ML18139A943
| ML18139A943 | |
| Person / Time | |
|---|---|
| Site: | Surry, North Anna  |
| Issue date: | 12/26/1980 |
| From: | Sylvia B VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton, Varga S Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.E.1.1, TASK-2.E.1.2, TASK-TM 999, NUDOCS 8012300585 | |
| Download: ML18139A943 (15) | |
Text
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VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGIN IA 23261 December 26, 1980 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation Attn:
Mr. Steven A. Varga, Chief Operating Reactors Branch No. 1 Division of Licensing U. S. Nuclear Regulatory Commission Washington, D. C.
- 20555
Dear Mr. Denton:
Serial No. 999 NO/LEN:ms Docket Nos. 50-338 50-280 50-281 License Nos. NPF-4 DPR-32 DPR-37 AUXILIARY FEEDWATER SYSTEM REQUIREMENTS SURRY UNITS 1 & 2 AND NORTH ANNA UNIT 1 We have reviewed your letter of September 25, 1979 that requests information concerning Surry auxiliary feedwater system design basis information and pump flow verification.
Our response to these items is provided in Attachment 1.
We believe that the attached response adequately addresses the NRC concerns for Surry Units 1 and 2.
On October 6, 1980, Vepco responded to the request for design basis informa-t_ion and pump flow verification for North Anna Vnit 1.
In that response the NRC was informed of a discrepancy discovered during the North Anna Unit 1 review.
Previous to our response, LER 80-061/0IT-O was issued explaining the discrepancy found in the pump flow calculations.
In verifying auxiliary feed pump 1-FW-P-3B flow through calculations, it was estimated that the pump would deliver 335.gpm to the steam generator instead of the design basis of 340 gpm.
This is to document that pump tests have since been performed and have verified 340 gpm are delivered to the steam generator.
Should you have any questions or require additional information, please contact ui:;.
Very truly yours,
~{~
B. R. Sylvia anager - Nuclear Operations and Maintenance
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Attachment cc:
Mr. Robert A. Clark, Chief Operating Reactors Branch No. 3 Division of Licensing Aoo/
1/1
,**<<.,I ATTACHMENT 1 AUXILIARY FEEDWATER SYSTEM DESIGN BASIS SURRY UNITS 1 & 2
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Question 1
- a.
Identify the plant transient and accident conditions considered in establishing AFWS f1 aw requiranents, 1 ncl uding the foll owing events:
- 1)
Loss of Main Feed (Lflf'W) 2 3
)
LMFW w/loss of offsite AC power
)
Lflf'W w/loss of onsite and offsite AC power
- 4) e1 ant cool down
- 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 pl ant protection acceptance criteria and corresponding technical bases used for each initiating event identified above.
The acceptance criteria stl:>uld address plant limits such as:
- 1) Maximun RCS pressure (P~V or safety valve actuation)
- 2)
Fuel tanperature er danage limits (DNB, PCT, maximum fuel central tenperature)
- 3)
RCS cooling rate limit to avoid excessive coolant shrinkage
- 4) Minimun stean generator level to assure sufficient stean gen-
- erator heat transfer surface to remove decay heat and/or a>ol down the primry system.
Response to l.a The Auxiliary Feect-tater Systan serves as a backup systan for supplying feect-tater to the secondary side of the stean generators at times when the feedwater sys tan is not available, thereby maintaining the heat sink capabilities of'the stean generator.
As an Engineered Safeguards Sys-tan, the Auxiliary Feedwater Systan is directly relied upon to prevent core danage and systan overpressurization in the event of transients such as a loss of ronnal feedwater or a secondary systan pipe rupture, and to provide a means fer plant cooldown following r.y plant transient.
Following a reactor trip, decay heat is dissipated by evaporating water in the stean generators and venting the generated stean either to the condensers through the stean d1111p or to the atmosphere through the stean generator safety valves or the power-operated relief valves. Steam generator water inventory must be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat ranoval process.
The water level is maintained under these circunstances by the Aux11i ry Feedwater System which delivers S'I emergency water supply to the stean generators.
The Auxiliary Feedwater Systan must be capable of functioning fer extended periods, allowing time either to restore nonnal feedwater flow or to proceed with an orderly cooldown of the plant to the reacter coolant tanperature where the Residual Heat Ranoval System can ass1111e the burden of decay heat ranoval.
The Auxiliary Feedwater 6685A
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System fl cw and the emergency water supply capacity must ~e sufficient to remove a>re decay heat, reactor a,olant pump heat, and sensible heat during the pl a~ cool dCMn.
DESI~ Q)ND!TIOOS The reactor pl ant conditions which impose safety-related perf onnance requiranents on the design af the Auxiliary Feedwater System are as foll arts for the Surry Units 1 and 2 pl ants.
Loss of Main Feedwater Transient Loss af main feedwater with offsite power available Station blackout {i.e., 1 oss af main feedwater without off site power available)
Rupture af a Main Steam Line Loss of all AC Power Loss of Coolant Accident (LOCA)
Cool ct>wn Loss of Main Feedwater Transients The design lass of main feedwater transients re those caused by:
Interruptions of the Main Feedwater System flCM due to a malfunction 1n the feedwater or condensate system Loss of off site power er blackout with the consequential shutdo,m of the systen pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a rapid reduction in stean generator water levels which results in a reactor trip, a turbine trip, and auxilf ary feedwater actuation by the protection system logic.
FollCJ1tfng reactor trip fran a high initial power level, the power quickly falls to decay heat levels. The water levels continue to decrease, progressively uncovering the sten generator tubes as 'decay heat is transferred and discharged in the fonn of sten either through the steam d1111p va hes to the condenser or through the ste1111 generator safety or power-operated relief valves to the atnosphere. The reactor cool ant temperature increases as the.residual heat in excess of that dissipated through the sten generators fs absorbed. With increased temperature, the vol1111e of reactor coolant expands m,d begins filling the pressur-izer. Without the addition of sufficient auxiliry feedwater, further expansion w111 result in water being discharged through the pressurizer safety and/r,r relief valves. If the temperature rise and the resulting 6685A
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volt111etric expansion of the primary coolant are permitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing overpressurization of the Reactor Coolant System and/or (2) the continu-ing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually 1n core uncovering, loss of natural circulation, and core damage.
If such a situation were ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pressure exceeds the shutoff head of the safety injection pumps, the nitrogen over-~ressure in the accumula-tor tanks, and the design pressure of the Residual Heat Removal Loop.
Hence, the timely introduction of sufficient auxiliary feedwater is necessary to arrest the decrease in the steam generator water levels, to reverse the rise in reactor coolant temperature, to prevent the pres-surizer from filling to a water solid condition, and eventually to establish stable hot standby conditions. Subsequently, a decision may be made to proceed with plant cooldown if the problem cannot be satis-factorily 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 generator safety valves or the power-operated relief valves. The calcu-lated transient is similar for both the loss of main feedwater and the blackout, except that reactor coolant pump heat input is not a consider-ation in the blackout transient following loss of power to the reactor coolant pump bus.
Rupture of a Main Steam Line Because the rupture of a main steam line may result in the complete blowdown of one steam generator, a partial loss of the plant heat sink is a concern. Main steamline rupture accident conditions are charac-terized initially by plant cooldown and, hence auxiliary feedwater is not needed during the early phase of the transient to remove decay heat from the Reactor Coolant System. Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop. Eventually, however, the Reactor Coolant System will heat up again and auxiliary feedwater flow will be required to be delivered to the unfaulted loop, but at somewhat lower rates than for the loss of feedwater transients described previously. Provisions must be made in the design of the Auxiliary Feedwater System to allow limitation, control, or termination of the auxiliary feedwater flow to the faulted loop as neces:sary in order to prevent containment overpressurization following a steamline break inside containment, and to ensure the m1nimll!I flow to the remainin~
unfaulted loops.
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Loss of A 11 AC Power The lass of all AC power 1s postulated as resulting fran accident" con-ditions wherein not only onsite and offsite AC pa.iter is lost but also AC emergency power 1s last as an assLmed cormion mode failure. Battery power for operation of protection circuits is ass1.J11ed available. The impact on the Auxiliary Feec:Mater System is the necessity for providing both an au~il i ary feeCMater PLmP pCMer and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown unti 1 AC p<Mer 1s restored.
Loss-of-Coolant Accident (LOCA)
The lass of coolant accidents do not impose on the auxiliaryfee<Mater system ~Y fl CM requirenents in addition to those required by the other accidents addressed in this response. The follCMing description of the snall LOCA is provided here for the sake of canpleteness to explain the role of the auxiliary feec:Mater system in this transient.
Small LOCA's are characterized by relatively slow rates of decrease in reactor coolant system pressure and liquid volune. The principal con-tri bution fran the Auxiliary Feedwater System foll owing such snall LOCAs is basically the sane as the system's function during hot shutdown or following spurious safety injection signal which trips the reactor.
Maintaining a water level inventory in the secondary side of the stec1T1 generators provides a heat sink for removing decay heat and establishes the capability for providing a buoyancy head for natural circulation.
The auxiliary feeCMater system may be utilized to assist in a system coolci>wn and depressurization folla.ting a small LOCA while bringing the reactor to a cold shutda.tn condition.
Cool cbwn The cooldown function perfonned by the Auxiliary Feectwater System is a partial one since the reactor coolant system is reduced fran nonnal zero load temperatures to a hot leg temperature of approximately 3SOOF.
The latter is the maximlJII temperature reccmnended for pl acing the Resi-dual Heat Rsnoval System (RHRS) into service. The RHR system completes the cool dCMn to cold shutdo.m conditions.
Coolct>wn may be reCJJired follCMing expected transients, follCMing an accident such as a main feedline break, or during a nonnal cool down prior to refueling or perfonning *reactor plant maintenance. If the reactor is tripped following extended operation at rated pCMer level, the AFhS 1s capable of delivering sufficient AFW to remove decay heat and reactor cool ant pLmp (RCP) heat foll owing reactor trip while main-taini ng the ste11n generator (SG) water level. FollCMing transients or accidents, the recannended cooldown rate is consistent with expected needs and at the sane time ct>es not impose additional requirements on the capacities of the auxiliary feedwater punps,.considering a sin~le failure.
In any event, the process consists of being able to dissipate plant sensible heat in addition to the decay heat produced by the reac-tor core.
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e Response to 1. b Table 18-1 sunmarizes the criteria which are the general design bases for each event, discussed in the response to Question 1.a, above.
Specific assLJnptions used in the analyses to verify that the design bases are met are discussed in response to Question 2.
The primary function of the Auxiliary Feectwater Sys ten is to provide sufficient h~at renoval capability for heatup accidents following reac-tor trip to renove the decay heat generated by the core and prevent systen overpressurization. Other plant protection systens are designed to meet short tenn or pre-trip fuel fai 1 ure criteria. The effects of excessive cool ant shrinkage are bounded by the analysis of the rupture of a main ste~ pipe transient. The maximlJTI f1CM requirenents deter-mined by other bases are incorporated into this analysis, resulting in no additional fl aw re~irenents.
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Condition or Transient Loss of Main Feedwater Statton Blackout Steamltne Rupture Loss of all A/C Power Loss of Coolant Cooldown
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TABLE 18-1 Criteria for Auxiliary Feedwater System Design Basis Conditions Classiftcat1on":
Condition II Conditt on II Condttton IV N/A Condition Ill Condition IV N/A Crtter1a*
Peak RCS pressure not to exceed design pressure.
No consequential fuel failures (same as LMFW) 10 CFR 100 dose limits Containnent design pressure not exceeded Note 1 10 CFR 100 dose limits 10 CFR 50 PCT limits 10 CFR 100 dose limits 10 CFR 50 PCT limits Additional Design Criteria
- Pressurizer does not become water so 11 d.
Pressurtzer does not become water solid.
Sc111e as blackout assuming turbine driven PllllP lOOoF/hr 5470f to 3500f
- Ref:
ANSI N18.2 (This infonnation provided for those transients perfonned tn the FSAR).
Note 1 Although thts transient establishes the basis for AFW pump* powered by a diverse power source.*this ts not evaluated relative to typical criteria since multiple failures must be assumed to postulate this transient.
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I Question 2 DesO"ibe the analyses and asslll!ptions and corresponding technical justi-fication used with plant condition considered in l.a above including:
- a.
Maxim1.J11 reactor power (including instrunent error allowance} at the time of the i ni ti ating transient or accident.
- b. Time--delay fran initiating event to reactor trip.
- c. Pl ant paraneter (s} which initiates AFWS fl ow and time de1 ay between initiating event and introduction of AFWS flow into stean genera-tor (s}.
- d. Minimllfl stean generator water level when initiating event occurs.
- e. Initial stean generator water inventory and depletion rate before and after AFWS flow convnences -- identify reactor decay heat rate used.
- f. Maxim1.J11 pressure at which stean 1s released fran stean generator(s) and against which the AFW punp must develop sufficient head.
- g. Minimun nunber of steill! generators that must receive AFW flow; e.g.,
1 out of 2?
2 out of 4?
- h.
RC flow a,ndition -- continued operation of RC p.imps or natural circulation.
- i. MaximlJ11 AFW inlet tenperature.
- j. Fo111J11ing a postulated ste1111 or feed line break, time delay ass1.J11ed to isolate break aid direct AFW fl ow to intact stean generator (s).
AFW p.imp fl CM capacity al launce to acC011111odate the time delay and maintain minimlJTI steilll generator water level. Also identify credit taken for primary systen heat renovel due to blowdown.
- k.
Vol1JT1e aid maximlJTI tenperature of water in main feed lines between stean generator(s) and AFWS connection to main feed 1 ine.
- 1. Operating condition of stean generator nonnal bl owdown foll owing ini ti ati ng event.
- m.
Primary and secondary systan water and metal sensible heat used for coo 1 dawn and AFW fl ow sizing.
- n. Time it hot standby and time to cool down RCS to RHR system cut in tanperature to s1 ze AFW water source inventory.
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e Response to 2 Analyses have been performed for the loss of main feedwater and the loss of offsite AC power to the station, the transients which define the AFWS performance requirements. These analyses have been provided for review and have been approved fn the _FS~R.
In addition,to the above analyses, calculations have been performed specifically for the Surry Units 1 and 2 plants to determine the plant
- cooldown flow (storage capacity) requirements. The LOCA analysis, as discussed in response l.b, incorporates the system flows requirements as defined by other transients, and therefore is not performed for the purpose of specifying AFWS flow requirements.
Each of the analyses mentioned above are explained in further detail in the following sections of this response.
Loss of Main Feedwater (Blackout)
A loss of feedwater, assuming a loss Clf Dower to the reactor coolant pumps, was performed in FSAR Section 14.2.llfor the purpose of showing that for a station blackout transient, the peak RCS pressure remains below the criterion for Condition II transients and no fuel failures occur (refer to Table 18-1). Table 2-1 sunmarizes the assumptions used fn this analysis. The transient analysis ass1J11es 10 seconds of steady state conditions prior to the loss of normal feedwater.
The time fr~
the 1 oss of feedwater until the reactor trip occurs is approximately 60 seconds wfth a 2 second delay assumed between the receipt of the low-low steam generator level signal and the beginning of rod motion.
The anal-ysis assumes that the plant is initially operating at 102% (calorimetric error) of the Engineered Safeguards design (ESD) rating shown on.the table, a very conservative assumption in defining decay heat and stored energy in the RCS.
The reactor fs ass1.111ed to be tripped on low-low steam generator level, allowing for level uncertainty. The FSAR shows that there is a considerable margin with respect to filling the pressurizer.
This analysis establishes the minimum auxiliary feedwater flow require-ment following transients which result in a primary system heatup. It also establishes train association of equipment so that this analysis remains valid assuming the most limiting single active failure.
Plant Cooldown Minimum flo:w requirements fran the previously discussed transients meet the flow requirements of plant cooldown. This operation, however, defines the basis for tankage size, based on the required cooldown duration, maximlJTI decay heat fnput and maxfmtm1 stored heat in the system.
As previously discussed in response lA, the auxiliary feed-water system partially cools the system to the point where the RHRS may canplete the*cooldown, f.e., 3SQOF in the RCS. Table 2-1 shows the assumptions used to determine the cooldown heat capacity of the auxil-iary feedwater system.
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I The cool down fs assuned to ccmnence at the maximun rated power I and maximun trip delays and decay heat source tenns are assuned when the reactor fs tripped.
Primary metal I primary water, secondary systen metal and secondary systen water are all included in the stored heat to be removed by the AFWS.
See Table 2-2 for the it ens constituting the sensible heat stored in the NSSS.
This operation fs analyzed to establish minim1111 tank size requirenents for auxiliar:t feedwater fluid source which are nonnally aligned.
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TABLE 2-1 Sllffllary of Assumptions Used in AFWS Design Verification Analyses Loss of Feedwater Transient (station blackout}
Cooldown
- a.
Max reactor power 102% of ESD rating 2488 MWt (102% of 2559 HWt)
- b. Time delay from 2 sec (delay after 2 sec event to Rx trip trip signal is reached to the beginning of rod motion)
- c.
AFWS actuation sig-lo-lo SG level NA nal/time delay for 1 minute AFWS flow
- d.
SG water level at (lo-lo SG level)
NA time of reactor trip 0% NR span r
- e. Initial SG inventory 53,700 lbm/SG (at 109,600 lbm/SG trip)
I 516.lOF Rate of change before
- N/A N/A
& after AFWS actuation decay heat See FSAR Figure 14.5.2.1-1
- f.
AFW pump design 1133 psia 1133 psia pressure
- h.
RC pump status Tripped I reactor trip Tripped
- f.
Maximt111 AFW l200F lOOOF temperature
- j. Operator action none N/A
- k.
MFW purge volume/temp.
142.7 ft3/441.0F 265 ft3 total/
(.
436.40F
- 1. Nonnal blowdown none asst111ed none assumed 6685A
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Sunmary of Assunpti ans Used in AFWS Design Ver if icati on* Analyses Loss of Feectwater Transient lstati on blackout l Cool dJwn
- m.
Sensible
- heat see cool da,m Table 2-2
- n.
Time at standby/time 2 hr/6 hr 2 hr/6 hr to cool dawn to RHR.
- o.
AFW f1 °" rate 500 gpn - constant variable (mi nimLIII requiranent) 6685A J
e TPBLE 2-2 S1.11111ary of Sensible Heat Sources Primary Water Sources (1n1ti ally at rated power tanperature and inventory)
- RCS f1 uid
- Pressurizer fl u1 d (11~1 d and vapor)
Primary Metal; Sources (1niti ally at rated power tenperature)
- Reactor mol ant piping, pumps and reactor vessel
- Pressurizer
- SteilTI generator tube metal and tube sheet
- SteilTI generator metal below tube sheet
- Reactor vessel internals Secondary Water Sources (1niti ally at rated power tanperature and inventory)
- SteilTI generator fluid (l1~1d and vapor)
- Main feectwater purge f1 uid between steilTI generator and AFWS piping.
Secondary Metal Sources (initially at rated power tanperature)
-All ste5n gener*atormetal above tube sheet, excluding tubes
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Question 3 Verify the AFW pumps in your plant will supply the necessary flow to the steam generator(s) as determined by items 1 and 2 above considering a single failure.
Identify the margin in sizing the pump flow to allow for pump recir-culation flow, seal leakage and pump wear.
Response to 3 Based upon the review of the flow requirements of Table 2-1, the station blackout with a loss of all AC power was selected as the most limiting condi-tion.
The parameters used are as follows:
- 1.
Steam Generator Pressure
- 2.
Flow Rate to:
A Steam Generator B Steam Generator C Steam Generator Total 1135 psig 259 gpm 249 gpm 240 gpm 748 gpm The required recirculation flow for the turbine driven pump is 35 gpm.
Seal leakage and pump wear are negligible.
As can be seen on Tab le 2-1, the AFW flow rate required to the steam generators is 500 gpm; therefore, the head margin available is 248 gpm.
This value takes into consideration pressure drops between the pump and the steam generator.
Therefore, the head margin available is adequate for pump recirculation flow, seal leakage and pump wear.