ML20063M589
| ML20063M589 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 09/10/1982 |
| From: | Devincentis J PUBLIC SERVICE CO. OF NEW HAMPSHIRE, YANKEE ATOMIC ELECTRIC CO. |
| To: | Karrigan J Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.1, TASK-TM SBN-324, NUDOCS 8209150258 | |
| Download: ML20063M589 (13) | |
Text
1 SEABRN STATION LJ PUBLIC SERVICE Engineedng Office:
Companyof New Hampshw e 1671 Worcester Road Framinoham, Mossochusetts 01701 (617) - 872-8100 September 10, 1982 SBN-324 T.F.
B 7.1.2 United States Nuclear Regulatory Commission Wa shi ng ton, D. C.
20555 Attention:
Ms. Janis B. Karrigan, Acting Chief Licensing Branch No. 3 Division of Licensing Re f e rence s :
(a) Construction Permit CPPR-135 and CPPR-136, Docke t Nos. 50-443 and 50-444 (b) USNRC Letter, dated February 12, 1982, " Request for Additional Information", F. J. Miraglia to W. C. Tallman (c)
PSNH Letter, dated March 12, 1982, " Response to 410 Series RAIs; ( Auxiliary Systems Branch)"
(d)
PSNH Letter, dated February 12,1982, " Implementation of TMI Action Plan Requirements of NUREG-0737", J. DeVincentis to F. J. Miraglia
Subject:
Emergency Feedwater System Flowrate Design Bases
Dear Ms. Karrigan:
In Reference (d), we committed to a re evaluation of the Emergency Feedwater System (EFW) flowrate design ba es pursuant to the requirements of NUREG-0737, Item 11 E.1.1.
In response to your Request for Aiditional Information [RAI 410.45(4)],
which you forwarded in Ref erence (b), we reiterated the above commitment in Re ference (c).
In fulfillment of this commitment, we have enclosed our re-evaluation of the EFW System (note that this is responsive to questions posed in the March 10, 1980, letter f rom D.
F. Ross (NRC) to all pending W and C-E License Ap plican t s).
Ve ry truly yours, YANKEE ATOMIC ELECTRIC COMPANY
\\
J foO J. DeVincentis Project Manager 1
ALL/dd 8209150258 820910 PDR ADOCK 05000443 A
Qu:stion 1 Identify the plant transient and accident conditions considered in a.
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 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 corresponding technical bases used for each initiating event identified above.
The acceptance criteria should address plant limits such as:
- 3) RCS cooling rate limit to avoid excessive coolant shrinkage
- 4) Minimum steam generator level-to assure sufficient steam gener-ator heat transfer surface to remove decay heat and/or cool down the primary system.
Response to 1.a The Emergency 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 Emergency Feedwater System is directly relied upon to prevent core damage and system overpressurization 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.
l 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 l
process. The water level is maintained under these circumstances by the Emergency Feedwater System which delivers an emergency water supply to the steam generators. The Emergency Feedwater System must be capable of l
functioning for extended periods, allowing time either to restore normal I
feedwater flow or to proceed with an orderly cooldown of the plant to the reactor coolant temperature where the Residual Heat Removal System can assume the burden of decay heat removal. The Emergency Feedwater System flow and the emergency water supply capacity must be sufficient I
to remove core decay heat, reactor coolant pump heat, and sensible heat 2288Q:1 i
l l
during the plant cooldown.
Tha Emargsney Fasdwater Systes 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 barrier to prevent leaka' e of fission products g
from the Reactor Coolant System into the secondary plant.
DESIGN CONDITIONS The reactor plant conditions which impose safety-re2ated performance requirements on the design of the Emergency Feedwater System are as follows for the Seabrook Units No. 1 and 2.
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 i
A turbine trip, MSIV closure or any other condition not mentioned above does not require operation of the Emergency Feedwater System.
There-fore, those transients impose no safety related design requirements on the EFS and are not discussed further.
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 l
in the feedwater or condensate system Loss of of fsite power (blackout) with the consequential shutdown of l
I the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a reduction in steam generator water levels which results in a reactor trip, a turbine trip, and emergency feedwater actuation by the protection system logic.
Following reactor trip from a high initial power level, the power quickly falls to decay heat levels. The water levels continue to decrease, progressively uncovering the steam generator tubes as decay i
heat is treneferred and discharged in the form of steam either through the steas y a p valves to the condenser or through the steam generator safety or
-,wer-operated relief valves to the atmosphere. The reactor coolant temperature increases as the residual heat in excess of that j
dissipated through the steam generators is absorbed. With increased i
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tarparatura, tha volume of reactor ecolant expands and begins filling the pressurizer. Without the addition of sufficient emergency feed-l
' ater, _ further expansion will result in water being discharged through w
the pressurizer safety and/or relief valves.
If the temperature rise and the resulting volumetric expansion of the primary coolant are per-mitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing overpressurization 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 r
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.
He nce, the timely introduction of sufficient emergency feedwater is necessary to arrest the decrease in the steam generator 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 be 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 generator safety valves or the power-operated relief valves. The calcu-lated transient is similar for bcth the loss of main feedwater and the blackout, except that reactor coolant pump heat input is not a consid-eration in the blackout transient following loss of power to the reactor coolant pump bus.
The station blackout transient serves as the basis for the minimum flow required for a single emergency feedwater pump for the Seabrook Plants.
The pump is sized so that a single pump will provide sufficient flow against the steam generator safety valve set pressure (with 3% accumula-tion) to prevent water relief from the pressurizer.
The same criterion is met for the loss of feedwater transient by the operation of any pump, where A/C power is available.
Secondary System Pipe Ruptures 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 emer-gency feedwater piping to an individual steam generator.
Another signi-ficant result of a feedline rupture may be the spilling of emergency feedwater to the faulted steam generator.
Such situations can result in the injection of a disproportionately large fraction of the total emer-gency feedwater flow (the system preferentially pumps water to the icwest pressure region) to the faulted loop rather than to the effective 2288Q:1
=
cterm gannrztors which ara at relctivaly high prassura. Tha systes design must allow for terminating, limiting, or minimizing that fraction of emergency f eedwater flow which is delivered to a f aulted loop or spilled through a break in order to ensure that sufficient flow will be delivered to the remaining effective steam generator (s). The concerns are similar for the main feedwater line rupture as those explained for the loss of main feedwater transients.
Main steamline rupture accident conditions are characterized initially by plant cooldown and, for breaks inside containment, by increasing containment pressure and temperature.
Emergency feedwater is not needed during the early phase of the transient but flow to the faulted loop will contribute to an exc?asive release of mass and energy to contain-ment. Thus, steamline rupture conditions establish the upper limit 'on emergency feedwater flow delivered to a faulted loop.
Eventually, how-ever, the Reactor Coolant System will heat up again and emergency feed-water flow will be required to be delivered to the non-faul'ted loops, but at somewhat lower rates than for the loss of feedwater transients described.previously. Provisions must be made in the design of the Emergency Feedwater System to limit, control, or terminate the emergency feedwater flow to the faulted loop as necessary in order to prevent containment overpressurization following 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 condi-tions wherein not only onsite and offsite AC power is lost but also AC emergency power is lost as an assumed common mode f ailure. Batte ry power for operation of protection circuits is assumed available. The impact on the Emergency Feedwater System is the necessity for providing both an emergency feedwater pump power and control source which are not dependent on AC power and which are capable of maintaining the plant at i
hot shutdown until AC power is restored.
Loss-of-Coolant Accident (LOCA)
The loss of coolant accidents do not impose on the emergency 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 f.or the sake of completeness to explain the role of the emergency 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 con-tribution from the Emergency 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 seconaary side of the steam generators provides a heat sink for removing decay heat and establishes the capability for providing a buoyancy head for natural c<.rculation.
The emergency feedwater system may be utilized to assist in a system 2288Q:1
4 cooldewn end depraecurizcticn follcwing a ccc11 LOCA while bringing tha reactor to a cold shutdown condition.
Cooldown The cooldown function performed by the Emergency Feedwater System is a partial one since the reactor coolant system is reduced from norm.1 zero load temperatures to a hot leg temperature of approximately 350*F.
The latter is the maximum temperature recommended for placing the Residual 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 such as a main feedline break, or during a normal cooldown prior to refueling or performing reactor plant maintenance.
If the reactor is tripped following extended operation at rated power level, the EFWS is capable of delivering sufficient EFW to remove decay heat and reactor coolant pump (RCP) heat following reactor trip while main-taining 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 emergency 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.
l 1
i l
l l
l 2288Q:1 i
R2 psn:a to 1.b
. Table 1B-1 summarizes the criteria which are the general design bases for each event, discussed in the response to Question 1.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 Emergency Feedwater System is to provide sufficient heat removal capability following reactor trip and to remove the decay heat generated by the core and prevent system overpressuriza-tion. 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 determined by other bases are incorporated into this analysis, resulting in no additional flow requirements.
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TABLE IB-1 Criteria for Emergency Feedwater System Design Basis Conditions 1
Condition or Additional Design Transient Classification
- Criteria Criteria Loss of Main Feedwater Condition II Peak RCS pressure not to exceed design pressure +10%.
No consequential fuel failures Station Blackout Condition II (same as LMFW)
Pressurizer does not fill.
1 emergency feed pump feeding 2 SGs.
Feedline Rupture Condition IV 10CFR100 dose limits.
Core does not uncover Peak RCS pressure not to exceed design pressure +10%
Loss of all A/C Power Condition II 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 Condition I 100*F/hr 557*F to 350*F Steamline Rupture Condition IV 10 CFR 100 dose limits DNBR above limit value Containment design pressure not exceeded
- Ref:
ANSI N18.2 (This information provided for those transients performed in the FSAR).
Note 1 Although this transient establishes the basis for EFW 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.
Quastien 2 Describe the analyses and assumptions and corresponding technical justi-fication used with plant condition considered in 1.a above including:
i 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 af ter 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 circulation.
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 steam generator (s).
AFW pump flow capacity allowance to accommodate the time delay and i
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.
i i
1.
Operating condition of steam generator normal blowdow,ollowing initiating event.
m.
Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
1 i
n.
Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.
Response to 2 4
Analyses have been performed for the limiting transients which define the EFWS performance requirements. These analyses have been provided for review in the Seabrook FSAR. The applicable ones include:
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Loss of Main Feedwater (Station Blackout)
-. Rupture of a Main Feedwater Pipe Rupture of a Main Steam Pipe Inside Containment In addition to the above analyses, calculations have been performed specifically for Seabrook Units No. 1 and 2 to determine the plant cool-down 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 emergency pump having a diverse (non-AC) power supply. The LOCA analysis, as discussed in response 1.b, incorporates the system flows requirements as defined by other transients, and there-fore is not performed for the purpose of specifying EFWS flow require-ments. Each of the analyses listed above is 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 15 2.6 for the purpose of showing that for a station blackout transient the peak RCS pressure remains below the criterion for Condition 11 transients and no fuel failures occur (refer to Table 1B-1).
Table 2-1 summarizes the assumptions used in this analysis. The analysis assumes that the plant is initially operating at 102% (calorimetric error) of the Engineered Safeguards design (ESD) rating shown on the table, a very conservativa assumption in defining decay heat and stored energy in the RCS. The reactor is assumed to be tripped on low-low steam generator water level, allowing for level uncertainty. The FSAR shows that there is a considerable margin with respect to filling the pressurizer. A loss of normal feed-water transient with the assumption that only one emergency feedwater pump is running also shows that there is considerable margin.
This analysis establishes the capacity of a single pump and also establishes train association of equipment so that this analysis remains valid assuming the most limiting single failure.
Rupture of Main Feedwater Pipe The double ended rupture of a main feedwater pipe downstream of the main feedwater line check valve is analyzed in FSAR Section 15 2 8.
Table 2-1 swamarizes the assumptions used in this analysis.
Reactor trip is assumed to occur when the faulted steam generator is at the low-low level setpoint (adjusted for errors). This conservative assump-tion maximizes the stored heat 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. As in the loss of normal feedwater analysis, the initial power rating was assumed to be 102% of the ESD rating. Emergency feed-water flow of 470 gpm was assumed to be delivered to the two non-faulted steam generators 1 minute af ter reactor trip. At 30 minutes af ter the break, it is assumed that the operator has isolated the EFWS from the break.
The criteria listed in Table IB-1 are met.
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This analysis dons establish ths capacity of a single pump, establishes requirassnts for layout to preclude indefinita loza of exsrgsncy feed-water to the postulated break, and establishes train association
' requirements for equipment so that the EFWS can deliver the minimum flow required in 1 minute assuming the worst single failure.
Rupture of a Main Steam Pipe Inside Containment Because the steamline break transient is a cooldown, the EFWS is not needed to remove heat in the short term. Furthermore, addition of excessive emergency feedwater to the faulted steam generator will affect the peak containment pressure following a steamline break inside con-tainment. This transient is performed at four power levels for several break sizes. Emergency feedwater is assumed to be initiated at the time of the break, independent of system actuation signals.
The maximum flow is used for this analysis. Table 2-1 summarizes the assumptions used in this analysis. At 10 minutes after the break, it is assumed that the operator has isolated the EFWS from the faulted steam generator which subsequently blows down to ambient pressure. The criteria stated in Table 1B-1 are met.
This transient establishes the maximum allowable emergency feedwater flow rate to a single faulted steam generator assuming all pumps oper-ating, establishes the basis for runout protection, if needed, and establishes 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 opera-tion, 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 1A, the emergency fee'd-water system partially cools the system to the point where the RHRS may complete the cooldown, i.e., 350*F in the RCS. Table 2-1 shows the assumptions used to determine the cooldown heat capacity of the emergency feedwater system.
The cooldown is assumed to commence at the maximum rated power, and maximum trip delays and decay heat source terms are assumed 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 EFWS.
See Table 2-2 for the items constituting the sensible heat stored in the NSSS.
This operation is analyzed to establish minimum tank size requirements for emergency feedwater fluid searce which are normally aligned.
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TABLE 3-1 Summary of Assumpti<.as Used in EfWS Design Verification Analyses loss of Feedwater Main Steamline Break Transient (station blackout)
Cooldown Main Feed!!ne Break (containment) a.
Han reactor power 102% of ESD rating 3650 MWt 1021 of ESD rating 0, 25, 50, 75, 102% of (1021 of 3579 MWt)
(102% of 3579 MWt) rated (percent -of
- 3425 MWt) b.
Time delay from 43 1 sec 2 sec 6 3 see variable event to Rx trip c.
EFWS actuation sig-lo-lo SG level NA low-low SG 1evel/
Assumed immediately f
nal/ time delay for 1 minute 1 minute 9 0 sec (no delay)
EfWS flow d.
SG water level at (low-low SG level)
NA (Iow-low SG level)
N/A time of reactor trip 01 NR span 01 NR span e.
Initial SG inventory 62.100 lbm/SG (at 91.270 lbm/SG 94.230 lbm/SG 1 SG consistent with power trip) at 544.6*F 83.970 lbe/Sd 3 SC's Rate of change before See FSAR N/A turnaround at 3760 see N/A l
& siter EfWS actuation Figure 15 2-10 6
Decay heat ANS + 20%
EFW pump design 1236 pois 1236 pois 1236 psia N/A I
pressure g.
Minimum # of SGs 2 of 4 N/A 2 of 4 N/A which must receive AFW flow h.
RC pump status Tripped at reactor Tripped Tripped at reactor trip All operating trip i
1.
Maximum EfW 120*F 120*F 120*F 100*F temperature J.
Operator action
- n..ne.
N/A 3'O min.
10 minutes 1
200 ft /446*F 800 ft3/ loop (for 3
k.
MFW purge volume /
200 f t3/446*F 200 f t3/
S/G and temperature 446*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.
Time at standby / time see cooldown 2 hr/5 hre see cooldown N/A to cooldown to RHR o.
EFW flow rate 470 gpa - constant variable 470 gpa af ter 750 spa to broken (min. requirement) 1 minute SG for 10 minutes 2288Q:1
TABLE 2-2 Summary of. Sensible Heat Sources Primary Water Sources (initially at ESD power temperature and inventory)
- RCS fluid
- Pressurizer fluid (liquid and vapor)
Primary Metal Sources (initially at ESD power temperature)
- Reactor coolant piping, pumps and reactor vessel
- Pressurizer
.i
- Steam generator tube metal and tube sheet
- Steam generator metal below tube sheet i
- Reactor vessel internals l
Secondary Water Sources (initially at ESD power temperature and invento ry)
- Steam generator fluid (liquid and vapor) s
- Main feedwater purge fluid between steam generator and EFWS piping.
Secondary Metal Sources (initially at ESD power temperature)
- All steam generator metal above tube sheet, excluding tubes.
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