ML19210B751
| ML19210B751 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 09/20/1979 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| IEB-79-05C, IEB-79-5C, NUDOCS 7911120246 | |
| Download: ML19210B751 (57) | |
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" UU Docket No. 50-346 EDISON License No. NPF-3 Serial No. 1-91 September 20, 1979 Mr. James G. Keppler Regional Director, Region III Office of Inspection and Enforcement U. S. Nuclear Regulatory Co= mission 799 Roosevelt Road Glen Ellyn, Illinois 60137
Dear Mr. Keppler:
Attachments A and B are Toledo Edison's follow-up responses to IE Bulletin 79-05C for the Davis-Besse Nuclear Power Station, Unit (DB-1) concerning reactor coolant pump operations. These discuss the analyses and development of preliminary operator guidelines for reactor coolant pu=p operations.
Attachment A is the continued evaluation of Short Term Action Item Number 2 discussed in Toledo Edison's letter of August 29,1979 (Serial No.1-85) . It is noted that this analysis was completed for a lowered loop reactor coolant system configuration and is a conservative evaluatien for the raised loop DB-1 design.
Attachment B is a revision to Section III of the " Analysis Su==ary in Support of an Early RC Pu=p Trip" also transmitted in our August 29, 1979 letter. This revision replaces entirely the previously submitted Section III.
Very truly yours, ff 2: 7._,
Richard P. Crouse Vice President - Energy Supply RPC/TJM cc:
Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Co==ission Washington, D. C. 20555
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Director Office of Inspection and Enforcement U. S. Nuclear Regulatory Com=ission g Washington, D. C. 20555 g C3 4 ,
i L ._777 THE TOLECO ECISCN COMPANY EDISCN PLAZA 300 MACISCN AVENUE TCLEDO. OMIO 43552 7 91112 0 Z 4- 6
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f Docket No. 50-346 License No. NPF-3 Serial No. 1-91 September 20, 1979 AITAQDENT A .
SW E N AL SMALL BREAK ANALYSIS 1312 278
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SUPPLEMENTAL SMA!! 3REAK ANALYSIS
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- 1. Introduction ,
Babcock & Wilcox has evaluated the effect of a delayed reactor coolant (RC) pu=p trip during the course of a r all loss-of-coolant accident. The results of this evaluation are centained in Section II of the report entitled "Analys's Su==ary in Support of an Early RC Pump Trip."1 (Letter R.B. Davis to B&W 177 Owner'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 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 system evolves to high void fractions.
- b. A prompt reactor ra olaat pu=p trip upon receipt of thc low pressure ESEAS signal provides (_ceptable LOCA consequences.
The following sections in this report are provided to supplement the information contained in refer ance 1. Specifically discussed in this report are: .
- a. The analyses to determine the time available for the operator to trip the reactor coolant pu=ps such that, under Appendix K assu=ptions, 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 cladding te=perature, evaluated with Appendix K assu=ptions, 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 pu=p trip at any time vill not exceed the 10 CFR 50.46 limits.
- 2. Time Available for RC Pump Trip Under Apeendix K Assumotions A spectrum of breaks was analy:ed to determine the et=e available for RC pu=p trip under Appendix K assu=ptions. The breaks analyzed ranged from 0.025 to 0.3 ft2 Av 9 camonstrated 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 fraction were predicted as early as 300 seconds for the 0.2 ft 2 break. For the r= aller breaks it takes much longer (hours) before the system evolves to high void fraction. Therefore, the time .iailable to trip the RC pump is minimum for the larger breaks. However, as will be shown later, for the larger s=all breaks
(>0.3 f t2), a very rapid depressuri:ation is achieved upon the trip of RC pu=ps at high system void fraction. This results in early CFI and LPI actuation, and 1312 ?30
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F a subsequent rapid core refill. Thus, only a small core uncovery ti=e will ensue. The fo., lowing paragraphs will discuss the available time to trip the RC 1
pumps for different break sizes. In performing this evaluation, only one HPI system was assumed available rather than the two EPI 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, 83, and 63%
void fractions for a 0.3 ft 2 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 LFI actuations. Following the RCP trip, the The core begins pressure drops rapidly and CFT is actuated at 250 seconds.
to refill 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 (equivalent to approxi=acely 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 puso trip at 95% void will not exceed the 22000F 11mit.
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 ti=e of 350 seconds. However, previous detailed s=all break analysis (reference 2) have shown that n 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 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 adiabatic heatur assurption.
From Figure 2, it can be concluded that a RC pump trip at 120 seconds will .
result in little core uncovery. For RC pe=ps trip at system void fractions
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i 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.
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 conse-quences even if Appendix K evaluation techniques are used.
For breaks larger than 0.3 ft ,2 a delayed RC pump trip at any time during the 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-able liquid volume in the vessel versus time fcr 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 nt 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 refi*1ed very rapidly with the actuation of CFT and LPI at approximately 420 and 450 seconds, respectively.
The core is refilled to an elevation 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 scconds 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 fracticas 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, 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 pu=p trip at any other time will not result in unacceptable con $equences.
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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, and 40% system void fractions for a 0.1 ft2 break at the RC pump discharge. The same discussions as tho.,e presented in sections 2.a and 2.b can be applied here. However, due to slower depres-surization of the system for this breal , complete core cooling is not pro-vided until the actuation of LPI's. As seen in Figure 7, the time to trip the RC pu=ps "without any core uncovery is approximately 250 seconds. In Figure 6, with the RC pumps operating the LPI's are actuated at approximately 2350 seconds. Tripping the RC pumps at any time before 2350 seconds vill 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 the consequences are acceptable.
- d. 0.075, 0.05 and 0.025 ft2 Breaks - Figures 8 and 9 show a co=parison of system void fractions for pumps running and pumps tripped 3 conditions. As seen in Figure 8, with the RC pu=ps tripped coincident with tla reactor trip, in the short term, the evolved system void fraction is greater than that with the RC pu=ps operative. The two curves cross at about 300 seconds.
Before this time, a RC pump trip will not result in unacceptable consequences since the system is at a lower void fraction than RC pumps trip case. There-fore, the time available for RC pumps trip with acceptable results is eati-mated at 300' seconds. As the system depressurizes and LPI's are actuated, the core vill be flooded, and a RC pump trip after this time will have ac-ceptable consequences. From the analyses performed, the LPI actuation time is esti=ated 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 argu=ent can be made using Figure 9. As seen from this figure, the time available to trip the RC pumps is approxi=ately 450 seconds. The LPI actuation time for this break size is esti=ated at approximately 4350 seconds. Therefore, the unacceptable times for RC pu=p 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 s= aller breaks. The system depressurization 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
, r void fraction below 50% and the core re=ains completely covered. A study of the 0.025 ft 2 break with 2 HPT'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, vill result in acceptable consequences. With one HPI available, a slower depressuriza-tion is expected but the system evolution to high void fraction vill still be very slow. Thus, the conclusion that a RC pump trip at any ti=e yields acceptable consequences for the 0.025 ft2 break holds whether one or two ETI's are assumed available.
The LPI actuation time for the 0.025 ft 2 break can he 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 ft break 2 under Appendix K assumptions.
- 3. Critical Time Window for RC Pumos 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 . indows 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 cecur 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 c sne order of two to three minutes after ESFAS.
- 4. " Realistic" Evaluation of Impact of Delayed RC Pumo Trip for a Small LOCA
- a. Introduction As discussed in the previous sections, there exists a combination of break sizes and RC pu=p trip times which will result in peak cladding te=peratures in excess of 2200F if the conservative requirements of Appendix K are utilized in the analysis. The analysis discussed in this section was performed utilizing
" realistic" assu=ptions and de=onstrates that a RC pu=p trip at any time vill not result in peak cladding te=peratures in excess of the 10 CFR 50.46 criteria.
1312 ?84
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- b. Method of Analvsis There are three overriding conservatisms in an Appendix K small break evalua-tion which maximizes cladding te=peratures. 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 is assumed to be at the LOCA limit value.
In performing a realistic evaluation of the effect of a delayed RC pu=p trip following a small 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 HPI and LPI systems were available. The axial peaking distribution was chosen to be representative of nor=al 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 nor=al operation, was utilized with these axial shapes. As such, the combination of radial and vorst axial peaking are expected to provide the e v % - expected kw/ft 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 highest 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 2 break in the pu=p dis-,
charge piping with the RC pump trip at the ti=e 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. TheCRAFT2results.forthiscaseandtheevaluationtechniques utilized are discussed in section II.B.5 of reference 1. A realistic peak e
ee6eem o em ge e
cladding temperature evaluation cf this case, which is discussed below, is ex-pected to yield roughly the highest peak cladding tc=perature for any break size and RC pump trip time. As shown in reference 1, uaximum 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 s= aller core uncovery times as der.onstrated in reference 1 and the preceeding sections. Use of 1.0 times the ANS decay heat curve would result in a similar reduction in core uncovery tine, approximate 1y .
200 seconds, for breaks in the 0.05 through 0.1 ft2 range. Thus, the core re-fill rate, uncovery ti=e, and peak cladding temperatures for the 0.05 ft2 case is typical of the worst casa values for the break spectrum.
- c. Results of Analysis Figure 14shows 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 90L The core initially uncovers and recovers approximately 375 seconds later. Using the previously discussed realistic assu=ptions the peak cladding te=perature for this case is below1900F. Therefore, the criteria of 10 CFR 50.46 is =et.
The te=perature response given above was developed in a conservative manner by comparing adiabatic heat up rates to maximam possible steady-state cladding temperatures. First, a temperature plot versus time is made up for each loca-tion on the hottest fuel assembly assuming that the assembly heats up M iabati-cally. Second, a series of FOAM" runs are made to produce the maxi =um sneac}y-state pin temperatures at each location as a function of core liquid volume.
FOAM calculates the =ixture 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 te=peratures 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 correlation5 , from the fluid temperature and steaming rate and the steady-state clad te=perature is obtained.
The FOAM data are then combined with the core liquid inventory history (derived from Figure 14:) to produce a maxi =um possible cladding te=perature as a function of time. This graph might be termed maxi =um steady-state cladding te=perature as a function of ti=e and decreases in value with time because the core liquid
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inventory is increasing. By cross plotting the adiabatic heat up curve with the maximum steady-state curve a conservatis e 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 evaluation techniques, there exists a combination of break site and RC pump trip times which result in a violation of 10 CFR 50.46 limits.
- b. Prompt tripping of th'e RC pumps upon receipt of a low pressure ESEAS signal will result in cladding temperatures which meet the criteria of 10 CFR 50.4y.
The nin4-" time available for the operator to perform this function is 2 to 3 minutes.
- c. Under realistic assu=ptions, a delayed RC pump trip following a s=all break will result in cladding te=peratures in ce=pliance with 10 CFR 50.46.
\ ,3 \ L 7 87
r REFERENCES 1 " Analysis Su==ary 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 Bulletin 79-05C Action 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 MRt Internals Vent Valve Plants, BAU-10064, Babcock
& Wilcox, April 1978.
5 R.H. Stoudt and K.C. Heck, THETAl-B - Computer Code for Nuclear Reactor Core Thermal Analysis - B&W Revisions to IN-1445, (Idaho Nuclear, C.J. Hocevar and T.W. Wineinger), BAW-10094, Rev.1, Babcock & Wilcox, April 1975.
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1 Figure 1 : 0.30 FT2 BREAK -@ P.D.,SYS'.'EM VOIO FRACTION VS TIME 100 - ,______-- .,
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>. I -.- - PUMPS OFF e 83% VOIO 40 7
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i i e 0 i e i i i i e 0 200 400 600 800 1000 Time, sec e
1312 289
Figure 2 : 0.30 FT2 BREAK @ P.D.,AVAILABLE LIQUID VOLUME IN RV VS TIME, 1 M?I AVAILAELE e
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- Figure 3 : 0.20 FT2 BREAK e P.D.,SYditM VOID FRACTION VS TIME 100 - , , , ,
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O 400 800 1200 1600 2000 Lg Time, see 1312 201
Figure 4 : 0.20 FT2 BREAK e P.D.,AVAILABLE LIQ. VOL.
IN RV VS TIME 1 HPI AVAILABLE 3000 -
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Figure;5 : 0.20 FT 2 BREAK e P.O. AVAILABLE LIQ. VOL.
IN RV VS TIME 1 HPI AVAILABLE 3200 -
3000 - ,
- I g--x--x PUMPS OFF e g 2800 -
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1200 1600 2000 O 400 800 Time, see 1312 393 4 M*
Figure 8 : 0.1 FT2 BREAK :- P.D., AVAILABLE LIQUID VOLUME
. IN RV VS TIME, 1 HPI AVAILABLE 3000 .
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Figure 7 : 0.1 FT2 BREAK c P.D., SYSTEM VOID FRACTION VS TIME LPI 100 - l c -
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/ -o-o PLNP TRIP s 40% VOID 20 -
O i e i 1 0 500 1000 1500 2000 2500 .
T.ime, sec
Figure B: SYSTEM VOIO FRACTION VERSUS TIME g .1 9-40 - s% f h /
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35 - 6 '# 3/ @% /
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Figure 9 : SYSTEM VOID FRACTION VERSUS TIME PUMPS RUNMING AND PUMPS TRIPPED WCDEL 40 -
s% g-h ,A 35 - O. 4 g
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0 0 100 200 300 400 500 Time, sec e
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EXTRAPOLATED LPI. FOR 0.025 BREAK 0
3000 4000 5000 6000 7000 8000 9000 0 1000 2000 .: -
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Figure 11 : CRITICAL REGION FOR RC PLNPS TRIP, BREAK SIZE VS TIME ,
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Figure 12 : " REALISTIC" CORE AXIAL PEAKING DISTRIBUTION-CASE 1 1.6 -
x 1.4 -
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a 1.2 -
3 x
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Figure 13* " REALISTIC" CORE AXIAL PEAKING DISTRIBUTION-CASE 2 ;
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x 1.4 -
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a u 1.0 -
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a 0.8 -
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0 20 40 60 80 100 120 140 160 Core Elevation (inenes) '
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Figure 14 : AVAILABLE LIQUID VOLUME VS TIME FOR 0.05 FT2 BREAK WITH 1.0 ANS DECAY CURVE 3000 -
"s u-2500 -
ai 5
o
> R 3 2000 -
LIQ.LEVELeCOREh _ _
j 9' ELEVATION
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S00 ' ' ' ' i 1600 1800 2000 2200 2400 2600 Time, sec . 1312 302
. .o Docket No.30-346 Licensa No. NPF-3 Serial No. 1-91 September 20, 1979 ATTACIDiENT B IMPACT ASSESSMENT OF A RC PUMP TRIP ON NON-IDCA EVENTS 1312 703 e
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' 79-as c W/77 .
LY k. SIMPACT ASSESS'ir' 7 OP A RC PO*P TRTT' III.
?!0'-LOCA EVCOS 0::
A. Introduction Some Chapter 15 events are characteri:cd by a primary system The Section 15.1 response similar to the one following a LOCA.
events that result in an increase in heat removal by the ;4condary system cause a primary system cooldown and deprest:urization, r.uch Therefore, an asressment of the conse-like a small break LOCA.
quences of an imposed RC pu=p trip, upon initiation of the loJ RC pressure ESFAS, was made for these events.
B. General Assesment of Punn Trin in Nen-LOCA Events that Several concerns have been raised with regard to the effect exhibit LOCA an early pu=p trip would have on non-LOCA events that Plant recovery would be more difficult, dependence, characteristics.
on natural circulation r.ede while achieving cold shutdewn vo'uld be highlighted, manual fill of the steam generators would be required, ,
and so on. However, all of these drawbachs can be acce==odated since Also, none of them vill on its own lead to unacceptabit. consequences.
restart of the pumps is reco= mended 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 pu=p trip could reduce the time to system fill /repressurization If or saf ety valve opening following an overcooling transient.
the time availabic to the operator for controlling EPI flow and the margin of subecoling were substantially reduced by the pu=p trip to where ti=ely and effective operato,r 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-pccially if any degree of return'to power is experienced.
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- 3. A more general concern exists with a large steam line break c conditions and whether or not a return to power is experienced following the RC pump trip. If a~ return to . critical is e.xperienced, natural circulation flow may not be sufficient to remove heat and to avoid core damage.
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Owerheating tvents were not considered in the impact of the RC pump trip since they do not initiate the low RC pressure ESFAS, In addi-and therefore, there would be no coincident pu:p trip. tion, these events typically do not result in an c=pty pressuri:cr Reactivity or the formation of a steam bubble in the primary system. In addi-transients were also not considered for the same reasons. tion, for overpressurization, previous analyses have shown that for e the prescure the worst case conditions, an RC pu=p trip will mitigat. 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 Ceneerns and Results
- 1. System Reeressurization In order to resolve this concern, an analysis was performed for a 177 FA plant using a MINITRAP m.edel based on the case set up for IMI42, Figure 3.1 shows the noding/ flow path ~
scheme used and Tabl,c3J provides s description of the nodes and flow paths. This case assumed that, as the result of a ' s=all steam line bresh (0.6 ft. split) or of some ce=bination of secondary side valve failure, secondary side heat demand This increase was increased from 100% to 138% at time zero. in secondary. side heat de=and is the smallest which results in a (high flux) reactor trip and is very st:ilar to the vorst moderate frequency overcooling event, a failure of the steam pressure regulator. In the analysis, it was assumed that following EPI actuation on low RC pressure ESTAS, main f eedwater is ra= ped down, MSIV's sh t, and the auxiliary feedwater initiated with a 40-second delay. This action was taken to stop the cooldown and the depressurizatic,n of the system as soon as possible after EPI actuation, in order to , minimize the time of refill and repressurization of the system. Both HPI pu=ps were assume 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 HPI initiation. The analysis shows that the In system behaves very si=ilarly with and without pu=ps. both cases, the pr,essuri:cr refills in about 14 to 16 ninutes from initiation of the transients, with the natural circula-13i2 305
# tion case refilling about one minute before the case with two of four pumps running (See Figures 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 (ref er to Figure 3.4).
- It is concluded that an RC pu=p trip folloving HPI 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 Although no case control EPI prior to saf ety valve opening.
with all RC pumps was made, it can be inf erred from the one loop ease (with pu=ps running) that the subcooled margin will The be slightly larger for the all pu=ps running case. pressurizer will take longer to fill but should do so by l'6 minutes into the transient. Figure 3.+ shows the coolant temperaturer (hot leg, cold. leg, and core) as a function of time for the no, RC pumps case. ,
- 2. Effect of Stee : Bubb'le on h*atural Circulatien Cooline For this concern, an analysis was performed for the sa=e generic 177 FA plant as outlined in Part 1, but assu=ing that DIR), the as a result of an unmitigated large SLB (12.2 ft.
excessive cooldown would produce void formation in the pri=ary
' system. The intent of the analysis was to also show the As in extent of the void f or=ation and where it occurred.
the case analyzed in Part 1, the break was sy==etric to both generators such that both would blow devn equally, taxi- Hing the cooldown (in this case there was a 6.1 f t. " break on ea loop). There was no liSIV closure during the transient on either stcan generater to taxi =1:e cocidown. 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 (bo(h pu=ps), tripped RC The AFW pumps (when applicable) and isolated the MFWIV's. was initiated to both gene /ators on the low SC pressure signal, with mini =um dcicy ti=e (both pu=ps operating) . This an ysis was perfor=cd twice, once assuming all RC pu=ps running, once- with all pumps being tripped on the HPI In actuation (nf ter ISTAS), with a short (s5 second) delay. both cases, voids were f ormed in the het icgs, but the dura-3312 306
~
- i tion and size vere =maller for the case with no RC pump trip (refer to Tigure 3.7).Although the RC pump operating case had a higher cooldown rate, there was less void f orma-The tion, resulting from the additional system mir.ing.
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 out1ct pressure and SC and pressuri:cr levels versus time are given for both cases in Figures 3.S. 3.9. This analysis shows that the system behaves similarly with anc! vithout pumps, although maintaining The RC pump flow does seem to help mitigate void formation. pump flow case shows a shorsar time to the stnre of pres-surizer refill than the natural circulation case (Figure 3.9), although the ti=e difference does not seem to be very large. Since the volume of the hot leg locp above the lowest point in the these steam fgemations candy cane portion is about 63 cubic feet, have the potential for blocking natural circulation in the hot leg loops. As a result of these findings and since TRAF had not been programmed to closely follow this specific condition, an additional steam It is based on the unmitigated 12.2 f t TRAP case was run. line break with RC pump trip, since this case represented the bound-ing event for steam formation. This case includt. a mor_e 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 secam for=ation on natural circulation in the loops could be observed. The noding and flow path sche =e used in this model is shown in Figure 3.10. Table 3.2 provides a description of these nodes and flow paths. Figure 3.11 details the hot leg - candy ecne - upper steam generator shroud noding and flow path model superim The flow path positions and over a scaled figure of those regions. sizes were carefully chosen to allow for countercurrent steam and This model is consistent liquid flow at the top of the candy cane. with that used for the small break LOCA analyses described in Sec-tion 6.2.4.2 of Ref. 5. . of this analysis showed steam formation only in the The resulta These steam volumes are pressurizer loop (refer to Figure 3.12). conservative since they include all of the steam that was calculated aa 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 13i2 307
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 transf er 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 unnecessary conscr-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 pressurizer loop to 45 to 100 lb/sec f res 250 to 360 seconds (ref er to Figure 3.13), with flow steadily increasing af ter this cima period. The flow in the non-pressurizer loop remained relatively unchanged at about 10031b/sec (refer to Figure 3.14). Core flow was maintained frem 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 subcooled margin achieved in the pressurizer loop (refer to Figure 3.16). Bec'h ste$n generators and the pressurizer established level and the system pressure was turned around from the HPI 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 F8AR, Section ISD) has shown that a RC pump trip will mir.igate the consequences of an EOL return to power condition bj reducing the cooldown of the primary system. The reduced cooldown substan-tially increases the suberitical margin which, in turn, reduces or eliminates return to power.
D. Conclusions and Summarv A general assessment of Chapter 15 non-LOCA even'ts identified three areas that warranted further investigation for tapact of a RC pu=p trip on ESFAS low RC pressure signal.
- 1. It was found that a pump trip does not significantly shorten the rLae to filling of the pressurizer and approximately the same time interval for operator action exists.
}3)2 !)Ob I .
- 2. For the maximum overcooling case analyzed, the RC pu=p trip increased the amount of void for=ation in the hot leg ', candy cane' of the pressurizer loop; however, natural circulat' ion was not completely blocked. The steam bubble was collapsed and full natural circulation was restored. Core cooling was 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 ef f ect.
. 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 frem the analyses.for the lowered loop pl.nts are applicable.
e e e O l
-19a-
HINITRAP2 NODE DESCRIPTION t NODE NU:!BER DESCRIPTION Reactor Vessel, Lover Plenum 1 Reactor Vessel, Core 2 Reactor Vessel, Upper Flenur. 3 Hot Leg Piping and Upper S. C. Shroud 4,10 5-7,11-13 Primary Steam Generator Tube Region Cold Leg Piping 8,14 9 Reactor Vessel Downconer 15 Pressurizer 16,24 Steam Generator Downcomer Steam Generator Lover Plenum 17,25 Secondary, Steam Cencrator Tube Region 18-20,26-28 21,29 Steam Risers 22,30 Main Steam Piping 23 Turbine 31 Containm ent MINITRAP2 PATH DESCRIPTION DESCRIPTION
- PATH NUMBER Core 1
Core Bypass 2 Upper Plenu=, Reactor Vessel 3 4,11 Hot Leg Piping Hot Leg Piping and Upper S. G. Shroud 5,12 6,7,13,14 Primary, Steam Generator RC Pumps 8,15 9,16 Cold Leg Piping Downcomer, Reactor Vessel 10 Pressurizer Surge Line 17 18,19.26,27 Steam Generator Downcomer 20.21,28,29 Secondary, Steam Generator 22,30 Aspirator r Steam Riser, Steam Generator 23,31 Main Steam Piping 24,32 Turbine Piping 25,33 34,35 Break (or Leak) Path 36,37 HPI 38,39,43,44 AFW Main Feed Pu=ps 40,41 *
- LP1 42 .
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MINITRAP2 NODE DESCRIPTION DESCRIPTION NODE NUMBER Reactor Vessel, Lower Plenum 1 Reactor Vessel, Core 2 Reactor Vessel, Upper Plenum 3 Hot Leg Piping (including ' Candy Cane') 4,10 ' Candy Cane' and Upper S. G. Shroud 32,33 Primary, Steam Generator Tube Region 5-7,11-13 Cold Leg Piping 8,14 Reactor Vessel Downcemer 9 Pressurizer 15 Steam Generator Downec=er 16,24 Steam Generator Lower Plenum 17,25 Secondary, Steam Generator Tube Region 18-20,26-28 21,29 Steam Risers Main Steam Piping 22,30 Turbine 23 Containment 31 MINITRAP2 PATH DESCRIPTION _ DESCRIPTION PATH NUM3ER Core 1 Core Bypass 2 Upper Plenum, Reactor Vessel 3 Hot Leg Piping 4,11 Upper Steam Generator Shroud 5,12 Top of Hot Leg ' Candy Cane' 45,'46,47,48 Primary Heat Transfer Region, S. G. 6,7,13,14 RC Pumps 8.15 Cold Leg Piping 9,16 Downcomer, Reactor Vessel 10 Pressurizer Surge Line 17 10.19,26,27 Steam Cznerator Downcomer and Plenes Secondary Heat Transfer Region, S. G. 20,21,28,29 Aspirator 22,30 Steam Riser, Steam Generator 23,31 - Main Steam Pi'p ing 24,32 Turbine Piping 25,33 34,35 Break (or Leak) Path HPI 36,37 AFW 36,39,43,44 Main, Peed Pumps 40,41 LPI 42 1312 3ii Table 3.2
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y m - - n (Eg) 'amni A at ng meals Figure 3.12 n 777 b ,] l ' l E-
+
e e
~
i w l _ m a , -
._ c=
w M
.-= ,=
z< z w w _ a_
.o - n z - e 2. - o
< m o -
m z o a .- e-- D J GAJ w = +
^
u, es m w w m o w o. a a, = . m es - o a: z r_e w w , o
- u. =
c a2 - m o. m n a zm o . C,
. V WD Q. ^ =
- u. o e e es a w, -
w u: >~ z
< N-- e-w cm = -
o u. x c
- a
- >.- w -
m N. a: z
- =
z ~ m c
< - w s ; =
o w w = -
- o , e-a.. =__ -
< u _, e- u.
. m o.
mw e< WW o, N w .s e a w a m o, - w v wm
- N D m
- a M -
= m m a .
m w w -
= w .
- n. -
, t h t t ' 8 O c, e es es = o es c3 c3 C O O C3 O O C3 o O c>
=> o c> c=. c=.
o. es. w ci
. ca.
c> m
,. w c=
(cas/t.'at) acij aue3 Xpuca ,y, dool , F.igure 3.13 . G
- 7',~)}
1512 e = *e **=-p-eo* *--e-=
e .. 6 . e e air e-w E - N W c w
>= >==
Z *st W c US >= Z = o 4 M = - C2: 2
>-= D
- em US o ^ v1
~3 LaJ A e w a - W o Z CI". =3 W W >= C *
>* * , to -
ua L E 92 J L2 v . o c3 :D J D %
- a3 6 o E o O -
W CI: >- ZN auC >= = w U h 3sd - CD C
.uC e
>= W --
C3 N. E un Z N CD C aut - eq CJ w n.
= Z >-
= Ch. - ,
CD a w J - 9 w Z.c . CL N w O Q >= . O e us - C w
- N t f ~_ t t
, e o o o o o o o o o C3 o o o o o o o o o o o o. a. o.
- o. o. o. o. .
o ,. N o to , N o co (cas/ mal) 3013 sues Apuca ,g, dool , Figure 3.14 1312 ?24
O s,
- e s
v i N= t.aJ C3 O E w Q".
- O
>= Z = "
C w hc -
>==
- aut ^
- Z haJ LaJ 6e w J C2: .c
= CO CD W to D 3 Z Q LaJ e C *
=rt G ZA E2:
>= N J CE:
CO VE
>= E >=
to w aut as D LaJ Ch. . E to >= 2 *== CE: N. 64 =3 >== baJ N C3
> - C H e.aJ C lir. = >= =
LD U C Q. asC - J La. C.3 M w - C b% >= t"J LaJ CN === t= CE: C 3 >= C === Z U V D
. - V
. I e
N e e f f f I% . , e o
, c= = a c= cm e e e J CD C C C C C C C o
=
=
. ca.
=
cm. e o. o o. e o. o cm. c= o. o o s CD N CD tt3 v F3
- N -
cas/Qi 'mol3 ajog Figure 3.15.. .
"9 W
. 1J -
% /
' L .)
e e * *
'N .
- n o. .
. a. e Q O Q .J J w
= =
a ~ -
~ = >.
= ~ z
- o. O V
a C.S w ..= < w a w .= , =
-. xo - -
O O at E W W M
~o w
e 4 ! I,
?
- w
.I wn -
a a w - w \ . x w w a a o t o ac w z
- o I m e = i N
fut W = M / a O O 4 - c cm - w w
= - o. a a #
i
- - u. W kJ a o o G W C.3
- w 2
m a, ._
,/ .
m- - z - < =w e i w == .- .o w
==
=
/
i o o c - g= g m -
/ o
.a a a. a= / =
e uJ O. " o O 4 - c .
'i w / c mw E.- m m -
w .w a. o / a o e
.o e _s x o e V w = =
- >. :s a. l H" of a O G M Q U m w *-
w
=N l
= . "c I - = c x = G =
u. o O g "., r w a w I - w n. m w C
- a. n w - ,- .
s d = m- w
- A o O -
w m x -
. z = w I I
m u .s o w a w P a o
< wm wmc w ~-
.a :
o o - e a e - m Io a 4 u w d5 m I
- w __ f - n o =
= a. Jo o G ,
/
1 l o c
' / w f I o
_-it , I w - o = o e a . o = e o o e o c
- g (J ) 'sajnlejadcraj,luetoo3
' / Figure 3.16 ...
e e M . s1 og {
~
s -
' . _a ,
' a =
a o
>= >=
- re a a LW W w w z w- .z w
= a o e n w
w - ~ < m m m s
. w w a w ~
a >- a. n.- m v w U OO G C w _ a
- n * <D .
;o O
<9
" a .
oc 5J e
-< a o 4 - ~
w w
. z m 0o 4 g= o oc
*W w - 0 DQ m
w m a o m a ^ w w D-o, 4l m
=
w a o - D
- w o w o=
w O 2
-s
" ^ o - o S -
a o. . O U o*aw a
. .- e a oo - 0 >-
.a o
- c. o ~
e a E O
- o e o e n. -
>- w -
o m
- - e m -
w a z o S
~
w t-o - o mw
- a w ,
.- w M > =
- w a .
z a .
~
= N. a w O
~ .
- a.
mW .O e a w w w m , ,.,. , e o w i a o_ - o
- a. -
o o o e o n m -
.a w
( u) tara) Jolejauas rJeals/Jazijnssajd , Figure 3.17 e, 1312 327 4 ,e. 9
m ~ Ww -
=
e ..- o_ g w M z o v, o _ z .
- a n a.
= w -
.- o - M z e4- =
to w W
= . =
V, WA C w a u - m - w CD D
> =o w v a
w o a. = M m . E
=~ =
to, Da= .- u w w w< *'* E w - co e A N. E = u co -.
.- - m w w e
- = a
.a- n. - -
w.w ca a - o m.c . W N w ce o a - u, o v _ m t e , - , , cm o e o c= e o e> o Q CD C CD o o o CD w a .C3 a a o o so <a a w n n n m - - o.
- o (Isd) aJnsssJd taltnD ajo3 Figure 3.18 1312 328 e &
a
*O- **N-W. m eg y , ,}}