ML19281A067

WARD-D-0169, Clinch River Breeder Reactor Plant,Response of Plant to Postulated Loss of One Heat Transport Loop W/ Failure of Both Reactor Shutdown Sys. Addresses Anticipated Event Causing Most Severe Loss of Heat Removal

ML19281A067
Person / Time
Site:	Clinch River
Issue date:	02/28/1977
From:	Markowski F, Mccall T EMVWE, WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19281A066	List: ML19281A064 ML19281A067
References
WARD-D-0169, WARD-D-169, NUDOCS 7902220010
Download: ML19281A067 (25)

Text

Tm WARD-D-0169 I

Clinch River Breeder Reactor Plant RESPONSE OF THE PLANT TO A POSTULATED LOSS OF ONE HEAT TRANSPORT LOOP WITH FAILURE OF BOTH REACTOR SHUTDOWN SYSTEMS FEBRUARY 1977 Prepared for the United States Department of Energy under contracts EY-76-C-15-2395 and EY-76-C-15-0003.

'[ - Any Further Distribution by any Holder of this Document or of the Data Therein to Third Parties Representing Foreign Interest, Foreign Govern-ments, Foreign Companies and Foreign Subsidi-artes or Foreign Divisions of U.S. Companies Should be Coordinated with the Director, Division of Reactor Research and Technology, United States Department of Energy.

@ Westinghouse Electric Corporation ADVANCED REACTORS DIVISION MADISON, PENNSYLV ANI A 15663 7 9 0 2 2 2 0o/d

WARD-D-0169 CLINCH RIVER BREEDER REACTOR PLANT RESPONSE OF THE PLANT TO A POSTULATED LOSS OF ONE HEAT TRANSPORT LOOP WITH FAILtRE OF BOTH REACTOR SHUTDOWN SYSTEMS m

Prepared By: -

\- D Wha T. B. McCall

'F. J. Markowski Approved By:

L. E. Str wbr dge E. H. Hemmerle February 1977 Advanced Reactors Division Westinghouse Electric Corporation Madison, Pennsylvania 15663 i

INFORMATION CONCERNING USE OF THIS REPORT PATENT STATUS This document copy, since it la transmitted in advance of patent clearance, la made available in confidence solely for use in performance of work under contracts with the U.S.

Department of Energy. This document is not to be published nor its contents otherwise -

disseminated or used for purposes other than specified above before patent approval for such release or use has been secured, upon request, from the Director, Chicago Patent Group, U.S. Department of Energy,0800 South Cass Avenue, Argonne,Illinols.

PRELIMINARY DOCUMENT This report contains information of a preliminary nature prepared in the course of work for the U.S. Department of Energy. This Information la subject to correction or modification upon the collection and evaluation of additionai data.

NOTICE This report was prepared as an account of work sponsored by the United States Government. Neither the U.S. Department of Energy nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any information, apparatus, product or process disciosed, or represents that its use would not infringe privately owned rights.

WESTINGHOUSE ELECTRIC CORPORATION ADVANCED REACTORS DIVISION BOX 158 MADISON, PENNSYLVANIA 15683 ii

TABLE OF CONTENTS Page I. INTRODUCTION 1 II. CONCLUSIONS 1 III. DISCUSSION 3 A. Comparison of HTS Loop Malfunctions 3 B. Plant Control System Control Modes 4 C. Plant Thermal-Hydraulic Response 5 D. Consequences 8 iii

I. INTRODUCTION This report evaluates the potential consequences of malfunctions which cause a decrease in heat removal capability of one Heat Transport System (HTS) loop.

To envelope these malfunctions, the analysis addresses the anticipated event which results in the most severe loss of heat removal in one HTS loop. The limiting anticipated event is a loss of electric power for both the primary and the intermediate coolant pumps. In addition, to assess plant design margins, this anticipated event is combined with a postulated coincident failure of both the primary and the secondary Reactor Shutdown Systems (RSS).

Therefore, the combined event is an Anticipated Transient Without Scram (ATWS). The analysis has been performed in part to respond to NRC Question 0001.440.

This report provides the following information: (1) justification that the loss of power to the primary and intermediate coolant pumps is the most limiting anticipated event involving loss of one HTS loop; (2) a description of the various control modes of the plant that could influence the response to this transient; (3) the thermal-hydraulic response of the plant to a loss of electric power to one HTS loop with failure of both shutdown systems, and (4) the evaluation of consequences of the event, including the potential for any structural failures, fuel pin cladding failures or radiological release.

II. CONCLUSIONS

1. The " anticipated event" consisting of the loss of electric power to the primary and intermediate coolant pumps in one HTS loop, when combined with failure of both shutdown systems, is the most severe ATWS initiated by a malfunction in one HTS loop. This is because (a) the continued operation of the other primary pumps would cause the checkvalve in the failed loop to close causing complete isolation of the failed loop and (b) the reactor core flow would immediately experience an approximately 30 percent reduction due to the cessation of flow in one primary loop.

1

2. Of the four control modes possible for the plant (see Figure 1) the ones that would result in the most severe thermal hydraulic plant responses for the ATWS event postulated are the Reactor Flux Control Mode and Manual Operation. These are the least likely control modes for power range operation, but the consequences have nevertheless been examined. In the Reactor Flux Control Mode the Plant Control System (PCS) would counteract all reactivity feedback effects and hold the nuclear power constant until the flux-to-flow rod block circuit would arrest the control rods. It is conservatively assumed that the reactor operator would take no corrective action. With the given system parameters the rod block circuit would respond at 2 seconds after the loss of power to the pumps had occured, at which time the control rods would not have moved by more than 0.3 inches.

Therefore, the Reactor Flux Control Mode leads to essentially the same transient as Manual Operation. In both aforementioned modes negative reactivity feedbacks would reduce the reactor power and a new steady state at reduced power would be established. In the Supervisory Control Mode and the Reactor Temperature Control Mode the reactor power would be rapidly reduced resulting in a mild transient with no significant safety consequences.

3. For the postulated event consistency of loss of electric power to the primary and intermediate pumps to one HTS loop, failure of both shutdown systems, operation in the Reactor Flux Control Mode, and no operator corrective action a new steady state at #85 percent full power (975 MWt) is reached in which all the core heat is rejected through the two remaining HTS loops at higher than normal but acceptable temperatures. No hot channel coolant boiling would occur. The highest sodium temperature at the core hot channel exit would be 15740F (at #15 seconds) with a margin of 2400F below ~

the local boiling temperature. The core hot channel exit temperature would decrease to 15500F after 8 minutes, and would settle to 15200F after 25 minutes. There would be no HTS structural failures. There could be some fuel pin cladding failures in random locations since the highest cladding temperature in the 2

0 fuel hot channel would be above 1600 F (clad failure criterion established in Section 15.1 of the PSAR) for 8.7 minutes. However, the maximum cladding temperature would be #16300F. However, if a few cladding failures were to occur the release of fission gas into the coolant would not be sufficient to result in a loss of coolable geometry. Since the reactor flow would be approximately 67% of normal, the gas release would not cause any significant fuel channel flow perturbations.

4. For the most limiting single loop ATWS, occuring during the most severe plant control mode (and the least likely to be used), and conservatively assuming no operator corrective action there would be no result in? fuel melting and no core disruption.

III. DISCUSSION A. Comparison of HTS Loop Malfunctions A single heat transfer train can be shut off by a failure in either the primary loop, the intermediate loop, or the stea:'i system. Failure in the primary loop is a more severe event than in the intermediate or steam systems because of the reduction in core flow. Examples of anticipated failures are:

1. Coastdown of one primary coolant pump. It could result from a failure of the electric power supply or by a failure of the pump motor.

2. A coastdown of one intermediate coolant pump, caused by events as outlined for the primary pump.

3. Inadvertent closure of a superheater and/or evaporator isolation valve or a feedflow control valve.

3

4. Inadvertent actuation of the sodium-water-reaction pressure relief system. During this event, the steam in the affected loop would be released through the steam dump valves, the modules would be isolated and flushed with nitrogen, and the intermediate sodium would be released into the sodium dump tank.

All of these everts would lead to loss of heat removal thru one HTS train. A coastdown of the primary and intermediate coolant pumps in one HTS Loop has been chosen for analysis as the most limiting anticipated event. It not only deactivates one of the three heat transport trains, it also reduces the reactor flow to 67% under the most adverse plant control mode and to approximately 73% in other modes.

As illustrated in Figure 1, a failure of the Reactor Shutdown Systems could be postulated to be of a mechanical or electrical nature. If mechanical failures (stuck rods) were postulated a sodium pump trip would be expected to occur without a reactor trip. This would lead to a Loss of Flow (LOF) CDA which is discussed in detail in the PSAR. If, however, both shutdown systems were postulated to fail electrically, the plant would continue to run on the two ranaining HTS loops. Then the plant response would depend on the control mode being used at the time of the event.

B. Modes of Operation for the Plant Control System (PCS)

Under normal circumstances the plant will always be run in the Supervisory Control Mode which includes the following two fully automatic options:

(1) the reactor-follow mode and (2) the load-follow mode. Only for testing, starter, shutdown, maintenance or repair would the subloops of the PCS or complete manual control be used. The load-follow mode is the one normally used. Figure 1 illustrates the various possible paths of the event depending on the plant control mode.

Figures 2 through 5 identify the main control loops in the PCS. All of the loops shown are used in the Supervisory Control Mode. The next lower level of control is the Reactor Temperature Control Mode. In this mode of operation, 4

the plant is separated into three almost self-contained blocks: reactor, HTS sodium loops, and steam supply system. In Figure 2, setpoints for load, sodium flow, and sodium temperature can be set manually. In Figure 3 the steam temperature feedback is disconnected, and the core exit temperature setpoint is set manually. In Figure 4 the steam pressure feedback is disconnected, and the sodium flow setpoints are set manually. All of the control loops shown in Figure 5 will normally be used in the Reactor Temperature Control Mode. Operation of the plant at this level of subloops will be rare compared to operation under full supervisory control. It will also be rare to go to the next lower level of subloops, the Flux Control Mode (see Figure 3); if the HTS loops also are put into the next lower level, it would be the Pump Speed Control Mode (see Figure 4). lne next lower level of control is manual operation. In this mode of operation, the operator manipulates most of the important plant parameters individually. The operator can run control rod banks in. Runout of control rods is subject to the CRDM rod block circuit which precludes rod withdrawal if too high a withdrawal speed is demanded and/or if abnormalities exist in specific plant variables.

There are two more rod block circuits. One is set to a high flux setpoint of 1.02 and the other is set to a high flux-to-flow setpoint of 1.10.

The flux control mode and the manual operation mode are not intended for use at full power. Even though the flux control mode is not intended for continuous full power operation, it has been selected as a conservative assumption for the analysis of this transient.

C. Thermal-Hydraulic Response of the Plant As explained in Section A above, a loss of power to the primary and

^

intermediate pumps of one HTS loop while the plant is operating in the flux control mode at full power is a conservative basis for the analysis. The thermal-hydraulic response of the plant to this postulated event is given in Section C.1. The response of the plant under other modes of operation is summarized in Sections C.2 and C.3. Sections C.1 through C.3 start with the assumption that there has been a loss of electric power to an entire HTS loop 5

(resulting with the primary and intermediate pumps coasting down) and postulate that both shutdown systems have failed. Therefore Sections C.1 through C.3 differentiate between the plant response based on the postulated accident for different plant control modes.

C.1 Reactor Under Flux Control As the reactor flow decreases and the coolant temperature increases the sum of the reactivity feedbacks would decrease the nuclear power. The reactor flux control system would respond by withdrawing control rods in order to restore the power. In less than 2 sec the flux-to-flow ratio would exceed 1.1 which would activate one of the rod block circuits. The reactivity added at that time by withdrawal of the control rods would be about 0.lc, which would make this case practically identical to the case of Manual Operation (see Figure 1). The sum of the reactivity feedbacks (sodium expansion, doppler broadening and uniform radial core expansion) would continue reducing the nuclear power and as a result, only a slight increase in the temperature rise across the reactor would occur. After about 30 minutes, the thermal reactor power would settle out at 85 percent of its nominal value and the two intact HTS loops would run at somewhat increased temperatures. The flow control system would adjust the individual loop flows to their original values and the reactor flow would stabilize at 67 percent. The current plant model shows that a turbine trip would occur on a low steam pressure signal at about 43 seconds. The steam produced is routed into the condenser and the deaerator.

The proper amount of feedwater flow (85 percent of full flow, based on the new steady state reactor power) would be maintained. The pressure drops across the superheaters would increase by 128 psi to 450 psi. The drum pressure would move from 1859 psi to a steady value of 2041 psi after going through a peak value of 2148 psi at #1.0 minute. This pressure increase would occur because each steam system would be running at 42.5 percent of total design flow (as compared to 33% normally). The two half-size feedpumps would readily accommodate the increased head demand and would run at 85 percent flow at approximately 97 percent of their nominal speed.

6

C.2 Reactor Under Temperature Control In this mode of operation the control scheme in Figure 3 without the steam temperature loop applies. A core exit teaperature loop is tied around the nuclear flux loop which acts as a fast inner loop. The core exit temperature setpoint is 10580F for full power. It has to be set manually for partload when the supervisory control is off. In Figure 2 the turbine control system is operating, but the sodium flow, sodium temperature and steam temperature demand signals are not calculated. In Figure 4 the flow demand signal is manual and the steam pressure feedback loop is disconnected. The controls in Figure 5 are fully used.

On loss of one HTS loop the core exit temperature passes through a short peak (increase of #900F) and returns to its setpoint in less than 2 minutes.

The outlet plenum temperature increases to 10530F and returns to its initial value of 10180F after 2.7 minutes. The reactor power stabilizes after 10 minutes at 74 percent. Each one of the two operating loops runs at 111 percent power which is reflected in slightly lower primary and intermediate cold leg temperatures. The primrry M intermediate loop flows are controlled to their original values by the flow contro; system. The turbine has tripped as it did in the Flux Control case (Section C.1). All the water side parameter v?Nes are close or iilentical to the corresponding values in the Flux Control case. The main di'ference is that for Flux Control the loops run at 128 percent power, whereas for the Temperature Control case they run at 111 percent. Since the supervisor) control loop is not on the plant will stay at 74 percent power until operator action is 1.3 ken.

C.3 Plant Under Supervisory C)ntrol In this mode of operation all af the control loops shown in the simplified diagrams on Figure 2 thru Figore 5 apply. In Figure 2 the supervisory controller provides the demani signals for primary sodium flow, sodium temperature and steam temperature. In Figure 3 the steam temperature trim on the core exit temperature demand signal is active. In Figure 4 the flow 7

The fuel assembly would increase from #13200F,to #15200F after having passed thru a maximum value of 15740F at #15 sec. There would be a margin from boiling of over 2000F since the saturation temperature at the core would be approximately 18200F. Furthermore, it is unlikely that there would be any significant number of fuel pin c*ladding failures. The statistical hot channel cladding midwall hotspot would exceed 16000F (accepted limit in the PSAR, Section 15.1) for 8.7 min., but the peak value will be only 16300F.

Some cladding failures could occur in random locations, but there would be no safety consequences. The failures would be expected to be incoherent because of the temperature distribution and varying properties of individual fuel pins. Release of fission gas would most likely occur slowly and the flow in the coolant channels would not be significantly perturbed. The conclusions above also apply to the radial blanket assemblies where the maximum hot channel coolant temperature is only <15000F and the margin from boiling is about 3000F.

The maximum HTS temperatures are also given in Table 1. It can be seen that the largest temperature increase would be 1360F and would occur in the primary hot leg which would react a maximum temperature of e11580F, An evaluation was made of the coolant boundary integrity for the conditions of Table 1. The criterion used was that the primary stress at the temperatures in Table 1 should be less than the yield stress at the same temperature.

The exit pipes from the superheaters consist of 2-1/4 Cr-1 Mo steel (SA-155).

The maximum temperature of the superheater exit steam pipe would be

1. 10200F.

The minimum (lower tolerance limit) yield stress at 10200F is approximately 19,500 psi. The hoop stress at 1850 psig is 9350 psi so there is at least a factor of two margin based on the minimum yield stress. The limiting temperature was estimated to be over 13000F so there is a margin from failure of e3000F in terms of temperature.

9

The superheater and evaporator shells and tubes will be fabricated from 2-1/4 Cr-1 Mo steel. The tubes of these components would be more highly stressed for the transient postulated. However, the primary stress would not exceed the yield stress until the temperature reached #14000F. Since the maximum temperature during the transient would be #10200F, there is a large margin from failure for the 2-1/4 Cr-1 Mo components.

The remaining components of interest are fabricated of carbon steel. These components are the steam drum, pipe from the drum to the superheater, recirculation piping, and steam pipe from the evaporators to the drum, Comparing predicted stresses and yield stresses as a function of temperature it is estimated that the temperature limit of the steam pipes would be

1. 12000F. Since the maximum temperature of any of the carbon steel components would be less than 6500F there is a very large margin from failure in terms of temperature.

Estimates have been made of the high temperature capabilities of thL primary and intermediate piping, and large margins are available for the postulated event. The estimated temperatures to which the system would remain intact are approximately 16000F for the PHTS and IHTS hot legs and 15500F for the cold legs. From Table 1 it can be seen that the margins from failure for the PHTS and IHTS piping are over 4000F.

Since there is such a small change (<250F) in the evaporator-steam urum loop temperature, the recirculation pumps would continue to operate. The PHTS pump would experience a transient of #1500F above normal temperatures and also appears capable of operating throughout the transient. For the intermediate coolant pump there is an insignificant temperature change of only 200F (IHTS cold leg).

The margins discussed above are summarized in Table 2.

In summary, there appear to be significant margins from structural failures for the loss of one heat transfer loop with failure of both shutdown systems and operation in the flux control mode.

10

TABLE 1 STEADY STATE COOLANT TEMPERATURES (OF) FOR POSTULATED LOSS OF ONE HTS LOOP IN FLUX CONTROL MODE AND FAILURE OF BOTH REACTOR SHUTDOWN SYSTEMS Coolant Initial Peak Final *

1. Primary Hot Leg Sodium 1017 1153 1129

2. Primary Cold Leg Sodium 753 799 790

3. Intermediate Hot Leg Sodium 958 1074 1054

4. Intermediate Cold Leg Sodium 673 693 688

5. Evaporator Exit Steam 628 647 641

6. Drum Exit Steam 625 644 639

7. Superheater Exit Steam 926 1018 1003

8. Fuel Assembly Hot Channel Sodium 1322 1576 1517

9. Radial Blanket Assy. Hot Channel 1273 1505 1455 Sodium

• Af ter 33 min 11

TABLE 2 MARGINS FROM STRUCTURAL FAILURE FOR LOSS OF ONE HTS LOOP IN FLUX CONTROL MODE AND POSTULATED FAILURE OF BOTH SHUT 00WN SYSTEMS Component Margin Primary Piping About 4500F Intermediate Piping About 4000F Superheater Steani Pipe About 3000F Evaporator Steam Pipe About 6000F Superheater and Evaporator About 3000F Shells and Tubes Core F.A.* Statistical Hot Channel About 2000F (to boiling)

R.B.A.* Statistical Hot Channel About 3000F (to boiling)

• F.A. = Fuel Assembly R.B.A. = Radial Blanket Assembly The margins quoted show by how much the peak temperatures resulting from this event would have to be increased in order to decrease the yield stress down to the existing stress.

)

12

A P

I O

R C T F P O M L U

P N

O N r I

T O

I S

e N

C A

T C

K w LO C T o AI T R U AC O 8 P P U A O E 0 N R T R E 5 l A A 8 l MP E

R R E F -

G N

u O E P

E WY 0 I F

, T ET O OB T AS t P

O A T

RER a .

N E l S PU p Y0T lA D 0 A m F

s AE RE u ES S

O I A P P E N S E M S O E C I N W CR T y O T OS E r M C I T K T N N I a

A C C R L R U 8A P C T m OO A i r

TR O O C T T E D 58 P E

A N A R E -

R E R O R Rf 0 e E

N E C E P

E Y T n PW X O WB O O OO U O F O P ' L O P f

OLL N SS F

o S T LU S E s OHF k s L T A

U T

o E ^A L R N O N 8 EP f U E T TI O I

P R o A

R C

A T

C T E H,O s E E L

O R USP C T G( e T

A P E OC R0 c R

M TR D O T

E P 4 7 OR n E

E N O A RY- e P

T OM R R B 0 u O RC E E T A q s O P W f e

S T O R s O

O S S C

A O P n E E N o C

N R C E M UE S QT E 1

E S R MSY ENS N E U e O N T T r TOL SCO I

T O A0A u Y Y I TE R ig SR R R C T T SS E E T L A C C AP F O

S O USP Y N H N WT U iR E R C O P 0 D EREM 0RC VT D O O

T E 4 A CT RN RYB - EE 0UT EO M A T D NG TF N P C R R 0 S U O LA U E E T T l N S P W WAfi E

S - P O O P N R O E sC S N P F C O NP C3C mw

E 4

'" TURBINE TURBINE STEAM 3 PCT / MIN + STEAM LOAD DEMAND +

RATE CONTROL THROTTLE FLOW' TURBINE 3,g LIMIT SYSTEM VALVE

~a l TURBINE LOAD STEAM OATA AT SUPERHEATER _ SODIUM FLOW -

PROFILE 4 _

FLOW _

s . -

SODIUM HDT PRESSURE POWER POWER .

B

, r LEG TEMPERATURE -

CALCULATOR PROFILE TEMPERATURE STEAM SIGN ALS FROM 3 LOOPS 1 TEMPERATURE  ;;; C PROFILE SIGNALS FOR 3 LOOPS Figure 2 Turbine Load Control System and Plant Supervisory Controller

C 8

STEAM TEMP. SODIUM TEMP.

C DEMAND B DEMAND (SEE FIG. 2) (SEE FIG. 2)

I' II

• THERMAL SODIUM STEAM STEAM + CORE EXIT + POWER' LOOPS GENERATORS TEMP. TEMP. REACTOR CONTROLLER CONTROLLER

- -il -ll -Ji w

NUCLEAR FLUX CORE EXIT TEMPERATURE STEAM TEMPERATURE AT TURBINE THROTTLE VALVE Figure 3 Steam Temperature Control

M O

R F

S E EL WP N T t OO l LO 8Ty F L ROL U A MR T H V A EH E

T T SO M

AW E O 4 I S F L

R O

MT AA E R T E S N E

G l

o

+ t r

n o

C Y

R MP E

T er A UO 1 A

I u M0OL I DM s s

R 0 E UP e P S M1 R 0O r E 0 O L E

P T S r N V I L A m V a L

E t e

W W L

T S

D O O T E

E L L O d P

S F F R H

n Y M M a R MP P U E U T

AIU M M 6 T A

1 0 T l MDU O P U

P 0

0 I

0 A o r

I R S S DM P D S E t P Y R Y E MIU M E E R n

U R DU E

T A R A O P P S A S S

o M

I M

E T

N S

P I

D E R

C R I M E P R I P

P U M w P R M o A~ = A E

l T E N

I T

S F

f-R O- r i m

u Y

E L

L

+ d o

RMWO AU S IOR R MDLT I

E y R OF ON r

'MWL L

P S a C

U O O R m 0 L T i a 0 F N r S O P C

M An \ '-

1 4 e

4 r

d u g

i

- F O 8 '-  :

l i

eEa Y

R EE E W L R T

L L

O )

T U O 2

OSS R L

F A E T RET SUlR N I

OA M U

I D

F E

P S0 5 SP 6 ET HRN TPO C

O S 4 RE S 1 PS p l q I C$E

,. G

w PUMP EXIT PRESSURE

-U 100 PSI

+

_[\ +

~(

FEEDPUMP SETPOINT * ~(j ~

+j i +h I>

STEAM DRUM PRESSURE z O 28 IN LEVEL SETPOINT

?  %)b a v n-Q O STEAM DRUM LEVEL 1

g m s

" 3:

+II

+

O j

+ ] j -

FEEDWATER CONTROL STEAM FLOW AT / ./ VALVE SUPERHEATER EXIT _j i _ji VALVE OPENING FEEDFLOW IN LOOP Figure 5 Feedwater Flow Control

1 0

2 8

1 1 6 1

t n

e v

E 4 p o

l 1

o L

e n

O 2 f I

1

)

o S s D s N o L

O C e E h I 0 S

(

t 1 r E o M

I f

T r e

w o

P I

8 r

t o

c a

e R

r I

6 a l

e c

u N

6 4 e r

u ig F

I 2

L - - -

0 0 0 0 0 0 0 0 2 0 8 6 4 2 1 1 2EE !E- e g 5m2

._ $wb

w 1250 b

1200 -

o[ 1150 u

3 E

e G 3 s

1100 1050 I I i l 1000 8 10 12 14 16 13 20 a 2 4 6 -

TIME (MINUTES) t Figure 7 Sodium Temperature at Exit of Core Average Channel for the Loss of One Loop Event

0 2

8 1

p o

o L

6 e 1

n O

f o

s s

4 o 1 L r

f o

l e

n 2

n 1

a h

)

S C E t T o U l f

N I

M e r

' 0 1

(

o E C MI f T o t

i x

E 8

t a

e r

t u

a r

e p

6 m

e T

m i

u I d 4 o S

8 e

r u

I 2 ig

- F O

0 0 0 0 0 0 0 0 6 5 4 3 1 1 1 1

_u.o E E's 3w C$$

8

C 4

1200 1100 HOT LEG C

o_

m 1000 -

e 5

e e

g 900 -

COLD LEG 800 -

t i I I ' '

700 0 2 4 6 8 10 12 14 16 18 20 TIME (MINUTES)

Figure 9 Primary 1101 Leg and Cold Leg Temperatures for Loss of One Loop Esent

O N

- CIS M

C O.e d

_ to L O

O d

o C

e >-W O d

@ M 4$ -

eo Mw s W> 0 S a d

4

> I lE c o w N g U i CC D

'4 -

~

M E Z M oa w -

N M E? >

o o =2 u

L Wo o -Z 3 C

D

>- - E t/j w

M >- .f MW #

W .J L1J O E

' ~2 g t-U H $

1 m O

?

=

8

- . c C

b

=

.9

- , a N

L l O O

g g 8 8 8

, m N -

(visa) aunssaud 1383-80 22

O N

- CD e-C k

W

- 2 ,

C C

J o

C O

_ e" w C

C J

w m

W b

+ o E b s

e o

" W $

3

_E w W

N b

=

.Na I.:.

ta W

N o

c o a m g

Q c o

• $ $ 8 a E R nflSd) 38nSS3Ud 1383-89 23

O N

ctB e

to C

U

-

• a i

e 4

o

_ @ C

- O w

O h

=

4

- N O

C/3 W $

+ o EI t

- O :ls c.

ens D

E ~

C P

O

= >

h n

A 23

~

CD . C.

a O

ti 9

N O

.3 6

N O

O C3 O O O O O O O O O O O CO LC3 W N c N e w w w w (Visd) Hunss3ud 1383-82 24

o N

- So e-

- Y

" =

c U

i

- " 3 o

b Lu H w 3 O a

Lu E

G O" no e C

E E

G e m.=

a

. .M.

L

- nt

- N

=

=

8

=

8 co 8 8, 8

N

,=

(33S/81) 31VU M013 WV31S 1383-83 25