Text

Proposed Tech Specs 3/4.4.8.1,3.4.8.3.1 & 3.4.8.3.2 Re RCS Pressure/Temp Limits
	ML20036A037
Person / Time
Site:	San Onofre
Issue date:	04/30/1993
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
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ML20036A035	List: ML20036A034; ML20036A037; ML20036A102;
References
	NUDOCS 9305070285
	Download: ML20036A037 (41)
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b ATTACHMENT A EXISTING TECHNICAL SPECIFICATIONS AND BASES UNIT 3 i

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9305070285 930430 PDR ADOCK 05000362

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.:w: :NG CONDI~* N FCR CpERAT*;N ANC $URVE LJNCE 4E t!; ewe %T5 DAGE

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3/44-5

>.: 5-U :ChN - LOCPS FILLEO.

3, a 4 5 7

COLO 5HUTDChN - LOOPS NOT FIL'.EO.

r 3,' 4 4 -

l 3/4. 4. 2 5AFETY VALVE 5 - CPERATING..

3/4 4-3 3/4. 4. 3 3RE55URIZER.....................

3/4 4-9 3/4.4.4 STEAM CENERATOR5.....

3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE 3/4 4 17 I

LEAKAGE DETECTION SYSTEMS.....................

3/4 4-13 OPERATIONAL LEAKAGE..............................

3/4 4-21.

3/4.4.5 CHEMISTRY...........................................

3/4 4-24 3/4.4.7 SPECIFIC ACTIVITY......................

3/4.4.3 PRES 5URE/ TEMPERATURE LIMITS 3/4 4-23 R EACTOR COOLANT SY STEM.....................

3/4'a-32 PEE 550RIZER - HEATUP/C00LOCVN..

CVERPRESSURE PROTECTION SYSTEMS 3/4 4-33 RCS TEMPERATURE < 302*F........

3/4 4-35 RCS TEMPERATURE I 302*F.......

3/4 4 36 3/4.4.9 STRUCTURAL INTEGRITY.......................

3/4 4-37 3/4.4'10 REACTOR COOLANT GAS VENT SYSTEM.......................

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4 5-1 l

3/4.5:1 SAFETY INJECTION TANK5...................................

ECCS SU85Y STEMS - T v g > 3 50' F................... ~.....

3 / 4 5 - 3 '

3/4.5.2 a

3/4 5-7.

ECCS SUBSYSTEMS - T,,,

350*F....................

l 2/4.5.3 L

3/4 5-3 3/4.5.4 REFUELING WATER STORAGE TANK...................

1' f

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y AMENCMENT No.

-e' SAN CNOFRE - UNIT 3 b

4 e.

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'1 IN00 LIST CF FIGURES PAGE FIGURE MINIMbH BORIC ACID STORAGE TANK VOLUME AND TEMPERATU 3/4'l-13

.3.1-1 AS A FUNCTION OF STORED BORIC ACID CONCENTRATION.......

CEA INSERTION LIMITS V5 FRACTION OF ALLOWASLE THERMAL 3/4 1-24 l

3.1 2 P0WER...................................................

u 3/42 DN8R MARGIN OPERATING LIMIT BASE ON COL 55..............

i 3.2-1 l

DNBR NARGIN OPERATING LIMIT BAS G ON CORE PROTECTION 3/4 2-8

?

3.2-2 CALCULATORS (COLS$ OUT OF5ERVICE)......................

f 3/4 3-40 DEGRAD G BUS V0LTAGE TRIP 5ETTING.......................

3.3-1 3/4 4-15 TURE WALL THINNING ACCEPTANCE CRITERIA.................

i 4.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIV 3.4-1 LIMIT VERSUS PERCDtT OF RAT E THERMAL POWER WITH THE PRIMARY C00LANT SPECIFIC ACTIVITY >1.0 pC1/4 RAM DOSE 3/4 4-27 l

EQUIVALENT I-131........................................

HEATUP RCS PRES $URE/ TEMPERATURE LIMITATIONS FOR 3/4 4-30 Li 3.4-2 0- 5 Y EARS.........

I C00LDOWN RCS PRES $URE/ TEMPERATURE LIMITATIONS FOR 3/4 4-30a

~

3.4-3 0-5 YEARS...............................................

3/4'4-311 l

RCS PRES 5URE/TD#ERATURE LIMITATIONS FOR 4-8 EFPY.......

3.4-4 l

r RCS PRESSURE / TEMPERATURE LIMITS MAXI M ALLOWABLE e

3/4 4-31a l

3.4-5 j

C00LDOWN RATES (4-8 EFPY)...............................

MINI M REQUIRED FEEDWATER INVENTORY FOR TANK T-121 FO 3/4 7-7 3.7-1 MAXI M POWER ACHIEV E TO 0 ATE..........................

~

5-2 EXCLUSION AREA.'...............................'c,i........

5.1-1.

5-3

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5.1-2 LOW POPULATION10NE............................

5-4 l

SITE SOUSARY FOR GASEQUS EFFLUENTS.........',............

-t 5.1-3 5-5..

5.1-4 SITE 80USARY FOR LIQUID EFFLUENTS......................

5.6-l UNITS 2'AS 3 FUEL MINI M BURMUP V5. INITIAL 5-12' l

ENRIQtENT FOR REGION II RACK 5..........................

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5.6-2 UNIT 1 FUEL MINI M SURSP V5. INITIAL "5-13 ENRIQ9ENT FOR REGION II _ RACK 5..........................

'b FUELLSTORAGE PATTERN 5 FOR REGION II RACKS...............-

l5-14, 5.6-3'

.s AMENDMENT Nos 77-3 I... sUNIT 3-- =

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. SAN'ONOFRE E

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FUEL STORAGE PATTERNS FOR REGION II AACKS 5.6-4 5-15 l

RECONSTITUTION 5TATION..................................

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6. 2-1 CFFSITE. ORGANIZATION....................................

6-4 6.2-2 UNIT ORGANIZATION........................................

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-INDEX LIST OF TABLES

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b 'I TABLE P' AGE I

4.3-7 ACCIDENT MONITORING I'NSTRUMENTATION SURVEILLANCE REQUIREMENTS...............................................

3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS.................................

3/4.3-58 3.3-12 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION --

DELETED 4.3-8 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -- DELETED L

3/4 3-66 3.3-13 RADI0 ACTIVE GASECUS EFFLUENT MONITORING INSTRUMENTATION....

4.3-9 RADI0 ACTIVE GASE005 EFFLUENT MONITORING' INSTRUMENTATION SURVEILLANCE REQUIREMENTS..................................

3/4 3 ]

4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO SE INSPECTED DURING INSERVICE INSPECTION.......................................

3/4.4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION............................

3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES...........

3/4 4-20 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY...........................

3/4 4 4.4-3 REACTOR COOLANT SYSTEM CNEMISTRY LIMITS SURVEILLANCE-REQUIREMENTS...............................................

3/4 4-23 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE...................

3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM - WITH0RAWAL SCHEDULE.t.................................................

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 TENDON LIFT OFF F0RCE.......................................

3/4'6-13 3.6-l CONTAINMENT ISOLATION VALVES.................................

3/4 6-21

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3.'7-1 MAIN $ TEAM 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 e

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. SAN ONOFRE - UNIT 3 XIX

' AMENDMENT NO.31

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aEaC R ~;CLANT SYS~EW PRESSURE / TEMPERATURE t'wI*3 3/a.4.3 qEac CR CCOLANT SYSTEM L!w! TING CONDITICN FOR OPERAT!CN

=

- with tne reactor vessel head toits tensioned", the Reactor Coolant-System (exceot the pressurizer) temperature and pressure snail be limited in 3.4.S.1 i

3.4-2, 3.4-3, 3.4-4 arc accorcance with the limit lines shown on Figures 3.4-5 curing hea testing witn:

A maximum heatup as specified by Figure 3.4-3 in any 1-hour period A maximum neatup of witn RCS cold leg temperature less than 153*F.

a.

60*F. in any 1-hour period with RCS cold leg temperature greater than 153*F.

A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period A maximum coolcown with RCS cold leg temperature less than 126*F.

b.

of 100*F in any 1-hour period with RCS cold leg temperature greater than 126*F.

A maximus temperature change of 10*F in any 1-h'our period during inservice hydrostatic and leak testing operations above the heatup c.

and cooldown limit curves.

A minimum temperature of 86*F to tension reactor vessel head bolts.

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d.

With the reactor vessel head bolts detensioned, the Reactor Coolant System

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(exceot 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 cetertaine the effects of the out-of-limit condition on the structural dntegrity.

of the, Reactor Coolant System; detemine that the Reactor i

l and pressure to less than withinthenext6hoursandreducetheRCST"YS11owing30 hours.

200*F and 500 psia, respectively, within the l

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

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s be less than 86*F.

AMEN 0 MENT NO.7f SAN ONCFRE - UNIT 3 3/4 4-28 e

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ete-mirac : :e itnin tne limits at least ente :er 30 minutes :ur g sjste-eat.:, :: ale: n, anc inservice leak anc nycrostatic testing Operati:ns.

T*e react:r vessel material ir-adiation surveillance scaciters 4. 4. 3. '. 2 snall :e -emc ec anc exsmined to cetermine nanges in saterial :rocerties, a:

se intervals recuirec ey 10 CFR 50 Accencix H in ac:orcance witn tre screcute The results of trase examinations sna11 ce used to accate in Ta:le 4,4-5 Figures 3.4-2 arc 3.4-3.

Recalculate tne Acjusted Reference Temcaratare_:ase:

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n tre greater of tne following:

t 7 e actual snift in reference temperature for plate C-6802-1 as l

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caterminec y incact testing, or The recicted snif t in reference temcerature-for welc seams 2-2034 2-2033, er 2-203C as ceterminec y 4equlatory Guice 1.99, Revisien 2," Radiation Emerittlement of Reactor vessel dateria s,'

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NOTL A MAXBdW CCOUXNfM RATE OF 100'F/HR 15 s

ALUNfED AT ANY TDdPERATURE A80VE 128 F j-

~ FIGURE 3.4-5 SONG,5 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWA8LE COOLDOWN RATES (4-8 EFPY) g-SAN ONOFRE - UNIT 3 3/4 4-31a AMEN 0 MENT No.71 1

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3. 3. 3. '. A: ' east one of tne following over:ressure :rotection systems-sea"

e ::E ABLE; e Snutcown Coolirg System Relief Va ve (85V9349).itn-a.

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A lift setting of 406 : 10 psig", anc Relief valve isolation valves 3HV9337, 3HV9339, 3kV9377. ar:

2) 3HV9378 open, or,

~5e Reactor Coolant System ce:ressurized witn an RC5 vent :f ;reate-

nan or equal to 5.5 square inenes.

AP t:CASIL!7Y:

M20E 4 nen tne temeerature of any one RC5 col: leg is less

nan or ecual to snat specified in Taele 3.4-3: M00E 5; MODE 6 witn tne rea:.:-

vessel neae on.

AC**:N:

Witn tne SCCS Relief valve inoperacle, recuce T,yg to less tnan a.

200*F, cepressurize and vent the RCS through a greater than or ecua!

to 5.5 square inen vent witnin the next 8 nours.

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ditn one or both 50C5 Relief valve isolation valves in a si gie

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c.

30C5 Relief Valve isolation valve pair (valve pair 3HV9337 are

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38V9339 or valve pair 3HV9377 and 3HV9378) closec, ocen tne :!:se:

<alve(s) witnin 7 days or reduce T to less than 200*F. ce:res-surize anc vent the RCS througn a @iater tnan or equal to 5.6 d c-vent nitnin the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , In the event eitner the 50C5 Relief valve or an RCS vent is used to sitigate an RCS pressure transient, a Special Reccet sna11 ce :reca e:

and sucaitted to the Conseission pursuant to Scecification 6.9.2 The report shall describe tne circumstances initi-witnin 30 days.

ating the transient, the effect of the 50C5 Relief valve or RCS vent on the transient and any corrective action necessary to prevent.

recurrence.

The provisions of' Specification 3.0.4 are not apolicable.

SURVEILLANCE REQUIREMENTS 4.4.3.3.1.1 The 50C5 Relief valve shall be comenstratec OPERABLE oy:

verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months when the SDCS Relief valve is a.

ceing used for overpressure protection that 50C5 Relief Valve isolation valves 3HV9337, 3HV9339, 3HV9377,*anc 3HV9378-are seen.

"The lif t setting pressure applicanle to valve temperatures of less than or I

equal to 130*F

.V AMENOMENT No.'3 SAN CNOFRE - UNIT 3 3/4 4-33 p

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sp: :: :::.a : 5 Ew ysg;oaE55 TEE 0'C*EC*:09 sv5 Ew!

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3.4 3.3.2

e ::E ABLE:

~ e Snute:wn C: cling System Relief Valve (PSV9349) witn-a.

1)

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

Relief valve isolation valves 3HV9337, 3HV9339, 3FV9377, arc 3Hv9378 open, or, A minimum of one :ressuri:er c:ce safety valve with a lift sett ;

c.

of 2500 psia 1%"

MODE a with RCs temperature above that specified in Table 3.4-3.

2;:L::AsiLI'v:

aC*:CN:

Witn no safety er relief valve OPERABLE, be in COLD SHUTCCVN and a.

vent tne RCS threugn a greater tnan or equal to 5.6 square inca vent witnin the next 8 nours.

In the event tne 50C5 Relief Valve or an RCS vent is used to I

b.

mitigate an RCS pressure transient, a Special Report sna11 e

>recared and submitted to the Conseission :ursuant to Scecifi:a-tien 5.9.2 witnin 30 cays.

The report small cascrite tre

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circumstances initiating the transient, the effect of the SCC 5 Relief Valve coce safety valve or RCS vent on the transient arc a j corrective action necessary to prevent recurrence.

5tEVE*LLANCE REOUIRE.ENTS w

4.4.8.3.2.1 The '50C5 Relief Valve snail ne comonstrated OPERA 8LE ty:

Verifying at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months that the SOCS Relief valve'-

a.

isolation valves 3HV9337, 3HV9339, 3Hv9377 and 3HV9378 are coen =nen the SDCS Relief Valve is being used for overpressure protection.

Verifying reitef valve satcoint at least once per 30 months wnen b.

tested pursuant to Specification 4.0.5.

The pressurizer. code safety valve has no additional surveillance 4.4.8.3.2.2 -

recuirements other than those. required ey Specification 4.0.5.

The RCS vent shall be verified to be open at least once :er 12 s: -s 4.4.8.3.2.3

=ren tne vent is being used for overpressure protection, except wnen tBe ve"t

ateway is provided with a valve whien is lected, sealed, or otherwise secee:

in ue open position, tnen verify tnese valves open at least once per 31 :ays, "The lif t setting pressure toglicable to valve tempe'ratures of less tnan Or equal to 130'F.

- The lif t setting pressure shall correspond to ancient conditions of tne'vabe at nominal operating temperature and pressure.

AMEN 0 MENT NO. P.

l 7

SAN CNOFRE - UNIT 3 3/4 4-35 n

~

-e

p: :: ::: W 5*5 Ew BASE 5

555JE/~E.WPERATURE O MIT5 (~:et nuec)

Jc and c:cidown limit :veves (Figures 3.4-2 ard 3.4-3) are Tre.nea:

-::s te :.rves.nien ere :recarec :y ceterminirg t e mest ::nservative

sse.

t, e t e :re insice or cutsica -all c:nt elli g, 'or.any aeatuo -ate 50% nr r :: ele:wn rate of uo to 100*F/hr.

T*e reatue are ::oi::+n

f.,: ::

nes ere :re:ared casec acon tne most limiting <alue of tre :recietec a: psted reference temperature at tne end of tne service :eriod inci:a:ec :-

ipres 3.4-2 anc 3.4-3.

The reactor vessel materials nave teen tested to cetermine tNeir ' nit'a React:r

tne results of these tests are snown in Ta01e 8 3/4.4-1.

R7Ocehitien and resultant f ast neutron (E greater than 1 Mov) irraciatien wi

y Therefore, an adjusted reference temcerat.re.

cause an increase in tne RT

asec seen tne fluence and $No.er and nickel c:ntent of tne saterial in.

uestion, can te precicted using FSAR Taele 5.2-5 and the recommencations :f Replatory Guide 1.99, Revision 2 " Radiation Escrittlement of Reactor Vessel The neatup and cooldown limit curves. Figures 3.4-2 and 3,4-3.

date ri al s. "

incluce credicted adjustments for tais snift in RT at the end of the acclic-aole service period, as well as adjustments for poskIble errors in tne pressure y

anc temcerature sensing instruments.

ine actual shift in RT of the vessel material will be estaolisned periccically curing operatikby removing and evaluating, in accorcance wi.tn ASTM Ela5-73 and 10 CFR Appendix H, reactor vessel material irradiati:n sur-

<e111ance specimens installed near tne inside wall of the reactor vessel in

)

The surveillance specimen withdrawal schedule is snown in tre ::re area. Since the neutron spectra at tne irradiation samples anc vesse:

Ta:!e 4.4-5.

~

insica radius are essentially identical, the measured transition snift fer a same'e can te acolied with confidence to the adjacent section' of tne reacte sessel taxing into account the location of tne sample closer to the core tran t e vessel wall by means of the Lead Factor.

The neatuo and coolcewn :veves cetermined from tne surveillance must :e recalculated when the delta RT

a:suleiscifferentfromthecalculatMaeltaRT for the equivalent ca:sJ e NOT ractation exposure.

The pressure-teeperature limit lines shown on Figures 3.4-2 and 3.4-3 for reactor criticality and for inservice leak and nydrostatic testing nave teen proviced to assure compliance with the minimum temperature requirements of appendix G to 10 CFR 50.

The maximus RT for all Reactor Coolant System pressure-retainiag materials, with the$ception of the reactor. pressure vessel, has Deen ceter-mined to be 90*F.

The Lowest Service Temperature limit line sacwn on.

L since Article N8-2332 (S e e-Figures 3.4-2 and 3.4 3 is based upon tnis RTAccencaof1972)ofSectionII 100*F f or pioing,. Dum:s recuires tne Lowest Service Temperature to be RTare valves. Below this maximum of 20% of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates are spray watcr temperature differential are provided to assure that tne pres -

surizer is operated within the design criteria. assumed for the fatigue analysis

)

performed in accordance with the ASME Code requirements.

I w

AMENOMENT NO. M L

SAN CNCFRE-UNIT 3 8 3/4 4-7 s

4

T a

IAHi[ B 3/4.4-l_

RfAC10R V[55(L 100GilNE55 E2 leaperature of Minimum Upper 5

Drop Charpy V-Notch Stielf Cv energy E

Weight

30

@ 50 for longitudinal Piece No.

Code No._

Material Vessel iocation Results it - Ib - ft - Ib Direction-it th 5

-20 28 64 115

']

215-01 C-6801-1 A533GASCL1 Upper Shell Plate

-20

-6 34 106 215-01 C-6801-2 A533GASCL1 Upper Shell Plate

-20 18 36 115 215-01 C-6801-3 A533GMCLI Upper Shell Plate

-30 32 62 115 215-02 C-6802-4 A533GASCL1 tower Shell Plate 0

36 64 110 215-02 C-6802-5 A533GASCLI Lower Shell Plate

-40 32 100 90 215-02 C-6802-6 A533GROCL1 Lower Shell Plate 215-03 C-6802-1 A533GaSCL1 Intermediate Shell

-20 56 100 95 215-03 C-6802-2 A533GMCL1 Intermediate Shell

-20 40 66 11.1 i

215-03 C-6802-3 A533GROCL1 Intermediate Shell

-10 44 80 101 I

i 203-02 C-6823

.A508CL2 Vessel !lange forging 0

-30

-15 NA i

209-02 C-6824-1 A500CL2 Closure Head flange

-40

-100

-100 NA i

se forging 205-02 C-6829-1 A508CL2 Inlet Mozzle forging 10

-35

-5 109 20' ^

C-6829-2 A508Cl2 Inlet Nozzle forging 0

-55

-35 156 20' -

C-6829-3 A508Cl2 Inlet Nozzle forging 10

-25 35 112 C-6829-4 A500Ct2 Inlet Nozzle forging 10

-30 25 108 20 Outlet Nozzle forging

-10

-30

-15 125 i

I Outlet Nozzle l'orging

-10

-20

-5 I II 2(1 3d C-6830 A500CL2 2t %

C-6830-2 A5080L2 10/

kl O C-6840-1 A533GRSCil Bottom llead lorus

-50

-10 0

m c--

2'

2 1 02 C-6841-1 A533GRBCli Bottoe llead Dome

-40 It) 20 99

~

r-m a

~

n 5

c l

c

,+

~

p-i f

i L

I k

i f

f

)

t k

ATTACHMENT B PROPOSED TECHNICAL SPECIFICATIONS AND BASES UNIT 3 r

9 b

I s

-[

El l

[

l INDEX J

LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE-HOT SHUTD0WN...........................................

3/4 4-3 COLD SHUTDOWN - LOOPS FILLED...........................

3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED.......................

'3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING..............................

3/4 4-7 3/4.4.3 PRESSURIZER............................................

3/4 4-8 3/4.4.4 STEAM GENERATORS.......................................

3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS........................

3/4 4-17 JPERATIONAL LEAKAGE..............................

3/4 4-18 i

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 LIMITS REACTOR COOLANT SYSTEM...........................

3/4 4-28 PRESSURIZER - HEATUP/C00LDOWN....................

3/4 4-32

.i OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE s 303 267oF...................

3/4 4-33 RCS TEMPERATURE > 303 267oF...................

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 C00 LING' SYSTEMS r

3/4.5.I' SAFETY INJECTION TANKS.................................

'3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,a 350aF.........................

3/4 5-3 o

3/4.5.3 ECCS SUBSYSTEMS - T, < 350oF.........................

3/4 5-7 o

3/4.5.4 REFUELING WATER STORAGE TANK...........................

3/4 5.

P

' SAN ON0FRE-UNIT 3 V

AMENDMENT.NO..

I

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 3.2-2 DNBR MARGIN OPERATING LIMIT BASED ON CORE PROTECTION CALCULATORS (COLSS OUT OF SERVICE)........................

3/4 2-8 3.3-1 DEGRADED BUS VOLTAGE TRIP SETTING.........................

3/4 3-40 4.4-1 TUBE WALL THINNING ACCEPTANCE CRITERIA....................

3/4 4-16 3.4-1 DOSE EQUIVALENT I-131 PRIMARY COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE THE PRIMARY COOLANT SPECIFIC ACTIVITY >1.0 pCi/ GRAM DOSE EQUIVALENT I-131..........................................

3/4 4-27 3.4-2 SONGSf3 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS FOR O'5" YEARS lUNTILT8]EFPYN:NORMAlf 0PERATION.................

3/4 4-30 3.4-3 C00LDOW" RCS PRESSURE / TEMPERATURE LIMITATIONS FOR 0 5 YEARS DELETED.................................

3/4 4 30; 3.4-4 SONGS 13LC00LDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS F0F "U

8"EFPY K NORMAQ0PERATION...................

3/4 4-31 3.4-5

$0NGSE.3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM. ALLOWABLE C00LD0WN RATES (

4?UNTIL8EFPY)KNORMA14LOPERATION.....

3/4 4-31a 3.4461 f SONGS 13 :LC00LDOWN ' RCS7 PR ESSURE / TEMPERATURE 1 LIM.ITATIONS

~ '

UNTILs8!E[PY P REMOTE:5 SHUTDOWN (0PERATION; n.. m.: g g.y @ /4T4i31b 314?7J

'50NGST3?RCS?: PRESSURE / TEMPERATURE! LIMITS" MAXIMUM l ALLOWABLE

~C00LDOWNLRATESf(UNTIls8:EFPY)MREMOTE' SHUTDOWN lf0PERATIONH3/64f31c 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 GASEOUS EFFLUENTS.......................

5-4 5.1-4 SITE BOUNDARY FOR LIQUID EFFLUENTS........................

5-5 l

-SAN ON0FRE-UNIT 3 XVII AMENDMENT NO.

i

INDEX LIST OF FIGURES FIGURE PAGE 516' 1?

/ UNITS:!;2 "AND::3 JFUEU: MINIMUM;BURNUPlVS]? INITIAL ENRICHMENT {: F.0R ; REGION.111 :- RAC KS..;............. gJe.; c. 2. g... 1. 5.12 5;6-2;::

E UNIT 311 FUEE MINIMUM TBURNUP)VSLJ INITI AL

' ENRICHMENT./ FOR.i;;REGIONi II o RACKS C.:...~..1..........-2:_......'ec.! 5-13 5.6-3 :

L. FUEESTORAGE PATTERNS {FOR" REGION (I D RACKSh......~..... :...

15;14 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.........................................

6-4 6.2-3 CONTROL ROOM AREA.........................................

6-6 t

I T

1 l

SAN ON0FRE-UNIT 3 XVIIa AMENDMENT NO.

t

F i

INDEX LIST OF TABLES TABLE PAGE 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS..............................................

3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS................................

3/4 3-58 3.3 12 "ADICACTIVE LIQUID EFFLUE#T MONITORINC INSTPUMENTATION DEETE41-4.3 8 RADI0 ACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENTATION SURVEILLANCE REQUIREMENTS DELETED 3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION...

3/4 3-66 4.3-9 RADIOACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................

3/4 3-68 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-I 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 gg REQUIREMENTS..............................................

3/4 4-23 e

4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND? ANALYSIS P R 0 G R AM.................................. '. '......... '.......

3/4 4-26 4.4-5 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM -

WITHDRAWAL SCHEDULE.......................................

3/4 4-29 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE.........

3/4 4-31bd 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 VALVES..............................

3/4 6-21 3.7-1 MAIN STEAM SAFETY VALVES..................................

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

3/4 7-3 SAN ON0FRE-UNIT 3 XIX AMENDMENT N0.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.8.1 With the reactor vessel head bolts tensioned *, the Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.1 3, 3.4-4, and 3.4-5p3..4 6;fandC3)4-7 during heatup, cooldown, criticality, and inservice leak and hydrostatic ~ testing with:

a.

A maximum heatup r specified by Figure 3.13s60[50*F in any 1-hour period with RCS cold leg temperature less than'^4&Uor7e A maximum heatup of 60of in any 1-hour period with'RCS"qualttoil59of.

cold' leg temperature greater than M31159aF.

b.

A maximum cooldown as specified by Figure 3.4-5 in any 1-hour period with RCS cold leg temperature lest than 4E61or}equalltol1720F. A l

maximum cooldown of 1000F in any 1-hour per'iod with'RCS" cold leg temperature greater than 4467172ef.

l c.

A maximum temperature change of 10*F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup.

and cooldown limit curves.

I d.

A minimum temperature of 86oF 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 60of in any 1-hour period.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural' integrity.

of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operations or be in at least HOT STANDBY within the next-l 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T and pressure to less than 200af and 500 psia, respectively,withinthefollyowing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

~

less than 86aF.

D SAN ON0FRE-UNIT 3 3/4 4-28 AMENDMENT N0.

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM b1"! TING CONDITION F0" OPEP^JION SURVEILLANCE REQUIREMENTS 4.4.8.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at leart 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 Figuresi; 3.4-2 and ar4-37314341throughf3f4h7. Recalculate-the Adjusted Reference Temperature bachd'6r?th5"gicite?~cf the fc11cv.inggin4ccordance%ithiRegulatorf Guide 11199RReVi' ion?2T"Radihtion:Embri ttlemen_t( of L Reactor:. Ves s eh Mate ri al s," ~ ~

s Mayl;1988.'

~~

~~'

2.

The actual shift in reference temperature for plate C 5802'1 :

determined by impact testing, or b.

The predicted shift in reference temperature for wcld scam 2 203,^,

2 203B, or 2 203C a; determined by Regulatory Cuide 1.99, Revisica 2,

" Radiation Embrittlement of Reactor Vessel Material;," May 1988.

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

1 4

1

l! ;;! l!! ! ! l ! t l! !!

l 1 ' ' ! l ?'!

I'

3500 LOWEST SERVICE

INSERVICE TESTS

HEATUP TEMP = 209 F 4

3000

. -.i.

...I...

...;. j..;....... ; ;......,

i I i....!.

F..l....;

i.. 4.i...._ a_.4. i.....&...;...&...&. i..

'.1............43..

j j

t i i

' ' ' ' l ji! '

i '

q '

...... 4..- j.. 4.. 4. 4..j... i... !.. 4.. '...j...j.. 4...t. 4 4.. 4.. 4... <...&

t.. 4....

^

j......i...i.. 4...

4 8

U)

_. 7,..

..y.

.. ~...,......,

..y.

.. 9........ -.

t i

2500

' -i-m m

Acceptable operating region - to the

{..... +. 4...+.4...

. a..!..i.. !

...~

i :.

U ri9ht of the inservice tests curve

.. i.. i u)

(Applicable in modes other than r ~ '! '"' '"~ '!..! "'*! '.'.' '. ",'.."..;...'.'.,;." ". 'i' '...

q) 1...

g Modes 1 and 2) i ' i

...9...-

y

.c

.{...

'...p.

T

...i..

a.c 2000 - # Acceptable operating region - to the

++4

.i_j.....t...p ; ;...

W right of the heatup curve in all modes.

..J..j...;...;..!...!

3...;......;...;..!. j...

i ' l l N

in addition,in Modes 1 and 2 operating a, ' ' ' ;

' !i

.4._..r.....

g region to the right of the core critical

, i.

-.3

~'i'" i' tr' i' '*~ T * '~'" " ~

e.

~8'

- - E'* E -- b ' '!---

- + - * -

U)

...................,..............................................{...t'55- ! 'i i ! ' ' 4.9..'

. 4... e,i... ' '. ' ' : 8

i W

i : i : ' i 1_

. 4...

4.~.

... ;. 4 4..

g j$0g

....m.....

4..,...#...,...

%... e #

8 i

a.

. 3..

.}...

C3 f

y

.g.

..p.

.g..

_.7..

4...,

.4.

..L

.4.,4....+..a

........?..r....!

! l...

i I:

F-

......t

....,... 7..

,~

4

......y...,...

9

. 4-..j

. p..?. 4. j.- [..

.h. 4. 4.... i. y,,. ;.. l 1000

... j.

..g...i

.......:...a...

i i

a..

i i

i r

... i... 4... 4... 4...... 4..........+... l... 4... 4,... i.... i. 4... 4...

......4....i....'..i....a....+...

i 3....;...

,a..a.. 4..;.y...j.+_i 4. 4. ; 4. 4.. 4 4 4-.;.

4.. :..... 4.. ; y f# CORE CR1TICAL...

i i

l i j '

i e

4 4

.e

.J MINIMUM t-

- +

-t '- + 4-I h-i- - -i-

+-f -

l BOLTUP

. !t..

..j... ;.

),.. ;.". !. [. i. !

.I TEMP = B6*F

.. 5.,

.~.L-..

..................... r.

j j 7

g g.

g

. l i i i iil i i i il i i i e l

l i ji Iie e i i i l i i i i l e 6 i a 15 0

50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE ( F)_4T.ci_

l FIGURE 3.4-2 l

SONGS 3 HEATUP RCS PRESSURE / TEMPERATURE LIMITATIONS F0" 1 UNTIL 8 EFPY NormaD 0peration SAN ONOFRE-UNIT 3 3/4 4-30 AMENDMENT N0.

i l

i i

1 i

4 1

i 1

(Figure 3.4 DELETED)

I l.:

l

!~

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

i f

'?

l ?

l l ! ! ! ! l I

'l

! !.I ! l 1 !

I l 1 1 ! !

LOWEST SERVICE COOLDOWN

~

TEMP = 209'F j

g

. i -

[

3000 --l l i i

1 1

....I t

i 2

.J,..

1..

.a..7...a...+

-7.

- -... g -

'.. ]...q._

s7 6

a _.A.

...p

.,1-,.:_.g m._..

m i

a......i 4. 4-4.d.

i i,

- i o. 4 4...i. a e... s..._............l...;.J...

....E...+.....+,.....4.4..4-.i.

O l

i 8

i 3

i E

. +.d...jl-pl, i

3 i

8

' i i i ! !

ll

.j _.p... l. 4._4

.+.4... i. ?

4.d..... -4 4.j}.d d. j.4-.i d.a.d._.l

~j O

w L4.!

_._ai i i!--

E i ! !

--i---t-t-p! y ----

3 2500 -r+d

... 7.... g._.....i 7.i.. e i- -

d-- t 1 "l j, I i !

I m

I i

_.p.4...L.

. 2. L+. 4._2 4 4_.4_d_4 4. d..t t J_1.ij.j_.j' j j t -+ *T

i { e t j L t + + + P-r-

g p

f.. j.

s.

T.. * ~+-F-~.~-+---'~F'

'I ul

. 4_

(L i

i gn

...j.....;...

.t

.}.....

}-

._.._y.._

..d..#_

w i

i :

i i __.. i..

C

...[_..1_3 a._ ;.

.ai 2 4 a1.._

+.+..

r i

i i

-+ t i d +. - +- t P +} j f

- t d

2000

' -+ ' *4 + f +! I. f f r-t + -+ - +' +i.e..i _. !. !.. a.i.a-4_.,. ;..,

-t-i 4.......{

. t. 4 ;......

!{

i _.,!

i T

. _....,.._..,._.....,._.y...

4.-

O

.......;..4...

g..

I i

e I

...........,..., _ p... p...i._. q_,.....; 7._.p

.. i

.r....I....re.. p.p.p-p,{..

O

..i.;u.,i.

w c

i

....!..t....i....!....i. l i ;

i i ; t~e" "

1 1

1 Unacceptable "

~"i~t~+ i-t" t"y~i-4

-i--Q.

o 1500 i---t. --i "--

Operating r

w l

f r

.......t...

Reg. ion j

g<

d_1.~j..

~ ;

Acceptable !.-]1...d_..J_.

l._ _ _.

ia. 4.a.w...._,.a.

_e

... v._;.a.a.a.....,i...i..i..g:....I....! _. !.,i :

i j

i j-l j 9

i....+.,._.~.'..#.4.,.

9.....i....i...,i }i

s. i. :

Operating.

4

+.4. 4-4. 4.

4.

i O

i, t i i T

1

_r_

-r e, y-Region

.l...T. 'E 8..... l-1000

-4 4..d

J t._.l.a-l.a.a l

}...

+ -.

1. 1 i 5

t _.. '. _

t..j.p2.a.s-2.a_4_

_ i a

.a.

...;_....+.+. i...;i i..l _.i._.l..i._.....

i!

...t 1._

a. 4. 4

.t...i_4.

_...#J'...p..#........_.l..t-....t..... ;t i I

!.L

' - - ~-

1-l

_[11;

.i _

....! J..,J_.i,. i...._...

3 u

j p

. j 4_ j,

y p%

I - I i,

....i...f...p 9..

4 I

h 4"- 4-- 4-t- - ; 4 -r e h -g]- p + 7 q y +-e -+-7 e-y p-t h j y- +- -

500 3

I I

~+ + -+ +- -- + 4* f 4 4 + -+-+-+ +{ 4 + + 4 4 4

+g I. - -

4 J

4 l -+i~4 -+ -

+ +-+ + t *d..d d...! d,'...-

MMWM

! d_..!

BOLTUP j ld1 L... 4. !

I I-TEMP = 86*F i.

l: '

j

.........44.....4...........4_+,.........i......!..4.44_.......

r 4

4...

i1i..i.ii.i.ii1 l.i.II'li4'i

i i ;

l 1

'..i1iiii i

l.:

i i

15 1

0 50 100 150 200 250' 300 350' 400

'l INDICATED RCS TEMPERATURE ( F)%_TS, 6

b FIGURE 3.4-4 SONGS 3-C00LDOWN RCS PRESSURE / TEMPERATURE LIMITATIONS *FOR 1 JUNTIUI 8 EFPY Noma.WuOperation"

. + -

I i

SAN.ON0FRE-UNIT'3' 3/4 4-31 AMENDMENT NO.

1 t

.i i

-I a

a i

W W

u 5

8 8

8 g

g g

l a

(

~

110 -

100 -

,o.

~

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a 8

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3o go 10

~

i i

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o 80 90.100 110 120 130 140 150 160 170 180 190 200- 210' INDICATED RCS TEMPERATURE ( F)hTM NOTE: 'A MAXIMUM C00LDOWN RATE OF-100aF/HR IS ALLOWED AT ANY TEMPERATURE AB0VE 4248F.172 7 -

FIGURE 3.4-5

. SONGS 3 RCS PRESSURE / TEMPERATURE LIMITS MAXIMUM ALLOWABLE-COOLDOWN RATES-(' " JUNTIL8EFPY)-

Nonuil2I0pgfatjpn i

' SAN ONOFRE-UNIT;3 3/4 4-31a AMENDMENT N0.

3500

,.g..,.g....g....g.;;;i.,;.g;,,.

LOWEST SERVICE COOLDOWN TEMP = 209'F s

e-

- i 4-

- v +- L -L t -

3000.

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9.. 9.. 9......9.....9..

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50 100 150 200 250 300 350 400 INDICATED RCS TEMPERATURE (*F)-Tc FIGURE 43.4-6 SONGS;3 COOLDOWNlRCS" PRESSURE / TEMPERATURE "LIMITATIONSsilNTIE 8; EFPY*

Remote Shutdown Operation SAN; 0N0FRE-UNIE3/

3/4 4-31bf l AMENDMENT.NO.

N I

120 s

s I

a s

s a

110 100 f

P m

E r

2 it e,

w n

<m z

a 3:

og 2

8o e

30 20 10

' ' I

'I

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

0 80 - 90 100 110 120 130 140 150 160 170 180 190 '200 210 INDICATED RCS TEMPERATURE ( F)-Tc s

NOTE C Ai MA.XIMONiC00LDOWN -i. RATE: : 0F:s100*F/NRf!S? ALLOWED -

-~

m-m_

m y

v -.

s SONGS;?31 RCSf PRESSURE /TEMPERATUREJ LIMITS.

f. MAXIMUM:l ALLOWABLE;COOLDOWNERATES((UNTILi ?EFPY) 8

~ Reinotei Shstdownl Operation ' '

t SAN [0N0FREv0NITt3;)

lg/4L4331cy fAMEN0MENTiNO) y

r Table 3.4-3 Low Temperature RCS Overpressure Protection Rance Operatina Period. EFPY Cold Lea Temperature, oF During During Heatup Cooldown 4-t#Until-8 (Normalf 0peration) s303j26,7-5 2 @ 250 Until 8L (Remote Shutdown 10peration)i J5 L.250

~

i

~

Hea' tup op'erations"are'n'otfnormallyf psfformediffomitheTRem6teFShutdoishT el s; '~

~

^

SAN ONOFRE-UNIT 3 3/4 4-31b-d, AMENDMENT NO.

'l

b

' REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE 5 M2 267aF LIMITING CONDITION FOR OPERATION 3.4.8.3.1 At least one of the following overpressure protection systems shall be OPERABLE:

a.

The Shutdown Cooling System Relief Valve-(PSV 9349) with:

1)

A lift setting of 406 10 psig*, and t

2)

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

-i 3HV9378 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:

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

t 200aF, depressurize and vent the RCS through a greater than or. equal r

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.

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) to less than 200aF, depressurize and within 7 days or reduce T,y,ter than or equal to 5.6 inch vent within vent the RCS through a grea j

the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .'

q 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 t

recurrence.

d.

-The provisions of Specificatio'n'3.0.4 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated'0PERABLE by:

i a.

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

'The lift setting pressure applicable to valve temperatures of less than or equal to 130eF.

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

~

i i

REACTOR COOLANT SYSTEM

.)

OVERPRESSURE PROTECTION SYSTEMS j

RCS TEMPERATURE >30I 267eF

LIMITING CONDITION FOR OPERATION 3.4.8.3.2 At least one of-the following overpressure protection systems shall be OPERABLE:

a.

The Shutdown Cooling System Relief. Valve-(PSV 9349) with:

1)

. A lift setting of 406 10 psig*,.and j

2)

Relief valve isolation valves 3HV9337, 3HV9339,'3HV9377,_and 3HV9378 open, or, 1

b.

A minimum of one pressurizer code safety valve with a lift setting of._

i 2500 psia 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 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

In the event lthe SDCS Relief Valve or an RCS. vent is used to mitigate l.

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 OF RCS vent on the transient and any cor.rective action necessary to prevent recurrence.

SURVEILLANCE-REQUIREMENTS

.q 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-the SDCS Relief Valve isolation valves 3HV9337, 3HV9339,'3HV9377 and 3HV9378-are open when i

the SDCS-Relief Valve is being used.for. overpressure l protection.

[

i b '.

Verifying relief valve setpoint-at least once per 30' months when tested pursuant 1to. Specification 4.0.5 4.4.8.3.2.2 The' pressurizer code: safety valve has no; additional surveillance:

I

~

requirements' other. than those required by.. Specification 4.0.5.

la

' " 8.3.2.3 The RCS veni shall be verified to'bc cpen at _lcast once per 12l,.

hour: eten the vent i; being used for everpressure protecticn, except when the.

vent pathway is provided with a valve which is locked, scaled, cr Otherwise.

secured in the Open pccitien, then ver4fy the c valve; cpen at least once per 31 daysr f

' The lift. setting pressure applicable. to valve temperatures of lesslthan--or equal;to'130of.

- The lif t setting pressure; shalll correspond -to ambient _ conditions of the valve j

at nominal operating temperature and pressure.

SAN,ON0FRE-UNIT 3 3/4.4-35' AMENDMENT NO.

W a

.. -a- -

_ _ _ _ -__ - _ _. _m____.;.__

F-REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

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 60oF/hr or cooldown rate of up to 100oF/hr. Thel limi t curves for Remote Shutdo'wn-. operation are. determined usin'g the. Total Loop Uncertainties (TLUs)~ for. temperature:and: pressure for. the. Remote. Shutdown Panel instruments in which the pressure TLUs are higher than those-for' the Control' Room shutdown instruments.

The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the service period indicated on Figures 3.12 and 3.4 3, and'they include adjustments -for instrument uncertainties, and: static and dynamic heads.

The reactor vessel materials have been were tested prior to reactor startup to determine their initial RT The results of these tests and the updates g7 resulting from the evaluation of material p'roperties in response. to Generic Letter 92-01,'" Reactor Vessel. Structural Integrity," Revision 1 are shown in Table B 3/4.4-1.

Reactor operation and" resultant fast' neutron'(E greater than 1 MeV) irradiation will cause an increase in the RT Therefore, an adjusted g7 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 Vessel 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 g7 adjustments for possde errors in the pressure and temperature sensing instrument + instrument ' uncertainties, and static and dynamic heads.

The actual shift in RT of the vessel material will be established g7 periodically during operation by removing and evaluating, in accordance with ASTM E185-73 and 10 CFR 50 Appendix H, reactor vessel material irradiation l

surveillance specimens installed near the inside wall of the reactor vessel in the core area.

The surveillance specimen withdrawal schedule is 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 of the Lead Factor.

The heatup and cooldown curves must be recalculated when the delta RT determined from the surveillance capsule is different from g7 the calculated delta RT for the equivalent capsule radiation exposure.

m The pressure-temperature limit lines shown on Figures 3.4-2 and 3.1 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.

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

i

~

REACTOR COOLANT SYSTEM l

BASES PRESSURE / TEMPERATURE LIMITS (Continued) l The maximum RT for all Reactor Coolant System pressure-retaining e1 materials, with the exception of the reactor pressure vessel, has been determined to be 90of.

The Lowest Service Temperature limit line shown on Figures 3.4-2, 3.4-4, and 3.4-36 is based upon this RT since Article NB-2332 (Summer Addenda of' 1972) of Section III of the ASME Boilorer and Pressure Vessel Code requires the Lowest Service Temperature to be RT

+ 100oF for piping, pumps and valves.

g7 Below this temperature, the system pressure must be limited to a maximum of 20%

of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The -Low Temperature:0verpress' re1 Protection.J(LTOP)Menable l temperatures?are~ based u

upon the. recommendations of' NUREG-0800 Branch : Technical-~ Position -(BTP)L RSB 5 2,~

Revision l1, ".0verpressurizationLProtection of.. Pressurized Water Reactors $hile Operating-at Low Temperatures." 'BTPzRSB 5-2,(Revision'11 defines.the. enable J

temperature asl"the; water temperature 'correspondinoito a metal. temperature"of"at least RT,

+ 90oF1at the~ beltline location (1/4t or 3/4t) that:is controllingjin theAppen$ix'G'limitcalculations.;"

l I

)

SAN ONOFRE-UNIT 3.

.B 3/4.4-7a' AMENDMENT.NO.

TABLE B 3/4.4-1

[

REACTOR VESSEL TOUGHNESS E

Temperature of Minimum Upper y

Drop Charpy V-Notch Shelf Cv energy Weight 0 30 0 50 for Longitudinal m

Piece No.

Code No.

Material Vessel Location Results ft - lb - ft - Ib Direction-ft Ib i

l C

L 215-01 C-6801-1 A533GRBCL1 Upper Shell Plate

-20 28 64 115 215-01 C-6801-2 A533GRBCL1 Upper Shell Plate

-20

-6 34 106 215-01 C-6801-3 A533GRBCL1 Upper Shell Plate

-20 18 36 115 g

215-02 C-6802-4 A533GRBCL1 Lower Shell Plate

-30 3240 6270 1168 215-02 C-6802-5 A533GRBCL1 Lower Shell Plate 0

3640 6470 1106 215-02 C-6802-6 A533GRBCL1 Lower Shell Plate

-40 3240 40080 9092 215-03 C-6802-1 A533GRBCL1 Intermediate Shell

- 2010 66110 400135

% 94 215-03 C-6802-2 A533GRBCL1 Intermediate Shell

-20 40 6670 1135 215-03 C-6802-3 A533GRBCL1 Intermediate Shell

-10 4460 80~

1015 4

203-02 C-6823 A508CL2 Vessel Flange Forging 0

-30

-15 NA 209-02 C-6824-1 A508CL2 Closure Head Flange

-40

-100

-100 NA Forging 205-02 C-6829-1 A508CL2 Inlet Nozzle Forging 10

-35

-5 109 205-02 C-6829-2 A508CL2 Inlet Nozzle Forging 0

-55

-35 156 205-02 C-6829-3 A508CL2 Inlet Nozzle Forging 10

-25 35 112 205-02 C-6829-4 A508CL2 Inlet Nozzle Forging 10

-30 25 108 205-06 C-6830-1 A508CL2 Outlet Nozzle Forging

-10

-30

-15 125 205-06 C-6830-2 A508Cl2 Outlet Nozzle Forging

-10

-20

-5 131 3

232-01 C-6840-1 A533GRBCL1 Bottom Head Torus

-50

-10 0

107 9

232-02 C-6841-1 A533GRBCLI Bottom Head Dome

-40 10 20 99 E-9 5

l l

DV":'

a

~,.

a l

e t

r ENCLOSURE 3 TECHNICAL SPECIFICATION PAGES CONTAINING THE CHANGES WHICH WERE PREVIOUSLY REQUESTED IN AMENDMENT APPLICATION N0. 97 (PCN-358) DATED DECEMBER 20, 1991,.

AMENDMENT APPLICATION N0. 101 (PCN-354) DATED SEPTEMBER 3, 1992, AND ARE BEING REQUESTED IN THIS LICENSE AMENDMENT APPLICATION NO.102 (PCN-359);

SAN ON0FRE UNIT 3

.o

V INDEX LIST OF TABLES TABLE PAGE l

4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE i

REQUIREMENTS..............................................

3/4 3-55 3.3-11 FIRE DETECTION INSTRUMENTS................................

3/4 3-58 3.3 12 PADMACTIVE LIQUID EFFLUENT "0NITORINC INSTRUMENTATION DELETED 6

P 1.3 8 RADICACTIVE LIQUID EFFLUENT MONITORINC INSTRUMENTATION f[;

SURVEILLANCE REQUIREMENTS DELETED S

3.3-13 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION... 3/4 3-66 p

4.3-9 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.................................

3/4 3-68 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION......................................

3/4 4-14 4.4-2 STEAM GENERATOR TUBE INSPECTION...........................

3/4 4-15 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES..........

3/4 4-20 1

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 gupP 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE ANDLANALYSIS 10 PR0 GRAM...................................................

3/4 4-26 ge g 4."

S REACT 0P VESSEL MATERIAL SURVEILLANCE PROGR^,M P'}

WITHDRAP"L SCHEDULE.......................................

3/1 ^ 29 19V 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE.........

3/4 4-31bd 18, al g'3 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 VALVES..............................

3/4 6-21 3.7-1 MAIN STEAM SAFETY VALVES..................................

3/4 7-2 3.7-2 MAXIMUM ALLOWABLE VALUE LINEAR POWER LE/EL-HIGH TRIP WITH I

INOPERABLE MAIN STEAM SAFETY VALVES DURING OPERATION WITH BOTH STEAM GENERATORS.....................................

3/4 7-3 4

SAN ON0FRE-UNIT 3 XIX AMENDMENT NO.

l i

i

c:

REACTOR COOLANT SYSTEM 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM i

pp LIMITINC CONDITION F0n OPERPIO" pcp 5A 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 shall be removed and examined, to determine changes in material properties, at pg

'3 Y 5

the interval fas required by 10 CFR 50 Appendix H. in accordance ith the sehedule in Tdble 1.1 5 The results of these examinations shall be used to update Figures 3.4-20andh4453L4.4fthFoughP354Q.

Recalculate the Adjusted Reference Temperature" based en~ thEgrdath"r 'c'f"thd "follec.ing:iin f accordance:with Regulatory Gui~ dell!.993 Revision?2i "Radi_~ tio(Embrittlementf of Reactor? Vessel g

a Materials,"1May11988:'

Wr-c.

The actual shift in reference temperature for plate C 5802 1 as fg-determined by impact testing, or b.

The predicted shift in reference temperature for cicld scam: 2 203A, 2 203B, Or 2 203C as determined by Regulatcry Guide 1.99, Revi;ien 2,

" Radiation Embrittlement of Reacter Vessel Materials," May 1988.

-l 1

~

SAN ONOFRE-UNIT 3 3/4 4-28a729 AMENDMENT _NO.

3ff

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS SMW >

RCS TEMPERATURE 5 M2 267oF LIMITING CONDITION FOR OPERATION pW ' '

3.4.8.3.1 No mo're' than !twoThigh? pressure; safety 1 injection pumps. shall. be ggg OPERABLE and'at'least'one^of'thefollowing overpres'sure' protection systems shall be'0PERABLEi a.

The Shutdown Cooling System Relief Valve (PSV9349) with:

1)

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

Relief valve isolation valves 3HV9337, 3HV9339, 3HV9377, and 3HV9378 open -orr r

or, b.

The Reactor Coolant System depressurized with an RCS vent of greater e

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 en~able temperatures specified in 1

Table 3.4-3; MODE 5; and~ MODE'6 Mth when'the reacter vessel t

head is on the? reactor: vessel?a~nd: the RCSiisEnotlvented; SN ACTION:

a.

With the SDCS Relief Valve inoperable, reduce T to less than g

2000F, depressurize and vent the RCS through a greater than or equal to 5.6 square inch vent within the next 8 hours9.259259e-5 days 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. opensthe ~other SDCSTRelief Valvelisolationi valvef pair within ' &ys 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 'or~redude"T /to less than 200oF,'

g 3

i depressurize and vent the RCS througfl 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 twolhigh fres'sursIshfety? injection [ pump's' OPERABLE, k>

securerthetthird high-pressure:-safety. injection ~piim:Fbyf racking 2out

,gg its motor circuit' breaker orLlocking'closeJ its? disc 1argel valve within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

^~

The lift setting pressure applicable to valve temperatures of less than or equal to 130oF.

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

o REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION ACTION: (Continued)

V' d.

In the event either the SDCS Relief Valve or an RCS vent is used to yff 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 recurrence.

lfCV e.

The provisions of Specification 3.0.4 are not applicable.

3ff SURVEILLANCE REQUIREMENTS 4.4.8.3.1.1 The SDCS Relief Valve shall be demonstrated OPERABLE by:

a.

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

b.

Verifying relief valve setpoint at least once per 30 months when tested pursuant to Specification 4.0.5.

4.4.8;3.1.2f

. i AttleastLonce? per?l2 hoursMthelthird high-pressureis~afety injection Lpump. shallibe " demonstrated L to f be(secured by!. veri fying; that:4its motor circuitLbreakerTisinottracked-in or:itsidischargeJvalveJisi ockedlclosedh The l

requi remento to" rack: out;;J thef thi rd L HPSI! pump 1breakerii.si sati sfied iwith jthel pump break. erLracked outT to its.idi.sconnec'ted critest' position'-

~~"

VAtfleastionce perl:12' hours,dthe':0PERABLE5SDCS7 Relief? Valve 4.4'8.3.113:1 isolation valveJpair? (salvelpair 3HV9337, and L3HV9339% ori. valve:. pair /3HV9377 Land 3g 3HV9378) thatJis-usediforloverpressure:protsction duelto.:the;other:SDCS Relief i

Valse 1 isolation valvefpair beingilN0PERABLE shall:be4 verified:to be in the' power-lock?open condition 1untilsthe!INOPERABLELSDCS Relief Valvetisolation valve pair is t returned to'0PERABLEistatu'sior the1RCS sisidepressurized and vented. The~

e 3

power-l_ock open requirementiis) satisfied?eitherc:with:the1AC breakers o'entfor p

valve pair l3HV93371and 3HV9339 or the; inverter?inputiand output 1breakersLopenffor valve pair 3HV9377:an'd 3HV9378, whichever: valve pairlis ' OPERABLE!

4.4.8.3.1.4 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 secured in the open position, then verif t these valves open at least once per 31 days.

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

S h

.I

REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The heatup and cooldown limit curves for'normalj ' peration (Figures 3.4-2 uf o

and 3.4-34) and the' cooldown:limiticurve for remote shutdown operativn- (Figure 3.4-6) are composite curves which were prepared by' determining the'most 9

~

conservative case, with either the inside or outside wall controlling, for any heatup rate of up to 60oF/hr or cooldown rate of up to 100cF/hr. The limit

%g curves for Remote Shutdown operation are determined -using:the Total. Loop Uncertainties (TLUs)lfor temperature and' pressure for:the Remote Shutdown ' Panel instruments in which the pressure 1TLus-are hiper than those~for. the: Control Room Phg,

}

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 4ndicated on Figures 3.12 and 3.' 3, andithey include adjustments for instrument uncertainties, and1 static and dynamic heads. '

pp The reactor vessel materials have been were tested priorL o reactor startup t

to determine their initial RT The results of these tests and'the. updates in N,d response to Generic Letter 92 dl, "ReactorLVessel; Structural: Integrity,"TRevision 3

un.

1 are shown in Table B 3/4.4-1. React'or operation and' resultant fa'st neutron"(E greater than 1 MeV) irradiation will cause an increase in the RT Therefore, ug7 an adjusted reference temperature based upon the fluence and copper and nickel 9

content of the material in question, can be predicted using FSAR Table 5.2-5'6 puf and the recommendations of Regulatory Guide 1.99, Revision 2, " Radiation yocd' Embrittlement of Reactor Vessel Materials." The heatup limit curve-(Figure 3.4-(sd

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 possig1me errors in the pressure and temperatwe su F sensing instruments instrument. uncertainties, and2 static and dynamic heads.

D, l

3d j

The actual shift in RT : of the vessel material will be established sn f9 periodically during operation by removing and evaluating, in accordance with ASTM l~

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 1

/

core area. The surveillance specimen withdrawal schedule is shown in-Table 1."

5 p:64

/"

maintained in'the FSAR.

Since the neutron spectra at the irradiation samples and 3

vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel taking into account the location of the sample closer to the core than the vessel wall by means of the Lead Factor. The heatup and cooldown curves must be recalculated when the delta RT determined from the surveillance capsule is s37 different from the calculated delta RT for the equivalent capsule radiation s31 exposure.

[/r The pressure-temperature limit lines shown on Figures 3.4-2 and 3.13 for reactor criticality and for inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements of gc%

Appendix G to 10 CFR 50.

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

. _ _ _ _ _ _ _ _ _ _ __

ML20036A037

Text

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