ML20249B039
| ML20249B039 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 06/17/1998 |
| From: | Jacob Zimmerman NRC (Affiliation Not Assigned) |
| To: | Dennis Morey SOUTHERN NUCLEAR OPERATING CO. |
| References | |
| TAC-MA1730, NUDOCS 9806220004 | |
| Download: ML20249B039 (13) | |
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UNITED STATES
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j NUCLEAR REGULATORY COMMISSION g'% * * * * *
- a WASHINGTON D.C. 30006 4001 June 17, 1998 Mr. D. N. Morey Vice President - Farley Project Southern Nuclear Operating Company, Inc.
Post Office Box 1295 Birmingham, Alabama 35201-1295
SUBJECT:
PROPOSED REVISION TO TECHNICAL SPECIFICATION BASES FOR JOSEPH M. FARLEY NUCLEAR PLANT, UNITS 1 AND 2 (TAC NOS. MA1730 AND MA1731)
Dear Mr. Morey:
By letter dated May 1,1998, Southern Nuclear Operating Company, Inc. (SNC) provided the U.S. Nuclear Regulatory Commission (NRC) staff revised pages for the Technical Specification (TS) Bases Sections 2.1.1 and 3/4.7.1.2, for the Joseph M. Farley Nuclear Plant, Units 1 and 2. The revision to the Bases section was the result of a safety evaluation performed by SNC that takes credit for the spent fuel pool boron amendment changes and incorporates the results of an analysis done to establish auxiliary flow requirements for a variety of plant I
l conditions.
As you are aware, the TS Bases are not part of the TSs as defined by 10 CFR 50.36. As such, changes to the TS Bases may be made in accordance with the provisions of 10 CFR 50.59.
l Should the proposed change involve an unreviewed safety question pursuant to 10 CFR 50.59(a)(2), or involve a change in the interpretation of implementa#n of the TS (i.e., constitute a TS change), then the proposed change is to be provided to the staff pursuant to the provisions of 10 CFR 50.59(c) and 10 CFR 50.90 for prior NRC review and approval.
For administrative purposes, the TS Bases change needs to be provided to the staff to enable all copies of the Farley TSs to be updated in a consistent and timely fashion, Ga 9806220004 980617 PDR ADOCK 05000348 P
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D. N. Mo' rey June 17, 1998 Enclosed is a copy of the Farley revised TS Bases pages B 2-1, B 2-2, and B 3/4 7-2 for Unit 1 and B 2-1, B 2-2, B 3/41-2, B 3/41-3, B 3/41-4, and B 3/4 7-2 for Unit 2. These have been-L
_ dated to correspond to the issue date of this letter.
Sincerely, ORIGINAL SIGNED BY:
Jacob 1. Zimmerman, Project Manager
. Project Directorate 11-2 Division of Reactor Projects - 1/ll.
Office of Nuclear Reactor Regulation
. Docket Nos. 50-348 and 50-364 DISTRIBUTION RCaldwell, Ril Docket File WBeckner
Enclosure:
Units 1 and 2 Revised PUBLIC ACRS
' Bases Section PDll-2 RF OGC G. Hill (4)
LPlisco,Ril cc w/ encl: See next page PSkinner, Rll JZwolinski OFFICE PDit M M PDil-2/LAl n PDil-2/p NAME-Mderman. 'LBerry. k( Hdrh DATE 4f//98-(N998
/198 '
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_ COPY.
M NO tNS NO '
'YES,b)'
! DOCUMENT NAME:. G:\\FARLEYtllAA1730. BAS OFFICIAL RECORD COPY T
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d D. N. Morey Enclosed is a copy of the Farley revised TS Bases pages B 2-1, B 2-2, and B 3/4 7-2 for Unit 1 and B 2-1, B 2-2, B 3/41-2, B 3/41-3, B 3/41-4, and B 3/4 7-2 for Unit 2. These have been dated to correspond to the issue date of this letter.
Sincerely,
^
J cob 1. Zim erman, Project Manager Project Directorate 11-2 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Docket Nos. 50-348 and 50-364
Enclosure:
Units 1 and 2 Revised Bases Section cc w/ encl: See next page
w Joseph M. Farley Nuclear' Plant cc:
Mr. R. D. Hill, Jr.
General Manager-Southern Nuclear Operating Company Post Office Box 470 -
Ashford, Alabama 36312 Mr. Mark Ajiuni, Licensing Manager Southem Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201-1295 Mr. M. Stanford Blanton Balch and Bingham Law Firm Post Office Box 306 1710 Sixth Avenue North Birmingham, Alabama 35201 Mr. J. D. Woodard Executive Vice President Southem Nuclear Operating Company Post Office Box 1295 Birmingham. Alabama 35201 State Health Officer Alabama Department of Public Health 434 Monroe Street Montgomery, Alabama 36130-1701 Chairman Houston County Commission Post Office Box 6406 Dothan, Alabama 36302 Regional Administrator, Region ll U.S. Nuclear Regubtory Commission Atlanta Federal Center 61 Forsyth Street, S.W., Suite 23T85 Atlanta, Georgia 30303 Resident inspector U.S. Nuclear Regulatory Commission 7388 N. State Highway 95 Columbia, Alabama 36319 L
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s 2.1 SAFETY LIMITS
-e BASES 2.1.1 REACTOR CORE The restrictions of this safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission.
products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within t%s nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.
DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through correlations which have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
The DNB thermal design criterion is that the probability of DNB not occurring on the most limiting rod is at least 95 percent (at a 95 percent
-confidence level) for any condition I or II event.
In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty.
Design limit DNBR values have been determined that satisfy the DNB design criterion.
Additional DNBR wargin is amintained by performing the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.
The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERMAL POWER, Reactor Coolant System pressure and average j
temperature which satisfy the following criteria:
]
The average enthalpy at the vessel-exit is less than the enthalpy of a.
saturated liquid (far left line segment in each curve).
b.
The minimum DNBR satisfies the DNB design criterion (all the other line segments in each curve).
Each curve reflects the most limiting result using either fuel with optimized fuel assembly fuel rod diameter or fuel with star.dard fuel assembly fuel rod diameter. The fuel with optimized rod diameter is analyzed using the WRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively._ The fuel with standard rod diameter is analyzed using the WRB-1 correlation with design limit DNBR values of 1.25 and 1.24 for the typical and thimble _ cells, respectively.
The hot channel exit quality is not greater than the upper limit of c.-
the quality range (including _ the effect of uncertainties) of the DNB correlations. This is not a limiting criterion for this plant.
]
1
-FARLEY - UNIT 1 B 2-1 Revised by NRC letter dated 3une 17,1$98 n
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SAF5.:TY LIMITS BASES The curves of Figures 2.1-1 and 2.1-2 are based on.the most limiting result using an enthalpy hot channel factor Fh which bounds the limit specified in the COLR, and a reference cosine with a peak-of 1.55 for axial power shape.
An allowance is included for an increase in F h at reduced power based on the expression:
N RTP F
Fm F,
1 + p, (1-P)
RTP where P is the fraction of RATED THERMAL POWER, and p g and F
pg are specified in the COLR These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial yower imbalance is within Uke limits of the f t
(delta I) function of the Overtemperature trip. when the axial power imbalance is not within tolerance, the axial power imbalance effect on the Overtemperature delta T. trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching ~the containment atmosphere.
The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial ooeration.
l l
l FARLEY - UNIT 1 B 2-2 Revised by NRC letter dat.edJune 17,1098
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a PLANT SYSTEMS BASES l
l 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The design of the auxiliary feed water system ensures that the Reactor Coolant System can be cooled down to less than 350*F (to commence cooldown with the residual l
heat removal system) from normal operating conditions in the event of any of the following incidents:
Loss of Normal Feedwater Loss of Off-site Power Feed Line Break Main Steam Line Break Accidental Depressurization of Steam Generators Steam Generator Tube Rupture (SGTR)
High Energy Line Break Small Break LOCA Normal Cooldown following a Reactor Trip Station Blackout Each motor driven auxiliary feedwater pump delivers a total of at least 285 gpm to all steam generators which are at a pressure of 1138 psia.
The minim 2m flow requirement for a motor driven pump is based on a high energy line break in the steam supply line to the steam driven auxiliary feedwater pump.
In this scenario, only one motor driven auxiliary feedwater pump will be the source of auxiliary feedwater.
For l
all other scenarios listed above, except Station Blackout, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.
The steam driven auxiliary feedwater pump delivers a total of at least 350 gpm to all steam generators which are at a pressure of 1138 psia. The minimum requirement for the steam driven pump is based on a station Blackout event.
In this scenario, the steam driven auxiliary feedwater pump will be the only source of auxiliary feedwater.
For all other scenarios listed above, except high energy line break in the steam supply line to the steam driven pump, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to the atmosphere concurrent with total 1
loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line l'ocation or other physical characteristics.
3/4.7.1.4
-ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 L
-limits in the event of e steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary.to secondary tube leak in the steam generator of the affected steam line. These values are consistent with the assumptions used in the accident analyses.
Revis'ad by NRC letter dated June 17,199p
'FARLEY-UNIT 1 B 3/4 7-2
2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety Limit prevent overheating of the fuel and I
possible cladding perforation which would result in the release of fission products to the reactor coolant. overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above.the coolant saturation temperature.
Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient..DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through correlations which have been developed to predict the l
DNS flux and the location of DNB for axially uniform and non-uniform heat flux
]
L distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the I
heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.
-The DNB thermal design criterion is that the probability of DNB not occurring on the most limiting rod is at least 95 percent (at a 95 percent confidence level) for any condition I or II event.
In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty.
Design limit DNBR values have been determined that satisfy the DNB design criterion.
Additional DNBR margin is maintained by performing the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g.,
rod bow and transition core) and to provide DNBR margin for operating and design flexibility.
The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERMAL POWER, Reactor Coolant System pressure and average temperature which satisfy the following criteria:
The average enthalpy at the vessel ex'it is less than the enthalpy of a.
saturated liquid (far left line segment in each curve).
b.-
The minimum DNBR satisfies the DNB design criterion (all the other line segments in each curve). Each curve reflects the most limiting result using either fuel with optimized fuel assembly fuel rod diameter or fuel with standard fuel assembly fuel rod diameter. The fuel with optimized rod diameter is' analyzed using the WRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively..The fuel with standard rod diameter is analyzed using the WRB-1 correlation with design limit DNBR values of 1.25 and 1.24 for the typical and thimble cells, respectively, The hot channel exit quality is not greater than the upper limit of c.
the quality range (including the effect of uncertainties) of the DNB correlations. This is not a limiting criterion for this plant.
FARLEY - UNIT 2 B 2-1 Revised by NRC letter datedJune 17, 1998
E SAFETY LIMITS BASES The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using an enthalpy hot channel factor Fh which. bounds the limit specified in the COLR, and a reference cosine with a peak of 1.55 for axial power shape.
An allowance is included for an increase in F h at reduced power based on the expression:
Fm = F, 1 + p, (1-P)'
N R1T F
R1T where P is the fraction of RATED THERMAL PoltER, and p g and F
p, are specified in the COLR These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the ft (delta Il function of the overtemperature trip. when the axial power imbalance is not within tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant system from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant system is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
FARLEY - UNIT:2' B 2-2 Revised by NRC letter dated 3une 17,1$8 j
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_ REACTIVITY CONTROL SYSTEMS BASES w
- MODERATOR TEMPERATURE COEFFICIENT (Continued)
Once the equilibrium boron concentration falls below 100 ppm, MTC measurements may be suspended provided the measured MTC value at an equilibrium boron concentration < 100 ppm is less negative than the 100 ppm MTC surveillance lindt specified in the COLR. The difference between this value and the limiting EOL MTC value conservatively bounds the anximum change in MTC between the 100 ppm equilibrium boron concentration (all rods withdrawn, RATED THERMAL POWER condition) and the licensed end-of-cycle, including the effects of boron concentration reduction, fuel depletion, and end-of-cycle coastdown.
The surveillsnee requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remain; within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.
3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541*F.
This limitation is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation l's within its normal operating range, 3) the P-12 interlock is above its setpoint, 4) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RTun temperature.
3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) separate flow paths, 4) boric acid transfer pumps, and 5) an emergency power supply from OPERABLE diesel generators.
With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions,after xenon decay and cooldown to 200*F.
The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 11,336 gallons of 7000 ppm borated water from the boric acid storage tanks or 44,826 gallons of 2300 ppm borated water from the refueling water storage tank.
Incorporates Amend. 120 & 127 FARLEY-UNIT 2 B 3/4 1-2 Revised by NRC Letter datedJune 17,1993
l REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS (Continued)
With the RCS temperature below 200*F, one injection system is acceptable without single failure consideration on the basis of-the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 180*F provides assurance that a mass addition pressure transient can be relieved by the operation of a single RHR relief valve. Two charging pumps may be capable-of injecting into the RCS for a short time to allow the pumps to be swapped.
This allows seal injection flow to be continually maintained, thus, minimizing the potential for RCP number one seal damage by reducing pressure transients on the seal and by preventing RCS water fram entering the seal.
Particles in the RCS water may cause wear on the seal surfaces and loss of seal injection pressure may cause the seal not to fully resent when pressure is reapplied.
Low temperature overpressure protection is most critical during shutdown when the RCS is water solid. Mass input transients can cause a very rapid increase in RCS pressure allowing little time for operator action to mitigate the event.
For these reasons, more than one pump should be made capable of injecting into the RCS only when the RCS is' in a non water solid condition and when both RHR relief valves are OPERABLE or the RCS is vented via an opening of at least 5.7 square inches. A 5.7 square inch opening is equivalent to the throat size area of two RHR relief valves.
The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN as specified in the COLR after xenon decay and cooldown from 200*F to 140*F.
This condition requires either 2,000 gallons of 7000 ppm borated water from the boric acid storage tanks or 7,750 gallons of 2300 ppm borated water from the refueling water storage tank.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
The limits on contained water volume and bc(ron concentration of the RWST also ensure a pH value of between 7.5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on wchanical systems and components.
Incorporates Amend. 110, 120, & 127 FARLEY-UNIT 2 B 3/4 1-3 Revised by NRC Letter dated]une 17, 1990
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a REACTIVITY CONTROL SYSTEMS RASES The OPERABILITY of one boron injection. system during REFUELING ensures that this system is available for reactivity control while in MODE 6.
3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable-power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident-analyses. OPERABILITY of the control rod position indicators is required to determine' control rod positions and thereby ensure compliance with the control rod alignment and insertion lindts.
For purposes of determining coupliance with Technical Specification 3.1.3.1, any inoperability of full length control rod (s), due to being immovable, invokes ACTION statement "a".
The intent of Technical specification 3.1.3.1 ACTION statement "a" is to ensure that before leaving ACTION statement "a" and utilizing ACTION statement "c" that the rod urgent failure alana is illuminated or that an obvious electrical problem is detected in the rod control system by adnimal electrical troubleshooting techniques.
Expeditious action will be taken to determine if rod innovability is due to an electrical problem in the rod control system.
The ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original design criteria are met.
Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued operation.
In addition, those safety analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.
The maximum rod drop time restriction is consistent with the assumed rod drop time used in the safety analyses. Measurement with T,yg greater than or equal to 541*r and with all reactor coolant pumps operating ensures that the naasured drop times will be representative of insertion times experienced during a reoctor trip at operating conditions.
Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.
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Incorporates Amend.-127 I
FARLEY-UNIT'2 B 3/4 1-4 Revised by NRC Letter datedUune 17,199 3
e d.
PLANT SYSTEMS BASES' 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The design of the auxiliary feed water system ensures that the Reactor Coolant system can be cooled down to less than 350*F (to. commence cooldown with the residual heat removal system) from normal operating conditions in the event of any of the following incidents:
Loss of Normal Feedwater Loss of Off-site Power Feed Line Break Main Steam Line Break Accidental Depressurization of_ Steam Generators Steam Generator Tube Rupture (SGTR)
High Energy Line Break Small Break LOCA Normal Cooldown following a Reactor Trip station Blackout Each motor driven auxiliary feedwater pump delivers a total of at least 285'gpa to all steam generators which are at a pressure of 1138 psia.
The minimum flow requirement for e motor driven pump is based on a high energy line break in the steam supply line to the steam driven auxiliary feedwater pump.
In this scenario, only one motor driven auxiliary feedwater pump will be the source of auxiliary feedwater.
For all other scenarios listed above, except station Blackout, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.
The steam driven auxiliary feedwater pump delivers a total of at least 350 gpm to all steam generators which are at a pressure of 1138 psia.
The minimum requirement for the steam driven pump is based on a Station Blackout event.
In this scenario, the steam driven auxiliary feedwater pump will be the only source of auxiliary feedwater.
For all other scenarios listed above, except high energy line break in the steam supply line to the steam driven pump, two out of three auxiliary feedwater pumps (motor or steam driven pump combination) are required to satisfy the flow demand.
3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> with steam discharge to th,e atmosphere concurrent with total loss of off-site power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.
3/4.7.1.4-ACTIVITY r-The limitations on secondary system specific activity ensure that the resultant
~off-site. radiation dose will be limited to a small fraction of 10 CFR Part 100 i
limits in the event of a steam line rupture. This dose also includes the effects-
"of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of 1
the affected steam line. These values are consistent with the assumptions used in i
Ethe accident analyses.
l FARLEY-UNIT 2 5 3/4 7-2. Revised by NRC letter dated]une 17, 1998
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