ML20083L221

Amend 106 to License NPF-38,changing App a TSs by Removing Rv Matl Specimen Withdrawal Schedule & by Updating RCS P/T Curves

ML20083L221
Person / Time
Site:	Waterford
Issue date:	05/08/1995
From:	Chandu Patel NRC (Affiliation Not Assigned)
To:
Shared Package
ML20083L224	List: ML20083L221 ML20083L228
References
NUDOCS 9505180202
Download: ML20083L221 (19)

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2 WASHINGTON, D.C. 30066-0001

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.ENTERGY' OPERATIONS. INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION. UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.106 License No. NPF-38 1.

The Nuclear Regulatory Cor, mission (the Commission) has found that:

A.

The application for amendment by Entergy Operations, Inc. (the licensee) dated December 14,-1993, as supplemented by letter dated March 3, 1995, complies with the standards and requirements of the Atomic Energy-Act of 1954, as amended (the Act), and the Commission'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 b

and. safety of the public, and. (ii),that such activities will be conducted in compliance with the Commission'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 Commission's regulations and all applicable requirements' have been satisfied.

9505180202 950508 PDR ADOCK 05000382 P

PM w

d 3 = *

• 2.

Accordingly, the' license is. amended by changes to the Technical

-Specifications as indicated in the attachment to this license amendment,-

and paragraph-2.C(2) of Facility Operating License No. NPF-38 is hereby amended to read as follows:

(2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised

- through Amendment No.106, and the Environmental Protection Plan -

contained in Appendix B, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

&hf Chandu P. Patel, Project Manager Project Directorate IV-I Division of Reactor Projects - III/IV-Office of Nuclear Reactor Regulation'

Attachment:

Changes to the Technical Specifications Date of Issuance: May 8, 1995 i

I 1

( -

ATTACHMENT TO LICENSE AMENDMENT NO.106 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 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 areas of change. The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE PAGES INSERT PAGES XXI XXI 3/4 4-3 3/4 4-3 3/4 4-5 3/4 4-5 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 3/4 4-31 3/4 4-31 3/4 4-32 3/4 4-32 3/4 4-34 3/4 4-34 B 3/4 4-7 B 3/4 4-7 8 3/4 4-10 B 3/4 4-10

INDEX LIST OF TABLES (Continued) l 18HLE E8EE 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................

3/4 3-40 3.3-9 REMOTE SHUTDOWN INSTRUMENTATION.......................

3/4 3-42 4.3-6 REMOTE SHUTDOWN INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................

3/4 3-43 3.3-10 ACCIDENT MONITORING INSTRUMENTATION...................

3/4 3-45 j

4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................

3/4 3-46 3.3-12 DELETED...............................................

3/4 3-56 4.3-8 DELETED...............................................

3/4 3-58 3.3-13 EXPLOSIVE GAS MONITORING INSTRUMENTATION............................

3/4 3-61 4.3-9 EXPLOSIVE GAS MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............. 3/4 3-65 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION...........................

3/4 4-15 4.4-2 STEAM GENERATOR TUBE INSPECTION.......................

3/4 4-16 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES......

3/4 4-20 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY......................

3/4 4-22 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.............................

3/4 4-23 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM...................................

3/4 4-26 WATERFORD - UNIT 3 XXI Amendment No. 9,50,5S,75,106

4

_INDEX

. LIST OF TABLES (Continued)

TABLE PAGE 3.7-1 STEAM LINE SAFETY VALVES PER L00P.....................

3/4 7-2 3.7-2 MAXIMUM ALLOWABLE LINEAR POWER LEVEL - HIGH TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS..........

3/4 7-3 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..........................

3/4 7-8 3.7-3 ULTIMATE HEAT SINK MINIMUM FAN REQUIREMENTS..........

3/4 7-14 4.8-1 DIESEL GENERATOR TEST SCHEDULE.......................

3/4 8-7 4.8-la ADDITIONAL RELIABILITY ACTIONS.......................

3/4 8-7a 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS....................

3/4 8-11 J

WATERFORD - UNIT 3 XXII Amendment No. 9, 23, 58, gg, 75

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loop (s)/trai1(s) listed below shall be OPERABLE and at least one reactor coolant and/or sautdova cooling loops shall be in operation.*

a.

Reactor Coolant Loop 1 and its associated steam generator and at least one associated reactor coolant pump,**

b.

Reactor Coolant Loop 2 and its associated steam generator and at least one associated reactor coolant pump,**

c.

Shutdown Cooling Train A, d.

Shutdown Cooling Train B.

APPLICABILITY: MODE 4 ACTION:

a.

With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

With no reactor coolant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

• All reactor coolant pumps and shutdown cooling pumps (LPSI pumps) may be deenergized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 4

(2) core outlet temperature is maintained at least 10*F below saturation temperature.

• • A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 272*F unless l

(1) the pressurizer water volume is less than 900 cubic feet or (2) the secondary water temperature of each steam generator is less than 100*F above each of the Reactor Coolant System cold leg temperatures.

WATERFORD - UNIT 3 3/4 4-3 Amendment No.106 i

a

., l REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE REQUIREMENTS 4.4.1.3;1 The required reactor coolant pump (s), if not in operation, shall be; determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.3.2. The required steam generator (s) shall be' determined OPERABLE by verifying the secondary side water level to be > 50% of wide range indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.1.3.3 At least one reactor coolant or shutdown cooling loop shall be verified to be in operation and. circulating reactor coolant at least once per i

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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WATERFORD - UNIT 3 3/4 4-4

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REACTOR COOLANT SYSTEM

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COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION

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3.4.1.4 At leatt two of,the loop (s)/ trains listed below shall be OPERABLE 'and.

-at'least ore reactor coolant and/or shutdown cooling loop shall be in:

e operatioa.*

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Reactor Coolant Loop-1-and its associated steam generator and at i

a.

least one; associated reactor coolant pump **,

b.

Reactor Coolant Loop 2 and.Its associated-steam generator and at least one associated reactor' coolant pump **,

c.

Shutdown Cooling Train A, d.

Shutdown Cooling Train 8.

d APPLICABILITY: MODE 5 with reactor coolant loops filled **.

ACTION:.

a.

With less than the above required reactor coolant and/or shutdown cooling loops OPERABLE or with less than'the required steam generator level, immediately initiate corrective action to return 3

the required loops to OPERABLE status or to restore the required 1

level as soon as possible.

b.

With no reactor coolant or shutdown cooling loop in operation, suspend all operations involving a reduction in boron 1

concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.

1 l

SURVEILLANCE REOUIREMENTS

'4.4.1.4.1 The required reactor coolant pump (s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.4.2 The required steam generator (s) shall be determined OPERABLE by verifying the secondary side water level to be ;t50% of wide ~ range indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1 4.4.1.4.3 At least one reactor coolant loop or shutdown cooling train shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

• All reactor coolant pumps and shutdown cooling pumps (LPSI pumps) may be deenergized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and i

(2) core. outlet temperature is maintained at least 10*F below saturation c

temperature.

L

• • A reactor coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or. equal to 272*F unless l

(1) the pressurizer water volume is less than 900 cubic feet or (2) the secondary water temperature of each steam generator is less than 100*F above each of the Reactor Coolant System cold leg temperatures.

WATERFORD - UNIT 3 3/4 4-5 Amendment No.106 i

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r REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.5 Two shutdown cooling loops shall be OPERABLE # and at least one shutdown cooling loop shall be in operation.*

APPLICABILITY:

MODE 5 with reactor coolant loops not filled.

ACTION:

a.

With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b.

With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.5 At least one shutdown cooling loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

1. 0ne shutdown cooling loop may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other shutdown cooling loop is OPERABLE and in operation.

• The shutdown cooling pump (LPSI pump) may be deenergized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core' outlet temperature is maintained at least 10'F below saturation temperature.

WATERFORD - UNIT 3 3/4 4-6

REACTOR COOLANT SYSTEM SURVEILLAhCE REQUIREMENTS (Continued)

Reactor Vessel material surveillance program - withdrawal schedule in FSAR Table 5.3-10.

The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3.

4 4

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WATERFORD - UNIT 3 3/4 4-29 Amendment No.106

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%CATED RCS TEMPERATURE, Tc, 'F FIGURE 3.4-2 WATERFORD UNIT 3 HEATUP CURVE REACTOR COOLANT SYSTEM PRESSURE-TEMPERATURE UMITS 0-15 EFPY CURVE BA515: PEAK SURFACE FLUENCE = 2.29 x 1019 rVcm2 e 20 EFPY WATERFORD - UNIT 3 3/4 4-30 Amendment No.106 m

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WATERFORD UNIT 3 COOLDOWN CURVE REACTOR COOLANT SYSTEM PRESSURE TEMPERATURE LIMITS 0-15 EFPY CURVE BASIS: PEAK SURFACE FLUENCE = 2 29 x 1019 n/cm2 e 20 EFPY l

WATERFORD - UNIT 3 3/4 4-31 Amendment No.106 l

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l WATERFORD - UNIT 3 3/4 4-32 Amendment No. E,1061 l

u REACTOR COOLANT SYSTEM t

~ PRESSURIZER HEATUP/COOLDOWN LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer shall be limited to:

a.

A maximum heatup rate of 200'F per hour, b.

A maximum cooldown rate of 200*F per hour, and

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c.

A maximum spray nozzle usage factor of 0.65.

APPLICABILITY: At all times.

1 ACTION:

a.

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine,that the pressurizer remains acceptable for continued operation or be in at least H0T STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the spray nozzle usage factor > 0.65, comply with requirements of Table 5.7-1.

SURVEILLANCE REQUIREMENTS 4.4.8.2.1 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

i 4.4.8.2.2 The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.

4.4.8.2.3 Each spray cycle and the corresponding AT (water temperature differential) shall be recorded whenever main spray is initiated with a AT (water temperature differential) of > 130'F and whenever auxiliary spray is initiated with a AT (water temperature differential) of > 140'F.

]

i WATERFORD - UNIT 3 3/4 4-33

~_.. _. -

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.8.3 Two Shutdown Cooling (SDC) System suction line relief valves and SI-4068) shall be OPERABLE with a lift setting of less than or equa(SI-406A l to 430 psia.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 272*F#, MODE 5, and MODE 6 when the head is on the reactor vessel and l the RCS is not vented through a 5.6 square inch or larger vent.

ACTION:

With one SDC System suction line relief valve inoperable in MODE 4, a.

restore the inoperable valve to OPERABLE status within 7 days, or depressurize and vent the RCS through at least a 5.6 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

i b.

With one SDC System suction line relief valve inoperable in MODES 5, or 6, either (1) restore the inoperable valve to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or (2) complete depressurization and venting of the RCS through at least a 5.6 square inch vent within a total of 32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />.

With both SDC System suction line relief valves inoperable, complete c.

depressurization and venting of the RCS through at least a 5.6 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

d.

In the event either the SDC System suction line relief valve (s) or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SDC System suction line relief valve (s) or RCS vent (s) on the transient, and any corrective action necessary to prevent recurrence.

e.

The provisions of Specification 3.0.4 are not applicable.

1. 260'F during inservice leak and hydrostatic testing with Reactor Coolant System temperature changes restricted in accordance with Specification 3.4.8.19 WATERFORD - UNIT 3 3/4 4-34 Amendment No. Gr72,106

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REACTOR COOLANT SYSTEM BASES The heatup and cooldown limit curves Figures 3.4-2 and 3.4-3 are i

composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 60*F per hour or cooldown rate of up to 100*F per hour. The heatup

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and cooldown curves were prepared based upon the_most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figures 3.4-2 and 3.4-3.

The limitations on the Reactor Coolant System heatup and cooldown rates are further restricted due to stress limitations in the Reactor Coolant Pump. As part of the LOCA support scheme, the Reactor Coolant Pump has a ring around the suction nozzle of the pump.

The support skirt is welded to the ring. Due to this design, the heatup and cooldown rates must be limited to maintain acceptable thermal stresses.

The reactor vessel materials have been tested to determine their initial RT,7; the results of these test are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT,7 Therefore, an adjusted reference temperature, based upon the fluence, copper and nickel content of the material in question, l can be predicted using FSAR Table 5.3-1 and the recommendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials."

TheheatupandcooldownlimitcurvesFigures3.4-2and3.4-3includepredictedl adjustments for this shift in RT at the end of the applicable service period, as well as adjustments f,or possible errors in the pressure and 7

temperature sensing instruments.

1 i

The actual shift in RT of the vessel material will be established periodically during operati$n by removing and evaluating, in accordance with ASTM E185-82 and 10 CFR Part 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the i

reactor vessel in the core area. The surveillance specimen withdrawal schedule is shown in FSAR Table 5.3-10.

Since the neutron spectra at the l

irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must i

be recalculated when the delta RT determined from the surveillance capsule is different from the calculated Ielta RT., for the equivalent capsule radiation exposure.

i 4

The pressure-temperature limit lines shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR Part 50.

WATERFORD - UNIT 3 8 3/4 4-7 Revi: d by 10tter d t:d h;=t 20, 1007 Amendment No. 106

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4 9

REACTOR COOLANT SYSTEM

, l BASES PRESSURE / TEMPERATURE LIMITS (Continued) l materials, with the., for all Reactor Coolant System pressure-retaining The maximum RT exception of the reactor pressure vessel, has been i

determined to_be 90*F.

The Lowest Service Temperature limit line shown on Section III of the ASME Boiler and Pressure V., since Article NB-2332 ofessel Cod Figures 3.4-2 and 3.4-3 is based upon this RT i

Service Temperature to be RT., + 100*F for piping, pumps, and valves.

Below this temperature, the system pressure must be limited to a maximum of 20% of 4

the system's hydrostatic test pressure of 3125 psia (as corrected.for elevation and instrument error).

5 The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of the shutdown cooling system relief valve or an RCS j

vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal

)

to 272*F. Each shutdown cooling system relief valve has adequate relieving l

capability to protect the RCS from overpressurization when the transient'is

)

either (1) the start of an idle RCP with the secondary water temoerature of the steam generator less than or equal to 100*F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with injection into a water-solid RCS. The limiting transient includes simultaneous, inadvertent operation of three HPSI pumps, three charging pumps, and all pressurizer backup heaters in operation. Since SIAS starts only two HPSI pumps, a 20%

margin is realized.

4 The restrictions on starting a reactor coolant pump in MODE 4 and with the reactor coolant loops filled in MODE 5, with one or more RCS cold legs less than or equal to 272*F, are provided in Specification 3.4.1.3 and 3.4.1.4 l to prevent RCS pressure transients caused by energy additions from the secondary system which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures. Maintaining the steam generator less than 100*F above each of the Reactor Coolant System cold leg temperatures (even with the RCS filled solid) or maintaining a large surge volume in the pressurizer ensures that this transient is less severe than the limiting transient considered above.

l

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WATERFORD - UNIT 3 B 3/4 4-10 Amendment No. M,106

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