R and GE8x8EB fuel. Therefore, SAFER /GESTR established that for all BP8x8R and GE8x8EB fuel designs the MAPLHGR values are not expected to be limited by LOCA/ECCS considerations. However, MAPLHGR values are still required to assure that the LHGR limits are not compromised and, consequently, fuel rod mechanical integrity is maintained.

4081K P

O EE.L.E.T 11 -

7 Sases Table 6 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters; co n THEAMAL POWER................... 3489 Mwt" wnich correscends to 105% of rated staan flow 0

14.87 x 10 lba/hr wnich Ves sel Steas Output...................

corresponds to 105% of cated steam flow Vessel Steas Oose Pressure............

1055 psia Oasign Basis Recirculation Line Break Area for:

2 a.

Large Breaks 3.1 ft,

2 b.

Small Breaks 0.10 ft.

-],

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN NINIMUM i

LINEAR HEAT IAL CRITICAL i

FUEL BUNDLE GENEMTION MTE P'

ING PCWER l

RJEL TYPE GEOMETRY (kw/ft)

FA OR RATIO Initial Core 8x8

13. 1

1. 4\\

1.18 i

i A sore detailed listing of input of each sodel and its so rce is presented SAR.

in Section II of Reference 1 and subsection 15.0-1 of the

%is power level tests the Accendix requirement of 102%.

core ation of heatus calculation assumes a bundle power consistant with oc LINEAR the highest powered red at 102% of its Technical Specificatic j

HEAT GENEMTION MTE limit.

i J.,

-l 8 3/4 2-2 l

LA SALLS - UNIT 1

POWER OISTRIBUTION SYSTEMS BASES 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 RATED THERMAL The flow biased simulated thermal power-upscale scram setting and POWER.

control rod block functions of the APRM 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 The or that > 1% plastic strain does not occur in the degraded situation.

scram settings and rod block settings are adjusted in accordance with the for-mula in this specification when the combination of THERMAL POWER and MFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational For any abnormal operating transient analysis evaluation with the transients.

initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnomal operational transient, the most liaf ting transients have been analyzed to determine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR), The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and The limiting transient yields the largest delt coolant temperature decrease.

When added to the Safety Limit MCPR, the required minimum operating MCPR.

lo..

sga:t Q limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3.2.

tM eevi7itT1

..a rs Th eva shown in FSAR Table 15.0-at are input to a GE-core dynamic behavio tra s n sed to evaluate escribed ote-8-M,,,(4' F";=*1.:.:.ti:n events computer program. The co M

..... ' 21.=N E R E:

m

...nte

...or.......

in m om.

Gloo;N The. outputs of Eprograssalong with the initial MCPR _(germe 9) m, y...._

m, 4u form the input for further analyses of the thermally limiting bunale%itn f,

j

WM W-H tm&nt '2:=1 %d=1': T'iC ::d:

The principal result of this evaluation is the reduction in MCPR caused b the d t

transient.

J The need to adjust the MCPR operating limit as a function of scras time arises from the statistical approach used in the implementation of the 00YN g

Generic statistical q

computer code for analyzing rapid pressurization events.

analyses were perfonned for plant groupings of similar design which considared the statistical variation in several parameters, i.e., initial power levei, CRD scras insertion time, and model uncertainty. These analyset, which are LA SALLE - L%1T 1 8 3/4 2-3 Amendment No.18

--- i

~

)

i INSERT G (T.S. Bases 3/4.2.3, Page B3/4 2-3)

When the Rod Withdrawal Error 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 RBM setpoint. The flexibility of the variable RBM setpoint/MCPR limit allows efficient use of the extended operating domain (ELLLA region), while maintaining transient protection with the more restrictive MCPR limit.

Analyses have been performed to determine the effects on CRITICAL POWER RATIO (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 performed assumed a single failure only and established the licensing bases to allow continuous plant operation with the analyzed equipment out of service. The following single equipment failures are included as part of the transient analyses input assumptions:

1) main turbine bypass system out of service,

2) recirculation pump trip system out of service,

3) safety / relief valve (S/RV) out of service, and

4) feedwater heater out of service (corresponding to a 100 degree F reduction in feedwater temperature).

For the main turbine bypass and recirculation pump trip systems, specific cycle-independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation (LCO) values are established to allow continuous plant operation with these systems out of service. A bounding end-of-cycle exposure condition was used to develop nuclear input to the transient analysis model.

The bounding exposure condition assumes a more top-peaked axial power distribution than the ncminal power shape, thus yielding a bounding scram response with reasonable conservatisms for the MCPR LCO values in future cycles. The cycle independent MCPR LCO values shown in Figure 3.2.3-lb for the main turbine bypass and recirculation pump trip systeias out of service are valid provided:

1) The cycle specific analysis for the Load Reject Without Bypass and Turbine Trip Without Bypass events yield MCPR LCO values less than or equal to 1.33 and 1.29 for Options A and B, respectively.

2) The cycle specific analysis for the Feedwater Controller Failure event yields MCPR LCO values less than 1.25 and 1.21 for Options A and B, respectively, when analyzed with normal feedwater temperature.

The analysis for main turbine bypass and recirculation pump trip systems inoperable allows operation with either system inoperable, but not both at the same time.

INSERT G i For operation with the feedwater heater out of service, a cycle specific analysis will be performed. With reduced feedwater temperature, the Load Reject Without Bypass event will be less severe because of the reduced core steaming rate and lower initial void fraction.

Consequently, no further analysis is 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 analysis for the feedwater controller failure event will be performed with a 100 degree F feedwater temperature reduction.

The calculated change in CPR for that event will then be used in determining the cycle specific MCPR LCO value.

In the case of a sinole S/RV out of service, transient analysis results showed that there is no impact on the calculated MCPR LCO 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 loop operation.

t 4081K

1

)

POWER OISTRIBUTION SYSTU45 BASES MINIMUM CRITICAL MwsR RATT XCont ued) described further in Refe ce duced generic Statistical Adjustment Factors whien have been acp plant and cycle specific 00YN results to yield operating limits which provide a 95% procanility with 95% confidence that the limiting pressurization event will not cause NPR to f all below the fuel cladding integrity Safety Limit.

As a result of this 95/95 accroacn, the average 20% insertion scram time igust be monitored to assure comoliance with the assumed statistical distribu-tion.

If the sean value on a cycle cumulative, running average, basis were to exceed a 5% significar:e level coscared to the distribution assumed in tne 00YN statistical analyses, the NPR limit must be increased ifnearly, as a function of the mean 20% scram time, to a more conservative value wnica reflects an NRC determined uncertainty penalty of 4.4%.

This penalty is applied to the plant soecific ODYN results, i.e. without statistical adjust-sent, for the limiting single f ailure pressurization event occurring at the limiting point in the cycle.

It is not apolied in full until the mean of all current cycle 2C% scras times reaches the 0.86 seconds value of Scocifica-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 > 0.86 secs, and the required actions taxen well before the running average exceeds 0.86 secs.

m The 55 significance level is defined in Reference 4 as:

Ig=

+ 1.65 (N / I N )

2, t

j i=L s

sean value for.atistical ser time dis ibution

=

where p l

to 205 insert m.dae 0 fe%

standard deviat of ao v stributi

= M O o t (,

cr

=

numeer of rods tes ad at SOC, i.e., al operaale Ng

=

rods i

i l

n l

t I N, a total nuncer of operan

ested in the l

i=1 '

current cycle

o. (oM l

The value for t used in Specification 3.2.

is +:131r'sec/nds wnich is t

g conservative for the following reason:

t For simplicity in formulating and isolamenting the t.C0, a conservative M

value for IN of 598 was used. This represents one full core data set g

i=1 at SCC plus one full core data set following a 120 day outage plus tnelve 1

10% 'of core,19 rods, data sets. The 12 data sets are equivalent to l

24 operating 'montns of surveillance at the increased surveillance

. V) frequency of one set per 60 days required by the action statements of l

Specifications 3.1.3.2 and 3.1.3.4.

I i

LA SALL! - UNIT 1 3 3/4 2 4 l

91b' POWER OISTRIBUTION SYSTEMS BASES MINIMUM CRITTCAL POWG RATIO (Continued)

References:

General Electric Comoany Analytical Model for Lass-of-Coolant 1.

Analysis in Ac:ordance with 10 CFR 50, Accencix X, NEE-20556, Novencer 1975.

. 3. '.i-f;rd, 'nlyti ni "n m,;; ;f '1;n *.;;;i; = ~ ;'. nica; ';r 2.

. OC A",.~. e. j M70 ("C0 M000;.

353. $..$.,$9..$.55. i$_ $.5 $."3_b. U_i 59', N..._Id. ' "_E51'"I '

";r th: *r=:f ut 'niy:f: :? : I' ;' -

'A00 01-A 0

t: r r;F =iptien, NC ::14~, J.n..,y l'd,0.

"h u mi, Tec ni ni 0;;;r "Qualification of the One-0imensional Core Transient Model for 3/.

Soiling Water Reactors" General Electric Co. Licensing Topical Report NGC 24154 Vols. I and II and NEE-24154 Vol. III as suo-plemented by letter dated Septancer 5,1980, fres R. N. Suchholz S.

(GE) to P. 5. Check (NRC).

3/4.2.4 LINEAR HDT GENGATION RATE The specification assures that the LINGR HEAT GOERATION RATE (L%R) in any red is less than the design linear heat generation even if fuel pellet The power spike penalty specified is based on densification is postulated.

the analysis presented in Section 3.2.1 of the GE topical report NEM-10735 l

Suoplement 6, and assumes a linearly increasing variation in axial gaps l

between core bottos and top and assures with a 95% confidence that no incre than one fuel red exceeds the design LINEAR HEAT GOERATION RATE cue to power l

spikin

3.

l aSalle. Codhty 5fation W'ts I and 3. SAF6R,[6(sTA-Loc.A L e s s. o F - ceolen t.

A cciJed An lyses," Creneral Eledric Comp y Ref e r+

NE DC.- 3 IS~lo P,DeJe416er /197.

4 "Generai Elech-ic S+ondard Applicob'on for Reactor N el,"

NE0 E-a9ois -P-A (Ides + oppreved revision )-

3 I

r. Extended opera +in3 oornain and Egw'pmed Out of service.

3 for L45elle. Ceu,aY Ahclea, Station Wi+5 / and A,"

NEDt -3/43.

Nevember 198'1.

v SALLE - UNIT 1

,3 3/4 2-6 l

i

)

INSTRUNENTATION BASES 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATICN The anticipated transient without scram (ATVS) recirculation cumo trip i

systes provides a means of limiting the consecuences of tne unlikely occurrence The response of the of a f ailure to scram during an anticicated transient.

clant to tais costulated event falls within the envelcoe of stucy events in General Electric Comoany Tooical Recort NE00-10349, catac Maren 1971 and NE00-24222, cated Cecemoer,1979, and Accencix G of the FSAR.

The enc-of-cycle encirculation pumo trip (EOC RPT) systas is a part of the Reactor Protection System and is an essential safety stoolement to the reactor trio.

The purpose of the EOC-RPT is to recover the loss of thermal The physical pnenomenon involved is margin wnien occurs at the end-of-cycle.

tnat tne void reactivity feecback due to a pressurization transient can acd

ositive reactivity to the reactor system at a faster rate tnan tne control Each EOC-RPT systas tries both recircula-rods r.dd negative scram reactivity.

tion pumos, reducing coolant flow in order to reduce tne void collaese in the The two events core during two of the most limiting pressuri:ation events.

for wnich the EOC-RPT protactive feature will function are closure of the reine stop valves and fast closure of the turbine control valves.

15% N A fast closure sensor frem each of two turbine control valves provides rput to the EOC RPT systas a fast closurt sensor from each of the other two Similarly, tunine control valves provides input to the second ECC-RPT system.

a cosition switch for each of two turbine stop valves provides input to one EOC-RPT systes; a position switch from each of the other two sten valves For each ECC-RPT system, the orovides input to the other EOC RPT system.

sensor relay contacts are arranged to fem a 2-cut of 2 logic for the fast closure of turoins control valves and a 2-out-of-2 logic for the turoine stop valves. The operation of either logic will actuate tne ECC-RPT system and trip both recirculation pumps.

Each ECC-RPT systes say be manually bypassed by use of a keyswiten wnich The manual bypasses and the automatic doerating is administratively controlled.

Sypass at less than 30% of RATED THERMAL PCWER are annunciated in the control roo3.

The EOC RPT system response time is the time assumed in the analysis tetneen initiation of valve motion and comolate suceression of the electric i.e.,190 ms, less the time allotted for sensor rescense, i.e.,10 ms, arc,less the time allotted for breaker are suportssion determined by tast, and as correlated to manuf acturer's test results, i.e., 83 ms, and plant pre-operational test results.

3/4.3.5 REACTOR CORE ISOLATICH CCOLING SYSTEN ACTUATION INSTRUMENTA The reactor core isolation cooling system actuation instrumentation is l

provided to initiate actions to assure acequata cars cooling in tne event of reactor isolation from its primary heat sink and the loss of fetewater flow to the reactor vessel without providing actuation of any of the emergency cort

~

's cooling equipment.

LA 5 LLE - UNIT 1 8 3/4 3-3

P-INSERT H (T.S. Bases 3/4.3.4. Page B3/4 3-3)

A generic analysis, which provides for continued operation with one or both trip systems of the EOC-RPT~ system inoperable, has been performed. The analysis determined bounding cycle independent MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Conoition for Operation (LCO) values which must be used if the EOC-RPT system is inoperable. These values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed transient with the RPT function inoperable. The analysis results are further discussed in the bases for specification 3.2.3.

4081K

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has been evaluated and been found to be acceptable provided the unit is operated in accordance with the single recirculation loop operation Technical Specifi-cations 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 jet pump performance on a prescribed scheduled for significant degradation.

Recirculation loop flow mismatch limits are in compliance with the ECCS The limits will ensure an adequate core flow LOCA analysis design criterion.

Where the recircu-coastdown from either recirculation loop following a LOCA.

lation loop flow mismatch limits can not be maintained during the recirculation 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 The loop temperature must also be within prior to startup of an idle loop.50*F of the reactor pressure vessel coolant tem Since the coolant shock to the recirculation pump and recirculation nozzles.

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.

The possibility of thermal hydraulic instability in a BWR has been inves-Based on tests and analytical models, tigated since the startup of early BWRs.it has been identified that the high This region maybe encountered map is the region of least stability margin. sequence exchanges, and as a result during startups, shutdowr,t, tion pump (s) trip event.

To ensure stability, single loop' operation is limited in a designated Single loop restricted region (Figure 3.4.1.1-1) of the power-to-flow map.

operation with a designated surveillance region (Figure 3.4.1.1-1) of the power-to-flow map requires monitoring of APRM and LPRM noise levels-I 3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety elief valves operate to prevent the reactor coolant system from being press ized above the Safety Limit of

" tetel ei 15 G Wn; C safety /

1325 psig in accordance with the ASME Code.

ME III relief valvegi; r; gird ;; li;;it-reactor pressure to withi a limWd Inse r f allowable values for the worst case upset transient, 1

Demonstration of the safety-relief valve lift settings will occur only

[e.

during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

B 3/4 4-1 Amendment No. 40 LA SALLE-UNIT 1

INSERT I (T.S. Bases 3/4.4.2 Pace B3/4 4-1)

Therefore, operation with any 1*? 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 boundry is known to exist at all times.

4081K

EMERGENCY CORE C00 LING SYSTEMS BASES ECCS-0PERATING and SHUTDOWN (Continued) the suppression pool into the reactor, but no credit is taken in the hazarde analyses for the condensate storage tank water.

With the HPCS system inoperable, adequate core cooling is assured by the OPERA 8ILITY of the redundant and diversified automatic depressurization systes In addition, the n actor core isciation and both the LPCS and LPCI systans.

cooling (RCIC) systas, a system for which no credit is taken in the hazards analysis, will automatically provide askaup at reactor opereting pressures se The HPCS out-of-service period of a reactor low water level condition.

14 days is based on the demonstrated OPERA 81LITY of redundant and divenified low pressure core cooling syntass.

The surveillance requirements provide adequate assurance that the HPCS Although all active components are system will be OPERA 8LE when required.

testable and full flaw can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test with reactor v injection requires reactor shutdown.

full to prevent water hammer damage and to provide cooling at the earliest some l

Upon failure of the HPCS systas to function properly, if mquired, the automatic depressurization system (ADS) autoestically causes selected safety-I 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 fue tesperature to less than 1200'F.

whenever nactor vessel pressure exceeds 122 psig even though lw pressure core cooling systems provide adequata core cooling up to 350 psig.

Six ADS automatically controls seven selected safety-relief valves.

valves are required to be OPERA 8LE cit %:;h 'M M::rc ;21M: =1;, tM; c It is therefore appropriate to permit one of the requi md valves to be out-of-service for up to 14 days without materially 4 4

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reducing systas reliability.

Mce the tocA aulysis doy/e ( A r6 J

in aM t/e

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SUPPRESSICN CHA>SER 3/4.5.3 ECCS to ensure that a sufficient supply of water is av This limit on suppression LPCS and LPCI systems in the event of a LOCA. chamber minimu The OPERA 8ILITY of the suppression recirculation cooling flee to the core. chamber in 07EAATIONAL COND Repair work night require making the suppression chamber inoperable.

This specification will permit those repairs to be l

in when the suppression chamber sust be made inoperable, including draining, CPERATIONAL CONDITION 4 or 5.

In OPERATICHAL CONDITION 4 and 5 the suppression chamber m below water volume is reduced because the reactor coolant is maintaine Since pressure suppression is not required below 212'F, the minimu lsa water volume is based on NPSH, recirculation volume, vortax pmvention p u 200*P.

2'-4" safety sargin for conservatism.

Amendment No.19 8 3/4 5-2 I

LA SALLE - UNIT 1

PLANT SYSTEMS BASES SNUBBERS (Continued)

Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan

as described in "Quality Control and Industrial Statistics" by Acheson i

J. Duncan.

Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the com-Snubbers so exempted pletion of their fabrication or at a subsequent date.shall be listed in the list the exemptions.

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature The requirement to monitor the snubber service life is included area, etc.).

to ensure that the snubbers periodically undergo a performance evaluation in These records will provide view of their age and operating conditions.

L statistical bases for future consideration of snubber service life.

3/4.7.'O MAIN TURBINE BYPASS SYSTEM

- h s ia t..Line ;,,,eee.iet ie. n oi e4 0" "J" C ee e.--.d in ::.e -

f::ft:t:r ::ntr:1?:r f:flur: :n:!y:f:.

L sso<r J LA SALLE - UNIT 1 B 3/4 7-5 Amendment No.18

INSERT J (T.S. Bases 3/4.7.10, Page B3/4 7-5)

A generic analysis, which provides for continued operation with the main turbine bypass system inoperabic, has been 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 bypass system is inoperable. The MCPR LCO values ensure that adequate reactivity margin to the MCPR safety limit exists in the event of the analyzed 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 accepting steam flow. The analysis results are further discussed in the bases for Specification 3.2.3.

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r CESIGN FEATURES 5.3 REACTCR CCRE FUEL 455E98L:55 5.3.1 The react:r ::re sn.sil c:ntain 754 fuel asserolies witn eacn fuel assemoly centaining 52 f Je1 recs and t'no water recs clad witn Zircaloy -2.

Esca fuel rec sna11 nave a nominal active fuel leng n of 150 inenes.

The initial core loading snall nave a maximum aversgo enrienment of 1.39 weignt carcent U-235.

Reloac fuel sna11 :e similar in ;nysical design to tne initial core leading.

OCNitCL 400 ASSE99L 53 T w rs + wop ss,'ble fyges oTeon+rol r 4d s, ona WWfi.ain 185 son..

red ass clies.Y-eeen.

5. 3. 2 The i sactor c:nsist a.ruciform array of stainless stae ning 143 inenes d

. ren carsica, S.C. powcer surrounced by a cruciform snace el a

sneata arul Fke sach d type contains le la<.hu o f obsceber mc+e r/s l o $ w hick x

At Fies t- (o incke $ are k $nlsm and &be remainder ig S C.

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5. 4

.A

. v R CCC LAN. SYSTEM

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.V _./

OESIGN 2RE55URE ANO *WPERATURE 5.4.1 The reactor coolant system is designed and shall be saintained:

a.

In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degrecation pursuant to the acclicaole Surveillance Requirements, 3.

For a pressure of:

L.

1250 psig on tne suction side of the recirculation cumns.

~

2.

1550 psig frem the recirculation pumo disenarge to the outlet side of the disenarge snutoff valve.

i 3.

1500 csig from the disenarse snutoff valve to the jet pumes.

I c.

For a temocrature of 575*F.

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VCLUME I

i

5. 4. 2 The total water and steam volume of tne reactor vessel and recirculation system is

21,000 cueie feet at a neminal 7,y, of $33*F.

5.5 WET 20RCLOGICAL TC%ER LOCATTCN l

I 5.5.1 The seteorological tower sna11 te located as snown on Figurs 5.1.1-1.

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l LA SALLE - UNIT 1 5-4 i

h

ATTACHMENT E SIGNIFICANT HAZARDS EVALUATION Commonwealth Edison proposes to amend Facility Operating License NPF-11 for LaSalle Unit 1 to support the cycle 3 core reload. The proposed revisions include =three basic types of changes; changes specific to the cycle 3 reload fuel and analyses including the new SAFER /GESTR-LOCA. Loss-of-Coolant Accident 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.

DESCRIPTION OF AMENDMENT REQUEST CONTENTS Commonwealth Edison has evaluated the proposed Technical Specifications and deternined that they do not represent a significant hazards consideration. The Technical Specification changes specific to the cycle 3 reload fuel and analyses include the following. Each of these is discussed I

with respect to the three questions of 10 CFR 50.92(c).

1) Incorporation of the Cycle 3 Minimum Critical Power Ratio (MCPR) limit, and new values and references resulting from the new ODYN methods.

2) Addition of the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits for the reload fuel and revision of the bases section to reflect the new bases of the MAPLHGR limits.

l

3) Addition of an LHGR limit specific to the GE8X8EB fuel.

j

4) Deletion of the requirement for the MAPLHGR reduction factor in single l

loop operation.

5) Revising the control rod assemblies section of the Design Features to be applicable to blades which contain hafnium as an absorber material, i

The Technical Specification changes resulting from analyses performed to expand the operating region and to allow certain equipment out-of-service include:

6) Changes to to the Average Power Range Monitor (APRM) flow biased simulated thermal power and Rod Block Monitor (RBM) upscale setpoints in dual and single loop operation and a requirement to clamp the RBM setpoint, due to the extended operating domain analysis.

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ATT. E 7) Changes to the End-Of-Cycle Recirculation pump Trip (EOC-RPT)

Technical Specifications to allow continued operation when the system is inoperable provided the MCPR limit is increased to the corresponding value.

8) Revise the safety / relief valve (S/RV) Technical Specifications to require action only after two S/RVs are found to be inoperable.

9) Changes to the Main Turbine Bypass System Technical Specifications to allow continued operation when one bypass valve is incapable of accepting steam flow.

The Technical Specification changes provided for clarification or as administrative changes include:

10) Deletion of the GEXL correlation and GETAB statistical model in the bases of the safety limit section.

11) Deletion of the footnotes which allow control rod 10-47 to use position 46 as the fully withdrawn position for cycle 2.

12) Deletion of the Special Test Exceptions for the recirculation loops and the confirmatory flow induced vibration test.

13) Revision to the Control Rod Program controls Technical Specification to require the RWM be demonstrated operable in operational condition 1, prior to reaching 20% power, when reducing thermal power.

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14) Correction of the Bases statement for Technical Specification 3/4 5.1 reflecting the input assumptions for the LOCA analysis regarding analyzed combinations of ADS System failures.

Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92(c), operation of LaSalle Unit 1 throughout Cycle 3 in accordance with the proposed changes will not:

i involve a significant increase in the probability or consequences of a.

an accident previously evaluated because:

1) The incorporation of the MCPR limits noted above is explicitly provided to establish limits on reactor operation which ensure that the core is operated within the assumptions and initial conditions of the transient analyses. Operation within these limits will ensure that the consequences j

of a transient or accident remain within the results of the anelyses. The l

probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected.

i 1

ATT. E 2) The incorporation of the proposed MAPLHGR limits establishes limits on reactor operation to ensure thermal-mechanical integrity of the fuel. In addition a SAFER /GESTR-LOCA analysis, using NRC approved methodology, was pertormed which demonstrated that MAPLHGR limits assure Peak Clad Temperatures which are approximately one half of the 10CFR50.46 requirements. The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected.

3) GE has calculated the LHGR limit for the GE8X8EB fuel using the GESTR-MECHANICAL code, which has been found acceptable by the NRC and demonstrates that with the new LHGR limit, the fuel design basis criteria are satisfied for GE8X8EB fuel. The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected.

4) A SAFER /GESTR-LOCA analysis using NRC approved methodology was performed for Single Loop Operation (SLO) and demonstrated that no MAPLHGR reduction was required, due to the increased margin in Peak Clad Temperature (PCT) over the previous LOCA analysis. The probability of an accident is not affected by this change because no physical systems or equipment which could initiate an accident are affected.

5) The change in the control blade description allows the use of control blades with hafnium metal. This is an improved design which extends control blade life and reduces the probability of blade cracking and potential loss of absorber material.

6) The proposed changes to the APRM setpointa increase the allowable operating region. This expanded operating region has been analyzed by GE using NRC approved methods to determine the required operating rastrictions (MCPR). The resulting MCPR limit is bounded by the proposed Cycle 3 MCPR. The RBM setpoints have been revised to ensure operation within the assumptions of the Cycle 3 Rod Withdrawal Error Analysis. The probability of an accident is not significantly increased because operation within these setpoints does not alter the normal operation of the equipment, whose failure have been previously analyzed.

7) Changes to allow the EOC-RPT to be inoperable require an increase in MCpR to ensure operation within the asr.umptions and initial conditions of the transient analysis, with this increase in the MCPR LCO equivalent protection is provided. The EOC-RPT is not assumed in the LOCA analysis; therefore, this change has no impact on the accident analysis.

ATT. E

_4_

8) The safety function of the safety / relief valves are only taken credit for in the overpressurization event. GE has performed an overpressurization analysis with the safety function of one S/RV out-of-service and showed that the change in pressure is small and adequate margin to the ASME code limit still exists.

9) This change, which allows one of the turbine bypass valves to be inoperable, requires an increase in the MCPR LCO to ensure operation within the assumptions and initial conditions of the transient analysis.

The MCPR LCO was determined assuming the entire turbine bypass system inoperable, adding extra conservatism. The probability of occurrence has not significantly changed as only one bypass valve out of five is allowed to be deliberately out-of-service. This change has no impact on the LOCA analysis since the turbine bypass system operability is not included in the assumptions to the analysis.

10) This change deletes information in the bases of the Technical Specifications that is overly detailed and has no affect on any systems or limits on reactor operation.

11, 12)

These changes involve deleting information that is no longer required as the exceptions are no longer valid at the beginning of Cycle 3.

13) This change clarifies the Technical Specification requirement to require the RWM to be operable before reaching 20% power. This change does not alter the operation of the system in any way.

14) This change is administrative in nature only and corrects an error in the l

bases discussion.

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4081K

l ATT. E b.

Create the possibility of a new or different kind of accident from any accident previously evaluated because:

1, 2, 3 & 4)

The proposed MCPR, MApLHGR, and LHGR limits represent limitations on core power distribution which do not directly affect the operation or function of any system or component. As a result, there is not impact on or addition of any systems or equipment whose failure could initiate an accident.

5) This change allows an improved design control blade to be installed as a replacement for the current blades. This improved blade design will function the same as the current blade design.

6) The proposed APRM and RBM satpoints represent changes to the core power and flow distribution and do not significantly affect the operation or function of any system or component. As a result, there is no significant impact on or addition of any system or equipment whose failure could initiate an accident.

7 & 9)There is no impact on or physical modification to the EOC-RPT and main turbine bypass systems and/or components whose failure could initiate an accident.

8) This change impacts a previously analyzed event as discussed in item a.

There is no impact or physical modifications to the system or components, whose failure could initiate an accident.

10, 11, & 12)

These changes are administrative in nature and have no impact on or addition to any system or equipment whose failure could initiate an accident.

13) This change clarifies the Technical Specification requirement to require the RWM to be operable before reaching 20% power. This change has no impact on any system or equipment whose failure could initiate an accident.

14) This change is administrative in nature only and corrects an error in the bases discussion.

4081K I

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ATT. E c.

Involve a significant reduction in the margin of safety because:

,1, 6, 7 & 9)

These changes have been analyzed to demonstrate that the consequences of transients or accidents are not increased, using the specified restrictions, beyond those previously evaluated and accepted at LaSalle.

The analyses show that the MCPR safety limit and steam dome pressure safety limit are not violated.

2, 4)

A new LOCA analysis has been performed using improved methodology, approved by the NRC.

This analysis resulted in large improvements in PCTs over the previous LOCA analysis, even wr.en the proposed changes are incorporated.

3) The LHGR limit is specific to GE8X8EB fuel and has been analyzed by GE and approved by the NRC.

5) This change allows for an improved control blade design to be installed, which has a decreased probability of cracking and absorber loss; thereby, increasing the margin of safety.

8) There is only a small increase in pressure and adequate margin to the safety limit for the steam dome pressure still exists, as well as margin to the ASME overpressurization limit.

10, 11, 12 & 13) l These changes are all administrative in nature, either deleting information that is no longer applicable, providing clarification to

+

current specifications, or information in the bases considered overly detailed.

14) This change is administrative in nature only and corrects an error in the bases discussion.

l Based on the above discussion, Commonwealth Edison concludes that the proposed amendments do not represent a significant hazards consideration.

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ML20148U192

Text

References:

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