ML20066B436
| ML20066B436 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/28/1990 |
| From: | Dyer J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20066B439 | List: |
| References | |
| NUDOCS 9101070143 | |
| Download: ML20066B436 (21) | |
Text
p !purg%,
UNITED STATES M
NUCLEAR REGULATORY COMMISSION g
e CASHING TON, D. C. 20666 L'.; \\,.... /
q,L SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY ThE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 2 AMENDHENT TO FACILITY OPERATING LICENSE Amendment No. 91 License No. NPF-10 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Southern California Edison Company, San Diego Gas and Electric Company, the City of Riverside, California, and the City of Anaheim, California (the licensee) dated November 8, 1990, complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the Atomic
, and the Comission's regulations 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; 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; 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.
'101070143 901228 PDR ADOCK 05000361 F
2 2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of facility Operating License No. NPF-10 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendit A and the Environmental Protection Plan contained in Appenaix B, as revised through Amendment No. 91, are hereby incorporated in the license.
Southern California Edison Company shall oper*.te the f acility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is ef fective as of the date of its issuance and must be fully implemented no later than 30 days f rom the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
- l. wf pt/
James E. Oyer, Director Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation Attachinent:
Changes to tne Technical Specifications Date of Issuance:
December 28, 1090 l
l 1
l l
l
. ~...
3
.w ATTACHMENT TO LICENSE AMEN 0 MENT NO. 91 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET NO. 50-361 Revise Appendix A Technical Specifications:by removing the pages identified below and inserting the enclosed pages.- The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of' change.
REMOVE INSERT XIX XIX 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2 3/4.7-3 3/4 7-3 B3/4'7-1 8 3/4 7-1 a
l:
i l'
t 3
l l
l
.r INDEX-LIST OF TABLES
~ TABLE
_?_AG E, -
3.3-10 ACCIDENT MONITORING INSTRUMENTATION......................
3/4 3-52 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.....,.......................................
3/4 3 54 3.3-11 FIRE DETECTION INSTRUMENTS - MINIMUM INSTRUMENTS OPERABLE 3/4 3-57 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION --
DELETED 4.3-8 RADI0 ACTIVE-LIQUID EFFLUENT MONITORING INSTRUMENTAT1vN SURVEILLANCE REQUIREMENTS -- DELETED 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION.,
3/4 3-65 4.3-9 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................
3/4 3-67 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION.....................................
3/4 4-14 4.4-2 STEAM GENERATOR TUSE INSPECTION..........................
3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.........
3/4 4-19 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY.........................
3/4 4-21 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE........
3/4 4-30a 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.............................................
3/4 4-22 L4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................................
3/4 4-25 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -
-WITHDRAWAL SCHEDULE......................................
3/4 4-28 4.6-1
-TENDON SURVEILLANCE......................................
3/4 6-12 4.6-2 TENDON LIFT-OFF F0RCE.................
3/4 6-12a 3.6-1
-CONTAINMENT ISOLATION VALVES.............................
3/4 6-20' 3.7-1
-MAIN STEAM SAFETY VALVES.................................
3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGi TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPEFsATION WITH BOTH STEAM GENERATORS...............................
3/4 7-3 SAN ONOFRE-UNIT 2 XIX AMENDMENT NO. 91
INDEX LIST OF' TABLES TABLE PAGE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..............................
3/4 7-8 3.7-5 SAFETY-RELATED SPRAY AND/OR SPRINKLER SYSTEMS............
c 4 7-31 3.7-6 FIRE HOSE STATIONS.....................,,................
3/4 7 33
- 4.8 1 DIESEL GENERATOR TEST SCHEDULE...........................
3/4 8-7 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS........................
3/4 8-11 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES.......................................
3/4 8-18 3.8-2 MOTOR OPERATED VALVES, THERMAL OVERLOAD PROTECTION BYPASS DEVICES...........................................
3/4 8-32 3.10-1 RADIATION MONITORING / SAMPLING EXCEPTION -- DELETED 4.11-1 RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM -- DELETED
)
4.11-2 RADIOACTIVE GASEOUS WASTE SAMPLING AND' ANALYSIS PROGRAM -- DELETED i
3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM -- DELETED 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMP.LES -- DELETED 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) -- DELETED B 3/4.4-1 REACTOR VESSEL TOUGHNESS.................................
B 3/4 4-8 S 7 COMPONENT CYCLIC OR TRANSIENT LIMITS.....................
5-8 6.2 MINIMUM SHIFT CREW COMPOSITION...........................
6-4 i
SAN ONOFRE-UNIT.-2 XX AMENDMENT NO 83 JAN 121990 4
e -,..
v-
,4_-
-,,,--.-,w_,r-,.,--
,-- er wegyv vr yiw www y v' t--
v t- -'
.s
=
3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES 1
l LIMITING CONDITION FOR OPERATION 3.7.1.1 All main steam safety valves shall be OPERABLE with lift settings as specified in Table 3.7-1.
A,PPLICABILITY:
MODES 1, 2 and 3.
ACTION:
a.
With both reactor coolant loops and associated steam generators in operation and with one or more main steam safety valves inoperable, l
operation in MODES 1, 2 and 3 may proceed provided, that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, either the inoperable valve is restored to OPERABLE status or the Power Level-High trip setpoint is reduced per Table 3.7-2; 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 SHUT 00WN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specificacion 4.0.5.
SAN ONOFRE-UNIT 2 3/4 7-1 AMENDMENT NO. 91
[
TABLE 3.7-1 5g MAIN STEAM SAFETY VALVES m
g VALVE NUMBER LIFT SETTING (1 1%)*
ORIFICE SIZE Z
Line No. 1 Line No. 2 m
a.
2PSV-8401 2PSV-8410 1100 psia 16 in2 b.
2PSV-8402 2PSV-8411 1107 psia 16 in2 c.
2PSV-8403 2PSV-8412 1114 psia 16 in2 d.
2PSV-8404 2PSV-8413 1121 psia 16 inz e.
. SV-8405 2PSV-8414 1128 psia 16 in2 f.
2PSV-8406 2PSV-8415 1135 osia 16 int g.
2PSV-8407 2PSV-8416 1142 psia 16 in2 h.
2PSV-8408 2PSV-8417 1149 psia 16 in2 i.
2PSV-8409 2PSV-8418 1155 psia 16 inz
$9 E
x y
The lif t setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
P S
_ _.. _. ~..,
TABLE 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVE 5 DURING OPERATION WITH BOTH STEAM GENERATORS Maximum Allowable Value Linear Maximum h,nber of Inoperable Safety Power Level-High Trip Valves on Any Operating Steam Generator (Percent of RATED THERMAL POWER) 1 98.6 2
86.3 3
74.0 4
61.6 SAN ONOFRE - UNIT 2 3/4 7-3 AMENDMENT NO. 91
c
.ll 3/4.7 PLANT SYSTEMS BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam safety valves (MSSVs) ensures that the secondary system pressure will not exceed 110% (1210 psia) of its design pressure of 1100 psia during the most severe anticipated system operational transient.
The total relief capacity available is greater than the maximum steam flow required after a turbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink.
1 The MSSV lift setpoints are staggered, as shown in TaDle 3.7-1, such that only those valves needed for pressure relief will actuate.
The MSSV lift settings and relieving capacities meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code, 1974 Edition, as described in the Overpressure Protection Report (UFSAR Appendix 5.2A).
The total available relieving capacity for all valves on all of the steam lines is 15,473,628 lbs/hr at 1190 psia.
A minimum of one OPERABLE safety valve per steam generator ensures that sufficient relieving capacity is available for removing decay heat.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduc-tion in secondary system steam flow and THERMAL POWER required by the reauced reactor trip settings of the Power Level-High channels.
The reduced reactor trip allowable values are derived on the following bases:
SP = 0) - (Y)(V) x 111.0 X
where:
SP = reduced reactor trip allowable value in percent of RATED THERMAL POWER.
V = maximum number of inoperable safety valves per steam line.
111.0 = Power Level-High Trip allowable value from Table 2.2-1.
X = Total relieving capacity of all safety valves per steam line in lbs/ hour (7,736,814 lbs/hr at 1190 psia).
Y = Maximum relieving capacity of any one safety valve in lbs/ hour (859,646 lbs/hr at 1190 psia).
SAN ONOFRE-UNIT 2 8 3/4 7-1 AMEN 0 MENT NO. 91 a-.-
PLANT SYSTEMS BASES 3/4.7.1.2 AUXILI ARY FEE 0n'ATER SYSTEM 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 total loss of of f-site power.
Each electric driven auxiliary feed ater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators.
The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators.
This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Ccolant System temperature to less than 350'F when the shutdown l
cocling system may be placed into operation.
l 3/4.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the condensate storage tank T-121 with the minimum I
water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by cooldown to shutdown cooling initiation, with steam discharge to atmosphere with. concurrent loss of offsite pcwer and most limiting single failure.
The OPERABILITY of condensate storage tank T-120 in conjunction with tank T-121 ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> including cooldown to shutd0wn cooling initiation, with steam discharge to atmosphere with concurrent loss of offsite power and most limiting single failure.
The contained water volume limits are specified relative to the highest auxiliary feedwater pump suction inlet in the tank for T-121, and to the T-121 cross connect siphon inlet for T-120.
(Water volume below these datum levels is not considered recoverable for purposes of this specification.) Vortexing, internal structure, and instrument error are considered in determining the tank levels corresponding to the specified water volume limits.
Prior to achieving 100% RATED ' THERMAL POWER, Figure 3.7-1 is used to determine the minimum required water volume for T-121 for the maximum power level (hence maximum decay heat) achieved.
/
4 i
SAN ONOFRE-UNIT 2 8 3/4 7-2 AMENDMENT NO.16
o UNITED STATES l'*[
j NUCLEAR REGULATORY COMMISSION o
B E
CASHINGToN, o. C. 20666 SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT NO. 3 AMEN 0 MENT TO FACILITY OPERATING LICENSE
{
Amendment No. 81 License No. NPF-15 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Southern California Edison Company, San Diego Gas and Electric Company, the City of Riverside, California, and the City of Anaheim, California (the licensee) dated November 8, 1990, complies with the 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 conformity with the application, the provisions of the Act, and the rules and regulations of the 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 cumpliance with the Comission's regulations; 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 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and
_ paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows:
(2) Technical Specifications The. Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 81
, are hereby incorporated in the license.
Southern California Edison Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and must be fully implemented no later than 30 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION.
MA- $.
JN' James E. Dyer, Director Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation-
Attachment:
Changes to the Technical Specifications Date of Issuance:
December 28, 1990
ATTACHMENT TO LICENSE _ AMENDMENT NO. 81 FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the areas of change.
REMOVE INSERT XIX XIX 3/4 7-1 3/4 7-1 3/4 7-2 3/4 7-2 3/4 7-3 3/4 7-3 B 3/4 7-1 B 3/4 7-1 u
a
_l INDEX
'l O.
LIST OF TABLES TABLE PAGE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................................
3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS.................................
3/4 3-58 3.3-12 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION --
DELETED 4.3 8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -- DELETED 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION....
3/4 3-66 4.3-9 RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................................
3/4 3-68 l
4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION....................................:...
3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION............................
3/4 4-15 3.4-1; REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...........
3/4 4-20 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY...........................
3/4 4-22 l
4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS...............................................
3/4'4-23 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE...................
3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL SCHEDULE...................................................
3/4 4-29 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..........
3/4 4-31b j
4.6-1 TENDON SURVEILLANCE........................................
3/4 6-12 4.6-2 TENDON. LIFT-OFF F0RCE......................................
3/4 6-13 3.6-1 CONTAINMENT ISOLATION VALVES...............................
3/4 6-21 3.7-1 MAIN STEAM SAFETY VALVES..................................
3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR-POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS.................................
3/4 7-3 SAN ONOFRE - UNIT 3 XIX AMENDMENT NO. 81 a... -. -.-
a
~
-^
INDEX LIST OF TABLES-1 TABLE PAGE 4.7 1 SECONDARY C00LANT' SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..........................................
3/4 7 9 3.7-5 SAFETY-RELATED SPRAY AND/0R SPRINKLER SYSTEMS....'.........
3/4 7-32 3.7-6 FIRE HOSE STATIONS........................................
3/4 7-34 4.8-1 DIESEL GENERATOR TEST SCHEDULE............................
3/4 8-7 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS.........................
3/4 8-11 3.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECT DEVICES..............................................IVE 3/4 8-18 3.8-2 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES PERMANENTLY BYPASSED..............................
3/4 8-32
- 4.11-1 RADIOACTIVE LIQUID SAMPLING AND ANALYSIS PROGRAM -- DELETED 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM --
DELETED g
3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM -- DELETED 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES - DELETED 4.12-1 MAXIMUM VALVES FOR THE LOWER LIMITS OF DETECTION -- DELETED B3/4.4-1 REACTOR VESSEL T0VGHNESS..................................
83/4 4-8 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS......................
5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION............................
6-5 1
SAN ONOFRE - UNIT 3 XX AMENOMENT NO. 73 JAN 121990
e
- 4_.
o 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.'7.1.1 All main steam safety valves shall be OPERABLE with lift settings as specified in Table 3.7-1.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
a..
With both reactor coolant loops and associated steam generators in operation and with one or more main steam safety valves inoperable.
l operation in MODES 1, 2 and 3 may proceed provided, that within 4' hours, either the inoperable valve is restored to OPERABLE status or the Power Level-High-trip.setpoint is reduced per Table 3.7-2; 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 />, b.
The provisions of Specification 3.0.4 are not applicable.
- SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional Surveillance Requirements other than those required by Specification 4.0.5.
SAN ONOFRE-UNIT 3 3/4 7-1 AMENDMENT NO. 81
v.
r E
TABLE 3.7-1 o.g MAIN STEAM SAFETY VALVES
.g VALVE NUMBER LIFT SETTING (11%)*
ORIFICE SIZE
--e t
Line No. 1 Line No. 2 m
i a.
3PSV-8410 1100 psia 16 in2 b.
3PSV-8402 3PSV-8411 1107 psia 16 in2 L
c.
3PSV-8403 3PSV-8412 1114 psia 16 in2 d.
3PSV-8404 3PSV-8413 1121 psia 16 in2 D
e.
3PSV-8405 3PSV-8414 1128 psia 16 in2 f.
3PSV-8406 3PSV-8415 1135 psia 16 inz i
g.
3PSV-8407 3PSV-8416 1142 psie 16 in2 i
h.
'3PSV-8408 3PSV-8417 1149 psia 16 in2 i..
3PSV-8409 3PSV-8418 1155 psia 16 in2 4
i 1g i
l E-j j
- The. lift setting pressure shall correspond to ambient conditions of the valve at nominal operating-j temperature and pressure.
o 1
+
i r
e b
l l
ai
.+
e e
O TABLE 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INDPERABLE HAlb 5 TEAM 5AFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS Maximum Allowable Value Linear Maximum Number of Inoperable Safety Power Level-High Trip Valves on Any Operatino Steam Generator (Percent of RATED THERMAL POWER) 1 98.6 2
86.3 3
74.0 4
61.6 1
SAN ONOFRE - UNIT 3 3/4 7-3 AMENDMENT NO. 8I
. =...
PLANT SYSTEMS
^
AUXILIARY FEE 0 WATER SYSTEM r
LIMITING CONDITION FOR OPERATION 3.7.3.2 At least three ' independent steam generator auxiliary feedwater pumps and essociated flow paths shall be OPERABLE with:
a.
Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and b.
One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.
APPLICABILITY:
MODES 1, 2 and 3.
ACTION:
a.
With one auxiliary feedwater pump inoperable, restore the required 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 /> or 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 two auxiliary pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUT 00WN 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 auxiliary feedwater pumps inoperable, immediately initiate 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:
a.
At least once per 31 days by:
1.
Testi.g the turbine driven pump and both motor driven pumps pursuant to Specification 4.0.5.
The provisions of Specification 4.0.4 are not applicable for the turbine-driven pump-for entry into MODE 3.
2.
Verifying that each valve (manual, power operated or automatic) in the flow path that is-not locked, sealed, or otherwise secured in position, is in its correct position.
3.
-Verifying that both manual valves in the suction lines from the primary AFW supply tank (condensate storage tank T-121) to each AFW pump, and the manual discharge line valve of each AFW' pump are locked in-the open position.
SAN ONOFRE-UNIT 3 3/4 7-4
+ + '
i 3/4.7 PLANT SYSTEMS i
. BASES 3/4.7.1 TURBINE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam safety valves (MSSVs) ensures that the secondary system pressure will not exceed 110% (1210 psia) of its design pressure of 1100 psia during the most severe anticipated system operational transient.
The total relief capacity available is greater than the maximum steam-flow required af ter a.urbine trip from 102% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink.
The MSSV lift setpoints are staggered, as shown in Table 3.7-1, such that only those valves needed for pressure relief will actuate.
The MSSV lif t uttings and relieving capacities meet the requirements of Section III of the ASME Boiler and Pressure Vessel Code,1974 Edition, as described in the Overpressure Protection Report (UFSAR Appendix 5.2A).
The total available relieving capacity for all valves on all of the steam lines is 15,473,628 lbs/hr at 1190 psia.
A minimum of one OPERABLE safety valve per steam generator ensures that sufficient relieving capacity is available for removing decay heat.
STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduc-tion in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels.
The reduced reactor trip allowable values are derived on the following bases:
SP = (X) - (Y)(V) x 111.0 X
where:
SP = reduced reactor trip allowable value in percent of RATED THERMAL POWER.
V = maximum number of inoperable safety valves per steam line.
l; 111.0 = Power Level-High Trip allowable value from Table 2.2-1.
X = Total relieving capacity of all safety valves per steam line in lbs/ hour (7,736,814 lbs/hr at 1190 psia),
l Y = Maximum relieving capacity of any one safety valve in 1bs/ hour (859,646 lbs/hr at-1190 psia).
1 SAN ONOFRE-UNIT 3 B 3/4 7-1 AMENDMENT NO. 81 L
\\
-.n..
-y
~. _ _
4 PLANT SYSTEMS 6
BASES 3/4.7.1.2 AUXILIARY FEE 0 WATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cocied down to less than 350*F from normal operating conditions in the event of-a total loss of offsite power.
Each electric-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators.
The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators.
This capacity is sufficient 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 shutdown cooling system may be placed into operation.
3/4.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the condensate storage tank T-121 with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by cooldown to shutdown cooling initiation, with steam discharge to atmosphere w!' h c ' current loss of offsite power and most limiting single failure, The OPERAeIL.TY of condensate storage tank T-120 in conjunction with tank T-121 ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> including cooldown to shutdown cooling-initiation, with steam discharge to atmosphere with concurrent loss of offsite power and most limiting single failure.
The contained water volume limits are specified relative to the highest auxiliary feedwater pump suction inlet in the tank for T-121, and to the T-121 cross connect siphon inlet for T-120.
(Water volume below these datum levels is not considered recoverable for purposes of this specification.) Vortexing, internal structure and instrument error are considered in determining the tank l
1evels corresponding to the specified water volume limits.
Prior to achieving 100% RATED THERMAL POWER, Figure 3.7-1 is used to determine the minimum required water volume for T-121 for the maximum powei' level (hence maximum decay heat) achieved.
l t
I l
-SAN ONOFRE-UNIT 3 B 3/4 7-2
-e--
is-m m
,- - --