Text

Proposed Tech Specs Re Core Operating Limits Rept,Control Rod Assemblies,Mcpr & LHGR
	ML20246E354
Person / Time
Site:	LaSalle
Issue date:	08/18/1989
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML20246E337	List: ML20246E332; ML20246E354;
References
	NUDOCS 8908290099
	Download: ML20246E354 (104)
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e AIIAC11MENLZ PROPOSED TECHNICAL SPECIFICATION CHANGES A. LIST AND DESCRIPTION OF CHANGES B. LASALLE COUNTY STATION UNIT 1 (NPF-11) PAGES C. LASALLE COUNTY STATION UNIT 2 (NPF-18) PAGES 0189T:8 8908290099 999g3g

'DR ADOCK 05000373 PDC

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- l AIIaCHMENL2 l ,

i A. LISL6NQJESfRIPll0N OF TECHtUCAL SPECIELCA110H_CEARGES The following section describes all areas requiring revisions and identifies the affected pages and sections of the Technical Specifications for each unit: 1 Unit i Technical Specifications EAGE._ROJS1 .0iSERIPl10R_E_fEARGES I Added new definition index entry for CORE OPERATING LIMITS Rf. PORT, renumbered all definition index entries afterwards.

II Renumbered all definition index entries.

XIX Deleted index entries for Figures 3.2.1-1, 3.2.1-2, 3.2.1-3,'3.2.3-la, 3.2.3-lb, and 3.2.3-2.

1-2 Added definition for CORE OPERATING LIMITS REPORT, renumbered all definitions afterwards.

1-3 to 1-7 Renumber all definitions.

3/4 2-1 In Sections 3.2.1 and 4.2.1; incorporated reference to CORE OPERATING LIMI15 REPORT in place of " Figures 3.2.1-1, 3.2.1-2 and 3.2.1-3" (MAPLHGR limits).

3/4 2-2 Deleted Figure 3.2.1-1 (MAPLHGR).

3/4 2-2a Deleted Figure 3.2.1-2 (MAPLHGR).

3/4 2-2b Deleted Figure 3.2.1-3 (MAPLHGR).

3/4 2-4 In Section 3.2.3; incorporated reference to CORE OPERATING LIMITS REPORT in place of MCPR limit explanations.

3/4 2-4a In Section 4.2.3; incorporated reference to CORE OPERATING LIMITS REPORT in place of " Figures 3.2.3 and 3.2.3-2" (MAPLHGR limits).

3/4 2-5 Deleted Figure 3.2.3-la (MCPR).

3/4 2-Sa Deleted Figure 3.2.3-lb (MCPR).

3/4 2-6 Deleted Figure 3.2.3-2 (MCPR Kr factor).

3/4 2-7 In Section 3.2.4; incorporated reference to CORE OPERATING LIMITS REPORT in ;. lace of LHGR values.

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' Unit l! Technical Specifications (cont'd) l PAGE NO.(S)- DESCRIPTION OF CHANGES 3/4 3 In Table 3.3.6-2; incorporated reference to CORE OPERATING

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(- LIMITS REPORT in place of Rod Block Monitor Upscale Setpoint relationships.

B 3/4 2-1 In Bases 3/4.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE", deleted last two paragraphs which discuss how APLHGR values were calculated for each existing fuel type;

- . Inserted statement; "The calculational procedure used to establish APLHGR values uses.GE calculational models which are approved and consistent with'the requirements of Appendix K 10 CFR 50".

B 3/4 2-3 In Bases 3/4.2.3, " MINIMUM' CRITICAL POWER RATIO",

incorporated reference to CORE OPERATING LIMITS REPORT in place of "Figu*e 3.2.3-la". Also deleted reference to two MCFR curves when Rod Hithdrawal Error is~the limiting transient.

B 3/4 2-3a In Bases 3/4.2.3, " MINIMUM CRITICAL POWER RATIO", remove the statements:

"The cycle independent MCPR LCO values shown in Figure 3.2.3-lb for the main turbine bypass and recirculation pump trip systems out of service are valid provided:

1) The cycle specific analysis for the Load Reject Hithout Bypass and Turbine Trip Hithout Bypass events yield MCPR LC0 values less than or equal to 1.33 and 1.29 for Options A and B, respectively.

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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."

Replace with the statement:

"The MCPR LCO values shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and j recirculation pump trip systems out of service are valid provided that these limits bound the cycle i specific results." <

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Unit 1 Technical Specifications (cont'd)

EAGEJQM1 - DESCRIPTION OF-CIMNGES' L

-B 3/4 2-5 In Bases 3/4.2.3, deleted reference to Kf factor.

5-4' In 5.3.1, " FUEL ASSEMBLIES", the definition has been changed-to read, "The reactor core shall contain 764 fuel assemblies. Each assembly consists of a matrix of Zircalloy clad fuel rods with-an initial composition of.

slightly. enriched uranium dioxide, U02. Fuel assemblies shall.be limited to those fuel designs approved for use in BHR's."'

5-4 In'5.3.2, " CONTROL R0D ASSEMBLIES", the definition has been changed to read,."The reactor core shall contain 185 cruciform shaped control rod assemblies. .The control' material shall be boron carbide powder (B4 C), and/or Hafnium metal. The control rod assembly shall have.a nominal axial absorber length'of 143 inches."

6-24. In the Administrative Controls, section 6.6, Reporting Requirements, add a new section after 6.6.5, entitled

" CORE OPERATING LIMITS REPORT".

Unit 2 Technical Specifications PAGE NO.(S) DESCRIPTION OF CHANGES I Added new definition index entry for CORE OPERATING LIMITS REPORT, renumbered all definition index entries afterwards.

II Renumbered all definition index entries.

XIX Deleted index entries'for figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.3-la, 3.2.3-1b, and 3.2.3-2.

1-2 Added definition for CORE OPERATING LIMITS REPORT, renumbered all definitions afterwards.

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4 Unit 2 Technical Specifications PAGE K3.iS1 DESCRIPTION OF CHANGES 1-3 to 1-7 Renumber all definitions.

3/4 2-1 In Sections 3.2.1 and 4.2.1; incorporated reference to CORE OPERATING LIMITS REPORT in place of " Figures 3.2.1-1, 3.2.1-2 and 3.2.1-3" (MAPLHGR limits).

3/4 2-2 Deleted Figure 3.2.1-1 (MAPLHGR).

3/4 2-2a Deleted Figure 3.2.1-2 (MAPLHGR).

3/4 2-2b Deleted Figure 3.2.1-3 (MAPLHGR).

3/4 2-4 In Section 3.2.3; incorporated reference to CORE OPERATING LIMITS REPORT in place of MCPR limit explanations.

3/4 2-4a In section 4.2.3; incorporated reference to CORE OPERATING LIMITS REPORT in place of " Figures 3.2.3 and 3.2.3-2" (MCPR limits).

1 3/4 2-5 Deleted Figure 3.2.3-la (MCPR).

3/4 2-Sa Deleted Figure 3.2.3-1b (MCPR).

3/4 2-6 Deleted Figure 3.2.3-2 (MCPR Kr factor).

3/4 2-7 In section 3.2.4; incorporated reference to CORE OPERATING LIMITS REPORT in place of LHGR values.

3/4 3-53 In Table 3.3.6-2; incorporated refarence to CORE OPERATING LIMITS REPORT in place of Rod Block Monitor Upscale Setpoint relationships.

B 3/4 2-1 In Bases 3/4.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE", deleted last two paragraphs which discuss how APLHGR values were calculated for each existing fuel type.

I Inserted statement; " The calculational procedure used to l' establish APLHGR values uses GE calculational models which are approved and consistent with the requirements of Appendix K, 10 CFR 50" 0189T:12 1

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( - SI-Unit 2 Technical Specifications (cont'd)

PAGE NO.(S)- DESCRIPTION OF CHANGES I

B 3/4 2-3 In Bases 3/4.2.3, " MINIMUM CRITICAL POWER RATIO",

incorporated reference to CORE OPERATING LIMITS REPORT in place of " Figure 3.2.3-la". Also deleted reference to two MCPR curves when Rod Withdrawal Error is the limiting transient.

B 3/4 2-4 In Bases 3/4.2.3, " MINIMUM CRITICAL POWER RATIO", remove the. statements:

"The cycle independent MCPR LCO values shown in Figure 3.2.3-lb for the main turbine bypass and recirculation pump trip systems out of service are valid provided:

1) The cycle specific analysis for the Load Reject Hithout Bypass and Turbine Trip Hithout 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."

Replace with the statement:

"The MCPR LCO values shown in the CORE OPERATING LIMITS REPORT for the main turbine bypass and recirculation pump trip systems out of service are valid provided that these limits bound the cycle specific results."

B 3/4 2-6 In Bases 3/4.2.3, deleted reference to Kf factor.

5-4 In 5.3.1, " FUEL ASSEMBLIES", the definition has been changed to read, "The reactor core shall contain 764 fuel assemblies. Each assembly consists of a matrix of Zircalloy clad fuel rods with an initial composition of clightly enriched uranium dioxide U02 . Fuel assemblies shall be limited to those fuel desigr:s approved for use in BWR's."

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,c, Unit 2 Technical Specifications (cont'd)

PAGE NO.(S1 DESCRIPTION OF CHANGES i

5-4 In 5.3.2, " CONTROL ROD ASSEMBLIES", the definition has been changed to read, "The reactor core shall contain 185 cruciform shaped control rod assemblies. The control material shall be boron carbide powder (B4 C), and/or Hafnium metal. The control rod assembly shall have a-nominal axial absorter length of 143 inches."'

6-24 In the Administrative Controls, section 6.6, Report Requirements, add a new section after 6.6.5, entitled

" CORE OPERATING LIMITS REPORT".

0189T:14

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INDEX Qy -

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..s. DEFINITIONS L.. SECTION PAGE j 1.0' DEFINITIONS 1 ~

1-1 j 1.1 ACTI0N.'...........................................................

4 1-1 1- i .1. 2 AVERAGE PLANAR EXP05URE...........................................

1 1.3': AVERAGE PLANAR LINEAR HEAT GENERATION RATE........................'

'1-1 1

4 1.4 CHANNEL CALIBRATION..........................'..................... 1-1

~t

. 1. 5 CHAN.NEL CHECK..................................................... 1-1 4

l 1. 6 CHANNEL FUNCTIONAL' TEST........................................... 1-1

' l-7.

[M 1.7- CORE ALTERATION.......................................

/. g CORE ofEM7tAIG l. writs ret %f7~. . ... . .

l J' l 1.fyCRITICALPOWERRATI0.............................................. 1-2 1-2 4

1.fDOSEEQUIVALENTI-131.............................................

1-2 1.hE-AVERAGEDISINTEGRATIONENERGY...................................

I i -'(2~~ 1. yf ' EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME. . . . . . . . . . . . . . . .

, s.

1-2 1 1. ENO-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME......... 1-2 W

I 1-3

1. 4 FRACTION OF LIMITING POWER 0ENSITY................................

! i. fr 1-3

! i 1. )4 FRACTION OF RATED THERMAL P0WER. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

' -; is 1-3

$ 1. */ FREQUENCY N0TATICN................................................

1-3

.'. 1./a GASEOUS RADWASTE SYSTEM................................. TREATMENT

1. IDENTIFIED LEAKAGE.......................'......................... 1-3 .

1.hISOLATIONSYSTEMRESPONSETIME.....................,.............. 1-3 u

1-3

[; 1. g LIMITING CONTROL RCD P ATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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1-4 1.)6 LINEAR HEAT GENERATION RATE.......................................

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1.pl LOGIC SYSTEM FUNCTIONAL TEST........ ............................. 1-4 l

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1. MAXIMUM FR ACTICM OF '. MITING POWER 0ENSITY. . . . . . . . . . . . . . . . . . . . . . . .. 1-4 j . ,

i . a 1.yIMINIMUMCRITICAL.P0wERRATI0...................................... 1-4

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('j )" 1. M OFFS ITE DOS E CALCULATION MANUAL. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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SECTION PAGE h g5E# . DEFINITIONS (Continued)

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@ '1.8OPERATIONALCONDITION-CONDITION.................................

'l 1 I 1.kPHYSICSTESTS......................................................

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[ 1.)K PRESSURE BOUNDARY LEAKAGE. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-5 b 1.kPRIMARYCONTAINMENTINTEGRITY.....................................

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1-5 4

1. X PROCESS CONTROL PR0 GRAM...........................................

sn- 1-5 i 1. M PURGE - PURGING...................................................

) p 1-5' 3 1.)ERATEDTHERMALP0WER...............................................

t vt 1 j 1. )$ REACTOR PROTECTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . .

1 sf 1-6 l-j . h3 1. )4 REPO RT ABLE EVENT. . . . . . . . . . . . . . . . . . . . . . . . . . . . J. . . . . . . . . . . . . . . . . . . . .

? .% 1 j 1. % ROD DENSITY.......................................................

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1-6 i! 1. 4 SECONDARY CONTAINMENT INTEGRITY...................................

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1-6 y

1. M SHUTDOWN MARGIN...................................................

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1-6 l- 1. 5 SOLIDIFICATION.................................................,..

1 e 1-7 j -

1. K SOURCE CHECK......................................................

y' *. 1-7 Lj -1. % STAGGERED TEST BASIS..............................................

9- p 1-7 4 II 1. 6 THERMAL P0WER..................................................... -

r s 1-7 TIME......................................

e 1. % TURBINE BYPASS RESPONSE n 1-7 lli fJ 1. PJ UNIDENTIFIED LEAKAGE. . . . . . . . . . . . . . -. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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% SYSTEM.............................. 1-7

l 1.44 VENTILATION EXHAUST TREATMENT .

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{ VEliTING....................................

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Amendment No. 23 l

II 9 LA SALLE - UNIT 1 .

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INDEX 3: A-LIST OF FIGURES

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FIGURE PAGE

.e l- 3.1.5-1 SODIUM PENTABORATE SOLUTION TEMPERATURE /

3 il -

CONCENTRATION REQUIREMENTS ........................ 3/4 1-21 h 3.1.5-2 SODIUM PENTABORATE (Na2 B ic0 1 s 10 H2O) l VOLUME / CONCENTRATION REQUIREMENTS ................. 3/4 1-22 4

1 .1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION l1 - RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, J- INITIAL CORE FUEL TYPES 8CRB176, BCRB219, and j

8CRB071 ........................... ............. . 3/4 2-2 1 3.2.1-2 MA AVERAGE PLANAR LINEA A GENERA" j RATE ( HGR) VERSUS AVE NAR SURE, FUEL TYPE 3/4 2-2a 4 RB299L / ..... ................

3.2.1-3 MAXIMUM AVERAGE @ LDf HEAT GENERATION l- RATE (MAP R1 VEW AGE PLANAR EXPOSURE, 1 FUEL TYPg BCp BC3 ...................... 3/4 2-2b i 3.2.3-la MINI L POWER RATIO (MCP VERSUS

[1 t AT : FLOW ....................... ........-... 3/4 2-5 l h~~

fi 3.2.3-lb- , MUM CRITICAL POWER RATIO (MCPR) VERSUS l t AT RATED FLOW FOR END OF CYCLE RECIRCULATION P TRIP AND MAIN TURBINE BYPASS SYSTEMS INOPERABLE ... /4 2-5a 1

i- 3-2 Kf FACTOR ......................................... 3/4 4

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) 3.4.1.5-1 CORE THERMAL POWER (% OF RATED).VERSUS TOTAL 1 CORE FLOW (% OF RATED) 3/4 4-4c j '

/ 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE li. VS. REACTOR VESSEL PRESSURE ....................... 3/4 4-18 1.

j 4.7-1 SAMPLE PLAN 2)' FOR SNUBBER FUNCTIONAL TEST ........ 3/4 7-32 j: B 3/4 3-1 REACTOR VESSEL WATER LEVEL ........................ B 3/4 3-7

+

B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T )

1- AS A FUNCTION OF SERVICE LIFE ..................... C 3/4 4-7 1

$ B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS . . . . . . . . . . . . . .. . . . . B 3/4 6-3a

.i t 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS j, AND LIQUID EFFLUENTS .............................. 5-2

.( 5.1.2-1 LOW POPULATION ZONE ....,.......................... 5-3' 1 , . .

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> LA SALLE - UNIT 1 XIX Amendment No. 60 ;

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DEFINITIONS lc> -

v CORE ALTERATION >

F 1.7 CORE ALTERATION shall' be the addition, removal', relocation or movement of i: /' fuel, sources, incere instruments or reactivity contrcls within the

) reactor pressure vessel with the vessel head removed and fuel in the

vessel. Suspension of CORE ALTERATIONS shall not preclude completiot, of

_ the movement of a component to a safe conservative position.

t h CRITICAL POWER RATIO

_ MD I 1.gf The CRITICAL POWER RATIO (CPR) shall be the ratio of that power'in the assembly which is calculated by application of the GEXL correlation to l

4 .

[ cause some point in the assemely to experience boiling transition, divided by the actual assemoly operating power.

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L DOSE EQUIVALENT I-131 *

e. 1.f,DOSEEQUIVALENTI-131shallbethatconcentrationofI-131, [

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' microcuries/ gram, which alone would procuce the same thyroid-dose as the --.

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quantity and isotopic mixture of I-131,1-132, I-133,1-134, and I-135 e actually present. The thyroid dose conversion factors used for this-calculation shall be those listed in Table III of TID 'I4844, " Calculation of Distance Factors for Power and Test Reactor Sites."

2 T-AVERAGE DISINTEGRATION ENERGY L

'(h. 1.14 I shall be the average, weighted in proportion to the concentration of {

j v ' 8' each radionuclides in the reactor coolant at the time of sampling, of the

- sum of the average beta and gamma energies per disintegration, in MeV, 7 for isotopes, with half lives greater than 15 minutes, making up- at least.

f 95% of the total non-iodine activity in the coolant.

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EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME.

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$ ! 1.)K The EMERGENCY

> interval from when the monitored CCRE COOLING SYSTEM parameter exceeds (ECCS) its ECCS RESPCNSE TIME s actuation i 1 -

setpoint at the channel sensor until the ECCS equipment is capable of .

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'jt performing its safety function, i.e., the valves travel to their required por*tions, pump discharge pressurts reach their required values, etc..

i Times shall include diesel generator starting and sequence leading delays k .

1 j where applicable. The response time may be measured by any series of a

sequential, overlapping or total steps such that the entire response time

[ ] is measured.

l- [i END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME L,1( The END-CF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPCNSE TIME shall be l

L 2 O that time interval to energization of the recirculation pump circuit breaker trip coil from when the monitored parameter exceeds its trip

k. setpoint at the channel sensor of the associated:

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$ I' a. Turbine stop valves, and t i Turbine control valves.

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The res;:ensa t' n :y be measured by any serie, of secuential, ovaricpping

[ or total steps eWn tnat the entire response time is measured.

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l CORE OPERATING LIMITS REPORT

!. 1. 8 The CORE OPERATING LIMITS REPORT is the unit-specific document that

!- provides core operating limits for the current operating reload cycle.

! These cycle-specific core operating limits shall be determined fot' each l reload cycle in accordance with Specification 6.6.A.G. Plant operation-within these operating-limits is addressed in individual specifications.-

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q DEFINITIONS I,e -

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FRACTIONOFLIMITINGPOWERDENS,IE 3L 1./The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR existing l 8- if at a given location divided by the specified LHGR limit for that bundle type. .

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)y FRACTION OF RATED THERMAL POWER ,

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[ 1./6The FRACTION OF RATED THERMAL' POWER (FRTP) shall be the measure POWER diviced by the RATED THERMAL POWER.

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a; FREQUENCY NOTATION 1.IThe. FREQUENCY NOTATION specified for the performance of Surveillance (~

- 16 Requirements shall correspond to the intervals defined in Table 1.1. .

'- GASEOUS RA0 WASTE TREATMENT SYSTEM A GASEOUS RA0VASTE TREATMENT SYSTEM shall be any system designed and 1.[d installed to reduce radioactive gaseous affluents by collecting primary . l-N-}

1;

- coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to

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release to the environment. ,

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p IDENTIFIED LEAKAGE

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l-i 1./10ENTIFIEDLEAKAGEshallbe:

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.i 1 a. Leakage into ec11ection systems, such as pump seal er valve -

i packing leaks, that is captured and conducted to a sump or

! collecting tank, or

.,' b. Leakage into the containment atmosphere from sources that are jt both specifically located and known either not to interfere with the operation of the leakage detection systems or not to L.

4- be PRESSURE EOUNDARY LEAKAGE.

1 ISOLATION SYSTEM RESPONSE TIME

[ I The ISOLATION SYSTEM RESPONSE TIME shall.be that time interval from when the 3

7 1.)6 monitored parameter exceeds its isolation actuation setpoint at the channel 5

sensor until the isolation valves travel to their_ recuired. positions.- fimes

' shall include diesel generator starting and sequence leading delays where -

i applicable. The response time may be measured by any series of sequential,.

overlapping or total steps such that the entire response time is measured.

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/. LIMITING CCNTROL RCD PATTERN l

1./A LIMITING CONTROL RCD PATTERN shall be a pattern which results in the y core being on a thermal hycraulic limit. i.e.. operating on a limiting

]) - value fer AFLHJR. LEGR. cr MOM.

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LA sat.LE - UNIT 1 1-3

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' DEFINITIONS I C ,

LINEAR HEAT GENERATION RATE 1.pf LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit l d length of fuel rod. Itlls the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL TEST A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic components, l 1.gvi.e.,

t all relays and contacts, all trip units, solid state logie elements, etc. of a logic circuit, from sensor through and including the actuated device to verify OPERABILITY. 1dE LOGIC SYSTEM FUNCTIONAL TEST may be performed by any series of sequential, overlapping or total systein steps i

l- sucn that the entire logic system is tested.

MAXIMUM FRACTION OF L!'ilTING POWER OENSITY 1.gTheMAXIMUMFRACTIONOFLIMITINGPOVERDENSITY(MFLPD)shallbethehighes -

0 value of the FLP0 which exists in the core.

MINIMUM CRITICAL POWER RATIO 1.pr The MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which l vl exist.s in the core. . -

OFFSITE 00SE CALCULATION MANUAL I

1.f# The OFFSITE DOSE CALCULATION MANUAL (00CM) shall contain the methodology I' j 14 and parameters used in the calculation of offsite cases due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid affluent monitoring alarm / trip setpoints.

OPERABLE - OPERABILITY 1.f4 A system, subsystem, train, component er device shall be OPERABLE cr have a OPERABILITY when it is capable of performing its specified function (s),

and when all necessary attendant instrumentation, controls, a normal and an. emergency electrical power source, cooling or seal water, lubrication !

or other auxiliary equipment that are required for the system, suosystem, train, component or device to perform its function (s) are also capable of

performing their related support function (s).

OPE %ATIONAL CONDITION - CONDITION 1.jg An OPERATIONAL CONDITION, i.e. , CONDITION, shall be any one inclusive el combination of mode switch position and average reactor coolant temperature as specified in Table 1.2.

PHYSICS TESTS-1.pIPHYSICSTESTSshallbethosetestsperformedtomeasurethefundamental 15 nuclear characteristics of the reactor core and related instrumentation and 1) described in Cha:::er la of the FSAR. O suth -i:ed under tse provisiens of 10 CFR 50.59. e- 31 atnerwise nor:.ec :) tte Co mission.

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. DEFINITIONS PRESSURE Bot!NOARY LEAKAGE 1.JFPRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolaole fault 4 in a reactor coolant. sy.tes component body, pipe wall or vessel yall. '-

PRIMARY CONTAllWENT INTEGRITY j

1.gPRIMARYCONTAINMENTINTEGRITYshalkexistwhen: '

N a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an CPERA8LE primary containment automatic: isolation systas, or ,

2. Closed by at least one manual valve, blind flange, or deactivated autcmatic valve secured in its closed position, except as provided in Tsble 3.6.3-1 of Specification, 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. Each primary containment air lock is OPERA 8LE pursuant to Specification 3.6.1.3.

d. The primary containment leakage rates are within the limits of Specification 3.6.1.2.

e. The suppression chamber is CPERABLE pursuant to Specification 3.6.2.1.

f. The sealing mechanism associated with each primary containment penetration; e.g., welds, bellows or 0-rings, is OPERA 8LE.

PROCESS CCNTROL PRCGRAM 1.)tf The PROCESS CCNTROL PROGRAM (PCP) shall contain the sampling, analysis, l JI and formulation determination by which SOLIDIFICATION of radioactive

wastes from liquid systems is assured.

j . PURGE - PURGING j 1.gPURGE or PURGING shall be the controlled process of discharging air or (

t 3> gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacam-j

, ont air or gas is required to purify the confinement.

RATED THERMAL POWER

.l l I 1.g RATED THERMAL POWER shall be a total reactor core tient transfer rate to

-I 53 the reactor coolant of 3323 MWT.

I REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval fres l 1.[3y when the monitored parameter exceeds its trip setpoint at the cha sensor until de-energi:ation of the scram pilot valve solenoids. The i

resporse time may ce :tasurea sy sny series s# s r: t . tis!. ovaria::in;~or total su:s suca t:at the entire response tica is zas;.reo.

o

'I Y.

LA SALLE - UNIT 1 1-5

1 I

. . I f

l 4;

{;)

DEFINITIONS 1

4 . . . .

REPORTABLE EVENT 1./AREPORTABLEEVENTshallbeanyofthoseconditionsspecifiedin l!

3 4 Section 50.73 to 10 CFR Part 50.

4 J ROD DENSITY 7

1.K36 ROD fraction DENSITY shall of the total be the of number number controlofrod control rod notches notches. inse led as a All roos fully 1 inserted is equivalent to 100% R00 DENSITY. *

[ SECONDARY CONTAINMENT INTEGRITY i l j 1.}8' SECONDARY CONTAINMENT INTEGRITY shall exist when:

d II a. All secondary containment penetrations required to be closed i during accident conditions are either:

1. Capable of being riosed by an OPERABLE secondary containment.

automatic isolation system, or 4

2. Closed by at least one manual valve, blind flange, or 1

3-deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of Specification 3.6.5.2.

Y

[ p""

b. All secondary containment hatches and blowout panels are closed 5 and sealed.

c. The standby gas treatment system is OPERABLE pursuant to 1

Specification 3.6.5.3.

d. At least one door in each access to the secondary containment is closed. '

l e. The sealing mechanism associated with each secondary containment penetration, e.g. , welds, bellows or 0-rings, is OPERABLE.

~

f. The pressure within the secondary ' containment is less than or equal to the value required by Specification 4.6.5.1.a.

1 SHUTDOWN MARGIN 17/ SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is

! % suberiticr.! or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown 1:" condition; cold, i.e. 68*F; and xenon free.

l ^

SOLIDIFICATION l

l 1.)df SOLIDIFICATION shall be the conversion of radioactive wastes from liquid p' 31 systems to a homogeneous (uniformly distributed), monolithic, immobili:ed

( solid with definite volume and shape, bounded by a. stable surf ace of

- distinct outline on all sides (free-standing).

t I

1-6 Amend ent No.23 LA SALLE UNIT 1 I l

j - ..

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IF - .

if DEFINITIONS p ' g.

o l SOURCE CHECK s1 I' ( A SOURCE CHECK shall be.*.he qualitative assessment of channel' .esponse -

j F 1.)# when the channel sensor'Is exposed to a radioactive source. ,

N

[, STAGGERED TEST BASIS ,

1. A STAGGERED TEST BASIS shall consist of: l r,; a. A test schedule for n systems, subsystems, trains or other g

designated components obtained by dividing the specified test 3-interval into n equal subintervals.

3:

i- b. The testing of one' system, subsystem, train or other designated

j. component at the beginning of each sucinterval.

I-I THERMAL POWER

[ 1.KTHERMALPOWERshallbethetotalreactorcoreheattransferratetothe .l

..j( F reactor coolant.

TURBINE BYPASS SYSTEM RESPONSE TIME

{.

hj ~ h)

[The TURBINE BYPASS SYSTEM RESPONSE TIME shall be time interval from when 1.)O the turbine bypass control unit generates a turbine bypass valve flow l

7-)

'.' signal until the turbine bypass valves travel to their required positions.

j- The response time may be measured by any series of sequential, overlapping j

) or total steps such that the entire response time is measured.

a ,

{

UNIDENTIFIED LEAKAGE l

f, 1.% UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

I+ ,

' VENTILATION EXHAUST TREATMENT SYSTEM 1.[dinstalled to reduce gaseous radiciodine or radioactive material in l.pa i A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and i ~

1 ! late form in effluents by passing ventilation or. vent exhaust gases j , thrcugh charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulate from the gaseous exhaust stream prior to the

1 release to the environment (such a system is not considered to have any.

i [ affect on noble gas affluents). Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT

[g 1

SYSTEM components.

I -

VENTING j t

b. VENTING sball be the controlled crocess of discharoine air or oas from a 1.[&confinecenttomaintaintemparar..e. pressure.!".7:iij, l l ! concentration or otner oce-ating concition, in suc, a anrer na . Wace ent air or gas is j !(g.y not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process.

.' 4 4

LA SALLE - UNIT 1 1-7

_ _____.-_____.-.m .___

F l

J l 3/4.2 ' POWER DISTRIBUTMN LIMITS 1 O

i i

1 3/4.2.1 . AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4

l

$ LIMITING CONDITION FOR OPERATION A

j i

3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits I l :h:m '- Ti;;;;; 2.2.1 1, 2.2.1 2 er.d 2.2.1 2. C APPLICABILITY: OPERATIONAL CONDITION 1,.when THERMAL POWER is greater than or 9;

.- equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits ' "i;;r:: 2. 2.1-1, 2. 2.1- 2 ;r.d 2. 2.1- 2,+ l' j' initiate corrective action within.15 minutes and restore APLHGR to within the 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 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 .

l' 7 57ed(1el M uc CORE CfEfAT)t4G LAM tTS REPoP_Tj SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits j tter " d 'r:: ri;;r:: 2. 2.1 '. , 2. 2.1- 2 ;r,i 2. 2.1- 2< = I

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

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

c. 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 with a LIMITING CONTROL ROD PATTERN for APLHGR.

l' 1

5 li I

(i, LA SALLE - UNIT 1 3/4 2-1 Amendment No. 58 l .

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a j .M.. POWER DISTRIBUTION LIMITS 3

] 3/4.2.3 MINIMUM CRITICAL POWER RATIO d LIMITING CONDITION FOR OPERATION I

1 p 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater.

g than the MCPR11mit d:tr '-M ' r-Q, crecht u ne_ cocr efew,NG UM5 CEtbRT hSingle Recirculation loop Operation BM re 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve B for of 110%) plus 0.01, times the kf determined from Figu .2.3-2.

setpo'

b. Two Recircui on Loop Operation 4

Figure 3.2.3-la rve A for a RBM setpoint of 106*' r Curve B for a RBM setpoint of 110%) t1 the kfdeterm from gure 3.2.3-2.

d c. Two Recirculation Leoo OceDatian /dntrTurbine Bypass Inocerable Figure 3.2.3-lb times the k e W eo from Figure 3.2.3-2, for two main turbine bypass system t

recirculation loop opernio th t

- inoperable per Specif ab( .7.10 (an BM setpoint determined per Specification Tat .

' 2 may be used).

l circulation Pump Two Recirculate Looo Operation with End-of-Cycle l d.

Trio Systen1Afioperaole j

(d] Figure 3-lb times the k with the end-of-cycle recircula for two re ' ulationloopoperatiob,determinedfromFigure3.2.3-a .

pump 4 ip system inoperable as directed by Specification 3.3.4.2 (any

" setpoint determined per Specification Table 3.3.6-2 may be used).

J .

APPLICABILITY: -

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to -

25% of RATED THERMAL P0WER. _

ACTION

a. With MCPR less tFan the applicable MCPR IImit as determined for one of the 1

alm

conditions /-5fec*ed * % Coff cPERAwG umrf 5 RQeSS Initiate corrective action within 15 minutes, and .

1. ;-

!I 2. Restore MCPR to within,the required limit within 2 hgurs.

2 Othemise, reduce THERMAL POWER to less th'an 25% of RATED THERMAL -

j 3. i.

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

9

/

i b. When operating in a conditica not ' Int"':d d:n, reduce THERMAL POWER 4 ~

4 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 . i a .

e a '

3/4 2-4 Amendment No. 58

' LA SALLE - UNIT 1

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jJ 4

i' .

[ .. POWER DISTRIBUTION LIMITS (Continued)

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l '" 3/4.2.3 MINIMUM CRITICAL POWER RATIO

[

j ' SURVEILLANCE' REQUIREMENTS 3

)- 4.2.3 MCPR, with:

l- a. t ave

'= 0.86 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. I,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 f'

j.. surveillance test required by Specification 4.1.3.2, shallbedeterminedtobeequaltDorgreater,thantheapplicableMCPRlimit5/'eeb9M

j. -dit.. m . n.d '; e..; " p :: 2. 2. 2-1 :-d 2. 2. 3-@ ihe @fE oPEMTM en hr 1

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

}

4- 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 j c. Init.iaily 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 oper-j ating with a LIMITING CONTROL ROD PATTERN for MCPR.

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POWER DISTRIBUTION LIMITS I 3/4.2.4 LINEAR HEAT GENERATION RATE -

g.

] - LIMITING CONDITION FOR OPERATION

(

a

c tou e/he l

-3.'2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not' exceed k He coK ..

t

a. 13.4 kw/ft' for fuel- typ . #[g%y ##W

~ L -A l 1. 8CRB173 I ~2. CM ~

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3. BP8%*J1,L el types:

b. 14.? /ft for 7

i . BC301A

[ 2. BC3208

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

9 ACTION:

?

j.

J Q With the LHGR of any fuel rod exceeding the limit, initiate corrective action' within 15 minutes and restore-the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER tp less than 25% of RATED THERMAL POWER within the next O

[ 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

F I

A:

O su 4

P l SURVEILLANCE REQUIREMENTS

].

4

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

i

a. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 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 1

I c. 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 j- on a LIMITING CONTROL ROD PATTERN for LHGR.

3 di p:

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3/4.2'POWERDISTRIBUTIONLIMQS

' BASES-The specifications of this section assure that the peak cladding temperature followingthepostulateddesignbasisloss-of-coolantaccidentwillnotexceed

-the 2200 F limit.specified in 10 CFR 50.46.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. The specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations.

The peak cladding temperature (PCT) following a postulated loss-of-coolant- ,

accident is primarily a function of the average heat generation rate of all the rods of a fuel assemoly at'any axial location and is dependent only secondarily on the red-to-red power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered red 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 Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. fy - - - - -

gq ,

" ,The calculational procedure used to establish the APLHGR values for the /

initial cycle and first reload fuel shown on Figures 3.2.1-1 and 3.2.1-2 are f ed on a loss-of-coolant accident analysis.

~

The analysis was perf,oJmed usin neral Electric (GE) calculational models which are consis t with the requi . ents of Appendix K to 10 CFR 50. A complete disc' ion of each code employed

the analysis is presented in Reference 1 ifferences in this. analysis comp d to previous analyses performed h Reference 1 are:

(1) the analysis assum a fuel assembly plaryar po consistent with 102% of the MAPLHGR snown in Figur .2.1-1, IhAiwproduct decay is computed ,

assuming an energy release rat f '. ion; (3) pool boiling is assumed i after nucleate boiling is lost du s gnation period; and (4) the effects of core spray entrainment an unter w rrent flow limitation as described in Reference 2, are inclu di he reflooding calculations.

The APLHGR values for th eload fuel show in Figure 3.2.1-3 are based on the fuel thermal-mechanica esign analysis. The roved 5f~~~'GESTR-LOCA analysis (Reference 3) formed before the startup o ycle 3, used bounding

.MAPLHGR values of 13. and 14.0 kw/ft, independent of no exposure. These MAPLHGR values are gher than the expected " thermal-mechan 1 MAPLHGR" for f both BP8x8R and 8x8EB fuel. Therefore, SAFER /GESTR establis that for all BP8x8R and G BEB fuel designs, the MAPLHGR values are not expec to be limited b OCA/ECCS considerations. However, MAPLHGR values are sti required-

)toass that the LHGR limits are not compromised; and, consequently, 1 rod

.[mec nical integrity is maintained.

J l

LA SALLE UNIT 1 B 3/4 2-1 mendment No. 58

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) .Qj;

(

I INSERT C~

h i.

v ..

l The calculational ~ procedure used to establish APLHGR values'uses SE'

. , calculational models which are approved:and consistent with the-lJ' requirements'of Appendix K to 16CFR59.

L.

i-i l..

3.

l' 2-e' O

7

).

,.m

, " ' h n

p i) t, s

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0:

,f 4

3-is '

e f) {

4; y

N:

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h a.

n

{, ',! }

, er b

= _ _.-__ ___ _- -_ . - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ - _ _ _ . _ - _ . - _ _ - _ _ _ - - _ _ - . _ - _ _ _ _ _ _ _ _ _ _ .___- .

7'.c r

P .

JPOWER DISTRIBUTION' SYSTEMS

{ n 4:

(W BASES d

4 3/4.2.2 APRM SETPOINTS p

t The fuel cladding integrity Safety Limits of Specification- 2.1 were based 1

on a power distribution which would yield the design LHGR at RATED THERMAL y POWER. The flow biased simulated thermal power-upscale scram setting and r control rod block functions of the APRM instruments for both two recirculation l ' loop operation and single recirculation loop operation must be adjusted to i ' ensure that the MCPR does not become less than the fuel cladding safety limit.

or that > 1% plastic strain does not occur in the degraded situation. The scram settings and rod block set _ tings are adjusted in accordance with the for-3 mula in this specification when the combination of THERMAL POWER and MFLPD f indicates a higher peaked power distribution to ensure that an LHGR transient j 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 J. cladding integrity SafetytLimit MCPR, and an analysis of abnormal operational

1. transients. For any abnormal operating transient analysis evaluation with the j; initial condition of the reactor being at the steady-state operating limit, J. it~is required that the resulting MCPR does not decrease below the Safety Limit j -

MCPR at any time during the transient assuming instrument trip setting given in P

b,)., Specification 2.2.

k To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tion on CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss i[,

of flow, increase in pressure and power, positive reactivity. insertion, and coolant temperature decrease. The limiting transient yields the largest delta li MCPk. When added to the Safety Limit MCPR, the required minimum operating

] limit MCPR of Specification 3.2.3 is obtained and presented infig ggg.pg l

' t the. R d Jithdra-al :...o 1- limits 9 ~ vided. Th s l'

,5 t..m ';.....% ..m.cm function of tha ~

c.c...,

. cK Monitor

~5 @#7 F (RBM) setpoint. Th ,

H+ mi 1 cased on the current

, RBM setpoint. The tf T'

< _ '* e RBM setpoint/MCPR limit allows efficient use o +": anced operating domain (t d f n , while main-

~

n' *

, .. enslent protection with the more restrictive MCPR hm .

Analyses have been performed to determine the effects on CRITICAL POWER g- RATIO (CPR) during a transient assuming that certain equipment is out of service.

i -A detailed description of the analyses is provided in Reference 5. The anal-i yses performed assumed a single failure only and established the licensing bases to allow continuous plant operation with the analyzed equipment out of L service. The following single equipment failures are included as part of the l! transient analyses input assumptions:

s i 1) main turbine bypass system out of service, 9

hm' 2) recirculation pump trip system out of service, 1

l LA SALLE UNIT 1 B 3/4 2-3 Amenament No. 58

i

~

I POWER DISTRIBUTION SYSTEMS j

BASES MINIMUM CRITICAL POWER RATIO (Continued) bhe C06 3) safety / relief valve (S/RV) out of service, and

, OPGAM7d'6 4) feedwater heater out of service (corresponding to a 100 degree F

, yppT5 reduction in feedwater temperature).

Lg6 PORT

- 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 estab,lished 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 g

nominal power shape, thus yielding a bounding scram response with reasonable

( conservatism for the MCPR LCO values in future cycles. The 'yc': W e Ment MCPR LCO values shown iPriwe 3.2.11b for the main turbine bypass and

[ov ' "h" ch M N k h ', * * * *

he cycle spe n analysis for the Load Reject Without Bypasim h f Without Bypass events yield MCPR va10es less than or eoual to . .29 for Oplions4 n B, respectively.

~

2) The cycle c4tif ana'i edwete ntroller Failure MCPR LCO values less than 1.25 I~fcH)pti A and B, respectively, when analy n d with ermal-feedwater-temperatur 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.

For operation with the feedwater heater out of service, a cycle specific l

> 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 becoce the limiting i transient for a specific cycle. Consequently, the cycle specific analysis for the feedwater controller failure event will be performed with a 100 degrec F feedwater temperature reduction. The calculated change in CPR for that event will then be used in determing the cycle specific MCPR LCO value.

In the case of a single S/RV out of service, transient analysis results showed that there is no impact on the calculated MCPR LCO value. The change in CPR for this operating condition will be bounded by reload licensing calcu-lations, and no further analyses a.e required. The analysis for a single S/RV out of service is valid in conjunction with dual and single recirculation loop operation.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient j LA SALLE UNIT 1 B 3/4 2-3a Amendment Ns. 58

p POWER DISTRIBUTION SYSTEMS 3

q E:)

BASES 4

i j MINIMUM CRITICAL POWER RATIO (Continued) 2 The value for T used in Specification 3.2.3 is 0.687 seconds which is B

conservative for the following reason

1 i For simplicity in formulating and implementing the LCO, a conservative n

value for IN of g 598 was used. This represents one full core data set i=1 4

at BOC plus one full core data set following a 120 day outage plus twelve 10% of core, 19 rods, data sets. The 12 data sets are equivalent to d 24 operating months of surveillance at the increased surveillance frequency of one set per 60 days required by the action statements of j Specifications 3.1.3.2 and 3.1.3.4.

q That is, a cycle length was assumed which is longer than any past or ,

3 contemplated refueling interval and the number of rods tested was maximized

, in order to simplify and conservatively reduce the criteria for the scram time at which MCPR penalization is necessary.

! H perpece c# the v. f-to F '~fgur: 3.2.2-2 i: to defi ne operating (

atotherthanratebcoreflowconditions. At less than 100% of ra [

N lim'

{d flow the uiredMCPRistheproductoftheMCPRandtheK[ed.

factor K .

i factors assure at the Safety Limit MCPR will nyt be viola K fac(ors were derived using MAL POWER and core fjl # correspond' o105%of j rated steam flow. y /t '

Q \ s The K factors were c ula ;rd for the maximum core flow rate

.] f 3

and the corresponding control line, the limijng

3. P0

along .e 405% of rated steam flow e's relative powE N as adjusted until the MCPR was slightly above th ety Limit. Using this relat R bundle power, the MCPRs were calc ed at different points along the 105% ob y ted steam flow control li corresponding to different core flows. The ratio o e MCPR calc .ed at a given point of core flow, divided by the operating 11. % HCPR, ermines the Kf. l 4 At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pumo speed and the j '

moderator void content will be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a

, considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum W

'i J r O

'f L LA SALLE UNIT 1 B 3/4 2-5 Amendment No. 58 l

l 1 L - --------------o

DESIGN FEATURES 5.3' REACTOR CORE FUEL ASSEMBLIES - lN' 5ar D 5.3.1 Thereactorcoreshallcontain764fuelassemblies,hith::;hfuel

== ==My c;at:iM aa 62 fuel rada and t:0 water iede-clad-with Zircaloy d .

Each f a:1 =d th:11 hr;: a a;;.insi ective fuel-length-ef-150-inches The initi:1 er; leeding shell hsve e . xirr evermae' eaeichment-oLI.49-weight- -

p: Meat U-235. #aland fuel-sha11 De sist11r-tn physical design-to-the4nitial cerc ?:; ding.

~

CONTROL ROD ASSEMBLIES crWMrm S/i'/4l 5.3.2 The reactor core shall contain 185 Icontrol rod assemblies. T6 . a re-te: pe:sible i.yper-of- control-rodem-one-consisting-of a cruciforstarr-sy-ef-

-stainless stevi tees- containinn 141_daches--of-boron-<artide,_B C#wder surre"ad*A by e cNc4f;= :.h;;;d- sta4alass-steel-sheath and the 4tecond-type centein:-143 --iaches ef 2:crbee-meterial-of-which-the-fiTTt-6-inches-are-hafni a nd the-remainderd s B.C. 4 ggf yg.f.;,/ ffgf/ /,e ggc,7 carn odg[gc) on '"N#8

/y[0RIOOLANT SYTTEh # ##A "& "haniumgefaf.

"# & cm frard "

5.[REAC '

DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of:

1. 1253 psig on the suction side of the recirculation pumps.

2. 1650 psig from the recirculation pump discharge to the outlet side of the discharge shutoff valve.

3. 1500 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575'F.

VOLUME 5.4.2 The total water and steam volume of the reactor vessel and re' circulation system is

21,000 cubic feet at a nominal T,y, of 533*F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

)

LA SALLE - UNIT 1 5-4 Amendment No. 58 L.-__-___--______-__-______--_______-_ _ _ _ _ _ _ . _ _ _ _ _ _ - __ _ _ _ _ _ _ - - _ _ -

INSERT D Each assembly consists of a matrix of Zircalloy clad fuel rods with an initial composition of.slightly enriched uranium dioxide, U0 2. fuel assemblies shall be limited to those fuel designs approved for use in BWR's.

0189T:20

. 3 <

t h )i ; y: '

A. -

3, -

3 V

. ADMINISTRATIVE CONTROLS 4

W Semiannual Radioactive Effluent Release Report (Continued)

( . :. . .

L The radioactive effluent release report shall' include the .following information for each type of solid waste shipped offsite during the (y report period: -

[ a. Container volume, a

l b. Total curie quantity (specify whether determined by 1 measurement or estimate),

b c. Principal radionuclides (specify whether determined by

{ measurement or estimate), *

$ d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

}

I a. Type of container (e.g. , LSA, Type A, Type B, Large Quantity),

and

] f. Solidification agent (e.g. , cement, urea. formaldehyde).

j The radioactive affluent release reports shall include unplanned 1 releases from the site to unrestricted areas of radioactive materials j 'in gaseous and liquid effluents on a quarterly basis.

tj '

h Tte radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting pe -f od.

lj

[ 5. Monthly Operating Report -

Routine reports of operating statistics and shutdown experience, 3 including documentation of all challenges to safety / relief valves, d shall be submitted on a monthly basis to the Director, Office of 3

Management Information and Program Control, US Nuclear Regulatory

[j . Commission, Washington, DC 20555, with a copy of the appropriate a Regional Office, to arrive no later than the 15th of each month j' following the calendar month covered by the report.

f Any changes 'to the OFFSITE 005E CALCULATION MANUAL shall be submitted i with the Monthly Operating Report within 90 days in which the change (s)

W was made effective. In addition, a report of any m&jor changes to f, the radioactive waste treatment systems shall be submitted with the 9 Monthly Operating Report for the period in which the evaluation was J reviewed and accepted by Onsite Review and Investigative Function.

B. Deleted ggg# 0

'L 3

d f (Q

!j! LA SALLE UNIT 1 6-24 Amendment No.23 i

4 i

INSERT B

6. CORE OPERATING LIMITS REPORT

a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following:

(1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1. s (2) The minimum Critical Power Ratio (MCPR) (including 20% scram time, tau (1P), dependent MCPR limits, and Kf core flow HCPR adjustment factors) for Technical Specification 3.2.3.

(3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4.

(4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in NEDE-240ll-P-A, General Electric Standard Application for Reactor Fuel (latest approved revision).

c. The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanical limits such as shutdown margin, and transient and accident analysis limits) of the safety analysis ,

d are met,

d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, for each reload cycle, to the U.S. Nuclear Regulatory Commission Document Control Desk with copies to the Regional Administrator and Resident Inspector.

i l

l 0189T:19

v y

1; a .- ,

J h

1~

INDEX-

. DEFINITIONS ~

u.

3-

~.

4'

" SECTION PAGE 4 1.0 ' DEFINITIONS 1-1 l

1.1 AC. TION.............................................................

1 1.2' AVERAGE PLANAR EXP05URE...........................................-

RATE........................ 1-1 t 1.3 AVERAGE PLANAR LINEAR HEAT GENERATION

, 1-1 i 1. 4 CHANNEL CALIBRATION...............................................

4 2 1-1 4 .1. 5 CHANNEL CHECK. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

1-1

, .1. 6 CHANNEL FUNCTIONAL TEST...........................................

j. #5N 1.7 CORE ALTERATION................................................

.. 1-2

-.. . . _ ..._ t . A- )

L 3s.s coa ot.sc+rw6 umtTS KD%cr...... 1-2

1. p? CRITICAL POWER RATI0. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

i s'

e ~1-2 '

~, 1.f ' DOSE EQUIVALENT I-131.............................................

.k il 1-2 a 1. $ l- AVERAGE DISINTEGRATION ENERGY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

(L 1-2 i

1. M EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME... . . . . . . . . . . . . . . -

1-2 1.hEND-OF-CYCLERECIRCULATIONPUMPTRIPSYSTEMRESPONSETIME.........

~

j. it DENSITY................................

1-3' i 1 5 FRACTION OF L W ING POWER L '/f 1-3

1. y FRACTION OF RATED THERMAL P0WER................................... i

/G 1-3

1. 3,5 FREQUENCY N0TATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

/7 1-3 1.lg GASEOUS RADWASTE TREATMENT SYSTEM.................................

n /f 1-3 1.yf IDENTIFIED LEAKAGE................................................

't

/1 1-3 'f TIME.................................... !

1.)4 ISOLATION SYSTEM RESPONSE

? zo 1-3

1. 4 LIMITING CONTROL ROD P ATTERN. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

lf i 1-4

f. 1.gLINEARHEATGENERATIONRATE.......................................

2 en 1-4 f 1. 7/I LOGIC SYSTEM FUNCTIONAL TEST. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

es 1-4

("_:

1 :

1. M MAXIMUM FRACTION OF LIMITING POWER DENSITY........................

zy 1-4 6 1. $ MINIMUM CRITICAL POWER RATI0......................................

g 1-4 !

MANUAL...................................

1.)4'0FFSITE DOSE CALCULATION

- --- "Emm o 1M& f2 I

$ -1

< }

(a

. 4

7. , , , ,

g- .

b . - .

3, ; , ..

i .

(qQ 1

INDEX j

p 3 DEFINITIONS 2

e 8.

,SECTION

[, ~

PAGE

.!' DEFINITIONS (Continued)

V 26 1-4

j. 1. g OPERABLE - OPERABILITY............................................

1-4

1. OPERATIONAL CONDITION - CONDITION.................................

j a 1-4 r 1.pT PHY SI CS T EST S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .'. . . . . l<'

1 . . et 1-5 1 1. % PRESSURE BOUNDARY LEAKAGE.......................................... ;'

i 30 1-5 j 1.J8PRIMARYCONTAINMENTINTEGRITY..................................... -

31 . 1-5 j .-

] 1. '$ti P ROCESS CONTRO L P R0 GRAM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i.

Q 3: 1-5 i

) 1.yf PURGE - PURGING................................................... i~

L ss 1-5 1.)IRATEDTHERMALP0WER...............................................

{j 'Yr .M 1-5

.; . 1. )!f REACTOR PROTECTION SYSTEM RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . .

1. 3{ 1-6 l

$, i 1. % REPORTABLE EVENT..................................................

1 3f. 1-6 j 1.)6 ROD DENSITY.......................................................

37 0

INTEGRITY...................................

1-6 dL

1. K SECONDARY CONTAINMENT -

{ 3s 1./T SHUTDOWN MARGIN....................................................

1-6 [

h 11 j

% 1-6 j- 1.) SOLIDIFICATION.....................................................

yo 1-7 j y~ 1.psSOURCECHECK......................................................

w i

1-7

i: 1 34f ST AGGERED TEST BASI S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

18.- 1-7

[ 1.yITHERMALP0WER....................................................

V 8;

g 1.g' TURBINE BYP ASS RESPONSE TIME. . . . . . . . . . . . . . . . . . . . . . . . . . . .1-7 ..........

4 vr 1-7 1 1.f5 UNIDENTIFIED LEAKAGE..............................................

l vr 1-7 j 1 J4 VENTILATION EXHAUST TREATMENT 4

SYSTEM..............................

f 1-7 4 1. M V ENT I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

2 lt p,

.. t ,7

'o'

\f ..

i is.

.P

s h LIST OF FIGURES FIGURE PAGE i 3.1.5-1 . SODIUM PENTABORATE SOLUTION TEMPERATURE /

j CONCENTRATION REQUIREMENTS ........................ 3/4 1-21 j 3.1.5-2 SODIUM PENTABORATE (Na2010016 10 H2O)

VOLUME / CONCENTRATION REQUIREMENTS ................. 3/4 1-22 1

1 -1 MAXIMUM AVERAGE PLANAR LINEAD. HEAT GENERATION RATE (MAPLHGR) VERSUS AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB176, 8CRB219, and 4

RB071 ......................... ........... .. 3/4 2-2 3.2.1-2 MAXI VERAGE PLANAR LINE Hi NE RATE

~

(MAPLHGR) SUS AVERAGE NAR . .. , FUEL TYPE

' BP8CRB299L. ... ... ..... .... ................. 3/4 2-2(a) 3.2.1-3 MAXIMUM AVERA PLA I HEAT GENERATION RATE ER S AV NAR EXPOSURE, FUEL (MAPLHGRp$0D TYBf BE 0C....... ................... 3/4 2-2b 3.2.3-la MIN 1 4L L POWER RATIO (MCPR RSUS IALJ FLOW ......................... ........ 3/4 2-5 3.2.3-lb MUM CRITICAL POWER RATIO (MCPR) VERSUS T I (v U

AT RATED FLOW FOR END OF CYCLE RECIRCULATION PUMP

'4 2-5a

? TRIP AND MAIN TURBINE BYPASS SYSTEMS INOPERABLE ...

a L .3-2 K f FACTOR ......................................... 3/4 ;

3.4.1.5-1 CORE THERMAL POWER (% OF RATED) VERSUS TOTAL CORE FLOW (% OF RATED) .................................. 3/4 4-Sc 3.4.6.1-1 MINIMUM REACTOR VESSEL METAL TEMPERATURE VS. REACTOR VESSEL PRESSURE ....................... 3/4 4-19 g

4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST ........ 3/4 7-33 m

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

B 3/4.4.6-1 CALCULATED FAST NEUTRON FLUENCE (E>1MeV) at 1/4 T AS A FUNCTION OF SERVICE LIFE ..................... B 3/4 4-7 B 3/4.6.2-1 SUPPRESSION POOL LEVEL SETPOINTS .................. .

B 3/4 6-3a E 5.1.1-1 EXCLUSION AREA AND SITE BOUNDARY FOR GASEOUS AND LIQUID EFFLUENTS .............................. 5-2 j.

5.1.2-1 LOW POPULATION ZONE ............................... 5-3 6.1-1 CORPORATE MANAGEMENT .............................. 5-11

[

6.1-2 UNIT ORGANIZATION ................................. 6-12 (4 . ,3 c 6.1-3 MINIMUM SHIFT CREW COMPOSITION ...................., 6-13 3

r.

LA SALLE - UNIT 2 XIX Amendment No. 41

--_~x--.___~

t l

'Q DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the addition, removal, relocation or movement of I fuel, sources, incore instruments or reactivity controls within the reactor prossure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of

[L the movement of ,a component to a safe conservative position. 7

'# A i

I CRITICAL POWER ratio '

l p 1./ The CRITICAL PUuR RATIO (CPR) shall be the ratio of that power in the L

i assembly which is calculated by application of the GEXL correlation to L

cause some point in the assembly to experience boiling transition, divided by the' actual assembly operating power.

p DOSE EQUIVALENT I-131

~

1./ 00SE EQUIVALENT I-131 shull be that concentration of I-131, l

} iD microcuries/ gram, which alone would produce the same thyroid dose as the i

quantity and isotopic mixture of I-131. I-132,1-133, I-134, and I-135 I actually present.. The thyroid dose conversion factors used for this

' calculation shd 1 be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

f E-AVERAGE DISINTEGRATION ENERGY l . (Ce l

l' 1.)6 E shall be the avertge, weighted in proportion to the concentration of each radionuclides in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 iainutes, making up at least 95% of the. total non-iodine activity in the coolant.

EMERGENCYCORECOOLINGSYSTEM(ECCS)$ESPONSETIME 1.yf The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that l time 9, interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of

[-

perf orming its safety function, i.e. , the valves travel to their required positions, pump discharge pressures reach their required values, etc..

2 Times shall include diesel generator starting and sequence loading delays where applicable. The response time way be measured by any series of seauential, overlapping or total steps such that the entire response time is mecsured. f END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME l

l /TheEND-OF-CYCLERECIRCULATIONPUMPTRIPSYSTEMPESPONSETIMEshallbe 1 1.)6 that time interval to energization of the recirculation pump circuit i breaker trip coil from when the monitored parameter exceeds its trip setpoint at the channel sensor of the associated:

! a. Turbine stcp valves, and I b. Turbine control valves.

The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

LA SALLE - UNIT 2 1-2

e N:

t

?

3

. .iU. ..

n.

q , gp 4

j. ,

INSERT A '

,)

i i

. CORE OPERATING LIMITS REPORT 9

l- 1.8" The CORE OPERATING LIMITS REPORT is the unit-specific document that provi'es d core operating limits for the current. operating' reload cycle.

1 These cycle-specific core operating limits shall be determined for each L reload cycle in accordance with Specification 6.6.A.6. plant operation within these operating limits is. addressed in indi'fdual specifications.

1 L

[-

]

0 j.

($) ,

1 i

rl

+

f.r.

s .

d

( .

l' l -

=

l

., . [

F,-

b' ':

N .

1.

f.

3 l1 l K

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}. .

(j i J- .. .

l 1

- . - - = ____m-__ _ _ _ _ __ _ _ _ _ _ ___ .

4 h4 4

)

,.h .(b j DEFINITIONS 3

S FRACTION OF LIMITING POWER DENSITY M

M

-I..JJfTheFRACTIONOFLIMITINGPOWERDENSITY(FLPB)shallbetheLHGRexisting l Q' IV at a given location divided by the specified LHGR limit for that bundle type.

ra y<

7' FRACTION OF RATED THERMAL POWER A,

4 4 The FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL

" 1.)iPOWERdividedbytheRATEDTHERMALP9WER.

f g FREQUENCY NOTATION s '

j ' Af The FREQUENCY NOTATION specified for the performance of Surveillance L-1.)te Requirements shall correspond to the intervals defined in Table 1.1.

g i i GASEOUS RADWASTE TREATMENT SYSTEM j 1.pf A GASEOUS RADWASTE TREATMENT SYSTEM shall be any system designed and (

7 installed to reduce radioactive gaseous effluents by collecting primary

i j coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to 3

release to the environment.

{3' ID8iNTIFIED LEAKAGI

{

1.)# IDENTIFIED LI:!AKAGE shall be:

9 ab i- 2. Leakage into collection systems, sxh as pamp sea! or valve 9 packing letts, thEt is captared and conducted to a sump or a collecting tink, or i}

j b. Leafage inte the c>ntainment atusrhers f som sourcas that ars both spedfka13y lochted and knwn eitter act 'o interfero y witn the operatiort of tno leakage dete.ction systes er not to a;

be PRESSERE BOUNDAR't LEAXAGE.

. ISOLATION SYSTEM RESPONSE TIME i

The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the J" 1.[gmonitoredparameterexceedsitsisolationactuationsetpointatthecha s.ensor until the isolation valves travel to their required positions. Times et"i include diesel generator starting and sequence loading delays where appiicable. The response time may be measured by any series of sequential, 8 overlapping or total steps such that the entire response time is measured.

1 f

l LIMITING CONTROL R00 PATTERN l

1./;/ A LIMITING CONTROL ROD PATTERN shall be a pattern wh value for APLHGR, LHGR, or MCPR.

(.. ,

n LA SALLE - UNIT 2 1-3

.-- _ ____ U

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

1 '

'((Ci

. DEFINITIONS i

1

] LINEAR HEAT GENERATION RATE j' 1.JAf LINEAR HEAT GENERATION RATE (LHGht) shall be the heat generation per el length of fuel rod. It'is the. integral of the heat flux over the heat -

y 4 transfer area associated with the unit length. ,

y .

) LOGIC SYSTEM FUNCTIONAL TEST'

.i -

- 1.g* A LOGIC SYSTEM FUNCTIONAL TEST shall be a test of all logic c

?,

etc. of a logic circuit, from sensor through and including the actuated j device to verify OPERA 8ILITY. THE LOGIC SYSTEM FUNCTIONAL TEST may be i performed by 'any series of sequential,. overlapping or total system steps r such that the entire logic system is tested.

4 j MAXIMUN FRACTION OF LIMITING POWER DENSITY l

4 (The MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD) shall be the highest 3

1.)4 value of the FLPD which exists in the core.

J MINIMUM CRITICAL POWER RATIO l

j. I 1

1.gThe MINIMUM CRITICAL POWER RATIO (MCPR) shall be the smallest CPR which 2 4 exists in the core.

j' f#

_ OFFSITE DOSE CALCllLATION MANUAL 3

1./ The 0FFSITE DOSE CALCULATION MAN'JAL (00CM) shall contain the r.athodology (

f d and parame*ers used in the calculation of offsite doses due to radioactive gaseaus and liquid efficenes and in the calculation of gaseous and ifquid

['

4 effluer.t monitoring ahrshrip . setpoints.

q '

[ OPERABLF - OPERA 3ILITY 3-i 1.[d OPERABILITY when it is capable of perfoming its scecified functio }

and when all necessary attendant instamentar. ion, c9ntro'is, a norms 1 and an emergency electrical ~ power source, cooling or seal water, lubrication 9

or other auxiliary equipment that are reouired for the syst2m, subrystet 1 train, component or device to perform its function (s) are also capab1e of performing their related support function ($).

OPERATIONAL CONDITION - CONDITION A An OPERATIONAL CONDITION, i.e., CONDITION, shall be any one inclusive i 1.g'lcombinationofmodeswitchpositionandaveragereactorcoolanttemperature e s j as specified in Table 1.2.

p I 4

PHYSICS TESTS i 4

1. PHYSICS TESTS shall be those tests performed to measure the fundamental

{ '

nuclear characteristics of th6 reactor core and related instrumentation

/ and 1) described in Chapter 14 of the FSAR, 2) authorized under the I provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

.I

-h.

i

'LA SALLE - UNIT 2 1-4

A DEFINITIONS PRESSURE BOUNDARY LEAKAGE

1. # PRESSURE BOUNDARY LEAKAGE shall be leakage through a non-isolable. fault l M in a reactor coolant system component body, pipe wall or vessel wall.

PRIMARY CONTAINMENT INTEGRITY 1.28f PRIMARY CONTAINMENT INTEGRITY shall exist when: l

a. All primary containment penetrations required to be closed during accident conditions are either:

1. Capable of being closed by an 0FERABLE primary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except' as provided in Table 3.6.3-1 of Specification, 3.6.3.

b. All primary containment equipment hatches are closed and sealed.

c. Each primary containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d. The primary containment leakage rates are "ithin the limits of Specification 3.6.1.2.

e. The suppression chamber is OPERABLE pursuant to Specification 3.6.2.1.

f. The sealing mechanism associated with eacn primary containment penetration; e.g., welds, bellows or 0-rings, is OPERABLE.

PROCESS CONTROL PROGRAM 1.FThe PROCESS CONTRCL PROGRAM (PCP) shall contain the sampling, analysis, I SI snd fonr:J1ation outermination by which SOLIDIFICATION of racicactive wastes frur,l! quid systees is assures. ,

PUME

PORGING }

1.dPUkGE or PURGM rn#11 be the contttiled process cf discharging' air or l l D ;ns t froir a co,1finaent to maintain temperature, pressure humidity, concentratitan or othet' cperating ecdition, in such a mannar that rep 1hte- f ment air or gas is required to purify +.hn confir.ement. 1 RATED THERMAL POWER .

1. X RATED THERMAL POWER shall be a total reactor core heat transfer rate to l )

U the reactor coolant of 3323 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME 1.WREACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from l 2Y when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of.the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping on total steps such that the entire response time is measured.

LA SALT.E - UNIT 2 1-5 a----Au._,_--- - - - - - - - - _ , _ _ _ _ _ _ _ _ _ _

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! DEFINITIONS I

p .

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h REPORTABLE EVENT

!. t

[ . ( A REPORTABLE EVENT shall'be any of those conditions specified in 3 o 1.)35 Section 50.73 to 10 CFR Part 50. ,

f RCD'OENSITY L .

J

[

6 ROD DENSITY shall be the number of control rod notches inserted as a 1.)# fraction of the total number of control rod notches. All rods fully

i. inserted is equivalent to 100% ROD DENSITY.

); SECONDARY CONTAINMENT INTEGRITY .

l 1. . SECONDARY CONTAINMENT INTEGRITY shall exist when:

a. All secondary containment penetrations required to be closed i during accident conditions are either: -l

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system, or

2. Closed by at least one manual valve, blind flange, or deactivated automatic damper secured in its closed position, except as provided in Table 3.6.5.2-1 of

.t

{uY Specification 3.6.5.2.

b. All secotidery containment hatches and blowout panels are closed f and sealed.

g:

[ c. The stancby gas treatment system is OPERA.BLE pursuant to i- Specification 3.6.5.3.

[ d. At inst che door in eAch &ccess to the secondary containment

)- is closed.

e. The seclirg e.echnism associated with each secondary containment

/ penetrating, e.g. , welds, bCilows, or 0-rings, is OPERABLE. ;

y . .

f. The pressura within the secondary containment is less than or equal to the vabe required by Specification 4.6.5.1. 2.

l i

i- FHUT00WN "ARGhi 0

3 1.yf SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is 39 suberitical or would be suberitical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth C which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68*F; and xenon free, a'

SOLIDIFICATION n

1.)6 SOLIDIFICATION shall be the conversion of radioactive wastes from liquid l

" 37 systems to a homogeneous (uniformly distributed), monolithic, immobilized 4{

'" solid with definite. volume and shape, bounded by a stable surface of distinct outline on all sides (free-standing).

LA SALLE - UNIT 2 1-6 Amendment No.ll

t. . .

I l

DEFINITIONS I .

SOURCE CHECK V

b 1.kA'SOURCECHECKshallbethequalitativeassessmentofchannelresponse l 40 when the channel sensor is exposed to a radioactive source.

?

STAGGERED TEST BASIS

[ 136ASTAGGEREDTESTBASISshallconsistof:

A test schedule for n systems, subsystems, trains.or other l

I a.

designated components obtained by dividing the specified test interval into n equal subintervals. '~

t r The testing of one system, subsystem, train or other designated b.

[ component at the beginning of each subinterval.

j f

h ; THERMAL POWER 4

l

[ -1.g THERMAL POWER shall be the total reactor core heat transfer rate to the

} W reactor coolant.

TUR8INE BYPASS SYSTEM RESPONSE TIME f .

M,'. l i 1.g The TUR8INE BYPASS SYSTEM RESPONSE TIME shall be time interval from when 6 the turbine bypass control unit generates a turbine bypass valve flow

~ signal until the turbine bypass valves travel to their required positions. I The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

it j UNIDENTIFIED LEAKAG,E, a:

l

[ 14 UNIDENTIFIED LEAKAGE thall be all leakge which is not ID54TIFTED LE b VENTILATIch EMAUST TPMAT_ -- MENT SYS_TE_M p

l lj 4 A VENTILAf t0N EX3AUST TREAT.Of SYSTEM r.hr.11 ea my syster; duigned and 1.pdirstalled to redxo gasecus ndicie41 ta or radioactive rineHa! f 2 partice- j lata form in affluents by passing rentiliticn or vent er.nnst gases j: through charcoal a6sorbers and/cc HEPA filters for the purpos,e of remo'*ing j

j iodines or p: articulates from the ge.saous exntoet streaa prior to tre release to the environment (such a system is.nct considered to have any

/ effect on noble gas effluents). Engineered Safety Feature (ESF) atmospheric H-cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT

[

W SYSTEM components.

i VENTING

/'

1 i 136 VENTING shall be the controlled process of discharging air or gas from a-4 confinement to maintain temperature, pressure, humidity, concentration or T m other operating condition, in such a manner that replacement air or gas is

%./ not provided or required during VENTING. Vent, used in system namw , does

?;

4 not imply a VENTING process.

A P-N LA SALLE - UNIT 2 1-7 P 4

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3/4.2 POWER DISTPJBUTION LIMITS

! 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 6

LIMITING CONDITION FOR OPERATION

$ 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type

. of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits

, d:r. f.7 "i;r n 2.2.1-1, 0.2.1-2, .r.d 2.2.1-0.

3.

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

J ACTION:

i With an APLHGR exceeding the limits :' "i;;rn 2. 2.1 -1, 2. 2.1-2 xd 2. 2.1-3,e initiate corrective action within 15 minutes and restore APLHGR to within the 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 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 .

hciQed * %e. CofE OPERATt% LitrhT5 REfbRjT 3

(,a l'

SURVEILLANCE REQUIREMENTS 4.2.1 #.11 APulGRs shall be verified to be equal to or less than the limits

, detem h & .2.es Tism <.; 0.0.0 1, 0.0.1 0 ad 0. 21-2; -

4

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

a.

b., Within 12 hodr$ efter completion of u TEEPAL POWER fr-:rease of at least 15% rA RATED *OiERHAL POWER, and

c. InitVally and at least once par 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months d en v.he reactor it, i

. operatir.g with a LIMITING CONTROL ROD MTTERN for APLHGR. i 5

i j.

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LASALLE-UNIT 2 3/4 2-1 Amendirent No. 41

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Amendment No. 41

4 i.

L .m POWER DISTRIBUTION LIMITS

! kV 3/4.2.3 MINIMUM CRITICAL POWER RATIO L

,- LIMITING CONDITION FOR OPERATION l 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit deux'nd ' xL-yec cd :n uc evonpaws umrrs R6FoAT l

[ a. Single Recirculation Loop Operation

Fi ure 3.2.3-la (Curve A for a RBM setpoint of 106% or Curve or.

BM setpoint of 110%) plus 0.01, times the kf determine rom '

Fig 3.2.3-2. ,

L b. Two Recirc tion Loop Operation Figure 3.2.3- Curve A for a RBM ye it 106% or Curve B for

!- a RBM setpoint o ) times k ined from Figure 3.2.3-2.

f

c. Two Recirculation Loop 0 on h Main Turbine Bypass Inoperable l armined from Figure 3.2.3-2, for two

Figure 3.2.3-lb ti the recirculation looli4pei n, wit e main turbine bypass system inoperable peOpse4'f tion 3.7.10 RBM setpoint determined.per -

Specification Tab .3.6-2 may be used).

tion Loop Operation with End-of-Cyc Recirculation Pump.

4 i o, d. Two Recirc '

! Qj Trip 5 em Inoperable 1- Fi 3.2.3-lb times the k determined from Figure 3. . -2, for two f '

i circulation loop' operation, with the end-of-cycle recircul on pump 4

trip system inoperation as directed by Specification 3.3.4.2 (an BM setpoint determined per Specification Table 3.3.6-2 may be used).

i.

APPLICABILITY:

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

b ACTION

a. With MCPR less than the app 7ictbit. MCPR limit as determined for one of the dava-conditierap. 9tct4M in A cnw cesAnMG umrrs l

{

I. initiate corrective action within 15 minutes, and

?

O 2. Rastore EPR to within the required limit witnin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . .

f' Otherwise, reduce THERMAL POWER to less than 25% of RATED f/ 3.

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

i When operating in a condition not i t I '!:d 5 0, reduce THERMAL l p b.

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 .

". g w.

9 I

LA SALLE - UNIT 2 3/4 2-4 Amendment No. 41 r

1

l l

I I~

I POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO l

1.

i

) SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with:

a. t"V" = 0.86 prior to performance oi' the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or l-

b. T,y, determined within 72. hours of the conclusion of each scram time

'l surveillance test required by Specification 4.1.3.2, I

shall be determined to be equal to or greater than the applicable MCPR limit 6rech9e/

dete.eined fre; ."ig re. 2.2.0-1 end 0.2.;g.$e ccREcpERATms urn:Ts geregt

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

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-L l c. 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 fi with a LIMITING CONTROL ROD PATTERN for MCPR.

4

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6 g-LA SALLE - UNIT 2 3/4 2-4a Amendment No. 41

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4 3/4.2.4 LINEAR HEAT GENERATION RATE O,

Ij- LIMITING CONDITION FOR OPERATION i

}?$ -

d 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall 'not exceeh65 x 4- cote seu%el y b OPERA w G u sits I D . 13.4 kw/ft.for fuel t :

REPotT.

d f 1 *

1. 8CRB176 ./'

2. B219 1 3. BP 9L 3-P b. 14. /ft for el types:

i N

'i 1. BC3000 g 2. BC320C d - APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

? ACTION:

J W

., rm With the LHGR of any fuel rod exceeding the limit, initiate corrective action l P (.? 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 next D 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ll li, 3 SURVEILLANCE REQUIREMENTS

't 3 '

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

c,

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

~

r 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 3: at least 1E% of RATED THERMAL POWER, and

?

R

, c. Intially and 2t least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating

,,j

. on a LIMITING CONTROL ROD FATTERN for LPGR.

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9 LA SALLE - UNIT 2 3/4 2-7 Amendment No. 41 4 .

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0 Ab IL BASES j k

l: .

i The specifications of this section assure that the peak cladding

} temperature following the postulated design basis loss-of-coolant accident I will not exceed the 2200'F limit specified in 10 CFR 50.46.

I

[ 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4-'

This specification assures that the peak cladding temperature following i k the postulated design basis loss-of-coolant accident will not exceed the limit 4

[ specified.in 10 CFR 50.46. This specification also assures that fuel rod .

L mechanical integrity is maintained during normal and transient operations.

l.

i The peak cladding temperature (PCT) following a postulated loss-of-coolant a accident is primarily a function of the average heat generation rate of all J the rods of a fuel assembly at any axial location and is _ dependent .only F

secondarily on the rod to rod power distribution within an assembly. The peak-L 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.

gEg C

(

h. The calculational procedure used to establish the APLHGR values for the

'in tal cycle and first reload fuel shown_on Figure 3.2.1-1 and 3.2.1-2 ar base e loss-of-coolant accident analysis. The analysis was perfo using General tric (GE) calculational models which are consistent wit e

requirements Appendix K to 10 CFR Part 50. A complete discus n of each-g code employed i e analysis is presented in Reference 1. D erences in this

, analysis compared t revious analyses performed- th Ref nce 1 are: (1) the 0 analysis assumes a fue sembly planar power siste with 102% of the j MAPLHGR shown in Figure 3. -1, 'r u decay is computed assuming

an energy release rate of 200 (2) fissioy(3 V/fissi boiling is assumed after j '~ nucleate boiling is lost during t su tion period; and (4) the effects of core spray entrainment and counta r nt flow limitation as described in i Reference 2, are included in t/(e ref

~

calculations.

The APLHGR values for oad fuel sho in Figure 3.2.1-3 are based i on the fuel thermal-mechani design analysis, improved SAFER /GESTR-LOCA 1 analysis (Reference 3) formed for Cycle 3 used bou g MAPLHGR ralues of

13. 0 and 14.0 ke/ft dependent of nodal exposure. Thes APLHGR values are f higher than the ected " thermal-mechanical MAPLHGR" for bo BP8x8R rnd

j. GE8x8EB fuel. erefore, SAFER /GESTR established that for all BR nnd !

GE8x8EB fu designs the MAPLHGR values are not expected to be lim by LOCA/E considerations. However, MAPLHGR values are still required ass that the LHGR limits are not compromised and, consequently, fuel ro chanical integrity is maintained. ..

l 4

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i i LA SALLE - UNIT 2 B 3/4 2-1 Amendment No. 41 i

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f. . The calculational procedure used to establish APLHGR i p values uses BE 1- requirements of Appendix M to 18CFR58. the- : calculationa f.

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- POWER DISTRIBUTION SYSTEMS h

BASES' b 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

POWER. ' The flow biased simulated thermal power-upscale scram setting and con -

trol rod block functions of the APRM instruments for both two' recirculation.

le ; -

loop operation and single recirculation loop operation must be adjusted to ensure h that the MCPR does'not become less than the fuel cladding safety limit or that; l ->.1% plastic strain does not occur in .the degraded situation.- The scram settings, L

and rod block settings are adjusted in'accordance with-the formula in this speci-i' fication when the combination of THERMAL POWER and MFLPD indicates a higher-I peaked power distribution to ensure that an LHGR transient would not be D increased,in the degraded condition.

3/4.2.3 MINIMUM CRITICAL POWER' RATIO The required operating limit MCPRs at steady-state operating conditions t ,

as specified in Specification 3.2.3 are derived from the established fuel cladding. integrity Safety Limit MCPR and an analysis.of abnormal _ operational

transients. . For any abnormal operating transient analysis evaluation with the. ~

L 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 E MCPR at any time during the transient assuming ies trument trip setting.given .

@,_' in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded '

during any anticipated abnormal operational transient, the most limiting 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

[ coolant temperature decrease. The limiting transient yields the largest delta

)- MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in +4pe: 2. 2. Sh.

c ~

m. cacaeurma wu es witurawal trror is 1/ Ing transient evenf M~ K ##

i li n:N-These i ati ction. M iock Monitor sen based on the current R94-(RBM) setpoint. The fw3

setpoint. The flexibil t mrdlii C " int /HCPR limit allows p efficient use o oded operating domain-(ELL ry:x) hile L m ga antient protection with the more restrictive MCPR 1 e. .

5 i 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.

i' The analyses performed assumed a single failure only and established tM licensing bases to allow continuous plant operation with the analyzed equipment out of service. The following single equipment failures are 7

included are part of the transient analyses input assumptions:

s.

L 1. main turbine bypass system out of service, i ( ,'i

2. recirculation pump trip system out of service, ,

LA SALLE - UNIT 2 B 3/4 2-3 Amendment No. 41 ,

- _mm ___._._______..m__m.___.m.m.____ _ . _ _ . _ _ _ - . _ . _ .

POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued)

.r

.Se Co#6 ) 3. safety /relicf valve (S/RV) out of service, and

}'

OfBKAT/N6] .,_f 4. feedwater heater out of service (corresponding to a 100 degree F -

ggg7- 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 nominal power shape, thus yielding a bounding scram response with reasonable conservatism for the MCPR LC0 values in future cycles. The-cycle id= Men MCPR LCO values shown irtfigme 3.2.0 Itr for the main turbine "hd* b *

[c" h [ N hf/c' feS h Y 1 ycle specific analysis for the Load Reject Without Bypass ~and ne4 rip Without Bypass events yield MCPR LCO valueLless-thanC T

equal to 1.S'3 arit-IrQ0ptions A andDerpYccivFly.

2. The cyc1 e analysis for Feedwater Controller Failure event

's M CPR LC0 values less than 1.25 and E21-for Options A and B, respectively, when analyzed with normal feedwater teniperatute.

J N-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.

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, Howner, the feedwater controller failure event becomes more severe with a feedwater heater out of service ano could become the limiting transient for a specific cycle. Consequently, the cycle specific analysis for the feedwater controller failure event wil 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 M JR LCO value.

In the case of a single S/RV Out of service, transient analysis results showed that there is no impact on the calculated MCPR 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.

The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-1 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate events are described LA SALLE - UNIT 2 8 3/4 2-4 Amendment No. 41

_ _ _ _ - _ _ _- -_- x

e ,

V lj '

i POWER DISTRIBUTION SYSTEMS

) %

y V j BASES f MINIMUM CRITICAL POWER RATIO (Continued)

The value for T used in Specification 3.2.3 is 0.687 seconds which is ij conservative for the Bfollowing reason:

f For simplicity in formulating and impleme'nting the LCO, a conservative n

i value for INg of 598 was used. This represents one full core data set I" i=1

- at BOC plus one full core data set following a 120 day outage plus twelve 4 10% of core, 19 rods, data sets. The 12 data sets are equivalent to e 24 operating inonths of surveillance at the increased surveillance

j. -

frequency of one set per 60 days required by the action statements of j Specifications 3.1'.3.2 and 3.1.3.4.

That is, a cycle length was assumed which is longer than any past or i contemplated refueling interval and the number of rods tested was maximized i in order to simplify and conservatively reduce the criteria for the scram time -

j at which MCPR penalization is necessary.

tlc g gvac vf t.he X 'ecter ;f Ti;;.r; 2. 2. 2-: i; t; if'= :ptr: tin;

]i limi otherthanrate[icoreflowconditions. At less than 100% of'ra a flow the r factor 3 factors assure red MCPR is the product of the MCPR and the K[e be viola eK(ors eK fac were derived using T theSafetyLimitMCPRwillg;orrespon to105%of 1 L POWER and co ; -

i rated steam flow. , /

/ -

[

j and the corresponding HE MW1DW The Kf factors were ca: la % ong .

or the maximum core flow rate 05% of rated steam flow

! controlling,thelimik . e's relative pow as adjusted until the MCPR f was slightly above th stety Limit. Using this rela bundle power, the i MCPRs were calc' d at different points along the 105% o- ed steam flow control li irresponding to different core flows. The ratio o t MCPR 4 cale ed at a given point of core flow, divided by the operating lim CPR, A gfermines the X,.

]

i i _

At THERMAL POWER levels less than or eoual to 25% of RATED THERMAL POWER, j the reactor will be operating at minimum recirculation pump speed ard the 1 moderator void content will be very small. For all designated control rod patterns which may be employed at this point, opere. ting plant experience indicates tMt the resulting MCPR value is in excess of requirements by a

, considerable margin. During initial start-up testing of the plant, a MCPR j~ evaluation will be made at 25% of RATED THERMAL POWER level with minimum {

4* recirculation pump speed. The MCPR margin will thus be demonstrated such that I future MCPR evaluation below this power level will be shown to be unnecessary. I e- The daily requirement for calculating MCPR when THERMAL POWER is greater than !

or equal to 25% of RATED THERMAL POWER is sufficient since power distribution i shifts are very slow when there have not been significant power or control rod

changes. The requirement for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in t

THERMAL POWER or power shape, regardless of magnitude, that could place

operation at a thermal limit.

i LA SALLE - UNIT 2 B 3/4 2-6 Amendment No. 41 4

- _ _ . . _ . - - - _ - _ _ _ .m. .____ . _ _ _ _ _ _ _ _ .__ _ - _ _ _ _ _ _ _ _ _ . . _

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSEMBLIES D#E 5.3.1 The reactor core shall contain 764 fuel assemblies,!.cith-::ch--feel assembly-containing 52- fuel rede end i. e water-rods-clad-with-Z4rcaley -2.

Each-fuel red eball haEEL.a--nominal-Set 4Ve #"o1 lennth of 150 inch 45. The initfai cere loaditag-sheH-have : maximum-average-enr4chment-of_1,89-weight p^r:;nt U-235.

enen Releed fuel shah-be :,imilar sw-phycical dec4gn-to-the-4nitial-

];;djp g, CONTROL ROD ASSEMBLIES cPaOM M V

5.3.2 The reactor core shall contain 185 control rod aisamblies. -There ere-two-poccihla tvnoc nf centre! reds, one :-onsethg-of-a-eetsc4 form-array-of stainless 5+o01 tuhoc COntti n $ ny ~ 143 --ioGhe6-Of-bc ron carbiderB Cr-powder,.

surround;d by e U UttTorm snaped steinless-steel-sheeth, :nd thbsecond-type-contein: 143 'nchac nf mbsc-rba meterial- of which the first-6-inches er hafnium :nd th*

conhol MG/crief s/10// d6 borM corbide hw&re~sinder cAa// ha ve a nomina (By C e is B])C- Tler Adnium nieYe/. & esdrol r 5.4 REACTOR COOLANT SYSTEM and0 ab.soder le#8JA o f /+3 he c4&

DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolant system is designed and shall be maintained:

a. In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of:

1. 1250 psig on the suction side of the recirculation pumps.

l 2. 1650 psig frot the recirculation pump dischar;e to the outlet l side of the discherge shutoff valve.

3. .% 00 psig from the discharge shutoff valve to the jet pumps.

c. For a temperature of 575*F.

y0LUME 5.4.2 The total watar and steam volume of the reactor vessel and recirculation system is + 21,000 cubic feet at a nominal T,y, of 533'F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1.1-1.

LA SALLE - UNIT 2 5-4 Amendment No. 41

44 3;

.i INSERT D l

-a Each. assembly consists of a matrix of Zircalloy clad fuel rods with a an initial composition'of.slightly enriched uranium dioxide. U02. Fuel ,

assemblies shall be limited to those fuel designs approved for use in BWR's. 4 4

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h .. ADMINISTRATIVE CONTROLS i- @~' "

B . . . .

Semiannual Radioactive Effluent Release Report (Continued) u

(; The radioactive affluent release report shall include tin following ,

f information for each type of solid waste shipped offst during the -l

~ report period: l i- j

a. Container volume,

)-

k b. Total. curie quantity (specify whether determined by 1 measurement or estimate),

p'

c. Principal radionuclides (specify whether determined by.

[j( measurement or estimate),

d.; Type of waste'(e.g., spent resin, compacted dry wasta, f evaporator bottoms),

p

e. . Type of container (e.g. , LSA, Type A, Type 8. Large Quantity),

I. aad i

i f. Solidification agent (e.g. , cement, urea formaldehyde).

The radioactive effluent release reports shall include unplanned.

[

1- releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

[

n s The radioactive effluent release reports shall include any changes d hy to the PROCESS CONTROL PROGRAM (PCP) made during the. reporting period.

S. Monthly Operating Report l

Roetina reportu o* 9perati g statistics and shutdawn' experience, kl inclodice cocumenttior of.all challenges to safety / relief valves, (e

j.

shall be subLitthd on a conthiy basis to tne GiNetor, Office of

a. Ninagement hformation stad Frogram Cor.trui,, US huclear Regu?atory Coesission, Washingtm, DC 10555, with a copy of the appropriate

] Regions 1 Offica, to arrive no later than the 15th of each month j .

folicAng.tnc calendar mor.th covsrad by the report.

Any c:hanges M the OFFSITE DOSE CALCULATION MANUAL shall be submitted

[ with.the Morthly Opert, ting Repo;t wttain 90 days in.waiG that change (s) 3 wari made effective. In Addition, a report of any major changes to 6, the raufoactive waste treatment systems shall be sutatitted with the i Monthly Operating Report for the period in which the evaluation was

g. _ reviewed and accepted by Onsite Review and Investigative Function.

(j B. Del'eted. gg l s .

H.

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LA SALLE - UNIT 2 6-24 Amendment No.11 Uf g

l _. _ - _ - _ _ -

INSERT B-

6. CORE OPERATING LIMITS REPORT '[

a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any- j remaining part of'a reload cycle for the following:

(1) The Average _ Planar Linear Heat Generation Rate (APLHGR) for :

Technical Specification 3.2.1. ;

(2) The minimum Critical Power Ratio (MCPR) (including 20% scram time, tau ('T), dependent MCPR limits, and Kf core flow MCPR ,

adjustment factors) for Technical Specification 3.2.3.

(3) The Linear Heat Gener tion Rate (LHGR) for Technical

-Specification 3.2.4.

1 (4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2.

b. The analytical methods used to determine the core operating limits j" shall be those previously reviewed and approved by NRC in NEDE-240ll-P-A, General Electric Standard Application for Reactor

, Fuel (latest approved revision). ,

c. The core operating limits shall be determined so that hil rpplichble limits (e.g.. , fuel thermal-mechanical lirait? such as shutnan mergin, and transient and accident analysis limits) of the safet) analysis urt- met,

d. The CORE OPERATING LIMITS REPORT, includiag any mid-cycle revisi.'ns or supplements thereto, shall be a ovided upon ! 3suance, for each reload cycle, to the U.S. Nuclear Regulatory Commissior. Document ')j Control Desk with copies to the Regional Administrator and Reside 1t I n spe c tor., J l

1 I

i 0189T:19

\___==______- l

ge q ~% .

AUAQMDIL3 SIGNIFICANT HAZARD EVALUATION-Commonwealth Edison proposes to amend facility Operatin'g License NPF-11.(LaSalle County Unit 1) and NPF-18 (LaSalle County Unit 2) to support the licensing under 10CFR50.59 of Unit 1 Cycle 4. Unit 2. Cycle 4, and future LaSalle County operating cycles. The purpose of the. proposed revision is to.

remove cycle-specific power distribution limits consistent with Generic Letter 88-16_and to appropriately revise other sections to facilitate future. reload licensing. Although a general description of the changes follows, more.

detailed discussion of the: changes and their technical bases can be found in

-Attachment 1.

DESCRIPTION OF AMENDMENT REQUESI The proposed changes in Appendix A Technical Specifications of racility Operating ~ License NPF-11 (LaSalle County Unit 1) and NPF-18 (LaSalle County Unit 2) are.as follows:

a) A definition for the Core Operating Limits Report is added.

b) The limits!for Average Planar Linear Heat Generation Rate, Minimum Critical Power Ratio, Linear Heat Generation Rate, and Rod Block Monitor Upscale'Setpoints, are removed and references to the Core Operating Limits Report art inserted.

c) The last two paragrar>hs in the bases for,'" Average *lanar Linear Fett Generation Rate", art yaoved. These pragraphs discuss how APLHGR values are calculated for th*. specific fuel types. Inserted is the statement.

"The calculational procedure used to establish APLHGR values uses GE c;1culational models which are consistent with the requirements of Appendix N,10 Cf R M". This chage is administrative in nature and does ;

not change the intrit of the basis, i d) Several paragraphs in the Bases for " Minimum Critical Power Ratio" are removed. These paragraphs contain unnecessary detali concerning MCPR-limit derivation as a function of non-rated flow conditions (Kf), whether Rod Withdrawal Error is the limiting cycle-specific event, and equipment

out-of-service analyses verification.

e) A new administrative reporting requirement entitled, " Core Operating Limits Report", is added to the existing reporting requirements.

-f) Some details are removed from the design features description of the fuel and control rod assemblies which may change during transition to approved, advanced designs.

0189T:15

- - ___ _ _-____ _ a

l l

MSIS FOR P10EQSED_51GNJFICANT HAZARDS _EVALUAUQ!i Commonwealth Edison has evaluated the proposed Technical Specification and determined that they do not represent a significant hazards consideration.

Based on the criteria for defining a significant hazards established in 10 CFR l 50.92(c), the proposed changes:

1. Do not involve a significant increase in the probability or consequences of any accident previously evaluated because:

a) to No plant protective functions are changed by this amendment. The e) amendment is essentially administrative in nature by removing cycle-specific and fuel bundle type specific power distribution limits from the Technical Specifications and placing them in a separate controlled document, the Core Operating Limits Report (COLR). The MCPR and MAPLHGR Bases section changes are also administrative. NRC approved methods will still be used to analyze reloads of NRC-approved fuel types to determine the results reported in the COLR in accordance with Technical l Specification Section 6.6.A.6. The surveillance requirements for these power distribution limits remain unchanged.

f) Previously analysed accidents are also not affected by the proposed chances to the fuel assemblies and control rod assemblies l sections of the design features. Since the changes do not allcw the use of any types of fuel assemblies and control rod essemblies that have not been previously evaluated for use ir, the reactor core. These changes are being made to eliminate the daign characteristics from the Technical Specifications since this l Deformation is fuel assembly and control rod assemoly sncific. I lise of new types of fuel assenblies and control rod astemblies is eval >Jated in accordarte with 10 CFR 10.92(c) and 10 CFR 50.59.

l

[ 2. Do not create the possibility of a new or different kind of accident from

i. any accident previously evaluated because:

l a) to It will not result in changes in operating mode or configuration I

e) of the reactor core or any plant components, thus the change does not create any new accident type. The current spectrum of reactor transients and accidents analyzed remains unchanged.

l l

0189T:16

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f): The' proposed change.to the fuel assemblies and control rod assemblies sections of the design features does not allow the use of any types of fuel assemblies and control rod assemblies that have not been previously evaluated'for use'in theireactor core.

L, thus.the' change does.not create-any new or_different accident mode. 'The current spectrum of reactor transient and accident analyses _ remain unchanged. As new designs are developed, appropriate-reviews will be conducted as required by 10 CFR 50.92(c) and.10 CFR 50.59 to ensure no new or different accident scenario is created.

l

3. Do not involve a significant reduction in the margin of safety because:

a) to The' methods for determining' fuel / core limits are not affected by e) this: change. ~No~ revisions in the parameters required to verify operation with existing safety-bases are involved (i.e., LHGR,.

MCPR and MAPLHGR are still the limits). No safety. limits are changed. The. change only removes cycle-specific'and fuel bundle

' type specific power distribution limits from the Technical Specifications and incorporates'these limits in the COLR. The plant will, continue to be operated under these same power distribution limits,'which will be calculated using NRC approved methods. The MCPR and MAPLHGR Bases section changes are

. administrative'and therefore no' impact on margin of safety.

f) The proposed change to the fuel assemblies and control rod assemblies section of the decign features does not allow the use of any types'of fuel assemblies and control rod assemblies that have'not been previously evaluated for use in the reactor core.

Any new type of fuel assemblies and' control rod assemblies will be evaluated before use in the reactor core. This. change only removes specific design characteristics of the' fuel assemblies and control rod assablies from 'he Technical Specifications, and does not impact any cbcrent Limi,iIg Condition for Operation or Surveillance Requirement.

Based on the'above discussion, Commonwealth Edison concludes that the "

proposed amendments do not represent a significant hazards consideration.

Ol89T:17

,e d'

I AIIACliMERL4 1 EXA!ELLCOBLOfERATING LIMITSJEEORIS A. LASALLE COUNTY STATION UNIT 1 RELOAD 2 (CYCLE 3)

B. LASALLE COUNTY STATION UNIT 2 RELOAD 2 (CYCLE 3) l l

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f REFERENCES .............................................'......... 11i o

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!: LIST OF FIGURES .......................'.......................... iv k

F- LIST OF TABLES .................................................. v s

1. 0 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (3/4.2.1) ...... 1-1 T-

- 1.1 ' ' TECHNICAL SPECIFICATION REFERENCE . . . . . . . . . . . . . . . . . 1-1

1. 2 DESCRIPTION .....'.................................. .1-1

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O 2. 0 ' MINIMUM CRITICAL POWER' RATIO-(3/4.2.3) ................... 2-1 l-

! 2.1 - TECHNICAL SPECIFICATION REFERENCE .................- 2-1

2. 2 - DESCRIPTION ....................................... 2-1

3. 0 LINEAR HEAT GENERATION RATE (3/4.2.4) ..................... 3-1 l

4 3.1 TECHNICAL SPECIFICATION REFERENCE'................. 3-1 I 3. 2 - DESCRIPTION ...................................... .3-1 o.

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4. 0 ' CONTROL ROD WITHDRAWAL. BLOCK INSTRUMENTATION (3/4.3.6) ....- 4-1) i: . 4.1 TECHNICAL SPECIFICATION REFERENCE ................. 4-1

j. 4. 2 . DESCRIPTION ........................................-4-1 t

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,.. 1. Commonwealth Edison Company Docket No. 50-373, LaSalle County j Station,' Unit i Facility Operating License, License No. NPF-il, a

8.

j- 2. Let er.from D. M. C.utchfield to All Power Reactor Licenses and 1 App 'icants, Generie Letter 88-16; Concerning the Removal of j Cyc 2-Specific Parameter Limits from Technical Specifications.

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FIGURE TITLE / DESCRIPTION PAGE 1.2-1 MAXIMUM AVERPGE PLANAR LINEAR HEAT GENERATION RATE 1-2 (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE FUEL TYPES 8CRB071, 8CRB176, BCRB219 1.2-2 MAPLHGR VS. AVERAGE PLANAR EXPOSURE FOR FUEL TYPE 1-3 BP8CRB299L 1.2-3 MAPLHGR VS. AVERAGE PLANAR EXPOSURE FOR FUEL TYPES 1-4 BC301A AND BC320B 2.2-1 POWER DISTRIBUTION LIMITS, MCPR VS.1: AT RATED FLOW, 2-2 RBM SETPOINTS.

2.2-2 . POWER DISTRIBUTION LIMITS, MCPR VS.1C AT RATED FLOW, 2-3 EOC-RPT INOPERABLE AND MAIN TURBINE BYPASS INOPERABLE 2.2-3 POWER DISTRIBUTION LIM 7TS, Kf FACTOR 2-4 I

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l~ 1.1 TECHNICAL SPECIFICATION REFERENCEt S Technical " specification 3.2.1.

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1.2 DESCRIPTION

The Maximum' Average Planar Linear Heat Generation Rates (MAPLHCR)

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$ The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)-

versus Average Planar Exposure for fuel type BP8CRB299L are 4 determined from Figure 1.2-2.

i.

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] 2. 2 DESCRIPTION:

4 j a. Sinale Recirculation Looo Operation q The MCPR limit when in Single Recirevlation Loop Operation is determined from Figure 2.2-1 (Curve A for a RBM setpoint of i 106% or Curve B for a.RBM setpoint of 110%) plus 0.01, times

) the Kf determined from Figure 2.2-3.

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b. Two Recirculation Loco Doeration

, The MCPR limit when in Dual Recirculation Loop Operation is determined from Figure 2.2-1 (Curve A for a RBM setpoint of 4

106% or Curve B for a RBM setpoint of 110%) times the Kf determined from Figure 2.2-3.

)- c. Two Recirculation Loco Doeration with Main Turbine Bvoess Inocerable The MCPR limit when in Dual Recirculation Loop Operation with the Main Turbine Bypass Inoperable is determined from Figure

2.2-2 times the Kf determined from Figure 2.2-3 per Technical i . '?j ; Specification 3.7.10 (any RBM setpoint determined per I Technical Specification Table 3.3.6-2 may be used).

I d. Two Recirculation Loop Doeration with End-of-Cvele

Recirculation Dumo Trio System Inocerable The MCPR limit when in Dual Recirculation Loop Operation with the End-of-Cycle Recirculation Pump Trip System Inoperable is

, determined from Figure 2.2-2 times the Kf determined from Figure 2.2-3, for two recirculation loop operation, with the end-of-cycle recirculation pump trip system inoperable as 4 directed by Technical Specification 3.3.4.2 (any RBM setpoint determined per Technical Specification 3.3.6-2 may be used).

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3. 0 LINEAR HEAT GENERATION RATE (3/4,2,4) 3.1 - TECHNICAL SPECIFICATION

REFERENCE:

17

/1 EMF : Technical Specification 3.2.4. .

3. 2 DESCRIPTION:

! a. The LHGR limit is 13.4 kw/ft for fuel types:

b 1. 8CRB176 P. 8CRB219 l

3. LPBCRB299L

b. The LHGR ' limit is 14.4 kw/ft for fuel types: -

l 1. BC301A

2. BC320B i

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", j T. 4. 0 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION (3/4.3.6) s

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( 4.1 TECHNICAL SPECIFICATION

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2 The Rod Block Monitor Upscale Instrumentation Setpoints are l} -

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CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS E

h TRIP' FUNCTION- TRIP SETPOINT ALLOWABLE VALUE 4

j 1. ROD BLOCK MONITOR A. UPSCALE A .

,] 1) TWO RECIRCULATION ft LOOP OPERATION B

fO) a. When using the MCPR 1 0.66 W + 37%** 1 0.66 W + 40%**

LCD from Curve A of Tl Figure 2.2-1 or the g curves from Figure 2.2-2.

3

?i b. When using the MCPR 1 0.66 W + 41%** 1 0.66 W + 44)**

] LCO from Curve B of Q Figure 2.2-1 or the curves from Figure 2.2-2.

(i A

l 2) SINGLE RECIRCULATION

LOOP OPERATION 8

a. When using the MCPR 1 0.66 W + 31.7%** I 0.66 W + 34.7%**

q LCD from Curve A of

Figure 2.2-1.

b. When using the MCPR 1 0.66 W + 35.7%** 1 0.66 W + 38.7%**

! LCD from Curve B of Figure 2.2-1.

i 3 ** Clamped, with an allowable value not to exceed the allowable value for recirculation loop flow (W) of 100%.

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EARE REFERENCES'...................................................... iii g

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' LIST OF-FIGURES ................................................. iv L LIST OF TABLES ..................................................- v

1. 0 ' AVERAGE PLANAR LINEAR HEAT GENERATION RATE (3/4.2.1) .....: 1 l J .1.1 - ' TECHNICAL ~ SPECIFICATION REFERENCE ................. 1-1

1. 2 . DESCRIPTION ........................................ 1-1.

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2. 0 MINIMUM CRITICAL POWER RATIO (3/4.2.3) ................... 2-1

2.1 TECHNICAL SPECIFICATION REFERENCE ................. 2-1

2. 2 DESCRIPTION ....................................... 2-1

3. 0 LINEAR HEAT GENERATION RATE 13/4.2.4) ..................... 3-1 r

3.1 TECHNICAL SPECIFICATION REFERENCE .........'........ 3-1

3. 2 DESCRIPTION ....................................... 3 '

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4. 0 - CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION (3/4.3.6) ... 4-1

l. 4.1 TECHNICAL SPECIFICATION REFERENCE ................. 4-1 l 4. 2 DESCRIPTION ....*................................... 4-1 L.

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. ., 1.- Commo'nwealth Edison Company Docket No. 50-374, LaSalle. County.

[ Station,' Unit 2 Facility Operating License, License No. NpF-11.;

t:

i:i- . 2.. Letter from D..M. Crutchfield to All power Reactor Licenses and I' Applicants,'. Generic Letter 88-16;. Concerning. the Removal of-

Cycle-Specific parameter Limits from Technical. Specifications.-

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[ FIGURE TITLE / DESCRIPTION PAGE W.

j 1.2-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1-2 W (MAPLHGR) VS. AVERAGE PLANAR EXPOSURE, INITIAL CORE fi FUEL TYPES 8CRB071, 8CRB176, 8CRB219 1.2-2 MAPLHGR VS. AVERAGE PLANAR EXPOSURE FOR FUEL TYPE 1-3 p BP8CRB299L 1.2-3 MAPLHGR VS. AVERAGE PLANAR EXPOSURE FOR FUEL TYPES 1-4 L BC300D AND BC320C t ..

2.2-1 POWER DISTRIBUTION LIMITS, MCPR VS. T AT RATED FLOW, 2-2 RBM SETPOINTS k*

'2.2-2 POWER DISTRIBUTION LIMITS, MCPR VS. T AT RATED FLOW, 2-3

, EOC-RPT INOPERABLE AND MAIN TURBINE BYPASS INOPERABLE 9

, 2.2-3 POWER DISTRIBUTION LIMITS, Kf FACTOR 2-4 4

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, . TITLE / DESCRIPTION E.

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E: .4.2-1 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION' ':4-1E '

SETPOINTS

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- -CORE OPERATING LIMfTS REPORT.

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, ' 1. 0 AVERAGE PLAN 4R LINEAR HEAT GENERATION RATE -(3/4.2.1)

' i.1 TECHNICAL SPECIFICATION

REFERENCE:

P h'_?h .

Technical Specification 3.2.1.

K 1. 2 DESCRIPTION:

h I - The Maximum Average Planar L'inear Heat Generation Rates (MAPLHGR) versus Average Planar Exposure for fuel types BCRB071, 8CRB176,

[ .

'and BCRB219 are determined.from Figure 1.2-1.

![ The Maximum Average Planar Linear. Heat, Generation Rate (MAPLHGR) versus Average Planar Exposure for fuel' type BP8CRB299L are p ' determined from. Figure 1.2-2.

[.

i - The Maximum _ Average Planar Linear Heat ' Generation Rates -(MAPLHGR) i_ versus Average Planar Exposure for fuel types BC300D and BC320C are determined from Figure 1.2-3.

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4 CORE OPERATING LIMITS rep 0RT

2. 0 MINIMUM CRITICAL POWER RATIO (3/4.2.3)

L

() 2.1 TECHNICAL SPECIFICATION

REFERENCE:

- Technical Specification 3.2.3.

V

2. 2 DESCRIPTION:

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a. Sinale Recirculation Looo Doeration l

The MCpR limit when in Single Recirculation Loop Operation is i' determined from Figure 2.2-1 (Curve A for a RBM setpoint 'of f 106% or Curve B for a RBM setpoint of 110%) plus 0.01, times the Kf determined from Figure 2.2-3.

, b. Two Recirculation Loop Doeration

! The MCpR limit when in Dual Recirculation Loop Operation is I

determined from Figure 2.2-1 (Curve A for a RBM setpoint of 106% or Curve B for a RBM setpoint of 110%) times the Kf

determined from Figure 2.2-3.

, c. Two Recirculation Loco ODeration with Main Turbine Bycass

! Inocerable The MCPR limit when in Dual Recirculation Loop Operation with the Main Turbine Bypass Inoperable is determined from Figure L

p7 2.2-2 times the Kf determined from Figure 2.2-3 per Technical 3 (;) Specification 3.7.10 (any RBM setpoint determined per 2

Technical Specification Table 3.3.5-2 may be used).

) d. Two Recirculation Looo ODeration with End-of-Cvele Recirculation Pumo " Trio System Inocerable l

j. The MCPR limit when in Dual Recirculation Loop Operation with

the End-of-Cycle Recirculation Pump Trip System Inoperable is 5 determined from Figure 2.2-2 times the Kf determined from l Figure 2.2-3, for two recirculation loop operation, with the end-of-cycle recirculation pump trip system inoperable as directed by Technical Specification 3.3.4.2 (any RBM setpoint determined per Technical Specification 3.3.6-2 may oe used).'

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CORE' OPERATING L7MITS REPORT.

1:3. 0 ' LINEAR HEAT GENERATION RATE (3/4.2.41' r ,

3.1 - TECHNICAL SPECIFICATION

REFERENCE:

. Technical. Specification 3.2.4.

3.' 2 . DESCRIPTION:

a. The LHGR limit is 13.4 kw/ft for fuel types: -

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2. 8CRB176

2. 8CRB219 3.- BP8CRB299L

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k b.- The LHGRl limit' is 14.4 kw/ft for fuel types:

l-0 1. BC300D

2. BC3POC

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.LA SALLE - UNIT 2 3-1 APRIL 1989 s

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[ CORE OPERATING LIMITS REPORT i

p* 4. 0 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION (3/4.3.6)

(u), ' 4.1 TECHNICAL ~ SPECIFICATION

REFERENCE:

g Technical Specification' Table 3.3.5-2.

4. 2 - DESCRIPTION

I .

[- .The Rod' Block Monitor Upscale Instrumentation Setpoints are

f. determined from the relationships shown in Table 4.2-1.

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[

s; TABLE 4.2-1 CONTRCL ROD WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS f

)~ .

?' TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE 1.~ ROD BLOCK MONITOR A. UPSCALE i

1) TWO RECIRCULATION LOOP OPERATION

(h, a. When using the MCPR 1 0.66 W + 37%** 1 0.66 W + 40%**

l' LCO from Curve A of

^'

Figure 2.2-1 or the curves from Figure 2.2-2.

b. When using the ACPR 1 0.66 W + 41%** 1 0.66 W + 44%**

- LCO from Curve B of j Figure 2.2-1 or the

}- curves from Figure 2.2-2.

Y

2) SINGLE RECIRCULATION LOOP OPERATION 3..

a. When us.lng the MCPR 1 0.66 W + 31.7%** 1 0.66 W + 34.7%**

LCD froia Curve A of C Figure 2.2-1.

j -. b. When using the MCPR 1 0.66 W + 35.7%** I 0.66 W + 38.7%**

k LCO from Curve B of L Figure 2.2-1.

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b, k ** Clamped, with an allowable value not to exceed the allowable value for i recirculation loop flow (W) of 100%.

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LA'SALLE - UNIT 2 4-1 APRIL 1989 9

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ML20246E354

Contents

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SUMMARY

1.2 DESCRIPTION

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SUMMARY

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ML20246E354

Text

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SUMMARY

1.2 DESCRIPTION

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SUMMARY

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