ML20126E403
| ML20126E403 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 12/22/1992 |
| From: | WOLF CREEK NUCLEAR OPERATING CORP. |
| To: | |
| Shared Package | |
| ML20126E395 | List: |
| References | |
| NUDOCS 9212290119 | |
| Download: ML20126E403 (10) | |
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Attachmerit II to NA 92-0122 Page 1 of 10 i
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HARKED-1JP TECl!NICAL SPECIFICATION BASES PAGES
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-l 9212290119 921222 PDR ADOCK 05000482 P.
A'ttachment 11 to f4A 92-01?2 l
Page 2 of 10 REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS _ (Continued)
With the RCS temperature below 200'F, one Boration 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 Boron 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 in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by the opera-tion of a single PORY or an RHR suction relief valve.
The boron capability required below 200*F is sufficient to provide a SHUT 00WN MARGIN of 1%.9/k af ter xenon decay and cooldown from 200*F to 140'F.
This condition requires either 2968 gallons of 7000 ppe borated water from the boric acid storage tanks or 14.071 gallons of 2400 ppm borated water from the RWST.
The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.
In the case of the boric acid tanks, all of the contained volume is considered usuable.
The required usable volume may be contained in either or both of the boric acid tanks.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 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 mechanical systems and components.
The OPERABILITY of one Boration System during REFUELING ensures that this system is avaiiabie for reactivity controi wniin in 800E s.
A 3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: (1) acceptable power distrihtion limits are maintained, (2) the minimus SHUTDOWN MARGIN is main-tained, s.nd (3) the potential effects of rod misalignment on associated acci-dent analyses are limited. 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 limits.
Verification that the Digital Rod Position Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital Rod Position Indicator is operating correctly over the full range of indication.
Since the Digital Rod Posit!on System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indicated ranges are picked for verificatier, of agreement with demanded position.
WOLF CREEK - UNIT 1 8 3/4 1-3 Amendment No, g -
- 1 to NA 92-0122 Page 3 of 10 3/4.3 !NSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained (3) sufficient redundancy is main-tained to permit a channel to be out of-service for-testing or maintenance, and (4) sufficient system functional capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and t ansient conditions.
The integrated operation of each of these systems is consistent witn the assumptions used in the safety analyses.
The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.
The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
A The Engineered Safety Features Actuation System Instrumentation Tr 4 Setpoints specified in Table 3.3-4 are the nominal values at which the b1 stables-are set for each functional unit.
A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is-within the band allowed for calibration accuracy.
Specified surveillance intervals and surveillance and maintenance outage times have bern determined in accordance with WCAP-10271, and Supplement 1, " Evaluation of Surveillance Frequencies and
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Out of Service times for the Reactor Protection Instrumentation System," sup-plements to that report, and the NRCs Safety Evaluation dated February 21, i
1985. WCAP-10271 Supplement 2 and WCAP-10271-P-A Supplement 2, Revision 1.
" Evaluation of Surveillance Frequencies and Out of Service Times for the l
Engineered Safety Features Actuation System " the NRC's Safety Evaluation dated i
February 22, 1989, and the NRC's Supplemental Safety Evaluation dated April 30, 1990.
Surveillance intervals and out of service times were determined based on maintaining and an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.
ESF response times specified in Table 3.3-5 which include sequential opera-tion of the RWST and VCT valves (Notes 3 and 4) are based on values assumed in the non-LOCA safety analyses.
These analyses take credit for injection of-borated water from the RWST.
Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening l
of the RWST charging pump suction valves. When the sequential operation of the RWST and VCT valves is not included in the response times (Note 7),the values specified are based on the LOCA analyses.
The LOCA analyses take credit for injection flow regardless of the source.
Verification of the response times specified in Table 3.3-5 5ill assure that the assumptions used for the LOCA and non-LOCA analyses with respect to operation of the VCT and RWST valves are valid.
WOLF CREEK - UNIT 1 8 3/4 3-1 Amencment No 2, 12, L
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.. 1 to NA 92-0122 Page 4 of 10 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above the safety analysis limit DNBR (1.32) during all normal operations and anticipated transients.
In H0 DES 1 and 2 with one reactor coolant loop not in operation this specification requires that the plant be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing decay heat even in the event of a bank withdrawal accident; however, single failure cor.siderations require that three loops be OPERABLE.
A single reactor coolant loop provides sufficient heat removal if a bank withdrawal accident can be prevented; i.e., by opening the Reactor Trip System breakers.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR loops be OPERABLE.
The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and cer. trol, u2G B
The restrictions on starting a reactor coolant pump in MODES 4 and 5 are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg temperatures.
3/4.4.2 SAFETY VALVES The pressurizer Code safety valves operate to prevent.the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs_per hour of saturated steam.
The relief capacity of-a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temneratures.
AmendmentNo,f[
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Page 5 of 10 3/4.4 REACTOR COOLANT SYSTEM BASES SAFETY VALVES (Continued)
During operation, all pressurizer Code safety valves must be OPERABLE to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip and also assuming no operation of the power-operated relief valves or steam dump valves.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASHE Boiler and Pressure Code.
T.as rm _4 3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation.
The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system.
The requirement that a minimum number of pressurizer heaters be OPER8 LE enhances the capability of the plant to control Reactor Coolant System pressure and establish natural circulation.
3/4.4.4 RELIEF VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
3/4.4.5 STEAM GENERATORS The Surveillance hauirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be main-tained.
The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.
Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion.
Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.
WOLF CREEK - UNIT 1 B 3/4 4-2
' 1 to NA 92-0122 Paye.6 of 10 PLANT SYSTEMS BASES ULTIKATE HEAT $1NK (Continued)
The limitations on minimus water level and maximum temperature are based on providing a 30-fay cooling water supply from the Essential Service Water pumps to safety-related equipment without exceecfing its design basis temperature and is con'sistent with the recommendations of Regulatory Guide ~1.27, " Ultimate Heat Sink for Nuclear Plants,f March 1974.
3/4.7.$ CONTROL ROOM EMERGE'NCY VENTILATION SYSTEM The OPERABILITY of the Control Ro'on Emergency Ventilation System ensures that:,(1)theambientairtemperaturedoesnotexceedtheallowabletemperature this system, andfor continuous-duty rating for the equipment and instrumentatio personnel during a(2) the control room will remain habitable for operations nd following all credible accident conditions. Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10 continuous hours in a 31-day period.it sufficient to reduce the buildup of moisture on the charcoal adsorbers and HEPA filters.
The OPERABILITYofthissysteminconjunctionwithcontrofroomdesignprovisions is based on limiting the radiation erposure to personnel, occupying the control room to 5. rems or less whole body, or its equivalent.
This limitation is consistent with the recuirements of General Design Criterion 19 of Appendix A, 10 CFR Part 50.
ANSI b510-1975 and N510-1980 will*be used as procedural guides for surveillance testing.erformances. Surveillance testing provides assurance t and component continue to be in accordance with performance speci-yylcationsfor olf Creek Unit 1, including,applicabl,e parts of ANSI H509-1976.
A 3/4.7.7 EMERGENCY EXHAUST SYSTEM - AUXILIARY BUILDING l
The OPERABILITY of the Emergency' Exhaust System ensures that radioactive materials leaking from the ECCS equipment within the Auxiliary Building following a LOCA are filtered prior to reaching the environment.
I with the heaters operating to maintain low humidity using automatic controlOperatio for at least 10 continuous hours in a 31-day period is' sufficient to reduce the buildup of moisture on the charcoal adsorbers and HEPA filters.
The opera-tion of this system and the resultant effect on offsite dosage calculations was assumed in the safety analyses.
ANSI N510-1975 and N510-1980 will be used as procedural guides for surveillance testing.
The surveillance requirements associated with the HEPA filters, charcoal adsorbers and heaters are stated in 4.9.13.
VOLF CREEK - UNIT 1 8 3/4 7 4 Ame.idmentNo.[
- 1 to NA 92-0122 Page 7 of 10 3/4.8 ELECTRICAL POWER SYSTEMS BASES 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, AND ONSITE POWER DISTRIBUTION The OPERABILITY of the A.C. and 0.C power sources and associated distri-bution systems during operation ensures that sufficient power will be available to supply the safety-related equipment required for: (1) the safe shutdown of the facility, and (2) the mitigation and control of accident conditions within the facility.
The minimum specified independent and redundant A.C. and D.C.
power sources and distribution systems satisfy the requirements of General i
Design Criterion 17 of Appendix A to 10 CFR Part 50.
The ACTION requirements specified for the levels of degradation of the i
power sources provide restriction upon continued facility operation commensurate with the level of degradation.
The OPERABILITY of the power sources are con-sistent with the initial condition assumptions of the safety analyses and are based upon maintaining at least one redundant set of onsite A.C. and 0.C.
power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss-of-offsite power and single failure of the other Unsite A.C. source.
The A.C. and D.C. source allowable out-of-service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources", December 1974.
When one diesel generator is inoperable, there is an additional ACTION requirement to verify that all required systems, sub-systems, trains, components and devices, that de diesel generator as a source of emergency power, pend on the remaining OPERABLE are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE.
This requirement is intended to provide assurance that a loss-of-offsite power event will not result in a complete loss of safety function of critical systems during the period one of the diesel generators is inoperable.
The tem verify as used in this context means to administrative 1y check by examining logs or other information to determine if certain components are out-of-service for main-tenance or other reasons.
It does not mean to perform the Surveillance Requirements needed to demonstrate the OPERABILITY of the component.
The OPERABILITY of the miaimum specified A.C. and 0.C. power sources and associated distribution systems during shutdown and refueling ensures that:
(1) the facility can be maintained in the shutdown or refueling condition for extended time periods, and (2) sufficient instrumentation and control capability are available for monitoring and maintaining the unit status.
A The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies," March 10, 1971, 1.108,'" Periodic Testing of Diesel Generator Units t
Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision-1, August 1977 as modified by Amendment No. 8 issued on May 29, 1987
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1.137, " Fuel-Oll Systems for Standby-Diesel Generators," Revision 1 l
October 1979.
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WOLF CREEK - UNIT 1 B 3/4 8-1 AmendmentNo.[
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. Attachment 11 to NA 92-0122 Page 8 and 10 3/4.9 REFUELING OPERATIONS BASES 3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
The limitation on Kaff of no greater than 0.95 is sufficient to prevent reactor criticality during refueling operations.
The locking closed of the required valves during refueling operations precludes the possibility of uncontrolled boron dilution of the filled portions of the Reactor Coolant System.
This action prevents flow to the RCS of unborated water by closing flow paths from sources of unborated water.
These Itaitations are consistent with the initial conditions assumed for the boron dilution incident in the safety analyses.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4 9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.
This decay time is consistent with the assumptions used in the safety analyses.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS The requiremants on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.
The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
The OPERABILITY of this system ensures the containment purge penetrations will be automatically isolated upon detection of high radiation levels within containment.
The OPERABILITY of this system is required to restrict the release of radioactive materials from the containment atmosphere to the environment.
3/4.9.5 COMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
WOLF CREEK - UNIT 1 8 3/4 9-1
Attachment !! to NA 92-0122 Page 9 of 10 BASES 3/4.9.6 REFUELING MACHINE The OPERABILITY reautrements for the refueling machine and auxiliary hoist ensure that: (1) manipulator cranes will be used for movement of drivs rods and fuel assemblies, (2) esch crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE FACILITY The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool areas ensures that in the event this load is dropped: (1) the activity release will te limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.
This assumption is consistent with the activity release cssumed in the safety analyses.
3/4.9.8 RESIOUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) loop be in operation ensures that: (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140*F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.
The minimum of 1000 gpa allows flow rates which provide additional margin against vortexing at the RHR pump suction while in a reduced RCS inventory condition.
8 The requirement to have two RHR loops OPERABLE whan there is less than 23 feet of water above the rcactor vessel flange ensures that a single failure of the operating RM loop will not result in a complete loss of RHR capability.
With the reactor vessel head removed and at least 23 feet of water above the reactor vessel flange, a large heat sink is available for core cooling.
- Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.
3/4.9.9 CONTAINMENT VENTILATION SYSTEM The OPERABILITY of this system ensures that the containment purge penetrations will be automatically isolated upon detection of high radiation levels within the containment.
The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.
WOLF CREEK - UNIT 1 B 3/4 9-2 Amendment No. 35
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Attachment II to NA 92-0122 Page 10 of 10 INSERT A When determining compliance with action statement requirements, addition to the RCS of borated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered to be 'a positive reactivity change.
INSERT B Addition of borated water with a concentration greater than or equal to the minimum required kWST concentration but less than the actual RCS baron concentration shall not be considered a reduction in boron concentration.
INMRT C Addition to the RCS of barated water with a concentration greater than or equal to the minimum required RWST concentration shall not be considered a positive reactivity change.
Cooldown of the RCS for restoration of operability of a pressurizer code safety valve, with a.- negative moderator temperature coefficient, chall not be considered a positive reactivity change provided the RCS is borated to the COLD SHUTDOWN, xenon-free conditions per specification 3.1.1.2.
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