ML19256E421
| ML19256E421 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 09/14/1979 |
| From: | Trimble D ARKANSAS POWER & LIGHT CO. |
| To: | Seyfrit K NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV) |
| References | |
| 1-099-12, 1-99-12, NUDOCS 7911050182 | |
| Download: ML19256E421 (52) | |
Text
- ' CENTRAL FILES PDR:HQ
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ARKANSAS POWER & LIGHT COMPANY.
POST OFFICE SOX 551 LITTLE ROCK. ARKANSAS 72203 (501)371-4000 September 14, 1979 l-099-12 Mr. K. V. Seyfrit, " rector Office of Inspect'an & Enforcement U.S. Nuclear Reg.latory Commission Region IV 611 Ryan Plaza Drive, Suite 1000 Arlington, Texas 76011
Subject:
Arkansas Nuclear One-Unit 1 Docket No. 50-313
'icense No. DPR-51 Jpplemental Response to IE Bulletin 79-05C (File: 1510.1)
Gentlemen:
Our letter of August 29, 1979, provided a report of LOCA analyses for a range of small break sizes and a range of time lapses between reactor trip and reactor coolant pump trip. Additional analysis was requested by the NRC Staff. These results are provided as Attachment 1.
Attachment 2 is a revision to Section III of Attachment 1 to our August 29, 1979, submittal. The revised areas are indicated *y vertical lines in the margin. This revised response is submitted p;rsuant to a request of the NRC Staff made at the September 13,1979, meeting of the B&W operating plants and the NRC. We note a further revision may be required dependent upon the outcome of ongoing investigations by B&W.
Very truly yours, bMbI David C. Trimble Manager, Licensing DCT:DGM:nak Attachments cc: Mr. Victor Stello, Jr., Director Mr. W. D. Johnson Office of Inspection & Enforcement U. S. Nuclear Regulatory Comm.
U. S. Nuclear Regulatory Comm. P. O. Box 2090 Washington, D. C. 20555 Russellville, Arkansas 72801 Mr. Harold R. Denton, Director .
009 Nuclear Reactor Regulation U. S. Nuclear Regulatory Comm. 7911050 !* -
Washington, D. C. y2gp5 mioote sours uriurics sysreu
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Attachment 1 -
O SUPPLEMENTAL SMALL BREAK ANALYSIS e
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- 1. Introduction Babcock & Wilcox has evaluated the effect of a delayed reactor coolant (RC) pump trip during the course of a small loss-of-coolant accident. The results of this evaluation are contained in Section II of the report entitled " Analysis Summary in Support of an Early RC Pump Trip."1 (Letter R.B. Davis to B&W 177 Oimer's' Group, " Responses to IE Bulletin 59-05C Action Items," dated August 21, 1979.)
The above letter demonstrated the following:
- a. A delayed RC pump trip at the tLme that the reactor coclant system is at high void fractions will result in t.nacceptable consequences when Appendix K cvaluation techniques are used. Therefore, the RC pumps must be tripped be-fore the RC system evolves to high void fractions.
- b. A prompt reactor coolant pump trip upon receipt. of the low pressure ESFAS
, signal provides acceptable LOCA consequences.
The following sections in this report are provided to supplement the information contained in reference 1. Specifically disc ussed in this report are:
- a. The analyses to determine the time as 211able for the operator to trip the
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reactor coolant pumps such that, under Appendix K assumptions, the criteria of 10 CFR 50.46 would not be violated.
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- b. The RC pump trip times for a spectrum o_ breaks for which the peak cladding temperature, evaluated with Appendix K assumptions, will exceed 10 CFR 50.46 limits.
- c. A realistic analysis of a typical worst c se to demonstrate that the conse-quences of a RC pump trip at any time will not exceed the 10 CFR 50.46 limits.
- 2. Time Available for RC Pump Trip Under Appendix K Assumptions A spentrum of breaks was analyzed to determine the time available for RC pump trip under Appendix K assumptions. The breaks analyzed ranged from 0.025 to 0.3 ft2 As was demonstrated in reference 1, the system evolves to high void frac-tions early in time for the larger sized breaks. Values in excess of 90% void fractior were predicted as early as 300 seconds for the 0.2 ft2 break. For the smaller breaks it takes much longer (hours) before the system evolves to high void fraction. Therefore, the time available to trip the RC pump is minidum for the larger breaks. However, as will be shown later, for the larger small breaks
(>0.3 f t 2), a very rapid depressurization is achieved upon the trip of RC pumps at high system void fraction. This results in early CFT and LPI actuation, and i2bk b) ~
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a subsequent rapid core refill. Thus, only a small core uncovery time will ensue. The following paragraphs will discuss the available time to trip the RC pumps for different break sizes. In performing this evaluation, only one HPI system was assumed available rather than the two HPI systems assumed in the ref-erence 1 analyses.
- a. 0.3 ft2 Break - Figures 1 and 2 show the system void fraction and available liquid volume in the vessel versus time for RC pump trips at 95, 8^, and 63%
void fractions for a 0.3 ft2 break at the RC pump discharge. For the pump trip at 95% void the system void fraction slowly decreases and then it drops faster following the CFT and LPI actuations. Following the RCP trip, the pressure drops rapidly and CFT is actuated at 250 seconds. The core begins to r- 11 at this time and, with LPI actuation at 300-seconds, the core is flooded faster and is filled to a liquid level of 9 feet (cquivalent to approximately 12 feet swelled mixture) at 370 seconds. The total core un-covery time is 170 seconds. Assuming an adiabatic heatup of 6.5'F/sec, as explained in reference 1, the consequences of a RC pump trip at 95% void will not exceed the 220F limit.
As seen in Figure 2 for the RC pump trip at 63% or lower void fractions, the available liquid in the core will keep the core covered above the 11 feet elevation for about 350 seconds, and above 12 feet elevation at all other ,
times. Therefore,. tripping the RC pumps at void fractions s 63% will not result in unacceptable consequences as the core will never uncover.
A RC pump trip at 83% void fraction demonstrates an uncovery time of 350 seconds. However, previous detailed small break analysis (reference 2) have shown that a 10 ft of mixture height in the core will provide sufficient core cooling to assure that the criteria of 10 CFR 50.46 is satisfied. For this case, the 10 feet of mixture height is provided by approximately 3500 ft5 liquid in the vessel. At this level in Figure 2, the core uncovery
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time is 220 seconds. Again, even with the assumption of adiabatic heatup over this period, the consequences are acceptable. It should be pointed out that if credit is taken for steam cooling of the upper portion of the , ,
fuel pin, the resulting PCT will be significantly lower then that obtained from the adie.batic heatup assumption.
From Figure ', it can be concluded that a RC pump trip at 120 seconds will result in little core uncovery. For RC pumps trip at system void fractions 1255 012 3Ja
-+ , a %sa
higher than 95% (at 200 seconds), the system will be at a lower pressure and with the CFT and LPI actuation there will be little or no core uncovery.
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Although core uncoveries are predicted for trips at 83% and 95% system void fractions, as shown earlier, the consequences are acceptable. Thus, a de-layed RC pump trip at anytime for this break will provide acceptable cons,e-quences even if Appendix K cvaluation techniques are uced.
For breaks larger than 0.3 ft2, a delt;ed Rd pump trip at any time during thu transient is also acceptable as the faster depressurization for these breaks will result in smaller delays between the pump trip and CFT and LPI actuation. Therefore, core uncovery times will be smaller than that shown for the 0.3 ft2 break.
- b. 0.2 ft2 Break - Figures 3 through 5 show the system void fraction and avail-abic liquid volume in the vessel versus time for RC pump trips ut 98, 73, 60 and 45% void fraction for a 0.2 ft2 break at the RC pump discharge. As seen in Figure 5, Ole RC pump trip at 45 and 60% void fraction does not re-sult in core uncovery. The available liquid volume is sufficient to keep the core covered above the 10 ft elevation at all times. For the trip at 98% void fraction in Figure 4, the core is refilled very rapidly with the actuation of CFT and LPI at approximately 420 ar. ' 550 seconds, respectively.
The core is refilled to an elevation of 9 feet at 460 seconds. The core an-covery time is in the order of 60 seconds, and the consequences are not sig-nificant. The RC pump trip at 73% void fraction as seen in Figure 4, re-sults in a 450 seconds core uncovery time. Although a 450 seconds uncovery time seems to result in unacceptabic consequences, if credit is taken for steam cooling and using the same rationale as that given Ior the RC pump trip at 83% system void in section 1.a, it is believed that the consequences will not be significant. Should the RC pumps be tripped at system voids less than 70%, there will be little or no core uncovery. However, for void fractions between 73% and 98%, there is a potential for a core uncovery depth and time which might be unacceptable. Thus, a time region can be de-fined in which a RC pump trip, ovaluated under Appendix K assumptions, could result in peak cladding temperatures exceeding the 10 CFR 50.46 cri-teria. This window is narrow and extends from 150 seconds (73% void) to 400 seconds (98% void) after ESEAS. A RC pump trip at any other time will not result in unacceptabic con' sequences. -
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- c. 0.1 ft2 Break - Figures 6 and 7 shows system void fractions and available liquid volume for trips at 90, 60, end 40% sy, ccm void fractions for a 0.1 ft2 break at the RC pump discharge. The rame discussions as those presented in sections 2.a and 2.b can be applied here. However, due to slower depres-surization of the system for this break, complete core cooling is net pro-vided until the actuation of LPI's. As seen in Figure 7, the time to trip the RC pumps without any core uncovery is approximately- 250 seconds. In Figure 6, with the RC pumps operating the LPI's are actuated at approximately 2350 see.onds. Tripping the RC pumps at any time before 2350 seconds will actuate the LPIs earlier in time. Therefore, unacceptable consequences are predicted for a delayed RC pump trip in a time range of 250 seconds to 2350 -
seconds. For any other time, all tha consequences are acceptc.ble.
- 4. 0.075, 0.05 and 0.025 ft2 Breaks - Figures 8 and 9 show a comparison of system void fractions for pumps running and pumps tripped 3 conditions. As seen in Figure 8, with the RC pumps tripped cc.ncident with the reactor trip, in the short term, the evolved system .oid fraction is greater than that with the RC pumos operative. The two curves cross at abov' 300 seconds.
Before this time, a RC pump trip will not result in unacceptabic consequences since the system is at a lower void fract .>n than RC pumps trip case. There-fore, the time available for RC pumps trip with acceptable results is esti-
, mated at 300 seconds. As the system depressurizes and LPI's are actuated, the core will be flooded, and a RC pump trip af ter this time will have ac- ,
ceptable consequences. From the analyses performed, the LPI actuation time is esthated at approximately 3000 seconds. Therefore, the region between 300 and 3000 seconds defines the time region in which a RC pump trip could result in unacceptable consequences.
For a 0.05 ft2 break, the same argument can be made using Figure 9. As seen from this figure, the time available to trip the RC pumps is approxicately 450 seconds. The LPI actuation time for this brea'c size is estimated at approximately 4350 seconds. ~
Therefore, the unacceptable times for RC pump trip is defined between 450 and 4350 seconds.
As discussed in reference 1, the system evolves to high void fractions very slowly for 0.025 ft2 or smaller breaks.
The system depressuriza* ion is very slow and it takes on the order of hours before the LPI's are actuated. A '
RC pump trip.at 2400 seconds for the 0.025 ft2 break results in a system 1255 014
void fraction below 50% and the core remains completely covered. A study of the 0.025 ft break 2 with 2 HPI's available shows with the RC pumps op-erative the system void fraction never exceeds 61%. The CFT is actuated at approximately 4800 seconds and the system void starts to decrease and available liquid volume in the RV starts to increase. Thus, the core will remain completely covered for any RC pump trip time and, thus, will resul't in acceptable consequences. With one RPI available, a slot ?r depressuriza-tion is expected but the system evolution to high void fraction will still be very slow. Thus, the conclusion that a RC pump trip at any time yields acceptable consequences for the 0.025 ft2 break holds whether one or two HPI's are assumed available.
The . actuation time for the 0.025 f t2 break can be. extrapolated using the available data of the other breaks. Figure 10 shows the extrapolated LPI actuation time at approximately 8000 seconds. Thus, a conservative unacceptable time region for pump trip can be defined between 2500 and 8000 seconds for the 0.025 ft2 break under Appendix K assumpcions.
- 3. Critical Time Window for RC Pumps Trip As discussed in section 2, there is a time region for each break size in which the consequenc. of the RC pump trip could exceed the 10 CFR 50.46 LOCA limit.
These critical time window, 9erc defined in section 2. Figure 11 shows a plot of the break size versus trip time RC pump which results in unacceptable conse-quences. The region indicated by dashed line; represent a boundary in which unacceptaole consequances may occut if the RC pumps are tripped. However, this region is defined using Appendix K assumptions. It should be recognized that this region, even under Appendix K assumptions, is smaller than what is shown 2
in Figure 11 as the 0.2 and 0.025 ft breaks may not even have an unacceptable region. The time available to trip the RC pumps can be obtained from the lower bound of this region and is on the order of two to three minutes after ESFAS.
- 4. " Realistic" Evaluation of Impact of Delayed RC Pump Trip for a Small LOCA
.a. Introduction -
As discussed in the previous sections, there exists a combination of break sizes and RC pump trip times which will result in peak cladding temperatures in excess of 2200F if the conservative requirements of Appendix K are utilized in the analysis. The enalysis discussed in this section was performed utilizing -
" realistic" assumptions and demonstrates that a RC pump trip at any time will not result in peak cladding temperatures in excess of the 10 CFR 50.46 criteria.
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- -- = - - . = ~ - .. .
- b. Method of Analysis There are three' overriding conservatisms in an Appendix K cmall break evalua-tion which maximizes cladding temperatures. These are:
(1) Decay heat must be based on 1.2 times the 1971 ANS decay heat curve for in-finite operation.
(2) Only one HPI pump and one LPI pump are assumed operable (single failure cri-terion).
(3) The axial peaking distribution is skewed towards the core outlet. The local heating rate for this power shape it assumed to be at the LOCA limit value.
In performing a realistic evaluation of the effect of a delayed RC pump trip following a small LOCA, the conservative assumptions described above were modi-fled. The evaluation described in this section utilized a decay heat based on 1.0 times the 1971 ANS standard and also assumed that both HPI and LPI systems were available. The axial peaking distribution was chosen to be representative of normal steady-state power operation.
Figures 12,and 13 show the axial peaking distributions utilized in this evalua-tion. These axial distributions were obtained from a review of available core follow data and the .results of manuvering analyses which have been performed for the operating plants. A radial peaking factor of 1.651, which is the maxi-mum calculated radial (uithout uncertainty) pin peak during normal operation, was utilized with these axial shapet. As such, the combination of radial and' worst axial peaking are expected to pt^ vide the maximum expected kw/f t values for the top half of the core for at least 90% of the core life. Since the worst case effect of a delayed RC pump trip is to result in total core uncovery with a subsequent bottom reflooding, maximum pin peaking towards the upper half of the core will produce the highes, peak cladding temperatures. Thus, this evaluation is expected to bound all axial peaks encountered during steady-state power operation for at least 90% of core life.
The actual case evaluated in this section is a 0.05 ft 2break in the pump di~s-charge piping ~with the RC pump trip at the time the RC system average void fraction. reaches 90%. As discussed in reference 1, RC pump trips at 90% system void fraction are expected to result in approximately the highest peak cladding temperatures. The CRAFT 2 tasults for this case and the evaluation techniques utilized are discussed in section II.B.5 of reference 1. A realistic peak
.I255 C!6
cladding temperature evaluation of this case, which is discussed below, is ex-pected to yield roughly the highest peak cladding temperature for any break size 6
and RC pump trip time. As shown in reference 1, maximum core uncovery times oi
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approximately 600 seconds occur over the break size range of 0.05 ft 2 through 0.1 ft2 using 1.2 times the ANS curve. Break sizes smaller than 0.05 ft2 and larger than 0.1 ft will 2 yield smaller core uncovery times as demonstrated inI reference 1 and the precceding sections. Use of.1.0 times the ANS decay heat curve would result in a similar reduction in core uncovery time, approximately ,
200 seconds, for breaks in the 0.05 through 0.1 ft2 range. Thus, the core re-fill rate, uncovery time, and peak cladding temperatures for the 0.05 ft 2 case is typical of the worst case values for tha break spectrum. .
- c. Results of Analysis .
Figure 4 shows the liquid volume in the reactor vessel for the 0.05 ft2 break
. with a RC pump trip at the time the system average void frac' ion reaches 907..
The core initially uncovers and recovers approximately 375 se :onds later. Using the previously discussed realistic assumptions the peak cladding temperature for this case is belowl 900F. Therefore, the criteria of 10 CFR 50.46 is met.
The temperature ._sponse given above was 2ceeloped in a conservative manner by comparing adiabatic heat.up rates to maximum possible steady-state cladding .
temperatures. First, a temperature plot versus time is made up for cach loca-tion on the hottest fuel assembly assuming that the assembly heats up adiabat!-
cally. Second, a series of FOAM4 runs are made to produce the maximum steady-state pin temperatures at each location as a function of core liquid volume.
FOAM calculates the mixture level in the core and the steaming rate from the portion of the core which is covered. Both the mixture height and steaming rate calculations are based on average core power. Fluid temperatures in the uncovered portion of the fuel rod are obtained by using the calculated average core steaming rate and by assuming all energy generated in the uncovered portion of the hot rod is transferred to the fluid. The surface heat transfer coeffi-cient is calculated, based on the Dittus-Boelter correlation , from.the F
fluid temperature and steaming rate and the steady-state clad temperature is obtained.
The FOAM data are then combined with the core liquid inventory history (derived from Figurel4') to produce a maximum possible cladding temperature as a function of time. This graph might be termed maximum steady-state cladding temperature as a function of time and decreases in value with time because the core liquid L2SS 0IU'
inventory is increasing. By cross plotting the adiabatic heat up curve with the maximum steady-state curve a tsnservative peak cladding temperature predic-tz.cn is obtained. -
- 5. Conclusions From this analysis, and the results in reference 1, the following conclusions have been drawn: -
- a. Using Appendix K evaluation techniques, there exists a combination of break size and RC pump trip times which result in a violation of 10 CFR 50.46 limits.
- b. Prompt tripping of the RC pumps upon receipt of a Im: pressure ESFAS signal will result in cladding temperatures which meet the criteria of 10 CFR 50.4y.
The minimum time available for the operator to perform this function is 2 to 3 minutes.
- c. Under realistic assumptions, a delayed RC pump trip following a small break will result in cladding temperatures in compliance with 10 CFR 50.46.
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REFERENCES I
" Analysis Summary in Support of an Early RC Pump Trip,"Section II of letter R.B. Davis to B&W 177 Owner's Group, Responses to IE Balletin 79-05C Action Items, dated August 21, 1979. -
2 Letter J.H. Taylor (B&W) "to Robert L. Baer, dated April 25, 1978.
3 . Letter J.H. Taylor to S.A. Varga, dated July 18, 1978.
4 B.M. Dunn, C.D. Morgan, and L.R. Cartin, Multinode Analysis of Core Flooding Line Break for B&W's 2568 MWt Internals Vent Valve Plants, BAW-10064, Babcock
& Wilco:., April 1978.
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5 R.H. E:oudt and K.C. Heck, THETAl-B - Computer Code for Nuclear Reactor Core Thermal Analysis - B&W Revisions to IN-1445, (Idaho Nuclear, C..T. Hocevar and T.W. Wineinger), BAW-10094, Rev. 1, Babcock & Wi'cox, April 1975.
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1 Figure 1 : 0.30 FT2 BREAK e P.D., SYSTEM VOID FRACTION VS TIME i
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Figure 2: 0.30 FT2 BREAK e P.D.,AVAILABLE LIQUID VOLUME IN RV VS TIME, 1 HPI AVAILABLE
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1255 022
t Figure 4 : 0.20 FT2 BREAK e P.D.,AVAILABLE LIQ. VOL.
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1255 024
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e Figure 7 : 0.l'FT2 BREAK e P.D., SYSTEM VOID FRACTION VS TIME LPI IT -
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1255 026
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e Figure 8 : SYSTEM V010 FRACTION VERSUS TIME i'
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Fi'gure 9 : SYSTEM V010 FRACTION VERSUS TIME PUMPS RUNNING AND PUMPS TRIPPE0 MODEL
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1255 028
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.n Figure 10 : BREAK SIZE VS LPI ACTUATION TIME
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Figure 11 : CRITICAL REGION FOR RC PUMPS TRIP,-BREAK SIZE VS TIME 0.3 -
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l Figure 12 : " REALISTIC" CORE AXIAL PEAKING DISTRIBUTION-CASE 1 i
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1255 033
Attachment 2 .
, III. , IMPACT ASSr.SSME :T OF A RC PMIP TRIP 0:: :0N-LOCA EVE!TS A. Introduction . '.
Some Chapter 15 cvents are characterized by a primary system response similar to the onc lowing a LOCA. The Section 15.1 cvents that result in an increase in heat removal by the secen'dary cystem cause a primary system cooldown and depressurization, much like a small break LOCA. Therefore, an assessment of the conse-qucnces of an imposed RC pump trip, upon initiation of the low RC pressure ESFAS, was made for these events.
B. General Assessment of Pump Trio in 1:en-LOCA Events Several concerns have been raised with regard to the effect that, an early ' pump trip would have on non-LOC '. events that exhibit LOCA characteristics. Plant recovery would be more dif ficult, dependence.
on natural circulation mode while achieving cold shutdcen wo*uld be highlighted, manual fill of the steam generators would be req;f red, and so on. Houever,. all of these drawbachs can be accommodatc3 since none of them v!11 on its own lead to unacceptable consequences. Also, restart of the pumps is recommended for plant control and cooldown "once controlled operator action is assumed. Out of this scarch, three major concerns have surf aced 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 saf ety valve opening following an overcooling transient. If
~
the time avcilable to the operator for controlling HPI flow and the margin of subecoling were substantially reduced by the pump trip to where timely and effcetive operator action could be questionable, the pump trip would become 1er.s desirable.
- 2. In the event of a large steam line break (maximum overcooling), the blowdownmayinduceasteambubbleint(pRCSwhichcouldimpair natural circulation, with severe consequences o'n the core, es-
~d' pecially if any degree of return'to power is experienced.
- 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.
1255 034
- RC pump trip since they do not initiate the low RC pressure ESFAS, and therefore, there would be no coincident pump trip. In addi-tion, these events typically do not result in an empty pressurizcr or the formation of a steam bubble in the primary system. Reactivity transients were also not considered for the same reasons. In addi-
~
tion, for overpressurization, previous analyses have shown that for the worst case conditions, an RC. pump trip will nitigate the pressure rise. This results from the greater than 100 psi reduction in pressure at the RC pu=p c>:it which occurs af ter trip.
C. Analysis of Concerns and Results
- 1. System ncoressurization In order to resolve this concern,' an analysis was perf ormed for a 177 FA plant using a MINITRAP model based on the case set up f or *D1I42, Figure 3.1 shows the noding/ flow path sr.hcme used and Tabl,c3.1 provides s description of the nodes" and flow paths. This case assumed that, as the result'of a small steam line break (0.6 ft. split) or of s n e combination of secondary side valve failure, utondary side heat demand was increased from 100% to 138% at time zero. This increase in secondary side heat dc=and 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 follouing 121 actuation on low RC pressure ESFAS, main feedwater is ramped down, MSIV's shut, and the auxiliary fecduater initiated with a 40-second delay. This arrion was taken to stop the cooldown and the d,cpressurization of the system as soon as possibic af ter IFI actuation, in order to ,
minimize the time of refill and repressurization of the system. Both HPI pumps were assumed to function.
The calculation was perforned twice, ones assuming two of the four RC pumps running (onc loop), and once assuming RC pump trip right af ter 11PI initiation. The analysis shows that the system behaves very similarly with and without pumps. In both cases, the pr.cssurizer refills in about 14 to 16 minutes from initiation of the transients, with the natural circula-
_ 16 _ 1255 035
=
- - + ~ - - . - - . ~
- tion case refilling about one minute e ore ec, . .
two of f our pumps running (Sec Figurcs 3.2,3.3). In both cases,
- the system is highly subcooled, from a minimum of 30*F to 120*T
- and increasing at the end of 14 minutes (refer to Figure 3.4).
It is concluded that an RC pump. trip following HPI actuation will not increase the probability of anusing a LOCA through the pressurizer code saf cries, and that the operator 17111 have the seme lead time, as well as a 1crge margin of subcooling, to control HPI prior to saf ety valve opening. Although no case with all RC pumps was made, it cat be infcrred from the one loop case (with pumps running) that the subcooled nargin will be slightly larger for the all pumps running case. The
.pressuriacr will take longer to fill but should do so by 16 ninutes into the transient. Figurc 3.tshows the coolant temperatures (hot icg, cold leg, and core) as a function of time for the no RC pumps casc. ,
- 2. Effect of Ste m Bubhle on Natural Circulation Cooline, For this concern, an analysis was perfor:$cd f or the same
'. generic 177 FA plant as outlined in Part 1, but assuming that as a result of an unmitigated large SLB (12.2 f t. DER), the
. cxcessive cooldown would produce void formation in the primary system. The intent of the analysis was to also show the As in extent of the void f ormation and where it occurred.
the case ana.tyzea in Part 1, the break was symmetric to both generators such that both would blow down equally, naximizing the cooldown (in this case there was a 6.1 f t. break on cach loop). There was no 1SIV closure during the transient on cither stcan generator to maxiniec cooldown. Als o,, the tur-bine bypass systen was assumed to operate, upon rupture, .
until isolation on ESFAS. ESFAS was initiated on low RC pressure ad also actuated HPI (bo(h pumps), tripped RC pumps (when applicabic) and isolat d the < JWIV's. The AFW was initiated to both generators on the low SG pressure signal, with mininum delay tk.c (both pumps operating).
This analysis was performed twice, once assuming all RC pumps running, once- with all pumps being tripped on the HPI actuation (after ESFAS), with a short (s5 second) delay. In both cases, voids were formed in the hot legs, but the dura-
- 7- -
1.255 036 . . ,
es.n.m.--- - - - -
- tion and size were smaller for the case with no R pump trip (ref er to Figure 3.7) . Although the RC' pump operating case had a higher cooldown rate, there was Icss void forma-tion, resulting f rom the additional, system mixing. The coolant tc=peratures in the pressurizer loop hot and cold legs, and the core, are shown for both cases in Figures 3.5,
- 3.6. The core outlet pressure and SG and pressurizer levels versus time are given for both cases in Figures 3.8, 3.9. This analysis shows that the system behaves similarly with and without pumps, although maintaining The RC pump flow does seem to help mitigate void formation.
pump flow case shous a shorter ti=e to the start of pres-surizer refill than the-natural c,irculation case (Figure 3.9),
although the time dif ference does not seem to be very large.
- Since the volume of the hot leg locp above the lowest point in the candy cane portion is about 63 cubic feet, these steam fgrmations have the potential for blocking natural circulation in the hot leg loops. As a result of these fiedings and since TRAP had not been programmed to closely follow this specific condition, an additional TRAP case was run. It is based on the unmitigated 12.2 ft steam line break with RC pump trip, since this case represented the bound-ing event for steam formation. This case included a more dctailed noding scheme and conservative bubble rise velocitics (5.0 f t/sec) to the upper regions of the hot legs such that the eff ect of steam formation on natural circulation in the loops could be observed.
The noding and flou path scheme used in this model is shown in Fibure 3.10. Table 3.2 provides a description of these nodes and flow paths. Figure 3.11 details the hot leg - candy cane -
upper steam generator shroud noding and flow path model superimposed over a ocaled figure of those regions. The finw path positions and sizes were carefully chosen to allow for countercurrent steam and liquid flow at the top of the candy cane. This model is consistent
' with that used for the small break LOCA analyscs described in Sec-tion 6.2.4.2 of Ref. 5.
The results of this aualysis showed steam formation only in the pressurizer loop (ref er to Figure 3.12). These steam volumes are conservative since they include all of the steam that was ;1culated as being entrained as bubbles in the liquid. Tne additional steam volumes calculated for this loop, compared with those shown-in
- Figure 3.7, are due to the additional boiling and steam separation j255 037
- that occurs in the candy cane as the liquid flow rates are reduced by steam formation and aided by metal heating. The lack of steam forma-tion in the non-pressurizer loop 'B' is attributed to.a correction in the metal heat transfer and metal' heat capacities calculated for the hot legs. The previous analysis erroneously included half of the steam generator tubes, based on the calculations from the ECCS CRAFT model. Since the TRAP code already accounts for the tube metal in ot< m generator model, this represen'ted an unnecessary conser-vatism sad it was dele ed from the model for this case.
This casr showed that the natural circulation flow was temporarily reduced. T5.is flow reduced in the pressurizer loop to 45 to 100 lb/sec from 250 to 360 seconds (refer to Figure 3.13), with flow steadily increasing after this time period. The flow in the non-pressurizer loop remained relatively unchanged at about 10CD lb/sec (ref er t o Figure 3.14) . Coro flow as maintained from 1000 to 2000 lb/sce and no void formation occurred (refer to Figures 3.15 and 3.16). The steam bubble was collapsed, natural circulation fully restored, and a greater than 50*F subcooled margin achieved in the pressurizer lcop (refer to Figure 3.16). Both steam generatsrs and the" pressurizer established level and the system pressure was turned around from the HPI flow by 14 minutes into the transient (refe to Figures 3.17 and 3.18).
- 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. }!idland FSAR, Section ISD) has shown that a RC pump trip will mitigate the consequences of an EOL return to power condition by reducing the cooldown of the primary system. The reduced cooldown substan-tially increases the suberitical margin which, in turn, reduces or clininates return to power.
D. Conclusions and Summary A general assensment 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 t:.at. a pump trip does not significantly shorten the time to filling of the pressurizer and approximately the same time interval foic operator action exists. ~
1255 038
- 2. For the maximum overcooling case analyzed, the RC pump trip increased the amount of void formation in the hot leg ' candy cane' of the pressurizer loop; however, natural circulation was not completely blocked. The steam bubble was collapsed and full natural circulation was restored. Core cooling unu maintained throughout the transient and no void formation occurred in the core.
- 3. The suberitical return-to-power condition is alleviated by the RC pump trip case due to the reduced overcooling'effect.
Based upon the above assessment and analysis, it is concluded that the consequences of Chapter 15 non-LOCA events are not increased due to the addition of a RC pump trip on ESFAS low RC pressure signal, for all 177 FA louered loop plants. Although there we're no specific analyses performed for TECO, the conclusions drawn from the. analyses for the lowered 1 cop plants are applicable.
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- NODE IUlfBER DESCRIPTION Reactor Vessel, Lower, Plenum 1 '
Reactor Vessel, Core' 2 Reactor Vessel, Upper Plenum 3 Hot Les Piping and Upper S. G. Sbroud 4,10 Primary, Steam Generator Tube Region
. 5-7,11-13 '
Cold Leg Piping 8,14 Reactor Vessel Downcomer 9
15 Pressurizer 16,24 Steam Generator Downr.omer 17,25 Steam Generator Lower Plenur.
18-20,26-28 Secondary, Steam Generator Tube Region 21,29 Steam"..isers Main Steam Piping 22,30 23 Turbine .
Containment ',
31 d
MINITRAP? PATH DESCRIPTION DESCRIPTION PATH hWIBER Core 1
2 Core Bypass 3
Upper Plenum, Reactor Vessel Hot Leg Piping 4,11 Hot Leg Piping and Upper S. G. Shroud 5,12 6,7,13,14 ' Primary, Steam Generator RC Pumps 8,15 Cold Leg Piping 9,16 Dotmcomer, Reactor Vessel 10 Pressurizer Surge Line 17 18,19,26,27 Steam Generator Downconer 20,21,28,29 ' Secondary, Steam Generator 22,30 Aspirator Steam Riser, Steam Generator 23,31 Main Steam Piping 24,32 Turbine Piping 25,33 34,35 Break (or Leak) Path HPI 36,37 -
38,39,43,44 AFW Main Feed Pumps 40,41 ~
42 LPI
. Table 3.1
. 1255 040
MINITRAP2 NODE D- .
DESCRIPTION NODE NUMBEP.
1 Reactor Vessel, Lower Plenum Reactor Vessel, Core 2
- Reactor Vessel, Upper Plenum -
3 4,10 Hot Leg Piping (including ' Candy Cane')
32,33 ' Candy Cane' and Upper S. G. Shroud 5-7,11-13 Primary, Steam Generator Tube Region 8,14 Cold Leg Piping 9 Reactor Vessel Downcomer 15 Pressurizer 16,24 Steam Generator Downcomer 17,25 Steam Generator Lower Plenum 18-20,26-28 .
Secondary, Steam Generator Tube Region '
21,29 Steam Risers -
22,30 Main Steam Piping 23 Turbine 31 Containment MINITRAP2 PATH DESCRIPTION PATH NUMBER DESCRIPTION
~ Core . .-
1 :,!..
2 Core Bypass 3 Upper Plenum, Reactor Vessel 4,11 Hot Leg Piping 5,12 Upper Steam Generator Shroud 45l46,47,48 Top of Hot Leg ' Candy Cane' 6,7,13,14 Primary Heat Transf er Region, S. G.
6,15 RC Pumps 9,16 Cold Leg Piping 10 Downcomer, Reactor Vessel Pressurizer Surge Line 17 18,19,26,27 Steam Generator Downcomer and Plenum 20,21,20,29 Secondary Ucat Transfer Region,'S. G.
22,30 Aspirator 23,31 Steam Riser, Steam Gc.. ator 24,32 Main Steam Pi'ing p 25,33 Turbine Piping 34,35 Break (or Leak) Path 36,37 HPI 38,39,43,44 AFW .
40,41- Main, Feed Pumps 42 LPI 1255 041 Table 3.2
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1255 047
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1255 048 e
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