Text

Proposed Tech Specs Re Operation W/Partial Feedwater Heating & Increased Core Flow Limits.Related Correspondence
	ML20207B380
Person / Time
Site:	Limerick
Issue date:	11/17/1986
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20207B375	List: ML20207B370; ML20207B376; ML20207B380; ML20207B389; ML20214E477;
References
	OL, NUDOCS 8611240198
	Download: ML20207B380 (8)
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE l

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 P8CIB163 3/42-4l l 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 SETPOINTS 3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RATIO 3/4 2-8 Table 3.2s3-1 Minimum Critical Power Ratio (MCPR) 3/4 2-8a versus Plant Operating Condition Figure 3.2.3-la Minimum Critical Power Ratio 3/4 2-10 (MCPR) Versus t at Maximum Core Flow

$ 100% Rated Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR) 3/4 2-10a Versus t at Maximum Core Flow < 105%

Rated and Maximum Feedwater Reduction

. $ 60% at Rated Conditions Figure 3.2.3-2 y Factor 3/4 2-11 3/4.2.4 LINEAR HEAT GENERATION RATE $/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 3/4 3-2 Instrumentation Table 3.3.1-2 Reactor Protection System 3/4 3-6 Response Times Table 4.3.1.1-1 Reactor Protection System . 3/4 3-7 BS11240198 061117 Instrumentation Surveillance PDR ADOCK 05000352 Requirements P PDR M E F 6 ff,f LUP4fN R eA

. POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO .

LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER PATIO (FCPR) shall be equal to or greater than the MCPR limit determirad using the appropriate figure taken from Table 3.2.3-1 times the l Kp shown in Figure 3.2.3-2, provided that the end-of-cycle iscirculation pump trip (ECC-RPT) system is OPERABLE per Specificatica 3.3.4.2, with:

t=

(Tave - 8)

t g -T B

where:

t A = 0.86 seconds, contiol rod average scram insertion time limit to notch 39 per Specification 3.1.3.3, I

1 ]b(0.052),

8 = 0.688 + 1.65[ n Ng I

i=1 n

I t ,y ,' , 9,3 Ng t, ,

n I Ng .

i=1 n =, number of surveillance tests perfomed to date in cycle, th !

l Ng = number of active control rods

measured in th's 4 survet,11ance test, 19 = average scram time to notch 39 of all rods measured in the i th surveillance test, and

,, N total number of active rods measured in Specification y = 4.1.3.2.a.

APPLICABILITY: ,

OPERATIONAL CONDITION 1, When TIENAL POWCR is greater than or equal'to 25% of PATED

'11Et%L ICWER.

LIMERICK - tt11T 1 3/4,2-8

, - ~ - - - - - _ _ . - _ _ - . _ _ - . - - - - _ _ . - _ * - - . - - -- '- f

TABLE 3.2.3-1 4

Minimum Critical Power Ratio (MCPR)

Versus Plant Operating Condition j Roted Feedwater Maximum Core MCPR Tamperature Reduction Flow (% of rated) Figure i From the Nominal, delta T* (*F) ,

F O < 100 3.2.3-la a p.

< 60 < 105 3.2.3-lb 1

i 1, .

1 .

i 1

}

i i

i

.l l

l , .

i I *This delta T refers'to the plan'ned reduction of feedwater j temperature at rated conditions from nominal rated feedwater t temperature during the prolonged removal of feedwater heaters i from service.

i s

y .

I L.I_ M.. E R _I C__K_ _ .U N _I T..1_ _ _ .-. _ ,_ _ _ .

___.8a

.._ _ __3 / _4 2 _. ___ _ ___

~

, POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

a. With the end-of-cycle recirculation pump trip system inoperable per Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable 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 the appropriate figure taken from Table 3.2.3-1 for EOC-RPT inoperable curve times the K g 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 to 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 .

Otherwise reduce THERMAL POWER sufficiently to restore MCPR to within the required limit or 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 performanc e 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 by Specification 4.1.3.2, chall be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-1 or 3.2.3-2: I

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

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

c. Init'ially 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.

\

LIMERICK - UNIT 1 3/4 2-9

l .4 2 = =l.42 1.40 = = 1.40

. l .3 8 - 1.38 1.36 = -1.36 1.3 4 = =1.34 ce b I .32 = f =1.32 EOC-RPT INOPERABLE 3 y / /

E l .3 0 = l.30 1.28= / 1.28

/

1.26 =

p / 1.26 I.24 - = 1.24 EOC-RkT O'PERABLE l.20 i i i ! ! i i .i ,, 1.20

0. 0.1 0.20.30.40.50.60.70.80.9I.0 T

MINIMUM CRITICAL POWER RATIO (MCPR)

VERSUS T AT MAXIMUM CORE FLOW 1100% RATED (RATED FEEDWATER TEMPERATURE) c l

FI GURE 3.2.3-l e ;

LIMERICK - UNIT I 3/4 2-10 k _. _ . - _ .A re e

l .42 = = 1.42 1.40 = = 1.40 1.38= =1.38-1.36 = =1.36 E0C-RPT INOPERABLE - g 1.34 = /y = 1.34 m

M

1.32 g I .32

f

/

l .30 = / = 1.30

/

E0C-RPT OPERABLE -

l .26 = -- = 1.26

l.24 I .?.4 -

I .22 = 1.22 1.20 i i a i i i i i a 1.20

0. 0.1 0.20.30.40.50.60.70.80.9I.0 T

MINIMUM CRITICAL POWER RATIO (MCPR)

VERSUS T AT MAXIMUM CORE FLOW s 105% RATED AND MAXIMUM FEEDWATER TEMPERATURE REDUCTION i 60*F AT RATED CONDITIONS ,

FIGURE 3.2.3-1b !

LIMERICK - UNIT I 3/4 7-10e

e TABLE 3.3.6-2 CONTROL COD BLOCK INSTRU ENTATION SETPOINTS ,

e TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE

1. ROD BLOCK MONTTOR 1 m. Upscale ,

1. flow biased < 0.66 W + 40%, with a maximum of. < 0.66 W + 43%. with a manimum of.

11. high flow clamped i 106% i 109%

b. Inoperative N.A. N.A.

c. Downscale 1 5% of RATED THERMAL POWER 1 3% RATED THERMAL POWER

2. APRM

a. Flow Biased Neutron Flus - Upscale 1 0.66 W + 42%* 1 0.66 W + 45%*

b. Inoperative N.A. N.A.

c. Downscale 1 4% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER

d. Neutron Flum - upscale. Startup i 12% of RATED THERMAL POWER 1 14% of RATED THERMAL POWER

3. SOURCE RANGE MONITORS

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

b. Upscale i 1 X10(5) cps 5 1.6 X10(5) cps

c. Inoperative N.A. N.A.

) d. Downscale 1 3 cps ** 1 1.8 cps **

4 INTERMEDIATE RANGE MONITORS

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

b. Upscale 5 108/125 divisions of i 110/125 divisions of full scale full scale

c. Inoperative N.A. N.A.

c. Downscale 1 5/125 divisions of full scale 1 3/125 divisions of full scale

5. SCRAM D7SCHARGE VOLUME

a. water Level-High 1 257* 5 9/16" elevation *** 1 257' 7 9/16" elevation

a. Float Switch

6. REACTOR COOLANT SYSTEM RECTRCULATION FLOW

a. Upscale i 111% of rated flow 1 114% of rated flow l

b. Inoperative N.A. N.A.

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

T. 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 bc maintained in accordance with Specification

) 3.2.2.

- May be reduced to 0.7 cps provided the signal-to-noise ratto is 1 2. ,

- - Equivalent to 13 gallons / scram discharge volume.

LIMERICK - UNIT 1 3/4 3-60

REACTOR COOLANT SYSTEM 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 l 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which monitoring is required unless baselining has previously been performed in the region since the last refueling outage.

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

_N _

.A _ _b _ = m - A I

POWER DISTRIBUTION LIMITS

, Mastb 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 f factors shown in Figure 3.2.3-2 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 Kg' '.

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 CPR 50, Appendix K, NEDE-20566, November 1975.

2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, NEDO-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. 1 dated November 6, 1986.

LIMERICK - UNIT 1 B 3/4 2-5

- . . .

ML20207B380

Text

1.32 g I .32

References:

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