ML20238F608

Amends 170 & 162 to Licenses NPF-35 & NPF-52,respectively, Revising Figures 3.4-2 & 3.4-3,Table 4.4-5 & Sections 3/4.4.9.3 & 3.5.3 of TSs

ML20238F608
Person / Time
Site:	Catawba
Issue date:	08/28/1998
From:	Berkow H NRC (Affiliation Not Assigned)
To:
Shared Package
ML20238F610	List: ML20238F608 ML20238F618
References
NUDOCS 9809040133
Download: ML20238F608 (29)

Text

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,j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. *= -1 d

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4 DUKE ENERGY CORPORATION NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE. INC.

DOCKET NO. 50-413 CATAWBA NUCLEAR STATION. UNIT 1 1

AMENDMENT TO FACILITY OPERATING LICENSE

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i Amendment No.170 License No. NPF-35

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility)

Facility Operating License No. NPF-35 filed by the Duke Energy Corporation, acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc. (licensees), dated September 15,1997, as supplemented by letters dated March 5, April 27, June 15, July 22, and August 10,1998, complies with the j

standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

j and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.

9809040133 980828 PDR ADOCK 05000413 P

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, 2. Accordingly, the license is hereby amended by page changes to the Technical l

Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through' Amendment No. 170, which are attached hereto, are hereby incorporated into this license. Duke Energy Corporation shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGUI.ATORY COMMISSION 7

Herbert N. Berkow, Director Project Directorate ll-2 Division of Reactor Projects - 1/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification -

Changes Date ofissuance:

August 28, 1998

.i ATTACHMENT TO LICENSE AMENDMENT NO. 170

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FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 Replace the following pages of the Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

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3/4 4-32 (Figure 3.4-2) 3/4 4-32 (Figure 3.4-2) 3/4 4-33 (Figure 3.4-3) 3/4 4-33 (Figure 3.4-3) 3/4 4-34 (Table 4.4-5) 3/4 4-34 (Table 4.4-5) 3/4 4-36 3/4 4-36 3/4 4-36a 3/4 4-37 3/4 4-37 3/4 5-8 3/4 5-8 B 3/4 4-9 8 3/4 4-9 B 3/4 4-14 83/44-14 B 3/4 4-15 B 3/4 4-15 4

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SECTION 8 31 I

3/4.4.3 PRESSURIZER......................... 3/4 4-9 i

3/4.4.4 RELIEF VALVES.......................

3/4 4-10 f

3/4.4.5 STEAM GENERATORS 3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION 3/4 4-17 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 3/4 4-18 l

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................

3/4 4-19 Operational Leakage....................

3/4 4-20 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES....

3/4 4-22 3/4.4.7 CHEMISTRY.........................

3/4 4-24 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 3/4 4-25 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS......................

3/4 4-26 3/4.4.8 SPECIFIC ACTIVITY.....................

3/4 4-27 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131.................

3/4 4-28 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..........................

3/4 4-29 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System 3/4 4-31 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 15 EFPY 3/4 4-32 i

FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 15 EFPY 3/4 4-33 l

CATAWBA - UNIT 1 VII Amendment No.170

i LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION EAGE TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE 3/4 4-34 Pressurizer........................

3/44-35 Overpressure Protection Systems..............

3/44-36 TABLE 3.4-3 REACTOR COOLANT PUMP OPERATING RESTRICTIONS FOR LOW TEMPERATURE OVERPRESSURE PROTECTION 3/4 4-36a 3/4.4.10 STRUCTURAL INTEGRITY 3/44-38 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 3/44-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection....................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T.,, t 350*F................ 3/45-4 3/4.5.3 ECCS SUBSYSTEMS - T.,, < 350*F...............

3/4 5-8 3/4.5.4 REFUELING WATER STORAGE TANK 3/4 5-10 i

3/4.6 CONTAINMENT SYSTEMS i

3/4.6.1 PRIMARY CONTAINMENT Contai nment Integri ty.................... 3/46-1 Containment Leakage....................

3/4 6-2 TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS.......

3/4 6-5 Containment Ai r Locks...................

3/4 6-8 Internal Pressure.....................

3/46-10 Air Temperature......................

3/46-11 Containment Vessel Structural Integrity..........

3/46-12 l

Reactor Building Structural Integrity...........

3/46-13 Annulus Ventilation System 3/46-14 Containment Purge Systems.................

3/46-16

. CATAWBA - UNIT 1 VIII Amendment No. 170 L________.__.___._.___________________

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MATERIAL PROPERTY BASIS UMITING MATERIAL: LOWER SHELL FORGING 04 UMITING ART AT 15 E9Y:

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CATAWBA - UNIT 1 3/4 4-32 Amendment No.170

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CATAWBA - UNIT 1 3/4 4-33 Amendment No.170

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,e REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 A Low Temperature Overpressure Protection System shall be OPERABLE with a maximum of one charging pump or one safety injection pump capable of injecting into the Reactor Coolant System, the cold leg accumulators isolated *,

Reactor Coolant pump operation limited as specified in Table 3.4-3, and either a or b below:

Two power operated relief valves (PORVs)# with a lift setting of less a.

than or equal to 400 psig (as left calibrated), allowable value of less than or equal to 425 psig (as found), or b.

The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABILITY: MODE 4 when the temperature of any Reactor Coolant System cold i

leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a.

With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 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 />, b.

With one PORV inoperable in MODES 5 or 6, restore the inoperable PORV 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 complete depressurization and venting of the Reactor Coolant System through at least a 4.5 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 />.

With both PORVs inoperable, complete depressurization and venting of c.

the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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• Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum Reactor Coolant System pressure for the existing Reactor Coolant System cold leg temperature allowed by the pressure / temperature limit curves provided in Specification 3/4.4.9.

1. Phen using the POP.Vs to meet the requirements of this Specification, each Reactor Coolant System cold leg temperature shall be greater than or equal to 65'F. When the Reactor Coolant pumps are secured, this temperature shall be measured at the residual heat removal heat exchanger outlet.

CATAWBA - UNIT 1 3/44-36 Amendment No.170 L

,e REACTOR COOLANT SYSTEM ACTION (Continued)

-d.

In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor Coolant System pressure transient, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence, The provisions of Specification 3.0.4 are not applicable, e.

f.

With more than the specified number of charging pumps and/or safety injection pumps capable of injecting into the Reactor Coolant System *,

immediately initiate action to verify a maximum of one charging pump or one safety injection pump is capable of injecting into the Reactor Coolant System, g.

With a cold leg accumulator not isolated as specified, within one hour, isolate the affected accumulator.

If the affected accumulator is not isolated within one hour, within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either J

increase Reactor Coolant System cold leg temperature to greater than i

285*F, or depressurize the affected accumulator to less than the maximum Reactor Coolant System pressure for the existing cold leg temperature allowed by Specification 3/4.4.9.

I h.

With Reactor Coolant pump operation not in accordance with the requirements of Table 3.4-3, imediately initiate action to limit pump operation as specified in Table 3.4-3.

TABLE 3.4-3 REACTOR COOLANT PUMP OPERATING RESTRICTIONS FOR LOW TEMPERATURE OVERPRESSURE PROTECTION Reactor Coolant System Maximum Number of Pumos Cold Leo Temperature Allowed in Ooeration i

E 67'F 1

2 68'F 2

2 73*F 4

• Two charging pumps may be capable of injecting into the Reactor Coolant System during pump swap operation for less than or equal to 15 minutes.

CATAWBA - UNIT 1 3/44-36a Amendment No.170

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REACTOR COOLANT SYSTEM M LANCE REQUIREMENTS i

4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

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when the PORV is being used for overpressure protection.

4.4.9.3.2 The Reactor Coolant System vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the vent (s) is being used for overpressure protection.

4.4.9.3.3 A maximum of one charging pump or safety injection pump shall be verified to be capable of injecting into the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.9.3.4 Each cold leg accumulator shall be verified to be isolated at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when accumulator isolation is required.

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• Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

l CATAWBA - UNIT 1 3/44-37 Amendment No. 170 1'

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EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T., < 350'F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump,#

b.

One OPERABLE residual heat removal heat exchanger, I

.c.

One OPERABLE residual heat removal pump, and

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l d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

l APPLICABILIH:

MODE 4.

ACTION:

l a.

With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no ECCS. subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T,., less than 350*F by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to j

the Conunission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation'and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

1. A maximum of one charging pump or one Safety Injection pump shall be OPERABLE l

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whenever the temperature of one or more of the Reactor Coolant System cold legs is less than or equal to 285'F.

CATAWBA - UNIT 1 3/45-8 Amendment No. 170

.e REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of the effective full power years (EFPY) of service life as indicated on the appli-cable heatup or cooldown curves. The service life period is chosen such that the limiting RTuo7 at the 1/4T location in the core region is greater than the RTNDT of the limiting unirrediated material.

The selection of such a limiting RTNDT assures that all cogonents in the Reactor Coolant System will be operated conservatively,in accordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initial RTuo7; the results of these tests are shown in Table B 3/4.4-1.

Reactor operation and resultant fast neutron (E greater than 1 Mev) irradiation can cause an increase in the RTuo7 Therefore, an adjusted reference temperature, based upon the fluence, copper content, and phosphorus content of the material _ in question, can be predicted using Figure B 3/4.4-1 and the largest value of ARTNDT-The adjusted reference temperature has been computed using the guidance of Regulatory Guide 1.99, Revision 2.

The heatup and cooldown limit curves in Figures 3.4-2 and 3.4-3 include predicted adjustments for the shift in RT at the end of the identified service life.

NDT Capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR Part 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in Table 4.4-5.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel. Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure vessel material by using the lead factor and the withdrawal time of the capsule.

The. heatup and cooldown curves must be re:alalated when the ARTuor determined from the surveillance capsule exceeds t'e calculated ARTnot for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G i

to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

l l

1 CATAWBA - UNIT 1 B3/44-9 Amendment No. 170

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure.

The metal temperature at the crack tip lags'the coolant temperature; therefore, the K g for the 1/4T crack i

for the 1/4T crack during steady-state during heatup is;10wer tha'n the Kig conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive

- thermal stresses and different K g's for steady-state and finite heatup rates i

do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is. considered.

Therefore, both cases have to be analyzed in. order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion-of the hea' tup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed. Unlike the situation at the vessel inside surface, the themal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses present. These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.

Furthemore, since the thermal stresses, at the outside are tensile ano increase with increasing heatup rate, a lower bound' curve cannot be defined..

Rather, each heatup rate of interest must be analyzed on an' individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are pro-duced as follows. A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and'the pressure limit must at all times be based on analysis of the most critical criterion.

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CATAWBA - UNIT 1 B 3/4 4-14 Amendment No.170 l

REACTOR COOLANT SYSTEM BASES j

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PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis perfomed in accordance with the ASME Code requirements.

LOW TEMPERATURE OVERPRESSURE PROTECTION l

The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G j

to 10 CFR Part 50 when one or more of the cold legs are less than or equal to 1

285'F.

Either PORV has adequate relieving cepability to protect the Reactor l

Coolant System from overpressurization when the transient is limited to either:

(1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50*F above the cold leg temperatures, or (2) the start of a Safety Injection pump and its injection into a water solid Reactor Coolant System.

The Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the perfomance

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of the LTOPS assuming various mass input and heat input transients and i

incorporates instrument uncertainties as well as corrections for Reactor Coolant Pump operation and the static pressure difference between the Reactor Vessel Beltline Region and the location of the pressure transmitters used for LTOP. Operation with a PORV Setpoint less than or equal to the maximum allowable value of 425 psig (as found) ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat input transients more severe than those assumed cannot occur, Technical Specifications require lockout of all but one Safety Injection pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50*F above primary temperature.

The Maximum Allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens perfomed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in T,able 4.4-5.

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I CATAWBA - UNIT 1 B3/44-15 Amendment No. 170 I

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UNITED STATES f

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4001

%...../

DUKE ENERGY CORPORATION NORTH CAROLINA MUNICIPAL POWER AGENCY nod PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 162 License No. NPF-52

1. The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility)

Facility Operating License No. NPF 52 filed by the Duke Energy Corporation, acting for itself, North Carolina Municipal Power Agency No.1 and Piedmont Municipal Power Agency (licensees), dated September 15,1997, as supplemented by letters dated March 5, April 27, June 15, July 22, and August 10,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; 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.

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, 2. Accordingly, the license is hereby amended by page 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-52 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.162, which are attached hereto, are hereby incorporated into this license. Duke Energy Corporation shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 60 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Herbert N. Berkow, Director J

Project Directorate ll-2 1

Division of Reactor Projects - 1/II j

Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes Date of issuance:

August 28, 1998 J

,a 1

l ATTACHMENT TO LICENSE AMENDMENT NO.162 FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Technical Specifications with the enclosed pages. The revised pages are dentified by Amendment number and contain vertical lines indicating the areas of change.

Remove Insert Vil Vil Vill Vill 3/4 4-33 (Figure 3.4-2) 3/4 4-33 (Figure 3.4-2) 3/4 4-34 (Figure 3.4-3) 3/4 4-34 (Figure 3.4-3) 3/4 4-35 (Table 4.4-5) 3/4 4-35 (Table 4.4-5) 3/4 4-37 3/4 4-37 3/4 4-37a 3/4 4-38 3/4 4-38 3/4 5-8 3/4 5-8 B 3/4 4-9 B 3/4 4-9 B 3/4 4-14 B 3/4 4-14 B 3/4 4-15 B 3/4 4-15

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION EASE 3/4.4.3 PR ESSUR I Z ER........................

3 /4 4 - 9 3/4.4.4 RELIEF VALVES.......................

3/4 4-10 3/4.4.5 STEAM GENERATORS 3/4 4-12 TABLE 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION 3/4 4-18 TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION 3/4 4-19 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................

3/4 4-20 Operational Leakage....................

3/4 4-21 TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES....

3/4 4-23 3/4.4.7 CHEMISTRY.........................

3/4 4-25 TABLE 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS 3/4 4-26 TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS......................

3/4 4-27 3/4.4'.8 SPECIFIC ACTIVITY 3/4 4-28 FIGURE 3.4-1 DOSE EQUIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >l pCi/ gram DOSE EQUIVALENT I-131.................

3/4 4-29 TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 3/4 4-30 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System 3/4 4-32 FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS APPLICABLE UP TO 15 EFPY 3/4 4-33 l

FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS -

APPLICABLE UP TO 15 EFPY 3/4 4-34 l

CATAWBA - UNIT 2 VII Amendment No. 162 l

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION P.8E TABLE 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM WITHDRAWAL SCHEDULE 3/44-35 P re s s uri ze r........................

3/44-36 Overpressure Protection Systems..............

3/44-37 TABLE 3.4-3 REACTOR COOLANT PUMP OPERATING RESTRICTIONS FOR LOW TEMPERATURE OVERPRESSURE PROTECTION 3/4 4-37a 3/4.4.10 STRUCTURAL INTEGRITY 3/44-39 3/4.4.11 REACTOR COOLANT SYSTEM VENTS 3/44-40 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS Cold Leg Injection 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T., t 350*F................ 3/45-4 3/4.5.3 ECCS SUBSYSTEMS - T, < 350*F................ 3/4 5-8 3/4.5.4 REFUELING WATER STORAGE TANK 3/4 5-10 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Contai nment Integri ty.................... 3/4 6-1 Containment Leakage....................

3/4 6-2 TABLE 3.6-1 SECONDARY CONTAINMENT BYPASS LEAKAGE PATHS.......

3/4 6-5 Containment Ai r Locks...................

3/4 6-8 Internal Pressure.....................

3/46-10 Ai r Tempe ra tu re......................

3/46-11 I

Containment Vessel Structural Integrity..........

3/46-12 Reactor Building Structural Integrity...........

3/4 6-13 Annulus Ventilation System 3/46-14 Containment Purge Systems.................

3/46-16 CATAWBA - UNIT 2 VIII Amendment No.162

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MATERIAL PROPERTY BASIS LIMI!1NG MATERIALS: INTERMEDIATE SHELL. B8605 2 l

LIMITING ART AT 15 EPPY:

1/4-L 112.6 7 3/4.L 96.0 7 l

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50 100 150 200 250 300 35D 400 450 500 l

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FIGURE 3.4-2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (Heatup rates up to 60*F/hr)

APPLICABLE FOR THE FIRST 15 EFPY (Without Margins For Instrumentation Errors)

CATAWBA - UNIT 2 3/4 4-33 Amendment No.162 i

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MATERIAL PROPERTY BASIS lad! TING MATERIALS DEERMEDIATE SHE11 B8605 2 LDd! TING ART AT 15 EFFY:

1/4-1, 112.6 7 3/4-t, 96.0 7 2,500-l 2,250 I

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FIGURE 3.4-3 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS (Cooldown rates up to 100'F/hr)

APPLICABLE FOR THE FIRST 15 EFPY (Without Margins For Instrumentation Errors)

CATAWBA - UNIT 2 3/4 4-34 Amendment No.162

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 A Low Temperature Overpressure Protection System shall be OPERABLE with a maximum of one charging pump or one safety injection pump capable of injecting into the Reactor Coolant System,.the cold leg accumulators isolated *, Reactor Coolant pump operation limited as specified in Table 3.4-3, and either a or b below:

Two power operated relief valves (PORVs)# with a lift setting of less a.

than or equal to 400 psig (as left calibrated), allowable value of less than or equal to 425 psig (as found), or b.

The Reactor Coolant System depressurized with a Reactor Coolant System vent of greater than or equal to 4.5 square inches.

APPLICABILITY:

MODE 4 when the temperature of any Reactor Coolant System cold leg is less than or equal to 285'F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a.

With one PORV inoperable in MODE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depressurization and venting of the Reactor Coolant System through at least a 4.5 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 />.

b.

With one PORV inoperable in MODES 5 or 6 restore the inoperable PORV

-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 complete depressurization and venting of the Reactor Coolant System through at least a 4.5 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 />, c.

With both PORVs inoperable, complete depressurization and venting of the Reactor Coolant System through at least a 4.5 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

• Accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum Reactor Coolant System pressure for the existing Reactor Coolant System cold leg temperature allowed by the pressure /temperaturelimitcurvesprovidedinSpecification3/4.4.9.

1. When using the PORVs to meet the requirements of this Specification, each Reactor Coolant System cold leg temperature shall be greater than or equal to 65'F.

When the Reactor Coolant pumps are secured, this temperature shall be l

measured at the residual heat removal heat exchanger outlet.

O i

REACTOR COOLANT SYSTEM ACTION (Continued 1 d.

In the event either the PORVs or the Reactor Coolant System vent (s) are used to mitigate a Reactor. Coolant System pressure transient, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence.

The provisions of Specification 3.0.4 are not applicable.

e.

f.

With more than the specified number of charging pumps and/or safety injection pumps capable ci injecting into the Reactor Coolant System *, immediately initiate action to verify a maximum of one charging pump or one, safety injection pump is capable of injecting into the Reactor Coolant System.

g.

With a cold leg accumulator not isolated as specified, within one hour, isolate the affected accumulator.

If the affected accumulator is not isolated within one hour, within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, either increase Reactor Coolant System cold leg temperature to greater than

'285*F, or depressurize the affected accumulator to less than the maximum Reactor Coolant System pressure for the existing cold leg temperature allowed by Specification 3/4.4.9.

h.

With Reactor Coolant pump operation not in accordance with the requirements of Table 3.4-3, immediately initiate action to limit pump operation as specified in Table 3.4-3.

TABLE 3.4-3 REACTOR COOLANT PUMP OPERATING RESTRICTIONS FOR LOW TEMPERATURE OVERPRESSURE PROTECTION Reactor Coolant System Maximum Number of Pumos Cold Lea Temperature Allowed in Ooeration 1 67'F 1

1 73'F

'2 2 95'F 4

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• Two charging pumps may be capable of injecting into the Reactor Coolant System during pump swap operation for less than or equal to 15 minutes.

CATAWBA - UNIT 2 3/4 4-37a Amendment No.162

REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

a.

Performance of an ANALOG CHANNEL OPERATIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per 31 days; b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months; and c.

Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the PORV is being used for overpressure protection.

4.4.9.3.2 The Reactor Coolant System vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the vent (s) is being used for overpressure protection.

4.4.9.3.3 A maximum of one charging pump or safety injection pump shall be verified to be capable of injecting into the Reactor Coolant System at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.4.9.3.4 Each cold leg accumulator shall be verified to be isolated at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when accumulator isolation is required.

l

• Except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

CATAWBA - UNIT 2 3/44-38 Amendment No.162

EMERGENCY CORE COOLING SYSTEMS 3/4.5.3 ECCS SUBSYSTEMS - T., < 350*F LIMITING CONDITION FOR OPERATION 3.5.3 As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

a.

One OPERABLE centrifugal charging pump,#

b.

One OPERABLE residual heat removal heat exchanger, c.

One OPERABLE residual heat removal pump, and d.

An OPERABLE flow path capable of taking suction from the refueling water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of operation.

APPLICABILITY: MODE 4.

ACTION:

With no ECCS subsystem OPERABLE because of the inoperability of a.

either the centrifugal charging pump or the flow path from the refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within I hour or be in COLD SHUTDOWN within the next 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no ECCS subsystem OPERABLE because of the inoperability of either the residual heat removal heat exchanger or residual heat removal pump, restore at least one ECCS subsystem to OPERABLE status or maintain the Reactor Coolant System T,., less than 350'F by use of alternate heat removal methods.

c.

In the event the ECCS is actuated and injects water into the Reacbr Coolant System, a Special Report shall be prepared and submitted to the Comission pursuant to Specification 6.9.2 within 90 days describing the circumstances of the actuation and the total accumu-lated actuation cycles to date.

The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

1. A maximum of one charging pump or one Safety Injection pump shall be OPERABLE l

whenever the temperature of one or more of the Reactor Coolant System cold legs is less than or equal to 285'F.

CATAWBA - UNIT 2 3/4 5-8 Amendment No. 162 L_____--------------_-__-_

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9 REACTOR COOLANT SYSTEM BASES

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PRESSURE / TEMPERATURE LIMITS (Continued) i Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTnny,dicated on the appli-at the end of the effective full power years (EFPY).of service life as in cable heatup or cooldown curves. The service life period is chosen such that the limiting RTnpy at. the 1/4T location in the core region is greater than the RTNDT of the limiting unirradiated material.

The selection of such a limiting RTNOT assures that all components in the Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

The reactor vessel materials have been tested to detemine their initial RTNDT;theresultsofthesetestsareshowninTableB3/4.4-1.

Reactor operation end resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RTuo7 Therefore, an adjusted reference

- temperature, based upon the fluence, copper content, and phosphorus content of the material in question, can be predicted using Figure B 3/4.4-1 and the

-largest value of ARTNDT-The adjusted reference temperature has been computed using the guidance of Regulatory Guide 1.99, Revision 2.

The heatup and cooldown limit curves in Figures 3.4-2 and 3.4-3 include predicted adjustments for the shift in RT at the end of the identified service life.

NDT Capsules will be removed in accordance with the requirements of ASTM E185-82 and 10 CFR Part 50, Appendix H.

The surveillance specimen withdrawal schedule is shown in Table 4.4-5.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel.

Therefore, the results obtained from the surveillance specimens can be used to predict future radiation damage to the pressure vessel material by using the lead factor and the withdrawal time of the capsule.

The heatup and cooldown curves must be recalculated when the-ARTuo7 detennined from the surveillance capsule exceeds the calculated ARTNDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Sec-tion III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

CATAWBA - UNIT 2 B3/44-9 Amendment No.162 l

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.o REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stresses at the inside of the wall that alleviate the tensile stresses produced by internal pressure. The metal temperature at the for the 1/4T crack crack tip lags the coolant temperature; therefore, the Kig for_the 1/4T crack during steady-state during heatup is lower than the Kig conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist such that the effects of compressive themal stresses and different Kig's for steady-state and finite heatup rates

-do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4T flaw is considered. Therefore, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce j

stresses which are tensile in nature and thus tend to reinforce any pressure stresses present.

These themel stresses, of course, are dependent on both the i

rate of heatup and the time (or coolant temperature) along the heatup ramp.

I Furthemore, since the themal stresses, at the outside are tensile and j

increase with increasing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are pro-

)

duced as follows.

A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data.

At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

CATAWBA - UNIT 2 B 3/4 4-14 Amendment No. 162

L REACTOR COOLANT SYSTEM BASES l

PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis l

perfomed in accordance with the ASME Code requirements.

LOW TEMPERATURE OVERPRESSURE PROTECTION The OPERABILITY of two PORVs or a Reactor Coolant System vent opening of at least 4.5 square-inches ensures that the Reactor Coolant System 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 cold legs are less than or equal to 285'F.

Either PORV has adequate relieving capability to protect the Reactor Coolant System from overpressurization when the transient is limited to either:

l (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50*F above the cold leg temperatures, or (2) the start of a Safety Injection pump and its injection into a water solid Reactor Coolant System.

The Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System (LTOPS) is derived by analysis which models the performance of the LTOPS assuming various mass input and heat input transients and incorporates instrument uncertainties as well as corrections for Reactor Coolant Pump operation and the static pressure difference between the Reactor Vessel Beltline Region and the location of the pressure transmitters used for LTOP. Operation with a PORV Setpoint less than or equal to the maximum allowable value of 425 psig (as found) ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV Setpoint which can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure.

To ensure that mass and heat input transients more severe than those assumed cannot occur Technical Specifications require lockout of all but one Safety Injection l

pump and all but one centrifugal charging pump while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of a RCP if secondary temperature is more than 50*F above primary temperature.

The Maximum Allowed PORV setpoint for the LTOPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens perfomed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 4.4-5.

(.

CATAWBA - UNIT 2 B3/44-15 Amendment No. 162