ML20195J902
| ML20195J902 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 11/23/1998 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20195J890 | List: |
| References | |
| NUDOCS 9811250112 | |
| Download: ML20195J902 (48) | |
Text
.. - . . -. . - - . . . . - - . . . . - . - . - . . . . . - - . . . - _ . .-. . . . . - - . . . ~ . - .- .. ..
(
l l
I i
l- ]
i i
J i
i ATTACHMENT A EXISTING TECHNICAL SPECIFICATION l TABLE 3.3.1-1 l-SAN ONOFRE UNIT 2 5
I l
f l
- l. 9811250112 981123TM PDR P
ADOCK 0500 "i p]4 ,
i
RPS Instrumentation-Operating i 3.3.1 Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED SURVE!LLANCE FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 1. Linear Power Level - High 1.2 SR 3.3.1.1 s 111.0% RTP SR 3.3.1.4 SR 3.3.1.6 -
SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9 SR 3.3.1.13 2.' Logarithmic Power Level - High(a) 2(b) SR 3.3.1.1 s .93% RTP l
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
.3 . Pressurizer Pressure - High 1,2 SR 3.3.1.1 s 2385 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 4. Pressurizer Pressure - Low (C) 1,2 SR 3.3.1.1 a 1700 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 5. Containment Pressure - High 1,2 SR 3.3.1.1 s 3.4 psig -
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 L eam Generator 1 Pressure-tow 1.2 SR 3.3.1.1 2 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 7. Steam Generator 2 Pressure-Low 1,2 SR 3.3.1.1 2 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 (continued)
(a) Trip may be bypassed when logarithmic power is > IE-4% RTP. Bypass shall be automatically removed when logarithmte power is s 1E-4% RTP. Trip may be manually bypassed during physics testing pursuant to LCO 3.1.12.
(b) When any RTCB is closed.
(c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between pressurizer pressure and the setpoint is maintained s 400 psia. Trips may be bypassed l
when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurizer pressure exceeds 500 vsia (the corresponding bistable allowable value is s 472 psia).
I l
I SAN ONOFRE--UNIT 2 3.3-8 Amendment No. +27, 142 P
RPS Instrumentation-Operating 3.3.1 Table 3.3.1-1 (page 2 of 2)
Reactor Protective System Instrumentation APPLICABLE MODES OR .
OTHER SPECIFIED SURVEILLANCE FUNCTION CON 0!TIONS REQUIREMENTS ALLOWABLE VALUE
- 8. Steam Generator 1 Level - Low 1,2 $R 3.3.1.1 a 20%
$R 3.3.1.9 SR 3.3.1.13
- 9. Steam Generator 2 Level - Low 1,2 SR 3.3.1.1 a 20%
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 10. Reactor Coolant Flow - Low (d) 1,2 SR 3.3.1.1 Ramp: s 0.231 psid/ rec.
SR 3.3.1.7 Floor: 2 12.1 psid SR 3.3.1.9 Step: s 7.25 psid SR 3.3.1.12 SR 3.3.1.13
- 11. Local Power Density - HighId) 1,2 $R 3.3.1.1 s 21.0 kW/ft SR 3.3.1.3 SR 3.3.1.4 SR.3.3.1.7 SR 3.3.1.9 SR 3.3.1.10
$R 3.3.1.11 SR 3.3.1.12
$R 3.3.1.13 ,
- 12. Departure From Nucl y ate Boiling 1,2 SR 3.3.1.1 a 1.31 Ratio (DNBR) - Low' SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5
$R 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13 (d) Trip may be bypassed when logarithmic power is < 1E-45 RTP. Bypass shall be automatically removed when logarithmic power is a 1E-4% RTP. During testing pursuant to LCO 3.1.12, trip may be bypassed below 5% RTP. Bypass shall be automatically removed when logarithmic power is a 5% RTP.
i SAN ON0FRE--UNIT 2 3.3-9 Amendment No. +27, 142 l
l
i I
i l
ATTACIIMENT B EXISTING TECIINICAL SPECIFICATION TABLE 3.3.1-1 SAN ONOFRE UNIT 3 l
l I
. - , ..- . . .- -.. _ . . . - . . - _ . . - - ~ _ - .. - - ..
i
.RPS Instrumentation-Operating 3.3.1 Table 3.3.1 1 (page 1 of 2)
Reactor Protective System Instrumentation' APPLICABLE MODES OR OTHER SPECIFIED SURVE!LLANCE '
FUNCTION CONDITIONS. REQUIREMENTS ALLOWA8LE VALUE
- 1. _ Linear Power Level - High ' 1.2 - SR 3.3.1.1 s 111.0% RTP SR 3.3.1.4 SR 3.3.1.6
'SR 3.3.1.7 l SR 3.3,1.8 SR 3.3.1.9
$R 3.3.1.13
- 2. ' Logarithmic' Power Level - High(a) 20) $R 3.3.1.1 s .93% RTP SR 3.3.1.7 SR'!.3.1.9 ,
- 3. Pressurizer Pressure-High 1,2 . SR 3.3.1.1 s 2385 psia
$R 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
' 4. . Pressurizer . Pressure - LowIC) - ' 1.2 SR 3.3.1.1 2 1700 psia !
SR 3.3.1.7 SR 3.3.1.9
$R 3.3.1.12 SR 3.3.1.13
- 5. Containment Pressure - High 1,2' SR 3.3.1.1 ~ s 3.4 psig 3R 3.3.1.7 5R 3.3.1.9 SR 3.3.1.13
- 6. Steam Generator 1 Pressure-Low 1.,2 .$R 3.3.1.1- a 729 psia SR 3.3.1.7 SR 3.3.1.9
, SR 3.3.1.13
- 7. Steam Generator 2 Pressure-Low 1,2- SR 3.3.1.1 2 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 L (continued)
"(a) Trip may be bypassed when THERMAL POWER is > IE-4% RTP. Bypass shall be automatically removed when
- . THERMAL POWER is s 1E 4% RTP. Trip may be manually bypassed during physics testing pursuant to j' LCO 3.1.12.
(b) , When any RTCB is closed.
j.
(c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided
! the margin between pressurizer pressure and the setpoint is maintained 5 400 psia. Trips may be bypassed
[,
when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurizer
. pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).
i[
1 SAN-ON0FRE--UNIT 3 3.3-8 Amendment No. 116
.. ~ . . . ~ . . . . . ~ . ~ . - - . .. _ - -... -. -. ......-.- - .
I i
RPS Instrumentation-Operating
- 3. 3.1 -
Table 3.3.11(page2of2)
Reactor Protective System Instrumentation
- i l.
APPLICABLE MODES OR OTHER SPECIFIED SURVEILLANCE 7 FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 8. : Steam Generator-1 Level - Low ~ 1,2 SR 3.3.1.1 a 20%
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 9. - Steam Generator 2 Level - Low 1,2 SR 3.3. 1 a 20%
SR 3.3.1.13
- 10. Reactor Coolant Flow - Low (d) . 1,2 SR 3.3.1.1 Ramp: s0.231psid/sec.
SR 3.3.1.7 Floors a'12.1 psid i Step: s 7.25 psid SR 3.3.1.9 SR 3.3.1.12.
'SR 3.3.1.13
- 11. Local Power Density - High(d) -1,2 SR 3.3.1.1 s 21.0 kW/ft SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.7 SR 3.3.1.9 l l
SR 3.3.1.10
SR 3.3.1.13 l
- 12. ' Departure From Nu I ate Boiling 1,2 SR 3.3.1.1 .2 1.31 j Ratio (DNBR)- Low SR 3.3.1.2 1 SR 3.3.1.3 SR 3.3.1.4 3R 3.3.1.5 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 i SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13 j (d) Trip may be bypassed when THERMAL POWER is < IE-4% RTP, Bypass shall be automatically removed when
- i. THERMAL POWER is a IE 4% RTP. During testing pursuant to LCO 3.1.12, trip may be bypassed below 5% RTP.
t ,
Bypass shall be automatically removed when THERMAL POWER is a 5% RTP. '
l.
!- )'
l l
r.
i.'
I l
- . SAN ON0FRE--UNIT 3 3.3-9 Amendment No. 116 l.
h
y4s, i'A - S %"--*
I i
ATTACIIMENT C PROPOSED TECIINICAL SPECIFICATION TABLE 3.3.1-1 SAN ONOFRE UNIT 2 (REDLINE AND STRIKEOUT) l i
l
l RPS Instrumentation-Operating l
3.3.1 I
Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED SURVEILLANCE FUNCTION CONDITIONS ALLOWABLE VALUE REQUIREMENTS
- 1. Linear Power Level -High 1,2 SR 3.3.1.1 s 111.0% RTP SR 3.3.1.4 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9 SR 3.3.1.13
- 2. Logarithmic Power Level - High(8) 2(D) SR 3.3.1.1 s .93% RTP SR 3.3.1.7 i SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 3. Pressurizer Pressure - High 1,2 SR 3.3.1.1 s 2385 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 4 Pressurizer Pressure - LowfC) 1,2 SR 3.3.1.1 a 1700 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 5. Containment Pressure - High 1,2 SR 3.3.1.1 s 3.4 psig SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 6. Steam Generator 1 Pressure-Low 1,2 SR 3.3.1.1 a 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 7. Steam Generator 2 Pressure-Low 1,2 SR 3.3.1.1 a 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 (continued)
(a) ' ; ,, e, L LJo..J ', ';;u t' - s; - . 1:
CJ ;.. ;'e .. a. ; u; soII s J 2+
i ';p ; ; ;; , . _ :: : "^
Trip must be enabled when logarithmic power is < 4E-5% RTP. Trip may l
be manually bypassed during physics testing pursuant to LCO 3.1.12.
(b) When any RTCB is closed.
l (c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided
( the margin between pressurizer pressure and the setpoint is maintained s 400 psia. Trips may be bypassed when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psia),
i SAN ON0FRE--UNIT 2 3.3-8 Amendment No. 4W,142 l
l' RPS Instrumentation-Operating 3.3.1' !
Table .3.3.1 1 (page 2 of 2) !
j Reactor Protective System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED '$URVEILLANCE FUNCTION
{
CONDITIONS REQUIREMENTS . ALLOWABLE VALUE '
[. ' Steam Generator ! Level - Low 1,2 - SR 3.3.1.1 a 20% i SR 3.3.1.7 l SR 3.3.1.9 I SR 3.3.1.13 i t^- .
- 9. . Steam Generator Z Level - Low 1,2 SR 3.3.1.1 a 20%
SR 3.3.1.7 .i l SR 3.3,1.9 I l SR 3.3.1.13 ;
~
i
- 10. Reactor Coolant Flow- Low (d) 1,2 $R 3.3.1.1 Ramp: 5 0.231 psid/sec. I
$R 3.3.1.7 Floor: a 12.1 psid SR 3.3.1.9 Step: s 7.25 psid 1 SR 3.3.1.12 )
j SR 3.3.1.13 j 11,' Local Power Density- HighIdI- ' 1,2 SR 3.3.1.1 s21.0kW/ft SR 3.3.1.3 i SR 3.3.1.4 j i'
SR 3.3.1.7 SR 3.3.1.9 i
$R 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 i SR 3.3.1.13 i
i
. 12. Departure From Nuql ate Boiling 1,2 $R 3.3.1.1 a 1.31 Ratio (DNBR) - Lowl $R 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4
$R 3.3.1.5' i SR 3.3.1.7
$R 3.3.1.9 2 SR 3.3.1.10 l'
$R 3.3.1.11 3R 3.3.1.12 SR 3.3.1.13 (d) ' o; :;, t; t,,;;. J L;. 1;;; t' ; ;;-. . : e , "" ;,;;.. ,5;;' t-; u . . ;;11, . ; .- .; j 2 I;;;. . ; ;;-; , . :: t 'a " " Trip must be enabled when logarithmic power is > 1.5E.4% RTP. During l testing pursuant to LCO 3.1.12, trip may be bypassed below 5% RTP. Bypass shall be ;.t ;; ;;11, removed when logarithmic po-er is a 5% RTP.
i r
I i-4 i'
SAN ON0FRE--UNIT.2 3.3-9 Amendment No. + N , 142
.w -
4 I
- l l
l l
l 1
i ATTACIIMENT D i l
PROPOSED TECIINICAL SPECIFICATION TABLE 3.3.1-1 SAN ONOFRE UNIT 3 (REDLINE AND STRIKEOUT) l
RPS Instrumentation-Operating 3.3.1 -l
, o Table 3.3.1 1 (page 1 of 2) i Reactor Protective System Instrumentation l
APPLICABLE MODES OR OTHER SPECIFIED SURVE!LLANCE FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE
- 1. Linear Power Level -High 1,2 SR 3.3.1.1 s 111.0% RTP SR 3.3.1.4 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9
- SR 3.3.1.13
- 2. Logarithmic Power Level - High(a) 2(b) 3g 3,3,g,g , ,93g pyp
'1 SR 3.3.1.7 i SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 3. Pressurizer Pressure - High 1.2 SR 3.3.1.1 s 2385 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 l
- 4. Pressurizer Pressure - LowICI 1.2 SR 3.3.1.1 a 1700 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 5. Containment Pressure - High 1.2 SR 3.3.1.1 s 3.4 psig SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 6. Steam Generator 1 Pressure-tow 1,2 ' SR 3.3.1.1 a 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 7. Steam Generator 2 Pressure-Low 1,2 SR 3.3.1.1 a 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 l (continued)
(a) 'c ,, .;, t. t,g m .: t ; 1:4."" 0,pa, ,M ' h a ; ri a", ...: u
'a:nP T.'::"",' t ICL:" , . 1: ' "J ." . T n;L rc;:",
.a h -;...;'i, t,;;,, , , , ,g f ,;,;; : :,n; ;. ... . :;
LCC 2,1,1:, Trip must be enabled when 'M:n"", " W:n logarithmf c power is * = 4E-5% RTP. Trip may be j manually bypassed during physics testing pursuant to LO 3.1.12.
-(b) When any RTCB is Closed.
l (c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between pressurizer pressure and the setpoint is maintained s 400 psia. Trips may be bypassed when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurtrer pressure exceeds 500 psia (the corresponding bistable allowable value is s 472 psta).
- - PCN 500 SAN ON0FRE--UNIT 3 3.3-8 Amendment No. 116 1
1 l
l l-l
L RPS Instrumentation-Operating i 3.3.1 Table 3.3.1 1 (page 2 of 2)
Reactor Protective System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED SURVEILLANCE FUNCTION CONDITIONS
- REQUIREMENTS ALLOWABLE VALUE
- 8. Steam Generator 1 Level - Low 1,2 SR 3.3.1.1 2 20% i SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 9. Steam Generator 2 Level - Low 1,2 SR 3.3.1.1 a 20%
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 10. Reactor Coolant Flow - Low (d) 1,2 SR 3.3.1.1 Ramp: s 0.231 psid/sec.
SR 3.3.1.7 Floor: a 12.1 psid -
SR 3.3.1.9 Stept s 7.25 psid SR 3.3.1.12 SR 3.3.1.13
- 11. Local Power Density - High(d) 1,2 SR 3.3.1.1 s21.0kW/ft SR 3.3.1.3 SR.3.3.1.4 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13
- 12. Departure From Nuc1 ate Boiling 1.2 SR 3.3.1.1 a 1.31 Ratio (DNBR) - LowI SR 3.3.1.2 SR 3.3.1.3 i SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13 (d) Trip ;y L: is ;,, 0 ;ust be enabled when ":^"'. ^=:a logarithmic power is *
- 1.5E-4%M N* RTP, Bypett r".... , ,, ' b ,;.- . ' :^"'. ^^.:" , . ': ' "a During testing pursuant to LCO 3.1.12, trip l may be bypassed below 5% RTP. Bypass shall be ;.;;-;.. s.'b removed when ' :^ ' ^% C" logarithmic power is
- a 5% RTP.
PCN 500 l.
l' l
l SAN 0N0FRE--UNIT 3 3.3-9 Amendment No. 116
1 1
i I
i l
i I
1 1
ATTACHMENT E PROPOSED TECHNICAL SPECIFICATION TABLE 3.3.1 1 i SAN ONOFRE UNIT 2 i l
RPS Instrumentation-Operating 3.3.1 Table 3.3.1 1 (page 1 of 2)
Reactor Protective System Instrumentation
]
I APPLICABLE MODES OR ,
OTHER SPECIFIED SURVE[LLANCE i FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE
'1. Linear Power Level - High 1,2 SR 3.3.1.1' s 111.0% RTP SR ?.3.1.4 .
SR 3.3.1.8 SR 3.3.1.9 SR 3.3.1.13
- 2. Logarithmic Power Level - Highf 'I' 2(b) SR 3.3.1.1 s .93% RTP SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 3. Pressurizer Pressure - High 1,2 SR 3.3.1.1 s 2385 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 ,
4 Pressurizer Pressure- Low (C} 1.2 SR 3.3.1.1 a 1700 psia .
SR 3.3.1.9 SR 3.3.1.12 ,
SR 3.3.1.13 I
- 5. Containment Pressure - High 1,2 SR 3.3.1.1 s 3.4 psig SR 3.3.1,7 SR 3.3.1.9 'i SR 3.3.1.13 ;
- 6. Steam Generator 1 Pressure-Low 1,2 SR 3.3.1.1 2 729 psia j SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 7. Steam Generator 2 Pressure Low 1.2 SR 3.3.1.1 2 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 (continued)
(a) Trip must be enabled when logarithmic power is < 4E-5% RTP. Trip may be manually bypassed during physics l testing pursuant to LCO 3.1.12.
(b) When any RTCB is closed.
i l' (c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between pressurizer pressure and the setpoint is maintained s 400 psia. Trips may be bypassed when pressurizer pressure is < 400 psia. Bypass shall be automatically removed before pressurizer pressure l' exceeds 500 psia (the corresponding bistable allowable value is s 472 psia).
A b SAN ON0FRE--UNIT 2 3.3-8 Amendment No. +27, 142
RPS Instrumentation-Operating 3.3.1 Table 3.3.1-1 (page 2 of 2) 1 peactor Protective System Instrumentation l
APPLICABLE MODES OR OTHER SPECIFIED SURVEILLANCE FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE i
t
- 8. ' team Generator 1 Level - Low !.2 SR 3.3.1.1 a 20%
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 9. Steam Generator 2 Level - Low 1.2 SR 3.3.1.1 a 20%
$R 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 10. Reactor Coolant Flow - LowId) 1.2 SR 3.3.1.1 Ramp: s0.231psid/sec.
SR 3.3.1.7 Floor: 2 12.1 psid SR 3.3.1.9 Step: s 7.25 psid l SR 3.3.1.12 SR 3.3.1.13 l
- 11. Local Power Density - HighId) 1.2 SR 3.3.1.1 s 21.0 kW/ft SR 3.3.1.3 I SR 3.3.1.4 i SR 3.3.1.7 '
SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13 l
- 12. Departure From Nucl ate Boiling 1.2 SR 3.3.1.1 a 1.31 Ratio (DNBR) - Low SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11 SR 3.3.1.12 SR 3.3.1.13 (d) Trip must be enabled when logarithmic power is > 1.5E-4% RTP. During testing pursuant to LCO 3.1.12, trip may be bypassed below 5% RTP. Bypass shall be removed when logarithmic power is 2 5% RTP.
SAN ON0FRE--UNIT 2 3.3-9 Amendment No. W . 142 l
ATTACIIMENT F PROPOSED TECIINICAL SPECIFICATION TABLE 3.3.1-1
,, SAN ONOFRE UNIT 3 4
S
RPS Instrumentation-Operating 3.3.1 Table 3.3.1-1 (page 1 of 2)
Reactor Protective System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED SURVE!LLANCE FUNCTION CONDITIONS REQVIREMENTS ALLOWABLE VALUE 1
- 1. - Linear Power Level - High 1,2 $g 3,3,1,1 s 111,og gyp SR 3.3.1.4 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9 SR 3.3.1.13
- 2. Logarithmic Power Level - High(a) 25U) SR 3.3.1.1 s .93% RTP SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 3. Pressurizer Pressure - High 1.2 SR 3.3.1.1 s 2385 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 4. Pressurizer Pressure - LowIC) 1.2 SR 3.3.1.1 a 1700 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.13
- 5. Containment Pressure - High 1.2 SR 3.3.1.1 s 3.4 psig SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13
- 6. Steam Generator 1 Pressure-Low 1.2 SR 3.3.1.1 a 729 psia SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 1
- 7. Steam Generator 2 Pressure-Low 1.2 SR 3.3.1.1 a 729 psia j SR 3.3.1.7 '
SR 3.3.1.9 SR 3.3.1.13 1
(continued) j (a) Trip must be enabled when logarithmic power is < 4E-5% RTP. Trip may be manually bypassed during physics testing pursuant to LC0 3.1.12. l (b) When any RICB is closed.
(c) The setpoint may be decreased to a minimum value of 300 psia, as pressurizer pressure is reduced, provided the margin between pressurizer pressure and the setpoint is maintained s 400 psia. Trips may be bypassed when pressurizer pressure is c 400 osia. Bypass shall be automatically removed before pressurizer pressure exceeds 500 psia (the correspono..,g hist 4He allowable value is s 472 psia).
SAN ON0FRE--UNIT 3 3.3-8 Amendment No. 116 l
l l
RPS LInstrumentation - Operating i
-3.3.1 l
Table 3.3.1 1 (page 2 of 2)
Reactor _ Protective System Instrumentation j
APPLICABLE MODES OR j OTHER SPECIFIED SURVE!LLANCE FUNCTION CONDITIONS REQUIREMENTS ALLOWABLE VALUE. !
l I l
l 8. 1,2' Steam Generator 1 Level - Low - SR 3.3.1.1 a 20%
l i
SR 3.3.1.7 SR 3.3.1.9.-
!" SR 3.3.1.13 f
L-~
' 9. Steam Generator 2 Level - Low 1,2 SR 3.3.1.1 a 20%
SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.13 j ' 10. Reactor Coolant Flow - Low (d) 1,2 SR 3.3.1.1 damps s0.231psid/sec.
SR 3.3.1.7. Floor: a 12.1 psid SR 3.3.1.9 Step- s 7.25 psid SR 3.3.1.12 SR 3.3.1.13
- 11. Local Power Dens'Ity - HighId) 1,2- SR 3.3.1.1 s 21.0 kW/ft
~
.SR 3.3.1.3 SR 3.3.1.4 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10
!- SR 3.3.1.11 SR 3.3.1.12 i SR 3.3.1.13 l
l~
- 12. ' Departure From Nucl ate Boiling , 1,2 SR 3.3.1.1 a 1.31 l Ratio (DNBR) - Low SR 3.3.1.2 3 SR 3.3.1.3
[
SR 3.3.1.4 SR 3.3.1.5 SR 3.3.1.7 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.11
! SR 3.3.1.12 I
SR 3.3.1.13 i
1.
l l2 '(d) Trip must be enabled when logarithmic power is > 1.5E-4% RTP. During testing pursuant to LCO 3.1.12, trip may I be bypassed below 5% RTP. Bypass shall be removed when logarithmic power is a 5% RTP.
i 4
i 7
!- SAN ONOFRE--UNIT 3 3.3-9 Amendment No. 116 h
y , ,_ _ , , . . _ . _ . _
l j
I l
l i
I ATTACIIMENT G PROPOSED SAFETY ANALYSIS REPORT TEXT SAN ONOFRE UNITS 2 & 3 1
l l
l 1
i
'l i
- - - . . . . - - - . _ ~ . _ . - . ~ ~ . . . . . - . . - . . - - . -.-..- _ - - - -
- r l
San onofre 243 FSAR Updatsd REACTIVITY AND POWER l- DISTRIBUTION ANOMALIES 1 l> 15.4 REACTIVITY ~ AND POWER DTETRIBUTTON ANOMALTES l
15.4.1 'HODERATE FREQUENCY INCIDENTS f
15-4.1.1 Uncontrolled CEA Withdrawal from a Suberitical or Low Power Condition 15.4.1.1.1- Identification of Causes and Frequency Classification
'The estimated frequency of a' control element assembly (CEA) withdrawal frem subcritical or low power conditions. classifies it as a moderate frequency l incident as defined in reference 1 of section 15.0. An uncontrolled withdrawal of CEAs is assumed to occur as-a result of a single failure in the control element drive mechanism (CEDM), control element drive mechanism control system (CEDMCS), or.as a result of operator error.
An analysis of the uncontrolled CEA' withdrawal, both from a saberitical and low power conditiens, is prisented here.
In accordance with the direction given in Sections 15.0 and 15.0.7, additional information which completes the presentation of this event is provided in Section 15.10.4.1.1.
15.4.1.1.2 Sequence of Events and Systems operation The withdrawal of CEAs from suberitical or low power conditions adds reactivity-to'the reactor cere, causing both the core power level and the core heat flux to increase together with corresponding increases in reactor. coolant temperatures and reactor coolant system (RCS) pressure. The withdrawal' motion of CEAs also produces a time-dependent redistribution of core power. These transient variations in core thermal parameters result in'the system's approach to the specified fuel design limits and RCS and secondary system pressure limits, thereby requiring 'the protective action of the reactor protection system (RPS) .
l The reactivity insertion rate accompanying the uncontrolled CEA withdrawal is dependent primarily upon the CEA withdrawal rate and the CEA worth since, at subcritical and lower power conditions, the normal reactor feedback mechanisms do not occur until power generation in the core is large enough to cause changes in the fuel and moderator temperatures. The reactivity insertion rate determines the rate of approach to the fuel design limits. Depending on the system initial conditions and reactivity insertion rate, the uncontrolled CEA withdrawal transient is terminated by either a high logarithmic power trip (for the subcritical initial condition), a high power level trip, high pressurizer pressure trip, a low departure from nucleate boiling ratio (low DNBR/VOPT), or a high local power density trip. The secondary system pressure increases following reactor trip and is limited by the steam generator safety valves.
-5 Table 15.4-1 and 15.4-2 give the sequence of events for the limitin CEA to 7, withdrawal transient from suberitical (10 ldPower) and low power % power) conditions, respectively. JThe sequence of events past 75.7 seconds shown in ,
l (table 15.4-1 is from cycle 1. -They are representative of the latest cycle (see G-s '
l .
section 15.0.7).
Ns-
'1/98 15.4-1 Revision 13
_. ____ - _ _ _ . - . _ _ - _ , _ ,_ _ , _ - ~ . .
. .- ~, . . . ~ . - . . .. . _ . . - - . _ - - . - . . . -
San Onofra 2&3 FSAR Updated REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4-1 SEQUENCE OF EVENTS FOR THE UNCONTROLLED CEA WITHDRAWAL FROM SUBCRITICAL CONDITIONS Setpoint or
. Time (sec) Event Value 0.0' Initiation of Uncontrolled Sequential CEA ---
Withdrawal. <
- '33 g, !] O Reactor Reaches Critically ---
g g ,3 -M-0-- Reactor Reaches High Logarithmic Power Trip Setpoint
% of Rated !
58 7 25,-Te Reactor Trip Generated ""~
$9 7 CEAs Begin to Drop ~
59 7 ?! C g s2t y.
Peak Reactor Core Power Reached #'
of 3410 MWt 59.% - m O' V Peak Reactor Core Heat Flux Reached
'+ 5. 2.
0 Tt of 3410 MWt 59.% -Hht-- Minimum DNBR Occurs > 1.31
?
?::k T.00 Tr ::er: Occur " ' " " ' -
00.0 Sini ;- Tr ;;uri;a. Ot::: 'Jc 1;;; "? ft'
'8 Original accident analysis assumed a 0.3 second delay between the time the reactor trip is generated and the CEAs begin to drop. Present analyses allow ,
up,to 1.01 second holding coil delay time as part of the overall averaga 3.4 second CEA drop time. Both SONGS Units 2 and 3 CEA drcp time measurements have very little margin to the current.- technical specification limit of 3.2 ,
seconds. As discussed in Section 15.0.2, three (3) CEA drop time curves were used for the safety analysis wit!i the longest delay time of 1.01 seconds for a drop time of 3.4 seconds.
Core operating limit supervisory system (COLSS)_
and core protection calculator system (CPCS) are to be modified to accommodate any loss or gain in thermal margins based on the measured average
}. CEA drop time in accordance with the revised Technical Specification 3.1.5.
The additional delay results in increases in peak core power and peak heat flux. However, the acceptance criteria for this event continue to be '
l satisfied and the c^onclusions of the analy' sis remain valid.
~
de
' e l
?
r 1/98 15.4-2 Revision 13
l' San Onofro'2s3 FSAR Updated r REACTIVITY AND POWER DISTRIBUTION ANOMALIES Table 15.4-2 SEQUENCE OF EVENTS FOR THE UNCONTROLLED CEA WITHDRAWAL V .--- ~ ~~ FROM LOW POWER CONDITIONS
.... .. / / <
Time (sec) Event Setpoint or Value
- f. 0.0 CEAW Initiated ---
3 150.5 High Pressurizer Pressure Trip 2475 psia s
.. Condition 151.4"' High Pressurizer Pressure /VOPT ---
! Reactor Trip Occurs 151.7"' Scram CEAs Begin to Drop ---
152.1 Pressurizer Safety Valves Open 2525 psia 152.9 Peak RCS Pressure 2640 psia 153.1 Peak Core Power 75.4% of 3410 MWt 153.8 Peak Core Heat Flux 61.8% of 3410 MWt (
j t 153.8 Minimum DNBR > 1.31 *
} 155.7 Pressurizer Safety valves Close 2400 psia '
\
"' Original accident analysis assumed a 0.3 second delay between the time the raactor trip is generated and the CEAs begin to drop. Present analyses allow up to 1.01 second holding coil delay time as part of the overall average 3.4 second CEA drop time. Both SONGS Units 2 and 3 CEA drop time
- measurements have very little margin to the current technical specification limit of 3.2 seconds. As discussed in Section 15.0.2, three (3) CEA drop time curves were used for the safety analysis with the longest delay time of
~1.01 seconds for a d cp time of 3.4 seconds. Core operating limit supervisory system (COLSS) and core protection calculator system (CPCS) are to be modified to acccmmodate any loss or gain in thermal margins based on the measured average CEA drop time in accordance with the revised Technical
\ Specification 3.1.5.
Sufficient conservatism exists in the CPC variable overpower trip (VOPT) to compensate for the additional delay. The
(
conclusions of the analysis remain valid.
. . , _ . _ _ . . . ~ ~~~~ ~ .~
f
) Y(kM. LU 5%i hb 1/98 15.4-3 Revision 13
--o ' ' ' '- * * " ' ~*"" -"*-*~*"- * ' ' ' " * ~ ~ ~ ~ - " " - " " " ~ ' " " ' " " ~ ~ ~ - " ~ * ' - - * ~ ^ * - "
d
. .,,M"**** _
- %.g _
i 1 ,
3 N Time (sec) Event 0 00 CEAW Indisted Setpoint or Value s'%
1 34 40 V0PT Tno Conddion , +
35 % of Rated 3480- Reactor Tno Occurs . }
35 81 Scram CEAs Socin to Orop .
}. . 35.82 Peak Core Power . 107.19 % of Rated .
Fuel Centerline Temperatura t.
< 4706 *F ,
35 98 Peak Core Heat Flux 47.66 % of Rated .
35 se Minimum ONBR
>1 31
'/'/
- l
~
h
^
j7
/ 4 a.y m ( !
~f
- +
i e ,
, 1 t
I, 1
.1 1
i E
k-4-
('
l 4
D
_ - .- - . - . . _ -. ~ _ . _ _ _ . - - - - - . _ . - ~ _ _- .- _ _ - _ -
i San Onofra 2&3 FSAR Updatsd REACTIVITY AND PohE2 DISTRIBUTION ANOMALIES 15.4.1.1.3 Core and System Performance
- A. Mathematical Model l
The nuclear steam supply system (NSSS) response to a' CEA withdrawal from saberitical or low power conditions was simulated using the CESEC III computer. program described in section 15.0. The thermal margin on DNBR in the reactor core was simulated using the CETOP-D computer program chapter 4. in section 15.0 with the CE-1 CHF. correlation described in described B. Input Parameters and Initial Conditions The input parameters and initial conditions used to analy:e the USSS response to a CEA withdrawal from suberitical and low pcwer conditions l are discussed in section 15.0. In particular, those parameters which were unique to the analysis for each event discussed belew are listed in tables 15.4-3 and 15.4-4 for a CEA withdrawal frem suberitical and low power conditions, respectively.
1.
CEA Withdrawal Event From Low Power-Conditions l
The initial conditions and NSSS characteristics assumed in this analysis have been determined to be the limiting set of conditions allowed by the limiting conditions for operation (LCos) in terms of providing the closest and fastest approach to the fuel design limits for a CEA withdrawal from low power level. The initial conditions
! which provide the closest and most rapid appreach to the fuel design 9 5 ss/', = limits correseced to a zero power core inlet temperature of(jIk'F, a ,sfo corein1tf1owoft{pgpofeestynflow, ane the mint =um RCS pressure '
'~
of 2000 psia. The initial RCS pressure is chosen to be the lowest ,_ -
llowed pressure within the LCos since.this,Pi c'~ - e tr2n;;est l
' ..g L h M y g sconse to r A -
utM ra m t+ 7 -~ a a- ** k ?=-Aim; i.;
y elafing ;;tuati;;. c' Li. 1 6 . prn r r4-=- r -= e e'2 r 2 tri; Us
$ *E Q_ point af 2'.22 reia * ~-Mir! ' rrert airti-s 1/ The initial core 3)N D- average axial power distribution ass ed in the analysis corresponds to an axial shape index (ASI) of[1c5c- Studies have shown that l
this initial shape undergoes a signi.1 cant cnd rapid shift to the l
top of the core during the transient. A one pin radial peaking factorof,p-St, including uncertainties, is also conservatively a ssumed for this analysis. The radial peaking factor is the highest
,,g,gf radial peak expected for any CEA configuration and time in core lifetime.
i .
.Q 3 1
I
}
3 1
1/98 15.4-4 Revision 13
San Onofra 2&3 FSAR Updated 4" REACTIVITY AND POWER MN DISTRIBUTION ANOMALIES Parametric 4nalyseshave indicated that the lowest initial power level @ 0'i suberitical core condit16), results in the closest and fastest apprcach to the ruel cesign limits during the CEA v withdrawal transient. Initially suberitical, :ero pcwer CEA withdrawaltransientsareterminatedbythehighlogarithmicpower)g level trip while those initiated frem a power level just abeve the "" *e high logarithmic power level trip bypass setting Cof 10"1 power are D c' " '
terminated by a high pressurizer pressure trip, CPC low DNER (v0PT) fo%...
trip, er high transient. At local powerpower levels density above @ trio at a much reactivity later time feedback - into the
._ go Q mechanisms prevail and provide a dampening etreet en the severity cf the transient.J The mc s positive mocerator temperature c: efficient fof +0.5 x 10 ok/k/*F is assumed for this analysis. Also, a conservative fuel temperature c: efficient multiplier of 0.75 was The regulating CEAs are initially in the fully inserted positien when the CIA withdrawal is initiated. Based en calculated CIA worths and the maximum CEA withdrawal rate of the CEA drive system, the assumed rate is conservative. J ror this ana;ys:.s, tne reactivitf' nsert:.on is the maximum expected rate of 1.1 x 10" 2 /s fer a CIA (
withdrawal fr:m low power conditiens. This rate correspends to approximately twice the largest insertion rate expected fr:m the sequential withdrawal of the CIA groups with 40% cverlap at the l Qmaximum
_ speed of 30 in/ minute.f -
Table 15.4-3 ASSUMPTICNS FOR THE UNCONTROLLED CEA WITE3RAWAL FRCM SUBCRITICAL CONDITIONS Parameter Value
-5 Initial Core Power Level. MWt 3.&-t v so 24 : -
Initial Inlet Coolant Temeerature. *F Ed ^^
~i~nitial Cere Mass Flow ram 1:' 1... 'h. M Yft*~~**
neWmn L)
Initial RCS Pressure, psia -2000 Moderator Temperature Coefficient, 10 "do/* F 0.5 Fuel Temperature Coefficient Multiplier ~4' O*86 ~
i y t a .,- ,- e s su
_ 3 ,. 4 Maximum Reactivity adM-irn Rate, x 10"ac/sec -.: 2N Insc-te n l
t 1/98 15.4-5 Revision 13 l
t l
i
San One'.ra 2&3 FSAR Updatsd REACTIVITY AND POWER DISTRISUTION ANOMALIIs Table 15.4-4 ASSUMPTIONS FOR THE UNCONTROLLED CEA WITHDRAWAL FROM LOW POWER CONDITIONS Parameter Value Initial Core Power Level,.HW .
r y
\c.5 MWt 333 1- 1 1 Initial. Inlet ~ Coolant Temperature. 'T Sform 1
-5::
3/6. Jet %>
Initial Core Mass Flow Rate _19 1hr 'hr_(9 5'/./ <.%d ,3e~ . 5:.:
Initial RCS Pressure, psia 2000 l Moderator Temperature Coefficient, 10"ac /
- F 0.5 Fuel Temeerature Coefficient Multiplier C'85 ~ ~5 Minimum CEA Worth at Trio, too
-5.15 Maximum Reacti vit.y .J i ~ Rate. x 10"ac/se: 1. C :.;
16serti.o For those transients classified under the category of reactivity and power distribution ancmalies, the uncontrolled CEA withdrawal at low power conditions has been shown to be the mest limiting transient l for maximum RCS pressure. The initial conditions and NSSS I characteristics used for this particular analysis are essentially _
identical to the above analysis.,FThe core inlet temperature assumed 1 l Fis the minimum value of 520'F. The lowest core inlet temperature is L:ed since this keeps the steam generator safety valves frem topening.f'In addition, no credit is taken f or the ture ne bypass valves, thereby maximi:ing the peak RCS pressure reached during the course of the transient. The initial parameters used in this analysis are shown in table 15.4-4. --
'{p-2.
CEA Withdrawal Event frem Saberitical Conditices CL V
- 5. GO'In d ,
An uncontrolled CEA withdrawal event from suberitical conditions was !
initiated at the power level ([f 10"% . The input parameters and initial conditions used to analyze this event were si=ilar to those of the low power CEA withdrawal analysis.
- g. ,,e
' IFor the subcritical conditions, the maximum reactivity addition rate !
of 1.9 x 10" oc/see was used. Also, minimum CEA worth at trip o'
(-4_.0 %co was_used it this analysis.,('The input parameters used in this analysis are shown in table 13.4-3.
C. Results The dynamic behavior of important NSSS parameters following a CEA withdrawal from suberitical and low pcwer conditions is presented in Figures 15.4-1 through 15.4-10.
1/98 15.4-6 Revision 13 i
l
San Onofra 2&3 FSAR Updated REACTIVITY AND PCWF.R DISTRIBUTION ANOMALIES 1.
CEA Withdrawal from Suberitical conditions N"
%d s h *7 The uncontrolled CEA withdrawal from suberitica conditions resulted in a reactor trip on high logarithmic power at 75.s seconds. The minimum DNBR calculated for this event initiated from the conditions of table 15.4-3 was greater than the design limit of 1.31. *he peak linear heat generation rate (PLHGR) was calculated to be .: . : :kw/f t which is in excess of the steady state acceptable fuel center Ine'- gy, f D melt (CTM) limit of 21 kw/ft. However, the fuel centerline temperature was less than 4706*F and, thus, the fuel is not predicted to melt. l The fuel cycle length extensions approved by the NRC for SONGS 2 and 3 required core designs containing erbia (Er:0 ) integral burnable 3
absorber in place of the traditional boren (B.C) discrete burnable absorber to control the MTC at the beginning of the cycle and the pin peaking factor throughout the cycle.
Center line fuel melting temperature is a criterion, which is adhered to in the safety analysis during plant transient conditions, when the Peak Linear Heat rate limit set in the Technical Specifications is exceeded. This temperature is also burnup dependent and is thus calculated on a cycle specific bases.
The NRC approval of the methodology for core designs containing erbium burnable absorbers (Reference 6) requested that the Erbia melt temperature limit be decreased linearly with Erbia content such that the limit at zero weight percent be equal to the UO melting temperature limit, and at the highest Erbia weight percent be at least 1 degree Celsius below the lowest melting temperature for which experimental data is available.
For the e design with Erbia integral burnable absorber rods, the maximum approved Erbia concentration is 2.5 wtt fuel, and the maximum assembly burnup is 60 GWD/MTU. These two factors create a center line fuel melting temperature limit of 4706*F.
Additionally, the peak RCS pressure is less than the design limit of 2750 psia. Table 15.4-1 presents the sequence of events for this event. Figures 15.4-1 through 15.4-5 present the NSSS response for core power, core heat flux, RCS temperatures, RCS pressure and steam generator pressure.
- 2. CEA Withdrawal from Low Power Conditions The low power CEAWs were analyzed to maximize the RCS pressure l increase and to maximize the potential for the fuel degradation. The
! initial conditions for the low power CEAW that maximizes peak RCS pressure are listed in table 15.4-4. h CPC low DNBR (V0PT) trip
- sa Feel
% Act Aegra dd.o' +A e.
%rnadr l
l 6/98 15.4-7 Revision 14 l
Stn onsfro 243 FSAR Updntcd
.I vipv4 .
REACTIVITT AND POWER j DISTRIBUTION ANOMALIES is credited to mitigate the consequences of this. event. A parametric study on the reactivit'y addition rate was performed to yield trip a coincident CPC low DNBR (VOPT) and,high pressurizer pressure in order to maximize the peak RCS pressuregA-high pressurizar ]
pressure trip and CPC low DNBR (VOPT) trip are generated at 151.4 j
seconds and the scram CEAs begin to ,' drop at 151.7 seconds Thtpeak)- ,
3 !
yessug40 ysia and occurs at _152 9a J aconds.f The. sequence I of events is presented in table 15.4 ~2. , "Figi:res 15.4-6 througn l 15.4-10 present the NSSS response for this event. ,"
~
i
- 15.4.1.1.4 Barrier Performance - '
A. Mathematical Model '
~
4 The mathematical model used for evaluation of bar:5.er performance is identical to that described in paragraph 15.4.1.1.3. , , , ,
l B. Input Parameters and Initial Conditions !
, i
' The input parameters and initial conditions used for evaluation of ;
1 barrier performance are identical to those described in paragraph 15.4.1.1.3. ' '
C. Results I
.., l
,- 1 l Figures 15.4-3 and 15.4-8 show the NSSS iesponse for'RCS pressure for a l CEA withdrawal from suberitical and low power 'cenditions. The peak RCS pressure for a CEA withdrawal from suberitical conditions is .less than that of the CEA withdrawal from lower power. The most li=iting case of RCS pressure for the CEA withdrawal at low power is 64 psia which is less than the design limit of 2750 psia.
15.4.1.1.5 Radiological' Consequences
. MSS .
The- radiological consequences due to steam releases from the secondary system are less severe than the consequences of the inadvertent opening of the atmospheric _ dump valve discussed in paragraph 15.1.1.4. ,,
15.4.1.2 Uncontrolled en withdrawal at power ," ,"
15.4.1.2.1_ Identification of Causes and Frequency Classification -
' The estimated frequency of a CEA withdrawal at power classifies it as a moderate frequency incident as defined in reference 1 of section 15.0. A CEA withdrawal is assumed. to occur as a result of a single failure in the CEDM or CEDMCS.
In accordance with the direction given in Section's 15.0 and 15.0.7, additional information which completes the presentation of this event is provided in Section.15.10.4.1.2. .
1/98 15.4-8 Revision 13
,. , _ , . . - __..__...__._._..-_.___.___.._-__......_.m . . _ _ -
-02610 i
150 I i i I
(
125 _
I h
.. 10e _
~
e a
nas l
W 75 - -
2 ,
) i 1
= . _ ~
l S
25 _
i 9 .
- [/ 0 I I I lti 0 20
/ ,
40 60 80 1c0 TDE. SECON05 Upda.ced SAN ONOFRE NUCLEAR GENERATING STA N Unita 2 & 3 i'
i CEA VITEDRAWAL FROM SUBCRITICAL l3 CORI PC'4R vs. TIE Figure 13.4-1 2/87 Revision 3
.. . . _ . _ . . _ - . . _ . . . . .- . - = - . - . _ . . . .. .
120 t I i j IM .\ _
i h -
g M - -
E E
- r. a E 3 1 a . !
> i G n - _ l E l 8 -
O I I I '\
0 -
20 40 60 8 100 TIME. SECCMD5 Y
L O
/J Updated 8AN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CEA WITHDRAWAL FROM SU3 CRITICAL 3 CORE HEAT FLUX vs. TLE Figure 15.4-2 2/87 Revision 3
._ 0 2 .' I Q. _ _ _
l l
l L l 2700 I I i i i
2500 -
l -
q lll I
= 1 W 2300 -
)
5 de _
1 1
g .
E
$ 2100 - --
>~
- 3 5 W _
-$ ^
5 1900 - -
1 l - h l
= l 1700 - -
1500 I I I -I ~
0 20 40 '60 80 -
100
_ TIME. SECCN05 Cs b -
.M4 l
'pdated SAN ONOFRE NUCLEAR GENERATING STATION Unia 2 & 3 CEA WITHDRAWAL FROM SUBCRITICAL RCS PRESStJRE vs. TIME I
I Figure 15.4-3 2/87 Revision 3
L i .' . ?
/
. 650
/
l i ! i
/
no -
/ a
- i 550 -
g - i E /
- * /
E /
1
=
5 5= -
,/ _
3
. 5 ,
$'450 5
- ~
! k
- W r
= -
I I I ~ l 350 0 20 40 60 . 80 100 ,
TIME. SECCN05
' - Updated
,A, SAN ONOFRE s
NUCLEAR GENERATING STATION I . Units 2 & 3 CEA WITHDRAWAL FROM. SU3 CRITICAL 3 RCS TEMPERATU?iS vs. TIPI Figure 15.4-4 2/87 Revision 3
i .
l 0 20IC l
l 1200 i i i , I l
l l
1000 _
l g 800 _
J I
w E
600 -..
5
- 3 3
W W
g =0 _ _
m 200 _ _
0 I I I '
0 20 40 60 80 100 TINE. $1CCN05 f
Upda:ed l
/
o -M SM4ONOfAE
, J,f' NUCt.EA& GENERATING STATION r ,
Units 2 & 3 CEA WITEDRAWAL FROM SUBCRTICIAL 3 STEAM CENERATOR PRESSURE vs. TIME t
Figure 15.4-5 2/87 Revisien 3
10 A ; ,. , ,
20 I i i e i 100 6 _
1 .0 -
S
- I E
g so - -
2 4 -
I w 40 - -
8 w
20 - -
\ l l 0 <
O 30 60 90 120 150 180 TIME, StC2 05 '
\
'l t
},,, Updaeed SAN c'!!)4.lE NUCLEAR GENE.R ~ LNG STATION Units 2 &
CEAWITEDRANALATLbPOWER ,
CORE POWER vs . TI.".r.,
Figure 15.4-5 2/87 Revision 3
- i. .. .
.,....c. .- -. - , - . . . - . - - _ . . - - . . -- -- ..- -. .. .. -.
/ C- a 120 i
I i - i i l i
100 - -
E s;r; .0 - -, l
- I E
t:
E 60 - _
'E 3
d
= = - -
w 5
u to - _.
a i
i l I 0
O 30 60 to 120 150 180 TIME. SECGIOS
)#
- C Updated SAN ONOFRE
- NUCLEAR GENERATING STATION Unit 2 & 3 CF.A WITHDRAWAL AT LOW POWF.R
.i 3 4
CORI EE.AT TLUX vs. TIME figure 15.4-7
'2/87 Revision 3
'v J.- l'
's i
1 1
I l
l
{
2600 i i i I
i' i
\
t 1 i
=
2500 ~
l-e .
l a l
C
- 2400 - - I 5
g W
5 0 2300 - --
E .
I I'
w
= -
g 2200 w
M W
- l l
= -
- 2100 .
l I
- l l
' I I I I 2000 .
~
0 30 60 90 120 0 180 i
~
f TIME,$2020$ -
R -
/
g Updated SAN ONOFRE NUCLEAR GENERATING STATION l Units 2 & 3 CEA WITHDRAWAL AT LOW POWER i RCS PRES 5URE vs. TI."
! Figure 15.4-8 2/87 Revision 3
l 1
I I I I i 4
42 0 e
.Z j 600 - _
w 5
I
=
g ua - _
E l
1 3
w ~
5 t T _
$60 - OUT 5 l ti<
W 7Avg _
540 -
T gy I I
$20-0 30 60 90 120 1 180 TIM SEC305 - -
[.
n.
r, ,
,\)
, RCS TEMPERATURES vs. TIME l3 figure L5.4-9 2/87 Revision 3
i A e o
g Q* b ~J i
l 1 0 s i 4 .
i 1000 . -
h5 - , /
l 940 . .
EN
= . l I 920 - .
m W
- 900 - .
l E
w lm 3
3 l'
I 5 aso .
m 840 =
820 = .
800 - -
1 1
780 O 30 60 90 12 150 180 TIME. SEC3 05 o
Ayi- Updated
{ SAN ONOFRE\
I NUCLEAR GENERATING STATION l Units 2 & 3 l
! CEA 'a'ITEDFXa*AL AT LOW PO'7.R i STEAM GEFERATOR PRESSURE vs. TU'.I 3 i
rigure 15.4-10 2/81 Revision 3
l i
i 1.4 ' s s ,
t h 1.2 -
2 _
o~ ,
w" ~
o 1 -
- u. _
O z
- O- 0.8 -
P -
O
' E
- u. 0.6 -
W w
h c.
0.4 -
W T
o 0.2 -
0
' ' l' 0 '
O 20' 40 60 80 100 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION UNrf5 2 & 3
~
CEA WITHDRAWAL FROM SUBCRITICAL POWER FRACTION vs TIME FIGURE 15.4-1 l
l i-
h 0.5 , ,
2 , ,
3 0.45 -
E ~
o 0.4 - -
^'
r 9
+ 0.35 -
- g. .
u.
0.3 -
y a
0.25- -
LL.
. s 0.2 -
w I 0.15
~
w -
e -
.y 0.1 -
w
~
< 0.05 -
w -
C '
0 0 ' 1' '
O O 20 40 60 80 - 100 TIME, SECONDS _
r SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 CEA WITHDRAWAL FROM SUBCRITICAL HEAT FLUX FRACTION vs TIME
, FIGURE 15.4-2
l 2220 ,
l g i . .
(O . 1 1 2200 -
l u.i -
i .c -
~
l @ -2180 -
. . e w
{ 2160 -
2 l $<n :
2140 -
- 2120 -
g _
z 0 -2100 -
8 o .
e 2080 -
O
& b j 2060 - -
W .
. c.- ,
2040 ' '
O 20 40 60 - 80
_100 TIME, SECONDS -
t SAN ONOFRE !
NUCLEAR GENERATING STATION UNITS 2 & 3 i
l CEA WITHDRAWAL FROM SUBCRITICAL '
l RCS PRESSURE vs TIME FIGURE 15 A-3 U ,
l -
,- _ s- - .. . . _ _ . . . _ . _ _ . - - _ . - _ _ - _ -
585 ,
TIN -
TAVG --
9 580 -
l: TOUT -
0 !!
W o . -
@ 575 -
i! -
cc : i 3
F-~
< ii
@_ 570 -
- ll\.1 ct \ ,;
2 i
- 3 \i; 't W
H 565 - ..-..'
2 i \ ' "' '
g ..
3 ,
O O t o 560 '
555 ' '
O 20 40 60 80 100 TIME, SECONDS -
SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 CEA WITHDRAWAL FROM SUBCRITICAL RCS TEMPERATURES vs TIME
, FIGURE 15.4-4
L i
1148 ,
I g R!GHT SG
< 1146 -
LEFT SG -
55 -
c.
[ 1144 --
o -
.cn L
d W
1142 -
c.
8 1140 -
W w 1138 -
w z -
0 2 1136 -
m 1134 -
W -
1132 ' ,
100 T E, SECON S SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 CEA WITHDRAWAL FROM SUBCRITICAL STEAM GENERATOR PRESSURE vs TIME FIGURE 15.4-5
s
-1 . 2 --
i 1-t I.
O '. 8 - -
c -
O
. 's ,
o i n
- k. 0.6 u
e 3
o n.
0.4
- 0. 2 -
0 - -
0 20 40 60 80 100 Time, Seconds SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 _
CEA WITHDRAWAL at LOW POWER POWER FRACTION vs TIME
, FIGURE 15.4-6 1.
h l
-.,w,--wr Tr 7 - - ==wr w
1 t
0.9. _ . . . ._ _ __.. __________ .
i ; .
t ,
,h O.45.' I'
- 0. 4 . ,
y ..
- 0.35
! c .
o r4 i
$ 0.3 e -
u - -
l w :
I u 0.25 - l s' .
.-4 m
- 0. 2 -
e e i
- j. m .
0.15 -
0.1 0.05 1
- 0. . . - - . - . - - -_.--
0 20 40 60 80 100 Time, Seconds l
SAN ONOFRE
, FIGURE 15.4-7 l
l t
I i
2300 , -
l' I
l 2250 J l
7
.a .
l I
g .
1
&'2200 - !
, l 64 -
a
.n . 4 e -
J 2150 m
O i
g .
,$ 2100 x .
4 x- ' l
.]
2050 -
2000 5 -- - - - - - - ---- - - - - - - - -
0 20 40 60 80 100 Time, seconds-SAN ONOFRE NUCLEAR GENERATING STATION l
UNITS 2 & 3 CEA WITHDRAWAL at LOW POWER RCS PRESSURE vs TIME FIGURE 15.4-8 I'
, +
0 r
i
- t. .- . . . , . - - . , , - .. - - - - . -
N 1 590 - . - -
- Core Outlet Temperature (F) *
! :--- Coolant Channel Temperature tr) 585 i- q Core Inlet Temperature (r) i
-l . 1 580 I ,
' j
-Q w
a -
i e i i
-oW 575 i sa . t e .
u ,
e -
Q. ' i
! 570 s ..
m '
O s
565
~
560 -
9 4
555 - - - - - - - - - - - - - . .. - - - - - . . .
0 20 40 60 80 100 Time, Seconds SAN ONOFRE NUCLEAR GENERATING STATION
~"
UNITS 2 & 3 .
CEA WITHDRAWAL at LOW POWER RCS TEMPERATURES vs TIME l FIGURE 15.4-9 i
i l
t
t I i i l I
l' 1152 - - )
~
LH SG Pressure (psia) l 1150f- ....- RH SG Pressure (psia). ,
1 i
1148 1 .
j
, - 11i 6 I
- g. .
.a
~
E 114 4 '
e 4
a 1142 -
m -
n e
4 -
!- A 1140 -
e en 1138 - 1 l
l l
1136 1134 l
1132 - - - - - - - - - - - - - - -- - - - - - - - - - - - - l 0 20 40 60 80 100 Time, Seconds i
l~
SAN ONOFRE NUCLEAR GENERATING STATION UNITS 2 & 3 CEA WITHDRAWAL at LOW POWER STEAM GENERATOR PRESSURE vs TIME
! FIGURE 15.4-10 i
f
, , , , - - ~