ML20211D009
| ML20211D009 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 10/16/1986 |
| From: | Joshua Wilson Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211D015 | List: |
| References | |
| NUDOCS 8610220078 | |
| Download: ML20211D009 (15) | |
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UNITED STATES
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g NUCLEAR REGULATORY COMMISSION g
-l WASHINGTON, D. C. 20555 "s
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4 LOUISIANA POWER AND LIGHT COMPANY DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 6 License No. NPF-38 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment, dated August 1, 1985 by Louisiana Power and Light Company (licensee), complies with standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in confomity with the application, the l
provisions of the Act, and the regulations of the Comission; C.
Thereisreasonableassurance(1)thattheactivitiesauthorizedby 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; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and al1~ applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. NPF-38 is hereby amended to read as follows:
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P
. (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 6, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in this license.
LP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
s James H. Wilson, Project Manager PWR Project Directorate No. 7 Division of PWR Licensing-B
Attachment:
Changes to the Technical Specifications Date of Issuance: October 16, 1986
C;tober 16, yggg i
ATTACHMENT TO LICENSE AMENDMENT NO. 6 TO FACILITY OPERATING LICENSE NO. NPF-38 DOCKET NO. 50-382 Replace the following pages of the Appendix A Technical Specificatioas with the enclosed pages. The revised pages are identified by Amendment nun.ber and 1',
contain vertical lines indicating the area of change. Also to be replaced are the following overleaf pages to the amended pages.
Amendment Pages Overleaf Pages 3/4 7-10 3/4 7-9 3/4 8-24 3/4 8-23 3/4 8-39 3/4 8-40 3/4 8-41 3/4 8-42 3/4 9-7 3/4 9-8 8 3/4 7-3 8 3/4 7-4
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PLANT SYSTEMS MAIN STEAM LINE ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.5 Each main steam line isolation valve shall be OPERABLE.
APPLICABILITY:
MODES 1, 2, 3, and 4.
ACTION:
MODE 1 With one main steam line isolation valve inoperable but open, POWER OFERATION may continue provided the inoperable valve is restored to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; 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 />.
MODES 2, 3, and 4 With one main steam line isolation valve inoperable, subsequent operation in MODE 2, 3, or 4 may proceed provided:
a.
The isolation valve is maintained closed, b.
The provisions of Specification 3.0.4 are not applicable.
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 COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIREMENTS 4.7.1.5 Each main steam line isolation valve shall be demonstrated OPERABLE by verifying full closure within 3.0 seconds when tested pursuant to Specification 4.0.5.
WATERFORD - UNIT 3 3/4 7-9
PLANT SYSTEMS 3/4.7.2 STEAM GENEP.ATOR PRESSURE / TEMPERATURE LIMITATION LIMITING CONDITION FOR OPERATION 3.7.2 The temperature of the secondary coolant in the steam generators shall be greater than 115'F when the pressure of the secondary coolant is greater than 210 psig.
APPLICABILITY:
At all times.
ACTION:
With the requirements of the above specification not satisfied:
a.
Reauce the steam generator pressure to less than or equal to 210 psig within 30 minutes, and 2
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 for continued operatic orior to increasing its temperatures above 200'F.
i SURVEILLANCE REQUIREMENTS 4.7.2 The pressure of the steam generators shall be determined to be less than 210 psig at least once per hour when the temperature of the secondary coolant is less than 115*F.
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WATERFORD - UNIT 3 3/4 7-10 Amendment No. 6
TABLE 3.8-1 (Continued) h 480 VOLTS POWER FROM MCCs o
8 OVERCURRENT PROTECTIVE DEVICES ITEM SYSTEM PROTECTION TRIP TINE-CURRENT c3 NO.
POWERED DEVICE TYPE SETPOINT CHARACTERISTIC RENARKS (NOTE 1) w 1
Safety Inj.
Primary Breaker EF 61 Notes 2, 3 Tank 1A Iso.
Val. ISI-Backup Fuse TRS 61 Note 4 V1505 Tk 1A (SI-331A) 2 Safety Inj.
Primary Breaker EF 61 Notes 2, 3 Tank 2A Iso.
w1 Val. ISI-Backup Fuse TRS 61 -
Note 4 m
V1507 Tk 2A (SI-332A) 3 LP-311 Primary Breaker EF 62 Notes 2, 3 Backup Fuse TRS 62 Note 4 4
RCS Loop 2 Primary Breaker EF 63 Notes 2, 3 SDC Iso. Val.
ISI-V1504A Backup Fuse TRS 63 Note 4 (SI-401A) 5 CARS Primary Breaker EF 64 Notes 2, 3 Suction Val.
2HV-F253A Backup Fuse TRS 64 Note 4 (CARS-201A) 6 Hydraulic Primary Breaker EF 64 Notes 2, 3 Pump For Val.
ISI-V1503A Backup Fuse TRS 64 Note 4 (SI-405A)
l TABLE 3.8-1 (Continued) c h
480 VOLTS POWER FROM MCCs (Continued)
E 7
OVERCURRENT PROTECTIVE DEVICES ITEM SYSTEM PROTECTION TRIP TIME-CURRENT c2 NO.
POWERED DEVICE TYPE SETPOINT CHARACTERISTIC REMARKS (NOTE 1)
H 7
Safety Inj.
Primary Breaker EF 65 Notes 2, 3 Tank 1B Iso.
Val. ISI-Backup.
Fuse TRS 65 Note 4 1
V1506 Tk 1B (SI-331B) 8 Safetv Inj.
Primary Breaker EF 65 Notes 2, 3 Tank 2B Iso.
)
Val. ISI-Backup Fuse TRS 65 Note 4 2
V1508 Tk 28 4
(SI-3328) 9 LP-310 Primary Breaker EF 66 Notes 2, 3 Backup Fuse TRS 66 Note 4 l
10 RCS Loop 1 Primary Breaker EF 67 Notes 2, 3 SDC Iso. Val.
1SI-V15028 Backup Fuse TRS 67 Note 4 (SI-4018)
I 11 CARS Primary Breaker EF 68 Notes 2, 3 Suction Val.
(
2HV-F254B Backup Fuse TRS 68 Note 4 g
(CAR-201B) i a
k 12 Hydraulic Primary Breaker EF 68 Notes 2, 3 Pump For Val.
o 1SI-V1501B Backup Fuse TRS 68 Note 4 (SI-405B)
.e
c TABLE 3.8-1 (Continued) l 120 VOLTS CONTROL POWER FROM PDPs OR MCCs (Continued)
S8 OVERCURRENT PROTECTIVE DEVICES ITEM SYSTEM PROTECTION TRIP SETPOINT (NOTE 1)
TYPE / TIME-CURRENT c-j 55 NO.
POWERED SHEET NO.
CIRCulT NO.
DEVICE CHARACTERISTIC REMARKS
-4 (NOTE 2) w 52 Motor Htr.
Primary 121 Ckt. 15 Breaker EE Leads AH-1 (3D-SB)
Backup 121A F4 Fuse TRS 53 Motor Htr.
Primary CWD 1141 Breaker TED 120/208V SWGR heater Leads bus, double breaker E-16 (38)
Backup CWD 1141 Breaker TED protection.
t Rt 54 Motor Htr.
Primary CWD 1142 Breaker TED 120/208V SWGR heater Leads bus, double breaker 9'
E-16 (30)
Backup CWD 1142 Breaker TED protection.
55 Cont. Fan Primary 121 Ckt. 17 Breaker EE Coolers Dampers Backup 121A F5 Fuse TRS l
56 Samp. Sys.
Primary 148A Ckt. 49 Breaker CD l
Sol. Valve l
2SL-F601 Backup 148A Ckt. 49 Fuse FRN (PSL-404A) i l
57 Samp. Sys.
Primary 148A Ckt. 45 Breaker CD j
Sol. Valve 37 2SL-F603 Backup 148A Ckt. 45 Fuse FRN g
(PSL-4048) a!
58 Samp. Sys.
Primary 133 Ckt. 35 Breaker EE r*
Recorder jF Panel Backup 133A F12 Fuse TRS m
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TABLE 3.8-1 (Continued) 120 VOLTS CONTROL POWER FROM PDPs OR MCCs (Continued) 95 OVERCURRENT PRuTECTIVE DEVICES ITEM SYSTEM PROTECTION TRIP SETPOINT (NOTE 1)
TYPE / TIME-CURRENT E
NO.
POWERED SHEET NO.
CIRCull NO.
DEVICE CHARACTERISTIC REMARKS O
(NOTE 2) w 59 Cont. Purge Pri:aary 133 Ckt. 1 Breaker EE Exh. Damper SV-D22 Backup 133A F5 Fuse TRS (CAP-202) and SV-D23 Primary 134 Ckt. 1 Breaker EE (CAP-201)
Backup 134A F2 Fuse ATM 60 Sol. Valve Primary 133 Ckt. 8 Breaker EE 7RC-F604 co A
(RC-323)
Backup 133A F3 Fuse TRS 61 Sol. Valve Primary 133 Ckt. 10 Breaker EE 7RC-F605 (RC-325)
Backup 133A F4 Fuse TRS 62 Sol. Valve Primary 148 Ckt. 29 Breaker CD ICH-E2504B (CVC-218B)
Backup 148A Ckt. 29 Fuse FRN 63 Sol. Valve Primary 147 Ckt. 27 Breaker CD ICH-E2503A (CVC-218A)
Backup 147A Ckt. 27 Fuse FRN 64 Sol. Valves Primary 150 Ckt. 25 Breaker TEB 3CC-P1501A1 (CC-665A) &
Backup CWD 280 F1 Fuse ATM 3CC P1505A1 (CC-679A)
g TABLE 3.8-1 (Continued) y 120 VOLTS CONTROL POWER FROM PDPs OR MCCs (Continued)
S 8
OVERCURRENT PROTECTIVE DEVICES c-ITEM SYSTEM PROTECTION TRIP SETPOINT (NOTE 1)
TYPE / TIME-CURRENT i'i NO.
POWERED SHEET NO.
CIRCUIT NO.
DEVICE CHARACTERISTIC REMARKS
]
(NOTE 2) 65 Sol. Valves Primary 150 Ckt. 27 Breaker TEB 3CC-P1503A2 (CC-666A) &
Backup CWD 282 F1 Fuse ATM 3CC-P1507A2 (CC-680A) 66 RCP1A Primary CWD 220 Fuse OTS Two fuses in series, Instrumentation one each, + and t'
and Backup CWD 220 Fuse OTS poles.
l Accessories 0
67 RCP2A Primary
.CWD 240 Fuse OTS Two fuses in series, Instrumentation one each, + and and Backup CWD 240 Fuse OTS poles.
Accessories 68 CEDM Cool.
Primary 149 Ckt. 14 Breaker TEB Valves
& Dampers Backup CWD 1145 F2 Fuse ATM 69 CEDM Cool.
Primary 150 Ckt. 20 Breaker TEB 1
Units Inlet i
Damper Backup CWD 1145 F1 Fuse ATM k
70 Sol. Valve Primary 150 Ckt. 5 Breaker TEB l
R 2C:!-F1514AB l
2 (RC-602)
Backup CWD 326 F2 Fuse ATM l
5 71 Sol. Valve Primary 135 Ckt. 11 Breaker EE 2
I P
7BM-P237 (GWM-101)
Backup CWD 401 F1 Fuse ATM cr,
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TABLE 3.8-1 (Continued) sc$
120 VOLTS CONTROL POWER FROM PDPs OR MCCs (Continued)
E 5
OVERCURRENT PROTECTIVE DEVICES ITEM SYSTEM PROTECTION TRIP SETPOINT (NOTE 1)
TYPE / TIME-CURRENT
[EE NO.
POWERED SHEET NO.
CIRCUIT NO.
DEVICE CHARACTERISTIC REMARKS (NOTE 2) w 72 Sol. Valve Primary 150 Ckt. 1 Breaker TEB SSI-F1563 l
(SI-342)
Backup CWD 499 F3 Fuse ATM 73 Sol. Valves Primary 150 Ckt. 26 Breaker TEB 3CC-P1502B1 (CC-6658) &
Backup CWD 281 F2 Fuse ATM 3CC-P1506B1 (CC-6798) w1 74 Sol. Valves Primary 150 Ckt. 28 Breaker TEB ao 1
3CC-P150482 (CC-6668) &
Backup CWD 283 F2 Fuse ATM 3CC-P1508B2 (CC-6808) 75 RCPIB Primary CWD 230 Fuse OTS Two fuses in series, Instrumentation one each, + and and Backup CWD 230 Fuse OTS poles.
Accessories 76 RCP2B Primary CWD 250 Fuse-OTS Two fuses in series, Instrumentation one each, + and and Backup CWD 250 Fuse OTS poles.
Accessories 77 Sol. Valve Primary 148 Ckt. 26 Breaker CD 2CA-E604B (ARM-109)
Backup 148A Ckt. 26 Fuse FRN
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REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING l
LIMITING CONDITION FOR OPERATION 3.9.7 Cranes in the fuel handling building shall be restricted as follows:
a.
The spent fuel handling machine shall be used* for the movement of l
fuel assemblies (with or'without CEAs) and shall be OPERABLE with:
l 1.
A minimum hoist capacity of 1800 pounds, and 2.
An overload cutoff limit of less than or equal to 1900 pounds,
- and, l
b.
Loads in excess of 2000 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool.
~
APPLICABILITY:
During movement of fuel assemblies in the fuel handling building, or with fuel assemblies in the spent fuel pool.
ACTION:
l a.
With the spent fuel handling machine inoperable, suspend the use of the spent fuel handling machine for movement of fuel assemblies and place the crane load in a safe position, i
b.
With loads in excess of 2000 pounds over fuel assemblies in the spent fuel pool, place the crane load in a safe position.
j c.
The provisions of Specification 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.9.7.1 The spent fuel handling machine shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of fuel assembly movement and at least once per 7 days thereafter by performing a load test of at least 1800 pounds and demon-strating the automatic load cutoff when the hoist load exceeds 1900 pounds, j
4.9.7.2 The electrical interlock system which prevents crane main hook travel j
over fuel assemblies in the spent fuel pool shal.1 be demonstrated OPERABLE t
within 7 days prior to crane use and at least once per 7 days thereafter l
during crane operation.
4 4.9.7.3 Administrative controls which prevent crane auxiliary hook travel 1
with loads in excess of 2000 pounds over the fuel assemblies in the spent fuel pool shall be enforced during crane operations.
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- Not required for movement of new fuel assemblies outside the spent fuel pool.
l WATERFORD - UNIT 3 3/4 9-7 Amendment No. 6 ll l
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REFUELING OPERATIONS 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION
'HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION 3.9.8.1 At least one shutdown cooling train shall be OPERABLE and in operation.*
APPLICABILITY:
MODE 6 when the water level above the top of the reactor pressure vessel flange is greater than or equal to 23 feet.
ACTION:
With no shutdown cooling train OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required shutdown cooling train to OPERABLE and operating status as soon as possible.
Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.9.8.1 At least one shutdown cooling train shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 4000 gpm** at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- The shutdown cooling loop may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.
- The minimum flow may be reduced to 3000 gpm after the reactor has been shut down for greater than or equal to 175 hours0.00203 days <br />0.0486 hours <br />2.893519e-4 weeks <br />6.65875e-5 months <br /> or by verifying at least once per hour that the RCS temperature is less then 135'F.
WATERFORD - UNIT 3 3/4 9-8
PLANT SYSTEMS
[
BASES 3/4.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the I
resultant cffsite radiation' dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture.
This dose also includes the effects of a coincident 1 gpm primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss-of-offsite electrical i
power. These values are consistent with the assumptions used in the safety analyses.
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3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVE 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 reactivity 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 Requirements are consistent with the assumptions used in the safety analyses.
3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator secondary pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitation 1
to 115'F and 210 psig is based on a steam generator RT of 40'F and is NDT sufficient to prevent brittle fracture.
Below this temperature of 115*F the system pressure must be limited to a maximum of 20% of the secondary hydro-static test pressure of 1375 psia (corrected for instrument error).
Should steam generator temperature drop below 115'F an engineering evaluation of the effects of the overpressurization is required.
However, to reduce the poten-tial for brittle failure the steam generator temperature may be increased to a limit of 200*F while performing the evaluation.
The limitations on the primary side of the steam generator are bounded by the restrictions on the reactor coolant system in Specification 3.4.8.1.
3/4.7.3 COMPONENT COOLING WATER AND AUXILIARY COMPONENT COOLING WATER SYSTEMS The OPERABILITY of the component cooling water system and its corresponding auxiliary component cooling water system ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions.
The redundant cooling capacity of i
these systems, assuming a single failure, is consistent with the assumptions 1
used in the safety analyses.
WATERFORD - UNIT 3 B 3/4 7-3 Amendment No. 6 i
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l PLANT SYSTEMS BASES 3/4.7.4 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level, temperature, and number of fans ensure that sufficient cooling capacity is available to either (1) provide normal cooldown of the facility, or (2) to mitigate the effects of.
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accident conditions within acceptable limits.
The limitations on minimum water level and maximum temperature are based on providing a 30-day cooling water supply to safety-related equipment without exceeding their design basis temperature and is consistent with the recommend-l ations of Regulatory Guide 1.27 " Ultimate Heat Sink for Nuclear Plants,"
l March 1974.
3/4.7.5 FLOOD PROTECTION The limitation on flood protection ensures that facility protective actions will be taken in the event of flood conditions.
The limit of elevation 27.0 ft Mean Sea Level is based on the maximum elevation at which the levee provides protection, the nuclear plant island structure provides protection to safety-related equipment up to elevation +30 ft Mean Sea Level.
3/4.7.6 CONTROL ROOM AIR CONDITIONING SYSTEM The OPERABILITY of the control room air conditioning system ensures that 4
I (1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent. This limitation is consistent with the requirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.
Operation of the system with the heaters on for at least 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> contin-uous over a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters.
Obtaining and analyzing charcoal samples after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of adsorber operation (since the last sample and analysis) ensures that the adsorber maintains the efficiency assumed in the safety analysis and 'is consistent with Regulatory Guide 1.52.
3/4.7.7 CONTROLLED VENTILATION AREA SYSTEM The OPERABILITY of the controlled ventilation area system ensures that radioactive materials leaking from the penetration area or the ECCS equipment within the pump room following a LOCA are filtered prior to reaching the environment.
The operation of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses.
WATERFORD - UNIT 3 8 3/4 7-4
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