ML20036A115

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Proposed Tech Specs 3/4.4.8.1,3.4.8.3.1 & 3.4.8.3.2 Re RCS Pressure/Temp Limits
ML20036A115
Person / Time
Site: San Onofre Southern California Edison icon.png
Issue date: 04/30/1993
From:
SOUTHERN CALIFORNIA EDISON CO.
To:
Shared Package
ML20036A108 List:
References
NUDOCS 9305100063
Download: ML20036A115 (38)
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{{#Wiki_filter:i 1 ATTACHMENT A i EXISTING TECHNICAL SPECIFICATIONS AND BASES UNIT 2 l I i I 9305100063 930430 PDR ADOCK 05000361 P PDR

_ m,, - ~ i e l i t _~ INDEX LIMITING CONDITION FOR OPERATICH AND SURVEILLANCE R l PAGE f SECT *04 3/4 4-3 l HOT SHUTCCWN. 3/4 4-5 COLD SHUTDCVN - Loops Filled........... 3/4 4-6 I COLD SHUTOCWN - Locps Not Fi11ed........................ 3/4 g,7 l SAFETY VALVES - OPERATING 3/4.4.2 3/4 4-3 3/4.4.3 PRE 55URIZER............................................. 3/4 4-9 3/4.4.4 S T E AM G EN E RAT O R5........................................ 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4-16 LEAKAGE DETECTICH 5YSTEMS............................ 3/4 4-17 O P E RAT I O N A L L E A KAG E.................................. 3/44*io 3/4.4.6 CHEMISTRY............................................... 3/4 4-23 3/4.4.7 SPECIFIC ACTIVITY....................................... ) PRESSURE / TEMPERATURE LIMITS 3/4.4.8 s-- 3/4 4-27 R EACT O R COO LANT SY 5T EM............................... 3/4 4-31 PRE 55URIZER - HEATUP/C00LOCWN........................ OVERPRESSURE PROTECTION SYSTEMS 3/4 4-32 RCS TEMPERATURE < 312*F............................ 3/4 4-33 ACS TEMPERATURE F 312*F............................ 3/4 4-34 3/4.4.9 ST RUCTU RAL INT EGRITY.................................... 3/4 4-35 3/4.4.10 REACTO R COO LANT GAS VENT 5Y ST EM......................... i 3/4.5 OMERGENCY CORE COOLING SYSTEMS 1 1 3/4 5-1 3/4. 5.1 -5AF ETY INJECTION TANK 5.................................. 3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350*F.......................... 3/4 5-7 3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F.......................... 3/4 5-8 3/4.5.4 REFUELING WATER STORAGE TANK............................ } ..J AMEN 0 MENT NO. 7 y SAN ONOFRE-UNIT 2

1 i e .' s - i.( LIST OF I'3' ES-sacs ,.) _'ABLE 3/4 3-52 ACCICENT McNITORING INSTRLMENTAT!"N. 3.3 '.3 ACCICENT WCN!TCRING INSTRLMESTAT*CN SURVE*LLASCE 3/4 3-54 4.37 t!C.!"! MINTS............ 3/4 3 17 !!il Isiic7'CN ;NSTRUMENTS MINIMUM INST 3LMEN73 CPERA!Li 3.3-11 8.ACICACT*VE LIOU*O EFFLUENT MONITORING INSTR 3.3 '.2 CELiiEO RA0!CACTIVE LIQUID EFFLUENT McNITORING INSTRU ~ SURVEILLANCE REQUIREMENTS - CELETED 4.3-5 3/4 3-!! RACICACTIVE 2ASECUS EFFLUENT HONITCAING INS 3.3-;3 RADICACTIVE GASECUS EFFLUENT MONITORING INST 3/4 3 67 4.3 9 SVRVEILLANCE AEQUIAEMENTS................................ MINIMUM NUM3EA 0F STEAM GENERATORS TO IE INS 3/4 4-14 4.4 1 INSIAVICE INSPECTICN..................................... 3/4 4.15 ST!AM GENERATOR TUBE INS 7ECTICN........................ 4.4 2 3/4 4-19 REACTOR C00LANT SYSTEM PRESSURE ISOLATION VALVE 3.4 1 3/4 4-21 REACTCR CCCLANT SYST EM CHEMI STRY......................... 3.4-2 3/4 4-3Ca LOW TEMPERATURE ACS QVERF AES$URE PROTECTICN 3.4-3 AEACICR CCOLANT SYSTEM CHEMISTRY LIMITS SURVE 3/4 4 22 4.4-i A!QUIREMENTS............................................. FR! MARY COOLANT SPECIFIC ACTIVITY SAMPLE AND 3/4 4 25 4.4'4 PA0 GRAM...............................c.................. REACTOR VESSEL MATERIAL SURVEILLANCE 7A0 GRAM - 3/4 4 26 4.4 5 W I TH0 RAWAL S CH EDU LI.................................... 3/4 5 12 TENDCM SURVEILLANCE...................................... 4.5-1 3/4 6-12a TEM 60NLIFT-0FFF0RCE.................................... 3/4 6 20 4.6 2 CONTAINMENT !$0LATION V ALVES............................. 3/4 7-2 3.5-1 HAIM ST EAM S AF ETY VALY E S.............................. 3.7 1 xAXIMUM ALLewAgt! yAlyt LINEAR POWER LEVEL-HIGM TRIP W INCPERA8LE MAIN STEAM SAFETY VALVES OURING 3.7 2 3/4 7-3 VITH BOTM ST EAM G EN E RATO R $...................... AMENCHENT No. O XIX SAN CN0FRE UNIT 2

i INCEX 1 LIST OF FIGURES PAGE f FIGURES MINIMUM 20RIC ACIO STORAGE TANK VOLUME A 3/4 1-13 AS A FUNCTION OF STORED BORIC ACIO CCNCENTRA 3.1-1 3/4 1-24 CEA INSERTICN LIMIT5..................................... 3.1-2 3/4 2-7 DNER MARGIN OPEMTING LIMIT BASED ON COL 55.... 3.2-1 DNBR MARGIN OPERATING. LIMIT BASED ON COR 3/4 2-3 CALCULATORS (COL 55 0UT OF 5ERVICE)................. 3/4 3-40 3.2-2 OEGRADED BUS VOLTAGE TRIP 5ETTING.....................:.. j

3. 1 3/4 4-15a TUBE VALL THINNING ACCEPTANCE CRITERI A...........

4.4-1 OOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIF 3/4 4-26 3.4-1 ACT IV ITY LI MIT.......................................... RCS HEATUP PRESSURE / TEMPERATURE t. IMITATIONS 3/4 4-29 3.4-2 FOR 4-8 EFPY............................................. RCS C00LDOWN PRESSURE /TEMPERATUkE L 3/4.4-30 3.4-3 4-8 EFPY................................................. MINIMUM REQUIRED FEE 0 VATER INVENTORY FOR 3 3/4 7-6A TANX T-121 FOR MAXIMUM POWER ACHIEVED TO D ATE s_f

3. 7-1 5-2 EXCLUSION AREA...........................................

5.1-1 5-3 LCW P O P U LAT IO N 10 N E................................ 5.1-2 5-4 SITE SOUNDARY FOR GASECUS EF FLUENT 5................ 5.1-3 5-5 l SITE BOUNDARY FOR LIQUID EFFLUENT 5.................... 5.1-4 UNITS 2 & 3 FUEL MINIMUM BURNUP V5. INITIAL E 5-12 5.6-1 FOR REGION II RACK 5...................................... UNIT 1 FUEL

  • MINIMUM BURMUP V5. INITIAL ENRIC 5-13 5.6-2

_F_08 REGION II RACK 5...................................... 5-14 FUEL STORAGE P ATTERMS FOR REGION II RACK 5..... 5.6-3 FUEL STORAGE PATTERN 5 FOR REGION II RACKS 5-15 5.6-4 RECONSTITUTION ST ATI0N............................. 5-2 0 F F S IT E 0RGANIIATION.............................. _ 6. 2-1 6-3 UNIT ORGANIZATION........................................

6. 2-2 6-4a CONT RO L ROOM AREA...............................

6.2-3 AMENDMENT NO. 87 XXI SAN ONOFRE-UNIT 2

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~ REACTOR C00 CANT SYSTEM PRESSURE /TEsp!RATURE LIMIT 5_

EACicR COOLANT SYSTEM L
w!T!NG COND*'!CN 'OR CPE4 ATION The Reactor Coolant Systen (except the pressuri:er) temcerature a pressure shall be limited in accordance with the lim 3.4.8.1

[ and inservice leak and hydrostatic testing with: A maximum heatup of 10*F in any one hour period with RC a. temperature less than 112*F. A maximum heat-period with RC cold leg temperature less than 163*F.on up of SO'F in angF. greater than 163 l l A maximum cooldown of 10*F in any one hour period with RC cold A maximum cooldown of 30*F in any b. temperatures less than 103*F. A one hour period with RC cold leg temperatures greater than 145'F. A maximum temperature change of less than or equal to 10*F in any one hour period during inservice hydrostatic and leak testing 'T c. J operations above the heatup and cooldown limit curves. A minimum temperature of 85*F to tension reactor vessel head bolts d. APPLICABILITY: At all times. ACTICN: With any of the above limits exceeded, restore the, temperature and/or p to within the limit within 30 minutes; perform an eng of the Reactor Coolant Systes; determine that the R ih and pressure to less than 200*F in the next 6 hours and reduce the ACS T,yg and 500 plTa', respectively, within the following 30 hours. 1 SURVEILLANCE REQUIREMENTS The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during s 4.4.8.1.1 heatup, cooldown, and inservice leak and hydrostatic testing operatio } v AMEN 0 MENT NO. 70 3/4 4-27 SAN ONOFRE - UNIT 2 J l

REACicR CCCLA.NT SYSTEM gs PRESSURE /TEMPEUTURE LIMIT 5 REAcicR COOLANT SYSTEM SURvitLLANCE REOUIREMENTS (Continued) The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in m 4.4.3.1.2 The results of these examinations shall be used to uodate Recalculate the Adjusted Reference Temperature based Table 4.4-5. Figures 3.4-2 and 3.4-3. en the greater of the following: The mean value of shif t in reference temperature for plates -a. C-6404-3", or The predicted shift in reference temperature for w b. Emerittlement of Reactor Vessel Materials," May 1988. 1 "Ine most limiting satorial in the reactor vessel in ac Calculative proce-Materials," May 1968, has changed and are plates C-6404-3. dures provided in the new guide should be used to obtain th shift in RT of C-6404-3 plates. in reference temcerature as determined by impact testing on the exist NOT v'. C-6404-2 surveillance material. AMENCHENT N0. 70 SAN ONOFRE. UNIT 2 3/# 4-27a

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I/ i l ii i ) jggg lIiiil If l/: (i il 11 l nf J 1 l Fl i I/ I / IIiiili g C. Ilj I / 1,f. III I Il ili l 1Ii' l lt g 11 Ii I /) i 1/ 1: l / 11 1/ fit! Iit! I i! l i in 1 1 1 I i / ll/l till lill l jegg i 1 a I l 'I / ! l/lI i tiI( liii l z l 1 t i f fI l iII Iii I iI l ii I 1ili iIti l i I i ' il ' l l 'I'' I 'F 1 l ff fI! iil1 =00 l I ii!1 ~ 14M1And "_g, # ll' ll',l TDe = as* r 0 O 50 100 150 200 250 300 M0 MO INDICATED RCS TDdPERATURE ( F) Figure 3.4-2 vl RCS HEATUP PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFP hENDMENTNO.70 3 SAN ONOFRE - UNIT 2 3/4 4-29 i

~ _ _ _ ~ o- ~ ....g... 6*'i* (' ....g. .i.g cece:ws ."F (1 */HR (103*F) 25C0 W =202 (30 A ( 145*F) h (100 /HR} 145 F) I l 'l i111' I i 11li! 'l I /lill' Illt

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l Il if J j J 1 't ilIi iIi r iJ II lI11 'l1 i i i 1' ll f/ 'l 01 ', I i/ i; i l T' E 2000 illt 1 1! i l I ! Il' l I i ( 11 e b ? / I l 1 (I I d 1 i i f I l ll l l I l I I I ~ / m 'Ii l111 I i i l iI i / ll d 15C0 1 i !/) il i l' ff l i I ii i [, / 6 11 Ii i it I l It 3 Ilil _ l/11 iiil til' Iilf i I I ' 7'I' I I' 3 1000 l I Ii i ilIiliil I11ItI I J g. I f II Ii 1l illi )i i{ l' I i g I ) l l I l! l Il1 500 l ( l d" gcg,wp = - wr J 0 O 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F) Figure 3.4-3 RCS COOLDOWN PRESSURE / TEMPERATURE LIMITAT AMENCHENT No. 70 SAN ONOFRE - UNIT 2 3/4 4-30 ~........ _ _.

4 Table 3,4-3 t I l J. r-Lew Temperature-RCS Overtressure Protection Range l Cold Lee Temeerature,

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l Operation Peried, EFPY i During .0uring f' l Heatuo Cooldewn 'l 4 to 10 -< 312 c'287 1 i l M l T l h i i l 4 1 l 1 1 ? J l. I AMENOMENT NO. 70 SAN ONOFRE - UNIT 2 3/4 4-30s i ' -~ ' ! i. ~- ' - ~ i. f ..............._ m u. f -- ~!

I- ~^ ~ ~ ~ z. ~ ^ - - :..., 1 REACTCg CC0tJNT SYSTEM j OVERPRESSURE PROTECTICN SYSTEMS-1 312*F 3C5IEMPERATURE LIMITING CCH0! TION FOR OPERATICH At least one of :ne foliewing overpress'ure protection systems snall 3.4.3.3.1 te OPER.2SLE:The Shutdevn Cooling System Relief Valve (PSV9349) with: a. A lift setting of 406 t 10 psig*, and f 1) l Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and l 2) 2HV9378 open, or, [ The Reactor Coolant System depressurized with an RCS vent of greater l l than or equal to 5.6 square inches. l b. MODE 4 when the temperature of any one RCS cold leg is less

nan or equal to that specified in Table 3.4-3; MODE 5; MODE 6 with the reacto APPLICAB!t.ITY:

l l vessel head on. ACTION: to less than With the SOCS Relief Valve inoperable, reduce T,yg 200*F, depressurize and vent the RCS through a greater than or eq a. i to 5.6 square inch vent within the next 8 hours. With one or both SOCS Relief Valve isolation valves in a single I b. SOCS Relief Valve isolation valve pair (valve pair 2HV9337 and 2HV9339 or valve pair 2HV9377 and 2HV9378) closed, open the closed to less than 200*F, depres-valve (s) within 7 days or reduce T,yg l surize and vent the RCS through a gritater than or equal to 5.5 j inch vent within the next 8 hours, In the event either the SOCS Relief Valve or an RCS vent is us i mitigate an RCS pressure transient, a.5pecial Report shall be prepa c. l 5 and submitted to the Commission pursuant to Specifica l ating the transient, the effect of the SOCS Relief Valve or RCS vent within 30 days. on the transient and any corrective action necessary to prevent l l The provisions of Specification 3.0.4 are not applicable. reo rre - e. d. SURVEIU.ANCE REQUIREMENTS L The SOCS Relief Valve shall be demonstrated CPERABLE by: 4.4.8.3.1.1 Verifying at least once per 72 hours when the SOCS Relief Valve is j being used for overpressure protection that SOCS Relief Valve a. i isolation valves '2HV9337, 2HV9?39, 2HV9377 and 2Hv9378 are open. "For valve temperatures less than or equal to 130*F. AMENCHENT NO. 70 SAN ONOFRE - UNIT.2 3/4 4-32 -.e a w r

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E
C CR CCOLANT SYSTEM

) - _.OVERPRESSURE PROTECTICN sys EM5_ RCS TEMPERATURE > 312*F LIHiTING CONDITION FOR CPERATION 3.4.3.3.2 At least one of the following overpressure protection sys: ems snali te CPERA3LE: The Shutdown Cooling System Relief Valve (PSV9349) with: a. 1) A lift setting of 406 10 psig", and 2) Relief Valve isolation valves 2HV9337, 2HY9339, 2HV9377 and 2HV9378 open, or, b. A minimum of one pressurizer code safety valve with a lif t setting of 2500 psia + 1%"". APPLICABILITY: H00E 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 vent the a. RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours. 'g b. In the event the SOCS Relief Valve or an RCS vent is used to mitigate an ) RCS pressure transient, a Special Report shall be prepared and submitted to the Ccmmission pursuant to Specification 6.9.2 within 30 days. The ~ report shall describe the circumstances initiating the transient, tne effect of the SOCS Relief Valve code safety valve or RCS vent on the i transient and any corrective action necessary to prevent recurrence. SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SOCS Relief Valve shall be demonstrated OPERABLE by: Verifying at least once per 72 hours that the 50C5 Relief Valve a. isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are o;e.n 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 requirementrother than those required by Specification 4.0.5. 4.4.8.3.2.3 The RCS vent shall be verified to be cpen at least once per 12 hours 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 i the open position, then verify these valves open at least once per 31 days. l "For valve temperatures less than or equal to 130*F.

    • The lif t setting pressure shall correspond to ar.bient conditions of the valve

) U at nominal operating temperature and pressure. AggscsENT No. M SAN ONOFRE - UNIT 2 3/4 4-33

'u _,m a-l REacTCR CCCONT sys;Eg ige 5 PRE 55UR_*/ TEMPERATURE LIMITS (C:ntinued) The u rup and c:oidewn limit curves (Figures 3.4-2 and 3.4-3) are c:e;osite u ves.nien -e-e prearec by determining the mest conservative case, with eitner the insice or outsice wall controlling, for any neatuo rate of up to The heatup and coolde-n :urves 50*F/hr or cooldown rate of up to 100*F/hr. i l f the predicted aofusted-were prepared cased upon the most limit ng va ue oreference tempe adjust. tents for possible errors in the pressure and temperature sensing l i n,s truments. The reactor vessel materials have been tested to determine their i Reactor o; era-the results of these tests are shown in Table B 3/4.4-1. i tiggi;ndresultantfastneutron(EgreaterthanIMov)irradiationwillcause RT l I Therefore, an adjusted-reference. temperature, based a i an increase in the RT er and nickel content of the material in question, uponthefluenceand$No. l can te predicted using FSAR Table 5.2-5 and the recommendations of Regulatory l Guide 1.99, Revision 2 " Radiation Embrittlement of Reactor Vesset Materials."- The heatup and cooldown limit curves, Figures 3.4-2 and 3.4-3, include predicted at the and of the applicable service period, l adjustments for this shif t in RTas well as adjustments for possi$N errors instruments. of the vessel material will be established The actual shif t in RT by removing and evaluating, in accordance with - ) pariodically during operatiggTASTM E185-73 and 10 CFR 50 A veillance specimens installed near the inside wall of the reactor vessel in The surveillance specimen withdrawal schedule is shown in Since the neutron spectra at the irradiation samples and vessel the core area. Table 4.4-5. inside radius are essentially identical, the measured transition shift for a sample can te 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 heatup and cooldown curves the vessel wall by means of the Lead Factor. determined from the surveillance must be recalculated when the delta RT capsule is different frem the calculatggT the equivalent capsule delta RTNOT radiation exposure. The pressure-tamperature 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 Appendix G-to 10 CFR 50. for all reactor coolant system pressure-retaining The maximum Ri materials, with the Nception of the reactor pressure vessel, has been deter-g mined to be 90*F. The Lowest Service Temperature limit line shown on since Article NS-2332 (Sumer rigure 3.4 2 and 3.4-3 is based upon this RTAddenda of 1972) of Sec i 100*F for piping, pumps requires the Lowest Service Temperature to be RTand valves. Belo maximum of 20% of the system's hydrostatic test pressure of 3125 psia. The limitations imposed on the pressuriter heatup and cooldown rates and . spray water temperature differential are provided to ass l .L' performed in accordance with the ASME Code requirements. AMENOMENT No. 70 SAN ONOFRE-UNIT 2 8 3/4 4-7 ~~

TABLE B 3/4.4-1 REACTOR VESSEL TOUGilNESS Temperature of Minimum Upper Orop Charpy V-Notch Shelf Cv energy n$ Weight 0 30 0 50 for Longiturlinal E-Piece No. Code No. Material Vessel Location Results It - lb - ft - lb Direction-f t lb h 215-01 C-6403-1 A533GRBCL-1 Upper Shell Plate 40 15 35 130 215-01 C-6403-2 0 20 25 133 m 215-01 C-6403-3 -10 20 45 131 215-03 C-6404-1 Intermediate Shell Plate -30 10 50 145 215-03 C-6404-2 -20 20 50 155 215-03 C-6404-3 -20 10 50 131 215-02 C-6404-4 Lower Shell Plate -10 -5 25 124 215-02 C-6404-5 -20 10 25 134 215-02 C-6404-6 -10 -20 0 151 238-02 C-6401 A508Cl-2 Vessel Flange Forging -10 -70 -35 148 m R 209-02 C-6402. Closure llead Flange -10 -90 -40 142 Forging h 205-02 C-6410-1 Inlet Nozzle Forging 20 -40 -35 130 205-02 C-6410-2 0 -20 -5 135 205-02 C-6410-3 0 -15 -15 140 205-02 C-6410-4 0 -65 -50 140 Outlet Nozzle Forging -100 -30 -10 140 C-6411-l' N "*r 205-06 C6411-2 0 -35 -10 140 205-06 no / 232-01 C-6424 h A533GRBCL-1 Bottom llead Torus -50 -20 10 122 Q 232-02 C-6425 $. Bottom llead Dome -50 -30 -20 136 n-vs 205-03 C-6428pr A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 205-03 C-64252 llj /" -30 -70 -50 174 j 205-03 C-64/633 y /" -30 -70 -50 174 l 205-03 C-64,28.-4 G/" -30 -70 -50 174 s j b. 205-07 C-64E3-1 "'I Outlet Nozzle Ext. -30 -40 -25 229 [ Forging 205-07 C-6429-1 / -30 -40 -25 229 -~ r % i 231-02 C-6430-1 ~A533GRBCL-1 Closure llead Peels +10 20 55 118 Ei' 231-02 C-6431-1 -20 10 50 100 231-02 C-6432-1 -10 -15 45 115 231-02 C-6432 Closure llead Dome -10 -15 45 115

i l l I l s l ATTACHMENT B PROPOSED TECHNICAL SPECIFICATIONS AND BASES UNIT 2 l l 1 i l f i I l l l I. I i'

i INDEX LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE HOT SHUTD0WN........................................... 3/4 4-3 l COLD SHUTDOWN - Loops Fi l l ed........................... 3/4 4-5 l COLD SHUTDOWN - Loops Not Filled....................... 3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING.............................. 3/4 4-7 l 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 MB 238o F................ 3/4 4-32 RCS TEMPERATURE > Me 238oF................ 3/4 4-33 3/4.4.9 STRUCTURAL INTEGRITY................................... 3/4 4-34 i 3/4.4.10 REACTOR C0OLANT GAS VENT SYSTEM........................ 3/4 4-35 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS t 3/4.5.1 SAFETY INJECTION TANKS................................. 3/4 5-1 r 3/4.5.2 ECCS SUBSYSTEMS - T,,, a 350oF......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,,, < 350aF......................... 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK........................... 3/4 5-8 / SAN ON0FRE-UNIT 2 V AMENDMENT N0.

1 l INDEX LIST OF TABLES TABLE PAGE j 3.3-10 ACCIDENT MONITORING INSTRUMENTATION....................... '3/4 3-52 l 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................. 3/4 3-54 3.3-11 FIRE DETECTION INSTRUMENTS-MINIMUM INSTRUMENTS OPERABLE... 3/4 3-57 i 3.3 12 PADI0 ACTIVE LIQUID EFFLUENT MON!TORING INSTRUMENTATION DELETED "3S PADI0 ACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENTATION SURVEILLANCE REQUIREMENTS DELETED 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-65a 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION i SURVEILLANCE REQUIREMENTS................................. 3/4 3-67 ) 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING l INSERVICE INSPECTION...................................... 3/4 4-14 .{ 4.4-2 STEAM GENERATOR TUBE INSPECTION........................... 3/4 4-15 3.4 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 4-30ad l 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE [ REQUIREMENTS.............................................. 3/4 4-22 l i 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 i 4.6-2 TENDON LIFT-0FF F0RCE..................................... 3/4 6-12a 3.6-1 CONTAINMENT ISOLATION VALVES............................. 3/4 6-20 3.7-1 MAIN STEAM SAFETY VALVES.................................. 3/4 7-2. 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS..................................... 3/4 7-3 SAN ONOFRE-UNIT 2 XIX -AMENDMENT NO. 1

.u~ ~ INDEX j LIST OF FIGURES i FIGURE PAGE 3.1-1 MINIMUM BORIC ACID STORAGE TANK VOLUME AND TEMPERATURES AS A FUNCTION OF STORED BORIC ACID CONCENTRATION.......... 3/4 1-13 4 3.1-2 CEA INSERTION LIMITS VSJFRACTION[0F{ALLOWABLEsTHERMAL i P 0W E R..................~.~ ;. :...............'................ 3/4 1-24 3.2-1 DNBR MARGIN OPERATING LIMIT BASED ON C0LSS................ 3/4 2-7 t 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 TRI P SETTING......................... 3/4 3-40. 4.4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA.................... 3/4.4-15a j 1 l 3.4-1 DOSE EQUIVALENT Ia131 PRIMARY COOLANT SPECIFIC ACTIVITY l LIMIT VERSUS;PERCENTf0FTRATEDlTHERMALIPOWER':WITHRTHE PRIMARYfCOOLANTsSPECIFIC.!ACTIVITQ~.t.7.T................. 110JCi/GRAMDOSE EQUIV.ALENTi 1-131f........ 'T/..... ;T. 3/4 4-26 j 3.4-2 SONGS?2tHEATUP~RCS PRESSURE / TEMPERATURE LIMITATIONS I FOR~ F UNTIL[8'EFPY-NORMAli0PERATION.................... 3/4 4-29 3:4' 3T ' [NOTV'J MM- % 3.4-3-4 $0NGS?2iC00LDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS FOR ~"' ' UNTIL 8 "EFPYpNORMALEOPERATION..................... 3/4 4-30 3 A-51 ^ ! SONGS?2TRCS7 PRESSURE / TEMPERATURE; LIMITS l MAXIMUM; ALLOWABLE? COOLDOWN RATES?(UNTIts,8?EFPY)'-NORMALT0PERATION w.z.a g/4{5306 ~ ~ ~v u 314-62 ESONGS125000LDOWN!RCSTPRESSURE/TEMPERATURETLIMITATIONS "U,NTIlq81EFPY-REMOT.ELSHUIDOW OPERATION. g.;...,.;.;.y ^ 3/47.4230b 3.4?7f ESONGST2(RCSlPRESSURE/ TEMPERATURE? LIMITS). MAXIMUM! ALLOWABLE C00LDOWNtRATESyUNTILf8sEFPYMREMOTE!SHUTDOWNf0PERAT_ ION).;[3/414f30c 3.7-1 MINIMUM REQUIRED FEEDWATER INVENTORY FOR l TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE........... 3/4 7-6A 5.1-1 EXCLUSION AREA.......................................... 5-2 l 1 i 1 5.1-2 LOW POPULATION 20NE..................................... 5-3 5.1-3 SITE BOUNDARY FOR GASE0US EFFLUENTS..................... 5, 5.1-4 SITE B0UNDARY FOR LIQUID EFFLUENTS...................... 5-5 SAN ONOFRE-UNIT 2 XXI AMENDMENT NO. 4

INDEX i LIST 10F7 FIGURES' '~ FIGURE 1 <"".1

1.... "

' iPAGE +.,. ' c3 ? a:w;s c &., .< + mg:;,, 5.6-1 UNITS 2 AND 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT i FOR REGION II RACKS........................................ 5-12-' i t 5.6-2 UNIT 1 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION II RACKS....................................... 5-13 j 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 i 6.2-1 0FFSITE ORGANIZATION.................................... 6-2 l [ 6.2-2 UNIT ORGANIZATION........................................ 6-3 l l 6.3-3 CONTROL ROOM AREA........................................ 6-4a { i t L i' n l E i l t I l l l i l SAN!0N0FREf0 NIT [2T, m.1 " ' ",2 G. XIIE, "ifA. MEN.DMENT.T N0h ~. ~ m m.m, L

I 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'boltsTtensioned*,-T-the Reactor l Coolant Systeni(eicept the' pfessurizer) ~~ temperature and ' pressure shall be limited in accordance with the limit lines shown on Figuresj3.4-2; and Figur-e 3.1 3, 3.4-4,f:3:4-5;(3.4-6kand13;4-7. during heatup, cooldown, criticality, bc! tup, and inservice leak and' hydrostatic testing with: a. f, maximu-' heatup of 10 aF in any One hour period.ith RC cold leg temperature less than 112aF.

f. maximum Scatup of 30af in any ane hcur period t.ith RC cold leg temperature les; than 153aF.

A maximum heatup of 60aF in any one 14 hour period with RCS cold leg temperature equalitolor greater than M 3L86 oF. b. ^ maximum cecidown of 10af in any one hour pericd t'ith RC cold leg temperature; le;; than 103 F. A maximum cooldown of 30aF"as specified.byLFigure 3;4-5 in any one 1-hour period with RCS cold leg' temperature les' ~ than 446 oriequ'al : to?143'E*F. A maximum s cooldown of 100aF in any one 1-ho'ur period with RCS coldEleg temperature greater than 446?143LoF. l c. A maximum temperature change of les; than er equal to 100F in any eneL1Lhour period during inservice hydrostatic and leak testing l operations above the heatup and cooldown limit curves. d. A minimum temperature of 86aF to tension reactor vessel head bolts. With theireactor vesselthead. bolts detensionedF the1ReactoriCoolantl) System (except: the: pressurizer)(temperatureyshallibe': limited to.a: maximum heatuphor cooldown ofc 60aF inla'ny?l-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 within the next 6 hours and reduce the RCS T and pressure to less than200aFand500 psia,respectively,withinthefoYfowing30 hours.

  • f With the' re' actor?vesseUhead? bolts ~detensioned;TRCS coldfleg temperature lmay be 11 ess ' than~.86oF.'

SAN ONOFRE-UNIT 2 3/4 4-27 AMENDMENT NO. i

i t i i I REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM i l SURVEILLANCE REQUIREMENTS l l 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be l 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, 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 l Figures 3.4-2 and 3.4 373.444"through*3;427 Recalculate the Adjusted Reference Temperature b;;6d'ch the grdat-er of the fcllcWing:linTaccordance i l with' Regulatory Guide ~ ~l~ 99,) Revision 12;Y" Radiation Embrittlement of Reactor Vessel Materials," May 1988. a-The mean,value of shift in reference temperature. for plates e c,o n,, s -9,~i@ e v v v b. The predicted shift in reference temperature for.;cid ; cam: 2 203A or 3 203B :: determined by Regulatory Cuide 1.99, Revision 2, " Radiation Embrittlement of Reacter V03 cl Material;," May inDO. asuv IThe most liniting material in the reacter ves cl in accordance..ith the new ~ Regulatory Cuide 1.99, Revision 2, " Radiation Embrittlement of-Reaetor Vessel Material;," May 1988, has changed and are plate: C 5404 3. Calculative procedure; provided in the new guide should be used to cbtain the mean value; i cf shift in RTgy of C 5404 3 plate:. Calculations-ere b;;cd on the actual

hift in reference temperature ;; determined by impact testing en the existing plate C 5404 2 surveillance material j

i SAN ON0FRE-UNIT 2 3/4 4-27a ~ AMENDMENT NO.

3500

ii! i i

.ie ie! ! !,eiei! eie! eii4 ei. T LOWEST SERVICE

  • INSERVICE TESTS
  1. HEATUP TEMP = 209 F E

J t' 1 a ...,.6-... ( t- -.g... f 3000 - --+-4 f-. :! b --I4 i - - + ' I -+- 4- .g...;...h.........;..< ..h....e.. '. . o. ;...

  • Acceptable operating region - to the j '

...e..g..-q.. ( i j ; ' '~ ' right of the inservice tests curve l (Applicable in rnodes other than i-

.'j~

f. Modes 1 and 2) - l -t " t, "'t " + ' + '"+ +-~ ' r " + " + ~ e i 2500 . * +.. 4.. i. .4.. - i...; i ' + g

  1. Acceptable operating region - to the

.j.. right of the heatup curve in all modes. In addition, in Modes 1 and 2 operating "j ' l"., g j n.y region to the right of the core critical i"T"j t" "f--'"+"4"+ f"f+"+ t". i "j"- c curve. - l ~ + ;- +. -' +-+ q U - 4 o -y - - + + - - - - - - 4--- m 2000 (D .1.. ;..... ;.. !- 4 I e ~l i uJ i 4

4.. 4... ;.. 4...... j........ ;... ;...;......+.....;.. 4...;...+..;...... +. 4..

g i ct ..v.,..e.., ..,....,!...+. .. v..g. e ...y e -..,.,.m.., oc .. g.. i .2 QJ ...'.......a... '.;. 4 - N g -[.- .7 .L-2.! 1500 -1 + + ...;...t... o g...y. 9... j__'.4. 4..; 1 i i i ; ' ! i i . : i .j.. ' ' <n

i

.. q....;...;. 4 g ..........._..,.......4 uf ! - = i. -....y......4.... + ;.d.,.[# CORE CRITICAL ct

_; 4 4......

...;... t. i O 1000 ...... i.. 4... t... i.. .. 4.. 4....e.. 4.. 4.. 4.. ..;.. 3....... 4.. ;... f- .l... ..a.. <C o .......;...i.......i._ ......;...o 3

..a. ;.

O ..m.. ........ ~'.... _

: i i
: i i.. :

...;.. < 3 25 ... 4... ;.. 4....... ;.. 4.-. 4.. 4.. 4.. 4.. ...;.. 4...; 4......;.. 4 ~ -4.- L.i- +-i. i, 500 MINIMUM i i BO!. TUP ..t...i...........i........;.4... .;....... ;...;... i.. ...;_.;...t...;..; '........i..1'....i.. i.....:....;....i....;..., . '....:....;....:....:........I....;....;....;.... TEMP = B6*F l 1 8 , ' ' ' i......!

i

' i ' ' i ' 9 p......p....4 1.,, ,I,...Ii;;

i,
; ; i, ; ;, I',,,! ;

e 15 0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ( F)iTc-FIGURE 3.4-2 SONGS'2'HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS F0" UNTIL 8 EFPY ^ Normal Operation SAN ON0FRE-UNIT 2 3/4 4-29 AMENDMENT NO. t l i

b (Figure 3.4 Not Used) i L I I l I j SAN:ONOFRE-UNIT 2"

3/414p29a?

! AMENDMENT;N0;

L i b 5 i e l ? '

I! ;f ;

35M 'it ? ? '

;;lt ;;;l

+'l' LOWEST SERVICE COOLDOWN i --f TEMP = 209'F ~ i +

  1. i i i i i ! i

.4 1 ! t ! I --L i 4 -j-- 4-- ! E - l-- 4 4- ? + a - - - - -- - *- 1 --y-- f --b 14- '- i 3000 -- I i i I; 4_ l _. - ' ' i ' j...I...L.d...f....I } ' i .....f.. E.. .......;.4...a... i... ......+ 4_4.;. 4,+4... 4..~i. +..;. L, l ;.. 4.. ' 8 ~. 1

i

[ ! !. }. 3 ! i 3 ' - j I i : : t r i.. r s . 4. 9 9 7..q e#.

4. 9 9 7. 7 4.

4.~....e p-g _9-. s y_ ~_ i 4.- l i i, i l. i ! j,5 - > i l J - -e.. 4....... ;.....a.4.....4............1..... '+-'ef-+jf!i f-- i i- -- 4 --p g- - j + 4 -- ms 3 2500 - --.j-- --Y- - i ii i[ .........;.. 4.... 7 4 n. .....;._....g_........_.,.....;.......9.... 4..i._!. 4. 1....!_.L. 4...i... j 4+!. i....i..,.L L.J..i L td.4 ;.-..L..- .J... .. - J-uJ l 3 j j g 3 g ..._..o. 4. 4..+.... 7..j 7...... 4.J.. 4..

; -- 7..

.._..2 -)

. ? i I

-.,i i ,c,:. I...a u....u. 4' _1.. i....j.. ' _ _._.......w.. - ~ ~. l U) ...i..._. g + O i . I j i i ' j i g - r +! l r r i r + t t i 4 - - f -f A-t- -t t -l t - 1.13 2000 +- t i l l l t ; i ! j i. - i i ~ C i W ~ ~ c -; +..p.. 4.4...y...+_+...+... .rt.9.14 y.-' Unacceptable i! Operating 4 P r-H L?d-]4-4 4 1-4-+-+ "- +-- T t T y q._ ~~........... !... !... I. i........ j........- Acceptable i y . 4., _...i. 4_Lg... . Region t 1 a- ..i..........._...! 4-' -- w+ -j -t -+ 7- +- + - + -4 +--{ 3 +- je-j-- @mty --j ~~ 1500 P 4- -t +! ! Reg'M g .4. +' + t - gj g.- ). !}! [ .. 4. .-..,. i !...._.9_..i.....,g.4..,..,_,.....,. _. ~..,,,. p., 4.. 9.. 9 4. 4. .g .. q.._..;.. l. I { ....p.9. p..p...i _. p...!....,.. 9................p...... 7.....p.g ---..,.__.,.+_._....!..+g...I... o.... i g uJ ,_i, i i l t 1 l ^.

  • I u

E 1 l . i * .i.e,, i. r! r jj - O i t i !...!. 4.. 4...i 4.... _ s ! l. 4... i ' ' ..L ....,......z... 1000 .._l.. ? g _4.. 7 7 4_.p.p.+_.y 7..._.p _...g y - -4 . 4...i._j +. 4...i i.!...j. i.i. 4...!. L ;...i...y.e-.p...i... ...im Z } .e f f,,: m!. ....,. 1..i i 'i +.4

4. 4. 4.

4....., 3 ' i ! l .r....: 4 ii - i :

....,....+....i....i...

999 4 p +_.p.t p 4._..+! _.l i._ _..i;_. ' i ' i ! i i I y..g .y.. 7,, j-A. $o0 -+ 4 i-+-+ + +-t i i 1 i i i - 1 MINIMUM 4 '-4-+ + 4-4 l4Ml4.j...[i l l._.t _4

4._.;

t : 2 i 5 i i_ ._4_...4._..., BOLTUP !, ! : ' ' ~.... '... _............!. l..+!...! .(..!...,l.- TEMP = 86*F i ! .--3..........y..,. .p...w.!. I 4 q.i' l .!__I j -+ i +!.4 + +.k., I,..u. i 3 l t .+.4. .i i.i i .!.i .!.I i.i.i.i

i. -.i.; I

.t.i.i. t,,,, i,, ti; e i 1 15 0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ( F)LTcl_ ~ FIGURE 3,4-4 13:4 4 SONGS 22 COOLDOWNTRCS PRESSURE / TEMPERATURE i LIMITATIONS'F00 a 1UNTIL 8 EFPY ~ Norma 110peration" I SAN ONOFRE-UNIT 2 3/4 4-30 AMENDMENT NO. t s i l

I l l i 120 g g g 3 iii i g 3 110 100 90 80 70 i C I I h w til g ( W 3 e O t 9O x O o 20 6 to 0 B0 90 100 110 120 130 140 150 160 170 180 190 200 210 l INDICATED RCS TEMPERATURE ( F)-Tc NOTE:f: A MAXINUM:COOLDOWN RATE!0F:100?F/HR:IS? ALLOWED 'AT:ANY: TEMPERATURE lABOVE143?F' ~ FIGURE!314$5 SONGS"2?RCSTPRESSURE/ TEMPERATURE ^ LIMITS

MAXIMUM ALLOWABLE! COOLDOWN ; RATES :!("UNTIL 8 ' EFPY)

~ ' ' '~ Nofmal{gratjon" ~ ~ SAN ONOFRE-UNIT 2 '. g/4[4-30{ TAMENDMENTLN0s i

~ t [ 3500 ,,,,I,i + I,,,3 - I,,,-, - - i! I,;,i 1 - 1,,,,I,,,,l,iii! i LOWEST SERVICE COOLDOWN TEMP - 209 F -.,...,..i. 9..i. 7 6,:._.:.:...p...i.. 4,,.. 4...,:...,.,.,i..,e...;i...;.... I i i - + 4-4 +j-- t-- t--+} 4 --+l i-4 J-- +l -

  • rtgg-*-+-

1---- + - + - - + --r l.

I I l l 1 l l

} P I 3000 i I l

  1. ..R....l._..............1....

.+.4.........+....I. ! ! I ij ' I i ' i ' j 2 i..i ...+.i._+i + +-; +_4 4 4.., 1 -: r.4 4.+ _.. j': t i i I I g j

  • 7 i

f 1 .+..!.%.. 1..:i.e .%.......?...t *I' ....i....:....?1...4...

i
  • +

i + l. + !,},j I .i ' i i I. . e._i l t M ^ l g i ,. 4_._+_._.- l l 1 ! i i + l 14 - +l +l - 1 2500 i ( l-4 t > + ?- 3 i 1 I i 1 t .,_A. _4. ,_.1.,. 7., + __p g _-..+3.- +.; .1,.-._.,l_. w i i l ! !il I !I ' !l C .......i....:....;......;..r..........i......t..4'.......n4,..,......3...+...., 3 i, i i . i - .j_ +.__i ? +j j7 4 ;,.t ._.l M [ _.....! a_4 +.f.+.- M ! s : 8 i il w i g j i

s-t

.. e..........a. 4. T _4 4....i...+!..- 4...s.- ! I e....~. ' } j j l i L qUnacceptable - .,9,. a m 4.. 4_+_ 4. _.- 4 -+_ ..! 4 4.. 4 4..j. 4. 4 4. 4. 4.. 4..;.4...+...p.4. 4..Operating

: ii

' I: : i i i j i! T W i j j 'i-i--4-4 ;..j Region ~ ~ ~~~ + ~f~t"T"i i~'~ i ' ! 3 y . t_H 7 ;., I. , i

- i

'I .. 4...!...:...;. 4...!.. m i i i ...... 2.:.. 2...L.. 3 I m -... ;..._.; 4 ;.J 4

.r.j 4..J.. 4.j..j_j..j..{.,. ; Acceptable

...i..... 4-M ..:. !..+..!...!.4...:. !. 4...!. _1_t.. 4.. 4.1_....}. !...!...t...{.. 4. Operating 350o 14...T :..L i..i.1. !.a. J....! Region .,i _......,4 c .1 14. .j 1 1 1 - t ii e ! - iiiiii t ...j... .7..-...j..q... p.y. 4.. , g..q. 4.4. 4. 4...+. 7 4..._,.. 4.~ 4...l,..}. 4. p.. ) 5 1 8 3 W .....4.4...+....i.... 4. 4....I.4...,I. '...,...p g. 8 l ........... +....... p b i ( ..,I.9..,i'i.! i j :' i ! i t :. l1 'ijji i .iI 1 .- - * - - + - - - r i 7... T*~t - W-I ' '+1-' ? --- t ---+: --P - t + ' 4'--*'+--'-1 ...+.4.-- = +.". 4... _4... e...; _,. ~ O 1000 p g z _...p.. _..g...j..... 4.._ 4......,. 4....... 4..,.. a l.. g...;...:.... 4 - !l ...........i.4...,..,,,.4..4.4,4,..I...4,......,...,.. i-i . i l l . y 4. 4 n..,,4. ,1.v!.4. e v..e v.y. + +..., I

f..

I l ~ 4.g.. 2 ..y.g.... - i i { i I t t .i.... ..,.....,4...y..A.,.................%.4.....+4.4....., { j -. -...-+.'......&-.+.- ~ ^ - t lI*l 500 -F - .s 4 I f 4. +l. 4..- i i + .-...e._-+ - - =. i-i i t-+ N. -+: 4-t t- -- MINIMUM 't +++t-t+111!d11lj....... _..

v. l. 4..j.!4........

l ' BOLTUP -4 TEMP = 86*F l 4 .................. _....,.... +.. +..... jt..- i. A ..a. l..,l _,i_;i.4. 4... ......L. ......,2...#._.. ..a...g!,.-.;:._ j1,,i,;; i.,,ii!!i! i I i 1 i L 15 i 0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE (*F)-Tc FIGU,RE13s4;6 i I SONGS *2 C00LDOWN LRCS7 PRESSURE / TEMPERATURE i ^ LIMITATIONS (UNTILtBtEFPY' Rem. o.te. i.ShutdowL,0P_e.mrat. ion. l 1 SAN"0NOF.RE UNITL2i, ., n ... n3,/~4. 34:30b; .n t ~ n .... ? AMENDMENT N0; l L l.

i W g g g g g g g g g g g g I a a e 5 s 3 g 3 u g 110 100 90 C 80 k LL. 70 LU f- <C g z l 3 oa 30 J O b 40 30 m 10 0 80 90 100 110 120 130 140 150 160 170 180 190 200 210 INDICATED RCS TEMPERATURE ( F)-Tc NOTEF F NAXIMUN C00LDOWN;RATEi:0FJ100?F/HR(I5iALLOWED "AT ANY! TEMPERATURE; ABOVE 151sE 1 1 FIGURE-;.3 4q7, SONGST2:.RCS? PRESSURE / TEMPERATURE ~ LIMITS MAXIMUMj.'ALLOWABLECOOLDOWN! RATES ((UNTIL;8lEFPY) ~RemotejhutdownL0ppration~' ~ ~ " ~ SAN'ONOFRE-UNIT?24 f3/4 4s30c7 LAMENDMENT NO.

Table 3.4-3 Low Temperature RCS Overpressure Protection Range Operatino Period. EFPY Cold Leo Temperature, oF During During Heatup Cooldown 1 to 10:Unti1L8 (Normal: Operation)_ s 342f238 s 28K221 Until81(RemoteJShutdownT0peration)j ~ -*i: 1 221 1 l ) 1 J i i i Heatup;'.' operations?;areTnottnorniall' jerformsd from~ the' RemotelShutdown y panels,.~ " " ~ SAN ONOFRE-UNIT 2 3/4 4-30ad AMENDMENT NO. l

l REACTOR COOLANT SYSTEM l l OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s M2 238oF LIMITING CONDITION FOR OPERATION 3.4.8.3.1 At least one of the following overpressure protection systems shall be OPERABLE: l The Shutdown Cooling System Relief Valve (PSV 9349) with: a. 1) A lift setting of 406 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. ACTION: l a. With the SDCS Relief Valve inoperable, reduce T,,, to less than l 200oF, 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 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,,,ter than or equal to 5.6 inch vent within to less than 200aF, depressurize and i I vent the RCS through a grea the next 8 hours. 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 initiating the transient, the effect of the SDCS Relief Valve or RCS vent on the transient and any corrective action necessary to prevent l recurrence. l d. The provisions of Technical 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 hours when the SDCS Relief Valve is being used for overpressure protection that SDCS Relief Valve isolation valves 2HV9337, 2HV9339, 2HV9377 and 2HV9378 are open. ' Thej lift; setting pressidr~e? applicable?to Fee valve temperatures o;f less than or equd ^ to~130eF. ^ ~ ~ ~ ' " I SAN ONOFRE-UNIT 2 3/4 4-32 AMENDMENT NO. j 1

. REACTOR COOLANT SYSTEM l 4 OVERPRESSURE PROTECTION SYSTEMS j RCS TEMPERATURE >MG 238oF LIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems e shall be OPERABLE: j The Shutdown Cooling System Relief Valve (PSV 9349) with: a. 1) A lift setting of 406 1 10 psig*, and 2) Relief Val've 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 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 er an RCS vent is used to mitigate i 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 er RCS vent on the transient and any corrective action necessary to prevent recurrence. SURVEILLANCE 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 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 requirements other than those required by Specification.4.0.5. 1.1.8.3.2.3 The RCS vent shall be verified to be Open at least once-per-43-houes when the vent is being used for everprc::ure :rctcction, except when the vent pathway is provided with a valve which is lected, scaled, or otherwise ccured in the open position, then verify the c valve Open at least once per'31 days. ' The;1ifts settingip~ressure? app ^licable1to-Fee valve temperatures of less than or M 6 130 F.' 4

    • 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.

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The heatup and cooldown limit curves forLnormar(operation (Figures 3.4-2 and 3,4-3; 3.444) andithe7do'oldown}1imitEcurve forJremote4 shutdown operation (Fi g u~re?3 :4 -6)f a re"co~mp o si t e~c u rie s wh i ch %e rejrep a re d " by'd e t e rmi n i n g " t h e 'mo s t conservstiYe"c~ase, with either the inside or outside wall controlling, for any heatup rate of up to 60aF/hr or cooldown rate of up to 100oF/hr. TheMimit curves 'for? Remote ~ Shutdown?bperation[areTdetsnninedissinglthe :nTotalfto'op Uncertainties"(TLUs)2for temperature!andl pressure:forithe1 Remote Shutdown; Panel instrumentsLi_n whichltheTpressure? TLUsiare higherJthan LthoseTfor'the! Control' Room shutdown ; instruments t The^ heatup and co'oldown' curves were ~ prepared based upon the~most limiting 7alue of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for pc::ible errors in the pressure and temperature sensing instrument binstrument uncertainties, Land static landl dynamic: heads. The reactor vessel materials have been were tested prior toireactor~startup to determine their initial RT The results"6f'these tests and4the updates m. resulting from1the{evaluationtoftmaterialEpropertiesFinf responsef to Generic Letter 92-01, " Reactor Vessel Structural? Integrity,"JRevisionT areshosn in Table"B"3/4.4-1~ ^ Readto' ~op'eFatibn"and~ resultant' fast"neutr6n~(E greater than -1 r 1 MeV) irradiation will cause an increase in the RT Therefore, an adjusted m. reference temperature, based upon the fluence and copper 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." The heatup limitic6rve"(Figsrel3?4 2). and the~ cooldown limit curves, Figures 3.4-3l4 andL3s4s6," include'predisted adjustments for this shift in RT at the end of'th~e~applibable service period, as well as g adjustments for pc:mic crrors in the prc sure and tempercturc sensing i n s t rume n t s; i n st rume nt' lince Fta i nt i e s, Dan d it ati c>and ' djn ami'c i h ea d s. The actual shift in RT of the vessel material will be established m periodically during operation by removing and evaluating, in accordance with 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 shown in 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 l be applied with confidence to the adjacent section of the reactor vessel taking i 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 m l the calculated delta RTm.for the equivalent capsule radiation exposure. l l The pressure-temperature limit lines shown on Figure 3.4-2 and 3.1 3 for I 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. l SAN ONOFRE-UNIT 2 B 3/4 4-7 AMENDMENT NO.

REACTOR ' COOLANT?SYSTQi BASES' PRESSURE / TEMPERATURE LIMITS (Continued) The maximum RT for all reactor coolant system pressure-retaining m materials, with the exception of the reactor pressure vessel, has been determined to be 90af. The Lowest Service Temperature limit line shown on Figures;3.4-2, and-3.4-3 4, and; 3.4-6L~is based upon this RT, since Article NB-2332 (Summer ~ Addenda of ~ 1972) of^Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RT, + 1000F 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. based upon <the' recommendati_ons-off NUREG-08001 BranchsTech)? enable L te TheLL'owiTemperature~0Verp'~ essure 7rotection?(LTOP r nicaliPositibn (BTP)LRSB 5-2, RevisionL 1, "0verpressurization. Protectionsofj Pressurized WaterlReactors While~ Operating:at Low! Temperatures."- ;BTP RSB;5-2,!. Revision l11.definesithe enable temperaturef as "the' water.. temperature correspondihg itof a(metal? temperaturef of f at'~ least RT; ::+190*F:ats the; beltline' location :(1/4t:or 3/4t):;thattislcontrolling in the = Appen$i x lG ilimi tL: cal cul ati ons. " ' ~ ' i 1 SAN.0N0FRE-UNIT 21 iBE3/414;7aL AMENDMENT?N0J l l

TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS Temperature of Minimum Upper w E Drop Charpy V-Notch Shelf Cv energy Weight 0 30 0 50 for Longitudinal c, y, - Piece No. Code No. Material Vessel location Results ft - lb - ft - lb Direction-ft Ib 5 m 215-01 C-6403-1 A533GRBCL-1 Upper Shell Plate 40 15 35 130 r 215-01 C-6403-2 A533GRBCLs1 UpperEShell1 Plate 0 20 25 133 .E 215-01 C-6403-3 A533GRBCL-1 Upperj S{elljP. late -10 20 45 131 Z 215-03 C-6404-1 A533GRBCLf1 Intermed. Shell Plate -30 M 40 50 80' 446 119 m 215-03 C-6404-2 A533GRBCL21 IntermedRShe11l Plate -20 10 20 70 50 80 M5113 215-03 C-6404-3 A533GRBCLp1 Intermed;iShellglate -20 M 7.0 50 80 - 4M 99" 215-02 C-6404-4 A533GRBCLF1 Lower Shell Plate -10 40 26 80 424 104 215-02 C-6404-5 A533GRBCLs1 LowerlShellMPlate -20 M $50 26 70 444 118 215-02 C-6404-6 A533GR8CLf1 .LoseyShell[P1a.t;e -10 -20l50 0 50 4M 124 238-02 C-6401 A508C1-2 Vessel Flange Forging -10 -70 -35 148 m 209-02 C-6402 A508Cp2 Closure ~ Head Flange -10 -90 -40 142 Forging-205-02 C-6410-1 A508C192 Inlet Nozzle Forging 20 -40 -35 130 205-02 C-6410-2 A508C192 InletlNozzlej! Forging 0 -20 -5 135 Inlet!l Nozzle}Worging Inlet 1. Nozzle Forging 0 -15 -15 140 205-02 C-6410-3 A508C1-2 205-02 C-6410-4 A5.08C1f2 0 .-65 -50 140 205-06 C-6411-1 A508C192 Outlet Nozzle Forging -100 -30 -10 140 205-06 C-6411 -A508C192 OstletjNozzlegorging. 0 -35 -10 140 232-01 C-6424 A533GRBCL-1 Bottom Head Torus -50 -20 10 122 232-02 C-6425 A533GRBCLH Bottom Head Dome- -50 -30 -20 136 h 205-03 C-6428-1 A508CL-1 Inlet Nozzle Forging S/E -30 -70 -50 174 5 205-03 C-6428-2 A508Cl$1 Inlet 1 Nozzle 7ForgingJS/E -30 -70 -50 174 205-03 C-6428-3 A508CLW1 InletjNozzlesForging(S/E-30 -70 -50 174 2P-205-03 C-6428-4 A508CLh1 Inlett.NozzleMorging1/E-30 -70 -50 174 i s.- e-.. ,.-w, s%, m. -..,n> m.ii e.-.--

TABLE ^ B '-3/424-1MContinued) ~l Temperature of Minimum Upper . gg Drop Charpy V-Notch Shelf Cv energy gh Piece No. Code No. Weight 0 30 0 50 for Longitudinal'- 79 Material Vessel Location Results ft - lb - ft - lb Direction-ft Ib l f ~ -205-07 C-6429-1 A508CL-1 Outlet Nozzle Ext. Forging -30 -40 -25 229 205-07 C-6429 A5.08CL-1 OutleUNozzlelExt?iForging -30 -40 -25 229 l h;j] IE E-231-02 C-6430-1 A533GRBCL-1 Closure Head Peels 10 20 55 118 EM 231-02 C-6431-1 A533GRBCL;1 ClosurehHeadiPeels -20 10 50 100 $4s 231-02 C-6432-1 A533GRBCL-1 Closure! Head: Peels- -10 .-15 45 115 Sh$ 231-02 C-6432 A533GRBCL-1 Closure' Head'Do d -10 -15 45-115 fffI u !='s iu ~ [6 . gst na h:p F., pf.1y )[ [3 - m 18m sa

4 l ENCLOSURE 3 i 1 [ TECHNICAL SPECIFICATION PAGES CONTAINING THE CHANGES WHICH WERE PREVIOUSLY-REQUESTED IN AMENDMENT APPLICATION NO. 113 (PCN-358) DATED DECEMBER 20, 1991, AMENDMENT APPLICATION NO. 117 (PCN-354) DATED SEPTEMBER 3, 1992, AND l ARE BEING REQUESTED IN THIS LICENSE AMENDMENT APPLICATION NO. 118 (PCN-335) { SAN ON0FRE UNIT 2 l l-I i l l l

INDEX LIST OF TABLES ) l TABLE PAGE 3.3-10 ACCIDENT MONITORING INSTRUMENTATION....................... 3/4 3-52 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................. 3/4 3-54 i i 3.3-11 FIRE DETECTION INSTRUMENTS-MINIMUM INSTRUMENTS OPERABLE... 3/4 3-57 i 3.3 12 RADI0 ACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENTATION -I DELETED 1.3 8 R^DI0 ACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENT" TION SURVEILLANCE REQUIREMENTS DELETED l \\ 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-65a 4.3-9 RADI0 ACTIVE GASE0US 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 4-30ad g# fc 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE......... 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQUIREMENTS.............................................. 3/4 4-22 i 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM................................................... 3/4 4-25 i 1.1 S REACTOR VESSEL MATERIAL SURVEILLANCE PROCPAM WITMDr^"^t SCNEDULE....................................... 3/1 a 28 59I ! 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 l 3.7-1 MAIN STEAM SAFETY VALVES.................................. 3/4 7-2 l 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS..................................... 3/4 7-3 SAN ON0FRE-UNIT 2 XIX AMENDMENT NO.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS l 1 REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 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 N N' shall be removed and examined, to determine changes in material properties,-at the interval Las required by 10 CFR 50 Appendix H. in accordance..ith the (g:5'l schedule in Table 1.15 The results of these examinations shall be used to update Figures 3.4-2 and 3A-3[3.424?throughT3 4-7. Recalculate the Adjusted-Reference Temperature based on~the"peate Of"the"followimpiin(accordanceLwith Regulatory ? Guide" 1;99,f Revi sic'n 12p'RadiationTEmbrittlementTof L Reactor Nessel ' ' ' ~~ ~ ' ~ Materials,."i May}1988. ' The mean,value of shift in reference temperature for plate; W(' a.

c. c a n. e.

v, i l

s. m. m.
1. A, m s m... - o o n. e.n.,

L. T L. m -. m a :,4. m.a, L.. :. r. +. : . m s. mm m.. m +mmmm,+,,. i-m .. r... . mr.... I cr 3 2938 as determined by Regulat-ery Guide 1.99, Revision 2, " Radiation Embrittlem^nt of Reactor V0;;cl Materials," May 1988. i i i t i i I l l l 1 i l l l I IThe =00t limiting material in the reactor vc;;cl in accordance '.ith the-new fCM' Regulatory Cuide 1.99, Revision 2, " Radiation Embrittlement of Reacter Ve;;cl e u..s+ m 4..,1. ,a 4,.. sn.oo,

u.., -

L..., m m.a.,..a.,. - m,, t m,

e. e n. n. a.o.
c..,1.,.,.,1.., +.. m. m 37db

..y. r.. n e n A,i w a r r.u.. A m A. : .n. + L.. m.n.. m. i y..Am.

e. k.. m. 1. A.hm.

. c_m.A

4. m m.i:

.mk+,:.....n. + L.. a. m. m, n. .,, l. o. n e.

m. i.

n e. n e. m r...... shift in RTmn-ef C SdO1 3 plates. Calculation; cre based en the actual shift in m.<.. ---m. +mmmm.,+...m. a m +.-m. 4 m a n., : --, + +. m, +. :... y m.. + n m. m 4, +. : y yiu.m c. 1,4 .sim . my. u.. u. um. ... uj . siiy u .ri C M n. M o e i e.. m : 1 1.., n o.m m..,+6.e.:,1 m v, ..s ism .i mus 1 l l i SAN ON0FRE-UNIT 2 3/44-37a[285 AMENDMENT NO. ys'f

i l REACTOR COOLANT SYSTEM j ( OVERPRESSURE PROTECTION SYSTEMS fCAP RCS TEMPERATURE 5 312 238oF >N LIMITING CONDITION FOR OPERATION 3.4.8.3.1 No' more' than' tw'ohigh-pressure safetylinjec~ tion pumps shall be f/M' ! OPERABLE-and at least'one of the following'ov'erpressure protection systems shall 358 be'0PERABLE: 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 open1--err F'#' or, gS 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 the'enableftemperatures specified in Table 3.4-3; MODE 5; and cg.- MODE 6 with when the"+ ^# "^"^'~ head is on theEreactor~ vessel 7and the RCS :is I ~ not* vented. N. i ACTION: With the SDCS Relief Valve inoperable, reduce T, to less than a. l 200oF, 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 l p'ower-lockiopen7the70therLSDCS Relief 2ValvelisolationT51ve: pair

Igg, to"Isss than~200oF, N

within"4L-dass'24th'oursi "or reduce"Tl[a greater than or equal to 5. l depressurize and' vent ~the RCS througI l l inch vent within the next 8 hours. 1 c.

With 'more' than twoLhigh.-pressure 1 safety: injection? pumpsf 0PERABLE, secure the? third high-pressure safetyjinjection' pumpiby racking lout fM j

l itsimotor circuitibreaker or3 locking closelitsidischarge valve within #gg' 8:hourst

  • The :^11f[130aF:settingjpressuref applicableito for valve temperatures of less than o f

equal to NI SAN ONOFRE-UNIT 2 3/4 4-32 AMENDMENT NO. 1 i

\\ REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The heatup and cooldown limit curves forinormall operation (Figures 3.4-2 gl and h4-3 3.4-4) and the:cooldown rlimit 'curkelfor: remote::shutdownfoperation 336 (Figure l3.4-6)'are Yompos~ite~curvss"whichTeFe"pfepared' b ~ determining ~ the"most conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 60eF/hr or cooldown rate of up to 100oF/hr. Theslimit l curves for LRemott!!Shutdownloperation'areVdetermined 'usingj the? Total: Loop ~ p Uncertainties >:-(TLUs). for temperature. and. pressure;for[theoRemote' Shutdown: Panel l instruments in which the pressureJTLUs;areihigherithan those for;the Control 1 Room 9 l shutdowntinstruments!, ~ The"heattlp and" cob 1down'cdeveslere prbpared base ~d*upon the most"limitinglalue of the predicted adjusted reference temperature at the end of the service period, and they include adjustments for pcccible error; in l the pressure and temperature sensing instrument; instrument? uncertainties Jand i static;and dynamic; heads; The reactor vessel materials have been wers tested priorttofreactbr~startup to determine their initial RT The results~of"these tests"andJtheDupdatei JN ug7 ' esultingi fromlthe7 evaluation of f materiaFpropsrties jinjesponse;toL Generic ~ r Letter' 92-01, "ReactorJVsssel StructuralfIntegrity,"1Revisioni*l"a're shown"in ~ 1 Table B'3/4:4-1; Reactor ~ operation'snd Fe'soltant fast ^nehtron (E greater than 4 1 MeV) irradiation will cause an increase in the RT Therefore, an adjusted ua7 reference temperature, based upon the fluence and copper 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." The heatup limitTcurve7(Figure 13.4-2) and the g9 cooldown limit curves, Figures 3.4-314;:a'ndf3.4-6, include ~predi~cted adjustments i for this shift. in RT at the end of the" applicable service period, as well as l adjustments for pos;ya7;ble errors in the pressure and temperature censing gf l in:trument; instrument' uncertainties,fand{staticiandfdynamicLheads. 33 ~ The actual shift in RT of the ' vessel material will be established uo7 periodically during operation by removing and evaluating, in accordance with 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 shc=n in Tabh 4A-5 maintained initheiFSAR. Since the neutron spectra at the irradiation M samples'and vessel inside~~ radius are essentially identical, the measured transition shift for a sample can be applied with confidence fT 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 ug7 surveillance capsule is different from the calculated delta RT,37 for the O9f l' equivalent capsule radiation exposure. The pressure-temperature limit lines shown on Figure 3.4-2 and 3.13 for N l reactor criticality and for inservice leak and hydrostatic testing have been I provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50. l SAN ON0FRE-UNIT 2 B 3/4 4-7 AMENDMENT NO. l}}

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