ML17285A178
| ML17285A178 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 01/06/1989 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17285A180 | List: |
| References | |
| NUDOCS 8901180315 | |
| Download: ML17285A178 (16) | |
Text
UNITED STATES MUGLEAR REGULATORY COMM I SS ION WASHINGTON, D. C. 20555 WASHINGTON PUBLIC POWER SUPPLY SYSTEM DOCKET NO. 50-397 NUCLEAR PROJECT NO.
2 AHENDYENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No.
NPF-21 The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Washington Public Power Supply System (the licensee),
dated December 21, 1988 complies wi.h the standards and requirements of the Atomic Energy Act of
- 1954, as amended (the Act), and the Commission's regulations set fortli ir: 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by tl;is 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 ComrIission's regulations set forth in 10 CFR Chapter I; D.
The issuance of 4his amendment will not be inimical to the commori 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.
8901180815 890106 PDR ADOCK 05000397 P
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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-21 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
64, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This amendment is effective as of the date of issuance.
Attachment; Changes to the Technical Specifications Date of Issuance:
January 6,
1989 FOR THE NUCLEAR REGULATORY COMMISSION 4~
George H. Knighton, Director Project Directorate V
Division of Reactor Projects - III, IV, V and Special Projects
ENCLOSURE TO LICENSE At1ENDNENT NO. 64 FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 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 irdicating the areas of change.
Also to be replaced are the following overleaf pages.
AMENDMENT PAGE 3/4 3-28 3/4 3-32 3/4 3-36 OVERLEAF PAGE 3/4 3-27 3/4 3-31 3/4 3-35
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t II
TABLE 3.3.3-1 (Continued)
EHERGENCY CORE COOLING SYSTEH ACTUATION INSTRUMENTATION TRIP FUNCTION B.
DIVISION 2 TRIP SYSTEH 1.
RHR B and C
LPCI HODE MINIMUMOPERABLE APPLICABLE CHANNELS Pg$
OPERATIONAL TRIP SYSTEM CONDITIONS ACTION a.
Reactor Vessel Water Level - Low Low Low, Level 1 b.
Drywell Pressure - High c.
Reactor Vessel Pressure-Low (LPCI Permfssfve) d.
LPCI Pump B Start Tfee Delay Relay e.
LPCI Pump Discharge Flow-Low (Hfnfmum Flow) f.
Hanual Initiation 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"0 a.
Reactor Vessel Water Level - Low Low Low, Level 1 b.
ADS Tfeer
.c.
Reactor Vessel Mater Level - Low, Level 3 (Perisfssfve)
LPCI Pump B and C Discharge Pressure-High (Pump Runnfng) e.
Manual Initiation f.
Inhfbft Switch 2
2 1/valve 1
1/pump 1/division 2/pump 2/division 1/division 1, 2, 3, 1, 2, 3
1, 2, 3, 5*
1s 2s 3) lo 2i 3i lo 2g 3i f, 2, 3 1, 2, 3 1o 2, 3 1, 2, 3 1, 2, 3 4*
5A 4A 5*
4*
5*
30 32 32 32 35 35 4*) 5" 30 30 32 33 32 31 34
t t,.
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TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION TRIP FUNCTION C.
DIVISION 3 TRIP SYSTEM HPCS SYSTEM MINIMUM OPERABLE CHANNELS PER TR-tP SYSTEM(a)
APPLICABLE OPERATIONAL CONDITIONS ACTION a.
b.
C.
d.
e.f.
g-2(b) 2(b) 2(c) 2(d) 2(d) 1'/division MINIMUM CHANNELS CHANNELS TO TRIP OPERABLE TOTAL NO.
OF CHANNELS Reactor Vessel Water Level - Low, Low, Level 2
Drywell Pressure High Reactor Vessel Water Level-High, Level 8
Condensate Storage Tanks Level-Low Suppression Pool Water Level-High HPCS System Flow Rate-Low (Minimum Flow)
Manual Initiation 1
2 3
1,2,3 1, 2, 3, 1, 2, 3, 1
2 3
1 2
3 12 3>>
APPLICABLE OPERATIONAL CONDITIONS 4~
")*
30 30 4*, 5" 32 4+, 5*
36 4*, 5*
36 4*, 5*
31 4", 5*
34 ACTION 0.
LOSS OF POWER 2/bus 1, 2, 3, 4*", 5*"
37 2/bus 2/bus 1, 2, 3, 4"", 5*"
38 3/bus 2/bus 1, 2, 3, 4"*, 5"*
38 1.
4.16 kV Emergency Bus Under-voltage (Loss of Voltage) 1/bus 2.
4.16 kV Emergency Bus Under-voltage (Degraded Voltage 2/bus Division 1 and 2) 3.
- 4. 16 kV Emergency Bus Undervoltage (Degraded Voltage Division 3) 2/bus 2/bus TABLE NOTATIONS (a)
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during periods of required surveillance without placing the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring, that parameter.
(b)
Also activates the associated division diesel generator.
(c)
Provides signal to close HPCS pump discharge valve only on 2-out-of-2 logic.
(d)
Provides signal to HPCS pump suction valves only.
When the system is 'required to be OPERABLE per Specification
- 3. 5. 2 or 3. 5. 3.
Required when ESF equipment is required to be OPERABLE.
¹ Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 128 psig.
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TABLE 3.3.3-2 (Continued)
ENERGENCY CORE COOLING SYSTEM ACTUATIOH INSTRtWENTATION SETPOINTS TRIP FUHCTIOH B.
DIVISION 2 TRIP SYSTEM 1.
Reactor Vessel Mater Level - Low Low Low, Level 1 b.
Drywell Pressure - Hfgh C.
Reactor Vessel Pressure-Low (LPCI Permfssfve) d.
LPCI Pump 8 Start Time Delay Relay e.
LPCI Pump Discharge F1ow-Low (Hfnfmum Flow) f.
Manual Inftfatfon 2.
AUTNATIC DEPRESSURIZATIOH SYSTEH TRIP SYSTEH "8" TRIP SETPOINT
> -129 inches*
< 1.65 psig
> 470 psfg, decreasing
< 5 seconds
> 80D gpm H.A.
ALLOWABLE VALUE
> -136 inches
< 1.85 psfg
> 450 psfg, decreasing
< 6 seconds
> 650 gpm H.A.
d.
Reactor Vessel Matel Level - Low Low Low, Level 1 ADS Timer Reactor Vessel Mater Level-low, Level 3, (Permissive)
LPCI Pump 8 and C Discharge Pressure-Hfgh (Pump Running)
Manual Initiation Inhibit Switch
> -129 inches*
< 105 seconds
> 13.0 inches~
> 125 psfg, increasing H.h.
H.A.
> -136 f.nches
< 117 seconds
> 11 inches
> 115 psig, increasing H.h.
H.h.
IV f
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TABLE 3.3.3-2 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS TRIP FUNCTION C.
DIVISION 3 TRIP SYSTEM 1.
HPCS SYSTEM a.
Reactor Vessel Water Level - Low Low, Level 2
b.
Drywell Pressure
- High c.
Reactor Vessel Water Level - High, Level 8 d.
Condensate Storage Tank Level - Low e.
Suppression Pool Water Level - High f.
HPCS System Flow Rate - Low (Minimum Flow) g.
Manual Intiation TRIP SFTPOINT
> -50 inches*
< 1.65 psig
< 54.5 inches*
~
> 448 ft 3 in.
,el evati on
< 466 ft 8 in.
elevation
> 1250 gpm R.A.
ALLOWABLE VALUE
> -57 inches
< 1;85 psig
< 56.0 inches
> 448 ft 0 in. elevation
< 466 ft 10 in. elevation
> 1200 gpm H.A.
D.
LOSS OF POWER 1.
4.16 kV Emergency Bus Undervoltage Loss of Voltage ¹¹ a.
Divisions 1 and 2
b.
Division 3 a.
4.16 kV Basis - 2870 t 86 volts b.
120 V Basis 82 k 2.5 volts a.
4.16 kV Basis - 3016 k 90 volts b.
120 V Basis - 87 + 2.5 volts 2870 k 172 volts 82 k 5 volts 3016 i 180 volts 87 k 5 volts 2.
4.16 kV Emergency Bus Undervoltage Degraded Voltage (Divisions 1, 2, and 3
aO b.
C.
4.16 kV Basis - 3632 + 108 volts 120 V Basi s - 104. 0 k 3. 0 volts 8 + 0.4 sec time delay 3632 k 216 volts 103.8 k 6.0 volts 8 f 0.8 sec time delay TABLE NOTATIONS "See Bases Figure B 3/4 3-1.
¹¹These are inverse time delay voltage relays or.instantaneous voltage relays with a time delay..
The voltages shown are the maximum that will not result in a trip.
Lower voltage conditions will result in decreased trip times.
[1 f
C/l CD TRIP FlNCTION CHAMHEL CHECK CHANNEL OPERATIONAL FUNCTIONAL CHANNEL CONDITIONS FOR WICM TEST CALIBRATIOH SURVEILLANCE RE UIP.
D TABLE 4.3.3. 1-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREMENTS C
o 8.
DIVISION 2 TRIP SYSTEM I.
Reactor Vessel Mater Level-Low Low Low, Level 1
b.
Drywell Pressure - High c.
Reactor Vessel Pressure-Low (LPCI Permissfve) d.
LPCI Pump 8 Start Time Delay Relay e.
LPCI Pump Dfscharge Flow-Low (Minimum Flow) f.
Manual Inftiatfon 2.
-AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "8"0 a.
Reactor Vessel Water Level-Low Low Low, Level 1 b.
ADS Timer c.
Reactor Vessel Water Level-Low, Level 3 (Permissive) d.
LPCI Pump 8 and C Discharge Pressure-High (Pump Running) e.
Manual Initiation f.
Inhibit Switch S
H.A.
N.A.
H.A.
H.A.
H.A.
S N.A.
N.A.
H.A.
N.A.
R R
R N.A.
R R
N.A.
N.A.
1, 2, 3, 4*, 5*
2 2
3 4*
5*
2, 2, 3, 4*, 5*
1, 2, 3, 4*, 5" lo 2o 3 2, 2, 3
2, 2, 3 2, 2, 3
0 1"
TABLE 4. 3. 3. 1-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE RE UIREHENTS nrm C
CAR CARI CHANNEL CHECK TRIP FUNCTION C.
DIVISION 3 TRIP SYSTEM 1.
HPCS SYSTEM a.
Reactor Vessel Water Level-Low Low, Level 2 b.
Drywell Pressure-High c.
Reactor Vessel Mater Level-High, Level 8 d.
Condensate Storage Tank"Level-Low e.
Suppression Pool Water Level " High f.
HPCS System Flow Rate-Low (Minimum Flow) g.
Manual Initiation D.
LOSS OF POWER 1.
- 4. 16 kV Emergency Bus Undervoltage
{Loss of Voltage) 2.
4.16 kV Emergency Bus Undervoltage (Degraded Voltage Division 1 and 2) 3.
4.16 kV Emergency Bus Undervoltage (Degraded Voltage Division 3)
S N.A.
N.A.
N.A.
N.A.
N.A.
N.A.
N.A N.A.
CHANNEL FUNCTIONAL TEST N.A.
N.A.
R N.A.
3 4A 5*
1 2
3
]
2 3
4'k 5*
3 4*
5A 3
4*
5*
1, 2, 3, 4*
5*
2 3
4A*
5IIIA 3
4SICIEl 5SIIA 3
4IRIA
. 5*IR.
OPERATIONAL CHANNEL CONDITIONS FOR WHICH CALIBRATION SURYEILL"NCE REQUIREO O
TABLE NOTATIONS ANot required to be OPERABLE when reactor steam dome pressure is less than or equal to 128 psig.
When the system is required to be OPERABLE per Specification 3.5.2.
"*Required when ESF equipment is required to be OPERABLE.
"""The secondary time delay 3 second relays are exempt from this monthly testing.
The secondary time delay relays associated with this logic will be functionally tested as part of the Logic System Functional Testing (Surveillance Requirement 4.3. 3.2)