Text

Proposed Tech Specs 3/4.4.8.1, Pressure/Temp Limits, 3.4.8.3.1, Overpressure Protection Sys - RCS Temp 302 F & 3.4.8.3.2, Overpressure Protection Sys..
	ML20073T096
Person / Time
Site:	San Onofre
Issue date:	07/06/1994
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
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ML20073T077	List: ML20073T072; ML20073T085; ML20073T096;
References
	NUDOCS 9407080159
	Download: ML20073T096 (39)
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ATTACHMENT A EXISTING TECHNICAL SPECIFICATimNS AND BASES UNIT 3 e

f 9407090159 940706 PDR P

ADOCK 05000362 PDR

l.

INDEX LIMITING CONDITION FOR ODERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE HOT SHUT 00WN ......................................... . 3/4 4 3 COLD SHUT 00WN - LOOPS FILLE 0......................... . 3/4 4-5 CO LD SHUT 00WN - LOOPS NOT F I LLE0. . . . . . . . . . . . . . . . . . . . . . . 3/4 4-6 3/4.4.2 SAFETY VALVES - OPERATING............ .................. .

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 LE AKAGE DETECTION SYSTEMS. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-17 O P E RAT I ONA L L E A KAG E . . . . . . . . . . . . . . . . . '. . . . . . . . . . . . . . . . 3/4 4-18>

n 3/4.4.6 CHEMISTRY............................................... 3/44-21) 3/4.4.7 %

SPECIFIC ACTIVITY....................................... 3/4 4-24 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM............................... 3/4'4-28 PRESSURIZER - HEATUP/C00LDOWN........................ 3/4 4-32 OVERPRESSURE PROTECTION SYSTEMS - - -

302*F.......................

RCS TEMPERATURE-< .... 3/4 4-33 RCS TEMPERATURE I 302*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 St ~. .

TANKS................................... 3/4 5-1 3/4.5.2-avg- T 350*F..........................

"$,CCS-SUBSYSTEMS i - 3/4 5-3 3/4.5.3- ECCS SUBSYSTEMS - T,yg 3 < 50*F........................ . 3/4 5-7 3/4.5.4- REFUELING WATER STORAGE TANK............................ 3/4 5-8 SAN ONOFRE - UNIT 3 V AMEN 0 MENT NO. 71

INDEX LIST OF FIGURES FIGURE PAGE 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 3.2-1 DNBR MARGIN OPERATING LIMIT BASED 0N COLSS.............. 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....................... 3/41 M40 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 i PRIMARY COOLANT SPECIFIC ACTIVITY >1.0,pci/ GRAM DOSE-EQUIVALENT I-131...............'........................ 3/4 4-27 3.4-2 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS FOR 0-5 YEARS............................................... 3/4 4-30 3.4-3 C00LDOWN RCS PRESSURE /fEMPERATURE LIMITATIONS FOR YEARS...............................................

0-5 3/4 4-30a 3.4-4 RCS PRESSURE / TEMPERATURE LIMITATIONS FOR 4-8 EFPY....... 3/4 4 3.4-5 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWA8LE C00LDOWN RATES (4-8 EFPY)............................... 3/4 4-31a 4 3.7-1 MINIMUM REQUIRED FEEDWATER INVENTORY FOR TANK T-121 FOR MAXIMUM POWER ACHIEVED TO DATE.......................... 3/4 7-7

~

5.1-1 EXCL9SION AREA................~......................... 5-2 Ayl g; 5.1-2 LOW POPULATION 20NE..................................... 5-3 s

5.1-3 SITE S0UEARY FOR GASEOUS EFFLUENTS..................... 5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS...................... 5-5 5.6-1 UNITS 2 AND 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICMENT 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-h FUEL STORAGE PATTERNS FOR REGION II RACKS............... 5-14 SAN ONOFRE - UNIT 3 XVII AMEN 0 MENT NO. 77

s. .

4 INDEX LIST OF FIGURES 1

FIGURE PAGE 5.6-4 FUEL STORAGE PATTERNS FOR REGION II RACKS RECONSTITUTION STATION..................................

~

5-15 6.2-1 0FFSITE ORGANIZATION.................................... 6-3 6.2-2 UNIT ORGANIZATION....................................... 6 6.2-3 CONTROL ROOM AREA....................................... 6-6 8,

4 5

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SAN.ONOFRE - UNIT XVIIa AMENDMENT NO. 77

_)

!NDEX LIST OF TABLES TABLE PAGE 4.3-7 ACCIDENT McNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS............................................... 3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS................................. 3/4 3-58 3.3-12 RADICACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION --

DELETED 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS - DELETED 3.3-13 RADICACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION.... 3/4 3-66 4.3-9 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SU RV E I L LANC E R EQUI REME NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 3-58 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION....................................... 3/44-14j 4.4-2 STEAM GENERATOR TURE INSPECTION............................ 3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES. . . . . . . . . . . 3/4 4-20 3.4-2 REACTOR COO LANT SYSTEM CHEMISTRY. . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 4-22 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REQU1'EMENTS............................................... 3/4 4-23 4.4 4 PRIMARY C0OLANT SPECIFIC ACTIVITY SAMPLE................... 3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRA - WITH0RAWAL SCHEDULE...................................M................ 3/4 4-29 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE.... . . .... 3/4 4-31b 4.6-1 TENDON SURVEILLANCE........................................ 3/4 6-12 4.6-2 TENDbN LI FT-0 F F F0 RC E . . . . . . . . . . . . . . . . . . . .3/4 . . 6-13 3.6-1 CONTAIMMENT' 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-HIGM TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS................................. 3/4 7-3 SAN ONOFRE - UNIT 3 XIX AMENDMENT NO.81

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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 ar.d pressure shall be limited in accordance with the limit lines shown on rigores 3.4-2, 3.4-3, 3.4-4, and 3.4-5 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup as specified by Figure 3.4-3 in any 1-hour period with RCS cold leg temperature less than 153*F. A maximum hoatup of 60 F ,in any 1-hour period with RCS cold leg temperature greater than 153*F.

b. A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period with RCS cold leg temperature less than 126'F. A maximum cooldown of 100 F in any 1-hour period with RCS cold leg temperature greater than 12S'F.

c. A maximum temperature change of 10*F in any 1-h'our 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 vessci nend 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 ef fects 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T a and pressure to less than 200Fand500 psia,respectively,withintheY811owing30 hours.

^With the reactor vessel head bolts detensioned, RCS cold leg temperature may be less than 86*F.

SAN ONOFRE - UNIT 3 3/4 4-28 AMENDMENT N0.71

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REAC'OR COOLANT SYSTEM 3/4.4.3 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION SURVEILLANCE REOUIREMENTS 4.4.3.1.1 The React r Coolant System temperature and pressure shall be-determined to be witi..in the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operationn, 4.4.8.1.2 The reactor vessel material irradiation surveillance specinens snall 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-i in Tacle 4.4 5. The results of these examinations shall be used to update Figures 3.4-2 and 3.4-3. Decalculate the Adjusted Reference Temperature based.

on the greater of tne following: Y

a. The actual shift in reference temperature for plate C-6802-1 as aetermined by impact-testing, or

b. .The predicted shift in reference temperature for weld seams 2-203A, 2-2038, or 2-203C.as determined by Regulatory Guide 1.99, Revision 2," Radiation Embrittlemer,t of Reactor vessel Materials," .

May 1938.

5 SAN ONOFRE - UNIT 3 3/4 4-28a - AMEN 0 MENT NO. 71

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FIGURE 3.4-2 SONGS 3 RCS PRESSURE / TEMPERATURE LIMITATION FOR 4-8 EFPY l

SAN ONOFRE - UNIT 3 3/4 4-30 AMEN 0 MENT NO, 71

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SAN ONOFRE - UNIT 3 3/4 4-30a AMENOMENT NO. 71 i

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' FIGURE 3.4-4 SONGS 3 RCS PRESSURE / TEMPERATURE q

LIMITATIONS FOR 4-8 EFPY i SAN ONOFRE - UNIT 3 3/4 4-31 AMENDMENT NO. M i

110

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100 -

90 -

80 -

m 70 -

D # ~

50 -

I40 M

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20 -

10 -

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80 ip0 100 110 120 130 140 150 IN06CATED RCS TDdPERAM @

NOTE: A WAXBduW COOLDOWN RATE OF 100'F/HR 15 Atl4WED AT ANY MPERATURE ABOVE 128'F FIGURE 3.4-5 SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (4-8 EFPY)

SAN ONOFRE - UNIT 3 3/4 4-31a AMEN 0 MENT NO. 71 3

1

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Table 3.4-3 Low Temoerature RCS Overoressure Protection Range Ooeratina Period, EFPY C,cid Leg Temeerature, 'F Juring During

,eatuo W Cooldown g

4 to 8 1 302 1 267

)

P SAN ONOFRE - UNIT 3 3/4 4-31b AMEN 0 MENT NO. 71 !

i l

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i cEACTOR COOLANT SYSTEM OVEQPRjjjURE PROTECTION SYSTEMS pts TEMPERATURE < 302'F ,

LIMITIM CONDITION FOR OPERATu _

3.4.P.3.1 No more than two high-pressure safety injection pumps shall-be OPERA 8LE and at least one of the following everpressure protection systems stall 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 3:W9337, 3HV9339, 3HV9377, and

3HV9378 open i or, g 1

i b. The F n'.y Coolant System depressurized with an RCS vent of greater than ne m ?1 to 5.6 square inches.

APPLICABILITY: N0E 4 when the temperature of any one RCS cold lag is less l than or equal to the enable temperatures specified in Table 3.4-3: MODE 5;-and j MODE 6 when the head is on the reactor vessel and the RCS is not vented.

F ACTION:

a. to less than With 200'F,the SDCS Relief depressurize andValve vent inoperable, reduce the RCS through a 9T,7 eater than or equal l

to 5.6 square inch vent within the next 0 hours0 days 0 hours 0 weeks 0 months ,

b. With one nr both SDCS Relief Valve isolation valves in a single SDCS Relief Valve isolation valve pair (valve pair 3HV9337 and 3HV9339 or 4 valve pair 3HV9377 and 3HV9378) closed, open the closed valve (s) or

. power-lock open the other SDCS Relief Valve isolation valve pair o

to less than 200F, depressurize and l within 24 hours, reduce T ,ter than or equal to 5.6 inch' vent within vent the RCS-through a grea the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l c. With more than two high-pressure safety injection pumps OPERABLE,-

l secure the third high-pressure safety injection pump by racking out its motor curcuit breaker or locking close its discharge valve

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

l

The lift setting 1.ressure applicable to valve temperatures of less than or equal-to 130'F.

AMEN 0 MENT NO. +b 9; SAN ONOFRE

- UNIT 3 3/4 4-33 l-

_, _ g

iiE4CTCR COCLANT SYSTEM OVERPRES5URE PROTECTION SYSTEMS I

RC5 TEMPERATURE > 302*F i ITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of the following overpressure protection systems shall y be OPERABLE:

a. The Shutdown Cooling System Relief Valve (PSV9349) with:

1) A lift setting of 406 2 10 psig*, and i

2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and t 3HV9378 open, or,

b. A minimum of one pressurizer code safety valve with a lift setting of 2500 psia : 1%".

APptl: ABILITY: M00E'4 with RCS temperature above that specified in Table 3.4 3. I ACT104:

a. With no safety or relief valve OPERABLE, be in COLD SHUT 00'aN and vent the RCS through a greater than or equal to 5.6 square inch vent within the nexc 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

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 th9 Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating-the transient, the effect of the SDCS

' Relief Valve code safety valve or RCS vent on the trantient and any corrective action necessary to prevent recurrence.

SURVEILLANCE REQUIREMENTS 4.4.8.3.2.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a. Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that tr.e SDCS Relief Valve isolation valves 3HV9337, 3HV9339, 3HV9377 a v 3HV9378 are open when the 50C5 Relief Valve is being used for overpi-ssure protection,

b. Verifying relief valve setpoint at least once per 30 months whea tested pursuant to 5pecification 4.0.5.

4.4.8.3.2.2 Thr: pressurizer code safety valve has no additional surveillance recuirements other than those required by Specification 4.0.5.

4.4.8.3.2.3 The RCS vent shall be verified to be open at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 secureo in the open position, then verify these valves open at least once per 31 days.

The lif t setting pressure sopMcable to valve tempe'ratteres of lest than or equal to 130*F.

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

SAN ONOFRE - UNIT 3 3/4 4-35 AMEN 0 MENT NO.71

_ _____ - - _ - - - - - - )

REACTOR COOLANT SYSTEM

, BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves (Figures 3.4-2 and 3.4-3) are ccmpesite curves hich were prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatuo rate of up to 60'F/hr or cooldown rate of up to 100*F/hr. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusteo reference temperature at the end of the service period indicated on Figures 3.4-2 snd 3.4-3.

The reactor vessel materials have been tested to determine their initial RT y ; the results of these tests are shown in Table B 3/4.4*1. Reactor opeNtionandresultantfastneutron(Egreaterthan1MeV)irradiationwill cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluence and h p.er and nickel content of the meterial in l cuestion, 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 and cooldown limit curves, figures 3.4-2 and 3.4 3, include predicted adjustments for this shift in RT at the end of the apolic-ableserviceperiod,aswellasadjustmentsforpobIbleerrorsinthepressure and temperature sensing instr ~ ats.

The actual shift in RT of the vessel material will be established periodically during operatikby removing and evaluating,* in accordance with ASTM E185-73 and 10 CFR Appendix H, reactor vessel material irradinion sur-veillance 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 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 o' the Lead Factor. The heatup : 9d cooldown curves ,

must ce recalculated when the delta RT determined from the surveillance capsule is different from the calculathdelta RTNDT for the equivalent capsule radiaticn exposure, i The pressure-temperature limit lines shown on Figures 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. '

The maximum RT g for all Reactor Coolant System pressure-retaining materials, with the Nception of the reactor pressure vessel, has been deter

mined to be 90'F. The Lowest Service Temperature limit line shown on 1

Figures 3,4-2 and 3.4-3 is based upon this RT since Article NB-2332 (Summer Addendaof1972)ofSection111oftheASMEBdIerandPressurevesselCode recuires the Lowest Service Temperature to be RT 100'F for piping, pumps andvalves.Sainwthistemperature,thesystempMs+uremustbelimitedtoa maximum of 20% of the system's hydrostatic test press.ure of-3125 psia, The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature different ki a.2 provided to assure that the pres-surizer is operated within the design criteria assumed for the fatigue analysis performed in eccordance with the ASME Code reauirements.

SAN ONOFRE-UNIT 3 8 3/4 4-7 AMENOMENT NO. 71

. . . . _ _ . . _ ._. ._..m . . _ . _ _ _ _ _ _ _ _ _ _ .___ __ .

r y TABLI B 3/4.4-1 z .

ti REACTOR VESSEL TOUGifNESS

?,

Temperature of Minimum Upper

??

E Orop Charpy V-Notch Shelf Cv energy 25 '*eight

. 0 30 0 50 for Longitudinal Piece No. Code No. Material Vessel location Resalts f t - Ib - f t - Ib Directinn-ft Ib w

215-01 C-6801-1 A533GRBCL1 Upper Shell Plate -20 28 64 115 !

215-01 C-6801-2 A533GRBCLI Upper Shell Plate -20 -6 34 106

-215-01 C-6801-3 A53?GRBCL1 Upper Shell Plate -20 18 36 115 215-02 C-6802-4 A533GRBCLI tower Shell Plate -30 32 62 115 l 215-02 C-6802-5 A533GRBCL1 Lower Shell Plate 0 36 64 110 ;

215-02 , C-6802-6 A533GRBCLI Lower Shell Plate -40 32 100 90 i t

.215-03 C-6802-1 A533GRBCL1 Intermediate Shell -20 56 100 95 215-03 C-6802-2 A533GRBCL1 Intermediate Shell -20 40 66 113 ,

215-03 C-6802-3 A533GRBCL1 Intermediate Shell -10 44 80 101

? 203-02 C-6823 A508CL2 Vessel Flange Forging 0 -30 -15 NA e

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 Farging 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 Cutlet Nozzle Forging -10 -30 -15 125

__ 205-06 C-6830-2 A508CL2 Outlet Nozzle Forging -10 -20 -5 131 G3 C-6840-1 A533GRBCt1 Bottom llead Torss -50 -10 0 107

][ 232-01 U'

232-02 C-6841-1 A533GRBCil Bottom flead Dome -40 10 20 99 G

E3

1

)

i i

j i

1 4

I i

l i

n 4

ATTACHMENT B s

PROPOSED TECHNICAL SPECIFICAT10NS AND BASES a

UNIT 3 4

I l

t 1

i i

l j

i

INDEX LIMITING CONDITION FOR OPERATION AND SURVElltANCE REQUIREMENTS SECTION PAGE a

HO T S H U T D0WN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/44-3 COLD SHUTDOWN - LOOPS FILLED.......................... 3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED...................... 3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING............................. 3/4 4-7 3/4.4.3 PRESSURIZER............................ .............. 3/4 4-8 3/4.4.4 STEAM GENERATORS...................................... 3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DET ECTION 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 3/4.4.8 PRESSURE / TEMPERATURE LIMIT 5 REACTOR COOLANT SYSTEM........................... 3/4 4-28 PRESSURIZER - HEATUP/C00LDOWN.................... 3/4 4-32 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 303 246*F................ 3/4 4-33 {pgI9%

RCS TEMPERATURE > 302 246*F................ 3/4 4-35 3/4.4.9 STRUCTURAL INTEGRITY.................................. 3/4 4-36 s 3/4.4.10 REACTOR COOLANT GAS VENT SiSTEM....................... 3/4 4-3/

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,y e 350af......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,y < 350af......................... 3/4 5-7 3/4.5.4 REFUELING WAT ER STORAGE TANK. . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 5-8 SAN ON0FRE-UNIT 3 V AMENDMENT NO.

_INDEX LIST OF FIGURES FIGURE PAGE 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 t=

3.2-2 DNBR MARGIN OPERATING LIMli 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 CRITER1A........................ 3/4.4-16 3.4-1 DOSE EQUIVALENT l-131 PRIMARY C00LAh! SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER UlTH THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 pCi/ GRAM DOSE EQUIVALENT l-131.............................................. 3/4 4-27 3.?-2 SONGS-~3 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS FOR I 0'S^ YEARS;UNTIL 20 EFPY-NORMAL OPERAT10N......................

3/4 4-30 lSuf 3.4-3 GOOLDOWN RCS PRESSURE / TEMPERATURE LIM MANONS-FOR 0-6-VEA RS D E L ET E D . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 1 30a 3.4-4 SONGS 3' C00LDOWN RCS PRESSURE / TEMPERATURE LIMIT *TIOM l FOR ^ 8 UNTIL 20 EFPY-NORMAL ~0PERATION.......................

3/44-31%uf 3.4-5 SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE !

C00LDOWN RATES (--4-8.UNTIL 20 EFPY)-NORMAL'0PERAT10N.........

3/4 4-31al5 7 3.4-6 - SONGS '3 C00LDOWN' RCS: PRESSURE / TEMPERATURE: LIMITATIONS l k

UNTILJ20EFPY-REMOTE-SHOTDOWN0PERATION.......................-3/4 3.4-7' ' SONGS 3 RCS PRESSURE / TEMPERATURE ^ LIMITS MAXIMUM" ALLOWABLE. !

COOLDOWN RATES L(UNTIL: 20. EFPY)-REMOTE SHUTDOWN OPERATION.'. . . . . 3/4 : 4-31cj $qS 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 BOUNDARY FOR GASE0US EFFLUENTS........................... 5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS............................ 5-5

-SAN ON0FRE-UNIT 3 XVII AMENDMENT NO.

_ _ _ _ _ _ _ _ _i

INDEX

! LIST OF FIGURES i

flGURE PAGE 5.6-1 UNITS 2 AND 3 FUEL MINIMUM BURNUP VS. INITIAL ENRICHMENT FOR REGION 11 RACKS............................ 5-12 l 1

4 5.6-2 UNIT 1 FUEL MINIMUM BURNUP VS. INITIAL 4 ENRICHMENT FOR REGION 11 RACKS............................ 5-13 I

,' 5.6-3 FUEL STORAGE PATTERNS FOR REGION 11 RACKS................. 5-14 a 5.6-4 FUEL STORAGE PATTERNS FOR REGION 11 RACKS RECONSTITUTION STAT 10N.................................... 5-15 l

6.2-1 0FFSITE ORGANIZATION...................................... 6-3 6.2-2 UNIT ORGANIZAT10N......................................... 6-4 6.2-3 CONTROL ROOM AREA......................................... 6-6 4

6 i

i i

SAN ON0FRE-UNIT 3 XVila AMENDMENT NO.

],

INDEY 1

LIST OF TABLES TABLE PAGE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................................. 3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS.................................... 3/4 3-58 3r3-12-RAD 40AGT4VE41@le--E4WENT-HOM4044NG-4NSTRUME44T AT404-i DE4-E4ED-1 a.' S RAD 40AGT4NE44WID--ERWENT M0!414044#G-INSTRUME44 TAT 40N SURVE414-ANC4-RE@l4EME44TS DELETED 3.3-13 RADIDACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION....... 3/4 3-66 4

4.3-9 RAD 10ACllVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................................... 3/4 3-68 l 4.4-1 MINIMUM NUMBTR OF STEAM GENERATORS TO BE INSPECTED DURING

INSERVICE INSPECT 10N..................... ..................... 3/4 4-14 a

4.4-2 3/4 4-15 STEAM GENERATOR TUBE INSPECT 10N...............................

3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.............. 3/4 4-20 3.4-2 REAC10R 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 PR0 GRAM....................................................... 3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITHDRAWAL i SCHEDULE...................................................... 3/4 4-29 i

3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE............. 3/4 4-31bd 4.6-1 TENDON SURVEILLANCE........................................... 3/4 6-12 4.6-2 TENDON LIFT-0FF F0RCE......................................... 3/4 6-13 I

l 3.6-1 CONTAINMENT ISOLATION VALVES.................................. 3/4 6-21 3.7-1 MAIN STEAM SAFETY VALVES...................................... 3/4 7-2 i

3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP l WITH IN0PERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS... ................................ 3/4 7-3 l

SAN ONOFRE-UNIT 3 XIX AMENDMENT NO.

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 sh:~ l be limited in l accordance with the limit lines shown on Figures 3.4-2, Pr4-31 3.4-4, and 3.4-5, 3.4-6, and 3.4-7 during heatup, cooldown, criticility, and inservice leak 1 and hydrostatic testing with:

a. A-madmwn-heatep- as speci fied by Figure 3.4 3 in-ay-1-hour period-with4GS cold = leg temper-ature-less-thawl&34r A meimum heatup of 60afr-in any 1-hour period with RCS cold leg temperature greater MI than 46 D or' equal to 860 F.

A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period b.

with RCS cold leg temperature less than 4M or equal to 147af. A l$Uf P maximum cooldown of 100*F in any 1-hour period with RCS cold leg temperature greater than 4M 147ef. lgyf

c. A maximum temperature change of 10af in any 1-hour aeriod during inservice hydrostatic and leak testing operations a)ove the heatup and cooldown limit curves.

i

d. A minimum temperature of 86cf 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 60af in any 1-hour period.

APPLICABILITY: At all times, t ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure i

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 and pressure to less than withinthenext6hoursand-reducetheRCST,7ollowing30 2000F and 500 psia, respectively, within the hours.

With the reactor vessel head bolts detensioned, RCS cold leg temperature may be less than 86cF.

SAN ONOFRE-UNIT 3 3/4 4-28 AMENDMENT NO.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM MMN4KG-00ND4-T4ON FOR OPF9M40N

, SURVEILLANCE RE0VIREMEN1S 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determi neti 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 Figures 3.4-2 and 3A-3 3.4-4 through 3.4-7. Recalculate the Adjusted Reference Temperature based en the-greater-of-the-f944ewin9+ in accordance with Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials," May 1988.

a. 'he-aetual-shi44-in-v+ference temperatute-fer-plate-G-6802-1-as determined by-ispaet-test 4n9r-or !

b. %e predieted-sM4-t-in-reference-temperature-for-we44 seams '2-303AT -

2-201Bv-er-2-203C as determined-by-Re9ulatory-Guide-4 r99, Re+i+ien--2, " Rad 441-ion-Embr4444 ement-of-Reaeter-Vessel-44a-tw4a1 s , "

May 1988.

SAN ONOFRE-UNIT 3 3/4 4-28a AMENDMENT NO.

i i

d NN i e iigu a i igiiiigiu e a giiii a a v y gT I s u ga i i u LOWEST SERVICE

INSERVICE TESTS # HEATUP h TEMP = 209'F l 1 3000 -

1 >

t ., 4

. .l . s ..

2 l

-s-i 70 .;- . . e . 4 1

W g < . . . . . . , ,. I... - , I. ,,

, I . ;., '; i >

- a' .

, w 2500 - '

i i

i ,

c

Acceptable operating region to the

., [ . 4 1... ++-...+- 4 ,

U right of the inservice tests curve , , ,

(Applicable in modes other than j

-i j .

W ' '

t c Modes 1 and 2) i

n. e 1

)'

i e 2000 # Acceptable operating region to the .l+. .

c.....4.. ~.4 ~ .

g right of the heatup curve in all modes. j In addition, in Modes 1 and 2 the i

i1 '

a -

i '

l operating region is to the right of the ~[. jT j [j 7

, m core critical curve. i ';

j1 i 05 - ,

.7 . . y . 7 4. .'L4 ....i.a ....5.. ...,...

j c 1500 -

, + i-i i j4 .

., .i , - . -

m.

g11

.7 ...,..

7.-.

w e, 3-t i !

. . 4 H

! , , 1! !

4 4 ...+,..7..'..7

.g 7 . i ,'

t

,..7-.. . , . , . .

9

. , . .i . ) J i .. j . i. .j ,

y 1000 .. ..

..l... '

f ! , -.

p.p - ,tj. ,

4 l . . p. 4..; . . . i . . ., . . ; +. 5. e < .!# CORE CRITICAL-t

..j;j. ; i

, . .p .. .. . ,

.. r . . s . .

.L f 500 -" - ~

,! -t -i-t- 1 MINIMUM "<bs+ot-b;ed-! '

r ;-++ o. <

, BOLTUP ,.

TEMP = 86*F -.

4 I ~, i +

1

_ _A _ , . i.

.- .- ~ i i.. ;

t i e i t i t i t i i i h j i d j g g g g h I

o 50 100 150 200 250- 300 350 400 F

INDICATED RCS TEMPERATURE ('F)-Te >

1 i

l FIGURE 3.4-2 l SONGS 3 HEATUP RCS PRESSURE / TEMPERATURE l

' LIMITATIONS FOR ' S UNTIt. 20 EFPY INA NormaF0peration' I

SAN ON0FRE-UNIT 3 3/4 4-30 AMENDMENT N0, wq- - _ - -++--p, q+ . . n ; .-

t l

l t

i I,

(Figure 3.4 3 - DELETED)

\

i 1

SAN ON0FRE-UNIT 3 3/4 4-30a AMENDMENT NO. 1 l

t !

. - - - N

y iy y v i e e a 1 i i g9 l ' '

l 'T

g ' ' 3 Y g 'f 1 3 LOWEST SERVICE COOLDOWN TEMP = 209'F 3000 -

6 1*

, . .I

<t i

i

. , , . . . . . . . - .4- '

O. s-i

! . e uj

I !

.g.. +

49 C i 8

. -;4 D 2500 -t -+ -*- " '-t-> - e+ t * + -- * - - - l

~ ' -

u) . , a g 4 3 -;

~

i i , ;

Q) I

...,,. .. i ..1, .y .4, j, j . . .

C ..'..

n. .

.j , ,.

-..;e7

, , 1 i C . , , . . . - . . ,s. ? .-

uJ

1, N 2000 -' '

?

' ~ t C .!. 4

!.;i y ! .1 .4.a .+..a 4

u) :

Unacceptable . ! !4 + t

[. 4. . . .. . . 4 . . . . . . . . - .. Operating ; .p j-j.. g/j1 m : .

Region ~ -

l Q.

l'4 ~

i 0 1500 ;

-l' Acceptable t H-N ; j 1 l-l

' ^

Operating ' * ' , j

. 9... .......+..m.4... .. .......f...p..... ......p.. 6...g...&.. ' -

a o i[j! i -

f j e <

Z ! 8 l-( ' 7 . . . ... 44 4 1000 . i ..__8. -- -- ,

-4

. ..; . 2 , . . . . +...... '

~.4 . .. . . . - . .

i . . .i . . . . . . .. . <

i. 1

. : i :

1 i 3-4-9-,

4 ,

. . 4 1..1. .. , , .. ; .

t,., ,.-+.. 6

3

' j.

i 50o . . v - 4. -4.. . . 4... . . 7 ......p.. . . _ !-

MINIMUM BOLTUP TEMP = 86"F e...-i.

.....5...._

$ 1

. . ,..r, 15 h i i iie i e t if I i t I h 1 i e it I - h1 0 50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F)-Tc' FIGURE 3.4-4 SONGS 3 COOLDOWN RCS PRESSURE / TEMPERATURE !

LIMITATIONS ~F0P,

S'UNTIL"20 EFPY Nonnal . 0peration [Su%

g SAN ONOFRE-UNIT 3 3/4-4-31 AMENDMENT NO.

3

)

l

! 120 iii i i i i i i i i -

i -

<- tio - -

p 8

100 -

go E M - -

~

L g 70 - -

w -

g 60 8 6 2a O 40 - -

30 -

20 -

10 - -

0

' I 'I I ' ' 'I 'I ' I ' I ' I ' I ' I '

B0 90 100 110 f20 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 ABOVE M4147'F l697b FIGURE 3.4-5 SONGS 3 RCS PRESSURE / TEMPERATURE-LIMITS MAXIMUM ALLOWABLE C00LDOWN RATES (-4-8 UNTIL.20 EFPY) l$tW1 NormalE0peration g I

SAN ONOFRE-UNIT 3 3/4 4-31a AMENDMENT NO.

j

r. . _

i.

i,

-1

3500 ,,,,,,, , , ,,,,,i..,,,,,,.i,,,,,,, .i....

! LOWEST SERVICE COOLDOWN ,

a I

2 TEMP = 209'F <

i ,

..,-1..-., ... . e, - _. .

, 3000 l

4 , >

4 t ; , t t' 3 1

y .. ; .

1 3 e . . . ..

. . , . h 1. .

] . [ f._

. .. ; 9 . .

Q 7 7_

- 4 .-

y) ,.

4. 4 , 4. . , i ..., .. 4. - 9.. . . . . . .. 4.. 4 . .. p . .p. i . 4 . . . p .. o . 4 .

4 . .s .

4 t

n.

- 2500 -- 4:

i i - - -

4 > - , .

1 - i1

- --1. :"

s -

1 W ! 'iq. . .j.. !< . q . .;. . l . . . . . . l . . , !

i i

...p.

e g., .

. . i . 2 i

.i e, D -

4 .

- - ' c . 3.) 1 , . . 4 , ,

u) !

u) w

. ..- . + .. & - i-.' :

1

- - . - a..+ . . + [ . ~ - -t. < 4 <

tt

! . Unacceptable j 2000 - Operating - -

w 4 t- 4 4 . Regbn <

g . .; ... .

4..:.. 3.....,

,.p....i +.* Acceptable p j . q .. q ..

- ;. - ; . . .

Operating -- <

g

]',, ! . . q . . , .. #. ' , - . l .4 2.

Region

, w 3500 .. 1 - .h t t . . .... . .e n i , _ . . . ; . L t . .

@ - ..d.4. .j.i.Q.. !. 4 3..!4 1 .l.4.i.h . . ._.....2.. .. ,

6(Af'1 o >

+

iii !

i .m , _ _ - . . _ .

w

! ! i i 4 '

i ]

H q..

' ..; 1

.g. . .q... ,

I i

.!.,. y . 7. . . ........,.e.., .

5 1000 -- , . ,, ..

z ,

i 5

a -

. 4 r g . . .i . . g . .,

. .,.5 .- . _ . . . . y ,

4-.. -

p.

ii .

) *f > fj* * + + 4 + i

f. .' '

500 - -

y ,

. o.

e

-} . '

l ! !

l MINIMUM _ ! >-l ! ;

j BOLTUP ! ! .. L i .. ; ... ! . . 3. . ... ... . . .. . . .....;..._., , , ,. , , ,

': ;~

TEMP = B6*F !: ;

g i . ' i .i: ~ .- ,

, r

-. L ,- , ,

ie i!Iiii ! !i ! l i lii ;lii!il i i i iiiie i I,iiu g

O SO 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ('F)-Tc i

l l

1 FIGURE 3.4-6 1

SONGS 3 COOLDOWW RCS PRESSURE / TEMPERATURE LIMITATIONS UNTib 20 EFPY l Remote. Shutdown _0peration Muf'4 P

i i

j SAN ONOFRE-UNIT 43' 13/4 4-31b' 1 AMENDMENT NO.

, , . , , - . - . .,-.v- , - - - - - - - , -

= . . . - . , - . - nw

120 i i i i i i , ,

i . .

i i .

i 110 -

100

,o . -

@ 80 - -

[ b '

b b 70 -

60 -

z 50 -

l g 40 -

911 30 -

20 -

10 -

i i .i ..i . i . i . i . i i . i i o

80 90 100 110 120 130 140 150 100 170 180 100 200 210 l

INDICATED RCS TEMPERATURE (*F)-Tc NOTE: A MAXIMUN C00LDOWN RATE OF 100*F/HR IS' ALLOWED

~

'AT.ANY TEMPERATURE AB0VE 155'F g

FIGURE 3.4-7 SONGS'3:-RCS PRESSURE / TEMPERATURE LIMITS M2 ti; M ALLOWABLE COOLDOWN RATES'(UNTIL.20 EFPY) l $tp,(

Remote Shutdown Operation I SAN.0NOFRE-UNIT 3i ~3/4 4-31c- AMENDMENT NO.

_ _ _ _ _ _ _ _ . _ _ _ _ d

TABLE 3.4-3 low Temperature RCS Overpressure Protectioh Rance 4

I Qgeratina Period. EFPY Cold Leo Temperature. *F During During Heatup Cooldown 4-t+ 8 Until 20 (Normal Operation) s 302 245 s 267 225 Nfd Until 20'(Remote Shutdown Operation)

5 225 i

l i

i i

t f

I

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

l SAN ON0FRE-UNIT 3 3/4 4-31b- d AMENDMENT NO.

i l

I REACTOR COOLANT SYS1EM

OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s M3 246cf Gu LIMITING CONDITION FOR OPERATION 3.4.8.3.1 No more than two high-pressure safety injection pumps shall be l OPERABLE and at least one of the following overpressure protection systems shall be OPERABLE

i

a. The Shutdown Cooling System Relief Valve (PfiV9349) with:

1) A lif t setting of 406 i 10 psig*, and

2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and

. 3HV9378 open or,

b. The Reactor Coolant System depressurized with an RCS vent of greater j than or equal to 5.6 square inches.

.A_PPL IC ABillTY: MODE 4 when the temperature of any cne 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:

l a. With the SDCS Relief Valve inoperable, reduce T,y to less than i

200*F, depressurize and vent the RCS through a greater than or equal

! to S.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b. With one or both SDCS Relief Valve isolation valves in a single SDCS Relief Valve isolation valve pair (valve pair 3HV9337 and 3HV9339 or valve pair 3HV9377 and 3HV9378) closed, open the closed valve (s) or power-lock open the other SDCS Relief Valve isolation valve pair

~

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , reduce T,y to less than 2000F, depressurize and vent the RCS through a greater than or equal to 5.6 inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. With more than two high-pressure safety injection pumps OPERABtE, secure the third high-pressure safety injection pump by racking out

its motor circuit bre,ker or locking close its discharge valve i within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

I

'The lif t setting pressure applicable to valve temperatures of less than or equal to 1300F.

SAN ON0FRE-UNIT 3 3/4 4-33 AMENDMENT NO.

RJACTOR COOLANT SYSTW OVERPRESSURE PROTECTION SYSTEMS RCS TEMPLRATURE >302 246'T i (IPilTING 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 1 10 psig*, and l 2) Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and

! 3HV9378 open corr or, GULA l

b. A minimum of one p*ressurizer code safety valve with a lift setting of 2500 psia i 14 .

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

veat the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ,

b. In the event the SDCS Relief Valve or-an--RC4-vent is used to l mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specifica-tion 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the ef fect of the SDCS Relief Valve code safety valve or-4C4-vent on the transient and any g corrective action necessary to prevent recurrence.

( SURVElLtANCE REQUIRE,MENTS 4.4.8.3.2.1 The SDCS Relief 7alve shall be demonstrated OPERABLE by:

a. Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the SDCS Relief Valve isolation valves 38V93',7, 3HV9339, 3HV9377 and 3HV9378 are open when the SDCS Relief VM u is being used for overpressure protection.

b. Verifying relief ulve 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.

4r44 Mr3-T he-R C4-v en t-sha41- be-ver4 f4 ed-to4e-open-a t4ea s t-onc+-per-lG-hou r-s when-the-ven t-i s-heing-u sed-for-overpressu re-pretec44sn rewept-when-the-vent pa thway-is-prev 4ded-wvt4%4*e-wMeh-i+-4ec4ed r-sealedror-otherwise-secur+d-in-the-open-pos+t4enrthen-ve r-ify-t hese-vM ves-open-a t4 ea st-once- pe r-M-da y s,

'The lif t setting pressure applicable to valve temperatures of less than or equal to 130oF.

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

SAN ON0FRE-UNIT 3 3/4 4-35 AMENDMENT NO.

l j

REACTOR COOLANT SYSTEM BASFS PRESSURE / TEMPERATURE LIMITS (Contin gdl The heatup and cooldown limit curves for normal operation (Figures 3.4-2 and 3.4-34) 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/nr or cooldown rate of up to 100*F/hr. The limit curves for Remotc 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 tht service period 4nd4+at+d-on449ur+s4r4-2-and-h4-3, and they include adjustments for ine,trument uncertainties, and static and dynamic heads.

The reactor vessel materials haw-been were tested prior to reactor startup to determine their initial RTsp; the results of these tests and the updates lX4%

resalting from the evaluation of material properties in response to Generic i Letter 92-01, " Reactor Vessel Structural Integrity," Revision 1 are shown in l Table B 3/4.4-1. Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation will cause an increase in the RTsa. Therefore, an adjusted reference temperature, based upon the fluence and copper and nickel content of the material in question, can be predicted using FSAR Table 5.2-6 6 and the l recommendations of Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vesse' Materials." The heatup limit curve (Figure 3.4-2) and the cooldown limit curves, Figures 3.4-24, and 3.4-36, include predicted adjustments for this shift in RT at the end of the applicable service period, as well as adjustments for poss+pMe-er+ ore in the-pre wre-an44emper+ture-sensing 4nst-rement+ instrument uncertainties, and static and dynamic heads.

The actual shift in RT sp of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation sur-veillance specimens installed near the inside wall of the reactor vessel in 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 he 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 e determined from the surveillance capsuleisdifferentfromthecalculateddeltaRT 33 for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figures 3.4-2 and 4r4-3 for reactor criticality and for inservice leak and hydrostatic testing have been l provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

SAN ONOFRE-UNIT 3 B 3/4 4-7 AMEGAEM NO.

REACTOR COOLANT SYSTEM g,AsEs PRESSU_RE/ TEMPERATURE LIMITS (Continued)

The maximum RT sy for all Reactor Coolant Svstem pressure-retaining materials, with the exception of the reactor pressure vessel, has been determined to be 900F. The Lowest Service Temperature limit line shown on figures 3.4-2, 3.4.4 and 3.4-3 6 is based upon this RT since Article NB-2332 (Summer Addenda of 1972) of Section 111 of the ASME Boil,e,r and Pressure Vessel Code requires the Lowest Service Temperature to be RTsn + 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 r.nd cooldown rates and spray water temperaturc differential are provided to assure that the pres-surizer is operated within the design criteria assumed for the f atic,ue 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, "Overpressurizatlen 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 metal temperature of at least RTug + 900F at the beltline location (1/4t or 3/4t) that is controlling in the Appendix G limit calculations."

SAN ONOFRE-UNIT 3 8 3/4 4-7a AMENDMENT NO.

i g TABLE B 3/4.4-1 z

o REACTOR VESSEL TOUGHNrSS Temperature of Minimum Upper

- T Drop Charpy V-Notch Shelf Cv energy

E Weight 0 30 @ 50 for Longitudinal Z Piece No. Code No. Material Vessel location Results l( - lb - ft - Ib Direction-ft lb w

215-01 C-6801-1 A533GRBCLI Upper Shell P? ate -20 28 64 115 215-01 C-6801-2 A533GRBCLI Upper Shell Plate -20 -6 34 106 215-01 C-6801-3 A533GRBCLI Upper Shell Plate -20 18 36 115 215-02 C-6802-4 A533GRBCLI Lower Shell Plate -30 3340 6270 1158 215-02 C-6802-5 A533GRBCL1 Lower Shell Plate 0 3640 6470 1106 215-02 C-6802-6 A533GRBCL1 Lower Shell Plate -40 3340 MG80 9092 215-03 C-6802-1 A533GRBCL1 Intermediate Shell -20 6680 100 9594 l9pfR m 215-03 C-6802-2 A533GRBCLI Intermediate Sheil -20 40 6670 1135 g 215-03 C-6802-3 A533GRBCL1 Intermediate Shell -10 4460 80 1015 l I 203-02 C-6823 A508CL2 Vessel Flange Forging 0 -30 -15 NA in 209-02 C-6824-1 A508CL2 Closure Head Flange -40 -100 ,JO 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 -1 -15 125 205-06 C-6830-2 A508CL2 Outlet Nozzle Forging -10 -20 -5 131 232-01 C-6840-1 A533GRBCLI Bottom Head Torus -50 -10 0- 107 g 232-02 C-6841-1 A533GRBCLI Bottom Head Dome -40 10 20 99 E

o 8

)

ENCLOSURE 3 TECHNICAL SPECIFICATION PAGES CONTAINING THE CHANGES WHICH WERE PREVIOUSLY REQUESTED IN AMENDMENT APPLICATION N0. 101 (PCN-354) DATED SEPTEMBER 3, 1992, AND ARE BEING REQUESTED IN THIS LICENSE AMENDMENT ki APPLICATION NO, 102 (PCN-359)

UNIT 3 50fk

INDEX l1ST OF TABLES TABLE PAGE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................. 3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENis............................... 3/4 3-58 3r3-12 R A N OAC44N E-L4 QU10-Ef fE0EN T-MON MOR I NG-14 STRUMENTAT404--

DE4E4 D- ()/

4.3 P RAD 10AGT4NE-L4 QUI") EFFL4E41-MONMOR44G-4NSTRUMEt4TA140N 78f SURVE4bbANGE-REQUIREMENTS DELE 4E0 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-66 4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTPON SURVEILLANCE REQUIREMENTS................................. 3/4 3-68 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECT 10N...................................... 3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECT 10N........................... 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 d PR0 GRAM.,................................................. 3/4 4-26 [#i 4.* 5 REAGTOR-VESSEL-MATE RML-SURVE4LLANGE-PROGRAlb--WMHOR AWAE SCH E 0 0 L E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . --3 /4-4-2 9 %l 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE......... 3/4 4-31bd 591 4.6-1 TENDON SURVEILLANCE....................................... 3/4 6-12 4.6-2 TENDON LIFT-0FF F0RCE..................................... 3/4 6-13 3.6-1 CONTAINMENT ISOLATION GLVES. ............................ 3/4 6-21 3.7-1 MAIN STEAM SAFETY VALVES.................................. 3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LEVEL-HIGH TRIP WITH INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS................................ 3/4 7-3 SAN ONOFRE-UNIT 3 XIX AMENDMENT NO.

REACTOR COOLANT SYSTEM t

3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM 144144 M-00t4NJ4014-FOR-O PF 4 AT404 .

f(.N SURVElllANCF Rf0VIREMENTS 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.J The reactor vessel material irradiation surve,illance specimens shall be removed and examined, to determine changet. in material properties, at the-inter-vals as required by 10 CFR 50 Appendix H.4n-secordance-with-the-schedule in-Table 4.15- The results of these examinations shall be used to update hj Figures 3.4-2, andh 4-3 3.4-4 through'3.4-7. Recalculate the Adjusted Reference Temperature based-on-the-greater--of-t4+e-4o14ewis9+ in accordance with Regulatory Guide 1.99 Revision 2, " Radiation Embrittlement of ".aactor Vessel Materials," May 1988.

{cff

a. T he-eetua4-s hi44-i n-r+4e renee-t em pera t u ref o r-p te t#-c-sao2-1-a s determined-by-impaet-les-tingr-or 3} 6 tw The-p redicted-shi4 t-ifsefe rence-tempera tu re-fo r-wel d-+eams-2-203 Ar 2-203 Br-e r-2-203C-a s-de t+ rm i ned-by-Regule te ry-Gu i de--1A9r-Rev4sion 2, "Radiat4en-Embr4t-t4ement-of-Reaetor-Wessel-Mater 4ah,1 May-10 ear i

SAN ONOFRE-UNIT 3 3/4 4-28a'29 nMENDMENT N]. < f, N

REACTOR COOLANT SYSTJM ggfS PRESSUREjTEMPERATURE LlHITS (Continued)

The heatup and cooldown limit curves for normal operation (Figures 3.4-2 d and 3.4-34) and the cooldown limit curve for remote shutdown operation (Figure q 3.4-6) are composite' curves which were prepared by determining the most conservative heatup rate ofcase, up towith either 60'F/hr orthe inside orrate conidown outside of upwall controlling,limitfor to 100*F/hr. The- any curves;for Remote Shutdown operation are determined using thecTotal Loop Uncertaintigs (TLUs) for. temperature and pressure for the Remote Shutdown ~ Panel- fg.

instruments"in which the; pressure TLus are higher than? those for: the Cottrol Room D 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 indicated on F!gures-312 anc 313; and they--include adjustments forlin$trument' uncertainties,: and. static and dynamic heads.

i The reactor vessel materials have been were tested prior to reactor!startup 0g i to determine their initial RTm; the results'~of'these tests and the' updates e resulting:from the evaluation of material properties;in response to: Generic l

l Letter 92-01, " Reactor Vessel Structural; Integrity," Revision 1 are shown in l

, Table B 3/4.4-1. Reactor operation and resultant fist neutron'(E greater than 1 l HeV) irradiation will cause an increase in the RTm . Therefore, an adjusted-

reference temperature, based upon the fluence and copper and nickel content of -

the material in question, can be predicted using FSAR Tahle 5.2-6 6 and the l h>cA!

l recommendations'of ReguN tory Guide 1.99, Revision 2, " Radiation Embrittlement of i Reactor Vess?1 Materials." Theheatuplimitfcurve1(Figure 3,'4~2)andthe -

id i rooldown limit curves, Figures 3.4-24; and 3.4-36, include predicted adjustments M'

'ar this shift in RT a l-adjustments for podet the endinof er+crs the the applicable pressure service period, as well as and tempe:ature-senshg Igt/

^

Mst+uments instrument uncertainties,Jand staticLand' dynamic heads. 9fi The actual shift in RT of the vessel material will be establi:,hed l

periodically during operation,by iemoving and evaluating, in accordance with g ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation sur- ltgj veille nce specimens installed near the inside wall of the reactor vessel in i the core area. The surveillanco specimen withdrawal schedule is shown-4*-

4Me4c4-6~ maintained :in .the' FSAR. Since the neutron spectra at the irrediation M S -

(d sairples and vessel inside radi's u are essentially identical, the measured j 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 m determined from the surveillance capsule is different from the calculated delta RTm for the equivalent capsule radiatic>n exposure.

The pressure-tem erature limit lines shown on Figures- 3.4-2 and-ar4-3 for reactor criticality and for inservice leak and hydrostatic testing have been N provided to assure compliance with the minimum temperature requirements of Appendix G to 10 CFR 50.

SAN ONOFRE-UNIT 3 0 3/4 4-7 AHENDMENT NO.

i

ML20073T096

Text

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