ML16138A417
| ML16138A417 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Arkansas Nuclear, Crystal River, Rancho Seco  |
| Issue date: | 09/14/1979 |
| From: | Heltemes C NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| To: | Ross D NRC - TMI-2 BULLETINS & ORDERS TASK FORCE |
| References | |
| NUDOCS 7910150678 | |
| Download: ML16138A417 (22) | |
Text
Distribution (Same as individual receiving responses to IE Bulletin 79-05C)
Docket File (5)
$7'6 I&E (6)
R. Audette B. Sheron R. Conran P. Norian L. Beltracchi R. Capra V. Benaroya A. Thadani R. Bosnak F. Williams S. Varga D. Crutchfield M. Fairtile C. J. Heltemes R. Mattson NRC PDR Z. Rosztoczy R. Frahm W. Jensen F. Odar B. Siegel B. Wilson S. Israel G. Mazetis M. Cunningham E. Case D. Eisenhut S. Hanauer T. Novak R. Reid R. Tedesco T. Wambach LPDR (5)
ACRS (16)
D. Garner G. Vissing C. Nelson L. Engle D. Diianni
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 September 14, 1979 Docket Nos.: 50-269, 50-270, 50-287, 50-302, 50-312, 50-313, 50-346 MEMORANDUM FOR:
D. F. Ross, Jr., Director Bulletins & Orders Task Force FROM:
C. J. Heltemes, Jr., Leader, Project-Management Group-----
Bulletins & Orders Task Force THRU:
Z. R. Rosztoczy, Leader, Analysis Group
'- R7 Bulletins & Orders Task Force
SUBJECT:
ERROR REPORTED BY BABCOCK & WILCOX IN ITS GENERIC REPORT "ANALYSIS
SUMMARY
IN SUPPORT OF AN EARLY RC PUMP TRIP" On September 7, 1979, Babcock & Wilcox (B&W) reported an error in its generic report entitled "Analysis Summary in Support of an Early RC Pump Trip."
The error is in Section III of the report, which presents an impact assessment of a RC pump trip for non-LOCA events.
IE Bulletin 79-05C required that each PWR licensee perform and submit a report of LOCA analyses for their plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip.
In addition, each licensee was to develop new guidelines for operator action, for both LOCA and non-LOCA transients, that take into account the impact of RCP trip requirements.
The subject analysis was performed by B&W and presented as a generic report applicable to all B&W 177 FA plants.
Each of the B&W operating plant licensees revtewel this report and endorsed its applicability in their responses to IE Bulletin 79-05C.
IE Bulletin 79-05C requires that RC pumps be tripped immediately upon a reactor 6
trip and an ESF actuation signal caused by RC system low pressure. This action is required to protect the core for a certain spectrum of small break si'zes, Since certain overcooling transients can cause the same conditions without having a LOCA, licensees were asked to perform an assessment of the RC pump trip for this non-LOCA condition.
In its assessment, B&W presented an analysis of what it considered the most limiting overcooling event, an unmitigated large steam line break. (SLB).. This analysis shows that the excessive cooldown would produce void formation in the RC system hot legs; however the size of the steam bubble volume and the duration of its presence was small and was not sufficient to affect the ability to cool the core on natural circulation.
The analysis shows a steam bubble volume of about 12 ft. 3 in the hot leg attached to the pressurizer surge line and about 5 ft. 3 in the other hot leg.
The duration is approximately 3 to 5 minutes, (The volume of the "candy cane" at the top of the hot leg is 63 ft. 3.)
i~ii0 7
- In reviewing this analysis, B&W discovered an error in the conversion of the steam mass to steam volume. The density used to convert the mass to volume was the average froth (two phase) density vice the actual steam density. When the proper values were used, the same analysis showed a volume of steam in the hot leg with the pressurizerattached was about 250 ft. 3 and about 150 ft. 3 in the other leg. B&W stated that it then performed some hand calculations using a bubble rise model which showed approximately 400 ft. 3 of steam in the hot leg with the pressurizer attached. B&W was not sure of the steam volume in the other loop. At that time, B&W stated that it would need more time to review the analysis and refine the model.
B&W committed to call back early the next week to review its results.
On September 12, 1979, another conference call was held between B&W and the staff. B&W stated, it had made refinements to its model and reran the analysis.
The changes included: (1) modifications to the sensible heat of the S/G tubes, (2) incorporati-onof a hajse separation model and- (3) division of the hot leg into 2 nodes. The results of this analysis showed approximately 400 ft. 3 of steam in the hot leg with the pressurizer attached and none in the other hot leg. The loop with no voiding remained between 400F to 80oF subcooled. The amount of steam in the hot leg with the pressurizer attached was sufficient to retard natural circulation. However, B&W stated, any formation of voids would be temporary and themake-upwater (from HPI) would collapse the steam bubble in approximately 9 minutes,' allowing natural circulation to commence in that loop. The loop without the pressurizer attached would not lose natural circulation during the course of the transient. B&W also stated that it saw no reason to change the guidelines it had developed for the B&W operating plant licensees.
The revised analysis would be given to the licensees on Thursday, September 13, 1979, for submittal to the NRC.
At a meeting with the B&W Owners' Group and B&W,-held on September 13, 1979, the staff presented an analysis done by Brookhaven National Laboratory, The analysis was a simulated overfeed transient done with the IRT computer program.
The initial conditions assumed;the reactor was at 100% power, 100% reactor coolant flow and pressurizer pressure of 2158 psig. At time '0" a turbine trip was initiated. It was also assumed that the ICS failed in such a way that both steam generators continued to be fed by the main feedwater system.
The analysis showed a very rapid drop in RCS pressure and a void fraction of 28% in the "'candy cane" within approximately 100 secs.
This transient could be more limiting than the DESLB analyzed by B&W. Rowever, there were questions concerning the number of single failures that would be necessary to produce this "run away main feedwater" transient. B&W stated that it would review this analysis and advise the NRC of the results of its review in a timely fashion.
The Analysis Group of the B&OTF has the lead on reviewing responses to IE Bulletin 79-05C. We will continue to keep you advised of the status of this situation.
C; Heleme,Jr, Leader P et Management Group
Enclosures:
Bulletins & Orders Task Force 1-B&W's "Impact Assessment of a RC Pump Trip on Non-LOCA Events" 2-Brookhaven N.L. Analysis for a B&W Overfeed Transient
' III. T fPACT ASSESSENT OF A RC PUMP TRIP ON
- O;-LOCA ETS A.
Introduction Some Chapter 15 events are characterized by. a primary system response similar to the one following a LOCA. The Section 15.1 events that result in an increase in heat removal by the secondary system cause a primary system cooldown and depressurization, much like a small break LOCA. Therefore, an assessment. of the conse quences of an imposed RC pump trip, upon initiation of the low RC pressure ESFAS, was made for these events.
B. General Assessment of Pumn Trio in Non-LOCA Events Several concerns have been raised with regard to the effect that an early pump trip would have on non-LOCA events that exhibit LOCA characteristics. Plant recovery would be more difficult, dependence.
on natural circulation mode while achieving cold shutdcwn would be highlighted, manual fill of the steam generators would be required, and so on.
- However, all of these drawbacks can be accomodated since none of them will on its own lead to unacceptable consequences.
- Also, restart of the pumps is not precluded for plant control and cooldown once controlled operator action is assumed. Out of this search,
-three major concerns have surfaced which have appeared to be sub stantial enough as to require analysis:
- 1.
A pump trip could reduce the time to system fill/repressurization or safety valve opening following an overcooling transient.
f the time available to the operator for controlling iPI flow and the margin of subcooling were substantially reduced by the pump trip to where timely and effective operator action could be questionable, the pump trip would become unacceptable.
- 2. In the event of al large steam line break (maximum overcooling), the blowdown may induce a steam bubble in the RCS which could impair natural circulation, with severe consequences on the core, es pecially if any degree of return to power is expericnced.
- 3.
A more general concern exists with a large steam line break at EOL conditions and whether or not, a return to power is experienced following the RC pump trip.. If a return to critical is experienced, natural circulation flow may not be sufficient to remove heat and to avoid core damage.
ver ting events were not considercen the impact of the RC pump trip since they do not initiate the low fuC pressure ESFAS, and therefore, there would be no coincident pump trip.
In addi tion, these events typically do not result in an empty pressurizer or the formation of a steam bubble in the primary system.
Reactivity transients were also not considered for the same easons.
In addi tion, for overpressurization, previous analyses have shown that for the worst case conditions, an RC pump trip will mitigate the pressure rise.
This results from the greater than 100 psi reduction in pressure at the RC pump exit which occurs after trip.
C.
Analysis of Concerns and Rsults
- 1.
System Repressurization In order to resolve this concern, an analysis was performed for a 177 FA plant using a MINITRAP model based on the case set up for T1I-2, Figure 3.1 shows the noding/flow path scheme used and Table3.1 provides s description of the nodes and flow paths. This case assumed that, as the result of a 2
small steam line break (0.6 ft.
split) or of some combination of secondary side valve failure, secondary side heat demand was increased from 100% to 138% at time zero.
This increase in secondary side heat demand is the smallest which results in a (high flux) reactor trip and is very similar to the worst moderate frequency overcooling event, a failure of the steam pressure regulator.
In the analysis, it was assumed that following HPI actuation on low RC pressure ESFAS, main feedwater is ramped down, MSIV's shut, and the auxiliary fecdwater initiated with a 40-second delay.
This action was taken to stop the cooldown and the depressurization of the system as soon as possible after IPI actuation, in order to minimize the time of refill and repressurization of the system.
Both HPI pumps were assumed to function.
The calculation was performed twice, once assuming two of the four RC pumps running (one loop),
and once assuming RC pump trip right after 1PI initiation.
The analysis shows that the system behaves very similarly with and without pumps.
In both cases, the pressurizer refills in about 14 to 16 minutes from initiation of the transients, with the natural circula case refilling about one minD before the case with two of four pumps running (See Figures 3.2,3.3). In both cares, the system is highly subcooled, from a mini;um of 300F to 1200F and increasing at the and of 14 minutes (refer to Figure 3.4).
It is concluded that an RC pump trip following HIPI actuation will not increase the probability of causing a LOCA through the pressurizer code safeties, and that the operator will have the same lead time, as well as a large margin of subcooling, to control HPI prior to safety-valve tapping-- Although no case with all RC pumps was made, it can be inferred from the one loop case (with pumps running) that the subcooled margin will be slightly larger for the all pumps running case. The pressurizer will take longer to fill but should do so by 16 minutes into the transient.
Figure 3.1shows the. coolant temperatures (hot leg, cold leg, and core) as a function of time for the no RC pumps case.
- 2. Effect of Steam Bubble on Natural Circulation Cooling A
For this concern, an analysis was perforzed for the same generic 177 FA plant as outlined in Part 1, but assuming that 2
as a result of an unmitigated large SL3 (12.2 ft.
DER),
the excessive cooldown.would produce void formation in the primary system.
The intent of the analysis was to also show the extent of the void formation and where it occurred. As in the case analyzed in Part 1, the break was symmetric to both generators such that both would blow down equally, maximizing the cooldown (in this case there was a 6.1 ft.2 break on each loop).
There was no MSIV closure during the transient on either steam generator to maximize cooldown.
Also, the tur bine bypass system was assumed to operate, upon rupture, until isolation on ESFAS. ESFAS was initiated on low RC pressure and also actuated HPI (boL pumps), tripped RC pumps (when applicable) and isolated the M1FWIV's.
The AFW was initiated to both generators on the low SG pressure signal, with minimum delay time (both pumps operating).
This analysis was performed twice, once assuming all RC pum-ps running, once with all pumps being tripped on the HPI actuation (after ESFAS), with a short (S second) delay.
In both cases, voids were formed in the hot legs, but the dura
-17
t:
and size were smaller for the se with no RC pump trip (refer to Figure 3.7).Although thc RC pump operating case had a higher cooldown rate, there was less void forma tion, resulting from the additional system mixing.
The coolant temperatures in the pressurizer loop hot and cold legs, and the core, are shown for both cases In Figure:: 3.5, 3.6.
The core outlet pressure and SC and pressurizer levels versus time are given for both cases in Figures 3.3, 3.9.
This analysis shows that the system behaves very similarly with and without pumps, although maintaining RC pump flow does seem to help mitigate void formation.
The pump flow case shows a shorter time to the start of. pres surizer refill than the natural circulation case (Figure 3.9),
although the time difference does not seem to be very large.
- 3.
Effect of Return to Power There was no return to power exhibited by any of the BOL cases analyzed above. Previous analysis experience (ref.
Midland FSAR, Section 15D) has shown that a RC pump trip will mitigate the consequences of an EOLreturn to power condition by reducing the cooldown of the primary system.
The reduced cooldown substantially increases the subcritical :margin which, in turn, reduces or eliminates return to power.
D.
Conclusions and Summary A general assessment of Chapter 15 non-LOCA events identified three areas that warranted further investigation for impact of a RC pump trip on ESFAS low RC pressure signal.
- 1. It was found that a pump trip does not significantly shorten the time to filling of the pressurizer and approximately the same time interval for operator action exists.
- 2.
For the maximum overcooling case analyzed, the RC pump trip increased the amount of two-phase in the primary loop; however, the percent void formation is still too small to affect the ability to cool on natural circulation.
- 3. The subcritical return-to-power condition is alleviated by the RC pump trip case due to the reduced overcooling effect.
Based upon the above assessmtent and analysis, it is con cluded that. the consequences of Chapter 15 non-LOCA events are not
incre due to the addition of, a.C trip on ESFAS low RC pressure signal, for all 177 rA owercd loop plants.
Although there were no specific analyses performed for TECO, the conclusions drawn from the analyses for thc lowered loop plants are applicable.
19 -
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B & W Overfeed Transient Event Sequence Event Time (Seconds)
Power 100%
Initial conditions Flow = 100% -
Initial conditions Pzr pressure = 2157.8 psia Initial conditions Turbine trip initiated 0
Turbine flow ramps to zero 0 -
2.5 sec.
.5 sec.
ICS Fails F.W. stays on Aux. F.W. ramps to 4.1%
1 - 4 sec.
Peak Pzr pressure = 2232 psia 5.4 sec.
Pzr empties (PPzr = 1589 psia) 42.7 sec.
ECCS trip (at P = 1615) 41.5 sec.
Pump trip (on ECCS trip) 41.5 sec.
ECCS delivery (after 6.4 sec. delay) 47.9 sec.
Upper head flashes 42.5 sec.
Candy cane flashes Between 45 and 50 sec.
Candy cane quality =.001 50 sec.
Max. candy cane quality =.0088
. 105 sec.
void =.28 (HEM)
Pressure falls Continues a
W.t OVERFEED TRANSIENT 9
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