ML20106E178

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Proposed Tech Specs Sections 1.0,4.3,4.4,4.5,4.6,4.7 & Associated Bases,Reflecting Conformance W/Ist Program
ML20106E178
Person / Time
Site: Pilgrim
Issue date: 10/30/1992
From:
BOSTON EDISON CO.
To:
Shared Package
ML20106E175 List:
References
NUDOCS 9211060180
Download: ML20106E178 (41)
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Text

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Attachment B Revised Technical Specification Pages 9211060100 921030 PDR ADOCK 03000293 p- PDR _ _

- . - - - . - - - . - . _ - - ~ . ~ . - - _ . - _ . . . . - - . . - . . ~ . - . - . .

1,0' DLUhlJ10!15 (Cont'd)

1. At least ono thor in each acces opening is closed, ,

1

2. The standby gas treatment system is operable. j
3. All automatic ventilation system isolation valves are operable or secured in the isolated position.

-i

0. Operatina Cx lg Interval between the end of one refueling outage  !

and the end of the next subsequent refueling outage. l 1

. 7. Egfuelina freggynciet

1. Refuelino 0qtagg Refue? 'ng outage is the period of time l between the shutdown of the unit prior to a refueling and the l startup of the plant af ter that refueling. For the purpose of designating frequency of testing and surveillance, a refueling ,

outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion of the  ;

previous refueling outage, the required surveillance testing  ;

need not be pet .ormed until 1.he .next. regularly scheduled . -

outage.

2. Etthglina Interval Refueling interval applies only to ASME .

Code,Section XI IWP and IWV surveDiance tests. For the purpose of designating frequency of these code tests, a refueling -interval shall mean at least once every 24 months.

Q. AlterqLion of the REAG10r Core - The act of noving any component in

! the region above the core support plate, below the upper grid and  !

within the shroud. Normal control rod movement with the control rod 3 drive hydraulic system is not defined as a core alteration. Normal i movement of in-core instrumentation is not def ti.ed as a core L alteration.

R. .Beactor_Yanel Pr3.isure - Unless otherwise indicated, reactor vessel aressures listed in the Technical Specifications are those measured ay the reactor vessel steam space detectors.

  • S. Ihermal Parameter 1 l

i 1. ZijnimunLCrf tiq1L.19wer -RatJ.g. (MCEE), the value of critical L power ratio associated with the most . limiting assembly in the - 7 L reactor core. Critical Power Ratio-(CPR) is the rat 'o of that-

-power in a fuel assembly, which-is calculated to cause some point in:the assembly to experience boiling transition,'to the actual assembly- operating power.

2. Ing.sition Bo.J.ljng - Transition bolling means the betling .

regime between. nucleate and film boiling. Transition boiling is.the regime in which h th nucleate and film boiling occur inter.alttently wib nesther type being completely stable.

3. 10MllcA!LtM.hf101 - The ratio of the fuel rod surface h9at l- flux to the h4at flux of an average' rod in an identical; +

l geometry feel assembly operating at the core average bundle

-power.

/

1 Amendment No. 15 J4 L

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, 1.0 DLOR1110M (Continued) 1 U. MryeillMce Freaucasy - f ach Surveillance Requirement shall be i performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval, lhe Surveillance frequency establishes the limit for which the specified time interval for Surveillance Requirements may be extended. it permits an allowable extension of the normal surveillance interval to f acilitate surveillance schedule and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance interynis beyond that specified for surveillances that are not performed during refueling outages. The limitation of Definition "U" is based on engineering judgment and the recognition that the most probable result of any particular surveillance being perfortned is the verification of conformance with the Surveillance Requirements. This provision in sulficient to ensure that the reliability ensured ther. ugh surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

V. htyfil.lMC.LLn12ty.al - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and examinations to be performed upon an instrument or component when it is required to be operable These tests may be waived when the instrument, component, or sptem is not required to be operable, but the instrument, component, or systcm shall be tested prior to being declared operable. The operating cycle interval is 18 months and the 25% tolerance given in Definition "U" is applicable. The refneling interval is 24 months and the 25% tolerance specified in definition "U" is applicable.

W. ElfD_llTDItssinn Wat_tr_1y1(se - A fire suppression water system shall consist of: a water source (s); gravity tank (s) or pump (s);

and distribution piting with associatui sectionalizing coatrol or isolation valves. Such valve: shall include hydrant por.t indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, hose staridpipe or spray system riser.

X. fdElered Test flajis - A staggered test basis shall consist of: (a) a test schedule for a systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals; (b) the testing of one system, subsystem, train or other designated components at the beginning of each subinterval.

Y. hurg_Cheth - A sourca check shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

i Amendment No. 425-89,-128 Sa

LIM 111NG CONDI UON FOR OPERATIQlj SMLVU1LAELB1QVIREMENT E. hattivitLhonlin E. hactivity AnonUn The reactivity equivalent of During the startup test progren the difference between the and startups following actual critical rod refueling outages, the critical configuration and the expected rod configurations will be configuration during power compared to the expected operation shall not exceed 1% configurations at selected t.K . If this limit is exceeded, operating conditions. These the reactor will be shut down comparisons will be used as until the cause has been base data for reactivity determined and corrective monitoring during subsequent actions have been taken if such power operation throughout the actions are appropriate. fuel cycle. At specific power operating conditions, the

f. If Specifications 3.3.A through critical rod configuration will D above cannot be met, an be compared to the orderly shutt ;wn shall be configuration expected based -

initiated and the reactor shall upon appropriately corrected be in the Cold Shutdown past data. This comparison condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, will be made at least every Specifications 3.3.A through D full power month, above do not apply when there is no fuel in the reactor vessel.

G. 'icram Disshusttyolume G. SgtaAJ111. charge Vohtme

1. The scram discharge volume 1. Scram discharge volume drain & vent valves shall drain and vent valves; be operable whenever more than one operable control a. Verified open at least rod is withdrawn, once per mor.th.
2. If any of the scram b. Test as specified in discharge volume drain or 3.13. These valves vent valves are made or may be closed found inoperable an orderly intermittently for shutdown shall be initiated testing under and the reactor shall be in administrative Cold Shutdown within 24 control, hours.
2. During each refueling interval verify the scram discharge volume drain and vent valves; a) Close within 30 seconds after receipt of a reactor scram signal and b) Open when the scram is reset.

Amendment No. 65 85

LitiLUM_CQHELHONS fOR OP@&lLOS M M LilL6NCE RE0VIRLIK!ilS.

. 3.4 1W10_0LL10VID CotLTROLSISIG 4.4 HaRQf!Lu14l0_00J11ROLFSTW bl!pli.qabi1itv: ej!pl[cability:

Applies to the operating Applies to the surveillance status of the Standby Liquid requirements of the Standby Control System. Liquid Control System.

(Lbitcliy.e: Qhjf.c_tjyg:  ;

To assure the availability of To verify the operability of a system with the capability the Standby Liquid Control to shutdown the reactor and System. f maintain the shutdown condition without the use of  ;

control rods.

SRtci ficatign: .Speqjitg.ajign:

A. Kong 31 System _Ay3]ld)jltty A.-- MgImALHyllem Availability i

1. During seriods when fuel The operability of the is in tie reactor and Standby Liquid Control prior to startup from a System shall be verified by cold condition, 1.he the perfonnance of the Standby Liquid Control following tests: .

System shall be.

2 operable, except as 1. when tested as specified .

specified in 3.4.8 in 3.13 verify that each below. This system need pump delivers at least not be operable when the 39 GPM against a system reactor is in the Cold head of 1275 psig.

Shutdown Condition, all operable control rods 2. As required below:

are fully inserted and Specification 3.3.A is a. Once each refueling .:

met, interval while i testing as specified in 3.13 verify the system relief valve .

set point of 1425  ;

psig i 43 psig.

95 .7 m

LIMITING C0HDITIONS FOR OPERATION- SURVEILLMU RE0VIREMENTS

'3.4 Ma@!!LUMilM0ERQL1YHfE 4.4 STANDBY L10VJD_C.QNTROL SYSIEM

b. At least once during each refueling-interval, while testing as specified l in 3.13, manually ,

initiate one of the "

Standby Liquid Control System loops and pump demineralized water -

into the reactor vessel.

This test checks explosion of the -

charge associated with-the tested. loop -

proper operation of-  ;

the valves, and pump: a capacity.- The- 4 replacement charges to be installed will i be selected from the san 9 manufactured batch as the tested charge.

c. When testing to satisfy requirement 4.4.A.2.b, both  ;

~

systems, including both explosive i valves, shall be-tested in the course ,

of two refueling i B. Operation with inolerable intervals.

i- .GER0Ht'ah:

l B. -Surveillancf wi &

l. From and after the JDoperable ComponRuh l

date that a redundant component is made or 1. When a'componenteis-L found to be found to be' t inoperable, inoperable, its Specification 3.4.A.1 - redundant component:1 -

, shall be considered shall be demonstrated .

~

. fulfilled and - to be operable . .

. continued operation immediately and daily l permitted provided thereafter unti.l the-that th9 component _is _ inoperable component' returned to an it repaired. ->

l operable condition .;

within seven days. -

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3 -

90- <: >

B _ a.  ; _ . _ w. . 2,_. .u._ .._..u-. _.

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CASES: i 3.4 & 4.4 SJR@.Y_UqtRD CONTROL SHIfjj A. The requirements for SLC capability to shutdown the reactor are identifitd via the station Nuclear Safety Operational Analysis (Appendix G to the FSAR, Special Event 45). If no more than one i operable control rod is withdrawn, tha basic shutdown reactivity '

requirement for the core is satisfied and the Standby liquid l Control system is not required. Thus, the basic reactivity l requirement for the cere is the primary determinant of when the l standby liquid control system is required. The design objective of 1 the standby liquid control system is to provide the capability of '

bringing the reactor from full power to a cold, xenon free shutdown condition assuming that none of the withdrawn control rods can be J inserted. To meet this objective, the Standby Liquid Control i system is designed to inject a quantity of baron that-produces a minimum concentration equivalent to 675 ppm of natural boron in the reactor core. The 675 ppm equivalent concentration in the reactor core is required to bring the reactor from full pcwer to at least a three percent Ak subcritical condition, cor 'dering the hot to cold reactivity difference, xenon poisoning etc. The system will inject- -l this boron solution in less than 125 minutes. The maximum time requirement for inserting the baron solution was selected to override the rate of reactivity insertion caused by cooldown of the i 4

reactor following the xenon poison peak. )

The Standby liquid Contral system is also required to meet l 10CFR50.62 (Requirements for Reduction of Risk from Anticipated Transients Without Scram (A1WS) Events for Light-Water-cooled . ,

Nuclear Power Plants). ihe Standby liquid Control system must have -i the equivalent control capacity (injection rate) of 86 gpm at 13 percent by wt natural sodium pentaborate for a 251" diameter j reactor pressure vessel in order to satisfy 10CFR50.62 requirements. This e ,uivalency requirement is fulfilled by a .

combination of conctatration, 810 enrichment and flow ratn of .l sodium pentaborate solution. A minimum 8.42% concentration and 54.5% enrichment of Bio isotope at a 39 GPM pump flow rate i satisfies the ATWS dule (10CFRE0.62) equivalency requirement. 1 Because the concentration / volume curve has been revised to reflect ,

the increased Bio isotopic enrichment, an additional requirement I has been added to evaluate the solution's capability to maet the original design shutdown criteria whenever the B10 enrichment -

requirement is not met. 3 Testing the pumps and valves in accordance with ASME B&PV Code 1 Section XI (Articles IWP and IWV, except where specific relief is <

granted) adequately assesses component operational readiness. The ,

only practical time to fully-test the liquid control system is - i during a refueling outage. Various components of the system are individually tested periodically, thus making more frequent testing

!- of the entire system unnecessary.

Amendment No. 102,-135 100- '

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1 BASLS: I 3.4 & 4.4 SJANOBY L10VID (Mi&OLS_512 (Cont'd)  ;

I  ;

B. Only one of the two standby liquid control pumping loops is needed for operating the system. One inoperable pumping circuit does not i immediately threaten the shutdown capability, and reactor operation +

can continue while the circuit is being repaired. Assurance that .

the remaining system will perform its intended function and that '

the long term average availability of the system is not reduced is l obtained for a one out of two system by an allowable equipment out of service time of one third of the normal surveillance frequency. i This method determines an equipment out of service time of ten '

days. Additional conservatism is introduced by reducing the ,

allowable out of service time to seven days, and by increased 4 testing of the ope able redundant component.

C. The quantity of B 10 storedintheStandbyliquidControlSgtem f Storage Tank is sufficient to bring the concentration of B in the reactor I; the point where the reactor will' be shutdown and to provide a minimum 25 percent margin beyond the amount needed to--

shutdown the reactor to allow for possible imperfect mixing--of the  ;

chemical solution in the reactor water. .

~

Level indication and alarm indicate whether the solution volume has-changed, which might indicate a possible solution concentration change. Test intervals for level monitoring have been established in consideration of these factors. Temperature and liquid level alarms for the system are annunciated in the control room. ,

The solution shall be kept at least 10 F above-the maximum -

saturation temperature to guard against boron precipitation.

Minimum solution tempereture is 48 f. This.is 10'F above the saturation temperature for the maximum allowed sodium pentaborate concentration of 9.22 Wt. Percent.

Each parameter-(concentration, pump flow rate, and enrichment) isf tested at an interval consistent with the potential for that parameter to vary and also to assure proper equipment performance, Enrichment testing is required when material is received and when chemical udition occurs since change cannot occur by any process  !

other tha, the addition of new chemicals to the Standby Liquid. 1 Control solution tank.

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- ,LIMITlfiG'C0t?! TION FOR OPERATION j.URVEll.LRfj[ L ELQQ M M W I

' 3.5 fMLat10_GMIAllLMLtiT C00.Lmn 4.5 CORE AND (qttlall4 MENT C001ING l 1Y1185- SlS10iS eppJ.kability Applicahil_ity Applies to the' operational status of Applies to the Surveillance the core and suppression pool Requirements of the core and cooling systems. suppression pool cooling systems which are required when the corresponding Lirriting Condition for operation is in effect.

1 DAiE11v3 @lE.11YS To assure the operability of the. To verify the operability of the core and suppression pool cooling core and suppression pool cooling systems under all conditions for systems under all conditions for which this cooling capability is an which this cooling capability is an  ;

essential response to station essential response to station abnortnali ties, abnormalities. i SEEll.iLiLtior1 Spfcif kalj.gn

, A. Cpre Sp.tay and.LPCI Systequ A, Carl.Satav and LPCI Systent f

1. Both core spray systems shall 1. Core Spray System Testing, be operable whenever irradiated fuel is in the vessel and prior Jim 13 Ertauency-to reactor startup from a Cold Condition, except as specified a. Simulated Once/ Operating in 3.5.A.2 below. Automatic - Cycle Actuation test,
b. Pump Operability When tested 4

as specified in 3.13- ,

verify that each core. r spray pump delivers at least.  ;

3300 GPM i

- against_a system head correspond-ing to a reactor-vessel ,

pressure of 104 psig

c. Motor 0perated. As specified Valve Operability in 3.13  ;

L d .' Core.5 pray Header l

Ap Instrumentation

- Amendment No._ 425-62,-1145 -135 _103-a 5

v -*b gy. - e,g ,qs.o,, q.ij w y- p9-7+:- m-g5 --p.y- y 3 s- p,7p- me-fy ye gri.f-- -p m ++ r 6 7 y v5-49g- w.am vr--g-pmem*' y *- r -Ws v*ww' - " 4'

Lililllh0 00l(QlllD1L[0R OPERAT10B jiURVEILLANCE RE01tLRJE"T .

3.5.A .CDIs .Sy ny_and_LPCI Systemi 4.5.A Core Spn y_Apd LFil Sygismi '

(cont'd) (cont'd)  ;

Check Once/ day 4

Calibrate Once/3 months Test Step Once/3 months .

2. From and after the date that 2. This section intentionally left one of the core spray systems blank is made or found to be ,

inoperable for any reason, 3. LPCI system Testing shall be as e continued reactor operation is follows:

permissible during the succeeding seven days, provided t, Simulated Once/ Operating that during such seven days all Automatic Cycle active components of the other Actuation core spray system and active lest components of the LPCI system and the diesel generators are b. Pump Wh9n tested operable. Operability as specified

  • in 3.13 i
3. lho LPCI system shall be verify that operable whenever irradiated each LPCI ,

fuel is in the reactor vessel, pump delivers and prior to reactor startup 4800 GPM at a ,

from a Cold Condition, except head across ,

as specified in 3.5.A.4 and the pump of-3.5.f 5. at least 380 "

ft

4. from and aftet the date that .

the LPCI system is made or c. Motor Operated As specified  ;

i found to be inoperable for any valve in_3.13  :

reason, continued reactor operability-operation is perm hsible only during the succeeding seven.

l days unless it is sooner made '

i operable, provided that during .

such seven days the containment cooling system (including 2 ,

LPCI pumps) and active .r components of both core spray systems, and the diesel generators: required for-l operation of such components if L no external' source of power l- were available shall be operable.

5. If the requirements of 3.5.A cannot be met, an orderly ,.

shutdown of the reactor shall be initiated.and the reactor-l shall be in the Cold Shutdown i Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

/ . .

I _ Amendment No. 425-625-1115-114 -135_ 7 104.

LlHlilNG CONDITION FOR OPERATIOf_4 LVRyl[LQti[JEgyJBQiDI1 ,

3.5.B Gmt13inment Coolina System 4.5.8 [9A111.rment Cgglina syst.es

1. Except as specified in 3,5,8.2 1. Containment Cooling system and 3.5.f.3 below, both Testing shall be as follows:

containment cooling system loops shall be operable 11em Frecuency

- whenever irradiated fuel is in the reactor vessel and reactor a. Pump When tested coolant temperature is greater Operability as specified than 212"f, and prior to in 3.13 verify reactor startup from a Cold that each Condition. RBCCW pump delivers 1700

2. from and after the date that GPM at 70 ft one containment cooling system TDH and loop is made or found to be each SSW pump inoperable for any reason, delivers 2/00 continced reactor operation is GPM at 55 ft permissible only during the TOH -

succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless such system loop is sooner made b. Valve As specified operable, provided that the Operability in 3.13 other containment cooling system loop, including 'ils c. Air test on once/5 years associated diesel generator, is drywell and operable. torus headers and nozzles

3. If the requirements of 3.5.8 cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l-L Amendment'No. 425 -44 5-114 -135 5 106

LIM 11EdO!!ullLQN FOR OPERA 11Q!i EUM1111ANCE REQR1RUiull C. HECL._S_Y.11en) C. HE[LSnte!!)

1. The HPCI system shall be 1. HPCI system testing shall be operable whenever there is performed as follows:

irradiated fuel in the reactor vessel, reactor pressure is a. Simulated Once/ operating greater than 150 psig, and Automatic cycle reactor coolant temperature is Actuation greater than 365 f; except as Test specified in 3.5.C.2 below,

b. Pump Oper- When tested
2. from and after the date that ability as specified the HPCI system is made or in 3.13 verify found to be inoperabic for any that the HPCI reason, continued reactor pump delivers operation is permissible only at least 4250 during the succeeding seven GPM for a days unless such system is system head sooner made operable, providing corresponding that during such seven days all to a reactor active components of the ADS pressure of syst em, the 1101C system, the 1000 psig iPCI system and both core spray systems are operable, c. Motor Operated As specified Valve Oper- in 3.13
3. If the requirements of 3.5.C ability cannot be met, an orderly shutdown shall be initiated and d. Flow Rate at Once/ operating the reactor pressure shall be 150 psig cycle verify reduced to or below 150 psig that the HPCI within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, pump delivers <

at least 4250 GPM for a system head corresponding to a reactor pressure of 150 psig The llPCI pump shall deliver at least 4250 gpm for a system head corresponding to a reactor pressure of 1000 to 150 psig.

Amendment No. 425 -44 5 108,-114,-136 107

1 LIMITING C0f01110N FOR OPERATION Mlay11LLANCE RE0Vjl@fjl 3.5.0 fleictor (ntf Isolation 4.5.0 Reactor Corthnldinn I Coolina (RCl(j_jiyltem - Coolina- [RCIC)_SJittm  !

1. 1ho RCIC system shall be 1. RCIC system testing shall be operable whenever there is performed as follows:

irradiated fuel in the reactor vessel, reactor pressure is a. Simulated Once/ operating  ;

greater than 150 psig, and -Automatic cycle i reactor coolant temperature is Actuation greater than 365'F; except as Test specified in 3.5.0.2 below. 1

b. Pump When tested c
2. from and after the date that the Operability as specified RCICS is made or found to be in 3.13 verify inoperable for any reason, that the RCIC .

continued reactor power pump delivers operation is permissible only at.least 400 during the succeeding seven days GPM at a t provided that during such seven system head days the HPCIS is operable, corresponding to a reactor

3. If the requirements of 3.5.D pressure of cannot be met, an orderly . .

1000 psig >

shutdown shall be initiated and the reactor pressure shall be c. Motor As specified reduced to or below 150 psig Operated in 3.13 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Valve Operability

d. Flow Rate at Once/ operating 150 psig cycle verify  ;

that the RCIC ,

pump delivers at least 400 GPM at a '

system head corresponding to a reactor pressure of  ;

150 psig The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor' pressure of 1000 to 150 psig.

LAmendmentNo. 425 109,-114,-135 108

MSLS:  !

l 3.5.8 Containment Coolina System i t

The containment cooling system for Pilgrim i consists of two independent loops  !

cach of which to be an operable loop requires one LPCI pump, two RBCCW pumas,  ;

and two SSW pumps to be operable. There are installed spares for snargin aaove the design conditions. Eachsysgemhasthecapabilitytoperformits l function; i.e., removing 64 x 10 Blu/hr (Ref. Amendment 18), even with some i system degradation. If one loop is out-of-service, reactor operation is ,

permitted for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. j With components or systons out of service, overall core and containment cooling reliability is maintained by the operability of the remaining cooling -!

equipment.

I Since some of the SSW and RBCCW pumps are required for normal operation, capacity testing of individual pumps by direct flow measurement is i impractical. Pump operability will be demonstrated during normal system operatior, and/or when system conditions allow capacity and performance testing in accordance with 3.13.

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' Amendment No. 135 115 e-

-9 wwwie -- ==-me- wn w-mera te,,eew eis , +- -, m-e a m,e,=esw- sr -p *c, e ma-wTT'-m-- + --7* +- r +--t-'f' T 'T"- "r-sw,--w w+r.c -we --yy v9vyrwv

115fS:

l 4.5 [ge_a_nL(_gnLairmnt (ooling_Sntgm3 Surveilla.rne I rmtueric_in The testing interval for the core and containment cooling systems is based on industry practice, quantitative reliability analysis, judgment and practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation would result in pumping cold water into the reactor vessel which is not desirable. Complete ADS testing during power operation causes an und sirable loss-of-coolant inventory, lo increase the availability of the core and containment cooling systems, the components which make up the system; i.e., instrumentation, pumps, valves, etc., are tested frequently. The pumps and mater operated valves are tested in accordance with ASML B&PV Code,Section XI (lWP and IWV, cr. cept where specific relief is granted) to assure their operability. The frequency and methods of testing are described in the Pf4P5 IST program. The Pl4PS 151 Program is used to assess the operational readiness of pumps and valves that are safety-related or important to safety. When components are tested and found inoperable the _

impact on system operability is determined, and corrective action or Limiting Conditions o' Operation are initiated. A simulated automatic actuation test once each cycle combined with code inservice testing of the pumps and valves is deemed to be adequate testing of these systems.

The surveillance requirements provide adequate assurance that the core and containment cooling systems will be operable when required.

Amendment No. 135 122

~ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ . . _ _ _

UlilllEl0RD111pm FOR OPERATION MyEllLANCE REQUIREMENTS i 3.6.D Safety and Relief Valvn 4.6.0 Safety and Relief Valv u t

1. During reactor power 1. Testing of safety and operating conditions and relief / safety valves shall be
  • prior to reactor startup from in accordance with 3.13.

a Cold Condition, or whenever reactor coolant pressure is 2. At least one of the greater than 104 psig and relief / safety valves shall be '

temgeraturegreaterthan disassembled and inspected 340 f, both safety valves and each refueling outage.

the safety modes of all relief valves shall be 3. Whenever the safety relief operable. The nominal valves are required to be setpcInt for the operable, the discharge pipe relief / safety valves shall be temocrature of each safety .

selected between 1095 and relief valve shall be logged 1115 psig. All relief / safety daily.

valves shall be set at this nominal setpoint i 11 psi. 4. Instrumentation shall be lhe safety valves shall be calibrated and checked as set at 1240 psig i 13 psi, indicated in Table 4.2.F.

2. If Specification 3.6.0.1 is not met, an orderly shutdown shall be initiated and the reactor coolant pressure shall be below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Note:

Technical Specifications 3.6,0,2 - 3.6.D.5 apply only when two Stage Target Rock SRVs are installed.

3. If the temperature of any safety relief discharge pipe exceeds 212 F during normal reactor power operation for a period of greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, an engineering evaluation shall be performed justifying continued operation for the corresponding temperature increases.

Amendment No.-421-56r.881-133,J139 _

126_.

Lili[JE(G CONDJJ10R$_[QR OPERAI10lt . SURVEILLANCE P[0VIREMENTS 3.7. A Primary Containacitt _(Con'.tl 4.7.A Primary Containment (Con't)

PI.llEELC.ontainnten.t_liplAllpn Val.yn .p rimary CQntainment isol,ation Valves 2.b. In the event any automatic 2.b.1 The primary containment Primary Containment Isolation isolation valves Valve becomes inoperable, at surveillance shall be least one containment performed as follows:

isolation valve in each line '

having an inoperable valve a. At least once per shall be deactivated in the operating cycle the isolated condition. (This operable primary requirement may be satisfied containment isolation by deactivating the inoperable valves that are pnwer .

valve in the isolated operated and condition. Deactivation means automatically initiated to electrically or shall be-tested for pneamatically disarm, or simulated automatic otherwise secure the valve.)* . initiation-and closure '

times.

b. Test primary containment isolation valves:
1. Verify power operated primary containment isolation valve operability as specified in 3.13.  :
2. Verify main steam isolation valve operability as specified in 3.13,
c. At least twice per week the main steam line power operated isolation valves shall be exercised by partial closure and subsee.uent reopening,
d. . Verify reactor coolant system instrument line flow check valve-operability as specified in 3.13.
  • lsolation valves closed to satisfy 2,b.2 Whenever a primary  ;

these requirements may be reopened containment automatic s on an intermittent basis under ORC isolation valve, is approved administrative controls. inoperable, the position of the isolated valve in each line having an inoperable valve shall be recorded daily.

~ Amendment No. 1136136 155a

~ .. . . . .

LIMITING CONDITION FOR OPERAT10B SE KILLANCE REQUIREMENIS 3.7 Primary Containme01 4.7 Primary Containment

3. PIgnyre Suppression Chamber - 3 Pressure S@pression Chamber -

. Reactor Buildina Vacupm Breakers Reactor Buildina Vacuum Breakers

a. Except as specified in 3.7.A.3.b below, two pressure suppression a. Verify operability of the chamber - reactor building vacuum pressure suppression chamber-breakers shall be operable at all reactor building vacuum times when primary containment breakers as specified in 3.13.

integrity as required. The setpoint of the differential b. Check the associated pressure instrumentation which instrumentation including set actuates the pressure suppression points for proper operation chamber - reactor building every three months. 1 breakers shall be 0.5 psig. I

b. From and after the date that one '

of the pressure suppression chamber - reactor building vacuum breakers is made or found to be  ;

inoperable for any reason, reactor operation is permissible only during the succeeding seven days unless such vacuum breaker is sooner made operable, provided that the repair procedure does not violate primary containment integrity.

4. DrvWell-Pressure Suporession
4. Drywell-Pressure Suporession Chamber Vacuum Breakers Chamber Vacuum Breakers
a. When primary containment is a. Periodic Operability Tests required, all drywell-pressure suppression chamber vacuum breakers shall be operable except during testing and as stated in Specifications 3.7.A.4.b, c and -

d, below. Drywell-pressure (1) Once each month each drywell-suppression chamber vacuum pressure suaoression chamber breakers shall be considered vacuum brea<er shall be sperable if: exercised and the operability of the valve 'and installed

[ 1) The valve is-demonstrated to open position indicators and alarms with the applied force of the verified, installed test actuator as indicated by the position (2) A drywell to suppression switches and remote position chamber differential pressure-indicating lights. decay rate test shall be conducted at least every 3 2).The valve shall return by gravity months.

when released after being opened by remote or manual means, to within 3/32" of:the fully closed-position.

Amendment No. 68 156 l

1 - -

- . - - - - - .~,_ - _. - _ . - - . --

LIMITING CONDlTION FOR OPERATIDE SURVEllLANCE RE0VIREMENTS .

3.7 Primary Containment 4.7 Primary Containment l

3) Neither of the two position alarm systems which annunciate on Panel C-7 and Panel 905 when any vacuum breaker opening exceeds 3/32" are in alarm,
b. Any drywell-suppression chamber b. During each refueling interval:

vacuum breaker may be non-fully closed as determined by the (1) Each vacuum breaker shall be position switches provided that tested to determine that the .

the drywell to suppression disc opens freely to the touch chamber differential decay rate and returns to the closed is demonstrated to be not position by gravity with no greater than 25% of the indication of binding.

differential pressure decay rate for the maximum llowable (2) Vacuum breaker position bypass area of 0.2ft switches and installed alarm systems shall be calibrated and

c. Reactor operation may continue functionally tested.

provided that no more than 2 of the drywell-pressure (3) At least 25% of the vacuum suppression chamber vacuum breakers shall be visually

breakers are determined to be inspected such that-all vacuum L inoperable provided that they breakers shall have been are secured or known to be in inspected following every

! the closed position. fourth refueling interval. If deficiencies are found, all

d. If a failure of one of the two vacuum breakers shall be installed position alarm visually inspected and systens occurs for one or more deficiencies corrected.

vacuum breakers, reactor operation may continue provided (4) A drywc11 to suppression that a differential pressure chamber leak rate test shall decay rate test is initiated demonstrate that the immediately and performed every differential pressure decay 15 days thereafter until the rate does not exceed the rate failure is corrected. The test which would occur through a 1 shall meet the requirements of inch orifice without the Specification 3.7.A.4.b. addition of air or nitrogen. $

5. Oxvoen Concentration 5. Oxygen Concentration
a. The primary containment The primary containment oxygen atmosphere shall be reduced to concentration shall be measured less than 4% uxygen by volume and recorded at least twice with nitrogen gas during weekly.  ;

reactor power operation with

-reactor coolant pressure above 100 psig, except as specified in'3.7.A.5.b.

1 1

l

-Amendment No, 87 157 1 . - . - . -. - _

LilijlllfG CORQ1iIONS fOR OPERATIQB SMym1MCE REQUIREMENIS 3.13 BCiLRYlLLLQDLlLSTIHQ 4.13 .1HSrRVICLc0nLitsiluc l iPPil(6011111: AdilLABilllY:

Applies to ASME Code Class 1, 2 and Applics to the periodic testing 3 or equivalent pumps and valves, requirements of ASME Code Class 1, 2 and 3 or equivalent pumps and valves.

OHlLL11YL: ODALC11YI:

To assure the operational readiness To assess the operational readiness of ASME Code Class 1, 2, and 3 of safety and safety-related pumps (Safety Related) or equivalent and valves by performant.e of (important to safety) pumps and inservice tests, valves.

SPLCimal103: LPE(l[1[AllM:

A. R4SERVICL(0DL 1ESilRG _ Of A. ULSIRylCL CODE TESTING Of WMPS AND LALYLS B)MPS.MU VALVES

1. Based on the facility 1. Inservice Code Testing Commercial Operation Date, activities shall be performed Inservice Code Testing of in accordance with Section XI  %

safety and safety-ralated of the ASMF Boiler and pumps and valves shall be Pressure Vessel Code and performed in accordance with applicable Addenda as required  ?

the ASME Boiler and Pressure by 10CFR50.55a(g), with the Vessel Code,Section XI exemptions and alternate

" Rules for Inservice testing that have been Inspection of Nuclear Power approved by the NRC pursuant Plant Components" Subsections to 10CFR50.55a(9)(6)(i).

lWP and IWV as reqdired by These exemptions and alternate 10CfR50.55a(g), except where testing are included in the specific relief has been FNPS Inservice Testing granted by the NRC pursuant Program, to 10CFR50.55a(g)(6)(i).

2. Test frequencies for Code Terminology when perfor.ning Inservice Test activities.

[ ode Terminolony Frenuencies Weekly 7 Days Monthly 31 Days Quarterly or 3 Mths 92 Days Semiannually /6 Mths 184 Days 9 Months 276 Days Yearly / Annually 366 Days Biannual /2 Yrs 732 Days

3. The provisions in Definitions (1.0) for REFUELING INTERVAL, SURVEILLANCE FREQUENCY, and SURVEILLANCE INTERVAL are applicable to Code testing and Amendment No. 205g

L I M I T I N G_C.0lGLT3fjlJOR .01G&I1011 EURVEILLANCf RE0VIREMEN'il 3.13 LilSLRVICE CODE TESTil4Q 4.13 INSERVICE CODE TESTING to the above frequencies for performing Code testing activities.

4. Performance of Code testing shall be in addition to other specified Surveillance Requiremi.v;s.
5. Nothing in the ASME Boiler and Pressure Vessel Code- shall sdpersede the remirements of Technical Specifications.

8 Amer.dment No.

205h {

H

u 3.13_'and 4,13 in. service _ Code Testin9

- The Limiting Conditions 'for Operation establishes ~ the requirement that inservice testing of ASME Code Class 1, 2, and 3: pumps. and valves shall be 1 performed in accordance with the periodically updated edition of Section-XI:of-the ASME Boiler _and Pressure Vessel Code and-Addenda as required by 10CFR50, Section 50.55a(g). These requirements apply except when relief has-been requested pursuantEto 10CFR50.55a(g)(6)(1) and granted by the NRC. The NRC may grant relief pur;uant to 10CFR50.55a(a)(3)(1),10CFR50.5Sa(3)(ii) or:

10CFR50.55a(g)(6)(i),

The detailed procedures for testing of pumps and valves are documented in the j PNPS Inservice Testing Program.

This specification includes a clarification of the frequencies for performing

. the testing activities required by Section XI of the ASME Boiler and Pressure

-Vessel Code and applicable Addenda. This clarification:is provided to ensure -

consistency in Surveillance Frequencies throaghout the Technical Specifications and to remove any ambiguities relative to the frequencies-for.

performing the required inservice testing activities. ,

Under the terms of this Specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and _ Pressure.

Vessel Code and applicable Addenda. For example:

  • Technical Specifications require components to be declared operable prior to entry into an operat.onal mode. The ASME 8&PV Code provision which allows pumps -and valves to be _ tested up toi one week: after return to normal operation is -superseded (and not. ,

allowed) by the more restrictive requirements of -Technical Specifications.

+ The allowancc for a valve to be incapable of performing its specified function for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> before being declared-inoperable is superseded (and not allowed) by the more restrictive:

Technical Specification definition of eperability which doesinot-allow a grace period.

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1.0 DEFNOTIONS_(Coat'd)

! 1. At least one door in each access opening is closed.

Ii .

2. L ' standby gas treatment syste n is operable.

I

3. All s.utomatic ventilation system isolation valves are operable or secured in the isolated position. ,

. 0. Operatino Cycle _ - Interval between the end of cne refueling outage '

ind tne end of the next subsequent refueling outage. ~

P. {t.h'f%Refueli'no Rf9venha Outace - Refueling cuthge is the period of time between

//. the snutcown of the unit prior to a refueling and the startup of 1

the plant after that refueling. For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled outage; however, where such outages occur within 8 months of the completion ~of the previous refueling outage, the required surveillance testing need not be perfo med until the next regularly scheduled cutage.

m Alteration of the Reactor Core - The act of mo'ving any component Q.

Tii the region above the core support plate, below the upper grid and within the shroud. Normal control rod movement with the control rod drive hydraulic system is not defined as a core alteration.

hormal movement of in-core instrumentation is not defined as a core alteration,

{

R.

Reactor Vessel Pressure - Unless otherwise indicated', reactor vessel pressures hstec in the Technical Specifications are those i

measured by the reactor vessel steam space cetectors.

, S. Thermal parameters Minimum critical Power Ratio (McW) - the value of critical 1.

power ratio associated with the most limiting assembly in the Critice.1 Icver Ratio (CER) is the ratto of that reactor core. 2 power in a fuel assembly, which is calculated to cause some point in the essembly to experience boiling transition, to the ,

actual assembly operating power.

2. ,rrensition Boiling - Transition boiling means the boiling Transition boiling regime between nucleate and fi.bn boiling.

is the rerf.e e in which both nucleate and film ' coiling ocer.r inter =itt%Lly with ueither type being completely stable.

3 Total peakinc Facter - The ratio of the fuel rod surface heat Y1ux to the teat flux of e.h average rod in an identical geometry fuel asce:bly operating at the core e.vera.ge bimile pover.

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1.0 QUINi}{QM (Continued)

V 0. Sprveillan10.lriquLnly - Each Surveillance Requirement shall-be performed within the specified surveillance interval with a maximum ('

allowable extension not to exceed 25 percent of the specified ,

surveillance interval. j r

The Surveillance frequency establishes the limit for which the 8 i'

sper.ified time interval for Surveillance Requirements may be i extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance schedule and {

consideration of plant operating conditions that may not be suitable i for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for surveillances that i are not performed during refueling outages. The limitation of Definition "U" is based on engineering judgment and the recognitinn that the most probable result of any particular surveillance being performed is the verification or conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not

significantly degraded beyond that obtained from the specified  !

l surveillance interval.

v. Survlllance i Interval - The surveillance interval is the calendar time between surveillance tests, checks, calibrations, and a

examinations to be performed upon an instrument or component when it is required to be operable. These tests may be waived when the Q' instrument, component, or system is not required to be operable, but the instrument, component, or system shall be tested prior to being declared operable. The operating cycla interval is 18 months and the 25% tolerance oiven in Definition "U" is applicable. TM eme x mv&s no 7xt as% 1xcxact ento w owbbso 2< ~v omms 4 * *N" W. Fire supprejijitn Wattr_1y11.es - A fire suppression water system

' shall censist of: a water scurce(s); gravity tank (s) or pump (5);

and distribution piping with associated sectionalizing control or isolation valves. Such valves shall include hydrant post indicator valves and the first valve ahead of the water flow alarm device on each sprinkler, hose standpipe or spray system riser.

(

l X. StLooered Test hsis - A' staggered test basis shall consist of: (a) a test schedule for n systems subsystems, trains, or other designated components obtained by dividing the specified test interval into n equal subintervals; (b) the testing of one system, l

subsystem, train or other designated components at the beginning of I each subiaterval.

Y. Source Check - A source check shall be tne qualitative assessment of channel response when the channel sensor is exposeu to a radioactive source.

l

~

(Y i

Revision 138 5a Amendment NC. 42, 89 128 t

LIMITINECONDITIONFOROPERAT80N SURVE_ILLANCE REQUIREMENT

'['

_i.

E. Reactivity Anomalies E. Reactivity Ance.alies

// The reactivity equivalent of l( ,

During the startup test program the difference between the and startups following refuel-i 4

k actual critical rod configur- ing outages, the critical rod '

ation and the expected con- configurations will be compared figuration during power operation shall not exceed 1% to the expected configurhtions 4 K. at selected operating conditions.

If this limit is exceed- These comparisons will be used ed, the reactor will be shut as base data for reactivity down until the cause has been monitoring during subsequent

' determined and corrective ac- power operation throughout the tions have been taken if such fuel cycle. At specific power actions are appropriate, operating conditions, the critical rod configuration will F. If Specifications 3.3.A be cunpared to the configuration expected based upon appropriately through 0 above cannet be met, corrected past data. This com-an orderly shutdown shall be parison will be made at least initieted and the reactor every full power month, shall be in the Cold.5hutdown conditierr within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Specifications 3,3. A through 0 above do not apply when '

there is no fuel in the reuctor vessel.

_ G. Scrain Discharce Volume G. Scram Discharge V.olume

\ 1. The scram discharge 1. .Th scram discharge volume volume drain & vent drain and vent valves valves shall be operable be verified-open-a-t-4e;she44 a+t-ence whenever more than one er stA- Tach-v44ve-shal '

operable control rod is se-syc4e4-oveteelyr---These.

withdrawn. va4ve: 2.:y be c4Med-htemf4-If any of the scram dis-

[tently-fee-test 4ng-vMw

2. 4daintstrativa-control charge volu a drain or vent valvr.s are rade or 2. During each refuelingutage- o,ew found iniperable an verify the scram discharge orderly shutdown shall be initiated and the reactor ( volume drain and vent valves; ~

shall be in Cold shutdown a) Close within 30 seconds within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after receipt of a reactor scram signal and

-b)' Open when the scram is reset.

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kl"llifil_CDNDITLQtJ1 EQR OPERAll.QN SURVQLLANCE REQ 9]ADtENTS _ _._ f g

3.4 5.TEj REY _([Qi,((Q_CQN TRQLSY.11[1]

4.4 $TANDBY L109I0 CGliRQL_$J11[1]

]

erlLlicatti.LLt3: h_pfli.C311IitV:  !

Applies to the operating status Applies to the surveillance 3 of the Standby Liquid Control requirements o# the Standby System. Liquid Control 1 stem. i QiLitCt ivg: Qhitc tive:

To assure the availability of a. To verify the operability of the system with the capability to Standby Liquid Control System, shutdown the reactor and maintain .

the shutdown condition without the use of control rods.

SDarific!1tign: Sucifieatlon:

A. Ngry l System Availability A. Normal System Availlbjlity f

l

1. During periods when fuel is The operability of tN Standby in the reactor and prior to Liquid Control System shall be
startup from a cold verified by the performance of condition, the Standby Liquid the following tests; i Control System shall be operable, except as specified 1. -A-t besteneun muth

_f~ in 3.4.B below. This system Gad-Mkip ! cap-shai+-tm .

need not be operable when the r - fer+e44na!!; tested.-by .

7' reactor is in the Cold FeeiretthtMg-l Shatdown Condition, all dehe"C i nd "MM--to-%e l operable control rods are tett-t e fully inserted and A s. e. ,> e/ de/*E l l

Specification 3.3.A is met. 2. At!=th. vr.ca 1ftrehg-tech.

openthg-+ysis.:..

a. Chgk that-%e-sys4em l

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t 4 s, t') './'/ g t m;; 3 c. e g g 39 Cou 37 m r_3, 4 sy t e~ % d-04-435 p.s.14 - -j-(. Revision 106 95 ,

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'LlHLU.t(G C01DlTlMS_EOS 0PlMTLON .

SURVEILLANLr V4UIRFl[GIS

'l 3.4 SIMDELLLQUID_.f01T_RQL SYSTEM  ;;

  • 4.4 5TA.NDELLLQUIRECONTROLS1SIG  !-

fehu,. - ve r, h /'Januallyinitiateone

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of the Standby Liquid [

/)/ /,cas / ocp dv dg Control System loops  !

eo d re 4 a f,h -

and pump demineralized wA,l e /=sfg

/s ",3r'% )/

y' water into the reactor T,

vessel.

J -

This test checks f FN ^ ' - d explosion of the charge associated with the tested loop, proper operation of the valves, and pump-C#'/ ,,d,f operatiili ty. The replacemert charges to ,

be-installed will be selected from the-same ,

manufactured batch as

, - , ,c - < w y- e v -g the tested charge.

(ab faofg /> S=Mh ef,4/,4,l %

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O u 9 h u '.- explosive valves, A--A.- U '.v. shall be tested in the-

,c%cs%oJAM. W u B. Quration_ylih_In0Rr? 'E B Ca SEUilNMd.nmepag

$ &&"Is '

gagg s n nents:

1. Frorn and after the dhte 1. When a component is found that a redundant .to be inoperable, fits component is made or redundant component shall found to be inoperable, be demonstrated to be Specification 3.4.A.1 operable immediately and.

shall be considered daily thereafter until the-fulfilled and continued . inoperable component is operation permitted repaired, provided that the-component is returned to an operable condition within seven days.

. 1 A.

. n.m

'Re'visica 106 96

1 i

MSES:

r~ ,

3.4 is 4.4 57EJLid@JJLCONTROLSYSTEM

\ A. The requirements for SLC capability to shutdown the reactor are identified via the station Nuc1Pir Safety Operational Analysis (Appendix G to the FSAR Special Event 45). If no clore than one t

operable control rod is withdrawn, the basic shutd:vn reactivity requirement for the core is satisfied and the Standby Liquid Control system is not required. Thus, the basic reactivity requirement for the core is the primary determir. The int of when design the of objective standby 11guld control system is required.

the standby liquid control system is to proviue the capability of bringing the reacter from full power to a cold, xenon-free shutdown l condition assuming that none of the withdrawn control rods can be inserted. To meet this objective, the Standby Liquid Control systee is designed to inject a quantity of boron that produces a  !

minimum concentration equivalent to 675 ppm of natural boron in the reattor core. 1hc 675 ppm equivalent concentration in the rec.ctor p

i core is required to bring the reactor from full power to at least a three percent Ak subtritical condition, considering the hot toThe system will cold reactivity difference, xenon poisocing etc. The maximum inject this boron solution in less than 125 minutes. time requireme override the rate vf reactivity insertion caused by cooldown of the teactor following the xenon poison peak.

The Standby liquid Coatrol system is also required to meet .

j 10CFR50.62 (Requirements for Reduction of Risk from Anticipated '

/ T ansients Hithout Scram (ATH5) Events for Light-Hater-Cooled l I The Standby Liquid Control system must have Huclear Power Plants).

the equivalent centrol capacity (injection rate) of 85 gpm at 13 '

percent by wt. natural sodium pentaboral:e for a-25." diameter reactor pressure vessel in order to satisfy 10CFR50.6?

requirements. This equivalenty4 requirement is fulfilled by.a enrichment and flow rate of combination of concentration, BA minimum 0.4TI concentration and scdium pentaborate solution.4 isotope at a 39 GPH pump flow rate 54.57, enrichment of 8 satisfies the ATits rtule (10CFR50.62) equivalency requirement.

I Because theM concentration / volume isotopic enrichmeat, curve an additional has been revised to refie requirement the increased B has been added to evaluate the solution's capability to meet the W enrichment original design shutdown criteria whenever the B requirement is not met.

OpfMense+WHpump-ope rab414ty-4 n#cahs-that-the-ranthly-test,,

N.$ gao v 1*-< omM e a thn-4ththe-t+s-t+-d ukg-eatThe bope r+thg-eyc4+<-

only 4+

practical time yww, wntrAent-to-mahtah-twep-perforetwee .

.c4 &&i M to fully test the liquid control system is during a refuelingVar gwuuu vca'

  • outage.

periodically, thus making more frequent testing of the entire w,co E Aryr/t system unnecessary.

s# no zuv& W y w a,., s e.ww- oe w ni n ,cisc a n.

y.emr adic wo i s Sewsw) .

Revision 148 100 Amendment No. M2, 135  !

%sm-e-,

-M16:

3.4 & 4.4 ilMLDEllWID CONTRAl_SYilff (Cont'd)

,She-muiimqm l

~._IT6fT33ter orr tttpe n et prTat1WW i n tendjd '

,W , pre wft th 'ss of sydium p 14torate ro)utfon via tk li fAtTig of lief yarve at low a prstsure. The,ufper Kmi t op e rol velveA tings pgcwfdes sysprdrotacdog/.

f rom.ef erpres;tr e. -

_y B. Only one of the two standby liquid control pumping loops is needed for operating the system. One inoperable pumping circuit does not immediately threaten the shutdown capability, and reactor operation can continue while the circuit is being repaired. Assurance that the remaining system will perform its intended function and that the long term average availability of the system is not reduced is obtained for a one out of two, system by an allowable equipment out of service time of one third of the normal surveillance frequency, This method '

decermines an equipment out of service time of ten days.

Additional conservatism is introduced by reducing the allowable out of service time to seven days, and by in:reased testing of the operable redundant component.

C. The quantity of BIO stored in the Standby Liquid Controi System Storage Tank is sufficient to bring the concentration of bun i the reactor to the point where the reactor will be shutdown and' to provide a minimum 25 percent margin beyrnd the amount needed c .

to shutdown the reactor to allow for possible imperfect mixing of the chemical solution in the reactor water.

~

'u Level indication and alarm indicate whether the solution vo'lume has chaaged, which might indicate a possible solution  ;

concentration change. Test intervals for level monitoring have been established in consideration of these factors. E -

Temperature and liquid level alarms for the system are s annunciated in the control room.

The solution shall be kept at least 10*f above the maximum saturation temperature to guard against boron precipitation.

Minimum solution temperature is 48'F. This is 10'F above'the -

saturation temperature for the maximum allowed sodlum J.

pentaborate concentration of 9.22 Ht. Percent.

Each parameter (concentration, pump flow rate, and enrichment) is tested at an interval consistent with the potential for that parameter to vary and also to assure proper equipment- '

performance. Enrichment testing is required t: hen material is received and when chemical addition occurs since change cannot '

, occur by any process other than the addition of new chemicals to the Standby 1.iquid Control solution ~ tank.

Revision 106 101 E

1,LSLTING CDEUJON FQERIRATION SMyII1LLN.QLRffRRLy[,N_L __

- r 4.5 CO3E ANQjyff,M LN3[g[ffg;3 G 3.5 CDELAD CONTA1MELCMLl!!G WlLSE 5.YSIf35 32c11eability 6221LfdhilltX Applies to the operational status of Applies to the Surveillance the core and suppression pool cooling Requirements of the core and systems, suppression pool cooling systems which }

( are required when the corresponding Limiting Condition for operation is in effect.

Obiective Qbiettive To assure the operability of the core To verify the operability of the core and suppression pool coolin9 sy;tems and suppression pool cooling systems ,

l under all conditions for which this under all conditions for which thir cooling capability is an essential culing capability is an essential response to station Abnormalities. response to station abnormalities.

50eci ficatian SAgdfieati23

[ A. Cat 2Jimr and t?CLSnten A. Core Snrev and t.P.CLSnt.en 'l '

1. Both core spray syst ws shall be  !

t 1. Core Spray Cystee Testing.

operable whenever irradiated fuel is in the vessel and prior to .Lr_ggggfn;y reactor startup from a Cold litra Condition, except as specified in Coce/ Operating 3.5.A.2 below, a. Simulated Automatic Cycle Actuation test.

Pump Operability- guehenth f - - -b.

smrv mrm As Mcu ko m Notor Operated Qu.ehmth As c.

3.13 vce>n Mr facs t.w gw / Valve Operability Mo"M M 4'3 N ,q p i n i w s s r m y 33x \

as,w sr A snmw kn2 [. N A)--f h{

y DW-Frie...g ~ '^""~^ 1');W~

'Spim ccmmucaite to x x4cnv bV yer A.;.Seu.t.

\ mstz .sas:uec c/ NY ssis 3 M 9- p p t e g.

\---~~--~---- 4-+y%ea-head te n H9and4*9-t+

ertettfre-ytm4 peessur: of 44.ps49, e gf. eCore Spray Header 6p Instrumentation n

l Revision 148 103 i Amendment No. 42, 62, II4, M S~

" ~ '

,7 LlHITING CONDITION FOR OPERATION SURVEILLANCr RFOUIRFSENT l3.5.A Core Sorav and LPCI Systemi 4.5.A Core Sotav ani.LP.CJ Syst,gm.1 l ,

(cont'd) (cont'd) '

Check oncelday.

Calibrate Once/3 months Test Step Once/3 months

2. From and after the date that one 2. This section intentionally left of the core spray black

[ or found to be ino, perable systemsfor is made any reason, continued reactor- 3. LPCI system Testing shall be as operation is permissible during .follows:

the succeeding seven days, provided that during such seven a. Simulated Once/ Operating days all active components of the Automatic Cycle other core spray system and active Actuation components of the LPCI. system and Test the diesel generators are operable, Tbc LPCI system shall be operable

b. Pump Operability 44c+%t%

[ 3. T wnenever irradiated fuel is in the .i reactor vessel, and prior to c. Motor Operated heJMenth 1 reactor startup from a Cold valve #3 NM Condition, except as specified in operability ^' 3 '3 ,

3.5.A.4 and 3.5J .5.

( d. Peep flow One* M-month

4. From and after the date that the ) e i.PCI sy5 tem is made or fcund to be fer5 '.PCI -pemp-M1--pesp-4800 /

inoperable for any re: son, 9pm+t? he:d across the p=p continued reactor operation is of-s t-++nt-3BO-ft, permissible only daring_the -%- A~- , ,

succeeding seven days unless it is zue.w acrc As tww/eo sooner made operable, provided .m J./3 verdy 7Anr 44w that during such seven days the g % ,gg;g, g containment cooling system jg g l_ (including 2 LPCI pumps) and active components of both core A m" W A W 3r m R.

^~

l spray systems, and the diesel --

generators required for operation of such components if no external source of power were available shril be operable.

l 5. If the requirements of 3.5.A cannot be met, an orderly shutdown of the recctor shall be initiated and the reactor shall be in the Cold Shutdown Ccndition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

I Revision 148 ,

Amendment No. 42. 62, III. II*. 135 104  !=

1 LIMlllEi_f0h0lllQ.N FOR10ffRATIQN SURVEILLANCE RE_Q!).lREMENT-l-3.5.B C2..tainagnt' n Cooling Sy11gm 4.5.8 faqtainment Coolino SyitJLm

-1. E:: cept as- specified in 3.5.8.2 1. Coretainment Cooling system ~ h and 3.5.F.3 below, both Testing shall be as follows: l i containment cooling system loops  ;

shall'be operable whenever 11gm ErgnpfDq1 i irradiated fuel is in the reactor e vessel and reactor coolant a, Ptnntr8rVaM Oncet$monttrs-  !*

temperature Is greater than OperabH+ty-212*F, and prior to reactor startup from a Cold Condition. b. Pttmp-Cepac-Ry M4ee-pump-

-Te1 Mach-fiBGCi . a4ntenante-

.? .

From and after the date that one ; pump--shd i and-every containment cooling systr.m leop deitver1700 vpm acwths-- ,

is made or found to be inoperable e-t-70-ftrfte.

for any reason, continued reactor Esth-SSHS ptnnp operation is permissible only , sha+HeHver-M0f}-

during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> spa-aMG-f-t, -TOH .

unless such system loop is sooner .

made operable, provided that the Air test on

c. Once/5. years

! other containment cooling system drywell and loop, including its associated torus headers diesel generator, is operable, and nozzles m

3. If the requirements of 3.5.8 cannot be met, an orderly

'W~') "> -

shutdawn shall be initiated and the reactor shall be in a Cold

/I " E* / v> A e ,, - /e s/ec/ . s

~#

Shutdown Condition within 24 / ofem U//df >

sf,c/A,/ ,A 3,g .

hours. '[ ' .N # ,

/ 6 Cc W jus p ,

ele /hr.1 I ?oo GAM j

  • f /* o Pf Yb hf so / _

eac4. ssiJ pe g \

oleAVers ? 7o0 crt>4 ,

af ss PF 7zw Y,

[

L

% 4 . Ilelve ks afc'et [eo, i

.r

@ 7 (,, 3./ 3 ,.

y

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l Revision 143 Amendment No. 42, 44, 174, 135 106

n LIMITING CONDIT[QS FOR,.QPERATION SURy[It.t.ANCE REOUIR*@ENT __,

C. fjPCI_Svstm C. HPCI Svs.t.m -]

1. The HPCI' system shall be operable 1. HPCI systen testing shall bel l s' whenever there is irradiated fuel p formed as follows:

in the reactor vessel, reactor pressure is greater than 150 a. Simulated Once/ operating -

psig, and reactor. coolant Automatic cycle temperature is greater than Actuation

. 365*F; except as specified in Test i 3.5.C.2 below.

b. Pump Oper- Once/ month--
2. From and after the date that the ability 'T l HPCI system is mace or found to be inoperable for any reason, c. Motor Operated OncthnenM continued reactor op'eration is Valve Oper- -As 9 cme /

permissible only during the abili ty 'b 3 /3:

succeeding seven days-unless such system is sooner made operable, dr-F4eHate-et--- Onct+3 w thr providing that during such seven 4CCO pitg-days all active components of the I

AUS system, the hCIC system, the c//. Flow Rata at Once/ operating-L.PCI system and both core spray 150 psig v .cyd.e.

systems are operable.

The HPCI pump shall deliver at least

3. If the requirements of 3.5.C 4250 gp'n for a system head

. cannot be met, an orderly ,

corresponding to a reactor' pressure of a shutdown shall be initiated and ' 1000 to 150 psig.

E= the reactor pressure shall be e 'Q- v- ,.

reduced to or below 150 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. /j w A. ,+ hah,/ . s p,Q -

'" 3 /3 vw riYQ f fo f. ;,< q l .

Mi E pump debt /ess e/ ,

\ lea a l- ya.so cnw ' f;,c w Sy vlum Ac.<c(cwe,e. spor,,,f,$. yl l.

l  ; a mwde ye.w.,, g Y'

-hoc > peg

,-,e ~-W Ls Ges/opamk) cy,je u.,.;q g,,t g ? '

Hf er ju.}da acc1 a/ Jea& .

%?.m GftH p,- a sys$ns .- Reac/

l , Coo'resfonAhf h a re n af, f 5 3"v'e h /.50fsi) ,

d. Revision'l4B

^ ~

Amendment No. #2, 44, 708, 774, 145 107 I

f i e'

b 1JMITING CONDITIOM FOR OPERATI0f S@.if]1LA. NICE _.EEQUIREMENT D l-3.5.0 Reactor 4.5.0 B.uitgr_ fore Isolation .Cqqling-

-18.CJ11Core ly.11 mIsolation Cooling LRflC) Sy.1.t.es l

1. The RCIC systen snall be operable 1. RCIC system testing shall be l whenever thWe is irradiated fuel perforced as follows:

in the reactor vessel, reactor pressure is greater than 150 a. Simulated Once/ operating psig, and reactor coolant Automatic cycle temperature is greater tnan Actuation 365'F; except as specified in Test .

3.5.L.2 below.

b. Pump OnceAnont4t-
7. . From and after the date that the RCICS is made or found to be Operability %N inoperable for any reason, c. Hotor -Once/neth-continued reactor power operation Operated 4: epeu'6'ed  !

is permissible only during the Valve I> 5. / 't succeeding seven days provided Operability that during such seven days the HPCIS is operabla. -4diew Role 2t- OnceAt-taeetM- -

KistFtsig-

3. If the requ-Irements of 3.5.0 cannot be met, an orderly ,/ /. Flow Rate at Onretoperdinsr shutdown shall be initiated and 150 psiL_ -v eysk-.,

the reactor pressure shall be

~, reduced to or below 150 pstg within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. /,Mhe RCIC pump least 400 shall gpm.for deliverhead a system at

/ corresponding to a reactor

/ pressure of 1000 to 150 psig.

] rm

/

/

/ _V XJ,s '

/./W4ao

. Ms 4,/ a.3 vgrceN/ <h

[ 3 /3 gedh 7%/ Ne /C.C C fuy c4lNe<~.s ' a .! feas{ ifao 6-frs1 ab A ..c p fe m Aea pl orer,cwr& rW a l'e a ebr~

\ / passw-< e f' / coo ps).

N--%._Ay% s A - v 7.

-y m

P{ y

,' &ce/opasn$p c.y e 4 &/k, t[ Yda f p .- .

TN A C- T C /uq c/el!V"r.S al j /ca..cf ifoo g,.fM af n .sy.s fen. -

f eq) corm spend fe, er

^

, newbr ym.sswe of / Sups >J ,

f Revision 148 Amendment No. 42, 109, Ild, 135

'Q p'f 108 r,

. . . =

, MRS:

3.5.B C.ontAinment Cr lid.g_ System },

The containment cooling system for Pilgrim 1 consists of two independert loops each of which to be an operable loop require, oos LPCI pump, two RBCCH pumps, and two SSH pumps to be operable. There are ?rstalled spares for margin above the design conditions. Each sys function; i.e., removing 64 x Btu 10 /hr gem(Ref.has Amendmentthe ca,sability to perform 18), even with some its s'-tem degradation. If one-loop is out-of-service, reactor operaiion is permitted for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Hith components or_ systems out-of-service, overall core and containment . ,

cooling reliability is maintained by the operability of the remaining cooling equipment.

Since some of the SSH and RBCCW pumps are required for normal g eration,

_capacitytestingofindividualpumpsbdirectflowmeasurementis ~

t fly tyt is a comparison of measurea pump _

[ypratticay'Qmp a orma pe pa Amsters capsh po p rform tests . mbined with / comp son to i e perforr ce of th p YTously ed pu These pumpe re to ed duritp3-

'ggratjp and prior ce tes will be egrated wi this or erforme V during refgaMng n pu e an be fl sted indiv ally. sts dur

\

\ p<fErcal Spirati will performe measuring t shutoff ..ad. Th the / '

pump (nder St wi e placed service and e of the y 'eviousl opera)Tng

_ system 1 be c red f t peps spc .Wr6 two pe'ses. red. Ie.! al flowp cation for

+ bis JVhot feasible e to chang syste onditi 3 e punfp dissh'a,rge prenWe will be mease ed and its er regnfremen 11 e ed.,destabli,WTlow at that ytfssure, j fu o,, , I, r"/i &

  • // 6e menxN* Y A"^ ^***"

r y,r-fam ejserabk a n clk e- J<n -e y s km c ond' bku s a t%W enfaerYy a n c/ ,oe(b.e am e e ;4s Ak;. , ;,

- c c o,d-  ;,y 8.i2 l

l i

i I

l D

1

! Revision 148 115 l

Amendment No. 135

'l l B_ASES: )

l

^- Core and Containment Coolina SystLms Surveillance Freautn,Ligi-4.5 l The testing interval for the core and containment- cooling systems is based on industry practice, quantitative reliability analysis, judgment and- -1 practicality. The core cooling systems have not been designed to be fully testable during operation. For example, in the case of the HPCI, automatic initiation during power operation woulo result in pumping cold water into the reactor vessel which is not desirable. Coplete ADS testing during power operation causes an undesirable loss-of-coolant inventory. To increase the aval' ability of the core and containment-cooling systems, the componente which make up the s 1.e. , instrumentation, pumps valves. etc., are tested fre u Tfie,ppmps Ano-mdTor, operated 13festion p a~ _ent_lyj.

h tTi,>m o

/.-

The pumps and motor operateu valves are tested in accordance with ASME B&PV Code,Section XI (IWP and IWV, except where specific relief is granted) to assure their operability. The frequency-and methods of testing-are described in the PNPS IST program. The PNPS IST Program is used to assess the operational readiness of pumps and valves which are safety related or ) ,

important to safety. When components are tested and fJund inoperable the -

impact on system operability is determined, and corrective action or limiting

)-

Conditions of Operation are initiated. A -simulated automatic -actuation test once each cycle combined with code inservice testing of the pumps and valves-is deemed to be adequate testing of these systems.

/

p l

r.

n lU l Revision 148 122 l Amendment No. l_35 i

l' I

- LIMllN.G CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS  ;

I 1

-3.6.0 Safety and Jelief Valves l 4.6.0 Safety and Relief Valves l

1. During reactor power operating 1. At-+ea*one-stfety vatre-snd-conditions and prior to reactor two-re44eUsafety-vahet the44 startup from a Cold Condition, te checked-or--r+pleeed-wl-th or whenever reactor coolant bensh-checked valves-once per j '

pressure is greater than 104 operating-cycle. 8.11-v44w +  !

psig and temperature greater w144-be-t ested-eve r-y-two-<yc l e s .

than 340*F, both safety valves and the safety modes of all 2. At least one of the relief valves shall be- relief / safety-valves shall be operable. The nominal setpoint disassembled and inspected each for the relief / safety valves refueling outage.

shallbeselectedbetween1095 and 1115 psig. All 3. Whenever the safety relief relief / safety valves shall be valves are required to be set at this nominal setpoint 2 operable, the discharge pipe ,

11 psi. The safety valves temperature of each safety shall be set at 1240 psig i 13 relief valve shall be logged psi. daily.

2. If Specification 3.6,0.1 is not 4. Instrumentation shall be met, an orderly shutdown shall calibrated and checked as be initiated and the reactor indicated in Table 4.2.F.

I coolant pressure shall be below 104 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. 5. Notwithilandimrthe-abover ;3 tr Note: Technical Specifications mie4mua r -safety-re11cf valves .

, 3.6,0.2 - 3.6,0.5 apply only that 5:ve h n ia tervira ch 34 1 when two Stage Target Rock SRVs 40 testedA the as-found. -

l -.

are installed, tend 4-t4en-dri nn sm+h e uio k and-Cye4e-h

3. If the temperature of any -

! safety relief discharge pipe exceeds 212'F during normal

(*, ,9

'" o'n " yY '" j %

reactor power operation for a re //e[% fe h v44'es 44*# -

period of greater than 24 y A 3,f y

)

hours, an engineering evaluation shall be performed j justifying continued operation -

for the corresponding l- temperature increases.

I L

Revision 155 Amendment No. 42, 56, 28, 133, 139 126

f SURVEfLLANCE REOUIREMENTS f U MITING CONDfl10NS FOR OPERATION 4.7.A Primary Containment (ConL 2

3.7.A Primary Contgjenent (Con!.tl

'/ Primary Containment Isolation Valvf.i Prinary Containment Isolation Valves In the event any automatic l2.b.1 The primary containment 2.b. isolation valves surveillaide Primary Containment Isolation shall be performed as follows:

I Valve becomes inoperable, at l least one containment isolation a. At least once per operating valve in each line having an cycle the operable primary i inoperable valve shall be containment isolation I deactivated in the isolated valves that are power condition. (This r'equirement operated and automatically may be satisfied by deactivating initiated shall be tested the inoperable valve in the for simulated automatic isolated condition. initiation and closure-Deactivation means to times. ,

electrically or pneumatically disarm, or otherwise secure the b.

g gaymy,&p m m

... . ca s t -onc+-per-qnrt+r4 ymres:

valve.)* a

1. .*' ner-3!!y ep:. ;wer va/iv Ahn #eper:ted prim ny l con 444AmentMat4 eft l ora 430 Acume o Mi*

awreiamsvr <yaaogV41V+5 4 :::Pt f0i t5:

M W' LWWGANY

' M SRW/m M

$Yh5'$?

.G Q ,..'a'2..".a'.:"...

, . . . . . , , , . . . . . 1..

^

2. Til; th: ti4 :t:1?

Vaby ech stoem *sel:tica n1vos-k.dMd=lly :nd verify-fu,g wag- +4ewr: ti = .

cw.ex6.hy n S M c/ D

  • S'J- c. At main least twice per week the steam line power

= ~%

operated isolation valves t

shall be exercised by partial closurs and' subsequent reopening.

I i .

W d. J,t 1:::t ,n:: p:r :; r444%

vgef,ry w rix m ar cyc1; the operability of p yff m a c co w

//U'4e ca:A' V4W f "r.=....,.NYs!.N[!'I

m. m :.ryg f ga o,2;rv. .e use J. L.. .m, 2.b,2 Hhenever a primary containment autoratic isolation valve, is l
  • Isolation valves closed to satisfy inoperable, the position of the these requirements may be reopened on isolated valve in each line-l an intemittent basis under ORC having an inoperable valve approved administrative controls, shall be recorded daily, i( 155a Revision 151 L Amendment No. 112. 136-

3.7 Primary Contdnment 4.7 Primary Containment

3. Pressure suppression Chamber - 3.

_ Pressure Suppression Chamber -

l[eactTr Buildino Vacuum Breakers Reactor Buildino Vacuum Breakers

a. E.xcept as specified in 3.7.A.3.b a.

below, two pressure suppression The- pres s u re-suppressionth ambe r--

chamber - reactor butiding vacu m reactor-buH dkg-vacuum-breekers-

.: breakers shall be operable at all and-associated-instnsnentation times when primary contairunent in- inc4eding-set-point shall be cher,ked-tegrity as required. The setpoint for-properoperetton-every-three monthsr.

of the differential prossure instru-rentation which actuates the pres- g,.,/fp ofefn/,,/d y J @

sure suppression chamber - reactor p ,# 7, g 4 9, building breakers shall be 0.5 psig.

N" d '~ A^ #'% N =

b. From and after the date~that one of "' "

the pressure suppression chamber -

  • 3'/3-reactor building vacuum breakers is stade or found to be inoperable for any reason, reactor operation is ,

lo . ( h 4 AG WSm< N pemissible only during the g j 4 fog ,/

succeeding seven days unless such vacuum breaker is sooner riade A'"4 f" 7" '/'"

*T # * "' " M operable, provided that the repair procedure does not violate primary contatrinent integrity. - -

4. Drywell-Pressure Suppression 4.

4 Drywell-Pressure Su)pression Chamber vacuum Breyers Chamber Vacuum Brea cers

a. When primary containment is a.

required, all drywell-pressure Periodic Operability Tests -

, suppression chamber vacuum breakers shall be operable -

except during testing and as - - -

stated in Specifications '

3.7.A.4.b, c and d below.

Drywell-pressure su,ppression chamber vacum breakers s'.all be considered operable if:

(1) The valve is demonstrated te (1) Once each wenth each dryvell-pressure open with the applied force of suppression charnber vacuum breaker the installed test actuator as shall be exercised and the operability indicated by the position of the valve and installed position switc5es and remote position indicating lights. irdicators and alams verified.

(2) The valve shall returr. by (2) A drywell to suppression chamber dif-gravity when released after being opened by remote or ferential pressure decay rate test -

manual means, to within 3/32" shall be conducted at least every 3 months.

of the fully closed position. ~

(3) Haither of the two position

__ alam systems which annunciate on Panel C-7 and Panel 905 when any vacum breaker opening l exceeds 3/32". are in alam. '

Amendment No. 68 155

. _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __________s

LIMli!NE CO30! TION FOR OPERATION __

SURVEILLA.<C RE2VidEMENTS -

3.7 Primary Containment 4.7 Primary Containment 4

b. Any drywell-suppression chamber b. Duringeachrefueling.$$dheY vacuum breaker may be non-fully closed as determined by the position (1) rach vacuum breaker shall be switches provided that the drywell tested to determine that the to suppression chamber differential disc opens freely to the touch decay rate is demonstrated to be not and returns to the closed greater than 25% of the differential position by gravity with no pressure decay rate for the maximum indication of binding.

allowable bypass area of 0.2ft',

c. Reactor operation may continue pro- (2) Vacuum breaker position vided that no more than 2 of the dry- switches and installed alarm well-pressure suppression chamber systems shall be calibrated and vacuum breakers are determined to be functionally tested.

inoperable provided that they are secured or known ce be in the (3) At least 25% of the vacuum ,

closed position, breakers shall be vicually g inspected such that all vacuum

d. If a fa114e of one of the two breakers shall have been installed position alarm systems occurs for one or more vacuum (n4/va g , inspected efueling_fgilowing cusov e. Ifevery fourth deficien- i breakers, reactor operation may cies are found, al1 vacuum continue provided that a breakers shall be visually

/ - differential pressure decay rate inspected and deficiencies

~

~s' test is initiated immediately and corfected.

performed every 15 days thereafter

_. until the failure is corrected. (4) A drywell to suppression chamber -

The test shall meet the leak rate test shall demonstrate requirements of that the differential pressure Specification 3.7.A.4.b. decay rate does not exceed the rate which would occur through a 1 inch ori,fice without the addition of air or nitroger..

5. Oxygen Contentration 5. Oxygen Concentrati-
a. The primary containment atmosphere The primary containment oxygen shall be reduced to less than concentration shall be measured 4% oxygen by volume with nitrogen and recorded at least twice gas during reactor power weekly. .

operation with reactor coolant .

pressure above 100 psig, except l>

as specified in'3.7.A.5.b. ty' 4 i

W Amendment No. 87 157.

.- - - . . . - -

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