ML20198B917

Proposed Tech Spec Changes Re First Reload

ML20198B917
Person / Time
Site:	LaSalle
Issue date:	10/22/1985
From:	COMMONWEALTH EDISON CO.
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ML19344C148	List: ML20198B912 ML20198B917 ML20198B931
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. - _ - _ . _ _ _ _ _ _ _ _ - _ . _ _ _ _ - _ _ _ - _ _ _ _ _ _ _ - _

ATTACHMENT _ B PROPOSED CHANGE TO APPENDIX A I

l TECHNICAL SPECIFICATION TO OPERATING LICENSE  !

(F-ll i

REVItFD PAGES: VI 3/4 2-4 XIX 3/4 2-5 2-1 3/4 2-6 8 2-1 3/4 3-39 l B 2-4 3/4 4-1  ;

8 2-5 3/4 4-la D 2-6 3/4 4-Aa (new page) b 2-7 3/4 4-40 (new page) 3/4 2-1 0 3/4 4-1 3/4 2-2 3/4 2-2(a) (new page) l 079'K 8511070362 851022 PDR ADOCK O$000373 P PDR

E INDEX I.j . LIMITINGCON0!TIONSFb'ROPERATIONANDSURVEILLANCEREQUIREMENTS l -

t l

SECTION PAGE 4

i 3/4.4 REACTOR C0OLANT SYSTEM ,

3/4.4.1 RECIRCULATION SYSTEM l

,, Recirculation Loops.......................................... 3/4 4-1 i

Jet Pumps.................................................... 3/4 4-2 Recirculation Loop F1cw...................................... 3/4 4-3 Idle Recirculation Loop Startup.............................. 3/4 4-4 i Wennel Edvaulie. Shbrith i

3/4.4.2 SAFETY / RELIEF VALVES.........T............................... 3/44-4*-l 3/4 4-5

~

3/4 4.3 f.EACTOR COOLANT SYSTEM LEAKAGE l Leakage Detection Systems.......... ......................... 3/4 4-6 i Operational Leakage.......................................... 3/4 4-7 h) 3/4.4.4 CHEMISTRY.................................................... 3/4 4-10 u 4.5 s,ECI,1C ACTIv!TY............................................ 3/4 4-u j 3/4.4.6 PRE 55URE/ TEMPERATURE LIMITS i

Reacto r Coo l ant Sy s tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 4-16 I

l Reactor Steam 0 cme............................'............... 3/4 4-20, l 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES............................. 3/4 4-21

'l 3/4.4.8 STRUCTURAL INTEGRITY.........................................

. 3/4 4-22 3/4.4.9 RESIDUAL HEAT REMOVAL

• Hot 5hutdown................................................. 3/4 4-23

. Cold Shutdown................................'............... 3/4/4-24 J/4.5 EMERGENCY CORE CCOLING SYSTEMS .

3/4.5.1 ECCS-0PERATING............................................... 3/4 5-1 i 3/4.5.2 ECCS SHUT 00WN................................................ 3/4 5-6 3/4.5.3 SUPPRE$i!Ci JNAMSER......................................... 3/4 5-5 n.

LA SALLE - UNIT 1 VI -

t .

INDEX '

i ,

l I LIST OF FIGURES -

( .

PAGE FIGURE ,

l 3.1.5-1 SODIUM PENTA 80 RATE SOLUTION TEMPERATURE / 3/4 1-21 CONCENTRATION REQUIREMENTS .............................

3.1.5-2 SODIUM PENTABORATE (Na28 t o0s. 10 H2O) 3/4 1-22 VOLUME / CONCENTRATION REQUIREMENTS ......................

3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, 807) l INITIAL CORE FUEL TYPES SCR100, OCR233. A'D-NCR8t% FCRS219; AND Jt's 3/4 2-2

-eGAME.................................................

3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS 3/4 2-5 t AT RATED FLOW ........................................

3/4 2-6 3.2.3-2 X FACTOR ..............................................

7 W

! 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE 3/4 4-18

) VS. REACTOR VESSEL PRESSURE . . . . . . . . . . . . . . . . . . . . . . . . . . . .

3/4 7-32 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST . . . . . . . . . . . . .

B 3/4 3-7 8 3/4 3-1 REACTOR VESSEL WATER LEVEL . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

8 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E)1MeV) at' 1/4 T 8 3/4 4-7 AS A FUNCTION OF SERVICE LIFE .......................... l 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS 5-2 AND LIQUID EF FLUENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

5-3 5.1.2-1 LOW POPULATICN ZONE ............'........................

6-11 s.1-1 . CORPORATE MANAGEMENT ...................................

UNIT 0;CANIZATION ......................................

6-12 6.1-2 i

6-13 6.1-3 MINIMUM SHIFT CREW COMPOSITION .........................

~ 3. 2.1 -2 MMIMuM MEME MN WERE HST WERATiod f

RATV (MAPL)M10 VER1LAS MVER$GE PtRNAR EIFOSuRE,

( ML Tt1E' ENCRB 299 t.,

.S h 2 2 M 3.t 1.5 -1 Cops THttsAL Fo(mER (% op Rawe vresus 3/9 q _9;;,

TOTAL Cogt FLOW l% 6FRAmp) .

XIX Amendment No.

LA SALLE - UNIT 1

p. _ .

- 2. 0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS ' .

THERMAL POWER. Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with.the reactor vessel steam done pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

i

. ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated '

flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.

THERMAL POWER, High Pressure and High Flow

/.07 ,.

2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less thanJdNI with two recirculation loop operation and shall not be less thanMwithgg l single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2.

ACTION:

/.07 With MCPR less than.1Mwith two recrculation loop operation or less than

/,d$ 1drTwith single recirculation loop operation and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least H01 SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require-ments of Specification 6.4 ,-.,.

REACTOR COOLANT SYSTEM PRESSURE -

2.1.3 The reactor coolant system pres'sure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: , OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION:

Wit'h the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least NOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.4.

LA SALLE - UNIT 1 2-1 Amendment No. .

' ~

2.1 SAFETY LIMITS BASES * -

2. 0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transsients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated

'. to occur if the limit is not violated. Because fuel damage is not directly observable. a step-back apjtenach is used to establish a Safety Limit such that l /,079he MCPR is not less tharT.lANC MCPR greater than 1A6' for two recirculation I.of- loop operation and L67 for single recirculation loop operation represents a conservative margin ' relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or l cracking. Although some corrosion or use related cracking may occur during the life o_f the cladding, fission product migration from this source is incre-mentally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signif-icant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power

. calculations at pressures below 785 psig or core flows less than 10% of rated -,

flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will alyays be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 10 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. 3Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 lbs/hr. Full scale ATLAS test data taken at pres-sures from 14.7 psia to 800 psia indicate that the fuel asserbly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.

LA SALLE - UNIT 1 8 2-1 Amendment No. -

F 1 I

Bases Table B2.1.2-1

. UNCERTAINTIES USED IN THE DETERMINATION .

OF kHE FUEL CLADDING SAFETY LIMIT

• Standard Deviation Quantity

(% of Point) 1.76 Feedwater Flow 0.76 Feedwater Temperature 0.5 Reactor Pressure 0.2 Core Inlet Temperature 2.5 Core Total Flow y7 Two recirculation Loop Operation 6.0 Single recirculation Loop Operation 3.0 Channel Flow Area 10.0 Friction Factor Multiplier Channel Friction Factor 5.0 Multiplier

,fv3" 8. 7 TIP Readings jg Two Recirculation Loop Operation 6.8 Single Recirculation Loop Operation R Factor M /. 4 ~.'

3.6 Critical Power

• The uncertainty analysis used to establish the-core wide- Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

The values herein appply to both two recirculation loop operation and single 18 recirculation loop operation, except as noted.

Amendment No.

B 2-4 LA SALLE - UNIT 1

r

~

8 Bases Table 82.1.2-2 NOMINAL VALUES OF PARAMETERS USED IN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT THERMAL POWER 4323 K,i 3293 MV Core Flow 100.5 M15/hr 102.5 MMy Dome Pressure 1010.4 psig -

-Ch = :1 E!:w Ar::

0.1000 ft 2 R-Factor '!!gh enrid.c.ent 1.043 Madisc. earict ut - 1.033 L;w anrict...ni - i . 030 1.033 - O GWD/t _

l. o s i - ? Gwo/e

1. 030 - 15 Gwo/+

1.033 - 20 GWD/t Y

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Bases Table 82.1.2-3 .

.4 RELATIV! BUNDLE POWER DISTRIBUTION

~

USED IN THE GETAB STATISTICAL ANALYSIS

' Percent of Fuel Bundles Within Rance of Relative Bundle Power Power Interval

^ 1. 525 iv 1.;7: M

. "2.47; wu 1.325 -e 1.42 w 1.475 .!y).

1.375 to 1.425 t 5. I 1.325 to 1.375 .ha- 7. 3 1.275 to 1.325 .a-3 7,7 .

1.225 to 1.275 && 9.1 1.175 to 1.225 4-17.3 1.125 to 1.175 . sea.11 8 _

1.075 to 1.125 he 4.7 1.025 to 1.075 . 1-e- 4/1 -

f,1.025 -

-Se-3-ql,5 1F6 5 P(  :

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Bases Table 92.1.2-4 R-FACTOR DISTRIBUTION USED IN GETAB STATISTICAL ANALYSIS 8x8 Red Array R-Factor

-Migh- . Meet" - .l.ow-.

Ene4shment- Er-ic Ment E-icMer,t Red'Secuence No.

1.043 -1.039 1.037 1.030- 1

-h0+3 - 1. 030 1.0 3 F 1.030 2

-kO42- h028- l.03 '; -h030- 3 h042- M28-1,037 6 030- 4

-bO38- 1.027-1,035 -h028- 5 1.038 -h027-l.035 4r028- 6 4-026 h026- 1.030 -bO24 7 11.024 +

_ h02&$ l.030 11. 020 - 8 through 64 f

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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. 2,1 -2

(

3.2.1 All AVERAGE PLANA LINEAR HEAT GENERATION RATESi (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1. The Ifmits of Figures 3.2.1-1"sha11 be reduced to a value of 0.85 times ~the two recirculation loop operation limit when in yy single recirculation loop operation.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or ~

equal to 25% of RATED THERMAL POWER.

4 ACTION:

and. 3,2,l- 2 WithanAPLHGRexceedingthelimitsofFigure3.2.1-1finitiatecorrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

(-

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits =

determined from Figures 3.2.1-1/ and 3,2,l-2 t

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

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

, c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is

) op'erating with'a LIMITING CONTROL ROD PATTERN for APLHGR.

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

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AVERAGE PLANAR EXPOSURE (mwd /t)

MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE' (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE INITIAL CORE FUEL TYPES BGM&3. BsRet3 ANO BEWH 8CR61'7h 8Clit.S 2.39 c M SC R.6c7/

Figure 3.2. -l

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MAP]GR V8. Average anar hposure P

g 7ue "yle BP8CRB299L M

G -13.00-Ei tt 12.50-12.00- '

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BP8CRB299L 10.00-9.50 i 1 ) ,

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Average Planar Exposure (mwd /t) -

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FIGURE 3.2.1-2

I -

POWER DISTRIBUTION LIMITS .

POWER DISTRIBUTION -LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO

} LIMITING CONDITION FOR OPERATION

' 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater

- than the MCPR limit determined from Figure 3.2.3-1 times the Kf determined

' - from Figure 3.2.3-2 for two recirculation loop operation and shall be equal

~

to or greater than the MCPR limit determined from Figure 3.2.3-1 + 0.01 times the K determined.from Figure 3.2.3-2 for single recirculation loop operation.

7 1 pr:vided th:t th: :nd-of-cyc4e-rec 4rsulation- pe=p trip (EOC ""T) systee is

--6PERABL2 per Specificetien 3.3.4.2.

APPLICABILITY: ,

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

18

~ ACTION '

. With th: :nd :f-:y:1 recirce!atica "" p t-4; sys + == i aar= *=h' = aa" Sp;;ificat4en 3.3.4.2, Operat4en-eay-continue-and-the-provisions-of Specif4 cat 4cn-3r0 4-are-not-applicable-provided-thatc*ithi- I heer, MCaa is detecained te be equ:1 to er gr;;t r th:n the MCP" 'irit'ather-4- ri;t-- 3.2.?-1 EOC 887 4-aperd'= _;ury , tirer th- y the"a

( Figere 3.2.0-2. o- .

-b With MCP2 less than the applicable MCPR limit determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS

' 4.2.3 MCPR, with:

1

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

for the cycle in accordance with Specification 4.1.3.2, or

b. I,y,, determined wi' thin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from Figures'3.2.3-1 and 3.2.3-2:

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

. ~

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and '

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

4 3/4 2-4 Amendment No.

i LA SALLE - UNIT 1

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.738.74 75 78 .77 78 .78 20 21 At .83 34 AS 28 Figure 3.2.3-1 MINIMUM CRITICAL POWER RATIO (MCPR) VER8US v AT RATED Fi.OW e

_ _ . _ _ _ _ __ _. _ _ _ _ _ _-- - _ _ - - - _ - _ _ - _ _ - _ - - _ _ _ . _ - - - . - - - - - - - _ . - _ _ . _ _ _ _ _ _ _ - - _ _ - - _-._-_.__---A

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Manual F1m' Contml Kr Curve #

1.00 0 90 20 30 ho 50 60 70 80 100 90

, Core Flow, % of Rated Core Flov -

t Kf Factor i Figure 3.2 3-2

~

INSTRUMENTATION END-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 OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table TIME 3.3.4.2-2 as shownand with'the in Table END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE 3.3.4.2-3.

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

ACTION:

f

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

1

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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. l$

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 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 <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. hg

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

d. With one trip system inoper'able, re' store the inoperable trip system -

to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or-tak: th; ACTION r: efr:d b S;n f fic:ti ;; 3.2.3. veduce. THERMAL PiWER -fo ["esslW

%n 30% 6 RRTED THERMAL POWER witMn the nex & hours,

e. With both trip systems inoperable, restore at least one trip system to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or t h: th: ACTION r: cir:d b" S;;;f fi::ti r. 3.2.3. Yedaee. THERMAL SWER -fo'less 9 hn 30% of RATED THERMAL. POWER wWn & nex4 6 hows, LA SALLE - UNIT 1 3/4 3-39 Amendment No.

- - -. -n, - - , . . . , , , ,

i l

3/4.4 REACTOR COOLANT SYSTEM  !

. . 1 3/4.4.1 RECIRCULATION SYSTEM -

-RECIRCULATION LOOPS

' LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation.

. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one reactor coolant system recirculation loop not in operation:

1. -Within 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />s:

a) Place the recirculation flow control system in the Master Manual mode, and b) Reduce THERMAL POL'ER te j[ 50% cf RATED THERM'L P0h'ER, and,-

2h jdF Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to)J.08.-trT per Specification 2.1.2, and, c) d7 Increase the MINIMUM CRITICAL P.0WER RATIO (MCPR) Limiting Condition for Operation by 0.01 per Specification 3.2.3, and, el)py Reduce the MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) limit to a value of 0.85 times the two recirculation loop operation limit per Specification 3.2.1, and, e.) ;F7 Reduce the Average Power Range Monitor (APRM) Scram and ~" l ,

Rod 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.

. At least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />s:

.a) Ver hat the APRM flux noise aver over 30 minutes does not SK peak to pe nrise, reduce the recirculation 1 the APRM flux noise is less than the 5% peak ea it, and, b) Ver' at .ttw core-plate AP noise et exceed 1 psi eak to peak; otherwise, reduce the recircu ~

loop flow until the AP noise is less than the 1 psi lim .

• See Special Test Exception 3.10.4.

LA SALLE - UNIT 1 3/44-1 Amendment No.

REACTOR COOLANT SYSTEM -1 LIMITING CONDITION FOR OPERATION (Continued) l w l ACTION: (Continued) i j!.)r! The provisions of Specification 3.0.4 are not applicable.

21s4 f Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

~

b. With no reactor coolant system recirculation loops in operation.

immediately initiate measures to place the unit in at least HOT a

SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.1.1 Each reactor coolant system recirculation loop flow control valve shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying that the control valve fails "as is" cn loss of hydraulic pressure at the hydraulic power units, and

b. Verifying that the average rate of control valve movement is:

s

1. Less than or equal to 11% of stroke per second opening, and

2. Less than or equal to 11% of stroke per second closing.

LA SALLE - UNIT 1 3/4 4-la Amendment No. f :

l 3/4.4 Reactor Coolant System

.. Thermal Hydraulic Stability 3'.4.1.5 APRM noise levels shall not exceed their established baseline noise levels by a factor of three.

APPLICABILITY: OPERATIONAL CONDITION 1, when operating in the Surveillance Region of Figure 3.4.1.5-1.

ACTION: With APRM noise levels three (3) times greater than their

. Surveillance Region baseline level:

a. Initiate corrective action within 15 minutes and restore the noise level to within the three-times-baseline limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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

SURVEILLANCE REQUIREMENTS 4.4.1.5.1 When operating in the Surveillance Region of Figure 3.4.1.5-1, verify that the APRM noise levels are less than or equal to three (3) times the baseline levels:*

a. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of entering the Surveillance Region,

b. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and

c. Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> af ter completion of a THERMAL POWER change of at least % of RATED THERMAL POWER.

d

• Baseline levels shall be established within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of entering the Surveillance Region if baseline levels have not been established since the last refueling.

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, I 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 d_ ring th; ... T_.! cy;?: cr'y, provided the unit is operated in accordance with the single recirculation loop operation Technical Specifications herein.

An inoperable jet pump is not,-in itself, a sufficient reason to declare a recirculation loop incperable, but it does present a hazard in c'ase 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 oump performance on a prescribed scheduled for significant degradation.

Recirculation loop flow mismi;ch limits are in compliance with the ECCS LOCA analysis design criterion. The limits will ensure an adequate core flow i coastdown from either recirculation loop following a LOCA. Where the recircu-lation loop flow mismatch limits can not be maintained during the recirculation gg ,

loop operation, continued operation is permitted in the single recirculation loop operation mode.

~

I In order to prevent undue stress on the vessel nozzles and bottom head 2

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 4

50*F.of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant i 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. ,,,g -P 7 C

4.Bodm page 3/4.4.2 SAFETY / RELIEF VALVES .

The safety valve function of the safety-relief valves operate to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 18 OPERABLE safety /

relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.

Demonstration of the safety-relief valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.

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

y e3 .. ..**= = = =a=e o e m .**E m - .m .----am.** - . ,

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.Thf pos;ibility cf thirmal hydraulic inattbility'in a BWR hrs b2sn investigated since the startup of early BWRs. Based on tests and analytical

~

models, it has been identified that the high power-low flow corner of the power-to-flow map is the region of least stability margin. This region maybe encountered during startups, shutdowns, sequence exchanges, and as a result of a recirculation pump (s) trip eyent. To ensure stability, a criteria has been placed on APRM noise levels (shall not exceed baseline values by a factor of

3) when in the region of possible instability.

m9 e

DOCUMENT 1817r

r ATTACHMENT C SIGNIFICnNT HAZARDS CONSIDERATION Commonwealth Edison proposes to amend Operating License tPF-ll and reload the LaSalle 1 reactor core in preparation for Cycle 2 operation. The proposed Technical Specification changes do not represent significant changes in acceptance criteria or safety margins and all changes have been made based on methods that have been previously accepted by the NRC. The reload core involves a new fuel type which must be licensed. The new fuel type has been analyzed with approved methods and meets the approved limits of GESTAR. The new fuel type presents no unreviewed safety questions because the bundle design has been approved by the NRC, and licensing of new bundle enrichments has been treated as a non-safety related change to GESTAR.

Under the provisions of 10 CFR 50.92 this means that the proposed amendment will not (1) involve a significant increase in the probability or occurrence of an accident previously evaluated; or (2) create the possibility for a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in the margin of safety.

The Technical Specification changes are proposed to incorporate the following:

For Cycle 2, the MCPR fuel cladding integrity safety limit was changed from 1.06 to 1.07 for two recirculation loop operation, and from 1.07 to 1.08 for single recircul'. tion loop operation. The safety limit is smaller for initial cores because the uncertainties in TIP readings and the R Factor are smaller.

The addition of a new MAPLHGR vs Exposure curve for the reload fuel type BP8CRB299L in the APLHGR Technical Specification. The MAPLHGR limits were provided by General Electric in the Supplemental Reload Licensing Submittal for LIC2.

The deletion of the EOC-RPT inoperable provision in the MCPR and EOC Recirculation Pump Trip System Technical Specifications. The EOC-RPT inoperable analysis was not justified for in the second cycle but may be included in future cycles.

The replacement of the existing MCPR curve with a revised curve which reflects the limiting transients for cycle 2. The MCPR limits were provided by General Electric in the Supplemental Reload Licensing Submittal for LIC2.

The replacement of the existing Kr curve with a revised curve which is based on LaSalle's rated core power and core flow. The original curve was a generic curve.

r The addition of a Thermal Hydraulic Stability Technical Specification to address the NRC concerns in this area. With the addition of this new -

specification, the single loop stability requirements in the Recirculation System Technical Specification have been deleted..

_ Based on the preceeding discussion, it is concluded that the consequences-of previously evaluated accidents will not be increased and the margin of safety will not be decreased by the proposed LIC2 reload and associated

-Technical Specification changes. Therefore, based on the guidance provided in the Federal Register and the criteria established in 10 CFR 50.92(c), the proposed changes do not constitute a significant hazards consideration.

l 0797K-