ML20244D423
| ML20244D423 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/11/1989 |
| From: | Knighton G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20244D429 | List: |
| References | |
| NUDOCS 8904210336 | |
| Download: ML20244D423 (30) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION p,
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- a WASHINGTON, D. C. 20655 e
s 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 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. NPF-10 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment to the license for San Onofre NuclearGeneratingStation, Unit 2(thefacility)filedby Southern California Edison Company (SCE) on behalf of itself and San Diego Gas and Electric Company, the City of Riverside, California and the City of Anaheim, California (licensees) dated November 7, 1988, and supplemental submittals dated December 29, 1988 and February 23, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, i
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 regulations of the Comission; C.
Thereisreasonableassurance(1)thattheactivitiesauthorized 4
by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be I
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.
i 89042103368904$1 DR ADOCK 05000361 PDC
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.. 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.Ck ) of-Facility Operating License No. NPF-10 is hereby amended to read as follows:
Q) Technical Specifications The Technical Specifications contained in Appendix A, and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 70, are hereby incorporated in the license. SCE 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 Georg Knigh Dir ctor Project Directorate V D ivision of Reactor Projects - III, IV, Y and Special Projects
Attachment:
Changes to the Technical Specifications Date of Issuance: April 11,1989
i 1
ATTACHMENT TO LICENSE AMENDMENT N0. 70 FACILITY OPERATING LICENSE NO. NPF-10 DOCKET NO. 50-361 i
I Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and contain marginal lines indicating the area of change.
Also enclosed are the following overleaf pages to the amended pages.
AMENDMENT PAGE OVERLEAF PAGE y
vi xx xix xxii xxi 3/4 4-3 3/4 4-4 3/4 4-5 3/4 4-5a 3/4 4-27 3/4 4-27a 3/4 4-29 3/4 4-30 3/4 4-30a l
3/4 4-32 3/4 4-31 3/4 4-33 3/4 4-34 B 3/4 4-1 B 3/4 4-2 B 3/4 4-6 B 3/4 4-5 B 3/4 4-7 B 3/4 4-8 B 3/4 4-9 Page 3/4 4-28 is reissued without change l
l
j l
INDEX
-1 LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS l
i SECTION PAGE HOT SHUTD0WN............................................
3/4 4-3 COLD SHUTDOWN - Loops' Filled............................
3/4 4-5 COLD SHUTDOWN - Loops Not Filled........................
3/4'4-6 3/4.4.2 SAFETY VALVES - 0PERATING...............................
3/4 4-7 3/4.4.3 PRESSURIZER.............................................
3/4 4-8 3/4.4.4 STEAM GENERATORS........................................
3/4 4-9 3/4.4.5-REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................
3/4 4-16 OPERATIONAL LEAKAGE..................................
3/4 4-17 3/4.4.6
' CHEMISTRY..............................................
3/4 4-20 3/4.4.7 SPECIFIC ACTIVITY.......................................
3/4 4-23 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM...............................
3/4 4-27 PRESSURIZER - HEATUP/C00LDOWN........................
3/4 4-31 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE < 312* F............................
3/4 4-32 RCS TEMPERATURE I 312 F............................
3/4 4-33 3/4.4.9 STRUCTURAL INTEGRITY....................................
3/4 4-34
'3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM.........................
3/4 4-35 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS..................................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tgg > 350 F..........................
3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350 F..........................
3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................
3/4 5-8 SAN ONOFRE-UNIT 2 V
AMENDMENT NO. 70
j 2
INDEX t
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i
SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY................................
3/4 6-1 z
CONTAINMENT LEAKAGE..................................
3/4 6-2 CONTAINMENT AIR L0CKS................................
3/4 6-5 INTERNAL PRESSURE....................................
3/4 6-7 AIR TEMPERATURE......................................
3/4 6-8 CONTAINMENT STRUCTURAL INTEGRITY.....................
3/4 6-9 CONTAINMENT VENTILATION SYSTEM.......................
3/4 6-13 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM.............................
3/4 6-14 I
IODINE REMOVAL SYSTEM................................
3/4 6-16 CONTAINMENT COO LING SYSTEM...........................
3/4 6-17 3/4.6.3 CONTAINMENT ISOLATION VALVES............................
3/4 6-18 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN M0NITORS....................................
3/4 6-26 ELECTRIC-HYDROGEN REC 0MBINERS........................
3/4 6-27 CONTAINMENT DOME AIR CIRCULAT0RS.....................
3/4 6-28 SAN ONOFRE-UNIT 2 VI
.,....~.. -
3 q.5 s
IflDEX
[.{
LIST OF T/.SLES
' TABLE PAGE 1.1 OPERATIONAL M3 DES.........................................
1-7 l'2
-FREQUENCY N3TATION...................
1-8
- 2. 2-l' REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOIliT~ LIMITS...
2-3~
.2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS.......... 5 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION.........................
3/4 3-3 3.3 REACT 04 PROTECTIVE INSTRUMENTATION RESPONSE TIMES.........
3/4 3-S.
o 4.3-1 REACTOR PROTECTIVE-INSTRUME!4TATION SURVEILLANCE.
REQUIREMENTS...................................;..........
.3/4 3-10' 3.3-3 ENGINEEREO SAFETY FEATURE ACTUATION SYSTEM-INSTRUMENTATION...........................................
3/4 3-14 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP. VALUES...........................................'.....
3/4 3-22 3.3-5 E:GINEERED SAFETY FEATURES RESPONSE TIMES.................
3/4 3-27 4.3-2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................
3/4 3-31 3.3-6 RADI ATION MONITORING ALARM INSTRUMENTATION................
3/4/3-35 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................
3/4/3-38 3.5-7 SEISMIC MONITORING INSTRUMENTATION........................
3/4 3-43 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................
3/4 3-44 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION..................
3/4 3-46
.4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................
3/4 3-47 3.3-9 REMOTE SHUTOOWN MONITORING INSTRUMENTATION................
3/4 3-49
/
4.3-6 REMOTE SHUTDOWN MONITORING INSTRUMENTATION SURVEILLANCE
. REQUIREMENTS..............................................
3/4 3-50 l
L S/.N O!.0FRE-UNIT 2 XIX AMENDMENT NO.16 I.
INDEX LIST OF TABLES TABLE PAGE 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...............................
3/4 3-57 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING. INSTRUMENTATION...
3/4 3-64 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS....................
3/4 3-66 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION..
3/4 3-69 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................
3/4 3-71 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 4-19 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 4.4-4 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 l
4.6-1 TENDON SURVEILLANCE......................................
3/4 6-12 4.6-2 TENDON LIFT-0FF F0RCE....................................
3/4 6-12a 3.6-1 CONTAINMENT ISOLATION VALVES.............................
3/4 6-20 3.7-1 STEAM LINE SAFETY VALVES PER L00P........................
3/4 7-2 1
3.7-2 MAXIMUM ALLOWABLE LINEAR POWER LEVEL-HIGF. TRIP SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS..............................
3/4 7-3 SAN ON0FRE-UNIT 2 XX AMENDMENT NO. 70
i P
- i 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............
3/4 7-31
,3.7-6 FIRE HOSE STATIONS........................................
3/4 7-33 4.8-1 DIESEL GENERATOR TEST.SCNEDULE...........................
3/4 8-7 i
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 OVERl0AD PROTECTION BYPASS DEVICES...........................................
3/4 8-32 3.10-1 RADIATION MONITORING / SAMPLING EXCEPTIONS................
3/4 10-6 4.11-1 RADI0 ACTIVE LIQUID WASTE SAMPLING AND. ANALYSIS PROGRAM...
3/4 11-2 I
i 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PR0 GRAM..................................................
3/4 11-9 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM............
3/4 12-3 3.12-2 REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN I
ENVIRONMENTAL SAMPLES....................................
3/4 12-7 4.12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECION (LLD)....
3/4 12-8 B 3/4.4-1 REACTOR VESSEL T0VGHNESS.................................
B 3/4 4-8 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS.....................
5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION...........................
6-4
\\
I SAN ONOFRE-UNIT 2 XXI AMENDMENT NO. 33 i
Q-INDEX LIST OF FIGURES 1
FIGURES PAGE 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURES AS A FUNCTION OF STORED BORIC ACID CONCENTRATION.........
3/4 1-13 3.1-2 CEA INSERTION LIMITS.....................................
3/4 1-24 l
3.2-1 DNBR MARGIN OPERATING LIMIT BASED ON C0LSS...............
3/4 2-7 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE).......................
3/4 2-8.
3.3-1 DEGRADED BUS VOLTAGE TRIP SETTING........................
3/4 3-40 4~4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA...................
3/4 4-15a 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT...........................................
3/4 4-26 3.4-2 RCS HEATUP PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY......................................
3/4 4-29 3.4-3 RCS C00LDOWN PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY.................................................
3/4 4-30 3.7-1 MINIMUM REQUIRED FEE 0 WATER INVENTORY FOR TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE............
3/4 7-6A 5.1-1 EXCLUSION AREA...........................................
5-2 5.1-2 LOW POPULATION Z0NE......................................
5-3 5.1-3 SITE BOUNDARY FOR GASE0US EFFLUENTS......................
5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS.......................
5-5 6.2-1 0FFSITE ORGANIZATION.....................................
6-2 6.2-2 UNIT ORGANIZATION........................................
6-3 6.2-3 CONTROL ROOM AREA........................................
6-4a SAN ONOFRE-UNIT 2 XXII AMENDMENT NO. 70
REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION l
3.4.1.3 At least two of the loop (s)/ train (s) listed below shall be OPERABLE and at least one Reactor Coolant and/or shutdown cooling loops shall be in operation.*
Reactor Coolant Loop 1 and its associated steam a.
at least one associated Reactor Coolant pump,** generator and b.
Reactor Coolant Loop 2 and its associated steam generator and at least one associated Reactor Coolant pump,**
c.
Shutdown Cooling Train A, d.
Shutdown Cooling Train B.
APPLICABILITY:
MODE 4 ACTION:
With less than the above required Reactor Coolant loops and/or a.
shutdown cooling trains OPERABLE, immediately initiate correc-tive action to return the required loops / trains to OPERABLE status as soon as possible; if the remaining OPERABLE loop is a shutdown cooling train, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
With no Reactor Coolant loop or shutdown cooling train in opera-tion, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop /
train to operation.
l I
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- All Reactor Coolant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10 F below saturation temperature.
- A Reactor Coolant pump shall not be started with one or more of the Reactor i
Coolant System cold leg temperatures less than or equal to that specified in Table 3.4-3 when the secondary water temperature of each steam generator is greater than 100 F above each of the Reactor Coolant System cold leg temper-
- atures, i
SAN ONOFRE - UNIT 2 3/4 4-3 AMENDMENT NO. 70 i
l 1
1 U-______-_.______.._.
n t
' REACTOR' COOLANT SYSTEM
! HOT SHUTOOWN-SURVEILLANCE REQUIREMENTS:
4.4.1.3.1 The required Reactor. Coolant' pump (s), if not in operation,. shall be
. determined to be OPERABLE once per 7 days by verifying correct breaker -
alignments ~and indicated power availability.
4.4.l.3.2 The required steam generator (s) shall'be determined 0PERABLE by verifying the seconcary side water level to be > 10*.' (wide range) at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.3.3 At least one Reactor Coolant loop or shutdown cooling train shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
1
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1 SAN ONOFRE-UNIT 2 3/4 4-4 L-______--__---_-___________-___--__
_ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - = - - - - -
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7 REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION
-3.4.1.4.1-a.-
At least one of the following loop (s)/ trains.. listed below shall be OPERABLE and in operation *:
1.
Reactor Coolant Loop 1 and its associated steam generator and at least one associated Reactor Coolant-Pump **
2.
Reactor Coolant Loop 2 and its associated steam generator.and at least one associated Reactor Coolant Pump **
3.
Shutdown Cooling Train A 4.
Shutdown Cooling Train B b.
One additional Reactor Coolant Loop / shutdown cooling train shall be OPERABLE, or The secondary side water level of'each steam generator shall c.
be greater than 10% (wide range).
' APPLICABILITY: MODE 5, with Reactor Coolant loops filled.
- ACTION:
a.
With less than the above required shutdown trains / loops OPERABLE or with less than the required steam generator level, immediately -
initiate corrective' action to return the required trains / loops to OPERABLE status or restore the required level as soon'as possible.
' b.
With no loop / train in operation, suspend all operations involving a reduction in boron concentration of'the Reactor Coolant System and immediately initiate corrective action to return the required loop /
train to operation.
"All reactor coolant pumps and shutdown cooling pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at-least 10'F below saturation temperature.
~
- A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to that specified in Table 3.4-3 when-the secondary water temperature of each steam generator is greater than 100'F above each of the Reactor Coolant System cold leg temper-atures.
t i
SAN ONOFRE, UNIT 2 3/4 4-5 AMENDMENT NO. 70
i REACTOR COOLANT SYSTEM COLD SHUTDOWN -' LOOPS FILLED SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The required Reactor Cooling pump (s) if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker align-ments and indicated power availability.
4.4.1.4.1.2 The required steam generator (s) shall be determined OPERABLE by verifying the. secondary side water level to be >10% (wide range)'at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.4.1.3 At least one Reactor Coolant loop or shutdown cooling train shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I SAN ONOFRE - UNIT 2 3/4 4-Sa AMENDMENT N0. 27
4 REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figure 3.4-2 and Figure 3.4-3 during heatup, cooldown, criticality, boltup, l
and inservice leak and hydrostatic testing with:
A maximum heatup of 10*F in any one hour period with RC cold leg a.
temperature less than 112 F.
A maximum heatup of 30*F in any one hour period with RC cold leg temperature less than 163 F.
A maximum heat-upof60*FinangonehourperiodwithRCcoldlegtemperature greater than 163 F.
b.
A maximum cooldown of 10 F in any one hour period with RC cold leg temperatures less than 103 F.
A maximum cooldown of 30*F in any one hour period with RC cold leg temperatures less than 145 F.
A maximum cooldown of 100 F in any one hour period with RC temperature greater than 145 F.
A maximum temperature change of less than or equal to 10 F in any c.
one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves, d.
A minimum temperature of 86 F to tension reactor vessel head bolts.
l APPLICABILITY: At all times.
ACTION:
With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY with-in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T,yg and pressure to less than 200 F and 500 psia, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
l SURVEILLANCE REQUIREMENTS l
4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.
i SAN ON0FRE - UNIT 2 3/4 4-27 AMENDMENT NO. 70
REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued) 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals _ required by 10 CFR 50 Appendix H in accordance with the schedule in Table 4.4-5.
The results of these examinations'shall be used to update Figures 3.4-2 and 3.4-3.
Recalculate the Adjusted Reference Temperature based on the greater of the following:
The mean value of shift in reference temperature for plates a.
C-6404-3*, or b.
The predicted shift in reference temperature for weld seams 3-203A or 3-203B as determined by Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.
I "The most limiting material in the reactor vessel in accordance with the new l
Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988, has changed and are plates C-6404-3.
Calculative proce-dures provided in the new guide should be used to obtain the mean values of shift in RT f C-6404-3 plates.
Calculations are based on the actual shift NDT in reference temperature as determined by impact testing on the existing plate C-6404-2 surveillance material.
SAN ONOFRE - UNIT 2 3/4 4-27a AMENDMENT N0. 70 l
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Figure 3.4-2 RCS HEATUP PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY SAN ONOFRE - UNIT 2 3/4 4 29 AMENDMENT NO. 70
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l 1000 V
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z 500 WINIMUW BOLTUP TEMP = 86'F 0 '
I
0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F)
Figure 3.4-3 RCS COOLDOWN PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY SAN ONOFRE - UNIT 2 3/4 4-30 AMENDMENT NO. 70
..pt
. s m
Table 3.4-3 QwTemperatureRCSOverpressureProtectionRange Operating Period EFPY Cold Leg Temperature,
'F During During.
Heatup Cooldown 4 to 10 1 312-1 287 1
1 SAN ONOFRE - UNIT 2 3/4 4-30a AMENDMENT NO. 70
1 REACTOR COOLANT SYSTEM PRESSURIZER - HEATUP/COOLDOWN
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LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer shall be limited to:
a.
A maximum heatup of 200*F in any one hour period, b.
A maximum cooldown of 200*F in any one hour period.
APPLICABILITY: At all times.
ACTION:
With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
SURVEILLANCE REQUIF:EMENTS 4.4.8.2.1 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system hsatup or cooldown.
4.4.8.2.2 The spray W9ter temperature differential shall be determined for use in Table 5.7-1 for each cycle of main spray when loss than 4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation.
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SAN ONOFRE - UNIT 2 3/4 4-31 AMENDMENT NO. 44
REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE I 312 F LIMITING CONDITION FOR OPERATION 3.4.8.3.1 At least one of the following overpressure protection systems shall be OPERABLE:
a.
The Shutdown Cooling System Relief Valve (PSV9349) with:
1)
A lift setting of 406 1 10 psig*, and 2)
Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open, or, b.
The Reactor Coolant System depressurized with an RCS vent of greater than or equal to 5.6 square inches.
APPLICABILITY: MODE 4 when the temperature of any one RCS cold leg is less than or equal to that specified in Table 3.4-3; MODE 5; MODE 6 with the reactor vessel head on.
-1 ACTION:
With the SDCS Relief Valve inoperable, reduce T,yg a.
to less than 200 F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
With one or both SDCS Relief Valve isolation valves in a single SDCS Relief Valve isolation valve pair (valve pair 2HV9337 and 2HV9339 or valve pair 2HV9377 and 2HV9378) closed, open the closed valve (s) within 7 days or reduce T,yg to less than 200 F, depres-surize and vent the RCS through a greater than or equal to 5.6 inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
c.
In the event either the SDCS Relief Valve or an RCS vent is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initi-ating the transient, the effect of the SDCS Relief Valve or RCS vent on the transient and any corrective action necessary to prevent recurrence.
d.
The provisions of Specification 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:
a.
Verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the SDCS Relief Valve is being used for overpressure protection that SDCS Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open.
- For valve temperatures less than or equal to 130*F.
SAN ONOFRE - UNIT 2 3/4 4-32 AMENDMENT NO. 70 l
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i REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE > 312 F LIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems shall be OPERABLE:
a.
The Shutdown Cooling System Relief Valve (PSV9349) with:
j 1)
A lift setting of 406 1 10 psig*, and l
2)
Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open, or, b.
A minimum of one pressurizer code safety valve with a lift setting of 2500 psia + 1%**.
APPLICABIL~ITY: MODE 4 with RCS temperature above that specified in Table 3.4-3.
ACTION:
a.
With no safety or relief valve OPERABLE, be in COLD SHUTDOWN and vent the RCS thro gh a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
b.
In the event the SDCS Relief Valve or an RCS vent is used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve or RCS vent on the transient and any corrective action necessary to prevent recurrence.
SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:
a.
Verifying at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> that the SDCS Relief Valve isolation valves 2HV9337,'2HV9339, 2HV9377 and 2HV9378 are open when I
the SDCS Relief Valve is being used for overpressure protection.
b.
Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.5.
4.4.8.3.2.2 The pressurizer code safety valve has no additional surveillance requirements other than those required by Specification 4.0.5.
4.4.8.3.2.3 The RCS vent shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the vent is being used for overpressure protection, except when the vent pathway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.
- For valve temperatures less than or equal to 130*F.
- The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.
SAN ON0FRE. UNIT 2 3/4 4-33 AMENDMENT NO. 70 l
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A 3.4.9 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.9 The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Specification 4.4.9.
APPLICABILITY: ALL MODES
, ACTION:
a.'
With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT. considerations.
b, With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.
I With the structural integrity of any ASME Code Class 3 component (s) c.
not conforming to the above requirements, restore the structural 4
integrity of the affected component to within its limit or isolate the affected component from service.
d.
The provisions of Specification 3.0.4 are not applicable.
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SURVEILLANCE REQUIREMFMTS 4.4.9 In addition to the requirements of Specification 4.0.5, each Reactor Coolant Pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.
l SAN ONOFRE-UNIT 2 3/4 4-34
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. 3/4'. 4 REACTOR ~ COOLANT SYSTEM-BASES 3/4.'4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor. coolant loops and.
associated reactor coolant pumps in operation,'and maintain DNBR greater than 1.31 during all normal' operations and. anticipated transients. As a result, in MODES 1 and 2 with one reactor coolant loop not in operation, this speci-fication requires that the plant be in at least HOT STAND 8Y within 1 hour-since'no safety analysis has.been conducted for operation with less than 4 reactor coolant pumps or less than two reactor coolant loops in operation.
In MODE' 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.
.In MODE 4. and in MODE 5 with reactor. coolant: loops filled, a single reactor coolant loop.or shutdown cooling train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops / trains (either RCS or shutdown cooling) be OPERABLE.
In MODE 5 with reactor coolant. loops not filled, a single shutdown cool-ing train provides sufficient heat removal capability for. removing decay. heat; but single failure considerations, and the unavailability of the steam genera-tors as a heat removing component, require that at.least two shutdown cooling trains be OPERABLE.
The operation of one React'or Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity' changes during boron concentration reductions in the Reac-tor Coolant System. The' reactivity change rate associated with boron reduc,
tions will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump in Modes 4 and 5 with one or more RCS cold legs less than or equal to that specified in Table 3.4-3 are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to
.10 CFR Part 50. The RCS will be protected against overpressure transients and will.not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 100*F above each of the RCS cold leg temperatures.
3/4.4.2 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia.
Each safety valve is designed to relieve 4.6 x 105 lbs per hour of saturated steam at the valve setpoint plus 3% accumulation. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown with RCS cold leg temperature greater than that specified in Table 3.4-3.
In the event l
that no safety valves are OPERABLE and for RCS cold leg temperature less than or equal to that specified in Table 3.4-3, the operating shutdown cooling relief I.
valve, connected to the RCS, provides overpressure relief capability and will i-prevent RCS overpressurization.
SAN ONOFRE-UNIT 2 8 3/4 4-1 AMENDMENT NO. 70
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REACTOR COOLANT SYSTEM 2ASES SAFETY VALVES (Continued)
During operation, all pressurizer code safety valves must be CPERABLE to prevent the RCS from being ' essurized above its safety limit of 2750 psia.
The c:" tined relief caoacit,. of these valves is sufficient to limit the System cressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pres-surizar Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the steam dump valves.
Demonstration of the safety valves' lift setti.ngs will occur only during shutdown and will be performed in accordance with the provisions of Section XI
-of the ASME Boiler and Pressure Vessel Code.
3/4.4.3 PRES $URIZER The limit on the maximum water volume in the pressurizer assures that the parameter is maintained within the normal steady-state envelope of operation assumed in the SAR. A steam bubble in the pressurizer ensures that the RCS is i
(
not a hydraulically solid system and is capable of accommodating pressure surges curing operation. The steam bubble.also protects the pressurizer code safety valves against water relief. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reacter Coolant System pressure and establish natural circulation.
3/4.4.4 STEAM GENERATORS The Surveillance Requirements for inspect' ion of the steam' generator tuees j
ensure that the structural integrity of this portion of the RCS will be main-j tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain i
surveillance of the conditions of the tubes in' the event that there is I
evidence of mechanical damage or progressive degradation due to design,
~
manufacturing errors, or inservice conditions that lead to corrosion.
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SAN ONOFRE-UNIT 2 8 3/4 4-2
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BASES REACTOR COOLANT ~ SYSTEM BASES CHEMISTRY (Continued) the chemistry within the Steady. State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.
Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural. integrity.of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits
-provides time for.taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
TDe surveillance' requirements provide adequate assurance that concentrations'in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.7 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that. the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appro-priately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the San Onofre site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than L0 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
SAN ONOFRE-UNIT 2 B 3/4 4-5 AMEN 0 MENT NO. 50
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REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)
Reducing T to less than 500 F prevents the release of activity should asteamgenerat@9 tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction.in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.8 PRESSURE / TEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.
These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 3.9.1.1 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the desigrs assumptions and satisfy the stress limits for cyclic operation.
During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel outside surface.
Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location when compared to the outside surface, the inside surface flaw may be more limiting. Consequently, for the heatup analysis both the inside and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.
During cooldown, the thermal gradients through the reacter vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and which are compressive at the reactor vessel outside surface.
Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface location. Since the neutron irradiation damage is also greatest at the inside surface location the inside surface flaw is the limiting location. Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.
SAN ONOFRE-UNIT 2 B 3/4 4-6 AMENDMENT NO. 70
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4 REACTOR COOLANT SYSTEM BASES
-PRESSURE /TEMPERATURELIMITS-(Continueg The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) are composite curves which were prepared by determining the most conservative case, with either'the inside or outside wall controlling, for any heatup rate of up to 60'F/hr or cooldown rate _of up to 100*F/hr..The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the'end of the service period, and they include adjustments'for possible errors in the pressure and temperature. sensing instruments.
The rea:: tor vessel materials have been tested to determine their initial RT the results of these tests are shown in Table.B'3/4.4-1.
tiURT;ndresuitantrastneutron(EgreaterthanIMev)irradiationwilicauseReactor opera -
a an increase in the RT Therefore, an adjusted reference temperature, based upon.thefluenceandUkp.er and nickel content of the material in question, can be predicted using FSAR Table 5.2-5 and the recommendations of Regulatory Guide-1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials."
- j The'heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted-adjustments for this shift in RT at the end of the applicable service period, aswellasadjustmentsforpossi$Nerrorsinthepressureandtemperaturesensing I
instruments.
The actual shift in RT periodicallyduringoperatiggTofthevesselmaterialwillbeestablished i
by removing and evaluating, in accordance with
)
ASTM.E185-73.and 10 CFR 50 Appendix H, reactor vessel material irradiation sur-veillance specimens installed near the inside wall of the reactor vessel in f
the core area. The surveillance specimen withdrawal schedule is shown in i
Table 4.4-5.
Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor j
vessel taking into account the location of the sample closer to the core than j
the vessel wall by means of the Lead Factor. The heatup and cooldown curves must be recalculated when the delta RT determined from the surveillance capsuleisdifferentfromthecalculatg'gTdelta RT the equivalent capsule i
NDT radiation exposure.
j The pressure-temperature limit lines shown on Figure 3.4-2 and 3.4-3 for
)
reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of L
Appendix G to 10'CFR 50.
The maximum RT for all reactor coolant system pressure-retaining materials, with the Nception of the reactor pressure vessel, has been deter-N mined to be 90*F.
The Lowest Service Temperature limit line shown on Figure 3.4-2 and 3.4-3 is based upon this RT since Article NB-2332 (Summer Addendaof1972)ofSectionIIIoftheASMElN1erandPressureVesselCode requires the Lowest Service Temperature to be RT 100'F for piping, pumps andvalves.Belowthistemperature,thesystempEs+uremustbelimitedtoa maximum of 20% of the system's hydrostatic test pressure of 3125 psia.
The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pres-surizer is operated within the design criteria assumed for the fatigue analysis performed in accords.;;e with the ASME Code requirements.
SAN ONOFRE-U. NIT 4 8 3/4 4-7 AMENDMENT NO.70 j
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i REACTOR COOLANT SYSTEM BASES i
PRESSURE / TEMPERATURE LIMITS (Continued)
The OPERABILITY of the Shutdown Cooling System relief valve or a RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from presture transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to that specified in Table 3.4-3.
The Shutdown Cooling System relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water tem-perature of the steam generator less than or equal to 100 F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with two HPSI pumps injecting into a water-solid RCS with full charging capacity and letdown isolated.
3/4.4.9 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i).
Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975.
3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM Reactor coolant system gas vents are provided to exhaust noncondensible gases from the primary system that could inhibit natural circulation core cooling folleving a non-design bases accident. The OPERABILITY of at least one reactoi coolint syster vent path from the reactor vessel head and the pressuriz e steam space ensures the capability exists to perform this function.
The design redundancy of the Reactor Coolant Gas Vent System serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control system does not prevent isolation of the vent path.
The function, capabilities, and testing requirements of the Reactor Coolant i
Gas Vent System are consistent with the requirements of Item II.b.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980.
SAN ON0FRE-UNIT 2 8 3/4 4-9 AMENDMENT NO. 70 l
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