ML20003B488
| ML20003B488 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 02/04/1981 |
| From: | NORTHERN STATES POWER CO. |
| To: | |
| Shared Package | |
| ML20003B484 | List: |
| References | |
| NUDOCS 8102120239 | |
| Download: ML20003B488 (40) | |
Text
__ ___ . __
Dirseto'r of NRR h February 4, 1981 Attachment 1 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Auxiliary Feedwater Pump Endurance Test 1
A 48-hour endurance test has been conducted on each of Prairie Island's four auxiliary feedwater pumps. Each pump performed satisf actorily during these tests. In the tests, water was initially pumped to the steam generators. Then as the pump flow was decreased the water was recirculated back to the co % ntate storage tanks. For both flow paths the pump's discharge pressure exceeded the shutdown pressure of the steam generators.
The bearing temperatures were monitored directly from thermocouples in the bearings (for exact location see Figures 1 and 2). These thermocouples are monitored by the plant's process computer. The pump's discharge pressure was also monitored by the computer. During the initial phase the flow to the steam generators also was recorded from meters for flow to each ,
steam generator on the control room auxiliary feedwater control board.
During the recirculation phase a local flowmeter was used to determine pump flowrates. The approximate' location of each sensor is shown on Figure
- 3. The humidity of the auxiliary feedwater pump room was not directly measured because the plant did not have the required instrumentation.
However, personnel conducting this test observed no change in the humidity of the room. This was checked during their hourly inspections. The
, temperature was recorded using a local thermemeter positioned near the pump tested (Figures 4 and 5).
The auxiliary feedwater pump flow was varied to conform with the decay heat removal characteristics of the reactor coolant system. h is was done to minimize the amount of demineralized water used and to perform a more realistic test. With the barrel type pump, low flow conditions can cause boiling of the liquid in the pump casing and vapor binding of the pums and is of much more concern than maintaining rated flow. The initial pump flowrate was 200 gpm, the rated flowrate of pump, and the water was pumped to the steam generators. After one hour the flowrate was decreased to 150 gym. This flow was held for seven hours before the flow of the pump was directed to recirculate back to the condensate storage tanks. When the pump was on recirculation the flowrate was lowered to 100 gpm, where it remained for the rest of the test.
Following the 48-hour test, the pumps were shut down and temperatures were allowed to approach ambient conditions. The pumps were then successfully restarted and run for an additional hour at *00 gpm. .
The plots of the pump and driver bearing temperature and room ambient temperature are attachad f. Figures 6 through 13). The difference in tempera-tures between the pumps is mostly due to the time of year when the tests were run. The turbine-driven pumps were tested in May when the ambient and cooling water temperatures were higher than they were in December when 21 motor-driven pump was tested. 12 motor driven pump was tested in August and the ef fect of the higher ambient and cooliq water temperature can easily be seen.
l l 1-1 8102110 1 l
The vibration measured at the coupling did not exceed 0.192 inch /second for the motor-driven and 0.314 inches /second for the turbine-driven pumps. These values are considered satisfactory.
It is concluded that all four auxiliary feedwater pumps performed satisf ac-torily during the test.
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Director of NRR February 4, 1981 Attachment 2 PRAIRIE ISI.AND NUCLEAR GENERATING Pl. ANT Turbine Driven Auxiliary Feedwater Pump AC Independence Design Change (80Y129) A. C. independence will be achieved by replacing existing motor operated valve 32264 (32265) with an air actuated control valve. In addition, the lube oil pump will be started periodically to maintain an oil film on all surfaces. Specifically this task will consist of the following:
- 1. Removing existing motor operated control valve and associated wire Connections.
- 2. Installation af a 3", f ail open, direct acting stainless steel globe valve. The air to this valve would be controlled with a D.C.
powered solenoid valve.
- 3. Reconnect 3/4" steam condensate line from existing valve upstream of new air valve.
- 4. Installation of new air line tapped from existing instrument air header.
- 5. Installation of an air receiver and double check valves in the supply air line to the air valve. This receiver will be sized so that enough air pressure is available to close this valve a minimum *of two times.
- 6. Installation of a timer to run the lube oil pump a minimum of once a day for five minutes. In addition, an annunciator will be installed in the main control board to alarm if the oil pressure is not reached after the oil pump start.
- 7. The controls in the Control Room and on ene Hot Shutdown Panel that operated the existing motor control valve would be used to operate the replacing air actuated valve. In addition, the indication in the Control Room would be modified such that the new air operated valve must be in the open position for the red light to be on and must be in the closed position for the green light to be on.
1 i 2-1 i
PROJECT ANALYSIS I. MECHANICAL 1.0 Scope For each unit, remove existing motor operated steam valve for the turbine driven auxiliary feedwater pump and replace with 3" air operated globe valve. In addition, an air receiver tank, a D.C. olenoid valve, and check valves are included in the air line to provide limited operation should the instrument air supply be unavailable. 2.0 Design and 0A Classifiction All components of this change shall be safety related Quality Assurance Type II and Design Class II except for the air operated valve and the D.C. solenoid valve which shall be Quality Assurance Type I and Design Class I. 3.0 Applicability of ANSI N45.2/10 CFR SO Appendix B Both ANSI N45.2 (all l'8 elements) and 10 CFR 50, Appendix B are applicable to both valves. 4.0 Procedural Coverage Fluor Power Services is using the Project Administration Manual as well as the Operating Instructions for Northern States Power Company and the Prairie Island Nuclear Generating Plant for this work. 5.0 Codes and Standards The motors shall conform to the following: (1) NEMA, MG-1 (2) IEEE Standard 323-1974 (3) IEEE Standard 344-1975 l (4) ANSI N45.2-1977 (5) NRC Regulation 10 CFR 50, Appendix B (6) ASME Section VIII II. ELECTRICAL 1.0 Scope ! DC control circuits will be established to operate the new air actuated valves. The logic for the new valves will be the same as it was for the motor valves with the following exceptions. l a. On loss of de control power or on less of air the l valve will assume the open position. 2-2 i
.]
- b. The permissive that required lube oil pressure before starting will be removed.
The logic for the lube oil pump will be revised as follows:
- a. An automatic timer will run the lube oli pump once a day for approximately five minutes.
- b. If during one of the above runs the lube oil does not reach the required pressure an alarm will alert the control room.
f
- c. The remaining logic will remain the same as before, i
i.e. start the pump when the steam valve receives the signal to open and also when the valve is open and the oil pressure is low. (Normally the shaf t drive oil pump will maintain suf ficient oil pressure.) ! 2.0 Design Classification Class IE Electrical Eq'uipment, except for timer on lube oil pump, which will be Class III. l 3.0 Applicability of ANSI /10CFR 50 i Applicable 4.0 Procedural Coverage t i Fluor Power Services is using the Project Administration Manual (PAM) as well as the QA Manual. 5.0 Codes and Standards
- 1. Institute of Electrical and Electronic Engineers (IEEE)
- 2. Insulated Cable Engineera Association (ICEA)
- 3. National Electric Manuf acturers Associatf ?n (NEMA) 4 American National Standarda Institute (ANSI)
- 5. American Society for Testing and Materials (ASTM)
- 6. IEEE 323-74 IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations
- 7. IEEE 383-74 IEEE Standard for Type Test of Class IE
- 8. Electrical Cables, Field Splices and Connections for Nuclear Power Cenerating Station 0 l
i i l 2-3 l l l
SAFETY EVALUATION REPORT (MECHANICAL) The turbine driven auxiliary feedwater pump is a component of a safeguard system. Since this modification involves a safeguard system, safety concerns were addressed with respect to the protection of the system and maintain-ing the pump's intended function upon loss of A.C. power. Specifically the following were identified as safety concerns with respect to this modification: A. Proper bearing lube oil pressure prior to and during start-up of the turbine driven auxiliary feedwater pump when A.C. power is unavailable. B. Heat transfer effect, feedwater level, and thermal expansion as a ) result of unplanned pump starts. C. Pressure drop across new valve and it's effect on the auxiliary feedwater pump turbine operation. D. Proper fabrication and erection of this modification. ! The following actions were taken to assure these concerns would not compromise the safety functions of the e.txiliary feedwater pump turbine: A. Time controlled operation of the A.C. driven lube oil pump for a minimum of once a week for five minutes eliminates the requirement for bearing lube oil pressure prior to start-up of the turbine driven auxiliary feedwater pump (Letter from Lloyd D Hanson - Pacific Pumps - dated 07/08/80 verifying this procedure). , B. This modification does not in-validate the FSAR safety evaluation report pertaining to heat transfer effects, feedwater level, and thermal expansion. C. Pressure drop across new valve is less than the pressure drop across the existing valve and causes no significant af fect on turbine inlet pressure. D. Piping loads for the modification were calculated, checked and reviewed. Pipe supports for which loading increased have been reviewed and modified where required. E. The piping f abrication and erection specification (HIA 10 6B) was revised to incorporate current codes and standards, and all contractor procedures were reviewed for compliance. F. Calculations were performed for stresses due to the dif ferential thermal expansion between the stainless steel valve and the carbon steel piping. Stress levels as a result of this were found to be acceptable within the limits found in the FSAR. G. The modification was reviewed with respect to the pipe rupture analysis using the guidelines set forth in the FSAR. It was determined that the present pipe rupture analysis is acceptable with ou t revision. 2-4
SAFETY EVALUATION REPORT (CONT. ) (ELECTRICAL) The electrical modifications to the turbine driven auxiliary feedwater pump consist mainly of changing the ac control circuitry of existing motor valve ta Ac control circuitry for the new solenoid operated valve. The control circuitry ot the lube oil pump was modified only to the extent of removing the permissive which allowed the new main steam supply valve to open. Specifically the following were identified as electrical safety concerns with respect to this modification: A. All electrical components and materials with the exception of the lube oil pump timer were classified as type IE. The lube oil pump timer was not classified type IE since the requirement for operation of the lube oil pump is only once a week and even a malfunction of th is t ime , which would operate once a day, would not prevent the turbine driven pump from running. B. All Class IE electrical components will be purchased, invoking the < latest IEEE standards in their manufacture. C. All new conduits being routed through Class I areas will be installed in accordance with Class I criteria. D. Separation will be maintained in incorporating the change. No. 11 Turbine Driven Aux. Feedwater pump was fed f rom Train A ac circuits and will be controlled f rom Train A de circuits. Likewise No. 22 Turbine Driven Aux Feedwater pump was and will be controlled f rom Train B. E. The de circuit draw on the battery will be less than one ampere. This has a negligible ef fect on the station batcery. 2-5
Director of NRR February 4, 1981 At t achment 3 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Auxiliary Feedwater Pump Suction Pressure Protection Design Change Design Change 80L579 will provide for trip of the Auxiliary Feedwater pumps upon loss of pump suction in accordance with an NRC letter of October 16, 1979 "NRC Requirements for Auxiliary Feedwater Systems at Prairie Island Nuclear Genersting Plant, Units I & II". This change will provide loss-of-suction protection for 11, 12, 21, & 22 Auxiliary Feedwater Pumps. Pump trip, with control room alarm and manual l reset, will occur if pump discharge pressure has not reached 850 psig within j 10 seconds after the AFW turbine steam inlet valve open limit switch actuates for a turbine-driven pump start, and after the oil pressure permissive start actuates for a motor-driven pump start. Trip and alarm I will also occur immediately if the pump discharge pressure drops below 850 psig during continuous operation. ll Setpoint Basis 4 I Auxiliary Feedwater Pump minimum ditierential head at maximum flow is 887 psig at 320 gpm. At flows greater than 320 gpa pump discharge pressure becomes erratic and decreases indicating unstable pump operation. Loss of pump suction will cause pump discharge pressure to decrease to
! zero. An 850 psig setpoint will protect the pump from loss-of-suction i destruction without limiting full flow conditions.
The elapsed time from start of the auxiliary lube oil pump until discharge
- presrire reaches 1750 psig is 6(+1) seconds for 21 AFW pump. Starting time j for a turbine-driven pump is somewhat less. A 10 second time delay on l
start will allow suf ficient time for the pump to reach rated RPM. No pump damage will occur if the pump is operated for 10 seconds with no suction pressure. l
; Hardware Design Considerations I
j The hardware for the control circuit will meet original specifications for safety-related components (Class IE). For the 11 & 22 turbine driven pumps 4 all control circuitry will be installed to operate on 125 volts DC as part of Design Change 80Y129 and will be Class IE, qualified in accordance with
! IEEE Standards 323, 1974 and 344.
i l Tubing between pump discharge piping and the pressure switch will be 3/8"
- 316 stainicss, .065" wall, and will be supported using existing tubing runs and supports.
I-3-1
4 Director of NRR February 4, 1981 j Attachment 4 i 1 d 4 t 4 1 PRAIRIE ISIAND NUCLEAR GENERATING PTAVf l Basis for Auxiliary Feedwater System Flow Requirements r (Response to Enclosure 2 of NRC letter of 10/16/79) 1 Prepared by Westieghouse Electric Corporation for Northern States Power Company 4-1 m .rr., - - , - - --,g , - - . - -.---y e , - . - ---,w---,,. - ~ ~ - - - , , a .- a - -- - --
Question 1
- a. Identify the plant transient and accident conditions considered in establishing AFWS ficw requirements, including the follcwing 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 stern isolation valve closure
- 7) Main feed line break
- 8) Main steam line break
- 9) Snall 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 sucn as:
- 1) Maximum RCS pressure (PORV or safety valve actuation)
- 2) Fuel temperature er drnage limits (DNB, PCT, maximum fuel central temperature)
- 3) RCS cooling rate limit to avoid excessive coolant shrinkage
- 4) Minimum steam generator level to assure sufficient stern gen-erator heat transfer surf ace 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 sice of the stern generators at times when the feedwater system is not available, thereby maintaining the heat sink
- capabilities of the stern generator. As an Engineered Safeguards Sys-tem, the Auxiliary Feedwater System is directly relied uoan 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 cooldcwn 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 tne l condensers through the steam dumo or to the atmospnere througn the steam generatcr safety valves or the power-operated relief valves. Steam ! generator water inventory must be maintained at a level sufficient to ensure adequate heat transfer and continuation of the decay heat removal process. The water level is maintained under these circumstances by the Auxiliary Feedwater System which delivers an emergency water supply to the steam generators. The Auxiliary Feedwater System must be capable of l functioning for extended periode s silawing time either to restore normal ' feedwater ficw or to proceed _th bi cederly cooldown of the plant to l 4-2
the reactor coolant temperature where the Residual Heat Removal System can assume tne burden of decay heat removal. The Auxiliary Feedwater System flow and the emergency water sucoly capacity must be sufficient to remove core decay heat, reactor coolant pump heat, and sensible heat durin; the plant cooldown. The Auxiliary Feedwater System can also ce used to maintain tne steam generator water levels above the tuces fol-lowing a LOCA. In the latter function, the water head in the stern generators serves as a barrier to prevent leakage of fission procucts 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 Prairie Island plants.
- 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 loss of Main Feedwater Transients The design loss of main feedwater transients are those caused by:
- Interruptions of the Main Feedwater System flow due to a malfunction in the feedwater or condensate system
- Loss of offsite power or blackout with the consequential shutdown of the system pumps, auxiliaries, and controls Loss of main feedwater transients are characterized by a rapid reduction in steam generator water levels which results in a reactor trip, a tur-bine trip, and auxiliary feedwater actuation by the protection system logic. Following reactor trip from high power, the power quickly f alls to decay heat levels. The water levels continue to decrease, progres-sively uncovering the steam generator tubes as decay heat is transferred and discharged in the form of steam either througn the steam duma valves to the condenser or through the steam generator safety or power-operated relief valves to the atmosphere. The reactor coolant temperature 4-3
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t 9 increases as the residual heat in excess of that dissipated througn the steam generators is aosorbed. With increased temoerature, the volume of reactor coolant expands and begins filling the cressurizer. Without the addition of sufficient auxiliary feedwater, further exoansion sill result in water being discharged througn the pressurizer safety anc relief valves. If the temperature rise and the resulting volumetric expansion of the primary coolant are permitted to continue, then (1) pressurizer safety valve capacities may be exceeded causing over-pressurization of the Reactor Coolant System and/or (2) the continuing loss of fluid fran the primary coolant system may result in bulk boiling in the Reactor Coolant System and eventually in core uncovering, loss of natural circulation, and core damage. If such a situation sere ever to occur, the Emergency Core Cooling System would be ineffectual because the primary coolant system pressure exceeds the shutoff head of the safety injection pumps, the nitrogen over-pressure in the accumulator tanks, and the design pressure of the Residual Heat Removal Loop. Hence, the timely introduction of sufficient auxiliary feeddater is necessary to arrest the decrease in the steam generator water levels, to reverse the rise in r? actor coolant temperature, to prevent the pres-surizer frem fillir3 to a water solid condition, and eventually to establish stable hot stanoby conditions. Subsequently, a decision may be made to proceed with plant cooldown if the oroblem cannot be satis-factorily corrected. The blackout transient differs from a simple loss of main feedwater in that emergency power sources must be relied upon to operate vital equio-ment. The loss of power to the electric driven condenser circulating water pumps results in a loss of condenser vacuum and condenser dumo 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 would be similar for both the loss of main feedwater 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. i The station blackout transient serves as the basis for the minimum ficw l required for the smallest caoacity single auxiliary feeddater pumo for the Prairie Island p1 ants. The pump is sized so that any single pump i will provide sufficient flow against the steam generator safety valve i set pressure (with 3% accumulation) to prevent water relief fecm the pressurizer. Secondarv System Pioe Ruotures The feedwater line rupture accident not only results in the loss of feedwater flow to the steam generators but also resu cs in the comolete l 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 aux-iliary feedwater out the break as a consequence of the fact that the 1 4-4 l
I l auxiliary feedwater brancn line may be connected to the main feedwater line in the region of tne postuiated break. Such situations can result in the spilling of a disproportionately large fraction of the total auxiliary feedwater flow because the system preferentially pumps water to the lowest pressure region in the f aulted loop rather than to the effective steam generator whicn is at a relatively nign pressure. The system design must allow for terminating, limiting, or minimizing that fraction of auxiliary feedwater flow wnich 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. 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. Auxiliary feedwater is not needed during the early phase of the transient but flow to the f aulted loop will contribute to the release of mass and energy to containment. Thus, steamline ructure conditions establish the upper limit on auxiliary feedwater flow delivered to a faulted loop. Eventually, however, the Reactor Coolant System will heat up again and auxiliary feedwater flow will be required to be delivered to the unf aulted loop, but at somewhat lower rates than for the loss of feedwater transients described pre-viously. Provisions must be made in the design of the Auxiliary Feed-water System to allow limitation, control, or termination of tne auxil-iary feedwater f'ow to the f aulted loop as necessary in order to prevent containment overpressurization following a steamline break insi;fe 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 con-ditions wherein not only onsite and offsite AC power is lost but also AC Sattery emergency power is lost as an assumed cormon mode failure. power for operation of protection circuits is assumed available. The impact on the Auxiliary Feedwater System is the necessity for 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. l l 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 otner accidents addressed in this response. The following description of the small LOCA is provided nere for the sake of completeness to explain the role of the auxiliary feedwater system in this transient. l Small LOCA's are characterized by relatively slow rates of decrease in l reactor coolant system pressure and liquid volume. The principal con-tribution from the Auxiliary Feedwater System following such small LOCAs is basically the same as the system's function during hot shutdown or 4-5
; following spurious safety injection signal which trips the reactor.
Maintaining a water level inventory in the secondary side of the steam generators provices a heat sink for renoving decay heat and estaclishes the capability for providing a buoyancy head for natural circulation. The auxiliary feedwater system may be utilized to assist in a system cooldown and depressurization following a small LOCA while bringing the reactor to a cold shutdown condition. Cooldown The cooldown function performed by the Auxiliary Feedwater System is a partial one since the reactor coolant system is reduced from nonnal zero load temperatures to a hot leg temperature of acoroximately 3500F. The latter is the maximum temperature recommended for placing the Resi-dual Heat Removal System (RHRS) into service. The RHR system completes the cooldown to cold shutdown conditions. Cooldown may be required following expected transients, following an accident such as a main feeoline 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 AFWS is capable of delivering sufficient AFW to remove decay neat 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 auxiliary feedwater pumos, 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 reac-tor core.
4-6
Resconse to 1.b Table 13-1 summarizes the criteria wnich are the general design bases for eacn event, discussed in the response to Question 1.a, aoove. Specific assumptions used in the analyses to verify that tne design bases are met are discussed in response to question 2. The primary function of tne Auxiliary Feedwater System is to provide sufficient neat removal capability for heatup accidents following reac-tor trip to remove the decay heat generated by the core and prevent system overpressurization. Other plant protection systems are designed to meet short term or pre-trip fuel f ailure criteria. The effects of excessive coolant shrinkage are bounded by the analysis of the rupture of a main steam pipe transient. The maximum flow requirements deter-mined by other bases are incorporated into this analysis, resulting in no additional flow requirements. t l l l 4-7 I i l _ _
TABLE 18-1 Criteria for Auxiliary Feedwater System Design Basis Conditions Condition or Additional Design Transient Classification
- Criteria
- Criteria Loss of Main Feedwater Condition 11 Peak RCS pressure not to exceed design pressure. No consequential fuel failures Station Blackout Condition II (same as LHFW) Pressurizer does not fill with I single motor driven aux.
feed pump feeding 1 SG, Feedline Rupture ~ Condition IV 10 CFR 100 dose limits. Core does not uncover c- Containment design pressure 5 not exceeded Steamline Rupture Condition IV 10 CFR 100 dose limits-Contalmnent design pressure not exceeded. Loss of all A/C Power N/A 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 N/A 100 F/hr 5470F to 3500F
*Ref: ANSI N18.2 (This infonnation 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 f ailures must be assuned to postulate this transient.
Question 2 Describe the analyses and assumptions and corresponding technical justi-fication used with eacn plant ccndition considered in 1.a aoove includ-ing:
- a. Maximum reactor power (including instrument error allowance) st the time of the initiati.ng transient or accioent.
- 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 genera-tor (s).
- d. Minimum steam generator water level when initiating event occurs.
- e. . Initial steam generator water inventory and depletion rate oefore and after 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.
- i. 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 temoerature 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.
- m. Primary and secondary system water and metal sensible heat used for cooldown and AFW flow sizing.
- n. Time at hot standby and time to cooldown RCS to RHR system cut in temperature to size AFW water source inventory.
4-9
Resconse to 2 Analyses have been perfcrmed for tne limiting transients anich define the AF'WS perfccmance requirements. These analyses nave seen proviced for review and have been approved in tne Applicant's FSAR. Specifi-cally, tney include: Loss of Main Feedwater (Station 31ackcut)
- Rupture of a Main Feecwater Pipe - Rupture of a Main Steam Pipe Inside Containment In addition to the aoove analyses, calculations have been performed specifically for the Prairie Island plants to determine tne plant cool-down flow (storage caoacity) requirements. The Loss of All AC Power is evaluated via a comoarison to tne transient results of a Blackout, assuming an availaole auxiliary pumo having a diverse (non-AC) pcwer supply. The LCCA analysis, as discussed in response 1.0, incorporates the system ficws requirements as defined by other transients, and there-fore is not performed for the purpose of specifying AF'4S ficw recuire-ments. Eacn of the analyses listed above are explained in furtner detail in the following sections of this response.
Loss of Main Feedwater (Blackcut) A loss of feedwater, assring a loss of power to the reactor coolant pumps, was performed in FSAR Section 14.1.10 for the purpcse of showing that for a station blackout transient, a single motor criven auxiliary feedwater pump delivering flow to one steam generator does not result in filling tne pressurizer. Furthermore, the peak RCS pressure remains below the critericn for Condition II transierts and no fuel f ailures occur (refer to Taole 13-1). Table 2-1 sumnarizes the assumptions useo in this analysis. The transient analysis begins at the time of reactor trip. This can be done because tne trip cccurs on a steam generator level signal, hence the core power, temperatures ano steam generator level at time of reactor trip do not depend on the event sequence pricr to trip. Although the time from the loss of feedwater until the reactor trip occurs cannot be determined frem this analysis, this delay is expected to be 20-30 seconds. 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 conservative assumption in cefining decay neat and stored energy in the RCS. The reactor is assumed to be tripped on Icw-low steam generator level. Steam generater level at the time of reactor trip was assumed to be 0% NRS for additional conservatism; to that, allowance for level uncertainty was also accounted for. The FSAR shows tnat tnere is a considerable margin with respect to filling the pressurizer. This analysis establishes the capacity of the smallest single pump and also establishes train association of equipment so that this analysis remains valid assuming the most limiting single f ailure. l 4-10 b
Ruoture of Main Feedwater Pice The double ended rupture of a main feedwater pipe downstream of the main feedwater line check valve is not analyzed in the FSAR. Tacle 2-1 Sum-marizes the assumptions used in the analyses performed for Prairie Island (Reference 1). Reactor trip is assumed to occar as a result of a safety injection signal based on low steam main pressure in either loop at 20 seconds into the transient. This conservative assumption maxi-mizes tne stored heat prior to reactor trip and minimizes the aoility of the steam generator to remove heat from the RCS following reacter 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. The analysis allows for 400 gpm auxiliary feedwater delivered to the intact loop within 10 minutes of the reactor trip (10 minutes for operator action to reroute flow paths and to start the auxiliary feedwater pumps). The criteria listed in Table 18-1 are met. This analysis may establish the capacity of single pumps, _ establishes requirements for layout to precluce indefinite loss af auxiliary feed-water to the postulated break, and establishes train association requirements for equipment so that the AFWS can deliver the minimum flow required in 10 minutes following operator actions assuming the worst single failure. Primary system heat removal due to blowcown is included in our analytical code model and is correctly simulated during the feed-line rupture analysisr. Ruoture of a Main Steam Pioe 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 f aulted steam generatcr will affect the peak containment pressure following a steamline break inside con-tainment. This transient is performed for several break sizes. Aux-iliary feedwater is assumed to be initiated at the time of tne break, independent of system actuation signals to provide the most conservative analysis with respect to containment pressure. Table 2-1 sumarizes the ' assumptions used in this analysis. The criteria stated in Table 13-1 are met. This transient establishes auxiliary feedsater flow rate to a single f aulted stear. generator assuming one pump operational and estaolisnes layout requirements so that the flow requirements may be met considering the worst single failure. Primary system heat removal due to blowdown is included in our analytical code model and is correctly simulated during the steamline rupture analysis. l Reference 1: "NSP/NRP Auxiliary Feedwater System Sizing." PIW-P-540/KW-P-571. Octooer, 1969. 4-11 l
Plant Cooldewn Maximum and minimum ficw recuirements frem the previously discussed transients meet the flow recuirements of clant coolcown. This opera-tion, hcwever, defines the basis fcr tankace size, based en the recuired cooldown duration, maximum decay heat inout and maximum stored heat in the system. As previcusly oiscussed in resocnse 1A, the auxiliary feec-water system partially cools the system to the point where the RHR5 may ccmplete the cooldewn, '.e., 350cF in the RCS. Table 2-1 shews the assumptions used to determine the cooldove heat capacity of the auxil-iary feecwater system. The cooldown is assumed to ccmmence at the maximum rated pcwer, and maximum trip delays and decay heat scurce 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 AF'WS. See Table 2-2 for the items ccnstituting the sensible heat stored in the NSSS. This coeration is analyzed to establish minimum tank size recuirements for auxiliary feedwater fluid source which are normally aligned. 4-12 f
TABLE 2-1 Stawaary of Assumptions U4ed in AfW5 Design Verification Analyses Loss of leedwater Main Stemeline Breat Transient { station blackout) Cooldown Main f eedline Break {cordalsment )
- a. Mas reactor power 102% of ESD rating 1683 W t 1021 of ESD rating 51 of 1650 MWt for iirst (102% of 1722 MWt) (102% of 1722 MWt) 100 sec. 11 thereafter
- b. Time delay from 1.5 sec 2 sec 20 sec 1.5 sec event to Ra trip 2 sec additional for i ,
rod movanent
- c. Af WS actuation sig- lo-lo SG level N/A Operator action / Assumed lasaediately.
nal/tlee delay for 1 minute 10 minutes after 0 sec (no delay) from AfWS flow reactor trip 301 design contalnment pressure and 1101 nuainal steam flow signal
- d. SG water level at 36.9 ft. N/A N/A level at hot aero power time of reactor trip 01 NR Span -
minus I loot. 39.875 ft. (24.611 NR Span).
- e. Initial SG Inventory 61,000 ths/5G 90,050 lha/56 92,500 limn /SG 154,000 Itan/5G 0* M Rate of change bef ore see figure 2 2 N/A see figure 2-1 N/a
& after AfW5 actuation i decay heat see figurs 2-3 N/A see figure 2-3 see f igure 2-3
- f. AfW pump design see cooldown 1133 psia see couldown see cooldown pressure i g. Mintaan # of SGs I of 2 N/A 1 of 2 I of 2 which must receive N W flow
- h. RC pump status Tripped 9 reactor trip Tripped Tripped 9 reactor trip All operating
, l. Maniaan Af W 1000f 1000F 1000f 1000f
! temperature J. Operator action / none assumed / N/A/ 10 min./ 10 min./
primary heat removal N/A N/A Yes Yes due to blowdown
- k. WW purge voltane/ temp. 100 f t3 /410.4 Stu/Ilm 100 f t3 / 100 f t 3/410.4 Btu /Itan 100 ft 3/404.6 Stu/lbe 432.0er
- 1. Narmal blowJown none assumed none assumed none assumed none assumed
- m. Sensible heat see cooldown Table 2-2 see coutdown see cooldown
- n. Time at standby / time see cooldown 2 hr/6 hr see cooldown see coulduwn to coutdown to RHR
- o. AfW flow rate 200 GPM - constant var iable 400 C#M at ten min- 200 GPM -
(min, requirement) utes from reactor trip 5946A ,
TABLE 2-2 Sunnary of senti:le Heat Scurces 3rimary Water Scurces (initia81y at ratec ;cwer temcerature anc inventcry)
- RCS fluic
- Pressuri:er fluid (liquid anc vaper)
Primary Metal Scurces (initially at rated power tem erature)
- Reactor c:olant pt:ing, pum s sna react:r vessel
- Pressurizer
- Stern generat:r tu e metal anc uce sneet
- Sterm generater metal ceicw tuce sneet
- Reactor vessel interr.als Sec:ncary Wa ar Scur:es (initially at ratec cwer tem:erature anc inventery)
- Steam generator fluid -(liquid anc vaccr)
- Main f eecwater curge fluic cetween stesa generat:r anc AFWS pi:ing.
Sec:ndary Metal Scurces (initially at rated :cwer tem:erature)
- All steam generatcr metal accve tuce sneet, exclucing tuces.
4 3 4-14
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