ML20062C188

Application for Amend to License DPR-26,changing Tech Specs to Establish Limiting Conditions for Operation & Surveillance Requirements for Overpressure Protection Sys

ML20062C188
Person / Time
Site:	Indian Point
Issue date:	11/03/1978
From:	William Cahill CONSOLIDATED EDISON CO. OF NEW YORK, INC.
To:
Shared Package
ML20062C185	List: ML20062C176 ML20062C188
References
NUDOCS 7811070163
Download: ML20062C188 (22)

Text

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UNITED STATES OF AVERICA hu 2 a R P N RY COMMISSICN In the Matter'of -

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)

CONSOLIDATED EDISCN CCt@ANY ) Docket tb. 50-247 OF NEW YORK, IFC. )

(Indian Point Station, ) -

Unit No. 2) )

l APPLICATICN EUR AMEtDelNT TO OPERATItG LICENSE

, Pursuant to Section 50.90 cf the Regulations of the Nuclear Regulatory Ccmnission (NBC), Consolidated Edison Capany of New York, Inc. (Con FAimn), as holder of Facility Operating License No. DPR-26, hereby applies for arrendment of the sechnical Specifications contained in Appendix A of that license. -

Specifically, we request that Technical Specifications 3.1, 3.3, Table 3-3, Table 4.1-1 and 6.9.2 be modified and that a new specification 4.16 .

be added to establish limiting conditions for operation (IrOs) ard survaillance requirenents for the recently installed Overpressure Protection Systen (OPS). This Application is being subnitted in response to the letter dated August 28, 1978 frtm Mr. A. Schwencer (NPO) to Mr. Willimn J. Cahill, Jr. (Con Edison) .

The specific proposed Technical Specification revisions are set forth in Attachnent A to this Application. A Safety Evaluation of the proposed changes is set forth in Attachnent B to this Application. This evaluation 1

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781107 oI43

.. t . __ _ . . _ _ .. . . - . . . .

P denonstrates that proposed changes do not represent a significant Pazards consideration and will not cause any change in the types or an increase in the anounts of effluents or any change in the authorized pow.r level -

of the facility.

A L

CXESni.TnA'IED EISON COMPME '

& NEW YORK, IIC.

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p ,

By: - [ ,. ~ /4 /

William J. Cahill, Jr.

Vice President .

Snha -ibEd and sworn tD before me this 3 'd day of Novenber, 1978.

6 5: <? c No1;ary Public ANGELA RODERTl

l. Notary Public. St:ts of New York No. 418503313 Qualified in Queens County Comm:ssiors Expires !. tar:n 30, ISEO T

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. - _ _ . . . . . _ _ . . _ . , . . . . _ _ _ _ . . _ , . . . ~ . . _ . ~ . _ . . _ _ . . _ . _ , . . . _ . . . _...m _ _m ATTACRENT A Technical Specification Page Revisions

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Consolidated Edison Company'of New York, Inc.

Indian Point Unit No. 2 ,

Docket No. 50-247 i November, 1978 l

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I TABLE OF CONTENTS (Continued)

Section Title Page 3.10 Control Rod and Power Distribution Limits 3.10-1 Shutdown Reactivity 3.10-1 ;

Power Distribution Limits 3.10-1 Quadrant Power Tilt Limits 3.10-4 '

Rod Insertion Limits 3.10-5 j Rod Misalignment Limitations 3.10-6 '

Inoperable Rod Position Indicator Channels 3.10-6 Inoperable Rod L!nitations 3.10 Rod Drop Time 3.10-7 Rod Position Monitor 3.10-7

Quadrant Power Tilt Monitor 3.10-7 Notification 3.10-8

3.11 Movable In-Core Instrumentation 3.11-1 !

, 3.12 Shock Suppressors (Snubbers) 3.12-1

[ 3.13 Fire Protection and Detection Systems 3.13-1 4 Surveillance Requirements 4.1-1  !

4.1 Operational Safety Review 4.1-1 4.2 Primary System Surveillance 4.2-1 l 4.3 Reactor Coolant System Integrity Testing 4.3-1 4.4 Containment Tests 4.4-1 Integrated Leakage Rate Test - Pre-Operational 4.4-1  ;

Integrated Leakage Rate Test - Post-Operational 4.4-2 ,

Report of Test Results 4.4-4 ,

Concinuous Leak Detection Testing via the Containment Penetration and Weld Channel Pressurization System 4.4-4

• Corrective Action 4.4-4 Isolation Valve Tests 4.4-4 i Residual Heat Removal Sy. stems 4.4-5  :

Annual Inspection 4.4-6 ,

Containment Modification 4.4-6 ,

~, 4.5 Engineered Safety Features 4.5-1 Safety Injection System 4.5-1 Containment Spray System 4.5-2 Hydrogen Recombiner System 4.5-2 Component Tests 4.5-3 4.6 Emergency Power System Periodic Tests 4.6-1 Diesel Generators 4.6-1 Diesel Feel Tanks 4.6-2 >

Station Batteries 4.6-2 4.7 Main Steam Stop Valves 4.7-1 4.8 Auxiliary Feedwater System 4.8-1 9 4.9 Reactivity Anomalies 4.9-1 4.10 DELETED

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4.11 DELETED 4.12 Shock Suppressors (Snubbers) 4.12-1 4.13 Steam Generator Tube Inservice Surveillance 4.13-1 Inspection Requirements 4.13-1 Corrective Measures 4.13-4 Reports 4.13-4 4.14 Fire Protection and Detection Systems 4.14-1 4.15 Radioactive Materials Surveillance 4.15-1

-4.16 Overpressure Protection System (OPS) 4.16-1l l Amendment No. 11 ,

4.

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LIST OF FIGURES i  ;

Safety Limits Four Loop Operation 100% Flow 2.1-1 i

Safety Limits Three Loop Operation 73% Flow 2.1-2 [

Reactor Coolant System Heatup Limitations 3.1-1 t i

Reactor Coolant System Cooldown Limitations 3.1-2 Reactor Coolant System OPS Setpoint Limit Curve 3.1-3 l

l .

Required Hot Shutdcwn Margin vs Reactor Coolant Boron [

Concentratics 3.10-1  !

j Hot Channel Factor Normalized Operating Envelope 3.10-2 l l

Insertion Limits,100 Step Overlap Four Loop Operation 3.10-3 l

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Insertion Limits,100 Step Overlap Three Loop Operation 3.10-4 h Target Band on Indicated Flux Difference as a Function of [

j Operating Power Level 3.10-5

! i j Permissible Operating Band on Indicated Flux Difference as i a Function of Burnup 3.10-6 r r

Reactor Coolant System Heatup Limitation 4.3-1 Facility Management and Technical Support Organization 6.2-1  !

l Facility Organization 6.2-2 t 4

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9 Amendment No. v s -v - -,a

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f AT trip setpoint for three loop operation has been set in accord- ,

ance with specification 2.3.1.B-4.

d. Reactor operation with one of the four loops out of service will be permitted for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the fourth loop can not be re-turned to service within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the reactor will be put in a 4

hoc s.tutdown condition using normal procedures. i

e. A reactor coolant pump may be started if:

(1) another reactor coolant pump is running, or (2) the secondary side temperature of each steam generator has been verified to be no more than 50'F greater than the l Reactor Coolant System temperature for each loop cold leg.  !

I 2. Steam Generator Two steam generators shall be capable of performing their heat trans-fer function whenever the reactor is critical and the average coolant i temperature is above 350*F.

2

3. f afety Valves

a. At least one pressurizer code safety valve shall be operable when- '

ever the reactor head is on the vessel except for hydrostatically testing the RCS in accordance with the applicable ASME Section XI Boiler and Pressure Vessel Code.

b. All pressurizer code safety valves shall be operable whenever the k-reactor is critical.

c. The pressurizer code safety valve lif t setting shall be set at 2485 psig with !1% allowance for error.

4. Overpressure Protection System (OPS)

a. The OPS shall be " armed" and " operable" whenever the reactor cealant system (RCS) temperature is below 267'F and the RCS is not depressurized 'and vented with an equivalent opening of > 2.00 square inches. The OPS pressurizer power operated relief valves (PORVs) shall have lif t settings within the limits of the PORV setpoint curve specified in Figure 3.1-3 whenever the OPS is re-

• quired to be operable.

Amendment No. 3.1-2 i

i

b. The requirements of 3.1.A.4.a may be modified to permit one PORV and/or its series MOV to be inoperable for a maximum of seven (7) consecutive days. If the single PORV and/or its series MOV is not restored to meet the requirements of 3.1.A.4.a within this seven (7) day period, or if both POKVs and/or their series MOVs are in-operable when required to be operable by 3.1.A.4.a, then, utiliz- <

ing normal operating procedures, either:

(1) the RCS must be depressurized and vented with an equivalent opening of > 2.00 square inches, or (2) the RCS must be heated and maintained above 320*F.

c. If the requirements of 3.1. A.4.b cannot be satisfied, then the plant may be brought to the cold shutdown condition only for the

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following limited conditions:

(1) the repairs cannot be accomplished within the seven (7) cay time period requirement of 3.1.A.4.b or cannot be performed under hot conditions, or (2) another action statement requires cooldown, or (3) protection and safety of plant personnel or equipment requires i cooldown.

d. In the event either the PORVs or the RCS vent (s) are used to miti- ,

gate an RCS pressure transient, a Special Report shall be prepared j and submitted to the Commission pursuant to Specification 6.9.2.f within 30 days. The report shall describe the circumstances initi-ating the transient, the effect of the PORVs or vent (s) on the '

transient and any corrective action necessary to prevent recurrence.

Basis When the bt{on concentration of the Reactor Coolant System is to be reduced the process must be uniform to prevent sudden reactivity changes in the reactor. Mixing of the reactor coolant will be sufficient to maintain a uni-form boron concentration if at least one reactor coolant pump or one residual

'd.

1;

, Amendment No. 3.1-3

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i heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the primary system volume in approximately one half hour. The pressurizer is of no concern because of the low pressurizer volume and because the pressurizer boron concentration will be higher than that of the rest of the reactor coolant.

Heat transfer analyses show that reactor heat equivalent to 10% of rated  !

power can be removed with natural circulation only( }; hence, the specified upper limit of 2% rated power without operating pumps provides a substantial f

safety factor.

Three loop operation is allowed over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to permit corrective

{' action to return the fourth loop to service and limit the number of unnecessary shutdown cycles. During these periods of three loop operation, the reactor ,

coolant system parameters will be maintained within the limits described for three loop operation in Section 2.1 and 3.1 of the Technical Specifications.

Each of the pressurizer code safety valves is designed to relieve 408,000 lbs.

per hr. of saturated steam at the valve set point. Below approximately 350*F l and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure.(

If no residual heat were removed by the Residual Heat Removal System the amount of steam which could be generated at safety valve relief pressure would be less than half the capacity of a single valve. One valve therefore provides adequate protection for over-pressurization.

The combined capacity of the three pressurizer safety valves is greater than the maximum surge rate resulting from complete loss of load (3) without a direct reactor trip or any other control.

' I Two steam generators capable of performing their heat transfer functlan will provide sufficient heat removal capability to remove core decay he,t af ter a reactor shutdown.

i The OPS is designed to relieve the RCS pressure for certain unlikely incidents to prevent peak RCS pressure from exceeding the 10 CFR 50, Appendix G, limits.

L  !

l Amendment No. 3.1-3(a)

heat removal pump is running while the change is taking place. The residual heat removal pump will circulate the primary system volume in approximately one half hour. The pressurizer is of no concern because of the low pressurizer

. volume and because the pressurizer boron concentration will be higher than. _ _ . _

I that of the rest of the reactor coolant.

Heat transfer analyses show that reactor heat equivalent to 10% of rated power can be removed with natural circulation only( } ; hence, the specified upper limit of 2% tated power without operating pumps provides a substantial safety factor.

Three loop operation is allowed over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period to permit corrective f action to return the fourth loop to service and limit the number of unnecessary shutdown cycles. During these periods of three loop operation, the reactor r

coolant system parameters will be maintained within the limits described for three loop operation in Section 2.1 and 3.1 of the Technical Specifications.

l Each of the pressurizer cede safety valves is designed to relieve 408,000 lbs.

, per hr. of saturated steam at the valve set point. Below approximately 350*F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure.

If no residual heat were removed by the Residual Heat Removal System the amount  ;

of steam which could be generated at safety valve relief pressure would be less than half the capacity of a single valve. One valve therefore provides adequate protection for over-pressurization.

The combined capacity of the three pressurizer safety valves is greater than the maximum surge rate t::alting from complete loss of load (3) without a direct reactor trip or any other control.

[

Two steam generators capable of performing their heat transfer function will .

i provice mufficient heat removal capability to remove core decay heat after a l reactor shutdown.

The OPS is designed to relieve the RCS pressure for certain unlikely incidents to prevent peak RCS pressure from exceeding the 10 CFR 50, Appendix G, limits.

l l

Amendment No. 3.1-3(a)

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When the OPS is " armed" it means that MOVs 535 and 536 are in the open posi-tion. This OPS " arming" can be accomplished ei*her automatically by the OPS when the RCS temperature is <267'F or manually by the control room operator.

The likelihood of occurrence of the Heat Input initiating even (i.e., SG/RCS temperature differential) for overpressure of the RCS when the RCS is below 267'F is significantly reduced if another RC pump is running, o r the steam generator secondary side temperature is no more than 50*F grc3t sr than the RCS temperature for each loop cold leg. The preventive measures for the Mass Input initiating event (i.e., Safety Injection pump flow) as well as the Heat Input initiating event have been fully described in the Reference 4) and 5) submittals to the NRC. (Also rrrer to specification 3.3.A, Safety Injection i ' and Residual Heat Removal Systems). ,

The OPS need not be operable when the RCS temperature is <267'F if the RCS is depressurized and vented with an equivalent opening of at least 2.00 square inches. This opening is adequate to relieve the worst case analyzed incidents and the PORVs need not, therefore, be operable.

An RCS temperature of 267'F is the minimum temperature at which an RCS hydro- l static test can be performed (i.e. , 277'F less 10*F for possible instrument errors - see Figure 3.4-1). Therefore, the OPS arming temperature of <267'F permits the performance of an RCS hydrostatic test without activating the OPS. Upon OPS inoperability, the RCS may be heated above 320*F. This RCS temperature is that value for which the RCS heatup and cooldown curves (Fig-ures 3.1-1 and 3.1-2) permit pressurization to the setting of the pressurizer code safety valves. Accordingly, with an inoperable OPS and an RCS temperature

>320*F, the pressurizer code safety valves will preclude exceeding the 10 CFR 50 Appendix G, limits.

Amendment No. 3.1-3(b)

9 Reference

1) FSAR Section 14.1.6

2) FSAR Section 9.3.1

3) FSAR'Section~14.1.1

4) Attachment 1 to the letter dated October 25, 1976, from Mr. William J.

Cahill, Jr. (Con Edison) to Mr. Robert W. Reid (NRC).

5) Attachments 1 and 2 to the letter dated February 28, 1977 from Mr. William J. Cahill, Jr. (Con Edison) to Mr. Robert W. Reid (NRC).

l l

l Amendment No. 3.1-3(c)

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- - , - - . _ _ . - _,4 m 41 AMENDMENT NO.

i i

3.3 ENGINEERED SAFETY FEATURES I

Applicability I

Applies to the operating status of the Engineered Safety Features.

Obiective To define those limiting conditions for operation that are necessary: '

(1) to remove decay heat from the core in emergency or normal shutdown situ-ations, (2) to remove heat from containment in normal operating and emergency  !

situations, (3) to remove airborne iodine from the containment atmosphere -

following a Design Basis Accident, (4) to minimize containment leakage to the environment subsequent to a Design Basis Accident, (5) to minimize the potential for and consequences of Reactor Coolant System pressure transients.

Specification '

The following specifications apply except during low temperature physics >

tests.

A. Safety Injection and Residual Heat Removal Systems

1. The reactor shall not be made critical, except for low temperature physics tests, unless the following conditions are met: I

a. The refueling water storage tank contains not less than 345,000 t gallons of water with a boron concentration of at least 2000 ppm.

b. The boron injection tank contains not less than 1000 gallons of a

' i 11 1/2% to 13% by weight (20,000 ppm to 22,500 ppm of boron) boric acid solution at a temperature of at least 145*F. Two channels of heat tracing shall be available for the flow path. Valves 1821 and 1831 shall be open and valves 1822A and 1822B shall be closed, except during short periods of time when they can be cycled to demonstrate their operability.

c. The four accumulators are pressurized to at least 600 psig and each contains a minimum of 800 f t3 and a maximum of 815 ft3 of r

water with a boron concentration of at least 2000 ppm. None of these four accumulators may be isolated.

d. Three safety injection pumps together with their associated piping i

and valves are operable.

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Amendment No. 3. 3-1 [

I .

. L i

I l e. Two residual heat removal pumps and heat exchangers together with their associated piping and valves are operable, f

f. Two recirculation pumps together with the associated piping and valves are operable.

g. Valves 842 and 843 in the mini-flow return line from the discharge of the safety injection pumps to the ERST are de-energized in the open position.

h. Valves 856A, C, D and E, in the discharge header of the safety injection header are in the open position. Valves 856B and F, in the discharge header of the safety injection header are in the closed position. The hot leg valves (856B and F) shall be closed

('

with their motor operators de-energized by locking out the circuit breakers at the Motor Control Centers.

J

1. The four accumulator isolation valves shall be open with their i

motor operators de-energized by locking out the circuit breakers at the Motor Control Centers.

J. Valve 1810 on the suction line of the high-head SI pumps and valves 882 and 744, respectively on the suction and discharge line of the residual heat removal pumps, shall be blocked open by de-energizing the valve-motor operators.

k. The refueling water storage tank low level alarms are operable and (m- set to alarm between 92,800 gallons and 99,000 gallons of water in the tank. .

2. During power operation, the requirements of 3.3. A-1 may be modified to i

allow any one of the following components to be inoperable at any one

-time. If the system is noc restored to meet the requirements of -

3.3. A-1 within the time period specifie , the reactor shall be placed ,

in the hot shutdown condition utilizin', normal operating procedures.

If the requirements of 3.3. A-1 are nc c satisfied within an additional 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> the reactor shall be placed in the cold shutdown condition utilizing normal operating procedures.

a. One safety injection pump may be out of service, provided the pump

-is -restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the remaining two pumps .are demonstrated to be operable.

Amendment No. 3. 3-2 g -e,,- --- , ---

i i

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b. One residual heat removal pump may be out of service, provided the  ;

pump is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the other  !

,L residual heat removal pump is demonstrated to be operable.

-c. One residual heat removal exchanger may be out of service provided that it is restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

d. Any valve required for the functioning of the system during and following accident conditions may be inoperable provided that it i is restored to operable status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and all valves in the -

system that provide the duplicate function are demonstrated to be operable.

e. One channel of heat tracing may be o'ut of service, for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

(- f. One refueling water storage tank low level alarm may be inoperable for up to 7 days provided the other low level alarm is operable. l f

3. No more than one safety injection pump may be energized when the RCS  ;

temperature is <267*F, unless:

a. the RCS is depressurized and vented with an equivalent opening of 5

>3.00 square inches, or

, b. at least one valve in the flow pa,th from the additional safety j 1

injection pumps to the RCS is closed and either locked (if manual)  !

or de-energized (if motor operated). ,

B. CONTAINMENT COOLING AND IODINE REMOVAL SYSTEMS

1. The reactor shall not be made critical unless the following conditions-are met:

a. The spray additive tank contains not less than 4000 gallons of f solution with a sodium hydroxide concentration of not less. than '

t 30% by weight. L

b. The five fan cooler-charcoal filter units and the two spray pumps, l with their associated valves and pr. ping, are operable. I

2. During power operation, the requirements of 3.3.B-1 may be modified ,

I to allow any one of the following components to be inoperable. If the system is not restored to meet the i e

i P

Amendment No. 3.3-3  !

, e-.,#w.e=+-w ww. ee% .-,.e .

t I

i l

I The OPS has been designed to withstand the effects of the postulated worst [

case for Mass Input flow (i.e., Single S.I. pump) without exceeding the 10 CFR 50, Appendix G, limits. Figure 3.1-3 shows the maximum setpoint curve of the l OPS PORVs which is sufficiently below the 10 CFR 50, Appendix G, limits such that PORV overshoots do not result in peak RCS pressures exceeding the 10 CFR  ;

i 50, Appendix G, limits. Thus it is acceptable to energize one S.I. pump when ..

the OPS is armed.

k More than one S.I. pump may be energized when the RCS temperature is <267'F if at least one valve in the flow path from the additional pumps to the RCS is closed and either locked (if manual) or de-energized (if motor operated), or l

< the RCS is depressurized and vented with an equivalent opening of at least 3.00 l

\ square inches. This opening is adequate to relieve tne RCS pressure transient resulting from a postulated start of all three (3) S. I. pumps. ,

i References ,

(1) FSAR Section 9  !

I (2) FSAR Section 6.2 (3) FSAR Section 6.2 l (4) FSAR Section 6.3 (5) FS AR Section 14.3.5 .

(6) FSAR Section 1.2 I (7) FSAR Section 8.2 I

(8) FSAR Section 9.6.1 ,

t (9) FSAR Section 14.3  ;

(10) Indian Point Unit No. 2 " Analysis of the Emergency Core Cooling System in Accordance with the Acceptance Criteria of 10CFR50.46 and Appendix.K l r

of 10CFR50", January 1977. '

i (11) Letter from William J. Cahill, Jr. of Consolidated Edison Company of New York, to Robert W. Reid of the Nuclear Regulatory Commission, dated July 13, 1976. Indian Point Unit No. 2 Small Break LOCA Analysis.

(12) Indian Point Unit No. 3 FSAR Sections 6.2 and 6.3 and the Safety Evalu-ation accompanying " Application for Amendment to Operating License" sworn to by Mr. William J. Cahill, Jr. on March 28, 1977.

Amendment No. 3.3-15

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TABLE 3-3 (CONTINUED) 1 2 3 4 5 NO. OF NO. OF MIN. MIN. OPERATOR ACTION CilANNELS CllANNELS OPERABLE DECREE OF IF CONDITIONS OF TO TRIP CllANNELS REDUNDANCY COLUMN 3 OR 4 CANNOT BE MET NO. FUNCTIONAL UNIT

3. OVERPRESSURE PROTECTION 3 2 2 I ****

SYSTE!! (OPS)

• • • • Refer to Specification 3.1.A.4.

i Anendment No.

g #

TABLE 4.1-1 (CONTINUED) {

Channel Des c rip t ion Check Calibrate Test Remarks l

24. Turbine First Stage Pressure S R M ,

'j. 25. Logic Channel Testing N.A. N.A. M I

1

26. Turbine Overspeed Protection Trip Channel (Electrical) N.A. R M

27. Control Room Ventilation N.A. N.A. R Check damper operation i

for accident mode with isolation signal

28. Overpressure Protection System (OPS) N.A. R

• Note: Specified intervals may be adjusted pl6s or minus 25% to accommodate normal test schedules.

S - Each shift M - Monthly Q - Quarterly S. A. - Semi-annually ,

D - Daily P - Prior to each startup if not done previous week W - Weekly R - Each Refueling Shutdown, but not to exceed 18 months, except for the first fuel cycle.

N.A. - Not Applicable i

• Within 31 days prior to entering a condition in which OPS is required to be operable and at monthly intervals thereaf ter when OPS is required to be operable. ,

1 i

i Amendment No.

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w we r - -

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4.16 Overpressure Protection System (OPS)

Applicability This specification applies to the surveillance requirements for the OPS pro-vided for prevention of RCS overpressurization. t i

Objective To verify operability of the OPS.

Specification o

A. When the OPS PORVs are being used for overpressure protection as required by specification 3.1. A.4, their associated series MOVs shall be verified to be open at least twice weekly with a maximum time between checks of 5 l a

(_.

days.  !

i B. When RCS venting is being used for overpressure protection as permitted by specifications 3.1. A.4 and 3.3. A.3.a. the RCS vent (s) shall be verified to be open at least daily. When the venting pathway is provided with a  ;

valve (s) which is locked, sealed, or otherwise secured in the open position,

• J then only these valves need be verified to be open at monthly' intervals.

C. Whenever the RCS temperature is below 267*F: -

1. the safety injection pumps required to be inoperable by specification 3.3. A.3 shall be demonstrated inoperable at monthly intervals by veri- .

fying lock-out of the pump circuit breakers at the appropriate 480V switchgear, or

[

2. if specification 3.3.A.3.a is being applied to satisfy specification t

3.3. A.3, the requirements of specification 4.16.B above shall be fol- .

Iowed, or

3. if specification 3.3. A.3.b is being applied to satisfy specification j 3.3. A.3, the appropriate safety injection valve (s) shall be demonstrated [

. inoperable at monthly intervals by verifying that manual valve (s) are  !

locked or that de-energized motor operated valve (s) have their circuit -

breakers locked-out at the appropriate motor control center (s). '

i Amendnent No. 4.16-1 1 i

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D. The remaining OPS surveillance requirements shall be as established in Table 4.1-1.

Basis These specifications establish the surveillance program for the RCS Overpressure l

Protection System (OPS) provided to reduce the potential for and mitigate the consequences of RCS pressure transients. This surveillance program is intended to verify operability of this system and will identify for corrective action any conditions which could prevent any portion of the system from performing ,

its intended function.

The PORVs and MOVs associated with the OPS are not included in this specifica-

[~

tion since the valve cycling and operability tests for these valves are per- ,

r formed in accordance with the applicable testing requirements of the ASME Code 6 Section XI and 10 CFR 50.55a.

Amendment No. 4.16-2

. .e

,- 3 I

SPECIAL REPORTS ,

6.9.2 Special reports shall be submitted to the Director of the Region I  !

Office of Inspection and Enforcement within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification: '

a. Each containment integrated leak rate test shall be the subject of a summary technical report including results of the local leak rate 6 test since the last report. The report shall include analyses and interpretations of tne results which demonstrate compliance in meet- '

ing the leak rate limits specified in the Technical Specifications.

. b. A report covering the X-Y xenon stability tests within three months

( -- upon completion of the tests,

c. To provide the Commission with added verifications of the safety and reliability of the pre-pressurized Zircaloy-clad nuclear fuel, a l limited program of non-destructive fuel inspections will be conducted. ,

The program shall consist of a visual inspection (e.g. , underwater TV, periscope, or other) of the two lead burnup assemblies in each region during the first, second, and third refueling shutdowns. Any condition observed by this inspection ,which would lead to unacceptable fuel performance may be the object of an expanded surveillance effort. f If another domestic plant which contains pre-pressurized fuel of a l similar design reaches fuel exposures equal to or greater than at

[~ j Indian Point Unit No. 2, and if a limited inspection program is or has been performed there, then the program may not have to be performed at Indian Point Unit No. 2. However, such action requires approval  !

of the Nuclear Regulatory Commission. The results of these inspections will be reported to the Nuclear Regulatory Commission.

d. Inoperable fire protection and detection equipment (Specification 3.13) .

e. Sealed sourcc leakage in excess of limits (Specification 4.15).

f. Operation of Overpressure Protection System (Specification 3.1. A.4.d) .

Amendment No. 6-19

r.

ATTACHMENT B Safety Evaluation

(

i

(

Consolidated Edison Company of New York, Inc.

Indian Point Unit No. 2 Docket No. 50-247 November, 1978

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f. - - ,.

Safety Evaluation l 8

l The proposed technical specification chenges, contained in Attachment A to this

] Application, are intended to assure operability of the Overpressure Protection System (OPS) and other overpressure mitigating measures to ensure that reactor coolant system (RCS) pressure transients will not exceed the 10 CFR 50. Appen- l dix G, limits for the Indian Point Unit No. 2 reactor vessel. The proposed [

revisions are based on the model technical specifications provided with Mr. A.

1 Schwencer's (NRC) August 28, 1978 letter to Mr. W. J. Cahill, Jr. (Con Edison). t The proposed changes to Section 3.1 of the Technical Specifications would establish limiting conditions for operation (LCOs) for starting a reactor f b

coolant pump (RCP) and for operability of the OPS. The Section 3.3 revisions f would establish LCOs for the safety injection pumps during shutdown modes to prevent inadvertent RCS overpressurization. The changes proposed for Table 3-3  ;

would likewise establish operability requirements for the OPS instrumentation '

t channels. The proposed revisions to Table 4.1-1 and the proposed requirements '

c presented as a new specification 4.16 would establish the surveillance program.

for assuring operability of the OPS and other RCS overpressure protection measures. The Section 6.9.2 modification would require reporting any opera- -

tion of the OPS for mitigating an RCS pressure transient. -

The proposed changes have been reviewed by both the Station Nuclear Safety

\ Committee and the Con Edison Nuclear Facilities Safety Committee. Both Com- '

mittees concur that the proposed changes do not represent a significant hazards I consideration andw' ill not cause any change in the types or an increase in the I amounts of effluents or any change in the authorized power level of the facility.

i j .

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