ML20099E608

Amend 83 to License DPR-35,changing Tech Specs to Incorporate Limiting Conditions for Operation of & Calibr Requirements for Instrumentation

ML20099E608
Person / Time
Site:	Pilgrim (DPR-035)
Issue date:	11/07/1984
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Boston Edison Co
Shared Package
ML20099E610	List: ML20099E608
References
DPR-35-A-083 NUDOCS 8411210445
Download: ML20099E608 (13)

Text

"

"Ecuq'o UNITED STATES e

! "^,,. g,i NUCLEAR REGULATORY COMMISSION 7/

gc WASHING TON, D, C. 20555 t'$

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BOSTON EDIS0N COMPANY

-DOCKET NO. 50-293 PILGRIM NUCLEAR POWER STATION

^

AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 83 License No. DPR-35 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Boston Edison Company (the licensee) dated August 9, 1984, as amended by letters dated September 21, and October 19, 1984, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Ciiapter I; B.

The facility will operate in conformity with the application, the provisions of the Act,.and the rules and regulations of the Comission; C.

Thereisreasonableassurance(1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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 amendment 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 Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-35 is hereby amended to read as follows:,-

8411210445 841107 PDR ADOCK 05000293 P

PDR

. B.

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 83, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

~

Domenic B. Vassallo, Chief

_.' Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: November 7,1984 9

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- i ATTACHMENT TO LICENSE AMENDMENT NO. 83-

- FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293' Remove.

Insert 58 58

'58a

'58a 58b 59 59 66a 66a 66b' 66b 66c 152 152 152a 152a 166 166 O

S e

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1 TABLE 3.2.f

-SURVEILLANCE INS 1RUMENTATION Hinimum # of Operable instrument Type Indication Channels' Instrument #

Pasameter and Range Notes'

~l 2

640-29A & B Reactor llater Level' Indicator 0-60" (1) (2)-(3)

~

/2' 640,25A & B Reactor Pressure Indicator 0-1200, psig-(1) (2) (3).

~

2 TRU-9044 Drywell Pressure Recorder 0-80 psia (1) (2) (3)-

' ~

\\

9j-TRG-9045-2 ORU-9044 Drywell Temperature Recor' der,. indica tor (1) '(2) (3)

I I'-9019 0-400*F 2

IRU-9045 Suppression Chamber Air Recorder, Indicator (1) (2) (3) 11-S0lf Temperature 0-400*F 2

LR-5538 Suppress ~lca Chamber Water Level Recorder 0-32" (1) (2) (3)

LR-50,49 4

e 1

NA Control Rod Position 28 Volt Indicating )

Lights

)

r

/

)

. (1) (2) (3)-(4)

'('

/-

i NA' Neutron Honitoring SRH, IRM, LPRH

)

/;.

7-O to 1007. power

)

~

/' TI-5021-Ol A

/

Suppression Chamber Water

' Dual Indicator /

',i

'1RU-5021-01A Temperature 7 Hul tipoint -Recorder ~

(4) (1 ) (2) (3)

./

q 30-230*F (Bulk / Local) o 2

1 4

'_11-5022-018 Suppression Chamber Hater Dual Indicator'/

yTRO-5022-OlB Temperature.

Hultipoint Recorder (4) (1) (2) (3)

' r

)

'30-230*f-(Bulk / Local)

I

l

.PI-5021" Drywell/Tocus Diff. Pressure Indicator

.25-> 3.0 psig (1) (2) (3) (4)

j I

(PI-5067A Dryuell Pressuie Indica tor

.25-* 3.0 psigi, LP,1-5067B Torus Pressure lud i ca tor -1.0-* +2.0

- (1) (2) (3)-(4) psig

?.

,.. / '

v Amendm.'nt flo. 82 58 5-

TABLE 3.2.f (Cont'd)

SURVEllt AtlCE INSTRUMElliATIOff Minimua # of Operable Instrument Type Indication Channels Instrument #

Parameter and Range Notes l

l/ Valve a) Primary Safety / Relief Valve Position a) Acoustic monitor (5) or (5) b) Thernocouple b) Backup 1/ Valve a) Primary Safety Valve Position Indicator a) Acoustic monitor (5) or (5) b) Thermocouple b) Backup 1/ Valve See Note (6)

Tall Pipe Temperature iliermocouple (6)

Indication

" LI 1001-604A Torus llater Level Indicator /Multipoint (4) (1 )- ( 2 ) (3)

LR 1001-604A (Hide Range)

Recorder 0-300"ll:0 2

+

LI 1001-604B Torus llater Level Indicator /Multipoint (4) (1) (2) (3) sLR 1001-6048 (Hide Range)

Recorder 0-300"!!,0 P PI 1001-600A Containment Pressure, Indicator /Multipoint (4) (1) (2) (3)

• }PR1001-600A (111 91: Range)

Recorder 0-225 psig 2

I PI 1001-6008 Containment Pressure, Indicator /Multl' point

'(4) (1) (2) (3)

,PR 1001-600B (High Range)

Recorder 0-225 psig I"PI 1001-601A Containment Pressure, Indica tor /liul tipoint (4) ( 1) (2) (3) 1 PR 1001-600A (Low Range)

Recorder -5'to 5 psig 2

PI 1001-6010 Containment Pressure, Indicator /Multipoint (4) (1) (2) (3)'

,PR 1001-6000 (Low Range)

Recoider -5 to c. psig "RIT 1001-606A Containment liigh Radiation lionitor/Multipoint 1

RIT 1001-606B (Drywell)-

Recorder (4)'(7)

• RR 1001-606A 1 to lx10' R/hr

, RR 1001-6068 Amethi.ren t lio. 83 58a i

TABLE 3.2.F-(Cont'd)

SURVEILLANCE INSTRUMENTATION Operable Instrument Type l Indication-Channels Instrument #

Parameter and Range Notes-l 1

RI 1001-607 Reactor Building Vent Indicator /Multipoint (4) (7)-

RR 1001-608 Recorder 10-' to.10* R/hr.

1 RI 1001-608 Main Stack Vent Indicator /Multipoint.

(4) (7)

RR 1001-608 Recorder 10-' to 10* R/hr I

RI 1001-610 Turbine Building ~. Vent Indicato.IMultipoint (4) (7)

RR 1001-608 Recorder 10 to 10*N/hr 3

Anientiment No. 83 Shb

Notes for' Table 3.2.F (1) With lessL han the minimum number of instrument channels, restore-the t

inoperable channel (s) within 30 days.

(2) With_the instrument channel (s) providing no indication to the Control recm, restore the indication to the control recm within seven days.

~

(3) If the requirements of notes (l) or (2) cannot be met, an orderly shutdown shall be initiated and the reactor shall be in tne Cold Shutcown Condition with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(4) These surveillance instruments are considered to be redundant to each other.

(5) At a minimum, the primary or back-up" parameter indicators shall be j

operable for each~ valve when the valves are reautred to be coerabl'e.

With both primary and backup

• Instrument cnannels inoceraole either ret'rn one u

(1) channel to operable status within 31 days or be in a snutdovn mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The following instruments are associated with the safety / relief and safety valves:

Primary Secondary I

Valve Acoustic Monitor Tail Pipe Temperature Therm 0 couple l

203-3A ZT-203-3A TE6271 i

203-3B ZT-203-38 TE6272 203-3C ZT-203-3C TE6273.

• 203-30 ZT-203-3D TE6276 l

203-4A ZT-203-4A TE6274-B 203-4B ZT-203-4B TE6275-8

• See Note (6)

(6) At a minimum, for thermocouples providing SRV tall _ pipe temperature, one i

of the dual thermocouples will be operable for each SRV when the valves are recuired to be operable.

If a thermocouple becomes incoeracle, it shall be returned to an operable condition within 31 days or the reactor snall be placed in a shutcown mode within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(7) With less than the minimum number of operable instrument channels, restore the inocerable cnannels to operable status within 7 days or prepare and submit a special recort to the Regional Director of Inspection and Enforcement within 14 days of the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the channels to operable status.

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imentre,- No. 83 59

'D T

PHPS lABLE 4.2.F (Cont.)

MINIMUM 1EST AND CAllBRATION FREQUENCY FOR SURVEILLANCE INSTRUMENTATION Instrument Channel Calibration Frequency Instrument Check l

I

13) lorus Water Level (Wide Range)

Each refueling outage once every 30 days l

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14) Containment Pressure Each refueling outage Once every.30 days l

l

15) Containment High Radiation Each refueling outage Once every 30 days l

l

16) Reactor Building Vent Radiation Monitor Each refueling otitage Once every 30 days l

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11) Main Stack Vent Radiation Monitor Each refueling outage Once every 30 days

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18) lurbine Building Vent Radiation Monitor Each refueling outage Once every 30 days l

4 9

1 Amendment No. 83 66a.

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9 PNPS Table 4.2-G Minimum Test and Calibrat' ion Frequency for ATW5 RPT/ARI Instrumentation Instrument I"St""*'"t(2)

Instrument Functional (2) Calibration (2)

Check Channel Test 1.

Reactor High Pressure (1)

Once/ Operating Once/ day Cycle-Transmitter Once/3 months -

Once/ day Trip unit 2.

Reactor Low-Low Water Level (1)

Once/ Operating Once/ day Cycle-Transmitter Once/3 months -

Once/ day Trip unit O

e b

gl Arendment to. 42$ 83 65b

!j pgp

(*

TABLE 4.2.ll Minimuni Test & Calibration Frequency fdr Drywell 3

Temperature Surveillance Instrinnentation 9

n.

8

3 Instrument Channels /

Calibration InstrumenL f

flominal Elevation Frequency Check O

80 Feet

'E'a'ch Refuel 1.ng ;

Once per Shift Outage 87 Feet Ea.ch Refueling Once per Shift Outage e

60 Feet Each' Refueling Once per Slij ft j

0utage n

i 4

41 Feet Each Refuelint)

Once per Shift

~

Outage 32 Feet Each Refueling Once per Shift Outage l,-

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?_IMITINGCONDITIONSFORdPERATION SURVEILLANCE REQUIREMENTS

?.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS

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-Acolicability:

Applicability:

( Applies to the operating status of the' primary Applies to the' primary and secondary and:stcondary containment systems.

containment integrity.

Objective:

Objective:

To assure th'e integrity of the primary and To verify the integrity of the primary secondary containment systems.

and secondary containment.

Scecification:

Suecification:

A.

Primary Containment.

A.

Primary Containment 1.

At any time that the nuclear system'is 1.

a.

The suppression chamber water pressurized above atmospheric pressure level and temperature shall or work is being done which has the be checked once per day.

' potential to drain the vessel, the l pressure suppression pool. water volume b.

Whenever there is indication and temperature shall be maintained of relief valve operation or

. within the rollowing limits except as testing which adds heat to th~'

speciried in 3.7.A.2 and 3.7.A.3.

suppression pool, the pool 3

P a.

Minimum water volume - 84,000 ft u1 non o d an also 4

b.

Maximum water volume - 94,000 ft observed and logged every 5 minutes until the heat addition c.

Maxin.um suppression pool bulk tempera-is terminated.

ture during normal contfnuous power operation shall be i 80 F, except as c.

Whenever there is indication of specified in 3.7. A.1.e.

~

relief valve operation with the bulk temperaturegfthesuppressionpool d.

Maxumum suppression pool bulk tempera-reaching 160 F or more and the primary ture during RCIC, gPCI or ADS opera-coolant system pressure greater than tion shall be 190 F, except as 200 psig, an external visual examina-specified in 3.7 A.1.e.

tion of the suppression chamber shall be conducted before resuming power e.

In order to continue reactor power operation.

operation, the suppression chamber d.

Whenever there is indication of poolbufktemperaturemustbereduced to < 80 F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, relief valve operation with the local

~

temperatureofthesuppgessionpool f.

If the suppression pool bulk tempera-T-quencher reaching 200 F or more, ture exceeds the limits of Specifica-an external visual exami. nation of the tion 3.7.A.1.d, RCIC, HPCI or ADS suppression chamber shall be conducted testing shall be terminated and before resuming power operation.

suppression pool cooling shall be initiated.

e.

A visual inspection of the suppresion chamber interior, including water g.

If the suppression pool bulk tempera-line regions, shall be made at each ture during reactor power operation major refueling cutage.

0 exceeds 110 F, the reactor shall be scrammed.

Amendment No. 83 152

LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.7 CONTAINMENT' SYSTEMS (Cont'd) 4.7 CONTAINMENT SYSTEMS (Cont'd)

h. During reactor isolation

f. The pressure differential

-l conditions, the reactor pressure between the drywell and vessel shall be depressurized suppression chamber shall be to less than 200 psig at normal recorded at least once each cooldownratesifthegoolbulk-

~ shift when the differential temperature reaches 120 F.

pressure is required.

i. Differential pressure between the

g. Suppression chamber water drywell. and suppression chamber ~

level shall be recorded at shall be maintained at equal to or least once each shift when greater than 1.17 psid, except as the differential pressure specified in j and k.

is required.

j. The differential pressure shall be established within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of placing the reactor in the run mode following a shutdown. The differential pressure may be reduced to less than 1.17 psid 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.

k. The differential pressure may be reduced to less than 1.17 psid for a maximum of four (4) hours for maintenance activities on the differential pressure control system and during required oper-ability testing of the HPCI system, the relief valves, the RCIC system and the drywell-suppression chamber vacuum breakers.

1. If the specifications of Item i, above, cannot be met, and the differential pressure cannot be restored within the subsequent six (6) hour period, an orderly shutdown shall be initiated and the~ reactor shall be in a cold shutdown condition in twenty-four (24) hours.

m. Suppression chamber water level shall be maintained between -6 to -3 inches on torus level instrument which corresponds to a downcomer submergence of 3.00 and 3.25 feet respectively.

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l.er:: ment No. 83 152a

'BASESr

' 3.7. A' & '4.7. A Primary Containment i

The integrity of the primary containment and operation 'of the core standby cooling system in combination limit the off-site doses to values less than those suggested

.in 10 CFR 100 in the event of a break in the primary system piping.. Thus, contain-mint integrity is specified whenever the potential for violation of the primary reactor system integrity exists ~. Concern about such.a violation exists whenever the reactor is critical and -above atmospheric pressure. An exception is made to this rcquirement during initial core loading and while the low power test program is -

'being conducted and ready access to the reactor vessel is required. There will be no pressure on the system at. this time, thus greatly reducing the chances of a pipe br:ak. The reactor may be taken critical during this period; however, restrictive op: rating. procedures will be-in effect again to minimize the probability of an accident occurring. Procedures and the Rod Worth Minimizer would limit control worth such that a rod d op would not result in any. fuel damage.

In building and standby gas treatment system,' which shall be operational during this time, offer a sufficient barrier to keep off-site doses well below 10 CFR 100 limits.

The pressure. suppression pool water provides.the heat sink' for the reactor primary system energy release following.a postulated rupture of the system. The pressure suppression chamber water volume must absorb tne associated decay and structural sensible heat released during primary system. blowdown from 1035 psig. Since all of thy gases in the drywell are purged into the pressure supression chamber air space during a loss-of-coolant accident, the pressure resulting from isothermal compression plus the vapor pressure of the liquid-must not exceed 62 psig, the suppression chamber maximum pressure. The design volume of the suppression chamber (water and air) was obtained by considering that.the, total volume of reactor coolant to be condensed is discharged to the suppression chamber and that the drywell volume is purged to the i

suppression chamber.

Using the minimum or maximum water volumes given in the specification, containment _

4 pr:ssure during the design basis accident is approxima}ely 45 psig which is below the maximum of 62 psig. Maximum water volume of 94,000 f3 results-in a downcomer sub-mergency of 4'-0" and the minimum volume" of 84,000 ft results in a submergence approximately 12-inches less. Mark I Containment Long Term Program Quarter Scale 3

j Test Facility (QATF) testing at a downcomer submergency of.3.25 feet and 1.17 psi wetwell to dry well pressure differential shows a significant suppression chamber load reduction and Long Term Program analysis and modifications are based on the l

above submergence andAP.

i Should it be necessary to drain the suppression chamber, provision will be made to maintain those requirements as described in Section 3.5.F BASES of this Technical Specification.

Experimental data indicates that excessive steam condensing loads can be avoided if i

the peak local temperature of the pressure suppression pool is maintained below 200 F

~

during any period of relief-valve operation with sonic conditions at the discharge Analysig has been performed to verify that the local pool tgmperature will exit.

stay below 200 F and the bulk poo,1 temperature will stay below 160 F for all SRV tran -

sients.

Specifications have been placed on the envelope of reactor operating con-ditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high pressure suppression chamber loadings.

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166.

Amene. ment No.'83 L

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