ML20136B828
| ML20136B828 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/29/1979 |
| From: | Crouse R TOLEDO EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| I-85, NUDOCS 7909120409 | |
| Download: ML20136B828 (55) | |
Text
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Docket No. 50-346 b Totreo License No. NPF-3 EDISON Serial No. I-85 i
August 29, 19 79 Mr.. James G. Kappler Regional Director Region III Office of Inspection and Enforcement
- 5. S. Nuclear Regulatory Commiss(on 799 Ecosevelt Road Clan Ellys. I111acis 60137 Bear Mr. Esppler:
Attachments A and B are Toledo Edison's responses to short term action items 2 and 3 of IE Bulletin 79-05C for the Davis-Besse Nuclear Power Station. Unit 1 concerning reactor coolant pump pperations.
These discuss the analyses and development of preliminary operator guidelines for reactor coolant pump operations.
In addition to these guidelines Toledo Edison is proposing an interia as well as a long term automatic reactor coolant pump trip (ARCPT). Details of design and the installation schedule of a trip system for the long teru action iten 1 will be submitted for NRC approval no later thaa October 31, i
1979. Attachment C provides details for an interia ARCPT system whose installation can be accomplished withis seven days of formal NRC approval.
Upon completion
.ef the interim installation the second licensed operator established in short term action item 1.8 would no longer be reqdred. We request
- jour approval.
Yours very truly.
//!:
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.:C D,. -.
VICE PRESJDENT. ENERGT SUPPLY IFC/TJM
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CC:
Director office of Nuclear Reactor Regulation 7
- 5. S. Nuclear Regulatory Commission g
unshington, D. C. 20555 g
g Director office of Inspe tim and Enforcement W. S. Nuclasr Ragubcory Commission 3
Unshington D. C. 2M55 190912 0 90 f 34 TOLICO EOSCN COAd7ANY ED$CN PLAZA 300
......we TOLEOC. CMC 43652
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Docket No. 50-346 License No. NPP-3 Serial No. 1-85 Augusc 29, 1979 ATTACHMENT A Short-Tern Action Item 2 Perform and submit a report of IACA analyses for your plants for a range of small break sizes and a range of time lapses between reactor trip and pump trip. For each pair of values of the parameters, determine the peak cladding temperature (PCT) which results. The range of values for each parameter must he wide enough to assure that the maximum PCT or, if appropriate, the region containing PCTs greater than 2200 degrees F is identified.
Response
This analysis is provided in Attach = mat B Short-Ters Action Item 3 Based on the analyses done smider Iten 2 above, develop new guidelines for operator action, for both IACA and non-LOCA transients, that take into account the impact of RCP trip requirements. For Babcock & Wilcox designed reactors, such guidelines should include appropriate requirements to fill the steam generators to a higher level, following ECP trip, to promote natural circulation flow.
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Response
Operator guidelines are currently being reviewed for station applicability and will be formalized by the issuance of a Babcock and Wilcox site instruction by August 31, 1979. These will be available on site for review.
The applicable procedures will be modified and station operators trained by September 13, 1979 as specified fa short term action item 4.
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l Docket No. 50-346 License No. NPF-3 Serial No. 1-85 August 29, 1979 ATTACHMD.T 3 l
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4HALYSIS SUMMART IN SUPPORT OF AN EARLY RC PUMP TRIP l
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comrs Page I.
DrTRODUCTI0tt................
1 II. SMALL BREAK ANALYSIS.......................
2 A. Introduction.........................
2 3.
System Response With RC Pumps murming 2
C.
Analysis Applicability to Davis-Besse 1 11 D.
Effect of Proept RC Pump Trip on Low Pressure ESTAS Signal............................ 13 E.
Conclusions
......................... 13 III. INACT ASSESSMENT OF A RC PUMP TRIP OIt 3016-LOCA EVENTS...... 15 I
A. Introduction......................... 15 3.
General Assessment of Pump Trip in Non-LOCA Events...... 15 C.
Analysis of Concerns and Results................ 16 D.
Conclusions sad S a mary
................... 18 o
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ANALYSIS SIMMART IN SUFFORT OF AN EARLY RC FUMP TRIP I.
INTRODUCTION B&W has evaluated the effect of a delayed BC pump trip during the course of small loss-of-coolant accidents and has found that an early trip of the
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BC pumps is required to show conformance to 10C1150.46. A s e ry of the LOCA* analyses performed to date is provided la Section II. This discussica includes:
1.
A description of the models neittmed.
2.
Break spectrum results with coatissous BC Pep Operatism.
3.
Break spectrum results with delayed ac pump trips including estimates of peak cladding temperatures.
4.
Justification that a prompt pump trip following ESFAS actuation on low i
EC pressure provides IACA sitigation.
An impact aasessment of the required pump trip on non-LOCA events has also been completed and is presented in Section III. This evaluation supports the use of a piasp trip following ISTAh actuation for LOCA mitigation since no detrimental consequences on non-LOCA events were identified.
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- II.
SMALL BREAK ANALYSES A.
Introduction Previous small break analyses have been performed assuming a loss-of-offsite power (reactor coolant pus:p coastdown) coincident with re-actor trip. These analyses support 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 transient in which the RC pumps remain 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. Specifically, section 3.
describes the systen response with the RC pumps r-f ag for BW's 177-FA lowered-loop plants. In-cluded in this section is the developasst of the model used for the analysis, a break spectrum sensitivity study, and peak cladding tem-parature assessments for cases where the RC pumps trip at the worst time.
section C demonstrates the applicability of the conclusions drava in section 3 to a 177-FA* raised-loop plant (Davis-Basse 1).
The effect of a prompt tripping of the RC pumps soon receipt of a low pressure ISTAS signal is discussed La sectica D.
Finally, sac-tion E stammarises the conclusions of this analysis.
B.
Systes Rossesse With 3C Ptuses *=ias 1.
_las"_ educt _lan Recent evaluations have been performed to en:ine the primary system response during==a11 breaks with the EC pumps operative.
During the transiaat with the RC pumps available, the forced circulation of reactor coolant will amistain the core at or near the saturated fluid temperature. Bowever, for a range of break 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 circulation. In fact, the RCS void fraction will increase to a value in excess of 90* in the short tera.
In e
- _ =
h_
_ m_
the long ters, the system void fraction will decrease as the RCS depressurizes HP1 flow increases, and decay heat diminishes.
With the RCS at a high void fraction, if all RC pumps are postu-laced to trip, the forced circulation wili' no longer be available and the residual liquid would not be suf fielent to keep the core covered. A cladding temperature excursion would ensue until core cooling is reestablished by the ECC systems. The following para-graphs summarizes the results of the analyses which were performed for the 177-FA lowered-loop plants, to develop the consequences of this transient.
2.
Method of Analvsis The esalysis method used for this evaluation is basically that de-scribed in section 5 of gAF-10104. Rev. 3. "B&W's ECCS Evaluation yog tal and the letter J.I. Taylor (B&W) te,S.A. Yarga (NRC), dated 2
Jely 18, 1978, which is applicable to the 177-FA lottered-loop plaats for power levels up to 2772 3 eft. The analysis uses the CRAFT 23
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oods to develop the history of the RCS hydrodynamics.
, Eowever, the CIAFT2 model used for thia' study is a modification of the small break swaluation model described la the above ref-erences. Figure 2-1 shows the CIAFT2 anding diagram for semil breaks free the above referenced letter. The modified CRAFT 2 radel consists of 4 modes to simulate the primary side,1 mode for the secondary side of the steen generator, and 1 mode representing the reactor buildbag. Figure 2-2 shows a -h== tic diagram of_. -
this model. Node 1 eestains the cold les pup Mwhrge piping,
',douncomer, and lower pleous. 3 ode 2 is the primary side of the SC l
nad the pap section piping. Mode 3 esatains the core, upper ple-miss, and the hot legs. 3 ode 4 is the pressuriser and modes 5 and 6 represent the reactor building and the SG secondary side, re-spectively. This 6 mode model is highly simplified compared to those utilized in past ECC3 analyses. It does, however, maintain RCS volume and ele ation relationships which are important to properly evaluate the systes response during a small break with the RC pumps running.
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__._.m___.__--_-
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. Section 3.7 demonstrates that this is the worst break location.
Key assue:ptions shich differ from those de-scribed in the July 18, 1978, letter are chose concerning the equip-ment availability and phase separation. 3ese are discussed below.
a.
Equipment Availability The analyses which were performed assumed that the RC pumps re-main operative after the reactor trips. For select cases, after the systen has evolved to high void fractions (approxi-mately 90T) the RC pumps were assumed to trip. Also, the in-pact of 1 versus 2 BF,1 systems for pump injection were awa=4ned.
The asjority of the analyses performed assered 2 HPI pumps.
Bowever, as is demonstrated later, even with 2 EPI pumps avail-
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able, el-Ading temperatures will exceed the criteria of 10 CTR 50.46 using Appendix K evaluation techniques. Therefore, fur-ther analysis with only 1 HFI pump would only be academic.
, b.
Phase Separation The. present ECCS evaluation model created to evaluate small breaks without RC pumps operative (quiescent RCS) uti-11aes the Wilson" bubblerise correlation for all primary sys-tea control volumes in the CRAPT evaluation. In this analysis, for the time period that the RC pumps are operative, the pri-mary system coolant is assumed to be homogeneous, i.e., no h
P ase separation in the system. In reality, the flow rates in the core and hot legs are low enough that slip will occur.
This will cause an increased liquid inventory in the reactor vessel compared to that calculated with the homogeneous model.
With the homogeneous assumption, core fluid is continuously i
circulated throughout the primary systes and a portion of that fluid is lost vi: the break. During the later stages of the transient, a slip model will result in fluid being trapped 1::
the resctor vessel and the hot legs.
The only method of losing liquid during this period will be by bciling caused by the core decay heat.
Thus, the assunption of hemogeniety for the period with the RC pumps operative is conservativs.
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Following tripping of the RC pumps and the subsequent loss-of-forced circulation, the system will collapse and separate.
The residual liquid will then collect in the reactor vessel and the loop seal in the cold leg suction piping. For this period of the transient, the "Jilson bubble rise codel is utili:ed.
The homogeneous assumption for the period with the RC pumps operating applies to nodes 1, 2, and 3 in the CRAFT model.
Node 4, the pressurizer, and node 6, the secondary side of the steam generators, utilize the Wilson bubble rise model throughout the transient as these nodes are not in the direct path of the forced circulation.
3.
Benchmarking of the 6 Node CRAFT Model Studies were performed to compare 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 (B&W) to S.A.
Varga (NRC), dated July 18, 1978. The break size selected for this comparison is a 0.025 ft2 break 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, cespara very well with chose from the more extensive (23 nodes) CM 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 utilized for this study is thus justified based on comparison of results to the more extensive small break model and desirable because of l
its economical run time.
4.
Analvsis Results The break sizes examined for this analysis ranged from 0.025 ft:
to 0.2 ft in area and are located in the pump discharge piping.
3reaks of this size do not result in a apid system depressuri-
- stion and rely predominantly upon the HP:s for nitigation.
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Table 2-1 summarizes the analyses perfor:ned for this evaluation.
The majority of the analyses perfor: sed utili:ed 2 HPI pumps through-out the transient.
~he effect of utilizing 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 Figure 2-6, the average system void fraction is defined as V3-Y2 Average system void, % =
x 100 9 1 V3 = total primary liquid volume excluding the pressuri-ser at time = 0, Y2 = total primary liquid volume escluding the pressuri-zar at time = t.
This parameter was utilized in place of the mixture height in that the coolant will tend to be homogeneously sized 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 water occurs. As shown in Figure 2-3, 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 asistains adequate heat transfer in the steam generators; the primary system thus depres-surises to a pressure (about 1100 psia) corresponding to the sec-ondary control pressure (i.e., set pressure of SG safety relief valves). After scos time (250 seconds for the 0.1 ft2 break), the system pressure will decrease as the break alsee relieves the core energy.
Figure 2-6 shows the evolution of the system void fraction; values in excess of 90% are predicted very early (300 seconds) into the transient. For the larger breaks the system high void fractions
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occur early in cine.
For the smaller breaks it takes in the order
{
of hours before the system evolves to high void fraction.
Core cooling is saintained during a small break with continuous RC pump
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i operation regardless of void fraction.
tem win depressurize and the enhanced performIn the long ter and I.PI) win result ance of the F.COS (h*PI in reduced syste:n void fraction.
Figure 2-7 illustrates rhis long ters system behavior f break.
or a 0.10 f t2 For this case, the LP S are operative at approx 1:nately seconds, and a substantial decrease in system void fraction An arbitrary pump trip af ter approximately 2700 sec results.
onds would not result in core uncovery.
The potential for core uncovery due to an RC pump trip is thus limited to a discrete time period d which the natural evolution of the systes produces high void f uring tions and prior to LPI actuation.
rac-For a 0.1 f t2 period is on the order of 2000 seconds.
break, this time For snauer breaks, this critical time could be a few hours even if the operator initiated I
a controlled cooldown and system depressurisation as recomme in the susu break guidelines.
e Although the analyses described above used 2 DI pumps t
of only 1 HPI pump available on the system void fraction evolu
, the effect while the RC pumps are operating is not significant.
on Figures 2-8 and 2-9 show the impact of one versus two DI pumps on sys sure and average void fraction transients for a 0 05 ft res-2 the RC pumps operative.
break vich As seen from these figures, the results with one R pump are not significantly different to the two DI pump case and are bounded by the spectrum approach utilized one UI peasp, the systes does depressurise more slowly (less ste With condensation) and a higher short terz equilibrium void fraction is achieved.
Also, recovery of the core fonowing a loss of the i
RC pumps would be significantly longer with only 1 HPI p able.
ump avail-The majority of the analyses provided in this report uses t wo HP:
pumps and denonstrates a core cooling problem with vorst ti:ne p trip given that assumption.
ump would only show a larger problem,As analysis of one EP! available sively considered.
such cases have not been exten-As de:nonstrated in section 3.l., the resolution of this problem, fer:ed sarly punp trip, provides assurance of core cooling for both one or two HPIs available cases. *herefore.
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there is no need for further pursuit of the single HPI available case.
The effect of the RCP tripping during the transient was studied by assuning that the pu::ps are lost when the system reaches 90% void fraction. Loss of the RC pumps at this void fraction is expected to produce essentiany the highest peak cladding temperature.
After the RC pumps are tripped, the fluid in the RCS separates and liquid fans to the lowest regions, i.e., the lower plenum of the RV and the pump suction piping. At 90% void fraction, the core will be totany uncovpred fonoving the RC pump trip. Thus, the time required to recover the core is longer than that for RC pump
, trips initiated at lower system void fractions. System void frac-tions in excess of 90: can possibix result in slightly higher tem-paratures due to the longer core refill 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 fractiott is reached, win be lover.
Table 2-2 shows the core uncovery tim 6 for the cases analyzed with the RC pumps tripping at 90% void fraction with 2 IFI pumps avail-able for core recovery. As shown, the core vin be uncovered for approximately 600 seconds for the breaks analyzed. Figures 2-10 and 2-u show the system pressure and void fraction response for the 0.075 ft2 break with a RC pump trip at 902 void fraction. As seen in these figures, the systen depressurises faster af ter the RC pump trip, due to the change in leak quality, and the void fraction decreases indicating that the core is being refined.
Figure 2-12 shows the core liquid level response fonowing the RC pump trip. The core is refilled to the 9 foot level with collapsed liquid approximately 625 seconds after the assumed pump trip.
Once the core liquid level reaches the 9 foot elevation, the core is axpected to be covered by a two-phase mixture and the cladding temperatura excursion would be terminated.
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5.
Effect of 1.0 ANS versus 1.2 ANS Decav /*urve An analysis was performed using the more realistic 1.0 ANS decay curve instead of 1.2 ANS decay curve. The study was done for a 0.05 ft2 break with 2 HPI;s available and pumps tripped at 90%
system void fraction.
Figures 2-13 and 2-14 show a comparison of system pres'.ure and average system void fraction for 1.0 and 1.2 ANS decay curves.
As seen in Figure 2-13, the system pressure for 1.0 ANS case begins to drop from saturation pressure 61100 psia) about 200 seconds earlier than the case with 1.2 ANS as a result of reduced decay heat. Also, the system will evolve to a lower average void fraction as shown in Figure 2-14.
After the 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-onds compared to 1.2 ANS case. This case demonstraces clat the effect of a delayed EC pump trip any be acceptable when viewed realistically. A peak cladding temperature assessment for this casa vill be provided in a supplementary response planned for September 15th, to the I&E Bulletin 7905-C.
6.
Effect of No a=~411sry Feedvster Analyses have also been performed with the RC pumps available and no auxiliary feedveter. These analyses all assumed 2 IFI pumps were available.
The system void fraction evolutions for these calculations were not significantly different from those
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discussed with airrf14=7 feedwater. Thus the conclusions of the cases with a--d14=ry feedwater apply.
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.2 Break Location Sensitiv1:v Studv A scudy was conducted to de:nens: rate tha: the break location utilized for the preceeding analyses is indeed the worst break location.
As stated previously, :he analyses were perfor: sed assuming that the break l
was ldgated in the bottom of the pump discharge piping. A 0.075 ft2 hot les break was analyzed to provide a direct comparison to a similar I
case in the cold leg. For this evaluation, the RC pumps were assumed I
to trip af ter the BCS void fraction reaches 90%. Figure 2.15 shows the i
average systes void fraction transient and the cora uncovery times for both the 0.075 ft2 hot and cold les breaks. As shown, the cold les break reaches 90 void fraction approximately ISO seconds earlier than the hot leg break. Also, the cold leg break yields a core un~ overy time of c
s 175 seconds longer than the het les break. The quicker core recovery time for the hot leg break is caused by the greater penetration of the WI fluid for this break.
For a cold leg break in the pump discharge piping, a portion of the HPI fluid is lost directly out the break asd is not available for cera refill. For a hot les break, the full DI flow is availabla for core refill. Thus, as shown by direct comparison and for the reasons given.above, hot les breaks are less severe t!an breaks in the pump discharge piping.
3 Peak Cladding Teneerature Assessment As described previously, a RC pump trip, at the time the RCS void fraction is 90*, will result in core uncovery times of approximately 600 seconds.
The peak cJadding temperatures for these cases were evaluated using :he sus 11 break evalua: ion model core power shape used l
co demonstrate compliance vi:h Appendix E and 10CyR30.l.6.
Also, an adiaba:1c hes:up assump:Lon during the time of core uncovery was utilized.
l This approach is ex:ressly :cuservative in that :he power shape and.
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local power rate (4v/ft) analyzed is not expected to o normal plant operation.
ccur during Furthermore, use of an adiabatic heacup assumption neglects any credit for the steam cooling th at will occur during the core refill phase and also neglects the a.ffset of any radiation heat transfer.
Using a decay heat power level based on 1.2 ANS at 1500 seconds, the cladding will heatup at a rate will be 6.5 F/S under the adiabstic assumption.
With a core uncovery period of 600 seconds and the adiabatic heatup assumption
, cladding temperatures will exceed tbs criteria of 10CFR50.46.
a Use of a acre realistic heat transfer approach with the extreme power shape utilized for this eval-nation is also expected to result in cladding temperat ure in excess of the criteria.
In order to ensure compliance of the 177 FA lowered loop plants to the criteria of 10CFR50.46 a prompt trip i p ng of the RC pumps is required.
Section 5. demonstrates that a prompt trip of the RC pumps upon receipt of a low pressure ESFAS signal will result la compliance to the criteria.
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&n evaluation of the peak cladding temperature usi ng a power shape encountered during normal operation for a realistic tran i s ent. response with delayed RC pump trip will be provided by September 15 1979.
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a C.
_Ana vsis Aeolicabiliev to Davis-Besse I The significant parametric differences between the raised-loop Davis-Besse I plant and the preceeding generic lowered-loop analysis are in the high pressure injection (HPI) delivery rate and the a:nount of liquid volume which can effectively be used to cool the core.
The liquid vcluse differential is due to the basic design difference; raised versus lowered loops. Because of the raised design, system water available after the RC pumps trip will drain into the reactor vessel.
For the lowered loop designs, the available water is split between the reactor vessel and the pump suction piping.
Thus, for the same average system void fraction, the collapsed core liquid level following an RC pump trip is higher for the raised loop design than for the lowered loop design'.
Figure 2-16 shows a comparison of the delivered HPI f1ow for the Davis-
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Beasa I plant and the lowered loop plaats.
As shown, for a similar number of EFI pumps available, the Davis-Besse I pumps will deliver more flow.
For the delayed pump trip cassa presented in section 3.4 of this report, the Davis-Besse I plaat will take approximately 450 seconds to recover the core as opposed to r600 seconds for the lowered-loop plants.
However, it is noted that the core recovery time is based l
on using two EPI's rather than one, as required by Appendix 1.
Use of only one HPI pump for Davis-Besse I will result in core uncovery times in excess of 600 seconds.
Ihe Davis-Besse I plant cannot be shown to be in compliance with 10CTRf0.l6 for a delayed RC pump trip.
Prompt reactor coolant pump trip is, therafare, necessary to ensure compliance of the Davis-Besse I plant with 10CTx50.46.
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.D.
Effect of Promet RC Puce Trio on Low Pressure ISTAS Signai As demonstrated by the previous sections, the ECC system can not be demonstrated to comply vi.10CT150.46 using present evaluation techniques and Appendix K assumptions under the assumption of a delayed RC pump trip. Thus, prompt tripping of the RC pumps 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.
Because ne diagnosis of the event is required by the operator and ESTAS initiation is alarmed in the control room, prompt tripping of the RC pumps can be assumed.
The effect of a prompt reactor coolant pump trip on an ESTAS signal has been avamined to ensure that the consequences of a small LOCA are I
bounded by previous anall break analyses which assume RC pump crip on 2
reactor trip.* As shown by Table 2-3 at the time of low pressure ESTAS initiation, kaaping the RC pumps running results in a lower average system void fraction. This occurs because the availability of the RC pumps results in Iower hot leg temperatures and thus less flashing in l
I the RCS at a given pressure. Thus, a prompt trip upon receipt of an i
ESFAS signal will result in a less severe systes void fraction evolution chan cases previously analysed assuming 2C pump on reactor trip.
E.
Conclusions The results of the analy:aa described in this section can be summarized as follows:
1)
If the RC pumps remain operative, core cooling is assured regardless of system void fraction.
2)
For breaks greater than 0.005 ft, the 2CS nay evolve to systen void fractions in excess of 901..
^
3)
At 40 minutes, the 0.025 ft2 break has evolved to only a 7% void fraction.
Thus, a delayed RC pump trip for breaks less than 0 02 will not result in core uncovery.
4)
The potential for high cladding temperatures for a small break transient with delayed RC pump trip is restricted to a ci ma period between that time where the system has evolved to a high void fraction and the cima of LPI actuation.
5)
Even with 2 EFI pumps available, tripping of the RC pumps at t e
worst time (90% void fraction) results in a core uncovery period which cannot be shown to comply with 10CTR50.46, if Appendix assumptions are utilised.
6)
A prompt RC pump trip upon receipt of a low pressure ISTAS will provide compliance to 10CTR50.46.
7)
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The above conclusions are applicable to both the 44W 177 FA k
e and raised loop NSS designs.
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III.
IMPAC" ASSES!':C:T OF A RC **?P in? 0;' ::ON-LOCA 7TCT*$
A.
Introduction Some Chapter 15 events afe characterized by a pri response similar to the one following a LOCA.
- mary system The Section 15.1 events that result in an increase in heat reoval by th systa cause a primary system cooldown and depres'urizati e secondary like a mall break LOCA.
s on, such Therefore, ma assessment of the conse-quences of an imposed RC pop trip. upon initiation of th low 3C pressure ESTAS. was made for these events e
5.
General Assessment of Pao Irio in Non-LOCA Evenes goveral concerns have been raised with regard t a early pump trip would have sa sea-LOCA events that exh o the effect that abaracteristics.
Plant recovery would be more difficult, A
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sa natural circulation mode while achieving cold shutdo dependence.
highlighted assual fill of the stesa generators would b wn would be 4
and so es.
Eshwwer, all ef these drawbacks' eas be accommodat d e required.
nome of the will on its eum lead to easeceptable ceaseq e since i
restart of the pumps is set precluded for plaat centrol a d uences. Also, esce sontrolled operater acties is assmed.
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n coeldown three major conceras have sarfaced which hav Out of this search.
stantial amough as to require analysist e appeared to be sub-
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i 1.
A pump trip esuld reduce the time to system fill /repr t
er safety valve speatas following an evercoolias transient essurisattoa the time available to the operater for controlli If the margia of subcooling were substantially r d ng NFI flow and trip to where timely and effective operator ae uced by the pump
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ction could be questionable, the pump trip would become unacceptable i
2.
In the event of a Israe steen line break
(==*i=
blowdown may induce a stema bubble fa tha RCS whi h
== overcooling) the natural circulation, with severe ecesequences en the cor c could impair pecially if any degree of return t.o power is experien e, as-3.
A acre general concern exists with a large st ced.
conditions and whether or not a return to power laema line break at IC following the RC pump trip.
experienced natural circulation !!aw may not be sufficient to remo to avoid core damage.
ve heat and 1
-U-3
r-Overheating events were not censidered in the impact of the RC pump trip since : hey do not initiate the lov RC pressure ESTAS, and therefore, there would be no coincident pump trip.
In addi-tion these events typically do not :esult in an empty pressuriser or the formation of a steam bubble in the primary system.
Reactivity transients were also not considered for the same reasons.
In addi-tion, for cvarpressuriza: ion, previous analyses have shown that for the worce case conditions, an RC pump crip will sitigate the pressure rise.
This results from the greater than 100 psi reduction in pressure at the RC pump exit which occurs af ter trip.
C.
Ansivsis of Concerns and Results 1.
Syst m teoressurization la order to resolve this concern, as analysis was performed for a 177 FA plant using a MINITRAp model based on the case set up for IMI-2, Figure 3.1 shows the noding/ flow path scheme used sad Tabl, ell provides a descriptise of the modes
- and flew paths.
This case assumed that, as the result of a small steam line break (0.4 ft.2 split) er of some combination of secondary side valve failure, secondary side heat danand 1
l was increased from 1002 to 134: at time sere.
This increase la secondary side heat dessad is the smallest which results in a (high flua) reactor trip and la very similar to the worst moderate frequency evereeeling gvent, a failure of the c
steam pressure regulator. Is the analysis, it was assmed f
that following EFI setuattee on law RC pressure ESTAS, mala feedwater is ramped deva, M8tV's obt, and the auxiliary feedwater inittsted with a 40-secsed delay.
This action was takaa to secy the cooldown and the depressurizatioa of the systen as soon as possible af ter WI actuation, in order to minimize the time of refill and repressurisation of the i
system. Both RPI peps were assm&d to function.
The calculation vos performed twice, once assuming two of the four RC pumps running (one loop), and once asseing RC pump trip right af:er EFI ini:iation.
The analysis shows : hat the systa behaves very similarly with and without peps.
In both cases, the pressurizer refilla in about 14 to 16 minutas from initiation of the transients, with the na: ural circula- ~
tion case refilling about one minute before the case w1:h
\\
two of four pu:sps running (See Figures 3.2.3.3). In both cases, the system is highly subcooled, from a mini:aua of 30*F and increasing at the end of 14 minutes (refer to Figur j
It 14 concluded that an RC pump trip following HPI actuation will not increase the probability of causing a 1.0CA throu the pressurizer code safeties, and that the operator viil hav the same lead time, as well as a large margin of subcoo e
, to control HPI prior to safety valve tapping.
Although no case with all RC pumps was made. it can be inferred from the one loop case (with pumps running) that the subcooled margi be slightly larger for the all pumps running case. The pressurizar will take longer to fill but should do so by 16 minutes into the transient.
Figure itshows the coolant temperatures (het leg, cold leg, and core) as a function of time for the 3, RC pumps case.
2.
Effect of Stees Bubb'le on Natural Circulation Cooling For this concern, as analysis was performed for the same generie 177 FA plaat as outlined la Part 1 but assuming that as a result of an monitigated larga SLE (12.2 ft.2 DER),the eueessive cooldova would produce void formation in the p system.
The intent of the analysis was to also show the extent cf the void formation and where it occurred. A the ease analysed ta part 1. the break was synnetric to both generators such that both would' blev deva equally, =awhizi the cooldown (in this case there was a 4.1 ft.2 break on each loop).
There was as MSIT closure during the tranaf ent on either stema generator te maximias cooldown.
{
Also, the tur-hise bypass systen was assursed to operate, upon rupture, satil isolation on ESTAs. 25FAS was inittsted on low pressure and also actuated NFI (bot'h pumps
- tripped RC pumps (when applicable) and isolatN the.%
"s.
The AW was initiated to both generators on the low SG pressure signal, with minimum delay time (both pumps operating).
This analysia was performed twice, once assuming all RC pumps running. once with all pumps bett.g tripped on the HP1 actuation (af ter ISTAS), with a short (* 3 second) delay.
In both cases, voids were formed in the hot legs but the dura-
~.
tion and cize were smaller for the case wi:h no RC pump trip (refer to Figure 3.1).Although the RC pump operating case had a higher cooldown rate, there was less void forma-tion, resulting f rce the additional system sixing.
The coolant teoperaturas in the pressurizer loop het and cold legs, and the core, are shown for both cases in Figures 3.5 3.6 The core outlet pressure and SG and pressuriser levels versus time are given for both cases in Figurea 3.8 3.9.
This analysis show that the system behaves very similarly with and without pumps, although maintaining RC pump flow does seem to help sitigate void formation.
i The pump flow case shows a shorter time to the start of pres-ouriser rafill than the natural circulation case (Figure 3.9),
although the time difference does not sees to be very large 3.
Ef t'oce of Return to power _
There was no return to power exhibited by any of the SQL esses analyzed above'.
Ptsvious analysis experience (ref.
Midland FSAR. Section 150) has shown that a IC pump mitigate the consequences of an EOL return to power c by reduclag the cosidown of the primary system.
The reduced eeoldown substantially incresses the suberitical margia which in turn, reduces or eliminates return to power.
D.
_Conclucions and Summary A general assessment of Chapter 15 non-10CA events identifie three areas that warranted further lavestigation for impact of a RC pump trip as ESTAS low RC pressure signal.
1.
It was found that a pump trip does not significantly shorten the time to filling of the pressuriser and approximately the same time interval for operator action exists.
2.
For the maximum overcooling case analyzed, the RC pump t increased the amount of two-phase in the primary loop; however, the percent void forsation is still too small to affect the abil.f ty to cool on natural circulation.
l 3.
The suberitical return-to-power condition is alleviated by the RC pap trip case due to the t duced overcooling affect l
Based upon the above assessment and analysis, it is cen-eluded that the consequences of Chapter 13 non-1CCA even:s are
(
1 1: -
_c-
v incree:ed due to,the addition of a RC pump trip on ISTAS low RC pressure signal, for all 177 FA lowered loop plants.
Although there were no specif f analyses performed for "EC3 the conclusions drawn from t nalyses for the lowered icop plants are applicable.
e 9
e e
e e
S S
9 8
e
~
- able 2-1.
Analvsis Scope tiith A N Available Continuous RC Break size,
_ Break location pump ooeration RC pump trip ? 902 void 2
(ft )
Cold let Hot let
I 0.05 I
I*
I I*
0.075 I
I I
I 0.10 I
I I
0.20 I
I e
Analysed with both 1.0 and 1.2 ANS decay curves.
e t
9 6 - - _ - - _ - - -.
- -- - ^- ~^
^-^ ^' -~~
- ~ " ^ ~ - ~ ~ ~ '
~ ~ ~ ^ ^ ^ ^ ^
l l
Table 2-2.
Impact Assessment of Break Spectrum With RC Pump Trip at 90: Void Break size (f 2)
Core tmcover r time (see) 0.10 550 0.075 625 0.05 575 t
Notes:
1.
TW HFIs available during the transient.
2.
Core uncovery cia is the time period following pump trip re-Wed to fill the inner RY with water to an elsvacion of
- 9. ft ta the core which is sp-proximately 12.f t when swelled.
o l
t
- 21.
~~
, -_ __--.--- - --- ~~~~^"^~ ~
l I
i
\\
Table 2-3.
Comparison of system void Fractions at ESTAS Sirnal j
System Percent void Break size, (ft2)
_ Pumps on
(*) Pumps tripped (3) 0.02463 0.0 0.04 4.47 0.05 0.04 0.055 6.74 0.07 8.06 0.075 0.90 0.085 S.45 0.10 2.17 7.97 0.15 10.70 0.20 6.78 9
e i
I I
1 3
i
,em-
' ~
-MINITRAPR N0DE DESCRfP* ION
=
NODE W BER DESCRIPTION 1,33 2,34 Reactor Vessel, Lower Plenus 3,35 Reactor Vessel, Core 4,10 Reactor Vessel, Upper Plenum 5-7,11-13 Est Les Piping 8,14 Primary, steam Generator s
9,32 Cold Les Piping 15 Reactor Vessel Dowscomer it 24 Pressuriser 17.23 Stese Generater Downconer 18-20,26-28 Steam Generator Lower Planum 21,29 secondary, steam Generator 22,30 Steam Risera 23 Main Steam Piping 31 Turbine Centainment l
MDf! TRAP 2 PATE DESCRirn0W PATE NED N
- DgSQLIPTION 1,2 45,46 Care 3,3,5,11,12.44 core typians 6,7,13,14 Ret Las Piping 1
t 8,15 Primary, steam Generater 9J6 BC Pumps 10,43 Ce34 Lag Piping 17 h, Reacter Yessel 18,19.26,27 Preneuriser surge Line 20,21,28,29 Steam Camarator De ecamer 22.30 Secondary, Stama Camerater i
23,31 Aspirator 24,32 Steam Riser 25,33 Steen Piping 34,35 Turbine Piping 34,37 Break (or Leak) Path 38,39,43,44 EFI 40,41
.AFV 42 Maia Feed Pumps 21 Table 3.1 b
i h t
- - - - - ~ - ~ ~
Figure 2-1. CRMT2 Nading Diagram fer Small arcu 1@l
~
1@l m
s 7
u
~
i.
9 d.
g as u
e Um s
j ca
)
<s
.s a
g wk 7
i j
3 h
a
$g su.. mesiu=
e...
m s,
se......
u D u s uss ~ u.=
me.
6:
OC XD
,y g
t_ @
n
, _k,k um.
na e u O
e.
. s.. e..
i wu hen- @
e.
.... u..u a
s
,e,.
\\
h identifiestien
_ _ _ Path Wo.
Ideettffesele
~
2 Devoteser 2
3 ewer Piemos 3,2 eers 3.4.18.19 Ret Les Pistaa 3
Core. Core Sypass. 3pper l
Piemos. 9pper Road 3.30 set leg. Upper 6.21 38 febes 4.34 Ret 143 Paping 7.23 3,23
- Scane Generaser Upper S
3ry awer Mead Core appass aned, as rebee (apper salt) s.23.24 cou les Paphs a.as as tw6es a ee anat) as.34.2s hnpo 8.38 SS teuer Esad 11.12.13.16.24.27 Cold Leg F1 piss 3.23.29 ce34143 Pspass (Pie, section) 12.31 Deameener 3842.20 sold to: Pspsas (Pump steenasse) 23 ~
1P1 13 9pper Dewneseer 28,39 Spper Dewucater
. (Atove the (,et poss2e Salt) 30 Presser 14er 31 Presceriser 33 Test Yalve 23 See441 ament I
33,34 Leak 4.Regers Fath 33,34 EF2 37 Costaineemt Sprays
- 24.
l
,.__..--.~~--~~~#~~
~
Figure 2-2.
CRAFT 2 N0 DING DIAGRAM FOR SMALL BREAKS (EN0DEMODEU CFT G
g i
O 2
6 5
3
? )
1 O
(
w r
seo g
M PATHS 8 8 9 wa<
M e No.
Ideettfiestion Path Wo..
Identification 1
Core 2
Pr1 mary 54 2
Lyg 3
Core. UP. Eat Lags 3,10.11 IFI 4
Presserizar 4
Ret Zags l-5 Cascainment 5
Pwp.
6 I
See d ry SG 6
Vent Yalve 7
Pressurizer 8.9 Leak & Return Psch l
CORE PRESSURE VS TIME,177 LL. 2772 nit.PbMPS ON 2
_ 0.025 FT BREkt 23 N00E E00EL 0.825 FT2 gREAK
~
8 N00E E00EL 18 i
g 2
,5 2
18 i
E a
E g4 2
3 12 10 3
0 564
- 1800 1508 2800 2500 Time, sec Figure 2 3 l
l l
i 1
PERCENT SYSTEN V0103 VS Time,PUNPS ON 100 80 m
en2 i
Se 5
%N
. 9 !L g3
,.,..p1 sto,.b -7.0 4s = s Tkto 2 "T. (t % .,r ,,s.r ..e5 i a 800 1208 1880 2000 2480 l Tlas, sec Figure 2-4 0 t I l
h + -~~ ^ i 1 BREAK SPECTRUN RC PRESSURE flTH THE RC PUNPS OPERATIVE AND 2 HPl PUNPS 2500 2000 i 1500 m".-. __, * *== -.."-- 0. 025 FT j 1000 2 \\ N s's~ ~- \\ 500 g ILPIN '* 0 05 FT g 2 hm= i 0.075 FT2
- \\
,0.2 FT2 0.10 FT2 I 0 8 500 1000 1500 2000 2500 3000 3500 Tlas, sec Figure 2-5 9
--- --- -- - - ~ ~ ~ ~ ~~- ~ ~~~ ~
BREAK $PECTRUN AVERAGE 373TEs V010 FRACT 81TM THE RC PURP3 OPERATIVE AM 2 HPI PUNf3 100 p -~~~~~~,v q o ~ p _ / 1 p# Re r ) .ey .i. 1 ~.
- .- /, +/
A 88 \\ + ~:
- ./
v C. !] / = f 5 A* *i} N t 3 p ' h ='"**,,,,e
- 1
/ s.C 2 l +/ /'./, #
- s ',,,,e t.
29 - a / i 4 gn c 8 444 No 1290 1840 2800 2480 2888 ( Time,sec o ? Figura 2-8 b 5 4 lf i,
's ,a -~ ~ 0.1 FT2 BREAK WITH CONTINUOUS RC PUNP ~ OPERATION AND 2 HPI PUNPS 108 7 ' LPI f C /
- (
/ 00 / 2000 \\ / \\ G I \\ 2 / \\ >. 84 / \\ 1500 2 a / \\ = 3 f s i E i \\, 6 %), %g
- - 40 p
1000 I I s,% I N LPI 23 500 l '~~ ~ 0 g j 0 500 1000 1500 2000 2500 3000 l Tiss, sec Figure 2 7 e / l AC PRESSURE FOR 0.05 FT2 BREAK AVAILA81.E I HPl VS 2 HPI'S 30 3
2 NPl's, PUNP DN, HONOGENEDUS 25
.- o - 1 H I, DN,NON0 GENE 0M
- 2 29 2
E. 15 i 4 l \\ [ 19 ~ ~~ N *T*S %*N i t N4% % * %,*%*ON' %. 5 su 1 0 588 1800 1500 2988 2500 3000 e t ,+ 9 Tlas, sec Figure 2-8 e t s t s i
AVERAGE SYSTEN V010 FRACTION FOR 0.05 FT2 AVAILABLE 1 NPI VS 2 NPl'S 100 r.am m o I / \\ l /.// t a / I / I i. / f ---2 NPI'3, PtNIP DN,NONDEENE003 i 88 / -o-1 NPI, PUNP DN,N050 GENE 003 f E / . / I E / ee / l 48 .j/ / / 29 l f f f 9 e s 4-8 400 800 1299 1800 2800 2400 2800 3200 Time, see I Figure 2 3 O e
2 RC PRESSURE FOR 0.075 FT,PUNPS OFF a 905 SYSTEM VOID 30 1 _ __ _ 2 HP I 'S, PUNT ON, 25 N050GENEOUS - - 2 NPl *S, PUE OFF o { 905 VOID, 2 PNASE 20 2 E 15 = i 10 %g 4.%.~e 5 .s*=%.'._ N.%.. g 0 500 1000 1500 2000 2500 3000 Time, sec Figure 2-10 L[ l m- -.-.m_g-,,,m, e, -,,-r-w-y wp-e-m d
3 AVERAGE SYSTEN V010 FRACTION FOR 2 0.075 FT,PUNPS OFF e 905 SYSTEN V010 100 , - - - - ~ ~ - - '==+% p' '% y ~ 80 i / 2 / / 50 / 3 l 's / ~ ,/ 2 HPI's, PUNP DN, HOP 0GENEOUS f 40 ,/ 2 HPl'S, PUNP 0FF e 905 V010, 2 / 2 PHASE / ~ 2e g 8 480 800 1200 1800 2000 2400 2800 Time, sec Figure 2. O 9 e e AVAILABLE LIQUID 70 LUBE VS TINE FOR 0.075 FT2 BREAK flTH 1.2 ANS DECAY HEAT CURVE 3000 m y { 2500 _I E 2000 = 5 S' LEVEL OF ACTIVE CORE 3 1500 ~ = 5 2 1000 = i RC PUNPS OFF 500 L 8 400 800 1200 1800 2000 Time, see Figure 2 12 \\ i 35 -
RC PRESSURE VS TINE FOR 0.05 FT2 BREAK EITH 1.0 AND 1.2 ANS BEFORC AND AFTER PUMP TRIP 2 _ _ _ _ 0.05 FT, 2 HPI'S 3000 1.2 ANS, PUNP ON 2 0.05 FT, 2 HPl'S ~* 1.0 ANS, PUNP ON I 2500 0.05 FT, 2 HPI'S ~' ~ 1.2 ANS, PullP OFF 2 ~ " 0.05 FT, 2 HPI'S 2000 1.0 ANS, PullP OFF = = a .lii 1500 F I ~ 'D % 1000 Q 500 .0 O 500 1000 1500 2000 2500 3000 Time, sec Figure '2-13 l
.a PERCENT SYSTEM V010 FRACTION FOR 0.05 FT2 BREAK flTH 1.0 AND !.2 ANS BEFORE AND AFTER PUMP TRIP 100 M 80 .p#." I /f ~ E' 80 _,0.05 FT, 2 NPl *3,PWP DN, 2 g, 1.2 ANS / 8 2 . 5 FT, 2 HPI's, PSP DN, E 40 3 1.8 ANS f 2 / _o O.05 FT, 2 HPl *3JWP OFF, / 1.2 ANS 20 2 - g 8.05 FT, 2 IIPI'3, PWP 0FF, 1.8 ANS O e 0 400 800 1200 1800 2000 2400 2800 Tlas, sec Figure 2 14 _ _. r N
AVERAGE SYSTEN V010 FRACTION VS TIME FOR A 0.075 FT2 BREAK, BREAK LOCATin COMPARISDN PUNPS OFF f 905 YOID 100 INC0VERY TIME = 625 IIC S0 / l D UNC0 VERT TIME = / &lN ~ 450 SEC 5 / *g$ 43 4/ I i 1& 5 </t _// ~ 3' i l 8 488 000 1200 1880 2000 2400 Time, see Figure 2-15 G 9 i
- I CONPARISDN OF DELIVERED HIGH PRESSURE INJECTION FLUID TO RV FOR PUNP ofSCHARGE BR k
1200 N, 1100 N <4 N 1000 \\,'?"/ \\ s00 g $800 \\ g g' 700 \\ -C 800 th f \\ j 500 \\
- p, 400 -%
E8 l##p \\ f 3M 290 LOO! I HPI in - NI 1 gu f 8 200 SN 1000 14W 18N 2280 2808 3800 I Pressors (psla) Figurs 2-18 e - - ~ ~ ~ ' ~ ' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~~
447 v L .[.-.:--; ~ M S l8
- p....@_......_..........
.tr e........o i s' ewww O q g l,r j I 1 tnov~64-N* g_i- ) o l g .w slt 18 ( g y t Q ~ - ~y re Q if 6 i l ~ J 4 _1 ,I O{- M d 21 It. S A4 Q L 11 g g W O ts 8C csv 0 gw W 2( je 2 13 LT-o j y g o r I ' As* i u l c i.- 6 i W c1 I W J ) t I g .O G [8 i ~ TO e e W l Figure 3.1 MINInAP2 Noding and Flow Path Scheme i P;;E:St;.'.1:"T. I.:3 ITE.'." C::.'II:A10.0 L !C'J13 I E!!!. '!:::US T!'!::; LENT It"EI( (102 FP. E':0 0F L t.~E,0.6 FT2 ST:A:::.l;;i E. !.T (C3t::0!:.0 :::El.l.TE Fl:EO .J. (RC Pil.Y iNtP) l i S.G. TUBE REC 10N - ooeogo a: FULL 0 o E g .5 0 3 40 g PRZR. FULL g;---- E3 e o* 0 e 0 k o O 0 } 30 ga*0 ~ 3 9 a s g 3 gm 8 3 a oD 6 3 20 'd. e o g O $g 5 0 e t a g gy E e 10 4 m: PRESSURIZER 0 $ o: STEAR GENERATOR 'A' a: STEAN SEhTRATOR 'B' E i r 0 2 4 8 3 la 12 14 16 Transient Time (Ninutes)~ Figure 3.2 . :.1 -
. g. l 1 _5 .e \\
== L N -= = 23 s -* s 5 l5 5 I. s m a!$ E ~uu \\\\ 588 I ~O 1 N S =55 N 2 's
=
E= s E5 \\ ls s ~ EE 'h, ^ f4f n Nl N s O dW 's ~ 03^ i N, s = ,u s 5E l'\\ \\ \\ "w - ogOs I h, ,s, 0- = \\ g _~E = g s s 5=o u s s's g = W N s m s ma sg \\ s o=m-gb a. s r 53" \\ a d.f x,s O 3*I \\ R 2;
- g\\
s Q w . -a we r L E. k %,\\ ( e sE^
- gS k
s i
==5 b b =
- s. t i
m .C.M "' G4 0 E
- a. - a G
4 0 n G 4 O 9 4 O G r _g.- 3o e o .a, .a e e v n (;;) 1.'A313!ah 1 Ja:cs:ura :::csis Ja:Iniss:Jd Figure 3.3 1 l -u- ~l l -J l
~ 9#g %;p4, 9,,o OD 4 OG 4 n em 4 g k l" "h E = s em 4 8 5*E
- 5 I
om 4 P g _ n GE g eG 4 a. v X e ga 4 a w a w es
- g g
3$ 5 n. l eG 4 s a: ~ ~ n. m wa= 1 oc 4 - as s *- + m a 5.y e 00 4 m u u *< m. = w au a (- as .E ~ "' 5 '. l i SC 4 ........ ~ 3 U t gc4 1 m !gg i GS 4 E a = 1 I + 3% N es 4 G b %e 5 l ee 4 r a =* as I ea 4 g = m - y: l em 4 if 'E): na l eD 4 3 ^ -r I h: lE OG 4 j (.
- O l
60 4 i ce 4 i e 4 I e 4 t
- ~
2 4 g l< g 9 4 w t S 4 i s4 /w e4 E l h E (3,) gjnt cjsdesi 20c1003 Figuic 3 4 . A3 -
\\ C " *.* : T
- i.* * '. ;.
.....i...., ' I '"/ f-b ? i: !!c l' '....: t i p 7.,,2 2 fT2 F?. On +. i E ,,... ". ". ',.. "..,,,,,,a r. g ir. ~.e.,,..... I T I '". '. r a " ~ ~ ~ * ' - - 700 p P.'ESSU:t tlEt A E".PTIES f l \\ I, NOT LEG *EGit:5 TEO. PHASE U\\\\ J.CNFISCO78.M S00 FLG65 CEGlus [ p i I y O o t i c Ko \\ NOT LEG SUBcactED 'i 8 \\ l N S. 503 I t o N ~ 9% Ss = I h i n s 3 40s F' s
- .s 1
a % 'p#o, 0-l \\ { a \\- at g$ \\ b e' NOT LES (PS2R LOOP) \\ 300 a: CORE \\ COLD LEG (PAZR LOOP) h i a: j 1 ~~: SATURATION LINE } t:: 9 9. 200 ~ 8 2 4 g 3. 10 h 1 Transient Time (Ninutes)# Figure 3.5 l... lj p
l CGSI.I.';i TE";E':',T J.~,53 'iE.'.:':0 1 :?:::::::1 i."' (107 Fa BE'il'r!,140 0F L IFE,l' 2 fi LY.."di E 0 T.';Ti ":I, ":".*. ! T i 0.'. !!3 $ !E.'.'". !!:E !.7E3 f. 90 TU;P T;;lF; 700 PRESSURIZER i EMPTIES ( p HOT LEG lt BEGli:S TTO. PHASE c". b \\. " ' 's 3', CC, tie FLOCD TANK j s s FLO: CECl!S i N s 9 8 8 'N A 1i O o s HOT LEG SUSC30 LED a o 5 500 A'A, o G a o A E \\ s a W ~ I E 6 D %'A a 8 i a C W% *' u 400 a o f. a m 0 o D ^' s l3 KEY a e, D i^ a g' : 300 e: HOT LES (PRZA LOOP) l G: CORE 6 s 2, a: COLD LES (PRZR LD0p) 6 --: SATURATION LINE l 200 2 4 6
- 3 10 Transient Tme (Einute:)
l Figure 3.0 : J
S IL ?. i. '.'.. L E 'iO' 'r s '*i.G US iT.f '.512::T 1 li' l ilC2 FP. u.l!.1:. SF LIFE.12.2 f t? L Z ;E C'.2 f.U.M. :.;. 'i 171 'ii: siit."L 1: 12 o I f.
- i 6
Il IS 10 1: IS I *I Ig 1 I S I t 3 8 3 s i, s u .Q E t 3 i n 8 p I, = / l k = .mq s l 4 s \\ l ,i t e s d. a g n g s \\ s 2-t 8 t ,i,bs 8 < Aa s s i n 8 W-0 2 4 5 8 i Transient Time (Ninutes) KEY Figure 3.7 O: 110T LES (PRZR) RC PU2P TRIP 0: HOT LEG 'B' LOOP RC PU'4P TRIP a: HOT LEG (PSZR LO;i') NO TRIP 0: HOT LEC 'B' LOOP NO TRIP l l 1"- I
\\ C.*"E O l'; '. !." '.' "; ' W ' 8.'N i;f * "~ f f "i i I "E (102: FP. 0~3::::::':0 CF L:TI.!?.2 772 ;;.j,g Er4D T.ua..ti,b:..illG:.1ED SiiA.u!.! c::EAK) MCC.- i 'f' t I 2000.- W o 5 1500 0 5 a .t o KEY 5 o 0: RC PUMP 3 TRIP 1000 a: NO PUEPS TRIP 6 0 o o O O 20 e o a O cP o o$0gQoo6 b O a o0 0 0 2 4 6 8 10 12 i Transient Time (Venutes) Figurs 3.8 __,,._--r
e E0@ Di I ., i Q o Q (e -3 I w g N p CD fb N =-
- u. m
-m-C 14 9=.J C3 >= >= na
== W _J e.- e a a. :- Q un
- (Q
>=== to ::s et .'3 >= w = . in. u,*J Nl = s. D. g,;p &I G L3 ( auw ?
- g se?
. :s u et so ~ m o d u. ~ s. a. .s E i S** ^ h" e >- ( * - i gg g m 4 c:.)~ a. ,,, o as a. i S m SC a. M xM( O Q M a w D U g,, A E w w w y M tas O o - - =. = $a .d., 4 m g .C.F E9d M E E C"'S s'as' O e t.d a n, J a > = > = <f 8"" Ma-
a c _=..a 2%WW --..a. 3- .E, E " " - w aa = I s,, M en as: ( l w s g W taA Idd W M W W >= 4.
- tie, y
O b b 64 6#, Eg a
aooe EE4 =st G M M uJ M W 3W w _.s M 3 4 34 4 4 .MA .and %N N -a d#9 em 4/5 v $h i 4 C 9 "* 1, o y I l x-sa O C t"J c V f" s ce g i' 0:st) :A31 M Q'*g'$ ~i Fleure 3.0 w l l a
i RITTRENCES I I B.M. Dunn, et al., "B&W's ICCS Evaluation Model," RAW-10104. Rev. 3, August 1977. 2 Letter, J.H. Taylor (B&W to S.A. Varga (NRC), July 18, 1978. 3 R.A. Bedrick, J.J. Cudlin, and R.C. Foltz, " CRAFT 2 - Fortran Program for Digital Sh1stion of a titinode Reactor Plant During Loss-of-Coolant," 3AU-10092. Rev. 2 April 1975. J.F. Wilson, R.J. Crenda, and J.F. Patterson, "The Velocity of Rising Steam in a Bubbline Two-Phase Mixture " ANS Transactions. 5, (1962). W e i e 9 l l . 49 - I - \\
Docket No. 50-346 License No. NPF-3 Serial No.1-85 August 29, 1979 ATTACHMDrr C Figure 1 is a schematic of the interim Automatic Reactor Coolant Pump Trip (ARCPT) logic. The Reactor Coolant Pump and breakers are automatically tripped from the existing isolated and independent incident level 2 Safety Features Actuation System (SFAS) trip signals. An incident level 2 SFAS signal occurs on either of the following two conditions:
- 1) Reactor Coolant System pressure less than 1600 psig
- 2) Containment pressure greater than 4 psig Installation Desian & Operation For detailed description of the Davis-Bessa.%: clear Power Station, Unit 1 (DD-1)
SFAS, refer to the DS-1 FSAR Section 7.3. Each Reactor Coolant Pump is tripped from a separate output logic composed of two logic channels both of which must be de-energized to actuate a SFAS trip (safety grade). The trip signal to the pimaps la run from an isolated relay output as a " control grade" signal. The trip signal is installed in parallel to the existing manual and autoestic trip switches of each pump. The SFAS trip signal once actuated, can be manually blocked to each pump allowing restart. If a RCP b"eaker fails to trip on SFAS incident level 2, the operator is required to manually de-energise the associated bus. Testina Each logic channel is capable of being tested and tripped without tripping a ptmp (full SFAS ch====1 actuation). In addition a full SFAS actuation can be performed to rest trip one pump. Upon fasta11ation during DB-1 operation, each logic channel will be tested. Actual pusy trip testing will be accomplished during the first outage when the RCPs are taken off-line. S V l _m..
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