ML20045D403

Forwards Results of Analyses Assuming Postulated Common Mode Failure of Sslc & Analyses Assuming Coincident Failure of Feedwater Control Sys

ML20045D403
Person / Time
Site:	05200001
Issue date:	06/18/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9306280337
Download: ML20045D403 (53)

Text

._.

Ah GE NucIcar Energy am nem cwm 175 Caff%r Awive. 5an Jaw CA 95125 i

June 18,1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule -

I & C Diversity issue, DFSER Open item 7.2.6-2

Dear Chet:

A bounding set of Chapter 15 events have been reanalyzed assuming a postulated common mode failure of the SSLC. These analyses were also done assuming the coincident failure of the feedwater control system. For the evaluation of this highly unlikely combination of breaks and failures, realistic LOCA evaluation models were applied to get a more accurate representation of the time available for operator action.

Due to a unique combination of analytical conditions for these analyses, these analyses were performed using the GE realistic LOCA evaluation model,-

TRACG. Since these analyses involve total core uncovery for an extended period of time, cooling of the fuel bundles by steam produced from boiling or flashing is greatly reduced. Also there is no spray cooling since ABWR does not have a core spray distribution sprager like previous BWRs. Therefore, for these cases radiation heat transfer is much more significant. Since the TRACG radiation heat transfer modelis more detailed than the SAFER model, it was decided that TRACG was the more appropriate LOCA evaluation model for this special application. Since the conditions identified above do not apply to the ~

standard LOCA analyses, evaluations of these cases using the SAFER LOCA evaluation model are still appropriate.

kldOUCG n

93o6280337 930618 PDR ADOCK 05200001

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-PDR

GE NucIcar Energy N N N A[U[ N m e cA 95 m The analyses are complete and the results are summarized in the enclosure.

.I The peak cladding temperatures for all cases were under the 2200 0F limit and

)

are listed below:

EVENT Maximum PCT 1

Steamline Break inside Containment 17110F Feedwater Line Break Inside Containment 13010F Shutdown Cooling Line Break Inside Containment 9530F.

HPCF Line Break inside Containment 9400F Bottom Drain Line Break inside Containment 6950F j

Turbine / Generator Trip with Bypass Valve Failure No Core Uncovery Further analyses of the above events were performed with the additional condition that the operator action was purposely delayed to increase the calculated peak cladding temperature to near the 2200 0F limit. The results of these additional analyses are also presented within the attached text.

This closes out the analysis portion of the I & C Diversity issue, DFSER Open Item 7.2.6-2 i

Sincerep,

}

Ys k Fox Advanced Reactor Programs cc:

Norman Fletcher (DOE)

Monty Ross (GE)

Frank Paradiso (GE)

Cal Tang (GE)

6/18/93 EVENT: STEAM LINE BREAK INSIDE CONTAINMENT This event is postulated to occur coincident with a undiscovered common mode failure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a steam line break inside containment, the main turbine and reactor feedpumps are tripped on a High Water Level 8 signal which is diverse from the SSLC. After the turbine is tripped, the reactor is scrammed from signals (diverse from EMUX) to the RPS from the turbine control system. It is postulated that because of EMUX common mode failure, the automatic isolation of the MSIVs is assumed to fail.

EOP ENTRY CONDITIONS:

The following alarms are provided by equipment independent of the EMUX. These are the entry conditions for emergency operating procedures expected for a LOCA inside the primary containment from instruments that are diverse from EMUX.

1. RPV WATER LEVEL LOW [ FIXED POsmON]

2. DRYWELL PRESSURE HIGH [ FIXED PosmON]

OPERATOR ACTIONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs developed from the RPV Control Guideline on High Drywell Pressure or RPV Water Level Low as an entry condition, and concurrently entering EOPs developed from the Primary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the following sets of actions are executed concurrently:

1.

RPV Control 1.

Initiate a manual scram if a scram has not been initiated.

2.

Initiate reactor isolation if it should have been isolated automatically but did not. (MSIV control is diverse from EMUX.)

3.

Restore and maintain RPV water level (water level signal is hardwired) above Level 3 using the CRD system, the condensate pumps and HPCF(B).

4.

If RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the active fuel.

5.

When RPV water level cannot be maintained above top of the active fuel, depressurize the reactor. (This action is not necessary as the reactor is depressurized through the break.)

11. Primary Containment Control:

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

2.

If necessary, initiate drywell spray using the fire protection system and the firewater addition mode of RHR(C ) for primary containment pressure control (drywell pressure signalis diverse from EMUX).

l CONTAINMENT RESPONSE The increase in suppression pool temperature due to vessel blowdown energy and decay heat is calculated to be approximately 70 'F over a time period of one hour. From this calculation, it is concluded that initiation of suppressior pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room.

The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure. Subsequent operator actions at the Remote Shutdown System willinclude initiation of the RHR suppression pool cooling function.

COMPARISION TO CHAPTER 6 LOCA ANALYSIS The analysis in Chapter 6 assumed the operability of 1 HPCF, RCIC,2 loops of LPFL and 8 ADS valves.

In addition, the automatic reactor isolation on Low Water Level 1.5 is assumed to be functional. The calculated peak clad temperature is 1025 F as given in Table 6.3-4.

LOCA ANALYSIS RESULTS For the main steamline break, scram was assumed to occur from tripping of the main turbine on a High Water Level 8 signal which is diverse from the SSLC. Also at the start of the LOCA, reactor feedpumps were tripped off by the Level 8 signal. It was also assumed that one CRD pump was running at the start of the LOCA and continues to operate but the CRD injection flow is not enough to overcome the break flow. The water level continues to drop and the core is uncovered. Reactor isolation was assumed to occur as a result of operator action upon entering the EOPs (assumed approximately 5 minutes after Low Water Level 1.5 is reached). Furthermore, it is assumed that the operator initiates HPCF(B) and places one condensate pump in operation to inject water into the RPV. This action is conservatively assumed to occur within 5 minutes after the start of the event. Thus RPV reflooding begins at this time since the reactor depressurizes through the break. Refer to Figures 1 through 3 for analysis results. These results of the analysis show that the peak clad temperature is 1711 F, well below the limit of 2200 F.

As a demonstration of the time available for operator action, an additional case was evaluated. For this case the same assumptions described above were used except that the operator did not isolate the reactor or initiate the HPCF(B) or the condensate pump injection until 7 minutes after the start of the event. The peak cladding temperature for this case was 2193 'F. Refer to Figures 4 through 6 for analysis results.

SUMMARY

For the postulated event of main steam line break inside containment coincident with an EMUX common mode f ailure, sufficient automatic control functions, information, and controls that are independent of EMUX are available in the main control room to mitigate the event, assuming that all operator actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown System.

1 WATER LEVEL 6.00E+01 5.00E+01 n

4.00E+01 v

3.00E+01 H

H 9

e 2.00E+01

.__)

LLJ

,q f 1.00E+01 V

0.00E+00 0.00E+00

~1.00E+02 2.00E+02 3.00E+02 1.00E+02 5.0CE+02 6.00E+02 TIME (SEC)

FIG 1. RPV WATER LEVEL - MSL BREAK

1 VESSEL PRESSURE 1

1.50E+03 l

l l

l l

3 3

3 1.00E+03 m

H EO CL LLJ

~

CC D

CO COg 5.00E+02 CL i

I I

I i

0.00E+00 0.00E+00 1.00E+02 2.00E+02 3.00E+02 4.00E+02 5.00E+02 6.00E+02 TIME (SEC)

FIG 2. RPV PRESSURE - MSL BREAK 4

m.

m m

--_-------__-r------,-------

1 PCT 2.50E+03 i

i i

i i

i

)

l l

2.00E+03 mw v

1.50E+03 g

a m

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- 1 LU Q_

1.00E+03 2-Ld' p_

g O

F 5.00E+02 g.

1 I

I I

I I

I

-0.00E+00 0.00E+00 1.00E+02 2.00E+02 3.00E+02 4.00E+02 5.00E+02 6.00E+02 TIME (SEC)

FIG 3. PEAK CLAD TEMP - MSL BREAK

1 WATER LEVEL 6.00E+01 i

i i

i i

i i

F 5.00E+01 m

f' 4.00E+01 v

g H

3.00E+01 p._.

H CO

~

O 2.00E+01 y

g J

y 1

1.00E+01 W

I 0.00E+00 O.00E+00 1.00E+02 2.00E+02 3.00E+02 4.00E+02 5.00E+02 6.00E+02 7.00E+02 TIME (SEC)

FIG 9. RPV WATER LEVEL - MSL BREAK

1 VESSEL PRESSURE 1 =5 i

j l

j i

i i

i i

i t

1.00E+03-H.

U2 Q_ -

v UJ 7g 03 uJ 5.00E+02

-QC Q_-

l li l

l lt

.M l'

0.00E+00 O.00E+00 1.00E+02

.2.00E+02 3.00E+02 4.00E+02 5.00E+02 6.00E+02 7.00E+02

. TIME (SEC)

FIG 5. RPV PRESSURE - MSL BREAK

.)

1 PCT i

i i

i i

l l

l

,l l

l 2.50E+03.

i 2.00E+03 m

1.50E+03 g

g D

F-M LLJ Q_.

1.00Et03 E

LLJ l-1 O

V

/

5.00E+02 g

LLJ g

I 0.00E+00 0.00E+00 l.00E+02 2.00E+02 3.00E+02 9.00E+02 5.00E+02 6.00E+02 7.00E+02 TIME (SEC)

FIG 6. PEAK CLAD TEMP - MSL BREAK o

6/18/93 i

i EVENT: FEEDWATER LINE BREAK INSIDE CONTAINMENT l

i This event is postulated to be a break of a feedwater line coincident with a undiscovered common mode failure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a feedwater line break inside containment, the reactor would be expected to automatically scram on High Drywell Pressure or RPV Low Water Level 3. Because of the assumed EMUX common rnode failure, these scram functions are postulated to fail, in addition, all ECCS systems are assumed not be available because of the postulated common mode failure except for HPCF(B) which has a diverse manual system initiation capability on the main control console. For the postulated line break, the RPV pressure will drop rapidly, resulting in closure of the turbine control valves by the pressure regulator and automatic scram signals (diverse from EMUX) from the subsequent turbine trip. A!so, when RPV water level drops to Low Water Level 2, the ATWS scram functions are automatically activated (diverse from EMUX) and initiate an automatic scram.

EOP ENTRY CONDITIONS:

l The following alarms are provided by equipment independent of the EMUX. These are the entry i

~

conditions for emergency operating procedures expected for a LOCA inside the primary containment from instruments that are diverse from EMUX.

I

1. RPV WATER LEVEL LOW [ FIXED POsmON)

2. DRYWELL PRESSURE HIGH [ FIXED POsmON)

OPERATOR ACTIONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs developed from the RPV Control Guideline on High Drywell Pres?nre Or RPV Water Level Low as an entry condition, and concurrently entering EOPs developed from th<

/ nary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the f ollowing sets of actions are executed concurrently

1.

RPV Control 1.

Initiate a manual scram if a scram has not been initiated.

2.

Initiate reactor isolation if it should have been isolated automatically but did not. (MSIV controlis diverse from EMUX.)

3.

Restore and maintain RPV water level (water level signal is diverse from EMUX) above Level 3 using the CRD system and HPCF(B).

4.

If RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the active fuel.

5.

When RPV water level cannot be maintained above top of the active fuel, depressurize the reactor. (This action is not necessary as the reactor is depressurized through the break.)

OPERATOR ACTIONS PER EOPS (continued) ll. Primary Containment Control:

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

2.

If necessary, initiate drywell spray using the fire protection system and the firewater addition I

mode of RHR(C ) for primary containment pressure control (drywell pressure signal is diverse l

from EMUX).

I l

CONTAINMENT RESPONSE The increase in suppression pool temperature due to vessel blowdown energy and decay heat is calculated to be approximately 70 *F over a time period of one hour. From this calculation, it is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room.

The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure. Subsequent operator actions at the Remote Shutdown System will include initiation of the RHR suppression pool cooling function.

COMPARISION TO CHAPTER 6 LOCA ANALYSIS The analysis in Chapter 6 assumed the operability of 1 HPCF,2 loops of LPFL and 8 ADS valves. In addition, the automatic reactor isolation on Low Water Level 1.5 is assumed to be functional. The calculated peak clad temperature is 1008 *F as given in Table 6.3-4.

LOCA ANALYSIS The complete circumferential break of a feedwater injection line was analyzed. For this case, HPCF(B) and one CRD pump were assumed to be available for RPV makeup. The CRD takes suction from the condensate system. The hotwell holds enough water to provide 4 minutes of rated feedwater flow.

No credit is taken for operator action and scram is assumed to occur at Low Water Level 2 from the ASVS scram functions. After scram one CRD pump continues to inject water into the RPV but CRD injection flow is not enough to overcome the break flow. The water level continues to drop and the core is uncovered. At 5 minutes into the event it is assumed that the operator takes action to manually initiate the HPCF(B). Furthermore, it is assumed that the operator isolates the reactor 5 minutes after Low Water Level 1.5 is reached. The reactor is depressurized through the break and hence an emergency depressurization is not required. Refer to Figures 1 through 3 for analysis results. The peak cladding temperature is 1301 F which is well below the 2200 *F limit. For this event the two-phase water level remains at the elevation of the feedwater nozzle (about 38 ft) for about 80 seconds. This is not explicitly shown on Figure 1 but is accounted for in the analysis.

As a demonstration of the time available for operator action, an additional case was evaluated. For this case the same assumptions described above were used except that the operator did not initiate the HPCF(B) until 12 minutes after the start of the event. The peak cladding temperature for this case was 2098 *F. Refer to Figures 4 through 6 for analysis results. As discussed above for Figure 1 the initial rise 1

of the two-phase level is not explicitly shown in Figure 4 either but is accounted for in the analysis.

l i

SUMMARY

For the postulated event of a feedwater line break inside the containment, sufficient automatic control functions, information, and controls that are independent of EMUX are available in the main control room to mitigate the event and maintain the fuel clad temperature below its limit, assuming that all operator i

actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown System.

i l

l l

l l

r l

l

l l

1 WATER LEVEL 5.00E+01 i

i 9.00E+01-f--

I b

\\

3.00E+01 7

o H

v g

s 2.00E+01 1

)

~

._uJ uJ

_J 1.00E+01 I

I I

0.00E+00 0.00E+00 3.00E+02 6.00E+02 9.00E+02 TIME (SEC)

FIG 1..RPV WATER LEVEL - FWL BREAK 4

.m..

.a....

m

--_g-..

o 4.w--

s.

I VESSEL PRESSURE 1.50E+05 i

i i

i i

t 1.00E+03 m

H 07 1

CL v

g

- 1 m

01 03 t.d 5.00E+02 M

Q_

1 0.00E+00 0.00E+00 3.00E+02 6.00E+02-9.00E+02 TIME (SEC)

FIG 2. RPV PRESSURE - FWL BREAK' 4

i

1 PCT 2.50E+03 i

i j

i i

i 2.00E+03 n

LL

~

v g -

t.50E+03 g

D F-1 LLJ Q_

1.00E+03 E

Ltj p_

O O

t-1 5.00E+02-g LLJ g

0.00E+00 O.00E+00 3.00E+02 6.00E+02 9.00E+02 TIME (SEC)

FIG'3. PEAK CLAD TEMP - FWL BREAK

L i

WATER LEVEL 5.00E+01 g

i i

i i

i i

i i

f l

4.00E+0i t

P 1

l v

r 3.00E+01 7

s

[

2.00E+01 l

1 w

_1 1.00E+01 I

0.00E+00 0.00E+00 2.50E+02 5.00E+02

-7.50E+02 1.00E+03 1.2SE+03 TIME-(SEC)

FIG 9. RPV WATER LEVEL - FWL BREAK-1 t

.e

k 4

1 VESSEL PRESSURE

1. M +05 l

l l

{

~

t 1.00E+03 m

Hm-Q_

W M

W

~

W u.]

5.00E+02 1

I Q-1 i.

'I I

0.00E+00.

O.00E+00

.2.50E+02 S.00E+02 7.50E+02 1.00E+03

_L.25E+03 TIME

-(SEC)

FIG 5. RPV PRESSURE - FWL BREAK

..., ~. _. -

k i

PCT 2.50E+05 i

i i

i i

i i

i i

i i

i i

i i

i 2.00E+05

)

m v

1.50E+03 g

M D

I-t T

g

/

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1 5.00E+02 g

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FIG 6. PEAK CLAD TEMP - FWL BREAK r

... ~

6/18/93 EVENT: SHUTDOWN COOLING LINE BREAK INSIDE CONTAINMENT This event is postulated to be a break of one shutdown cooling line of RHR coincident with a undismvered common rnode failure of the Essential Multiplex System (EMUX) in such a manner that all valid anc correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a shutdown cooling line break inside containment, the reactor would be expected to automatically scram on High Drywell Pressure. Because of the assumed EMUX common mode failure, this scram is postulated to fail. In addition, all ECCS systems are assumed not be available because of the postulated common mode f ailure except for HPCF(B) which has a diverse manual system initiation capability on the main control console. The reactor feedpumps are assumed to fail at the beginning of the event. For the postulated line break, reactor scram is assumed to occur on a Low Water Level 2 ATWS scram signal which is diversed from the SSLC. The MSIVs should close when RPV water level drops to Level 1.5 but because of the assumed common mode failure, the MSIVs will not close automatically.

EOP ENTRY CONDITIONS:

The following alarms are provided by equipment independent of the EMUX. These are the entry conditions for emergency operating procedures expecter' W a LOCA inside the primary containment from inst;uments that are diverse from EMUX.

1. DRYWELL PRESSURE HIGH [ FIXED POSmON)

OPERATOR ACVONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs developed from the PPV Control Guideline on High Drywell Pressure as an entry condition, and concurrently entering EOPs developed from the Primary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the following sets of actions are executed concurrently:

1.

RPV Control 1.

Initiate a manual scram if a scram has not been initiated.

2.

Initiate reactor isolation if it should have been isolated automa3cally but did not. (MSIV controlis diverse from EMUX.)

3.

Mairitain RPV water level (water level signal is diverse from EMUX) above Level 3 using the CRD system, the condensate pumps and the HPCF(B).

4 If RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the sctive fuel.

5 When RPV water level drops below top of the active fuel, perform an emergency depressurization. (This action is not necessary as the reactor is depressurized through the break.)

OPERATOR ACTIONS PER EOPS (continued) l

11. Primary Containment Control:

l l

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

l 2.

If necessary, initiate drywell spray using the fire protection system and the firewater addition mode of RHR(C ) for primary containment pressure control (drywell pressure signal is diverse from EMUX).

CONTAINMENT RESPONSE ihe increase in suppression pool temperature due to vessel blowdown energy and decay heat is calculated to be approximately 70 F over a time period of one hour. From this calculation, it is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room.

The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure. Subsequent operator actions at the Remote Shutdown System willinclude initiation of the RHR suppression pool cooling function.

COMPARISION TO CHAPTER 6 LOCA ANALYSIS The analysis in Chapter 6 assumed the operability of 1 HPCF, RCIC,2 loops of LPFL and 8 ADS valves.

In addition, the automatic reactor isolation on Low Water Level 1.5 is assumed to be functional. The calculated peak clad temperature is 1008 *F as given in Table 6.3-4.

LOCA ANALYSIS l

The complete circumferential break of a shutdown cooling suction line was analyzed. For this case HPCF(B), one CRD pump and one condensate pump were assumed to be available for RPV makeup.

The CRD takes suction from the the condensate system (hotwell) or from the CST.

No credit is taken for operator action and scram is assumed to occur at Low Water Level 2 from the l

ATWS scram functions. After scram one CRD pump continues to inject water into the RPV but the CRD flow is not enough to overcome the break flow. The water level continues to drop and the core is l

uncovered. Upon entry into the EOPs, the operator is assumed to isolate the reactor 5 minutes after reaching the automatic isolation setpoint of Level 1.5. Furthermore, it is assumed that the operator initiates HPCF(B) and places one condensate pump in operation to inject water into the RPV. This action is conservatively assumed to occur at 5 minutes after the start of the event. (The operator is instructed to control water level above Level 3 with availab!e injection systems.) Thus RPV flooding begins at this time since the reactor is depressurized through the 'areak. Refer to Figures 1 through 3 for analysis results.

The calculated peak cladding temperature is 953 F which is well below the 2200 *F limit.

As a demonstration of the time available for operator action, an additional case was evaluated. For this case the same assumptions described above were used except that the operator did not initiate the HPCF(B) or the condensate pump until 12 minutes after the start of the event. The peak cladding temperature for this case was 1988 F. Refer to Figures 4 through 6 for analysis results.

j

SUMMARY

For the postulated event of a shutdown cooling suction line break inside the containment, sufficient automatic control functions, information, and controls that are independent of EMUX are available in the main control room to mitigate the event and maintain the fuel clad temperature below its limit, assuming that all operator actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold i

shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown l

System.

1 i

r

1 WATER LEVEL 5.00E+0i i

i j

g i

i i

1 9.00E+01

~

n

~

3.00E+01

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Z o

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H h

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1.00E+01 0.00E+00

~0.00E+00 1.00E+02-2.00E+02 5.00E+02 4.00E+02 5.00E+02 6.00E+02 TIME-

(SEC)

FIG 1. RPV WATER-LEVEL - SDCL BREAK J

-_______-_.--.-_._.-__.___-__---,,_--.--._u-_..__.-.-__--..._..____..__.x=-

.-L.. -

i 1

VESSEL PRESSURE 1.50E+05 i

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g g

g i

i i

me 1

1.00E+03 m

w g

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Q_

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0.00E+00 O.00E+00 1.00E+02 2.00E+02 3.00E+02 9.00E+02 5.00E+02 6.00E+02 TIME (SEC)

FIG 2. RPV PRESSURE - SDCL BREAK-

'I

i 1

PCT 2.s a+03 i

2.00E+03

^

Lt v

1.50E+03 g

QC LU

' Q_

1.00E+03 E

Lij p

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5.00E+02 g

Ld 1

0.00E+00 O.00E+00 t.00E+02 2.00E+02 3.00E+02 9.00E+02 5.00E+02 6.00E+02 TIME (SEC)

FIG 3. PEAK CLAD TEMP - SDCL BREAK

1 WATER _. VEL g

g g

i i

g 5.00E+0!

i i

i i

i i

h 4.00E+01 i

v 3.00E+01 o

H g

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.LU t.tj

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I I

I I

I I

I 0.00E+00 0.00E+00 2.00E+02 9.00E+02 6.00E+02 8.00E+02

-1.00E+05' TIME (SEC) 1 FIG 4. RPV WATER' LEVEL - SDCL BREAK 1

f 4

.m.

-c~

,4

=.

-,-.,_,__-__._____m___.___._m m

1 VESSEL PRESSURE i

i i

3 i

3

)

l l

l l

l l

1.50E+05 i

i

~

~

r 1.00E+03 m

h CL v

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Ul LLI 5.00E+02

.y-1 0.00E+00 O.00E+00 2.00E+02 et.00E+02 6.00E+02 8.00E+02 1.00E+03 TIME (SEC)

FIG 5. RPV PRESSURE - SDCL BREAK

I1

'I

1 i

30+

E 0

i N

I 0

1 l

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2 0

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8 i

3 l

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0 I

1 0

_ _ ~ F - - _ -

+

F E00 0

3 3

3 3

2 0

0 0

0 0

0

+

+

+

+

+

E E

E E

E E

0 0

0 0

0 0

5 0

5

0. -

0 0

2 2

1 1

5 0

-. fJ _- J

^u ggDF<OuQ2up_ OO M<)uQ i,

l 6/18/93 EVENT: HPCF LINE BREAK INSIDE CONTAINMENT This event is postulated to be a break of HPCF(B) injection line coincident with a undiscovered common mode failure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a HPCF(B) line break inside containment, the reactor would be expected to automatically scram on High Drywell Pressure. Because of the assumed EMUX common mode failure, this scram is postulated to fail. In addition, all ECCS systems are assumed not be available because of the postulated common mode failure and the break postulated in HPCF(B). The reactor feedpumps are assumed to fail. For the postulated line break, reactor scram is assumed to occur on a Low Water Level 2 ATWS scram signal which is diversed from the SSLC. The MSIVs should close when RPV water level drops to Level 1.5 but because of the assumed common mode failure, the MSIVs will not close automatically.

EOP ENTRY CONDITIONS:

The following alarms are provided by equipment independent of the EMUX. These are the entry conditions for emergency operating procedures expected for a LOCA inside the primary containment from instruments that are diverse from EMUX.

1. DRYWELL PRESSURE HIGH [ FIXED POSITION)

OPERATOR ACTIONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs developed from the RPV Control Guideline on High Drywell Pressure as an entry condition, and concurrently entering EOPs developed from the Primary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the following sets of actions are executed concurrently:

1.

RPV Control 1.

Initiate a manual scram if a scram has not been initiated.

2.

Initiate reactor isolation if it should have been isolated automatically but did not. (MSIV controlls diverse from EMUX.)

3.

Maintain RPV water level (water level signal is diverse from EMUX) above Level 3 using the CRD system and the condensate pumps. Since the postulated break is located at the HPCF(B) injection line which is below Level 1 but above top of the active fuel, water will spill out the break and into the drywell.

4 If RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the active fuel.

5 When RPV water level drops below top of the active fuel, perform an emergency depressurization. (This action is accomplished by re-opening the MSIVs if closed and opening the turbine bypass valves.)

OPERATOR ACTIONS PER EOPS (continued) l

11. Primary Containment Control:

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

2.

If necessary, initiate drywell spray using the fire protection system and the firewater addition mode of RHR(C ) for primary containment pressure control (drywell pressure signal is diverse from EMUX).

CONTAINMENT RESPONSE The increase in suppression pool temperature due to vessel blowdown energy and decay heat is calculated to be approximately 70 'F over a time period of one hour. From this calculation, it is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room.

The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure. Subsequent operator actions at the Remote Shutdown System will include initiation of the RHR suppression pool cooling function.

l l

COMPARISION TO CHAPTER 6 LOCA ANALYSIS The analysis in Chapter 6 assumed the operability of RCli,2 loops of LPFL and 8 ADS valves. In addition, the automatic reactor isolation on Low Water Level 1.5 is assumed to be functional. The calculated peak clad temperature is 1008'F as given in Table 6.3-4.

LOCA ANALYSIS l

The complete circumferential break of the HPCF(B) injection line was analyzed. For this case only one CRD pump and one condensate pump were assumed to be available for RPV makeup. The CRD takes suction from the the condensate system (hotwell) or from the CST.

No credit is taken for operator action and scram is assumed to occur at Low Water Level 2 from the ATWS scram functions. After scram one CRD pump continues to inject water into the RPV but it is not j

enough to overcome the break flow. The water level continues to drop and the core is uncovered. Upon entry into the EOPs, the operator is assumed to isolate the reactor 30 seconds after reaching the automatic isolation setpoint of Low Water Level 1.5. (For this event the short isolation time is conservatively used to maximize the system presure and break flow.) Furthermore, it is assumed that the operator places one condensate pump in operation to be available for injecting water into the RPV once the system has been depressurized. This action is assumed to occur at 5 minutes after the start of the event. (The operator is instructed to control water level above Level 3 with available injection systems.)

When water level drops below the top of active fuel, the operator is assumed to re-open the MSIVs and perform an emergency depressurization via the turbine bypass valves (per EOPs) to depressurize the reactor and thus permit condensate system flow to the RPV. Refer to Figures 1 through 3 for analysis results. The calculated peak cladding temperature for the first peak is 940 F which is well below the 2200 *F limit. At 25 minutes into the event it is assumed that water injection into the RPV by the condensate pump ceases because the water inventory in the hotwell has been depleted. The CRD pump is manual!y aligned to the CST and continues to inject water into the RPV but it is not enough to maintain the water level. At 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event the calculated peak cladding temperature is only 1706 *F. This shows that there is sufficient time for the operator to go to the remote shutdown panel and initiate ECCS injection.

As a demonstration of the time available for operator action, an additional case was evaluated. For this case the same assumptions described above were used except that the operator did not initiate the condensate pump until 20 minutes after the start of the event and did not depressurize the RPV until 20 minutes after the water level had reached the top of the active fuel. The peak cladding temperature for this case was 1465 F. Refer to Figures 4 through 6 for analysis results.

l

SUMMARY

For the postulated event of HPCF(B) line break inside the containment, sufficient automatic control functions, information, and controls that are independent of EMUX are available in the main control room to mitigate the event and maintain the fuel clad temperature below its limit, assuming that all operator actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour :Jeriod. Subsequently, reactor cold shutdown conditions aM post accident recovery operations can bv initiated using the Remote Shutdown System.

l f

I l

r 1

WATER LEVEL i

i i

i g

g l

l g

5.00Et01 i

i 4.00E&O1 m

l-L-

LL v

3.00EF01 Z

O H

1-

.g 2.00E+01 g

Q._

j LLJ LLJ

_J 1.00E+01 l

l l

l I

l l

t

0.00EK)0 1.00E+03 2.00E+03 3.00E+03 1.00E+03

-.5.00E+03-6.00E43 7.00E+03 TIME (SEC)

FIG-1. RPV WATER-LEVEL - HPCF LINE BREAK

~

l 1

VESSEL PRESSURE 1.50Et03 i

1.00E+03 m

5-+

EO L

Q_

v w

tr D

EO CD

. L1_J 5.00E+02 gr -

Q t

~

~

I I

I I

I 0.00E+00 O.00E+00 1.00E+03 2.00E+03 3.00E+03 9.00E+03 5.00E+03 6.00E+03 7.00E+03 TIME (SEC)

FIG ~2.

RPV PRESSURE - HPCF LINE BREAK q

1 PCT 2.50E+05

)

)

)

)

i i

i

.i 2.00E+03 n

v 1.50E+05 l

gx

/

F--<

LLJ g

1.00E+03 O

~

/

I j

,/

5.00E+02 p

g

/

1 I

I 0.00E400 I

0.00E+00 1.00E+03 2.00E+03 3.0CE+03 4.00E+03 5.00E+03 6.00E+03 7.00E+03 TIME (SEC)

FIG 3. PEAK CLAD TEMP - HPCF LINE BREAK

. m.

m.m

.m.. m.

m.i

1 WATER LEVEL 5.00E+01 i

4.00E+01

~

n l--

u LL v

3.00E+01 8

1x

,I i

2.00E+01 g

L1 g

Lt.J

__J 1.00E+01

~

~

I 0.00E+00 O.00E+00 1.00E+03 2.00E+03 3.00E+03 4.00E +03 -

5.00E+03 S.00E+03 7.00E+03 TIME (SEC)

FIG 4. RPV WATER LEVEL - HPCF LINE BREAK

I VESSEL PRESSURE 1.50E+03 i

i g

g l

g l

i i

i i

1.00E+03 m

l H

-U3 CL

~

w

~

U~)

u, LLI 5.00E+02 e

CL e

I l

i i

i i

i i

i i

0.00e+00 0.00E+00 1.00E+03 2.00E+03 3.00E+03 4.00E+03 5.00E+03 6.00E+03 7.00E+03 TIME (SEC)

FIG 5. RPV PRESSURE - HPCF LINE BREAK

i 1

PCT 2.50E+03

)

i 2.00E+03

^

+

v g

1.50E+03 7

y

/_

D F--

E LL)

Q 1.00E+03 Z

LL]

W Q

O t-E 5.00E+02 n'

/

w Q-0.00E+00

.0.00E+00 1.00E+03 2.00E+03 3.00E+03 t00E+03 5.00E+03 6.00E+03 7.00E+03 TIME (SEC) 4 FIG 6. PEAK CLAD TEMP - HPCF LINE BREAK.

i

6/18/93 EVENT: BOTTOM DRAIN LINE BREAK INSIDE CONTAINMENT This event is postulated to be a break of RPV bottom drain line coincident with a undiscovered common mode failure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost.

AUTOMATIC ACTIONS After a RPV bottom drain line break inside containment, the reactor would be expected to automatically scram on High Drywell Pressure. Because of the assumed EMUX common mode failure, this scram is postulated to fail. In addition, all ECCS systems are assumed not be available because of the postulated common mode f ailure except for HPCF(B) which has a diverse manual system initiation capability on the main control console. The reactor feedpumps are assumed to fail at the beginning of the event. For the postulated line break, reactor scram is assumed to occur on a Low Water Level 2 ATWS scram signal which is aiversed from the SSLC. The MSIVs should close when RPV water level drops to Level 1.5 but because of the assumed common mode failure, the MSIVs will not close automatically.

EOP ENTRY CONDITIONS:

The following alarm is provided by equipment independent of the EMUX. This is the entry condition for emergency operating procedures expected for a LOCA inside the primary containment from instruments that are diwerse from EMUX

1. DRYWELL PRESSURE HIGH [ FIXED POSmON]

OPERATOR ACTIONS PER EOPS The expected principal operator actions are given herein. All control functions and process parameters are provided by equipment independent of EMUX.

Upon entering the EOPs developed from the RPV Control Guideline on High Drywell Pressure as an entry condition, and concurrently entering EOPs developed from the Primary Containment Control Guideline on Drywell Pressure High alarm as an entry condition, the following sets of actionc are executed concurrently:

1.

RPV Control 1.

Initiate a manual scram if a scram has not been initiated.

2.

Initiate reactor isolation if it should have been isolated automatically but did not. (MSIV controlls diverse from EMUX.)

j 3.

Maintain RPV water level (water level signal is diverse from EMUX) above Level 3 using the CRD system, the condensate pumps and the HPCF(B).

4 If RPV water level cannot be maintained above Level 3, maintain RPV water level above top of the active fuel.

5 When RPV water level drops below top of the active fuel, perform an emergency depressurization. (This action is accomplished by re-opening the MSIVs if closed and opening the turbine bypass valves.)

.=

i l

i OPERATOR ACTIONS PER EOPS (continued) il. Primary Containment Control:

1.

Initiate wetwell sprays using the fire protection system and the firewater addition mode of RHR(C) for primary containment pressure control.

2.

If necessary, initiate drywell spray using the fire protection system and the firewater addition j

mode of RHR(C ) for primary containment pressure control (drywell pressure signal is diverse from EMUX).

CONTAINMENT RESPONSE The increase in suppression pool temperature due to vessel blowdown energy and decay heat is conservatively bounded by the other inside containment LOCA events and therefore will be less than 70 *F over a time period of one hour. From this calculation, it is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room. The RHR suppression pool i

cooling function is assumed to be not avaitable because of the EMUX common mode failure.

Subsequent operator actions at the Remote Shutdown System willinclude initiation of the RHR suppression pool cooling function.

~

i COMPARISION TO CHAPTER 6 LOCA ANALYSIS The analysis in Chapter 6 assumed the operability of 1 HPCF, RCIC,2 loops of LPFL and C ADS valves.

In addition, the automatic reactor isolation on Low Water Level 1.5 is assumed to be functionat. The calculated peak clad temperature is 1008 *F as given in Table 6.3-4.

LOCA ANALYSIS I

1 The complete circumferential break of the RPV bottom drain line was analyzed. For this case HPCF(B),

one CRD pump and one condensate pump were assumed to be available for RPV makeup. The CRD takes suction from the the condensate system (hotwell) or from the CST.

No credit is taken for operator action and scram is assumed to occur at Low Water Level 2 from the ATWS scram functions. After scram one CRD pump continues to inject water into the RPV but the CRD flow is not enough to overcome the break flow. The water level continues to drop and the core is uncovered. Upon entry into the EOPs, the operator is assumed to isolate the reactor 30 seconds after reaching the automatic isolation setpoint of Low Water Level 1.5.' (For this event the short isolation time is conservatively used to max mize the system pressure and break flow.) Furthermore,it is assumed that the operator initiates HPCF(B) which begins injection into the RPV and places one condensate pump in operation to be available for injecting water into the RPV once the system has been depressurized. This action is assumed to occur at 5 minutes after the start of the event. (The operator is instructed to control water level above Level 3 with available injection systems.) When water level drops below the top of active fuel, the operator is assumed to re-open the MSIVs and perform an emergency depressurization via the turbine bypass valves (per EOPs) to depressurize the reactor and thus permit condensate system flow to the RPV. Refer to Figures 1 through 3 for analysis results. The calculated peak cladding temperature is 695 *F which is well below the 2200 'F limit.

i As a demonstration of the time available for operator action, an additional case was evaluated.. For this case the same assumptions described above were used except that the operator did not initiate the.

HPCF(B) or the condensate pump until 20 minutes after the start of the event and did not depressurize.

the P PV until 20 minutes after the water level had reached the top of the active fuel. The peak cladding temi erature for this case was 1877 *F. Refer to Figures 4 through 6 for analysis results.

SUMMARY

For the postulated event of RPV bottom drain line break inside the containment, sufficient automatic control functions, information, and controls that are independent of EMUX are available in the main control room to mitigate the event and maintain the fuel clad temperature below its limit, assuming that all operator actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown System.

l l

4 i

WATER LEVEL 5.00E+01 i

4.00E+01 n

F--

L.

LL m

3.00E+01 h

7 -

M.

~

H H

ED 2.00E+01 d

Q_.

i t1J

~

LLI

~

_J 1.00E+01 l

l l

l l

l 1

1 1

t 1

i i

0.00E+00 0.00E+00 1.00E+02 2.00E+02 3.00E+02 9.00E+02 5.00E+02 6.00E+02 7.00E+02 8.00E+02 TIME (SEC)

FIG 1. RPV WATER LEVEL - BDL BREAK-

t

?

t i

VESSEL PRESSURE 1.50E+03 i

i

~

I 1.00E+05 m

(f)

Q-v W

M Dg 07 LiJ 5.00E+02 T

m

-LL

+

t I

0.00E+00 0.00E+00 t.00E+02 2.00E+02 3.00E+02 9.00E+02 5.00EK)2 6.00E+02 7.00E+02 8.00E+02

- i TIME

( SEC )-

FIG 2. RPV PRESSURE - BDL BREAK

=

s L

t

-[

l

\\

PCT l

2. m 3 2.ooE+o3

-u.

m 1.50E+03 gg D

l --

E w

Q_

1.00E+o3 rw i

}--

1

(

1 1

5.00E+o2 g

w-

.a.

-0.00E+o0 7

o.coe oo t.ooema 2.ooe+o2-3.ooe+o2 4.ooe+o2

s. ooc +o2 6.ooe+oa 7.ooe+oa 8.coe+o2 TIME

-(SEC)

I FIG 3. PEAK CLAD TEMP - BOL BREAK i

i

l l

1 WATER LEVEL 5.00E+0L i

i i

i i

i i

i i

i 4.00E+01 m

i H

L j

I1-j v

3.00E+01 4

7 i

D H

t--

~t g

?

2.00E+01 1

_J

-w LLJ I

I~

1.00E+01 i

l i

l I

I i

t t

t i

f t

i 0.M+M I

0.00E+00

.3.00E+02 6.00E+02 9.00E+02 1.20E+03 1.50E+03 l

TIME (SEC)

FIG 4. RPV WATER LEVEL - BDL BREAK

1 VESSEL PRESSURE 1.50E+03 i

1.00E+05 m

H (n

CL tJ g

CD LLI 5.00Et02 LX Q-I I

I I

I 0.00E+00 0.00E+00 3.00E+02 6.00E+02

-9.00E+02

'1.20E+03 1-.50E+05 TIME (SECT FIG 5. RPV PRESSURE - BDL BREAK 4

l

=. -

1 PCT 2.50E+03 i

l

)

l l

i i

i i

i 2.00E+03 m

v 1.50E&O3 g

.y

~D l--

Z LU r'

Q_

.1.00E+03 j'

Z-4 uJ I--

Q i

i O

F i

5.00E+02 i

M u_1 w

~

l I

I I

~

0.00E+00 1

0.00E+00 3.00E+02 6.00E+02 9.00E+02 1.20E+03 1.50E+03 4

TIME-(SEC)

FIG 6. PEAK CLAD TEMP - BDL-BREAK 4

}

L 6*M

=

teL-a.

m-a h.

-+- e

'k+

rm

+=

a

.w.

.... -- a.

+-

i 6/18/93 EVENT: 15.2.2.2.1.3 GENERATOR LOAD REJECTION WITH FAILURE OF ALL BYPASS VALVES,

(

15.2.3.2.1.3 TURBINE TRIP WITH FAILURE OF ALL BYPASS VALVES These events are postulated to occur coincident with a undiscovered common mode failure of the Essential Multiplex System (EMUX) in such a manner that all valid and correct EMUX control and monitoring data transmissions are lost. The reactor response to these two events are similar.

AUTOMATIC ACTIONS Upon a turbine / generator trip, the reactor scrams immediately. The scram signals generated by turbine stop valve or control valve instruments are hardwired to RPS. The SRVs open on spring setpoint since it is assumed that the SRVs cannot be open by its normal relief mode due to the postulated common mode f ailure. One CRD continues to injects water into the RPV since it is normally in operation.

EOP ENTRY CONDITIONS:

The following alarm condition is provided by equipment independent of the EMUX and is an entry condition to the emergency operating procedures:

1. RPV WATER LEVEL LOW [ FIXED POSITION),

Column 8.18F-14.

OPERATOR ACTIONS For this event scenario, one CRD pump and HPCF(B) are used to inject water into the RPV. It is assumed that because of the failure of the bypass valves, the bypass valves can not be reopen for decay heat removal.

1.

Enter EOPs developed from the RPV Control Guideline, upon receiving the RPV Water Level Low alarm.

2.

Restore and maintain water level (water level signal is diverse from EMUX) above Level 3 using one CRD pump, HPCF(B) and one condensate pump if reactor pressure decreases below its shutoff head.

3.

11 water level cannot be maintained above Level 3, control water level above top of active fuel.

4.

If water level drops below top of active fuel, perform an emergency depressurization (because of the assumed failure of the bypass valves and the control capability of ADS and SRVs, this operation is not possible).

CONTAINMENT RESPONSE The increase in suppression pool temperature due to decay heat is calculated to be approximately 40 *F over a time period of one hour. From this calculation, it is concluded that initiation of suppression pool cooling using RHR will not be required for the first hour of this postulated scenario in which all operator actions are limited to those performed in the main control room. The RHR suppression pool cooling function is assumed to be not available because of the EMUX common mode failure. Subsequent operator actions at the Remote Shutdown System willinclude initiation of the RHR suppression pool cooling function.

l l

l l

COMPARISON TO CHAPTER 15 ANALYSIS The analysis in Chapter 15 only simulated the first five seconds of the turbine / generator trip with bypass failure event. The primary purpose is to analyze the effect on fuel thermal margins.

ANALYSIS RESULTS The analysis assumed the use of one CRD pump and HPCF(B) for RPV makeup. After scram one CRD pump continues to inject water into the RPV but the CRD flow is not enough to overcome the boil-off.

Furthermore, it is assumed that the operator initiates HPCF(B) which begins injection into the RPV and places one condensate pump in operation to be available for injecting water into the RPV orce the system has been depressurized. This action is assumed to occur at 5 minutes after the start of the event. Since the reactor pressure does not decrease below the shutoff head of the condensate pump, the operator will not be able to operate the condensate pumps for makeup to the RPV. The SRVs are assumed to open on the spring relief setpoint and discharge steam to the suppression pool. The turbine bypass valves are assumed to f ail and hence cannot be used for pressure control and decay heat removal. Refer to Figures 1 through 3 for analysis results. Since there is no core uncovery calculated for this case, the peak cladding temeprature remains at the saturation temperature throughout the event.

As a demonstration of the time available for operator action, an additional case was evaluated. For this case the same assumptions described above were used except that the operator did not initiate the HPCF(B) or the concensate pump until 20 minutes after the start of the event. The peak cladding temperature for this case was 939 *F. Refer to Figures 4 through 6 for analysis results.

SUMMARY

For the postulated event of turbine trip or generator trip with total failure of the bypass valves coincident l

with an EMUX common mode failure, sufficient information and controls that are independent of EMUX are available in the main control room to mitigate the event, assuming that all operato' actions will be limited to the main control room for the first hour. Sufficient water inventory is available for decay heat removal during that one hour period. Subsequently, reactor cold shutdown conditions and post accident recovery operations can be initiated using the Remote Shutdown System.

l l

1 blATER LEVEL s.00E+0i 4.00E+01 F-LL 3.00E+01-

~

W Z

(-)

H F-g

[

2.00E+01 o_

.a

. u.J

__1 1.00E+01 3

0.00E+00 0.00E+00 2.00E+02 4.00E+02 6.00E+02 8.00E+02-t.00E+05 TIME (SEC)

FIG 1. RPV WATER LEVEL - TURB./ GEN. TRIP

1 VESSEL PRESSURE 2 "5 i

l l

j j

i i

i i

i i

i i

i 1.00E+03 m

o

~

LLI LLJ 5.00E+02 O-l l

l l

l I

0.00E+00.

I I

I I

l 0.00E+00 2.00E+02 9.00E+02 6.00E+02 8.00E+02 1.00E+03 TIME (SEC)

FIG 2. RPV PRESSURE - TURB./ GEN. TRIP

1 PCT 2.50E+03 i

i i

i i

i e

i i

i 2.00E+03

^

LL

~

v 1.50E+03 g.

g D

l-

'M L2J Q-1.00E+03 Z

Ld p_

i O

~

O t-i i

1-5.00E+02 g-LU e

0.00E+00 O.00E+00 2.00E+02 9.00E+02.

6.00E+02 8.00E+02 1.00E+03 TIME (SEC)

FIG 3. PEAK CLAD TEMP - TURB./ GEN. TRIP t

>m-we w=-

4 Mt+-

~

r' e

4de4-4mer-

W rMer U w*'i

+-

H

r~

WD h,

n "m--

1 I

WATER LEVEL i

i i

t i

i i

g j

l j

5.00E+01 i

4.00E+01 n-

}--

L-LL v

3.00E+01 F-

>-4

[

2.00E+01 g

__I t.1.J

_I 1.00E+01

~

l l

t 1

l l

1 I

I I

I 0.00E+00 3.00E+02 6.00E+02 9.00E+02 1.20E+03 1.50E+03 TIME (SEC)

FIG'4. RPV WATER LEVEL -'TURB./ GEN. TRIP

1 VESSEL PRESSURE 1.50E+03 i

3 i

i e

b g_

5.00E+02 CL l

I 0.00E+00 O.00E+00 3.00E+02 6.00E+02 9.00E+02 1.20E+03 I.50E+03 TIME (SECL FIG 5. RPV PRESSURE - TURB./ GEN. TRIP.

a-p 1

PCT 2.50E&O3 i

i i

i i

i j

i i

4 2.0%+03

^

v g-1.50E+03 g

p M

LLI Q

1.00E+03 t

T LL1 O

t-

__ t __ _ ______ __ _

T 5.00E+02 g

t LLI.

Q_

-0.00E+00 O.00E+00 3.00E+02 6.00E+02 9.0CE+02

-1.20E+03 1.50E+03 TIME

-( SEC ) -

FIG 6. PEAK CLAD TEMP -'TURB./ GEN. TRIP T

- - _ _ - _ _ - _. _ = _ - = _ _ _ _ - - - -. - _ _ - - _.- - - -, _ _ _ _ _ - _ _ _ _. _ _...

.