ML20101A428

Amends 95 & 89 to Licenses NPF-35 & NPF-52,respectively, Revising TS in Response to Guidance of Generic Ltr 90-06 to Enhance Reliability of PORVs & Block Valves & Providing Addl low-temperature Overpresure Protection

ML20101A428
Person / Time
Site:	Catawba
Issue date:	04/14/1992
From:	Matthews D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20101A430	List: ML20101A428 ML20101A434
References
GL-90-06, GL-90-6, NUDOCS 9204220140
Download: ML20101A428 (12)

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UNITED STATES i

NUCLEAR REGULATORY COMMISSION

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DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION Mt00ARIVERELECTRICCOOPERATIVE,INC.

DOCKET NO. 50 413 CATAWBANUCLEARSTATION,UNITJ AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 95 License No. NPF-35 1.

The Nuclear Regulatory Comission (the Comission) 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 Power Company, acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc.

(licensees) dated May 9, 1991, as supplemented on February 6, 1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Connission'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 (1) 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 Comission's regulations set forth in 10 CFR Chapter 1; l

l D.

The issuance of this amendment will not be inimical to the comon defense and security or to the health and safu / of the public; and l

E.

The issuance of this ameridment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have l

been satisfied.

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Accordirgly, the license is tereby amended by page changes to the Technical Specifications as indicated in thc attachtnent to this license onencraerit, and Parograph 2.C.(?) of facility Operating License No. NPT-35 is hereby en,end. i to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 95, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Company 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 Adb David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects.1/ll Office of Nuclear Reactor Regulation

Attachment:

Technical Specification

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Date of Issuance: April 14, 1992 1

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DUKE POWER COMPANY NORTH CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUH1CIPAL POWER AGENCY 00CKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 89 License No. NPF-52 1.

The N s. lear R*quietory Comission (the Comission) has found that:

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating iicense No. NPF-52 filed by the Duke Power Company, acting for itself, North Carolina Municipal Power Agency No I and Piedmont Municipal Power Agency (licensees) dated May 9,1991 as supplemented on February 6,1992, complies with the standcrds and requirements of the Atomic Energy Act of 1954, as anended (the Act), and the Comission'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 s

Comission; C.

There is reasonable assurance (1) 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 Comission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendment will not be inimical to the common defense end 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 Comission'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 attachn.ent to this license amendnent, and Paragraph 2.0 (2) of facility Operating License No. IlPF-52 is hereby arter:ded to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Arrendment t!o.

89, onc the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Cornpany shall cperate the facility in accordance with it Technical Specifications and the Environmental Protection Plan.

3.

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

F0P THE I40 CLEAR PEGULATORY COMMIS$10N WN

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V-Davio B. Matthews, Director Project Directorat? 11-3 Division of Reactor Projects-1/11 0'fice of !<uclear Reactor Regulation Attachinent:

Technical Specification Changes Date of issuance: April 14, 1992 l

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d ATTACHMENT T0__ LICENSE AMENDHENT N0. 95 FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO 50-413 e

E TO LICENSE AMENDMENT NO. 89 FACIL_1TY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

Remove Pages insert Pages 3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 3/4 4-37 3/4 4-37 3/4 4-38 3/4 4-38 8 3/4 4-2 8 3/4 4-2 B 3/4 4-3 B 3/4 4-3 8 3/4 4-3a B 3/4 4-3a s

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REACTOR COOLANT SYSTEM 3/4.4,4 RELIEFVALVEj LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABILITY:

H0 DES 1, 2, and 3.

ACTION:

a.

With one or more PORV(s) inoperabin, because of excessive seat leak-age, within I hour either restore the PORV(s) to OPERABLE status or closa the associated block valve (s) with power maintained to the block valve (s); otherwise. be in at least HOT 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.

With one or two PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); restore the PORV(s) to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT 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 in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c.

With three PORVs inoperable due to causes other than excessive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore at least one PORY to OPERABLE status or close their associated block valves and remove power from the block valves and be in Hf1T STANDBY within the next 6 hot.rs and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d.

With one or more block valvd s) inoperable and not closed, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore the block valve (s) to OPERABLE siaIus, or place its associated PORV switch (es) in the 'close' position.

Restore at least one block valve to OPERABLE status within the next hour if three block valves are moperable; restore any remaining inoperable block valve (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in HOT STAND 8Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The pvovisions of Speification 3.0.4 are not applicable.

e.

l CATAWBA - UNITS 1 & 2 S/4 4-10 Amendment No. 95 Unit 1)

Amendment No. 89 Unit 2)

REACTOR COOLANT SYSTEM Sj!RJEILLANCEREQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be denionstrated OPERABLE at least once per 18 months by:

a.

Performance of a CHANNEL CALIBRATION, and b.

Operating the valve through one complete cycle of full travel *.

I 4.4.4.2 Each block valve shall be demonstrated OfERABLE at least once per 92 days by operating the valve tnrough one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of ACTION b. or c. in Specification 3.4.4.

4.4.4.3 The safety related nitrogen supply for the PORVs shall be demonstrated l

OPERABLE at least once per 18 months by:

a.

Manually transferring motive power from the normal (air) supply to the emergency (nitrogen) supply, b.

Isolating and venting the normal (air) supply, and c.

Operating the valves through a complete cycle of full tr m 1.

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• In order to simulate environmental effects representative of operating conditions SR 4.4.4.lb should be conducted when the reactor coolant system temperature is greater than 200 F; however this SR shall not be performed in H0 DES 1 or 2.

l CATAWBA - UNITS 1 & 2 3/4 4-11 Amendment No. 95 Unit 1 Amendment No. 89 Unit 2

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the followinq

.ressure Protection Systems shall be OPERABLE:

a.

Two power operated relief valves (PORVs) with a lift setting of less than or equal to 450 psig, 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 Teg is less than or equal to leS'F, MODE 5 and MODE 6 when the head is on the reactor vessel.

ACTION:

a.

With one PORV ihuperable in H0DE 4, restore the inoperable PORV to OPERABLE status within 7 days or complete depretsurization and venting of the Reactor Coolant System through at least 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 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 vtnting 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 />.

d.

In the event either the PORVs or the Reactor Coolant System vent (s) l are used to mitigate a Reactor Coolant System 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 PORVs or Reactor Coolant System vent (s) on the transient, and any corrective action necessary to prevent recurrence, e.

The provisions of Specification 3.0.4 are not applicable.

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CATAWBA - UNITS 1 & 2 3/4 4-37 Anendment No. 95 (Unit 1)

Amendment No. 89 (Unit 2)

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, ac 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, k

.4.4.9.'3.2 The Reactor Coolant System vent (s) shall be verified to be opan 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.

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.

d-CATAWBA - UNITS 1 & 2 3/4 4-38 Amendment No.95 (Unit 1)

Amendment No.89 (Unit 2)

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e ret.CTOR COOLANT $YstEM BASES SAFETY VALVES (Continued) relief capability and will prevent overpressurization.

In addition, the Overpressure Protection System provides a diverse means of protection against overpressurization at low temperatures.

During operation, all pressurizer Code safety valves must be 0,ERABLE to prevent the Ruactor Coolant System from being pressurized above its Safety Limit of 2' psig.

The combined relief capacity of all of these valves is greater tn.. the maximum surge rate resulting from a complete loss-of+ load assuming no Reactor trip until the first R(actor Trip System Trip Setpoint is recched (i.e., no credit is taken for a direct Reactor trip on the loss-of-load) and al a assuming no operation of the power-operated relief valves or steam dump valves.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.

3/4.4.3 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the nurmal steady-state envelope of operation assumed in the SAR.

The limit is consistent with the initial SAR assumptions, t

The 12-hour periodic surveillance is sufficient to ensure that the parameter I

is restored to within its limit following expected transient operation.

The maximum water volume also ensures that a steam bubble is formed and thLs the Reactor Coolant System is not a hydraulically solid system.

The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability l

of the plant to control Reactor Coolant System pressure and establish natural l

circulation.

3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to re-i lieve Reactor Coolant System pressure during all design transients up to and including the design step load decrease with steam dump.

Each PORV has a l

remotely operated block valve to provide a positive shutoff capability should l

I a relief valve"t>ecome inoperable.

Tho OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions:

1) Manual control of PORVs to control Reactor Coolant i

System pressure.

This is a function that is used for the steam generator tube I

rupture accident coincident with a loss of all offsite power and for plant shut-l down.

2) Maintaining the integrity of the reactor coolant pressure boundary.

This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage.

3) Manual control of the block valve to unblock an isolated PORV to allow it to be used for manual control of Ractor Coolant System pressure and isolate a PORV with excessive seat leakage.

4) Automatic control of PORVs to control CATAWBA - UNITS 1 & 2 B 3/4 4-2 Amendment No. 95 (Unit 1)

Amendment No. 89 (Unit 2)

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a R_E CTOR ?00LANT SYSTEM Q

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Ip 5ft 0 GENERATORS (Continued) reactor coolant syatem pressure except for limited periods where the PORV has tv en 'solated due to excessive seat leakage and except for limited periods

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he PORV and/or block volve is closed because of testi*g and is fully nf being returned to its normal alignment at any tir. provided that this

covered by an approved procedure.

This is a fu. tion that reduces to the coue safety vaivas for overpressurization events. 5) Manual

' a block valve to 1solate a stuck-open PORV.

Testing of the PORVs

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.e emergency N supply from the Cold leg Accumulators.

This test f

7 des d.es that the vaTves in the supply line operate satisfactorily and t

nonsafety portin of tne instrument air system is not ne:assary for

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JI operation.

e pi s c 3R.4.b.'EAM GENERATORS T = burveillance Requirements for inspection of the steam generator tubes J

m.sure that the structural integrity of this portion cf the Rractor Coolant

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Tystem will be maintai d.

The program for inservice inspection of steam generator tubes is bas on a modification of Regulatory Guide L83, Revision 1.

inservi.a inspection of steam generator tubing is essential in order to main-f tain surveillance of the conditions of the tubes in the avent that there is evidence ci mechanical damage or progresf ve degradction due to design, manu-facturing errors, or inservice conditions that 1 cad to corrosion.

Inservice f

inspection cf steam generator tubing also provides a Meens of characterizing the nature and cause of any 'ube degradation so that corrective measures can be taken.

The BM process (v tetnod equivalent) to the inspection methcd described in Topiccl Raport BAW-2045(P)-A will be u: ed.

Inservice inspection of steam gen-erator sleeves is also required to ensure RCS integrity.

Because tha sleeves introduce changes in the wall thickness and diameter, they ruduce the sensitivity of eddy current testing, therefore, special inspection methods mrt be usc%

A method is dascribed in Topical Report 3AW-2045(P)*A with supporting validation data that demonstrates the insp9ctability of toe sleeve and underlying tube.

As regited by NRC for licensees authorized to use this repair process, Catawba com-inits to validate the sdequacy of a6y systu that is used for periodic inservice inspectiens of the sleeves, and will evaluate and, as deemed appropriate by Duke Power Company, implement testing methods as octter methods are developed and validated for commercial use.

The plant is expected to be operated in a uanner such that the secondary 3

coolant will be saintained within thase chemistry licits found to result in negligible cortesion of the stees generitor tubes.

If the secondary coolant chemistry is not maintained within these limits, locall ed corrosion may likely result in stress corrosion cracking.

The es tent of cracking during plant opera-tion would be limited by the limitation ol steant generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (reactor-to-seccudary leakage = S00 gallons per day per steam generatur).

Cracks having a reactor-to-secondary leakuge less than this limit during operation will have an adeo.uate margin of safety to withstand the loitds imposed during nomal operation and by postulated accidents.

Opsrating plants have demonstrated that reactor-to-secondary ?eakage of 500 gallons per day per steem gerarator can readily be datected by raciation r:onitors o1r steam generatur bicwdown.

Idage in excess of this lirait will require plant shutdown and an unscheduled k action, during which the leaking tubes will be locatea and repaired.

CATAWBA - UNITS 1 & 2 8 3/4 4-3 Amendment No.95 (Unit 1)

Amendment No.89 (Unit 2)

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REACTOR COOLANT SYSTEM i-BASES STEC1 GENERATORS (Continued)

Wastaae-typa defects are unlik.ely with proper chemistry treatment of the secondary coolant.

Hovn :r, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Repair will b9 required for all tubes with imperfections exceeding the repair limit of 40" of the tube nominal wall thickness.

For Unit 1, defe:tive tubes which fall under the alternate tube plugging criteria do not have to be repaired.

Defec-tive steam generator tubes can be repaired by the installation of sleeves which span the area of degradation, and serve as a replacenent pressure boundary for the degraded portion of the tube, allowing the tube to remain in service.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect wastage type degradation that has penetrated 20% of the original tube wall thickness.

Whenever the results of ar,y steem generator tubing inservice inspection fall into Category C-3, these results will be reported to the Commission pur-suant to Specification 6.9.2 prior to resumption of plant operation.

Such cases will be considered by the Commission on a case-by-case basis and may result in a requirement for analysis, laboratory examinations, tests, additional eddy-current inspection, and revision of the Technical Specifications, if necessary.

If a tube is sleeved due to degradation in the F* distance, then any defects in the tube below the sleeve will remain in service without repair.

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The Leakage Detection Systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.

These Detection Systems are consistent with the recommendations of Regulatory Guide 1.45, " Reactor Coolant Pressure Boundary Leakage Detection Systems,"

May 1973.

CATAWBA - UNITS 1 & 2 B 3/4 4-3a Amendment No 95 (Unit 1)

Amendment No.89 (Unit 2) r