ML17291A713
| ML17291A713 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 03/27/1995 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17291A714 | List: |
| References | |
| NUDOCS 9504050255 | |
| Download: ML17291A713 (13) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055~001 WASHINGTON PUBLIC POWER SUPPLY SYSTEM
~DOCKET NO.
5 -3 UCLEAR PROJECT NO.
2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
135 License No. NPF-21 The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by the Washington Public Power Supply System (licensee) dated October 31,
- 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 2.
D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment and paragraph 2.C.(2) of Facility Operating License No. NPF-21 is hereby amended to read as follows:
9504050255 950327'>
PDR ADOCK 05000397 P
PDP
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{2)
Technical S ecifications and nvironmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
f35 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
This amendment is effective iaeediately and will be implemented prior to restart from the spring 1995 refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
Harch 27, 1995 J
es W. Cl if rd, Senior Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation
TTACHMENT TO LICENSE AMENDMENT MENDMENT NO.
1 TO FACILITY OPERATING LICENSE NO. NPF-21 DOCKET NO. 50-397 REMOVE Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
The corresponding overleaf pages are also provided to maintain document completeness.
INSERT 3/4 3-71 3/4 3-73 3/4 3-74 3/4 4-7a B 3/4 4-la 3/4 3-71 3/4 3-73 3/4 3-74 3/4 4-7a B 3/4 4-la
C/7 C)
INSTRUMENT TABLE 3.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION REQUIRED NUMBER OF CHANNELS MINIMUM APPLICABLE CHANNELS OPERATIONAL OPERABLE CONDITIONS ACTION Cn Im I
C CrO Crd I
1.
Reactor Vessel Pressure 2.
Suppression Chamber Water Level 4.
Suppression Chamber Water Temperature 5.
Suppression Chamber Air Temperature 6.
Drywell Pressure 7.
Drywell Air Temperature 8.
Drywell Oxygen Concentration 9.
Drywell Hydrogen Concentration 10.
11.
Suppression Chamber Pressure 12.
Condensate Storage Tank Level 13.
Main Steam Line Isolation Valve Leakage Control System Pressure 2/sector 1/sector 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2 1,
2
&0 80 80 80 80 80 80 80 80 80 80 80
~0 4
f
ACTION 80 Table 3.3.7.5-1 (Continued)
ACCIDENT MONITORING INSTRUMENTATION ACTION STATEMENTS a 0 b.
ACTION a ~
b.
With the number of OPERABLE accident monitoring instrumentation channels less than the Required Number of Channels shown in Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
With the number of OPERABLE accident monitoring instrumentation channels less than the Minimum Channels OPERABLE requirements of Table 3.3.7.5-1, restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
81 With the number of OPERABLE accident monitoring instrumentation channels less than required by the Minimum Channels OPERABLE requirement, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:
Initiate the preplanned alternate method of monitoring the appropriate parameter(s),
and In lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.
WASHINGTON NUCLEAR UNIT 2 3/4 3-73 Amendment No. +6%1"-
TABLE 4.3.7.5-1 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS C/l C)
C I
Kl I
C:
INSTRUMENT 1.
Reactor Vessel Pressure 2.
4.
5.
6.
Primary Containment Pressure 7.
Drywell Ai.r Temperature 8.
Drywell Oxygen Concentration 9.
Drywell Hydrogen Concentration 10.
ll.
Suppression Chamber Pressure 12.
Condensate Storage Tank Level 13.
Main Steam Line Isolation Valve Leakage Control System Pressure 14.
Neutron Flux:
RCIC Flow 16.
HPCS Flow 17.
LPCS Flow Suppression Chamber Water Level Suppression Chamber Water Temperature Suppression Chamber Air Temperature CHANNEL CHECK CHANNEL CALIBRATION APPLICABLE OPERATIONAL CONDITIONS 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2 1,2
~ RO q
REACTOR COOLANT SYSTEM 3/4.4. 2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 a) Thc safety valve function of at least 12 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function liftsettings:~
2 safety/relief valves 0 1150 psig +)Xi-%
4 safety/relief valves 6 1175 psig +IX/-3X 4
safety/relief valves O 1185 psig +lX/-3X 4
safety/relief valves e 1195 psig +IX/-3X 4
safety/relief valves 0 1205 psig +i%/-3X APPLICABILITY:
OPERATIONAL CONDITIONS 1, and 2, when THERMAL POWER is greater q
T RRE I RRTEO THERHRL OTHER.
b) The safety valve function of at least 4 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function lift settings:~
2 safety/relief valves e 1150 psig +1%/-3X 4
safety/relief valves 8 1175 psig +1%/-3X 4
safety/relief valves 0 1185 psig +lZ/-3X 4
safety/relief valves 0 1195 psig +1%/-3X 4
safety/relief valves 0 1205 psig +IX/-3X APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3, when THERMAL POWER is '.ess TEO T EIHRL PO ER.
ACTION:
a.
With thc safety valve function of one or sore of the above required safety/relief valves inoperable, bc in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLO SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With one or lore safety/relief valves stuck open, provided that suppression pool average water teiperaturc is less than 904F, close the stuck open safety/relief valve(s); if unable to close the open
- The liftsetting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
WASHINGTON NUCLEAR - NIT 2 3/4 4-7 Amendment No.
BO
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY/RELIEF VALVES LIMITING CONDITION FOR OPERATION ACTION: (Continued) valve(s) within 2 minutes or if suppression pool average water temperature is 110'F or greater, place the reactor mode switch in the Shutdown position.
SURVEILLANCE RE UIREMENTS MASHINGTON NUCLEAR UNIT 2 3/4 4-7a Amendment No.
~1 35
3/4.4 REACTOR COOLANT S STER ES 3/4.4.1 RECIRCULATION SYSTEM Operation with cee reactor recfrculatfon loop inoperable has been evaluated
~nd teen found te he ecceptoHe yrivfded the un1t fe opereted fn eccordence arith the single recirculation loop operation Technical Specifications herein.
An inoperable get pap 4s not, '4n itself, I Sufficient reason to declare e recirculation loop inoperable, be it does, in case If a design-basis-accident, increase the blowdown area and reduce the capability of refJood$ ng the core; thus, the requireaent for shutdown of the facilityHth a get puap inoperable.
Jet pmp failure can be detected by aonitoring get pmp perforaance on a prescribed schedule for significant degradation.
R<<circulation loop flow aisaatch liaits are in colpliance with the KCCS I.OCA analysis design criteria.
The limits will ensure an adequate core flow coastdcwn from either recirculation loop followinp a LOCA.
lthere the recircula-t',on loop f1ow Iisaatch liaits cannot be aaintained during two recirculation loop operation, continued operation is peraitted in.the single recirculation 1ocp operation Node.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop teaperatures shall be ~ithin 50"F of each othe".
prior to startup of an idle loop.
The loop teIeperature aust a1so be within 50OF of the reactor pressure vessel coolant temperature to prevent therial sho:~ t. the recirculation puap and recirculation nozzles.
Since the coolant it tlute Ictus of the vessel is at a lower temperature than the coolant in the leper regions of the core, undue stress on the vessel would result if the teILperature difference was greater than 145OF.
3/4.4.2 SAFETY/RELIEF VALVES The safety valve capacity is designed to licit the priaary system pressure, Cncluding transients, in accordance with the requireoents of the ASIDE Boiler and Pressure Vessel Code,Section III, X971, Nuclear Power Plant ccepanents (up
~
to and including S~r 1971 Addenda).
The Code allows a peak pressure of I&Xof design pressure (l250 (design)
X 1.10 ~ 7375 psig aaxiaw) under upset conditions.
In addition, the Code specifications require that the lowest valve aetpoint be at or below design pressure and the highest valve setpoint'e set Io that total accumulated pressure does not exceed DtN of the design pressure.
c The safety va1ve sizing evaluation ass~s credit for operation of the acrm protective systea which say be tripped by one of Oro sources; i.e., a direct position switch or neutron flux ignal.
The direct acrm signal 4 derived froa position switches aaa~
ce the aain stamline isolation valves CNSIV's) or the turbine stop valve, ee Aea geessure Notches mounted on the pep va1ve of the Curbing control valve lychaulic actuation ayatea.
The posi-tion switches are @cheated when the respective valves are closing, and follow 4ng 10L'rave1 ot full atroke.
The pressure switches are actuated when a fast closure of the ctetrol valves is initiated.
- Further, no credit fs taken for paar operation of the pressure relieving devices.
Credit is only taken for WASHINGTON lOCLEN NIT 2 I 3/4 1-l Aiendaent No. 62
EACTOR COOLANT SYS v
BASES 3 4.4.2 SAFETY REL VAL ES (Continued) the dual purpose safety/relief valves in their ASHE Code qualified mode (spring lift) of safety operation.
The overpressure protection system must accommodate the most severe pres-surization transient.
There are two major transients that represent the most severe abnormal operational transient resulting in a nuclear system pressu're rise.
The evaluation of these events with the final plant configuration has shown that the HSIV closure is slightly more severe when credit is taken only for indirect derived scrams; i.e.,
a flux scram.
Utilizing this worse case transient as the design basis event, a minimum of 12 safety/relief valves are required to assure peak reactor pressure remains within the Code limit of 110X of design pressure.
Testing of safety/relief valves is normally performed at lower power with adequate steam pressure and flow. It is desirable to allow an increased number of valves to be out of service during testing.
Therefore, an evaluation of the HSIV closure without direct scram was performed at 25X of RATED THERHAL POWER assuming only 4 safety/relief valves were operable.
The results of this evaluation demonstrate that any 4 safety/relief valves have sufficient flow capacity to assure that the peak reactor pressure remains well below the code limit of 110X of design pressure.
Demonstration of the safety/relief valve lift settings will be performed in accordance with the provisions of Specification 4.0.5.
3 4.4.3 REACTOR COOLANT SYSTEH LEAKAGE 3 4.4.3. 1 LEAKAGE DETECTION SYS EHS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"
Hay 1973.
The primary containment sump flow monitoring system monitors the UNIDENTIFIED LEAKAGE collected in the floor drain sump with a sensitivity such that 1 gpm change within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> can be measured.
Alternatively, other methods for measuring flow to the sump which are capable of detecting a change in UNIDENTIFIED LEAKAGE-of 1 gpm within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with an accuracy of + 2X may be used, for up to 30 days, when the installed system is INOPERABLE.
MASHINGTON NUCLEAR UNIT 2 B 3/4 4-la Amendment No...,,135