ML19331C693
| ML19331C693 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 08/31/1980 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML19331C687 | List: |
| References | |
| NUDOCS 8008190395 | |
| Download: ML19331C693 (14) | |
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.IHis DOCUMENT C0tlTAINS POOR QUAUTY PAGES I
DUKE POWER COMPANY '
MCGUIRE NUCLEAR STATION UNITS 1 and 2 ATTACHMENT 2 Response to Enclosure 2 of NRC letter of March 10, 1980,
" Basis for Auxiliary Feedwater System Flow Requirements" l l
August, 1980 I,
I f 8008190Y
, Ouestien 1 ,
- a. Identify the plant transient and accident conditiens considered in establishing AFWS ficw requirements, including the following events:
- 1) Less of Main Feec (LMFW)
- 2) LMFW w/ loss of offsite AC power
- 3) LMFW w/ loss of onsite and offsite AC pcwer
- 4) Plant cooldown
- 5) Turbine trip with'and without bypass
- 6) Main stern isolation valve closure
- 7) Main feed line break
- 8) Main stern line break
- 9) Small break LCCA
- 10) Other transient or accident conditions not listed above. .
- b. Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.
The acceptance criteria should address plant limits such as:
- 1) Maximum RCS pressure (PORY or safety valve actuation) .
- 2) Fuel temperature or damage limits (OrlB, PCT, maximum fuel cen-tral temperature)
- 3) RCS cooling rate limit to avoid excessive coolant shrinkage
- 4) Minimum steam generatcr level to assure sufficient steam genera-tor heat transfer surface to remove decay heat and/or cool down the primary system.
Rescanse to 1.a The Auxiliary Feedwater System serves as a backup system for supplying feedwater to the secondary side of the steam generators at times when the feedwater system is not available, thereby maintaining the heat sink capabilities of the steam generator. As an Engineered Safeguards Sys-tem, the Auxiliary Feedwater System is directly relied upon to prevent core damage and system overpressurization in the ever.t of transients I such as a loss of normal feedwater or a secondary system pipe rupture, and to provide a means for plant cooldown following any plant transient.
Following a reactor trip, decay heat is dissipated by evaporating water in the steam generators and venting the generated stesa either to the condensers through the steam dump cr to the atmosphere through the steam generator safety valves or the power-operated relief valves. Steam generatcr water inventory must be maintained at a level sufficient to ensure acequate heat transfer and ccntinuation of the decay heat removal process. The water level is maintained under these circumstances by the Auxiliary Feedwater System which delivers an emergency water supply to the steam generators. The Auxiliary Feedwater System must be capable of functiening for extended periods, allowing time either to restore normal feedwater ficw or to pecceed with an orderly cooldown of the plant to a47SA w
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the reactor ::olant temperature where the Residual Heat Removal System can assume the turden of decay neat removal. The Auxiliary Feedwater System ficw and the emergency water supply cacacity must be sufficient to remove core decay neat, reacter coolant pump heat, and sensible heat during the clant c:cid:wn. The Auxiliary Feedwater System can also be used to maintain tne steam generator water levels acove the tuces fol-lowing a LOCA. In the latter function, the water head in the steam generators serves as a barrier to prevent leakage of fission products from the Reactor Coolant System into the secondary plant.
DESIGN CONDITIONS The reactor plant conditions which impose safety-related performance requirements on the design of the Auxiliary Feedwater System are as follows for the McGuire Unit No. 1.
Loss of Main Feedwater Transient Loss of main feedwater with offsite power available Station blackout (i.e., loss of main feedwater without offsite power available)
Secondary System Pipe Ruptures Feedline rupture
- Steamline rupture Loss of all AC Power Loss of Coolant Accident (LOCA)
Cooldown Loss of Main Feedwater Transients. .
. The design loss of main feedwater transients are those caused by:
Interruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system
- Loss of offsite power or blackout with the consequential shutdcwn of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a reduction in steam generator water levels wnich results in a reactor trip, a turbine trio, and auxiliary feedwater actuation by the protection system logic.
Following reactor trip frem a high initial power level, the power quickly falls to decay heac levels. The water levels continue to decrease, progressively uncovering the steam generator tubes as decay heat is transferreo and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety Or power-ocerated relief valves to tne atmosanere. The reactor
- olant temperature increases as the resicual heat in excess of that dissipated thr u;n tre steam generators is aoscr:ec. With increased temperature, the volume Of reacter coolant expancs and begins filling a 75A
the cressuri:cr. Without the addition of sufficient auxiliary feed- .
water, furtner exoansion will result in water being discnarged through the pressuri:er safety and/or relief valves. If :he tem:erature rise and the resulting voicmetric exoansion of :ne crimary coolant are :er-mitted to continue, then (1) pressuri:er safety valve cacacities may be i exceeded causing overaressurizatien of the Reactor Ccolant System and/or (2) the continuing loss of fluid from tne crimary coolant system may i result in bulk boiling in the Reactor Coolant System ano even:ually in core uncovering, loss of natural circulation, and core dnage. If such a situation were ever to occur, the Emergency Core Cociing System wculd be ineffectual because the primary coolant system cressure exceeds the snutoff head of the safety injection pumos, the nitrogen over-pressure in the accumulator tanks, and the design pressure of the Residual Heat Removal Loos. Hence, tne timely introduction of sufficient auxiliary feedwater is necessary to arrest the decrease in the steam cenerator water levels, to reverse the rise in reactor coolant temperature, to prevent the pressurizer from filling to a water solid condition, and eventually to establish stable hot standby conditions. Subsequently, a decision may 50 made to proceed with plant cooldown if the problem can-not be satisfactorily corrected.
The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equip-ment. The loss of power to the electric driven condenser circulating water pumps results in a loss of condenser vacuum and condenser dump valves. Hence, steam formed by decay heat is relieved through the steam enerator safety valves or the power-operated relief valves. The calcu-
$atedtransientissimilarforboththelossofmainfeedwaterandthe ;
b}ackout, except that reactor coolant pump heat input is not a consid-eration in the blackout transient following loss of power to the reaccor '
coolant pump bus.
Secondary System Pice Ruotures The feedwater line rupture accident not only results in the loss of feedwater flow to the ste n generators but also results in the ecmolete bicwdcwn of cne sten generator within a short time if the rupture should occur dcwnstream of the last nonreturn valve in the main or aux-iliary feedsater piping to an individual steam generator. Another sig-nificant result of a feedline rupture may be the spilling of auxiliary feedwater to the f aulted sten generator. Such situations can result in
- ne injecticn of a disarcocrtionately large fract5 of the ::tal aux-iliary feedwater ficw (tne system creferentially cumos water o the lowest pressure region) to the faulted icop rather than to che effec-tive steam generators which are at relatively high cressure. Tne system design must allow for terminating, limiting, cr minimizing that fractica of auxiliary feecwater ficw which is delivered to a f aulted loco or sailled througn a break in ceder to ensure that sufficient flow aill be delivered to the remaining effective sten generatcr(s). The concerns are similar for the main feedwa:er line rupture as those exclained for the loss of main fee: water transients.
1275A
Main steamlire rupture accident ::nditions are chart:terized initially by cl ant tocidewn and, fer breaks inside containment, ty increasing containment pressure and tem:erature. Auxiliary feedwater is not needed during the early phase of :ne transient but flow to the faultec loc?
will contricute to an excessive release of mass and energy to contain-ment. Thus, steamline ruoture c:nditions establish the u;per limit en auxiliary feedwater flew celivered to a f aulted icoo. -ventually, how-ever, tne Reactor Coolant System will neat uo again and auxiliary feed-water flow will be required to be delivered to the non-faulted locos, but at scmewhat icwer rates than for the loss of feedwater transients described previously. Provisions must be made in the oesign of the Auxiliary Feedwater System to limit, control, or terminate the auxiliary feedwater flow to the f aulted loco as necessary in creer to prevent containment overpressurization felicwing a steamline break inside con-tainment, and to ensure the minimum flow to the remaining unf aulted loops.
Loss of All AC Power The loss of all AC power is postulated as resulting from accident con-ditions wherein not only onsite and offsite AC ccwer is lost but also AC emergency power is lost as an assumed common mode failure. Sattery power for operation of protection circuits is assumed available. The impact on the Auxiliary Feedwater System is the necessity fer providing both an auxiliary feedwater pump power and control scurce which are not dependent on AC power and wnich are capable of maintaining the plant at hot shutd:wn until AC power is restored.
Loss-of-Coolant Accident (LOCA)
The loss of coolant accidents do not impose en the auxiliary feedwater system any ficw requirements in addition to those required by the other accidents addressed in this response. The following descripticn of the small LCCA is provided here for the sake of completeness to explain the role of the auxiliary feecwater system in this transient.
Small LOCA's are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volume. The principal con-tribution from the Auxiliary Feedwater System following such small LOCAs is basically the same as the, system's function during het shu::cwn or foll: wing s:uricus safety inf e::icn signal snicn tri:s the sa tor.
'<2intaining a sater level invent:ry in the sec;ndary site of the steam generat:rs pr:vides a neat sink for removing decay heat anc establishes the :a: ability for providing a cuoyancy head for natural cir:ulation.
The auxiliary feedwater system may be utilized to assist in a system coola wn and decressuri:ation following a small LOCA while bringing the reactor :: 3 cold shute:wn ::ndit'en.
Coold:wn e
The ::alccwn uncti:n performed by the Auxiliary Fee: water System is a c ar t i 2'. cre si nce t e rea:::r ::: lint system is -ecuce: f :m normal :er0 7 0ad ten;eratJres to 2 h0 le; te90erature of 10 r:X'matel, 33CcF, 1CEA
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5-he latte- is the maximum temoerature rec:mrended for placing the Residual Hea Remova' System (RHRS) int: service. The RMR sys;em ccm-pletes the c:oi:enn to ::ic snutd:wn conditicns.
Cooldown may be required folicwing ex:ected :-ansients, fellcwing an accident such as a sin feedline brea'<, :r curing a normal cooldewn prior to refueling or performing reacter slant maintenance. If the reactor is trip:ed following extended creration at rated power level, the AFWS is capable of delivering sufficient AFW to remove decay heat and reacter coolant cump (RCP) heat foll wing reactor trip wnile main-taining the steam generator (SG) water level. Following transients or accidents, the recommended cooldown rate is consistent with expected needs and at .the sete time does not impose additional requirements on -
the capacities of tne auxiliary feedwater pumps, considering a single failure. In any event, the process consiscs of being able to dissipate plant sensible heat in addition to the decay heat produced by the reactor care.
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l l 175A
Res:ense to 1.b Tacle 13-1 summari:es the criteria which are One general design bases f:r each event, dis:ussed in :ne resacnse to Question 1.a, 3:ove. Spe-cific assumptiens used in :ne analyses to verify :na: :ne design bases are met are discussed in response to Question 2.
The primary function of the Auxiliary Feedwater System is to orovide sufficient heat removal cacability folicwing reactor trio and to remove the decay heat generated by the core and prevent system over-pressuriza-tion. Other plant orotection systems are designed to meet short term or pre-trip fuel failure criteria. The effects of excessive coolant -
shrinkage are evaluated by the analysis cf the, rupture of a main steam pipe transient. The maximum ficw recuirements determined by other bases are incorporated into this analysis, resulting in no additional ficw requirements.
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TABLE 18-1 O
m Criteria for Auxiliary Feedwater System Design Basis Conditions Condition or Additional Design
_T ransi ent Classification
- Criteria Cri_teria _
Loss of Main feedwater Condition 11 Peak RCS pressure not to Pressurizer does not become exceed design pressure. No water solid consequential fuel failures Station Blackout Condition II (same as LMFW) Pressurizer does not becone water solid Steomline Rupture Condition IV 10CFR100 dose limits containment design pressure not exceeded feedline Rupture Condition IV i- 10 CFR 100 dose limits. No hot leg bulk boiling prior Containment design pressure to transient turnaround not exceeded Loss of all A/C Power N/A Note 1 Same as blackout assuming turbine driven pump Loss of Coolant Condition III 10 CFR 100 dose limits 10 CFR 50 PCT limits Condition IV 10 CFR 100 dose limits 10 CFR 50 PCT limits Cooldown N/A 1000F/hr 5570F to 3500F
- Ref: ANSI N18.2 (This information provided for those transients performed in the FSAR).
4 Note 1 Although this transient establishes the basis for AFW pump powered by a diverse power source, this is not evaluated relative to typical criteria since multiple f ailures must be assuned to postulate this transient.
3 Question 2 s
Describe the analyses and assumstions and corresconding technical justi-fication used with plant condition considered in 1.a above including: ,
- 3. Maximum reactor pcwer (including instrument error allowance) at the time of the initiating transient or accident,
- b. Time delay from initiating event to reactor trip.
- c. Plant parameter (s) which initiates AFWS ficw and time delay between initiating . vent and introduction of AFWS flow into steam generator (r,. ,
- d. Minimum steam generator water level when initiating event occurs.
- e. Initial steam generator water inventory and depletion rate before and after AFWS flow commences -- identify reactor decay heat rate used.
- f. Maximum pressure at which steam is released from steam generator (s) and against which the AFW pump must develop sufficient head.
- g. Minimum number of steam generators that must receive AFW flow; e.g.,
1 out of 27 2 out of 47 ,
- h. RC flow condition -- continued operation of RC pumps or natural ci rcul ati on.
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- 1. Maximum AFW inlet temperature.
J. Following a postulated steam or feed line break, time delay assumed to isolate break and direct AFW flow to intact steun generator (s).
AFW pump flow capacity allowance to accommodate the time delay and maintain minimum steam generator water level. Also identify credit taken for primary system heat removel due to blowdown.
- k. Volume and maximum temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line.
- 1. Ocerating condition of steam generator normal blowdown following initiating event.
- m. Primary and secondary system water and metal sensible heat used for l
cooldown and AFW ficw sizing.
- n. Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory, i
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Resconse t: 2 Annlyses have been cerfccmed for the limiting transients whicq define One AFWS performance requirements. These analyses have :een prcvided for review and nave been approved in the A::plicant's FSAR. Speci-fically, they include:
Loss of Main Feedwater (Station Blackout)
- Rupture of a Main Feedwater Pipe
- Rupture of , Main Steam Pipe Inside Centainment In addition to the above analyses, calculations have been performed specifically for McGuire Unit No. I to determine the plant cooldown flow (storage capacity) requirements. The Loss of All AC Power is evaluated -
via a comparison to the transient results of a Blackout, assuming an available auxiliary pump having a diverse (ncn-AC) power supply. The LOCA analysis, as discussed in response 1.b, incoroorates the system flows requirements as defined by other transients, and therefore is not performed fcr the purpose of specifying AFWS flow requirements. Each of the analyses listed above are explained in further detail in the fol-icwing sections of this response.
Loss of Main Feedwater (Blackout)
A loss of feedwater, assuming a loss of power to the reactor coolant pumps, was performed in FSAR Section 15.2.9 for the purpose of showing that for a station blackout transient the ceak RCS pressure remains below the criterion for Condition II transients and no fuel f ailures 6ccur. Table 2-1 summarizes the assumotions used in this analysis. The transient analysis begins at the time of reacter trip. This can be done because the trip occurs on a steam generator level signal, hence the core power, temoeratures and steam generator level at time of reactor trip do not depend on the event sequence prior to trip. Althoug the time frem the Icss of feedwater until the reacter trip occurs cannot be determined frca this analysis, this delay is expected to be 20-30 sec-ands. The analysis assumes tnat the plant is initially operating at 102*, (calorimetric error) of the Engineered Safeguards design (E50) rating shown en the table, a very conservative assumption in defining decay heat and stored energy in the RCS. .The reactor is assumed to be trioped en low-Icw steam generator water level, allcwing for level uncertainty. The FSAR shows that there is a censiderable margin with res:ect to filling the pressurizer.
Ruoture of .ain Feedwater ?ine The double anded rupture cf a main feedwater pipe dcwnstream of the main feedwater line check valve is analy:ed :n FSAR, Section 15.4.2.2.
Taoie 2-1 s .mmari:es :ne assumptions used in :nis analysis. Reacter trip is assumed to :ccur den the f aulted generatcr is at the icw-icw level set:oint (acjus:ed f:r errors). This :ensarvative assumoticn maximizes the store: nea ori;r to -eac::r trio anc minimi:es the asil-ity of the steam genera:Or to remove eat f-:= :ne RCS foll: wing reacter tri: cue :: 2 ::nse-va:i.>ely small ::tal s:eam generator inver.::ry. As 2075A .
in :ne i:ss of n:rmal #eecuater analysis, tre initia! ;;ae- ca ing was assumed to :e 1227 cf the ESO -ating. Auciliary #eecwa:er f w Of 450 can aas issumed t: be ceiiv ere: :: -he 3 non fa;lte: 5:eim generaters i minute af ter reactor trip. The criteria listed in TI:le 13-1 are met.
This analys s estaclisnes recuirements fcr laycut to crec'ade indefinite loss of auxiliary feedwater to tne postulated break, and estaclishes train association recuirements for ecuipment so that the AFWS can deliver the minimum ficw required in 1 minute assuming the worst single failure. 3 Rupture of a Main Steam Pice Inside Containment 1 Secause the steamline break transient is a cooldown, the AFWS is not needed to remove heat in the short term. Furthermore, addition of excessive auxiliary feedwater to the f aulted steam generator will affect i the peak cntainment pressure folicwing a steamline break inside con-tainment. This transient is cerformed at three pcwer levels for several break sizes. Auxiliary feedwater is assumed to be initicted at the time of the break, independent of system actuation signals. The maximum ficw :
is used for this analysis, censidering a case where runout protection -
for the largest pump fails. Table 2-1 summarizes the assumptiens used in this analysis. At 10 minutes af ter the break, it is assumed that the operator has isolated the AFWS from the faulted steam generater which subsequently blews down to ambient pressure. The criteria stated in Table 13-1 are met.
This transient establishes the maximum allewable auxiliary feedwater ficw rate to a single f aulted steam generator assuming all pumps opera-i ting, establishes the basis for runout protection, if neeced, and estab-lishes lavout requirements so that the ficw requirements may be met considering the worst single failu.re.
l Plant Coold:wn l
Maximum and minimum ficw requirements frca the previously discussed transients meet the flew recuirements of plant c oldewn. This opera-tien, however, defines tne basis for tankage size, based on the required cooldown duration, maximum decay heat input and maximum s:cred heat in the system. As creviously discussed in res:cnse 1A, the auxiliary feed-water sys em cartially c:cis the system to the poin: anere the RHR$ may c:colete :Pe :::id:an, : .e. , 35Cc7 in tne RCS. Table 2-1 shcws tne l assum:tions used to determine tne ccold:wn neat capacity of the aux-iliary fee: water system.
The cocidcwn is assumed to ccmmence at the maximum rated pcwer, and maximum tric delays and Cecay heat scurce terms are assumed wnen the
-eact:r is trip;ec. Primary metal, primary water, secondary system metal and se::ndary system water are all included in One stored heat to l be ren ~.ed by !"e AfWS, 30e II:le 2-2 f:r One items unsti:Uting tne l
sensiDie 9 eat s:Ored in :ne ,1333.
l l Tqis ::erati:n :s inaiy ed :: estab' ;;h rinimum tank :ize re:ui ement:
( # :" aux ' ' 2 ") #ee:wa e# #';i f s0ur e a9icn are Crma'If ali;ned.
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TA8tE 2-1 Suunary of Assumptions Ussi in AfW5 Design Verification Analyses loss of Feeduster Mata 5teaaline 8 seas Transient (st.tlon blackoup Cooldown Mala feedline_ Sreat _ (gentainment) _
- a. Mas reactor pnwer 1021 of E50 rating 3651 mt 0 1021 af E50 ratin9 10. 102% of rated (102% of 3579
- t) (1921 of 3579 Ntl (percent of le25 MJt)
- 6. I6me delay from 2 sec 2 sec 2 sec vernable event to As trip
- c. MWS actuation sig. 10 1o % level NA low-low % level Assumed imediately nel/ time delay for I minute I minuts O set (no delay)
AFW5 flow
- d. SG water level at low-low SG water NA low-low % water level N/A tline of reactor trip level minus 5 percent alnus 12 percent of of span . span
. t
- e. Initial % inventory 49.050 lbm/5G (at 65.205 Itum/ % 89.830 lbm/ ruptured SG consistent =6th power trip) at 544.60F 18.570 Itse/ intact % ,
pate of change before See F54R N/A turnaround at 2254 sec N/A
& af ter ATWS actuation Figure 15.2.8 1 ,
with offsite :over i turnaround at 1986 sec .
without offsite power Decay heat F5AR figure 15.l.8-1 F5AR figure 15.1.8 1 FSAR Figure 15.l.8 1 F5AR Figure 15.1.A.I *
- f. AFW p. sap design 1225 psia 1225 psia 1225 psia N/A pressure 9 Mini,*:s i of %s 2 of 4 N/A 3 of 4 N/A =
whlih must receive AfW flow '
- h. RC peanp status irlpped at reactor trip Tripped Operating and tripped All nperating '
at reactor trip I. Ma Imum AfW ' 1208F 1600F 120T equal to main feed temperature temperature J. Operator actton none N/A none
, 10 minutes ftfw) ftI /Inop (for MFW purge volume / 3 '
- t. 135 f t /443.30F 150 ft3 / 50 ft 1/443.)"F 5/G and temperature 440*F dryout time)
- 1. Normal blowdown none assumed none assumed none assumed none assumed '
- m. Sensible heat see cooldown Table 2-2 see cooldown N/A
- n. T 6sse at stanJtsy/t ime 2 hr/4 hr 2 hr/4 hr 2 hr/4 hr N/A to cooldown to AIR
- o. MW flow rate 450 GPM - constant variable 450 gia - constant 1400 GPM (constant) to let , . r.n..i r m no a.<*a - ' - -
TABLE 2-2 Summary of Sensible Heat Sources Primary Water Sources (initially at ESD power tem:erature and inventory)
- RCS fluid
- Pressuri:er fluid (liquid ar.d vacce)
Primary Metal Sources (initially at ESD power temperature)
- Reactor coolant piping, pumps and reactor vessel
- Pressurizer
- Steam generator tube metal and tube sheet
- Steam generator metal below tube sheet :
- Reactor vessel internals .
Secondary Water Sources (initially at ESD power temperature and inventory)
- Steam generator fluid (liquid and vapor) .
- Main feedwate. purge fluid between stern generator and AFWS piping.
Secondary Metal Sources (initially at ESD power temperature)
- All stesu generator metal above tube sheet, excluding tubes.
Questien 3 Verify tnat the AF'a' pumos in your plant will suoply the necessary ficw to the stern generator (s) as determined by itens 1 and 2 above con-sidering a single f ailure. Identify tne margin in si:ing the pump ficw to alicw for pump recirculation ficw, seal leakage and pump wear.
Response to 3 a) The AFW pumps will supply the necessary flow to the steam generators -
considering a single failure.
b) A margin of 2% is available to account for pump wear.
c) Pump recirculation flow is automatically isolated f611oving any automatic pump start.
d) Both trains are designed for a total leakage of 2 gpa during the design basis event (feedwater line break).
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