ML19246C524
| ML19246C524 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 06/06/1979 |
| From: | Gammill W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19246C522 | List: |
| References | |
| NUDOCS 7907260149 | |
| Download: ML19246C524 (26) | |
Text
,
Os* **Gv fs
.?
UNITED STATES
{
)y
'g NUCLEAR REGULATORY COMMISSION g
W ASHING TON, D. C. 20555
%.v /
PUBLIC SERVICE COMPANY OF COLORADO DOCKET NO. 50-267 FORT ST. VRAIN NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 21 License No. DPR-34 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The applications for amendment by Public Service Company of Colorado (the licensee) dated December 11, 1978, September 8, 1978, April 2, 1979, November 16, 1977 and October 5,1978, comply with the stan-dards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this emendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendnent is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.D(2) of facility Operating License No. DPR-34 is hereby ar anded to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 21, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Soecifications.
358 270 7907266Mg
2-This license anandment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION bb k>
[
h-(
William P. Gammill, Assistant Director for Standardization and Advanced Reactors Division of Project Management Office of Nuclear Reactor Regulation i ttachment:
'hanges to the Technical Specifications Date of Issuance: June 6, 1979 t
358 27i
ATTACHMENT _
AMENDMENT NO. 21 TO FACILITY OPERATING LICENSE DPR-34 Docket No. 50-267 Remove existing pages 4.2-3, 4.2-4, 4.2-20, 4.2-21, 4.3-6, 4.3-11, 4.5-1, 4.5-2, 4.10-1, 4.10-2, 4.10-3, 5.1-3, 5.2-7, 5.2-8, 5.2-17, 5.2-18, 5.5-1, 5.5-2 and 5.5-2(a) with the attached revised pages bearing the same numbers.
The changed areas are reflected by marginal lines.
Also, add pages 4.2-22, 4.10-2(a), 5.2-21 and 5.10-3 which contain new material.
358 272
L.2-3 These accumulators centain sufficient water to permit circulater s_
- sst-down without circulator damage if both the normal and the backup bearing water supplies should fail. The minor water makeup requirements for the normal besring we.ter system is provided by the bearing water makeup pu=ps.
Snecification LCO L.2. 3 - Turbine Water Removal Pumn, Limitin Conditioqs for Oreration There shall be one operable turbine water re=cval pump during pover operaticn.
Basis fer Srecification LCO h.2.3 One turbine wr.ter removal pump has sufficient capacity to remove t'e water from two circulator water turbines. This is adequate for a safe shutdown cooling.
Srecification LCO h.2.h - Service Water Pumes, Limitin Conditions for Oreration At least two service water pumps and the associated pump pit shall be operable f.uring power operation.
Basis for Srecification LCO b.2.h The availability of the service water system ensures the capability of supplying essential components with cooling water, as described in FSAR Sections 1.k, 10.3, and Ih.h.
Erecificcticn LCO L.2. 5 - Circulatine Water Makeur Syste=, Limitine Conditions for Creration At least two circulating water makeup pumps connectible to the essential bus shall be operable during power operation.
Basis for Scecification LCO k.2 5 Circulating water system makeup to the service water and fire protection system provides adequate makeap water to safely shut the reacter down from c
Jv lb e
L.2 h sny ner:.al operating condition.
For further explanation see FSAR lectiona 1.h, 10.3 and 14.h.
Specification LCO h.?.6 - Firevater Pumes, Limitin. Conditions rur Operation The engine-ariven fir pump, motor driven fire pump, and associated pump pits shall be operable and there shall be at least 325 gallens of fuel in storage during power operation.
Basis for Soecification LCO h.2.6 Either of the fire pumps operating in conjunction with either firewater booster pump provides adequate capacity to operate a circulator water turbine and supply emergency cooling water for safe shutdown cooling. With the 325 gallons of fuel in storage, the engine driven fire pump can operate at rated conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> which is adequate time to have more fuel delivered to the site. For further erplanation see Final Safety Analysis Report, Sections 1.4, 10.3, and 14.4.
S ecific ation LCO h.2.7 - PCR7 Pressurizatien Limiting Cenditions l
for 01eration The PCRV shall not be pressurized to more than 100 psia unless:
a) The PCRV safety valve installation is operable, and there is less than 5 psig between the rupture disc an1 relief valve, and both inlet block valves are locked open.
- b)
All primary and secondary penetration closures and hold dcun plates are in place and operable, per Specification LCO h.2 9
- c)
The interspaces between the primary and seccndary penetration closures are maintained at a pressure greater than primary bD0
]' lil7 system pressure with purified helian gas.
- During the initial low power physics testing ($ 0.1% of rated thermal power)
~
with the PCRV pressurized and when helium circulators C and D are in the pre-nuclear Felton wheel configuration, exceptions to provisions b) and c) of this LCO vill be the installation of the secondary closures and pressurization of the primary secondary closure interspace of C and D helium circulator penetrations.
Amendment No. 21
4.2-20 Specification LCO 4.2.16 - Diesel-Driven Pumos for IACM, Limiting Conditions for Operation DELETE THIS SPECIFICATION IN ITS ENTIRETY Amendment No. 21 330 4/3 n -, -
i 4.2-21 Specification LCO 4.2.17 - Diesel-Driven Generator for ACM, Limiting Conditions for Operation The reactor shall not be operated at power unlese the ACM diesel-generator is operable, including the following:
1.
One fuel oil transfer pump from the fuel oil storage tanks to the diesel fuel oil day tank is operable.
2.
The associated switchgear and motor control center are operable.
3.
There are at least 19,000 gallonn of fuel total in storagc.
The diesel-generator set may be inoperable for up to 7 consecutive days per nonth or a total of 21 days in a three month period for performance of maintenance, with the reactor at power.
Basis for Specification LCO 4.2.17 The ACM diesel-generator provides power independently of the plant electrical distribution network to various valves, lighting, and pieces of equipment. That equipment provides an alternate means of maintaining PCRV cooling during the Loss of Forced Circulation situation described in the Final Safety Analysis Report, Section 14.10.
The 10,000 gallons of fuel provides for one week operation of the generator with full ACM load, which is adequate time for obtaining additional fuel from off site sources.
Specification LCO 4.2.18 - Primary Coolant Depressurization Limiting Condition for Operation The reactor shall not be operated at power unless a flow path for depressurization of the primary system exists that includes the HTFA, Helium Purification Cooler, Helium Purification Dryer, Low Temperature Gas-to-Cas Exchanger, LTA, and associated valves and piping to the reactor building exhaust ducting.
Basis for Soecification LCO 4.2.18 In the event that pe rmanent loss of forced circulation occurs, it is necessary to depressurize the primary coolant system.
Start of depressuri-zation following onset of loss of circulation is initiated as a function of prior power levels, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> from full power operation, and is ccmpleted in approximately seven hours. Depressurization is completed by venting the purified gas to the atmosphere.
358 276 Amendment No. 21
4.3-6 Specification LCO 4.3.9 - High Pressure Helium Supply System DELETE THIS SPECIFICATION IN ITS ENTIRETY A=ead=ent No. 21 c
4.3-11 TABLE 4.3.10-1 (Continued)
CLASS I HYDRAULIC SNUBBERS (Continued)
Boiler Feed Snubbers BFS-54 BFS-398 BFS-435 SFS-526 BFS-563 BFS-763 BFS-138 BFS-400 SFS-437 BFS-528 BFS-564 BFS-764 BFS-139 BFS-402 BFS-451 BFS-529 BFS-566 BFS-796 BFS-142 BFS-412 BFS-477 BFS-530 BFS-572 BFS-820 BFS-l*9 BFS-416 BFS-479 BFS-532 BFS-573 BFS-823 BFS-152 BFS-420 BFS-498 BFS-534 BFS-577 BFS-824 BFS-153 BFS-421 BFS-500 BFS-536 BFS-614 BFS-843 BFS-297 BFS-422 BFS-501 BFS-537 BFS-641 BFS-844 l BFS-352 BFS-425 BFS-516 BFS-553 (2)
BFS-679 BFS-397 BFS-434 BFS-523 BFS-556
, EFS-711 Boiler Feed Snubbers - Emergency BFS-14E BFS-57E BFS-167E BFS-229E BFS-298E BFS-430E BFS-ISE BFS-74E BFS-181E BFS-243E BFS-398E BFS-431E BFS-16E BFS-76E BFS-197E BFS-244E BFS-399E BFS-432E BFS-26E BFS-77E BFS-2035 BFS-245E BFS-405E BFS-442E BFS-29E BFS-89E BFS-204E BFS-?.57E BFS-414E BFS-444E BFS-30E BFS-122E BFS-210E BFS-260E BFS-417E BFS-31' BFS-141E BFS-216E BFS-263E BFS-419E BFS-47E BFS-142E BFS-218E BFS-264E BFS-421E BFS-53E BFS-143E BFS-219E BFS-268E BFS-422E BFS-56E BFS-158E BFS-?28E BFS-269E BFS-423E Hydraulic Oil Snubbers HCS-1 HCS-14 HOS-29 HOS-46 HOS-61 HCS-76 HCS-2 HOS-15 HOS-30 HCS-48 HCS-63 hCS-77 HOS-3 HCS-16 HCS-31 HOS-49 HCS-64 HCS-78 HCS-4A HOS-17 HCS-33 HOS-50 HCS-65 HCS-79 HOS-4B HCS-18 HCS-34 HOS-51 HOS-66 HOS-80 HCS-5 HCS-19 HCS-35 HCS-52 HCS-67 HCS-81 HCS-6 HCS-20 HOS-36 HCS-53 HCS-68 HCS-82 HCS-7 HCS-21 HCS-37 HOS-54 HCS-69 HCS-83 HCS-8 HCS-22 HCS-38 HCS-55 HOS-70 HCS-84 HOS-9 HCS-23 HOS-39 HCS-56 HOS-71 HCS-85 HCS-10 HCS-24 HOS-40 HCS-57 HOS-72 HCS-86 HOS-ll HCS-25 HCS-41 HOS-58 HOS-73 HOS-87 HCS-12 HOS-27 HOS-42 HCS-59 HCS-74 HCS-88 HCS-13 HCS-2b HOS-45 HCS-60 HOS-75 bJ0 ro q-g L/o Amendment No. 21
h.5-1 k.5 CONFINEMENT SYSTD! - LIMITIT, COVDITIO iS FC9 CFERATIC:1 Arplicability Applies to the minimum cperable equipment of the reactor building (corfinenent), and the ventilation system.
Objective To assure the operability of the confinement systems.
Specification LCO k.5.1 - Reactor Buildine, Limitine Conditiens for Oeeration The plant shall not be operated at power; reactor vessel internal maintenance shall not be performed with irradiated fuel in the PCRV; or irradiated fuel handling shall not be performed within the reacter building unless:
a) Reactor Building Integrity is maintained as follows:
1.
Personnel access to the building is controlled.
2.
The reactor building pressure is sub-atmospheric.
3.
The reactor building louvers are closed and the
" pressure set point" is at 3 inches of water or less, except that reactor building louver groups may be opened one at a time quarterly for surveillance testing while the reactor is at power provided that the pre-equisites of SR 5.5.2 are adhered to.
4.
When the truck doors to the truck bay are open, the reactor floor hatch, the deck hatch and all personnel doors in the truck bay are closed.
When the reactor floor hatch and/or the deck hatch are 5.
the truck doors and external personnel doors in
- open, 7CO 7 7 Q)
L/
the truck bay are closed.
JJU b) Two of the t'aree reactor building exhaust fans are operable.
4,
h.5-2 Basis fer Srecification LCO h.5.1 The integrity of the reactor building and operation of the ventilating system in cumbination ILnit the off-site doses under normsl and abnormsl conditions. In the unlikely event of a majcr release of activity frcm the PCRV, the combination of the reactor building and ventilation systen vould to keep off-site doses well below 10 CFR 100 limits (see FSAR Section act lb.10.3.h).
The pressure in the reactor building is held slightly belev atmospheric Exfiltration would occur only above a wind velocity of about pressure.
30 mph.
Wind conditions v;. thin the range of 0 to 25 mph prevail at the site about 98% of the time. The mechanical turbulence from vind speeds of 25 mph or higher would result in a dilutien better than during lesser vini speed conditions for any nuclides exf1.rsted frou the reactor building.(FSAR Section 6.1.h.2)
The purpose of the pressure relief device is to maintain the integrity of the reactor building by relieving the pressure inside the building when it equals oi exceeds 3 inches of water.
In the unlikely avent cf the occurrence af a rapid increase of pressure inside the building of or exceeding 3 inches of water, the louvers vould open, relieving the pressure, and then be autc=*1tically clcsed at approximately atmcspheric pressure (or they can be manually closed), restering the int egrity of the reacter building (see FSAR 6.1.3.L) ar.d maint2ining the pctential loses frca the occurrence to as lov as practicable.
The building ventilaticn system caintains the reactor building pressure slightly subatacepheric and reduces the amcunt of radicactivity released to :he environment, during normal cperation or accident ccnditions.
J)U lod
4.10-1
~
Specification LCO 4.10.1 - Room Isolation Dampers, Three Room Control Complex. Limiting _ Condition for Operation The HVAC Room Isolation Dampers of the control room, auxiliary electric room and the 480 volt switchgear room, shall be operable during reactor power operation.
If the dampers become inoperable and cannot be made operable within 72 'aours, the reactor shall be shut down in an orderly manner.
Basis for Specification LCO 4.10.1 The HVAC room isolation dampers for the control room, auxiliary electric room and the 480 volt switchgear room, provide the required area isolation for maintaining an effective concentration of Halon after actuation of the Halon fire suppression system.
Specification LCO 4.10.2 - Halon Fire Suppression System, Three Room Centrol Complex, Limiting Condition for Operatica The Halon Fire Suppression system for the control room, auxiliary electric room, and the 480 volt switchgear room shall be operable during reactor power operation.
If the Halon system becomes inoperable and cannot be made operable within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the reactor shall be shut down in an orderly manner.
Basis for Specification LCO 4.10.2 The Halon system provides fire suppression capability for the c ntrol room, auxiliary electric room, and the 480 volt switchgear room.
Halon is a non-toxic, halogenated chemical fire suppressant. The Halon system is a total flooding extinguishing system divided into three sections. One section supplies the 480 volt switchgear room, the second section supplies the control room, and the third the auxiliary electric room.
Total flooding of these areas will extinguish an active fire without requiring abandonment of the area.
The auxiliary electric room Halon System Section is automatically initiated by the simultaneous actuation of a detector in zone #2 and #3 of Table 4.10-3; the 480 volt room Halon System Section is automatically initiated by the simultaneous actuation of a detector in fire detection zone #5 and #6 of Table 4.10-3.
The Halon Syst m Section for the 480 volt switchgear reum may be manually initiated by a switch located just outside the door of the rocm. The Halon System for the auxiliary electric room may be manually initiated by a switch located just cutside of the room.
The Halon System Section of the control room is manually initiated by a switch located in the control room.
In the event that electrical power is not available to initiate Halon System operation, each Halon storage bottle is provided with a manually operated release mechanism
}
}
which will release the bcttle contents when operated.
,n Amendment No. 21
4.10-2 Specification LCO 4.10.3 - Smoke Detectors and Alarms for Three Room Control Complex and Congested Cable Areas, Limiting Condition ior Operation The reactor shall not be operated at power unless the minimum number of smoke detectors for each of the zones listed in Table 4.10-3 are operable. If the minimum number of detectors of each zone are not operable, the following actione shall be taken:
An individual shall be designated to inspect the area or areas with inoperable detectors once per hour.
The inoperable detectors shall be made operable within thirty (30) days or the reactor shall be shut down in an orderly manner.
Basis for Specification LCO 4.10.3 The smoke detection and alarm system provides detection and alarm the 480 vcit capability for the control room, auxiliary electric root, switch gear room, the congested cable areas located at the "G" and "J" column rows and seJected reactor building EVAC return air ducts in various areas which are not normally manned.
In additienthe system will automatically initiate operation of the Halon fire suppression system in the auxiliary electric room or the 480 volt switch-gear room upon actuation of a detector in both of the zones in each room.
The system alerts the operator to the possibility of a fire in the congested cable areas and to the necessity of investigation of conditions in these areas.
)3b Amendment No. 21
4.10-3 Specification LCO 4.10.4 - Fire Barrier Penecration Seals, Limiting Condition for Operation All fire barrier penetration seais shall remain intact.
If a fire barrier penetration seal is disturbed, a continuous fire watch shall be posted on either side of the disturbed seal.
Basis for Specification LCO 4.10.4 There are a number of fire barrier penetration seals installed between In vital plant areas wiere cables penetrate walls which act as fire stops.
order to prevent tic. spread of a fire from one vital area to another, cable If penetrations have been sealed with various fire retardant materials.
the material of a fire barrier must be disturbed for maintenance, establishing a fire watch on either 3ide of the barrier assures early notification of a potential fire hazard.
Specification LCO 4.10.5 - Fixed Water Spray S/ stems for the Auxiliary Electric Room, 480 Volt Switchgear Room, and Congested Cable Areas The fixed wacer spray systems providing supplementary fire protection in the 480 volt switchgear room, auxiliary electrical equi pment room, and congested cable areas of the reactor building side of the "J" wall and turbine building side of the "G" vall shall be operable during reactor power operation.
If the spray systems become inoperable and cannot be made operable within 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, the reactor shall be shut down in an orderly manner.
Basis for Specification LCO 4.10.5 The fixed water spray systems provide backup fire suppression in the congested cable areas.
In the 480 volt switchgear room and the auxiliary electric room, the systems are a backup to the installed Ealon system.
If the fixed water spray systems ware inoperable, the smoke detectors and alarm system of LCO 4.10.3 would be available for detection and alarm of a fire.
In all areas where it is installed the fixed water spray system is a backup to normal fire fighting techniques.
7r J)9 7-,
ud)
U Amendment No. 21
5 1-3 abcve, and calibrated once a year. Operable reserve shutdown s
hoppers shall have an actuating bottle pressure >,1500 psig.
e) The reserve shutdown hopper pressure switches shall be calibrated at the sa=e interval that they are removed from the reactor for maintenance.
Basis for Srecification SR 5.1.2 The reliability of the reserve shutdown system to perform its function vill be maintained by a control system precsure test and actual off-line rupture tests conducted in the hot service facility or other suitable facility.
The control system pressure test demonstrates the ability to pressurize the hoppers and indicates the operability of the cont:ol system canponents.
A successful test vill increase the hopper pressure about 10 psi above reactor pressure. This differential is well below the minimum 115 psi l
differential required to burst the disc.
The off-line tests consist of actual aisc ruptures and poison drops.
These vill be used to determine the reliability of the differential burst pressure of the disc, and the tenda-cy of the poisen material to hang up cr deteriorate in the hoppers over extended periods of time.
This test information vill be used to verify the capability to shut down the reactor in an emergency situ? tion. The reserve shutdown system hoppers operat e in ts n subsyst ems.
"he first consists of the seven hoppers in refueling regions 1, 2. 5, 7, 22, 26 and 3h; the second subsystem is compr ed of the remaining thirty hcppers in the remaining refueling regions.
4 Safe control of the reactor by the reserve shutdevn system can be accomplished with one of the seven hoppers inoperative, and one of the remaining 30 hoppers incperative. A differential pressure of from 585 to 315 psi is available from the helium supply bottle with a pressure > 1500 psis 8 28,4 33 Amendment No. 21
5 2-7 a.
Tensile specimens are not included, since the liner is not a load carrying member but only a ductile me=brane.
b.
No thermal control specimens nave been provided, since there is no appreciable tempereture cycling of the liner.
The liner =aterials vill nor= ally be kept at or below 150 F during all plant operation.
Tests performed on this liner material (see FSAR Section 5.7 2.2) have indicatea that no observable changes in material characteristics developed during an exposure to a fluence equivalent to the first five years of power operation. Further, these te.its de=cnstrated no significant damage after a fluence equivalent to 30 years of power operation. Tbe testing program prescribed for the Fort Et ain liner is in canpliance with the AS{E Boiler and Pressure Vessel C.
Section III N-110.
The interval for specimen removal and Lesting subsequent to the fifth refueling cycle cay be adjusted based en the analysis of prior resultn.
Scecification SR 5.2.6 - Plateout Probe Surveillance One platecut probe shall be removed for evaluation coincident with the first, third, and fifth refueling, and at intervals not to exceed five refueling cycles thereafter.
If, during the second or fourth refueling cycle, or any refueling cycle folleving the fifth refueling, the primary c oolant noble gas activity (ga--' + beta) should increase by 25% over the average activity of the previous three =cnths at the same reactor power level and the primary coolant activity is greater than 25% of design, the plateout probe shall be renoved at the end of that refueling cycle.
The probes shall be analyzed for 'OSr inventory in the reactor circuit.
131 The probes removed shall also be analyzed for 7,
7co q--
JJO 4dD
4 5.2-8 Basis for Specification SR 5.2.6 The plateout probes are located in penetrationc extending into steam generator shroads and then into the gas stream of each coolant locp. One sample is accumulated by continuously bypassing a small portion of the core cutlet coolant stres= through diffusion tubes and sorption beds located in tne probe body. Another sample can be accumulated by continuously bypassing a portion of the circulator outlet coolant stream through the probe. The core outlet sample can be used to determine the concentrations of fissien products in the coolant stream entering the steam generator; the circulator outlet sa=ple provides information about the amount of cleanup in each pass around the circuit.
90Sr and the results shall be used The probes shall be analyzed for 90Sr inventory in the reactor circuit to determine to establish the total compliance with LCO h.2.8.
Results of probe analyses ahc11 be canpared 90Sr which were made between probe with the calculated esti=ates of 1311 shall be rade to determine the degree r emovals. The analysis for of conservatism of the assumptiens made regarding the circulating and plate 1 out iodine in the primary ecolant circuit.
The interval for probe remaval and analysis subsequent to the fifth refueling cycle may be adjustei based upon the anal) is of prior results.
Srecification SR 5.2.7
'4ater Turbine Drive Surveillance Caapcnents of the helium afreulater water turbine drive system shall be tested as follows:
a) One circulator and the associated water supply valving in each loop will be functionally tested by operation on water turbine drive using fee & tater, condensate, and boosted condensate (supplied to the firewater oooster pumps et fire pump discharge pressure), annually.
)]rh gy,j Amend ent No. 21
5.2-17 Unlike most designs of emergency systems of conventional nuclear power plants, the components of the Safe Shutdown System of the Fort St. Vrain plant are utilized and operated during normal operation of the plant. This includes the helium circulators.
The performance of the helium circulators is continuously monitored during operation, t.e., compressor differential pressure, steam turbine steam flows, bear:.ng water flows, buf fer helium flows, and shaf t vibration.
Examination at the time of the first turb ne generator overhaul, and at 10-year intervals thereafter, 1, ufficient to monitor the condition of the helium circulator. The first turbine generator ' tear-down' or overhaul usually occurs af ter one year running to check the total assembly. Additional overhauls occur at 4 to 5-year intervals, for the life of the plant, to check components.
The helium compressor and steam turbine blading should experience minimal wear in its running environment, and, with this length of service before inspection, will have undergone sufficient etress cycling to accurately indicate service life.
Specification SR 5.2.19 - IACM Diesel-Driven Pumps Surveillance DELETE SPECIFICATION SR 5.2.19 IN ITS FNTIRETY t-o on,
3:10 tU/
Amendment No. 21
5.2-18 Specification SR 5.2.20 - ACM Diesal Driven Generator Surveillanyy Requirement al The diesel driven ACM generator shall be checked weekly by startine, and obtaining design speed and voltage, b) The generator shall be tested monthly under load for a minimum of two hours. The load under this condition shall be at least 100% of design ACM equipment full load.
Basis for Specification SR 5.2.20 A weekly check of the Alternate Cooling Method generator to demonstrate its capability to start and a monthly test of the generator under load pro-vides adequate assurance that the Alternate Cooling Method generator will be available to supply electrical power under the highly degraded, loss of forced circulation situation.
Specification SR 5.2.21 - Hand Valve and Transfer Switch, Surveillance Requirements Those pneumatically and electrically operated valves and electrical transfer switches that must be manually positioned to implement the ACM l
shall be tested twice annually at an interval between tests to be not less than four (4) months, nor greater than eight (8) months.
Basis for Specification SR 5.2.21 In the event that the.ACM i
-t be implemented, it is necessary to positior. pneumatically and electric ally operated valves manually and to reposition electrical transfer switches. The test frequency and interval specified will assure operability in the event such operation is required.
358 288 Amendment No. 21
5 5-1 55 CONFINDER SYSTE4 - SURVEILIANCE REQUIREMENTS Applicability Applies to *.he surveillance of the reactor building (confinement) and the reactor building ventilation system.
Objective To ensure that the structure and camponents of the reactor building and ventilation systems are capable of minimizing the release of radio-activity to the atmosphere during potential abnormal conditions.
Specification SR 5 5.1 - Reactor Building, Surveillance Requirements The instrumentation which monitors the reactor building sub-atmospheric pressure vill be runctionally tested once every month and calibrated once a year.
Basis for Specification SR 5 5.1 The reactor building atmosphere is norrally =aintained slightly belov atmospheric pressure by the ventilation system (see FSAR Section 6.1 3 2).
This requirement minimizes the amount and consequences of airborne activity released from the plant under most conditions (see FSAR Section lb.12.8).
The leak rate of the bu.4 fing itself is not a significant parameter as is shown in FSAR Section 6.1.h.2.
Srecificatien SR 5.5.2 - Reactor Buildins Pressure Relief Device, Surveillance The reactor building overpressure relief system difi 'rential pressure switches 1 hall ta functionally tested on a monthly basis and calibrated annually.
The louver groups shall be individually exercised quarterly.
Amendment No. 21 358 289
5.5-2 Quarterly louver testing =ay be performed while the reactor is in opera-tion only if the following prerequisites are adhered to:
1.
Reactor shall be under nor=al steady s tate operating conditions.
2.
Pri=ary coolant pressure is within the normal envelope for existing conditions.
3.
Reactor building ventilation system is operating per Technical Speci-fications.
4.
No radioactive gas weste releases are in progress, nor is fuel handling being performed.
5.
No airborne activity above background as indica ted by the building ac tivity monitors.
6.
Area radiation monitors and local alar s are operab:e per Technical Sp ecifica tions.
7.
No surveillancc testing is being perfor=ed on the reactor ventila-tion system or the radiation monitoring systems.
8.
Only one segment (group o f louvers) of the louver system shall be tested at any given tine.
9.
Coc=unication shall exis t be tween personnel perforning the tes ts and the control room op erators.
10.
Capabili ty shall exis t to manually shut the louver panels.
11.
Tes ting o f the louver sys ten shall no t exceed a to tal duration o f six (6) hours in any one quarter.
12.
Non-cocpliance with any of the above conditions will require testing to be discontinued and the louver system will be re.mned to no rmal.
The reactor building relief (louver) sys te= shall be exercised annually.
358
'oa
~,L Amendment No. 21
5.5-2(a)
~
Basis for Specification SR 5.5.2 The reactor building pressure relief device is designed to protect the building in the event that pressure in the reactor building exceeds the turbine building pressure by 3 inches of water. The device consists of louvers installed in a number of individual modules operated by mechanical linkages to pneumatic actaators (see FSAR Section 6.1.3.4).
The specified test frequency shall ensure the operability of the reactor building relief system.
Specification SR 5.5.3 - Reactor Buildine Exhaust Filters, Surveillance The exhaust filters in the reactor building ventilation system shall be tested as follows:
a)
Samples from tha charcoal filters shall be laboratory tested after each 4400 hours0.0509 days <br />1.222 hours <br />0.00728 weeks <br />0.00167 months <br /> of operation of the unit, or following painting, fire, or chemical
- release in any ventilation ?.one communicating with the unit.
The results of laboratory carbon sample analysis from the unit shall show > 90% radioactive methyl iodide removed when tested in accordante with ANSI N510-1975 (130 C, 95% R.H.).
b) A halogenated hydrocarbon test shall be performed once per calendar year cr after each replacement o f a charcoal adsorber bank or af ter str;ctural maintenance on the filter housing.
Halogenated hydrocarbon removal by the charcoal filters shall be
_ 99%, when conducted at nor t.11 flow conditions in accordance with the applicable portions of ANSI N510-1975.
- Defined as any material which could reasonably be expected to interfere with the charcoal to adsorb methyl iodide.
Amendment No. 18 bbf []} '
4.2-22 Specification LCO 4.2.19 - Firewater Booster Pumps, Limiting Conditions for Operation There shall be one operable firewater booster pump during power operatiot..
Basis for Specification LCO 4.2.19 One firewater booster pump has suf ficient capacity to supply the This water for any one of the four helium circulator water turbines.
is adequate for safe shutdown cooling.
Amendment No. 21 358 29 1
4.10-2 (a)
TA3LE 4.10-3 SMOKE DETECTORS AND AL\\3MS NUM3ER (..'
- ".1:lD.UM !D3E2 ZONE DETECTORS OPERA 3LE
- 1) Control Roon 6
4
- 2) Anv411ary Electric Roon 3
2 31 Auxiliary Electric Roon 3
2
- 4) Auxiliary Electric Roon Return Air Duct 1
0*
5} 480 Volt Switchgear Raon 3
2
- 6) 480 Volt Switchgear Roon 3
2
- 7) Reactor " tiding "J" Wall aevation 4756' to 4791' 4
2
- 8) Reactor Building "J" Wall Elevation 4791' to 4829' 4
2
- 9) Reactor Building
'J" Wall Elevation 4829' to 4849' 2
1
- 10) Reactor Building "J" Wall Elevation 4849' to 4881' 2
1
- 11) Turbine Building "G" t all Elevation 4791' to 4811' 2
1
- 12) Turbine Building "G" Wall Elevation 4811' to 4829' 2
1
- 13) Reactor Building HVAC Return Air Duct at Elevation 4932' 1
1
- 14) Reactor Building MVAC Return Air Ducts 4
2 153 Reactor Building HVAC Return Air Duet 6
4
- To be returned to serrice as soon as practicable ifter loss of function.
358 293 Anandment No. 21
5.2-21 Specification SR 5.2.23 - Firewater Ecos".er Pump Surveillance _
Each firewater booster pump shall be tested annually by providing motive power to one water turbine drive in conjunction with the performance In addition each pump shall be functionally tested quarterly.
of SR 5.2.7.
The associated instruments and controls shall functionally be tested quarterly and calibrated annually.
Basis for Specification SR 5.2.23 During accident conditions described in Final Safety Analysis fire-Section 14.4.2.1, one of the firevater booster pumps and one
- Report, water pump are required to provide adequate core cooling. The specified testing interval is sufficient to ensure proper operation of the pump and associated controls.
358 291 Amendment No. 21
5.10-3 Specification c 5.10.6 - Foxed Water Spray System, Surveillance Requirement The manually operated valves actuating the fixed water spraf system providing supplementary fire protection in the 480 volt switchgear room, auxiliary electric room, and congested cable areas shall be opened annually. The flow path from the manual isolation valves to and including the spray nozzles shall be verified to be open annually.
Basis for Specification SR 5.10.6 Annual opening of the manually operated valves and verification of available flow path is sufficient to demonstrate capability to operate if required.
4mg JDU 7 pl D u
Amendment No. 21