ML20086S388

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Amends 122 & 111 to Licenses NPF-10 & NPF-15,respectively, Revising RCS P-T Limits & LTOP Enable Temps to Be Effective Until 20 EFPY of Operation & Makes Minor Editorial Changes
ML20086S388
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 07/18/1995
From: Fields M
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20086S393 List:
References
NUDOCS 9508010207
Download: ML20086S388 (52)
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Text

{{#Wiki_filter:. LJ yems I% UNITED STATES j NUCLEAR REGULATORY COMMISSION f WASHINGTON, D.C. 20616 4001 %,.....f SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE. CALIFORNIA THE CITY OF ANACIM. CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.122 License No. NPF-10 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Southern California Edison Company, et al. (SCE or the licensee) dated April 30, 1993, as supplemented by letters dated July 6,1994, and January 27, 1995, ccmplies 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; i C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the puo'.ic, and (ii) that such activities will be conducted in compliance with the Comission's regulations; 1 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. 4 9508010207 950718 PDR ADOCK 05000361 P PDR;

,l. + m. / 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 Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 122, 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 to be implemented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMISSION Mel 8. Fields, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attachmer. : Changes to the Technical Specifications Date of Issuance: July 18, 1995 I 1

2. s: -e ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.122 TO 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 Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness. REMOVE INSERT v v X I V. XIX yk! XXI XXII 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-29a 3/4 4-30 3/4 4-30 3/4 4-30a 3/4 4-30b 3/4 4-30c 3/4 4-30a 3/4 4-30d 3/4 4-32 3/4 4-32 3/4 4-33 3/4 4-33 8 3/4 4-7 B 3/4 4-7 B 3/4 4-7a B 3/4 4-8 8 3/4 4-8 B 3/4 4-8a l l l i i

l+ C s INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE HOT SHUTD0WN............................................ 3/4 4-3 COLD SHUTDOWN - Loops Fi11ed............................ 3/4 4-5 COLD SHUTDOWN - Loops Not Fi11ed........................ 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 s 256'F............................ 3/4 4-32 RCS TEMPERATURE > 256*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 l 1 3/4.5.1 SAFETY INJECTION TANKS.................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,y, 2 350*F.......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,y, < 350*F.......................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................ 3/4 5-8 \\ SAN ON0FRE-UNIT 2 V Amendment No. 70,122 i

L INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE i 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT i CONTAINMENT INTEGRITY................................ 3/4 6-1 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 CONTAINHENT STRUCTURAL INTEGRITY..................... 3/4 6-9 CONTAIMENT VENTILATION SYSTEM....................... 3/4 6-13 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM............................. 3/4 6-14 IODINE REMOVAL 5YSTEM................................ 3/4 6-16 CONTAINMENT COOLING 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 HYOROGEN REC 0MBINERS........................ 3/4 6-27 CONTAI MENT DOME AIR CIRCULAT0RS..................... 3/4 6-28 6 SAN ONOFRE-UNIT 2 VI

2 m' = INDEX LIST OF TABLES i TABLE PA_GE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION......................... 3/4 3-52 t 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................................ 3/4 3-54 l 3.3-11 FIRE DETECTION INSTRUMENTS................................... 3/4 3-57 3.3-13 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION..... 3/4 3-65 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING 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 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 4-30d 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS................................................ 3/4 4-22 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..................................................... 3/4 4-25 4.4-5 DELETED 4.6-1 TENDON SURVEILLANCE......................................... 3/4 6-12 4.6-2 TENDON LIFT-0FF FORCE....................................... 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-HIGH TRIP WITH j OPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION i WITH BOTH STEAM GENERATORS.................................. 3/4 7 ' SAN ON0FRE - UNIT 2 XIX Amendment No. 91,114,121,122

g 1 LIST OF TABLES TABLE PA[iE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM 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 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 RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM -- DELETED 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AtlD ANALYSIS PROGRAM -- DELETED 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM -- DELETED 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES -- DELETED 4".12-1 MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) -- DELETED B 3/4.4-1 REACTOR VESSEL TOUGHNESS 8 3/4 4-8 ) 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS 5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION 6-4 I i SAN ONOFRE-UNIT 2 XX AMENDMENT NO. 83

g, INDEX LIST OF FIGURES fit;URES PfE 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 VS FRACTION OF ALLOWABLE THERMAL P0WER.................................................... 3/4 1-24 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 VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 vCi/ GRAM DOSE EQUIVALENT I-131........................................ 3/4 4-26 3.4-2 SONGS 2 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY-NORMAL 0PERATION........................... 3/4 4-29 3.4-4 SONGS 2 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY-NORMAL 0PERATION........................... 3/4 4-30 3.4-5 SONGS 2 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY)-NORMAL 0PERATION................................................ 3/4 4-30a 3.4-6 SONGS 2 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY-REMOTE SHUTDOWN 0PERATION.................. 3/4 4-30b 3.4-7 SONGS 2 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY)-REMOTE SHUTDOWN 0PERATION....................................... 3/4 4-30c 3.7-1 MINIMUM REQUIRED FEEDWATER 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 B0UNDARY FOR GASE0US EFFLUENTS...................... 5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS....................... 5-5 SAN ONOFRE-UNIT 2 XXI Amendment No. B7, 122

c INDEX LIST OF FIGURES FIGURES PAGE 5.6-1 UNITS 2 & 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS...................................... 5-12 5.6-2 UNIT I FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS...................................... 5-13 5.6-3 FUEL STORAGE PATTERNS FOR REGION II RACKS................ 5-14 5.6-4 FUEL STORAGE PATTERNS FOR REGION II RACKS RECONSTITUTION. STATION................................... 5-15 6.2-1 0FFSITE ORGANIZATION..................................... 6-2 6.2-2 UNIT ORGANIZATION........................................ 6-3 6.2-3 CONTROL R00M AREA........................................ 6-4a i i l l i SAN ON0FRE-UNIT 2 XXII Amendment No.122 l

S..,' i BMGJOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDIT!rg{, FOR OPERATION 3.4.8.1 With the reactor vessel head bolts tensioned *, the Reactor Coolant I System (except the pressurizer) temperature and pressure shall be limited in accordance yith the limit lines shown on Figures 3.4-2, 3.4-4, 3.4-5, 3.4-6, and 3.4-7 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: i a. A maximum heatup of 60*F in any 1-hour period with RCS cold leg l temperature greater than or equal to 86*F. i b. A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period with RCS cold leg temperature less than or equal to 160*F. A maximum cooldown of 100*F in any 1-hour period with RCS cold leg temperature greater than 160*F. c. A maximum temperature change of 10*F in any 1-hour period during l 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. ) With the reactor vessel head bolts detensioned, the Reactor Coolant System (except the pressurizer) temperature shall be limited to a maximum heatup or cooldown of 60*F in any 1-hour period. 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 hours and reduce the RCS T and pressure to less than 200*F and 500 psia,respectively,withinthefollowYng30 hours. i i

  • With the reactor vessel head bolts detensioned, RCS cold leg temperature i

may be less than 86*F. I SAN ON0FRE - UNIT 2 3/4 4-27 Amendment No. M,122

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM t SURVEILLANCE REOUIREMENTS 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 inseryt:e leak and hydrostatic testing operations. 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50 Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-4 through 3.4-7. Recalculate the Adjusted Reference Temperature in accordance with Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988. i r P i i SAN ONOFRE - UNIT 2 3/4 4-28 Amendment No. 70,121,122

-,s t n v f i saco LOWEST SERVICE o INSERVICE TESTS

  1. HEATUP i

TEMP = 209'F ^ l{ i e ! i i1.I f ? ...t i ;... ...p..., 3oon iiit i i

  • Acceptable operating region to the right of the inservice tests curve Tl

~[T"I"'

1

~"~~ "~"' (Applicable in modes other than !,;'^ Modes 1 and 2) - l-2500

  • +-

+<

  1. Acceptable operating region to the

. j._...;.........;,,,; -I right of the heatup curve in all modes. ~ e in addition, in Modes 1 and 2 the ~"' ~" operating region is to the right of the T g core critical curve. l- "~ ~ g

  • +T+-+'--~-l-*-*-----------

2000 i m I w .8. c. i e g g b l 1500 ga M l u) a w [ .# CORE i o ,non . CRITICAL, w i p g 5 i z l j. a MINIMUM BOLTUP TEMP. 86'F ...A ,,..f... .l.i.iI. ..f....l.. l....l.... o 50 too 150 200 250 300 350 400 .i INDICATED RCS TEMPERATURE (?F)-Tc Figure 3.4-2 i l SONGS 2 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Normal Operation SAN ONOFRE - UNIT 2 3/4 4-29 Amendment No. 70,122 l l

t. l l I (Figure 3.4 Not Used) 1 3 l 1 l I SAN ON0FRE - UNIT 2 3/4 4-29a Amendment No.122 l

g-g t e s i i 3500 ...g....,... g....,... LOWEST SEFMCE COOLDOWN TEMP = 209'F 3000 I 8 3 : ..,s.. 2..; t ! . i; s + l !;ii32. g ; i . 3 r f f,,... i ......;-.y:....i....f....! 7....!. 7,e.- ........a. .., !... l 3 4 j } .[ j -} }[ - l } i -t......... 7 g 23ao ..._._......J..il j -- !j-l s.. Q. .. j. r w ...._.........................._............t.._.. l j g ,... '... '.,..,j j jl8 ! i ; - ' - g ctr- - --- +- ! I f ,I._ i.___.____.___._ W 2000 ! I i f ~ 'i

  • d;f.

l- .i l Unacceptable - - - ~ ~ - - - Operating d Region sr t o l Acceptable g 1500 Operating ~s-wc a Regen + g i o a w I g s. 2000 - l 9 I o e 2 4 5 4 500 - / min 6 MUM BOLTUP i f TEMP = 66'F A ,f;,;,i,;,.I,,,,l,,.,I,.. 1.,..I,,,, 15 0 50 100 150 200 250 300 360 400 I INDICATED RCS TEMPERATURE (*F)-Tc t e FIGURE 3.4-4 i SONGS 2 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Normal Operation 1 3 SAN ONOFRE - UNIT 2 3/4 4-30 Amendment No. M,122

=. i 120 i g g g 5 3 i 3 3 g 110 100 so iE e0 E u. ~ n w f m g 8 o e 30 20 - 10 - o I t ,t I t t e i t e e e 80 90 100 110 120 130 140 150 160 170 180 190 200 210 INDICATED RCS TEMPERATURE (*F)-Tc NOTE: A MAXIMUM C00LDOWN RATE OF 100*F/HR IS ALLOWED AT ANY TEMPERATURE AB0VE 160*F FIGURE 3.4-5 SONGS 2 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY) Normal Operation SAN ONOFRE - UNIT 2 3/4 4-30a Amendment No.122

y 4 ., e *~ e' 3500 ,g,,,,,,,,,,,,,,,,,,,,,,,,,,,,, LOWEST SERVICE COOLDOWN TEMP = 209'F ,j s ! ! i ....;...+.. s.. 9.;..!.J..;;.;I ..s g i !. a. l....).....i.. .L...... ,.;,,3..;,,.,1,,_ _, s....... i . i. i .........f.. j,.. h.........!...j. 9.;.. ;.. 4... ..,,.l...,, ;,, J,,, _, 7...!..,t...!. :....i... :...;:.. ;.......;.;. i 0-2500 .2 a...l... ....!...j.... ..c, W ...:... a..a. ' C 3 ..............I......I......f.......,I..,. k. M .. 4...} j...!.. 4...;. !..,j.......; 3..,,...i.. W C l....j.... *...L. O.- 2000 C

Unacceptable

~ w Operating b! C

Region D

e M s e .l Acceptable 1500 C Operating e 1 l Region O W I >= 4 i g i t Q '000 - s a i l = l 4 \\ 1 l 500 - min 1 MUM BOLTUP TEMP = 86'F ,I,, ,,i,,,,e,,,.i,,,,e,,,,i,,T,e,,,,' ,3 O 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F)-Tc FIGURE 3.4-6 SONGS 2 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Remote Shutdown Operation SAN ONOFRE - UNIT 2 3/4 4-30b Amendment No.122

l 1 P 120 .i 7" i i i i i i ? 150 i l 100 t w so 70 E E p. s0 w i E 50 E z g c l c s g 30 - o 20 10 n e t ,e I t I t i i t t t 80 90 100 110 120 130 140 150 160 170 180 190 200 210 'NDICATED RCS TEMPERATURE (*F)-Tc i NOTE: A MAXIMUM COOLDOWN RATE OF 100*F/HR IS ALLOWED AT ANY TEMPERATURE AB0VE 168'F i FIGURE 3.4-7 SONGS 2 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY) i Remote Shutdown Operation i SAN ONOFRE - UNIT 2 3/4 4-30c Amen.s ont No.122

.o Table 3.4-3 Low Temperature RCS Overoressure Protection Ranae l I Operatina Period. EFPY Cold Leo Temoerature. *F I During During Heatuo Cooldown Until 20 (Normal Operation) s 256 s 238 l s 238 l Until 20 (Remote Shutdown Operation) i

  • Heatup operations are not normally performed from the Remote Shutdown panels.

SAN ON0FRE - UNIT 2 3/4 4-30d Amendment No. M,122

f. REACT 0a COOLANT SYSTEw PRE 55URI2ER - HEATUD/COOLDOWN LIMITING CONDITION FOR OPERATION 3.4.8.2 The pressurizer shall be limited tc. A maximum heatup of 200'F in any one hour pt riod, a. b. A maximum cooldown of 200'F in any one hour geriod. APPLICABILITY: At all times. A ' ION: with the pressurizer toecerature limits in excess of any of the above limits, restere the temperature to within the limits within 30 minutes; perfors an engineering evelvetion to cetermine the effects of the out-of-limit condition on tne structurel integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDSY within the next 6 hours and reduce the pressuriger pressure to less than 500 psig within the following 30 hours. SURVEILLANCE REQUIREMENTS 4.a.8.2.1 The pressurizer tesceratures shall be detensined to be within the limits at least once per 30 minutes during systes heatup or cooldown. a.4.8.2.2 The spray water toeparature differential shall be determined for use in Table 5.7-1 for each cycle of main spray when less than 4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation. SAN ON0FRE - UNIT 2 3/4 4*31 AMENDMENT NO. M

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 256*F l 1 LIMITING CONDITION FOR OPERATION 3.4.8.3.1 No more than two high-pressure safety injection pumps shall be OPERABLE and at least one of the following overpressure protection systems 3 shall be OPERABLE: j a. The Shutdown Cooling System Relief Valve (PSV9349) with: 1) A lift setting of 406 i ',0 psig*, and 1 2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 open i or, i 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 the enable temperatures specified in Table 3.4-3; MODE 5; and MODE r ~5en the head is on the reactor vessel and the RCS is not vented. ACTION: With the SDCS Relief Valve inoperable, reduce T,y, to less than i a. 200*F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours. 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) or power-lock open the OPERACLE SDCS Relief Valve isolation valve pair J within 24 hours, or reduce T,y, to less than 200*F, depressurize and vent the RCS through a greater than or equal to 5.6 inch vent within the next 8 hours. c. With more than two high-pressure safety injection pumps OPERABLE, secure the third high-pressure safety injection pump by racking out its motor circuit breaker or locking close its discharge valve within 8 hours.

  • The lift setting pressure applicable to valve temperatures of less than or equal to 130*F.

SAN ON0FRE - UNIT 2 3/4 4-32 Amendment No. Mrl45,122

M.* i 4 1 REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE > 256*E i LIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems shall be OPERABLE: The Shutdown Cooling System Relief Valve (PSV9349) with: a. 1) A lift setting of 406 i 10 psig*, and 2) Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and I i 2HV9378 open I or, b. A minimum of one pressurizer code safety valve with a lift setting of 2500 psia i 1%**. APPLICABILITY: MODE 4 with RCS temperature above that specified in Table 3.4-3. ACTION: With no safety or relief valve OPERABLE, be in COLD SHUTDOWN and a. vent the RCS through a g reater than or equal to 5.6 square inch i vent within the next 8 hours. b. In the event the SDCS Relief Valve is used to mitigate an RCS l pressure transient, a Special Report shall be prepared and submitted h to the Commission pursuant to Specification 6.9.2 within 30 days, i The report shall describe the circumstances initiating the transient, the effect of the SDCS Relief Valve code safety valve on l 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: Verifying at least once per 72 hours that the SDCS Relief Valve a. isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open when the SDCS Relief Valve is being used for overpressure protection. l 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 l requirements other than those required by Specification 4.0.5. f

  • The lift setting pressure applicable to valve temperatures of less than or l

j equal to 130*F.

    • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE - UNIT 2 3/4 4-33 Amendment No. 70,122

.,g. REACTOR COOLANT SYSTEM [ 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. a b. With the structural integrity of any ASME Code Class 2 component (s) I 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. c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural 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. ~ SURVEILLANCE REQUIREMENTS 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. i l l l SAN ONOFRE-UNIT 2 3/4 4-34 i

.y,. ~, REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The heatup and cooldown limit curves for normal operation (Figures 3.4-2 and 3.4-4) and the cooldown limit curve for remote shutdown operation (Figure 3.4-6) 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 limit curves for Remote Shutdown operation are determined using the Total Loop Uncertainties (TLUs) for temperature and pressure for the Remote Shutdown Panel instruments in which the pressure TLUs are higher than those for the Control Room shutdown instruments. 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 instrument uncertainties, and static and dynamic heads. The reactor vessel materials were tested prior to reactor startup to determine their initial RT,,; the results of these tests and the updates resulting from the evaluation of material properties in response to Generic Letter 92-01, " Reactor Vessel Structural Integrity," Revision 1 are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (F greater than 1 Mev) irradiation will cause an increase in the RT Therefore, an adjusted l i or. reference temperature, based upon the fluence and copper and nickel content of t 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 i of Reactor Vessel Materials." The heatup limit curve (Figure 3.4-2) and the cooldown limit curves, figures 3.4-4 and 3.4-6, include predicted adjustments for this shift in RT, at the end of the applicable service period, as well as adjustments for instrument uncertainties, and static and dynamic heads. l The actual shift in RT of the vessel material will be established periodically during operati,by removing and evaluating, in accordance with on ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is maintained in the FSAR. 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 vessel taking into account the location of the sample closer to the core than 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 capsule is different from the calculatId delta RT, for the equivalent capsule l radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. SAN ON0FRE - UNIT 2 B 3/4 4-7 Amendment No. 70,121, 122

'.T* REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) materials, with the., for all reactor coolant system pressure-retaining The maximum RT exception of the reactor pressure vessel, has been determined to be 90*F. The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-4, and 3.4-6 is based upon this RT since Article NB-2332 l (Summer Addenda of 1972) of.Section III of the ASME BNier and Pressure Vessel Code requires the Lowest Service Temperature to be RTer + 100*F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a 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 pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. The Low Temperature Overpressure Protection (LTOP) enable temperatures are based upon the recommendations of NUREG-0800 Branch Technical Position (BTP) RSB 5-2, Revision 1, "0verpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures." BTP RSB 5-2, Revision 1 defines the enable temperature as "the water temperature corresponding to a + 90*F at the beltline location (1/4t or metal temperature of at least RT* Appendix G limit calculations." 3/4t) that is controlling in the l 1 l 4 SAN ON0FRE - UNIT 2 B 3/4 4-7a Amendment No. 122 l l l

0N TABLE B 3/4.4-1 y REACTOR VESSEL TOUGHNESS E Temperature of Minimum Upper ? Drop Charpy V-Notch Shelf Cv energy M Weight 9 30 9 50 for Longitudinal i Piece No. Code No. Material Vessel Location Results ft - lb - ft - lb Direction-ft lb C5 215-01 C-6403-1 A533GRBCL-1 Upper Shell Plate 40 15 35 130 215-01 C-6403-2 A533GRBCL-1 Upper Shell Plate 0 20 25 133 ~ 215-01 C-6403-3 A533GRBCL-1 Upper Shell Plate -10 20 45 131 215-03 C-6404-1 A533GRBCL-1 Intermed. Shell Plate -30 40 80 119 215-03 C-6404-2 A533GRBCL-1 Intermed. Shell Plate -20 70 80 113 215-03 C-6404-3 A533GRBCL-1 Intermed. Shell Plate -20 70 80 99 215-02 C-6404-4 A533GRBCL-1 Lower Shell Plate -10 -40 80 104 215-02 C-6404-5 A533GRBCL-1 Lower Shell Plate -20 50 70 118 215-02 C-6404-6 A533GRBCL-1 Lower Shell Plate -10 50 50 124 238-02 C-6401 A508Cl-2 Vessel Flange Forging -10 -70 -35 148 ? 209-02 C-6402 A508Cl-2 Closure Head Flange -10 -90 -40 142 l Forging 205-02 C-6410-1 A508Cl-2 Inlet Nozzle Forging 20 -40 -35 130 205-02 C-6410-2 A508Cl-2 Inlet Nozzle Forging 0 -20 -5 135 205-02 C-6410 '1 A508Cl-2 Inlet Nozzle Forging 0 -15 -15 140 205-02 C-6410- 4 A508Cl-2 Inelt Nozzle Forging 0 -65 -50 140 205-06 C-6411-1 A508Cl-2 Outlet Nozzle Forging -100 -30 -10 140 205-06 C-6411-2 A508Cl-2 Outlet Nozzle Forging 0 -35 -10 140 [ 232-01 C-6424 A533GRBCL-1 Bottom Head Torus -50 -20 10 122 232-02 C-6425 A533GRBCL-1 Bottom Head Dome -50 -30 -20 136 [ E 205-03 C-6428-1 A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 205-03 C-6428-2 A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 C 205-03 C-6428-3 A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 205-03 C-6428-4 A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 a-

TABLE B 3/4.4-1 (Continued) $2 Temperature of Minimum Upper le Drop Charpy V-Notch Shelf CV energy E4 Weight 9 30 9 50 for Longitudinal E8 Piece No. Code No. Material Vessel location Results ft - lb - ft - lb Direction-ft lb c: 205-07 C-6429-1 A508CL-1 Outlet Nozzle Ext. -30 -40 -25 229 l Forging 205-07 C-6429-1 A508CL-1 Outlet Nozzle Ext. -30 -40 -25 229 Forging l 231-02 C-6430-1 A533GRBCL-1 Closure Head Peels +10 20 55 118-231-02 C-6431-1 A533GRBCL-1 Closure Head Peels -20 10 50 100. 231-02 C-6432-1 A533GRBCL-1 Closure Head Peels -10 -15 45 115 231-02 C-6432 A533GRBCL-1 Closure Head Dome -10 -15 45 115 co Ua A A ? 8it i an no l ~~ e

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UNITED STATES NUCLEAR REGULATORY COMMISSION N f WASHINGTON, D.C. 2006Hm01 "s.,...../ I 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 AMENDMENT TO FACILITY OPERATING LICENSE I Amendment No.111 License No. NPF-15 l 1. The Nuclear Regulatory Comission (the Commission) has found that: A. The application for amendment by Southern California Edison j Company, et al. (SCE or the licensee) dated April 30, 1993, as supplemented by letters dated July 6, 1994, and January 27, 1995, 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; 1 B. The facility will operate in conformity with the application, the i provisions of the Act, and the rules and regulations of the l Comission; I C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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. b

  • . I 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 Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 111, 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 to be implemented within 30 days from the date of issuance. FOR THE NUCLEAR REGULATORY COMMISSION Mel B. Fields, Project Manager Project Directorate IV-2 Civision of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: July 18, 1995 l

= ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.111 TO 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 Amendment number and contain marginal lines indicating the areas of change. The corresponding overleaf pages are also provided to maintain document completeness. REMOVE INSERT i V V XVII XVII XVIIa XVIIa XIX XIX 3/4 4-28 3/4 4-28 3/4 4-29 3/4 4-29 3/4 4-30 3/4 4-30 3/4 4-30a 3/4 4-30a 3/4 4-31 3/4 4-31 3/4 4-31a 3/4 4-31a 3/4 4-31b 3/4 4-31c 3/4 4-31b 3/4 4-31d 3/4 4-33 3/4 4-33 3/4 4-35 3/4 4-35 B 3/4 4-7 B 3/4 4-7 B 3/4 4-7a B 3/4 4-8 8 3/4 4-8 l

hw ^ v INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PMf I 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 l 3/4.4.4 STEAM GENERATORS........................................ 3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................ 3/4 4-17 OPERATIONAL LEAKAGE.................................. 3/4 4-18 3/4.4.6 CHEMISTRY............................................... 3/4 4-21 3/4.4.7 SPECIFIC ACTIVITY....................................... 3/4 4-24 1 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM............................... 3/4 4-28 PRESSURIZER - HEATUP/C00LDOWN........................ 3/4 4-32 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 246*F............................ 3/4 4-33 RCS TEMPERATURE > 246*F............................ 3/4 4-35 3/4.4.9 STRUCTURAL INTEGRITY.................................... 3/4 4-36 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM......................... 3/4 4-37 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS.................................. 3/4 5-I 3/4.5.2 ECCS SUBSYSTEMS - T,,,2 350*F.......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,y, < 350'F.......................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................ 3/4 5-8 SAN ON0FRE - UNIT 3 V Amendment No. 71,111

s '. - 4 l l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE-3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT i CONTAINMENT INTEGRITY................................ 3/4 6-1 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 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS l CONTAINMENT SPRAY SYSTEM............................. 3/4 6-15 IODINE REMOVAL SYSTEM................................ 3/4 6-17 l CONTAINMENT COOLING SYSTEM........................... 3/4 6-18 3/4.6.3 CONTAINMENT ISOLATION VALVES............................ 3/4 6-19 1 3/4.6.4 COMBUSTIBLE GAS CONTROL HYDROGEN M0NITORS.................................... 3/4 6-27 ELECTRIC HYDROGEN REC 0MBINERS........................ 3/4 6-28 CONTAINMENT DOME AIR CIRCULAT0RS..................... 3/4 6-29 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES...................................... 3/4 7-1 AUXILIARY FEEDWATER SYSTEM........................... 3/4 7-4 CONDENSATE STORAGE TANKS.............. 3/4 7-6 ACTIVITY.............................. 3/4 7-8 MAIN STEAM LINE ISOLATION VALVES................ 3/4 7-10 ATMOSPHERIC DUMP VALVES......................... 3/4 7-10a l. l l SAN ONOFRE - UNIT 3 VI AMENDMENT NO. 79 i L

lbg: 'T ~ ~ INDEX LIST OF FIGURES FIGURE EAE 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND' TEMPERATURE AS A FUNCTION OF STORED BORIC ACID CONCENTRATION.......... 3/4 1-13 3.1-2 CEA INSERTION LIMITS VS FRACTION OF ALLOWABLE THERMAL P0WER..................................................... 3/4 1-24 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 i 3.3-1 DEGRADED BUS V0LTAGE TRIP SETTING......................... 3/4 3-40 4.4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA.................... 3/4 4-16 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE PRIMARY COOLANT SPECIFIC ACTIVITY >l.0 Ci/ GRAM DOSE EQUIVALENT I-131.......................................... 3/4 4-27 3.4-2 SONGS 3 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY-NORMAL 0PERATION............................ 3/4 4-30 3.4-3 DELETED 3.4-4 SONGS 3 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS 4 UNTIL 20 EFPY-NORMAL 0PERATION............................. 3/4 4-31 j 3.4-5 S0llGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY)-NORMAL OPERATION........... 3/4 4-31a 3.4-6 SONGS 3 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY-REMOTE SHUTDOWN 0PERATION................... 3/4 4-31b 3.4-7 SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY)-REMOTE SHUTDOWN OPERATION., 3/4 4-31c 3.7-1 MINIMUM REQUIRED FEEDWATER INVENTORY FOR TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE............................ 3/4 7-7 5.1-1 EXCLUSION AREA................................ 5-2 5.1-2 LOW POPULATION Z0NE....................................... 5-3 5.1-3 SITE B0UNDARY FOR GASE0US EFFLUENTS....................... 5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS........................ 5-5 SAN ONOFRE - UNIT 3 XVII Amendment No. 77,111

1 3 = LIST OF FIGURES F:GURE PE 5.6-1 UNITS 2 AND 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS.......................... 5-12 5.6-2 UNIT 1 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS..... 5-13 5.6-3 FUEL STORAGE PATTERNS FOR REGION II RACKS............... 5-14 5.6-4 FUEL STORAGE PATTERNS FOR REGION II RACKS RECONSTITUTION STATI0N.................................. 5-15 6.2-1 0FFSITE ORGANIZATION.................................... 6-3 6.2-2 UNIT ORGANIZATION....................................... 6-4 6.2-3 CONTROL ROOM AREA....................................... 6-6 SAN ON0FRE - UNIT 3 XVIIa Amendment No. N,111

Ts INDEX LIST OF TABLES TABLE PAGE 1.1 OPERATIONAL MODE5.......................................... 1-7

1. 2 FREQUENCY N0TATION.........................................

1-8 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS.... 2-3 2.2-2 CORE PROTECTION CALCULATOR ADDRESSABLE CONSTANTS........... 2-5 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION......................... 3/4 3-3 3.3-2 REACTOR PROTECTIVE INSTRUMENTATION RESPONSE TIMES.......... 3/4 3-8 4.3-1 REACTOR PROTECTIVE INSTRUMENTATION SURVEILLANCE REQUIREMENTS.......................................... 3/4 3-10 3.3-3 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............................................ 3/4 3-14 3.3-4 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP VALUE5................................ 3/4 3-22 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIME5.................. 3/4 3-27 4.3-2 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLMENTATION SURVElLLANCE REQUIREMENTS.................. 3/4 3-31 3.3-6 RADIATION MONITORING ALARM INSTRUMENTATION................. 3/4 3-35 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................... 3/4 3-38 3.3-7 SEISMIC MONITORING INSTRUMENTATION......................... 3/4.3-43 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................... 3/4 3-44 3.3-8 METECR0 LOGICAL MONITORING INSTRUMENTATION.................. 3/4 3-46 4.3-5 METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................... 3/4 3-47 3.3-9 REMOTE SHUTDOWN MONITORING INSTRUMENTATICN................. 3/4 3-49 4.3-6 REMOTE SHUTDOWN MCNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................'.............. 3/4 3 ~3 3.3-10 ACCIDENT MONITORING INSTRUMENTATION........................ 3/4 3-5' SAN ONCFRE-UNIT 3 XVIII

p es, INDEX LIST OF TABLES TABLE PEE f 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................................................ 3/4 3-55 l 3.3-11 FIRE DETECTION INSTRUMENTS.................................. 3/4 3-58 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION..... 3/4 3-66 j 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION l SURVEILLANCE REQUIREMENTS................................... 3/4 3 ' 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING i INSERVICE INSPECTION........................................ 3/4 4-14 I 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 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS................................................ 3/4 4-23 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS l PR0 GRAM...................................................... 3/4 4-26 4.4-5 DELETED l 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE........... 3/4 4-31d l i 4.6-1 TENDON SURVEILLANCE......................................... 3/4 6-12 4.6-2 TENDON LIFT-0FF FORCE....................................... 3/4 6-13 1 3.6-I 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 IN0PERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS.................................. 3/4 7-3 SAN ON0FRE - UNIT 3 XIX Amendment No. 81,103,110,111

E 4 1 i INDEX LIST OF TABLES TABLE EAGE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..................... 3/4 7-9 l 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 l 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 PERMANENTLY BYPASSED............... 3/4 8-32 4.11-1 RADI0 ACTIVE LIQUID SAMPLING AND ANALYSIS PROGRAM -- DELETED 4.11-2 RADI0 ACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM -- DELETED 3.12-1 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM -- DELETED 3.12-2 REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES -- DELETED 4.12-1 MAXIMUM VALVES FOR THE LOWER LIMITS OF DETECTION -- DELETED 9 B3/4.4-1 REACTOR VESSEL TOUGHNESS................. B3/4 4-8 5.7-1 COMPONENT CYCLIC OR TRANSIENT LIMITS........... 5-8 6.2-1 MINIMUM SHIFT CREW COMPOSITION.............. 6-5 SAN ONOFRE - UNIT 3 XX AMENDMENT NO. 73

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^ , 9 -+- T ri, t-g . 7.,,; ,7 - q g 100 _L m 4 ..:_;.4. ' j _L 4_ 4_.__,l '. 1.H,..,'-.Ai. r:i i - ---- 3 8 g. .2 7p. g .u._._ }..t. h,.. _ y. g_.. l- 't i-M-' -*---,, -. i . _..... _ -., _ _. -.., n. .. _r.-_-... g p _ _..u. 8 'iiL .;... _L I '.J_;..!_ _ ; s i.J.. z ACCEPTAB LE._ _-N_ q _L. 4 1 i 3 OPERATION -- -t-H 2 #' 50 g .~_~j.q,_. .4.' 4 -l, H 4 ;. ;... 4 ..__..i. l *- L - ~_t_- ' _,r + . ;_'. --- t- + ,..t. 3 .i... 4 4;jy,..._._p._J., .; 4.,g ..,.a.q.,r. g.7 4 o 1, j. .u.......,._....l...p. .i ; 9 9,. _ y.,..,,.. w ,i.1,.;..:.+.,1 4.. 2. p. ..t.. m 4_ _E : 4.I - g ._. H.__. g. .:... 4. - 4g_. 4_. } j. {..r_T ' _ - t r,. 4, _ - :-e t....".... _ _ -. g O 20 30 40 50 ' 80 70 80 90 100 PERCENT OF RATED THERMAL POWER i FIGURE 3,41 t DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity > 1.0pCi/ gram Dose Equivalent 1-131 1 SAN ONOFRE-tmIT 3 3/4 4 77

n REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1 With the reactor vessel head bolts tensioned *, the Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-4, 3.4-5, 3.4-6, and 3.4-7 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with: a. A maximum heatup of 60*F in any I-hour period with RCS cold leg l I temperature greater than or equal to 86*F. b. A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period with RCS cold leg temperature less than or equal to 147'F. A maximum l cooldown of 100*F in any 1-hour period with RCS cold leg temperature greater than 147'F. I c. A maximum temperature change of 10*F in any 1-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. With the reactor vessel head bolts detensioned, the Reactor Coolant System (except the pressurizer) temperature shall be limited to a maximum heatup or cooldown of 60*F in any 1-hour period. APPLICABILITY: At all times. ACTION: With any of the above limits exceeded, restore the temperature a,d/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 within the next 6 hours and reduce the RCS T and pressure to less than 200*Fand500 psia,respectively,withinthe7ollowing30 hours. l l

  • With the reactor vessel head bolts detensioned, RCS cold leg temperature may l

be less than 86*F. l SAN ON0FRE - UNIT 3 3/4 4-28 Amendment No. M,111

7- .u i REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS j REACTOR COOLANT SYSTEM SURVEILLANCE RE0VIREMENTS 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. 4.4.8.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, as required by 10 CFR 50 Appendix H. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-4 through 3.4-7. Recalculate the Adjusted i Reference Temperature in accordance with Regulatory Guide 1.99, Revision 2, i " Radiation Embrittlement of Reactor Vessel Materials," May 1988. t l ) i i l l l SAN ON0FRE - UNIT 3 3/4 4-29 Amendment No. M,111 i

n . o Y U g 5 u 5 g u s y y s y 3 y g y g g g a y a y y a g g g LOWEST SERVICE

  • INSERVICE TESTS
  1. HEATUP TEMP - 209'F -

2000 s .. s.. a +.. _.. s [ l y m g ,. e Acceptable operating region to the l ..y... . - ~... - D right of the inservice tests curve s (Applicable in rnodes other than j W Modes 1 and 2) C e m

  1. Acceptable operating region to the w

nght of the heatup curve in all rnodes. l b! In addition. in Modes 1 and 2 the a ~ operating region is to the nght of the j u) core cracal curve. y)w . t a C 1500 8 0-O w Q 9 0 3000 3

  1. CORE CRITICAL.

......... ] 500 ~ IANiMUM 90LTUP TEMP. 86*F .S ....l... iI....l.. .f...iI. ...l... 1.... o 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F)-Te FIGURE 3.4-2 SONGS 3 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Normal Operation 1 SAN ONOFRE - UNIT 3 3/4 4-30 Amendment No. M, til f

'.: 6, 1 i I (Figure 3.4 DELETED) zi i tl I SAN ON0FRE - UNIT 3 3/4 4-30a Amendment No. M,III

u. e. I ,g y g y 3500 COOLDOWN t.OWEST SERVICE TEMP - 209'F a f r : 3000 i ? . 8.. .q... f_ w ... +.. ' ~ ' " C '??~"'---*~"'"'*' 2500 ---j-' .. [ ... ;.1... ;..... ...2.., M t Ii4 .f l t .s. g W d 2000 E ...y. Unacceptable M Operating Region c C. gggg 1500 e Operating i -! i, R@ c3 .. l .z.......5,........ { ,v. 5000 . i.: i ..m , a n. A a... g .e.4... 1 RSNIMUM SOLTUP TEMP = 86*F ... ~ i g .1....t.. 1....t....i....t.... ....i... o so too tso 200 ano 300 sao ao 15 INDICATED RCS TEMPERATURE (*F)-Tc FIGURE 3.4-4 SONGS 3 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Normal Operation SAN ONOFRE - UNIT 3 3/4 4-31 Amendment No. M,111

.,v 1 ) i i i 120 t 110 100 i so ^c w z t w 70 h l m l m q 40 m m 10 = a i,i i i,i.i,e. .-i. i. i. i. o 80 90 100 110 f20 - 130 140 150 160 170 100 190 200 210 INDICATED RCS TEMPERATURE (*F)-Tc NOTE: A MAXIMUM C00LDOWN RATE OF 100*F/HR IS ALLOWED i AT ANY TEMPERATURE ABOVE 147'F FIGURE 3.4-5 SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY) Normal Operation i SAN ONOFRE - UNIT 3 3/4 4-31a Amendment No. M,111 1 b ,.,y .-6 p

e, gI gI gv gs gu g**[ M y a s v v a e a e a e a v r a v y LOWEST SERVICE COOLDOWN TEMP = 209'F i .....g ..y....... g 2 I ? g i g i. i.. - 1. .~ ,........4...........-.....~...-4..4....4... -l '.

  • -8'

' ~ - 2500 . i ; i i t ' !"= w i.. E i i

4 :. (. 4 i, f. 4....

. ;.... g 1-3 ....................5,..... O

r....;
4. 4...

g i W Unacceptable ct l 1 Operating -~ 2000 l 1 Regbn w l u. 4..... g g e-Operating i 3 i. g . s.. ... pggion w I . -... [... ..c.. W isoo C ..............4,.............4...... 1 J .......s.._.. I O 4 W >(= 8.. l 4.....................4....._... .. ~...... O 1000 Z i ,.4 ..~.. i 500 \\h 1 MINIMUM ! F t BCX. TUP

._4,..;..;......;..._............................

j TEMP 46*F g ii'!;i1....l... 1....t....l...i ..itii, i 15 o so 100 150 200 250 300 360 400 INDICATED RCS TEMPERATURE (*F)-Tc \\ FIGURE 3.4-6 8 SONGS 3 COOLDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS UNTIL 20 EFPY Remote Shutdown Operation I SAN ONOFRE - UNIT 3 3/4 4-31b Amendment No.111

,;v 2 L.- t 120 i i i - e i i i . - i 11o / too So g n u.L m m C 60 t z t so b f 4o O 3o 20 l to i.i.i..i. i i. 80 90 100 110 120 130 140 150 160 170 180 190 200 110 INDICATED RCS TEMPERATURE ('F)-Tc NOTE: A MAXIMUM C00LDOWN RATE OF 100*F/HR IS ALLOWED i AT ANY TEMPERATURE ABOVE 155'F FIGURE 3.4-7 t SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS t MAXIMUM ALLOWABLE C00LDOWN RATES (UNTIL 20 EFPY) i Remote Shutdown Operation SAN ONOFRE - UNIT 3 3/4 4-31c Amendment No.111

w I: Tabla 3.4-3 Low Temperature RCS Overoressure Protection Ranae Operatina Period. EFPY Cold Leo Temoerature. *F During During Heatuo Cooldown Until 20 (Normal Operation) s 246 s 225 Until 20 (Remote Shutdown Operation) s 225

  • Heatup operations are not normally performed from the Remote Shutdown panels.

) SAN ONOFRE - UNIT 3 3/4 4-31d Amendment No. 74 111 l

REACTOR COOLANT SYSTEM PRESSURIZER - HEATUP/COOLDOWN 6 LIMITING CONDITION FOR OPERATION i 3.4.8.2 The pressurizer shall be limited to: A maximum heatup of 200*F in any 1 hour period, a. t b. A maximum cooldown of 200'F in any I 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 emains acceptable for continued operation or be in at least HOT STANDBY i within the next 5 hours ano reduce the pressurizer pressure to less than 500 psig within the following 30 hours. SURVE:' aNCE REOU::EuENTS 4.4.8.2.1 The pressurizer temceratures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. t 4.4.8.2.2 The spray water temperature differential shall be determined for use in Table 5.7-1 for eacn cycle of main spray when less than 4 reactor coolant pumps are operating and for each cycle of auxiliary spray operation. l i l i SAN CNOFRE - UNIT 3 3/4 4-32 AMENCMENT NO. I3

b REACTOR COOLANT SYSTEM l OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 246*F l LIMITING CONDITION FOR OPERATION 3.4.8.3.1 No more than two high-pressure safety injection pumps shall be OPERABLE and at least one of the following overpressure protection systems shall be OPERABLE: a. The Shutdown Cooling System Relief Valve (PSV9349) with: l) A lift setting of 406 i 10 psig*, and 2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and 3HV9378 open l 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 the enable temperatures specified in Table 3.4-3; MODE 5; and MODE 6 when the head is on the reactor vessel and the RCS is not vented. ACTION: With the SDCS Relief Valve inoperable, reduce T,y, to less than a. 200*F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours. i b. With one or both SDCS Relief Valve isolation valves in a single SDCS l Relief Valve isolation valve pair (valve pair 3HV9337 and 3HV9339 or valve pair 3HV9377 and 3HV9378) closed, open the closed valve (s) or t power-lock open the other SDCS Relief Valve isolation valve pair within 24 hours, or reduce T,y, to less than 200'F, depressurize and i vent the RCS through a greater than or equal to 5.6 inch vent within the next 8 hours. c. With more than two high-pressure safety injection pumps OPERABLE, secure the third high-pressure safety injection pump by racking out its motor curcuit breaker or locking close its discharge valve within 8 hours. j J

  • The lift setting pressure applicable to valve temperatures of less than or equal to 130*F.

SAN ON0FRE - UNIT 3 3/4 4-33 Amendment No. 7+r94,ll!

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION: (Continued) 1 d. In the event either the SDCS Relief Vallve or an RCS vent is used to l mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Connission pursuant to Specification 6.9.2 within 30 days. The report shall describe the circumstances l initiating the transient, the effect of the SDCS Relief Valve or RCS vent on the transient and any correctieve action necessary to prevent recurrence. e. The provisions of Specification 3.0.4 are not applicable. l SURVEILLANCE RE0VIREMENTS i 4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated OPERABLE by: a. Verifying at least once per 72 hours when the SDCS Relief Valve is l being used for overpressure protection that SDCS Relief Valve isolation valves 3HV9337, 3HV9339, 3HV9377, and 3HV9378 are open. b. Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.S. 4.4.8.3.1.2 At least once per 12 hours, the third high-pressure safety 4 injection pump shall be demonstrated to be secured by verifying that its motor circuit breaker is not racked-in or its discharge. valve is locked closed. The requirement to rack out the third HPSI pump breaker is satisfied with the pump breaker racked out to its disconnected or test position. 4.4.8.3.1.3 At least once per 12 hours, the OPERABLE SDCS Relief Valve isolation valve pair (valve pair 3HV9337 and 3HV9339, or valve pair 3HV9377 and 3HV9378) that is used for overpressure protection due to the other SDCS Relief Valve isolation valve pair being INOPERABLE shall be verified to be in the power-lock open condition until the INOPERABLE SDCS Relief Valve isolation valve pair is returned to OPERABLE status or the RCS is depressurized and vented. The power-lock open requirement is satisfied either with the AC breakers open for valve pair 3HV9337 and 3HV9339 or the inverter input and output breakers open for valve pair 3HV9377 and 3HV9378, whichever valve pair is OPERABLE. 4.4.8.3.1.4 The RCS vent shall be verified to be open at least once per l 12 hours

  • when the vent is being used for overpressure protection.

l

  • 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.

SAN ONOFRE - UNIT 3 3/4 4-34 AMENDMENT NO.94

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE > 246*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: 1) A lift setting of 406 i 10 psig*, and 2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and 3HV9378 open I or, l b. A minimum of one pressurizer code safety valve with a lift setting of 2500 psia i 1%**. APPLICABILITY: 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 through a greater than or equal to 5.6 square inch vent within the next 8 hours, b. In the event the SDCS Relief Valve is used to mitigate an RCS l pressure transient, a Special Report shall be prepared and submitted I 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 on l the transient and any corrective action necessary to prevent recurrence. SVRVElllANCE RE0VIREMENTS ) 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by: a. Verifying at least once per 72 hours that the SDCS Relief Valve isolation valves 3HV9337, 3HV9339, 3HV9377 and 3HV9378 are open when 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. l

  • The lift setting pressure applicable to valve temperatures of less than or equal to 130*F.

4

    • The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

SAN ONOFRE - UNIT 3 3/4 4-35 Amendment No. M,111

,,.~, REACTOR COOLANT SYSTEM 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 tne affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature reauired by NDT considerations. b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above reouirements, restore the structural integrity of the affected component (s) to within its limit or 4solate the affected component (s) prior to increasing the Reactor Coolant System temperature aDove 200*F. c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component to within its limit or isolate t.9e affecteo component from service. c. 'he provisions of Specification 3.0.4 are not applicable. SURVEILLANCE REOUIREMENTS 4.4.9 In aedition to the reauirements of Specification 4.0.5, each Reactor Coolant Pumo flywneel shall be inspected per the reccmmendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975. O SAN ONOFRE-UNIT 3 3/4 4-36

m 1 \\ . ~, REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The heatup and cooldown limit curves for normal operation (Figures 3.4-2 and 3.4-4) and the cooldown limit curve for remote shutdown operation (Figure 3.4-6) 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 limit curves for Remote Shutdown operation are determined using the Total Loop Uncertainties (TLUs) for temperature and pressure for the Remote Shutdown Panel instruments in which the pressure TLUs are higher than those for the Control Room shutdown instruments. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted 1 reference temperature at the end of the service period, and they include adjustments for instrument uncertainties, and static and dynamic heads. The reactor vessel materials were tested prior to reactor startup to determine their initial RT,; the results of these tests and the updates resulting from the evaluation of material properties in response to Generic Letter 92-01, " Reactor Vessel Structural Integrity," Revision 1 are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the flue,.nce and copper and nickel content of the material in question, can be predicted using FSAR Table 5.2-6 l l and the recommendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials." The heatup limit curve (Figure 3.4-2) and the cooldown limit curves, Figures 3.4-4 and 3.4-6, include t at the end of the applicable predictedadjustmentsforthisshiftinRT*[nstrumentuncertainties,and service period, as well as adjustments for static and dynamic heads. The actual shift in RT of the vessel material will be establithed periodicallyduringoperaI[onbyremovingandevaluating,inaccordancewith ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation l surveillance specimens installed near the inside wall of the reactor vessel in the core area. The surveillance specimen withdrawal schedule is maintained in the FSAR. 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 vessel taking into account the location of the sample closer to the core than 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 capsule is different from the calculatId delta RT., for the equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 for reactor l criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. j SAN ON0FRE - UNIT 3 B 3/4 4-7 Amendment No. 71,110,111

  • ~.

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) L The maximum RT for all Reactor Coolant System pressure-retaining materials, with tie' exception of the reactor pressure vessel, has been L determined to be 90*F. The Lowest Service Temperature limit line shown on since Article NB-2332 l (Summer Addenda of 1972) of Section III of the ASME Fo3.4-2, 3.4-4 and 3.4-6 is based up Figures Code requires the Lowest Service Temperature to be RT., + 100*F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia. i The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements. The Low Temperature Overpressure Protection (LTOP) enable temperatures are based upon the recommendations of NUREG-0800 Branch Technical Position (BTP) RSB 5-2, Revision 1, "0verpressurization Protection of Pressurized Water Reactors While Operating at Low Temperatures." BTP RSB 5-2, Revision 1 defines the enable temperature as "the water temperature corresponding to a l + 90*F at the beltline location (1/4t or 1 metal temperature of at least RT* Appendix G limit calculations." 3/4t) that is controlling in the j SAN ON0FRE - UNIT 3 8 3/4 4-7a Amendment No.111 l

TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS Temperature of Minimum Upper A Drop Charpy V-Notch Shelf Cv Energy Weight 9 30 9 50 for Longitudinal i Piece No. Code No. Material Vessel location Results ft - lb - ft - lb Direction-ft lb g 215-01 C-6801-1 A533GRBCL1 Upper Shell Plate -20 28 64 115 215-01 C-6801-2 A533GRBCL1 Upper Shell Plate -20 -6 34 106 215-01 C-6801-3 A533GRBCL1 Upper Shell Plate -20 18 36 115 215-02 C-6802-4 A533GRBCLI Lower Shell Plate -30 40 70 118 215-02 C-6802-5 A533GRBCL1 Lower Shell Plate 0 40 70 116 215-02 C-6802-6 A533GRBCLI Lower Shell Plato -40 40 80 92 l 215-03 C-6802-1 A533GRBCL1 Intermediate Shell -20 80 100 94 215-03 C-6802-2 A533GRBCLI Intermediate Shell -20 40 70 115 g 215-03 C-6802-3 A533GRBCLI Intermediate Shell -10 60 80 105 { 203-02 C-6823 A508CL2 Vessel Flange Forging 0 -30 -15 NA 209-02 C-6824-1 A508Cl2 Closure Head Flange -40 -100 -100 NA Forging 205-02 C-6829-1 A508Cl2 Inlet Nozzle Forging 10 -35 -5 109 205-02 C-6829-2 A508CL2 Inlet Nozzle Forging 0 -55 -35 156 205-02 C-6829-3 A508CL2 Inlet Nozzle Forging 10 -25 35 112 205-02 C-6829-4 A508Cl2 Inlet Nozzle Forging 10 -30 25 108 205-06 C-6830-1 A508CL2 Outlet Nozzle Forging -10 -30 -15 125 205-06 C-6830-2 A508CL2 Outlet Nozzle Forging -10 -20 -5 131 E 232-01 C-6840-1 A533GRBCLI Bottom Head Torus -50 -10 0 107 [ 232-02 C-6841-1 A533GRBCL1 Bottom Head Dome -40 10 20 99 f+ .?? h I s m.}}

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