Text

Amends 68 & 62 to Licenses NPF-35 & NPF-52,respectively, Modifying Tech Specs to Add Requirements for Steam Generator power-operated Relief Valves
	ML20246B587
Person / Time
Site:	Catawba
Issue date:	08/15/1989
From:	Matthews D; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20246B591	List: ML20246B587; ML20246B607;
References
	NUDOCS 8908230403
	Download: ML20246B587 (11)
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UNITED STATES

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- DUKE POWER COMPANY NORTH CAROLINA ELECTRIC MEPPERSHIP CORPORATION

~ SALUDA RIVER ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-413 CATAWBA NUCLEAR STATICK, UNIT 1 AMENCMENT TO FACILITY OPERATING LICENSE Amen &ent No. 68 License No. NPF-35 1.

The Nuclear Fegulatory Commission (the Commission) has found that:

A.-

The application for amenhent 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 Decenter 4,1987, as supplemented December 7,11, and 29,1967, March 29, May 4 and 18, June 16, July 1, August S and 24,'and December-15, 1968, and June 12 and 28, 1989, complies with the staridards 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; 2.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Conmission; 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 conpliance 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 5'1 of the Ccomission's regulations and all applicable requirements have been satisfied.

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Accordingly, the'-license.is hereby anended by 'page changes to the Technical

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' Specifications as' indicated in the~ attachments to this license amendment, 4

andL Paragraph 2.C.(2) of Facility-Operating License No. NPF-35. is. hereby

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amended to read as follows:

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L (2) Technical Specifications = /

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<The Technical-Specifications contained in Appendix A, as revised through Amendment No.'- 68, are' hereby incorporated into the-license.

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.The licensee shall operate the facility:.in:accordance with the.

Technical Specifications and the Environmental Protection Plan.

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Thisilicense amendment is effective as of its date. of issuance.

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FOR THE NUCLEAR REGULATORY COMMISSION y.

Original sign'd by:

e David B. Matthews, Director' Project Directorate-11-3 Division of Reactor Projects-I/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes.

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LDate of Issuance:

August-15, 1989 0FFICIAL RECORD COPY

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UNITED STATES

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i NUCLEAR REGULATORY. COMMISSION WASHINGTON. D. C 20555

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l DOCKET hC. 50-414

. CATAWBA' NUCLEAF. ETATION, LtIT 2 AliENDMENT TO FACILITY OPERATING LICENSE Amerdment No. 62 License flo. NPF-52 1.

The f:uclear Regulatory Comission (the Commission) has found that:

The application for amendment to the Catawba Nuclear Station, Unit 2 A..

.(the facility) Facility Operating License No. NFF-52 filed by the Duke Power Company acting for itself, North Carolina-Municipal Power Agency No. I and Piedmont Municipal Power Agency, (licensees) dated December 4,1907, as supplemented Decenber 7,11, and 29,1987, March 29, May 4 and 18, June 16, July 1, August 8 and 24, and December 15, 1988, and June 12 and'28, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commissicn'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 Comission; 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; C.

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 Comission's regulations and all applicable requirements have been satisfied.

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i 21 Accordingly, the111 cense is hereby amended by page changes-to the Technical Specifications as indicated in the attachments to this license amendment,

- and Paragraph 2.C.(2) of Facility' Operating License No. NPF-52-is~ hereby amended to read as follows:

(2)' LTechnical Specifications

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K The' Technical Specifications contained in Appendix A, as revised through. Amendment No. ' 62, are hereby incorporated into the license.1 The licensee shall operate the facility'in accordance~with the Technical Specifications and the' Environmental. Protection Plan.

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

FOR THE NUCLEAR REGULATOPv COMMISSION-.

m Original signed by:

David B. Matthews, Director

. Project Directorate 3 Division of Reactor' Projects-1/II Office of Nuclear Reactor Regulation

Attachment:

Technical Specification Changes.

Date'of' Issuance: August 15, 1989 r

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. ATTACHMENT TO LICENSE AMENDMENT NO. 68 FACILITY OPEFATING LICENSE fl0. NPF-35 DOCKET flC. 50-413 AliD TO LICENSE Af'ENDMENT t!0. 62 F/CILITY CPERATING LICENSE NO. NPF-52 C0CKET N0. 50-414 Feplace 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. The corresponding over -

leaf-pages are also previded to caintain document completeness.

Amended Page Overleaf Page 3/4 7-9a (new page) 3/4 7-9 3/4 7-10 8 3/4 7-2 B 3/4 7-1 B 3/4-7-2a (new nace)

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PLANT SYSTEMS

- QNDENSATESTORAGESYSTEM

- LIMITING CONDITION FOR OPERATION

. 3.7.1.'S The Condensate Storage System (CSS) (CA Condensate Storage Tank, Upper Surge and Condenser'Hotwell) shall be OPERABLE with a contained water volume of at least 225,000. gallons of water.

. APPLICABILITY:

MODES 1, 2, and 3. (Unit 2)

ACTION:

With' the CSS inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a.

Restore the CSS to OPERABLE status or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 heurs, or b.

Demonstrate the OPERABILITY of the standby nuclear service water pond as ~ a backup supply to the auxiliary feedwater pumps and restore the CSS'to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

' SURVEILLANCE REQUIREMENTS

4. 7.1. 5 The CSS shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying the contained water volume is within its limits.

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STEAM: GENERATOR POWER OPERATED' RELIEF VALVES

. LIMITING CONDITION FOR OPERATION

. 3. 7.1. 6 Three steam generator power-operated'r elief valves (PORVs) and

' associated remote manual. controls,. including the safety-related gas supply c

' systems, shall'be OPERABLE; APPLICABILITY MODES 1, 2, 3, and 4.*

-ACTION:

With one'.less than the required steam generator PORVs OPERABLE, a.

restore the inoperable steam generator PORV to OPERABLE status

'within 7 days;:or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within.the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and place the required Residual Heat Removal loop in operation for. decay heat removal.

b.

.With two_less than the required steam generator PORVs OPERABLE, restore at-least one of the inoperable steam generator PORVs to'.

OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or.be in at least' HOT STANDBY within

'the next.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .and in H0T SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ' and place;the' required Residual Heat Removal loop in operation for decay.. heat removal.

SURVEILLANCE REQUIREMENTS 4.7.1.6 Each steam generator PORV and associated remote manual controls including the safety-related gas supply systems shall be demonstrated OPERABLE:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that at least one of the two nitrogen bottles associated with each PORV has a pressure greater

.than or equal to 2100 psig, and b.

. At least once per 18 months and prior to startup following any refueling shutdown by verifying that all steam generator PORVs will operate through one cycle of full travel using remote manual controls and safety-related gas supply.

When; steam generators are being used for decay heat removal.

-CATAWBA - UNITS 1 & 2 3/4 7-9a Amendment No. 68 (Unit 1)

Amendment No. 62 (Unit. 2) m h_____i__

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' PLANT'SYSTEMg H,

3/4.7.2oSTEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION-l LIMITING CONDITION >FOR OPERATION

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-3.7.2; The--temperatu'res of both'the. reactor and secondary coolants in the steam' generators shall be' greater.than 70*F.t: hen the pressure of either.

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coolant in the. steam generator is greater than 200 psig.

APPLICABILITY:

At all times.

ACTION:

-With thefrequirements of.the.above specification not satisfied:

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a.

Reduce the steam generator pressure of'the applicable side-to i

=les's than or equal to 200 psig within 30' minutes, and l

b.

. Perform an engineering' evaluation:to determine the effect of the-overpressurization on the structural integrity.of.the

.. steam' generator.

Determine that the steam generator remains--

acceptable forl continued operation prior to increasing its-temperatures above 200"F.

SURVEILLANCE REQUIREMENTS'

'4.7.2 The. pressure'in'each side'of the steam generator shall be' determined ~

.to be less than 200 psig at least'once per hour when.the temperature of either the reactor'or! secondary coolant is'less than 70 F.

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~ CATAWBA - UNITS 1 & 2 3/4 7-10 1

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!3/4.7 PLANT SYSTEMS BASES g.

l-3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFET VALVES L

The OPERABILITY-of the main steam line Code safety valves ensures that the Secondary System pressure will be limited to within 110% (1304 psig) of

its design pressure of 1185 psig during the most' severe anticipated system

' operational transient.

The maximum relieving capacity is associated with a Turbine trip from valve wide-open condition coincident with an assumed loss l

of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve. lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure -

Code, 1971 Edition.

The total relieving capacity for all valves on all of the steam lines'is 16.85 x 106 lbs/h which is 105% of the total secondary steam flow of 16.05 x 106 lbs/h at 100% RATED THERMAL POWER.

A minimum of two OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in Secondary Coolant System steam flow and THERMAL POWER required by the reduced Reactor trip settings of the Power Range Neutron Flux channels. The Reactor Trip Setpoint reductions'are derived on the following bases:

For four loop operation sp = (X) - (Y)(V)

(109)

Where:

SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL

POWER, V = Maximum number of inoperable safety valves per steam line, Power Range Ne.utron Flux-High Trip Setpoint for four loop 109

=

operation, Total relieving capacity of all safety valves per steam X

=

line in Ibs/ hour, and Maximum relieving capacity of any one safety valve in Y =

lbs/ hour CATAWBA - UNITS 1 & 2 B 3/4 7-1

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' : PLANT SYSTEMS ~

BASES 1

.3/4.7.1.2.

j AUXILIARY FEEDWATER SYSTEM l

. The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor

' Coolant System can-be cooled down to less than 350 F from normal operating conditions in the event of'a feedwater line break accident with a worst case D

single active failure.

1 The Auxiliary Feedwater System is capable of delivering a total feedwater flow of at least 492 gpm at a pressure of 1210 psig to the entrance of at least l

two of the steam generators.

This capacity is su?ficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the Residual Heat Removal System may be placed into operation.

3/4.7.1.3 SPECIFIC ACTIVITY The limitations on Secondary Coolant System specific activity ensure that the resultant offsite radiation dose will be limited to a small fraction of 10 CFR Part 100 dose guideline values in the event of a steam line rupture.

This dose also includes the effects of a coincident 1 gpm reactor to secondary tube leak in' the steam generator of the affected steam line.

These values are consistent with the assumptions used in the safety analyses.

3/4.7.1.4 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blow down in the event of a steam line rupture.

This restriction is required to:

(1) minimize the positive reac-tivity effects of the Reactor Coolant System cooldown associated with the blowdown, and (2) limit the pressure rise within containment in the event the steam line rupture occurs within containment.

The OPERABILITY of the main steam isolation valves within the closure times of the Surveillance Require-ments are consistent with the assumptions used in the safety analyses.

3/4.7.1.5 CONDENSATE STORAGE SYSTEM The OPERABILITY of the Condensate Storage System with the minimum water volume ensures that sufficient water is available to maintain the Reactor Coolant system at HOT STANDBY conditions for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months followed by approximately 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months cooldown with steam discharge to the atmosphere concurrent with total

-loss-of-offsight power..The contained water volume limit includes an allowance for water not usuable because of tank discharge line location or other physical characteristics, 3/4.7.1.6 STEAM GENERATOR POWER OPERATED RELIEF VALVES l

The Surveillance Requirement for the Main Steam power-operated relief valves (PORVs) nitrogen suppl,ies ensures that the PORVs will be available to mitigate the consequences of a steam generator tube rupture accident CATAWBA - UNITS 1 & 2 B 3/4 7-2 Amendment No. 68 (Unit 1)

Amendment No. 62 (Unit 2)

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. PLANT SYSTEMS-BASES p

STEAM GENERATOR POWER OPERATED RELIEF VALVES (Continued) p<

concurrent with loss of offsite power.

This assumes that the PORV on the ruptured steam gfnerator is unavailable, and that the other two are used to cool the Reactor Coolant System inventory to less than the saturation temperature of the ruptured steam generator.

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' Concurrent with the requirement'that a specific number of PORVs be OPERABLE-is-the requirement that. the associated PORV block' valves upstream be e

open or OPERABLE.

Should an associated PORV block valve be closed and inoperable, the PORV downstream of that block valve.hould also be considered inoperable and the applicable ACTION statement shall be entered until such time that the block valve is opened or returned to 0PERABLE status.

Additionally,.if a PORV is inoperable and open, then the requirements of Technical Specification 3.6.3, Containment Isolation Valves, would apply.in addition to Technical Specification'3.7.1.6.

'3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The 1. imitation on steam. generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits.

The limitations of 70*F and 200 psig are based on a steam generator RT of 60 F and are sufficient NDT to prevent brittle fracture.

CATAWBA - UNITS 1 & 2 B 3/4 7-2a Amendment No. 68 (Unit 1)

Amendment No. 62 (Unit 2)

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ML20246B587

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