NOC-AE-11002703, Supplement to the License Amendment Request for Approval of a Revision to the South Texas Fire Protection Program Related to the Alternative Shutdown Capability

Supplement to the License Amendment Request for Approval of a Revision to the South Texas Fire Protection Program Related to the Alternative Shutdown Capability

ML11221A230
Person / Time
Site:	South Texas
Issue date:	08/01/2011
From:	Bowman C South Texas
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NOC-AE-11002703, STI: 32903186, TAC ME6346, TAC ME6347
Download: ML11221A230 (24)

Text

Nuclear Operating Company South Texas Pro/ectElectric GeneratingStation PC. Box 289 Wadsworth. Texas 77483 V August 1,2011 NOC-AE-1 1002703 G25 10CFR50.90 10CFR50.48 STI: 32903186 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units 1 & 2 Docket Nos. STN 50-498, STN 50-499 Supplement to the License Amendment Request for Approval of a Revision to the South Texas Project Fire Protection Program Related to the Alternative Shutdown Capability (TAC Nos. ME6346 and ME6347)

Reference:

1. Letter from G. T. Powell, STPNOC, to NRC Document Control Desk, "License Amendment for Approval of a Revision to the South Texas Project Fire Protection Program Related to the Alternate Shutdown Capability," dated June 2, 2011 (NOC-AE- 11002643) (ML11161A143)

2. Letter from Balwant K. Singal to Edward D. Halpin, STPNOC, "South Texas Project, Units 1 and 2 -License Amendment Request for Approval of a Revision to the Fire Protection Program Related to the Alternate Shutdown Capability unacceptable for Review with opportunity to supplement," dated July 22, 2011 (AE-NOC- 11002123) (ML112010150)

In reference 1, STP Nuclear Operating Company (STPNOC) submitted a licensee amendment request (LAR) for approval of a revision to the South Texas Project (STP) Fire Protection Program (FPP) related to the Alternative Shutdown Capability.

Per Reference 2 of this letter, The Nuclear Regulatory Commission (NRC) notified STPNOC that additional information is necessary to enable the NRC staff to make an independent assessment regarding the acceptability of the proposed amendment request in terms of regulatory requirements and the protection of public health and safety and the environment.

NOC-AE-1 1002703 Page 2 The enclosure to this letter provides supplemental information to the LAR to support the NRC staff request.

The supplemental information does not impact the No Significant Hazards Determination provided in the LAR There are no regulatory commitments in this letter.

If there are any questions regarding this amendment request, please contact Ken Taplett at (361) 972-8416 or me at (361) 972-7566.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on / *1/

'-Date 62400 4 -

ý2lf Charles T. Bowman General Manager, Nuclear Safety Assurance

Enclosure:

Supplemental Information to the LAR for approval of a revision to the South Texas Project (STP) Fire Protection Program (FPP) related to the Alternate Shutdown Capability

NOC-AE-1 1002703 Page 3 cc:

(paper copy) (electronic copy)

Regional Administrator, Region IV A. H. Gutterman, Esquire U. S. Nuclear Regulatory Commission Morgan, Lewis & Bockius LLP 612 East Lamar Blvd, Suite 400 Arlington, Texas 76011-4125 Balwant K. Singal U. S. Nuclear Regulatory Commission Balwant K. Singal John Ragan Senior Project Manager Catherine Callaway U.S. Nuclear Regulatory Commission Jim von Suskil One White Flint North (MS 8B1) NRG South Texas LP 11555 Rockville Pike Rockville, MD 20852 Ed Alarcon Senior Resident Inspector Kevin Pollo U. S. Nuclear Regulatory Commission Richard Pena P. 0. Box 289, Mail Code: MNI 16 City Public Service Wadsworth, TX 77483 C. M; Canady Peter Nemeth City of Austin Crain Caton & James, P.C.

Electric Utility Department 721 Barton Springs Road C. Mele Austin, TX 78704 City of Austin Richard A. Ratliff Texas Department of State Health Services Alice Rogers Texas Department of State Health Services

Enclosure NOC-AE-1 1002703 Page 1 of 21 Supplemental Information to the LAR for Approval of a Revision to the South Texas Project (STP)

Fire Protection Program (FPP) related to the Alternative Shutdown Capability

References:

1. Letter from G. T. Powell, STPNOC, to NRC Document Control Desk, "License Amendment for Approval of a Revision to the South Texas Project Fire Protection Program Related to the Alternative Shutdown Capability," dated June 2, 2011 (NOC-AE-1 1002643) (ML11161A143)

2. WCAP 11397-P-A, "Revised Thermal Design Procedure," April 1989

3. Letter from Balwant K. Singal to Edward D. Halpin, STPNOC, "South Texas Project, Units 1 and 2 -License Amendment Request for Approval of a Revision to the Fire Protection Program Related to the Alternate Shutdown Capability unacceptable for Review with opportunity to supplement (TAC NOS. ME6346 AND ME6347)," dated July 22, 2011 (AE-NOC-1 1002123) (ML112010150)

NRC Request for Additional Information( Ref. 3)

Please provide information to supplement the license amendment request in order to facilitate a comparison between the RETRAN analyses performed to support the Appendix R licensing request and a typical RETRAN analysis performed for a licensing basis event. Include information to compare key analytic inputs, and a summary description of the RETRAN results for the Appendix R analyses, including a discussion of the key figures of merit, limiting cases and results.

STPNOC Response The fire hazard analysis to support the four selected operator actions discussed in Section 3.2 of the enclosure to the referenced license amendment request were based on the RETRAN computer code. The four actions are:

1. Close both pressurizer power-operated relief valve (PORV) block valves 2: Place centrifugal charging pumps in PULL-TO-LOCK (PTL)

3. Initiate feedwater (FW) isolation

4. Place startup feedwater pump (SUFP) in PTL A comparison between the RETRAN analysis performed to support the fire hazards analysis licensing request and a typical RETRAN analysis performed for a licensing basis event is provided below for each of the above actions.

Enclosure NOC-AE- 11002703 Page 2 of 21

1. Close Both Pressurizer Power-Operated PORV Block Valves The RETRAN analysis performed to support the fire hazards analysis licensing request for evaluating the requirement to close both pressurizer PORV block valves assumed a fire-induced spurious opening of one pressurizer PORV coincident with the operator initiated reactor trip when leaving the control room. The purpose of closing the PORV block valves is to protect against an uncontrolled depressurization of and loss of Reactor Coolant System (RCS) inventory resulting in an Engineered Safety Features (ESF) actuation signal based on low pressurizer pressure. The acceptance criteria for the fire hazard analysis was that an ESF safety actuation signal is considered a process variable outside what would be predicted for a normal loss of a-c power event as required by Appendix R,Section III.L. 1.

The RETRAN analysis assumes the pressurizer PORV spuriously opens when the operator trips the reactor. The pressurizer PORV is assumed to stay open for 5 minutes before the block valve is closed. The analysis shows that the operator has approximately 61 seconds before a safety injection (SI) signal would be expected due to a spurious opening of one pressurizer PORV. A time sequence of events for this analysis is provided on Table 1-1. A plot of reactor power, pressurizer pressure, hot leg temperature and cold leg temperature versus time is provided on Figures 1A-D. As stated in Reference 1, operators have demonstrated that the block valves can be closed before leaving the control room in time to preclude an uncontrolled depressurization of and loss of RCS inventory.

An equivalent typical RETRAN analysis performed for a licensing basis event analysis is provided in Section 15.6.1 of the Updated Final Safety Analysis Report (UFSAR),

"Inadvertent Opening of a Pressurizer Safety or Relief Valve." For this event, the reactor is assumed to be at full power at which time a pressurizer PORV or safety fails open. The analysis in the UFSAR assumes a failure of a pressurizer safety valve. The transient continues until a reactor trip occurs terminating the decrease in Departure fromNucleate Boiling (DNBR) margin. The acceptance criteria for this event described in the UFSAR is that the DNBR remains above the safety analysis limit throughout the transient. The DNBR analysis uses the Revised Thermal Design Procedure methodology described in Section 4.4.1.1 of the UFSAR and in WCAP 11397-P-A (Reference 2). A time sequence of events is provided on Table 15.6-1 of the UFSAR. A plot of nuclear power, pressurizer pressure and core average temperature versus time is provided on Figures 15.6-1 and 15.6-2 of the UFSAR.

Table 1-2 provides a comparison of the key analytical inputs for the RETRAN analysis performed to support the fire hazards analysis licensing request and the equivalent typical RETRAN analysis performed for a licensing basis event described in the UFSAR.

Enclosure NOC-AE-1 1002703 Page 3 of 21

2. Place Centrifugal Charging Pumps in PTL The RETRAN analysis performed to support the fire hazards analysis licensing request for evaluating the requirement to place the centrifugal charging pumps in PTL assumed a fire-induced spurious opening of the pressurizer auxiliary spray valve coincident with the operator initiated reactor trip when leaving the control room. The purpose of placing the centrifugal charging pumps in PTL was to preclude depressurizing the RCS which could result in an ESF actuation signal and to protect against the indicated pressurizer water level going off-scale high. The acceptance criteria for the fire hazard analysis is that an ESF safety actuation signal and the indicated pressurizer water level going off-scale high is considered a process variable outside what would be predicted for a normal loss of a-c power as required by Appendix R,Section III.L. 1 and III.L.2.b respectively.

The RETRAN analysis assumes the auxiliary pressurizer spray valve spuriously opens when the operator trips the reactor. The analysis conservatively assumes an initial flow of 150 gallons per minute (21 lbm/sec) of flow from the centrifugal charging pumps through the open auxiliary pressurizer spray valve into the pressurizer which increases as pressurizer pressure decreases. A turbine trip and FW isolation is also assumed to occur at this time.

The Reactor Coolant Pumps (RCPs) are assumed to trip two seconds later. The analysis also conservatively assumes auxiliary feedwater (AFW) starts on a steam generator (SG) low-low signal at 19 seconds to maximize the RCS depressurization. The operators are then assumed to secure auxiliary spray 600 seconds (10 minutes) after the initiation of the event. The results show that SI actuation occurs at 146 seconds, and the indicated pressurizer water level goes off-scale high at 465 seconds. A time sequence of events for this analysis is provided on Table 2-1. A plot of pressurizer pressure, indicated pressurizer water level, hot leg temperature, cold leg temperature, pressurizer auxiliary spray flow and SI flow versus time is provided on Figures 2A-F. As stated in Reference 1, operators have demonstrated that the centrifugal charging pumps can be placed in PTL before leaving the control room in time to preclude an uncontrolled RCS depressurization and the indicated pressurizer water level going off-scale high.

The centrifugal charging pumps are also required to achieve and maintain cold shutdown conditions as required by Appendix R, Sections III.2.a and III.L.5. Placing the centrifugal charging pumps in PTL ensures that a fire-induced spurious closure of the Volume Control Tank isolation valve will not result in damage to these pumps. A RETRAN analysis was not performed to address this requirement.

An equivalent typical RETRAN analysis performed for a licensing basis event analysis is provided in Section 15.6.1 of the Updated Final Safety Analysis Report (UFSAR),

"Inadvertent Opening of a Pressurizer Safety or Relief Valve." Table 1-2 provides a comparison of the key analytical inputs for the RETRAN analysis performed to support the fire hazards analysis licensing request and the equivalent typical RETRAN analysis performed for a licensing basis event described in the UFSAR.

Enclosure NOC-AE- 11002703 Page 4 of 21

3. Initiate FW Isolation The RETRAN analysis performed to support the fire hazards analysis licensing request for evaluating the requirement to initiate FW isolation assumed a fire-induced spurious starting of the SUFP coincident with the operator initiated reactor trip when leaving the control room.

The purpose of closing the FW isolation valves is to protect against the indicated pressurizer water level going off-scale low which would be a violation of Appendix R,Section III.L.2.b.

In addition, the SG water level would go off scale high due to overfilling the SGs. Overfilling the SGs is considered a process variable outside what would be predicted for a normal loss of a-c power which is a violation of Appendix R,Section III.L. 1.

The RETRAN analysis assumes the operators trip the reactor followed by a turbine trip 3.5 seconds later. The operators then close the main steam isolation valves 35 seconds after the reactor trip which trips the main FW pumps. With off-site power available, the startup FW pump starts in 50 seconds. The analysis also conservatively assumes AFW starts at 50 seconds to maximize the RCS cooldown and exacerbate the SG overfill. The analysis assumes the operators trip the RCPs 120 seconds after the reactor trip. The SUFP pump and AFW is secured 540 seconds (9 minutes) after the initiation of the event. The results of the analysis show that the indicated pressurizer water level goes off-scale low in 285 seconds and the SG water level goes off-scale high in 295 seconds in all four SGs. A time sequence of events is provided on Table 3-1. A plot of hot leg temperature, cold leg temperature, pressurizer pressure, indicated pressurizer water level and indicated steam generator water level versus time is provided on Figures 3A-E. As stated in Reference 1, operators have demonstrated that the main feedwater isolation valve can be closed before leaving the control room in time to preclude the indicated pressurizer water level going off-scale low or the indicated steam generator water level going off-scale high.

An equivalent typical RETRAN analysis performed for a licensing basis event analysis is described in Section 15.1.2 of the UFSAR, "Feedwater System Malfunctions Causing an Increase in Feedwater Flow." For this event, the following conditions are evaluated:

1. Accidental opening of one feedwater (FW) control valve at hot zero power conditions resulting in an increase of 225% of nominal flow to one steam generator (SG) with conservatively low FW temperature of 70' F.

2. Accidental opening of all four FW control valves at hot zero power conditions. One turbine driven feed pump is at run-out flow with a conservatively low FW temperature of 700 F.

3. Accidental opening of one FW control valve at hot full power conditions resulting in a step increase to 200% of nominal FW flow to one SG. The rod control system is either in automatic or manual mode.

Enclosure NOC-AE-1 1002703 Page 5 of 21

4. Accidental opening of all FW control valves at hot full power conditions resulting in a step increase to 150% of nominal FW flow to all SGs. A 25 British Thermal Units per pound mass (Btu/lbm) FW enthalpy reduction is assumed. The rod control system is in manual mode.

The UFSAR further states that the zero power events are bounded by the full power cases and therefore not presented.

For the analysis presented in the UFSAR, a failure of one FW control valve is assumed resulting in increased FW flow to one steam generator. When the SG water level in the faulted loop reaches the high-high level setpoint, all FW isolation valves and FW control valves are automatically closed and the SG feed pumps are tripped. This prevents continued addition of FW. In addition, a turbine trip is initiated. Following turbine trip, the reactor will automatically trip due to turbine trip. The acceptance criterion for this event is that the DNBR remains above the safety analysis limit throughout the transient. The DNBR analysis uses the Revised Thermal Design Procedure methodology. A second but less limiting criteria discussed in the UFSAR is that fuel temperature remains below the fuel melting temperature. The analysis demonstrates this criterion is satisfied since the peak power does not exceed 118% power. A time sequence of events is provided on Table 15.1-1 of the UFSAR. A plot of nuclear power, DNBR, vessel average temperature, and pressurizer pressure temperature versus time is provided on Figures 15.1-1 and 15.1-2 of the UFSAR.

Table 3-2 provides a comparison of the key analytical inputs for the RETRAN analysis performed to support the fire hazards analysis licensing request and the equivalent typical RETRAN analysis performed for a licensing basis event described in the UFSAR.

4. Place SUFP in PTL The RETRAN analysis performed to support the fire hazards analysis licensing request for evaluating the requirement to place the SUFP in PTL assumed a fire-induced spurious opening of a main feedwater isolation valve coincident with the operator initiated reactor trip when leaving the control room. The purpose of SUFP in PTL is to protect against the indicated pressurizer water level going off-scale low which would be a violation of Appendix R,Section III.L.2.b. In addition, the SG water level would go off scale high due to overfilling the SG. Overfilling the SG is considered a process variable outside what would be predicted for a normal loss of a-c power, which is a violation of Appendix R,Section III.L. 1.

The RETRAN analysis assumes the operator trips the reactor followed by an automatic turbine trip 3.5 seconds later. The operators then close the main steam isolation valves 35 seconds after the reactor trip, which trips the main feedwater pumps. The main feedwater isolation valves to loops 2, 3, and 4 close at this time. The main feedwater isolation valve to loop 1 does not close due to a fire-induced spurious signal. With off-site power available, the SUFP starts in 50 seconds after the reactor trip. The analysis also conservatively assumes AFW

Enclosure NOC-AE-1 1002703 Page 6 of 21 starts at 50 seconds after the reactor trip to maximize the RCS cooldown and exacerbate the SG overfill. The analysis assumes the operators trip the RCPs 120 seconds after the reactor trip. Results show that the indicated SG water level in loop 1 goes off scale high approximately 130 seconds after the reactor trip. The indicated pressurizer water level goes off-scale low 257 seconds after the reactor trip. A time sequence of events is provided on Table 4-1. A plot of hot leg temperature, cold leg temperature, pressurizer pressure, indicated pressurizer water level and indicated steam generator water level versus time is provided on Figures 4A-E. As stated in Reference 1, operators have demonstrated that the main feedwater isolation valve can be closed before leaving the control room in time to preclude the indicated pressurizer water level going off-scale low or the indicated SG water level going off-scale high.

The equivalent typical RETRAN analysis performed for a licensing basis event analysis is described in Section 15.1.2 of the UFSAR Section, "Feedwater System Malfunctions Causing an Increase in Feedwater Flow," which is discussed in Item 3 above. Table 3-2 provides a comparison of the key analytical inputs for the RETRAN analysis performed to support the fire hazards analysis licensing request and the equivalent typical RETRAN analysis performed for a licensing basis event described in the UFSAR.

Table 1-1 Time Sequence of Event For A Fire-Induced Spurious Opening of One Pressurizer PORV Coincident With Operator Initiated Reactor Trip Time (sec) Event 0 Reactor Trip 0 Turbine Trip 0 FW Isolation 0 MSIV Closure 0 Pressurizer PORV Opens 2 RCP Trip 61.5 SI Actuation 300 Operator Closes Pressurizer PORV Block Valves

Enclosure NOC-AE-1 1002703 Page 7 of 21 Table 1-2 Comparison of Key Analysis Inputs for a Fire-Induced Spurious Opening of (1) One Pressurizer PORV Coincident With Operator Initiated Reactor Trip, and (2) Pressurizer Auxiliary Spray Valve Coincident With Operator Initiated Reactor Trip PARAMETER UFSAR CHAPTER 15.6.1 FIRE HAZARD ANALYSIS ASSUMPTION ANALYSIS ASSUMPTION Initial Power Nominal full power plus instrument Nominal full power bias (3853 MWt)

Initial RCS Average High end of allowed operating band Actual nominal Temperature (593°F) plus instrument bias (5921F)

Initial Pressurizer pressure Nominal minus instrument bias Nominal (2250 psia)

RCS flow Best Estimate Flow Best Estimate Flow Moderator Temperature Most positive NA (reactor trip Coefficient occurs @ Time = 0)

Doppler Coefficient Small (absolute value) NA (reactor trip occurs @ Time = 0)

Table 2-1 Time Sequence of Event For A Fire-Induced Spurious Opening of Pressurizer Auxiliary Spray Valve Coincident With Operator Initiated Reactor Trip Time (sec) Event 0 Reactor Trip 0 Turbine Trip 0 FW Isolation 0 Pressurizer Auxiliary Spray activated 2 RCP Trip 19 Low Low SG Level 21 AFW Actuation 146 SI Actuation 465 Pressurizer Level > 100%

600 Operator Action to Terminate Auxiliary Spray

Enclosure NOC-AE-1 1002703 Page 8 of 21 Table 3-1 Time Sequence of Events For a Startup Feedwater Pump Spurious Actuation Scenario with Feedwater Isolation Valves Open (Main Feedwater Isolation at 9 Minutes After Reactor Trip)

Time (Seconds) Event Notes 0 Reactor Trip 3.5 Turbine Trip 35 MSIVs Closed Operator Action (includes 5 seconds for MSIV closure) 35 Main Feedwater Pumps Trip 50 SUFP Starts 50 AFW Starts 120 RCPs Trip Operator Action 285 Pressurizer Water Level @ 0%

295 SG Water Level @ 100% NRS (narrow range span) 540 AFW Flow Terminates Operator Action 540 Feedwater Isolated Operator Action 1315 Pressurizer Water Level > 0%

2980 Max. SG Water Volume 7401 ft3 Notes:

1. AFW and SUFP stop 9 minutes after reactor trip.

ft3

2. Maximum SG water volume is -7500

Enclosure NOC-AE- 11002703 Page 9 of 21 Table 3-2 Comparison of Key Analysis Inputs for a Startup Feedwater Pump Spurious Actuation Scenario With Feedwater Isolation Valves Open PARAMETER UFSAR CHAPTER 15 FIRE HAZARD ASSUMPTION ANALYSIS ASSUMPTION Initial Power Nominal full power plus instrument Nominal full power bias (3853 MWt)

Initial RCS Average High end of allowed operating band Actual nominal Temperature (593°F) plus instrument bias (592°F)

Initial Pressurizer pressure Nominal minus instrument bias Nominal (2250 psia)

RCS flow Best Estimate Flow Best Estimate Flow Reactivity Feedback Maximum EOL NA (reactor trip occurs @ Time = 0)

FW Temperature Bottom of allowed operating band 440'F prior to (390°F) reactor trip 370'F after reactor trip

Enclosure NOC-AE-l 1002703 Page 10 of 21 Table 4-1 Time Sequence of Events for a Spurious Opening of One Main Feedwater Isolation Valve Time (Seconds) Event Notes 0 Manual Reactor Trip Transient Starts 3.5 Turbine Trip 35 Manual MS Isolation All 4 MSIVs Closed Manual Operator Actions 35 Turbine Driven Main FW Pumps Trip 3 MFIVs Closed (Loop 1 Remains Open) 35 3 MFIVs Closed (Loop 1 Remains Open) 50 AFW Pumps start Flow to All SGs 50 Startup Feed Pump Starts Flow to Loop 1 Only 120 All RCPs Trip Manual Operator Actions 125 High-High SG Water Level SG 1 > 98.30%NR z130 SG Water Level @ 100% NRS 229 SG 1 Filled SG Water Solid 257 Pressurizer Water Level < 0%

Steamline Water Solid.

265 Steamline 1 Filled Liquid Release from I SG PORV

Enclosure NOC-AE-1 1002703 Page 11 of 21 Figure 1A Core Power vs. Time Spurious Opening of One Pressurizer PORV Coincident With Operator Initiated Reactor Trip I6 K

0 i U jo 2000 3000 40J00 50M0 Time (sec)

Figure 1B Pressurizer Pressure vs. Time Spurious Opening of One Pressurizer PORV Coincident With Operator Initiated Reactor Trip 2400

. 2200

• • 2000

, 1400 1200 Time

Enclosure NOC-AE- 11002703 Page 12 of 21 Figure 1C Hot Leg Temperature vs. Time Spurious Opening of One Pressurizer PORV Coincident With Operator Initiated Reactor Trip

.-- 630 620 F-,

600 50 5a0

-"570 Time (sec)

Figure 1D Cold Leg Temperature vs. Time Spurious Opening of One Pressurizer PORV Coincident With Operator Initiated Reactor Trip

.-.-- 580 5.50 540 520 500

-- 480 460 3o,300 Time (sec)

Loop 1 Loop 2 Loop 3 Loop 4

Enclosure NOC-AE-1 1002703 Page 13 of 21 Figure 2A Pressurizer Pressure vs. Time Spurious Opening of Aux Spray Valve With Charging Flow

,2400 2200 4,0

"-t 1500 1400 .

1200 ... .

0 1 20 2oo 00 4000 IV, Time (sec)

Figure 2B Indicated Pressurizer Water Level vs. Time Spurious Opening of Aux Spray Valve With Charging Flow 10

, 570

" C* 60 . ... .. ... ... .. .. . . . . . . . . . . . . . . . .

• 5)

Ct 50 - *. .. . .. ..

40

Enclosure NOC-AE-l 1002703 Page 14 of 21 Figure 2C Hot Leg Temperature vs. Time Spurious Opening of Aux Spray Valve With Charging Flow

... 630 620?

0I00 0 0o0O J000 I I ,00, 000 Trime (sec)

Figure 2D Cold Leg Temperature vs. Time Spurious Opening of Aux Spray Valve With Charging Flow 560570 58 X- . ..... ' '

S540 S520.

E .. .500 . . . . . ... .. . .. .

0 000 2000 3000 4000 5000 Time (sec)

Looa9 2 5- . -...

Lo. 3 .

LoL 4

Enclosure NOC-AE-1 1002703 Page 15 of 21 Figure 2E Pressurizer Auxiliary Spray Flow vs. Time Spurious Opening of Aux Spray Valve With Charging Flow 25 20 S15

~10 0

Time (sec)

Figure 2F Safety Injection Flow vs. Time Spurious Opening of Aux Spray Valve With Charging Flow J

<jOC E

a U

o *~

C) 2000 5000 Time (sec)

Lo0P 1 00oP 2 Lo0p 3) 1~.z 4

Enclosure NOC-AE-1 1002703 Page 16 of 21 Figure 3A Hot Leg Temperature vs. Time Main FWIV Open With Startup Feedwater Pump 640 -

w 620-

3 06 1 . .... . . . . . . . . . . . . . . . . . .

• 600-E 58 01 ... . . . . . . . . . . .. .. . .. . . .. .. . ..

o

- 56. -

542 . ,

0 500 1000 1500 2000 2500 Time (s)

Figure 3B Cold Leg Temperature vs. Time Main FWIY Open With Startup Feedwater Pump 560 - . . . .. . . . . . . . ..

C-E 540 . .. . . .. . . . .. . .. . .

0_ 520 - - . . . . .. . .. . .. . . . . . . . . . . . . . . . . ..

500 0 5o 1000 Time (s)1500 2000 2500 Loop 1 Loop 2 Loop 3

__ - Loop 4

Enclosure NOC-AE-l 1002703 Page 17 of 21 Figure 3C Pressurizer Pressure vs. Time Main FWIV Open With Startup Feedwater Pump 2400

" 2200

- 2000 Cf)

Cn Q_) 1800 1600 C-1400 1200 Figure 3D Indicated Pressurizer Water Level vs. Time Main FWIV Open With Startup Feedwater Pump 60

~-50 30 20 130 0

Enclosure NOC-AE-1 1002703 Page 18 of 21 Figure 3E Indicated Steam Generator Water Level vs. Time Main FWIV Open With Startup Feedwater Pump 100 -

S40. .................................................

Q) 24 0-- .. . . . . ...... . . . . . . .. . .

C 0-)

0 Time (s)

Enclosure NOC-AE-1 1002703 Page 19 of 21 Figure 4A Hot Leg Temperature vs. Time Spurious Opening of One Main Feedwater Isolation Valve 630 L- 620

' 610 S600 CL 590 E

a_.)

G.> 580

__ 570 o 560 550 Figure 4B Cold Leg Temperature vs. Time Spurious Opening of One Main Feedwater Isolation Valve 580 U-o 560 - 77 CD 540 - .. .. .. .. .. .. . . . . . . . . . . . . . . . . . . . .

E

,- 520-o5 500 - .............

480 1 I 1 2 50 100 150 200 250 i

300 Time (s)

Loop 1 Loop 2 Loop 3 Loop 4

Enclosure NOC-AE-1 1002703 Page 20 of 21 Figure 4C Pressurizer Pressure vs. Time Spurious Opening of One Main Feedwater Isolation Valve 2300 2200 ca-D 2100 73 U)

(- 2000 r)

"- 1900 73 C_)

A 1800 1700 150 300 Time (s)

Figure 4D Indicated Pressurizer Water Level vs. Time Spurious Opening of One Main Feedwater Isolation Valve 60

<L) 40 QL)

Cf)

S10 0

Time (s)

Enclosure NOC-AE-1 1002703 Page 21 of 21 Figure 4E Indicated Steam Generator Water Level vs. Time Spurious Opening of One Main Feedwater Isolation Valve 100 5 (I)

.__J 0,L)

- - - - - - L oop 3'*='

050 t00 150 200 250 300 Time (s)

Loop 1 Loop 2 Loop 3

____-Loop 4