Text

Amend 3 to License NPF-39,changing Tech Specs to Allow Plant Operation W/Partial Feedwater Heating & W/Increased Reactor Core Cooling Water Flow Rates Up to 105% of Rated Flow
	ML20211F116
Person / Time
Site:	Limerick
Issue date:	02/17/1987
From:	Butler W; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20211F120	List: ML20211F116; ML20211F124;
References
	NUDOCS 8702240502
	Download: ML20211F116 (21)
	v • d • e

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pm Rf Cg UNITED STATES o

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.,qg o NUCLEAR REGULATORY COMMISSION g

I WASHINGTON. D. C. 20555

- c' f fe PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 3 License No. NPF-39 1.

The Nuclear Regulatory Comission (the Comission) has found that i

A.

The application for amendment by Philadelphia Electric Company (thelicensee)datedNovember 17, 1986 as atended on December 22, 1986 and as supplemented on January 2 and January 29, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

Accordingly, the license is amended by changes to the Technical Specifications l

2.

as indicated in the attachment to this license amendment, and paragraph 2.C.(?)

l l

of Facility Operating License No. NPF-39 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 3

, are hereby incorporated into this l

Philadelphia Electric Company shall operate the facility license.

(

in accordance with the Technical Specifications and the Environmental Protection Plan.

ho h 05000352 B70217 p

PDR

-2 3.

The license is further amended by deleting paragraph 2.C(13).

4.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

_/s/ Robert E. Martin for Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing Attachments:

1.

Change to the License 2.

Changes to the Technical Specifications Date of Issuance: February 17, 1987 s

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OG 3M q

/EM A V)

E ' Butler s

ten tin:lb 7/ll /87 1/// /87

// /87 L/j7/87 e

, 3.

The license is further amended by deleting paragraph 2.C(13).

4.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Bhe[

L&

, Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing Attachments:

1.

Change to the License 2.

Changes to the Technical Specifications Date of Issuance:

February 17, 1987 l

l l

1 l

ATTACHMENT TO LICENSE AMENDMENT NO. 3 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 1.

Revised Page 6 of Facility Operating License No. NPF-39 2.

Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf page(s) provided to maintain document completeness.*

Remove Insert y

v*

vi vi 3/4 2-7 3/4 2-7*

3/4 2-8 3/4 2-8 3/4 2-8a 3/4 2-9 3/4 2-9 3/a 2-10 3/4 2-10 3/4 2-10a 3/4 3-59 3/4 3-59*

3/4 3-60 3/4 3-60 3/4 3-60a 3/4 4-1 3/4 4-1*

3/4 4-2 3/4 4-2 B 3/4 2-3 B 3/4 2-3*

B 3/4 2-4 8 3/4 2-4 B 3/4 2-5 B 3/4 2-5 i

l L

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,, ;, u., _

l l

1 (10) Reactor Enclosure Cooling Water and Chilled Water g

1 Isolttic.r. Velves (Section 6.2.4.2, SER cr.d 55Ek-3)

The licensee shall, prior to startup following the first refuel-ing outage, provide automatic and diverse isolation signals to the reactor erclosure couling water inboard ano outboard isolation valves in the supply and return lines to the recirculation pumps ard the drywell chilled water outboard isolation valves in the supply and return lines.

(11)HydrogenRecombinerisolation(Section6.2.4.2, SER and 55ER-1 and 55ER-3)

The licensee shall, prior to startup following the first refuel-irs cutage, install and test an additional automatic isolation valve in each of the hydrogen recortbiner lines penetrating the primary containment.

(12)RenoteShutternSystem(Sections 7.1.4.4, 7.4.2.3, SEk ano SEction 7.4.2.3, 55ER-3 and SSER-5)

I.

The licensee shall, prior to startup following the first refueling outage, have completed modifications to the existing remcte shutdown system to provide a redundant safety-related method of achieving safe shutdown conditions without lifting leads or adding jumpers.

The modifications to be completed shall be those described in the licensee's letters dated April 18 and 22, 1985 which allow for the operation of the B RHR pump, the B RHR Sil pump and the B ESW pump from the respective punp breaker compartarents by the installation of transfer switches. The licensee shall perform necessary tests prior to startup following the first refueling outage to demonstrate the operability of the modified system.

(13) (Deleted) l l

i Amendment No. 3 i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE AP'LICABILITY...............................................

3/4 0-1 3/4.0 P

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..........................................

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.....................................

3/4 1-2 3/4.1.3 CdNTROLRODS 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 Position Indication..........................

3/4 1-13 Control Rod Drive Housing Support........................

3/4 1-15 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer......................................

3/4 1-16 1

Rod Sequence Control System.........................,.....

3/4 1-17 l

Rod Block Monitor........................................

3/4 1-18 i

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................

3/4 1-19 i

Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements........................

3/4 1-21 Figure 3.1.5-2 Sodium Pentaborate Solution I

Volume / Concentration Requirements...

3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...............

3/4 2-1 Figure 3.2.1-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB278............

3/4 2-2 LIMERICK - UNIT 1 v

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE POWER DISTRIBUTION LIMITS (Continued)

Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB248...........

3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB263...........

3/4 2-4 Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB094...........

3/4 2-5 Figure 3.2.1-3 Maximum Average Planar Linear' Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB071...........

3/4 2-6 3/4 2.2 APRM SETP0lNTS..........................................

3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RATI0............................

3/4 2-8 Table 3.2.3-1 Minimum Critical Power Ratio (MCPR) versus Plant Operating Condition.......

3/4 2-8a figure 3.2.3-la Minimum Critical Power Ratio (MCPR)

Versus i at Maximum Core Flow $ 100%

Rated..................................

3/4 2-10 Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR)

Versus I at Maximum Core Flow < 105%

Rated and Maximum Feedwater ReHuction 5 60% at Rated Conditions..............

3/4 2-10a Figure 3.2.3-2 K

Factor..............................

3/4 2-11 f

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 Amendment t.o. 3

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-upscale scram trip setpoint (S) and and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the following relationships:

TRIP SETPOINT ALLOWABLE VALUE S < (0.66W + 51%)T S < (0.66W + 54%)T Sj1(0.66W+45%)T Sj$(0.66W+42%)T R

R where: S and S are in percent of RATED THERMAL POWER, RB W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY.

T is applied only if less than or equal to 1.0.

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

ACTION:

With the APRM flow biased neutron flux-upscale scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the Allowable Value column for S or S

as above determined, initiate corrective action within 15 minutes a$,adjustSand/ ors to be consistent with the Trip Setpoint values

gp within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or redute 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 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased neutron flux-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required:

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 MFLPD greater than or equal to FRTP.

d.

The provisions of Specification 4.0.4 are not applicable.

With MFLPD greater than the FRTP during power ascension up to 90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel.

LIMERICK - UNIT 1 3/4 2-7 l

i

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit determined using the appropriate figure taken from Table 3.2.3-1, times the K shown in Figure 3.2.3-2, provided that the end-f of-cycle recirculation pump trip (E0C-RPT) system is OPERABLE per Specifica-tion 3.3.4.2, with:

(Iave I )

B T

~I A

B where:

A = 0.86 seconds, control rod average scram insertion T

time limit to notch 39 per Specification 3.1.3.3, B = 0.688 + 1.65[

]b(0.052),

I T

N I

i=1 n

I t,y, =

4=1 N ig j

n I

N g i=1 n = number of surveillance tests performed to date in cycle, th Ng = number of active control rods measured in the i surveillance test, 4 = average scram time to notch 39 of all rods measured 1

th in the i surveillance test, and N1 = 4.1.3.2.a. total number of active rods measured in Specification APPLICABILITY:

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

4 LIMERICK - UNIT 1 3/4 2-8 Amendment No. 3

TABLE 3.2.3-1 Minimum Critical Power Ratio (MCPR)

Versus Plant Operating Condition Rated Feedwater Maximum Core MCPR Temperature Reduction Flow (% of rated)

Figure #

From the Nominal, delta T* ('F) 0 1 100 3.2.3-la 1 60 1 105 3.2.3-lb i

l l

This delta T refers to the planned reduction of feedwater temperature at rated conditions from nominal rated feedwater temperature during the prolonged re-moval of feedwater heaters from service.

LIMERICK - UNIT 1 3/4 2-8a Amendment No. 3

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION a.

With the end-of-cycle recirculation pump trip system inoperable pr Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within I hour, r

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 taken from Table 3.2.3-1 for EOC-RPT inoperable curve times the K f

shown in Figure 3.2.3-2.

b.

With MCPR less than the applicable MCPR limit as identified in ACTION a above, 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.

I 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 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 taken from Table 3.2.3-1 times the Kf shown in Figure 3.2.3-2.

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 MCPR.

d.

The provisions of Specification 4.0.4 are not applicable.

l f

LIMERICK - UNIT 1 3/4 2-9 Amendment No. 3

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.,,.,in.i. O. O.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 r s r MINIMUM CRITICAL POWER RATIO 04 4) VERSUS I AT MAXIMUM CORE FLOW < 100% RATED (RATED FEEDWATER TEMPERATURE) FIGURE 3 2.3-la 1 !'\\ / ) LIMERICK - UNIT 1 3/4 2-10 Amendment No. 3 r s] ~ -.., - -

4 1,4 2 = 1.42 1.40-1.40 1.38-1.38-1.36 = 1.36 EOC-RPT INOPERABLE-g /y 1.34 l.34 = M cr: 1.32 = l.32 g o / l.30- / -1.30 / l.28-l.28 E0C-RPT OPERABLE s 1.26= 1.26 1.24 - l.24 1.22 = 1.22 1.20 1.20 3 3 g a i 3 3 i i 0. 0.1 0.20.30.40.50.60.70.80.91.0 T MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS T AT MAXIMUM CORE FLOW < 105% RATED AND MAXIMUM FEEDWATER TEMPERATURE REDUCTION < 60*F AT RATED CONDITIONS FIGURE 3.2.3-1b LIHERICK - UNIT 1 3/4 2-10a Amendment No. 3

TABLE 3.3.6-1 (Continued) CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS Declare the RBM inoperable and take the ACTION required by ACTION 60 Specification 3.1.4.3. ACTION 61 - With the number of OPERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour. With the number of OPERABLE channels less than required by the ACTION 62 Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour. With the number of OPERABLE channels less than required by the ACTION 63 Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block. NOTES With THEPMAL POWER > 30% of RATED THERMAL POWER. With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. (a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER. (b) This function shall be automatically bypassed if detector count rate is > 100 cps or the IRM channels are on range 3 or higher. (c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher. (d) This function is automatically bypassed when the IRM channels are on range 3 or higher. (e) This function is automatically bypassed when the IRM channels are on range 1. LIMERICK - UNIT 1 3/4 3-59

i r-TABLE 3.3.6-2 x 2 CONTROL R00 BLOCK INSTRUMENTATION SETPOINTS [* TRIP FUNCTION TRIP SETPOINT ALLOWA8LE VALUE EE 1. ROD BLOCK MONITOR El a. Upscale sa i. flow biased < 0.66 W + 40%, with a < 0.66 W + 43%, with a maximum of, saximum of, ii. high flow clamped < 106% < 109% b. Inoperative N.A. N.A. c. Downscale > 5% of RATED THERMAL POWER > 3% of RATED THERMAL POWER 2. APRM a. Flow Biased Neutron Flux - Upscale < 0.66 W + 42%* b. Inoperative N.A. ~< 0.66 W + 45%* N.A. us c. Downscale > 4% of RATED THERMAL POWER > 3% of RATED THERMAL POWER 30 d. Neutron Flux - Upscale, Startup {12%ofRATEDTHERMALPOWER {14%ofRATEDTHERMALPOWER t us J. 3. SOURCE RANGE MONITORS C' a. Detector not full in N.A. N.A. b. Upscale < 1 x 105 cps < 1.6 x 105 cps c. Inoperative R.A. N.A. d. Downscale > 3 cps ** > 1.8 cps ** 4. INTERMEDIATE RANGE MONITORS a. Detector not full in N.A. N.A. b. Upscale < 108/125 divisions of < 110/125 divisions of Tull scale Tull scale c. Inoperative N.A. N.A. d. Downscale > 5/125 divisions of full > 3/125 divisions of full icale scale R 5. SCRAM DISCHARGE VOLUME R a. Water Level-High 5 257' 5 9/16" elevation *** $ 257' 7 9/16" elevation 5 a. Float Switch E us I

TABLE 3.3.6-2 (Continued) r- !b CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS M [* TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE EE 6. REACTOR COOLANT SYSTEM RECIRCULATION Cl FLOW sa a. Upscale < 111% of rated flow < 114% of rated flow b. Inoperative R.A. H.A. c. Comparator i 10% flow deviation $ 11% flow deviation 7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. N.A. "The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W). The trip setting of this function must be maintained in accordance with Specification 3.2.2.

- May be reduced to 0.7 cps provided the signal-to-noise ratio is > 2.

i'

- - Equivalent to 13 gallons / scram discharge volume.

E E .E

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with: a. Total core flow greater than or equal to 45% of rated core flow, or b. THERMAL POWER less than or equal to the limit specified in Figure 3.4.1.1-1. APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*. ACTION: a. With one reactor coolant system recirculation loop not in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and initiate measures to place the unit in at least HOT SHUTDOWN within 12 hours. b. With no reactor coolant system recirculation loops in operation, immediately initiate action to reduce THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1 within 2 hours and initiate measures to place the unit in at least STARTUP within 6 hours and in HOT SHUTDOWN within the next 6 hours. I c. With two reactor coolant system recirculation loops in operation and total core flow less than 45% of rated core flow and THERMAL POWER greater than the limit specified in Figure 3.4.1.1-1: 1. Determine the APRM and LPRM** noise levels (Surveillance 4.4.1.1.3): a) At least once per 8 hours, and b) Within 30 minutes after the completion of a THERMAL POWER 4 increase of at least 5% of RATED THERMAL POWER. 2. With the APRM or LPRM** neutron flux noise levels greater than three times their established baseline noise levels, immediately initiate corrective action to restore the noise levels to within the required limits within 2 hours by increasing core flow to greater than 45% of rated core flow or by reducing THERMAL POWER to less than or equal to the limit specified in Figure 3.4.1.1-1.

See Special Test Exception 3.10.4.
- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - UNIT 1 3/4 4-1 i

REACTOR COOLANT SYSTEM l l SURVEILLANCE REQUIREMENTS 4.4.1.1.1 Each pump discharge valve shall be demonstrated OPERABLE by cycling each valve through at least one complete cycle of full travel during each startup* prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER. 4.4.1.1.2 Each pump MG set scoop tube mechanical and electrical stop shall be demonstrated OPERABLE with overspeed setpoints less than or equal to 109% and 107%, respectively, of rated core flow, at least once per 18 months. 4.4.1.1.3 Establish a baseline APRM and LPRM** neutron flux noise value within the regions for which monitoring is required (Specification 3.4.1.1, ACTION c) within 2 hours of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage. l l l l l

If not performed within the previous 31 days.
- Detector levels A and C of one LPRM string per core octant plus detectors A and C of one LPRM string in the center of the core should be monitored.

LIMERICK - UNIT 1 3/4 4-2 Amendment No. 3 l

POWER DISTRIBUTION LIMITS BASES TABLE B 3/4.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS Plant Parameters: Core THERMAL POWER. ................. 3430 MWt* which corresponds to 105% of rated steam flow Vessel Steam Output................... 14.86 x 108 lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure............. 1055 psia Design Basis Recirculation Line Break Area for: 2 2 4.1 ft, 1.0 ft a. Large Breaks 2 2 2 2 1.0 ft, 0.07 ft, 0.09 ft, 0.02 ft b. Small Breaks Fuel Farameters: PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO ( Initial Core 8x8 13.4 1.4 1.20 A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 15.0.2 of the FSAR.

This power level meets the Appendix K requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

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

POWER DISTRIBUTION LIMITS BASES I 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.06, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evalua-tion with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWi" ?ATIO (CPR). The type of transients evaluated were loss of flow, increase pressure and power, positive reactivity insertion, and coolant tempera se decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR of 1.06, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in Figures 3.2.3-la and 3.2.3-lb. l 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 progr events is described in NED0-24154 g The code used to evaluate pressurization events is described in NED0-10802(p) and the program used in non pressurization The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundlewithtgsinglechanneltransientthermalhydraulicTASCcodedescribed. in NEDE-25149 The principal result of this evaluation is the reduction in MCPR caused by the transient. The purpose of the K factor of Figure 3.2.3-2 is to define operating f limits at other than ratea core flow conditions. At less than 100% of rated flowtherequiredMCPRistheproductoftheMCPRandtheK(edduringaflow factor. The K factors assure that the Safety Limit MCPR will not be viola increase transient resulting from a motor generator speed control failure. The K factors may be applied to both manual and automatic flow control modes. g The K, factors values shown in Figure 3.2.3-2 were developed generically and are applicable to all BWR/2, BWR/3, and BWR/4 reactors. The K factors were f derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. For the manual flow control mode, the K factors were calculated such that f 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 K. f LIMERICK - UNIT 1 B 3/4 2-4 Amendment No. 3

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 K, factors shown in Figure 3.2.3-2 are conservative for the General Electric B6111ng Water Reactor plant operation because the operating limit MCPRs of Specification 3.2.3 are greater than the original 1.20 operating limit 1 MCPR used for the generic derivation of K. f 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 indi-cates 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 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 require-ment 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. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NED0-10802, February 1973. 3. Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors, NEDO-24154, October 1978. 4. TASC 01-A Computer Program for the Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980. 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. I dated November 6, 1986. LIMERICK - UNIT 1 B 3/4 2-5 Amendment No. 3}}

ML20211F116

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