ML20043B257
| ML20043B257 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 05/31/1988 |
| From: | Cheng H BROOKHAVEN NATIONAL LABORATORY |
| To: | |
| Shared Package | |
| ML20042D069 | List: |
| References | |
| FOIA-90-13 NUDOCS 9005250159 | |
| Download: ML20043B257 (18) | |
Text
,
fl 1h'dp g C-1 IJ -Ln W[ -[ BROOKHAVEN NATIONAL LABORATORY ASSOCIATED UNIVERSITIES, INC.
{lIl~ 'w J a Upton Long Island. New York 11973 l
l l (516) 282s Department of Nuclear Energy FTS 666/ - 2444 ;
June 6, 1988 ,
E. L. Jordan Office of the Analysis and Evaluation of Operational Data Nuclear Regulatory Commission ..
Maryland National Bank Building 7735 Old Georgetown Road Room 3203/ Mail Stop .3302 Bethesda, MD 20814
Dear.Mrp)gr n,
Attached please find our report, which summarizes the effort conducted by BNL at the request of AE00 to study the recent Lasalle-2 event using our Engi-neering Plant Analyzer (EPA). We feel that the level of detail provided by the EPA was really underscored through the analysis of this event. The EPA models were able to predict the thermal hydraulic instability, which was re-
. sponsible for the. growing power oscillation.
Our analyses included in the attachment were generated without the bene-fit of the detailed plant / incident related data, which we recently received from AE00. Thus, while we were able to' accurately reproduce the event, we feel that. further improved analyses could yet be' conducted if you desire. We are- presently having discussions with your staff to explore these possibili-ties. .
These rapid response, low cost computations show the EPA's versatility and emphasize the very useful role that this capability provides. We look forward to your continued interest in our work.
Sincerely yours, CM .
W. Y. Kato Chairman WYK/ctf Distribution Attached
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9005250159 900328 l PDR FOIA PDR HIATT90-13 TELEX: 6852516 BNL DOE FACSIMILE: (s16)282 3000. FTS 666 3000 CABLE: BOOOKLAB UPTONNY
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$2 ..
f D'istribution:
E. S. Beckjord, NRC/RES 0
.H.-S. Cheng, BNL S. Fabic, NRC/AEOD ..
J. G. Guppy BNL G.. F. Lanik, NRC/ AE00 T. M. Novak, NRC/AE00 J. E. Rosenthal, NRC/ AE00 ,!
D. F. Ross, NRC/RES H. H. Scott, NRC/RES i B. Sheron, NRC/RES 9
L.-M. Shotkin NRC/RES T. P. Speis, NRC/RES
-0 W. Wulf f BNL -
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't SIMULATIONS OF THE LASALLE 2 INCIDENT WITH THE BNL PLANT ANALYZER H. S. Cheng Informal Report May, 1988 t
i Plant Analyzer Development Group Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 J7&.;
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TABLE OF CONTENTS i
-i ABSTRACT . . . .. . . . .. . . . . . . . . . . . . . 11 s !
LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . iv l i
.~. y
' LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . :
1.0 INTRODUCTION
. . . . . . . . . . . . . . . . . . . . . . . . 1 I L
- 2 2.0 EVENT DESCRIPTIONS . . . . . . . . . . . . . . . . . . . . .
1 i
2.1 Initial Conditions . . . . . . . . . . . . . . . . . . . 2 i 2.2 Reference Case . . . . . . . . .. . . . . . . . . . . 4 2.3 Case 2 (Flow Reduction by Increasing Core Exit Pressure Drop) . . . . . .. . . . . . . . . . . . . . . . . . . 5 2.4 Case 3 (Flow Reduction, Followed by Recirculation Pump Restart) . . . . . . . . . . . . . . . . . . . . . . . 7 2.5 Case 4 (Flow Reduction, RCP Restart and MSIV Closure) . 8 3.0 ANALYSIS AND RESULTS . . . . . . . . . . . . . . . . . . . . . 9 3.1 Case 1 (Reference Case) . . . . . . . . . . . . . . . . . 9-3.2 Case'2 (Flow Reduction by Increasing Core Exit Pressure Drop).. . . . .. . . . . . . . . . . . . :. . ., . . . . 9 3.3 Case 3 (Flow Reduction, Followed by Recirculation Pump Restart) . . . . . . . . . - . . . . . . . . . . . . . . 10
- 3. 4- Case 4 (Flow Reduction, RCP Restart and MSIV Closure) . 10 4.0 DISCUSSIONS . . . . . . . . . . . . . . . . . . . . . . . . 11 i
4.1 The Effect of Feedwater Control System - Case 5-, . , . 11 4.2 The Impact of Void Reactivity Coefficients - Case 6 ., 11 5.0
SUMMARY
AND CONCLUSIONS . . . . . . . . . . . . . . . . .. . 12
6.0 REFERENCES
. . . . . . . : . . . . . ... . . . . . . . . . 13 V
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4 LIST OF TABLES Table Eage.
1 Feedback Coefficients Used . . . . . . . . . . . . . . . 3 2- Sequence of Events for Reference Case . . . . . . . . . . 4 3- Sequence of Events for Case 2 . . . . . . . . . . . . . . 6 ;
4 Sequence of Events for Case 3 . . . . . . '. . . . . . . . 7 .
5- Sequence of Events for Case 4 . . . . . . . . . . . . . . 8 i
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i 2.0 EVENT DESCRIPTIONS a
2.1 Initial conditions The proper initial conditions are essential for the analysis of this event. A steady-state run was first made to obtain the desired initial conditions. The initial conditions obtained by the BNL Plant Analyzer are summarized as follows:
o Rated Thermal Power 3323 MWt ,
o Rated Core Flow Rate ,
108 Mlb/hr o Reactor Power 2808 MWt (85%)
o Core Inlet Flow Rate 81 Mlb/hr (7 5'/.).
o System Pressure. 1007 pela o Steamline Flow Rate 11.3 Mlb/hr o Feedwater Flow Rate 11.3 Mlb/hr o Recirculation Drive Flow Rate 28.9 Mlb/hr o Recirculation Pump Speed 1590 rpm o Downcomer Liquid Level 562 in, o Core Average Void Fraction 41.7%
o Core Average Fuel Temperature 1113 *F o Core Average Coolant Temperature 545 *F o Core-Inlet Subcooling 20 *F o Feedwater Temperature 402 *F Reactivity feedback plays an important role in a BWR for both' steady state and transient analyses. Table 1 summarizes the feedback l
l coefficients used in the present work, which were obtained from the.
earlier work on the void feedback [Cheng, 1977) and the' Doppler feedback [Cheng, 1978).
Axial power distribution is known to affect the core instability
! of a BWR [Yokomizo, 1987). The axial power profile used in this work is shown in Figure 1.
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l 4 2.2 Reference Cast -
0 The reference caseoscillation is defined as the best-estimate simulation of '
L the LaSalle 2 power incident without the reported flow L
reduction. The transient was initiated from the initial conditions l
described in Section 2.1 by a recirculation pump trip with a partial feedwater heater loss. This case was run for 10 minutes without any intervention from keyboard. Table 2 lists the sequence of events for '
the reference case.
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TABLE 2 j
l Secuence of Events for Reference Case
.1 Event / Action Time (m) l l !
1 -5.0 i
! 1. Steady State at 85% power and 75% flow
~
l 1 2. Recirculation pumps tripped and heaters isolateded 0.0 l l
Reactor power dropped to minimum (28%) 0.08 l
! 3. !
l 0.2 l 4. Reactor power recovered to 40% due to heater loss l 1 0.5 l
! 5. Core flow reached natural circulation (29%) !
! 3.0 l 6. Reactor power reached 60% & " beat" phenomena began !
! 7. Modulated limit cycle oscillations continued- 9.0 l l
I End of transient 10.0 l l 8. l This case exhibits a limit cycle oscillation in reactor power and core flow with " beat" phenomena. The detailed results of this case are presented in Section 3.1 and shown in Figures 2 through 5. -The limit: cycle oscillation conditions, reached under natural circulation conditions and maintained indefinitely, are shown as Point' A in the
. power vs. flow map for LaSalle 2 [Kaufman, 1988) of Attachment 1.
64 . s. '
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l------------------_ --------__----_------------------------__------_l 1 l l TABLE 3 !
I i l Seouence of Events for Case 2 l t l l Event / Action Time (m) l l l
! 1. Steady State at 85% power'and 75% flow -5.0 I .
l >
l- 2. Recirculation pumps tripped and heatere failed 0.0 i I l l 3. Reactor power dropped to minimum (28%) 0.08 l I l 1 4. Reactor pcuer recovered to 40% due to heater loss 0.2 l l l l S. Core flow reached natural circulation (29%) '
0.5 l l l l 6. Reactor power reached 60% & " beat" phenomena began 3.0 i l 7. Modulated limit cycle oscillations continued 4.8 l I i l 8. Interactive application of the multiplier started 4.9 l 1 l-l (Initially, the multiplier was only 1.05)- l 1 l l 9. Enhanced limit, cycle oscillatione continued as 5.9 l I !
I core flow was being reduced l l l l 10. ' Unstable diverging oscillations started when 6.0 i i
l the core flow was reduced to 20% at 50% power i
+
L l 1 l 11. Power and flow started to show growing oscillationc 6.1 l 1
l 12. Reactor power reached 118%, and reactor tripped 6.9 l 1
l l 13. End of transient 9.0 1 l_----_--_____-----_-_----_-_-____----_____--_-_--_----_------___--l!
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2.5- Case 4 (Flow Reduction. RCp Restart and MSIV Closure) ]
This is an extension of Case 3 in that all four MSIVe were closed 1
- after the pump restart but before the reactor trip. Thus Case 4 was ;
intended to see the effect of-MSIV closure on the power oscillations. !
Table 5 zummarizes the sequence of events for Case 4.
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____________________________________________________________________t i
1 TABLE 5 l j l- l !
Secuence of Evente for Case 4 j l l L l !
] Event / Action Time (m) l l
! I i 1. Steady State at 85% power and 75% flow -5.0 l l l l l 2. Recirculation pumps tripped and heaters failed 0.0 l !
l l l 3. Reactor power dropped to minimum (28%) 0.08 l l l l 4. Reactor power recovered to 40% due to heater lose 0.2 i l I' l 5. Core flow reached natural circulation (29%) 0.5 l I l t 6. Reactor power reached 60% & " beat" phenomena began 3.0 l 1 - !
i 7. Modulated limit cycle oscillatione continued 4.0 !.
l I
! 8. Interactive application of the multiplier started ' 4. 9 - l l l l (Initially, the multiplier was only 1.05) l l
i L l 9. Unstable diverging oscillations started when 6.0 t l 1 l l the core flow was reduced to 20% at 50% power i
! 10. Both recirculation pumps were restarted 6.5 i l I l i 11. MSIVe were closed 6.84 !
l l l l 12. Reactor tripped by MSIV closure signal 6.9 l
-l l l 13. End of transient 9.0 l j I l l
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t 10 temperature reactivity are secondary to tha void reactivity in this event as shown in Fig. 11.
The Plant Analyser predicted, for Browns Ferry 1, a two-phase multiplier of 6.2 for the form loss at coro exit, while GE's design methods predicted a value of 19.3 (Kavanagh, 1969). Thus, by using the GE's correlation for the two-phase multiplier and two multipliers of 1.3 each to account for the uncertainties in the core exit form lose coefficient and the two-phase multiplier, we arrived at a core exit pressure drop five times larger than the reference value for Browns Feery 1.
Since the operator did attempt to restart the recirculation pumps, although failed, the next case was intended to see the impact-of the pump restart.
3.3 Case 3 (Flou Reduction. Followed by pecirculation Pumn Restart)
The results for Case 3 are presented in Figures 12 through 17.
Figure 12 shows the reactor power and core flow response of this cace.
One sees clearly the effect of the pump restart on the core flow response. The pump restart resulted in a rapid increase in core flow-and appeared to be effective in damping the power oscillation, but the positive void reactivity due to the increased core flow also led to rapid increase in reactor power so as to trip the reactor sooner than the previous case as seen in the zoomed display of the power oscillation (Fig. 13). Figure 14 chows the behavior of the total reactivity of- Case 3, and Fig. 15 presents that of the void reactivity. Once again, it was the void reactivity which was driving the transient. The response of the recirculation pump speed is shown in Fig. 16, and that of the system pressure is presented in Fig. 17.
The next case was intended to see if MSIV closure right after the l pump restart would aggravate the situation.
3.4 Case 4 (Flow Reduction. RCP Restart and MSIV Closure)
The results for Case 4 are presented in Figures 18 through 21.
In this case, the recirculation pumps were restarted at 6.5 min and l the MSIVs were closed at 6.84 min. The reactor was automatically l ecrammed by the MSIV closure signal before the reactor power reached
-ite cetpoint as shown in Fig. 18 and 19. Figure 20 shows the total (net) reactivity behavior of Case 4 and Fig. 21 presents the steam flow reeponse, which clearly indicates the MSIV closure. It appeare that the MSIV closure would not aggravate the power oscillation and, indeed, would help trip the reactor sooner.
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12 j 5.0 EEMMARY AND CONCLUSIONS l i
The BHL Plant Analyzer has been used to simulate the recent i LaSalle 2 power oscillation incident. By driving the reactor into the i unstable region'of the power / flow map, we were able to reproduce the l
diverging power oscillation as observed in the LaSalle 2 event. The ~
following conclusions can be drawn from the present analysis: 1
- 1. T1.e most likely cause for the growing power oscillation in l the LaSalle 2 event was a gross power / flow mismatch brought 1 about by the recirculation pump trip together with the loss i of feedwater heating. The reactor appeared to be in the unstable operating region. ~
- 2. Restarting the recirculation pumps would be effective in damping the power oscillation, but would also cause the reactor power to increase rapidly so as to trip the reactor sooner. It would seem advisable to restore the pumps to only partial speed instead of full speed so that the reactor would not be tripped.
- 3. MSIV closure would not aggravate the power oscillation as the MSIV closure signal would trip the reactor in a few seconds.
- 4. The void feedback was the driving potential for the power oscillation in the LaSalle 2 incident. In general, the more negative the void coefficients, the more pronounced is the power oscillation.
- 5. Doppler and moderator temperature feedback played only a secondary role in the LaSalle 2 event.
- 6. The feedwater control system could not be responsible for the growing power oscillation in the LaSalle 2 event.
- 7. The BNL Plant Analyzer is capable of analyzing the nuclear I thermal hydraulic stability of BWRs.
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TOTAL REACTIVITY 1.8 i i i i
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INL Flant Analyzer 96-MAY 8817l82 Figure 4. Total Reactivity Behavior of Reference Case l FEEDLIATER TEMPERATURE l 459.8 i i i i l . .
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m . l l D 350.8 ' .'.
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BNL Plant Analvfer; ^"*
06-HAY-8817:92 I
Figure 5. Feedvater Temperature Response of Reference Case
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BHL Plant Analyzer 99MAV-8898:34 Figure 8. Total Reactivity Behavior of Case 2 >
YOID REACTIVITY 2.8 ! ! ! !
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, Title (HIN)
BHL Flant Analvfer B1 HAY 8898:34 Figure 9. Void Reactivity Behavior of Case 2
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20 REACTOR POE R 1,s. HIXTURE FLOWRATE.AT CORE. INLET is,e n
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M 1, g.l; o N
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g,g so 2 :: m _; J W o g,g '- . I-b ]
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0.0 1.0 2.0 3.0 4.0 5.0 60 7.0 B.0 9.0 TlHE (HIN)
ENL Flant Analyzer 99-MAY8818:04 Figure 12. Reactor Power and Core Flow Response of Case 3 REACTOR POER 1
l 1.2 1 M 1,s.:; .
w 2 :: ::
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3 . 3 5.0 5.5 6.0 6.5 7.0 7.5 T.lHE(HIN)
BHL Plant AnalyzeY ' " 89-MAY8810:04 Figure 13. Zoomed Display of the Power Oscillation of Case 3
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RECIRCULATION PUMP SPEED
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0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 -
TIME (MIN) ;
BHL Plant Analven 89MAY8819:04 Figure 16. Recirculation Pump Speed Response of Case 3 SYSTEM PRESSURE -
12,8 ' i . , i g i i n
-g18,8-' (. .
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' (MIN.)
EHL Flant Analyzer" '- -
89MAY8818:R4 Figure 17. System Pressure Response of Case 3 i
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24 TOTAL REACTIVITY 1.8 e ! ! !
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g 0.5 ,. ..
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,1, g5 = 0 5.5 6.0 6.5 7.0 7.5 TIME (MIN)
BMI, Plant Analyzer 99MAY8819:27 Figure 20. Total Reactivity Behavior of Case 4 FLOLIRATE AT STEAHLINE ENT11ANCE gg,g__ i i i. i i i .
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r 15.8 "
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BNI,PlantAnalyz'eY" 89MAY8818:27 Figure 21. Steam Flow Response of Case 4 l
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REACTOR POER 1,t i > > >
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5.0 5.5 6.0 6.5 7.0 7.5 TlHE (HIN)
INL Flant Analyzer 89 HAY 8811:22 Figure 24 Power Oscillation Between 5 and 7.5 Min, of Case 5 REACTOR POE R s,g- ., .,
FEEDWATERFI.0WRATE
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BNL Flant Analyzer '-
99-HAY 8811:22 l
Figure 25. Power Oscillation and Feedwater Flow Oscillation of Case 5 l
28 REACTOR POER i 1.2 ; ; ;
MIXT lJRE FLO)lRAT.E. AT CORE. INLET 18,8 I n
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0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 TIME (HIN) -
INL Flant Analyzer 99 NAY 88 12lB2 Figure 27. Reactor Power and Core Flow Response of Case 6 REACTOR POWER 1.2 1
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# 09-NAY8B12l82 Figure 28. Zoomed Display of the Power Oscillation of Case 6
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ATTACHMENT i )
l A. RATURAL CIRCULATION
- 3. WW RECIRC. PUMP $ PED TALVE MINIMUM POSITION (0! OPENINC)
)
.C. WW RECIRC. PUMP $ PEED YALVE MAXIMUM POSITIDH ,,
D. RATED RICIRC. PUMP $ PED VALVE MINIMUM POSITION (01 OPENINC)
E. ]
ANALYTICAL I4WER LIMIT OF AUTOMATIC LOAD POLL &=TRC RECICH 1
. P. ANALYTICAL UPPER LIMIT OT AUTOMATIC LOAD POLI 4VING RICION, RATED . i POWER TLOW :
C. LOWEST ALLOWABLE FLOW AT RATD POWER (I001 P, 75:T) i
.H. RICHEST ALLOVA?LE TWW AT RATD POWER (100% P,105% T) .
KACTX: EXTEND D LOAD LINE RIGION THIJP: INCREASD CORE TWW REGION 1
~
OWN% nd.s of LFh6[mox vabe, posh 120
- @@ + O C. lose. ve h es +Trauhr t h(b speed *
- 33, , @ ,g open ve\ves b i 9 5'/> to m b # ,
@w@,@ WNh% red 3 6 deitre.d. ,
goo . rod. liu c rN ,
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8 ' s j CASE i t M IenLINE 60 -
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CASE 2 yf -
30 Q P
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20 CAVITATION REC 10N 10 -
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0 O 10 20 30 40 50 60 70 80 90 100 110 120 C0F.E FLOW (2)
--.- - -..--- --_ -_ _ _ _ - - - - - . - - - . . . - ., , ,-- - - - , - - - - - - - ,-we