ML17261A209
| ML17261A209 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/14/1980 |
| From: | White L ROCHESTER GAS & ELECTRIC CORP. |
| To: | Crutchfield D Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML17250A418 | List: |
| References | |
| TASK-10, TASK-RR NUDOCS 8007220403 | |
| Download: ML17261A209 (30) | |
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ACCESSION NBR ~ 8007220403 DOC,DATE ~ 80/07/14 NOTARIZED:
NO 0
KET FACIL:50<<244 Robert Emmet -Ginna Nuclear Plant~
Unit 1~ Rochester G
05000244 AUTH INANE AUTHOR AFFILIATION "WHITEgL~ D, Rochester Gas L Electric Corp.
RECIP ~ NAME RECIPIENT AFFILIATION CRUTCHFIELDgD ~
Operating Reactors Branch 5
I
SUBJECT:
Forwards r esponse to NRC 791022 1 tr requesting addi info re auxiliary feedwater sys flow requirements. Response to request for info re evaluation of auxiliary feedwater sys requirements was provided 791128
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~O UNITEO STATES NUCLEAR REGULATORY COMMiSSION WASHINGTON, D. C. 20555 iMEMORANDUM FOR:
TERA CORPORATION FROM'S NRC/TIDC/DOCUMENT MANAGEMENT BRANCH SOBQEC'E:
Special Document Handling Requirements C31.,
Please use the following special distribution list for the attached document.
2.
The attached document requires the following special considerations:
Cl Do not send oversize enclosures to the NRC PDR.
C.ig(re>
Eeahy~e oversize enclosurefwas received please return for Regulatory File storage.
0 Proprietary information send only non-proprietary portions to the NRC PDR.
Ic7 Other: (specify) cc:
DMB Files TIDC DMB A t o zed Signature
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3K ROCHESTER GAS AND ELECTRIC CORPORATION
~ 89 EAST AVENUE, ROCHESTER, N.Y. 14649 TELEPHONE AREA CODE TIE 546-2700 July 14, 1980 Director of Nuclear Reactor Regulation Attn:
Mr. Dennis M. Crutchfield, Chief Operating Reactors Branch 55 U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
NRC Requirements for Auxiliary Feedwater Systems R. E..Ginna Nuclear Power Plant Docket No. 50-244
Dear Mr. Crutchfield:
This letter is in response to a letter from Darrell Eisenhut dated October 22, 1979.
The letter requested that we evaluate our auxiliary feedwater systems against the applicable requirements contained in Enclosure 1.to that, letter.
The letter also requested that we respond to a generic request for additional information regarding auxiliary feedwater system flow requirements contained in Enclosure 2 to that letter.
We provided our response to the requirements of Enclosure 1
on November 28, 1979.
The information requested in Enclosure 2
is provided in the attachment to this letter.
Sincerely yours, L. D. Whi
, Jr.
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guestion 1
a.
Identify the plant transient and accident conditions considered in establishing AFWS flow requirements, including the following events:
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.
coo1down 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 corresponding technical bases used for each initiating event identified above.
The acceptance criteria should address plant limits such as:
1)
Maximum RCS pressure (PORV or safety valve actuation) 2)
Fuel temperature or damage limits (DNB, PCT, maximum fuel central temperature) 3)
RCS cooling rate limit to avoid excessive coolant shrinkage 4)
Minimum steam generator level to assure sufficient steam gen-erator heat transfer surface to remove decay heat and/or cool down the primary system.
Res onse 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 overpressur ization in the event of transients 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 steam either to the condensers through the steam dump or to the atmosphere through the steam 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 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 functioning for extended
- periods, allowing time either to restore normal feedwater flow or to proceed with an order ly cooldown of the plant to 4916A
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the reactor coolant temperature where the Residual Heat Removal System can assume the burden of decay heat removal.
The Auxiliary Feedwater System flow and the emergency water supply capacity must be sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown..
The Auxiliary Feedwater System can also be used to maintain the steam generator water levels above the tubes fol-lowing a LOCA.
In the latter function, the water head in the steam generators serves as a barr ier 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 Robert E. Ginna plant.
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)
Rupture of a Main Steam Line Loss of all AG 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 H
Loss of offsite power to the station with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a rapid reduction in steam generator water levels which results in a reactor trip, a tur-bine trip, and auxiliary feedwater actuation by the protection system logic.
Following reactor trip from high power, the power quickly falls to decay heat levels.
The water levels continue to decrease, progres-sively uncovering the steam generator tubes as decay heat is transferred and discharged in the form of steam either through the steam dump valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere.
The reactor coolant temperature 4916A
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increases as the residual heat in excess of that dissipated through the steam generators is absorbed.
With increased temperature, the volume of reactor coolant expands and begins filling the pressurizer.
Without the addition of sufficient auxiliary feedwater, further expansion will result in water being discharged through the pressurizer safety and relief valves.
If the temperature rise and the resulting volumetric expansion of the primary coolant are permitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing over-pressurization of the Reactor Coolant System and/or (2) the continuing loss of fluid from the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in 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-pressure in the accumulator
- 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 loss of offsite A/C power transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equipment.
The loss of power to the electric driven con-denser 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 calculated transient is similar for both the loss of main feedwater and the blackout, except that reactor coolant pump heat input is not a consideration in this transient due to loss of power to the reactor coolant pump bus.
The loss of offsite A/C power to the station and auxiliaries serves as the basis for the minimum flow required for the smallest capacity single auxiliary feedwater pump for the Robert E. Ginna Plant.
The pump is sized so that any single pump will provide sufficient flow against the steam generator safety valve set pressure (with 3X 'accumulation) to pre-vent water relief from the pressurizer.
The same criterion is met for the loss of feedwater transient by the operation of any two pumps, where.
A/C power is assumed available.
Rupture of a Main Steam Line 8ecause 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.
The main steamline rupture accident conditions are char-acterized initially by plant cooldown, and hence, auxiliary feedwater flow is not needed during the early stage of the transient to remove decay heat from the Reactor Coolant System.
Provisions must be made in the design of the auxiliary feedwater system to allow termination of flow to the faulted loop and to provide flow to the intact steam genera-tors during the controlled cooldown following the steamline break accident.
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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 power is lost but also AC emergency power is lost as an assumed common mode failure.
Although this accident scenario is not a design basis for the Robert E. Ginna plant, several features are incorporated into the design to mitigate the effects of such a sequence.
In the event of complete loss of AC power, DC power operated valves are automatically opened to admit steam from two steam generators to power a steam turbine driven auxiliary feedwater pump, which delivers flow to both steam generator s through a common header.
All normally closed valves in the auxiliary feedwater supply lines are fail-open air operated valves.
Loss-of-Coolant Accident (LOCA)
The loss of coolant accidents do not impose on the auxiliary feedwater system any flow requirements in addition to those required by the other accidents addressed in this response.
The following description of the small LOCA is provided here for the sake of completeness to explain the role of the auxiliary feedwater 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 contribution from the Auxiliary Feedwater System following such small LOCAs is basically the same 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 steam generators provides a heat sink for removing decay heat and establishes the capability'or providing a buoyancy head for natural circulation.
The auxiliary feedwater system may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the reactor to a cold shutdown condition.
Cooldown The cooldown function performed by the Auxiliary Feedwater System is a
partial one since the reactor coolant system is reduced from normal zero load temperatures to a hot leg temperature of approximately
- 350oF, The latter is the maximum temperature recommended for placing the Resi-dual Heat Removal System (RHRS) into service.
The RHR system completes the cooldown to cold shutdown conditions.
Cooldown may be required following expected transients, following an accident or prior to refueling or plant maintenance.
If the reactor is tripped following extended operation at rated power level, the AFWS is capable of delivering sufficient AFW to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while maintaining the steam generator (SG) water level.
Following transients or accidents, the recommended cooldown rate is consistent with expected needs and at the same time does not impose additional requirements on the capacities of the auxiliary feedwater
- pumps, considering a single failure.
In any
- event, the process consists of being able to dissipate plant sensible heat in addition to the decay heat produced by the reactor core.
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Response
to 1.b Table 1B-1 summarizes the criteria which are the general design bases for each event discussed in the response to question l.a above.
Specific assumptions 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 Feedwater System is to provide sufficient heat removal capability for heatup accidents following reac-tor trip to remove the decay heat generated by the core and prevent system overpressurization.
Other plant protection systems are designed to meet short term or pre-trip fuel failure criteria.
The effects of excessive coolant shrinkage are evaluated by the analysis of the rupture of a main steam pipe transient.
The maximum flow requirements deter-mined by other bases are incorporated into this analysis, resulting in no additional flow requirements.
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TABLE 18-1 Criteria for Auxiliary Feedwater System Design Basis Conditions Condition or Transient Loss of Hain Feedwater Classification*
Condition II Criter ia+
Peak RCS pressure not to exceed design pressure.
No consequential fuel failures Additional Design Criteria LHFW with loss of offsite Condition II A/C power (same as LHFW)
Pressurizer does not fillwith 1 single motor driven aux.
feed pump feeding 1 SG.
Loss of all A/C Power Loss of Coolant Cooldown Condition III Condition IV N/A Note 1
10 CFR 100 dose limits 10 CFR 50 PCT limits 10 CFR 100 dose limits 10 CFR 50 PCT limits 100oF/hr 547oF to 350oF
- Ref:
ANSI N18.2 (This information provided for those transients performed in the FSAR).
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 failures must be assumed to postulate this transient.
If the turbine pump is operable, this transient is bounded by the LHFW.
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question 2
Describe the analyses and assumptions and corresponding technical justi-fication used with plant condition considered in l.a above including:
ae b.
c.
d.
e.
g.
Maximum reactor, power, (including instrument error allowance) at the time of the initiating transient or accident.
Time delay from initiating event to reactor trip.
Plant parameter(s) which initiates AFWS flow and time delay between initiating event and introduction of AFWS flow into steam genera-tor(s).
Minimum steam generator water level when initiating event occurs.
Initial steam generator water inventory and depletion rate before and after AFWS flow commences identify reactor decay heat rate used.
Maximum pressure at which steam is released from steam generator(s) and against which the AFW pump must develop sufficient head.
Minimum number of steam generators that must receive AFW flow; e.g.,
1 out of 2?
2 out of 4?
h.
RC flow condition continued operation of RC pumps or natural circulation.
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Maximum AFW inlet temperature.
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 minimum steam generator water level.
Also identify credit taken for primary system heat removel due to blowdown.
Volume and maximum temperature of water in main feed lines between steam generator(s) and AFWS connection to main feed line.
Operating condition of steam generator normal blowdown following initiating event.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
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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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 in the Ginna FSAR.
In addition to the above'nalyses, calculations have been performed specifically for the Robert E. Ginna plant to determine the plant cool-down flow (storage capacity) requirements.
The LOCA analysis, as dis-cussed in response 1.b, incorporates the system flow requirements as defined by other transients, and therefore is not performed for the purpose of specifying AFWS flow requirements.
Each of the analyses listed 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 of power to the reactor coolant
- pumps, was performed in FSAR Section 14.1.9 for the purpose of showing that for a station blackout transient, a single motor driven auxiliary feedwater pump delivering flow to one steam generator does not result in filling the pressurizer.
Furthermore, the peak RCS pressure remains below the criterion for Condition II transients and no fuel failures occur (refer to Table 1B-1).
Table 2-1 summarizes the assumptions used in this analysis.
The transient analysis begins at the time of reactor trip.
This can be done because the trip occurs on a steam generator level signal, hence the core power, temperatures and steam generator level at time of reactor trip do not depend on the event sequence prior to trip.
Although the time from the loss of feedwater until the reactor trip occurs cannot be determined from this analysis, this delay is expected to be 20-30 seconds.
The analysis assumes that the plant is initially operating at 102K (calorimetric error) of the Engineered Safe-guards design (ESD) rating shown on the table, a very conservative assumption in defining decay heat and stored energy in the RCS.
The reactor is assumed to be tripped on low-low steam generator level, allowing for level uncertainty.
The initiating logic for the AFWS is shown on drawing 882D612, sheet 8 which was provided to you with our letter dated November 22, 1978.
The FSAR shows that there is a
margin with respect to filling the pressurizer.
A loss of normal feedwater transient with the assumption that the two smallest auxiliary feedwater pumps and reactor coolant pumps are running results in even more margin.
This analysis establishes the capacity of the smallest single pump and
'also establishes train association of equipment so that this analysis remains valid assuming the most limiting single failure.
Plant Cooldown Minimum flow requirements from 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, maximum decay heat input and maximum 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 complete the cooldown, i.e.,
350oF 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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The cooldown is assumed to commence at the maximum rated power, and maximum trip delays and decay heat source terms are as'sumed when the reactor is tripped.
Primary metal, primary water, secondary system metal and secondary system water are all included in the stored heat to be removed by the AFWS.
See Table 2-2 for the items constituting the sensible heat stored in the NSSS.
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This operation is analyzed to establish minimum tank size requirements for auxiliary feedwater fluid source which are normally aligned.
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TABLE 2-1 Summary of Assumptions Used in AFWS Design Verification Analyses Transient a..
Max reactor power b.
Time delay from event to Rx trip c.
AFWS actuation sig-nal/time delay for AFWS flow d.
SG water level at time of reactor trip e.
-Initial SG inventory Rate of change bef ore 8 after AFWS actuation decay heat f.
AFW pump design g.
Minimum ¹ of SGs which must receive AFW flow Loss of Feedwater station blackout 102K of ESD rating (102K of 1520 MWt) 2 sec lo-lo SG level 1 minute (lo-lo SG level)
OX NR.span 44,000 ibm/SG (at trip)
See FSAR Figure 14.1.9-1 ANS + 20K 1133 psia lof 2 Cooldown 1326 MWt 2 sec 85,575 ibm/SG 8 518oF N/A ANS +
20%%d 1133 psia N/A h.
RC pump status i.
Maximum AFW temperature Tripped 8 reactor trip Tripped 80oF Operator action k.
MFW purge temp.
none 427.3oF m.
Sensible heat see cooldown n.
Time at standby/time 2 hr/4 hr to cooldown to RHR 1.
Normal bl owdown none assumed N/A 64 ft3/
427.7oF none assumed Table 2-2 2 hr/4 hr o.
AFW flow rate 200 GPM - constant (min. requirement) variable 4916A
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TABLE 2-2 Summary of Sensible Heat Sources Primary Water Sources (initially at rated power temperature and inventory)
- RCS fluid
- Pressurizer fluid (liquid and,vapor)
Primary Metal Sources (initially at rated 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 rated power temperature and inventory)
- Steam generator fluid (liquid and vapor)
- Main feedwater purge fluid between steam generator and AFWS piping.
Secondary Metal Sources (initially at rated power temperature)
- All steam generator metal above tube sheet, excluding tubes.
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Question
.3 Verify that the AFW pumps in your plant will supply the necessary flow to the steam generator(s) as determined by items 1
and 2 above consid-ering a single failure.
Identify the margin in sizing the pump flow to allow for pump recirculation flow, seal leakage and pump wear.
Res onse to 3
Drawing 33013-544, Revision 5 schematically shows the major features and components of the Auxiliary Feedwater System for the Robert E.
Ginna Plant.
Three prints and one microfilm card of this drawing are being provided.
Flow rates for all of the design transients described in Response 2 have been met by the system for the worst single failure.
The. flows for those single failures considered are tabulated for the various transients in Table 3-1, including the following:
A.
A/C Train Failure B.
Turbine Driven Pump Failure C.
11otor Driven Pump Failure D.
AFWS check valve failure (failure to close on reverse flow).
The capacities of the AFW pumps and flow characteristics of the AFW piping have been reviewed to assure that the design bases can be met.
The shutoff head of the pumps is above the safety valve setpoint of the steam generator so that recirculation flow is not required.
Periodic pump testing assures that any seal leakage or pump wear is within acceptable limits.
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TABLE" 3-1 Auxiliary Feedwater Flow (1,3) to Steam Generators Following an Accident/Transient with Selected Single Failure -
GPH Single Failure Accident/Transient Elec. Train Failure TD Pump Failure MD Pump FAilure CV(2)
Failure 1.
Loss of Main FW 2.
LMFW w/loss of offsite A/C power 3.
Coo 1 down
>600
>600
>600
>400
>400
>400
>600
>600
>600
>800
>800
>800 Notes:
(1) Items 1 thru 3 are minimum expected flows.
(2) Including only those CVs in the AFWS.
Failure" is interpreted as failure to close on reverse flow; failure of the CV to open to permit flow >n the'normal direction is not considered.
(3) The flows given in this table assume that three auxiliary feedwater pumps are operable.
If only two pumps are operable in accordance with Technical Specification 3.4.2.b columns A,
C and D may be reduced by 400
- gpm, column B
by 200 gpm.
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