05000361/LER-1998-003, LOCA Evaluation of Safety Significance of Failure of Emergency Sump Valve Linestarter (LER 1998-003)
| ML20217Q582 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 04/06/1998 |
| From: | Flournoy W, Remick T, Rustaey A SOUTHERN CALIFORNIA EDISON CO. |
| To: | |
| Shared Package | |
| ML20217Q561 | List: |
| References | |
| NUDOCS 9804130040 | |
| Download: ML20217Q582 (52) | |
| Event date: | |
|---|---|
| Report date: | |
| 3611998003R00 - NRC Website | |
text
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Page1 LOCA Evaluation of Safety Significance of Failure of l
Emergency Sump Valve Linestarter
{
(LER 1998-003) l l
l j
Prepared by :
7 M
Date:
4/6 /'1 f Tom Remick Prepared by :
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rJ6 M 4 Date: Y/(/9 #
William Floumoy Prepared by:
d b-Date:
M4/f8 Paul Barbour P
Reviewed by :
[/B/p,(2057A67 Date: 9/S/92 Abid Rustasy Approved by:
b Date: Y Vick #dzareth 8
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{BR**iB8!EE8jgt p
Page 2 TABLE OF CONTENTS i
pESCRIPTION PAGE S U M MARY.........................................................
3 2" RC S B REAK......................................................
5 INTRODUCTION - 2" RCS BREAK.................................
5 IN P UTS - 2" RC S B REAK........................................
5 ASSUMPTIONS - 2" RCS BREAK..................................
7 REFERENCES - 2" RCS BREAK...................................
8 COMPUTER CODES - 2" RCS BREAK..............................
8 ANALYS I S - 2" R C S B R EAK......................................
9 Containment Pressure Response - 2" RCS Break................
9 ECCS Evaluation - 2" RCS Break............................
10 CONCLUSION - 2" PIPE BREAK.................................
11 3" R C S B R EAK.....................................................
1 8 INTRODUCTION - 3" RCS BREAK................................
18 INPUTS - 3" PIPE B REAK.......................................
18 ASSUMPTIONS - 3" RCS BREAK.................................
19 REFERENCES - 3" RCS BREAK..................................
20 COMPUTER CODES - 3" RCS BREAK.............................
20 ANALYSIS - 3" RC S BREAK.....................................
21 CONCLUSION - 3" RCS BREAK..................................
22 LARGE BREAK LOSS OF COOLANT ACCIDENT.........................
28 INTRODUCTION - LBLOCA.....................................
28 IN PUTS - LB LOCA............................................
28 ASS UMPTIONS - LBLOCA......................................
29 REFERENC ES - LBLOCA.......................................
29 COMPUTER CODES - LBLOCA..................................
30 ANALYSIS - LB LOCA.........................................
30 i
injection Mode..........................................
30 Recirculation Mode.......................................
31 Time To RAS............................................
31 Time From RAS To S% RWST Level.........................
32 Time To Core Uncovery...................................
32 CONC LUSION - LB LOC A.......................................
33
Att chm:nt 2 l
Paga 3 l
l SUMMARY LER 1998-003 relates to an occurrence of an inoperable containment emergency sump outlet valve. To support an assessment of safety significance for LER 1998-003, Nuclear Fuel Management (NFM) performed an engineering evaluation of the postulated worst case accident scenario for this event, the Loss of Coolant Accident.
The limiting LOCA scenario is the Large Break LOCA (LBLOCA). The consequences of the LBLOCA scenario bound all the postulated LOCA scenarios. However, based on input from the Probability Risk Assessment (PRA) group several different break size scenarios are evaluated as follows:
1)
The 2" break (approximately 0.025 ft break area) 2 II)
All break sizes greater than the 2" break Separate assessments for these two categories of LOCA break sizes are included in this evaluation. The purpose of this evaluation is to determine the minimum time to core damage after the actuation of Recirculation Actuation Signal (RAS).
Simplifying assumptions were used to minimize the time to core damage including:
Starting of the swing HPSI pump on SIAS.
Assuming only two charging pumps are available.
Using a maximum constant containment spray pump flow.
Assuming the minimum useable level in RWST is 5%.
No credit is taken for Operator action to secure any ECCS or Containment Spray pump after failure to enter the recirculation mode of operation (to extend the time to RWST depletion), or to initiate a plant cooldown via the atmospheric dump valves or the shutdown cooling system. This minimizes the time after RAS that is availiable for Operator action to mitigate the event.
The location of the break was assumed to be in the bottom of the RCS hot leg to minimize the RCS liquid volume available to cool the core after loss of ECCS. This evaluation conservatively assumed that core damage occurs when the RCS level drops 4
below the top of the active core. However, there is additional time available between starting to uncover the core and the actual onset of core damage.
The 2" and 3" Break LOCA cases were evaluated by NFM using ABB's CENTS computer code. While CENTS is not approved for LOCA analyses, the numerical methods and nodal scheme are consistent with those utilized in CEFLASH, a computer code used for LOCA analyses. Therefore, CENTS is a reasonable computational tool for evaluating NSSS response characteristics for small RCS breaks. The CENTS input data base for these cases represents nominal 100% power operation.
Attachm:nt 2 Pcgs 4 i
The time for containment spray actuation was determined for the 2" Break LOCA case using Bechtel's computer code COPATTA, which is an approved code 'or these type of analyres. Initiation of containment spray was conservatively assumed to be coincident with SIAS for the 3" LOCA case.
Fpr the Limiting LBLOCA, information from UFSAR Sections 6.3 and 15.6.3.3 and the Accident Analysis Design Basis Document, Section 4.3.11 was used to perform the l
evaluation. The limiting LBLOCA evaluation bounds all break sizes greater than 3".
This event is dcminated by the break size making the sequence of events essentially the same as the UFSAR case. Initiation of containment spray was conservatively assumed to be coincident with SIAS.
The following times to core uncovery were determined:
1) 144 minutes for the 2" break 11)
For breaks larger than 2":
55 minutes for the 3" break 31 minutes for the large break LOCA i
i i
l
[
l 4
Attichm nt 2 Pcga 5 l
2" RCS BREAK INTRODUCTION - 2" RCS BREAK An engineering evaluation was performed to determine the expected response of the plant to a postulated 2 inch diameter break in the reactor coolant system (breck area of 2
0.025 ft ). The break was postulated to occur in the bottom of the hot leg. This evaluation included consideration of the actua; plant configuration for the period between January 6,1998 and January 24,1998, when the Train A containment sump l
valve (HV-9305) line starter was Jammed, but not known to be inoperable. Because of the unknown failure of the HV-9305 line starter, the Train B Component Cooling Water system was removed from service during this period for maintenance.
Containment Spray pump operation is the largest contributor to the depletion of the Refueling Water Storage Tank (RWST) during this event. A containment pressure analysis was performed using the COPATTA computer code to determine when the containment high-high pressure setpoint would be reached. This time was then used in the ECCS evaluation as the time at which the Containment Spray pump was initiated.
When the postulated hot leg LOCA has injected sufficient water from the RWST to reduce the RWST level to 18.5%, a Recirculation Actuation Signal (RAS) is generated.
At this time, the failure of HV-9305 would become known to the operators. The failure
{
of HV-9305, combined with the Train B maintenance activities, mean that only charging flow would enter the vessel once the RWST decreased below the ECCS suction lines (approximately 5% RWST level).
The charging pump suction is assumed to be aligned to the Boric Acid Makeup (BAMU) j Tanks, and charging flow is injected to the RCS throughout the event.
This engineering evaluation determined the time needed generate the RAS signal, the time needed to reach the point at which no HPSI flow was available (approximately 5%
RWST level), and determined the RCS and core response to the event.
i To support this engineering evaluation, the ABB CENTS code was used. This code is not approved by the NRC for LOCA analyses, but was judged to be a reasonable evaluation tool for this type of small RCS break.
INPUTS - 2" RCS BREAK 1.
HPSI flow is per the CENTS basedeck, and is consistent with RGR Table Vil-1 (Reference 2). To minimize the time to RWST depletion, the flow from one HPSI pump was doubled to determine the 2 HPSI pump flow (neglecting the additional line losses that would occur).
I
l Attachm:nt 2 t
Paga 6 2.
The RWST volume (from 92% to 5%) is as follows l
per Reference 1, Figure 2.1-1:
RWST. toieri =
x 491,207 gallons = 361,037 gallons 92 100 %
j 18.5% - 5%
RWST,,,,,3 3 x 491,207 gallons = 66,312 gallons
=
i 1000.
I 3.
Containment Spray nozzle flow is 2136 gpm per RGR ltem IX.003 (Refarence 2).
4.
The Containment High Pressure setpoint used in the analysis is 5 psig per RGR ltems V!B.015 and VIB.021 (Reference 2). This signal starts containment cooling.
S.
The Containment High-High Pressure setpoint used in the analys:s is 15 psig, which is conservative th the RGR ltem VIB.017 analysis value. This signal starts flow through the containment spray nozzles.
6.
The containment pressure analysis uses the following break flow rates and RCS pressures, per UFSAR Figures 15.6-115 and 15.6-116 (Reference 3). The associated enthalples are taken from the ASME Steam Tables (Reference 4).
2 The UFSAR Figures are for a 0.025 ft cold leg break, but are also assumed to be representative of a hot !eg break of the same size. The enthalpy values used are representative of a hot leg break.
Time Break flow (Ib/hr)
Enthalpy (Btu /lb)
Comments (seconds)
O to 200 18.0E5 613 Water @ 600*F & 2300 psia 200 to 1200 9.0E5 613 Water @ 600*F & 2300 psia 1200 to 2400 2.52E5 1185 Sat Steam @ 1200 psia 2400 to 4800 2.16ES 1185 Sat Steam @ 1200 psia 4800 to end 1.44E5 1200 Sat Steam @ 800 psia l
i Attrchm::nt 2 Pcgs 7 7.
Two 44 gpm charging pumps, with suction from the BAMU Tanks, are available throughout the event. This is the minimurn number of pumps required to be operable per Technical Specification 3.1.9. The charging pumps are not impacted by the Train B CCW maintenance. The total charging flow is 88 gpm, which is consistent with RGR ltem 111.025 (Reference 2). Over the 250 minute analysis interval, the total charging flow is:
Total Charging Flow = 2
- A# E"I x 250 min = 22,000 gallons mm 3.
Per Licensee Controlled Specification 3.1.104, a minimum level of 5000 gallons is required in each BAMU Tank. Makeup to the BAMU Tanks is available from the 25,000 gallon Boric Acid Storage Tank (T-069). If necessary, additional boric acid can be prepared and added to the Boric Acid Storage Tank. Thus, suffici6nt boric acid is available to support the charging pumps.
9.
The nominal Recirculation Actuation Signal setpoint of 18.5% level in the RWST t
is used. This is consistent with RGR ltem VIB.020 (Reference 2).
ASSUMPTIONS - 2" RCS BREAK l
1.
Train B HPSI, LPSI, and Containment Spray are not available due to 1
maintenance on the Train B Component Cooling Water system.
l 2.
The swing HPSI pump will conservatively be assumed to be initiated concurrent I
with the autostart of the Train A HPSI pump, to maximize the rate of RWST depletion. This is consistent with the guidance provided in procedure SO23 3 (Reference 4).
l 3.
For the purposes of the containment pressure analysis, one train of containment cooling and one train of containment spray are assumed to be available. This is consistent with the unavailability of CCW train 3.
4.
Per procedure 8023-12-3 (Reference 5) the RCPs are assumed to be secured by operator action when the operational limits of 1430 psia (in the RCS) or 40*F subcooling are reached.
5.
It is assumed that the HPSI and Containment Spray pumps will lose suction flow when the RWST level has decreased to 5%.
6.
Conservatively, operator actions to cooldown the plan +. using the Atmospheric Dump Valves are not credited.
Attichm:nt 2 Pcgn 8 REFERENCES - 2" RCS BREAK 1.
Calculation J-BHB-029, Revision 0, CCN-1, "RWST Minimum Level to Maintain Safety Analysis Assumptions, including instrument Uncertainties" 2.
RGR-U2-C9, Revision 0 (including Change Notes 1 and 2), " SONGS Unit 2 Cycle 9 Reload Ground Rules" 3.
SONGS 2&3 Updated Final Safety Analysis Report, Revision 13.
4.
ASME Steam Tables 5.
Procedure SO23-12-3, Revision 14, " Loss of Coolant Accident" 6.
CENPD 282-P-A, " Technical Manual for the CENTS Code", dated February 1991.
7.'
Letter # ST-98-203 from I.C. Rickard (ABB) to P.D. Myers (SCE), " Transmittal of Evaluation of 0.025 fta and 0.05 ft: hot leg breaks with the CENTS code," April l
6,1998.
COMPUTER CODES - 2" RCS BREAK i
1.
The ECCS evaluation was performed using a developmental version of the ABB i
CENTS code (which has improved code stability). The CENTS code (Reference
- 6) is NRC approved only for use in non-LOCA transients. While CENTS is not approved for use in LOCA analyses, ABB determined it to be the most appropriate code to use for this evaluation. The SONGS code base deck used has not yet been formally approved, however ABB does not know of any discrepancies which would affect the results of this evaluation. The CENTS reactor kinetics best estimate decay heat model was used.
2.
The containment pressure evaluation was performed using the Bechtel Corporation COPATTA code. The COPATTA code was derived by Bechtel from the CONTEMPT program and is described in section 6.2.1.1.3.1.C. of the SONGS Units 2&3 UFSAR. The COPATTA program was used by Bechtel during the design and licensing of SONGS Units 2 and 3, and is now used under license by SCE in support of continuing plant operations.
Attachmant 2 Pags 9 ANALYSIS - 2" RCS BREAK Containment Pressure Response - 2" RCS Break in order to determine when the Containment Spray pump would initiate, the containment pressure response to a 2" hot leg break event was determined using the COPATTA program. The basic containment model, including performance characteristics of the emergency air cooling units (ECUS) and containment spray (CS) system was the same as that used in current analyses of record for containment peak pressure analyses reported in section 6.2.1 of the SONGS Units 2&3 UFSAR.
The results of the containment pressure responso to the 2 break event are shown in Figure 1. As indicated on the figure, the containment pressure reaches 15 psig at 1200 seconds (20 minutes), initiating spray flow from the one available train of containment spray. The analysis also showed that one train of ECUS would be operating by 170 seconds (about 3 minutes) following occurrence of the pipe break. The analysis assumed that RWST would decrease to a level of 5% in approximately 145 minutes, at which time the containment spray flow was discontinued. This time differs slightly from 4
the 140 minute time subsequently determined by the ECCS analysis, j
A small change in the timing of spray flow termination would not substantially change the shape of the subsequent pressure versus time curve. The general trend of containment pressure at 145 minuter is not affected by small changes in the time of i
spray flow termination. The discontinuities in the pressure curve at 200,2400, and I
4800 seconds are due to step changes in break flow and/or enthalpy used in the simplified input modeling of the break flow. The sharp change in slope of the curve at 1200 seconds is due to a combination of a change in break flow and the start of containment spray.
1 Att: Chm:nt 2 Pcg;> 10 ECCS Evaluation - 2" RCS Break Table 1 ShowS the Sequence of events for thiS event. Figures 2 through 6 Show the response of Selected plant parameters during the event.
1 l
Table 1 - 2 Inch RCS Break - Sequence of Events Time After Break Time Action Comments After sec min S
0 0
Break occurs in 2 inch pipe (0.025 Plant status at event irutiation:
ft' break area)
Train B HPSI, LPSI, and Containment Spray not available due to Train B CCW maintenance 0
0 2 charging pumps inject to RCS 2 charging pumps are the minimum required to be from BAMU available per Technical Specr5 cation 3.1.9. The
)
l charging pumps are nc*
acted by the Train B CCW maintenance.
Makeup to the BAMu tanks is available from the boric acid storage tank.
107 1.8 Reactor trip occurs 118 2
operators secure all four RCPs Per procedure So23-12 3 the RCPs are assumed to be secured when the operationallimits of 1430 psia and 40*F subcooling reached (both limits are reached at approximately 118 seconds).
138 2.3 2 HPSI pumps start ConsrNatively assume the swing HPSI pump is initiated concurrently with the autostart of the Train A HPSI pump, to maximize the rate of RWST depletion.
170 2.8 Containment Emergency Cooler Only one train credited, due to lack of second CCW starts train.
1200 20 Containment Spray flow from 1 CS nozzle flow is 2136 gpm for the ECCS analysis (for pump starts, based on CoPATTA RWST depletion) which is conservative to the value j
analysis of cor,tainment pressure used in the containment pressure analysis.
response to this event 3400 57 SITS start to discharge RCS pressure 615 psia.
Pressurizer and upper head empty, loops essentially solid at 490*F.
RCS is cooling secondary side, which is bottled up at ~760 psia (reverse heat transfer).
Total Si flow spikes as SIT flow starts and stops as a function of SIT pressure and RCS pressure.
4800 80 Containment pressure reaches 20 This is the maximum containment pressure that occurs psig while Containment Sprays are running.
7200 120 0
RAS setpoint ree.hed.
18.5% level on RWST.
Operators discover that HV-g305 fails to open.
i 1
l l
AtttChnunt 2 Pega 11 Table 1 - 2 inch RCS Break - Sequence of Events Time After Break Time Action Comments After see min S
8400 140 20 HWST reaches 5%, ECCS RCS pressure -280 psia.
analysis assumes that operators Voiding in pressurizer, upper head, and upper secure 2 HPSI and Containment region of core (level above top of hot leg).
Spray pumps. only remaining Loops stdl solid, at -320*F.
flow into RCS is 2 charging pumps RCS is cooling secondary side, SG side is botded (suction from BAMU tanks, which up at -250 psia (reverse heat transfer),
are refilled as necessary from the Boric Acid Storage Tank) 9700 162 42 SIT discharge ends RCS pressure 160 psia (minimum pressure for event).
System.it saturation conddions.
Systert, begins to repressurize.
Break flow 70 lb/sec.
SG Pressure -200 psia 10,140 169 49 System has heated up to Break flow has dropped from 230 lbm/sec (at 8400 saturation conditons, break flow seconds) to 42 lbm/sec (at 10,140 sec) becomes saturated break flow 10.500 175 55 SG tubes draining begins 13,500 225 105 System has repressurized to 470 Break flow has increased to -80 lb/see due to the psia increase in pressure.
Core remains covered.
SG Pressure 450 psia 13,700 228 108 Break uncovers Break mass flow rate decreases -25 lb/sec as break is now passing steam rather than liquid.
Flow spikes back to 80 lb/see when break passes liquid.
Core remains covered.
15,870 264 144 Actve core uncovers RCS pressure 550 pria.
Break flow 27 lb/sec.
Charging flow 12 lb/sec, System at saturation conditions.
Secondary side (at 530 psia) is cooling RCS (forward heat transfer).
Containment at 33 psig and 272*F.
CONCLUSION - 2" PIPE BREAK The Core remains Covered for at least 144 minutes after receipt of the RAS Signal.
l l
Attichm:nt 2 Prg312 Figure 1 CONTAINMENT PRESSURE RESPONSE 2" PIPE BREAK LOCA-SONGS UNITS 2&3 COPATTA ANALYSi$
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I Attachm:nt 2 Pcgs 13 Figure 2 2 INCH RCS BREAK - BREAK FLOW RATE J
Time (sec)
Event 0-138 Break occurs, Reactor trips, RCPs are manually tripped,2 HPSI pumps are started. Break flow decreases with decreasing RCS pressure.
900 - 8400 HPSI flow into the RCS promotes increase in break flow.
8400 5% RWST level is reached, HPSI and Containment Spray pumps are secured 9700 SITS discharge ends, RCS begins to repressurize, break flow is increased 13700 Break flow changes from liquid to steam as level drops below the bottom of the hot leg
>13700 Spiking in break flow due to covering and uncovering break location i
~
c.on 300 250 200 If 150 -
1 lC 100 50 -
~
0 0
3200 6400 9600 12800 16000 time (sec) l
1 l
Attachm::nt 2 Page 14 Figure 3 2 INCH RCS BREAK - SAFETY INJECTION FLOW RATE (From HPSI and Safety injection Tank)
Time (sec)
Event 138 2 HPSI pumps are started 3400 SITS start to discharge 8400 RWST level reaches 5%, HPSI pumps are secured, SIT flow maintains level 9700 SITS discharge ends 0.025 600 500 400 7 I
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Attachm:nt 2 Page 15 Figure 4 2 INCH RCS BREAK - PRESSURIZER PRESSURE Time (sec)
Event 0
Break Occurs 107 Reactor trips 138 -8400 2 HPSI pumps start, reducing rate of RCS depressurization 8400 5% RWST level is reached, HPSI pumps are secured 9700 SITS discharge ends, RCS begins to repressurize 0.025 3000
. 6 2500
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0 3200 6400 9600 12800 16000 time (sec)
Attechm:nt 2 Page 16 Figure 5 2 INCH RCS BREAK - REACTOR VESSEL MIXTURE LEVEL
)
i Time (sec)
Event
(
0 -118 Break Occurs, Reactor trips, RCPs secured, all of which drop level 138 2 HPSI pumps are started 600 Level begins to turn around as a result of HPSI 3400 SITS start to discharge, break flow increases 8400 RWST level reaches 5%, HPSI pumps are secured, SIT flow maintains level 9700 SITS discharge ends, level begins to drop 10,000 -12,900 Oscillations in level due to interaction between cold water in downcomer and charging and boiling conditions in core, and due to I
reflux condensation in Steam Generators 15,870 Core uncovers Figure 6 0.025 30
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{
}
top OF CORE 15 E
1 10 5
e 0
O 3200 6400 9600 12800 16000 time (sec) 1
Attachm:nt 2 Page 17 2 !NCH RCS BREAK - TOTAL REACTOR COOLANT SYSTEM MASS 0
0.026 800
- - i.
4 500 400 a
I mb 300 100 a
a a l, a a !a
!a A
R E
O 3200 6400 9600 12800 10000 time (sec) t
Attachmtnt 2 Page 18
)
3" RCS BREAK INTRODUCTION - 3" RCS BREAK An engineering evaluation was performed to determine the expected response of the plant to a postulated 3 inch diameter break in the reactor coolant system (break area of 2
0.05 ft ). The break was postulated to occur in the bottom of the hot leg. This evaluation included consideration of the actual plant configuration for the period between January 6,1998 and January 24,1998, when the Train A containment sump valve (HV-9305) line starter was jammed, but not known to be inoperable. Because of the unknown failure of the HV-9305 line starter, the Train B Component Cooling Water system was removed from service during this period for maintenance.
When the postulated hot leg LOCA has injected sufficient water from the RWST to reduce the RWST level to 18.5%, a Recirculation Actuation Signal (RAS) is generated.
At this time, the failure of HV-9305 would become known to the operators. The failure of HV-9305, combined with the Train B maintenance activities, mean that only charging flow would enter the vessel once the RWST decreased below the ECCS suction lines (approximately 5% RWST level).
The charging pump suction is assumed to be aligned to the Boric Acid Makeup (BAMU)
Tanks, and charging flow is injected to the RCS throughout the event.
This engineering evaluation determined the time needed generate the RAS signal, the time needed to reach the point at which no HPSI flow was available (approximately 5%
RWST isvel), and determined the RCS and core response to the event.
To support this engineering evaluation, the ABB CENTS code was used. This rode is not approved by the NRC for LOCA analyses, but was judged to be a reasonable evaluation tool for this type of small RCS break.
INPUTS - 3" PIPE BREAK 1.
HPSI flow is per the CENTS basedeck, and is consistent with RGR Table Vll-1 (Reference 2). To minimize the time to RWST depletion, the flow from one HPSI pump was doubled to determine the 2 HPSI pump flow (neglecting the additional line losses that would occur).
Attechm:nt 2 Page 19 2.
The RWST volume (from 92% to 5%) is as follows per Reference 1, Figure 2.1-1:
RWST92% tois.3% =
.SY. x 491,207 gallons = 361,037 gallons 100 %
RWST :.5% r.5% " 18.5% - 5% x 491,207 gallons = 66,312 gallons 100 %
3.
Containment Spray flow is 2136 gpm per RGR ltem IX.008 (Reference 2). A value of 2200 gpm will conservatively be used.
4.
Two 44 gpm chsrging pumps, with suction from the BAMU Tanks, are available throughout the event. This is the minimum number of pumps required to be operable per Technical Specification 3.1.9. The charging pumps are not impacted by the Train B CCW maintenance. The total charging flow is 88 gpm, which is consistent with RGR ltem lll.025 (Reference 2). Over the 250 minute analysis interval, the total charging flow is:
- AA E"I Total Charging Flow =
x 250 min = 22,000 gallons Min 5.
Per Licensee Controlled Specification 3.1.104, a minimum level of 5000 gallons is required in each BAMU Tank. Makeup to the BAMU Tanks is available from the 25,000 gallon Boric Acid Storage Tank (T-069). If necessary, additional boric acid can be prepared and added to the Boric Acid Storage Tank. Thus, sufficient boric acid is available to support the charging pumps.
6.
The nominal Recirculation Actuation Signal setpoint of 18.5% level in the RWST is used. This is consistent with RGR ltem VIB.020 (Reference 2).
ASSUMPTIONS - 3" RCS GREAK 1.
Train B HPSI, LPSI, and Containment Spray are not available due to maintenance on the Train B Component Cooling Water system.
2.
The swing HPSI pump will conservatively be assumed to be initiated concurrent with the autostart of the Train A HPS! pump, to maximize the rate of RWST depletion. This is consistent with the guidance provided in procedure SO23-12-3 (Reference 4).
s am
{
Pcg3 20 3.
Conservatively assume that the Containment Spray pump initiates when the SIAS is generated (a detailed containment pressure analysis was not performed for this break size).
4.
Per precedure SO23-12-3 (Reference 4) the RCPs are assumed to be secured by operator action when the operational limits of 1430 psia (in the RCS) or 40*F subcooling are reached.
5.
It is assumed that the HPSI and Containment Spray pumps will lose suction flow when the RWST level has decreased to 5%.
6.
Conservatively, operator actions to cooldown the plant using the Atmospheric
{
Dump Valves are not credited.
REFERENCES - 3" RCS BREAK 1.
Calculation J-BHB-029, Revision 0, CCN-1, "RWST Minimum Level to Maintain Safety Analysis Assumptions, including Instrument Uncertainties" 2.
RGR-U2-C9, Revision 0 (including Change Notes 1 and 2), " SONGS Unit 2 Cycle 9 Reload Ground Rules" 3.
SONGS 2&3 Updated Final Safety Analysis Report, Revision 13.
4.
Procedure S023-12-3, Revision 14," Loss of Coolant Accident" 5.
CENPD 282-P-A, " Technical Manual for the CENTS Code", dated February 1991.
6.
Letter # ST-98-203 from I.C. Rickard (ABB) to P.D. Myers (SCE), " Transmittal of 2
2 Evaluation of 0.025 ft and 0.05 ft hot leg breaks with the CENTS code " April 6, 1998.
COMPUTER CODES - 3" RCS BREAK j
1.
The ECCS evaluation was performed using a developmental version of the ABB CENTS code (which has improved code stability). The CENTS code (Reference
- 6) is NRC approved only for use in non-LOCA transients. While CENTS is not approved for use in LOCA analyses, ABB determined it to be the most appropriate code to use for this evaluation. The SONGS code base deck used has not yet been formally approved, however ABB does not know of any discrepancies which would affect the results of this evaluation. The CENTS reactor kinetics best estimate decay heat model was used.
l Attichm:nt 2 Pcgs 21 I
ANALYSIS - 3" RCS 13REAK Table 2 shows the sequence of events for his event. Figures 7 through 11 show the response of selected plant pararneters dur ng the event.
1 Table 2 - 3 Inch RCS Break - Sequence of Events Time After Time Action Comments Break After RAS sec min (Min) 0 0
Break occurs in 3 inch pipe Plant status at event initiation:
2 (0.05 ft break area)
Train B HPSI, LPSI, and Containment Spray not available due to Train B CCW maintenance.
0 0
2 Charging pumps inject to 2 charging pumps are the minimum required RCS from BAMU to be available per Technical Specification 3.1.9. The charging pumps are notimpacted by the Train B CCW maintenance.
Makeup to the BAMU tanks is available from the boric acid storage tank.
40 0.7 Reactor trip occurs 65 1.1 Operators secure all four Per procedure SO23-12-3 the RCPs are RCPs assumed to be secured when the operational limits of 1430 psia and 40'F subcooling reached (both limits are reached at approximately 65 seconds).
100 1.7 2 HPSI pumps start Conservatively assume the swing HPSI pump Containment Spray pump is initiated concurrently with the autostart of starts the Train A HPSI pump, to maximize the rate of RWST depletion.
Conservatively assume 1 Containment Spray pump is initiated concurrently with HPSI, to maximize the rate of RWST depletion.
280 4.7 SG tubes start to drain 300 5
RCS reaches plateau pressure RCS at-850 psia.
Secondary at -830 psia.
Secondary side is cooling the RCS (forward heat transfer).
2200 37 SG tubes empty 2460 41 Level reaches bottom of hot Two phase conditions at break.
leg System depressurizes.
2760 46 SITS start to discharge RCS pressure 615 psia.
Pressurizer and upper head empty.
SG empty.
Hot legs essentially empty.
Total Si flow epikes as SIT flow starts and stops as a function of SIT pressure and RCS pressure.
SG Pressure ~860 psia.
Attachm:nt 2 Pegs 22 Table 2 - 3 inch RCS Break - Sequence of Events Tirne After Time Action Comments Break After RAS sec mm (Min) 2600 47 Break is covered again RCS is cooling secondary system (reverse heat transfer).
t 3620 60 SIT injectun r rminates, RCS Pressure -195 pala.
a minimum RCS Pressure reached 4050 68 Hotleg is full RCS Pressure -450 psia.
6250 104 0
RAS setpoint reached.
18.5% level on RWST.
Operators discover that HV-9305 fails to open.
7000 117 13 RWST reaches 5%, ECCS RCS pressure ~380 psia.
analysis assumes that RCS is cooling secondary side, which is operators secure 2 HPSI and botded up at -610 psia (reverse heat transfer).
Containment Spray pumps.
Only remaining flowinto RCS is 2 charging pumps (suction from BAMU tanks, which are i
refilled as necessary from the
{
Boric Acid Storage Tank)
>7000
>117
>13 System heats up and RCS depressurizes.
depressurizes to saturation, draining of hotlegs begin 7120 119 15 System reaches saturation RCS pressure ~260 psia, begins to repressurire.
SG Pressure -600 psia.
8300 138 34 Break uncovers Break flow reduces from 160 lb/sec to 48 lb/sec.
RCS Pressure ~480 nsia, begins to depressurize.
SG Pressure ~540 psia.
9540 159 55 Active core uncovers RCS Pressure -375 psia.
SG Pressure ~500 psia.
l CONCLUSION - 3" RCS BREAK The core remains covered for 55 minutes after receipt of the RAS signal.
Attachmtnt 2 Pcgs 23 Figure 7 3 INCH RCS BREAK - BREAK FLOW RATE Time (sec)
Event 0-100 Break occurs, Reactor trips, RCPs are manually tripped,2 HPSI pumps are started. Break flow decrease with decreasing RCS pressure.
300 RCS reaches plateau pressure, break flow begins to increase 2460 Level reaches bottom of hot leg, steam replaces liquid flow at break location, therefore break flow decreases 2800 Break is covered, liquid replaces steam flow at break location, therefore break flow increases 7000 5% RWST level is reached,2 HPSI and Containrnent Spray pumps are secured 8300 Break flow changes from liquid to steam as level drops below the bottom of the hot leg 0.05 300 m Or\\
200 l
1
\\
3.
M J 1
100 50 0
0 3000 6000 9000 12000 15000 time (sec) 1 l
Attachm:nt 2 Pcg3 24 Figure 8 3 INCH RCS BREAK - SAFETY INJECTION FLOW RATE (From HPSI and Safety injection Tank)
Time (sec)
Event 100 2 HPSI pumps start 2760 SITS begin to discharge 3620 SITS discharge ends 7000 RWST reaches 5% level, operators secure both HPSI pumps l
l 0.05 600 g
=
400 I
300 a
200
' b "' '
100 4
0 0
3000 6000 9000 12000 15000 time (soc)
T Attrchm:nt 2 Peg 2 25 Figure 9 3 INCH RCS BREAK - PRESSURIZER PRESSURE l
Time (sec)
Event i
0 Break occurs l
40 Reactor trips 100 2 HPSI pumps start l
2460 Level at the bottom of hot leg, steam replaces liquid flow, RCS begins to depressurize 2800 Break is covered again l
3620 SITS discharge ends i
7000 RWST level reaches 5%, HPSI is terminated 8300 Break uncovers, pressure begins to drop 0.05 3000
... =.,
2500 2000 I
1500 1000 500 -
l O
0 3000 6000 9000 12000 15000 time (sec)
l Attcchm:nt 2 Pag 2 26 Figure 10 1
3 INCH RCS BREAK. REACTOR VESSEL MIXTURE LEVEL j
Time (sec)
Event
)
100 2 HPSI pumps start 2760 SITS begin to discharge 3620 SITS discharge end 7000 5% RWST level is reached, HPSI is terminated 9540 Core is uncovered and remains uncovered i
1 1
0.06 30 i-r 25 2
0 I
20 k
E 1
y 15 10 7
m 5
h
,t,
.t.,,
,t.
,I.
~
0 3000 6000 9000 12000 15000
)
um. < c)
.m
Attachmsnt 2 Pcg3 27 Figure 11 3 lNCH RCS BREAK - TOTAL REACTOR COOLANT SYSTEM MASS 1
0.08 600 i.
f 1
500 400 E
c 300 200 100
.l.
,t.
.t, 0
'M)00 6000 9000 12000 15000 time (sec)
Attachm:nt 2 Pago 28 LARGE BREAK LOSS OF COOLANT ACCIDENT lNTRODUCTION - LBLOCA An evaluation of Large Break Loss of Cooling Accident (LBLOCA) was performed during plant operation from January 6,1998 to January 24,1998 when Train A containment sump outlet valve 2HV9305 was inoperable. During this period, the Train B CCW was out of service for maintenance. Hence, Train B ECCS was inoperable.
With Train A containment sump outlet valve inoperable and Train B ECCS inoperable, recirculation flow from the sump to the core would not be available post-LOCA unless the Train A sump valve was restored to operability, Train B ECCS was restored to operability, or Train A HPSI was cross-connected to Train B sump vaive to provide flow.
j Following Recirculation Actuation Signal (RAS), ECCS flow would be available for a j
short time by depleting the RWST below the RAS setpoint.
The evaluation is limited to the LBLOCA recirculation mode, since the injection mode is not affected by the failure of the containment sump outlet valve. The evaluation determines the minimum time to RAS following LBLOCA for the plant conditions and I
equipment operable during the affected period. Then, the time post-RAS to reach 5%
RWST level is determined. Finally, the time to core uncovery assuming no HPSI flow is calculated assuming Si flow from 2 charging pumps.
J The minimum time to RAS, time to 5% RWST level, and time to core uncovery are hand calculations (see Section 3.0) assuming no HPSI pump flow to the core. Long term core cooling following LBLOCA is normally assured by operation of one HPSI pump during the recirculation mode.
INPUTS - LBLOCA Time To RAS and Time To 5% RWST Level 1.
HPSI pump flow = 850 gpm @ 50 psia Ref. 2, Table Vll-1 A 2.
LPSI pump flow = 3805 gpm @ 50 psia Ref. 2, Table Vll-3 3.
Containment spray pump flow = 2136 gpm @ spray nozzle Ref. 2, IX.008 4.
RWST Volume To RAS (92 - 18.5%) = 361,037 gal Ref. 3, Figure 2.1-1 5.
RWST Volume To RAS (18.5 - 5%) = 66,312 gal Ref 3, Figure 2.1-1 I
Attachmint 2 Pega 29 Time To Core Uncoverv 1.
Water volume above core (bottom of hot leg to top of core)
= 603 ft* (extrapolated from midloop volumes)
Ref. 5, Table 7 2.
Water density = (1/.017482) Ibm /ft* @ 70 psia Ref. 4 3.
Charging pump flow rate = 88 gpm Ref. 2, lll.025 4.
Decay Heat = 167 x 10' Btu /hr @ 1 hr after reactor trip Ref. 6, pg. 5 5.
Heat of vaporization = 907.8 Btu /lbm @ 70 psia Ref. 4 ASSUMPTIONS - LBLOCA 1.
Actual RWST level during the period was 92% per Operations.
1 2.
The swing HPSI pump will conservativel: be assumed to be initiated concurrent with the autostart of the Train A HPSI pu,1p, to maximize the rate of RWST depletion. This is consistent with the guidance provided in procedure SO23-12-3 (Reference 7).
3.
Train B HPSI, LPSI, and Containment Spray pumps are inoperable. This is the plant condition described in LER 1998-003.
4.
Two charging pumps are automatically started on SIAS. This is normal system operation on SIAS. The third swing charging pump would be started by the operators, but its flow is conservatively neglected to minimize the time to core uncovery. Since the charging pumps take suction from the BAMU tanks, and not the RWST, they do not deplete the RWST.
i 5.
The suction source for the charging pumps is assumed to be the BAMU tanks, backed up with the BAST.
l REFERENCES-LBLOCA 1.
UFSAR SONGS Units 2&3, Revision 13 2.
RGR-U2-C9, Revision 0 (including Change Notes 1 and 2), " SONGS Unit 2 Cycle 9 Reload Ground Rules" 3.
RWST Minimum Level To Maintain Safety Analysis Assumptions, including Instrument Uncertainties, J-BHB-029, Rev. O CCN-1
f l
Pago 30 4.
1967 ASME Steam Tables 5.
RCS Heatup Following Loss of bDC, N-0220-029, Rev. 0 6.
SONGS 2/3 Single Fuel Assembij Decay Heat, N-1020-043, Rev. 0 7.
Procedure SO23-12-3, Revision 14," Loss of Coolant Accident" COMPUTER CODES - LBLOCA No codes were used to perform this evaluation.
ANALYSIS - LBLOCA Injection Mode Following a LBLOCA, Train A HPSI pump, LPSI pump, and containment spray pump are actuated on SIAS and CSAS. Train B HPSI pump, LPSI pump, and containment spray pump are not available due to Train B CCW out of service. The operator is assumed to start the swing HPSI pump which is aligned to Train A immediately after SIAS. This will minimize the time to deplete the RWST. The HPSI pump, LPSI pump, and containment spray pump take suction from the RWST.
Train A and Train 8 charging pumps are actuated on SIAS and take suction from the BAMU tanks. Letdown is isolated on the SIAS. It is assumed that the third swing charging pump is conservatively not started by the operators to minimize time to core uncovery.
Since one Train of ECCS is operable during the injection mode, the UFSAR LBLOCA safety analysis is valid for the injection mode. The LBLOCA injection mode (includes blowdown, refill, and reflood) is not affected by the failure of the Train A containment sump outlet valve 2HV9305. The Peak Clad Temperature (PCT) for LBLOCA occurs during the injection mode at approximately 270 seconds (UFSAR Figure 15.6-106).
Hence, the PCT for LBLOCA is not affected by the failure of 2HV9305.
f Attrchmant 2 Pcgs 31 Recirculation Mode The post-LOCA recirculation mode is initiated automatically when the RAS is reached.
The RAS initiation setpoint is at 18.5% RWST level. When RAS is reached, the Train A LPSI pump trips and the Train A containment sump outlet valve 2HV9305 is supposed to open. In the case being evaluated, it is assumed the sump outlet valve 2HV9305 fails to open. The Train B sump outlet valve 2HV9304 opens but the Train B l
HPSI pump, LPSI pump, and containment spray pump are inoperable due to the Train l
B CCW being out of service.
l On RAS, Train A LPSI pump trips, but Train A HPSI pump, the swing HPSI pump, and l
the containment spray pump continue to run and take suction from the RWST because sump suction.is unavailable for Train A (failure of 2HV9305).
{
Time To RAS l
The minimum time to RAS (18.5% RWST level) is calculated by depleting the RWST l
volume from 92% (actual initial RW9T level during the period) to 18.5% (RAS setpoint) by 2 HPSI pumps,1 LPSI pump, and 1 containment spray pump.
Safety analysis values are assumed for HPSI pump, LPSI pump, and containment l
spray pump flow rates. The flow for 2 HPSI pumps is conservatively assumed to be twice the safety analysis flow of 1 HPSI pump.
l l
Time to RAS = RWST Volume To RAS /
[ 2 x HPSI pump now + LPSI pump flow + spray pump flow ]
Using values from Design input, L
Time To RAS = 361,037 gal / [ 2 x 850 gpm +3805 gpm + 2136 gpm ]
= 47.2 min 1
l l
l L
i
AttachmInt 2 Pag 3 32 Time From RAS To 5% RWST Level The minimum additional time to 5% RWST level is calculated by depleting the RWST from 18.5% RWST level (RAS setpoint) to 5% RWST level assuming 2 HPSI pumps and 1 containment spray pump running. The operator is assumed to trip the HPSI pump and spray pump when RWST level reaches 5% to prevent vortexing and pump damage.
Using the values from Design input, Time From RAS To 5% RWST Level = 66,312 gal / [ 2 x 850 gpm + 2136 gpm ]
= 17.3 min Time To Core Uncovery After the operator trips the HPSI pumps at 5% RWST level, the only SI flow remaining to the core is from 2 charging pumps. The time to core uncovery (TTCU) is calculated i
by assuming the minimum mass above the core is depleted by the bolloff rate minus the charging pump flow rate.
The mass above the core conservatively assumes that the RCS level is at the bottom of the hot leg when the HPSI pumps are stopped. The decay heat used is at the time HPSI flow is terminated (approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after event). The boiloff rate is i
conservatively assumed at 70 psia (maximum containment backpressure of 55 psla) without credit for charging subcooling.
Time To Core Uncovery (TTCU) = Mass Above Core / [ Boiloff Rate - Charging Rate ]
Using values from Design input, Mass Above Core = (603 ft') x (1/.017482 lbm/ft3) = 34,492 lbm l
Charging Flow Rate = (88 gpm)(1/7.48 ft'/ gal)(1/60 min /sec)(1/.017482 lbm/ft')
= 11.2 lbm/sec Boiloff Rate = 1.67 x 10s Btu /hr / 407 A Btullbm = 1.84 x 10 lbm/hr = 51.1 lbm/sec 5
Time To Core Uncovery (TTCU) = 34,492 / (51.1 - 11.2) = 864.5 sec = 14.4 min 1
f Attachm:nt 2 Pegs 33 l
l Table 3 Large Break LOCA - Sequence of Events Time Time After Action Comments After RAS i
Break (Min) l (Min) 0 Break Occurs 0
2 HPSI pumps and 1 LPSI pump Assume swing HPSI is initiated earty, startinjection to RCS and 1 HPSI flow to same line for given Containment Spray (CS) pump pressure in UFSAR case. LPSI flow for starts spraying to containment from given pressure in UFSAR case. CS at RWST.
2136 gpm nozzle flow per Reload SITS initiate injection.
Groundrules.
2 Charging pumps inject to RCS from BAMU i
1 SITS discharge ends 47 0
RAS setpoint reached, LPS! trip 18.5% level on RWST
{
47 0
Operators discover that HV 9305 f
fails to open
\\
64 17 2 HPSI and CS secured Reached 5% level on RWST and pumps secured to prevent damage 78 31 Top of core uncovered Assumes two charging pumps operating l
l CONCLUSION - LBLOCA The results for the LBLOCA evaluation are:
Time To RAS
= 47.2 min Time To 5% RWST Level
= 17.3 min Time To Core Uncovery
= 14.4 min The total time from RAS to core uncovery is the sum of the latter two times, or 17.3 +
14.4 = 31.7 minutes. This represents the time required for operator action to establish HPSI flow to augment the charging flow and prevent top of core uncovery.
I ATTACHMENT 3 Small Break Loss-of-Coolant Accident (SBLOCA) with incomplete Recirculation Actuation Signal (RAS) - A Simulator Case Study I
1 SMALL BREAK LOSS-OF-COOLANT ACCIDENT (SBLOCA) WITH INCOMPLETE 1
RECIRCULATION ACTUATION S:GNAL (RAS)- A SIMULATOR CASE STUDY PURPOSE To validate assumptions made with regards to operator actions on a SBLOCA with incomplete RAS, a simulator scenario was prepared to simulate the conditions that existed on Unit 2 on January 13-14,1998 (i.e., inoperable Train A containment emergency sump outlet valve,2HV9305, and a Train B Salt Water Cooling Heat Exchanger tube leak repair outage).
1 The scenario was run on April 2,1998, with two crew groups who were in their third day of simulator training. One crew group performed the scenario in the morning and one in the afternoon. Neither crew group was aware this day's training had any unique purpose, even after the scenario was run. Based on feedback after completion of the j
scenario, both crews felt this scenario was new and quite interesting.
The simulator setup was as follows: Train B Saltwater Cooling (SWC) Heat Exchanger 2E002 cleared for tube leak repair, Train B Emergency Core Cooling System (ECCS) pumps had the direct current (DC) control power open (to preclude inadvertent start when SWC was not available). High Pressure Safety injection (HPSI) pump 2P018 and Component Cooling Water (CCW) pump 2P025 were both aligned to Train A. The Unit was simulated to be at full power.
TIMELINE LAYOUT The timelines that follow are shown with Time =0 as time of RAS actuation. Times are in minutes. The first timeline is Crew 1 and the second timeline is Crew 2.
SUMMARY OF ACTIONS PRIOR TO RAS The scenario was initiated by triggering a SBLOCA, causing multiple alarms related to loss of inventory and radiation release in containment. Both crews manually tripped the reactor before an automatic trip setpoint was reached. Shortly after the manual trip an automatic Safety injection Actuation Signal (SlAS)/ Containment Cooling Actuation Signal (CCAS) and Containment Isolation Actuation Signal (CIAS) occurred (in addition to the usual Emergency Feedwater Actuation Signal [EFAS] 1& 2 actuations). Both crews performed the Standard Post-Trip Actions (SPTAs) although, due to having only one train of ECCS, both crew Shift Supervisors (SS's) directed manually starting the second HPSI pump (2P017) to maximize safety injection flow (this is in accordance with the LOCA Emergency Operating Instruction (EOI)).
Both crews promptly completed the SPTA's, diagnosed a LOCA in containment, and transitioned to the LOCA EOl, SO23-12-3. In addition, both crew SS's declared a Site Area Emergency about 10 minutes after the reactor trip, per EPIP SO123-Vill-1, 1
" Recognition and Classification of Emergencies," Table B3. The rest of Emergency Plan response was simulated for this scenario. The Nuclear Operations Assistant (NOA) would at this time initiate offsite notification and recall. The break size was sufficient to depressurize the RCS to about 900 psia initially, and caused the RCS loops to cool below Steam Generator saturation temperature quite rapidly.
Containmerd Spray was actuated automatically about 10 minutes after the trip. One crew SS dispatched operators to turn control power back on for CCW pump 2P026, and HPSI pump 2P019 to make them available. (This action was not well thought out as CCW pump 2P026 auto-started and the CCW Hi Temp alarm immediately came in due to the Emergency Cooling Units (ECUS) pumping heat back into the CCW system).
The SS then directed CCW pump 2P026 be stopped, and rescinded his direction to turn control power on for HPSI 2P019.
At this point, both crews had Train A containment spray, plus HPSI 2P017 and 2P018 running, feeding from the Refueling Water Storage Tanks (RWSTr). Approximately 25 minutes into the event, the Safety Injection Tanks (SITS) began to inject. After most actions were completed, the crews took a break and debriefed their actions up to this point (this allowed the simulator to continue to run and lower RWST level to closer to the RAS setpoint at a time when there was little operator action required). After the critique, the crew took over again with RWST level about 30%, and SIT's empty. The containment pressure had peaked at about 15 psi and was now trending down to about 11 psi (Iow enough to be secured per EOl procedure). One crew considered securing Containment Spray prior to RAS, but decided against it.
TIMELINE OF ACTIONS AFTER RAS Note: The simulator could not model 2HV9305 failure, so a redundant Train A isolation valve 2HV9303 was modeled to not open.
CREW #1 T=0 RAS Train A and B actuation occurred. Control Board operator directed to verify proper RAS actuation, and identified 2HV9303 Emergency Sump outlet did not open. Crew attempted to manually open 2HV9303 (from Control Room (CR)).
T=1 Crew recognizes 2HV9303 is outside containment and requests operator be dispatched to locally open the valve [ simulator instructor did not allow this success path, as it would not work for 2HV9305)
T=3 Verified still had a flowpath from the RWST for HPSI 2P017/018, and elected to continue to run all Train A ECCS pumps (did not consider Containment Spray termination]. Crew problem solving began using P&lD's and simplified drawings.
2
3 T=12 Crew recognizes that HPSI 2P018 can have its suction aligned to Train B Emergency Sump and still remain aligned to Train A for CCW pump / motor cooling.
T=16 SS calls Operations Support Center (OSC) to dispatch operator to open HPSI suction Train A/ Train B cross-tie, S21204MUO11 (after reviewing flowpath on prints with RO).
T=21 HPSI throttle-stop criteria met and direction given to throttle HPSI flow.
T=23 Crew noted loss of flow on Containment Spray pump 2P012, gave direction to override SIAS and stop 2P012.
T=26 Crew noted loss of flow on HPSI 2P017 and 2P018, gave direction to override i
SIAS and stop 2P017 and 2P018. Crew now recognizes there is only 132 gpm
(
charging flow from Boric Acid Makeup (BAMU) tanks.
j T=29 ARO prompts starting HPSI 2P019 after isolating ECUS and restarting CCW 2P026. SS concurs and action is initiated.
T=33 Crew restarts HPSI 2P018 after receiving word that MUO11 is open. Verifies proper HPSI flow. Abandons effort to turn control power back on HPSI 2P019.
T=35 Crew discusses long term need for cooling, need to get SWC B back or 2HV9303 open to allow Containment Spray to cool emergency sump.
End of Crew #1 Scenario Critique CREW #2 T=0 RAS Train A and B actuation occurred. Board operator directed to verify proper RAS actuation and identifies 2HV-9303 did not open. Crew attempted to manually (from CR) open HV-9303.
T=1 Problem solving starts. CO recommends stopping Containment Spray. Shift Technical Advisor (STA) recommends stopping one HPSI pump to extend RWST drawdown time. SS sends operator to breaker for 2HV9303 to investi(late and cycle. SS/ Control Room Supervisor (CRS)/STA agree on plan to stop Containment Spray and one HPSI pump.
T=5 Crew overrode and stopped Containment Spray, after verifying all EOl termination criteria met. Crew overrode and stopped HPSI 2P018 and verified Safety injection flow still satisfactory. Team dispatched to locally open 2HV9303.
i 3
o
l T=9 ARO recommends opening HPSI 2P018 suction cross-tie, S21204MUO11, while pointing to simplified ECCS drawing on CR-57. SS/STA/CRS discuss and agree to leave 2P017 running, shut 2S21204MUO10 and open S21204MUO11 l
(isolating 2P017 to Train A and 2P018 to Train B suction).
T=11 Operator dispatched to close 2S21204MUO10 and open 2S21204MUO11. SS gets Technical Support Center (TSC) concurrence with plan, and declares and logs 50.54X entry.
T=29 HPSI 2P018 started on Train B suction, crew verified proper Sl flow and then stopped 2P017. RWST level still at about 14%.
T=35 Crew discusses long term cooling plan, Shutdown Cooling (SDC) entry criteria met, so plan is to align LPSI 2P015 to SDC for cooling and ccntinue to use HPSI 2P018 suction from emergency sump to keep RCS inventory up (combination of RAS and SDC). This combination does not require E002 RTS or 2HV9303 opened.
End of Crew #2 Scenario Critique SUMMARY Both crews immediately recognized an improper RAS actuation, and promptly came to same solution to align HPSI 2P018 suction to Train B sump and leave the rest of HPSI 2P018 aligned to Train A. Fifteen minutes was allotted for travel and alignment of the valve, resulting in restoration of tne Safety injection flow in 33,and 29 minutes respectfully. Neither crew pursued the maintenance repair aspects of 2HV9303 or 2E002 RTS. In non-EPIP training scenarios, it is the crews problem-solving skills alone that are being evaluated, and they tend to pursue options they have control of first. In full-blown EPIP scenarios, the SS will promptly call for restoration of any OOS equipment at the onset of the accident (drill), and immediately refer any failures during the scenario to the TSC and OSC for support. It is believed that in a real event, the crew would aggressively pursue restoration of OOS equipment anytime an accident occurred.
l 4
ATTACHMENT 4 Probabilistic Risk Assessment - 2HV9305 Linestarter Failure l
l l
1 i
t Attachnient 4 Page 1 of 12 PROBABILISTIC RISK ASSESSMENT 2HV9305 Linestader Failure PURPOSE The purpose of this study is to determine the increase in core damage and large early release risk from inoperability of valve 2HV9305 from January 6,1998 to January 24, 1998.
BACKGROUND LER-98-003 documents an event where valve 2HV9305 was determined to be inoperable from 1142 hours0.0132 days <br />0.317 hours <br />0.00189 weeks <br />4.34531e-4 months <br /> on January 6,1998 to 1133 hours0.0131 days <br />0.315 hours <br />0.00187 weeks <br />4.311065e-4 months <br /> on January 24,1998.
The valve was inoperable due to a failure of the mechanical interlock on valve's reversing linestarter.
Valve 2HV9305 is located inside containment in the containment sump post-LOCA recirculation flow path (see Figure 1). This valve is normally closed and receives a signal to open on a recirculation actuation signal (RAS). Failure of the valve to open would have prevented the success of the Train A post-LOCA recirculation flow path.
The San Onofre Living PRA model assumes that post-LOCA recirculation flow is required in small (3/8" to 2" diameter breaks), medium (2" to 6" diameter breaks), and large (6" diameter to double-ended break) LOCAs. The bases for this assumption are plant specific MAAP analyses which assume the most likely plant response (e.g., both trains of ECCS injecting) and worst case break location.
1 METHODOLOGY The increase in core damage and large early release risk from inoperability of valve j
2HV9305 was evaluated using the San Onofre Living PRA model contained in the Safety Monitor. Actual plant configurations during the period January 6,1998 to January 24,1998 including the unavailability and alignment of key plant components impodant to risk were considered in the assessment. During the period January 6, 1998 to January 24,1998, three events contributed significantly to the risk increase from inoperability of 2HV9305: Train B CCW heat exchanger E002 out of service for a tube repair for 27.08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> on 1/13/98, heat treatment of the Train B SWC/CCW for 6.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> on 1/16/98 and 7.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> on 1/24/98.
The Safety Monitor risk model contains the following conservatisms which impact the assessment of the risk increase from the 2HV9305 inoperability:
1
1 Page 2 of 12 1.
Recovery from the heat treat events prior to core damage is not modeled.
2.
Recovery of 2HV9305 prior to core damage is not modeled.
3.
Recovery of CCW heat exchanger 2E002 prior to core damage is not modeled.
4.
Cross-tie of post-LOCA HPGl suction injection paths prior to core damage is not modeled.
These recovery actions were not modeled either because engineer; ig analysis to support their success were not available or the actions required to perform the recoveries were not described in written plant procedures. Based on guidance in NUREG-1335, " Individual Plant Examination: Submittal Guidance", "The [NRC) staff, however, expects that all assumed or modeled recovery actions will have written l
procedures. Most often the [NRC] staff has received justifications for the assumptions of success for non-proceduralized actions based solely on time available for such actions. The [NRC) staff does not believe this type of argument to be correct."
However, since these recovery actions are likely and at least one was utilized in two simulator exercises modeling this event, a sensitivity analysis was performed to determine the risk impact of success of each and combinations of these recovery actions. The failure probability for each recovery action is calculated assuming the i
action was proceduralized or within the skill of plant personnel. This best estimate 9.ilure probabilities are then determined based on engineering judgement and available plant data including simulator runs. Given that different plant personnel would be performing the recoveries after event diagnosis by control room operators, all the actions are assumed to be independent and will not individually impact the likelihood of other recovery actions.
ANALYSIS The actual plant configurations during the period January 6,1998 to January 24,1998 were analyzed in the Safety Monitor for two cases: (1) assuming availability of 2HV9305 and (2) assuming unavailability of 2HV9305. The difference in the risk results between the two cases represents the increase in risk due to unavailability of 2HV9305.
The Safety Monitor model was then modified to include the four recovery actions identified above. Since these recovery actions are not proceduralized, sensitivity analyses were performed for each and combinations of these recovery actions to determine the risk impact of the recoveries on plant risk.
The failure probability of each recovery action was determined using the ASEP human reliability analysis method documented in NUREGICR-4772 (Ref.1). It was assumed for the purposes of this analysis that the recovery actions were proceduralized or with!n Page 3 of 12 I
the skill of the plant operating staff. The results of the human reliability analysis are provided in Table 1.
Plant management reviewed the human reliability analysis for each recovery action and selected a more conservative value to account for some or all the action not being proceduralized (see Table 2). Also, simulator runs modeling the failure of 2HV9305 and unavailability of CCW heat exchanger 2E002 provided additional information on timing and the likely recovery action which the operators would select (a summary of the simulator runs are provided in another attachment).
Since these recovery actions have varying likelihoods of success depending on the event timing, the recovery actions were credited only in small LOCAs (3/8" to 2" diameter LOCAs). Large and medium LOCAs were conservatively assumed to lead to core damage prior to the success of any of the recovery actions based on the limited time available for recovery from deterministic thermal-hydraulic analyses.
4 The recovery actions were not credited for time periods other than those where the Train B post-LOCA recirculation train redundant to 2HV9305 was known to be unavailab% (i.e, during the repair of CCW heat exchanger 2E002). The recovery of random failures in other time periods is very difficult and time-consuming since some of the random failures in the other time periods cannot be recovered by these recover) actions. This represents a significant conservatism in the analysis.
ASSUMPTIONS The following key assumptions were used in the risk assessment:
1.
Based on input from operations, recovery from the CCW/SWC heat treatments prior to core damage was assumed. The time available to recover the inoperable CCW/SWC train in heat treat prior to core damage for a small LOCA is approximately 250 minutes. The operators would be expected to exit the heat treat configuration soon after the LOCA initiation and restore CCW/SWC capability within 20 minutes. Based on the large margin between the time available and time required, this recovery action was assumed to occur.
When heat treating a SWC train the Inop flag alarms are actuated by procedure
. as a reminder of high temp on the SWC system. Usually the CCW train is off so as to not heat up the entire CCW system to 103*F. Upon the SIAS the CCW pump would start and actuate the CCW Hi Temp alarm, a further prompt to take i
action.
l
Atta6 ment 4 Pagt ' of 12 For the unit under heat treat, the LOCA will cause an auto reactor and turbine trip. If there is not a loss of all offsite power, the loss of turbine heat and the circ flow will rapidly cool the intake back down with no operator action required.
For the heat treat on the other unit, the above alarms would prompt the crew to secure the heat treat on the other unit and initiate swapping the SWC pump -
breaker to the opposite unit intake. To do this the operators would first need to secure the ECCS pumps and CCW train (override and stop from CR). Then overide and stop the SWC pump, go to 50' electrical, rack out one breaker, rack in the other, start the SWC pump, and restart the CCW pump and the ECCS pumps.
2.
The recovery actions considered are assumed to be independent and do not individually impact the likelihood of other recovery actions. The plant personnel performing each of recovery action would be different, thus concems about staff manning and miscommunication are not applicable. The diagnosis error rate for each of these recovery actions is negligible compared to the action error rate for these actions. Therefore, it can be conservatively assumed that the control room staff performs the diagnosis for each of these recovery actions sequentially. Once the diagnosis occurs and plant personnel are implementing the actions, the actions are assumed to occur in parallel since each involves a different organi7.ation (i.e., maintenance mechanics would restore the CCW heat exchanger, maintenance electricians would troubleshoot and open 2HV9305, and an auxiliary operator would realign the HPSI suction flow path).
RESULTS The increase in core damage risk from inoperability of valve 2HV9305 from January 6, 1998 to January 24,1998 without crediting any recovery actions is estimated to be 2E-5. Crediting each of the recovery actions independent of the others results in the core damage risks graphs provided in Figures 2,3, and 4. When all the recovery actions are credited with the best estimate values selected in Table 2, the core damage risk increase is estimated to be 6E.6. The likelihood that at least one recovery action will succeed is 96%.
The increase in large early release risk from inoperability of valve 2HV9305 was estimated to be 3E-8 assuming recovery of the heat treat and 3.3E-8 assuming no recovery of the heat treat events. However, based on the large redundancy in containment cooling systems (4 emergency f6.n coolers as well 2 trains of containment spray), this event would not have significantly impacted containment cooling capability.
Any one emergency containment fan cooler or train of containment spray would have been capable of preventing containment failure based on MAAP analyses performed for the IPE. The inoperability of 2HV9305 would only have impacted the availability of
i l
l Page 5 of 12 a single containment spray train. Therefore the increase in large early release risk from this event is assumed negligible.
CONCLUSIONS j
l The increase in core damage release risk from inoperability of valvi,2HV9305 from January 6,1998 to January 24,1998 was small. The increase in large early release risk from inoperability of valve 2HV9305 was insignificant.
REFERENCES l
1.
Swain, A.," Accident Sequence Evaluation Program Human Reliability Analysis Procedure", NUREGICR-4772, February 1987.
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ATTACHMENT 5 Failure Analysis Report 98-005, Failure Analysis of the 2HV9305 Motor Starter 0