ML20236Q317
| ML20236Q317 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/13/1987 |
| From: | Ellison R, Parece M BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20236Q289 | List: |
| References | |
| 51-1170352-01, 51-1170352-1, NUDOCS 8711190310 | |
| Download: ML20236Q317 (13) | |
Text
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BWNP 20440 5 (12/86)
S Babcock & Wilcox ENGINEERING INFORMATION RECORD a uccemen company 1170352-01 Document identifier 51-Title FPC Diesel Loading Evaluations PREPARED BY:
REVIEWED BY:
Martin V. Parece Randy H. Ellison Name Name Signature M 8 [/
Date ///////7 Signature (ME b hA-Date 4'//4 f 7
/
l
/
Technical Manager Statement: Initials Reviewer is Independent.
Remarks:
This document reflects changes to the Crystal River 3 diesel loading evaluations performed in Rev. 00 of this file. The information presented here is a supplement to the Rev. 00 file and is added to address NRC comments received during an NRC/FPC/GCI/B&W meeting on 11/5/87.
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G During an Operatimal Safety Team Inspectim by the NRC at Crystal River 3, the 1NRC identified an inconsistency between the energency diesel generator (EDG) loadings,and the surveillance testing performed at the plant. %e team also identified the overload condition for the EDGs.was less than the mavi==
expected loading..Be NRC requested a review of the' inconsistency and a
. r niktion of that' review to the NRC in C=ta-E.
Gilbert Ownnnwealth, Inc.
-(GCI) began addressing the diesel inading issue for Florida Power Corporation (FPC).- 2e calculatims evaluated the actual electrical loadings due to -
instrumentation and other arrianant Inadings expected.to be automatically l) loaded following a loss of offsite power in conjunction with an Engineered Safeguards Actuatim System (ESAS) actuation. B e loadings are higher on EDG "A", so a failure of EDG "B" was postulated for this evaluation. By.
f
.p postulating"B" diesel failure, the "A" diesel will be required to handle all of q
the load fra safety systems for a ' longer period of time. Mus, if the "A" diesel can be shown to have acceptable inadings, the "B" diesel, by entwrison can also be shown to have acceptable inadings. Among the largest individual loads are those naarciated with the high pressure injection (HPI), low prwatm injecticm (IPI), bi41 ding spray (BS), and energency feedwater (EFW) pungs.
Since for various scenarios each puup will have a different inading, it was danidad to perform an evaluation of the diesel Inadings hamad on event scenarios. GCI developed a series of scenarios to bound the FSAR accidents.
Be scenarios had as a rmnnn assunption'a loss of offsite power, and thus a diesel start, at the initiation of the event. %e flow rates==c+ad and the aaanciated horse power requirements for the individual punps were examined to arrive at ocriservative inadings for each event. Due to the nature of the task,
)
=avimization of diesel inading, see of the scenarios develop differently than J
dameribed in the Crystal River 3 FSAR. Raamd on this, the scenarios which I
diverge. form the PSAR scenarios will not be evaluated for specific I
thermodynamic rairisies since the design basis of the plant remains the same as that r==i= M in the FSAR.
A meeting between the NRC, FTC, GCI, and B&W personnel was held on Novmber 5, 1987. During the meeting, additional clarification on the NRC's opinion on a pv_,--M emergency feedwater punp trip was received. S e NRC's current l
position is that an autmatic trip of an emergency feedwater puup during large break IOCAs, when EFW flow is not required, is not acceptable. %erefore, a revision to the original issue of this file is being made to address this change. Additionally, the NRC asked for clarification on the peak biilding pressure expected during the steam line break event scenario. A calculation of this consistant with the FSAR method is presented and indicates a lower prwuaim than that shown in the FSAR.
(
Io 32-12 7 03 52.0h l
x Diammaian of Scenarios:
'Ihe scenarios &osen by GCI4Ld. a wide range of thermal hydraulic canditions. Msemina of this,. the events prwide a good basis for determining 3
the range of diesel loadings a-+=1 for design basis. accidents at the Crystal River 3 plant.. For each event type, the scenario is prwided with a description of the --+ad fim rates frun the major pumps. Again, the r=adar is reminded that the intent is to determine conservative diessi' loadings and-1not define additional design basis events for. the Crystal River 3 plant.
In all cases where EFW is actuated, it is ma==ad that ficw will be required to L
rea d the natural circulation steam generator (SG) level. 'Ihis value has been increased fran the B&W generic 50% SG level to allow for =miP M t calibration and instrument ura d alinty. 'Ihe SG level for all power levels will be required to be increased to the natural circulation point since, following reactor trip, the steam generator levels collapse to a liquid pool. Since sad event is a===ad to exist coincident with a loss of offsite power -(IDOP), EFW will be initiated on the loss of reactor coolant (RC) pungs.
IARGE IREAK IDCA i
'Ihe large break IDCA is typified by a rapid primary systan C-raizatim as I
the break flow rwMnaa the primary systen inventory. 'the rapid
)
depeaamwization initiates HPI and IPI en the low reactor coolant pressure functim of the ESAS. Additionally, the BS will be actuated by the ESAS high-high reactor biilding pressure function when the reactor biilding pressure reases the 30 psig setpoint. 'Ihe primary systen pens response has been predicted by approved evaluation inndals in Reference 1.
'Ihe contairanent response is described for different break sizes in the Crystal River 3 FSAR.
Pbr large break IOChs, the primary systen pressure will be approximately 100 psi at approximately 15-25 seconds after the rupture.
'Ihe' actual timing is dependent upon the actual break size and location. In all cases, the pressure is sufficiently low that IFI flow is expected once the pung is 1*iad onto the EDG. Since the primary systen pressure will remain low during the entire event, it is ma==ad that the IEI flow will be at capacity. 'Ihis will maximize the diesel Inadings for this event.
During large break IOCAs, the primary systen voids rapidly, t1==4 r==ica11y f
g da m rling the primary and secondary system. Virtually no heat transfer, frun the primary to the secondary, nmns after the first few sect:1ds of the event.
In fact,' the IDCA analysis assumes EFW flow to provide a sewimry systen heat source for energy addition to the primary. Emergency feedwater will be initiated m loss of RC pungs or ESAS. Since the steam generators will not be shaming, the EFW flow nust only account for level incraaaam, based on fill i
rate control by EFIC, and recirculation. 'Ihe EFW flow required can be calculated based m the fill rate desired. Using an area of approximately 66 f
ft2 for the secondary tube region and dowrn nar, it will take approximately 330 a
. gallans of water to raise the level 8 inches. Since the mav4== fill rate for EFW is 8 indes per minute, the fill rate alone will require a flow of 330 gpn.
'Ibe recirculation flow rate is taken as 200 gpn for either one or two pungs operating. 'Iherefore, the total required flow rate is 860 gpn (330+330+200).
With the two EFW punps sharing the flow desarri, the motor driven EFW punp flow will be half of this total, or 430 gpn. When steam pressure drops below 3
x 82-1170352 01 o
800 psig,.the' fill rate is limited to 2 inches per minute.. Ihis results in a total f1w requirement of 365 gpe ( [33(M330]*[2/8]+200 ).
'Ihe turbine driven emergency feedwater pump, EFP-2,.can deliver the required flw at this pressure. However, ~ if EFP-1 were required to provide this entire flw, it is less than the flow rate for EFP-1 aasn=ad for the initial phase of the event.
'Iherefore, maanning EPP-1 nust supply 430 gpn during the entire event will be cosimervative.
l Within the first 30 seconds of the event, the HPI, IPI, and BS pungs will be f
Wad to be inadad onto the energency diesel generator. Licensing. analysis aasn=as that the HPI flw reaches the reactor ocolant system by 35 seemds.
j Due to the lw pressures, the HPI system will be in nearly a runaut omditim.
l
'Ihis f1w has been determined to be 600 gpn, hamad on the pump performance curves. Also, p.-thwilling of the valves in the HPI lines will prevent runaut of the HPI puups. 'Ihe IFI systen will be up to full flow plus l
recirculation, since the primary cystan pressure will be below the 100 psia IPI l
design point-(3000 gpm per puup at 100 psig). R111 f1w of 3150 gpa is l
aasn =ad. 'Ihe bi41 ding spray punp will also be in operation with its flow l
beginning at 56 seconds.and full flow reached by 71 seconds. milding coolers are aa==ad operational within 25 seconds of initiation which is consistent, with FSAR analysis assumptions.
'Ihus, the aryiir===1t required to be operating and the appropriate flow rates are venenarized below for the IBUX:A event.
HPI -
600 gpn IPI -
3250 gpn, including 100 gpn recirculation 1600 gpn BS EFW -
EFP-1 flow of 430 gpn INITEMEDIATE IREAK IOCAs Intermediate break IOCAs are defined as those break sizes which result in a final event candition of severely dagaissed primary system pressures to the point of IPI flow, but do not exhibit the piasmana associated with a large break IDCA (i.e. blowdown, refill, and reflood). 'Ihus, during the early portims of the event, the primary and secondary systans are still coupled.
'Ihis means that EFW flow will result in heat transfer early in the event. Once the primary systen levels decrease below the secondary sufficiently, the only heat transfer whidi can occur is between the seomdary systan liquid and the primary system steam. When the primary system depressurizes below the secondary systen pressure, the heat transfer frun primary to secondary is reversed and the secondary system acts as a heat source and not a heat sink.
At this point, the fill rate-in the steam generators will easily be met as very little staaming is nmirring. As the event evolves frun a seomdary cooling event'into a ence thruxjh oore cooling with HPI and IPI, the diesel j
loading will change. 'Ihus, the EDG loadings for this event will range from j
those for the small break IOCA, to be di-aad later, to the loadings for the j
large break IOCA, dae ribed earlier. '1he only differences are the timing of the agiirmarit actuations and not the actual loadings themselves. At all times during this scenario, the diesel loadirxjs will be bounded by that specified for the large break IDCA scenario.
Y
u 82.-1170352 OL Due to the extended nature of the events, the EShS high-high reactor Nilriing ponemure motpoint will be reached later in the event, as ocupared to large.
bonak IOChs. A review of the various IOCA analyses for this type of break size indicates that the BS will normally be required prior to IPI flow initiation.
'Iherefore, the aastaption that BS is a load during the entire scenario is'a ocnservative aammption and will lead to higher diesel IrwHngs. 'Ihe resulting equipnent flows following the depressurization to the IPI flow range are
==marized below.
HPI.-
600 gpn IPI -
3250 gpn 1600 gpn BS EFW.
430 gpn (worst case ocndition when IPI, and BS are operating) i I
STEAM IINE MEhK EVENT
'the steam line break event is a severe overcoolirq event caused by the failure of a main steam line.' ' 'Ihe increase in steam flow and the depraamtrization of the r== uddry systen inczmaaaa the primary to rain.udary heat transfer. 'this
]
will lead to a primary systen cuda. action.
Assuming a IDOP at the initiation of the event, results in a trip of the reactor, trip of the RC pumps, trip of f
the main feedwater punps,. and trip of the turbine.
'Ihe EFIC system will initiate EFW haamd on the loss of the RC pungs. Also, the break will result in EFIC isolating the steam generators haamd on low SG pressure. 'Ihis signal will occur nearly instantaneously as the pressure wave frun the opening of the break moves through the secondary systan. 'the EFW flow will be ocntrolled to raise the SG 1evel to the natural circulation setpoint of 65% cm the operating range. If the break is inside ocmtainnent, EFW is fed to the unaffected SG while the affected SG is allowed to boil dry. If the break is outside ocntainnent, the break will be isolated frun both generators, and EFW will be delivered to both SGs.
'Ihe EFW flow required can be calculated haamd on the fill rate desired and the 2
steaming expected. Using an area of approximately 66 ft for the rm udary tube region and downocner, it will take approximately 330 gallcos of water to raise j
the level 8 inches.. Since the maximn fill rate for EFW is 8 inches per minute,. the fill rate alone will require a flow of 330 gpn. 'Ihe staaming rate can be estimated as 470 gpn total based en a 3% FP decay heat assunpticm. 'Ihis assumpticm is justified since the decay heat will have &m :4 to approximately s
3% within the first 2 minutes, Reference 6.
'Ihe 470 gpn can be calculated by ma==irq an approximate 1100 BIU/lbn enthalpy increase fran EFW injection until it is reladaad as saturated steam through the MSSVs. 'Ihus the total flow required is a sum of these two values plus a recirculation value. 'Ihe recirculation flow rate is taken as 200 gpn for either one or two pungs operating. 'Iherefore, the total required flow rate is 1000 gpm (330+470+200).
With the two EFW punps sharing the flow desarrl, the motor driven EFW punp flow will be half of this total, or 500 gpn. Although this flow rate will decrease during the event, especially after the natural circulation level has been reached, it is conservative to assume a constant flow of 500 gpn for the event.
5
w 82-H 70852 'O~1 l
i 1
-If the turbine driven EFW punp is ma==ad to fail, instead of the "B" EDG, the motor driven ptmp will be required to provide the total desired now. However, the flow frtan one EPW pung to one SG, with the other isolated, is 940 gpn including a 200 gan recirculation flow. 'Ihis limitation is due to a change in
)
the EFIC logic whidt provides flow limitation for overcooling and tube impingement concerns. 'Ihus, for the failure of the turbine driven puup,'the flow danand frtan the notor driven puup will be 940 gym.
For a steam line break outside contalment, both SGs will be available-after isolation, so the required flow will iiu.a by the fill rate requirement cm the extra steam generator. 'Ihe total required flow rate will then be 1330 gpn l
I (330&330+470&200). Splitting this flow between two EFW pungs will place a flow requirement of 665 gpn en the motor driven punp. If the turbine driven punp is-aa==ad to fail-for the same break location, the motor driven pump will try to fulfill the needs of both steam generators for both fill rate and steaming.
However, the motor driven punp will be limited by the total flow available frtan the EFW syntesa hamad cn previous GCI analysis of the EFW system. 'Ihis flow rate, at full siu 4ary systen praamim is approximately 1070 gpa. 'Ihus, the motor driven EFW pung will have a flow rate of 1070 gun for the case of a steam i
line break outside contalment with a failure of the turbine driven EEW puup.
' Since this flow rate does not meet the EFIC denarrl, the refill period will be slightly extended. 'Ibe fill rate is within the allowable EFIC band at all timaa, even with the EEW flow rate less than the demand.
A review of the mass and energy relaaaaa for this evertt was made to determine if the BS is actuated. Based cm the conservative FSAR assumptions, BS would not be expected for this event. Additionally, an i=aadiate MEW punp trip coincident with reactor trip and RC puup trip at the beginning of the event ensures mass and energy relaaaaa to the containment will be less severe than that predicted in the PSAR. A value of less than 23 psig was calculated. '!his calculation is r===Aed in Appendix A of this report. Consequently, the reactor bi41 ding pressure will reunain below the high-high reactor biilding pr=aans setpoint.
'Ibe HPI flow can be caneervatively nammad to be at the =av4== flow of 600 gpn while the system is depressurized. In actuality, the HPI flow will be dependent upon-the system pressure arx1 will continue to decrease during the event. Once the primary systen repressurizes, the HPI flow will be haamd on the a system praamim of approximately 2500 psig. 'Ihe a===rtion by GCI of a change in HPI flow at 10 mimtes is conservative for diesel leadings and will bound that expected for this type scenario. 'Ihe agiir= ant required to operate during this event are listed below.
HPI - < 10 minutes 600 gpn
> 10 minutes 295 gpa IPI -
100 gpn recirc No flow since not actuated BS EEW -
Break inside cxmtainnent 500 gpn Break cutside ocntainment 665 gpm Break inside w/ EFP-2 failure 940 gpn break outside w/ EFP-2 failure 1070 gpn b
32-1170352 01 FEMGmE UNE MERK
'the feedwater line break event'is very similar to the IDFW event. Again since this is an overheating event, the HPI and IPI will not be initiated unless the break is inside cantainnent. If the break is outside containment, the HPI and l
IPI pags will not be actuated.. 'Ihe mass and energy releases for this event are insufficient to cause the biilding sprays to actuate. 'Ihis can be easily
- verified, since most of the fluid will exit the break starting frtan a ahled
.or saturated liquid state in the feedwater system. Since the latent heat of l.<
vaporization is not present, approximately 500 BIU/lba at design steam canditions, the faarhater line break will put out about 40-45% less energy.
'Iherefore, the feedwater line break biilding pressure will be bounded by the steam line break case, and will runnin under the BS actuation setpoint.
Unrestricted HPI flow for this event will result in the primary systen remaining pressurized at the pra==wizer safety valve setpoint and the system will evenh=11y go solid. 'Ihe HPI will then puup against this prenant until operator action to cool the plant down and control HPI flow. EFW will be initiated on low steam generator pressure. Also, the feedwater and steam syntans will be isolated by the same signal. 'Ihe flow frtan EPP-1 will be controlled by EPIC to raise the steam generator level of the unaffected SG.
Iong term cooldown will be Maad on the cooldown rata desired by the operator, the aWsaric dump capacity and the decay heat levels. 'Ihe v4==nt required to operate for this event are listed below.
HPI -
Flow Maad on RCS pramann if break inside containment IPI -
If break inside containment, recire flow. No flow if break outside containment No flow since not initiated BS EEW -
Flow Maad cui danand for steaming and level increase
)
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82-1170352 01 Paferences
' 1.
BAW-10103A, Rev. 3, "EOCS Analysis of B&W's 177-FA Lowered-I. cop NSS -
Revision 3", Dated Jtily 1977.
2.
BAW-1976, "h11 Break Ioss of Coolant Accident Analysis for B&W 177FA Iowered Icop Plants in Response to NUREG-0737, Item II.K.3.31", Dated Sept:enbar 1986.
- 3..
86-1103119-00, "Cantalment Response to a Small Break IOCA", Oantract 620-0012,0013, Task 04-001-001, July 10, 1979.
4.
86-1103119-01, " Containment Response to a Small Break IDCA", Contract 620-0012,0013, Task 04-001-001, namnhar 4,1979.
5.
32-1159004-00, " Task AS-4 Operator Actions /N Cire", contract 582-7454, Task 999, February 1987.
6.
'32-1150622-00, "ANS 5.1 (1979) Decay Heat Envelopie",- Contract 582-7151, Task 502, Seg*mnhar 26, 1985.
Crystal River 3, Final Safety Analysis Report, Chapter 14, Table 14-25,"
7.
Revision 8, 7/87.
O
t 3-82-1170352 01-j l
Appenibc A hiildig Prassure Fbilowing l-A Steam Line Break Event A review of the FSAR analysis for the bi41 ding pressure response following a steam line break event was made. 'Ihe pressure pdad in the FSAR is hmmed cn an niilihrium corxiition of an instaittiu-cus release of all the mass and energy for the event with no credit taken for. heat transfer to the heat sinks in the biilditig. 'Iha solution technique is to assume a te ture in the e
biild%g and verify that the steam, liquid and air energies result in the proper building canditions. 'Ihis involves the calculation of the quality of the water at the aamunad canditions.
Specifically, an air tasperature is aamunad and its resultant energy content is calculated. 'Ihe energy content of the air is subtracted frem the total energy added to contalment and a quality is determined frm these ocnditions. 'Ibe energy content of the steam and liquid is then determined. Finally, a volume check is made to determine if the mass of steam at its partial pressure can fill the entire cantalment volume. It is assumed that the volume occupied by the liquid is small and its volume reduction can be neglected.' 'Ihis results in a slightly higher pressure than if the volume of the liquid mass was subtracted fra the cxxitaiment volume. 'Iha devalr= ant of the equations is shown below.
Steam p ugm l.ies are taken fr a the ASME steam tables. Input data is taken fra the FSAR, or derived from that data.
Ototal " Osteam + Oliquid + Oair liquid) x = quality = Mass
/(Mass
+h Mass
= x
- Mass x = Mass y Massg g
D iquid " Il~*)
- total
- hg + Mass,g* c
- ST Q = x
- Mass
- hg + (1-x)
- Mass y
AT = the change in temperature from initial to final conditions Q' = Q - Q,g = Q - Mass,g* c *dT y
h Q'/ Mass total f
h h
g g
At a given temperature, the air energy can be calculated. 'Ihe quality is then evaluated and then the total energy checked. Because of the manner used, the total energy will always work out correctly. 'Ibe final check is that the steam occupies the correct volume. 'Ihis is determined by multiplying the steam mass by the specific volume for the assumed cxxxiitions. 'Ibe building te ture is varied until the calculated steam volume matches the contalment e
free volume. 'Ihis agnilibrium temperature yields the steam partial pressure j
(P at) and the air partial pressure (idaal gas law). Sumation of the partial s
pressures yields the containment billding pressure.
j
N 82-1170352 01 The tables cn'the following page shows tim calculations ard irput. She base irput is takan' fmn the PSAR, Table 14-25. The feedwater mass is determined by a===4ng a five seccmd ramp to zero flow follcwing the IOOP at time zero. This mass is shown in the revised mass and energy table. The total energy is rd==d by the mass ctange multiplied by the average mass energy fmn the PSAR.
1his -ints for the energy in the feedwater and the energy renovea fmn the 1
R3 as the feedwater flows through the steam generator.
Fim the calculations, the peak bsildiry pressure will be approximately 22.66 psig. This rsenins below the PSAR value of 28, and will ensure that the bi41 ding sprays are not actuated during this event.
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82-1170352 01 Table 1 FSAR Data from Table 14-25 Energy Mass, lb Millicms of BIUs Staan Generator Inventory (fouled) 62,600 36.0 Feedwater F1w (6.5s full flw plus 30,600 13.6 28s coastdown to 0% flw)
Reactor Coolant System Energy 76.0 Transferred Available Mass in Feedwater Line 23,750 10.5 Between Feedwater Ocmtrol Valves and Steam Genarator' Steam Fl w from Unaffected Steam 9,250 11.1 Generator (until trip) 126,200 147.2 Average energy of fluid = 1166.4 BIU/lbn
)
I b
i I-ll J
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32-1170352 01 Table 2 Mjusted Mass ard Energy Resultirg from IDOP and SIB Energy Mass, lb Millions of_gBjs Steam Generator Inventory (foulad) 62,600 36.0 Feedwater Flow (5s coastdown to 3,733 1.6 0% flow)
Reactor Coolant Systen Energy 46.3 Transferred i
Available Mass in Feedwater Line 23,750 10.5 Between Feedwater Omtrol Valves ard Steam Generator i
Steam Flow frun Unaffected Steam 9,250 11.1 Generator (until trip) 99,333 115.9 j
Average energy of fluid = 1166.4 HIU/lba Energy of Feedwater is based on average 444 BIU/lbu Energy from IK:S is difference between FSAR and new total a===Ng an average energy of 1166.4 BIU/lbn for all fluids.
1 1
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r 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 6 y
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