ML19209A129
| ML19209A129 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 09/14/1979 |
| From: | Stewart W FLORIDA POWER CORP. |
| To: | James O'Reilly NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| References | |
| NUDOCS 7910020595 | |
| Download: ML19209A129 (53) | |
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, >l J.' ~ \1 f<lu : 2 S Flon.da Power C D A P O A A i s O iv Sseptember 14, 1979 File: 3-0-3-a-3 Mr. J. P. O'Reilly, Director U.S. Nuclear Regulatory Commission Office of Inspection and Enforcement, Suite 3100 101 Mariett. Street Atlanta, CA 30303
SUBJECT:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 I.E. Bulletin 79-05C
Dear Mr. O 'Reilly:
Enclosed is our response to short term action Items 2 and 4 on the subject Bulletin. The response to Item 2 is the additional information we cocimitted to provide you by our earlier response dated August 24, 1979.
If you have any questions concerning these responses, please contact this office.
Very truly yours, FLORIDA POWER CORPORATION M
W. P. Stewart Manager, Nuclear Operations Enclosure .
WPShewR03 D6 cc: Director
- Office of Inspection and Enforcement U.S. Nuclear Regulatory Commission -*
i136 09-Washington, D.C. 20555 Director Office of Nuclear Reactor Regulation ,
U.S. Nuclear Regulatory Commission
- Washington, D.C. 20555 ,;q':w'g.,p?5 General Office 3201 Tnirty-foure. street soutn . P O Box 14042. St Petersburg, Florida 33733 813 - 866-5151 701002 o sfr p
Item 2 Perform and submit a report of LOCA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip. For each pair of values of the parameters, determine the peak cladding temperature (PCT) which results. The range of values for each parameter must be wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 220 degrees F is identified.
Response: -
On August 24, 1979, Florida Power Corporation submitted the B & W report entitled " Analysis Summary In Support of an Early RC Pump Trip" as our response to Item 2 above. At that time we indicated that additional work related to Item 2 was underway at B & W and that FPC would submit this information on September 14, 1979. In that regard, enclosed is a copy of the report " Supplemental Small Break Analysis" which is being submitted in response to Item 2.
Also enclosed is a revised copy of Section III, Impact Assessment of a RC Pump Trip on Non-LOCA Events, of the B & W report " Analysis Summary In Support of an Early RC Pump Trip". This section has been revised to correct errors discovered by B & W in its calculational techniques which affected the results provided in our August 24, 1979 submittal.
The NRC staff was notified by B & W on September 7,1979 about this problem and additional discussions of the problem were held with the NRC staff on September 11, 1979 and September 13, 1979. As a result of these discussions with the NRC staff it was agreed that the revised Section III would be submitted on September 14, 1979 and B & W would perform additional analyses to insure that the 12.2 sq. ft. steam line break case is the worst case event for the non-LOCA Analysis. The results of this additional work will be submitted by Florida Power Corporation upon its completion.
WPShewR03 f D6 t1-36 i10'
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e SUPPLEMEtiTAL SMALL BREAK AtlALYSIS e
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- 1. Introductton eJ J@ 1-[ 6 .
Babcock 6 911cox has evaluated tbc effect of a delayed reactor coolant (RC) pump trip during the course of a small loss-of-coolant accident. The renults of this evaluation are contained in Section II of the report entitled " Analysis Summary in Support of an Early RC Pump Trip."3 (Letter R.B. Davis to B&W 177 Owner's Group, " Responses to IE Bulletin 59-05L Action Items," dated August 21, 1979.)
I The above letter demonstrated the following:
- a. A delayed RC pump trip at the time that the reactor coolant system is at high void fractions will result in unacceptable consequences when Appendix K evaluation techniques are used. Therefore, the RC pumps must be tripped be-fore the RC sys ca 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 folinwing sections in this report are provided to supplement the information contained in ref : tace 1. Specifically discussed in this report are:
- a. The analyses to determine the time available for the operator to trip the reactor coolant pumps such that, under Appendix K assumptions, the criteria of 10 CFR 50.46 would not be violated.
- b. The RC pump trip times for a spectrum of breaks for which the peak claddin .
temperature, evaluated with Appendix K assumptions, will exceed 10 CFR 50.46
, limits.
- c. A realistic analysis of a typical worst case 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 Availabic for RC Pump Trip Under Appendix K Assumptions
. A spectrum of breaks was analyzed to determine the time availabic for RC pump trip under Appendix K assumptions. Thebreaksanalyzedrangedfrom0.025to0.)
ft2 As was demonstrated in reference 1, the systes evolves to high void frac-tions early in time for the larger s %cd breaks. Values in excess of 90% void..
fraction were predicted as early as 300 seconds for the 0.2 f t2 break. For the.
~~
smaller breaks it takes much longer (hours) before the system evo1vcs to high void fraction. Therefore, the time available to trip the RC pump is minimum for the larger breaks. Ilowever, as vid hc shown later, for the larger small breaks
(>0.3 ft ),
2 a very rapid depressurization is achieved upon the trip of RC pumps at high system void fraction. This results in early CFT and LP1 actuation, and
~1136 112 m
a subsequent rapid core refill. Thus, only a small core uncovery time will ensuc. The following paragraphs will dis-uss the availabic time to trip the RG pumps for different break sizes. In performing this evaluatius, only one llPI system was assumed availabic rather than the two liPI systems assumed in the ref-crence 1 analyses.
- a. 0.3 ft2 Break - Figures 1 cnd 2 show the system void fraction and available liquid volume in the vessel versus time for RC pump trips at 95, 83, and 63%
void fractions for a 0.3 ft 2 break at the RC pump discharge. For the pump trip at 97A wid u.c aptu void fraction slowly decreases and then it drops faster following the CFT and LPl actuations. Following the RCP trip, the pressure drops rapidly and CFT is actuated at 250 seconds. The core begins to refill at this time and, with LPl actuation at 300 seconds, the core is flooded faster and is filled to a liquid Icvel of 9 feet (equivalent to approximately 12 feet swelled mixture) at 370 seconds. The total core un-covery time is 170 seconc's. 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. Ilowever, 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 1600
< ft3 liquid in the vessel. At this level in Figure 2, the core uncovery
(
time is 220 seconds. Again, even with the assumption of adiabatic heatup over this period, the consequences are acceptab]c. 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 M o'm the adiabatic heatup assumption. --
From Figure 2, 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 1136 113 M O
.~ , .
higher than 95% (at 200 seconds), the system will be at a lower pressure and with the CFT and LPI actuation there will be ]Itt.le or no core uncover >*.
. Although core uncoverics 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 thi.s break will provide acceptable conse-quences even if Appendix K. evaluation techniques are used. -
For breaks larger than 0.3 ft2, a delayed RC pump trip at any time during the transier.t 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-able liquid volume in the vessel versus time for RC pump trips at 98, 73, 60 and 45% void fraction for a 0.2 ft2 break at the RC pump discharge. As seen in Figure 5, the 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 and 450 seconds, respectively.
The core is refilled to an clevation of 9 feet at 460 seconds. The core un-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 unacceptable consequences, if credit is taken for steam cooling and using the same rationale as that given for 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. ' Thun, n time region can be de-fined in which a RC pump trip, evaluated 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 180 seconds (73%, void) to 400 seconds (98% void) after ESFAS. A RC pump trip at any other time will .
not result in unacceptable consequences... l D D i
1136 tt qgy eJO_a!Li a
- c. 0.1 ft2 lircak - Figures 6 and 7 showr system void fractions and available -
liquid volume for trips at 90, 60, and 40% system void fractionsafor ft2 break at the RC pump dischargo. 0.1 The same discussions as those presented in sections 2.a and 2.b can he applied here.
llowever, due to slower depres-surization of the system for this brhak, complete core cooling is not 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 econds. 250 s In Figure 6, with the RC pumps operating the LPI's are actuated at a 2350 seconds. pproximately 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, a3L the consequences are acceptable .
- d. 0.075, 0.05 and 0.025 ft 2 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 coincident with the reactor trip, in the short term, the evolved system void fraction is g than reater that with the RC pumps operative.
The two curves cross at about 300 seconds.
Before this time, a RC pump trip will not result in unacceptabl e consequences since the system is at a lower void fraction than RC pumps tripTherc- .
case fore, the time available for RC pumps trip with acceptable results s esti-i 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 h ave ac-
'ceptable consequences.
From the analyses performed, the LPl actuation time is estimated at approximately 3000 seconds.
Therefore, the region between 300 and 3000 seconds defines the time region in which a RC u pump trip c 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 approximate 450 seconds. (
.approximately 4350 Theseconds.
LPI actuation time for this break size is estimated at Therefore, the unacceptabic times for RC pump trip is defined between 450 and 4350 seconds. '
As-discussed in reference 1, the system evolves to high void fractions v slowly for 0.025 ft2 ery or smaller breaks.
The system depr.cssurizadon is very.
slow and it takes on the order of hours before the LPI's RC pump trip at 2400 seconds for the 0.025 f t 2
. A are actua break results in a system 1136 115
I
. J void fraction below 50% and the core remains completely covered. A study of the 0.025 ft2 break with 2 HPI's available shows with the RC pumps op- -
erative the system void fraction never excceds 61%. The CFr is actuated
- at approximately 4800 seconds and the system void starts to decrease and availabic liquid volume in the RV starts to increase. Thus, the core will
~
remain completely covered for any RC pump trip time and, thur, will result in acceptable consequences. With one HpI available, a slouer depressuriza-tion is expected but the system evolution to high void fraction will still be very slow. Thus, the conclusion that a SC 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 LPI actuation time for the 0.025 ft2 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 assumptions.
- 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 consequences of the RC pump trip could exceed the 10 CFR 50.46 LOCA limit.
These critical time windows were 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 lines represent a boundary in which unacceptable consequences may occur 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 in Figure 11 as the 0.2 and 0.025 ft2 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.
t i
- 4. " Realistic" Evaluation of Impact of Delayed.RC
. Pump Trip for a Small LOCA -}h )}h
-} .a. Introduction -
As discussed in the previous sections, there exists a combination.o.f 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 ar:e utilized
- in the analysis. The analysis discussed in this section was performed utilizing I " 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.
~
- b. Method of Analysis There are three overriding conservatisms in an Appendix K small 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 llPI pump and one LPI pump arc assumed operabic (single failure cri-terion). (3) The axial peaking distribution is skewed towards the core outlet. The local heating rate for this power shape is a.ssumed to be at the LOCA limit value. In performing a realistic evaluation of the effect of a delayed RC pump trip following a rmal1 LOCA, the conservative assumptions described above were modi _ fied. The evaluation described in this section utilized a decay heat based on 1.0 times the 1971 ANS standard and also assumed that both llPI and IP1 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 (without uncertainty) pin peak during normal operation, was, utilized with these axial shapes. As such, the combination of radial and worst axial peaking are expected to provide 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 hottom reflooding, maximum pin peaking towards the upper half Thus, this of the core will produce the highest peak cladd_ing temperatures. cvaluation is expected to bound all axial peaks encountered during steady-state, power operation for at least 90% of core life. 2 The actual case evaluated in this section is a 0.05 ft break in the pump dis , charge piping with the RC pump trip at the time the RC system average void fraction reaches 90%. As discussed in reference 1, RC punip trips at 90% system void fraction are expected to result in approximately the highest peak cladding,
- temperatures. The CRAFT 2 results for this case and the evaluation Jechniques A realistle peak utilized are discussed in section II.B.5 of reference 1.
1136 117
P00R ORGIRL cladding temperature evaluation of this case, which is discussed beloE, is ex-pected r.to yield rougaly the highest peak cladding temperature for any break siac and RC pump trip time. As shown in reference 1, maximum core uncovery times of approximately 600 seconds occur over the break size range of 0.05 ft2 through 0.1 ft2 using 1.2 times the ANS curve. Break sizes smaller than 0.05 ft2 and larger than 0.1 ft2 will yield smaller core uncovery times as demonstrated in 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? cane is typical of the vorst case values for the 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 fraction reaches 907.. The core initially uncovers and recovers approximately 375 seconds later. Using
- the previously discussed realistic assumptions the peak cladding temperature for this case is below1900F. Therefore, the criteria of 10 CFR 50.46 in met.
. The temperature response given above was developed 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-t' ion on the hottest fuel assembly assuming that the assembly heats up adiabati-cally. Second, a series of FOAM4 runs are made to produce the maximum steady-state pin temperatures at cach location as a function of co- liquid volume. FCAM calculates the mixture level in the core and the st, ..ag rate from the portion of the core which is covered. Both the mixtura height and steaming rate calculations are based on average core power. Fluid temperatttres in the uncovered portion of the fuel rod arc obtained by using the calculated average
, core steaming rate and by assuming all energy <;cncrated 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-Boclter correintions , from the fluid temperature and steaming rate and the steady-stato clad temperature is obtained.'
The FOAM data are then combined with the core liquid inventory history (derived from Figurc 14.) to produce a maximum possible cladding temperature an.a function of time. This graph might be termed maximum steady-state cladding temperaturc . as a function of time and decreases in value with time because the co're liquid 1136 i18' g
' n ., .
inventory is increasinr;. lly cross plotting the adiabatic heat up curve ulth the maximum steady-state curve a conservative peak cladding temperature predic-tion is obtained.
- 5. Conclusions From this analysis, and the results in reference 1, the following' conclusions have been drawn:
- a. Using Appendix K cvaluation 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 low pressure ESTAS signal will result in cladding temperatures which meet the criteria of 10 CTR 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.
e 8 N- e 8 =9 m e i 1136 119 .
?
e 9 REl'ElU NCES , 1 " Analysis Summary in Support of an Early RC l' ump Trip," Section II of letter ! R.B. Davis to B&W 177 Owner's Group, Responses to IE Bulletin 79-05C ActJon l Items, dated August 21, 1979. - 2 Letter J.H. Taylor (B&W) to p.obert 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
& Wilcox, April 1978.
5 R.11. Stoudt and K.C.11cck, TilETAl-B - Computer Code for Nuclear Reactor Core Thermal Analysis - B&W Revisions to IN-1445, (Idaho Nucicar, C..I. Ilocevar and T.W. Wincinger), BAW-10094, Rev.1, Babcock & Wilcox, April 1975. M. . ' 1136 120
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e t, t Figure 1 : 0.30 FT2 BREAK e P.D., SYSTEM VOID FRACTION VS TIME 100 - ----------% j / 1,.' >.& . eae Ono O% 8 a 80 - I f F -
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T"DA5%> ~ y y/o./ I
- -- PUMPS RUNNING v 60 $
o 1 - . - PUMPS OFF e 95% VOID e i 3 I m g N 40
- - PUMPS OFF e 83% VOID g
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I i -x-x PUMPS OFF e 63% VOID 20 - 0 i i i i i i ' ' ' 200 400 600 800 1000 O Time, sec ee O
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1 1 ji Figure 2: 0.30 FT2 BREAl' e P.D., AVAILABLE LIQUIO - VOLUME IN RV VS TIME, 1 HPI AVAILABLE a 3000 _ S -
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$ o ll II o , I -
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F29ure 3 : 0.20 FT 2 !!RFAK e P.D.,SYSID1 VOID IRACTION VS LIME 100 - o p,~o-a ,,'*~ ~ ' j i-t
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-x x - -x x- PUMPS OFF e 7b% VOID 40 -
------ PUMPS OFF e 60% VOID t
-x x- PUMPS OFF e 45% VOID 30 -
i -
- i 20 - ' ' ,
0 400 800 1200 1600 2000 Time, sec 1136 123
/ Figure 4 : 0.20 FT2 BREAK e P.D.,AV/iILABLE LIQ. VOL. - IN RV VS TIME 1 HPI AVAILABLE
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3000 - R R / 9 9 / x x / N $ , I j/ i l u k~ r,<*x' g p- x-s/ _
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.11 : 5 124 i '
O .J . . . O 400' 800 1200 1600 2000 Time, sec -
F29ule .5 : 0.20 F1 2 BREAK P.D. AVAll Alt! F l. TO. V(ik . IN RV VS TIME 1 HPI AVAILABLE 3200 _ 3000 _ , 4 4 l 2800 _ _. g._ x_. x PUMPE OFF co 45% VOIO
! g PUMPS OFF e_
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. '1 1 3 6 1 2 5~
_ a Figure 6 : 0.1 FT 2 BREAK O' P.D., AVAILABLE LIQUID VOLUME -
- IN RV VS TIME, 1 HPI AVAILABLE s%
3000 .
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s f o/ --- PUMI?S OFF 5 % e 40% VOID __ l-o' ~ ~ 9; a.
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O i i i _ _ t. . O 400 .800 1200 1600, 2000 Time, sec
Figure 7 : 0.1 F12 BREAK e P.D., SYSTEM - VOID FRACTION VS TIME 100 _ LPI r l
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60 - M / g R I \ S l'/ E - $m a (( ~~" I y 40 - o -x-x-PUMPS TRIP e 90% VOID I j I - - PUMP 1 RIP e 60% VOID
-o-o PUMP TRIP e 40% VOIO 20 -
0 . , i i 0 i 500 1000 1500 2000 2500 Tine , sec t
- m. e fi36 127
. Figure 8: SYSTEM VCID FRACTION VERSUS Tl!4E 4
9' 40
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59
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35 D / g (g - -/ s A
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0 50 100 150 ~ 200 250 300
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1136. 128' S
a Figure 9 : SYSTEM V0l!! FRACTION VERSUS TIME
- PUMPS RUNNING AND PUMPS TRIPPE0 MODEL 40 n g
s ? 35 &
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. Time, sec - -
1136 129 . O
u l C ca Figure 10: ECEAK SIZE VS LPI ACTUATION TIME 0.4 .a I O.3 4 0 0.2 - 5 O.1 - EXTRAPOLATED
.LPI. FOR 0.025 BREAK O ' ' ' '
' ' ' ! e 1000 2000 3000 4000 5000 6000 ,7000 8000 9000 T2me, sec I0 .
q ? I - n
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Figure .11 : CRITICAL REGION FOR RC PUMPS TRIP, BREAK SIZE VS TIME e m
~~
0.3 - i b '
" \_% \ .
\- N s
k-2 a 0.2 - -
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m 5 y - w CRITICAL REGICN NOTE: TIME T = 0.IS REACTOR TRIP 0.1 - e TIME l . 0 0 . . . . ... . . . . . . ... 100 1000 10000 Tme, m
- cm m
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- I III. IMPACT ASSESSME':7 0F A RC PUMP TRIP ON NON-LOCA EVENTS 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 primary system cooldown and depressurization, much like a small break LOCA. Therefore, an assessment of the conse-quences of an imposed RC pu=p trip, upon initiation of the low RC pressure ESFAS, was made for these events.
B. General Assessment of Pump Trin in Non-LOCA Events Several concerns have been raised with regard to the eff ect 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 suutdown wo'uld be highlighted, manual fill of the steam generators would be required, and so on. However, all of these drawbacks can be acco=modated since none of them will on its own Icad to unacceptable consequences. Also, restart of the pu=ps 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 systen fill /repressurization or safety valve opening following an overcooling transient. If the time available to the operator for controlling HPI flow and
. the margin of subcooling were substantially reduced by the pu=p trip to where timely and effective operator action could be questionable, the pump trip would become less desirable.
- 2. In the event of a 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 experienced.
- 3. A more general concern exists with a large steam iine break at EOL conditions and whether or not a return to power is experienced following the RC pump trip. If a return to cr,itical is experienecd, natural circulation flow may not be sufficient to remove heat and to avoid core damage. jj}f ]}}
Ov;crhcating events were not considered in the impact of the RC pump trip since they do not initiate the low RC pressurc ESFAS, and therefore, there would be no coincident pump trip. In addi-
~
tion, these events typically do not result in an m pty pressurizer or the formation of a steam bubble in the primary syst s. Reactivity transients were also not considered for.the same r.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 pu=p exit which occurs af ter trip. C. Analysis of Concerns and Results
- 1. System Reorossurization 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 TMI22, Figure 3.1 shows the noding/ flow path scheme 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 f t. split) or of some combination of secondary side valve failure, secondary side heat demand was inercased 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 cualysis, it was assumed that following HPI actuation on low RC pressure t.SFAS, main ,
feedwater is ramped down, MSIV's shut, and the auxiliary feedwater initiated with a 40-second delay. This action was taken to stop the cooldown and the d,cpressurization of the system as soon as possible af ter HPI 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), had once assuming RC pump trip right after HPI initiation. The analysis shows that the system behaves very similarly with and without pu=ps. In both cases, the pr.cssurizer refills in about 14 to 16 minutes from initiation of the transients, with the natural circula-1136 136 6 s_
i
# tion case refilling about one minute before the case with - two of four pumps running (Sec Figures 3.2,3.3). In both cases, the system is highly subcooled, from a minimum of 30*F to 120*F and increasing at the end of 14 minutes (ref er to Figure 3.4).
It is concluded that an RC pump trip following HPI actuation will not increase the probability of causing a LOCA through the pressurizer code saf eties, and that the operator will have-the same lead time, as well as a large margin of subcooling, to control HPI prior to saf ety valve opening. Although no case teith all RC pumps was made, it can be inf erred from the one loop case (with pu=ps 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 M shows the coolant temperatures (hot leg, cold leg, and core) as a function of time for the no RC pumps case. ,
- 2. Effect of Steam Eubble on Natural Circulation Coolint For this concern, an analysis was performed for the same
- generic 177 FA plant as outlined in Part 1, but assuming that as a result of an unmitigated large SLB (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 Benerators such that both would blow down equally, maximizing 2
the cooldown (in this case there was a 6.1 f t. break on each loop). There was no MSIV closure during the transient on either stcan 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 pressureandalsoactua[edHPI(bo(hpumps),trippedRC pumps (when applicable) and it.clated the MFWIV's. The AFW
~
was initiated to both generators on the low SC pressure signal, with minimum delay time (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-1136 137 _ 12 _ . g
; tion and size were smaller for the case with no RC pump trip (refer to Figure 3.7).Although the RC pump operating case had a higher cooldown rate, there was less void forma-tion, resulting from the additional system mixing. The coolant te=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 shown that the system behaves 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 dif f erence 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 formations have the potential for blocking natural circulation in the hot leg loops. As a result of these findings and s'ince 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 f t steam line break with RC pump trip, since this case represented the bound-ing event for steam formetion. This case included a more detailed noding scheme and conservative bubble rise velocities (5.0 f t/sec) to the upper regions of the hot legs such that the effect of steam formation on natural circulation in the loops could be observed.
The noding and flow path scheme used in this model is shown in F Qire 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 scaled figure of those regions. The flow 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 fr. the small break LOCA analyses described in Sec-tien 6.2.4.2 of Ref. 5. The results of this analysis showed steam formation only in the pressurizer loop (ref er to Figure 3.12). The'se' steam volumes are conservative since they include all of the steam'that was calculated as being entrained as bubbles in the liquid. The additional steam volumes calculated for this loop, compared with those shown in Figure 3.7, are due to the additional boiling and steam separation llbb l3b
1 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 its steam generator model, this represented an urnecessary conser-vatism and it was deleted from the model for this case.
This case showed that the natural circulation flow was temporarily reduced. This flow reduced in the pressuriner loop to 45 to 100 lb/sec f rom 250 to 360 seconds (ref er to Figure 3.13), with flow steadily increasing af ter this time per.iod. The flow in the non-pressurizar loop remained relatively unchanged at about 100Dlb/sec (refer to Figure 3.14). Core flow was maintained from 1000 to 2000 lb/sec and no void formation occurred (ref er to Figures 3.15 and 3.16). The steam bubble was collapsed, natural circulation fully restored, and a greater than 50*F subecoled margin achieved in the pressurizer loop (refer to Figure 3.16). Both steam generators and the pressurizer established level and the system pressure was turned around from the HPl flow by 14 minutes into the transient (refer 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. Midland FSAR, Section 15D) 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 subcritical margin which, in turn, reduces or eliminates return to power.
D. Conclusions and Smumary - A general assessme..t 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 thme 1.!36 139 to fil ins of the Pressurizer and approximately the same time interval
. 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 fell natural circulation was restored. Core cooling was maintained througfiout 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 overceoling'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 lowered loop plants. Although there were no specific analyses performed for TECO, the conclusions drawn from the. analyses for the lowered loop plants are applicable. f
.. g amma i136 140
HINIThAP2 NODE (DESCRIPTION NODE NUMBER DESCRIPTION 1 Reactor Vessel, Lover , Plenum 2 Reactor Vessel, Corc~ 3 Reactor Vessel, Upper Plenum 4,10 Hot Leg Piping and Upper S, G. Shroud
. 5-7,11-13 Primary, Steam Ccncrator Tube Region 8,14 Cold Leg Piping 9 Reactor Vessel Downcomer 15 Pressurizer 16,24 Steam Generato1 Downconer 17,25 Steam Generator Lower Plenum 18-20,26-28 Secondary, Stea= Generator Tube Region 21,29 Steam Riscrs 22,30 Main Steam Piping 23 Turbinc 31 containment MINITRAP2 PATH DESCRIPTION PATH NUMBER DESCRIPTION 1 Core 2 Core Bypass 3 Upper Plenum, Reactor Vessel 4,11 Hot Leg Piping 5,12 Hot Leg Piping and Upper S. G. Shroud 6,7,13,14 Primary, Steam Generator 8,15 RC Pumps 9,16 Cold Leg Piping 10 Downcomer, Reactor Vessel 17 Pressurizer Surge Line 18,19,26,27 Steam Generator Downcomer 20,21,28,29 Secondary, Steam Generator 22,30 - Aspirator 23,31 Steam Riser, Steam Generator 24,32 Main Steam Piping 25,33 Turbine Piping 34,35 Break (or Leak) Path 36,37 HPI 38,39,43,44 AFW 40,41 Main Feed Pumps
~
4 2' ~ LPI - -. Tabic 3.1 1136 141
I MINITRAP2 NODE DESCRIPTION e DESCRIPTION NODE NUMBER Reactor Vessel, Lower Plenum 1 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 Cold Leg Piping 8,14 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 Main Steam Piping 22,30 23 Turbine 31 Containment MINITRAP2 PATH DESCRIPTION DESCRIPTION PATH NUMBER 1 Core 2 Core Bypaes 3 Upper Plenum, Reactor Vessel 4,11 Hot Leg Piping 5,12 Upper Steam Generator Shroud 45,'46,47,48 Top of Hot Leg ' Candy Cane' 6,7,13,14 Primary Heat Transfer Region, S. G. RC Pumps 8.15 Cold Leg Piping 9.16 Downcomer, Reactor Vessel I 10 Pressurizer Surge Line 17 18,19,26,27 Steam Generator Downcomer and Plenum 20,21,28,29 Secondary Heat Transfer Region, S. G. 22,3D Aspirator , 23,31 Steam Riser, Steam Generator
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.' ~~ " Figure 3.11 Li r fr
-48b- 1136 15"7 P
- TOTAL STEAM BUBBLE VOLUME VS TRANSIENT TIME .
2 (102% FP, 12.2 FT DOUBLE-ENDE0 UNMITIGATED STEAMLINE BREAK, RC PUMP TRIP) , g , MAXIMUM =366 FT2e300 SECONDS
~
360 o
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3 o ', m: 6 . 5 240 i- 3 E 2 3
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- ; 3 '
w g a o - G 5 -s 120 - G -
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- (LOOP 'A'-PRZR)
~
2 4 6 8 10 12 14 0 u ' O' Transient Time (minutes) U1 4
LOOP 'A' CAN0Y CANE FLOW YERSUS TRANSIENT TIME
' 2 (102% FP, 12.2 FT 00U8tE-ENDED UNMITIGATED l
~ ;
. STEAMLINE. BREAK, RC PUMP TRIP) 16,000 14,000 - -
n E 12,000 - 5 - E \
=
{= 10,000 [ . n E E CORE FLOOO TANK
- a =
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',' w FLOW BEGINS b{S PRESSURIZER EMPTIES
' 6,000 -
(STEAM BUBBLES FORM IN HOT LEG) k y S ,
.3 4,000 -
- ' 2,000 -
g u 0.000 & 0 2 4 6 0 10 12 14
- Transient Time (Minutes) tn (11
i 1
. g w . m = - in W cm w WW z-w__ .. c.o.
MW
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< m = z *^
WD m n o M c> - DW C1. = u> c2 - c . m Z m CD - w w - . E
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(cas/ mal) solj aue3 Apuc3 ,8, doo1 Figure 3.14 M
~
1136 i56
CORE FLOW VERSUS TRANSIENT TIME (1025 FP, 12.2 FT2 DOUBLE-ENDED l UNMITIGATED STEAMLINE BREAK, RC - 80,000 PUMP TRIP) 70,000 - 60,000 - o
, { 50,000 -
- i;; 2
=
; i w
40,000 - G . [ a u 30,000 - 20,000 - - 10,000 - - A _ 0,000 ' ' 0 2 4 6 10 g 8 12 14 Transient Time (minut'es) N
$ ^
^ w e CL O O O O J e8 LA.D CC Z CE N -
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= 4 w CC LD /
>= C.D w w /
.-- M NJ a /
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.rf
- Ca O O 4 - W c t.aJ = Cr* CD w l su CL N CD 3E: - -
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- t=-
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. a 3c=. w w so O 4 - ,,
J N La.a CD N 5 C3 C3 e- = Q - C/3 CD CC l *
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C3 CC f Z CL lQ Q & - N
- 2 /
- l l O /
/ m
_- I I hi N I C"3 C3 C3 O CD C3 o . O C3 C.3 r- . m Lr> O,e o o c3 N (3 ) 'saJnlej adujal lue1003
, Figure 3.,16 ,
a f136 158
e p 4 o
- e - -
}
CE CE
. o* o ;
6-- 0-af auf 1 CE CC CC w W w > N Z Z
- w w CE C.D LD D ^
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. w w CC W CC 6- Q- D==
a.- u> v u>
~ .. .. ..
E D O G w a S <
, -
.s o OG v> -m o Do o . CE < - . w w O P 2as w o o - > w o c w a - a z o . w ^ a. o - o 5 g g => .w Ce es .o n o - - o > o a. o E o - m g
- c. mw o a. o - o a w m a o g w a o E z -
w e-- o e-o - = . =: z 4 => < ,e w +- u> m- . . W o z m . eJ m ce - -H w D N.- n. . = CC w -l) e . U V> %% O un u g, e o E"
- a. v i f e o
o o o c, c,,, m - o n (13) laAa1 JolejauaD tucals/Jaz[Jnssaid ~ - Figure 3.17 1136 159' 6 ,e cm w w ~ a w - .4 6- CD .- n z - w r _ z m o o z _ a o n w o. . .m- n - ^ z e
- tA >= =
o u,J - m w m m _, m => uJ En o C = =o o. - o - E
- u. V w C3 cc a: O oN m >- - se m u. e +-
w < K N W # EL.
- m ,a:
o o e .- - = w w - m -z c: .a - m. - a a => u. - o- n ~ m - - w N w ,e ec o - - o - m a v , - m I f f g , o o o o 4' e o o o Cs o o C3 C3 C3 o o o o n a m o o o o o o,e n m m u - - o. o (Isd) aJnssaJd talan0 ajoa . - . Figure 3.18 ~ 1136 160 9 Response to '.E. Bulletin 79-05C Short-Term Actions Item 4: Revise emergency procedures and train all licensed reactor operators and senior above. operators based on the guidelines developed under Item 3 Response 4: The emergency procedures have been reviced and all licensed reactor operators and senior operators have been trained based on the guidelines developed under I.E. Bulletin 79-05C, Item 3. / WPShewR03 D6 1136 161 : >}}