Text

Proposed Tech Specs Re Core Operating Limits Rept & cycle- Specific Parameter Limits
	ML20011F657
Person / Time
Site:	Limerick
Issue date:	02/23/1990
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20011F655	List: ML20011F654; ML20011F657;
References
	NUDOCS 9003070091
	Download: ML20011F657 (39)
	v • d • e

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t ATTACHMENT 2 LIMERICK GENERATING STATION Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 ,

PROPOSED TECHNICAL SPECIFICATIONS CHANGES List of Attached Change Pages Unit 1 ' Unit 2 i i y v vi- vi ~

xxvii xxvii 1 1-2 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-8a 3/4 2-10 3/4 2-9 3/4 2-12 3/4 2-10 3/4 3-60-' 3/4 2-12 3/4 3-60a 3/4 3-60 B 3/4 2-1 3/4 3-60a B 3/4 2-4 B 3/4 2-1 B 3/4 2-5 B 3/4 2-2 6-18 B 3/4 2-4 6-18a B 3/4 2-5 6-18 6-18a

'j 9003070091 900223 h PDR ADOCK 05000352 #p P. PDC ,

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INDEX

., ' m ' DEFINITIONS

-SECTION

1. 0' DEFINITIONS PAGE t

'1.1- ACTI0N....................................................... 1-1 1.2 AVERAGE PLANAR EXP0SURE...................................... 1 .1.3 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 1-1 1.4 CHANNEL CALIBRATION.......................................... 1-1 1.5- CHANNEL CHECK................................................ 1-1

.1.6 CHANNEL FUNCTIONAL TEST...................................... 1-1 .

l' . 7 C0RE' ALTERATION.............................................. 1-2 1.7a' CORE OPERATING LIMITS REP 0RT................................. 1-2 1.8 CRITICAL POWER RATI0......................................... 1 1.9~ DOSE EQUIVALENT I-131........................................ l'- 2 1.10 'E-AVERAGE DISINTEGRATION ENERGY.............................. 1-2 1.11 EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME........... 1-2 1.12 END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME.... 1-2

-1.13 FRACTION OF LIMITING POWER DENSITY........................... 1-3 1.14 FRACTION OF RATED: THERMAL P0WER.............................. 1-3

.1.15 1 FREQUENCY N0TATION........................................... 1-3 1.16 IDENTIFIED LEAKAGE........................................... 1-3 1.17 ISOLATION. SYSTEM RESPONSE TIME.........,...................... 1-3' 1.18 LIMITING CONTROL R0D PATTERN................................. 1-3 1.19 ' LINEAR HEAT GENERATION RATE.................................. 1-3 1.20 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 1.21 MAXIMUM FRACTION '0F LIMITING POWER DENSITY................... 1-4 l

LIMERICK - UNIT 1 i

. INDEX-

' .7

,. LIMITING CONDITIONS FOR OPERATIONS AND SURVEILLANCE REQUIREMENTS

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SECTION PAGE

'3/4.0 APPLICABILITY................................................... 3/4 0-1 3/4.1- REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.......................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability....................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times..................... 3/4 1-6 Control Rod Average Scram Insertion Times..................... 3/4 1-7 Four Control Rod Group Scram Insertion Times.................. 3/4 1-8 Control Rod Scram Accumulators. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 Control Rod Drive Coupling.................................... 3/4 1-11 Control Rod Pos i t ion Indication. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , .3/4 1-13 Control Rod Drive Housi ng Support. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-15 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Rod Worth Minim 1zer........................................... 3/4 1-16 Rod Block Monitor............................................. 3/4 1-19 .

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-'19 Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements.............................. 3/4 1-21 Figure 3.1.5-2 Deleted...(LEFT BLANK INTENTIONALLY)...... 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................... 3/4 2-1 Information on pages 3/4 2-2 thru 3/4 2-6c has been INTENTIONALLY OMITTED, refer to note on page 3/4 2-2...... 3/4 2-2 1 LIMERICK - UNIT 1 v

INDEX t-

-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) 3/4 2.2 APRM SETPCINTS.......................................... 3/4 2-7 3/4-2.3 MINIMUM CRITICAL POWER RAT 10............................ 3/4 2-8 Information on pages 3/4 2-10 thru 3/4 2-11 has been INTENTIONALLY OMITTED, refer to Note on page 3/4 2-10... 3/4 2-10 3/4.2.4 LINEAR HEAT GENERATION RATE............................. 3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............... 3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation..................... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times...................... 3/4 3-6 Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements........................ 3/4 3-7 LIMERICK UNIT 1 vi

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ADMINISTRATIVE CONTROLS.

SECTION PAGE 6.5.2. NUCLEARREVIEWBOARD(NRB)

Function..................................................... 6-9 Composition.................................................. 6-9 Alternates.................................,................. 6-10 Consultants.................................................. 6-10 Meeting Frequency............................................ 6-10 Quorum....................................................... 6-10 Review....................................................... 6-10 Audits....................................................... 6-11 Records...................................................... 6-12 6.6 REPORTABLE EVENT ACTI0N.......................................... 6-12 6.7 SAFETY LIMIT VIOLATION........................................... 6-12 ,

6.8 PROCEDURES AND PR0 GRAMS.......................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1

-ROUTINE REP 0RTS.............................................. 6-15 -

Startup Report.............................................. 6-15 Annual Reports.............................................. 6-15 Monthly Operating Reports................................... 6-16 L

1 Annual Radiological Environmental Operating Report.......... 6-16 l

Semiannual Radioactive Effluent Release Report.............. 6-17 CORE OPERATING LIMITS REP 0RTS............................... 6-18a 6.9.2 SPECIAL REP 0RTS............................................. 6-18a

! 6.10 RECORD RETENTION............................................... 6-19 6.11 RADIATION PROTECTION PR0 GRAM................................... 6-20 1

6.12 HIGH RADIATION L

l AREA............................................' 6-20 LIMERICK - UNIT 1 xxvii

in DEFINIJ10NS .

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. CCRE ALTERATION l ,

l1.7L-CORE ALTERATION shall be the addition, removal, relocation or movement of fuel, sources.,

or reactivity-controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs IRMs, TIPS, or special movable detectors

.is not considered a CORE ALTERATION. Susoension of CORE' ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that ptovides the core operating limits for the current operating reload cycle. These cycle-specific core-operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO

,1. 8 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 cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUlVALENT l-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which

.alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

Y-AVERAGE DISINTEGRATION ENERGY 1.10 Y shall be the average, weighted in proportion to the concentration of each

- radionuclide 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 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function i.e., the valves

. etc.

travel to their required positions, pump discharge pressures reach their required values, Times shall include diesel generator starting and sequence loading delays where applicable.

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

total steps such that the entire response time is measured.

. END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME

-1.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be t to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a. Turbine stop valves, and

b. Turbine control valves. ,

l LIMERICK - UNIT 1 1-2

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3/4.2POWERDISTRIBUTIONLIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE -

LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of

. fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be within

' limits based on applicable APLHGR limit values which have been determined by approved methodology for the respective fuel and lattice types for two recirculation loop operation. When hand calculations are required, the APLHGR for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limiting value for the most limiting lattice (excluding natural uranium) as shown in the CORE OPERATING LIMITS REPORT (COLR). During single loop operation, the APLHGR for each fuel type- ,

shall not exceed the above values multiplied by the reduction factors shown in the !

COLR. ;

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

1 With an'APLHGR exceeding the limiting value, 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 . '

1 SURVEILLANCE REQUIREMENTS _

i 4.2.1 All APLHGRs shall be verified to be equal to or less than the limiting value ,

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

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 R00 PATTERN for APLHGR,

d. The provisions of Specification 4.0.4 are not applicable.

' LIMERICK - UNIT 1 3/4 2-1

i: ' of Figures on pages' 3/4 2-2 thru 3/4 2-6c have been removed from Technical Specifications, and relocated to the CORE OPERATING LIMITS REPORT.

Technical Specifications pages-3/4 2-3 thru 3/4 2-6c have been INTENTIONALLY OMITTED.

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F . -' . .? PDWER DISTRIBUTION LIMITS y 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION ;t 13.2.3' The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater

.than the MCPR limit times the Kr, both values shown in the CORE OPERATING LIMITS REPORT,-provided that the end-of-cycle recirculation pump i trip (E00-RPT) 'l system is OPERABLE per Specification 3.3.4.2, with: '

T = (Tave TB)-

TA - TB where: ,

TA =_0.86 seconds, control rod average-scram insertion ,

time limit to notch 39 per Specification 3.1.3.3, I

= N TB 1 )ija(0.016),

0.672 + 1.65 ( n I

N j -

i=1 i

n E

Tave " i=1 i'i e n

E N-g i=1 i 3 ^

n -=

number of surveillance tests performed to date in~ cycle, N1 = number _of active control rods measured in the ith surveillance test.

.. t Tj = . average scram time to notch 39 of all rods measured in the'ith surveillance test, and

_H_=

1 total number of active rods measured in Specification ;

4.1.3.2.a. -

APPLICABILITY:

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

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' .1 ' POWER DISTRIBUTION LIMITS l- _

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i.IMITINGCONDITIONFOROPERATION-(Continued)

ACTION

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue provided that..within-1

, hour, MCPR is-determined to be greater than or equal to the MCPR limit as a function of the average scram time (shown in the CORE OPERATING LIMITS

' REPORT) E0C-RPT inoperable curve, times the Kf shown in the CORE p 0PERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit shown in the' CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes-and-restore MCPR to within.the required-limit within 2 hoers or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER W M in the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS -4.2.3 MCPR, with:

i

a. .t = 1.0 prior to performance of the initial scram time measurements for-the cycle-in accordance with Specification 4.1.3.2, or-

b. t as defined in Specification 3.2.3 used.to determine the limit .

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test '!

required test required by Specification 4.1.3.2,- .

shall be determined to be equal to or greater than the applicable MCPR limit I determined from the CORE OPERATING LIMITS REPORT.

l

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

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 R0D PATTERN for MCPR.

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d. The provisions of Specification 4.0.4 are not applicable.

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LIMERICK - UNIT 1 3/4 2-9

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r Figures on pages 3/4 2-10 thru 3/4 2-11 . .

L have been removed from Technical Specifications, and relocated to the CORE OPERATING LIMITS REPORT. ,

l Technical Specifications pages 3/4 2-10s thru 3/4 2-11 ,

have been INTENTIONALLY OMITTED. '

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4. > POWER DISTRIBUTION LIMITS h 4-

~3/4.2.4 LINEAR HEAT GENERATION RATE !

LIMITING CONDITION FOR OPERATION p !

[ .l 3.2.4 1he LINEAR HEAT GENERATION RATE (LHGR) for each fuel type shall not exceed the value in the CORE OPERATING LIMITS REPORT.

L >

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

[ equal to 25% of RATED THERMAL POWER.

i ACTION

1 L With the LHGR of.any fuel rod exceeding the limit, initiate corrective action H

.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 ur ,

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 .

SURVEILLANCE REQUIREMENTS l

4.2.4 LHGRs shall be determined to be equal to or less than the limit:

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

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 on a LIMITING CONTROL R0D PATTERN for LHGR. i

d. TM provisions of Specification 4.0.4 are not applicable. '

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1 LLIMERICK - UNIT 1 3/4 2-12 "

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-TABLE 3 3 6-2' .

CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS '* !

' RIP FUNCTION . TRIP SETPOINT ALLOWABLE'VALUE , ,-

1. ROD BLOCK MONITOR **** I-

a. Upscale-

1) During two recirculation loop operation a) Flow Biased * - 5 0.66 W + (N-66)%, with a < 0.66 W + (N-63)%, with a maximum of, maximum of, b) High Flow Clamped < N% < (N+3)%

2) During single recirculation loop operation a) Flow Biased * $ 0.66 W + (N-72)%, with a 5 0.66 W + (N-69)%, with a maximum of, maximum of, b) High-Flow Clamped < N% < (N+3)%

b. Inoperative N.A. N.A.

c. Downscale > 5% of RATED THERMAL POWER > 3% of RATED THERMAL POWER T

- 2. AFRM

a. Flow Biased Neutron Flux - Upscale

1) During two recirculation loop operation < 0.58 W + 50%* < 0.58 W + 53%*

2) During single recirculation loop operation < 0.58 W + 45%* _

< 0.58 W + 48%*

b. Inoperative N.A. M.A.

c. Downscale > 4% of RATED THERMAL POWER > 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale, Startup 312%ofRATEDTHERMALPOWER 314%ofRATEDTHERMALPOWER SOURCE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale < 1 X 105 cp3 .

< .1.6 X 105 cp3 ,

c. Inoperative N.A. N.A.

d. Downscale > 3 cps **

> l.8 cps **

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. N.A

b. Upscale < 108/125 divisions of $ 110/125 divisions of full-scale full scale

c. Inoperative N.A. N.A.

d. Downscale > 5/125 divisions of full scale-

_ -> 3/125 divisions of full scale

, 5. SCRAM DISCHARGE VOLU58E_

a. Water Level-High -< 257' 5 9/16" elevation *** < 257' 7 9/16" elevation

a. Float Switch

.iMERICK - UNIT 1 '

3/4 3-60 ,

._._. _ _ . . . . . . _ . . _ . _ . . . . . _ _ _ - . . _ _ . _ . __ . _ _ _ - . _ _ _ _ _ _ ___ _ _ . _ ~ . _ _ . _ . _

TABLE 3.3.6-2 (Continued) - - -

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS ,

TRIP FUNCTION TRIP SETPOINT . ALLOWABLE VALUE

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale < 111% of rated flow < 114% of rated flow

b. Inoperative N.A. N.A.

c. Comparator 5 10% flow deviation 5-11 % flow deviation

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. N.A.

lhe rod block function varies as a function of recirculation loop drive flow (W). The trip setting of the Average Power Range Monitor rod block function must be maintained in accordance with Specification 3.2.2.

May be reduced provided the Source Range Monitor has an observed count rate and signal-to-noise ratio on or above the curve shown in Figure 3.3.6-1.

Equivalent to 13 gallons / scram discharge volume.

- - - The value of N is shown in the CORE OPERATING LIMITS REPORT in accordance with Specifications 6.9.1.9 thru 6.9.1.12.

t LIMERICK - UNIT 1 3/4 3-60a ,

. ,w , . ., . - , - ~ , , , - - , - , , . ,,..,e , ~_.- . _ _ - . - - - - . _ - -

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3/4.2 POWER O!STRIBUTION LlHITS BASES -

I 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10CFR50.46 and that the fuel design analysis limits specified in NEDE-24011-P-A (Reference 2) will not be exceeded.

Mechanical Design Analysis: NRC approved methods (specified in Reference 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding p:wer history, meet the fuel design limits specified in Reference 2. No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis MAPLHGR limit. 4 LOCA Analysis: A LOCA analysis is performed in accordance with 10CFR50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits comply with the ECCS limits specified in 10CFR50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant.

The MAPLHGR limit as shown in the CORE OPERATING LIMITS REPORT is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR limit.

Only the most limiting MAPLHGR values are shown in the CORE OPERATING LIMITS REPORT for multiple lattice fuel. Compliance with the specific lattice MAPLHGR cperating limits, which are available in Reference 3 is ensured by use of the process computer.

~

The MAPLHGR limits shall be reduced to the value shown in the CORE OPERATING LIMITS REPORT times the two recirculation loop operation limit when in single recirculation loop operation. The constant factor shown in the CORE OPERATING LIMITS REPORT is derived from LOCA analyses initiated from single loop operation to account for earlier boiling transition at the limiting fuel node compared to the standard LOCA evaluations.

1

~ LIMERICK - UNIT 1 B 3/4 2-1

.POWERaDISTRIBUTION LIMITS t

< BASES *

3/4.2.3 MINIMUM CRITICAL POWER RATIO i

The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR,and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial conditions of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction 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 evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed in Reference 2.

The purpose of the Kr factor shown in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow conditions. At less than 100%

of rated flow the required MCPR is the product of the MCPR and the Kr factor.

-The Kr factors assure that the Safety Limit MCPR will not be violated during a flow increase transient resulting from a motor-generator speed control failure. The Kr factors may be applied to both manual and automatic flow control modes.

The Kr factors values shown in the CORE OPERATING LIMITS REP 0ru were developed l generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The Kr factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. .

For the manual flow control mode, the Kr factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. The i ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kr. ,

l LIMERICK - UNIT 1 B 3/4 2-4

'i POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)

For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow.

The Kr factors shown in the CORE OPERATING LIMITS REPORT are conservative for the General Electric Boiling Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit MCPR used for the generic derivation of Kr.

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, ,

the reactor will be operating at minimum recirculation pump speed and the 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 startup testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power icvel will be shown to be unnecessary. 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 shifts are very slow when there have not been significant power or control rod changes. The requirement for calculating MCPR when a limiting control rtd pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

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References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K. NEDE-20566, November 1975.

2. " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (latestapprovedrevision).

3. " Basis of MAPLHGR Technical Specifications for Limerick Unit 1", NE00-31401 February 1987 (as amended).

4. Deleted.

5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 1 Cycle 1. NEDC-31323, October 1986 including Errata and Addenda Sheet No.1 dated November 6,1986.

LIMERICK-- UNIT 1 B 3/4 2-5 t

ADMINISTRATIVE CONTROLS SEMIANNUAL RADIDACTIVE EFFLUENT RELEASE REPORT (Contin 0ed)' -

The Semiannual Radioactive Effluent Release Report to be submitted 60 days af ter January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous ef fluents are given in Regulatory Guide 1.109, Rev.1, October 1977.

The Semiannual Radioactive Effluent Release Reports shall include the following information for each type of solid waste (as defined in 10 CFR Part 61) shipped offsite during the report period:

a. Container volume,

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

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacteddrywaste,evaporatorbottoms),

e. Type of container (e.g., LSA Type A, Type B, large Quantity), and

f. SOLIDiflCATION agent or absorbent (e.g., cement; urea formaldehyde).

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period. .

The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

LIMERICK - UNIT 1 6-18

ADMikISTRATIVE CONTROLS CORE OPERATING LIMITS REPORT 6.9.1.9 Core Operating Limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the CORE OPERATING LIMITS REPORT for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1,

b. The MlhlMUM CRITfCAL POWER RATIO (MCPR) for Specification 3.E.3,

c. The Kr core flow adjustment factor for Specification 3.2.3,

d. The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4,

e. The upscale flow biased Rod Block Monitor setpoint and the upscale high flow clamped Rod Block monitor setpoint of Specification 3.3.6.

6.9.1.10 The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

a. HEDE-24011-P-A " General Electric Standard Application for Reactor fuel" (Latestapprovedrevision).

6.9.1.11 The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS .

limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident-Inspector.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

'llMfRICK - UNIT 1 6-1Ba

i -

INDEX

' DEFINITIONS SECTION 1.0 DEFINITIONS PAGE 1.1 ACTION....................................................... 1-1 1.2 AVERAGE PLANAR EXP05URE...................................... 1-1 1.3 AVERAGE PLANAR LINEAR HEAT GENERATIONRATE................... 1-1 1.4 CHANNEL CALIBRATION .....................................,,,, 31 L

1.5 CHANNEL CHECK................................................ 1-1 t

1.6 CHANNEL FUNCTIONAL TEST...................................... 1-1 .

1.7 CORE ALTERATION.............................................. 1-2 1.7a CORE OPERATING LIMITS REP 0RT................................. 1-2 l 1.8 CRITICAL POWER RATI0......................................... 1-2 1.9 DOSE EQUIVALENT I-131........................................ 1-2 1.10 T-AVERAGE DISINTEGRATION ENERGY.............................. 1-2 1.11 EMERGENCY CORE COOLING SYSTEN (ECCS) TIME...........

RESPONSE 1-2 1.12 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE 1-2 TIME....

1.13 FRACTION OF LIMITING POWER DENSITY........................... 1-3 -

1.14 FRACTION OF RATED THERMAL P0WER.............................. 1-3

-1.15 FREQUENCY N0TATION........................................... 1-3 1.16 IDENTIFIED LEAKAGE........................................... 1-3 1.17 ISOLATION SiSTEM RESPONSE TIME............................... 1-3 1.18 LIMITING CONTROL R0D PATTERN................................. 1-3 1.19 LINEAR HEAT GENERATION RATE.................................. 1-3 1.20 11 LOGIC SYSTEM FUNCTIONAL TEST................................. 1-4 1.21 MAXIMUM FRACTION OF LIMITING POWERDENSITY................,,, 1,4 LIMERICK-UNIT 2 i

mmm INDEX

!= 4, e,

- . LIMITING CONDITIONS FOR OPERATIONS AND SURVEILLANCE REQVIREMENTS L

SECTION PAGE 3/4.0 AP P L I C AB I L I T Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 / 4 0- 1 3/4.1 REACTIVITY CONTROL SYSTEM _S 3/4.1.1 SHU100WN MARG I N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 /4 1 - 1 3/4.1.2 RE ACT IVI TY AN0 Mall ES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability....................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times..................... 3/4 1-6 Control Rod Average Scram Insertion Times..................... 3/4 1-7 Four Control Rod Group Scram Insertion Times.................. 3/4 1-8 Centrol Rod Scram Accumul ators. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-9 Controi Rod Drive Coupling.................................... 3/4 1-11 Control Rod Position Indication. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-13 Control Rod Drive Hou sing Support. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer........................................... 3/4 1-16 Rod B l ock Moni tor. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-18 .

3/4.1.5 STANDBY LIQUID CONT ROL SYSTEM. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements.............................. 3/4 1-21 Figure 3.1.5-2 (LEFT BLANK INTENT!0HALLY)................ 3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................... 3/4 2-1 Information on pages 3/4 2-2 thru 3/4 2-6a has been INTENTIONALLY OMITTED refer to note on page 3/4 2-2...... 3/4 2-2 i

LIMERICK-UNIT 2 y

3 Y, c. INDEX LlHITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS :

SECTION PAGE POWER DISTRIBUTION LIMITS (Continued) l 3/4 2.2 APRM SETP0lNTS.......................................... 3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RAT 10............................ 3/4 2-8 Table 3.2.3-1 Deleted. (INTENTIONALLY LEFT BLANK).. 3/4 2-8a Information on pages 3/4 2-10 thru 3/4 2-11 has been INTENTIONALLY OMITTED, refer to note on page 3/4 2-10... 3/4 2-10 3/4.2.4 LINEAR HEAT GENERATION RATE............................. 3/4 2-12 ;

i 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............... 3/4 3-1 .

Table 3.3.1-1 Reactor Protection System Instrumentation..................... 3/4 3-2 Table 3.3.1-2 Reactor Protection System Response Times...................... 3/4 3-6 >

Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements........................ 3/4 3-7 W

LIMERICK-UNIT 2 vi

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, C.

INDEX

i. y .

ADMINISTRATIVE CONTROLS -

SECTION PAGE 6.5.2 NUCLEAR REVIEW BOARD (NRB)

Function..................................................... 6-9 Composition.................................................. 6-9 Alternates................................................... 6-10 Consultants.................................................. 6-10 Meeting frequency............................................ 6-10 Quorum....................................................... 6-10 Review....................................................... 6-10 Audits....................................................... 6-11 Records...................................................... 5-12 6.6 REPORTABLE EVENT ACT10N.......................................... 6-12 6.7 SAFETY LIMIT V10LAT10N........................................... 6-12 6.8 PROCEDURES AND PR0 GRAMS.......................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REP 0RTS............................................. 6-15 -

Startup Report.............................................. 6-15 Annual Reports.............................................. 6-15 Monthly Operating Reports................................... 6-16 Annual Radiological Environmental Operating Report.......... 6-16 Semiannual Radioactive Effluent Release Report... .......... 6-17 CORE OPERATING LIMITS REP 0RT................................... 6-18a 1 6.9.2 SPECIAL REP 0RTS............................................. 6-18a ,

6.10 RECORD RETENTION............................................... 6-19 i

6.11 RADIATION PROTECTION PR0 GRAM................................... 6-20 l

6.12 HIGH RADIATION AREA............................................ 6-20 LIMERICK-UNIT 2 xxvii l

L

I DO '1Ni IlO6 CORE Ad ERATION

.1.7 CORE ALTERATION shall be the addition, removal, relocati~on or movement of fuel, sourcess or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Normal movement of the SRMs. IRMs. TIPS, or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.

CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.12. Plant operation within these limits is addressed in individual specifications.

CRITICAL POWER RATIO 1.8 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 cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, 1-133. I-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation.of Distance Factors for Power and Test Reactor Sites."

T-AVERAGE DISIWTEGRATION ENERGY

-1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide 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 minutes, making up at least 95% of the total noniodine activity in the coolant.

EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME l

1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time inter when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor i

until the ECCS equipment is capable of performing its safety function i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that tim to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker from initial movement of the associated:

a. Turbine stop valves, and b.. Turbine control valves.

LIMERICK-UNIT 2 1-2

q 2, r.

f

'3/4.2 POWER DISTRIBUTION LIMITS

- 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE --

1 LIMITING CONDITION FOR OPERATION l

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 shown in the CORE OPERATING LIMITS REPORT for two recirculation loop operation. During single loop operation, the APLHGR for each fuel type shall not exceed the above values ,

multiplied by the reduction factors shown in the CORE OPERATING LIMITS REPORT. '

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

to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limiting value, 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 .

w

~

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the limiting value: .

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 R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable,

!IurotCK-UNIT.? 3/4 2-1

' 'f $ . , ..._ !

l i

h l

Figures on pages i 3/4 2-2 thru 3/4 2-6 ;

have been removed from Technical t Specifications, and relocated to the CORE OPERATING LIMITS REPORT.

Technical Specifications pages 3/4 2-3 thru 3/4 2-6a have been INTENTIONALLY OMITTED.

e S

t

[

e b

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LIMERICK UNIT 2 3/4 2-2 1

'y 1

~ POWER DISTRIBUTION LIMITSi - i h"I3/4'. 2. 3 MINIMUMCRITICALPOWERRATIO

~ ~

~

ELIMITING CONDITION FOR OPERATION

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

,, ~Ethe MCPR limit determined using the appropriate figure in the CORE' OPERATING LIMITS !

REPORT- times the Kr shown in the CORE OPERATING LIMITS REPORT, provided that :

the end-of-cycle recirculation pump trip (EOC-RPT)-system is OPERABLE per !

Specification 3.3.4.2, with: !

L l

t t= (Tave T' B) ;

TA - TB ;

!-: where:

TA = 0.86 seconds, control rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, N

TB

=

i -)ija(0.052),

0.688 +' 1.65 [ nN E g i=1 n

E-tave = i=1 Ng tj ,

n Ng E

i=1 n = number of surveillance tests performed to date in cycle, Ni = number of active control rods measured in the ith surveillance. test, Tj = average scram time to notch 39 of all rods measured' in the i th surveillance test, and N1--= total number of active rods measured in Specification 4.1.3.2.a.

APPLICABILITY:

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

LIMERICK-UNIT 2 3/4 2-8

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1

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-r THIS PAGE INTENTIONALLY LEFT BLANK i

i e

I

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1 l-Limerick - Unit 2 3/4 2-8a W

.e,-- , , - * -

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'[0WERDISTRIBUTIONLIMITS LIMITINGCONDITIONFOROPERATION(Continued)

ACTION: ;

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue provided that, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , MCPR is determined to be greater than or equal to the MCPR limit as

",. a function of the average scram time shown in the appropriate figure in the CORE OPERATING LIMITS REPORT, for EOC-RPT inoperable curve, times the Kr shown in the CORE OPERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit shown in the CORE OPERATING LIMITS REPORT, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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 .

SURVEILLANCE REQUIREMENTS '

4.2.3 MCPR, with:

a. t = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or

b. t as defined in Specification 3.2.3 used to determine the limit -

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from the appropriate figure in the CORE OPERATING LIMITS REPORT times '

the Kr shown in the CORE OPERATING LIMITS REPORT.

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 R0D PATTERN for MCPR.

d. The provisions of ;ipecification 4.0.4 are not applicable.

LIMERICK - UNIT 2 3/4 2-9 e _,

.,4 ..

r e

Figures on pages 3/4 2-10 thru 3/4 2-11 have been removed from Technical Specifications, and relocated to the CORE OPERATING LIMITS REPORT.

1 Technical Specifications pages 3/4 2-10a and 3/4 2-11 have been INTENTIONALLY OMITTED.

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l- . 4 l

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!< LIMERICK - IINIT 2 1 ? 10

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!W "" : POWER DISTRIBUTION LIMITS

, =

s

g , 3/4.2.4-- LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION - ,

L !

13.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not. exceed the value.in the CORE OPERATING LIMITS REPORT.

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

equal to 25% of RATED THERMAL POWER. '

t

(.

ACTION:- '

With the LHGR of any_ fuel rod exceeding the limit, initiate corrective action ,

within 15 minutes and restore the LHGR to within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or -

'o reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 i r hours. -

1 i

t SURVEILLANCE REQUIREMENTS i.

4.2.4 LHGRs shall be determined to be equal to or less than the limit: ~

L --

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 on a LIMITING CONTROL ROD PATTERN for LHGR. '

d .- TheprbvisionsofSpecification4.0.4arenotapplicable, i

i k

l !

LIMERICK-UNIT 2 3/4 2-12 l

cad-

TABLE 3,3.6-2 .

~

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS -;

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE . :

1. ROD BLOCK MONITOR **** I

c. Upscale

1) During two recirculation loop operation -

a) Flow Biased * $ 0.66 W + (N-66)%, with a < 0.66 W + (N-63)%, with a maximum of, , maximum of, b) High Flow Clamped < N% < (N+3)%

2) During single recirculation loop operation a) Flow Biased * ,

3 0.66 W + (N-72)%, with a < 0.66 W + (N-69)%, with a maximum of, maximum of, b) High Flow Clamped < N% < (N+3)%

b. Inoperative N.A. N.A.

c. Downscale > 5% of RATED THERMAL POWER > 3% of P.ATED THERMAL POWER

2. APRM

a. Flow Biased Neutron Flux - Upscale

1) During two recirculation loop operation < 0.58 W + 50%* 5 0.58 W + 53%*

2) During single recirculation loop operation 5 0.58 W + 45%* $ 0.58 W + 48%*

b. Inoperative N.A. N.A.

c. Downscale > 4% of RATED THERMAL POWER > 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale, Startup 312% of PATED THERMAL POWER 514%ofRATEDTHERMALPOWER

3. SOURCE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale 51 X 105cps 5 1.6 X 105 cps '

c. Inoperative N.A. N.A.

d. Downscale > 3 cps **

_ > 1.8 cps **

ilMERICK - UNIT 2 3/4 3-60 1

, , , . . . , , , . . - . ~ -#-,- , , , ~ .,,r . . ,, .. .,, , . ~ . , ._ , . , . _ . .W

~

TABLE 3 3.6-2 (Continued)

CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS , !

TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. N.A.

b. Upscale < 108/125 divisions of < 110/125 divisions of full scale full scale

c. Inoperative N.A. .N.A.

d. Downscale > 5/125 divisions of full > 3/125 divisions of ful scale scale

5. SCRAM DISCHARGE VOLUME

a. Water Level-High

a. Float Switch 5 257' 7 3/8" elevation +++ < 257' 9 3/8" elevation

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale 5 111% of rated flow

b. Inaperative $ 114% of rated flow N.A. N.A.

c. Comparator < 10% flow deviation < 11 % flow deviation

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. N.A.

The rod block function varies as a function of recirculation loop drive flow (W). The trip setting of the average power range monitor rod block function must be maintained in accordance ,

with Specification 3.2.2.

For initial fuel loading and start-up the count rate may be reduced to 0.7 cps provided the signal-to-noise ratio is > 2.

Equivalent to 13.56 gallons / scram discharge volume.

- - - The value of N is shown in the CORE OPERATING LIMITS REPORT in accordance with Specifications 6.9.1.9 thru 6.9.1.12.

LIMERICK-UNIT 2 3/4 3-60a .

9 *

-s . , . ~ r x, . ,-. , . ~ - . , , - - - ,. _ , r,,, . . - - ._ ~ _

a

..+ o.

3/4.'2 POWER DISTRIBUTION LIMITS

~

BASES -

The specifications of this section assure that the peak cladding temperature followin 2200@ho@g thespecified sF limit postulated design in 10 CFRbasis loss-of-coolant accident.will not exceed the 50.46.

3/4 2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 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 assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. TheAVERAGEPLANARLINEARHEATGENERATIONRATE(APLHGR)limitasshowninthe CORE OPERATING LIMITS REPORT is the LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in the CORE OPERATING LIMITS REPORT.

The calculational procedure used to establish the APLHGR shown in the CORE OPERATING LIMITS REPORT is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR Part 50. A complete discussion of each code employed in the analysis is presented in Reference 1. Differences in this analysis compared to previous analyses can be broken down as follows,

a. Input Changes

1. Corrected Vaporization Calculation - coefficient's in the vaporization correlation used in the REFLOOD code were corrected. -

2. Incorporated more accurate bypass areas- The bypass areas in the top guide were recalculated using a more accurate technique.

3. Corrected guide tube thermal resistance.

4. Correct head capacity of reactor internals heat modes. ,

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.

2. Incorporate NRC pressure transfer assumption - The assumption used in the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

l LIMERICK-UNIT 2 B 3/4 2-1

p; .

POWOt DISTRIBUTION LIMITS n BASES ;

i AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) c A few of the changes affect the accident calculation irrespective of CCFL. These

[ changes are listed below.

[ a. Input Change y 1. Break Areas - The DBA break area was calculated more accurately. !

b. Model Change 3 t

1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

A list of the significant plant input parameters to the loss-of-coolant accident

. analysis is presented in Bases Table B 3/4.2.1-1.

The MAPLHGR limits shall be reduced to the value shown in the CORE OPERATING LIMITS REPORT times the two recirculation loop operation limit when in single recirculation loop operation. The constant factor shown in the CORE OPERATING LIMITS REPORTS is derived from LOCA analyses initiated from single loop operation to to account for earlier boiling transition at the limiting fuel node compared to the standard LOCA evaluation.

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 neutron flux-upscale scram trip setpoint and flow biased neutron flux-upscale control rod block functions of the APRM instruments must be adjusted to -

ensure that the MCPR does not become less than the fuel cladding Safety Limit MCPR or that-> 1% plastic strain does not occur in the degraded situation. The scram and rod l

block setpoints are adjusted in accordance with the formula in this specification I

when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

l

. LIMERICK - UNIT 2 B 3/4 2-2

[

PDWEft* DISTRIBUTION LIMITS l8ASES' 3/4.2.3 MINIMUM CRITICAL POWER RATIO i

.The required operating limit MCPRs at steady-state operating conditions as r

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

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction 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 evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior transient computer program. The codes used to evaluate transients are discussed in Reference 2.

The purpose of the Kr factor shown in the CORE OPERATING LIMITS REPORT is to I define operating limits at other than rated core flow conditions. At less than 100%

of rated flow the required MCPR is the product of the MCPR and the Kr factor.

The Kr factors assure that the Safety Limit MCPR will not be violated during a

, flow increase transient resulting from a motor-generator speed control failure. The Kr factors may be applied to both manual and automatic flow control modes.

The Kr factors values shown in the CORE OPERATING LIMITS REPORT were developed l generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The Kr factors were derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. .

For the manual flow control mode, the Kf factors were calculated such that for the maximum flow rate, as limited by the pump scoop tube set point and the corresponding THERMAL POWER along the rated flow control line, the limiting bundle's relative power was adjusted until the MCPR changes with different core flows. The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determines the Kr.

l LIMERICK-UNIT 2 B 3/4 2-4

J0 y DISTRIBUTION LIMITS

, BASES MINIMUM CRITICAL POWER RATIO (Continued) for. operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow. ,

The Kr factors shown in the CORE OPERATING LIMITS REPORT are l conservative for the General Electric Boiling Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 ,

operating limit MCPR used for the generic derivation of Kr. ;

At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, ,

the reactor will be operating at minimum recirculation pump speed and the 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 startup testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary. 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 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 THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated.

References:

1. General Electric Company Analytical Model for loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566 November 1975.

2. " General Electric Standard Application for Reactor Fuel", NEDE-24011-P-A (latest approved revision).

3. Deleted.

4. Deleted.

5. Increased Core Flow and Partial Feedwater Heating Analysis for Limerick Generating Station Unit 2 Cycle 1, NEDC-31578P, March 1989 including Errata and Addenda Sheet No. I dated May 31, 1989.

6. General Electric Boiling Water Reactor Extended Load Line Limit Analysis for Limerick Generating Station Unit 2 Cycle 1, NEDC-31677P, March 1989.

LIMERICK-UNIT 2 B 3/4 2-5

" ADMINISTRATIVE CONTROLS

. e

'$EMIANNUAL RAD 10 ACTIVE EFFLUENT RELEASE REPORT (Continued) ..

The Semiannual Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed MEMBER OF THE PUBLIC from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous calendar year to show conformance with 40 CFR Part 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid ,

and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1, October 1977.

The Semiannual Radioactive Effluent Release Reports shall include the following

-information for each type of solid waste (as defined in 10 CFR Part 61) shipped offsite during the report period:

a. Container volume,

b. Total curie quantity (specify whether determined by measurement or estimate). -

c. Principal radionuclides (specify whether determined by measurement or estimate),

d. Source of waste and processing employed (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Type of container (e.g., LSA, Type A. Type B Large Quantity), and 7

f. SOLIDIFICATION agent or absorbent (e.g., cement; urea formaldehyde).

The Semiannual Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting ;

period. .

The Semiannual Radioactive Effluent Release Reports shall include any changes made l during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the ODCM, as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

LIMERICK - UNIT 2 6-18

~

1.

g,y f. l. . ..

e c Mo' e ,

r: ADMINISTRATIVE CONTROLS . :

i n'

. CORE OPERATING LIMITS REPORT F

6.9.1.9' Core operating limits shall-be established prior to each reload cycle, or prior to any' remaining portion of a reload cycle, and shall be documented in the CORE

, 0PERATING LIMITS REPORT for the following-

~ a. The AVERAGE PLANAR LINEAR HEAT GEhERATION RATE (APLHGR) for Specification j" g 3.2.1, i

b.'TheMINIMUMCRITICALPOWERRATIO(MCPR)forSpecification3.2.3, ;

Je c. The Kr core flow adjustment factor for Specification 3.2.3,

d. The LINEAR HEAT GENERATION' RATE (LHGR) for Specification 3.2.4,

-e. The upscale flow biased Rod Block Monitor setpoint and the q scale high flow clamped Rod Block monitor setpcint of Specification 3.3.6.

+ .

6.9.1.10 The' analytical methods used to determine the core' operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

.a. HEDE-24011-P-A'" General Electric' Standard Application for Reactor Fuel" (Latest approvec revision).

3 : 6.9.1.11.The cwe operating limits'sh'all be' determined such that all applicable

. limits (e.g.,_ fuel thermal-mechanical limits, core ther. mal-hydraulic limits. ECCS .

-limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and

! accident analysis 1.imits) of the safety analysis are -met.

6.9.1.12 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or

' supplements, shall be provided upon issuance for each reload cycle to the NRC

? Document Control Desk with copies to the Regional Administrator and Resident !

? Inspector. -

'SPECIAL REPORTS 6.9.'2 Special reports shall be submitted to the Regional Administrator of the ;

. Regional Office of the NRC within the time period specifieci for each report.

i m > LIMERICK-- UNIT 2 6-18a

ML20011F657

Text

References:

References:

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