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Simulations of BWR Power Oscillations W/O Scram, Informal Rept
	ML20042D087
Person / Time
Site:	LaSalle
Issue date:	08/04/1988
From:	Cheng H; BROOKHAVEN NATIONAL LABORATORY
To:	; NRC
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SIMULATION OF BWR POWER OSCILLATIONS WITHOUT SCRAM n

H. S Cheng Informal Report August 4, 1988 Final Version Brookhaven National Laboratory

, Department of Nuclear Energy Plant Analyzer Development Grou0p Upton, New York 11973

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s SIMULATION OF BWR POWER OSCILLATIONS WITHOUT SCRAM H. S. Cheng Brookhaven National Laboratory ABSTRACT Postulated failure to scram events following the occurrence of limit cycle power oscillations in a BWR have been simulated with the BNL Engineering Plant Analyzer (EPA, Ref. 1). In case of a total scram failure (with feedwater flow available) following a power oscillation event, the reactor power is predicted by the EPA to experience sustained oscillations between 10% and 500% of full power. In these simulations, the power oscillations reach a limit cycle and do not diverge, due to the inherently negative reactivity feedback.

1.0 INTRODUCTION

The recent LaSalle-2 power oscillation incident has received considerable attention. The consequences of the actual LaSalle-2 event at the plant were not serious since the reactor was shut down automatically, no fuel was damaged and no radiation was released.

However, the incident raised concerns about the thermal-hydraulic instability of BWRs.

These results were generated with the'same Brown's Ferry model used in a previous report (Ref. 2). As in the previous study, since Brown's Ferry is being simulated and not LaSalle-2, and since the flow reduction under natural circulation conditions is being initiated via a coding artifact, these results must be viewed as preliminary. However, they do provide qualitative information.

Power oscillations in a BWR have been simulated previously by the BNL EPA (Ref. 2). Postulated events that are simulated here utill e the EPA to investigate the following concerns: 1) what if the reactor failed to trip due to a scram failure; and 2) is there a possibility of divergent oscillations?

2.0 EVENT DESCRIPTION The postulated failure to scram event simulated herein was initiated using the same plant description and initial conditions as those reported in Ref. 2. These conditions are not identically representative of the state of the LaSalle-2 plant since the input f description used here is an adaptation of a Brown's Ferry simulation, which was readily available.

Since a Brown's Ferry representation is being used, important parameters such as the reactivity feedback coefficients and the l

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neutronic state of the core do not duplicate the LaSalle-2 reactor at the time of the actual event. Furthermore, Brown's Ferry is a l BWR-4 without flow control valves, while LaSalle-2 is a BWR-5 with l flow control valves. In these preliminary studies, the reduction i in the natural circulation flow rate just prior to the onset of growing power oscillations was introduced via a coding artifact.

The pressure drop form loss coefficient at the core outlet was arbitrarily increased by 30%.

Thus, these results must be interpreted as containing qualitative information, rather than containing quantitative results representative of a specific BWR plant. Further studies using more realistic conditions will be conducted in the near future under the BWR stability program.

3.0 SIMULATION RESULTS BWR Power Oscillation Event Without Scram This case is an extension of Case 2 of Ref. 2 with the additional assumption of scram failure. The postulated event sequence is summari::ed in Table 1. The EPA simulated results are shown in Figures 1 to 16.

Figure 1 shows the sustained power oscillations (between 10%

and 500%) after the reactor experienced growing fluctuations and the scram system was disabled. This type of behavior is caused by the effect of a hydrodynamic density wave traveling up the channel and resulting neutronic feedback through the reactivity effect of the voids (Ref. 3). It appears for the case studied here, that these oscillations would continue indefinitely.

The effect of these sustained power oscillations on the fuel elements can be inferred, in part, from the behavior of the minimum critical power ratio (MCPR) as shown in Figure 3. If the value of MCPR tends to 1.0, an approach to boiling transition is indicated.

During the period of the power oscillations, the average MCPR is about 1.6, which is well above the normal range of operating limits for BWRs. However, the MCPR tends to approach unity momentarily at ' times , implying that the fuel integrity might be challenged.

It should be noted that the empirical Hench-Gilles correlation (Ref. 4) employed for the MCPR calculation is probably more uncertain under such transient conditions than under normal operating conditions (i.e., the measurement data base for this correlation does not include these high frequency power oscillations). The system pressure also exhibits i 10% f oscillations during the power oscillations as s'..own in Figure 4.

Although the pressure control system is active during this simulation, these oscillations are within the safety relief valve controller setpoint limits used. The influence of the pressure

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TABLE 1 Secuence of Events for Postulated Scram Failure Case Event / Action Time (m)

1. Steady state at 85% power and 75% flow -5.0

2. Recirculation pumps tripped and FW heaters failed 0.0

3. Reactor power dropped to minimum (28%) 0.08

4. Reactor power recovered to 40% due to heater loss 0.2

5. Core flow. reached natural circulation (29%) 0.5

6. Reactor power reached 60% & " beat" phenomena began 3.0

7. Modulated limit cycle oscillations continued 4.8

8. Core flow reduction started and scram disabled 4.9

9. Enhanced limit cycle oscillations continued as- 5.9 core flow was being reduced

10. Growing oscillations started when the core flow 6.0 was reduced to 20% at 50% power'

.11. Reactor power reached sustained oscillations 7.1 between 10% and 500% of full power

12. These power oscillations continued indefinitely 20.0 and transient was terminated i

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oscillations is correctly included in the calculated power response through the EPA modeling.

In order to see the frequency of the power oscillations, Figure 1 was noomed between 6 and- 9 minutes as shown in Figure 5. The frequency prior to 7 minutes is about 0.29 He (corresponding to a period of 3.4 s) and the frequency after 7 minutes is approximately 0.2 H: (corresponding to a period of 5 s). It is noted that the period of 3.4 seconds is longer than that observed during the LaSalle-2 ' event. Important parameters affectibg the frequency (period) are the drif t velocity correlation and the coolant transit time. A small increase in the drift velocity computed using the EPA- drif t flux model would decrease the calculated period.

Similarly, the occasional flow reversals (as shown in Figure 2),

which are computed after 7 minutes, would increase the coolant transit time, which would increase the period. The zoomed display of the MCPR behavior between 6 and 9 minutes is presented in Figure 6.

The oscillations in the core average centerline fuel temperature and the clad wall temperature of node 14 (location of axial power peak) are shown in Figures 7 and 8, respectively. The corresponding zoomed displays of the temperatures between 6 and 9 minutes are presented in Figures 9 and 10, respectively. The centerline fuel temperature was predicted by the EPA to oscillate between . 1200

F and 2300*F, while the clad wall temperature was oscillating between 546*F and 566*F. The maximum centerline temperature was well below the melting temperature of oxide fuel (approximately 5000 *F). Additionally, as indicated in a previous GE study (Ref 5),

The EPA results can give an indication of the reactivity swings of the reactor core during these power peaks. Figure 11 shows the total reactivity behavior, indicating that as much as 0.9$ of reactivity was added to the core at times during these oscillations. On the other hand, several dollars of negative reactivity are predicted to be added to the core during the power oscillatione. This was primarily due to the negative void.

reactivity as slown in Figure 12, although the Doppler reactivity and the moderator temperature reactivity also provide contributions as shown in Figures 13 and 14, respectively. The large negative void reactivity resulted from vapor generation due to the large power oscillations. This is evident in the core average void behavior as shown in Figure 15. The EPA simulation predicted that the core average void fraction was oscillating between 40% and 76%

during the power oscillations. The core average fuel temperature was also oscillating between 850 *F and 1300*F during the same period as shown in Figure 16.

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4.0 CONCLUSION

S AND RECOMMENDATIONS '

The BNL Engineering Plant Analyzer has been used to cimulate a postulated failure to scram event following the occurrence of growing power oscillations in a BWR, These results were generated with the same Brown's Ferry model used in a previous report (Ref. .'

2). As in the previous study, since Brown's Ferry is being ,

simulated and not LaSalle-2, and since the flow reduction under 4 natural circulation conditions is being initiated via a coding artifact, these results must be viewed as preliminary. However, they do provide qualitative information.

Should the scram system fail during the growing power oscillations simulated here, the reactor power is predicted to undergo sustained oscillations between 10% and 500% of full power.

Based on the results and assumptions of the present analysis it is concluded that: 1) a failure to scram event would not lead to unlimited divergent oscillations because of the inherently negative -

reactivity feedbacks, 2) substantial margin exists between the predicted core average maximum fuel centerline temperature and that of the melting temperature of oxide fuel, and 3) the predicted swings in the fuel cladding temperatures indicate that no significant threat to fuel integrity should occur..

The present EPA calculations indicate that under certain extreme conditions large power / flow oscillations may occur in BWRs.

Additional EPA studies using more realistic conditions will be conducted in the near future under the BWR stability program. In order to determine the sensitivity of these results to the specific i reactor design and operating conditions, additional EPA sensitivity ;

l calculations. will be performed to determine worst-case initial l conditions and results.

5.0 REFERENCES

1. Wulff, W., Cheng, H. S., Lekach, S. V. and Mallen, A. N.

(1984),- "The BWR Plant Analyzer," Final Report, Brookhaven

National Laboratory, NUREG/CR-3943, August 1984.

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2. Cheng, H. S. (1988), " Simulations of the LaSalle-2 Incident with BNL Plant Analyser," Informal Report, Brookhaven National Laboratory, May 1988.

3. March-Leuba, J., Cacuci, D. G., and Perez, R. B., (1986),

" Nonlinear Dynamics and Stability of Boiling Water Reactors:

Part 2 -

Quantitative Analysis," Nuclear Science and Enaineerinc, 11, pp. 124-136, January 1986.

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4. Hench J. E. and Gilles J. C. (1981), " Correlation of Critical Heat ~ Flux Data for Applications to Boiling Water Reactor i Conditions,'" EPRI-NP-1898, Electric Power Research Institute,- l June 1981.

5. " Assessment of BWR Mitigation of ATWS, Volume II," NEDO-24222, ,

. General Electric, Proprietary !

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BWR INSTABILITY SIMULATIONS WITH BNL ENGINEERING PLANT ANALYZER 1

H. S. Cheng and W. Wulff Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 presented on December 6,1988 at U. S. Nuclear Regulatory Commission White Flint Building l Rockville, Maryland

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4 OUTLINE 1 PRIMARY CONCLUSIONS 2 SIMULATION OF LASALLE INSTABILITY 21 KEY PARAMETERS AFFECTING INSTABILITY 22 PARAMETER SELECTION FOR SIMULATION WITH EPA 23 SIMULATION RESULTS 24

SUMMARY

OF SENSITIVITIES TO PARAMETER l-UNCERTAINTY ESTIMATES L

3 COMPARISON OF EPA /RAMONA RESULTS 4

l l- BROOKHAVEN NATIONAL LABORATORY l)l)l A5500ATED UNIVERSITIES, INC.(llll

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9 OUTLINE (CONT.)

4 CONSEQUENCES FROM OPERATOR ACTIONS AND SYSTEM FAILURES 41 RESTORATION OF RECIRCULATION FLOW BEFORE' SCRAM 42 MSIV CLOSURE DURING OSCILLATIONS ;

43 LASALLE EVENT WITH SCRAM FAILURE 5 ATWS EVENTS 51 RCP TRIP, HEATER FAILURE, TT, THREE VARIATIONS 52 RCP TRIP, HEATER FAILURE, MSIV CLOSURE 6

SUMMARY

OF EPA-LASALLE ACTIVITIES TO DATE 7 RECOMMENDATIONS i

BROOKHAVEN NAll0NAL LABORATORY l} g)l :

ASSOCIATED UNIVERSITIES, INC.(Illl

inAcazauuwwa rmumme;a r. 2m;tamrzu1:wamusu amsm aeu te,3 1 PRIMARY CONCLUSIONS SIMULATION WITH EPA SHOWS THAT:

11 LASALLE B.E. CONDITIONS LEAD TO LIMIT-CYCLE OSCILLATIONS 12 LASALLE CONDITIONS WITHIN UNCERTAINTY ENVELOPE PRODUCE OSCILLATIONS LEADING TO AUTOMATIC SCRAM 1 3- THERM 0 HYDRAULIC INSTABILITY AT LASALLE IS CAUSED BY THE COMBINATION OF RADIAL POWER PEAKING PLUS FLOW REDUCTION DUE-TO RCP TRIP, PLUS FW TEMPERATURE REDUCTION DUE TO FW HEATER FAILURE 14 . AMPLITUDE OF. POWER OSCILLATIONS REMAINS-BOUNDED EVEN AFTER SCRAM FAILURE (10 TO 500% OF FULL POWER)

CLAD TEMPERATURES OSCILLATE LESS THAN 260 F BROOKHAVEN NATIONAL LABORATORY l} l)l A5500ATED UNIVERSITIES, INC(llll

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1 PRIMARY CONCLUSIONS (CONT.)

15 SLOW RESTORATION OF RECIRCULATION FLOW RATE STABLIIZES POWER OSCILLATIONS UNDER ALL CONDITIONS 16 REACTOR SHOULD BE-SCRAMMED WHEN OSCILLATIONS OCCUR

(+ UNCERTAINTY OF TRANSITION ~ FROM LIMIT-CYCLE TO RAPIDLY GROWING OSCILLATIONS)

(EPA CONFIRMS NRC BULLETIN) 17 PEAKING FACTORS AND 11CPR SHOULD BE MONITORED AND CONTROLLED

(+ STRONG IMPACT ON INSTABILITY HAZARD AFTER COMPLETED RCP TRIP)

BROOKHAVEN NATIONAL LABORATORY l}lgl j AS$00ATED UNIVERSITIES, INC.(llll

2 SIMULATION OF LASALLE INSTABILITY 21 KEY PARAMETERS AFFECTING REACTOR INSTABILITY 211 SIGNIFICANCE OF PARAMETER SELECTION ISSUE OF LASALLE + PLANT-SPECIFIC PARAMETERS REFERENCE TO STAR TREK DATA + PLANT SPECIFIC

- QUANTITATIVE ANALYSIS NEEDED TO IDENTIFY PRIMARY CAUSES OF INSTABILITY AMONG KNUWN CAUSES BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(Blll

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21 KEY PARAMETERS AFFECTING REACTOR INSTABILITY (CONT.) i 212 THERM 0 HYDRAULIC PARAMETRS AFFECTED SIMILARITY OBSERVED GROUP (S) EFFECTS POWER / WALL HEAT FLUX Npeg (D;) 2-8 TIME DELAY INSTABILITY CORE FLOW y,Npg, NRE # ' '

Wu STABILITY SUBC00 LING W 1-8 TIME DELAY

_lWSTABILITY (POWER)

DRIFT VELOCITY W FREQUENCY D

POWER INSTABILITY INLET FORM LOSS Kg STABILITY EXIT FORM LOSS Kg EMW ,

2 Hg p, (K/pcR ) FUEL FREQUENCY INSTABILITY BROOKHAVEN Nail 0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(1lll 1

21 KEY PARAMETERS AFFECTING REACTOR INSTABILITY (CONT.)

213 FISSION POWER PARAMETERS AFFECTED SIMILARITY OBSERVED GROUP (S) EFFECTS I

Vlil0 FEEDBACK POWER ,

INSTABILITY DIRECT GAMMA HEATING POWER INSTABILITY POWER PEAKING FACTORS RADIAL Fa MCPR AX1AL FZ INSTABILITY l

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SIMILARITY GROUPS STILL NOT KNOWN.

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l BROOKHAVEN NATIONAL LABORATORY l}lg A5500ATED UNIVER5lilES, INC.(I(lI i

1 VEND 0190?EA1Y 3FORVATON EETD 2 SIMULATION OF LASALLE INSTABILITY (CONT.)

22 PARAMETER SELECTION FOR EPA SIMULATION OF LASALLE 3-9-1988 EVENT EPA EPA APRIL 1988 SIMUL. NOV. 1988 SIMUL.

POWER PEAKING 10 13 VOID FEEDBACK -278p/% -278[/I i t'

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BROOKHAVEN NATIONAL LABORATORY l)l)l ASSOCIATED UNIVERSITIES, INC,(1lll

202 PARAMETER SELECTION (CONTO)

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c i

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t.u ._

m ,

g ,,

Lu 5--

388.0 i i i i i i i 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 80 '

TIME (MIN)

BNL P1ani Ana1yzer 81-DEC-88 13:28 t

ii i

l REACTOR P06ER 1.Z ;,

y m g,g .

w 2

o 9.8-'

9. l e

9.2-. .. .

i 9.6 i i i i i i i i 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 6.0 9.0 i TIME (HIN)

BNL Plant Analyr.er 81-DEC-98 13:29 !

REACTOR POWER 1.2 m 1,g..

w <

2 ,,

~

8.8-l

,f w 8O.IhIIlll!/I kilfI I j{ll kfIl -

E

,,J L lbOdbhbhpddh),jnhebj( i i e ,,

9.2-.

[ -

m ,,

B.8 ; ; i 6.0 6.5 7.0 7.5 8.0 TIME (HIN)

BNL Plant Analyr.or 81-DEC-88 13:28

MIXTURE FLOWRATE AT CORE INLET -

g,. . .

n r

8.8 .

~

J ,,

c .

'~

~

6.P '

ca

4.8_. --

X e,n- .c=.g.:M.yvey-m = .*r,!.Ctet?-h

'~

2 "

2.8' k o 8

.J 9.8 i i i i i i i- i 0.0 1.0 2.0 3.0 40 5.0 6.0 70 6.0 9.0 TINE (MIN)

BNL Plant Analyser 81-DEC-98 13:28 i MIXTURE FLONRATE AT CORE INLET 4.0 r .

N 8"

3.5_. --

.-J g ,,

m ,,

,p .,

3.8 ' j j qI ff .

x 2.5-h )

(A M b h .i ,

~

z -

O A

2.0 i i i 6.0 6.5 70 T.5 8.0 TIME (MIN)

BML Plant Analyzer 81-DEC-88 13:28

I CORE-AYERASE VOID FRACTION g,. .

. n l w 98.8_. ..

l z 4.

l w

e o

a 46.8-w I ,

y; n. r. 2

.a m,_,,,. i, y,g n w. , - - + , _-_

c:a o 42.8-:

l 48.8 i ; ; ; ; ; ; ;-

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 60 90 TIME (MIN)

BNL Plant Analyzer 81-DEC-9813:28 MININUM CRITICAL POER RATIO 4.8 l

3.8_. .-

i m ,

,, +

o 2. 8_'. ..

s' .w

,-, :.y. -;;;;, m .. - - - -

_a....._an.,

l t

i 1.8- 7 8.8 ; ; ; ; ; ; ; i 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TIME (MIN)

BNL Plant Analyzer 81-DEC-88 13:28 ,

CLAD WALL TEMPERATURE N0DE 14 7.0 ; '

, l i

q

- 6 . 5',,' ,

O 6.8 , t c3 , ,

'* 5.5y ,,

x 5.8 . .-

m

4.5-; .-

w '

w l

4.0 i i i i i i i i 0.0 1.0 2.0 3.0 40 5.0 6.0 7.0 6.0 9.0 TIME (MIN)

IML Plant Analyzer 81-DEC-8813:28 l

FUEL TEMP. AT CENTERLINE NODE 14 -

28.0 -

n ,

u.

O '-

15. 8~ '

G '

%g "

--4 - . . . . . . . , - , , ,

M 18.8_. ..

m l C 1 w b

1 5.0 ; ; ; ; , ; ; ;

0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TIME (MIN)

BNL Plant Analyzer 81-DEC-88 13:28

l i

COLLAPSED LIQUID LEVEL 5.0 l

1 I

n 2.5_ _- _ _ _ _ . . _ _ _

w

] ,

\

i 8.8_. ._ l

.-4 <

j W

> ( l 1

W ' '

a -2.5~ 7 1

-5.0 ; ; ; ; ; ; ; ;

0.0 1.0 2.0 30 4.0 5.0 6.0 7.0 6.0 9.0 ,

1 TIME (NIN) '

BML Plant Analyser 81-DEC-98 13:28 SYSTEN PRESSURE 11.0 l

n e

~

18.5_. ._

M fa.e se ,, ,,

18.8 '

a ._

O A -

^::

9.5- k?

m '

(%

9.8 i i i ; i t i i 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TIME (MIN)

BNL Plant Analyser 81-DEC-88 13:28 N

- . , . - . - . . ,-n e -e -

4 9

SUMMARY

OF SENSITIVITY STUDIES EFFECT PERIOD MEAN MEAN STABILITY PARAMETER OF OSC. POWER FLOW i WALL SHEAR (+86%) -0 01 S -5 11' -3 4%* DESTABILIZING (HTFS TO M-N-J) l ROD RADIUS (+111) 40 31 S +9% +0 9% STABILIZING GAP COND. (+46%) -0 07 S -1 5% -0 08% DESTABILIZING RADIAL Fo (+20%) -0 28 S -2 2% -1 3% DESTABillZING I l

AXIAL Fo (+50%) -0 28 S -72 -2 6% DESTABILIZING-VOID COEF. (+34%) +0 36 S +6% -0 46% STABILIZING (LESS NEGATIVE) l LOSS COEF. KE (+20%) -0 09 S 0 0% -0 5% DESTABILIZING K1 STABILIZING TFW RED. (+40%) +0 01 S +2 3% +0 26% DESTABILIZING (SUBC00 LING) l DIR. HEATING (+63%) -0 17 S -3 1% -1 3% DESTABILIZING

< VGJ > (88%) -0 08 S +4 0% +1 21 DESTABILIZING

' PERCENT CHANGES ARE GIVEN IN TERMS OF CORRESPONDING VALUES AT FULL POWER BROOKHAVEN NAll0NAL LABORATORY l)l)l A5500ATED UNIVERSITIES, INC.(llll i

. i 3 COMPARISON OF EPA WITH RAMONA RESULTS 31 COMPARIS0N OF KEY SIMULATION PARAMETERS ;

PARAMETER VALUE IN PARAMETER RAMONA EPA POWER PEAKING 1 9/1 3 1 9/1 3 (VARIABLE POWER SHAPE)

VOID FEEDBACK -229p/1 -27 8 /% ,

GAMMA HEAllHG l4% g% '

FUEL Kp MATPRO MATPRO l

57 5 7 KW/(M2g)

HP G

LOSS COEFF.*

INLET EXIT 2-0 WALL SHEAR BLASIUS/BECKER M-N-J l RELATIVE VELOCITY B-J SLIP DRIFT l FW TEMPERATURE DROP 90 F 50 F l

' PROPRIETARY DATA -

l

/

BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC.(llll

. t man y -

y a,

. G, 3 COMPARIS0N OF EPA WITH RAMOWA RESULTS (CONT.) ; ',

32 COMPARISON OF KEY RESULTS RAMONA EPA POWER

46% 48%

CORE FLOW

34% 29 5%

PERIOD OF OSCIL. 2 60 S 2 54 S POWER At1FLITUDE LIMIT CYCLE (MAX.) iSI *7 5%

' VALUES AFTER RCP TRIP

/

BROOKHAVEN NAll0NAL LABORATORY l} l)l A5500ATED UNIVERSITIES, INC.(llll

I BROWNS FERRY INSTABILITY ROBLYSIS CORE FLOW RATE o

y 12.0 CORE OUTLET

--CORE lia.ET 10.0 -

1 i

- 9.0 -

v)

, N C.9 M

- 8.0 -

ta3 H

(C Q: 7.0 -

O

]

L 6.0 -

l 5.0 -

4'0 -

l ; _w__. _ _ _ - . .-cc. J.~,ANWWAMi#diNM e

i 3.0 , , , , , , ., , ,

I 0.0 25.0 50.0 75.0 100.0 125.0 150.0 175.0 200.0 225.0 250.0 TIME IS) 4" -

% ot 1 + w - - - _ . - _ - _- __ - -___----__t<w-_ ___ _ _ _ _-___a-_1_____--___m an

i o :

-=

z o u _ m'

~

[d W E o

u -

o

[- 7 ~

.R '

~$

E %.. ., '

c i

p .. . o.

He 4% -S

-w .s* -

)2

-o on n.

D -

v C . o -) '

sa m> -N w z-

-s c- z C H b) ew o x c:

u - ~8' u.

m z

2 a e.

x R m

o

'8

.a

~M M o o

i T . . . ,

9 9 5 9 9 9 n

- o o o o o o d

. 83M0d 3A11U138

+

i 4 CONSEQUENCES FROM OPERAIDR--ACTIONS AND SYSTEMS _ FAILURES .

41 RESTORATION OF RC FLOW ;

i 42 RESTORAT(DN OF RC FLOW AND HSly CLOStlRE 43 LASALLE AND SCRAN FAILURE l

l BROOKHAVEN Nail 0NAL LABORATORY l} l)l

1. ASSOCIATED UNIVERSITIES, INC.(llll n

4

i 1

l 1

> e i

l i

)

l 4

1 84 1 RES10 R U 011 0F RC FLOW.

l 1

BROOKHAVEN Nail 0NAL LABORATORY l)l)l

~

l A5500ATED UNIVERSITIES, INC,(lIll

a 20 ;

REACTOR POLER 1,2-;, . .e

- . ; HIJTURE K0$AT.E of CQRE. INLET _ is,g n

,, I N 1, g.:: .. N w - -

3,3 pa 2 :: m w ;. J .-4 o 3,g '. . E A -

- 6.0 -

- il [ 4,3 3

.-.....m.. ,. ., x F4 0.2- 2 W

, o A :' s4 g,g h.

g,g0.0 ;;;;t 0j ; . 2.0

; j ; 3.0

; ; j 4.0

; ; ; 5.0 j ; ; ; 6.0

j ; ; 7.0

w-H4 8.0, j 9.0

j;;;;i TINE (HIN) -

BNL Flant Analn er 89-MAY8818:04 FJgure 12. Reactor Power and Core Flow Response cf Case 3 PEA 0709 P0i'fR

1. 2.: .- -.4. : ; -.4-. - !

4-~.-++ p N 1, g.:; :'.

W -

2 :, :.

O g,g ,,

g ,,

06< j l w g,4 , Ji .

W el 0.2- --

W .. \ ,

M .. . . . . . :

g,g .3 ,

3 .3 3 5.0 5.5 6.0 6.5 7.0 7.5 TlHE (HIN)

EHL Plant Anainer 99 NAY 88 18l84 Figure 13. Zoomed Display of the Power Oscillation of Case 3 ,

o* o 2? l t

RECIRCULATION PUMP SPEED :

28.8 i . .. . ., e, . . i

' l, 15.4'

O .

O ,

4 .

i 10.0 ~ l M ,,

E ,,

i 2 5.9T :7 g,g ; ; ; ; j ; ; ; ; j ; ; . ; j ; ; ; ; j ; ; i ; j ; ; ; ;1 . . g . . . . g . . . . -

0.0 1.02.0 3.0 4.0 5.0 6.0 7.0 0.0 9.0 TlHE (H!N)

EHL Flant Analyter 99-MAY88la:H Figure 16. Heciren)ation Pump Speed Rer.ponse of Case 3 SYSTEM PRESSURE L 12.8 i

e. i mu i

!. . %%. g.

n.

- 18.8A.

g . _ _ _ _ . _ _ _

8.0 '

g g ..

H ,,

X 6.9T ;

w h m ,,

, p

;; j ;;;;j ;;;; j ;;;; j ;;;; j ;; . ;;;;;; j ;;;; j ;;;;,

4,g 0.0 1.02.0 3.04.0 5.0 6.0 7.0 8.0 9.0 TlHE (HIN) "

BNL Flant Analyzer l 99MAY-8818:04 ,

Figure 17. System Pressure Response of Case 3

T' , 4 jr:n '-

4 .

S 3 ft ti . 2 RESTORATION OF RCP FLOW AND MSIV CLOSURE i

)

l BROOKHAVEN NAll0NAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC.(1lll

, _ + . . . . . . .

. ;;yed % ;:s y

, m .

23 REACTOR POE R

1. 2 , MlUyRE FLO)! RATE. AT CORE.IN.I.ET. is,g n

. .. I N 1,g; . e N W ::

8,8 so

'2 O g,g--

' g 14 W

E' A -- 6. . -

N h h,k i 4.8 x=e. : . ::: n :

, x W g.2 :'

. 1 W , ,o m :: W g,g j ;; ; ; j ; ;; , j ; ; ;, j ;; ;; j ;; ; ; t ; ;; ; j ;-; ;; j . ; , ;; g,g ti, ,

0.0 , ; ; ;1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TlHE (HIN) ,

BNL Flant Analner 89MAY8819:27 rigure 18, R< ac t o: eower enc' core riow nesponce or case 4 l REACTOR POER

!.2 4- _ m 4_, ,_ u . -. , 3 ;_

q .

M 1,g; , .,

W :: ::

2 "

i :-

O g,g- i j :

4 -

5 L :

E,,hMOM4M$h w .

(

< i , i.

W g

0. 2 ,- ,,

A :

g,g' ,. ,. ,,g . . . . ,. ,. .,3, . ,. ,. ,. . . ,. ,. ,. . . . ., ,.

3 3

, , e, 5.0 5.5 6.0 6.5 7.0 7.5 TlHE (HIN)

EHL Flant Analner 99MAY8818:27 l

Figure 19. Zoomed Display of the Power Oscillation of Case 4 i

e

- - . - - - - , - a_ . .

-___-_.____.______.____.__m_______-__

.o .

1 24 TOTAL REACTIVITY 1.9 e i e ; .

A ,. .. i a ,,3 H . i ,

f l- ..

N ?

o

< 9. 5 ,, ,,

tu m , ,.

,},g .

j ; . ; ; j ; ; ; ; t ; ; ; ; ;-t ; -i ; (-

5.0 5.5 6.0 6.5 7.0 7.5 TlHE (HIN)

BNL Flant Anily.er 09MAY8810'27 Figure 20. Total Reactivity Behavior of Case 4 i g

FLOL1 RATE AT STEAMLINE ENTRANCE N . EpJ +-+4-,~ 4 . -4. w % # 6 + ' .

,i ,i r 15.B . ..

N CQ . .

r 10.0 ;_

n ,, ,,

"~~

2 b) '

o A 5. 9 , ..

Li. ;, ;,

j . , j ; ; . ; j ; ; . ; j ; ; . ;t+s+t ; ; ; ; f 4 . .T 7N-g,g0.0 ;;;;1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TlHE (HIN)

BNL Plant Analyzer 99 MAY 88 10:27 Figure 21. Steam Flow Response of Case 4

i

[

43 LASALLE WITH SCRAM FAllyRE (i) POWER AND CORE FLOW t

(ii) MAXIMUM FUEL TEMPERATURE +

(iii) MAXIMUM Cl. ADDING TEMPERATURE (iv) MIXTURE LEVEL IN D.C. ;

i i

BROOKHAVEN NATIONAL LABORATORY l)l A5500ATED UNIVERSITIES, INC.(llll -

1 l

REACTOR P064ER 5.

4.8 w

I o

3.

w

2.8 s-e .

" 1.8 l w l

m i

~

S.

0.0 5.0 10 0 15.0 20.0 TINE (MIH)

RNL Plant Analper 03-DEC-98 12:23

!' REACTOR P06 Fin 5.0 - :- -- - - - * -

l l ,,

l

4.8 W i . ! J,

x < I g i p o

l

3.8~~ l '~

w l

> 2. 8-l .-

~

s- <

K

" 1.8-'

w 1 9.8

uvbLJuvubuubbLuvuu!

i i i i 7.0 7.2 74 76 7.8 S.0 TIME (MIN)

BML Plant Analper 83-DEC-88 12:23

e- 1

)

. i I

REACTOR POER !

l 5.0 ,

4.8 1 w , i 2 , <,

l o <

l 3.8-: 1 1 w

2.8_.

1 p , ,

c

" 1. 8""

w B.8 d bVObV hL)UObV i i i i O.

C 19.0 19.2 19.4 19 6 19.8 20.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 12:23 a

REACTOR F0ER 5.C

,, l .

w 't . 8-- I --

, 3 -

r 2 ,,

q ,

o 3.8-: --

w ,

2.8_.

p ,

e .

t

\

1.0, w

m

..: ; , i 19.0 19.1 19.2 19.3 19.4 TIME (MIN)

BNL Plant Ana19zer 83-DEC-88 12:23

t MIXTURE FLONRATE AT CORE INLET 18.0 r ,,

N se 8.8-" -

a ,

c

~

6.0" ,

( c3 9.6 0.0 5.0 10.0 15.0 20.0 TIME (MIN)

_ BHL Plant Analyzer 63-DEC-88 12:23 e

MIXTURE FLCHRATE AT CORE li4LET

[

10.6

=

{= %

w. ' "

.x "

- r2 '

t J e g,,- . .-

qff)h(1fhrH"]fifffhh*Ff x e.e_.

w I

,i i l i

,w ._

r ,

o

-5.6 ; ; ;

6.0 6.5 7.0 7.5 8.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-98 12:23 i

MIXTURE FLONRATE AT CORE INLET 18.0 n .

r N

8.8 ..

T a . t c

~ -

6.8-:

o ,

4.8-:n piqfp)Pfri f ij'tf jitqfp rtf n[

x r

2.8~ ,

[ "I i iliiI l I ,

Ii 1 -}i[L l 9i If f I;:

... ; i i 18.0 18.5 19.0 19.5 20.0 ,

TINE (MIN)

BNL Plant Ao lyzer 83-DEC-88 12:23 MIXTURE FLONRATE AT CORE INLET n

10.0 r ,

I N

8.8_. ._

l .

1

^

l E l s.e- - --

a .. .

4.8-ff 5

5 f f e?t}rpqs*'"'frf ffffffp-

= 2.e-a ,,

j iii i i L II/I} I !j dLlhlI,f ki b .

L 0.0 ; i i 8.0 8.5 90 9.5 10.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 12:23

FUEL TEMP. AT CENTERLINE NODE 14 g,;

n u.

~ '

20.O .

c3 c3 15.8_

x

" 18.8-tu 5.-

5.0 ; ; i 4 0.0 5.0 10.0 15.0 20 0 TIME (MIN)

BML Plant Analyser 83-DEC-88 12:23 FUEL TEMP. AT CENTERLINE NODE 14 25.0

^ '

t 1 7

u '

I I

(

~ 28.8_. l ._

l

! l c3 !

l 15.8 j .

{ I Iili I

x

18.8-c

J ; ,

s-5.5 i ;

6.0 7.0 0.0 9.0 TIME (MIN)

BNL Plant Analyzer 03-DEC-88 12:23

LL; . .. aua u.u aax.:.m -.-. .... . .

- - -....-----n.-----.-----~-

CLAD WALL TEMPERATURE NODE 14 6.5--

n

'_ 5.B_. ..

cs 5.6 .'

5.4_. --

w r 5.2-

' '~

uJ s_. ..

5. i ; ;

0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BNL Plant Analper 83-DEC-88 12:23 CLAD NALL TEMPERATURE NODE 14 5 . 8- -

n u-

~

5.?_. --

O ,

ca . , ,g .

s .

k h a 5.5-'

r -

Lu 5-5.4 i i 6.0 70 8.0 9.0 TIME (HIN)

BNL Plant Analyzer 83-DEC-88 12:23 i

MINIMUM CRITICAL POER RATIO 4.0 3.8_. ._

m o 2.B_. ._

c

~

1.5-9.8 i i 0.0 5.0 10 0 15.0 TINE (MIN)

BHL Plant Analysc.r 03-DEC-88 12:23 HINIMUM CRITICAL POWR RATIC 4.0 3.B_. ._

m -

O 2*0_. ._

r 1.8~

't* ,

l} ff ,,

~

9.0 i i ,

6.0 7.0 8.0 9.0 TIME (MIN)

BML Plant Analyzer 83-DEC-88 12:23

CORE-AVERASE YOID FRACTION 70.0 m 1 s-.

2: W.B .

w a -

58.8 - .

4 w 3 48.8-' ' ' i V U o

x 38.8 ;

0.0 5.0 10.0 15.0 TINE (MIN)

BNL Plant Analyzer 83-DEC-88 12:23 CORE-AVERASE YOID FRACTION 78.8

^ .

68.B_. -

g { ,_

o .

{ ! I

- 48.8-;

ell ,

o ,

38.8 ; .

6.0 7.0 ao 9.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 12:23

er.. .

TOTAL REACTIVITY 1.C lMAAAAA a.e-g

>- -1.8-s-.

l

-2.B_ ,

-3. 8_.

o e

tu -4.&-

~

l

-5.0 i i i 0.0 5.0 10.0 15.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 12:23 TOTAL REACTIVITY

1. 8=

n l m B.8- lJI ,

- l -

>- -1.0-'

t-- I

- -2.0,; _

s_. -3.O_'. . _

o l L1J -4.8-'- ! -

m

-5 . 0 - 'i i 6.0 7.0 8.0 9.0 TIME (MIN)

BML Plant Analyzer 83-DEC-88 12:23

5 ATW EVENTS (NRR SELECTION)-

51 ATWS WITH RCP TRIP, HEATER FAILURE AND:

511 TURBINE-TRIP W/0 BYPASS O!ITH POWER OSCILLATIONS DEVELOPED) 512 TURBlWE TRIP WITH BYPASS. AND FWP TRIP 513 TURBINE TRIP WITH BYPASS AND NO FWP TRIP 52 ATWS WITH RCP TRIP HEATER FAILURE MSly CLOSURE (WITH POWER OSCILLATIONS DEVELOPED) i

v- ,;

l l

1 S.1 1 TURBINE TRIP W/0 BYPASS DURING POWER OSCILLATIONS l

l l-1 t

BROOKHAVEN NATIONAL LABORATORY l} lg g A5500ATED UNIVERSITIES, INC(I(Il

RECIRCULATION FLOWRATE 7 38,0 ' '

n 25.8-: --

z s ,

na 29.8-a r 15.8_: ,_

z 18.8_. ._

o l

u. 5.8 : ^

_91 :: - ?- 1_ _== ._.

B.0 i i i i. . ;

0.0 2.0 4.0 6.0 B.O 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 14:49 RECIRCULATION PUMP SPEED s

15.0 c3 C ~~

19.8-w ,

X i

E 5.8-- .-

a , ,

l 0.0 i ; i ;

L 0.0 2.0 4.0 6.0 L.D 10.0 12.0 TIME (MIN) 5tiL Plant Analyzer 83-DEC-88 14:49 I i

l' L.

FEEDWATER FLOWRATE 15.0

^

l ,

g _

% l

" 18.8-'

'I

.-.8 , ,, 1

, 5.6_. ._

4 o -

a , e l

-0.8 i i i i i L 00 2.0 40 6.0 8.0 10.0 12.0 TIME (NIN)

BHL Plant Analper 83-DEC-88 14:49 FEEDWATER TEMPERATURE

. 580.C l ^

! L

~ 458.B_. ._

u.J , ,,

m ,

l "g 480.8_'. __

c m

uJ

s-350.B-', -

uJ g

300.0 i i i i i l

l 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN) l l BNL Plant Analper 83-DEC-88 14:49 l

l l

i

^

REACTOR POWER 5.0 ,

4.8_.

2 o

3.8 ,:

w

2.8_.

s-

_ _ _ - :4 0.0 i i i i i i 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 63-DEC-88 14:49 REACTOR POWER 5.6 4.B_. .

s -

o 3.8-:

w .

2.8-'. --

s-m U (y((f(k(( h LLlV bl bO

..e i ; ;

6.0 6.5 7.0 7.5 8.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 14:49

b HIXTURE FLONRATE AT CORE INLET 18.0 n ,, ,,

z N

8.0_. ._

A ,

J g . .

~

6.0" ' '"

ca , ,

2.

' 0.B i i ; ;

i O.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BML Plant Analyzer 83-DEC-88 14:49 MININUM CRITICAL POWER RATIO 4.0 3.8.. ..

m ,,

. 4, a 2.8_', _

r 1.5- Ij 0.0 i i i i ;

0.0 2.0 4.0 6.0 0.0 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 14:49 l

1 l

FUEL TEMP. AT CENTERLINE NODE 14 25.0 n

u.

~ 29.B_. .

o i ca

15.0_, ,_ ,

x

" 18.8-'

r <

ua 5--

5.0 ; ; ; ; ;

0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BML Plant Analyzer 83-DEC-88 14:49 CLAD NALL TEMPERATURE NODE 14 l 6 . 8-- --

n S.8_.

=ain C- -

l:

!- Q (

l C J L

a 5.6-( - _

l .

5.4J. .-

m n 5.Z~

~

uJ s_. .

5.0 ; i ; i i 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 14:49 1

SYSTEM PRESSURE 14.0

13.8- ' --

N sa ' -

12.8-a 11. 8_- . ..

o

18.8_( g ._

x -

9.8_. .-

en 8.0 i i ; ;

O.0 2.0 4.0 6.0 8.0 10.0 12.0 T!HE (MIN)

BNL Plant Analyzer 83-DEC-88 14:49 COLLAPSED LIQUID LEVEL MIXTURE LEVEL 5.0 . . . , , 5,8 4.8_. ._ 4.8 n ,

n s- s--

u_- <

u_

3.8- -

- 3.9 J _

- . - _ _ __, l ji.{;Q .

w 2.8-- . _ 2.8 m W w a ._s

1. 8~ ' 1,9 0.0 ; ; ; ; ; 8.8 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-98 14:49 1

1

SUPPRESSION POOL TEMPERATURE 258.0 ,,

n

' 288.8_.

\

w 3 158.8 ..

s-iM.8_. ._

s:=

g , ,

CL r 50.8-'

w

'~

5-0.8 i i i i i 0.0 2.0 40 6.0 8.0 10.0 12.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 14:49 i

SUPPRESSION POOL LEVEL 15.0 n 18.B_. ._

.x , ,

5.B_. __

.--a w

w

_i 8.8-' '-

-5.8 i ; i i ;

0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BHL Plant Analyzer 03-DEC-88 14:49

6 DRY ELL PRESSURE ETWELL PRESSURE 17.8 . . . . ,

17.8~

n n 16.0_, , _ 16.8

M 64

~

15.8-; -

- 15.8 ~

w ^

, w

m14.8-. + . 14.8a= "

M '

64 M to W 13. 8~ '

13.8cx*

cm ct. --

m 12.8 i i i ; ; .

12.8 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 14:49 DRYELL TEMPERATURE WTELL TEMPERATURE 128.8 . . . . ,

128.8 188.B_. ._ 188.8 n , , -n o o

" 88.8-f -

-M8~

. , p , .

r 68.8-7 . .' 68.8s*

W - -

w E-- + s-.

48.8~ ~ 48.8 b

28.8 i ; i ; 28.8 0.0 2.0 4.0 6.0 8.0 10.0 12.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 14:49

i l

c

=

512 TURBINE TRIP WITH BYPASS AND WITH FWP TRIP DURING POWER OSCILLATIONS am BROOKHAVEN NATIONAL LABORATORY l} gyl ASSOCIATED UNIVERSITIES, INC.(1 til

l REACTOR POE R 5.o i 1

w 4.8-. ._

m .

I o .

3.8-: l

" l

2.8-.

s-- -

t

" 1.6

w JM2!h'!l

'N!!?ll& ,

8.0 , . ; ; ; ;

0.0 5.0 10.0 15.0 20.0 25.0 TINE (MIN)

ML Plant Analyzer 83-DEC-88 15:55 REACTOR POER 5.C y 4.8_. ._

s . :

o 3.8-: --

w .

2.8_.

s g \-

a 1.6-1 -

w JUdbdbkl0LdghggL,yt ,

8.a ; ; ;

6.0 6.5 7.0 7.5 0.0 TINE (HIN)

ML Plant Analyzer 83-DEC-88 15:55

l i

l SYSTEM PRESSURE 14.0 - '+ --

1 m \

c 1 3

~ 12.8- '

'~

I ,t: ,

18.O_ .

J ._

x -

M m

8.8- i i i i .

0.0 5.0 10.0 15.0 20.0 25.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-98 15:55 SYSTEM PRESSURE 14.0 M

- 12.8-' ,,

10.0_. ._

. = . _ _ . . . . y;;,i X

m A

8.8 i ; i i -

5.0 6.0 7.0 8.0 9.0 10.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 15:55

l REACTOR POWR 5.0 l

4.8 w

I

x 1 o ,

m 3,g_ . .-

w

> 2.6_.

p ,,

e a i,g_. 1 .-

' UUbd DVb VdVbsbVVvLvV)

Y r

... i ; ;

7.0 7.5 0.0 8.5 9.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 15:55 REACTOR POWR 1.6 j

8.B_.

W ,,

g ,,

o m 9.6- ' "-

w , ,,

8.4_. ..

- 1 l

h h \ \

l l i0L

~

8.2" i hL0 ;L l;-

w I ii ,

m -

0.0 i .

i i i 10.0 11.0 12.0 13.0 14.0 15.0 TINE (MIN)

BNL Plant Analyzer 03-DEC-88 15:55

t 4 l

l MIXTURE FLOWRATE AT CORE INLET 18.0 n

3- .

N

, 8.8_. ..

a ,,

n:

~ --

6.8-' i o

4.8_ -

x --

y 2. 0,,,

h ,;;;r-i ata m*>i,,man f ,

g, _. , .n n u ,

.c .... ,. , - , . E r r

' 9.0 i i i 0.0 5.0 10.0 15.0 20.0 TIME (HIN)

BNL Plant Analyzer 03-DEC-88 15:55 HININUM CRITICAL POWER RATIO 4.6 - ~ - - - -

3.B.. ..

m I "

l 'llp l

M 2.B_'. l <

g l l l[i ] i 11 [ _

s: ,

1.0-0.0 i ; i 0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-00 15:55

J

. I e

FUEL TEMP. AT CENTERLINE NODE 14 -

l 25.0 '

^ l La_ .

' l

'~

29.6 . .-

l l

C3 .

.j c) <-

15.8_ ._

I x

1 l

" 18. 8~

~

g L w ,

i s--

L l S.6 i i i 3

0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BHL Plant Anelyzer ,

83-DEC-88 15:55 CLAD WALL TEMPERATURE NODE 14 6.8 7 --

n ,

' 5.8 ._

~ ~,:

' ~

5 .6- ' , , - '

-i - L x 5.c ..

w

'~

x- '

5.2~

w s._ ..

5.8 l i i 0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 15:55

COLLAPSED LIQUID LEVEL MIXTURE LEVEL i 5.0 - - . 5,8 Ik _,

n 8.8_. q . . 8.8 a i o s._

t--

u. '

u_

w w i

-5.B_. ,

_-5,8 i

.-4 ._4 uJ uJ

> > i u; m ,i ua

_ -18.8-l ~' " y -18,8,_4 1

i

-15.8 ; ; ; -15.8 0.0 5.0 10.0 15.0 20.0 T!NE (HIN)

M L Plant Analper 83-DEC-88 15:55 l COLLAPSED LIQUID LEVEL MIXTURE LEVEL ,

6,0 . , i . 6,8' 4.8-:

4.8 m . ,

n

,_. , _ - . _ _ . . ....<et- ,,

u_ 2.8-', 2.8 u.

8.8_. ._

B.8 A ,

A uJ u;

> -2.8_. ._ -2.8 3 u) , , u.J

-4.8-'. .- -4,8

-6.8 i ; .

i ; -6.0 5.0 6.0 70 80 9.0 10.0 TIME (MIN)

ML Plant Analper 83-DEC-88 15:55

4 f

SUPPRESSION POOL TEMPERATURE 258.0 ,

t,

' 288.8_.

w "3 150.8 ..

5--

196.0_. ._

m w -

CL x 50.6~

'~

w 5--

B.0 ; ; ; ;

0.0 5.0 10.0 15.0 20.0 25.0 TIME (HIN)

BNL Plant Analyser 03-DEC-88 15:55 SUPPRESSION POOL LEVEL 18.8

^ 5.B_. ._

z s.s_. ._

w w --

._. -5.0-;

-10.0 ; ; ; ;

0.0 5.0 10.0 15.0 20.0 25.0 TIME (NIN)

BNL Plant Analyzer 83-DEC-88 15:55

1 DRYELL PRESSURE WETWELL PRESSURE 17.0 . . > - - 17.8 n n

16.8 .- 16.8 #

a g -

m n

~ 15.8 .: -- 15.9 ~

w 'gN8 . w

m14.8-. 14.8o" M M M M W 13.8~ ~ 13.8 W m m m n 12.0 ; ; ; 12.8 0.0 5.0 10.0 15.0 20.0 25.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-88 15:55 DRYWELL TEMPERATURE NETELL TEMPERATURE 128.0 . . . . 128.8 188.B_. -_- 1 ,_ 188.8 n ,.

u u 88.8- -

N.8 ~

68 B_:

. ' 68.8 x::

, x::

LU UJ 5-. s

48. 6~ ~ 48.8 20.0 i i i . 23.8 0.0 5.0 10.0 15.0 20.0 25.0 TIME (HIN)

BHL Plant Analyzer 83-DEC-88 15:55 1

513 TURBINE TRIP WITH BYPASS BUT NO FWP TRIP DURING POWER OSCILLATIONS s-l l

BROOKHAVEN NATIONAL LABORATORY l} l} l l A5500ATED UNIVERSITIES, INC(Illl l

g

-],

s I

l SYSTEM PRESSURE l

14.0-1 n.

13.R: ,

a '

m g, g_ . ..

l 11.8 l l l l l l ,

ca -

18.0_. _

x 9.8-' . ._

m 8.8 ; i 0.0 5.0 10.0 15.0 TIME (MIN)

BML Plant Analyzer 83-DEC-88 15:12 SYSTEM PRESSURE 14.0 13.8-:

a '

m g,g-. ..

I)

18.0;. ._

x -

9.8_. ". .

m 8.8 ; ; i 6.0 7.0 8.0 9.0 10.0 l

TIME (MIN)

BHL Plant Analyzer 83-DEC-88 15:13

t REACTOR POW R 5.0 ,,

4.8_. _

3.8-: !-

, w , ';

2.8-: I

. i -

[ 1.8~ . .

f-

=

g,g .

- s -i u- .

O.0 5.0 10.0 15.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 15:12 REACTOR POW R 5.0

" 4.8 .

[_

s :

o 3.8-: --

w :

[ 2.8-. ._

l

~

0.0 ; ; ;

b\) fgg f 6.0 6.2 6.4 6.6 6.8 7.0 TIME (MIN)

BHL Plant Analyzer 03-DEC-88 15:12 1

4 FUEL TEMP. AT CENTERLINE NODE 14 38.8 n

u.

~

25.8 .

I cs ca

28.8_.

x . i

c 15.8-tu b

19.8 i ; i 0.0 50 10.0 15.0 TIME (MIN)

BNL Plant Analyser 83-DEC-98 15:12 CLAD WALL TEMPERATURE NODE 14 6.8-l ~ ;-

n 5.8-- .-

ta..nl iL L L a t L a Lu uL ,u a LL ui cs '

lyprne.ryyr 9y snv um w m'y Q 5.6-t- 4 a '_

5.4-. .'-

m n 5.2~'

~

tu s_.

') . O , i 0.0 5.0 10.0 15.0 TIME (MIN)

BNL Plant A m lyzer 83-DEC-88 15:12

l y . . . - . . . . . , . . . - - - , .

MIXTURE FLONRATE AT CORE INLET 18.0 x

N sua 8.8_. ..

a c

~

6.P '

ca

' B.O i

ggen i 0.0 5.0 10.0 15.0 TIME (HIN)

ML Plant Analyzer 83-DEC-88 15:12 MININUM CRITICAL POWER RATIC 4.0 3.8_. --

em ",

m u 2.8_. ..

r

~

1.8- '

hik llk !

l ll l 8.8 i ; -

0.0 5.0 10.0 15.0 TIME (MIN)

ML Plant Analyzer 83-DEC-88 15:12 i

l l

l. . . . . . . . . . . . . . _ . . . . . . . . . . .

I SUPPRESSION POOL TEMPERATURE 250.0 ,,

n

' 288.8_. .-

w 158.t- '

a 6-a

" 188.6-. --

m W

CL

'~

x '

58.8~

tu G-.

9.0 i i 0.0 5.0 10.0 15.0 TIME (MIN)

BXL Plant Analyzer 83-DEC-88 15:12 SUPPRESSION POOL LEVEL 15.0 18.6 - ..

n z

- 5.8 ..

A tu 8.8-,. .-

tu

.J ' '-

-5.8~

-10.8 i i 0.0 5.0 10.0 15.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 15:12

i DRYELL PRESSURE ETWELL PRESSURE 17.0 . . 17.8 n , n .

16.8

, , _ 16.8 M

M l ,

m . m

~ 15.8-:

- 15.8 ~

LAW 4W//.377A/,"/.""#4W *

" 14.8- 14.8=>"

=a '

M -

H M M

~ 13.8 W W 13.8T '

m "

m n .

m 12.8 i i 12.8 0.0 50 10.0 15.0 TINE (H3N)

BNL Plant Analper 83-DEC-88 15:12 DRYWELL TEMPERATURE ETWLL TEMPERATURE 128.C- . . 128.8 188.8__ .

188.8 m ,

n u ,

u

" 98.8- - -

- 98.8 ~ ,

c 68.8-7 . 68.8n*

W '

W t-- ,

s-48.8~ j 48.8

~

28.8 i i 28.8 0.0 5.0 10.0 15.0 TIME (NIN)

BNL Plant Anal per 83-DEC-88 15:12 l

l l

O i

52 MSly CLOSURE DURING POWER OSCILLATION BROOKHAVEN NATIONAL LABORATORY l} g)l ASSOCIATED UNIVERSITIES, INC. (I til

52 MSIV CLOSilRE (1) RC FLOW 8 RCP SPEED ,

(11) FW FLOW & TEMPERATURE (iii) POWER AND CORE FLOW (iv) MAXIMUM FUEL A TEMPERATURE R MAXIMUM CLAD TEMPERATURE (v) MIXTURE AND COLL. L10lllD LEVELS IN DC (vi) P0OL LEVEL a TEMPERATURE (vit) TEMPERATURES IN DRY AND WETWELLS BROOKHAVEN NATIONAL LABORATORY l} g)l A5500ATED UNIVERSITIES, INC(llll

i j

SYSTEM PRESSURE I 14.0 j

< 1 l

" I

n.8-'

1 l

$ l l 6

18.8_

L M

A 8.0 i ;

0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BNL Plant Analyzer 83-DEC-08 14:13 COLLAPSED LIQUID LEVEL MIXTURE LEVEL 5.6 . . . 5,8 L u, f

^ B.8-. .' B.8

_ n s_.

a_ ,, ,,

w

- w

-5.8_. .. -5.8 A a ua uj

-18.6-

-18.8

-15.0 i ; i -15.8 0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BHL Plant Analyzer 83-DEC-88 14:13

A 5

. i 4

9 MININUM CRITICAL POER RATIO 4.0 4 3.8_. ,_

2.6_.

l l

1.6-'

8.0 . i i 0.0 5.0 10.0 15.0 20.0 i

TINE CHIN)

BNL Plant Analyzer 83-DEC-88 14:13 e

1 SUPPRESSION POOL T.EVEL 10.C a 5.8_ .

m:

B.8 - - .-

w l >

l D -<

-5. 8-I l ,,

1 ,,

-10.0 , ; ;

0.0 5.0 10.0 15.0 20.0 TIME (MIN)

BNL Plant Analyzor 83-DEC-88 14:13

9

% I REACTOR POER 5.0 - ,

g 4.8-. .-

o

3.8 " --

tu , ,,

2.6- .-

5-e

" 1.8~

' '~

8.8 i i i i i 0.0 5.0 10.0 15.0 20.0 25.0 30.0 TIME (MIN)

BNL Plant Analper 83-DEC-88 14:13 SUPPRESSION POOL TEMPERATURE l

250.8 l

n

"- 288.8_. ._

l , :

tu l

l

158 8- - --

m 5- <.

188.B_: .-

m y ,, ,

cb x 58.8~

' '~

id 5-3.6 i i i i i 0.0 5.0 10.0 15.0 20.0 25.0 30.0 TINE (MIN)

BNL Plant Analper 83-DEC-88 14:13

REA0 TOR POE R 5.0 -

W 4.8_. ._

2 o

3.8-:

w

> 2.B_.

F , i e ;

a .. ,

l .

W

Q i ..

e.:

NM0;,digMI(i"NLblu M dnN-i d:

5.0 6.0 7.0 8.0 9.0 10.0 TIME (MIN)

ML Plant Analyzer 83-DEC-98 14:13 REACTOR PONER 1.: "

8.8 . .-

3 o

8.6- - '-

w

> 8.42 .-

a g ,. \ y i hn -

w I II I li n

9.0 i i i ; i i i 10.0 11.0 12.0 13.0 14.0 15.0 16.0 17.0 18.0 TIME (MIN)

ML Plant Analyzer 83-DEC-88 14:13

MIXTURE FLOWRATE AT CORE INLET

, 18.0 r

m 8.0- . .

A ,

e

~

6.8- ' '-

ca

4.8- .-

x ;

2.0-M -

" i

' O.8 i i i 0.0 5.0 10.0 15.0 20.0 TINE (MIN)

ML Plant Analyzer 03-DEC-88 14:13 MIXTURE FLOWRATE AT CORE INLET 5.8 -

x N

m 4.6 '- i i

I .-

ll

- \ \

\

I l

~

3.8-'

h!I((!l . - l i t

ca .

, i i ,,

2.8- I

%x il x . 1, 2 1,,-

o a .

' B.0 i ; i i i 6.0 6.5 7.0 7.5 8.0 8.5 9.0 TIME (MIN)

ML Plant Analyzer 83-DEC-88 14:13 1

4 FEEDWATER FLONRATE <

15.0 n

x !

N

~

19.8~

a , ,,

c w

, 5.6_. ,_

o a , ,

LL.

~

8.0 i i i U.0 5.0 10.0 15.0 20.0 TIME (MIN)

BHL Plant Analper 83-DEC-88 14:13 FEEDWATER TEMPERATURE <

588.0

^ ,, : .

' 488.8 '

~

w I m 388.8-' '-

m '

H ,

288.6- - .-

m Lu CL.

c ggg,3-. -

La g_. ,,

8.0 i i 0.0 5.0 10.0 15.0 TIME (MIN)

BHL Plant Analper 83-DEC-88 14:13

1 l

FUEL TEMP. AT CENTERLINE NODE 14 25.0 n

u.

~ 29.9_. .-

cs ca 15.8_ ._

18*6-'

E g

M4WMW4M ,

C--

5." i i i 0.0 5.0 10.0 15.0 20.0 TINE (MIN)

M L Plant Analyzer 83-DEC-88 14:13 CLAD WALL TEMPERATURE NODE 14 6.5--

n 5.8_. ._

c' 5.6-' #' ' ! -

H $

x 5.4_

w e 5.2-uJ s -

5.8 ; i i 0.0 50 10.0 15.0 20.0 TIME (MIN)

M L Plant Analyzer 83-DEC-98 14:13

{ . ,

6

SUMMARY

OF LASALLE RELATED EPA ACTIVITIES TO DATE

- > 21 TRANSIENTS HAVE BEEN SIMULATED:

1 FOR LASALLE REFERENCE SIMULATION 1 FOR RAMONA VS. EPA COMPARISON 10 FOR SENSITIVITY STUDY 2 FOR ASSESSING EFFECTS FROM CONTROL SYSTEMS 5 FOR LASALLE-RELATED TRANSIENTS (9 - 25 NIN SIMULATED FOR EACH TRANSIENT) 8 MAN WEEKS WITH EPA VS. 8-12 MAN MONTHS WITH MAINFRAME 1 1/2 FOR COMPUTING 1 1/2 FOR LITERATURE / DOCUMENT SEARCH 2 FOR HIPA MODIFICATIONS / HARDWARE REPAIR 3 FOR INFORMAL DOCUMENTATION COMPUTING COST IS INSIGNIFICANT

. EXPERIENCE / COMPETENCE NEEDS:

FAMILIARITY WITH: EPA BWR PLANT SIMULATION BOILING INSTABILITY BROOKHAVEN NATIONAL LABORATORY l}ljl ASSOCIATED UNIVERSITIES, INC.(llll

g.

7 RECOMMENDATIONS 71 NRR ISSilES TO BE RESOLVED 8 6

< E 8 ANALYSIS TOOL: b E &

I. IDENTIFICATION OF BWR CONDITIONS FOR:

A. CORE-WIDE INSTABILITY (IN PHASE)

. ONSET X X

- TRANSITION, LIMIT-CYCLE /LARGE AMPL. X X OSCll. OSCll.

B. REGION-WIDE INSTABILITY (0llT-0F-PHASE)

ONSET X X

. TRANSITION, LC0/LA0 X C. REGION-WIDE INSTABILITY WITHOUT DETECTABLE CORE-WIDE INSTABILITY X 11 IDENTIFICATION OF COMPONENT FAILURE CAUSES ASSOCI ATED WITH POWER INSTABILITY ATWS - MSly CLOSURE: SUPPRESSION POOL X ATWS - TT W/0 BYPASS: SUPPRESSION POOL X BROOKHAVEN NATIONAL LABORATORY l)l3l A5500ATED UNIVERSITIES, INC.(llll

l 1

I 7 RECOMMENDATIONS (CONT-)

72 EPA SIMULATIONS l EPA ASSESSMENT WITH PEACil-BOTTON 8 VERMONT-YANKEE j POWER OSCILLATIONS EPA ASSESSMENT WITil FRIGG INSTABILITY TESTS SIMILARITY / SCALING GROUPS FOR FUEL & FISSION IDENTIFICATION OF STABILITY BOUNDARIES FOR i

CORE-WIDE, IN-PilASE POWER OSCILLATIONS SIMULATION OF SUPPRESSION POOL RESPONSE TO POWER OSCILLATIONS WITH STEAM LINE ISOLATION 1

l l

l BROOKHAVEN Nail 0NAL LABORATORY l} g)l AS500ATED UNIVERSITIES, INC.(llll

i l

l 7 REC 011MENDAT10NS (CONT.)

73 NUFREQ ANALYSES MODIFY AND ASSESS 3-D NEUTRON KINETICS DETERMINE "0N-SET' STABILITY BOUNDARIES (SEE 7 1, l.A AND B.)

74 RAf10NA ANALYSES MODIFY RAlt0NA CODE: REPLACE SLIP BY DRIFT

- RAMONA ASSESSMENT WITH FRIGG TESTS, PEACH-BOTTOM TESTS, CA0RSO DATA DETERMINE STABILITY BOUNDARIES (SEE 7 1, l.A AND B.) .

BROOKHAVEN Nail 0NAL LABORATOR)l} g} l ASSOCIATED UNIVER$1 TIES, INC.(llll

ML20042D087

Contents

Text

1.0 INTRODUCTION

4.0 CONCLUSION

5.0 REFERENCES

SUMMARY

SUMMARY

SUMMARY

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ML20042D087

Text

1.0 INTRODUCTION

4.0 CONCLUSION

5.0 REFERENCES

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Navigation menu

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