Text

Proposed Tech Specs Re CR Block Instrumentation,Standby Liquid Control Sys Operability in Mod 5,scram Discharge Volume Valve Testing,Optional Method of Scram Testing & Defination of Core Alteration
	ML20072K005
Person / Time
Site:	Limerick
Issue date:	08/22/1994
From:	; PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	;
Shared Package
ML20072J997	List: ML20072J994; ML20072K005;
References
	NUDOCS 9408290308
	Download: ML20072K005 (26)
	v • d • e

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TSCR 94-33-0

" Control Rod Block Instrumentation" J

Revised Technical Specifications Pages 9408290308 940822 PDR ADOCK 05000352 P PDR w_ __ _ ._ __ .-

TABLE 4.3.6-1 CONTROL R0D BLOCK INSTRUMENTATION SURVETLLANCE REQUIREMENTS CHANNEL OPERATIONAL -

CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH

_[ TRIP FUNCTION CHECK TEST CALIBRATION") SURVEILLANCE REOUIRED g

$ 1. ROD BLOCK MONITOR 9 1*

, a. Upscale N.A. Q(*) R

b. Inoperative N.A. Q") N.A. 1*

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,5 c. Downscale N.A. Q") R 1*

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2. APRM

a. Flow Biased Neutron Flux-Upscale N.A. Q SA 1

b. Inoperative N.A. Q N.A. 1, 2, 5***

c. Downscale N.A. Q SA 1

d. Neutron Flux - Upscale, Startup N.A. Q SA 2, 5***

3. SOURCE RANGE MONITORS R a. Detector not full in N.A. M""*) , W") N.A. 2, 5

b. Upscale N.A. M""*) , W") R 2, 5 y c. Inoperative N.A. M ""* ) , W ") N.A. 2, 5

$ d. Downscale N.A. M "" , W ") k 2t, 5

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. W N.A. 2, 5

b. Upscale N.A. W R 2, 5

c. Inoperative N.A. W N.A. 2, 5

d. Downscale N.A. W R 2, 5

5. SCRAM DISCHARGE VOLUME

a. Water Level - High N.A. Q R 1, 2, 5**

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale N.A. Q SA 1

b. Inoperative N.A. Q N.A. I

c. Comparator N.A. Q SA 1

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. R") N.A. 3, 4 l

. TABLE 4.3.6-1 (Continued)

CONTROL R0D BLOCK 1NSTRUMENTATION SURVEILLANCE RE0VIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

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(b) Deleted.

(c) Includes reactor manual control multiplexing system input.

o For OPERATIONAL CONDITION of Specification 3.1.4.3.

o* With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

o** Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

(d) When in OPERATIONAL CONDITION 2.

(e) The provisions of Specification 4.0.4 are not applicable provided that the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the IRMs are on Range 2 or below during a shutdown.

(f) When in OPERATIONAL CONDITION 5.

(g) L e provisions of Specification 4.0.4 are not applicable provided that the surveillance is performed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the Reactor Mode Switch has been placed in the shutdown position.

LIMERICK - UNIT 1 3/4 3-62

INSTRUMENTATION SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be OPERABLE:

a. In OPERATIONAL CONDITION 2*, three.

b. In OPERATIONAL CONDITION 3 and-4, two.'

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

ACTION:

a. In OPERATIONAL CONDITION 2* with one of the above required source range monitor channels inoperable, restore at least three source range monitor channels to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

SURVEILLANCE RE0VIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:

a. Performance of a:

1. CHANNEL CHECK at least once per:

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in CONDITION 2*, AND b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in CONDITION 3 or 4.

2. CHANNEL CA!.IBRATION** at least once per 24 months,

b. Performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days.

c. Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 3.0 cps *** with the detector fully inserted.

With IRM's on range 2 or below.

- Neutron detectors may be excluded from CHANNEL CALIBRATION.

- - 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.

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LIMERICK - UNIT 1 3/4 3-88 l l

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REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Cg_ntinued) ,

b. Performance of a CHANNEL FUNCTIONAL TEST at least once per 7 days,

c. Verifying that the channel count rate is at least 3.0 cps:* ,

1. Prior to control rod withdrawal,

2. Prior to and at least oncd'per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, and

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

d. Verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during, that the RPS circuitry " shorting links" have been removed during:

1. The time any control rod is withdrawn,** or

2. Shutdown margin demonstrations.

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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. l These channels are not required when sixteen or fewer fuel assemblies, adja-cent to the SRMs, are in the core.

o*Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK - UNIT 1 3/4 9-4 ,

. TABLE 4.3.6-1 CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REOUTREMENTS C CHANNEL '0PERATIONAL i5 CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 5 TRIP FUNCTION CHECK TEST CALIBRATION (a) SURVEILLANCE RE0VIRED S 1. R0D BLOCK MONITOR i a. Upscale N.A. Q C) SA 1*

CC) gi b. Inoperative N.A. Q N.A. 1*

M c. Downscale N.A. Q C} SA 1*

n

2. APRM

a. Flow Biased Neutron Flux-Upscale N.A. Q SA 1

b. Inoperative N.A. Q N.A. 1, 2, 5***

c. Downscale N.A. Q SA 1

d. Neutron Flux - Upscale, Startup N.A. Q SA 2, 5***

3. SOURCE RANGE MONITORS

a. Detector not full in N.A. M(d)C*) WCf} N.A. 2, 5 t' b. Upscale N.A. M(d)(*),WCf) R 2, 5

c. Inoperative N.A. M(d)(*),W(f) N.A. 2, 5

d. Downscale N.A. M(d)(*) W, Cf) R 2,1 5

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4. INTERMEDIATE RANGE MONITORS

a. Detector not full in N.A. W N.A. 2, 5

b. Upscale N.A. W R 2, 5

c. Inoperative N.A. W N.A. 2, 5

d. Downscale N.A. W R 2, 5

5. SCRAM DISCHARGE VOLUME

a. Water Level - High N.A. Q R 1, 2, 5**

6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

a. Upscale N .' A . Q SA 1

b. Inoperative N.A. Q N.A. I

c. Comparator N.A. Q SA 1

7. REACTOR MODE SWITCH SHUTDOWN POSITION N.A. R(8) N.A. 3, 4

, TABLE 4.3.6-1 (Continued)

CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE RE0VIREMENTS TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.

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(b) Deleted.

(c) Includes reactor manual control multiplexing system input.

With THERMAL POWER 2 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.

- - Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.

(d) When in OPERATIONAL CONDITION 2.

(e) The provisions of Specification 4.0.4 are not applicable provided that the surveillance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the IRMs are on Range 2 or below during a shutdown.

(f) When in OPERATIONAL CONDITION 5.

(g) The provis ons of Specification 4.0.4 are not applicable provided that the i

surveillance is performed within I hour after the Reactor Mode Switch has been placed in the shutdown position.

LIMERICK - UNIT 2 3/4 3-62 l.

INSTRUMENTATION SOURCE RANGE MONITORS LIMITING CONDITION FOR OPERATION 3.3.7.6 At least the following source range monitor channels shall be,0PERABLE:

a. In OPERATIONAL CONDITION 2*, three.

b. In OPERATIONAL CONDITION 3 and 4, tw6.

APPLICABILITY: OPERATIONAL CONDITIONS 2*#, 3, and 4.

ACTION:

a. In OPERATIONAL CONDITION 2* with one of the above required source range monitor channels inoperable, restore at least three source range monitor channels to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. In OPERATIONAL CONDITION 3 or 4 with one or more of the above required source range monitor channels inoperable, verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within I hour.

SURVEILLANCE REQUIREMENTS 4.3.7.6 Each of the above required source range monitor channels shall be demonstrated OPERABLE by:

a. Performance of a:

l. CHANNEL CHECK at least once per:

a) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in CONDITION 2*, and b) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in CONDITION 3 or 4.

2. CHANNEL CALIBRATION ** at least once per 24 months.

b. Performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days. ,

c. Verifying, prior to withdrawal of control rods, that the SRM count rate is at least 3.0 cps *** with the detector fully inserted.#

With IRM's on range 2 or below. l

- Neutron detectors may be excluded from CHANNEL CALIBRATION.

- - 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.

During initial startup test program, SRM detectors may be partially withdrawn prior to IRM on-scale indication provided that the SRM channels remain on scale above 100 cps and respond to changes in the neutron flux.

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LIMERICK - UNIT 2 3/4 3-88 l

REVELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued)

b. Performance of a CHANNEL FUNCTIONAL TEST at least once per 7 days.

c. Verifying that the channel count rate is at least 3.0 cps:*

1. Prior to control rod withdrawal,

2. Prior to and at least oncs~per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS, and

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

d. Verifying, within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during, that the RPS circuitry " shorting links" have been removed during:

1. The time any control rod is withdrawn,** or

2. Shutdown margin demonstrations.

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.

These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs, are in the core.

- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.

LIMERICK - UNIT 2 3/4 9-4

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TSCR 94-34-0

" Standby Uquid Control System Operability in Mode 5" Revised Technical Specifications Pages t

4 REACTIVITY CONTROL SYSTEMS r

3/4.1.5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 SCTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With standby liquid control system otherwise inoperable, restore the system to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

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

1. The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.

2. The available volume of sodium pentaborate solution is at least 3160 gallons.

3. The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis. l l

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REdCTIVITYCONTROLSYSTEMS 3/4.1.5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION f

3.1.5 The standby liquid control system, consisting of a . minimum of two pumps and ,

corresponding flow paths, shall be OPERABLE.

JPPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 ACTION:

a. In OPERATIONAL CONDITION 1 or 2:

1. With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within-the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With standby liquid control system otherwise inoperable, restore ,

the system to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE0VIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:

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

1. The temperature of the sodium pentaborate solution is within !

. the limits of Figure 3.1.5-1. ;

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2. The available volume of sodium pentaborate solution is at least 4537 l gallons. !

3. The temperature of the pump suction piping is greater than or equal )

to 70*F.

LIMERICK - UNIT 2 3/4 I-19 ;

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TSCR 94-37-0 1 i

" Scram Discharge Volume Valve Testing" i Revised Technical Specifications Pages i

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' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by e demonstrating:

a. The scram discharge volume drain and vent valves OPERABLE at least once per 24 months, by verifying that the drain and vent valves:

1. Close within 30 seconds after receipt of a signal for control rods to scram, and

2. Open when the scram signal is reset.

b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation at least once per 92 days.

LIMERICK - UNIT 1 3/4 1-5

' REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) 4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by ,

demonstrating:

a. The scram discharge volume drain and vent valves OPERABLE at least once per 24 months, by verifying that the drain and vent valves:

1. Close within 30 seconds after receipt of a signal for control rods to scram, and

2. Open when the scram signal is reset.

b. Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST of the scram discharge volume scram and control rod block level instrumentation at least once per 92 days.

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T TSCR 94-39-0

" Optional Method of Scram Testing" ,

Revised Technical Specifications Pages t

REACTTVITY CONTROL SYSTEMS

. I CONTROL R0D MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds'.

APPLICABILITY: OPERATIONAL CONDITIONS I and 2. "

ACTION: l

a. With the maximum scram insertion time of one or more control rods I exceeding 7 seconds:

1. Declare the control rod (s) with the slow insertion time inoperable, and

2. Perform the Surveillance Requirements of Specification 4.1.3.2c.

at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER with reactor coolant pressure greater than or equal to 950 psig, following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.

b. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "1" or "2" as follows:

1.a Specifically affected individual control rods shall be scram time tested at zero reactor coolant pressure and the scram insertion time from the fully withdrawn position to notch position 05 shall not exceed 2.0 seconds, and 1.b Specifically affected individual control rods shall be scram time tested at greater than or equal to 950 psig reactor coolant pressure prior to exceeding 40% of RATED THERMAL POWER.

2. Specifically affected individual control rods shall be scram time tested at greater than or equal to 950 psig reactor coolant pressure.

c. For at least 10% of the control rods, with reactor coolant pressure greater thhn or equal to 950 psig, on a rotating basis, and at least once per 120 days of POWER OPERATION.

LIMERICK - UNIT 1 3/4 1-6

POWER DISTRIBUTION LIMITS :

LIMfTVNG 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 provided that, within I hour, MCPR is determined to be greater than or equal to the rated MCPR limit as a function of the average scram time (shown in the CORE OPERATING LIMITS REPORT) E0C-RPT inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit adjusted by the MCPR(P) and MCPR(F) factors as 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 .

c. With the main turbine bypass system inoperable per Specification 3.7.8, 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 rated MCPR limit as a function of the average scram time (shown in the CORE OPERATING LIMITS REPORT) main turbine bypass valve inoperable curve, adjusted by the MCPR(P) and MCPR(F) factors as shown in the CORE OPERATING LIMITS REPORT.

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 4.1.3.2a and during reactor startups prior to control rod scram time tests in accordtace with Specification 4.1.3.2.b.l.b, 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, shall be determined to be equal to or greater than the applicable MCPR limit, including application of the MCPR(P) and MCPR(F) factors as determined from 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 l 15% of RATED THERMAL POWER, and 1

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 i with a LIMITING CONTROL ROD PATTERN for MCPR.

d. The previsions of Specification 4.0.4 are not applicable. !

l LIMERICK - UNIT 1 3/4 2-9 l

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REACTIVITY CONTROL SYSTEMS

_ BASES __ z __ _ _

3/4.1.3 CONTROL R005 The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod drop accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more restYictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than the eight allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shutdown for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a rate fast enough to prevent the MCPR from becoming less than the fuel cladding safety limit during the limiting power transient analyzed in Section 15.2 of the

. FSAR. This analysis shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the specifi-cations, provided the required protection and MCPR remains greater than the fuel cladding safety limit. The occurrence of scram times longer then those specified should be viewed as an indication of a systemic problem with the rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious problem.

Scram time testing at zero psig reactor coolant pressure is adequate to ensure that the control rod will perform its intended scram function during startup of the plant until scram time testing at 950 psig reactor coolant pressure is performed prior to exceeding 40% rated core thermal power.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1 then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control roos with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the

. accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

LIMERICK - UNIT 1 8 3/4 1-2

REACTIVITY CONTROL SYSTEMS C0 TROL R00 MAXIMUM SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.2 The maximum scram insertion time of each control rod from the fully withdrawn position to notch position 5, based on deenergization of the ,

scram pilot valve solenoids as time zero, shall not exceed 7.0 seconds.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2. -

ACTION:

a. With the maximum scram insertion time of one or more control rods exceeding 7 seconds:

1. Declare the control rod (s) with the slow insertion time inoperable, and

2. Perform the Surveillance Requirements of Specification 4.1.3.2c.

at least once per 60 days when operation is continued with three or more control rods with maximum scram insertion times in excess of 7.0 seconds.

Otherwise, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. The provisions of Specification 3.0.4 are not applicable.

S,URVEILLANCE REQUIREMENTS 4.1.3.2 The maximum scram insertion time of the control rods shall be demon-strated through measurement and, during single control rod scram time tests, the control rod drive pumps shall be isolated from the accumulators:

a. For all control rods prior to THERMAL POWER exceeding 40% of RATED THERMAL POWER with reactor coolant pressure greater than or equal to 950 psig, following CORE ALTERATIONS or after a reactor shutdown that is greater than 120 days.

b. For specifically affected individual control rods following maintenance on or modification to the control rod or control rod drive system which could affect the scram insertion time of those specific control rods in accordance with either "1" or "2" as i follows:

1.a Specifically affected individual control rods shall be scram !

time tested at zero reactor coolant pressure and the scram !

insertion time from the fully withdrawn position to notch position 05 shall not exceed 2.0 seconds, and !

1.b Specifically affected individual control rods shall be scram j time tested at greater than or equal to 950 psig reactor coolant pressure prior to exceeding 40% of RATED THERMAL POWER.

2. Specifically affected individual control rods shall be l scram time tested at greater than or equal to 950 psig !

reactor coolant pressure. l

c. For at least 10% of the control rods, with reactor coolant pressure greater than or equal to 950 psig, on a rotating basis, and at least once per 120 days of POWER OPERATION. )

LIMERICK - UNIT 2 3/4 1-6 l

-- ~ . . . - - - - . - . - . -. - - . -

POWER DISTRIBUTION LIMITS

LIAITING 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 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 appropriate figure in the CORE OPERATING LIMITS REPORT, for E0C-RPT inoperable curve, times the Kg shown in the CORE OPERATING LIMITS REPORT.

b. With MCPR less than the applicable MCPR limit shown in the CORE OPERATING LIMITS i 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 . !

c. With the main turbine bypass system inoperable per Specification 3.7.8, operation may continue provided that, within I 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) main a tur'vine bype s valve inoperable curve, times the Kr shown in the CORE- '

OPERATING LIMITS REPORT. i SURVEILLANCE REQUIREMENTS F.2.3 MCPR, with:

a. 7 - 1.0 prior to performance of the initial scram time measurements ,

for the cycle in accordance with Specification 4.1.3.2.a and during reactor startups prior to control rod scram time tests in accordance. >

with Specification 4.1.3.2.b.1.b, or '

b. y 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 i by Specification 4.1.3.2, l 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. e the Kg 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 l RATED THERMAL POWER, and .i

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R00 PATTERN for MCPR. ,

d. The provisions of Specification 4.0.4 are not applicable.

l i

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

REACTIVITY. CONTROL SYSTEMS ;

BASES , , _ _ l 1/4.1.3 CONTROL RODS The specification of this section ensure that (1) the minimum SHUTDOWN MARGIN is maintained, (2) the control rod insertion times are consistent with those used in the accident analysis, and (3) the potential effects of the rod '

drop accident are limited. The ACTION statements permit variations from the basic requirements but at the same time impose more rest?ictive criteria for continued operation. A limitation on inoperable rods is set such that the resultant effect ,

on total rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic problems with rod drives will be investigated on a timely basis.

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechanical interference, operation of the reactor is limited to a time period l which is reasonable to determine tne cause of the inoperability and at the same !

time prevent operation with a large number of inoperable control rods. l Control rods that are inoperable for other reasons are permitted to be !

taken out of service provided that those in the nonfully-inserted position are consistent with the SHUTDOWN MARGIN requirements.

The number of control rods permitted to be inoperable could be more than (

the eight allowed by the specification, but the occurrence of eight inoperable '

rods could be indicative of a generic problem and the reactor must be shutdown l for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subtritical at a :

rate fast enough to prevent the MCPR from becoming less than the fuel cladding i safety limit during the limiting power transient analyzed in Section 15.2 of the FSAR. .This analysis shows that the negative reactivity rates resulting '

from the scram with the average response of all the drives as given in the specifications, provided the required protection and MCPR remains greater than ,

the fuel cladding safety limit. The occurrence of scram times longer then i those specified should be viewed as an indication of a systemic problem with '

the. rod drives and therefore the surveillance interval is reduced in order to prevent operation of the reactor for long periods of time with a potentially serious prchlem.

Scram time testing at zero psig reactor coolant pressure is adequate to ensure that the control rod will perform its intended scram function during startup of the plant until scram time testing at 950 psig reactor coolant pressure is performed prior to exceeding 40% rated core thermal power.

The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment when required.

Control rods with inoperable accumulators are declared inoperable and Specification 3.1.3.1; then applies. This prevents a pattern of inoperable accumulators that would result in less reactivity insertion on a scram'than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor. l LIMERICK - UNIT 2 B 3/4 1-2 1

e i

f TSCR G4-43-0

" Definition of Core Alteration" Revised Technical Specifications Pages i

1 l

j L

DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitor.s, local power range monitors, intermediate range monitors, traversing incore proDes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe 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 6.9.1.12. 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 E0VIVALENT 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, I-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."

00WNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.

E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radicnuclide 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 SY, STEM ([CCS) RESPON51 TIE 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor 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.

LIMERICK - UNIT 1 1-2 e

' I DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:

a) Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b) Control rod movement, provided there are no fuel assemblies in the associated core cell. i Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe 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 l 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.

l 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, I-132, I-133,1-134, and I-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."

E-AVERAGE DISINTEGRATION ENERGY 1.10 5 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 interval from when the monitored parameter exceeds its ECCS actuation set- '

point at the channG1 sensor 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 i applicable. The response time may be measured by any series of sequential, :

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

LIMERICK - UNIT 2 1-2 l

l t

DEFINITIONS -

END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME 1.12 The END-0F-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME shall be that time interval 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. .

This total system response time consists of two components, the instrumen- ,

tation response time and the breaker arc suppression time. These times I may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.

FRACTION OF LIMITING POWER DENSITY 1.13 The FRACTION OF LIMITING POWER DENSITY (FLPD) shall be the LHGR !

existing at a given location divided by the specified LHGR limit for that bundle type.

FRACTION OF RATED THERMAL POWER 1.14 lhe FRACTION OF RATED THERMAL POWER (FRTP) shall be the measured THERMAL POWER divided by the RATED THERMAL POWER.

FRE0VENCY NOTATION 1.15 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.1.

IDENTIFIED LEAKAGE 1.16 IDENTIFIED LEAKAGE shall be:

a. Leakage into collection systems, such as pump seal or valve packing leaks, that is captured and conducted to a sump or collecting tank, or '

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of the leakage detection systems or not to be PRESSURE B0UNDARY LEAKAGE.

ISOLATION SYSTEM RESPONSE TIME 1.17 The ISOLATION SYSTEM RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its isolation actuation setpoint at the channel sensor until the isolation valves travel to their required positions. '

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.

LIMITING CONTROL R0D PATTERN i

~

1.18 A LIMITING CONTROL R00 PATTERN shall be a pattern which results in the core being on a thermal hydraulic limit, i.e., operating on a limiting >

value for APLHGR, LHGR, or MCPR.

LINEAR HEAT GENERATION RATE 1.19 LINEAR HEAT GENERATION RATE (LHGR) shall be the heat generation per unit '

length of fuel rod. It is the integral of the heat flux over the heat.

transfer area associated with the unit length. ,

LIMERICK - UNIT 2 1-3

ML20072K005

Text

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