ML19346A195
| ML19346A195 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway  |
| Issue date: | 06/03/1981 |
| From: | Petrick N STANDARDIZED NUCLEAR UNIT POWER PLANT SYSTEM |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM SLNRC-81-39, NUDOCS 8106050372 | |
| Download: ML19346A195 (41) | |
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e-D_4 JUN 04)981d SNUPPS
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Sandardized Nucieer Unit
.O Power Plant System 36 Nicholas A. f strick 5 choke ctmery Road g'~~
Executive Director Rockvine, Meryland 20eso (301) 869 8010 June 3, 1981 SLNRC 81-39 FILE: 0290 SUBJ: Auxiliary Feedwater System Mr.HaroldR.Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Dockets: STN 50-482, STN 50-483, STN 50-486 Ref. 1. SLNRC 81-010, dated February 19, 1981, SNUPPS Auxiliary Feedwater System Meeting
- 2. NRC Sumary of Meeting on February 12, 1981 with Callaway and Wolf Creek Applicants, dated February 18, 1981
Dear Mr. Denton:
On February 12, 1981, a meeting with the NRC Staff was conducted to re-view the Auxiliary Feedwater System. The referenced correspondence identified the additional information that was required in order for the NRC to complete its review of the system. This letter provides the requested information as delineated below.
1.
The attached table 10.4-13A provides an evaluation of the SNUPPS Auxiliary Feedwater System against each recommendation in Standard Review Plans 10.4-9 and Branch Technical Position ASB 10-1. This table will be incorporated into the FSAR in a future revision.
2.
The attached table 10.4-13B provides an evaluation of the SNUPPS design against NRC recommendations on Auxiliary Feedwater Systems contained in a March 10, 1980 NRC generic letter. This table will be incorporated into the FSAR in a future revision.
- 3. to the NRC generic letter of March 10, 1980 requested certain information relative to Auxiliary Feedwater System flow requirements. This information is attached to this lett1r as i
Enclosure B.
l 4.
SNUPPS was requested to provide a discussion and verification
@f l
that the automatic transfer of auxiliary feedwater suction supply from the condensate storage tank to the essential service water 5
system can occur fast enough to prevent damage to the AFW pumps
/
in the event the common suction supply valve is inadvertently f//
closed or blocked. The response to this request is included in 81060503?)._
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I SLNRC 81-39 Page Two the SNUPPS position relqttive to NRC recommendation GS-4 (see attached table 10.4-138). Also the attached FSAR changes (see insert for page 10.4-50F expand the description of low suction pressure switchover of supply sources.
5.
SNUPPS was requested to provide a discussion of the potential for adverse thermal effects on the main feedwater piping when cold auxiliary feedwater is injected into the main feedwater line.
l l
The effects of several occurrences of filling hot (504F) feedwater l
piping with cold (50F) auxiliary feedwater on feedwater pipe design life has been investigated and determined to-be negligible.
This conclusion is based on a design comparison with the results
(
of ASME Class 1 piping fatigue analysis of more severe Reactor Coolant System temperature transients. Specifically, the fatigue analysis due to at least 70 occurrences of high head safety injec-tion with temperature transients greater than 500F results-in Code acceptable usage factors.
Further, these analyses are performed on j
thicker wall stainless steel pipe which are conservative conditions l
with respect to the feedwater piping.
This conclusion is also based on the fact that many operating plants use cold auxiliary feedwater for normal cooldown, which subjects the feedwater piping to a similar transient to that dis-cussed above, without adverse effects.
Transient fatigue analysis for Class 2 piping is not required by the ASME Code. The range of pipe stress has been determined to I
be within Code allowable limits based on operating temperature extremes. All feedwater piping from the auxiliary feedwater con-nection to the steam generator is impact tested and normalized to remove residual stresses from fabrication.
In addition, periodic volumetric inservice inspection is performed on this piping in accordance with ASME Code requirements. SNUPPS considers the l
existing stress analysis sufficient for determination of structural adequacy in accordance with the ASME Code.
6.
SNUPPS was requested to evaluate the potential for adverse conse-quences caused by failures associated with the gas storage tanks used as valve operator accumulators. The effects of failure of the largest nozzle on the gas storage tanks were considered. The resulting maximum room pressure is less than 1 psi which is well l
within design capabilities of the structure.
7.
The SNUPPS auxiliary feedwater system automatic initiation and flow indication meet or exceed the requirements of NUREG-0737, II.E.1.2.
f 8.
The 48-hour pump endurance test is addressed in Table 10.4-138 (Sheet 7). A water-hammer test using normal plant procedures will j
be conducted at all SNUPPS plants.
f SLNRC 81 39 Page Three 9.
The reliability study of the SNUPPS Auxiliary Feedwater System is being transmitted to the NRC by a separate letter.
In addition to the February 12 meeting discussed above, a meeting.was held on April 29, 1981 with the NRC. The main feedwater system was among the topics discussed in this latter meeting. A planned design change to the main feedwater system is the addition of a small, motor-driven feedwater pump in parallel with the main feedwater pumps.
This new pump will be used during normal startup and shutdown instead of the auxiliary feedwater pumps as had been planned. The attached FSAR.
changes on the Auxiliary Feedwater System delete t.he use of the sys-tem during normal plant operations. Changes to the descriptions of the main feedwater system will be made at a later date.
The primary reason for adding the new pump is so that heated feedwater can be used during normal plant operations and thereby reduce the likelihood of steam generator nozzle cracking (refer to IE Bulletin 79-13).
Ver truly yours,.
b<%C Nicholas A. Petrick RLS/dck/lb18
Enclosures:
A. Draft FSAR Changes B. Response to Enclosure 2 of NRC 3/10/80 letter cc:
J. K. Bryan, UE G. L. Koester, KGE
- 0. T. McPhee, KCPL W. A. Hanson, NRC Callaway j
T. E. Vandel, NRC Wolf Creek l
l l
SNUPPS Enclosure A to SLNPC 81-39 10.4.9 AUXILIARY FEEDWATER SYSTEM The auxiliary feedwater system (AFS) is a-reliable source of water for the steam generators.
The AFS, in conjunction with safety valves in the main steam lines, is a safety-the function of which is to remove thermal related system, energy from the reactor coolant system by releasing second-The AFS also provides emergency ary steam to the atmosphere.
Removal of water following a secondary side line rupture.
heat in this manner prevents the reactor coolant pressure from increasing and causing release of reactor coolant through the pressurizer relief and/or safety valves.
The auxiliary feedwater system may also be used following a reactor shutdown in conjunction with the condenser dump to cool the reactor valves or atmospheric relief valves, coolant system to 350 F and 400 psig, at which temperature the residual heat removal system is brought into operation.
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10.4.9.1 Design Bases 10.4.9.1.1 Safety Design Bases SAFETY DESIGN BASIS ONE - The AFS is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles (GDC-2).
SAFETY DESIGN BASIS TWO - The AFS is designed to remain functional after an SSE or to perform its intended function internal following a postulated hazard, such as a fire, misuile, or pipe break (GDC-3 and 4).
SAFLTY DESIGN BASIS THREE - The safety funtions can be assuming a single active component failure coin-performed, The system require-cident with the loss of offsite power.
ments may be met with a complete loss of ac power (GDC-34).
SAFETY DESIGN BASIS FOUR - The AFS is designed so that the active components are capable of being tested during plant Provisions are made to allow for inservice operation.
inspection of components at appropriate times specified in Section XI.
the ASME Boiler and Pressure Vessel Code, 10.4-46
~
SNUPPS SAFETY DESIGN BASIS FIVE - The AFS is designed and fabricated consistent with the. quality group classification assigned by Regulatory Guide 1.26 and the seismic category assigned by Regulatory Guide 1.29.
The power supply and control functions are in accordance with Regulatory Guide 1.32.
SAFETY-DESIGN BASIS SIX - The AFS, in conjunction with the condensate storage tank (nonsafety-related) or essential service water system, provides feedwater to maintain suf-ficient steam g.enerator level to ensure heat removal from the reactor coolant system in order to achieve a safe shut-down following a main feedwater line break, a main steamline break, or an abnormal plant situation requiring shutdown.
The auxiliary feedwater system is capable of delivering full flow when required, auxiliary feedwater (refer to chapter 15.0).after detection of any accident requ SAFETY DESIGN BASIS SEVEN - The capability to isolate com-ponents or piping is provided, if required, so that the AFS safety function will not be compromised.
This includes isolation of components to deal with leakage or malfunctions and to isolate portions of the system that may be directing flow to a broken secondary side loop.
SAFETY DESIGN BASIS EIGHT - The AFS has the capacity to be operated locall water control, y as an alternate, redundant means of feed-(
in the unlikely event that the control room
)
must be evacuated.
10.4.9.1.2 Power Generation Design Bases The AFS has no power generation design bases. The Condensate and Feedwater System is designed to provide a continuous feedwater supply to the steam generators during startup, normal plant operation, and shutdown. Refer to Section 10.4.7.
10.4.9.2
System Description
10.4.9.2.1 General Description The system consists of two motor-driven pumps, one steam turbine-driven pump, and associate piping, valves, instru-ments, and controls, as shown on Figure 10.4-9 and described in Table 10.4-12.
Figure 10.4-10 shows the piping and instrumentation for the steam turbine.
Each motor-driven auxiliary feedwater pump will supply 100 percent of the feedwater flow required for removal of decay heat from the reactor.
The turbine-driven pump is sized to i
supply up to twice the capacity of a motor-driven pump. This capacity is sufficient to remove decay heat and to provide adequate feedwater for cooldown of the reactor coolant system at 50 F/hr within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> of a reactor trip from full l
power.
10.4-47 Rev. 3 4/81
SNUPPS Normal flow is from the condensate storage tank (CST) to the auxiliary feedwater pumps.
Two redundant safety-related back-up sources of water from the essential service water system (ESWS) are provided for the pumps.
For a more detailed description of the automatic sequence of events, refer to Section 10.4.9.2.3.
Thecondensatestoragebankcapacityallowstheplant to remain at hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and then cool down the primary system at an average rate of 50 F per hour to a temperature of 350 F.
Figure 10.4-11 provides the required makeup rate to the steam generators to maintain their level based on removing decay heat from a fully irradiated reactor core.
Initially, sensible heat is removed from the reactor coolant system to reduce the temperature from a full-power operation average temperature of 588 F to a nominal het shutdown tempera-ture of 500 F.
Subsequently, to bring the reactor down to 050 F at 50 F/hr, an initial makeup rate of 500 gpm is required.
Refer to Section 9.2.6 for a description of the condensate storage system.
In order to remove decay heat by the steam generators, auxiliary feedwater must be supplied to the steam generators in the event that the normal source of l
'^"' '
feedwater is lost.
The minimum required flow rate is 470 gpm.
Provisions are incorporated in the AFS design to allow for periodic operation to demons rate performance and struc-tural and leaktight 11tegrit.
Leak detection is provided by visual examination and in he floor drain system described in Section 9.3.3.
10.4.9.2.2 Component Description l
Codes and standards applicable to the AFS are listed in Tables 3.2-1 and 10.4-12.
The AFS is designed and con-structed in accordance with quality groups B and C and seismic Category I requirements.
MOTOR-DRIVEN PUMPS - Two auxiliary feedwater pumps are driven by ac-powered electric motors supplied with power from indepen-dent Class lE switchgear busses.
Each horizontal centrifugal pump takes suction from the condensate storage tank, or alternatively, from the ESWS.
Pump design capacity includes continuous minimum flow recirculation, which is controlled by restriction orifices.
TURBINE-DRIVEN PUMP - A turbine-driven pump provides sys-tem redundancy of auxiliary feedwater supply and diversity of motive pumping power. The pump is a horizontal centrifugal I
10.4-48
SNUPPS i
unit.
Pump bearings are cooled by the pumped fluid.
Pump design capacity includes continuous minimum flow recirculation.
Power for all controls, valve operators and other support systems is independent of ac power sou,rces.
Steam supply piping to the turbine driver is taken from.
two of the four main steam lines between the containment penetrations and the mais steam isolation valves.
.Each of the steam supply lines to the turbine is equipped with a locked-open gate valve, normally closed air-operated globe valve with air-operated globe bypass to. keep the line warm, and two nonreturn valves.
Air-operated globe valves'are equipped with de-powered solenoid valves.
These steam sup-ply lines join to form a header which leads to the turbine through a normally closed, de motor-operated mechanical trip and throttle valve.
The main steam system is described in Section 10.3.
The steam lines contain provisions to prevent the accumu-i lation of condensate.
The turbine driver can operate with steam inlet pressures ranging from 100 to 1,250 psia.
Ex-haust steam from the turbine driver is vented to the at-mosphere above the auxiliary building roof.
PIPING AND VALVES - All piping in the AFS is seamless car-j bon steel.
Welded joints are used throughout the system, except for flanged connections at the pumps.
The piping from the ESWS to the suction of each of the auxiliary feedwater pumps is equipped with a motor-operated butterfly valve, an isolation valve, and a nonreturn valve.
Each line from the condensate storage. tank is equipped with a motor-operated gate valve and a nonreturn valve.
Each motor-driven pump discharges through a nonreturn valve and a locked-open isolation valve to feed two steam generators through individual sets of a locked open isolation valve, a normally open, motor-operated control valve, a check valve followed by a flow restriction orifice, and a locked-open globe valve.
The turbine-driven pump discharges through a nonreturn valve, a locked-open gate valve to each of the four steam generators through individual sets of a locked-open isolation valve, a normally open air-operated control valve, followed by a nonreturn valve, a flow restriction orifice, and a locked-open globe valve.
The turbine-driven pump discharge control valves are air operated with de-powered solenoid valves.
At each connection to the four main feedwater lines, the auxiliary feedwater lines are eq;ipped with check valves.
The system design precludes the occurrence of water hammer in the main feedwater inlet to the steam generators.
For a de-scription of prevention of water hamme r, refer to Section 10.4.7.2.1.
10.4-49 i
l i,
SNUPPS 10.4.9.2.3 System Operation "L.'2" CTA"TU" "uring startup, th; cunilicry f : duct:r nd centr:1 valve cre
- d under ::nu:1 :::tr:1 te purp supply fcedwatc; f c; the conden;;t; st;;;;c tank t; th
- te : ;:ncr t ::
ntil :ufficicnt steca i; availabic t:
Oper t: the turbin; driven _ncin f : duct; p;;;:.
NORMAL PLANT OPERATION - The AFS is not required during The pumps are placed in the auto-normal power generation.
matic mode, lined up with the condensate storage tank, and are available if needed.
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centr:1 t cupply r:ter fre th: ::nd n::t: :t ::;: t:n': te th:
ten: ;:ncreter:.
Auniliary f::ductor f1:u t: ::ch etenn gener:t:r i: ::;ulated hy th; ;;ntrol valvc;.
Otc;;
gener:ted in thi: n:nner i: hyp :::d t; the ncin-;;ndon:cr.
The muniliary fe:drater punp; crc u;;d until rcacter coolant temper:tur; drep: t: 250 r, et which point th: :::idu:1 h::t n:v:1 cycter ir placed in : rvie: and furth:r : 01: dern th: ::::t:r.
EMERGENCY OPERATION - In addition to remote manua'.-actuation capabilities, the AFS is aligned to be placed into service automatically in the event of an emergency.
Anyone of the following conditions will cause automatic startup of both motor-driven pumps:
Two out of four low-low level signals in any one a.
l l
Trip of both main feedwater pumps Safeguards sequence signal (initiated by safety c.
injection signal or loss-of-offsite power)
Class IE bus loss of voltage sequence signal (i.e.
d.
I loss-of-offsite power)
The turbine-driven pump is actuated automatically on either of the following signals.
Two out of four low-low level signals in any two i
a.
steam generators Class IE bus loss of voltage sequence signal (i.e.
b.
loss-of-offsite power)
The common water supply header from,the condensate storage tank contains a locked-open, twelve-inch, butterfly isolation valve. Correct valve position is verified by periodic In case of failure of the water supply from surveillance.
the condensate storage tank, the normally closed, motor-7 operated butterfly valves from the ESWS are automatically opened on low suction header pressure. Valve opening time start time are coordinated to assure adequate and pump suction pressure with either onsite or offsite power available.
10.4-50 Rev. 3 4/81 W
l l
SNUPPS If a motor-driven pump supplying two of the three intact
~
steam generators fails to function, the turbine-driven pump will automatically start when a low-low level is reached in two of the four steam generators.
During all of the above emergency conditions, the normally open control valves are remote manually operated.
During all of the above imergency conditions, the normally open motor-driven pump control valves are automatically operated to limit runout flow under all secondary side pressure conditions.
This is required to prevent pump suction cavitation at high flow rates.
The turbine-driven pump design includes a lower NPSH requirement.
Therefore, the turbine-driven pump control valves are remote manually operated.
Low pump discharge pressure alarms will alert the operator to a secondary side break.
The operator will then determine which loop is broken by observing high auxiliary feedwater flow, using control room flow indication, and close the appropriate discharge control valve.
This can be accomplished within 10 minutes after pump start.
Refer to Chapter 15.0.
10.4.9.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety
~
design bases in Section 10.4.9.1.1.
SAFETY EVALUATION ONE - The AFS is located in the auxiliary building.
This building is designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external mis-siles, and other appropriate natural phenomena.
Sections 3.3, 3.4, 3.5, 3.7(B), and 3.8 provide the bases for the adequacy of the structura? design of the auxiliary building.
SAFETY EVALUATION TWO - The AFS is designed to remain func-tional after a SSE.
Sections 3.7(B).2 and 3.9(B) provide the design loading conditions that were considered.
Secticns 3.5, 3.6, and 9.5.1 provide the hazards analyses to ensure that a safe shutdown, as outlined in Section 7.4, can be achieved and maintained.
For a more complete description of motor qualification, refer to Sections 3.10(B) and 3.11(B).
SAFETY EVALUATION THREE - Complete redundancy is provided and, as indicated by Table 10.4-13, no single failure will compromise the system's safety functions.
All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0.
The turbine-driven pump is energized by steam drawn from two main steam lines between the containment penetrations and the main steam isolation valves.
All valves ano controls 10.4-51 Rev. 3 4/81
SNUPPS necessary for the function of the turbine-driven pump are
~
Turbine bearing energized by the Class IE dc power supplies.
lube oil is circulated by an integral shaft-driven pump.
Turbine and pump bearing oil is cooled by pumped auxiliary feedwater.
SAFETY EVALUATION FOUR - The AFS is initially tested with the program given in Chapter 14.0.
Periodic operational testing is done in accordance with Section 10.4.9.4.
Section 6.6 provides the ASME Boiler and Pressure Vessel Section XI requirements that are appropriate for the
- Code, AFS.
SAFETY EVALUATION FIVE - Section 3.2 delineates the quality group classification and seismic category applicable to this Table 10.4-12 shows that the system and supporting systems.
components meet the design and fabrication codes given in Section 3.2.
All' the power supplies and control function necessary for safe function of the AFS are Class IE, as described in Chapters 7.0 and 8.0.
SAFETY EVALUATION SIX - The AFS provides a means of pumping sufficient feedwater to prevent damage to the reactor follow-ing a main feedwater line break inside the containment, or a as well as to cool down the main stramline break incident, reactor coolant system at a rate of 50 F per hour to a temperature of 350 F, at which point the residual heat Pump capacities, as shown in removal system can operate.
and start times'are such that these objectives Table 10.4-12, Restriction crifices located in the pump discharge are met.
lines and automatic flow control valves for the motor-driven pumps limit the flow to the broken loop so that adequate cooldown flow (470 gpm) can be provided to the other steam generators for removal of reactor decay heat and so that containment design pressure is not exceeded.
Pump discharge head is sufficient to establish the minimum recessary flowrate against a steam generator pressure corresponding to the lowest pressure setpoint of the main steam safety valves.
The maximum time period required to start the electric motors and the steam turbine which drive the auxiliary feedwater pumps is chosen so that sufficient flowrates are established within the required time for primary system protection.
Refer to Chapter 15.0.
SAFETY EVALUATION SEVEN - As discussed in Sections 10.4.9.2 and 10.4.9.5 and Chapter 15.0, adequate instrumentation and control capability is provided to permit the plant operator to quickly identify and isolate the auxiliary feedwater flow to a brokan secondary side loop.
Isolation from nonsafety-including the condensate related portions of the system, is provided as described in Section 10.4.9.2.
storage tank, 10.4-52 Rev. 3 4/81
SNUPPS SAFETY EVALUATION EIGHT - The AFS can be controlled from either the main control room or the auxiliary shutdown panel.
Refer to Section 7.4 for the control description.
10.4.9.4 Tests'and Inspections Preoperational testing is described in Chapter 14.0.
The performance and structural and leaktight integrity of system components is demonstrated by periodic operation.
The AFS is testable through the full operational sequence l
that brings the system into operation for reactor shutdown i
and for DBA, including operation of applicable portions of the protection system and the transfer between normal and standby power sources.
l The safety-related components, i.e., pumps, valves, piping, and turbine, are designed and located to permit preservice
~
and inservice inspection.
10.4.9.5 Instrumentation Applications The AFS instrumentation is designed to facilitate automatic operation and remote control of the system and to provide continuous indication of system parameters.
l Pressur'e transmitters are provided in the discharge and suction lines of the auxiliary feedwater pumps.
Auxiliary feedwater flow to each steam generator is indicated by flow indicators provided in the control room.
If the condensate supply from tha storage tank fails, the resulting reduction of pressure at the pump suction is indicated in the control room.
Flow transmitters and control valves with zemote control stations are provided on the auxiliary fecdwater lines to each steam generator to indicate and allow control of flow at the auxiliary shutdown panel and in the control room.
Flow controllers for the motor-driven pump control valves position the valves to limit the flow to a preset value through the full range of downstream operating pressures.
Table 10.4-14 summarizes AFS controls, alarms, indication of 3
status, etc.
j
[Redundantcondensatestoragetank level indication and alarms are provided in the control room. The backup L
indication and a' arm use auxiliary feedwater pump suction s
I pressure by converting it to available tank level. Both alarms provide at least 20 minutes for operator action (e.g. refill the tank) assuming the largest capacity auxiliary (feedwater pump is operating.
10.4-53 Rev.
4/81, s
1 SHUPPS TABLE 10.4-13A i
DESJGN COMPAhISONS TO RECOMMENDATIONS OF STANDARD gy REVIEW PLANT IO.4.9 REVISION 1,
" AUXILIARY FEEDWATER SYSTEM'(P R)"
AND BRANCH TECHNICAL POSITION ASB 10-1 REVISION 1,
" DESIGN GUIDELINES FOR AUXILIARY FEEDWATER SYSTEM j
PUMP DRIVE K1D POWER SUPPLY DIVERSITY FOR 1
PRESSURIZED WATER REACTOR PLANTS" i
SNUPPS POSITION I.
SRP 10.4.9 RECOMMENDATION j
ACCEPTANCE CRITERIA:
1 GeneralDesignCriteri[2, l
as related to struc-Complies.
The system is located in a seismic tures housing the system and the system itself Category 1 structure that is tornado, missile, being capable of withstanding the ef fects of and flood protected.
Refer to Section 3.1.3.
1 natural phenomena such as earthquakes, torna-does, hurricanes, and floods.
{
Ceneral Design Criterion 4,,with respect to Complies.
The system components are located structures housing the sysee and the system in individual rooms that will withstand the i
itself being capable of withstanding the ef-effects of flooding, missiles, pipe whip, and fects of exte nal missiles and internally gen-jet impingement forces associated with pipe erated missiles, pipe whip, and jet impinge-breaks.
Refer to Section 3.1.3 and Figure i
ment forces associated with pipe breaks.
1.2-11.
General Design Criterion 5, as related to the complies.
There is no sharing between units I
capability of shared systems and components of auxiliary feedwater systems or components.
important to safety to perform required safety Refer to Section 3.1.3.
l functions.
l General r+
Agn Criterion 19, as related to the complies. 'The system can be controlled from design o taility of system instrumentation either the main control room or the auxiliary and cont, as for prompt hot shutdown of the shutdown panel as a redundant means of feed-l reactor and potential capability for subsequent water control.
Refer to Section 3.1.3.
I cold shutdown.
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i SNUPPS TABLE 10.4-13B I
DESIGN COMPARISONS TO NRC RECOMMENDATIONS i
ON AUXILIARY FEEDWATER SYSTEMS CONTAINED l
IN THE MARCH 10, 1980 NRC LETTER j
i SNUPPS POSITION j
A. SHORT TERM RECONMENDATIONS The limiting conditions for operation related 1
Recommendation CS The licensee should propose j
l.
modifications to the Technical Specifications to to the auxiliary feedwater system will be l
limit the time that one auxiliary feedwater system addressed in the proposed Technical Specifications pump and its associated flow train and essential for the SNUPPS plants. The proposed Technical i
instrumentation can be inoperable. The outage Specifications will be submitted approximately time limit and subsequent action time should be one year before the scheduled fuel load for the first SNUPPS unit and will be based on NUREC-0452, as required in current Technical Specifications;
- 3. " Standard Technical Specifications for 1.e., 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, respectively.
Rev.
j ~
Westinghouse Pressurized Water Reactors."
l Recommend' tion CS The licensee should lock This item is not applicable to SNUPPS because the i
2.
a open single valves or multiple valves in series design does not include single valves or multiple i
in the au'xiliary feedwater system pump suction valves in series that could interrupt auxiliary j
piping and lock open other single valves or feedwater pump suction or all auxiliary feedwater flow.
multiple valves in series that could interrupt 1
i all auxiliary feedwater system flow. Monthly inspections should be performed to verify that j
these valves are locked and in the open position.
l These inspections should be proposed for incorpo-ration into the surveillance requirements of the l
l plant Technical Specifications. See recommendation CL-2 for the lonP,er-term resolution of this I
concern.
i
{
3.
Reccamendation CS The licensee has stated that Throttling auxiliary feedwater flow to avoid water it throttles auxiliary feedwater flow to avoid hammer will not be utilized. The system design water hammer. The licensee should reexamine the precludes the occurrence of water hammer in the l
practice of throttling auxiliary feedwater system steam generator inlet as described in Section 10.4.7.2.1.
j flow to avoid water hammer. The licensee should verify,that the auxiliary feedwater system will supply on demand sufficient initial flow to the necessary steam generators to assure adequate j
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i SNUPPS TABLE 10.4-135 (Sheet 4)
DESIGN COMPARISONG TO NRC RECOMMENDATIONS ON AUXILIARY FEEDWATER SYSTEMS CONTAINED IN THE MARCH 10. 1980 NRC LETTER 6 (con't)
.i that the valves are properly aligned The licensee should propose Technical Specifications to assure that prior l
to plant startup following an extended cold shutdown, a flow test would be per-l formed to verify the normal flow path from the primary auxiliary feedwater
}
system water source to the steam gener-i ators. The flow test should be conducted with auxiliary feedwater system valves j
in their normal alignment.
I 7.
Recommendation CS The licensee should verify The SNUPPS auxiliary feedwater system is designed that the automatic start auxiliary feedwater system so that automatic initiation signals and circuits l
l signals and associated circuitry are safety grade, are redundant.and meet safety-grade requirements.
l If this connot be verified, the auxiliary system Refer to Section 7.3.6.
automatic initiation system should be modified in I
the short-term to meet the fuctional requirements listed below. For the longer term, the automatic initiation sygnals and circutis should be upgraded I
to meet safety-grade requirements as indicated in b
Recommendation GL-5.
f (1) ihe design should provide for the automatic initiation of the auxiliary feedwater system i
4 iAlow.
l (2) The automatic initiation signals and circuits should be designed so that a single failure i
will not result in the loss of auxiliary I
feedwater system function.
i i
I j
I
4 1
SNUPPS 1
TABLR 10.4-138 (Sheet 5)
DESIGN COMPARISONS TO NRC RECOMMENDATIONS 2
ON AUXILIARY FEEDWATER SYSTEMS CONTAINED IN THE MARCH 10, 1980 NRC LETTER f
l 7 (con't) l (3) Testability of the initiation signal and circuits shall be a feature of the design.
1 (4) The initiation signals and circuits should be powered from the emergency buses.
I (5) Man 2al capability initiate the auxiliary feedwater system from the control room should l
be implemented so that a single failure in the i
manual circuits will not result in the loss of system function, j
(6) The alternating current motor-driven pumps and valves in the auxiliary feedwater system should be in61oded in the automatic actuation (simul-i taneous and/or sequential) of the loads to the f
emergency buses.
)
(7) The autoestic initiation signals and circuits shall be designed so that their failure will not result in the loss of manual capability I]
to initiate the auxiliary feedwater system j
from the control room.
i Recommendation CS The licensee should install a See response to CS-7 above, j
8.
system to automatically initiate auxiliary feedwater system flow. This system need not be safety-grade; however, in the shcrt-term, it should meet the criteria listed below, which are similar to item j
2.1.7.a of NUREC-0578. For the longer term, the automatic initiation _ signals and circuits should he upgraded to meet d&fety-grade requirements, as i
indicated in Recommendation CL-2.
(
5 i
L
+
N i
1 SNUPPS 1
TABLE 10.4-138 (Sheet 6)
DESIGN COMPARISONS TO NRC RECOMNENDATIONS 1
ON AUXILIARY FEEDWATER SYSTEMS CONTAINED l
IN T!!E MARCH 10, 1980 NRC LETTER t
8 (con't) l
(!) The design should provide for the automatic initiation of the auxiliary feedwater system 4
1 flow.
(2) The automatic initiation signal and circuits l
should be designed so that a single failure will not result in the loss of auxiliary l
feedwater system function.
i i
(3) Testability of the initiating signals and circuits should be a feature of the design.
(4) The initiating signals and circuits should l
be pqwered from the emergency buses.
f (5) Manual capability to initiate the auxiliary feedwater system from the control room should be retained and shylj be implemented so that a single f ailure in the manual circuits will j
not result in the loss of system function.
(6) The alternating current powered motor-dirven t'
pumps and valves in the auxiliary feedwater system should be included in the automatic i
actuation (simultaneous and/or sequential)
I of the loads to the emergency buses.
(7) The automatic initiation signals and circuits should be designed so that their failure will l
not result in the loss of manual capability l
to initiate the auxiliary feedwater system l
from the control room.
B.
ADDITIONAL SHORT-TERM REC 099tENDAT10NS Recommen'dation - The licensee should provide re-The existing SNUPPS design provides the following l
C.ondant levei indication and low level alarms in redundant control room indication for condensate 1.
,,-e e,,se.e g
g
l 1
SNUPPS I
f' TABLE 10.4-138 (Sheet 7) i 1
l 1 (con't) l the control room for the auxiliary feedwater storage tank level.
system primary water supply, to allow the operator a) LI-4A shown on Fig. 9.2-12.
J to anticipate the need to make up water or transfer b)
PI-24A. PI-25A or PI-26A - Class IE auxiliary i
to an alternate water supply and prevent a low feedwater pump suction pressure indication pump suction pressure condition from occuring.
shown on Fig. 10.4-9.
j The low level alarm setpoint should allow at l
least 20 minutes for operator action, assuming Direct correlation between pump suction pressure and tank level is achieved by simple conversion. Exclusion that the largest capacity auxiliary feedwater of dynamic piping losses from the conversion results system pump is operating.
in a conservative determination of tank level.
Redundant control roce, tank level alarms are as follows:
a) LAHL-7 shown on Rig. 9.2-12.
b) LAL Class IE auxiliary feedwater pump low suction pressure aura shown on Fig. 10.4-9.
Setpoints for both alcras will allow at least 20 i
minutes for operator action assuming that the largest i
capacity auxiliary feedwater pump is or9 rating.
i l
2.
Reccassendation (This recommendation has been SNUPPS will perform a 48-hour, in situ endurance i
revised from the original recomumendation in test on all auxiliary feedwater pumps as part of the l
NUREG-0611)- The licensee should perform a 48-hour pre-operational test program.
l endurance test on all auxiliary feedwater system j
pumps, if such a test or continuous period of operation has not been accomplished to date.
i l
Following the 48-hour pump run, the pumps should j
be shut down and cooled down and then restarted
]
and run for one hour. Test acceptance criteria i
should include demonstrating that the pumps j
remain within design lisits with respect to bearing / bearing oil temperatures and vibration i
and that pump room ambient conditions (temper-ture, humidity) do not exceed environmental I
qualific'ation. limits for safety-related equip-
,1 ment in the room.
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a SNUPPS TABLE 10.4-13B (Sheet 9 )
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! (con't) i I
serving as backup to automatic auxiliary j
system initiation 2.
Recommendation CL Licensees with plant design The alternate water supply (essential service water) 1
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~
in which all (primary and alternate) water supplies connects to the auxiliary feedwater pump suction to the auxiliacy feedwater systems pass through piping downstream of the single, normally locked-valves in a single flow path should install re-open valve in a single flots path from the primary dundant parallel flow paths (piping and valves).
water source (condensate storage tank). Valves from the alternate supply automatically open on low pump Licensees with plants in which the primary aux-suction pressure. Refer to the response to CS-2 iliary feedwater system water supply passes through valves in a single flow path, but the alternate auxiliary feedwater system water supplies connect to the auxiliary feedwater system pump suction piping downstream of the above valve (s) from i
the alternate water supply upon low pump suction pressure.
The licensee should propose Technical Specifications j
to incorporate appropriate periodic inspections to verify the valve postions.
i 3.
Recommendation CL At least one auxiliary feed-The SNUPPS design meets this recommendation. Refer i
water system pumpf and its associated flow path and to the response to CS-5.
essential instrumentation should automat? ally initiate auxiliary feedwater system flow,nd be capable of being operated independently of any j
alternating current power source for at least two i
{
hours. Conversiono)ydirectcurrentpowerto alternating current power is acceptable.
4.
Recommendation CL Licensees having plants As discussed in the response to CS-4 and CL-2 above.
with unprotected normal auxiliary feedwater the SNUPPS design includes automatic transfer to the system supplies should evaluate the design of alternate water source. The alterrate source (essential their auxiliat s feedwater systems to determine service water) is protected form tornados and is f rom 4
j i
i l
SNUPPS I
TABLE 10.4-138 (Sheet 10) 1 10 (con't) seismic Category I.
if automatic protection of the pumps is necessary l
The time l
following a seismic event or a tornado.
j available before pump damage, the alarms and in-l dications available to the control room operator.
j and the time necessary for assessing the problem l
and taking action should be considered in deter-mining whether operator action can be relied on Consideration should be to prevent pump damage.
j given to providing pump protection by means such as automatic switchover of the pump suctions to the alternate safety-grade source of water.
i automatic pump trips on low suction pressure, or upgrading the normal source of water to meet seismic Category I and tornado, protection require-3 1
ments.
As stated in the response to CS-7. the auxiliary Recommendation CL The licensee should upgrade feedwater system automatic initiation signals and 5.
the auxiliary feedwater system automatic initiation l
circuits are safety grade, l
signals and circuits to meet safety-grade require-
- ments, j
i i
i i
i 4
h i
l 1
l 4
I TABLE 10.4-14 AUXILIARY 5TEDWATER SYSTEM INDICATING, ALARM, AND CONTROL DEVICES i
(
Control Room III Alarm Indication / Control Control Room Local j
l Condensate storage tank suction valve position X
X ESW suction valve position X
X M
i condensate storage tank suction header pressure X
X X
X Low pump suction pressure X
X X
Low pump discharge pressure Pump flow control valve X
X operation
(
Pump flow control valve X
X position Auxiliary feedwater flow
. X X
Auxiliary feedwater pump turbine trip & throttle X
X valve position Auxiliary feedwater pump X
X turbine speed f'
Auxiliary feedwater pump X
~
turbine low lube oil pressure l
Auxiliary feedwater pump turbine high lube oil X
temperature (1) Local control here means the auxi,lia y shutdown panel.
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~
ENCLOSURE B to SLNRC 81-39 Question 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 offiste 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.'evel to assure sufficient steam gen-erator heat transfer surface to remove decay heat and/or cool down the primary system.
Response 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 generators. As an Engineered Safeguards System, the Auxiliary Feedwater System is directly relied upon to pre-vent core damage and primary 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 l
plant transient.
i 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 The water level is maintained under these circumstances by the process.
feedwater system, or if the feedwater system is not operable, by the Auxiliary Feedwater System which delivers an emergency water supply to
Enclosure B Page 2 the steam generators. The Auxiliary Feedwater System must be cabable of functioning for extended periods, allowing time either to restore normal feedwater flow or to proceed with an orderly cooldown of the plant to conditions where the Residual Heat Removal System can be placed into operation for continued decay heat removal. The Auxiliary Feedwater System flow and the emergency war?r supply capacity are sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat during the plant cooldown. The Auxiliary Feedwater System can also maintain the steam generator water levels above the tubes following a LOCA.
In the latter function, the water head in the steam generators serves as a barrier 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 SNUPPS units.
Loss of Main Feedwater Transient Loss of main feedwater with offsite power available Station blackov 1.e., loss of main feedwater without offsite power available)
Secondary System Pipe Ruptures Feedline rupture Steamline rupture i
l Loss of all AC 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 l
in the feedwater or condensate system I
Loss of offsite power or blarkaut with the consequential shutdown of the system pumps, auxiliarie, 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 l
trip, and auxiliary fesec.tur actuation by the protection system logic.
l Following reactor trip from a high initial power level, the power quickly falls to decay heat levels. The water levels continue to decrease, pro-l gressively uncovering the steam generator tubes as decay heat is trans-l ferred and discharged in the form of steam either through the steam dump
Page 3 valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere. 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' pressurizer from filling to.a water solid i
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 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 f 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 power-operated relief valves. The calculated transient is similar for both the loss of main feedater and the blackout, 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.
Secondary System Pipe Ruptures The feedwater line rupture accident is postulated to result in the loss of feedwater flow to the steam generators and the complete blowdown of one steam generator within a short time, assuming the rupture to be downstream of the last nonreturn valve in the main or auxiliary feedwater piping to an individual steam generator. Another consequence of a feed-line rupture may be the spilling of auxiliary feedwater to the faulted steam generator. The system design allows for limiting the 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 i
to the remaining effective steam generator (s).
Main steamline rupture accident conditions are characterized initially by plant cooldown and, for breaks inside containment, by i;' reasing con-tainment pressure and temperature. Auxiliary feedwater is not needed during the early phase of the transient but flow to the faulted loop will contribute to an excessive release of mass and energy to contain-ment. Thus, steamline rupture conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop. Eventually, how-ever, the Reactor Coolant System will heat up again and auxiliary feed-water flow is required to be delivered to the non-faulted loops, but at somewhat lower rates than for the loss of feedwater transients described previously. Provisions are made in the design of the Auxiliary Feedwater System to limit the auxiliary feedwater flow to the faulted loop as necessary in order to prevent containment overpressurization following a steamline break inside containment, and to ensure the minimum flow to the remaining unfaulted loops.
i 2
l
.,1
Enclosure B Page 4 Loss of All AC Power T'.e loss of all AC power is postulated as resulting from accident con-ditions wherein all onsite and offsite AC power supplies are lost.
Battery power for operation of protection circuits is assumed available.
The a turbine driven auxiliary feedwater system pump power and control sources are not dependent on AC~ power and are capable of maintaining the plant at hot shutdown until AC power is restored.
Loss-of-Coolant Accident (LOCA)
Loss of coolant accidents do not impose any flow requirements on the auxiliary feedwater system beyond those required by the other accidents addressed in this response. The following description of the small LOCA is provided 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 con-tribution from the Auxiliary Feedwater System following such small LOCAs is 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 capabil-ity for providing a buoyancy head for natural circulation.
The auxiliary feedwater system may be utilized to assist in a system cooldown and de-pressurization following a small LOCA while bringing the reactor to a cold shutdown condition.
Cooldown The Auxiliary Feedwater System is a safety-related system that may be used in conjunction with the steam generators and atmospheric relief valves to reduce the reactor coolant system temperature from normal l
zero load temperatures to a hot leg temperature of approximately 3500F, l
under conditions when offsite power is not available. The RHR system l
l completes the cooldown to cold shutdown conditions.
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 1.a, above. Spec-ific 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 a safety-related heat removal capability following reactor trip. Other plant protection systems are designed to meet short term or pre-trip fuel failure criteria. The effects of excessive coolant shrinkage are eval-uated 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.
Enclosure B Page 5 Question 2 Describe the analyses and assumptions and corresponding technical justi-fication used with plant conditions considered in 1.a above including:
a.
Maximum reactor power (iiicluding instrument errors allowance) at the time of the initiating transient or accident.
b.
Time delay from initiating event to reactor trip.
I l
c.
Plant parameter (s) which initiates AFWS flow and time delay between initiating event and introduction of AFWS flow into steam generator (s).
d.
Minimum steam generator water level when initiating event occurs.
e.
Initial steam generator water inventory and depletion rate before and after AFWS flow comences -- 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.
l 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 cir-culation.
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 maintain minimum steam generator water level. Also identify credit taken for primary system heat removal due to blowdown.
k.
Volume and maximum temperature of water in main feed lines between steam generator (s) and AFWS connection to main feed line.
1.
Operating condition of steam generator normal blowdown following initiating event.
l Primary and secondary system water and metal sensible heat used for I
m.
I cooldown and AFW flow sizing.
I Time at hot standby and time to cooldown RCS to RHR system cut in n.
l temperature to size AFW water source inventory.
Response to 2 l
'nalyses have been performed for the limiting transients which define the l
AFWS performance requirements. These analyses are described in the FSAR and below:
l l
Enclosure B Page 6 Loss of Main Feedwater (Blackout)
A loss of feedwater, assuming a loss of power to the reactor coolant pumps, is described in FSAR Section 15.2.6.
It is shown that for a station blackout transient the peak RC3 pressure remains below the criterion for Condition II transients and no fuel failures occur (refer to Table 18-1). Table 2-1 summarizes the assumptions used in this anal-ysis. The analysis assumes that the plant is initially operating at 102% (calorimetric error) of the Engineered Safeguards Design (ES0) 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 water level, allowing for level un-certainty. As shown in the FSAR, there is a considerable margin with respect to filling the pressurizer for a loss of normal feedwater tran-sient with or without power to the reactor coolant pumps.
Rupture of Main Feedwater Pipe The double ended rupture cf a main feedwater pipe downstream of the main feedwater line check valve is analyzed in FSAR, Section 15.2.8.
Table 2-1 summarizes the assumpt & 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 conKrvative assumption 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 cor,servatively 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. Auxiliary feedwater flow of 563 gpm was assumed to be delivered to the three non-faulted steam generators. The criteria listed in Table 18-1 are met.
4 The analysis establishes the capacity of single pumps, establishes re-quirements for layout to preclude indefinite loss of auxiliary feedwater to the postulated break., and establishes train association requirements for equipment so th e the AFWS can deliver the minimum flow required in 1 minute assum'eng the worst single failure.
Rupture of I hain Steam Inside Containment Because the steamline break transient is a cooldown, the AFWS is not needed to remove heat in the short term. Furthermore, addition of excessive auxiliary feedwater to the faulted steam generator will affect the peak containment pressure following a steamline break inside con-tainment. This transient is performed at fiva power levels for Saveral break sizes. Auxiliary faedwater is assumed to be initbte
- che time of the break, independent of system actuation signhh.
ihe maximum flow is used for this analysis. Table 2-1 summarizes the assumptions used in this analysis. At 30 minutes after the break, it is assumed that the operator has isolated the AFWS from the f aulted steam generator which subsequently blows down to ambient pressure. The criteria stated in Table 18-1 are met.
Enclosure B Page 7 This transient establishes the maximum allowable auxiliary feedwater flow rate to a single faultea 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 operation defines the basis for minimum required condensate storage tank level based on the required cooldown duration, maximum decay heat input and l
maximum stored heat in the system. Table 2-1 shows the assumptions used to determine the cooldown helt capacity of the auxiliary feedwater system.
The cooldown is assumed to cc:nmence at the maximum cated 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 AFWS. See Table 2-2 for the items constituting the sensible heat stored in the NSSS.
l l
l l
I l
Enc'iosure B Page 8 Question 3 Verify that the AFW pumps in your plant will supply the necessary flow to the steam generator (s) at-determined by items 1 and 2 above consider-ing a single failure.
Identify the margin in sizing the pump flow to allow the pump recirculation flow, seal leakage and pump wear.
Response to 3 The auxiliary feedwater flow rates identified in items 1 and 2 above represent the minimum expected for analysis purposes and are based on the limiting single failure for each event.
The sizing of the pumps includes an allowance fcr the manufacturer's recommended recirculation flow.
Seal (i.e. pocking) leakage is expected to be minimal and will be verified to be within acceptable limits during periodic testing. Further, since the periodic testing is the only normal operating time for these pumps, flow degradation due to pump wear is expected to be negligible.
RLS/dck/lb21
. ~
4 TABLE 18-1
~
CRITERIA FOR AUXILIARY FEEDWATER SYSTEM DESIGN BASIS CONDITIONS Condition Additional Design Transient Classification
- Criteria Criteria or Loss of Main Feedwater Condition II Peak RCS pressure not to exceed design pressure +101.
l No consequential fuel failures Station Blackout Condition II (same as LMFW)
Pressurizer does not fill l
Feedline Rupture Condition IV 10CFR100 dose limits.
Core does not uncover Containment design pressure not exceeded Same as blackout assuming Loss of all A/C Power N/A Note 1 Containment design pressure turbine driven pump m t exceeded 1
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 1000F/hr l
Cooldown N/A 5570F to 3500F
- ANSI N18.2 (This information provided for.those transients performed in the FSAR).
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 Note 1:
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TABLE 2-2
SUMMARY
OF SENSIBLF HEAT SOURCES Frimary Water Sources (initially at rated power temperature and inventory)
RCS fluid Pressurizer fluid (liquid and vapor)
Primary Metal Sources (initially at rated oower temperature)
Reactor coolant piping, pumps and reactor vessel l
Pressurizer Steam generator tube metal and tube sheet Steam generator metal below tube sheet Reactor vessel internals Secondary) Water Sources (initially at rated power temperatu inventory Steam generator fluid (liquid and vapor)
Main feedwater purge fluid between steam generator and AFWS piping.
l Secondary Metal Sources (initially at rated power temperature)
All steam generator metal above tube sheet, excluding tubes.
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