ML19254G983
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DJITRIBUTION:
LCENTRAL Fli.E iiRR READING ASB READING February 25, 1960 "0:0!WiDU." FOR:
T. Novak, Chief Reactor Syster.s 3 ranch, DSS F" '
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". 'iattne:s, 'iection Leader Auxiliary Syster.s Jrtnca, DSS SU.$ JECT:
ARl FLOW DESIG1 BASIS Attached are responses from Calvert Cliffs and AN0-2 in answer to the AP.1 generic staff request (also attached) for information to verify that
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operating plant AR4 system flows are adequate for design basis transients and accidents.
RSi$ is requested to evaluate these responses since theseransients and accidents are paft of FSAR Chapter 15 review. Please subnit your evalua-tions or questions direct to the ORPM; however, please provide copies to ASB. Previous review work by S. Newberry of the Till-l response to this r
question should be considered to provide review consistency. This work is part of Task Action Plan II.E.1.
For your information,AA00-2 has ARJ auto initiation, but Calvert Cliffs is presently manaally initiated but has been requested to modify to auto-matic initiation.
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P. R. Matthews, Section Leader E
Auxiliary Systens Branch Division of Systems Safety cc: w/3ttachments V. Benaroya C. Liang V. Leung cc: w/o attachrents W. LeFave E. Connerg=
G. Vissing
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Basis for Auxiliary Feecuatei System Flow Recuirements As a result of recent staff revicis of operating plant Auxiliary Feec-water Systems (AFWS), the staff concludes that the design bases and criteria provided by licensees for establishing AFWS requirements for flow to the steam generator (s) to assure adequate removal of reactor decay heat are not well defined or documented.
We require that you provide the following AFWS flow design basis infor-mation as applicable to the design basis transients and accident con-ditions for your plant.
1.
a.
Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:
d 1)
Loss of Main Feed (LMFW) 2)
LMFW w/ loss of offsite AC power 3)
LMFW w/ loss of onsite and offsite AC power
- 4) Plant cooldown
- 5) Turbine trip with and without bypass
- 6) Main steam isolation valve closure
- 7) Main feed line break
- 8) Main steam line break
- 9) Small break LOCA
- 10) Other transient or accident conditions not listed above b.
Describe the plant protection acceptance criteria and corres-ponding technical bases used for each initiating event identi-fied above. The acceptance criteria should address plant Jimits such as:
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- Maximum RCS pressure (PORV or safety valve actuation)
- Fuel temperature or damage limits (DNS, PCT, maxir.in fuel central temperature)
- RCS cooling rate limit to avoid excessive coolant shrinkage
- Minimum steam generator level to assure sufficient steam generator heat transfer surface to remove decay heat and/or cool dcwn the primary system.
2.
Describe the analyses and assumptions and corresponding technical justification used with plant condition considered in 1.3. above including:
a.
Maximum reactor power (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 flow and time delay between initiating event and introduction of AFWS flow into steam generator (s).
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.
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l'aximum pressure at which steam is released from steam generator (s) and acainst which the AF'..' pump must develop sufficient head.
9 Minimum number of steam generators that must receive AFW flow; e.g. 1 out of 2?, 2 out of 4?
h.
RC ficw condition - continued operation of RC pumps or natural circulation.
- i. 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 steam generator (s). AFW pump flow capacity allowance to accommodate the time delay and maintain m'inimum steam generator water level.
Also identify credit taken for primary system heat removal 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.
Operating condition of steam generator normal blowdown following initiating event.
m.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
n.
Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.
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a 3.
Verify that the AF.? cumps in your plant will supply the necessary fic > to the steam generator (s) as detemined by items 1 and 2 above considering a single failure.
Identify the margin in sizing the cump f. low to allow for pump recirculation flow, seal leakage and pump wear.
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t iu2s1 APKANSAS PCWER S. LIGHT CCMPANY PCST CFFICE SCX 551 UT3.E ACCX. AFC.NSAS 72203 C501) 371-.".000
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January 31, 1980 2-010-24 1
Director of Nuclear Reactor Regulation ATTil: Mr. Darrell G. Eisenhut, Acting Director i
Operating Reactors U.S. Nuclear Regulatory Commission Washington, D. C.
20555
Subject:
Arkansas Nuclear One - Unit 2 Docket No. 50-368 License No. NPF-6 Emergency Feedwater System (File:
2-1510.1)
Gentlemen; i
In response to your letter of November 6,1979, pertaining to the Emergency i
Feedwater System for Arkansas Nuclear One - Unit 2, the attached information is orovided.
More specifically, Attachment I addresses your recer:mendations in enclosure 1 and Attachment 2 addresses your enclosure 2.
For future reference, the actuation system for emergency feedwater on Unit 2 is the Emergency Feedwater Actuation System (EFAS) not the Engineered Safety Actuation System (ESFAS).
Very truly yours,
.DU 0~ f David C. Trimble Manager, Licensing CCT:0EJ:nak Attachments f
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VEveEA MCCLE SOUN UTLitES SYSTEM j
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.,s SYSTE!! FL:'.! RE7JIRE"ENTS Desian Bases for EF5 Punn Cacacity The design hase requirerent for Erergency Feedwater System (EFS) pung capacity, as stated in FSAR Secticn 10.3.9 is that each punp must be capable of delivering sufficient energency feedwater to the stean cenerator(s) to preserve their function as a sec:ndary heat sink for ncrnal shutdc..ns and the ;umo must also provide sufficient feedwater in c::bination with pressuri er sprays or the' RCS safety valves to preclude overpressuri:ation of the RCS for feedwater line break accidents.
Emercency Feeduater Puno Cesinn Point Based on the above requirements, the folicwing c:nditions were used to size each EFS pump.
1.
Steam generator water level is to be naintained when either steam generator is being used to remove up to 2.95% full pcwer in decay heat.
2.
The maximum steam cenerator pressure against which the EFS pump must. provide sufficient ficw is 1220 psia, wnich is 1105 of the steam generator design pressure.
3.
Sucticn pressure available to the EFS pump corresponds to the one foot water level in condensate storace tanks 2TalA or B.
4 Each EFS pump recirculation line capacity is a maxicun of 75 cpa.
5.
The maximun energency feedwater tsoperature is ICO*F.
The resulting rating of each EFS pump is 575 gpm at 2300 ft.
The not ficw rate supplied to either stean cenerator is 500 ppm at 1220 psi.
For conservatisn, 385 gpm is used in analyses as the nininum EFS ficwrate to either steam generator. The 435 gpn ficw rate is the assuced EFS ficw rate for the FSAR Chapter 15 Safety Analyses.
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Eaq.2 ed Information Ec;;rdinc t:i--- - ;se Transients cnd Accident Conditicrs.
The f:: :..inc informati:n is revidad in ac: rdance with the f4RC request for additi:nal inf:rmation regardinc e ergoncy feedwater systsa ficw requirements.
1.a.
Identify the plant transient and accident conditions considered in establishing AFils ficw renuire..ents.
- 1) Less of Hain Feedwater (LMRI)
'ithough not a design base event for detennining EFS pump capacity,,
s..a adequacy of EFS ficw to maintain steam generator heat renoval capability is shown by FSAR analysis 15.1.8.
2)
LMR! with loss of offsite AC power.
Although not a design base event for determining EFS flewrate re-quirements, the adequacy of available EFS flew is shewn by FSAR analys is 15,. l.9.
3)
LMR! with loss of cnsite and offsite AC pcuer.
Although not a design base event for deternining EFS punp capacity, the required ficu rate is the sane as that for a LMRI with loss of offsite AC pcuer.
In the remote case of failure of normal, perferred and energency electrical power, the required flow is delivered by the turbine driver EFS Oump.
- 4) Plant Cooldewn.
Aia design base event fnr si:ing the ER! oumps, each ER! pump has sufficient capacity to ensure adequate fl:w tn naintain staan generator
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uater level when either steen generator is being used to recove reactor decay heat, RCP heat, and primary and secondary systen water and metal sensibic heat.
These requirements are illustrated in Figure 1.
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Althcugh not a design base event for da:ccmir.ing EFS pun capaci ty, the adequacy of EFS flow for turbine rip with:ut bycass is shown by FSAR Analysis 15.1.7 A turoine trip with the steam bypass system available will not result in actuation of the EFS.
5)
Main Steam Isolation Valve Closure.
This event is enveicped by the less of main feedwater event discussed in FSAR Analysis 15.1.8.
- 7) l'ain Feedline Break.
As a design base event for sizing the EFW punps, the cases in FSAR Section 15.1.14 denonstrate that the capacity of each EFW punp is sufficient in c:nbination with either the pressuri:er sprays if AC power is available,nr the PCS safety valves if AC pcwer is lost,to prevent overpressurization of the reactor c0olant system.
3)
!1ain Steam Line Break.
Although not a design basis for determining EFS pump capacity, FSAR Section 15.1.14 verifies that EFW punp capacity is adecuate to maintain a water level' in the intact steam generator during the transient, thus preserving the sec:ndary heat sink.
9)
Small Break LOCA.
Although not a design basis event for determining EFS punp capacity, analysis of this event shows that the EF.! systen will maintain sufficient cass in the steam generator (s) to maintain then as effective heat sinks.
10)
Other transient or accident conditicns not listed above.
a) Plant Startup EFS ficw requirement is less than that required for plant cooldown.
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b) Hnt sti.ndby and hot snutde,in Although not a design base event for deter-ining EFS
- p cracity, the ERJ system is placed in operation to naintain stea Janeratnr water level.
Pump ficw requirement is less than that required for plant cooldewn.
1.b.
Describe the plant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.
- 1) Main Feedline Break Criteria a) The RCS pressure dcei; not exceed 110 percent of the design value.
b) The DNSR in the limiting ccoiant channel in the core shall not be less than 1.3 c) The peak local power density in the limiting fuel pin in the core shall not be sufficient to initiate centerline fuel nelting.
There are no specific cooldewn rates or steam generatnr water level acceptance criteria for this event.
2)
Plant Cooldown, a) Steam generator nornal water level is maintained in either steam generator until RCS temperature is cooled to 350*F and shutdcun cooling has been initiated, b) RCS Cooling rate is linited to no nere than 100*F per hour based en thermal stress considerations.
c) Maximum RCS pressure dces not exceed the tanperature dependent Technical Specificaticn limits based on low temperature over-pressure protection.
2.
Describe the analyses and assumotions and correspcnding technical justi-fication used with plant ccnditions censidered in 1.a above includino:
!iaxinum reactor pcuer (including instrucent error alleuance) at the a.
tire of the initiating transient or_ accident.
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The reactor pcwer, includi-- instr;.nn: c c e, a t th e :". - a the initiating event is cnnter.atively ass. ed to be 29':: P.;t,
which is 103 percent of licensed c:re zwer.
b.
Tine delay from initiating event to reactor trip.
For the main feedline break case, 58.5 seconds is assumed.
c.
Plant parameter (s) which initiates ARiS ficw and time delay between initiating event and intrnduction of AR15 ficw into steam generator (s).
L:w steam generator level coincicen': witn no icw pressure trip present and Icw staan generat:r level c/ cident with a differential pressure between the two steam genera:crs, with the nigher pressure steam gen-erator..being ' feed, are the parameter which automatically initiates EFW ficw.
For the main feedwater line break case, EP.4 flow is assumed to enter the intack steam generator at 123.5 seconds.
d.
liinimum' steam generator water level when initiating event occurs.
For the feedwater line break case, the initiaT stean generator inventory is 181610 lbm. This c:rres;0nds to a level at the high end of the in-dicating range chosen such that reactor trip on high nressure will occur simultaneously with a reactor trip en low staan generator level.
Such a choice assures that 1) the most limiting RCS transient prior to reactor trip has occurred and that 2) a. minimum u<nnd w v 'naat sink...
is assu.ed for the subsetuent cooldewn of the RCS.
e.
Initial steam generator water inventory and depletion rate before and after AR-!S flow cor.ences - identify reactor decay heat rate used.
Initial steam generator inventories and depletion rates have the greatest impact on the feedwater line break case with' respect to assuring that 1) the most liniting RCS transient pricr to reactor trip has occurred, and that 2) a worst case secondary heat sink exists for the subsequent c idnwn of the RCS.
Table 15.1.14-21 of FSAR Section 15.1.14 shcus the cenerator depletion rate for the feeriwater line break case. Additionally, the ca:e presen+.ed in Section 15.1.14 demonstrates that ERI pump cacacity is sufficient in c mbination with either the pressurizer sprays or the RCS safety valves to prevent overpressurization of the RCS.
Once ERl flou enters the intact steam generator, sufficient EFirl pump capacity exists to remove decay heat and maintain steam generator water level.
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"axinum pressure at wnich steam is released fcr st:a
- enerator',
- ;
and against wnich the AFU punp nust devcio; sufficient nead.
Each EFU will develop sufficient head against a pressure of 1220 psi which is 110 percent of steam generator design pressure.
9
!!ininum number of steam generators tnat nust receive AFU flew; e.g.
1 out of 2?, 2 out of 4?
Only cne of two steam generators is required to assure adequate removal of decay heat during all plant acciient and operating conditions.
h.
RC flow conditien - continued operation of RC purps or natural circulatien.
For the feedwater line break cases, reactor coolant pumps are assumed operating except when offsite AC power is unavailable.
During plant coold:wn, reactor coolant punps are assumed operating,
- i. Itaxinum AF.1 inlet temperature.
The naxirum EFS inlet tenperature is assumed to be 100 F.
j.
Folicwing a postulated steam or feed line break, tire delay assumed to isolate break and direct AR1 flow to intact stean generator (s).
AR1 punp ficw capacity allowance to acconnodate the time delay and caintain minimum steam generator water level.
Also identi fy credit taken for primary system heat removal due to bicudewn.
For the main feedwater line break analysis (FSAR,15.1.14), the conditions necessary to identify and isolate the affected steam generator occur at 57.5 seconds subsequent to the initiation of the event.
EFW flow enters the intact stean generator at 123.5 seconds.
!!o credit ias taken for primary heat removal due to blowdcun.
k.
Volune and maximum temperature of water in nain feed lines between steam generator (s) and AFHS ccnnection to nain feed line.
Initial main feodwater temperature is assumed to be 452*F, which
n ~e: pends te full iced pian: :: :,:':ns.
For the feedwater line n< case, main feec..a:Or flew is :s nmed t: be automa:ically reduced
'" en EFW ficu is assumed ta enter the to :2-o ;er:ent in 20 sec:nds.
s: San :nerator no credi-is tnen for the volure of feedwater that would normally be availcole in the feedline between the steam generator ar.d the ERi system c:nvection.
1.
Operating ccnditien of steam cenerator nornal blowdown following initiating event.
Stean generator normal blewdcwn is not ccnsidered subsequent to the initiating event.
During plant accident conditions bicwdown is isciated upon a main steam isolation signal (11515) resulting from icw steam generator pressure.
Primary and secondary system water and retal sensible heat used for m.
cooldown and A?W ficw sizing.
6 1.54 x 10 BTU / F Tine at hot standby and tire to cooldewn RCS to RHR system cut in n.
tcmperature to si:e AR water source inventory.
The ccndensate storage tank water volure is adequate to enable one hour of hot standby operation folicwed by a three to four hour plant cool-dcun to shutdcwn cooling systen initiation temperature.
3.
Verify that the ARI pumps in your plant will supply the necessary ficw to the stean generator (s) as determined by items 1 and 2 above considering a single failure.
Identify the cargin in sizing the pump fic$ to allow for pump recirculation flew, seal leakace and pump wear.
The ERI system will supply the necessary flow to naintain the functior of the staam generators as effective heat sinks.
Each ERf pump can provice the required system ficwrate to either stean generator.
Redundant valves, piping, pump 3, and control systens and diverse cenponent power supplies ensure that given a single failure, sufficient feedwater is supplied to the steam generators.
FSAR Table 10.4-11 and Appendix 3A provide additional inforeatien with respect to the ability of the EFS to perform its design function in the event of a single failure.
The allowance for ERI punp recirculation is 75 gp, per pump.
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