ML20197C728
| ML20197C728 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/19/1997 |
| From: | SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20197C687 | List: |
| References | |
| NUDOCS 9712240260 | |
| Download: ML20197C728 (86) | |
Text
{{#Wiki_filter:- ..
+ :
Pressurizer i
',; 3.4.9 [
3.'4 REACTOR COOLANT SYS'EM (RCS)
.f 3.4.9 Pressurizer f f
LCO 3.4.9 - The pressurizer shall be OPERABLE with: l
- a. ' Pressurizer water volume s 900 ft3 ; and
[;
'F Two groups of pressurizer heaters OPERABLE with the
,apacity of each group e 150 kW and capable of being powered from :n emergency power supply.
7 APPLICABILITY: MODES 1, 2, and 3. ; FACTIONS- ! F CONDITION-- REQUIRED ACTION- COMPLETION TIME : i A. Prtssurizer water A.1 Be in MODE 3 with 6 hours ; volume not within reactor trip breakers i limit, open. ; E . i A.2 Be in MODE 4. 12 hours i B. One required group of- B.1 --Restore requirei 72 hours ; pressurizer heaters group of pressurizer inoperable. heaters to OPERABLE status. , t C. ;Mequired Action and C.i Be in MODE 3. 6 hours 4 associated Completion-Time of- Conditton B AtQ + not met, . . C.2 Be in MODE 4. 12 hours U" M aa g g7 g j _ SAN ONOFRE UNITJ21 3.4+25- Amendment No.1127 ,
-- i if ' '[
, . , ,2- , _ _ ,; . v _
,J _ _, - . _ , .._
pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water volume s 900 ft3 . 12 hours SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters a 150 kW. w. SAN ON0FRE--UNIT 2 3.4-27. Amendment No. 127
Pressurizer B 3.4.9 BASES BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief. If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psig. The requirement to have two groups of pressurizer heaters : ensures that RCS pressure can be maintained. The : pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled. Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat. APPLICABLE in MODES 1, 2, and 3, the LCO requiremet., for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer, in making this assumption, the analyses neglect the small fraction of noncondensable gases normally present. Safety analyses presented in the UFSAR do not take credit for pressurizer hecter operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NVREG 0737 (Ref. 1), is the reason for their inclusion. The requirement for emergency power supplies is based on NUREG-0737 (Ref. 1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, tirr.e period after a loss of offsite r,0wer. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in the accident analyses. (continued) SAN ON0fRE--UNIT 2 B 3.4-47 Amendment No. 127
Pressurizer i B 3.4.9 l s BASES ! o _ APPtir.ABLE ' The pressuracr satisfies Criterion 3 of the NRC Policy : SAFE 1Y ANALYSES Statement. l (continued) C0 The LCO requirement fgr the pressurizer to be OPERABli with water volume < 900 f t ensures that a steam bubble exists. ! Limiting the uaximum operating water volume preserves the -i steam space for pressure control. The LCO has been , established to minimize the consequences of potential overpressure transients. Requiring the presence of a steam
- bubble is also consistent with analytical assumptions.
1he LCO requires two groups of OPERABLE pressurizer heiters, ! each with a capacity a 150 kW and capable of being powered , from an emergency power supply. The exact design value of 150 kW is derived from the use of three neaters rated at 50
- kW each. The amount needed to maintain aressure is ,
dependant on the ambient heat losses. Tae minimum heater capacity required is sufficient to maintain the RCS near normal onerating pressure when accounting for heat losses through che pressurizer insulation. By maintaining the ! preseure near the operating conditions, a wide subcooling , margi.. to saturation can be obtained in the loops, APPLICABILITY The need for pressure control is most aertinent when core : heat can cause the greatest effect on RCS temperature : resulting in the greatest effect on aressurizer level and RCS pressure control. Thus, Applica)ility has been designated for MODES I and 2. The Applicability is also provided for MODE 3. The purpose is to prevent solid water i RCS operatiun during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LCO does not apply to MODE 5 (Loops filled) because LC0 3.4.12 " Low TemperatureOverpressureProtection(L10P) System," applies. . The LCO does not apply to MODES 5 and 6 with partial loop operation. In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply, in the event of a loss of offsite power, the initial conditior.s of these MODES gives (continued) ] SAN ON0fRE--llNIT 2 8 3.4-48 Amendment No. 127 , s - ,- y n u ,,,e c ,-:-, -- . , , - n,,. n e,- cv, -,, ---
. - - - . . , ,w-... --r-,,--, - , . -
4 ATTACilMENT 2 EXISTING TEttlNICAL SPECIFICATIONS AND BASES UNIT 3
l Pressurizer l 3.4.9 -! 3.4 REACTORCOOLANTSYSTEM(RCS) ~! l 3.4.9 Pressurizer -! l LC0 3.4.9 .The pressurizer shall be OPERABLE with: .
- a. Pressurizer water volume s 900 ft ; and 3 fr
- b. Two groups:of pressurizer heaters OPERABLE with the !
-capacity of each group = 150 kW and capable of being !
powered from an emergency power supply.- ;
-APPLICABILITY: MODES 1- 2. and 3. !
SACTIONS i CONDITION- REQUIRED ACTION. COMPLETION TIME- l A. Pressurizer water A.1 Be:in MODE 3 with 6 hours
' volume not within reactor trip breakers
' limit. open.
- A.2 Be in MODE 4. 12 hours
+
'B. One-required group of B.1 Restore required 72 - hours -- [
pressurizer heaters group of pressurizer ; inoperable. heaters to OPERABLE ? sta us. i C. Required Action and C.I. Be in MODE 3. 6 hours i associated Completion * ; Time of Condition B lLNQ
- not met. :
C.2 Be in MODE'4. 12 hours & l 1 i k
' SAN'ON0fREl -UNIT 3 3.4 26- Amendmant No. 116 ,
P 4 + i.p.'., :.-,..-,-m.-- , . , , - e e v m , , m v ,-,.ap-- - e- ,w,.wr-..+ ,a.ny-- ,, .c , n, r w ,- eswrn
Pressurizer 3.4.9 SVRVEllLANCE REQUIREMENTS SURVEILLANCE FREQUENCY , SR 3.4.9.1 Verify pressurizer water volume 5 900 ft3. 12 hours SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters e 150 kW. 1 SAN ON0fRE--UNIT 3 3.4-27 Amendment No, 116
Pressurizer B 3.4.9 BASES BACKGROUND (Pressuriter safety valves) can control pressure by (continued) steam relief rather than water relief. If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to-water relief, the maximum RCS pressure might exceed the Safety Limit of 2'/50 psig. The requirement to have two groups of pressurizer heaters ensures that RCS pressure c."i be maintained. The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled. Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat. APPLICABLE in MODES 1, 2, and 3, the- LCO recuirement for a steam bubble-SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present. Safety analyses ) resented in the UFSAR do not take credit for pressurizer Teater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. Although the heaters are not specifically used 1.1 accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG 0737
-(Ref.1), is the reason for their inclusion. The requirement for emergency power supplies is based on NUREG 0737 (Ref. 1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in lthe accident analyses.
(continued). SAN ON0fRE- UNIT 3 8 3.4;47- Amendment No.- 116
i Pressurizer B 3.4.9 ;
-BASES.
f APPLICABLE The. pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement. (continued) l 1 LC0 TheLC0requirementfgrthepressurizertobeOPERABLEwith water volume < 900 f t ensures that a steam bubble exists. Limiting the maximum operating water volume preserves the j steam space for pressure control. The LC0 has been established to minimize the consequences of potential i overpressure transients. Requiring the presence of a steam , bubb_le is also consistent with analytical assumptions, i t The LCO requires two groups of OPERABLE pressurizer heaters, l each with-a capacity m 150 kW and capable of being powered , from an emergency power supply. The exact design value of i 150 kW is ' derived from the use of three heaters rated at 50 kW cach. The amount needed to maintain )ressure is ! dependant on the ambient heat losses. 11e minimum heater i capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses i through the pressurizer insulation. By maintaining the !
-pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops.
e .- i APPLICAblLITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest' effe:t on aressurizer level and RCS pressure control. Thus, Applica)ility has been i designated for MODES I and 2. The Applicability is also . provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid t pressure rises caused by normal operational perturbation, ; such as reactor coolant pump startup. The LC0 does not apply to MODE 5 (Loops _ Filled) because LCO 3.4.12. " Low-Temperature Overpressure Protection (LTOP) System," applies. The LCD does not apply to MODES 5 and 6 with partial loop operation. In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters-capable of being powered
'from an emergency power supply, in the event of a loss of offsite power, the initial conditions of these MODES gives ,
(continued) SAN ONOFRE UNIT-3 8 3.4-48 Amendment No. 116 .
b~ A > - A - ,< +A- + - - , 4 a--E-,-,-A,G k-&3, 444 -mAwadA. -4moe -mard4* - -+-- 4 b O
+
1 l i 5 ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATIONS AND BASES UNIT 2 4 f 4
-. ,, -.w..,. ,,i.r- , y + .- -- -- - ., _+.- .
Pressurizer 3.4.9 3.4 REACTORCOOLANTSYSTEM(RCS) 3.4.9 JPressurizer LC0 3.4.9 - The' pressurizer shall be OPERABil with:
- a. Pressurizer water vel =cleve] 5 900 ft267%; and
- b. Two groups of pressurizer h"ters OPERABLE with the capacity of each group a '# "W and capable of being g powered from an emergene" .. wen supply.-
- APPLICABilliY:- MODES 1,._2, and 3..
-ACTIONS
-CONDITION REQUIRED ACTION _ COMPLETION TIME A. Pressurizer ' water A.1 Be in MODE 3 with 6 hours vehmelevel not within reactor trip breakers limit." open. l AND A.2 Be in MODE 4. 12' hours B. -One reqtiired group of B.1 Restore required .72 hours pressurizer heaters group of pressurizer inoperable, heaters to OPERABLE status.
C. Required Action and C.1 Be in MODE 3. 6 hours associated Completion Time of Condition B ANQ
.not met.-
C.2 Be-in H0DE 4. 12 hours
., a- :
SAN ON0FRE -UNIT 2' : 3;4 26 Amendment No. 127 __._,,__._______m--__ -_______.__.___. . _ _ _ -___-__.A..----.-
._ _ _ ~. . -. _ . _ . . _ . _ _-
Pressurizer 3.4.9 SVRVEILLANCE REQUIREMENTS SVRVEILLANCE FREQUENCY [ SR 3.4.9.1 Ve 12 hours hpify pressurizer water vehelevel 5 900 57%. SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters e 150 kW. SAN ON0FRE- UNIT 2 3.4-27 Amendment No. 127
Pressurizer
,' B 3.4.9 BASES BACKGROUND (Pressurizer safety valves) can control pressure by
'(continued) steam relief rather than water relief. if the volume limits ;
were exceeded prior to a transient that creates a large i pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of f 2750 psiga. The requirement to have two groups of pressurizer heaters ! ensures that RCS pressure can be maintained. The 1 l pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled. Inability to control RCS
. pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat. ,
in MODES 1, 2, and 3, the LC0 requirement for a steam bubble APPLICABLE SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety . analyses are performed in lower MODES. All analyses , performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the > pressurize,. In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present. Safety analyses presented in the UFSAR do not take credit for pressurizer heater operation; however, an implicit : l initial condition assumption of the safety analyses is that the RCS is operating at normal. pressure. Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 1), is the reason for their inclusion. The requirement for emergency power supplies-is based on
- NUREG 0737 (Ref. 1). The intent is to keep the reactor -
coolant in a subcooied condition with natural circulation at , hot, high pressure conditions for an undefined, but extended, time period after a loss of offsite power. While loss of offsite power is a coincident occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period is not evaluated in .
-the-accident-analyses.
t b (continued)
- 1
~ SAN ON0FRE ~-UNIT'2 B 3.4-47 Amendment No. 127 s
Pressurizer
," B 3.4.9 BASES APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement.
(continued) LC0 The LCO requirement for th water vehelevel s 9004ty prassurizer 57% ensurestothat beaOPERABLE steam bubblewith exists. Limiting the maximum operating water volume preserves the steam space for )ressure control. The LC0 has been established to minimize tie consequences of potential overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions. The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW and capable of being powered from an emergency power supply. The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain )ressure is dependant on the ambient heat losses. T1e minimum heater capacity required is sufficient to maintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops. APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on pressurizer level and RCS. pressure control Thus, Applicability has been designated for MODEb I and 2. The Applicability is also provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LCO does not apply to MODE 5 (Loops filled) because LC0 3.4.12, " Low Temperature Overpressure Protection (LTOP) System," applies. The LCO does not apply to MODES 5 and 6 with partial loop operation, in MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer heaters capable of being powered from an emergency power supply. In the event of a loss of offsite power, the initial conditions of these MODES gives (continued) SAN ON0fRE--UNIT 2 B 3.4-48 Amendment No. 127 L-
e o h h i ATTACliMENT 4 PROPOSED TECilNICAL SPECIFICATIONS AND BASES UNIT 3 l
. i Pressurizer 3.4.9 l
3.4 REACTORCOOLANTSYSTEM(RCS) 3.4.9 Pressurizer- l LCO 3.4.9 The pressurizer shall be OPERABLE with:
- a. Pressurizer water vehelevel s W4257%; and'- :
, i
- b. Two groups of pressurizer heaters OPERABLE with the capacity of each group a 150 kW and capable of- being powered from an-emergency power supply. ,
APPLICABilliY:. ' MODES 1, 2, and 3. iACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A.- Pressurizer water . A.1 Be in MODE 3 with 6 hours vehelevel not within reactor trip breakers - limit." open. ! MQ : A.2 Be in MODE 4. 12 hours i i B. One required group of B.1 Restore requirec 72 hours pressurizer heaters group of pressuruer ; inoperable. heaters to OPERABLE status. ; C.: Required Action and. C.1 Be in MODE 3. 6 hours , associated Completion i Time of Condition-B MQ not' met.- C.2 Be in MODE 4. 12 hours i 1
-- _' SAN ON0fREv UNIT 3 3.4-26 Amendment No.-116 i 1
yy 9-, M ---+5 y,,yapyy,- %.py..yw-7..y,9-g- yw-n9 y,v g.+ y- , g
.. - .- . _ - . - .- - - _ . ~ _
Pressurizer > 3.4.9 SURVE!LLANCE REQUIR[MENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Vepifypressurizerwatervolumelevel5900 12 hours ; (4-57%. SR 3.4.9.2 Verify capacity of each required group of 92 days pressurizer heaters e 150 kW. SAN ON0fRE- UNIT 3 3.4-27 Amendment No. 116
.- . - - - _ . . .- - .- . - _ = _ _
l Pressurizer B 3.4.9 BASES BACKGROUND (Pressurizer safety valves) can control pressure by (continued) steam relief rather than water relief. If the volume limits were exceeded prior to a transient that creates a large pressurizer insurge volume leading to water relief, the maximum RCS pressure might exceed the Safety Limit of 2750 psiga. l The requirement to have two groups of pressurizer heaters ensures that RCS pressure can be maintained. The pressurizer heaters maintain RCS pressure to keep the reactor coolant subcooled. Inability to control RCS pressure during natural circulation flow could result in loss of single phase flow and decreased capability to remove core decay heat. APPLICABLE in MODES 1, 2, and 3, the LC0 requirement for a steam bubble SAFETY ANALYSES is reflected implicitly in the accident analyses. No safety analyses are performed in lower MODES. All analyses performed from a critical reactor condition assume the existence of a steam bubble and saturated conditions in the pressurizer. In making this assumption, the analyses neglect the small fraction of noncondensable gases normally present. Safety analyses ) resented in the UFSAR do not take credit for pressurizer 1 eater operation; however, an implicit initial condition assumption of the safety analyses is that the RCS is operating at normal pressure. Although the heaters are not specifically used in accident analysis, the need to maintain subcooling in the long term during loss of offsite power, as indicated in NUREG-0737 (Ref. 1), is the reason for their inclusion. The requirement for emergency power supplies is based on NUREG 0737 (Ref. 1). The intent is to keep the reactor coolant in a subcooled condition with natural circulation at hot, high pressure conditions for an undefined, but extended, time period after a los.i of offsite power. While loss of offsite power is a coinn dent occurrence assumed in the accident analyses, maintaining hot, high pressure conditions over an extended time period _is not evaluated in the accident analyses. (continued) SAN ON0fRE- UNIT 3 8 3.4 47 Amendment No. 116 ew y -
---"r- -
Pressurizer
- \
8 3.4.9 BASES APPLICABLE The pressurizer satisfies Criterion 3 of the NRC Policy SAFETY ANALYSES Statement. ; (continued) LCO The LCO requirement for th water volumelevel 5 900-f4g pressurizer SM ensures that ato be OPERABLE steam bubble with . exists. Limiting the maxi' r operating water volume preserves the steam space i . pressure control. The LCO has been established to minimize the consequences of potential , overpressure transients. Requiring the presence of a steam bubble is also consistent with analytical assumptions. The LCO requires two groups of OPERABLE pressurizer heaters, each with a capacity a 150 kW and capable of being powered from an emergency power supply. The exact design value of 150 kW is derived from the use of three heaters rated at 50 kW each. The amount needed to maintain pressure is dependant on tSe ambient heat losses. The minimum heater capacity required is sufficient to (Lintain the RCS near normal operating pressure when accounting for heat losses through the pressurizer insulation. By maintaining the pressure near the operating conditions, a wide subcooling margin to saturation can be obtained in the loops. APPLICABILITY The need for pressure control is most pertinent when core heat can cause the greatest effect on RCS temperature resulting in the greatest effect on )ressurizer level and RCS pressure control. Thus, Applica)ility has been designated for MODES I and 2. The Applicability is also provided for MODE 3. The purpose is to prevent solid water RCS operation during heatup and cooldown to avoid rapid pressure rises caused by normal operational perturbation, such as reactor coolant pump startup. The LCO does not apply to MODE 5 (Loops filled) because LC0 3.4.12. " Low Temperatere Overpressure Protection (LTOP) System," applies. The LC0 does not apply to MODES 5 and 6 with partial loop operation. In MODES 1, 2, and 3, there is the need to maintain the availability of pressurizer henters capable of being powered from an emergency power supply, in the event of a loss of offsite power, the initial conditions of these MODES gives (continued) SAN ON0fRE--UNIT 3 8 3.4-48 Amendment No. 116
A11ACHMENT 5 PROPOSED UfSAR SECTIONS FOR fEEDWATER SYSTEM PIPE BREAK, CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS) MALFUNCTION AND INADVERTENT OPERATION Of DIERGENCY CORE COOLING SYSTEM (ECCS) DURING POWER OPERATION t 9
i t San Onofro 2&3 FSAR ! Updated DECREASE- IN HEAT RD40 VAL BY THE I SECONDARY SYSTEM (TURBINE PLANT) 15.2.3 LIMITING EAULTS in order te determine the weret cese RCS peek pressere for feedweter system pipe breeks en essessment of reduced-euxitstry feedweter flewo wee conducted. The feedweter eystem pipe bresh event we5 enelyzed te demonstrete thet e sufficient 5ecendery heat 5 ink exists during the trensient with 500 gpm APN ficw, end ; that primery pressure centrel-is meinteined. The resulte ef'the i anelysis demonstreted thet the RCO peek pressure is bounded by
- the Cycle 1 enelysis. Therefere, the representetive enely5is !
presented-in this section corre5 pend 5 te-Cycle 1 which was conducted using thu euxiliery flow essumptions indientedr ; 15.2.3.1 Feedwater System Pipe Breaks i 15.2.3.1.1 Identification of Causes and Frequency . Classification The estimated frequency of a feedwater system pipe break classifies it as a limiting fault incident as defined in Reference 1.of section 15.0. A feedwater system pipe break may . occur due to a pipe failure in the main feedwater / stem. 15.2.3.1.2 Sequence of Events and Systems Operation A feedwater system pipe break may produce a total loss of normal feedwater and a blowdown of one steam generator. If normal plant electrical power is lost, this superimposes a loss of primary ' coolant flow, turbine load, pressurizer pressure and level control, and steam bypass control. The culmination of these events is a rapid decrease in the heat transfer capability of both steam generators and eventual elimination of one steam . generator's heat transfer capability. The result is an RCS heatup and pressurization. The NSSS is protected during this transient by the pressurizer safety valves and the following reactor trips (1) steam generator low water level, (2) steam generator low pressure, (3) high pressurizer pressure, (4) low 4 DNBR and (5) high conta. ament pressure. Depending on the i particular initial conditions, any one of these trips may terminate this transientijlThe NSSS is also protected by main steam isolation valves, the feedline check valves, the steam generator safety valves, and the auxiliary feedwater system which serve to maintain the integrity of the secondary heat sink
- following reactor trip. In this analysis, however, the most adverse = single active fadiure assumed 11s equivalent to the failure of the elec;ric driven auxiliary feedwater pump associated with the intact steam generator. The operator can
- initiate a controlled plant cocl-down using the atmospheric steam
-dump valves.any time after reactor trip occurs. The analysis
- presented herein conservatively assumes that operator action is delayed until 30 minutes after the first initiating event. Table- ,
- w ce- -e- .. , . ,., --n . , . - - -- ~v. r 5 e-, , , , , r.,,- - - , . . . , - , , , . , - - ---w..- .-- .,
l Scn Onofre 2&3 FSAR
* . Updated ;
DECREASE IN HEAT REMOVAL BY THE : SECONDARY SYSTEM (TURBINE PLANT) ! 15.2-8 gives the sequence of events that occurs following a
' feedwater system pipe break to the final stabilized condition.
The sequence Of evente pest 1000 eecendo ehoe.. in teble 10.2-0 15 Isei cycle 1. This is sepie5entstive of the-l&te5t cycle '5ee ' sectien 10.0.7)r - 15.2.3.1.3' Core and System Performance _ ( 15.2.3.1.3.1 Mathomatical Model. The HSSS response to a ! feedwater system-pipe break was simulated using the CESEC-III ' computer program described in section 15.0 along with the s ' blowdown model described below. _A detailed description of the i method of analysis, the initial conditions, and the input l
- parameters is presented in Appendix 15E. Using the core heat >
flux and core inlet conditions calculated by CESEC-III, the thermal margin on DNBR in the reactor. core was simulated using - the CETOP-D, the thermal margin on:DNBR in the reactor core was simulated using the CETOP-D computer program described in section , 15.0 with the CE-1 CHF correlation described in Chaptec 4. l Blowdown of the steam generator nearest the feedwater line break was modeled assuming frictionless critical flew calculated by the , Henry-Fauske.correlationm.- The enthalpy of the blowdown is assumed to lx3 that of saturated liquid until no liquid remains,
- at_which time saturated steam di scharge is assumed. This model conservatively underestimates the blowdown energy and- ;
- overestimates the died arge. rate, thereby-leading to a more rapid
' blowdown and thus minimizing the steam generator heat removal
- capability.
A-sensitivity study was performed to determine the influence on peak RCS pressure of the rate of decrease of effective heat 't
- transfer area in the ruptured steam generator. The effective heat transfer area is assumed to decrease linearly (from design value to zero) as the steam generator mass decreases (from selected value to zero). Thus, decreasing the mass interval over
- which the rampdown is assumed to occur implies a more rapid loss of heat' transfer.in-the ruptured steam generator. This study showed that maximizing the rate of decrease of heat transfer area ;
maximizes the peak-RCS-pressure. _Therefore, a conservatively , high rate of loss of heat transfer is assumed. 4
+
r,,,.. - ~ ~ -, , ,.m.m-- u .mn --. ~ - A ,-, g -v- -,-e..nnw--,~, , .n,,,rr.--.-en., ,, -r,'n,w,v--.-,.-,,m,, ,Ane,n, ,,,,-,,,.,,w,,,..r.,.r,r an
San Onofro 2&3 FSAR
+
Updated DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) 15.2.3.1.3.2 Input Parameters and Initial Conditions. The input parameters and initial conditions used to analyze the NSSS [' response to a feedwater pipe break are discussed in Section 15.0. In particular, those parameters which were unique to the analysis discussed below are listed in Table 15.2-9. ' The initial conditions for the principal process variables monitored by the COLSS were varied within the reactor operating space given in Table 15.0-4 to determine the set of conditions i that would produce the most adverse consequences following a j feedwater' system pipe break. The full spectrum of break areas ' was considered up to a break size of the combined area of the flow distributing nozzles in the feedwater ring. For each break, the initial steam generator liquid inventory and the initial pressurizer pressure were adjusted within the plant operating , space to maximize the mismatch between core power and steam generator heat removal capacity prior to the CEAs dropping into the core. This mismatch will thus, maximize the peak RCS ' pressure and pressurizer volume. In order to eliminate any impact of uncertainty in the calculated water level in the ruptured steam generator, no credit was taken for low water level trip in the ruptured steam generator. This delays the reactor ' trip, prolonging the RCS heatup and increasing the peak RCS pressure. Loss of AC is assumed to occur at the time of turbine trip. This causes the RCS pumps to coastdown, resulting in higher peak RCS pressure. In addition, in response to loss of non-emergency ac power upon trip, turbine stop valves are assumed to close immediated - Core inlet temperature and flow had negligible effects on the peak RCS pressure for a given blowdown rate. However, maximizing the core inlet temperature also maximizes the steam generator pressure, which increases the maximum blowdown rate. The maximum inlet temperature of 560 F also maximizes the RCS energy content and thereby increases the radio 2ogical releases associated with steam generator safety valve and atmospheric steam dump valve flows. The pressurizer control system is maintained in the : automatic mode so that it suppresses the pressure transient before trip. This delays the time of reactor trip, prolonging the RCS heatup and increasing the over-pressurization. However, this-control system mude had a small impact on peak RCS pressure. Of those systems and components called upon to mitigate the consequences of a feedwater system pipe break; i.e., pressurizer and steam generator safety valves, feed line check valves, auxiliary feedwater system, and reactor protective system, failure of the pressurizer or steam generator safety valves, or the feed line check valves, is not considered credible. With respect to the reactor protective system, the most reactive CEA is conservatively assumed'to be stuck in the fully withdrawn v o ,7..--..:, er. - --:- y , r- .m ,. ,. -.m- ---.--.-r,-,,-.--,--,-- - - - - - - . . - , - ,
Scn Onofro 2&3-FSAR ! Upd0ted DECREASE IN HEAT REMOVAL BY THE . SECONDARY SYSTEM (TURBINE PIANT) [ t position. Therefore,-the worst acti*ie single failure, in addition to the stuck CEA, is the failure of one out of the two auxiliary feedwater pumps. This failure leads to larger radiological releases through the steam generator safety valves with onl" one-half the auxiliary feedwater flow available. 15.2.3.1.3.3 Results. The dynamic behavior of impartant parameters following a feedwater system pipe break for a break size of 0.2 ft.', which gives the maximum of the peak pressure, i is presented in figures 15.2-37 through 15.2-560. The rigures 15.2-00, 15.2-41, ;5.2-40 end 15.2-45Lthrough 15.2-55 ere frew Cycle 1. They 6re -representetive of the lete5t cycle (5ee sectier, 10.0.0). ' 1
- -- .- . - . _ - - . . . - . _ -- .= - _ - . ,
San Onofro 2&3 FSAR Updated , DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) i The rupture of the main feedwater line is assumed to ! instantaneously terminate feedwater flow to the steam generators. ; Critical flow is assumed to be instantaneously established from i the steam generator connected to the ruptured feedline due to the . break location between the steam generator and thn check valve. Check va,1ve closure prevents flow from the intact steam generator to the break. The first 95r0 38?0 seconds are characterized'by a l l gradual heatup of the primary and secondary systems due to the absence of subcooled feedwater flow to the steam generators. - During this stage, the steam generator connected to the ruptured line loses its heat transfer capability due to the depleted , inventory. This initiates an RCS-to-steam generator power mismatch, producing a large insurge to the pressurizer, causing its pressure to exceed the high pressure trip setpoint at 6570
$870 seconds. A decrease in the steam generator water level initiates a reactor trip oa low steam generator water level simultaneously. Reactor trip followed by turbine trip occurs at 95Ts 38.9 seconds. Pressurizer pressure continues to increase, passing the pressurizer safety valve setpoint of_esas 2550 psia at 9576 39.0 seconds. Loss of normal onsite and offsite j electrical power is assumed to occur simultaneously with the turbine trip, causing the reactor coolant pumps to coast down.
The pressure turns around after reaching a maximum of e9+s 2893'1 psia in the RCS at 40re A2iG seconds. The core heat flux has decayed sufficiently by this time to reduce the RCS-to-steam generator power imbalance. By 407039(0 l seconds, the steam generator safety valves open, limiting the steam generator pressure to a maximum of tt40 1167;2 psia. By 49Ttf2i6 seconds, the power imbalance reverses, with the steam generator removing more energy than the core produces. The pressurizer safety valves close at +&r4 52;5 seconds as the l prima'ry coolant temperature decreases. The auxiliary feedwater flow reaches the intact steam generator by 0079 93;0 seconds. l Reverse steam flow from, the intact to the ruptured steam l generator and to the break causes the secondary pressure to decrease below the main steam isolation signal (MSIS) setpoint of 675 psia at 212.7C22510 seconds closing the main steam isolation l valves (MSIVs). Closure of the MSIVs causes the secondary pressure and temperature in the intact steam generator to rise, decreasing the differential temperature (RCS-to-steam generator) and reducing the heat transfer rate. This causes the core average temperature and RCS pressure to rise by 300 seconds and to reach a steady state by 500 seconds. By see A3550' seconds, the steam generator safety valves open again .nd continue to relieve-steam to the atmosphere until the atmospheric dump valves , are opened by the operator at 30 minutes. The plant is then cooled to 350 F at which time shutdown cooling is initiated.
- The results indicate that the feedwater system pipe break event will not result in a peak RCS pressure which exceeds the faulted
Scn Onofro 2&3 FSAR Updnted DECREASE IN HEAT REMOVAL BY THE SECONDARY SYSTEM (TURBINE PLANT) , stress pressure limit of 3,000 psia. 15.2.3.1.4 Barrier Performance 15.2.3.1.4.1 Mathematical Model. The mothematical model used for evaluation of barrier performance is identical to that described in paragraph 15.2.3.1.3. 15.2.3.1.4.2 Input Parameters and Initial Conditions. The input parameters and initial conditions used for evaluation of barrier performance are identical to those described in Paragraph 15.2.3.1.3.- 15.2.3.1.4.3 Results, rigures 15.2-48 and 15.2-49 are steam generator and pressurf zer safety valve flowrates versus time for _the feedwater system pipe F ak transient. By 30 minutes, when tlus atmospheric dump valves are opened, the steam generator safety valves will have discharged no more than 4+r0 % 74;800 pounds of-steam. Approximately 934,000 pounds of steam would be discharged through the atmospheric dump valves during the 3.2 hours of cooldown, giving total steam release to the atmosphere of 1,000,0007,1T,008;800 pounds. The steam generator connected to l the ruptured feedwater line discharges 151,033 pounds of fluid to containment. The pressurizer safety valves release 2325 pounds l . of steam to the quench tank. 15.2.3.1.5 Radiological Consequences The radiological consequences of this event are less severe than the consequences of-the main steam line break discussed in Paragraph 15.1.3.1.B. l
,..y .- --
tevW y -vW -
-T---
re S3U OU0110 n E ASVM ndpcaep G33W3VS3 IN H3VI M3HCAVI' EA 1H3 S3DONGV8A SAS13H )108EIN3 dI'VNI(
.- ,. a
. o ,, e n r.,__,,
s, 3s w s 6 vu utaw 4 m 4$41464 w'w , w ,, u w sw r
**i Q N1f ,4L *) ,i a ,v ,4h * %/ T T'P A ,'* ,i19a wOw ,ul*/v uu w T,6,aTP w
- T, TP T
'T'/1lU TT*,1 9f ? D ? T*,.f
,.m eu ,-uu ,..,.m.w m .. ,m
,m,.,._ ,, 1 u, s c,,n n , , ne
,m ,
r f , ,'! n % e ap. TP A n , , i" 5f n i' N n , , , i' nT A5, 1 .,r- .r-- O.6 dsna ,,n ,n-ne u o v ,, ,.-r--r,-,- onnuo c nr,rn ___
.& U , , n %" )* n n D T' m 0,'1 7 u,_ __ ,w_-_.,__..
n n ,,n T o ,,/T T*,!!N ,Y n e n n u'- n,nem nn,,n w7 - - - - e-- v ---r o ,n e- - - *uc,nv * , -nnAnv -- a - - - , , ve
,s .h, . ne ,,.
,, , - - __. c,, ,
, t r a , n ,,n v , , n,, nesn,e ,.,, ._
,,,,o,,.
y y. ___ _ _ _ _ . . _ , _ _ . D i' n 3.,,11 n n ,,n T o )" n ," m __._,___ __ _ _ . _ _ ,, ,_ , ,_
,., n e e n , n Hr, A ,, ,,nec.rev,n, ,,va , . , , , ac,o
,,__r,,_, __ ___
T* n , , n ,1* T n , , n \* , . ,.n _e 49*.9 nagg on,n,K. , , _ ,ADTAne ___-. 1inecn,reni , r >- ~r ac,n R g ,,7-T T v a ,r ,. n,. _ , .n_, -e - - --
^ d /* 9ra n,-,#.
--, , _ _ ___7 , , , , _ , , , _ .
H n ,- m o v, n,,er,n o,,n n , , e , ,. .n , nuu qcM
&' m _m. -
y- ,
,- , . . . . _ _ 7_,-- , . , , _ . _ .
, T'FU e ,"( n b , , , 1* ,t T/T nN 7,, 9 n ,nTn ys .-6 _ ~ . ._,_
e ,,n o_ m n n , e n r o , n ,_ nevn,.r AovAne r,nn aevo edau yg .-5 , n n i. u,e . , . necn.
, _ _ . , _ n. n,en,e
, _ . _ ,w , u
, e .f e aceg yc.v, s n n i. n, nom _gaungan, ,__
, ,necn,a - _ _ _
, , .f n aeve y __ ., -
n,,n,e ps.p ,,neen,rene co,n,r Ae,Ane ,,nem
,,mm
,tna aevo h
w M .
! f o V ,, tim 111 T TneenT,,nT 4--- -
,F ,r bn,n
,c a.,,-,---
A 9/ M 11n b L t .," .L z J\
^
J,4 1 M se.p ,g,, n U,11 _ - , _ _ _ _ , _ _ _ . . , , . , , _ , _ _ _ - , n ,. , m-n i'" , , n , D )* ,f T O U T n i* [ Ad,Ane * ^ * * ^ NM?e
,enen W. , - . _ . _ . .
,mn e n n..,r
__7.,--..
,snnuu,ne
,,,sn,_e , ,.nn n,num nn,,n,u m n,.-_.nem n n , ,3 , o ,. n _ ,.
,nu ,,neen,_n
_ _. w___
,,va
,,r acen
,_c7,, -_c ,. . . 7 ,,c
, n , ,n e 9 , n,, D ,,n ,.,0,*( ,. , , ? 1' , D 9 n n
,.r.n nemen,n ne no,,, n,nem ___
,n,__ ,4.a, v, n , ,n , , _ , _ _ _ . ,, ____ _ _ _ ,
AO,Ane
,e ,t ,_,i".Tir,, , _ _ _ - f n T ,'f T , , D ,
m, __.,_ .
,c , n a n n u ,, ,enm ,,,n
,,,,o ,, n,nem
, , _ _ _ _ , nv nn,,n,c,. -
,mcf., m I ., ,. ,,emnm
. . --. , r ,,- - - ,- ,, - , , , - --
rrnn,n ueee ,,, ,,en c,nem _m
,r,,
, ,,m
)gg n.. . TU9n2 ,. _ . , n ___, ,- _.,, , n , , n_7 n 9n , _ _ ,_,- -
,,,,U,,\*
cs , s n m_ n v , t t n a _7 ._.
.f_ 7__.
,_- _ r ... - . _ _ . . - . ...
n . a . s -- - r- ~ - - -
. . . . _ . g
,_, r a , _ . -c -, n , . . - - - , . .- ,--,n - -- . - ,_-.~,.. - .- - . - - , , - ,--n_---.---
O. s o u o u o ; 2 e M E i.S W ndpegep 033V3VS3-IN H3VI H3H0AVT 8A 1H3 9300NOVHA SAS13H )4OWeIN3 dI'VNI( _=._ m. - ;. . _.g._.
. . _r_ 4 . _ . . -
*J V s,n.1 ,omem' . ,V _ m
, . _ - --.m ,.n,_
,.,m..,..,,
# .u n ,- , . ieu . . .o . . , . ,. _ mu. , , . , , . , e e ,,.,.mu
., ,m mu n , r , ,. . .u ..
,.- .,,..n . . . , , - , ,, . .,
, , , , * ,f, $Wu ,...,.,L,
*A
- I D , , m m %* , %r T 7* f li_.t. % e m %8f ,* A s , ,1*
,N .e _ , _ _ , _ , ._ . .Ao, 4, m i' N %/ , ,, i' %/ $
-{.A ,4,u. ,. y__ _ _ _ , _ .
na m , o , m ,. ,am,,e ___ , , ,, , m- , , _ - . _ _ _ , u ,mm e a, _,,u}3 ,, __,
,,._ ,, s n , , , , , o , , ,-
c,nem im,Ame ,m
,-,c.-.. nma, , , , , , _ . , , _ _
%8 %/ Mf , i im iA , ,. t* m N , ,n,nmu
. . , , _ , , .. % %t %t , , , , b
,.,mmm ram.m m ,. , , _ .
o ,, n , n m u,, s m m , , ,,b .,,,,,,,,mn, ,vo ,,__
- a. _ _ _ . . _ - _ , _ , ,, _ _ _ _ _ , ,_ , _ .
,. , r n , , r o v o , n e n e, ,, ,_ ,- o,,ovro,e ocenmen o m.m e m,m, n i ,or
,,,m_ , ,,
qm_ ,1* M 1"* m , , ,,.m ,-___, m o , i' ,__ ,.,7.- b m,. ,m T o44%, o _ _c
,,n
,' ,- r-i* ,it s m
4* T II ,P r 1* , , m r-,,_,..
%8 , 1.te
, s n ,r , , ,,,m n e- ,r , ,rr n , , o , m ,,,o,
, e m e m , , , - mm o e n , m m m , , r,- mm ,,
,n n,ma.
4 r_a_ ,, s_ o_ ,, e, r. o. m.m e m,m,,n nm,or ,e nvam,,m,,,mn. ,.,m,,, ,, , _ , , _ cnuoc
,.,, ,, ,e,_ , o ,,%,
%, ,- %/ , tim o e 1, ,-__-
tte _,,4 _ ,_ ,, o A m if
,_ ,, _ _ _ _ , Ir
*/ % ,eu ia , , tim a
___"*,_,,,=e-_,, _Am,r _ _ _ . . , , , 1* , 3 wo , A , ,, ,m ,,,m ,.,,m,,, n ,m ,,,,, ,n o , eam,,,,%o,,m,, ,, y,
,,m m_e , .,
gm,,n- e r
. ue nr ,reenoomn -----r ,r , , n m , >,* , m , , ,. s,.m.,,
- - ,- v ,, , , - - -
,,,,mm
,s, m
s ,mu. m,-.-_ma
,,mm , n ms . . _ _ _mm,m noen ,m,
,,,m eo,m,. ___,
r, _ o ,, o , r e , .c- ,,__
.,,,,-,f o ,,,
,,,m g ag g apo_r.. ,,
, ems m, ,.mv e m,m,,n e
,m, o n,ma ., __
,,mm m,
,. e m ,m ,,n e . ,m,m nam ,.o,,,,n
, , _ _ ,em,, _ . - _
e n a m ,-- i r e-_mr ,-, ,e _ _._ r_ e ,s m m._ -
.g;mrnof _ , . . - _ ,
o , .n _ _m_ _ If , m4a%- , mt ,,_- sam, , ,,_..,,_,__ o ,i o ,$r v ero
- _.. ,mm i m
,m
%* e f o,-n
,m s - rr7rc ,- - - - r ,- -- , - - - - - - - -r- , _ , . . _ _ _ ,
,/ m ,11%/ n , T , hf m 1* if o , , T/,li,11%/ n o 1* m o,,[ , %/ c e %/ T ,r,-._, 4* ,, m T fit Q_T.
_ _ _ _ , _ _ ,. __, _ ,,- _b o, ,, _, -, - ,
,t t o , b , , , e ,t o e m%, . %t,, 1* , , m ,ltm
- _o_e _n .,._. mn _ , _oAmTobm ___.__- , ,, tim
- _ , , , Yr__,,_id
%8,F 1 %, T %/ , , ,
. S. ,m
,._n ,___ _' m_ o .v. ,,, , , , , , - __ _ _ _ , _ _ ,
m,m ,_,,,, ,, ,_ e_ v, o ,,,
,,,e ,miremm n a m , , , , , o , , m ,-- , e..,.<.
,,,m ,m,,,,ne,m,,em, ,,,re m A m,, , o ,- m n ,, ,,,o ,,b s n .
~ San Onofre 2&3 FSAR LUpdated _ _
3 DECREASE IN HEAT REMOVAL BY THE-SECONDARY SYSTEM (TURBINE PLANT) 015?273 8 g-wgnmppe?W"7 ksimmisAssofi rvanti%r N'$ TibT(dedwater i theTr ?spWEsis epipr aisak'3EVehf T11663 Avsnt' - SetpoidE sec ' orsyalue 1 RM BOPthiRio fs thelMalalTe sdQstier[liilns E"jdp; i 3910. - HisMPrWssdriseMPressdrsjjTEiplCondiLti6n7l Q2450
- ps'ia !
~
3 BIG: AffictlsdES t;siEGsdsf aLtoflEnip Ele s;21bni [315000 38X9 ,TElpIBfe'sKeysEOpsh7j DZy-3 ' No'rsiEOn?~sitslPoidsrTlL6st.; [7 5 3910 pre s s u si fe rfS a f e.t y!Valies!S t_ir titolopsh7 F12550- , psla Mi3 - EssNssdf[ F.e eldOAVs r2Xdt6a'E16n[8 lyha1D ti B2Q22 i 3919 CEAs]Bsgin]f63Dr6p M E55 L4.2E6 Pss_kiRCS[PEsssdF#JoeEhrsNps;fE [28.93)) 48MO PsaKiS t eanGGenb rit6E Pk3dshfeioccuRb gps is K1,16]?2 5R5 PrissdE1:eRSifetf(VilVe'i8#6sel# psia M2200 52J51 Mixim6RPr~esbEfzersL1Mid;V61bm6?occurug [31435D
'f t8
#379 AOkifihy IFesdWstsfEEntei'si16Es6t}Stlsss 7hi-i Ge.neratoz;
;22510 - St"sisEGensidEdijL6WTPfessd M TElpIcohditl6h M675
,pndiMS!SSInitiat.edupsla 226VQ hisiH@Ess@Is61stl6dINslysslB.si1h - [icl6;ss-23 Ed' _
Cf6hiplsti!!c16sGEsI6EMaihisEsssEIi61st16ft NTE63 Vali.a essTeiniin. ati.ng"Blowdowntfr. 6mi~th.e3Intist
. c ~ - ~ --
1
. Steam;LGenerator ~
^
conne_c tedi tognt a ctge:edline190 bm _
~ * .-,,,,,y,.-+r- -,,v+.- y , v- v :.,, y , _ ---.
. - . . - - - . . ~ - . . , - . . - . . . - . - . - . .- . . . . . . -
a San Onofre 2&3;FSAR
-..- Updated ,
DECREASE IR HEAT REMOVAL BY- THE -i SECONDARY SYSTEM (TURBIME PLANT)- , 3357Q pliiM s t'sisiissf sty [VslyesI0psH!661ths%Ihtd5tf [E1133 Steam 1GeneratoMpsia ! l#0My bpefatB s;[gid3 yalves t,M?opehsithsi?Atm6sphsr16YS toj begiln[Planticoo.idownitoishbtdowd fsasyDump Coo. ling idI228 . shWta5ish!c6Bir6pIIsrtritiid DF Ei ! i s e f k 4 i-1 a + E 4 R 1 i 4. t b f .,
, . .. :e . . . - ...;.,-.....--.- - -. - - - . . . - . , , , . - -
' San Onofre-2&3 FSAR Q: Updated ._ .
DECREASE IN HEAT REMOVAL BY THE
-SECONDAR'.* SYSTEM (TURBINE PLANT) m _ t_ , _ - - , n .4 4QUdU 4 94 9 dm #
saa t es s e . re= , a s e n ,Saw me e r1 1,ne,ht_ss m m h a t e n te _m s k ns nn n h n s es ; n=JWVGE 4 4 Val d K V4T 4 Aau & LL Vnn 4 L4T V 4 J i_ L4'4 & 4&L WIuen.ET I n_______t e,_ 1___ E Q E G AUG bU M__ V Q ek LJ U 1 L m___ v _ _ _ s a,_s t udu3 Q1 uGvUi , .Yv U .h vnm -_ _ _ _ _ , nnb
, _ 1 A_ 1 _ 9 ,i _ L n_ __ 1 _ __ t e ri _ _ _ _ _ _ __ t_ _ _ _ _ v1 d444hdQd 4 d4dU h bVN d Q44 b d U4La U h Q h bd h U f E
-gm a a- 4 m i m .. _
e as e w 4 u e vusw e eu w w u-__ r,3 _ . n _ a. _ s ne
- mm m
-t 6 A V vt 61u w w p 4v 4d6e6 3 A 2 - 1 ML pg _ _ _ _ M__ '
4_44b4QA ___ _.___L _ _ _ _ _ _ _ _ _ _ _ a 44 V b U Q4L4 kJU 4 4 G L Q b U h 44UGGMkU$ g/ G A G
, _ _ > c s. < um,. n___ _ _ _ _ _ _ _ _. __ _ -
A344 b 4 Q4 (Tb d & & U G G 64 & G $ }/ O 4 Q e= _a.__ m_----..A...-_ #9 f#J_2_ L F 1 A*$ wswwvs a wanyw a u w u s w vN w s s .6- w a w a a w gav e i _ s , i L4WI E$ _ s m-_-_ ____t..__ _ n_ _ rr* =-n s a _ s u s __ s _ _ pVe -s e U4 5 L4L&g/U M Q b bA & G V b/ G 6 4 4 bd U&4 6 4'd LA 4 6 A f 4 4 U M.__ i m.I 4 sa_ _ _ ___ ,,m. ,_, _ s m m__ . < . - - nn n4 44 A4uusu van nv A bu Qb 4 A A ts g mp - v.vv
,,n _ - . _ . n___ _ __ n___ c__ _ , n.__t _ __ , _ _ _ _ _ ___t .
d b U Q4ik _ N QGG V b/ d E b debd d d G b UIN d EEV UbQbEYU n_______; _ _ __ m_. , , _ _ _ m__ _s m___t _ _ _ ,__m _____t * , , _ , _
& & G G G L4 L 4 de U & & & G_ G_ G _ _%d_ A_ _U _ , bb/ & & b & b/ 4 V J G b C&M GM L b/aHQ b b 4'4 61 b& U n___
- s - _ . _ _ A n- *
.. L _ _ _ 1 m _. _ _ t -_ <r - __ ____L K4CGGLAL44.UL ' UVU4 b kJ & 4 b A kJ A VJ G bC4d 4&Abdg/U& Q 64 V U r__A.mi-__
s w w u rv u w w a. . t : -_ n _ _ ~ i, w ,_ - 4J 4 4 e w uswun 4 65 w u p rot
& w uvea
,2 L 3 - 1 , _ _ . t _ __t nn_ _ _ _ _ r.___ _ __ _t _ _ _ _ v_______t _ _ _ _ _ s en amn
,4 4 4 b 4 Q ab 44AbQbb V b UQ44 kJU & 4 U L G b bl & 4 4 4 V U 4 4 b bl A g g ' 4 V J $ V =J V it'Ttt M ____I $ _ - m f'1 _ _ _i _ __ L _ _ . _ ,t _ _ _ _ ' L_._ % __ _ _ _ . _ _ _ . _ _ _ _ _ A__ _
Ak&Jb A L A Q L g__ &LUbenQbU& b Q g) Q b A 6 g E hG G M4i& J.4 & y %/4 4 U
, , _ ; , _ _i m_- _ _ _ _ _ _
E G A AUL A 6 bkdLigJ $ $plid se-_ t- _ _ _ E s _ _ _ _ _ _i mi_______ _1 mn _ _ __ m___ ____t _ __
&l %=A AkkAd U M bl d- UG 444U bA J. A b4 y y G LJ V b UWLA VU4&C&Qbbl&
m _ t_ __ 4 L4&JU G i s
-. _.p.e -
,-ie .,-. _
San Ono2re 2&3 FSAR o _ Updated DECREASE IN HEAT REMOVAL BY ThE SECONDARY SYSTEM (TURBINE PLANT)_ Tabis?1Sj213.9 [ *jjR 'JASSUMPTIONS"FORVTHE"FEEDWATER" SYSTEM
- PIPE" BREAM ParisstkE Unit valiis Id'itisE46rsP6WefEMwt ;M " ;;l3478 Ihi,tiil"21n1ht(C6ola'tlTemperaEdreM'E n F ,"1560
;I ni t'iallCo re] Ma s s? Flokj Rit e",E gpm P 73567400 261tia13CS$Pressurs',ipsis E ,12100 I nit i alis.t e am2Ge rie sa t6ff Pre s s dre',;fpsis f' ,11 0 9 4 9
^
M6dsratBr;;;Tsmp;ef atses1Co6f f fci'ent^kl0-A m - :10 { 0 AP/M FdsETempefatdfs C6effi:disntiMdit'ipliss 9 JM25 Hihis;dm7CEMWorthEstlTr'ipMQhii F P lV610 Ste d Bipsss7C6ntr61ESystem fl*iIriopsrltivs
, Pre s sofi z'e GPfe s s d r si Cont rolTj S ps telii Autoisatic!Mo.d;d Preisufi zsr$syslf Contfbl! Sys teis AutomsticTMode FeedwsterfIins}BruskiXFesL]t: [ ~~);012 I rilti alS I.n;t a;d tis tle am j?; Gen e r a t6 rTInvch{o fp]j y jl8070_00 lbm Ad'x111h rpMe ddwa t e r31 oms 55 uinihg 21C Pump .KF146rd Only,igpm L 115;2 3f51l NdmbdF;16fiU(EdbsM Assdise di Pluggedf pe f [_ ]10,0.0 SteamiGenerator
.PSVjSsltVJilnt'jMsla e 52550 Thifi'al5PrsisdfiTeRVo16meifff3 : ^^ " 2913
1.1 i i . . i 1 - 0.9 -
-o g^ 0.8 -
u. O- 0.7 - 5 p 0.6 - O ,
<t cc 0.5 -
u. d w 0.4 - 3 o CL 0.3 -
-$ 0.2 -
O U 0.1 - 0 O 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE 4 NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Core Power vs. Time Figure 15.2-37
1.1 i i i i i S- - 1 7> E 0.9 - - O 8 0.8 - - F Q 0.7 - - t
- 0.6 - -
x o { 0.5 - - h 0.4 - - I ' w 0.3 - - e< m 0.2 - - w . , -< 0.1 - w ' C ' ' ' ' ' O O O O 10 20 30 40 50 60 TIME, SECONDS 1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Core Average Heat Flux vs. Time Figure 15.2 38
L *, 1 E b 2900 , , , , , A , 2800 - w
'E D 2700 - -
E w .- [ 2600 - - m-2500. - -
- 2400 - -
2300 - - o. 2200 7, -
.H O 2100 - -
I tr 2000 O 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Reactor Coolant System Pressure vs. Time
] Figure 15.2-39 l __
b 650 1 I I l l
/ '. TIN 640 -
f
. TAVG- - -
- u. -
TOUT - 0 w 630 - O g 620 - - w ...... g o- 610 - F-600
~'..-..
w- 590 - f - F-
$ 580 -
5 8 O 570 - 560 - 550 0 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Ursits 2 & 3 Feedwater System Pipe Break Coolant Temperaturt.s vs. Time Figure 1.w.2-40
~ . _ . . - . - _ _ . _ - - _ .. _.. . ._ . _ .
- o.-
t h 1400 , , , , 1350- - - 0 1 1300 - -
~uJ
~2
, y 1250 - -
d
>~ -1200 - -
x w 1150 - -
@ 1100 - -
u m-1050 - - m W 1000 - - m Q. 950 - - 900 0 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Pressurizer Water Volume vs. Time Figure 15.2 , v- - . _ + w w
L.. e 4 1200 , i i i i. RIGHT SG
.< LE.FT SG 1150' -
E UI l
- i. I 1100 -
/ _
-$ l w :
[ .1050 -
/ _
8 . l
< -1000 -
- x. .
iu z w.
\ - >!
950 O - _ 2
. h H.
c.o 900 - -
' ' i i i 850 0 10 20 30 40 50 6G TIME, SECONDS 4
SAN ONOFRE NUCLEAR GENERATING STATION r. Units 2 & 3 Feedwater System Pipe Break Steam Generator Pressure vs. Time Figure 15.2-42'
4000 , , , , ,
/ AFFECTED SG
/ INTACT SG --
3000 -
/ -
O -' w , 2000 T a ,i 1000 a ; . u.- w h 0 -- - 9 2- -1000 - - I H to
-2000 -
-3000 0 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Steam Nozzle Flow vs. Time Figure 15.2-43
-n- , - - , - - - y y , -~v , x-w--
0 , , , . i AFFECTED SG
-200 -
INTACT SG - - fo -400 - N - w
$. -600 - -
9 .
-800 - -
3 o i -1000 - -
-W
-1200 - -
3 o -1400 - - w w
- u. -1000 - -
-1800 - -
-2000 0 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Feedwater Flow vs. Time Figure 15.2-44 _
w 2800 , , , , , 2700 - - 5 E 2600 - ul C'
@ 2500 -
m w E 2400 - x w 2300 - o m 0 m 2200 - CL 2100 - 2000
'O 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Pressurizer Pressure vs. Time Figure 15.2-45
I 180000 _, , , , , AFFECTED SG 160000 .. INTACT SG - 2 ' g 140000 - - vi ' Q-'120000 2 y100000 - - O fx 80000 - - O 60000 - - 5 8 v> 40000 - .. 20000 - 0 0 10 20 . 30 40 50 60 TIME, SECONDS < SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Secondary Liquid Mass vs. Time Figure 15.2-46
~.
3: 40000' , , ,- , , 9 u. 35000 - - M o cc h 30000 - - O E 25000 - - si 9 20000 - - _a siE 3 15000 - - A m N' O ' ' ' ' ' o 10000-0-- 10 20 30 40' 50 60 TIME, SECONDS 1 SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Core Avg. Inlet Flow vs. Time Figure 15.2-47
4
..=
4 i
-w. . -700 , , , , ,
(G ' h- AFFECTED SG INTACT.,SG
~
*- 600 -
I \ - e . l \
/
500 - t W - ! 5 400 - ) > - ! i ~ ,LJ LL !
'g 300 -
l m I O 3 k m 200 - w z ;
$ 100 -
lj 2 t b& 0 ' ' ' ' I
'b '
- 0 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Steam Generator Valve Flow vs. Time Figure 15.2 48-I
300 , i i i i o L< , R) 2 co 250 - - J o-d 200 - - w b
> 150 - -
[J u. M 100 - - K W b! C
@ 50 - -
m W E O O 10 20 30 40 50 60
- TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Pressurizer Safety Va lve Flow vs. Time Figure 15 2 49
1800 , , , , , o-W 1600 - - g 1400 - - h'1200- - - d w 1000 - - z
~3 w 800. - -
0 C D 600 - - m C w 400 - -
-!S C
{ m 200 - - E 0' # - CL
-200 O 10 20 30 40 50 60 TIME, SECONDS SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 Feedwater System Pipe Break Pressurizer Surge Line Flow vs. Time Figure 15.2 50
I 3 'd San Onofre 2&3 FSAR Updated INCREASE.IN REACTOR COOLANT SYSTEM INVENTORY 15.5 INCREASE IN REACTOR ~ COOLANT SYSTEM INVENTORY 15.5.1 MODERATE FREQUENCY INCIDENTS 15.5.1.1- Chemical and Volume Control Svstem Malfunction 15.5.1.1.1 Identification of Causes and Frequency Classification The estimated frequency of a chemical and Volume control System (CVCS) malfunction cle sifies it as a moderate frequency incident as defined in Reference 1 of section 15.0, A CVCS malfunction that produces an unplanned increase in reactor coolant inventory may be caused by equipment or electrical malfunction or operator error that erroneously activates e cherging pnme bhsXorimoVe
#tandbM[chh@in D umps 26MdsdrbssssE15tdokniflbW. The CVCS malfunction is assumed to occur without increasing or diluting the primary coolant initial boron concentration. The case of a CVCS malfunction that produces a boron dilution is presented in paragraph 15.4.1.4.
15.5.1.1.2 Sequence of Events and System Operation Under normal operating conditions [stEpo;Wbs, the' Pressurizer Level Control System (PLCS) would detect the respondsit6j&sh increase in pressurizer level end increese SE2fhcrsasing the letdown flow to compensete for the increesed cherging Pump flow mEintainiths pt6grhmmsdSleVel. Ths fs {s reitWcQ pYs s siir i'z's W1' eve llme a s u r ement and L cont ko b chshns is , :{ea ch; havings aspro ce's h/signd 1%1nd i ca tor $ and aflow-leveluandshigh-levelialarm 'associhted5withMt> Lon1p!6ne shannelsat;?a stime3m selectedibs? the; opera t orMisVini contro1% o fsths PLCSE (TheOPLCSOcan2 befo' ma intiain / prbcMammeddpre spe rat edl ei'thersini manu'al f od automat sdrimsrilevel~.$UsusllpKAtOisastione ' i c 9 to chasging pdmphisfrunningnlp.atched)by21std.ownlandifeact6ricoo1Ent pumpfsealbblee.doffl Menuel control of the cherging pump 5 end of the letdown velves is eveilable to the opereter at ell times. "owever, et leest one cherwing ewup end e minissu istdown flew of 30 gel / sin ere-im Operetica et ell times. A mendel switching stetica 10 provided-for eech cherginy peng. Cech centrol 5tetion p;cvides foi
= closing ei tripping the motor control end for selecting eutemetic, senuel, cr.stendby Operetion for the cherging snap.
Th10 diversity of contsc15 weens thet e single CVCO welfunctiva, cither elect 11cel er vperetcr errer, cen ectivetc Only one edditionel cherging pmue.
T _ J 4 4 -
- San Onofre 2&3 FSAR Updated m INCREASE'IN REACTOR COOLANT !
- SYSTEM INVENTORY- :
.,1_,_ J_ m. m, ,_ _ _ _ _ _ _ _ _ , _ _ _ , _ m ___m,_m . , _m>___ _ _ _ , _ _ _ ..-.,a n 4 ba4 64 4U s aewa 4 44 AuGu u G 4 muuU , . bug 4uvuu d Q b a 44y A U buvn44 vQ4vU nvuau .
t _ ___ _i_, _ m _ _ __ _ __ __m_ , _ _ _ mt _ . ____ _ _ _ a _t . 1, __ _ mt . !
&JU _ us4 Q4J 4 U bv.vvmyG44GQbU Avs bd4U 444waUQGUw w&&Q& y444y AAvn. 4 444 G wenld re5. , Ult in en increese in the pressurizer-pre 55 die end !
t , _t . , _ _ _ _ , _, --._,,__,1- m_2 _, , t A U V U 4. e _ TV44U A4 Q 4 4 4 %J 4 4 g / & U G G LA & 4 4, U & AUyG4 G 4 (J A & Q 4 4G %d 5 b U Y b U b4 p Q44 f m___,__ _ - - - __m_. _ ___ m_ . _, 1- . _ 1- . m _ m.,,,, b 4 4 Q A, bJ A 4 4 y g/ k&&ug/ Alt \/ b b/ & Q G&E b & A g./g> U 64 g . bC A 4LL4 4 & Q b Ad dy b44U V Y V eJ _ _ _ , 1__ __1 , _ _ _ mt _ _ _ _ ____ , _ _ _ _ , ______________i- . _ _, _-.__i-___ , ' 4U Q 4 & b4 5 A b b 4 b/ 4 4 e & 44U& G , Q A U____ b n bl 4UVU4 4HU Q G %d A U&i&G & 4 b Q&3\d bbi4 4 b & h/ 4
)
chennels, , eech heving e process signel iridicetor erd e low levU1
.__a t , _ t_ _ _ _ _ , _ , _ _ _ _ _ ._m _ 2 _am . .-
o &.u .44y.4 , 4UvC4 Q4Q&m. Q G G v v a Q s. U u n d s.4 4 Ab. 4
. Lt _ - - - - - - m mL _L 1t L , .t , - - _ _ , - , _ _ , 32 A _ _m t- , - - mL - _L *
+4 , b445 U V U44 6 b&&Q 6 bd &U &&4bjl4 AUVG4 b4 4 %4 , 44k b b&4U 1 m_ < >
G 4 %J 4 4 Q 4 b h 4 }J V4 4 Q & \J 4 4 4 kJ
----- _ _t , _ _ _ _ , . _ _ _ , . .t . _ , - __ _ _ _ , _ _ _ _, , m_ _._ _ _ . _ , ___
,_ ts umyG g 64aC 444 y &4- 4 G v u. G aya4Q4 n444 wa4 G 4 4 U A y 4 4.U G 244 b4sU y A U G G u A 4 4.U h - . .,___m__ -_, , , - _ _ _ _ . , _ . . . . m_ <_2 , _________m__ _ , _ _ _
& & U Q hr G 4 G n444 &UQM4b 4 44 Q44 4 4 4 bs 4. U Q G U 4 44 b4&G AJk/ 4 4 4 4 4\J 4MQ G G 64Q&4G&Uh to the steem region to increese - the pressuriser pressure t.o the.
. __ , - - _ _ _ _ _ _ . _ . _ _ _ , _ _ _t .
d gJk# 4 4 4 b \# & %)U &4U & Q b 4 4 43 Q 4 A4y&4 jJ & G G G b& A 4 de U A }J4 U G G kd A G A U Q b b %d A b & 4 gJ
.signel_before the presenriser fills,
,m - -- 4 , . _ , _ - -, ~ , . ._,- _ _ _ _ _ _ _ , _ _ _ , _ -_ , -- - ---_24 .- . m _e_
&4 Y U 4 4 4A b 4 &U K AJwW 4G ak & 4 44AU 4HQ44 u Q 4 4Hb/b6U g Q&Abd 4 4 %# be A U b4 4 6 4G b Q 3% U 4 4 for.the ene4. vising Of 611 the pressuriser hesters en high level,
___2 .__t - -- , _ _ _ , _ _ _ _ __ __ , m_ _ , - .____ -- . - - t , -.t G &Ak444v _ _b _A -Uk4 - _4 mm 6 4G b Q Te U 4 4 & k/ A b A 4 g/pd 4 4 4 kJ bJ & b44U L A A Q k kJ 4 4 4 kJ gd %d4M).J G \/44 444kj44
, _ _ _ _ , m t. _ _ _ . _ _ , _ _ _m _ _ . _ _ _ . _ _ _ _ , _ _ _.._m _ _ _ ,i,,,,,,,i ____,a _m , 2 , t- _
j- ak U V U 4 g b&4U 4 4 44 be 4 U Q &.. G b U Q3H G big /g#4 g G J G b UAM \ 41 V V V / nk/ 64 4 b6 Gb444 L/ U i________t_- 1- - _ _m _ 1_ , _ _ ___m_. , _ . - t , _t _ __ _ _ _ _ . _ _ . AJ k b/ k4kJ & 4 b b b/ Q G b Q b# 4 U U k/44b44 6 4 bl44 AJ J Q 4 4 4 kJ & 4 - }J & U G G b& & 4 de U & g4 A U G G 64 & G reecter trip vr cpereter ection. "
._ errorsicans
,. Sev_er_slyfa._
~ -
u u l.ts_5or?m. _ c._b~e,t postulated ~t th_a gill w~ w7d~ e.~d , a_ to..- ra
~hargin~geandfl downsflow+~?and finnturnwsul~tgin misma,tch?betwe. . ensc
+ ~ . c n et. , = -vel? andf .pressu_reRj . a.
, izerfle For[exampl~e,1with an,tincrea~singyp~re. ssur i
theR PLCS Kini manual $th, e#,ie td6wn%v.alve?c6uld m . 2 m - c . s c16s e_Mr6ducin_y
~ .. .
letdownnflowyt. . ~ ~ - .of zer.o, m gwhil~etcharging4 flow . - - s r_emai - .
.n.s i. c o n s t a n t W_ d ,n...
m . , . . . -
;th;e~ts amev mode f y ais tandb~yd cha g r . in.g[ pump;couldsstartfasstn megresult 6fia ss' ingle 7equipme nttf aulty orlope r'a to $str6 r ;Ti$thiW event 2 M th2c6nstan Mletd6w Q lown pse'ssurizerilevelM ouldlin6reasM Ths311
~ s M th.sYsVedtEthst Would,hiltihCE6dsfit's?ffs4WE6EPT,~lEcideHEMEbin~e_ntory)s dead 5t.oKtheMno _ aser___
v
~ , -
occu 1- + ther~a~utomaticimodew_st+vrapidV,incre m _ m 6 Thi s s,incidenticaus e s~y
. ~the insRCS _
l pre s suri.zertl. . .evel
. . h .. ~c L
ti ho,n,tro.11ent._o$ ope ra t ingscha rg ing( pumps fail,6a_nd(tran_ n (as sumi_ngs thasmit m ti @at.s si'gn, x ingstandbyh.__d aF,bothgareg.m# m an m blose3thesletdownMontrolivalieg
~ -- m s m x _ m
-defective m 1
.be+Nannunci~atedrifsthe?failu~e?wastoffascales AIf6Wilev_Me_ve.lsinstrument~tchanne_lMandsl
-~ ~c .
& lowdevel?alarmVw - . .
r b M s6sisilo W @A %igh71eveR andsthshsashish?hig d1hvelW iafmEwilli ~
.- B, ~ eta,shu,nciate, s,bihthetalteririst_eWh,snds_na. ssactu_a91evsl . _._" ~_~ "
Anc~reases,u lnithlssdis cuss ionttheislternateechan~nelWrefers - -n n-da Moe fe.l,evdMd.c.. .i, .o. .. Q.wc. u.N.a.nne.M.mv.t. h. .i. .t.-tM e v - s - . .c
,lh, , .asi. .h.. o,o M.<b....e..e. h...N.. ,e.. ..l e. s. c..t. ie. d_#
m c #w < w - m e
., ,, ,, ,-_ , _ ,, , , - _ _ . . . _ m , . , , , _ _ . , . , , . - _ < _ ,. __
.i.
1 San Onofre 2&3'FSAR' Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY
- ThWilhrt fal?prWasuElisf?Tev'sWraf assuinsd Wo!bsij ns tS bs16w J ths highMevelll alarm 7setpointisuch:Ktheti the;1 alarm Jl's! prellentil shortly af ter2the; bsginninge ofi the Deventb JIndicationstincludektheVsecond phannelNffpressuriserflevelhchyginglandlletdown[flowira_tefadd Allicharging;pumpsirunn.ing.:
ThWopsfhli6f?fi'{ekpebt'sdff67Yslsp6ndiltW thsfsla rms!aEd fliddfest fohs: eitherf by Eswitching i to i thelsecondilevelX channel %for L controlif or: bp2stoppingithelchargingfpumpsimanuall;yborfbyfrestoringsletdown? Th'siptimar ? prs'asdieffsflimited7b'pTth~eTHihhfPr6ss6fizeiEPressu;rs.
;(H P P )htdipia nd ip r e s s u ri z e ds a f e tie s , dorf bylope ra to ria c tionit o terminateithefeventi The secondary pressure would norgielly 'ee lis limited by the turbine bypess velves. If the stesui bypess control systewi is in the Rienvel iuOde, the secondery pressure wedld be lisiited by the s t e ein v enereter 5efety velve5. s t shd? 6 ppa s s"cb6 tirdlis $s t"siQ( S BCS E pa lves $ o r2by] theis t e ain3ensra to rda f e t yjyalve's fi f fS BCsii;s hnot available.:
The consequences of a single 66mponsnC6fIsVstiesi malfunction of a codiponent er systuni following e CVC3 wielfunction thfsieVsnt are discussed in paragraph 15.5.2.1. 15.5.1.1.0 Core and System Performance i A. Mathematical Model Ths7 NSSSilsspodssIf6? ths9 CVCS imsl fUnct10nNAs /sih615 tied usin githeSCES EC-II I dciompO te rip r o gf am[de s cribedjin
~
S_e_ctionfl5]01 A bounding calculation was done for the CVCS malfunction to determine if the potential existed for filling the pressurizer before operator action can terminate the charging-letdown flow imbalance. The total volume added by the charging-letdown flow-imbalance _over a % ~ 15sminute period, which is the l conservative time assumed for operator action, was determined. Finally, a liquid volume increase is produced by condensing the steam-in the pressurizer due to the proportional-sprays.- This volume was t conservatively assumed to be the liquid volume that i would result from condensing the total initial steam region in the pressurizer. The sum of these two
' i s
I San Onofre 2&3 FSAR Updated- : INCREASE IN REACTOR COOLANT ! SYSTEM INVENTORY volumes, plus the initial pressurizer liquid volume,. , produces _the pressurizer liquid volume based on initial conditions that could be produced during a CVCS . malfunction. , B.^ Input Parameters and Initial Conditions : The input parameters anc f.nitial- conditions used to analyze the NSSS response to a this CVCS malfunction are
' discussed in Section 15.0. The results presented below ere 6ppliceble threnghent the reecter opereting 5pece defined in section 15.0. Thysrticdl'affUth0su pa Eame t ers[dhihuestibithisYan alys ilsla r e211ls t edilETibis 15K5dli SinceisthETpridry'doolantE koac t6rT tilplaicaused i by?phissiffe'itikhe an?indfeaseilnspriharp iht?56 fdrasths boolanWinvento ryhandi noti byfrea dtoripoweriincreaseMHO .
powerNeoolantitemperaturedorfDNB2transientdis producedjeforentheftripI MiHisisiiif theWiViftiis1MCSIpissiWGFsEiss51EsHiHEh6 Iongeststimelpbssible/for6fi'111ng Whicht makss3 theipressurej ahilielwoMl theypres shrizer,} aelontahtripw
- Howeve r,4 maximi zingfinitiale RCipres stile icads 'sfa6 fe'a rlie rp t pip J andia ll'owsleno6 ghRiiime ff o rstihefda r gind pumpsS t of r ai se -L RCS g pr e s.sdrefa ga ini be f o r6[ ope ra t o r actions 1sicreditedhiThelpeakiRCS?pressureEon kepressdritat;ionlisborseithardonltheitkih2 ~
C. Results The following scenario describes the sequence of events that would occur during the limiting CVCS malfunction. The increase in reactor ~ coolant inventory initiated by 4 the- startup of a th'e CVCS charging pumpsjssdijf6sM6f latidoQii produces an over-pressurization of the reactor
-coolant system (RCS).- The increasing pressurizer ^
pressure activates-the-proportional sprays which slow the pressure increase by condensing steam in the pressurizer. The incree5ing pre 55nz1zer. pressure eventuelly ectivotus all of the propeitionel spreys.
~
The addition of water to the RCS by both sprays and ' charging. increases the pressurizer liquid volume and
-r Y;
~
San Onofre 2&3 FSAR
~ Updated INCREASE IN REACTOR COOLANT SYSTEM' INVENTORY hence raices the water level in the pressurizer.- The rate of-filling is slow enough so that operator action in se ).5 minutes to terminate the charging-letdown flow 1
l
- imbalance is sufficient to prevent filling the
-pressurizer. The total' increase of water level in the pressurizer during the first.99 15 minutes-of the ,
transient is approximately-4M BO% of .the available steam volume. After this occurs, the pressuriser pressure increeses et e more repid rete. If h)~sisM13eHspfsy Q sfi]M# jiiMKilllably the RCS pressure. increases so that a high-pressurizer pressure reactor trip is-requiredfd Ths RCS coolant contraction on trip decreases the. rate of liquid level rise in the pressurizer so that operator action in 90 15 minutes is sufficient to prevent filling and produces an immediate_ reduction of the RCS pressure. The maximum RCS pressure is limited by the high-pressurizer pressure reactor trip and the steem genereter pfimsEy safety valves to 110%.of design pressure. Also, the steam generator safety valves limit the main steam system pressure to~within 110% of design. Therefore, the integrity of the RCS and main steam system is maintained.- 4 The CVCS malfunction transient is slow enough so that the core protection calculators will assure that the minimum DNBR is greater than 1.1^^ 13J1) throughout the ' C/CS malfunction transient indicating no violation of the fuel thermal limits. Thi$lhyWsliiiE?6ehsV16f?6'ffthsPslydf fidsEEINS$SIpifssiEsEs
'following EaTCVCShnalf uri6flonkaredhownsinp16dresif15F5s iiithr6 ugh il 5 :5 .12 MadthRsMghehs:eMfisfMntslisigived
[An!7abl @ 51532) 15.5.1.1.4 Barrier Performance A. Mathematical Model
=
The mathematical model used for evaluation of bc rier performance is identical-to that described in paragraph 15.5.1.1.3. k
* --4
. . . .. ~
s. San Onofre-2&3 FSAR Updated-. INCREASE IN REACTOR COOLANT SYSTEM INVENTORY
-B. Input Parameters and Initial Conditions The input parameters and initial conditions used for evaluating barrier-performance are identical to those .
described in paragraph 15.5.1.1.3. , C. Results FsisisiiifiYsF)!
!7dipsiit^FssifdiV6EsF#siiNVMEF3Flissiy?T@finkthe safety! valve @isifcompletelyCcondensed Muendi lt anh andi no tM elg a s edit.o;jthel a_tmo sphe r e) The steam releases to atmosphere through the steam genarator safety valves and maximum RCS pressure-reached during the CVCS-malfunction transient are not worse than those of the loss of condenser vacuum shown in paragraph 15.2.1.3. This is due to the less severe primary and secondary transient with a CVCS malfunction. The slewne55 ef thu tren5ient ellews the repid N000 coeldown, produced by the decicesing core power dee to the high-pre 55nsirer presense reector trip, to occur before the pre 55nsirer sefety velves Open.
Aveilebility of the turbine Lype55 wenld reseve the need for coeling the N000 vie the etse5pheric steem dume velvcs. 15.5.1.1.5 Radiological consequences The radiological consequences of this event ddsitC5fsER ysllssis ffoissth6Tsec6ndsfpfsystsM are less severe than the consequences of thoaeiduelto the inadvertent opening of the atmospheric' dump valve discussed in paragraph 15.1.1.4. 15.5.2 INFREQUENT INCIDENTS -- 15.5.2.1 Chemical and Volume Control Svstem Malfunction with a Concurrent Sinale Failure of an Active Comnonent 15'.5.2.1.1 Identification of Causes and Frequency Classification The estimated froquency of a'CVCS malfunction with a concurrent single failure of an active component classifies this incident as aniinfrequent incident as defined in reference 1 of Section 15.0. The cause of-the CVCS malfunction'is discussed in paragraph 15.5.1;1.1. Various active component single failures were
1-San Onofre 2&3 FSAR Updated. _. INCREASE IN REACTOR COOLANT SYSTEM INVENTORY considered to determine which. failure has-the most adverse effect following a CVCS malfunction. The eingle feilures considered were (1) etertup of the third cherging pump end (2) feilure of the-letdown system. The effect of both of the5e single feilure5 is to increese the RCS inventory et e fester rete. The stertup Of the third cherging pump produces the scst edverse effect following e CVC3 melfunction due to e more repid increese in RC3 ~ inventory. The"QdrstKsidgl'e?a6tiveifsirufeZiaEthsIlbsRofiloffsite phWeCh titheftitnel o Grea ctors tkip) 15.5.2.1.2 Sequence of Events and System Operation The systems and reactot trip that operate following a CVCS malfunction with stertup of the third cherging pump singlefactlis failure are the same as those described in paragraph 15.5.1.1.2 fellowing & CVC3 melfunction. , 15.5.2.1.3 Core and System Performance A. Mathematical Model-The mathematical model used for evaluating core and system performance is identical to that described in
- paragraph 15.5.1.1.3.
B. Input Paranieters and Initial Conditions The input parameters and initial conditions used for evaluation of core and system performance infresp6nse to iths'i CVCS ? msl f und tl6n?With?A isingleR a c tivei f a llu re areidiscussedfin: paragraph 15:02 Those{parametets im ique? tol this t ana lysisla f ehlis t 6d vin ? Table il5; 523 E TheJsinglelactive1 failure?isdtherlbssLoffoffsitelpowei attthe,timeiof treactoritrip.-(Minimizing?initialiRCS. pressuresdelapsjthefhighThressureitripjaddicausesia higherlpeak3CSjpressurs? ere identicel to those descrihed in peregreph 15.5.1.1.3 C. Results The dynamic behavior of the NSSS following a CVCS malfunction with 16ssE6f[offsit's7posisrZsWtheitinis26f
)(ip ccncerrent stertep cf the third cherging pump is similar to that following a CVCS malfunction, which is described in paragraph 15.5.1.1.3. While the rete of l
s San Onofre'2&3-FSAR Updated
- INCREASE IN REACTOR. COOLANT SYSTEM. INVENTORY.
filling the pressuriser is vueeter fer the CVCS l . malidnction with e single feilure then for_the CVC3 melfunction,- Operator action will correct the CVCS
-malfunction and prevent filling the pressurizer even if
-suchf action-is delayed until-se 15 minutes af ter first l
, indicatiou_of the event.- e The peak RCS and main steam system _ pressures-sys weeld . be within 110% of design ensuring that the integrity of the RCS and main. steam system is maintained following a CVCS ma1 function with lbssE6fMffsitVJ6MiiidiQ$hssfimf pfEfp4ct6r! # 1pi eeneursunt stertup of the third cherging penip. The minimum DNBR is greater than 1.19 Till indicating no violation of the fuel thermal limits.
. ;The idpnsmiBibeh'sWibEfb fVtWeyfj'ghl fi'datiMNSS S 7psfass Esjs following2atCVCS!malfu'nctioniwithisinglesfailuresis
~
jshown41n1Piguresh5(5ii3jthreddh)ll5)5-2.4Mankthe' s egtiepc eio fie ven t sjfi sigivbrCin dabl eM5j 5342 15.5.2.1.4 Barrier Performance A. Mathematical Model The mathematical model used for evaluating barrier performance is identical to that described in paragraph 15.5.1.1.3. _B. Input-Parameters and Initial Conditions The input parameters and initial conditions used for evaluating. barrier performance are identical to those ' described in paragraph 15y572ilT3 15.5.1.1.0. C. Results As in the CVCS malfunction, described in paragraph 15.5.1.1.4,- the steam released to containment or atmosphere-and maximum RCS_ pressure reached are no worse than that released in the loss of condenser vacuum discussed-in paragraph 15.2.1.3. 15.5.2.1.5 Radiological Consequences The' radiological consequences of this event are less severe than _. _ _ ~
e San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT ' SYSTEM INVENTORY the consequences of the inadvertent opening of a steam generator atmospheric dump valve as discussed in paragraph 15.1.2.4. 15.5.3 LIMITING FAULTS There are no limiting faults resulting from an increase in RCS inventory.
5 Y
- SanTonofre 2&3_FSAR-Updated- .
INCREASE IN REACTOR COOLANT SYSTEM-INVENTORY: pmwpq;ysymrwwm.m;imkwseNA.BEEM.u wxa.a%nz. Lina &#uuant s nmyyes- ..x*5.A' 5 fMESEMMAsstnapTroNs7 OR
.wn~-naa. F THE'CVCS!htFUNe'TI6N
? TANALYSIS PirAEster Bs56@t155 t
J51tisX[poisMop461sysMMstl' M18
#6tEihlytR661%hENeispuistHRe7KQ M4 C_ ore;. mas . sg1_pg_ra_t _lfi.E1611b_m,s/.h_t e 113],]
, -s w.p.. ..# . . . > . . . ,c.c. .,.,,;. u.
.vy. . y, ,g.e u u .
2t300 Re,sp c t. ~ , y.gg m.m.w ,.lp ,. . tqq <, f.orgcoo an gayatem,_g..,. qmipressureigpsi.1,,m a ' Mo^de EtoRt(smp6fstiffsIEhe f fi^cleift'f*! EIA Rio delta W K
-e..,, s ,,m -s pp eracoe.. .,.f f,.i. .%.c i e,, . tmm . . .n imul..t.i,pl, 7
Do% .. ..l u n.,w, ier _., 1),j25 s . _C.E.K_TW6..r.. ti_h._?.b_h_Yti.i. p.T._f.04 -
. - - . - E_f. _f_0
- SBCSMmods Mshi2sl
-- m PPCSimo_d4 M_an. . Ma_l.
CVGMfibsfsisks(ehl
- Charging /sgpm- 135 L.et. .d.. o. s_d. Nm'._RC_P?bleid. 6.ff. &"yps.
. . ~ - 4' (I n i tJa l??P. F,#s
-- - s sWi.~teMVo1.Uh.sF.f
~ a t3 b_MS.m*9.9.1.
a 3..
- f. k %..w b m'fh f a m.n e'wgac' f6-w' a';
5 -- has a uuw 7 it um -t - t- - ev 7- -g-- -
.m .-4 ert n+mr. n, e- -
< Nr
t l San Onofre 2&3 FSAR Updated I!1 CREASE Ill REACTOR COOLANT SYSTEli IINENTORY
~
jTABLEJ15*.542 k m .
- a. ',. s I~
, SEQUENCE QF 'EVENTSjlOR* TIIE ' CVCET MALPtJ11CTION EVENT i-Ti me '," Setunint of See Event Value CO . Charginef-floWLniaximizid,Viet' d6Wn?RCP F;135fandT4 hieedoff flow: minimized,1gpm i
214;;1; liibh'pressuriz6r? pressure:tr1s1 "243.7, ejeneratedBpsia 21571 QE?K;beginit'ofdiop 240.0 Ste'ani[ generator:;safetiesLbegin f tof.~open,1 ,L1'12C psia 2GC1; Stoani generator" peak Tpressure,;, psia p :1146 , 732.3 Pressurizer safeticolopen~,;fsla *
, L255.0 73213 Peak:RCSipresspre,;psin '2629 73.6;;,3 MaximuniTpro'sadri'zeri:11gsid:volhme,fft8 F _ <1465.7 736i3 Pres'aurizer2safetiesTclosespsia e 1220!)
901 Ope rator i takes ;;;contro11 o f.: plant !;. l---
2 San Onofre 2&3 FSAR Updated INCREASE I11 REACTOR COOLANT SYSTEM INVENTORY i 01 W . TABLE 15 5-3 ,
! ASSUMPTIONS POR THE CVCS'MAttifNOTION f Mrttysis D , . . m. _
WITICSINGLE' FAILURE PaFameter As siimptloD Initiali;Icohe' power?l~eVelk MQt 3i470 Core inlet".'coolhnt!t'amrieratiirsh _ F 542 CofeImass~2 flow ratM E4 6.L'1bm/hr 143.7 ReactorLcoolantisyst6m pressure,,psin 2,000 ModeratorL;temperatur(costficient,iEf4 0.0 delta K/K Dop'pl e ricoe f f ici ent fmul t i~pli e r 1;25 CEAL Worth 1.onstrip,..l10** E6f;.0 SBCSifmode Mantial PPCS? mode Manual CVCS'flowTm16matchi; Charging,-gpm 135 Letdown t; RCEblesdof f, gpm A hQtiaEPressurifzerDVolume,-ift 3 , , 91.3 ht/ J 1 tM" ( 4 _4 S
.-.-, .-w . ..,-.r.. . , ,.m,, , - ,
o f San Onofre 2&3 FSAR Updated INCREAfsE IN REACTOR COOLANT i 3YSTEM INNENTOR" f 4
, "j: ' ,e
? TABLE 115 & 4
, EffUENCE '0F7 EVENTS ^ FOR2THE"CV.CE,MALFUNCTIONiEVENT
[ ,$;, . JITH SINGLE FAILURE ii :' ~, w. .a Time l! Setpoint Et c Event or value F 10 charejing5 flow' maximized,.ll letdown M RCP flow [135ZanR4 minimi~ zed,4gpm 500;9 HighipressurizeEpEessiire tribigeneratsd and,. loss:, ofG of f si teLpoWer,, psia 2 -2437 60179 CEAsibegiriit~oTdrop F-u 603?4 Pressuri serfsafeticis : open, '; psia >
,l:2550 603;8 Peak: RCS. l pressure,J psia i: ;2629 607.O HakimumTp2 fess.urizerlliquid! volbnisi f t 2 '.;<1465l.7 608..S Pea' Npressure,fpsia F .2200 612'.;;7 S team igeni(rator; sa fetiesl. begin" to[6peni C 1112.2 psia
$17?S S t e ani' g e n eVn to rl p e a kll p rn s s b r e ,ip s la o 11'4.6 901 Cpe ra t 6rB take s / contf o'lio flTpl ant ---
,m. m
, . , ...r_, - ,-,, y- ------c- - - -
P CVCS Malfunction i " I I I I I i 1 i
- n. 0.8 - -
1 E> 76 0.6 - - tc +
'O
.5 3
I. 0.4 - - i n. e o o 0.2 - - 0 - - O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Core Power vs Time Ficure 15 51
CVCS Malfunction i I I I I I I I r 1 1 D.
~ ~
0.8 - -
-E C 0.6 - - ~
O u. i 0.4 - - I e o 0 0.2 - - 0 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Core Heat Flux vs. Time Figure 15.5 2
1 CVCS Malfunction . 2700 , , , , , , , , i 2600 - - t 2500 - -
.g a 2400 - -
ai
!5 h
& 2300 -
B
$ 2200 - -
g e J 2100 - - 2000 - - 1900 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Pressurizer Pressure vs. Time Figure 15.5 3
- -- . . = . -
CVCS Malfunction 2700 , , , , , , , , 2600 -
.9 2500 - -
8. af 2400 - - E 2300 - - e 2200 - - scc 2100 - t 2000 - - 1900 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction RCS Pressure (at Cold Leg Discharge)vs Time Figure 15.5-4
CVCS Malfunction 0.03 , , , , , , , , Total Moderator ----- O.02 - Doppler -
, . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .S.c rafn . ; . . . . . . ,
0.01 - 0 - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - 5.
-1 -0.01 - -
gi
.= .2: 0.02 - -
N. E o -0.03 - - b
-0.04 - -
-0.05 - -
-0.06 - - - . -
-0.07 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Core Reactivity vs Time Figure 15.5-5
CVCS Ma! function 640 , , , , , , , , T-out T-avg ----- T-In - 620 - - 600 - - u. p 580 - - 3 --...........,
;-l " 7Tr ::r..ur.,,ur.,_,u,,,s,, p ,,,,.,,
b c : 540 520 - - 500 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction RCS Temperature vs. Time Figure 15.5 6
O e l CVCS Malfunction i 1600 , , , , , , , , 1400 - - 1200 - - -
~
u 8 . 1000 - g - B E 800 - - 5 5 u g 600 - - 4 E n. 400 - - 200 - - 0 O 100 200 300 400 500 600 700 800 900 Time, seconds ' SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Pressurizer Liquid Volume vs Time
- Ficure 15.5-7 - -- _ . ~ . _ . . . . _ . _ . . . ~ - . _ . _ - - .-_ _ - - _
CVCS Malfunction 250 , , , , , , , , 200 - - 1 ai k 150 - -
.6 LL.
m
.T 5
fe 100 - -
's
!3 e
CL 50 - - 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CYCS Malfunction Pressurizer Safety Flow Rate vs. Time Figure 15.5 8 1
-,n -, --w- , - , - -
t CVCS Malfunction 20 , , , , , ,
. Charging l Letoown plus oleeacu -----
16 - - 1 ( 14 - - - 12 - -
.9 10 - -
S 8 - - t
.a E
m 6 - - en
.E E"
2 4 - - 0 2 - - . 0 O 100 200 300 400 500 600- 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GFNEP.ATING STATION Units 2 & 3 ~ CVCS Malfunction Charging and Letdowrt%:P Bleedoff Flow Rate vs. Tirne Figure 15.5 9
CVCS Malfunction 1300 , , , , , , , , 1200 - - 1100 - -
.G
- 8. 1000 - -
900 - + V) 800 - - 700 - 600 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS htalfunction Steam Generator Pressure vs Time Figure 15 510
- - - - _ - _ _ _ _ _ _ - _ _ _ _ _ - - _ _ _ _ _ - - _ _ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ ___ .- ]
O CVCS Malfunction 1 2500 , , , , , , , , j l 2000 - _
$ 1500 - _
l6 C 1000 - _ E ro O 500 - _ en t 0 - -
' ' ' ' ' i i i 500 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION '
Units 2 & 3 CVCS Malfunction Steam Generator Steam Flow Rate vs Time Figure 15.5 11 r , - - - ,.v. . - . . - - - .
f CVCS Malfunction 2500 ' ' ' ' i i , , i . 2000 - . i e' 1500 - . a N ' CC f 1000 - - s 500 - ' 0 0 100 200 300 400 500 600 700 000 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction Steam Generator Feed Flow Rate vs Time . Figure 15.512
O
- 1 CVCS Malfunction with Single Failure I i i i i i i I 1 -
t 1 0- 0.8 - - I E W- 0.6 - - cc E
' . 0.4 - -
ie e 8 0.2 - 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Core Power vs. Time Ficure 15.513
'e CVCS Malfunction with Single Failure l i i i i i i i i : 1 - - l 1 CL q 0.8 - - E . To
.C 0.6 - -
~
o c
.9 LL f 0.4 - -
E m O I O O 0.2 - - 0 O 100 200 300 400 500 600 700 800 900 Time, seconds - P SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Core Heat Flux vs. Time Figure 15.5 14
I : CVCS Malfunction with Single Failure . 2700 , , , , , , , , 2600 - - i 2500 - - 2400 - -
.o h
&- 2300 - - '
t
.N 2200 - -
5 2100 - - ' 2000 - 1900 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE ' i NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Pressurizer Pressure vs Time Figure 15 515 -
4 CVCS Malfunction with Single Failure 2700 , , , , , , , , i i 1 2600 - -
.g 2500 - -
a 8 g 2400 - d[ IB
.c 2300 - - .8 O
E 2200 - - 32 o O CD 0 2100 - 2000 - 1900 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CYCS Malfunction with Single Failure , RCS Pressure (at Cold Leg Discharge vs Time Ficure 15.5.:6
1 CVCS Malfunction with Single Failure 0.03 , , , , , , , , Total Moderator ----- 0.02 - Doppler - Scram 0.01 - 0 ._________________ f. {. -0.01 - - i ib -0.02 - - 8 e [ E -0.03 - _ 8
-0.04 - -
-0.05 - _
-0.06 - - - -
-0.07 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Core Reactivity vs. Time Ficure 15.517
t CVCS Malfunction with Single Failure 640 , , , , , , , , T-out T-avg ----- T-in ""- 620 - - 600 - - t-k p 580 - - g -- ............. b y ; ,.... .. W ;! g v. , ' 560 : ge i- to : O : W ! 540 2"""""""""""""""""""-' - 520 - - l 500 0 100 200 300 400 500 600 700 800 900 Time, seconds - SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 C\'CS Malfunction with Single Failure RCS Temperatures vs Time
~
Ficure 15.518
CVCS Malfunction with Single Failure 1600 , , , , , , , , 1400 - - 1 1205 - - e e -. 8 y 1000 - - a
~6 3 800 - -
5 e Y 600 - - g O O' 400 - - 200 - - 0 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Pressurizer Liquid Volume vs Time Figure 15.519
CVCS Malfunction with Single Failure 250 , , , , , , , , 200 - - 1
$ 150 - -
b
.E 5
E 100 - - 5 V i n. 50 - - 0 O 100 200 300 400 500 600 700 800 900
-Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure
_ Pressurizer Safety Flow Rate vs Time Figure 15.5 20
CVCS Malfunction with Single Failure 20 , , , , , , , i Chargin i.eidowrrptas bleedo'g -- f ----+ 18 - - k 16 - - 1 x k 14 - - 12 - - 10 - - 8 - - J E m 6 - - I
.5 4 - -
0 2 - - i 0 O -100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE - NUCLEAR GENERATING STATION Units 2 & 3-CYCS Malfunction with Single Failure Charging and Letdowrt /RCP BleedofT Flow Rate vs. Time Ficure 15.5 21
CVCS Malfunction with Single Failure 1300 , , , , , , , , 1200 - - ( 1100 - -
.G E 1000 - -
i E 900 - - b 800 - - 700 - - 600 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Steam Generator Pressure vs. Time Figure 15 5 22
. ~ _ . .- - _.. .. . -- - - _ - . _ - _ _ .
P
- i. i l
CVCS Malfunction with Single Failure 2500 , , , , , , , , 2000 - - 1 1500 - - i Ei W 1000 - - E E 05 0 500 - - (n 0 - - - 500 O 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOFRE NUCLEAR GENERATING STATION Units 2 & 3 CVCS Malfunction with Single Failure Steam Generator Steam Flow Rate vs Time Ficure 15.5 23
O t CVCS Malfunction with Single Failure 2500 , , , , , , , , 2000 - - 8 1500 - - i 16 m
}
1000 - - u. 500 - - 0 - 0 100 200 300 400 500 600 700 800 900 Time, seconds SAN ONOF! NUCLEAR G ENERATI. :iTATION Units 2 & 3 CYCS Malfunction with Single Failure Steam Generator Feed Flow Rate vs. Tim figure 15.5 24 .
{ 6 San Onofre 2&3 FSAR Updated INCREASE IN REACTOR CN?LANT SYSTEM INVENTORY 15.5 INCREASE IN REACTOR COOLAllT SYSTEM INVENTORY 15.5.1 MODERATE FREQUENCY INCIDENTS 15.5.1.2 Inadvertent Oneration of the ECCS Durina Power oneration buring power Operation, the psc55ure Of the RC3 (2250 psi &P exceeds the 5hutoff heed of the 55fety injection pump 5 or the ' cpening pressure of the 5&fety injection tenks. Therefore, & 5purious setety injecte n sctuation signel will not ceuse in-jection- of emerger.cy cool-ing fluid -into the RCS during power operation. 15.5.1;2sl IdentifibationTo~flCabses[and_ Frequency Classificatiori The? estimated'fiequency"ofLan"inadverteht7opsrationEof1ths
- cmergency core cooling system _(ECCS). classifies it as a moderate frequency incident as. defined'in: Reference l'of:section 15'.0'.* An inadvertantioperation of ECCS.that produces P.q unplanned increase An reactor coolant 11nventory.may be; caused'by4perator-error that erroneouslyf. actuates!a safety injection. actuation; signal--(SIAS).
Tho 1nadvertent SIAS-activates all'three charging pumps, isolates letdtMn flow,1 startsethe boric ' acid : makeup : (BAMU) pumps,ashifts charging pump 3 suction!tosthel highly borated'BAMU tanks, starts the safety injection ~ pumps,; andri'solates instrument air to hontainment.- Primary 3 coolant boron concentrationfincreasesIby the injection of. highly borated water from the charging pumps.; During power' operation the pressure-of.the RCS:(2250-psiai
'xceeds the shutoffLhead of theisafety injection pumps:andEthe e
opening pressure of the safety injectionitanks,s sohthi.s equipment has no effectnoitthistanalysis. 15.5?lj2i2 5egnence oflEventsfandiSystemiop;eration The?lithiting~moderats frequenep? incident F (l'.e.,Fthe1eventTthat Would lead ;to ithe;most rapid ^1ncrease 'in pressurizer Llevel)llis ,ali Anadvertent, actuation of SIAS.- SinceJcharging:; flow exceeds Actdown ? flow,c pressurizer level?(liquid volumo)?begins to increase. . TheEcharging pump suction line. will; he ~ cleared?6f los bosongmakeup.waterLafteria'short time andLtheniwil.1;isupply highly boratedtwater.s The borated water injects 1nogative reactivity
~
intosthescore, causingLa.-poker._ decrease 1and'resulting-RCS temperature decrease,whichicauses shrinkage ofithe RCSsvolUids. Thistshrinkageitendsitoicoun.teract_theieffectsof;thelincreaselin
5 San Onofre 2&3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY RCS ? inventoryTf rcW chakgingl and[makss!the3 evsntjlessj adverssithari theJ CVCS;;; malfunction ; event.~ TheLconsequences;'of'alsingleDcomponent orfsystem malfunction foll@ wing 7this1eventfare dis:ussed iniparagraph.15.5.2.2.- 157/ji11203 Core ;and': System) P6rformadce The7 core. andi system 7performarice 3arsmeters7followingEan int.dvertentioperationLof the ECCS-during, power:; operation wodid?bs less adverse than those!followingja;;CVCSimalfunctionfwhich is dkscribeddin:Lparagraph: 15.5 1.1.' . 1l M (li274 Barrier 1 Performance fthe! barrier performance psrametersTfoll'owinglan ~1nadverteht oprtration of; s thef ECCS :durirb7 power -operation would be -lass adverse thanfthoseLfollowing the; loss of;condenserLvacuumfevent discussedLiniparagraph 15f.2.1;.3; 15;511l.225 Rsldiological; Consequences Tho? radiological consequences.due toisteam releasestfros ths secondary system ~arefless severe'than;those dueLto'the inadvertent openi'ng.of;thefatmospheric dump' valve discussedTin paragraph;J15.1'li4. . 15.5.2 INFREQUENT INCIDENTS Inadve7 tent Ooeration Vof the' ECCS Durina Powid
- 15. 5.[ 2 .' 2 Doeration with a-Concurrent'S.jncle Failure of anV Active Comconent
~
15,5.21211 Ident1 fi'ca tiohTo f J Caus es isndl Fre'quencp Classification TheCest'imated frequencyioffan71nadvertentTeperatibn rofftheiECCS Ulthiconcurrent1 single failure classifies?itTas;sh' infrequent eventiincident-as defined 1in Refcrenceilvof section 15.0. 'The 7 single..Lf ailure si's ;the11oss;of o f f site lpowerL att the . timeGof reactor trip.
;15(5?2.2;2 Sequencejof(Evsntsland(System Operat' ion Theisps temsL and [snbsequent l:reactoritfripLthst;-;; opera teif611bWing Tan
ab ^6 San Onofre 2r:3 FSAR Updated INCREASE IN REACTOR COOLANT SYSTEM INVENTORY Jna diisytint@6peisti6nT6f ?thi[ ECCS ;;;du rin gl pose rTopess t idd?Wi thfa' poncurrentysinglefa"etiveJfailuredaresthelsameiasithoseldescribed Anlparagraphi:-1515[2Q12] IN 151512f2)13 CorsladdISystss!PifformW6ce Thisc6re Endd? sys temi psr f 6rssiidei@aYsms tsrstif oll'oWing Tid a 'ina dver te n tfor,e r a tiont;o f t th e d ECCS lTdu ring [ poWe r s opera ti otif'sith7s iconcurient2i singl e f active Tf a11ure tNouldibe 'lessT. adverse ': thsn ithods fo11oWing?a.LCVCS i malfhi.icti6niWith! Main'g1htactive[failurekas
#e s c rlibe.diinE pa ra g raph s1M 5{2117 1515121214 BsifildilPefforsaiiO4 TEs7 bsfffsfFpsff6rssnbsTp sisms tefsfif6116WiiiVfsiQins diisEEehti '
ope ra t ionCo f ;(the h ECCS ? during Mowe r[ ope ra t 16niwi thi a$ s ing1 e @sh tlifs f a 11 urea wouldi be ino llgrea ter7 than J thos'eb forithefi'os silo fi.conderiser lva cuumla ssde s cribed ;;id.2 para gr_aph1151211M 31 157572I2Y5 BsdlBibgida1XC6dishsehEss ThsTradi61odi'ch1?coris e quen ess i dde ? t6?'s te as? r si;ess ss? f f6siths secondary;systemfarelless se'vereithanZthoseidueKtolthe lnadvertent? openingsfitheiMmospheric@n:mp[961Eeidiscusssldl.iri
~
haragraphj;.15M110.4j; f 15.5.3 LIMITING FAULTS < ! There are no limiting faults resulting from an increase in RCS inventory. ,}}