ML20093C884
| ML20093C884 | |
| Person / Time | |
|---|---|
| Site: | Catawba  (NPF-24-A-001, NPF-24-A-1) |
| Issue date: | 09/24/1984 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20093C887 | List: |
| References | |
| NUDOCS 8410100777 | |
| Download: ML20093C884 (8) | |
Text
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p UN TED STATES
-j %.,. -ji i NUCLEAR REGULATORY COMMISSION
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DUV,E POWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION l
SALUDA RIVER ELECTRIC COOPERATIVE, INC.
l DOCKET NO. 50-413 l
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CATAUBA NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Anendment No. 1 License No. NPF-24 1.
The nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment to.the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-24 filed by the Duke Power Company, acting for itself, North Carolina Electric Member-shi.p. Corporation and Saluda River Electric Cooperative, Inc.,
(licensees) dated July 31, 1984, and supplemented August 17, 24, and 29',
Energy Act of 1954, as amended (the'Act)quirements of the Atomic 1984, complies wi.th the standards and re and-the Commission's regulations as set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with 'the application, as amended, the provisions of the Act, and the 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 license'acendment will r.ct be inimical to the conmon defense and security or to the health and safety of the public; 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.
2.
Accordingly, the license is hsreby enended by page changes to the Technical Specificatiens as indicated in the attachments to this license amendnent and paragraph 2.C.(2) of Facility Operating License No. NPF-24 is hereby amended to read as follows:
8410100777 840924 PDR ADOCK 05000413 p
.(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Aren&?nt t'c.1, and the Environnental Protection Plan con-tained in Apperdix 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.
The designated requirements of the following Technical Specifications in Appendix A are not applicable during fuel load and precritical operations:
(a)
T.S. 3.1.2.1 Baration Systems - Flow Path - Shutdown -
OPERABLE emergency power source not required.
(b)
T.S. 3.1.2.3 Reactivity Control Systems - Charging Pump :
Shutdown - OPERABLE-emergency power source not required.
(c)
T.S. 3/4.3.2 Engineered Sr"ety Features Actuation System
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Instrumentatinn - In Tables 3.3-3, 3.3-4 and 4.3-2, Items 1 (Safety Injection - Emergency
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Diesel Generator Operation),15 (Emergency Diesel Generator Operation - Diesel Building l
Ventilation Operation, Nuclear Service Water Operation), and 17 (Diesel Building Ventila-tion Operation) are excepted.
In Table 3.3-5.,
Items 2.a.9 (Emergency Diesel Generatoe -
Operation), 3.a.9 (Emergency Diesel Generator Operation), 4.a.9 (Emergency Diesel Generator i
Operation), and 13.d. (Emergency Diesel Generator Operation) as well as Notes (1) and (4) (Diesel generator starting and sequence loading delays included) for Response Times are excepted.
(d)
T.S. 3.7.i.2.a.
Auxiliary Feedwater System - capability of being powered fron emergency buses not required.
(e)
T.S. 3.7.6 - ACTION Control Room Area Ventilation System -
- b. for MODES 5 ar.d 6. OPERABLE emergency power source not required.
(f)
T.S. 3.8.1.1.b.,
A.C. Sources - Operating - OPERABLE diesel 4.8.1.1.2, 4.8.1.1.3, generators not required.
4. 8.1.1..i (g)
T.S. 3.8.1.2.b. and A.C. Sources - Shutdown - OPERABLE diesel 4.8.1.2 generator not required.
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'3.~ This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY' COMMISSION
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Elinor G. Adensam,. Chief Licensing Branch No. 4 Division of Licensing Attachtnent:
Technical Specification-Changes Date of Issuance: September 24, 1984 1
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ATTACHl4EllT TO LICEliSE AltENDliENT fl0. 'l FACILITY OPERATING LICENSE NO. NPF-24 DCCKET NO. 50-413 L
l Riplace the following pages of the Appendix "A" Technical Specificatiens tiith' the enclosed pages.
The revised pages are. identified by Amendment number and contain a vertical line indicating the area of change.. The correspending over-leaf pages are also provided to maintain document completeness.
Amended Overleaf Page Page
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3/4 :7-4 3/4 7-3 l
B3/4 7-2 B3/4 7-1 g.
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TABLE 3.7-2' l,"
STEAM LINE SAFETY VALVES PER LOOP si VALVE NUMBER LIFT SETTING (
1%)*
ORIFICE SIZE e
c:
[j Loop A Loop 8 Loop C
. Loop D 1.
ISV-20 1SV-14 ISV-8 ISV-2 1175 psig_
14.18 in.2 2.
ISV-21 ISV-15 ISV-9 ISV-3
. 119'O psig 14.18 in.2 1205 psig 14.10 in.2'
'3.
ISV-22 1SV-16 ISV-10 ISV-4 4.
ISV-23 ISV-17 ISV-11 ISV-5 1220 psig 14.10 in.2 5.
ISV-2/.
ISV-18 ISV-12 ISV-6 1230 psig "14.18 in.2 E
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.'^fhe lift setting pressure shall correspond to ' ambient conditions of the valve at nominal operating temperature'and pressure.
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,PLAtT SYSTEMS j
p AUXILIARY FEEDWATEP. SYSTEM e_
' LIMITING CONDITION FOR OPERATION
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3.7.1.2 At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall'be OPERABLE with:
Two motor-driven auxiliary feedwater pumps, each capable of being-a.
powered from separate emergency busses,.and b.
One steam turbine-driven auxiliary feodwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILI'TY:
MODES 1, 2, and 3.
ACTION:
With one auxiliary feedwater pump inoperable, restore the required a.
auxiliary feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> o'r be i~n 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 two auxiliary feedwater pumps inoperable, be in at least HOT
- STANDBY witin 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 />.
With three auxiliary feedwater pumps inoperable., immediately 'initiatIe
c.
corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 4.7.1.2.1 Each auxiliary feedwater pump shall be demonstrated OPERABLE:
At least once per 31 days on a STAGGERED TEST BASIS by:
a.
1)
Verifying that each motor-driven pump develops a total dynamic head of greater than or equal to 3470 feet at a fict of greater than or eaual to 400 gpm; 2)
Verifying that the steam turbine-driven pump develops a total dynamic head of greater than or equal to 3550 feet at a ficw of greater than or equal to 400 gpm when the secondary steam supply pressure is greater than 600 psig and the auxiliary feedwater pump turbine is operating at 3600 rpm.
The provisions
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of Specification 4.0.4 are not applicable for entry into MODE-3; 9
. CATAWBA - UNIT 1 3/4 7-4 Amendment No. 1
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-a 4-3/4.7.1 TUE5INE CYCLE-3/4.7.1.1 SAFETY YALVES
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The :?Er *EILITY' of the ai. steam lir.e Ccde s afety salves ens ures that
'e Se:b-dary Systar pressure will be litited to w'tM?n 11:% (13C4 psig) of its casign pressure of 1185 ;sig during the most severe ar.ticipated system cperational transient.
The raxirt; relieving capacity'is associated with a
.TL-tine trip from valve wide-open condition coincicent with an assumed loss of 'ccndenser heat sink '(i.e., no steam bypess to the c:ndanser).
l The specified valve lif t settings and relieving capacities are in-L a5cordance with the requirements of Section III of the ASME Boiler and Pressure Ccde,-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 100 lbs/h at 100% RATED THERMAL POWER.
A minimum of two C EE BLE sa'ety cal'.as par steam gs s atcr ensures that sufficiant reliacing 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 Dy 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) x (109)
X I
Where:
SP = Reduced Reactor Trip Setpoint in percent of RATED THERMAL
- POWER, V = Maximum number of inoperable safety valves per steam line, 109 Power Range Neutron Flux-High Trip Setpoint for four loop
=
cperation, X
Total relieving capacity of all safety.alves per steam
=
line in Ibs/ hour, and Maximum relieving capacity of any one safety valve in Y =
lbs/ hour CATAWEA - UNIT 1 B 3/4 7-1 j
PLANT SYSTEMS EASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Ccolant System can be cooled down to less than 350 F from normal operating ccaditions in the event of a feedsater line break accident with a worst case single active failure.
The Auxiliary Feedwater System is capable of delivering a total feedwater flow of at least 492 gpm at a pressure of 1210 psig t'o the entrance of at least two of tiie steam generators.
This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temp'erature 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 specif c 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 seccndary tube leak in the steam generator of the affected steam.line.
These values are consistent with the assumptions used in the safety ari~alyses.
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 litie 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 OPEP. ABILITY of the main steam isolation valves within the clocure times of the Surveillance Require-ments are consistent with the assumpt.ons used in the safety analyses.
3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The licitation on steam generator pressure and temperature endures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress litaits.
The limitations of 70 F and 2L psig are based on a steam generator RT f 10 F and are sufficient NDT to prevent brittle fracture.
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l CATAWBA - UNIT 1 B 3/4 7-2 Amendment No. 1