ML19332B245
| ML19332B245 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 08/21/1979 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19332B231 | List: |
| References | |
| ISSUANCES-SP, NUDOCS 8009260352 | |
| Download: ML19332B245 (53) | |
Text
, Docket No. 50-289 9- (Restart)
Licensee's Exhibit No.
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Report, " Analysis Summary in Support of an Early RC Pump Trip," (August 21, 1979) 1 6g y 4 THIS DOCUMENT CONTAINS 9999 POOR QUALITY PAGES
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ANALYSIS Stwy.ARY IN SITPPC1t! oF .
AN FJJl.Y RC PUMP ~?.!? -
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ERRATA , , , ,
On Page 2. Second Pa'ragraph ~
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- Section A.3 should be Section C
- Section A.2 should be Section B
- Section A.4 should be Section D -
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- Section A.5 should be Section E "
Page !1 umber 4 s'hould be changed to Page Number 5.
Page. Number 5 should be changed to Page Number 6.
Page Number 6 should be cht.nged to Page Number 4. .
I Page Number 12 should be changed to Page Number 11. '
1 Page Number 11 should be changed to Page Number 12.
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ellALYSIS SIMMART 3 SUPPORT 07 AN EARLY RC Pt3fP TRIP e
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I. INTRODUCTION . . . . . . ..................... 1
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II. SMALL B2ZAK ANAI.TSIS . . . . . . . . . . . . . . . . . . . . . . . 2 l A. Intzuduction . . . . . . . . . . . . . . . . . . . . . . . . . 2 B. System Response With RC Pumps h mning ............ 2 )
i C. Analysis Applicability to Davis-Besse 1 ........... 11 {
D. Effect of Prompt RC Pump Trip on Iov Pressure ESTAS Signal . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E. Conclusions ......................... 13 j III. IMPACT ASSESSMENT OF A RC PUMF IRIP CN NON-LOCA E7ENTS . . . . . . 15 A. Introduction . . . . . . . . . . . . . . . . . . . . . . . . . 15
- 3. General Assessment of Pump Irip in Non-LOCA Events . . . . . . 15 C. Analysis of Concerns and Rasults . . . . . . . . . . . . . . . 16 D. Conclusions and Summary . . . .,. . .,. . . . . . . . . . . . 18 l
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, _ dNALYSIS SUMMART Di SUPPCPO OF AN EARE.T RC PUMP TRIP I. D:TRODUCTICN BW has evaluated the aff act of a delayed RC ptany trip during the coursa of small loss-of-coolant accidents and has found that an early trip of the RC pumps is required to show conformance to 1CCFR50.46. A summa.y of the LOCA analyses performed to data is provided 12 Section II. This, discussion includes:
- 1. A description of the models util1=ad.
- 2. Break spectrum results with continuous RC Pump Operation.
- 3. Break spectrum results with delayed RC pump trips including estimates !
l of peak etadd hi temperatures.
- 4. Justification that a prompt pump trip following ESPAS actuation on lov RC pressure providas LOCA mitigation. ,
l An inspect assessment cf the required pump trip on non 'JCA avants has also been completed and is prasanted in Section III. Ihis evaluation supports the use of a pump trip folleving ESPAS actuation for LOCA P.1:1gatica since no de :1aantal consequences on non-LOCA events were identifiad.
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II. SMALL 3REAK ANALYSIS .
A. Increductica Previous small break analyses have been performed assuming a loss-of-offsi:e power (reactor coolant pump coastdown) coincident with re-actor trip. These analyses supper: the conclusion that an early RC pump trip for a LOCA is a safe condition. However, a concern has been identified regarding the consequences of a small break transiant i=
which the RC pumps rema'n operative for some time period and then are lost by some means (operator action, loss-of-offsite power, equipment failure, etc.). This section contains the results of a study to further understand how the small break LOCA transient evolves with the RC pumps operative. Sper4es117, see:1on 3. describes the system response with the RC pumps running for B&W's 177-FA lowered-loop plants. In-cluded in this section is the development of the model used for the
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analysis, a break spectrum sensitivity study, and peak ehdding tem-parature assessments for cases where the RC pumps trip at the worst time. l l
Sectinn A.3 demonstrates the applicability of the conclusicis
( drawn in section A.2 to a 177-TA raised-loop plant (Davis-Bessa 1). .
The effect of a, prompt tripping of the RC pumps upon receipt of a low pressure ES?AS signal is discussed in section A.4. Finally, see-tion A.5 summarizes the conclusiens of this analysis.
- 3. System Reseense With RC pu=es R= '9st
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- 1. Introduction
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Recent evaluations have been performed to s e a the primary system response during small breaks with the RC pumps operative.
During the transient with the RC pumps available, the forced .
1 circulation of reactor coolant will maintain the core at or near 1 1
the saturated finid temperature. However, for a range of break j sizes, the reactor coolant system (RCS) will evolve to high void )
fractions due to the slow system depressurization and the high liquid (low quality fluid) discharge through the break as a re-sult of the forced circulatien. In fact, the RCS void frac: Lou will increase to a value in excess of 90: in the short ter=. In 2-see .
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the long term, tha sy:;taa void it:ction .rt11 d:creas2 as tha E.CS
. RCS 'depressurizes, HP1 flow increases, and decay heat Ad-d-d=hes.
i With 'the RCS at a high void fraction, if all RC pumps are postu-laced to trip, the forced circulation vill no longer be available and the residual liquid would not be sufficient to kaep the core covered. A *1=dddng temperature excursion would ensua until core cooling is reestablished by the ICC systems. The following para-graphs summarizes the results of the analyses which were performed for the 177-TA lowered-loop plants, to develop the consequences of this transient. !
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- 2. Method of Analvsis - I
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The analysis method used for this evaluation is basically that de-scribed in section 5 of BAW-10104. Rev. 3, "3&W's ECCS Evalua. ion Model"1 and the letter J.E. Taylor (3&W) to S.A. Varga (NRC), dated -
July 18, 19782 , which is applicable to the 177-FA lowered-loop plants for power levels up to 2772 Mit. The analysis uses the CIAT 23 coda to develop the history of *de RCS hydrodynamics. !
However, the CRAFT 2 modal used for this study is a modification '
of the small brea'k evaluation model described in the above ref-erances. Figure 2-1 shows the CRAIT2 noding diagram for small breaks
- rom the above referenced letter. The modified C3A m l model consists of 4 nodes to simulate the primary side, 1 node for I the secondary side of the steam generator, and 1 node representing t!.e reactor building. Figure 2-2 shows a schematic diagram of this model. Node 1 contains the cold leg pump discharge piping, downcomer, and lower plenum. Node 2 is the primary side of the SG and the pump suction piping. Node 3 contains the core, apper ple-nun, and the hot legs. Node 4 is the pressuri=ar and nodes 5 and
- 6 represent the reactor building and the SG sarand=ry side, re-spectively. This 6 node model is highly simplified co= pared ec those utilized in past ECCS analyses. It does, however ain ain RCS volume and elevation relationships which are important to properly evaluate the system response during a small break vi h the RC pu=ps t- --bg.
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Fallowing tripping cf ths RC pump 3 and the subtequent lost-of-
. forced circulation, the system will collapse and separate.
The residual liquid will then ec11ect in the reactor vessel and the loop seal in the cold les suc. ion piping. For this period of the transient, the Wilson bubble rise model is u** ed.
The homogeneous assumption for the period with the RC pumps operating applies to modes 1, 2, and 3 i= the C2 AFT model.
Node 4, the pressurizar, and node 6, the secondat7 side of the steen Sensrators, utilize the Wilson bubble rise model throughout the transient as these nodes are not in the direct path of the forced cittulation.
- 3. Benchmarking of the 6 Node CRAFT Model Studies were performed to compara the results of the 6 node model to the more extensive evaluation model for B&W's 177-FA lowered-loop plants as described in the letter J.H. Taylor (3&W) to S.A.
Varga (NRC), dated July 18, 1978. The break size selected for this comparison is a 0.025 ft2break at pump discharge. This break represents the largest single-ended rupture of a high energy line (2-1/2 inch sch 160 pipe) on the operating plants. The break can be viewed as " realistic" or the worst that would be ex-pected on a real plant. Figures 2-3 and 2-4 are the results of this comparison. System pressure and percent void fraction shown in Figures 2-3 and 2-4, respectively, compara very well with those from the more extensive (23 nodes) CHAM small break model. As seen in these figures, the difference is not significant and is less than a few percent. The computer time for this 6 node model is, however, significantly decreased. The model ut*1* ad for this study is thus justified based on comparison of results to the more extensive small break model and desirable because of its economical run time.
- 4. Analvsis Results The break sizes = = dved for this analysis ranged from 0.025 ft 2 to 0.2 ft2 in area and are located in the m discharge piping.
Breaks of this si=e do not result in a Tapid systa= depressuri-zation and rely pred H "=ntly upon the EPIs for sitigation. l l
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Tablo 2-1 summarizes the analyses performed for this evaluation.
The majority of the analyses performed utilized 2 HPI pumps through-out the transient. The effect of u*dving 1 HPI pump is discussed in this section.
Figures 2-5 and 2-6 show the system pressure and average system void fraction transients for the break spectrum analyzed assuming continuous RC pump operation and 2 EPI's available. In Figura 2-6, the average system void fraction is defined as V3-V2 Average system void, I = y x 100 1
where V1 = total primary liquid volume excluding the pressuri-ser at tima = 0, V2 = total primary liquid voluna - l a g the pressuri-ser at time = t. -
This paramatar was utilized in place of the mixture height in that
, the coolant vill tend to be homogeneously mixed with the RC pumps operative. Under these assumptions, the core is cooled by forced circulation of two-phase fluid and not by pool boiling as in the case where the RC pumps are not running and separation of steam and varar occurs. As shown in figure 2-5, the system pressure re-sponse is basically independent of break size during the first several hundred seconds into the transient. This occurs because the forced circulation of reactor coolant maintains adequate heat transfar in the steam generators; the primary system thus depres-surizas to a pressure (about 1100 psia) cos-ponding to the see-ondary control pressura (i.e., set pressure of SG safety relief valves). After some time (250 seconds for the 0.1 ftt break), the
- l system pressure vill decrease as the break alone relieves the core !
energy.
Figure 2-6 shows the evolution of the system void fraction; values i in excess of 90% are predicted very early (300 seconds) into the l
t transient. For the larger breaks the system high void fractions occur early in ime. For the smallar breaks it takes in the crder of hours before the system evolves to high void fracticn. Core cooling is maintained during a small break with conti=ucus RC pu=s umano e
- The breaks analyzed in this section are assumed to be located in the cold leg piping between the reactor coolant pump discharge and the reactor vessel. Seccion 3.7 demonstrates that this is the worst break location. Kay assumptions which differ from those de-scribed in the July 18, 1978, letter are chose concerning the equip- ;
ment availability and phase separation. These are discussed below.
- a. Zouipment Availability The analyses which were performed assumed that the RC pumps re-For select cases, main operative after the teactor trips.
j after the system has evolved to high void fractions (approzi-Also, the in-nately 90%) the RC pumps were assumed to trip.
l pact of 1 versus 1 EPI systems for pump injection were examined. ]
The majority of the analyses performed assumed 2 EPI pumps. I However, as is demonstrated later, even with 2 EPI pumps avail-able, cladding temperatures will exceed the criteria of 10 CIR Therefore, fur-50.46 using Appendiz K evaluation techniques.
ther analysis with only 1 HPI pump would only be acade=ic.
- h. Phase Seearatien .
The present ICCS evaluation model created to evaluate l
small breaks without RC pumps operative,(quiescent RCS) uti-lizes the Wilson bubblarise correlation for all primary sys-
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I In this analysis, tem control volumes in the CRAIT evaluation.
for the time period that the RC pumps are operative, the pri-i.e., no mary system coolant is assumed to be homogeneous, In reality, the flow ratas phase separation in the system.
in the core and hot legs are low enough that slip vill occur.
This will cause an increased liquid inventory in the reactor
- vessel compared to that *=1 ~1= tad with the homogeneous model.
With the homogeneous assumption, core fluid is continuously circulated throughout the pri=ary system and a portion of that fluid is lost via the break. During the later stages of the N eient, a slip =cdel will result in fluid being trapped in the reactor vessel and the hot legs. The only method of losing liquid during this period vill be by boiling caused by he core decay heat. Thus, the assumption of homogeniety for the period with the RC pumps operative is conservattve.
s In the long term, the sys-operation regardless of void fraction.
tem vill depressuri=e and the achavu ed parformance of *he ZCCS (EPI and LPI) vill result in reduced systaa void fraction.
2 Figure 2-7 illustrates this long tarm system behavior for a 0.10 ft break. For this case, the LPIS are operative at approximately 2300 seconds, and a substantial decrease in system void fraction results. -
An arbitrary pump trip af ter approximately 2700 seconds would not result in core uncovery. The potent,1al for core uncovery due to an RC pump trip is thus limiend e, a discrets time period during which the natural evolution of the system produces high void frac-tions and prior to LPI actuation. For a 0.1 ft2 break, this time period is on the order of 2000 seconds. Tor sas11er_ breaks, this critical time could be a fav hours even if the operator i=1tiated a controlled cocidown and system depressurization as ree-aded .
in the small break g"dd='dnes.
Although the analyses described above used 2 EPI pumps, the effset of only 1 EPI pump available on the system void fraction evolution while the RC pumps are operating is not significent. Figures 2-8 and 2-9 show the impact cf one versus two EPI pumps on system pres-sure and average void fraction transients Sr a 0.05 ft2 break with the RC pumps operative. As seen from thess figures, the results with one EPI pump are not significantly different to the two EPI With pump case and are bounded by the spectrum approach utilized. -
one EPI pump, the system does depressurize more slowly (less steam condensation) and a higher short term equilibrium void fraction is achieved. Also, recovery of the core follovirg a loss of the RC pumps would be significantly longer with only 1 EPI pump avail-able. .
The majority of the analyses provided in this report uses two EPI pumps and demonst: stas a core cecling probism with verst time pump trip given that assumption. As analysis of one H?I available cases would only show a larger problem, such cases have not been exten-sively considered. As demonstrated in section 3.4, the resolutien of this problen, forced early pu=p trip, provides assu.ance of Therefore, core cooling for both one or two EPIs available cases.
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there is no need for further pursuit of the single EPI available case. l The effect of the RCP tripping during the transient was studied by assuming that the pumps are lost when the system reachas 90 void fraction. Loss of the RC pumps at this void fraction is expected to produce essen 1any the highest peak cladding temperatura.
After the RC pumps are tripped, the fluid in the RCS separates and liquid falls to the lowest regions, i.e., the lower plenum of the 17 sad the pump suction piping. At 90* void fraction, the cora vill be tocany uncovered following the RC pump trip. Thus, the time required to recover the cora is longer than that for RC pump trips initiated at lower system void fractions. System void frac-tions in excess of 90: can possibly result in slightly higher tam-peraturas due to the longer ccre redill times that may occur.
However, the peak cladding temperature results are not expected -
to be significantly different as the system pressure and core de-cay heat, at the time that a higher void frae:1on is reached, v'"
l be lover. , ,
Table 2-2 shows the cara uncovery time for the cases analyzad with the RC pumps tripping at 90* void fraction with 2 HPI pumps avail-
. able for core recovery. As shown, the cora vill be uncovered for approximately 600 seconds for the breaks analyzed. Figures 2-10 and 2- u show the system pressure and void fractics response for the 0.075 fc2 break with a RC pump trip at 90: void fraction. As seen in these figures, the system depressurizas fastar afcar the RC pump trip, due to the change in leak quality, and the void fraction decreases indicating that the core is being refilled.
, Figure 2-12 shows the cora liquid level response following the RC l pump trip. The core is refined to the 9 foot level with conapsed liquid approxtmately 625 seconds after the assumed pump trip.
Once the core liquid level reaches the 9 foot elevation, the cora is expected to be covered by a two-phase mixture and the cladding temperature excursion would be ta- ihmted.
= - . . . m. .- . . . . . . . . . . . . . . . . .
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, 5 .- Effect of i.O ANS versus 1.2 ANS Decav Curve An analysis was performed using the more realisric 1.0 ANS decay The study was done for a curve instead of 1.2 ANS decay curve.
0.05 ft 2 break with 2 EPI;s availabis and pumps crdpped at 90:
71gures 2-13 and 2-14 show a comparison of system void frae:1on.
system pressure and average system void frac. ion for 1.0 and 1.2 As seen in Figure 2-13, the system pressure ANS decay curves.
for 1.01.NS case basins to drop from satura*. ion pressure 61100 psia) about 200 seconds earlier than the case with 1.2 ANS as a Also, the system will evolve to a result of reduced decay heat. After the .
lower average void fraction as shown in figure 2-14.
pumps trip at 90 system void fraction, the case with 1.0 ANS decay curve has a shorter core uncovery time by approximately 200 sec-ands compared to 1.2 ANS case. This casa demonstrates that the '
effect of a delayed E pump crip may be acceptable when viewed realistically. A Peak M AA4ng camperature assessment for t W case vill be provided in a suppiamentary response pinnned for September 15th, to the I&E Ju11stin 7905-C.
- 6. Effect of No Auxiliarv Feedvater Analyses have also been performed with the E punps available These analyses all assumed 2 EFI
~ and no a M idsry feedvatar.
The systen void fracrion evolutions for pump's were available.
these a=le"1=tions were not significantly dddfarent from those discussed with auxiliary feedvatar. Thus the conclusions of the cases with a'*14 =7 feedvater apply.
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.2 Break Location Sensitivity ll:ed A study was conducted to demonstrate that the break location uti Ar for tha preceeding analyses is indeed the vorst break location.
k I stated previously, the analyses were performed assuming that the brea A 0.075 ft was ligated in the bottaa of the pump disctarge piping.
hot les break was analyzed to provide a direct comparison to a similar For this evaluation, the RC pumps were assumed case in the cold leg.
Figure 2.15 shows the to trip after the RCS void fraction reaches 90%.
i for average system void fraction transient and the core uncovery t mes 2
As shavn, the cold leg break both the 0.075 ft bot and cold les breaks. hot reaches 90lll void fraction approximately 150 seconds earlier than the .
leg break. Also, he cold lag break yields a core uncovery time of
-he quickar core recovery 175 seconds longer than the hot les break.
f the time for the bot les. break is caused by the greater penetration o For a cold leg break in the pump discharge BPI fluid for tht.s break.
piping, a portion of the EP! fluid is Inst directly out the break and is not available for core refill. ' For a hot leg break, the full EPI Thus, as shown by direct comparison flow is available for core refill.
and for the reasons given above, het leg breaks are less severe than breaks in the pump discharge piping. .
8 a d Cladding Temeerature Assessment _ -
As described previously, a RC pump trip, at the tb the RCS void fraction is 902, will result in core uncovery times of approximately The peak 4 add *ng temperatures for these cases were 600 seconds.
d evaluated using the small break evaluation model core power shape use Also, an to demensrrate compliance with Appendix I and 10CFRf0.46.
ilised.
adiabatic beacup assumption during the time of core u=covery was ut This approach is extremely conservative in that the power stage and
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. I C. p=1* sis Antlicabilitv to Davis-3rns1 The significant parametric differences between the raised-loop l is Davis-Basse I plant and the proceeding generic lowered-loop ana ys d the amount are in the high pressure injection (EPI) delivery rate an l
of liquid volume which can effectively be used to cool the core. l fferenca; l The liquid volume differential is due to the basic design di Baeare of the raised design, system raised versus lowered loops.
the reactor votar available af ter the RC pumps trip will drain into vessel.
For the lowered loop designs, the svaflable water is split
- hus, for between the reactor vessel and the pump suction piping.
d core liquid level the same average system void fraction, the collapse ~
following an RC pump trip is higher for the raised loop design th the lowered loop design.
Davis.-
Figure 2-16 shows a comparison of the delivered HPI flow for the
. As shown, for a similar Besse I plant and the lowered loop plants. .
deliver-number of EPI pumps available, the Davis-Besse I pumps vill For the delsyed pump trip cases presented in section 3.4 more flow. l 450 of this report, the Davis-Besse I plant will taka apprcximate y h lowered-seconds to recover the core as opposed to :600 seconds for t e However, it is noted that the core recovery time is based loop plants. Use of on using two HPI's rather than one, as required by Appendix K.
y times -
only one HPI pump for Davis-Besse I will result in core uncover The Davis-Basse I plant cannot be shown to be in excess of 600 seconds.
in comptinnes v1th ICCT250.46 for a delayed RC pump trip.
Prompt reactor coolant pump trip is, therefore, necessary to ensure compliance of the Davis-Basse I plant vi h 10NO.46.
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- . j local power rate (kw/ft) analyzed is nec expected to occur during
=Ad=k=ri e heacup P e he - re, use of an normal plant operation.
h ill occur assumption negisets any credit for the steam cooling t at w l
during the core refill phase and also neglec:s the effect of any !
Using a decay heat power level based on 1.2 radiation heat transfer. 6.5 ANS at 1500 seconds, the cladding vill heatup at a ra:e vill be With a core uncovery period of i F/S under the adiabatic assumption. l 600 seconds and the adiabatic haatup asstamption, eladding tempera l
Use of a more realistic hea vill exceed the criteria of 10CFt50.46. l ilized for this eval- \
transfer approach with the extreme power shape ut '
s of nation is also expected to result in c3=AA4ng temperature in exces
- i' In order to ensure compliance of the 177 FA lowered the criteria.
loop plants to the criteria of 10CFR50.46 a prompt tripping of the Section 3. demonstrates that a prompt trip of
-RC pumps is required.
i l vill resul I the RC pumps upon receipt of a low pressure ESTAS s gna in compliance to the criteria.
An evaluation of the peak cladding tamparature using a power shape i t response encountered during normal operation for a realis:ic trans en 15, 1979.
with delayed RC pump trip vill be provided by September O
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.D. Effect of Promet RC Pume Trip on T.aw Pressere ISTAS Signal As demonstrated by the previous sections, the ICC system can not be denonstrated to comply with 1 % 7R50.46 us1=g present evaluation techniques and Appenr!h K assumptions under the assumption of a i
l delayed RC pump trip. Thus, promnt =1pping of the RC pu=ps is necessary to ensure conformance. Operating guidelines for both LOCA and non-LOCA events have been developed which require prompt tripping of the RC pumps upon receipt of a low pressure ESTAS signal.
1 Because no diagnosis of .he event is required by the operator and ? STAS initiation is =1=M in the control room, prompt =1pping of the RC
)
pseps can be assumed. 1 The effect of a prompt reactor coolant pump =1p on an ISTAS signal has 1 been - 4 to ensure that the consequences of a small LOCA are l bounded by p;evious small. break analyses 2which assume RC pump trip on .
reactor trip.- As shown by Table 2-3 at the time of low pressure ESTAS l initiation, keeping the RC pumps running results in m lower average )
system void fraction. This occurs because the availability of the RC pumps results in lower hot leg temperatures and thus less flashing in the RCS at a given pressure. Thus, a prompt trip upon receipt of an f I
)
ESTAS signal vill result in a less severe system void fraction evolution I l
than cases previously analyzed assuming RC pump on reactor trip.
l E. Conclusions The results of the analyzes described in this section can be suszarized as follove:
- 1) If the RC pumps remain operative, core cooling is assured regardless of system void frac ion.
- 2) For breaks greater than 0.005 ft , che RCS may evolve to systa= void fractions in exccas of 90::.
2
' 3) At 40 =inutes, ths 0.025 ft break has ev:1ved ta only c 47: v:id fract:o e Taus, a delayed 3C pump trip for breaks less than 0.025 f t I will not tesult in core uncovery.
- 4) ne potential for high cladding temperatures for a small break transient with delayed F.C pump trip is restricted to a time period betseen that time whers the sytten has evolved to a high void fraction and the time of L?I actuation.
5)
Even with 2 EP! pumps available, tripping of the EC pumps at the worst time (90 void fraction) results in a core uncovery period which cannot be shown to comply with 10CTK50.46, if Appendix K assumptions are utilized.
- 6) A prompt RC purup trip upon receipt of a low pressure ESTAS signal will provide compliance to 10CTR50.46.
7) ne above conclusions arm applicable to both the B&W 177 7A lowered and raisad loop 3tSS'desigrs.
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ON ':0N-LOCA 1~/D CS_
b!. IMPACT __ Ass *SSTr OF A RC !c:o TRIS A. Introduction tem Some Chapter 15 events are characteri=ed The Section 15.1 by a primary sys response similar to the ons fenoving a LOCA. by the secondary events that result in an increase in heat i=ation,smoval much system cause a primary system cooldosa and depress like a small break LOCA. i of the quences of an imposed RC pump trip, upon initiat on lov RC pressure ESTAS, was made for these events.
3.
General Assessnent of Pume Trip in Non-LOCA Eventsffect _ that Several concerns have been raised with h regard xhibit LOCA to the e en early pump trip would have on non-LOCA eventsdependence. t at e Plant recovery would be more difficult, characteristics. d shutdown would be on natural circulation mode while achieving col i d highlighted, manual fill of the steam generators would be req
. However, an of these, drawbacks can beces. accousnodated Also, and so on.
none of them win on its own' lead to unacceptable l andconsequen cocidown restarr of the pumps is not precluded for plant Out of contro this search,
- ence centroned operator action is assumed. d to be sub- .
. three major concerns have surfaced which have appeare stantial enough as to require analysis: i ion
- 1. A pump trip could reduce the time to system i fill /repressur If or safsty velve opening ionoving an overcooling HPI flow trans andent. l the time available to the operator for controningd by the pu=p l the margin of subcooli=g were substantiany reduceld be trip to where timely and effective operator action cou -
questionable, the pump trip would become unacceptable. ) h 2.
In the event of a large steam line break.(mmHmum, i overcoo blevdown may induce a steam bubble in the RCSes-which cou l J
natural circulatten, vich severe consequences on the core, /
if any degree of return to power is experienced.
pae' = ' 17 "!OL f 3.
A more generni concern exists with a large steam i line d breakl conditions and whether or not a return to power is exper ence f If a' return to critical is experienced, fonowing the RC pump trip. heat and f i
natural circulation flow may not be sufficient to re=ove I to avoid core damage.
M- O J
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l Overheating events vars not considered in the impact of :h3 BC pump trip since they do not initiate the lov RC pressure ISTAS, and therefore, there would be no coincident psamp rrip. In addi- .
tion, these events typically do not result in an empty pressurizer or the formation of a steam bubble in the prd=ary system. Rese:ivi:y :
transients were also not considered for the same reasons. In addi- 1 tion, for overpressurization, previons 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 redue:1on in l pressure at the RC pump exit vnich occurs af:er trip.
l C. Analysis of Ceneerns and Results 1
- 1. System Repressu-1:stion l
In order to resolve this concern, an analysis was perfcrmed i for a 177 FA plant using a !CNITRA.P model based on the case l set up for TMI42, Figure 3.1 shows the noding/ flow path i scheme used and Tab 1pl.1 provides s description of the nodes" and flow paths. This case assumed that, as the result of a e
small steam line break (0.6 f t." split) or of some combination of secondary side valve failure, secondary side h'est demand was inersased from 100% :o 138:: at time zero. This increase )
in secondary side hea: demand is the smalles which results
- in a (high fluz) reactor trip and is' very similar to the worst moderate frequency overcooling event, a failu:e. of the steam pressure regulator. In the analysis, it was assumed t that following EI actuation on low RC pressure ESTAS, main feedvater is ramped down, MSIV's shut, and the anv"4=ry ;
feedvater initiated with a 40-second delay. This action was ,
taken to stop the cocidown and the depressuriza:Lo,n of the systen as soon as possible after EPI actuation, in order ,co minimize the time of refill and repressurizatien of the ,
system. Both HPI pumps were asstedd to fune:1cn. }
The calculation was performed twica, once assuming two of the :
four RC pumps running (one loop), and once assu=ing RC pump trip right after HPI initiation. The analysis shows that the ,
system behaves very si-1'arly with and without pu=ps. In k~
both cases, the pressurizer refills in abou: 14 to 16 =inutes
~
from initia: ion of the transients, with the na: ural circula- i -:.'
g-t.
?-
i
1
., h l
. tion case refining about coa minuto befsro th3 caca vir. l m of four peaps running (See Figures 3.2,3.3). In both cases, f 3C'T to 120*7 the system is highly subcooled, from a minimum o 4).
and increasing at the and of 14 minutes (refer to Figure 3.
- It is concluded that an RC pu=p *::1p following F2Ih actuation vin not increase the probabili.*y of causing a LOCA ill throug have the prusurizar code saf eties, ard that tne operator b v ling, to the same lead time, as van as a large margin of su coo Although no case control EPI prior to safety valve tapping.
with an RC pumps was made, it can be inf erred from the one loop case (vich pumps running) that the subcooled The-marg I be slightly larger for the au pump's running case.
pressurizar vill taka long e to fill but should do so by 16l Figure 3.tshows the coolant minutes into the transient. i f temperaturns (hot leg, cold leg, and core) as a funct on o .
time for the ge, eRC pumps case. l i
- 2. Eff ect of Stes= Bubble en Natural Circulation Coelfac For this concern, an analysis was puformed for ,the same generic 177 FA plant as outlined in Part 1, but DIR), assuming the th as a result of an unmitigated large SL3 (12.2 ft.
excessive cooldown would produce void fermation in the primary system.
The intent of the analysis was to also show the
~
As in extent of the void for=ation and where it occurred. both the case analyzed in Part 1, the break was symmetric to generators such that both would blow down equally, -m
- i=i 2 break on each the cooldown (in this case there vas a 6.1 ft.
loop). There was no ESIV. closure during the transient on ehe cocidown. Also,, the tur-either steam generator to bine bypass system was assumed to operate, upon rupture, .
ESTAS was initiated on low RC until isolation en ISTAS.
pressure and also actuated EFI (bajh pumps), The tripped AF;T RC pumps (when applicable) and isolated the ET7's.
was initiated to both generators on the lov SG pressure signal, with mini =um delay time (both pumps operating).
This analysis was perforned ,tvice, once assu=ing all RC pumps runni=g, once with au pumps being tripped en In the E (s5 second) delay.
actuation (af ter ESFAS), with a shcrt both cases, veids were formed in the hot legs, but the durn-
_ 17 - . I
1 I
l tion and siza vero smaller for tha casa with no RC pump j
trip (refer to Figure 3.7).Although the RC pump operating '
case had a higher cooldown sts, there was less void fo:ma-De tion, resulting from the additional system =44"s.
coolant temperatures in the pressurizar loop het and cold legs, and the core, are shown for both cases in Figures 3.5, 3.6. ne core ou:1st pressure and SG and pressuriser levels versus time are given for bcch cases in Figures 3.8, 3.9. nis analysis shows that the system behaves very sim*1mely with and wi:hout pumps, although maint-4"4"!
ne hC pump flow does sees to help mitigate void formation.
of pres-pump flow case shows a shorter time to the star:
' suriser refill than the natural circulation case (Figure 3.9),
although the time difference does not seem to be very large.
- 3. Effect of Return to power here was no return to power exhibited by any of the BCL cases analyzed above'. Previous analysis experience (ref.
Fidland FSAR, Section 15D) has shown that a RC pump ::1p vill zitigate the consequences of an EDL return to power condi:1ci.
D e reduced by reducing the cocidown of the primary system.
cooldown substandall'y increases the suberitical margin which, in turn, reduces or ald d =:es return to power. ,
D. Conclusiens and Summarv A general assessment of Chapter 15 non-LCCA events identified three areas that warranted further i=vestigation for impact of a RC pump trip on ESFAS low RC pressure signal.
i
- 1. I: was found tiat a pump trip does not significan:17 shorte J the time to filling of the pressuri=ar and apprordgately the same time interval for operator action exists.
dme overcooling case analyze.I, the RC pump ::1p
- 2. For the increased the amount of two-phase in the primary loop; however, the percent void fornatied is still too small to aff act the ability to cool on natural circulation.
- 3. n e suberi:ical return-to-power conditica is alleviated by the RC pump trip case due to the reduced overcooling effect.
is con-Based upon the above assessment and analysis,1:
i cluded that the censequences of Chap:er 15 non-LC0J events are no:
- Seem o
e C' pump trip on E5FAS increated duo to tho addition of a R low RC pressure signal, fcr cil 177 FA lowered icop plants.
Although there were no specific analyses performed d loop for 2 00, j the conclusions drawn from the analyses for the lowere plants are applicable.
e O
e e
\
O e
9 O
e G
9 e
me
- 19 -
m..m,
Table 2-1. %=1vsis Scene With AN Ave.ilable.
Continuous E E puurp trip 9 90*. void etuso coeration_
Break location _
2 HPI 3 , , , ,
Cold leg Hot leg 2 HPI Q (ft2) _
I 0.025 I '
2*
I I* I I
0.05 I I
- 0.075 I I I
0.10 I I
0.20 I i
e Analyzed with both 1.0 and 1.2 ANS decay curves.
O e
e e
b e
i
k Spectrum l
' Table 2-2. Lspect Assessment of Brea'" rip at 90 Void _
With RC h 7.
e Core uncove n d = (sec)_
3reak size (fr l _
550 0.10 625 0.075 .
575 0.05 Two EPIs available during the Notas: 1 transient.
- 2. Cors uncovery time.is the time period following pump trip re-quired to fill the inner RV l with water to an elevation of '
- 9. ft in the core which is ap-proximately 12.ft when sws11ed.
4 4
f-
?
.i.d E,
)
- i y
t f f
-~
- . [
)
c m e m on of System Void 7:scrions -
Tabis 2-3 at ESTAS Sizmal System void fraction ._
at ESTAS 3reak size, Pusos on_
p s trisoed -
(fc2 ) _
0.0 0.02463 4.47 0.04 0.04 0.05 6.74 0.055 8.06
- 0.07 0.90 0.07 5 8.45 0.085 7.97 2.17 0.10 10.70 0.15 6.78 0.20 e
e o
. . ~ . . _ - . .
l __
- MINTnAP2 NCDE DESCRIPHON
! NODE NDf3ER DESCRIPU CN -
1,33 Reactor Vessel, Lower Planus 2,34 Reactor vessel, Cora 3,35 Reactor Vessel, Upper Planum 4,10 Bot Les Piping 5-7,11-13 Primary, Steam Generator 1 8,14 Cold Les Piping I 9,32 Reactor Vessel Downconer 15 Pressurizer 16,24 Steam Generator Downcomer 17,25 Steam Generator Lower Planum 18-20,26-28 Secondary, steam Generator 1 21,29 Steam 11sers 22,30 Main Steam Piping 23 Turbine 31 Conem h ant 1
l l
i I
MINITRAP2 PATH DESCRIPTICN PATH Nm!3ER DESCRIPTION 1,2 Core 45,46 Core Bypass ,
3,5,5,11,12,44 Bot Leg Piping .
l 6,7,13,14 Primary, steam Generate- I 8,15 1C Pumps . l 9,16 Cold Lag Piping 10.43 Downcomer, Reactor Vessel 17 Pressurizer Surge Line 18,19.26,27 Steen Generator Downcomer 20,21,28,29 secondary, Steam Generator 22.30 Aspirator 23,31 Steam 11ser 24,32 Steam Piping 25,33 Turbine Piping .
34,35 Break (or Leak) Path 36,37 EFI 38,39,43,44 APW 40,41 Main Teod Pumps .
42 LPI -
Table 3.1 l
l
-n-
Figurs 2-1. M Noding, Diagram ic: Sas113: cat- +
1""" ===
[OA g g , ert as ,_
9v] = = - m
~'
a
' - ' s '- ,
8'
- 1s t [
, b', - as e
~
@ ( s, %
< a
- =.==
cs <s 9 s ,
g @
,. W W ,, 3@f ,,% f3 - E_ - .
1
= (W W", _,J
,} w ' s 1 s
6
. I
- W m- ) s 3.nas adessa.maa teen s.s sem as an es m g_
= eTu,,
T ,a_e _ _,__
eao O . . ..,.._,_. ,.
j '*** h I*, *s~.N. *I.
T. "
l' se a ==
h (Vsh sash-
- .4 ==*
@ sea.s., e.a s,es se C Path Wo.
Ydentifiesef en _
yod, No._ 7dentiftention _
h 1,3 Bat 1** 7171:3 31sweesser 3,4,28.23 Bot lac. UPP*r 2
sewar 71emum 5,20 30 fuW 2 Core. Care 3ypass, Vpper 6,21 3 Flensa, typer Read 7.22 3SM" Core 3ypas s,
4,24 set 14g 71 ping 3 Staan Centrator Upper f,23,24 Cold 143 P171st 3,33
- Beed, se Tubes (Upper Bag) 20,34,25 Puups Cold Lag FiP i=5 SC Tubes (laver EaM) 31.12,15,16,26,27 6,26 sc laver used 17,21
+M 3,13 Cold tag Piping (?nus, Seccian) LII f 12.19 23 Cold Les Piping (71 ssp Discharge) 28,29 DPP*? W ***f 20,12.20 Pressurizu
. 23 Vyper Devocamer 30 I*** v"1
Cabove the~ C of Beazle Salt) 32 E 21 Wuu 33,34 33,36 IF1 2amea4neeme Castainment sprays 37 i e e TYQ o 0b* oU * @u o . S .$ lidl
--=%.--..-,
rien. 2-2. CRAFT 2 N0 DING DIAGRAM FOR SMALL BREAK (5. NODE MODEL) r CFT y ._, G - 1-4 5 2 6
@ 3 "L -
; +4- ,
' J 1
m L ku @ .
- w u .
m @ s LEAK PATHS 8 & 9--
*@ C
-O g k*y .
Identificatien Path No. Node No._ Identifiestion _ Core 1 PD Piping. DC, LP 2 L7I 1 Primary SC EFI 2 3,10,11 core, UP, Bot Legs 4 Pot tags 3 Pumps 4 Pressurizar 5 cour=4 ~ : 6 Vent Valve 5 d = y SG Pressuizer 6 9- 7
. 8,9 Laak & Return Psch i
J- ,
*e.-. ., ,
CORE PRESSURE VS T15E,177.LL, 2772 Ett PU ES M i BREAK
" 0 025 FT 25H00E300EL l l
20 l. 0.025 FT2 BREAK 3 N00E MODEL l I 18
^ \
e I
=
~~
w
". 16
. l
- '=
l 3 14 E. . 12 , 10 - 2000 2500 1000 1500 O 500
* 'Tlas, sac .
Figurs 2-3 e e l l 1
i . t l ( PERCENT SYSTEM V010s VS TIME, PUMPS ON 100 , 80 e 2 (! 60 - E 1 (g g ep9'-G ' i 40 -
..gfi.-;f*.C n (S 20 -
4
*- f-i.e -
g f* - i-
, 2000 2400 1200 1600 400 800 0
Tiss, see Figure 2-4 , 0 9 . e
. o _
BREAK SPECTRUM-RC PRESSURE WITH THE RC PUMPS OPERATIVE M ANU 2 2500-2000 , I as
~.. 1500 m
; 2 h;
E ar== L----* " - -
- .- 0*025 FT r~ %
y 1000
\
% 2 0.05.FT
"\. , .
- 0 075 FT 2
500;-t
\, gtph %,,~x~ ~
* . 2 -1 0.10 FT
,%'N. N i . 0.2 FT 2 ,
W M 1500 2000 500 1000 Time, see . Figure 2-5 e e
- :o -
l 1 -
BREAK SPECTRUM-AVERAGE SYSTEM V010 FRACTIO 11TH THE RC PURPS OPERATIVE AND 2 HPI PUNPS 100 , ,,..... .. m; o_ y 7 0 2 g* , ,
= = ~
.i o
/o ,f e
s'# 80 -1; /y %~~f ., w % ,'
. 7 :! t oAl 4 /
=: %7 % / e
/ .
2 : =*o %* t Q# ., g 60 -! / +/ %A/
- o %/
5 Ti. iJ 4/
/
/
$/
s '. -
% A0 ' : l +/+ / s.$$-Q1p*'..
E
=
- i,/,/
/
o+ ,./- 20
/,
/.
i , , . s.,---. - 2000 2400 2800 800 1200 1600 0 400 Time, sec Figure 2-6
0.1 FT 2 BREAK WITH CONTINUQUS RC PUMP OPERATION AND 2 HPI PtrdPS 1 2500 l 100 tPI
- ~ ~ ~ - _ - _ _ _ _. !
/ -
6
/ t 2000 80 -
/ \s en j %
. *--r \
; I
' N g
- 1500 h 60 -
I g = a I 5-
# \ -
i s
! 5
; 40 -%.J t *g _
- 1000 E
/ N ~
- I N
- I N LPI 500 2D - \ %
g
~~'.4- .
0 . , . . 5 --- 0 1500 2000 2500 3000 0 500 1000 Time, sec Figure 2 7 e e e e*
~ * * ** **+ ,w6 meg og,, , , , ,
2 BREAK RC PRESSURE FOR 0.05 FT AVAll.ABl.E 1 HPI YS 2 HPI'S 30
. 2 HPI's, PdMP DN, HONOEENEDUS 1 HPI, PUMP DH,HOMOSENEDUS o
25 - i e 1 N
= 20 . -
l timsP E. 15 i
= \
E
=
\ o
- s -- - o r 10 - m ' -Qg
- N *h d g% c~ N ,
5 0 2000 2500 3000 O 500 1000 1500 1 Time, sac , l Figure 2-8 l 1 e l l
-n-l . ..
1
3 1 2 AVERAGE SYSTEM V010 FRACTION FOR 0.05 FT AVAILABLE 1 HPI VS 2 HPI'S l l 100
.~o-.
* /go ---
80 - l/*/[ y ,.
/
/
2 HPI'S, PUMP DN, HOMOGENEOUS i
- f. f
- BD -
/
/ e - 1 HPI, PUMP DN,HOMOGENE0US l
P.
*/
N /
/ -
E 40 -
/
3 ./ h . /
/ '
/ ,
20 . O ' - ' ' -
- 800 1200 1600 2000 2400 2300 3200 0 400 Time, see Figure 2-9 ,
. l i
l l l 6 I i l t I
-- -- ^ - - _ _ _ __
.e_
I RC PRESSURE2 FOR 0.075 FT ,PUNPS OFF e 905 S 30 g
, , ,_ 2 HPI'S, PUMP DN, MotoGENEQUS 25 -
2 HPl'S, PUMP OFF e 905 V010, 2 PHASE
$: 20
.2 -
E- 15 - E t
- t ~
a. I0 - %s 4.s N % 5 ---- _ N. %%
. =
2500 3000
, 1500 2000 500 1000 0 .
Tiss, sec Figure 2-10 . e l , l l AVERASE SYSTEM VOID FRACTION FOR 0.075 FT2 , PUMPS OFF e 905 SYSTEM Y010 100 ----- ____
/
,s'~ ~.~
80 - f i ,N'%,% *
* / .
/
/
3 /
/ .
60 - a f E / 2 HPI's, PUMP OH HONOGENEDUS b-
~ t
/
2 HPI'S, PUMP 0FF e 905 VotD, j 40
,/ 2 PHASE j ,/
/
20 - 0 2400 3200 2000 1200 1600 400 800 O Time, see Figure 2-11
l l AVAILABLE LIQUID VOLUBE VS TIME i FOR 0.075 FT 2 BREAK WITH 1.2 ANS
]
DECAY HEAT CURVE i 7 3000 n l
^
c,
- ' 2500 I
E 2000 S LEVEL OF ACTIVE CORE
- l 3
1500 I 2 . S ,
= 1000 t .
RC PUNPS OFF 1 500 w -- 1600 2000 800 1200 O 400 Tlas, sec Figura 2-12
.- l
. 35 - ,
1 1 2 RC PRESSURE VS TIME FOR 0.05 FT BREAK IITH 1.0 AND 1.2 ANS BEFORE AND AFTER PUNP TRlP - 0.05 FT2 , 2 HPI'S
~~~ 1.2 ANS, PuuP OH 3000 -
0.05 FT2 , 2 HPI'S
*" 1.0 ANS, PURP DN O.05 FT2 , 2 HPI'S ;
-
- 1.2 ANS, PUNP OFF 2500 0.05 FT2 , 2 HPI'S j
- W -" 1.0 ANS, PUMP OFF j l
. 2000
=
1 l E 3 1500 r . I i L
\-- - -
1000
.. ,[, ,-
% *~*~., %
(4, ; 500 I a
~
M Ol~ 2000 2HO
~
15p . 500 1000 , ; 0 . . 4 Time, sec - ! Figure 2-13 e I
- ~ ~ - --
2 PERDENT SYST'EM ER PUMPYO10 TRIP FRACTION FO BREAK IITH 1.0 AND 1.2 ANS BEFORE AND l 100 g '
*; W k s %
SG . /./'Y
* ,y -
sf . 5 2 E - _ _ _ 0.05 FT , 2 HPI'S, PUMP DN, 60 1.2 ANS
- g. 2 P DN,
; _ _ 0.05 FT , 2 HPI's,PUM f 1.0 ANS
/ MP OFF, g 40 -
/ 0.05 FT2 , 2 HPI'S,PU o _ o 2 / ,
1.2 ANS
,/ 0.05 FT2 , 2 HPI'S, PUMP OFF, 20 -
1.0 ANS s 2400 2800 0 1 1600 2000 800 1200 O 400 Tlas, see
. Figurs 2-14 e
0
l 2 AYERAGE FPEAK , BREAK l.CCATION COMPARISDN PUMP SYSTEM YO10 FRACT!DN V 1 100 p UNC0VERY TillE = 625
" " - SEC
[
- 80 -
/Y UN00VERY TIME =
f g 4 /.e 450 SEC 30 - gh// es E
/~ 9
/* .. -
40 -
- 4,le5
=
E 20 - f/ - . 0 - 2000 2400 1200 1600 400 800 O~ Time, see Figure 2-15
~- __
COMPARISDN OF DELIVERED HIGH PRESSURE INJECTlDN FLU 10 TO RV FOR PURP DISCHARGE SR
\
1200 - N , 1100 - N \<#p0 1000 - N# ! \?'/ _ 900 - N 800
- g
=~ 700 -
\
\ #
~
# '# p 600 \ 0 ' Sp,
\
j 500
~
\
( g 400 >%m h isED ggg, ,
~
\
300 N~% , LQF LQQp 1 HP)
- ~s 200 N- .
100 -
,j ,
2200 2600 3000 1000 1400 1800 0 200 600 Frassure, (psla) Figure 2-16 e e e
O sw i W M-- 21 1
. '--D<
. r.* * * * * * * ' . ,
> Q l5 @*
l k i,........,,. _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . .. s m, @
- g f 1 (moV->4- ,,,,L .
I*
- 6) '
%. n W II -
\.
12L
- g" .31 ,y 3 II , S 20 t 8 0 g is j -
1 et w p .-
- 4 eg he .a 0 . Vs J A+
J7 IL 3 @ (.- If @
"q . g- "
i /6 g t s[ <- - cst o.) d' ,e e
.. 2s IS + s.
7 a -
.fU e 1
M
.. P$
- e_ nLt
' 6
'. '.. n D
.( .
2.. .Q., s
.O G -
is
,i
. TO - .
e 1 e - I i
- I .
g I t l F4ua 3.1 . MDCTRAP2 Noding and Flow Path Scheme I D ~9~ D @ @ ~D h'
- 40
- i _dbi d L1 }ljil$
[ l - ._ --
v - r~iE::SU.~;i2:i; A:] STEA ' CE:2P.A1CP LIC'JID LEVEL V !3US T!*.'.::SIENT TI:'I 2 (102 FP. EIG OF LIFE,D.G FT 5T:A::t1::1 U.'.;AT, (CDL':2!!:3 COEf: ATE Fl:EO.). (RC PlliP 1 HIP) l i l l l l l l 3.6. 1UDE RECION Ooeogo 0 ' FULL 0 o G 3 0 3 40 - g P!!ZR. FULL g U
~~
E e G c g O O O O
'g 30 -
g g C
. m u
3 . 6 g o
~ gs o u- c c m o8 5 20 aU S U G
- o e
- U e
- 6 KEY E &
10 - B G: PRESSURIZER ; U 6 o: STEAM EEHERATOR 'A' , A: STEAM SENERATOR 'B'
- 9 , , , , ,. , .,
0 2 4 8 8 10 12 14 16 ' ~ Transient Tin:e (Minutes)' i Figure 3.2 l
- e. ,
i' l
'l
, :. 1
.n- -
. if I_.--
i V. ~ ~ = _ __
!E:
eme.* --
. .. , g- o e omw t
4.
.< . = _ .__
MM w o o
\I >= >=
= <
N = w ww wm z ox
,\
_, \ w
- w w
~
\ ne u u m as <=
m_ -
=. - >
w a w <w w z a
.c =
l . w w a.- = w v
. t.
j\ a g %
\
wm w - w s , s \ . ~ .
=
m 4 w % s i ss t e o <J - = m m ,w s i W ,
= = Q I
%s A w
w .c x _. , s s - w z- s sj s s ss s .
- - ., . . n w .
ua n. ls s N .
\
e
= w
= -
a.
- c 1
s%. Ns E, N s e\. s e e z. s s 2' _., m o=.
- . s 's s "
. w
- N
~ E E=E g b "
h= *w b w8 3= Ns -E=-sz c m%Ags -
= w _=
.^ - s ma-
--o \s
= - s ew= gs s o a u 58, s=== -*
a< w a e o m 3 l s\ k,k N s\ Wa
= = .
s
$!e g Ww Ww s
N D i* % D *
% ".6. "w >=- G4 0 5
a w a 05 w ~ 8 e 4 .O a
. = G m - v m G 4 G 4 G G
$ . ,=
3-
= =
~
., ., u
., e (23) insi pinh!1 jo:ca:u a acsis'J.s:!)nss:Ja :
Figure 3.3 42 - l
i i _. i
. es c ^
& g i
om c ~ b W" . I em c 5 a ,#
. em o ". .W
=. .=. t em o ~
. =. s
- ~
~
,e oa e t, -E m a.:
=
m me i em e o m m a < 5 g om C " m .= - m =^ em o a=3: - 2 " Rm
= z w
^
ec e.
=""" E=
g 2 oo 4 - - 2
. .=..
g
- ee < -
eoo I i
- gyQ l I
ao c W .
=
E eo e -
=~m a:
w m m 1 3 ee q - e o e
= =. =
w v I em e - w
& k
- M Et i ee o 5=
o ~ I oo o g 8S* i em o I es c - , I C3 4 I e c I g CE C i e 4 - ~ I ED 4 . . I @4 con s lg e e r , o e o o : o o e o o 3 o o M W af3 v M N (J.) ajnicJad:::31 luc l cog Figuic 3.4 .- l 1
~ ---- ~ ~ _ . - -
/
- CCU.':T TE"?iL*.TC.*.1 "TT.;'.'S TT..~.:.:iE::7 i:. -
( I ll'/.- FF. BF.Dilcit'*; 0F LIFF, 12.2 F1 2 Dnn.ig C".If.r. ('J::"ITIC'.TED) F"9 P.'J"T!J",E,, STE!."'.1::: pg r.a '~ n- - raio)
~ ' '
700 , I a P"ESSURIZEE E*2 TIES
's ! HDT LEG BEGINS TIO. PHASE
\ I l* C05.*i FLOD; T EK 50Df~\s i
K\ FLOiS DEGINS
\
\ I HOT LEG SUBCGCLED 0{be\ g N
l y
~
500 - g $\s 5 8\
- EN '
E bs gg
~
=
400 - [s>r"% -
.\
B d a\s/e/e'"\ ci y
\
\
\
5 o \ . KEY B o \ e: HOT LEG (PRZR LOOP) 3 \' m: CORE 300 - 4: COLD LEG (PRZR LOOP) g$
-- : SATURATION LINE $
s 9 . 200
- 8. 10 12 2 4 6 D
Transient Time ('Jinutes[ Figure 3.5 44 - l l
. ~
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n .- l t i (ta:it) IaAal J71d *S*$ 43 - Flo,ure 3.0
.- i RETERINCES_
l l d l " 3rd-10104, Rev. 3, "3&W's ECCS Ivaluation Mo e , 1 3.M. Dunn, et al._ August 1977. 18, 1978. 2 Latter, J.R. Taylor (3&W to S.A. Varga (NRC), July Program for 3 1.A. Hedrick, J.J. Cudlin, and R.C. Toltz, "CRA~."*2 -fCTortran lant," l Digital Simula: ion of a Multinode Reactor Plant Du:ing Imss-o - oo RNJ-10092. Rev. 2, April 1973.
The vslocity of Rising Steam
" J.F. Wilson, R.J. Granda, and J.F. Patterson, 5, (1962).
in a Bubbline Two-?hsse Mixture," ANS Transactiour, e
)
l e e 1 i e
. 49 -
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