ML17331A514
| ML17331A514 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 11/03/1980 |
| From: | Hunter R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:00300C, AEP:NRC:300C, NUDOCS 8011100394 | |
| Download: ML17331A514 (16) | |
Text
REGULATOR iVFORMATION DISTRIBUTION S f EM (RIDS)
ACCESSION NBR:8011100394 DOCOOATK: 80/11/03 NOTARIZED! NO DOCKET FACIL:50-315 Donald C ~
Cook Nuclear Power Planti Unit 1~ Indiana L
0500 1
50-316 Donald C ~, Cook Nuclear Power Plant~
Unit 2~ Indiana 8,
5000316 AUTH BYNAME AUTHOR AFFILIATION HUNTERER ~ S ~
RKC IP ~ NAME REC IP IKiVT AFF Il.IATION DENTONPH ~ RE Office of Nuclear Reactor Regulationr Director pt
SUBJECT:
Forwards responses to questions contained in Encl 2 to OG, Ei senhut 791030 Itr re flow design basis of auxiliary feedwater
- sys, DISTRIBUTION CODE)
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INDIANA 5 MICHIGAN ELECTRIC=CQMPAH.Yaw P. O. BOX 18 BOWLING GREEN STATION NEW YORK, N. Y. 10004 Donald C.
Cook Nuclear Plant Unit Nos.
1 and 2
Docket Nos. 50-315 and 50-316 License Nos.
OPR-58 and DPR-74 Auxiliary Feedwater System Flow Design Basis
<<UO lwQr 7 r>
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November 3, 1980
- AEP: NRC: 00300C.
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regu)atory Commission Washington, O.
C.
20555
Dear Mr. Oenton:
The attachment to this letter contains our responses to the questions contained in Enclosure 2 to Mr. Eisenhut's October 30, 1979 letter concerning the flow design basis of the Auxiliary Feedwater System.
Very truly yours, R.
S.
unter Vice President cc:
R.
C. Callen G. Charnoff John E. Dolan R.
S. Hunter R.
W. Jurgensen O. V. Shaller - Bridgman
~ NRC Region III Resident Inspector at Cook Plant - Bridgman goo)
ATTACHMENT TO AEP:NRC:03QOC Responses are for D. C.
Cook Units 1
8 2, except as noted, using the same format as enclosure 2 to Mr. Eisenhut's October 30; 1979 letter.
Res onse to uestion 1:
The Auxiliary Feedwater System (AFS) serves as an emergency backup system for supplying feedwater to the secondary side of the steam generators at times when the main feedwater system is not available, thereby maintaining an adequate heat sink.
As an Engineered Safeguards
- System, the AFS is one of the mitigating systems which would prevent possible core damage and system over-pressurization 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 and decay heat removal 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 system or to the atmosphere through the steam generator safety valves or the power-operated relief valves.
Steam generator water inventory is maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal process.
The water level is main-tained under these circumstances by the AFS which delivers an alternate water supply to the steam generators after the main feedwater pumps are tripped.
The AFS is capable of functioning for extended
- periods, thereby allowing time either to restore normal feedwater flow or to proceed with an orderly primary system cooldown to conditi'ons where the Residual Heat Removal System (RHR) can be employed for decay heat removal.
The AFS flow and water supply capacity are sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown.
l.a The plant transient and accident conditions which impose safety-related performance requirements on the design of the AFS including AFS flow requirements are as follows:
A.
Loss of Main Feedwater Transient i} Loss of main feedwater with offsite power available ii Station blackout (i.e., loss of main feedwater without off-site power available)
B.
Secondary System Ruptures i}
Main Steamline rupture ii} Main Feedline rupture (Unit 2 FSAR analysis only)
C.
Loss of all AC Power 0.
Loss of Coolant Accident (LOCA)
E.
Cooldown
A.
Loss of Main Feedwater Transients The design Loss of Main Feedwater transients are those caused by:
-'nterruptions of the Main Feedwater System flow due to a malfunction in the main feedwater or condensate system.
-- 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 turbine trip, feedwater isolation, and auxiliary-feedwater actuation by the Reactor Protection System logic.
Following reactor trip from full power, the power quickly falls to decay heat levels.
The steam generator water levels continue
'o decrease, progressively 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 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.
8'ithout the addition of sufficient auxiliary feedwater, continued unarrested temperature rise could potentially cause challenges to the RCS overpressurization protection system (relief/safety valves).
- Hence, the timely introduction of sufficient auxiliary "feedwater mitigates the decrease in the steam generator water
- levels, reverses the rise in reactor coolant temperature, prevents the pressurizer from filling to a water solid condition, and establishes sta5le hot standby conditions.
The loss of offsite AC power transient differs from a simple loss of main feedwater in that emergency power sources are relied upon to operate vital equipment.
The loss of power to the electric driven condenser circulating water pumps results in a loss of con-denser vacuum and condenser dump valves.
Hence, for this case steam formed by the 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 hlackout, except that reactor coolant pump heat input is not a consideration in the blackout transient following loss of power to the reactor coolant pump bus.
B.
Secondar S stem Ru tures Main steamline rupture accident conditions are characterized initially by plant cooldown and, for breaks inside containment, by increasing containment pressure and temperature.
Auxiliary feedwater is not needed during the early phase of the transient and flow to the faulted loop will contribute to an excessive release of mass and energy to containment.
Thus, steamline rupture conditions establish the upper limit on auxiliary feed-water 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 loops, but at somewhat lower rates than for the loss of feedwater transients described previously.
Provisions are made in the design of the AFS to limit, control, and/or terminate the auxiliary feedwater flow to the faulted loop as necessary in order to limit containment pressure rise following a steamline break inside containment and to ensure the flow to the remaining unfaulted loops is sufficient to terminate the RCS heatup.
The feedwater line rupture accident not only results in the loss of feedwater flow to the steam generators but also results in the complete blowdown of one steam generator within a short time if the rupture should occur downstream of the last nonreturn valve in the main or auxiliary feedwater piping 'to an individual steam generator.
Another significant result of a feedline rupture may be the spilling of auxiliary feedwater out the break as a
consequence of the fact that the auxiliary feedwater branch line is connected to the main feedwater line.
Such situations can result in the loss of a fraction of the total auxiliary feedwater flow because the system preferentially pumps water to the lowest pressure region in the faulted loop.
The AFS design allows for terminating, limiting, and minimizing that fraction of auxiliary feedwater flow. which is delivered to a faulted loop or spilled through a break in order to ensure that sufficient flow will be delivered to the remaining effective steam generator(s).
The design flow requirements are similar for the main.feedwater line rupture as those explained for the loss of main feedwater transients.
C.
L'oss of All AC Power The loss of all AC power is postulated as resulting from accident conditions wherein not only offsite AC power is lost but also onsite AC emergency power is lost.
DC power supplied by the redundant onsite safety related station batteries for operation of protection circuits is assumed available.
For the Cook Plant, the Turbine Driven Auxiliary Feedwater Pump Train operates inde-pendent of onsite and offsite AC power.
This AFS train is DC powered by its own dedicated safety related battery.
The impact of a total loss of all,AC power on the AFS is accounted for by
providing both an auxiliary feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at hot shutdown until AC power is restored.
D.
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 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.
E.
Cooldown The. cooldown function performed by the AFS is a partial one since the reactor coolant system is reduced from normal no load temperatures to a hot leg temperature of approximately 350oF.
The latter is the maximum temperature recommended for placing the RHR system into service.
The RHR system completes the cooldown to cold shutdown conditions.
Cooldown may be required following expected transients, following an 'accident, or to accomplish a normal cooldown prior to refueling or performing reactor plant maintenance.
If the reactor is tripped following extended operation at rated power level, the AFS is capable of delivering sufficient AFN to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while main-taining adequate steam generator (SG) water inventory.
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 AFS, 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.
Table 1E-1 summarizes the criteria which are the general design bases for each event, discussed in the response to guestion l.a, above.
Specific assumptions used in the analyses to verify that the design bases are met are discussed in response to guestion 2.
The primary function of the AFS is to provide sufficient heat removal capability for heatup accidents following reactor 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 fuel failure cri'teria.
The effects of excessive coolant shrinkage are evaluated by the analysis of the rupture of a main steam pipe transient.
The maximum flow requirements determined by other bases are incorporated into this analysis, resulting in no additional flow requirements for the AFS.
TABLE lB-1 Criteria for Auxiliary Feedwater System Design Basis Conditions Condition or Transient Loss of Main Feedwater (LMFW)
Classification*
Condition II Criter ia*
Peak RCS pressure not to exceed design pressure.
No consequential fuel failures.
Additional Design Criteria LMFW with loss of offsite Condition II (same as LMFW).
Pressurizer does not fillwater solid with a single motor driven auxiliary feed pump feeding 2 SGs.
Steamline Rupture Feedline Rupture (Unit-2 FSAR Analysis)
Loss of all AC Power Loss of Coolant Condition IV Condition IV Condition III 10 CFR 100 dose limits.
Containment design pressure not exceeded.
RCS design pressure not exceeded.
10 CFR 100 dose limits.
Note 1
'IO CFR 100 dose limits.
Same as blackout assuming turbine driven pump operates Cooldown Condition IV 10 CFR 100 dose limits.
100oF/hr 547oF to RHR
- 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.
With the turbine driven pump train operab'le from its'wn DC power source, this transient is bounded by the LMFW.
Res onse to uestion 2:
Analyses have been performed for the limiting transients which define the AFWS performance requirements.
These analyses have been provided for review and have been approved in the Donald C.
Cook Nuclear Plant FSAR.
Specifically, they include:
- Loss of Main Feedwater/Station Blackout (similar for flow design basis)
- Rupture of a Main Feedwater Line (Unit 2 FSAR analysis)
- Rupture of a Main Steam Line Inside Containment The Loss of All AC Power is discussed via a comparison to the transient results of a Station Blackout, assuming an available auxiliary feedwater pump having an independent DC power supply.
The LOCA analysis, as discussed in the Response 1.b above, incorporates the system flow requirements as defined by other transients, and therefore is not per'formed 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.
In addition to the above analyses, calculations have been performed specifically to determine the require-ments for plant cooldown flow for the purpose of determining the storage capacity of the condensate storage tank.
Loss of Main Feedwater/Station Blackout similar for flow desi n basis 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 two steam generators 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 18-1).
Table 2. 1 and Table 2.2 summarize the assumptions used in this analysis for Cook Unit 1 and Unit 2, respectively.
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 at most 20-30 seconds.
The
'nalysis assumes that the plant is initially operating at 102Ã (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.
In Unit I, the reactor is assumed to be tripped on steam flow/
feed flow mismatch coincident with low steam generator water level, allowing for level'ncertainty.
In Unit 2, the reactor is assumed to be tripped on low-low steam generator level, allowing for level uncertainty.
The FSAR
- 7'-
ana1ysis shows that there is a considerable 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 with respect to filling the pressurizer.
Ru ture of Main Feedwater Pi e Unit 2 FSAR Anal sis The double ended rupture of a main feedwat'er pipe downstream of the main feedwater line check valve is analyzed for Cook Unit 2.
Table 2.2 includes the assumptions used in this analysis.
The assumption is made that reactor trip occurs when the steam generators are at the low level
- setpoint, adjusted for errors, coincident with steam flow/feed flow mis-match and the faulted loop is assumed to be empty.
This conservative assumption maximizes the'primary system heat input prior to reactor trip and minimizes the ability of the steam generator to remove heat from the RCS following reactor trip due to a conservatively small total steam generator inventory.
The initial power rating was assumed to be 102% of rated power.
The auxiliary feedwater is injected into the unfaulted steam generators one (I) minute after the reactor trip setpoint is reached.
The qriteria listed in Table 1B-1 are met.
This analysis establishes the capacity of single
- pumps, requirements for layout to preclude indefinite loss of auxiliary feedwater to the postulated
- break, and train association requirements for equipment so that the AFHS can deliver the minimum flow required assuming the worst single failure.
This applies to both Units I and 2 which have identical AFS.
Ru ture of a Main Steam Pi e Inside Containment Because the steam line break transient is initially a cooldown, the AFS is not needed to remove heat in the short term.
Furthermore, addition of excessive auxiliary feedwater to the faulted steam generator could affect the peak containment pressure following a steam line break inside containment.
This transient in Unit 2 is performed at four power levels for several break sizes.
Auxiliary feedwater is assumed to be initiated at the time of the break.
Maximum flow is used for this analysis, considering failure of the runout protection for the largest pump.
Tables 2-1 and 2-2 include the assumptions used in this analysis.
At 10 minutes after the break, it is assumed that the operator has isolated the AFS from the faulted steam generator which subsequently blows down to ambient pressure.
The criteria stated in Table 18-1 are met.
This transient establishes the maximum allowable auxiliary feedwater flow rate to a single faulted steam generator assuming all pumps operating, the basis for runout protection, and the layout requirements so that the flow requirements may be met considering the worst single failure.
Plant Cooldown Maximum and 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 AFS partially cools the system to the point where the RHR system may complete the cooldown.
Tables 2-1 and 2-2 show the assumptions used to determine the cooldown heat capacity of the auxiliary feedwater system.
These assumptions were used in the cooldown calculations for D.
C.
Cook Unit 1.
They are also applicable to Unit 2 since the two units have an identical AFS and condensate storage tanks.
The cooldown is assumed to commence at rated
- power, and maximum trip delays and decay heat source terms are assumed when the reactor is tripped during the cooldown process.
Primary metal, primary water, secondary system metal and secondary system water are all included in the stored heat to be removed by the AFS.
See Table 2-3 for the items constituting the sensible heat stored in the NSSS.
TABLE 2-1 Sugary of Assumptions Used in AFWS Design Verification Analyses For D.
C.
Cook Unit 1 Transient Loss of Feedwater Station Blackout)
Cooldown Main Steamline Break Containment a;
Max Reactor Power 102K of ESD Rating 3250 MWt (102% of 3391 MWt) 102% of ESD Rating (102K of 3391 MWt)
- b. Time Delay From Reactor Trip Signal To Rod Motion 2 Seconds 2 Seconds c.
AFWS Actuation Signal/
Low-Low SG Level/
NA Time Delay for AFWS Flow 1 Minute AFWS Initiated At 10 Seconds.
d.
SG Water Level At Time of Reactor Trip 05 NR Span NA (Low SG Level
+
Steam-feed Mismatch)
NA
- e. Initial SG Inventory 63,000 lbIH/SG (At Trip) 87,935 1 bm/SG 117,500 1 bm/SG 9 512.1 F
Rate of Change Before 8 After Actuation Decay Heat
- f. AFW Pump Design
/
h.
RC Pump Status
- i. Maximum AFW Temperature
- j. Operator Action k.
MFW Purge Volume/
Temperature See FSAR Figure 14.1.9-1 ANS + 20K 1187 psia 2of4 (4 of 4 Receive Mater)
Tripped 9 Reactor Trip 100 F
None 200 ft per loop/
3 431oF (bounding volume}
NA 1187 psia NA Tripped 80 F
NA None Assumed NA ANS + 20K NA All Operating 100 F
10 Min.
1250 ft per loop/
435oF (boundtng volume}
10-TABLE 2-1 (CONTINUED) j>
Transient 1.
Normal Blowdown
- m. Sensible Heat
- n. Time At Standby/
Time To Cooldown To RHR Loss of Feedwater Station Blackout None Assumed See Cooldown 2 hr/4 hr Cool down None Assumed Table 2-3 2 hr/4 hr Main Steamline Break Containment None Assumed
TABLE 2-2 Summary of Assumptions Used in AFWS Design Verification Analyses for D. C.
Cook Unit 2 Transient Loss of Feedwater Station Blackout Cooldown*
Hain Feedline Break Main Steamline Break Containment a.
Maximum Reactor Power
- b. Time Delay from Rx Trip Signal to Rod Motion c.
AFWS Actuation Signal/Time Delay for AFWS Flow d.
SG Mater Level at Time of Reactor Trip
NA 1 Minute OX NR Span NA (Low-Low SG Level) 56,442 ibm/SG (At Trip) 87,925 ibm/SG 9 512.1oF 102% of ESD Rating 3250 HMt (102K of 3556.1 HWt) 102K of Rated Power (102K of 3403 HWt) 2 Seconds (20.1 Sec.
From Event to Rx Trip)
Low-Low SG Level/.
1 Minute 3 9 20K NR Span 1 9 Tube Sheet (Low SG Level +
Steamfeed Mismatch) 97,914 1 b}11/SG 0, 30, 70, 1025 of Rated (I of 3403 HMt).
2 Seconds (Time from Event to Rx Trip is Variable)
Assumed Immediately 0 Sec.
(No Delay)
Depending on Power Level
- . Rate of Change Before 8
After AFWS -Actuation Decay Heat
- f. AFW Pump Design
- g. Minimum No. of SGs Which Must Receive AFW Flow See FSAR Figure 14.1.9-1 See FSAR Figure 14.1-6 ANS + 2GX 1187 psia 2of4 1187 psia Turnaround 1700 Sec.
NA See FSAR Figure 14.1-6 See FSAR Figure 14.1-6 1187 psia 2of4 NA (3 of.4 Receive Water) ik h.
RC Pump Status
- i. Haximum AFW Temperature
- g. Operator Action 80 F
None Tripped 9 Reactor Tripped Tpip 120 F
Tripped 9 Reactor Trip 120 F
None All Operating
.Equal to Hain Feedwater Temperature 10 Min.
TABLE 2-2 (CONTINUED)
Transient k.
MFM Purge Volume/Temperature Loss of Feedwater Station Blackout 79.4 ft per Loop/
431oF Cooldown*
None Assumed Main Feedline Break 78.6 ft per Loop/
431oF Main Steamline Break Containment 800 ft /Loop (For Dryout Time)
Bounding Volume)
. Normal Blowdown
- m. Sensible Heat
- n. Time At Standby/Time To Cooldown to RHR 0
None Assumed See Cooldown 2 hr/4 hr Table 2-3 See Cooldown 2 hr/4 hr 2 hr/4 hr.
None Assumed None Assumed None Assumed
~ Cooldown analyzed for Unit 1 and repeated here for information.