Text

Proposed Tech Specs Re Overpressure Mitigation Sys
	ML13329A726
Person / Time
Site:	San Onofre
Issue date:	05/13/1991
From:	; SOUTHERN CALIFORNIA EDISON CO.
To:	;
Shared Package
ML13326A771	List: ML13326A770; ML13329A726;
References
	NUDOCS 9105170311
	Download: ML13329A726 (37)
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nOI Fig. 1 RETRAN Nodal Scheme For Calculating Criticlo Flow Rate TDV TDV 1

(1)

Time-dependent volume (TDV) is a boundary volume in which the fluid conditions are specified as a Function oF time.

(2)

Critical Flow model, flow area and dischorqe coelFicient are specified at Junction 1.

Fig. 2 SONGS 1 RETRAN EMS Anolysis Moolel 3

PIoIRV PZR 2

4 Pv s

D RCS 1

Chorging Flow

Fig. 3 SONGS 1 RV-206 Water Flow Discharge Versus Temperature RETRAN Input : upstream pressure = 580 psia valve flow area = 1.28 sq. inch discharge coefficient = 0.75 600 isoenthalpic expansion choking model 550 500 N'

450 E

2 400-*

o 350-*

x 300 250 200 150 A

RETRAN 0

100 Data (Ref.)

50 0

50 100 150 200 250 300 350 400 Subcooling, F Ref.: Table 1, Flow Testing of Crosby 2 1/2 J4 JB-35-TD, Type E Relief Valve, Test Report Number 4637, July 10, 1990

VIII. ATTACHMENT - Listing of RETRAN Input Data for Calculating Critical Flow Rate - Listing of RETRAN Input Data for SONGS 1 OMS Analysis Model 13

STING OFINP*

A FOR CASE 1

1 RETRAN CALCULATION OF CHOKED FLON 2

3 4

5 1

1 1

1 6

1 TDV 81 1

JUN 1 1

TDV 82 1

7 1

P = 580 PSIA 1----------------->1 1

8 1

T m 90 F 1

C a 0.75 1

1 9

M 1

1 0

1 1

10 w

T

= 482.6 F)

A v 1.28 SW.IN 11 SAT C

12 13 14 15 w

16 w

CODE VERSION: RETRANOt-HOD4 U

17 w

U 18 19 20 21 22

--. w.en-neMMM PROBLEM CONTROL AND DESCRIPTION DATA O1000Y U******U****

23 24 25 Y----Y---1-----2-----3-----4-----5-----6-----7-----8-----9----10---

26 27 010001 0

6 2

1 2

0 2

1 0

0 28 0

29 30 -

11----12----13----14----15----16----17----18----19----20---

31 U

32 010002 0

0 0

0 33 U

34 U

35 21----22----23----24----25----26----27----28----29----30---

36 U

37 010003 0

0 0

0 1

0 0

0 38 39 40 K--------------31----32----333----34----33----36----37 38 ---- 39----40---

41 42 010004 0

0 1

0 0

0 43 U

44 U

45 46 M

MINOR EDIT VARIABLES -

02000Y UeweenIwwwwwwwwwUns 47 U

48 U

Y 49 K

50 020001

NPUU, 1

ICHK, 1

PRES, 1

TEMP, 1

PRES, 2

TEMP, 2

51 U

52 U

53 M

TIME STEPS DATA -

03KXXO

- - M**********n 54 K

55 56 57 U

NHIN NMAJ NDMP NCHK DELTH DTHIN TLAST 58 IOL

59 030010 @ 20 100 9999 0

0.01 0.0 10.0 60 030020 50 500 9999 0

0.01 0.0 1000.0 61 62 M

63 M

64 if

IMifififime******

TRIP CONTROLS -

04XXXO ***wwwwwwwwwn

ifn 65 66 M

67 i

68 M

TRIP SIG 69

4XXXO ID ID IXI IX2 SETPT DELAY 70 71 040010 1

1 0

0 1.0 0.0

PROBLEM RUNNING TIME 72 i

73 i

74 ifn e

VOLUME DATA -

05XX Y (Y91) afs*ftnianfeminnfnfan 75 76 77 i

78 i

1 2

3 4

5 6

7 8

9 10 79 Y

IS IR P

T X

VOL ZVOL ZH FLOL FLONA 80 i

(H) 61 050011 0

1 0.0 0.0 0.0 1000.0 25.0 25.0 50.0 10.0 82 050021 0

2 0.0 0.0 0.0 1000.0 25.0 25.0 50.0 10.0 83 ASSUMPTION --------

84 a

85 i-----

VOLUME DATA -

05XXXY (Y=)

86 87 i

11 12 13 14 15 16 88 V

DIAMV ELEV INEQ VRAIN VLHTC MESH 89 M

90 050012 10.0 0.0 91 050022 10.0 0.0 92 IASSUMPTION) 93 i

94 i

95 i

96 e

TIME-DEPENDENT VOLUMES 107XXYY) ife*iwinf*w*fenfn 97 i

98 a

99 i

N TIME PRES TEMP AVG.

MIXL 100 (SEC)

(PSIA)

IF)

X (FT) 101

f 102 070101 2

0.0 580.0 90.0 0.0 25.0 103 070102 1.OE+6 580.0 90.0 0.0 25.0 104 i

105 070201 2

0.0 25.0 70.0 0.0 25.0 106 070202 1.OE+6 25.0 70.0 0.0 25.0 107 M

108 M

109 i

110 JUNCTION DATA 08XOXY (Yl)

- if*****ififi*

111 if 112 113 i

114 i

1 2

3 4 5

6 7

8 9

10 115

XXXY VI VO IP IV HP AJUN ZJUN INERTA FJUNF FJUNR 116 i

117 080011 1

2 0 0 0.0 8.8889E-3 25.0 1.0 0.0 0.0 118

(

11.28 IN*IN) <-------

ASSUMPTION ------- >

119 N

120 nn JUNCTION DATA -

08OXXXY (Y=2) 121 N

122 N

123 N

11 12 13 14 15 16 17 18 19 20 21 124 Y JVERT CHOK JCA MIX DIAHJ CNTR REGH 2PHS ANGL INDEX ISP 126 080012 0

1 0

0 0.0 0.75 0

-1 0.0 0

0 127 128 AJUN

CNTR = 8.8889E-3

0.75 129

=

u 0.96 SQ.

INCH 130 N

131 u

END OF DATA 132

STING OFINPU FOR CASE 1

1

= RETRAN ANALYSIS OF SONGS 1 OS TRANSIENTS HEAT ADDITION 2

4 5

U BASIC MODEL NEN RV206 FLON 1469 GPH) 6 w

CHARGING FLOM = 0.0 GPM U

7 U

(420 GPM) a DELTA T 5 F OR 30 MN 9

n 10 U

11 w PZR (NODE 1):

RETRAN NODING DIAGRAM 12 U

VOLUME, FT**3 = 1300 13

PRESSURE, PSIA =

415 1

1 14 U

TEMPERATURE, F = SAT JUN 3 1 NODE 3 1 15 1

1---><--->1 IPRT) 1 16 R

RCS (NODE 2):

1 NODE 1 1 IPORV) 1 1

17

VOLUME, FT**3 = 5450 1 (PIR) 1 1

18 PRESSURE, PSIA = 415 1

1 19 U

TEMPERATURE, F = 140 1

1 JUN 4 20 1 JUN 2 1

21

PRT (NODE 3) -

TDV:

1 (RV-206) 1 22 PRESSURE, PSIA = 14.7 1

1----><-----1 23 TEMPERATURE, F x 70 1

NODE 2 1

24 U

1 (RCS) 1 (CHARGING) 25 U PORV CHARACTERISTICS (JUN 3):

1 1<------ JUN 1 26 U

LIFT SETPOINT u 525 PSIA 27 U -

DELAY

= 0.6 SECOND 28 U

STROKE TIME a 1.9 SECONDS 29 CO M AC

=0.583 30 U

FLON (LIQUID)

= ISOENTHALPIC MODEL 31 U

32 U RV 206 CHARACTERISTICS (JUN 4):

33 U

LIFT SETPOINT a 515 PSIG/530 PSIA (AT THE VALVE) 34 U

VALVE FLOM AREA = 1.28 INCH**2 35 U

DISCHARGE COEF. = 0.42 36 FULL CAPACITY

= 469 GPM 37 U

ISOENTHALPIC EXPANSION CRITICAL MODEL 38 39 U

RETRAN RV-206 40 U

IPSIA)

POSITION 41 42 U

0.0 0.0 43 U

502.0 0.0 U 28 PSI HEAD BETHEEN PZR AND RV-206 44 U

503.0 0.9 45 U

552.0 1.0 46 U

1.OE+6 1.0E+6 47 U

48 U CHARGING FLON (JUN 1):

49 U

FLONRATE

= 420 GPM 50 TEMPERATURE a 145 F (MAXIMUM PRESSURIAZATION) 51 U

52 U

53 U

54 U

n 55 U

56 CODE VERSION: RETRAN02-MOD4 U

57-U n

58 U

'7

S9@

60 61 U

62 newen-een.een PROBLEM CONTROL AND DESCRIPTION DATA -

O1000Y U'swa*e*se 63 64 65 Y----Y-----1-----2-----3-----4-----5-----6-----7-----8-----9----10---

66 U

67 010001 0

-11 2

5 3

0 1

4 0

2 68 69 A

70 -

11---- 12----13----14---15----16----17

18-- 19 ----

20---

71 72 010002 3

1 0

0 0

0 1

0 0

0 73 74 U

75 U----------21----22----23----

24-29----

30---

76 77 010003 0

0 0

0 1

0 0

0 1

0 78 I

79 U

80

----------31----32----33----34----35----

37 38 ----

39----40---

81 82 010004 0

0 1

0 0

0 83 84 U

85 86 MINOR EDIT VARIABLES -

02000Y 87 U

88 U

y 89 U

90 020001

PRES, 1

TEMPs 1

PRES, 2

TEMP, 2

PRES, 3

TEMP, 3

91 020002 NP**,

1 HP**, 3 NPa*,

4

COUT,

-1

COUT,

-2 92 U

93 U

COUT,

-2

HEAT ADDITION, MEGANATTS 94 U

95 m

TIME STEPS DATA -

03)XO *KK*(0**w 96 U

97 U

98 U

99 N NN NHAJ NDHP NCHK DELTH DThIN TLAST 100 U

101 030010 10 100 9000 0

0.01 0.0 30.0 102 030020 50 100 9000 0

0.01 0.0 1000.0 103 U

104 U

105 U

106 UeUUU*UUUIEeneweenUe*se TRIP CONTROLS 04XXXO wwUo*an*

107 U

108 U

109 U

110 U

TRIP SIG 111

400(

ID ID IXI IX2 SETPT DELAY 112 U

113 040010 1

1 0

0 30.0 0.0 U PROBLEM RUNNING TIME 114 U

115 040020 2

1 0

0 0.0 0.0

FOR CHARGING FLON 116 040020 2

1 0

0 0.0 1.OE+6

FOR CHARGING FLOM

- w THIS CARD IS A REPLACEMENT CARD. M 117 118 040030 3

4 1

0 525.0 0.6

PORV LIFT SETPOINT IR__

_ _ _1

t20 040040 4

4 1

0 502.0 0.1

RV-206 LIFT SETPOINT 121 t22 040050 5

1 0

0 0.0 0.0

FOR NON-CONDUCTING HX 123 L24 125 126 ans

Menesheenewsu VOLUME DATA 05XXXY IY=1)

ns *ns ****nen**

L27 L28 m

129 a

130 a

1 2

3 4

5 6

7 8

9 10 131 VIB IR P

T X

VOL ZVOL ZN FLONL FLONA 132 i

(N)

L33 050011 0 0 415.0 0.0

-1.0 1300.0 50.0 50.0 50.0 50.0 L34 050021 0

0 415.0 108.98 0.0 5 4 5 0, 0 50.0 50.0 50.0 50.0 135 4

1140 F)

ASSUMPTION ---->

136 050031 0

1 14.7 70.0 0.5 2.OE+6 100.0 0.0 100.0 1.OE+6 137 i

138 e-----

VOLUHE DATA -

05XXXY Vyg) 139 i

140 i

11 12 13 14 15 16 L41 i

Y DIAMl ELEY INEQ VRAIN VLHTC MESH 142 i

L43 050012 20.0 0.0 L44 050022 20.0 0.0 L45 i

I ASSUMPTION) 146 050032 50.0 0.0 L47 i

L48 i

L49 M

L50 i

e w

e TIME-DEPENDENT VOLUME 07XXYY 151 if 152 if L53 i

IRIN TIME PRES TEMP AVG.

MIXL 154 i XXYY (SEC)

(PSIA)

IF)

X (FT) 155 i

L56 070101 2

0.0 25.0 70.0 0.5 0.0 L57 070102 1.OE+6 25.0 70.0 0.5 0.0 L58 i

159 if L60 m

JUNCTION DATA -

08XXXY (Y=1) Manamenewomenames 161 i

L62 i

L63 m

L64 i

1 2

3 4

5 6

7 8

9 10 L65 i XXXY VI VO IP IV HP AJUN ZJUN INERTA FJUNF FJUNR L66 i

L67 080011 0 2 1 0 0.0 1.0 1.0 0.0 0.0 0.0 168 080021 2 1 0 0 0.0 1.OE+6 0.0 1.OE-6 0.0 0.0 L69 080031 1 3 0 1 0.0 1.0 0.0 1.OE-6 0.0 0.0 L70 080041 2 3 0 2 0.0 8.8889E-3 1.0 1.OE-6 0.0 0.0 L71 11.28 SQ. INCHI 172 M

L73 if****

JUNCTION DATA O8XXXY IY=2) 174 i

L75 176 11 12 13 14 15 16 17 18 19 20 21 177 V JVERT CHOK JCA MIX DIAMJ CNTR REGH 2PHS ANGL INDEX ISP 178 i

179 08001

-1 0

0 0.0 0.0 0

-1 0.0 0

0

0 180 080022 0

-1 0

0 0.0 0.0 0

-1 0.0 0

0 181 080032 0

1.

0 0

0.0 4.0486E-3 0

-1 0.0 0

0 182 NOTE: CNTR M AJUN a 4.0486E-3 FTi*2 I = 0.583 INCH*m2) 183 184 080042 0

1 0

0 0.0 0.42 0

-1 0.0 0

0 185 NOTE: CNTR

AJUN = 3.7333E-3 FT**2

= 0.538 INCHN*2) 186 187 188 VALVE DATA 11XXX0 m

189 190 191 TRIP 192

)XO ID IACV IAC2 PCV CV1 CV2 CV3 193 194 110010

-3 1

0 0

0 195 196 110020 1000

-3 0

0 0

197 U

198 U

199 -----------------------

GENERAL DATA 12XKYY

e**en****was age 200 201 U

202 TIME NORM.

203 U XXYY N

(SEC)

CURVE 204 0

205 120101

-3 0.0 0.0 U PORV 206 120102 1.9 1.0 207 120103 1.0E+6 1.0 208 U

209 N

PRES.

NORM.

210 w

IPSIA)

CURVE 211 U

212 120201

-5 0.0 0.0 U RV-206 213 120202 502.0 0.0 214 120203 503.0 0.9 215 120204 552.0 1.0 216 120205 1.OE+6 1.0 217 U

218 U

219 U

TIME NORM. HEAT 220 N

(SEC)

ADDITION 221 i

222 120301

-6

-1.OE+6 0.0 223 120302 0.0 0.0 224 120303 0.1 1.0 225 120304 20.0 1.0 226 120305 20.1 0.0 227 120306 1.0E+6 0.0 228 U

229 230 FILL TABLE -

13)XYY ane*U*en*answen*een 231 232 U

233 U

234 TABLE 1 (JUN 1 -

CHARGING FLOW) 235 U

236

.237 U

TRIP JX TIME FLUX H

PSIA 238 U XXYY N

ID ITIME) JY (SEC)

(GPM/FT**2) 20

239 240 130101

-2 2

0 1

0.0 420.0 115.0 1000.0 241 130102 1.0E+6 420.0 115.0 1000.0 242

( (145 F) 243 244 245 246

- N4'*eH**menw****

HEAT EXCHANGER 21X)XYY

a 247 N

248 249 N

250 IHTX ITHTXQ JYOL INTYPE M5 N6 251 ICNTL BLK) 252 210101

-2 1000 2

7 0.0 0.0 253

!54 N

255

!56 CONTROL SYSTEN HODELINS 70YXXX 258 N

!59 NO. OF CNTL NO. OF MAX TIME 260 N

INPUT BLOCKS CONTROL BLOCKS STEP SIZE

!61 162 701000 3

3 0.01 263 264 N

265 CONTROL INPUT BLOCK DEFINITION

!66 N

267 N

268 XXX IDC SYMBOL IREG CGAIN CIC 269 N

270 702001 1

TINK, 0

1.0 0.0 271 702002 2

TRPT, 5

1.0 1.0E+6 272 702003 3

PRES, 1

1.0 415.0 273 N

?74 N

275 CONTROL BLOCK DEFINITION 276 N

277 278 N

IDC ITYPE INCI INC2 GAIN CP1 CP2 CIC CHIN CHAC 279 N

280 703001

-1 Ss54.

1 2

1.0 1.0 -1.0 -1.0E+6 281 703002

-2

FNG,

-1 3

-30.0 0.0 0.0 0.0

HEAT IN 282 M )

283 703003

-3

FNG, 3

2 1.0 0.0 0.0 0.0

RV-206 284 POSITION 285 P

286 ----------------------------END OF DATA-:

287

ATTACHMENT 1 EXISTING TECHNICAL SPECIFICATIONS 3.1 REACTOR COOLANT SYSTEM 3.1.1 MAXIMM REACTOR COOLANT ACTIVITY APPLICABILITY:

Applies to measured maximum activity in the reactor coolant system at any time.

OBJECTIVE:

To limit the consequences of an accidental release of reactor coolant to the. environment.

SPECIFICATION:

The specific activity of the reactor coolant shall de limited to:

1. 1 1.0 iAC1/gm DOSE EQUIVALENT 1-131.

2.

& 100/E 4C1/gm.

ACIIQM:

A. With the specific activity of the reactor coolant determined to be >1 14Cl/gm DOSE EQUIVALENT 1-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or

>60 4CI/gm DOSE EQUIVALENT I-1lJ or >100/E ACilgm, be in at least HOT STAND8Y with the average temperature of the reactor coolant (Tayg) less than 535'F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

B. With the specific activity of the reactor c lant

> 1.0 *:Cilgm DOSE EQUIVALENT 1-131 or >100/E pCilgm, perform the sampling and analysis requirements of item la.4.a of Table 4.1.2 until the specific activity of the reactor coolant is restored to within its limits.

C. The provisions of Specification 3.0.4 are not applicable.

SAN ONOFRE - UNIT 1 3.1-1 AMENDMENT NO: 29. 38. 70, 83, 91, 96, 130

Snecific Activity The limitations on the specific activity of the reactor Coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed the guidelines of 10 CFR Part 100 following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity > 1.0 4Ci/gm DOSE EQUIVALENT 1-131, accommodates possible iodine spiking phenomena which may occur following changes in THERMAL POWER.

Reducing Tavg to < 535'F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the reactor coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

Increased surveillance for performing isotopic analyses for iodine is required whenever the DOSE EQUIVALENT 1-131 exceeds 1.0.Ci/gram and following a significant change in power level to monitor possible iodine spiking phenomena to assure the activity remains < 60 sCilgm DOSE EQUIVALENT 1-131.

The assumptions and results of these calculations are documented in "Safety Evaluation by the Office of Nuclear Reactor Regulation." Docket No. 50-206, dated April 1, 1977.

SAN ONOFRE - UNIT 1 3.1-2 AMENDMENT NO:

29. 83. 96, 130

3.1.2 OPERATIQNAL COMPONErTS APPLICABILITY:

Applies to the operating status of the reactor coolant system equipment and related equipment. For the applicable surveillance requirements, see Table 4.1.2.

OBJECTVE:

To identify those conditions of the reactor coolant system necessary to ensure safe reactor operation.

SPECIFICATIONS:

A. At least one pressurizer safety valve shall be OPERASLE or open when the reactor head is on the vessel, except for hydrostatic tests.

B. The reactor shall not be made critical or maintained critical unless both pressurizer safety valves are OPERASLE.

C. During MODES I and 2 and in MODE 3 with reactor trip breakers closed, all three reactor coolant loops and their associated steam generators and reactor coolant pumps shall be in operation.

With less than the above required coolant loops in operation. be in at least HOT STAND8Y with reactor trip breakers open within I hour, except as modified by Specification 0 below.

0.

The limitations of Specification C may be suspended as follows:

1. During MODES 1 and 2, operation may be conducted with 0, 1. 2 or 3 reactor coolant pumps operating at less than 5% of RATED THERMAL POWER for purposes of conducting low power physics testing.

2. During MODES 1 and 2 and in Mode 3 with reactor trip breakers closed, operation may be conducted for less than 24 consecutive hours with one or two reactor coolant pumps operating if THERMAL POWER is less than 10% of RATED THERMAL POWER.

E. During MODE 3 with the reactor trip breakers open, the following specifications shall apply:

1. At least two of the reactor coolant loops listed below shall be OPERASLE:

a. Reactor coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop 8 and its associated steam generator and reactor coolant pump.

c.

Reactor coolant loop C and its associated steam generator and reactor coolant pump.

SAN ONOFRE -

UNIT 1 3.1-3 AMENDMENT NO:

43. 77.

103. 130

2. At least one of the above reactor coolant loops shall be in operation.'

3. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SUTC N within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required reactor coolant loop to operation.

F. During MODE 4, the following specifications shall apply:

1. At least two of the reactor coolant loops/RESIDUAL HEAT REMOVAL (RHR)

TRAINS listed below shall be OPERABLE:

a. Reactor Coolant loop A and its associated steam generator and reactor coolant pump.

b. Reactor coolant loop 8 and its associated steam generator and reactor coolant pump.

c. Reactor Coolant loop C and its associated steam generator and reactor coolant pump.

d. Residual heat removal (RHR) pump G-14A and on' associated RHR TRAIN.

0.

Residual heat removal (RHR) pump G-148 and one associated RHR TRAIN.

2. At least one of the above loops/trains shall be in operation.**

All reactor coolant pumps may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor Coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

All reactor Coolant Pumps and residual heat removal pumps may be deenergized for up to one hour provided (a) no operations are permitted that would Cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 400F below saturation temperature.

SAN ONOFRE - UNIT 1 3.1-4 AMENOMENT N0:

77, 130

3. With less than the above required loops/trains operable immediately initiate corrective action to return the required loops/trains to operable status as soon as possible; if the remaining operable loop/train is an RHR train, be in Cold Shutdown within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4. With no loop or train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return one required loop or train to operation.

G. During Mode 5 with reactor coolant loops filled, the following specifications shall apply:

1. At least one residual heat removal (RHR) train shall be OPERABLE and in operation*, and either

a. One additional RHR train shall be operable,** or

b. The secondary side water level of at least two steam generators shall be greater than or equal to 256 inches (wide range).

2. With less than the above required loops/trains operable, or with less than the required steam generator level, immediately initiate corrective action to return the required loops/trains to operable status.or to restore the required level as soon as possible.

3. With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR train to operation.

The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

- One RHR train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing, provided the other RHR train is operable and in operation.

SAN ONOFRE - UNIT 1 3.1-5 AMENDMENT NO:

77, 125, 130

H. During MO0E 5 with reactor coolant loops not filled, the following specifications shall apply:

1. Two RESIDUAL HEAT REMOVAL (RHR) TRAINS shall be OPERASLE* and at least one RHR TRAIN shall be in operation*.

2.

With less than the above required RHR TRAINS OPERABLE. immediately initiate corrective action to return the required RHR TRAINS to operable status as soon as possible.

3. With no RHR TRAIN in operation, suspend all operations involving a reduction in boron concentration of the reactor coolant system and immediately initiate corrective action to return the required RHR TRAIN to operation.

I. A reactor coolant pump shall not be started with the RCS pressure j 400 psig unless:

1. the pressurizer water level is less than 80%, or

2. the potential for having developed reactor coolant system temperature gradients has been evaluated.

J=5:

One pressurizer safety valve is sufficient to prevent over pressurizing when the reactor is subcritical, since its relieving capacity is greater than that required by the sum of the available heat sources, i.e., residual heat, pump energy and pressurizer heaters.

Prior to reducing boron concentration by dilution with make up water either a reactor coolant pump or a residual heat removal pump is specified to be in operation in order to provide effective mixing.

During boron injection, the operation of a pump, although desirable, is not essential.

The boron Is injected into an inlet leg of the reactor coolant loop.

Thermal circulation which exists whenever there is residual heat in the core and the reactor coolant system is filled and vented, will cause the boron to flow to the core.

One RHR TRAIN may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing Provided the other RHR TRAIN is operable and in operation.

The RHR pump may be de-energized for up to one hour provided (a) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (b) core outlet temperature is maintained at least 40*F below saturation temperature.

SAN ONOFRE - UNIT 1 3.1-6 AMENDMENT NO:

77, 102, 130

4ack of further mixing cannot result in areas of reduced boron concentration within the Core. Prior to criticality the two pressurizer safety relief valves are specified in service in ordyr to conform to the system relief capabilities.()

The plant is designed to have all three reactor coolant loco operational during normal power operation (0OtS I and 2).

Under these conditions, the ONS ratio will not r below 1.30 after a loss of flow with a reactor trip.( J)( ) 4 one reactor coolant loop not in operation, this specification requires that the plant be in at least NOT STANO8Y with reactor trip breakers open within one hour (for the significance of the trip breaker position, see below).

However, exception is taken whenever reactor Power is less than 10% of RATED THERMAL PCER. Heat transfer analyses show that reactor heat equivalent to 8 of RATED THERMAL POWER can be removed with natural circulation only: hence, for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months the specified upper limit of 10% of RATED THERMAL PCWER with I or 2 reactor coolant pumps operating provides a substantial safety factor.

In MODES other than MODES I and 2, functional redundancy in the core heat removal methods (not necessarily system redundancy) is specified to satisfy single failure considerations.

Functional redundancy, as applied to the San Onofre Unit I power plant, includes use of diverse heat removal methods. Furthermore, single failure considerations apply only to active components.

For operation in MO0E 3 under all design basis conditions, it

'has been determined that one reactor coolant (RC) loop generally provides the required decay heat removal capability, the only exception to this being the control rod bank withdrawal from subcritical accident, when the 0N8 design basis may not be met.

Since power to the gripper and lift coils of the control rod drive mechanism is carried through two reactor trip circuit breakers connected in series with the coils, both breakers must be manually closed before any control rod motion out of the core can take place. In light of this design feature, these Technical Specifications require that all three RC loops be in operation in MOD0 3 if the reactor trip breakers are closed.

Whenever the reactor trip breakers are open.

the design feature would prevent any control rod motion, even though single failure considerations*

require that at least two loops be operable.

For the same reasons and subject to the same limitations that are stated in the preceding paragraph. exception is taken whenever reactor power is less than 10% of RATED THERMAL POWER.

Single failure considerations apply to active components.

SAN ONOFRE - UNIT 1 3.1-7 AMENOMENT N0:

77, 102, 103, 130

In MOCES 4 and 5, the Technical Specifications permit functional redundancy in the core heat removal methods (not necessarily system redundancy) to satisfy single failure Considerations.

Functional redundancy, as applied to the San Onofre Unit I power plant, includes use of diverse heat removal methods.

In MOCE 4 and MODE 5 (reactor coolant loops filled), a single reactor coolant loop or RHR TRAIN provides sufficient capability for removing decay heat; but single failure considerations' require that at least two Methods (either RCS loop or RHR TRAIN) be OPERABLE.

In MO0E 5 (reactor coolant loops not filled). a single RHR TRAIN provides sufficient heat removal Capability for removing decay heat; but single failure considerations,& and the unavailability of any of the steam generators as a heat removing component, require that at least two RHR TRAINS be OPERABLE.

The operation of one reactor coolant pump or one residual heat removal pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be wi thin the capability of operator recognition and control(

The limitation on reactor coolant pump operation with the RCS pressure 1 400 psig ensures that the RCS will be protected from pressure transients whigh could exceed the limits of Appendix G to 10 CFR Part 50 5).

A pressurizer water level of less than 80% ensures that the start of a reactor coolant pump. with a temperature differential of 100*F will not result in 10 CFR Part 50 Appendix G limits being exceeded.

There are several means available for determining that there is not a temperature differential of > 50'F between the secondary and primary systems with t 400 psig primary system pressure. These methods may include but are not necessarily limited to the following:

1) Converting steam line pressure indication into maximum temperature of steam generator fluid.

2) Tagging RCP switches with shutoff temperatures.

Single failure considerations apply to active components.

SAN ONOFRE - UNIT 1 3.1-8 AMENOMENT NO:

43, 102, 103, 130

3) Assuring adequate time for temperature gradients to dissipate.

4) Filling steam generators with water of known temperature.

REFERENCES:

(1) Final Engineering-Report and Safety Analysis. Sections 9 and 10.

(2) Final Engineering Report and Safety Analysis, Paragraph 10.2.

(3) Supplement No. I to Final Engineering Report and Safety Analysis. Section 3, Question 9.

(4) NRC letter dated June 11, 1980, from 0. G. Elsenhut to all operating pressurized water reactors.

(5) Letter to A. Schwencer from K. Baskin dated October 12, 1977.

SAN ONOFRE - UNIT 1 3.1-9 AMENDMINT NO: 77, 102, 130

S~

~

ANOPR~S~iF LIMITATIN 3.1.3 CO l

of tFTUheu a

renuo Coo ant Syse APPLICABILt:

Applies to heatup and cooldown of the reactor coolant system.

To mantain the structural integrity of the reactor coolant system throughout the lifetime of the plant.

SPFICF~_ATIQN:

A. Reactor pressure and heatup and cooldown of the reactor coolant system during the first 16 yeac of eruivalent full power operation shall be limited in accordance with Figures 3.1.3a and 3.1.3b. Thereafter, limits shall be based onl neutron exposure equivalent to not less than 16 years of full power operation, and Figures 3.1.3a and 3..3b shall be updated accordingly (by formal license amendment application).

Figures 3.1.3a and 3.1.16 shall be updated in accordance with the following critlf-nd-procedures:.

(1)

The methods of Appendix G, "Protection Against Nonductile Failure". to Section III of the ASe Boiler and Pressure Vessel Code shall be used to obtain the allowable pressure-temperature relationships for the reactor coolant system.

(2)

The curves in Figure 3.1.3C shall be used in predicting the reference nil-ductility temperature increase, RTNOT unless measurements on the irradiation specimens show RTNOT greater than those predicted by the curves. in which case a new curve having the same slope as the original shall be constructed.

C. The pressurizer heatup rate of 100*F/hour and cooldown rate of 200*F/hour shall not be exceeded.

0. The reactor shall not be brought to a critical condition until the pressurtemffperaturs state is to the right of the criticality limit line as shown in Figures 3.1.3a.

Mu:

The initial Reference Nil Ouctility Temperature (RTNOT) for all reactor vessel material based on Charpy V-notch data, drop weight tests. and conservative estimates is 82 F or less.

The RTNOT at the 114 thickness location (location of Appendix G reference flaw tip) increases as a function of cumulative neutron exposure up to approximately 240*F for the coare region of the reactor vessel after 30 years of operation.

Technical Specification 3.20.A(l) should be reevaluatedpfor continued applicability of the low pressure PORY overpressure setpoint at any time the heatup and cooldown curves are changed.

NRC Standard Review Plan Branch Technical Position M4TES 5-2.

Change No: 14 SAN ONOFRE - UNIT 1 3.1-10 AMENOMENT NO:

38, 102, 130

A sixteen (16) equivalent full power year service period was chosen for the operational limits given in this specification because at the and of this period the limiting RTNOT of the reactor vessel at the 1/4 thickness location is aproximately 217'F in the core region. This RTNOT is at least 509F above the RTNOT Of all other regions in the primary reactor coolant system.

The highest RTNOT of the core region material is determined by adding the radiation induced 4ATar-for-the applicable, time period to the original RTNOT shown in the Table 3.1.3.1.

The fast neutron (E > IMev) fluence at 1/4 thickness and 3/4,thickness vessel locations is given as a function of full power service life in Figure 3.1.3d. Using the applicable fluence at the end of the year period and the copper content of the material in question, the RTNOT is obtained from Figure 3.1.3c.

Values of dRTNOT may continue to be determined in this manner unless measurements on the irradiation specimens show ARTNOTS greater than those predicted by the curves for the equivalent capsule exposure.

Allowable pressure-temperature relationships for various heatup and cooldown rates are calculated using methods derived from non-mandatory Appendix G in Section III of the ASME Soiler and Pressure Vessel Code, and discussed in detail in Reference 1.

The results of these calculations are provided in Reference 2.

The design heatup and cooldown rates for the pressurizer are 100*F/hour and 200*F/hour, respectively.

The vertical line portion of the criticality limit given in Figures 3.1.3a is at the minimum permissible temperature for the 2485 psig in-service hydrostatic test as required by Appendix G to 10CFR Part 50. The non-vertical portion of the criticality limit Is shifted 40'F to the right of the heatup curve as required by Appendix G to 10CFR Part 50.

REFERENCES:

(1) "Pressure Temperature Limits* Section 5.3.2 of Standard Review Plan, NUREG-751087. 1975.

(2) 5. E. Yanichko, et al. *Analysis of Capsule F from the Southern California Edison Company San Onofre Reactor Vessel Radiation Surveillance Program". WCAP 9520, May 1979.

angeN T No.: 31023 SAN ONOFRE - UNIT 1 3.1-11 1AMNOMEN NO: A1 2 3

w4ATERI A PRO PETY aAS IS A 7

/1&T 1. 1ATI ON I CONTROL41NG MATERIAL:

INTERM4EDIATE SHEI.LL COPPER CONTENT: 0.18 WT P1,03PNORU,11 CONTEXT: 0.0114 V

RTNOT INITIAL: 550f A..

T AFTER 16 EFPY: 217'3 3000 RTOT INITIAL: 32" UP TO IS EFPY AMC CONTAINS MARGINS OF '100f &NO 60 PSIG FOR PO31SILL.

INSTRUMENT ERRORZS

'.0 a W 2000 LA 200 3IM05 On CONTR LINGMATE IA I TNEDIAT E

)

COCPER CONSENT: 0.18 WT T~c 1

111 A

II.

.A

..0

~

PO RUN CO.T2Na:

N:t1 1000 LI:O92T A130 CONTROLLIBASG OAERAN IMSER SHE PM03PHORSMCOR0ETATI.01TEST CURVE~~EMEATR (3L330tFOf)ATPRAESU 70 ~ ~

~

O 609/M SORTERSRVCICEE UP~~E 100 UP TFP 1n COTAM MARGI 6 FIC 0 60-ITO P3 1t I

II 200 l0 o

INIAEAD TWETU (OFIT L aio uiidtatonsApia o h accountt 1

frcouEm flangegregion pe SAN ~

~

Aneni GNR to 10I CT 50-2AEOEN O

2 3

3000

- &7ERIAL PROPERqTY 3ASIS l

CONTROLLING MTERIAL: INrERMeIATE SHELL COPPER CONTENT: 0.1& wT'F POSPH4ORUS CONTENT: 0.014 wt4 "NOT INITIAL: 55cy RTNOT AFTER I8 (FPY:

1/14.

2170F TO 100*F/4fR FOR THE SERVICE PERIOD UP T0 16 EFPY AnO CONTAINS MARGINS oR I0 0F ANO 60 PSIG FOR PSSISLI INST412UtIWT ERR.

Region modified to1711111 1 1111 aoa C

account for closure flange region per 000

+#

Appndix Gto 10C7R50 0

t00 200 300 I400 500 INOICATED TEMPERATURE RAT FIGURE 3.1.3b San Cuofre Unit No. 1 Reactor Coolant Syse C oldown Limitations Apricable for the FirsI 6 ETP1.

Ocange No:

14 FSAN ONOFRE 3

bUNIT 1 S 3.1-13 AMENOENT NO:

92, 130

400

____--300 200 150 4 80 -u%

Cu mu, am %.' Wul 0

Ls% Cu astr, Me% Os stu a%

u "a, austo was 20 L8%C

-&LC C

0.1o% so ust, u~s% O was 10l 1

5 0ot 2

5 10*

FLUENCE ( NA:Mgat > ME)

FIGURE 3.1.3c Effect of Fluence and Copper Content on ART For Reactor Vessel Steels Exposed to Irradiatiog0It 550OF SAN ONFRE -UNIT 1

Cang No: 14 3.1.14 AMENOMENT NO:

102, 130

10 1.0 x 1020 SURFACE 4.2 X 0o 2

1/4 THICXNES 0

7.1 X o04 2

101 a

s to s

20 as o30 3s SERVICE LIFE (EFFECTIVE FULL POWER YEARS)

Figure 3. 1. 3d Fast Neutron Fluence (E > I MEV) as a Function of Full Power Service Life SAN ONFRE -UNIT 1

Change No: 14 3.1-15 AMENDMENT NO:

130

TABLE 3.1.3.1 REACTOR VESSEL TOUGHNESS DATA (UNIRRApIATEO)

-4 Miaimm Average 50 ft-lb/35 ll Upper She I Material Cu P

NoTf Tea (f)

RTn UpprSef Component Code No.

Type

01)

(1)

()

Long.

Trans.

T Long.

Trans.

CI. ld. Dome 74604 A3028 60(a) 112 132 72 72.5 Peel Segment 141605-1 A3028

-10 114 134 74 70.5 Peel Segment N11605-2 A3028

-10 90 110 50 122 Peel Segment W7605-3 A3020

-10 108 128 68 85 Peel Segment W1605-4 A3028

-10 120 140 80 74 Peel Segment 11605-5 A3020

-10 26 46 10 109 Peel Segment W7605-6 A3028

-10 102 122 62 88 Md. Flange 1602 A336 mod 60(a)

(b) 60 Ves. Flange 11603 A336 mod 60(a)

(b) 60 Inlet Nozzle NJ611-1 A336 mod 60(a)

(b) 60 Inlet Nozzle 1611-2 A336 mod 60(a)

(b) 60 Inlet Nozzle W7611-3 A336 mod 60(a)

(b) 60 Outlet Nozzle N7610-1 A336 mod 60(a)

(b) 60 g

Outlet Nozzle W7610-2 A336 mod 60(a)

(b) 60 j

Outlet Nozzle 1610-3 A336 mod 60(a)

(b) 60 Upper Shell W7601-3 A3028 0.15 0.014

-10 48 68 8

98.5 g

Upper Shell W7601-6 A3028 0.16 0.012

-30 64 84 24 104 Upper Shell W7601-7 A3028 0.15 0.014

-20 52 72 12 95.5 A. Estimated per NRC Standard Review Plan Branch Technical Position HTEB 5-2.

b. Only 10*F Charpy V-notch data available. Conservative estimates for NDTT and RlnoI were used.

C TABLE 3.1.3.1(cont'd)

REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATEO) minimu Average 50 ft-lbl35 Elt Upper Shelf material Cu P

IIOTT lejja f

Upp Eer She-l Component Code No.

Type (a

(CF)

Long.

Trans.

Long.

Trans.

Inter. Shell WT601-l A3028 0.1 0.013 0

57 120(a) 60 94 15 Inter. Shell W7601-8 A3028 0.18 0.012 10 93 100(a) 40 91 19 Inter. Shell M7601-9 A3028 0.18 0.014 0

64 115(a) 55 102 72 Lower Shell W7601-2 A3028 0.17 0.013

-20 74 94 34 97 Lower Shell W7601-4 A3028 0.14 0.014

-10 91 1Il Si 94 Lower Shell 1601-5 A302B 0.14 0.014 10 122 142 82 87.5 Bot. Nd. Peel W7607 A302B

-20 62 82 22 91 Bot. ld. Dome W7606 A3028 60(b) 99 119 60 86 Weld 0.19 0.017 0(b) 29(a) 0 90 HAZ 0(b)

-14(a) 0 101

a. Actual not estimated.

3.

b. Estimated per MRC Standard Review Plan Branch Technical Position NTEB 5-2.

We

-se Do

3.1.5 PRESSURIZER RELIEF VALVES APPLICAILIT:

MO0ES 1, 2 and 3.

BJECTTVE:

To ensure reliability of the power operated relief valves (PORVs) and their associated block valves.

SPECIELCATION:

Two PORVs and their associated block valves shall be OPERABLE.

ACIIGN:

A. With one or more PORV(s) inoperable, within I hour either restore the PORV(s) to OPERABLE status or close. the associated block valve(s) and maintain the block valys(s) in the closed position; otherwise, 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 in COLD SHUTDON within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

S. With one or more block valve(s) inoperable, within I hour restore the block valve(s) to OPERABLE status; otherwise, be in at least NOT STANDSY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

C. The provisions of Specification 3.0.4 are not applicable.

&SAL:

The power operated relief valves (PORVs) operate to-relievit RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The air supply for both the relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.

REFERENCE (1)

NRC letter dated July 2. 1980, from 0. G. Eisenhut to all pressurized water reactor licensees.

SAN ONOFRE - UNIT 1 3.1-21 AMENOMENT NO:

58, 59. 83, 130

3.2 CHEMICAL AND VOLUME CONTROL SYSTEM Applies to the operational status of the chemical and volume Control system.

To identify those conditions of the chemical and volume control system necessary to ensure safe reactor operation.

SPECICATION:

A. When fuel Is in the reactor, the following chemical and volume control system conditions shall be met:

(1)

One charging pump or the test pump shall be OPERABLE.

However, when the RCS pressure is < 400 psig and pressurizer water level is greater than 50%, a maximum of one of the two centrifugal charging pumps shall be OPERABLE. The inoperable centrifugal charging pump shall have the motor circuit breaker removed from the electrical power supply circuit and shall be condition tagged.

(2)

One boric acid transfer pump or the boric acid injection pump shall be OPERABLE.

(3) A solution of at least 3450 pounds of boric acid in not less than 3500 gallons of water at a temperature of 140'F or higher, with at least one heater OPERABLE. shall be in the boric acid tank.

(4)

System piping and valves shall be OPERABLE to the extant of establishing two flow paths for boric acid

.tanks.

(5)

During periods when borated water is in the refueling cavity, the requirements-in A.(1) through A.(4) may be waived provided that an alternate source of borated water is available to establish at least one flow path to the core for boric acid injection which can be initiated from the control room. The minimum capability for boric acid addition shall be equivalent to that supplied by a charging pump from the refueling water storage tank.

8. The reactor shall not be made critical unless the following additional conditions are met:

(1)

One additional charging pump or test pump OPERABLE.

(2)

One additional boric acid transfer pump or boric acid injection pump OPERABLE.

(3)

Electrical heat tracing for boric acid piping OPERABLE.

SAN ONOFRE - UNIT 1 3.2-1 AMENDMENT NO:

102, 130.

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3.20 OVERPRESSURE PROTCTTON SYSTEMS APPLICAILITY:

Applies to operability of the overpressurization protection systems.

To preclude the potential for exceeding 10 CFR 50, Appendix G, in the event of a pressre-transient while water-solid.

SPECIFICATION:

A. When the RCS pressure is 1 400 psig* and pressurizer water level is greater than 50%, at least one of the following overpressure protection systems shall be OPERABLE:

(1)

Two power operated relief valves (PORVs) with a lift setting of 1 500 psig,'* or (2) A reactor coolant system vent(s) of 2.

1.75 square inches.

ACT1N:

8. With one PORV inoperable when required in accordance with Specification A above. either restore the inoperable PORV to OPERA8LE status within seven days or depressurize and vent the RCS through a 1.75 square inch vent(s) within the next eight hours: maintain the RCS In a vented and tagged condition until both PORVs have been restored to OPERA8LE status.

C. With both PORVs inoperable when required in accordance with Specification A above, depressurize and vent the RCS through at least a 1.75 square inch vent(s) within eight hours; maintain the RCS In a vented and tagged condition until both PORVs have been restored to OPERASLE status.

0. In the event either the PORVs or the RCS vent(s) are-used to mitigate a potential 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 indicating transient, the effect of the PORVs or vent(s) on the transient and any corrective action necessary to prevent recurrence.

The placing in service of the OHS at j 400 psig is intended to assure that protection is provided whenever temperature is below 3600F.

The alarm to arm the CHS being keyed to pressure assures that inadvertent opening of the PORVs does not occur due to placing the OMS into service with RCS pressure above the 500 psig initiation setpoint.

- The 500 psig setpoint is based on the current heatup and cooldown curves for 16 EFPY. The setpoint requires reevaluation for acceptability any time the curves are changed.

SAN ONOFRE - UNIT 1 3.20-1 AMENOMENT NO:

102, 130.

EMS1:

The OPERASILITY of two PORVs or an RCS vent opening of greater than 1.75 square inches ensures that the RCS will be Protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when the initial RCS pressure is j 400 psig and the pressurizer water level is greater than 50%, assuming a single failure of one PORV and no operator action for 10 minutes.

Either PORQ has adequate relieving capability to protect the RCS from over~ressuri.

Zation due to a design basis transient as described in submittal to the NRC dated October 12, 1977.

Tagged as it refers to the RCS vent, means tagged in accordance with current Southern California Edison procedures for tagging of equipment which must not be operated.

SAN ONOFRE -

UNIT 1 3.20-2 AMENOMENT NO:

102, 130.

4.1.6 PRESSURIZER RELIEF VALVES APPuLIALTY:

Applies to the power operated relief valves (PORVs) an their associated block valves for MOCES 1, 2 and 3.

OBJECTIVE:

To ensure the reliability of the PORVs and block valves.

SPECIFICATIONi A. Each PORV shall be demonstrated OPERABLE:

1. At least once per 31 days by performance of a C4ANNEL TEST. which may include valve operation, and

2. At least once per 18 months by performance of a CHANNEL CALIBRATION.

8. Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel, unless the block valve is being maintained closed in order to meet the requirements of Specification 3.1.5.A.

C. The backup nitrogen supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by transferring motive power from the normal air supply to the nitrogen supply and operating the valves through a complete cycle of full travel.

BAI11:

The power operated relief valves (PORVs) operate to relieve RCS pressure below the getting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The air supply for both the relief valves and the block valves is capable of being supplied from a backup passive nitrogen source to ensure the ability to seal this possible RCS leakage path.

REFERENCES:

(1) NRC letter dated July 2. 1980, from 0. G. Elsenhut to all pressurized water reactor licensees.

SAN ONOFRE - UNIT 1 4.1-24 AMENDMENT NO:

58. 109, 130

4.20 QEPRESSURE PROTECTTON SYSTEMS APPLICA8ILI:

Applies to OPERABILITY of the overpressuritzation Protection IBJECTIVE:

To verify that the overpressure protection systems will respond Promptly and properly if required.

SPECIFICATION:

A. Each power operated relief valve (PORV) shall be demonstrated operable by:

(1) Adjusting the pressure control bistable setpoint such that the PORVs are actuated and the annunciators alarm within 31 days prior to returning to a water-solid condition following a COLD SHUTDOWN with the RCS depressurized.

(2) Performance of a CHANNEL TEST within 31 days prior to enabling the low pressure overpressure mitigation setting of the pressurizer PORVs on cooldown.

(3) Performance of a CHANNEL CALISRATION on the PORV actuation channel at least once per 18 months.

(4) Verifying that position indications on the PORV isolation valves indicate that thi valves are open at least once per week when the PORVs are being used for overpressure protection.

The surveillance requirement to verify OPERABILITY of the POR~s provides assurance that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CPR Part 50 when the initial RCS pressure is

.j 400 psig.

Either PORV has adequate relieving capability to protect the RCS from overpressurization due to a design basis transient as discussed in Reference 1.

(11 Letter to A. Schwencer from X. Baskin dated October 12, 1977.

SAN ONOFRE -UNIT 1

4.20-1 AMENOMENT NO:

102, 130'

ML13329A726

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