Text

Forwards Responses to Action Items & marked-up Pages from Proof & Review Tech Specs Resulting from Discussion During & Subsequent to 861020-24 Tech Spec Meetings
	ML20213G962
Person / Time
Site:	Vogtle
Issue date:	11/14/1986
From:	Bailey J; GEORGIA POWER CO.
To:	Youngblood B; Office of Nuclear Reactor Regulation
References
	GN-1184, NUDOCS 8611180430
	Download: ML20213G962 (42)
	v • d • e

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

Georgis Power Company Pbst Offica Box 282 Waynesboro, Georgia 30830 Telephone 404 554-9961 404 724-8114 Southern Company Services, Inc.

Post Office Box 2625 Birmingham, Alabama 35202 Telephone 205 870-6011 VOgtle Project November 14, 1986 Director of Nuclear Reactor Regulation File: X7N16 Attention: Mr. B. J. Youngblood Log: GN-ll84 PWR Project Directorate f4 Division of PWR Licensing A U. S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBER 50-424 CONSTRUCTION PERMIT NUMBER CPPR-108 V0GTLE ELECTRIC GENERATING PIANT - UNIT 1 TECHNICAL SPECIFICATIONS

Dear Mr. Denton:

Enclosed for your staff's review are responses to action items and marked-up pages from the Proof and Review Copy of the VEGP Unit 1 Technical Specifications resulting from discussion which took place during and subsequent to the October 20-24, 1986 Technical Specification meeting in Bethesda.

If your staff requires additional information, please do not hesitate to call.

Sincerely,

. A. Bailey

}

Project Licensing Manager JAB /caa Enclosure xc: R. E. Conway NRC Regional Administrator R. A. Thomas NRC Resident Inspector J. E. Joiner, Esquire D. Feig B. W. Churchill, Esquire R. W. McManus (w/o enclosure)

M. A. Miller (2) L. T. Gucwa B. Jones, Esquire (w/o enclosure) Vogtle Project File G. Bockhold, Jr.

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!. 8611180430 861114 I

! PDR ADOCK 05000424 l A PDR t

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Action Item During the October 20-24 meeting we were asked to address the manner in which the seal injection throttle valves are protected from an inadvertent position change. These valves are enclosed within protective covers which are locked under administrative control. Thus positive protection is afforded to prevent inadvertent movement of the valves.

Action Item At the October 20-24 meeting we were asked to address a letter from D. O. Foster to J. Nelson Grace dated June 4,1986 which stated that GPC will incorporate certain guidelines into their diesel fuel chemistry control procedure. Since the inorganic zine coating of the diesel fuel oil tanks has developed into an issue in the realm of the FSAR, we will have to defer any discussion regarding the Technical Specifications until this issue has been resolved in the realm of the FSAR.

Action Item During the meeting in Bethesda with the NRC the question was raised with respect to the acceptability of the allowabig float voltage of > 2.10 volts for each connected cell specified in Table yd8-2. The specific question was whether or not this float voltage could maintain the battery in a charged condition. The technical specification requirements are that the parameters specified in Table 4.8-2 be verified once per 92 days and within 7 days after a battery discharge. If the float voltage is outside the limit (> 2.13 volts) _

the battery may be considered OPERABLE provided that the float voltage is within the allowable value (> 2.10 volts) and that the float voltage be restored to within the limit within 7 days. We have contacted our battery vendor and they have indicated that connected cells with a float voltage of

> 2.10 volts and less than the limit of 2.13 volts could maintain the battery at a floating charge condition for the 7 days.

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Action Item During the October 20-24, 1986 Technical Specification meeting in Bethesda we were asked to provide additional justification for our proposed revision to item 1 on page 6-8 of Section 6.0 of the Technical Specifications. In response, we offer the following:

Specifying minimum unplanned radioactive releases for PRB review provides a clearly defined threshold to initiate the review and ensures that PRB, SRB, and corporate management resources are not devoted to reviewing trivial releases. The specific minimum quantities proposed for inclusion in Specification 6.4.1.6.1 were taken from NUREG-0472, Rev. 2, " Radiological Effluent Technical Specifications for PWRs." These quantities defined the minimum unplanned radioactive releases for which reports to the NRC were required. GPC has judged that these quantities are appropriate for initiating elevated internal reviews and has incorporated them into the Hatch Units 1 and 2 Technical Specifications. Wording similar to that in the Hatch Specifications is proposed for VEGP.

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SURVEILLANCE REQUIREMENTS M b) Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 10 seconds," energizes the auto-connected emergency (accident) loads through the load sequencer and operates for greater than or equal to 5 minutes while its generator is loaded with the emergency loads.. After energization, the steady-state voltage and frequency of the emergency busses shall be maintained at 4160 +240,

-410 volts and 60 + 1.2 Hz during this test; and I c) Verifying that all automatic diesel generator trips, except engine overspeed low lube oil pressure, high jacket water

' temperatures and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a Safety Injection Actuatio,n signal.

7) Verifying the diesel generator erstes for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

i During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of thi test, the diesel generator shall

! be loaded tn an indicated 7600 4e-7700 kW,** and during the j - . remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be

, loaded to an indicated 6800-7000 kW.** The generator voltage

! and frequency shall be 4160 40, - 410 volts and 60 i 1.2 Hz

g within X seconds after the sta t signal; the steady-state

. Ne r - generator voltage and frequency shall be maintained within these limits during this test.# Within 5 minutes after completing j this 24-hour test, perform Specification 4.8.1.1.2g.6)b);##

l 8) Verifying that the auto-connected loads to each diesel

generator do not exceed the 2000-hour rating of 46R kW; 1ED

9) Verifying the diesel generator's capability to:

a) Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a j simulated restoration of offsite power,

i b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.

All engines starts for the purpose of surveillance testing as required by 4.8.1.1.2 may be preceded by an engine prelube period as recommended by the manufacturerer to minimize mechanical stress and wear on the diesel engine.

- This band is meant as guidance to avoid routine overloading of the engine.

Loads in excess of this band or momentary variations due to changing bus loads shall not invalidate the test.

Failure to maintain voltage and frequency requirements due to grid

! disturbances does not render a 24-hour test as a failure.

i ##If Specification 4.8.1.1.2g.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator

may be operated at '- "8-4 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temper-ature has stabilized. "[--~

L./h fod y uge)by Sunenlawe Ryuimert

'/.8.l.I.2d.S V0GTLE - UNIT 1 3/4 8-7 .

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Justification Item 8 on page 3/4 8-7 should be revised to specify a load of 7000 KW since this is the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating for the VEGP diesel generators.

9 REACTOR COOLANT SYSTEM PROOF & REWEW C 3/4.4.11 REACTOR COOLANT SYSTEM VENTS LIMITING CONDITION FOR OPERATION 3.4.11 Two reactor vessel head vent paths each consisting of two vent valves a#d

cof/rg/ahepowered from emergency busses, shall be OPERABLE and closed.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a. With one of the above reactor vessel head vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the vent valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With two reactor vessel head vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the vent valves in the inoperable vent paths, and restore at least one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.1 Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open position, i

g

b. Cyclingeachventvalve[throughatleastonecompletecycleof full travel from the control room, and

! c. Verifying flow through the Reactor Coolant System vent paths during venting.

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l V0GTLE - UNIT 1 3/4 4-37

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d Justification At the October 20-24, 1986 meeting in Bethesda we were asked to address the control valves in the head vent flow paths. In response, we offer the revised specification which defines an operable flow path in terms of two vent valves and a control valve.

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TABLE 3.8-f

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SAFETY-RELATED MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES FUNCTION VALVE NUMBER Nuclear Service C1g. Twr. A Return 1HV-1668A Nuclear Service C1g. Twr. A Fans Bypass 1HV-1668B Nuclear Service C1g. Twr. B. Return 1HV-1669A Nuclear Service C1g. Twr. B Fans Bypass 1HV-16698 1HV-1806 CTMT Air Cool A4001 CW Inlet 1HV-1807 CTMT Air Cool A4004 CW Inlet 1HV-1808 CTMT Air Cool A4003 CW Inlet 1HV-1809 CTMT Air Cool A4002 CW Inlet 1HV-1822 CTMT Air Clr A4001 Cire Wtr Outlet 1HV-1823 CTMT Air Cool A4004 CW Outlet 1HV-1830 CTMT Air Clr A4003 Cire Wtr Outlet 1HV-1831 CTMT Air Clr A7006 CW Outlet 1HV-1974 Aux Comp CW Trn B Return Iso 1HV-1975 Aux Comp CW Trn A Return Iso 1HV-1978 Aux Comp CW Trn B Supply Iso 1HV-1979 Aux Comp CW Trn A Supply Iso

Reactor Coolant Pumps Thermal Barrier 1HV-2041 ACCWS Outlet Header

(. 1HV-2134 Reactor Cavity C1g coil A4001 Inlet Iso Reactor Cavity C1g Coil A4002 Inlet Iso 1HV-2135 1HV-2138 Reactor Cabity C1g coil A4001 Outlet Iso 1HV-2139 Reactor Cavity C1g Coil A4002 Outlet Iso 1HV-12114 CR Outside Air Intake Iso 2HV-12115 CR Outside Air Intake Iso 1HV-12118 CB CR Filter Units N7001 Inlets 1HV-12119 CB CR Filter Units N7002 Inlets 1HV-12128 CB CR Filter Units N7001 Outlets 1HV-12129 CB CR Filter Units N7002 Outlets 1HV-12130 CB CR Return Air Fans B7001 Inlets IHV-12131 CB CR Return Air Fans B7002 Inlets 1HV-12727 CB SR Battery Rm Exh B7002 Damper 1HV-12742 CB SF Battery Rm Exh B7001 Damper 1HV-12748 CB !F Battery Rm Exh B7003 Damper 1HV-12749 CB SF Battery Rm Exh 87004 Damper 1HV-19051 Thermal Barrier Cooling Wtr RCF 001 1HV-19053 Thermal Barrier Cooling Wtr RCP 002 1HV-19055 Thermal Barrier Cooling Wtr RCP 003 '

1HV-19077 Thermal Barrier Cooling Wtr RCP 004 1HV-8920 Safety-Injection Pump Mintflow Isolation 1HV-8803A Baron Injection Tank Inlet Isolation 1HV-8803B Boron Injection Tank Inlet Isolation 1HV-2624A CTB Post LOCA Purge Exhaust Iso 1HV-26248 CTB Post LOCA Purge Exhaust Iso CTMT Bldg Norm Purge Supply Iso

( 1HV-2626A

1 TABLE 3.8-2 SAFETY-RELATED MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION BYPASS DEVICES FUNCTION VALVE NUMBER CTB Norm Purge Supply Iso 1HV-2627A CTMT Bldg Norm Purge Exhaust Iso 1HV-2628A CTB Norm Purge Exhaust Iso 1HV-2629A Aux FDW Pump Turbine 1HV-5106 Conds Stor TK V4002 to Pump P4001 IHV-5113 Conds Stor TK V4002 to Pump P4002 1HV-5118 Conds Stor TK V4002 to Pump P4003 1HV-5119 Aux FDW Pump P4001 Discharge Trn C 1HV-5120 Aus FDW Pump P4001 Discharge Trn C IHV-5122 Aux FDW Pump P4001 Discharge Trn C IHV-5125 Aux FDW Pump P4001 Discharge TRN C 1HV-4127 Aux FDW Pump P4002 Discharge Trn B IHV-5132 Aux FDW Pump P4002 Discharge Trn B 1HV-5134 Aux FDW Pump P4003 Discharge Trn A INV-5137 Aux FDW Pump P4003 Discharge Trn A IHV-5139 Aux FDW Pump P4002 Miniflow IFV-5154 Aux FDW Pump P4003 Miniflow 1FV-5155 Charging Pump B Discharge IHV-8438 Charging Pump A Suction 1HV-8471A 1HV-84718 Charging Pu p B Suction Charging Pump A Discharge 1HV-8485A Charging Pump B Discharge IHV-8485B Charging Pump Miniflow Iso to RWST IHV-8508A, B 1HV-8509A, B Charging Pump Miniflow Iso to RWST 1HV-9380A CTMT ATM Unit 1 SVCE Air I

IHV-93808 CTMT ATM Unit 1 SVCE Air

! Trn B Aux FDW Pump Rm Inlet Damper )

INV-12005 Trn A Aux FDW Pump Rm Inlet Damper 1HV-12006 l' DGB Exb Fan B7001 Disch Damper (Trn A) 1HV-12050 DGB Exh Fan 87003 Disch Damper (Trn A) 1HV-12051 f DGB Exh Fan B7005 Disch Damper (Trn A)

' 1HV-12052 DGB Exh Fan B7002 Disch Damper (Trn B)

IHV-12053 DGB Exh Fan B7004 Disch Damper (Trn B) 1HV-12054 DGB Exh Fan B7006 Disch Damper (Trn B)

- 1HV-12055

' HV-8000A, B PORV Blockline Reg. Hx Tube Dutlet to RCS Alternate Chg HV-8147 Reg. Hx Tube Outlet to RCS Normal Chg HV-8146 HV-8100, 8112 No 1 Seal Leakoff g HV-8103, B, C, D RCP No 1 Seal from Chg f

HV-8111A, B, 8110 Chg Pump Miniflow LV-112C, B VCT Discharge Header SIS RWST Discharge to Chg/SI Pump Suction LV-112E, D CVCS Boric Acid Filler to Charging Pump Suction HV-8104 HV-8485A, B Chg Pump Discharge Header I - - - -

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1 I TABLE 3.8-f SAFETY-RELATED j

HOTOR OPERATYb VALVES THERMAL OVERLOAD

' PROTECTION BYPASS DEVICES ,

I I FUNCTION VALVE NUMBER 1

HV-8105; 8106 PD Pump Discharge HV-8807A, B; 8924 HHSI Suction to Chg/SI Suction HV-8801A, 8 BIT Discharge 2

HV-8808A, B, C, D Accumulator Discharge i

HV-8811A, B Containment Emergency Sump Isolation l RHR Suction from RWST j

HV-8812A, B HV-8809A, B RHR Discharge Header

RHR Hx ko. 1 Outlet to Charge Pump

! HV-8804A RHR Hx No. 2 Outlet to SI Pumps i

HV-88048 RWST Discharge Header to SI Pumps HV-8806 SI Pump Suction Isolation HV-8923A, 8 HV-8813; 8814 SI Pump Miniflow i

HV-8821A, B SI Pump Crosschannel i

SI Pump Discharge to Cold Legs '

l HV-8835 RHR Pump Discharge to Hot Legs l

HV-8840 SI Pump Discharge Header i

HV-8802A, B RHR Suction from RCS Hot Legs 1, 4 HV-8701A, B; 8702A, B

!- RHR Miniflow i( FV-610; 611 HV-8716A,.B RHR Cross Connect j Spray Pump Containment Emergency Sump Isolation

' HV-9002A, 8 HV-9003A, B Spray Pump Containment Emergency Sump Isolation HV-9017A, 8 Spray Pump Suction from RWST HV-9001A, B Spray Pump Discharge Header HV-8994A, B Spray Additive Tank Discharge

, NSCW Pump Discharge Isolation i HV-11600 NSCW Pump Discharge Isolation HV-11605 NSCW Pump Discharge Isolation HV-11606 NSCW Pump Discharge Isolation HV-11607 NSCW Pump Discharge Isolation i HV-11612 HSCW Pump Discharge Isolation i HV-11613 Control Bldg. ESF Chiller Temp. Control le TV-11675A TV-11740A Control Bldg. ESF Chiller Temp. Control

{ PV-2550A Piping Penetration Room to Atmosphere Piping Penetration Room to Atmosphere PV-2551A TDAFP Speed Governing SV-15133 TDAFP Steam Supply Isolation f

HV-3009 TDAFP Steam Supply Isolation l

HV-3019 Charging Pump Discharge Boron Injection i HV-8116 TDAFP Trip and Throtle Valve

{

, PV-51129 TDAFW Pump Room Air Inlet l

HV-12010 l

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- - - -----. - _,-.-. - --.- _ , - --,, _ ,.- - - - _ .,-, - , n - _ . - _ _ - _ - - -

i Justification At the October 20-24 meeting we agreed that the table of safety-related motor operated valves thermal overload protection bypass devices will remain in the Technical Specifications.

104.7 v=Ils b 4raos A+6, Ios3 n/k Ar /nin C sad lx t clis & hak D af a heky <=*e *w,:em D. C. SOURCES

.lempvohm of 70*f i PROOF & R&g g SURVEILLANCE REQUIREMENTS (Continued)

b. At least once per 92 days and withi i 7 da with battery terminal voltage belo# .....ys. .after a battery discharge

. . , or battery overcharge with battery terminal voltage above 140 volts, by verifying that:

1) The parameters in Table 4.8-2 meet the Category B limits,

2) There is no visible corrosion at either terminals or connectors, or the connection resistance of these items is less than

u. . ; ohm, and Sox i od,

3) The average electrolyte temperature of twelve connected cells is above 55* F.

c. At least once per 18 months by verifying that:

1) The cells, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration,

2) The cell-to-cell and terminal connections are clean, tight, and coated with anticorrosion material,

3) The resistance of each cell-to-cell and terminal connection is less than or equal to EE ^ .. 10 8-Fohm, and SOY t04

4) The battery charger will supply at least 400 amperes for system A and B, 300 amperes for system C, and 200 amperes for system D-at 125 volts nominally for at least 4thours.

d.

8 At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status all of the actual or simulated emergency loads for the design duty cycle when the battery is subjected to a battery service test;

e. At least once per 60 mo'nths, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60-month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1d.; and

.f . At least once per 18 months, during shutdown, by giving performance discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on previous performance tests, or is below 90% of the manufacturer's rating.

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V0GTLE - UNIT 1 3/4 8-12

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REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE M

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 ~

each of theReactorabove limits Coolant by:System leakages shall be demonstrated to be within

a. Monitoring the containment atmosphere radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;seousorparticulate[)

b.

Monitoring the containment normal sumps and reactor cavity sump inventory :nd di:;h;r;; at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

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Performance of a Reactor Coolant System water inventory balanen at' least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and -

dt. Monitoring 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . the Reactor Head Flange Leakoff System at least once per l 4.4.6.2.2 l Each Reactor Coolant System Pressure Isolation Valve specified in Table its limit: shall be demonstrated OPERABLE by verifying leakage to be within 3.4-1 ggg g .

At 1 :st cace per la .7,;n t h s ,

/5M'"'

2 d )f. Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leaka testin performed in the previous 9 monthsfcyc for' Va " 5has not been MV-870/ A /O and HV-870Z A/6l

^rior=to-retttenirng-the=vehe- te service =foHowing= maintenancer repair-er-replac; .;nt werk-on=the= valve,-and:-

dr----Within=24= hours-fel4 wing valve: actuation:dueste-atttomet4e-ee-manuat.

Mtiemor-f4ew-through the-valver e.

As outlined in the ASME Code,Section XI, paragraph IWV-3427(b).

The or 4. provisions of Specification 4.0.4 are not applicable for entry into MODE 3 V0GTLE - UNIT 1 3/4 4-21

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a. & /s.r/ mce due arc 4 n/ceby say' (1esk4rfiyg _

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Justification The proposed wording for 4.4.6.2.2 addresses the concerns expressed by the NRC staff at the October 20-24 meeting in Bethesda and, at the same time, maintains consistency with our response to FSAR Question 210.48 and subsection 3.9.6 of the VEGP SER.

In addition, we feel that the exception for valves HV-8701 A/B and HV-8702 A/B is warranted on the basis that these are the RHR suction isolation valves and they are equipped with position indication in the control room. They are also interlocked to prevent both of them from being opened when RCS pressure is greater than approximately 377 psig and to automatically close before RCS pressure exceeds 750 psig. Also, the RHR system is equipped with a high pressure alarm at the main control board at the discharge of each RHR pump as shown on Figure 5.4.7-1 of the VEGP FSAR.

6 me

15 Insert for 3/4 3-25 Sensor Total Error Functional Unit Allowance (TA) Z (S) Trip Setpoint Allowable Value

4. Containment Radiation (a)

1. Area Low Range NA NA NA 15 mr N,4 (RE-002, RE-003) hrg gocomr[/hr

! 11. Ventilation NA NA NA (C) b4 Particulate Activity (RE-2565A) 111. Ventilation Iodine NA NA NA (C) M A.

Activity (RE-2565B)

IV. Ventilation Caseous Activity (RE-2565C)

NA NA NA (C) Nd O' Ovan3 rehehy operations

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d by,y,g';h phowera f- So h m ergenhon.

lutck 7ks ir an l.s//wl/6e isHio/ dehwr/sd se/pomf enA .

a ,, wie a or & yround w aeya. /wef. Sedgrand/

pec/# %r jg}l ob 8 CUIll O f 6 f Ct W & //4 llJ O ( ) 3 //. 2 ,/, :

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TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

SENSOR TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWA8LE VALUE U 3. Containment Isolation '

a. Phase "A" Isolation

1) Manual Initiation N.A. N.A. N. A'. N.A. N.A.

2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

3) Safety Injection See Functional Unit 1. above for all Safety Injection Trip Setpoints and Allowable Values.

4) Containment Area Radiation / ,) / y/

(Hfgh Range)(RE-0005, (N. A.f /N.A.Jg (V g '" " '

N.A.J .. JN. A.)y)

RE-0006) y

b. Containment Ventilation Z Isolation

1) Manual Initiation N.A. N.A. N.A. N.A. N.A.

2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.

Logic and Actuation Relays

3) Safety Injection See Functional Unit 1. above for all Safety Injection Trip Setpoints and Allowable Values.

4) ainment Area

~

later Later Later er Radioa ty (Low b fbE Range) (RE- E-0003) en.H h ef t. 5) Containment' Ventilation -

fu Radiation ,

g3/4 S-25 i Particulate Activity Later Later Later Later (RE-2565A) Q

,, ii Iodine Activity Later Later Later m

(RE-25658 E 111 Gas ctivity La'ter Later Later Later

-2565C)

O 1

Justification Footnote a is proposed for the containment area low range radiation monitors in order to provide for a setpoint which is appropriate for refueling operations.

Footnote b is proposed in order to allow for adjustment of the setpoint as background level changes. The initial setpoint of 1000 mr/hr should be appropriate for the initial cycle, after which the setpoint will be based on the background level. The setpoint can be adjusted to the sppropriate level without requiring a change to the Technical Specifications.

Footnote c is proposed for the setpoints for the skid-mounted ventilation radiation monitors in order to allow for slight changes in RCS leakage, pinhole leaks in fuel assemblies, etc. without requiring a change to the Technical Specifications. In any event, the limits of Specification 3.11.2.1 will not be exceeded.

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~~ V0GTLE-UNIT 1 6-4

PLANT SYSTEMS PROOF & REVIEW COPY SURVEILLANCE REQUIREMENTS (Continued) c.

At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

1) -M + ion Verifying that the :.+:m ru; system satisfies the in place testing acceptance criteria of greater than or equal to 99.95% filter efficiency while operating the system at a flow rate of # e99 cfm 210% and performing the following tests: f

/6f000 (a) A visual inspection of the control room emergencyOr Hqdiar

. r up system shall be made before each DOP test or activated carbon adsorber section leak test in accordance with Sec-tion 5 of ANSI N510-1980.

~ (b)" An in place DOP test' for the HEPA filters shall be performed in accordance with Section 10 of ANSI N510-1980.

(c) A charcoal adsorber section leak test with a gaseous halo-genated hydrocarbon refrigerant shall be performed in accordance with Section 12 of ANSI N510-1980.

2)

Verifying within 31 days after removal that a laboratory analysis of a representative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing cri-terion of greater than or equal to 99.8% when tested with methyl iodideat80'Cand70%relativehumidity.

3) Verifying a system flow rate of 25,000 cfm + 10% during system '

operation when tested in accordance withjAN3I N510-1975.

d. [ Sec/hrt /2 o f After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Section 13 of ANSI N510-1980 meets the laboratory testing criterion of greater than or 70% equalhumidity.

relative to 99.8% when tested with methyl fodide at f0*C and 3

e. At least once per 18 months by:

1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the system at a flow rate of 25,000 cfm + 10%;

/6,000

2) Verifying that on a Control Room Isolation Test Signal, the system automatically switches into an emergency U;; h; ht k 2 mode of operation with flow through the HEPA filters and charcoal adsorber banks; V0GTLE - UNIT 1 3/4 7-15

PLANT SYSTEMS W & REV N COPY SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch Water Gauge at less than or equal to a pressurization flow of gg o cfm relative to adjacent areas during system operation;

4) Verifying that the heaters dissipate 118 i 6 kW when tested in accordance with Section 14 of ANSI N510-1980; and

5) Verifying that on a Control Room / Toxic Gas Isolation test signal, the control room isolation dampers close within 6 seconds and the system automatically switches into an isolation mode of operation with flow through the HEPA filters and charcoal adsorbers.

f.

Aft'er'eich complete or partial replacement of a HEPA filter bank, by verifying that.the HEPA filter banks remove greater than or equal to 99.95% of the DOP when they are tested in place in accordance with ANSI NS10-1980 while operating the system at a flow rate of

& 000 cfm i 10%; and

/F i,Ow i

g. After each complete or partial replacement of a charcoal adsorber bank, by verifying that the charcoal absorbers remove greater than or equal to 99.95% of a halogenated hydrocarbon refrigerant test gas when tested in-place in accordance with Section 12 of ANSI N510-1980 while operating the system at a flow rate of 19:400 cfm i 10%.

%,000

{

V0GTLE - UNIT 1 3/4 7-16

Justification The flow rate specified in ci, el, f and g was revised to 16,000 CFM on the basis that this is the value appropriate for the Unit 1 control room with the wall separating the Unit 1 side from the Unit 2 side in place.

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

LIMITING SAFETY SYSTEM SETTINGS

$gggg BASES' Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels. The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active. The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active. ." : cr:dit c:

gg r

tt Stur

'^r per:ti:n of 16 tri;. :::::ict:d eith citherth: Inter;;di:" :P -

MN S ;;; Mnne! " th: :::ident-:nMy:::; he;; ;;r, th;ir functi e.a4 poi'N st " rp: !fi:d trip ::ttiqi: i: r:';ut d by thi, :p;cification h t : t nce th: ;;rd' r:lidility of th; Reecter Irctee44cn ( _t: L Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribu-tion, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds), and pressure is within the range between the Pressurizer High and Low Pressure trips. The Set-point is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping dela detectors, (2) pressurizer pressure,ysand from(3)the corepower axial to thedistribution.

loop temperature With .

normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reacto" trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g. , no fuel pellet melting and less than 1% cladding strain) under all possible i

overpower conditions, limits the required range for Overtemperature AT trip, and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat genera-tion rate (kW/ft) is not exceeded. The Overpower AT trip provides protection i

to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

V0GTLE - UNIT 1 B 2-5 SEP 19 506

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44,mmem.,*-r- ---.m==me--w------m.- e.--ww, -,i_----++ + - .- .- ,4.__ _

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.-%w ----- -- + . - ,. , -, __.. - a. - .,..+.

Am.asmonia--.,eia-e-.~ = -- = .- -

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aw-w,.--====..w w-ne s-iW ==.. 4-e- -+- w,w - + -i. -- -_. - -..w w.---.

--- g- --e- g -..-, - , -, u -<, - , - -, * - -,e ,7.- -

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Justification The proposed revision to page B 2-5 is offered in response to a request from the Reactor Systems Branch review to address the use of the source range instrumentation to mitigate the consequences of an inadvertent rod bank withdrawal in Modes 3, 4, and 5.

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Justification The following marked-up pages are provided in an effort to make the numbering of the Proof and Review Copy consistent with plant procedures which have already been established. This will help to minimize the chances for erroneous references to the Technical Specifications in plant procedures.

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3 f. 1,

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TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

' TRIP ANALOG ACTUATING MODES FOR E CHANNEL DEVICE WHICH Q CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE ea FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED Reactor Trip System Interlocks (Continued) 18.

J g Power Range 7 Neutron Flux, P-10 N.A. R(4) R N.A. N.A. 1, 2 j ef urbine Impulse Chamber

. Pressure, P-13 N.A. R R N.A. N.A. 1 l w .

i 1 19. Reactor Trip Breaker Pl. A. N.A. N.A. M(7, 11) N.A. 1, 2, 3 ,

8 Y a 4,5 8 c$

20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. N(7) 1, 2, 3 ,

a Logic _ a a 4,S

21. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(15),R(16) N.A. 1, 2, 3 ,

a 4,5 _]

5

j TABLE 3.3-3 (Continued) i ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS E

' SENSOR TOTAL ERROR E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

U TRIP SETPOINT ALLOWABLE VALUE I

y 4. Steam Line Isolation l

l a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

! b. Automatic Actuation :.ogic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays c.

Containment Pressure--High-2 3.0 .

- 0.71 _ 1.5 _15.0

< psig _15.8

( psig

d. Steam Line. Pressure--Low 13.1

( [10.71] [1.5] >585 psig >567 psig*

R

e. Steam Line Pressure - 5.0 0.5 0

-<-100 psi /s -<-123 psi /s**

Negative Rate--High M 5. Turbine Trip and Feedwater m

Isolation

) bf Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

Actuation Relays cd Steam Generator Water 5.1 2.18 1. 5 <78.0% of -<79.8% of narrow

Level--High-High (P-14) narrow range range instrument l instrument span.

span.

Og Safety Injection See Functional Unit 1 above for all Safety Injection Trip Setpoints and Allowable Values.

6. Auxiliary Feedwater De

o

.f pt Manual Initiation N.A. N.A. N.A. N.A. N.A.

E ay Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. g and Actuation Relays move 4s PM' I'l *h

h TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

SENSOR TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

[ 6. Auxiliary Feedwater (Continued) be Steam Generater water Level--Low-Low

1. Start Motor-Driven 17.0 12.18 1. 5 >17.0% of >15.3% of Pumps Harrow range Harrow range instrument instrument span.

span.

2. Start, Turbine- 17.0 12.18 1. 5 >17.0% of >15.3% of Driven Pump Harrow range Harrow range y instrument instrument l w ,

span. span.

$ c# Safety Injection See Functional Unit 1. above for all Safety Injection Trip Start Motor-Driven Pumps Setpoints and Allowable Values.

1

. dp Loss of or degraded 4.16 kV l ESF Bus Voltage Degraded Degraded

1. Start Motor-Driven Pumps N.A. N.A. N.A. >3598 volts with >3548 volts with 20 s time delay 520stimedelay Loss Loss 529I2 volts with 52912 volts with i <0.8s time i <0.8s time delay deTay em llU

! E

TABLE 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS m

i

SENSOR

' TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE g

Degraded Degraded

2. Turbine Driven Pump N.A. N.A. N.A.

>3598 volts with >3548 volts with 20 s time delay 520stimedelay loss Loss

' M 2 volts with 5'2912 volts with i <0.8s time i <0.8s time

! deTay delay

! e. E Trip of All Main Feedwater N.A. N.A. N.A. N.A.

Pumps, Start Motor-Driven N.A.

w Pumps 1 E Manual Initiafion

> 7. Semi-Automatic Switchover to y Containment Emergency Sump

a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

, and Actuation Relays

b. RWST Level--Low-Low N.A. N.A. N.A. 274 in. from Coincident With 254 in. from tank base tank base i (35.6% of instru- (32.5% of instru-ment span) ment span)

Safety Injection See Functional Unit 1. above for asi saret5rinjection~ Trip 3 Setpoints and Allowable Values.

8. Loss of Power to 4.16 kV ESF Bus N -

~~'

a. 4.16 kV ESF Bus N.A. N.A. N.A. >2912 volts 8" Undervoltage-Loss of Voltage >2912 volts sith a < 0.8 Gith a < 0.8 2 second time second time E__

delay. delay. g

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I{h l

SURVEILLANCE REQUIREMENTS

!' TRIP E ANALOG ACTUATING MODES Z CHANNEL DEVICE MASTER SLAVE FOR WHICH w CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED S. Turbine Trip and Feedwater Isolation M

I b x Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q' 1, 2 Logic and Actuation Relays R c E Steam Generator Water S R M N.A. N.A. N.A. N.A. 1, 2

[. Level-High-High (P-14)

% a p: Safety Injection See Functional Unit 1 above for all Safety Injection Surveillance Requirements.

6. Auxiliary Feedwater g f x Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 a #: Automatic Actuation N.A. N.A N.A. N.A. M(1) N(1) Q 1,2,3 e +* Logic and Actuation d P'M Relays b p: Steam Generator Water Level-Low-Low *

1) Start Motor-Driven Pumps S R M N.A. N.A. N.A N.A 1,2,3

2) Start Turbine Driven .

Pump 5 R M N.A. N.A. N.A N.A 1,2,3 QSee Specification 4.3.3.6 g $a MN

j TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m SURVEILLANCE REQUIREMENTS TRIP E ANALOG ACTUATING MODES Q CHANNEL DEVICE MASTER SLAVE FOR WHICH w CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST i

.IS REQUIRED c .&. Safety Injection See Functional Unit 1. above for all Saf'ety Injection Surveillance Requirements.

ds: Loss of or Degraded N.A. R M N. 4. M.A. N.A. N.A 1,2,3 4.16 kV ESF Bus Voltage ,

! E $ Trip of All Main Feed- N.A. N.A. N.A. R N.A. N.A. N.A 1, 2 j .

i R

h. water PumpsManual ini4iaH0n -

) 7. Semi-Automatic Switchover to Containment Emergency Sump

a. Automatic Actuation N.A. N.A. N.A. N. A.' N(1) M(1) Q 1, 2, 3, 4 Logic and Actuation Relays

b. RWST Level-Low-Low

S R M N.A. N.A. N.A. N.A 1,2,3,4 Coincident With Safety - ' ~ ~ ~-~

j Injection y~~~._. _.

_ ~~ ~~

l nctional Unit 1. above for all Safety Injection Surveillance Requirements.

~~~- .. . . -

Loss of Power to

8. ~ ~ ~ ~ ~ - - -

4.16 kV ESF Bus

a. 4.16 kV ESF Bus N.A. R N N.N. N.A. N.A. N.A. 1,2,3,4 Undervoltage-Loss of Voltage o

"See Specification 4.3.3.6 -

PROOF & REVIEW COPY

1 TABLE 4.3-2 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

{,

SURVEILLANCE REQUIREMENTS TRIP

, z ANALOG ACTUATING MODES Q CHANNEL DEVICE

~ CHANNEL MASTER SLAVE FOR WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY FUNCTIONAL UNIT RELAY SURVEILLANCE

, CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

8. Loss of Power to i

4.16 kV ESF Bus (continued)

b. 4.16 kV ESF Bus N.A. R. M N.A, N.A. N.A. N.A. 1, 2, 3, 4 Undervoltage-Grid

Degraded Voltage Iw 9. Engineered Safety .

l } Features Actuation

w System Interlocks M a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1, 2, 3 j Pressure, P-11

b. Reactor Trip, P-4 N. A. N.A. N.A. R N.A. N. A. N.A. 1, 2, 3

10. Control Room Ventilation Emergency Mode 1

Actuation

a. Manual Initiation N.A. N. A. N.A. R N.A. N.A. N.A. All Modes

b. Automatic Actuation

, Logic and Actuation Relays N.A. N.A. N.A. N.A. M(1) N.A. N.A. All Modes dpr Intake Radiogas -

Monitor S R M N.A. N.A. N.A. N.A.

(RE-12116, RE-12117) All Modes C. .A Safety Injection See Functional Unit 1 above for all Safety Injection Surveillance Requirements TABLE NOTATION (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

I PROOF & REVIEW COPY

TABLE 4.3 4 A pctnenar eney SEISMIC HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ANALOG CHANNEL ACCESSIBLE CHANNEL CHANNEL OPERATIONAL DURING INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST MODES

1. Triaxial Time-History Accelerographs

a. Free Field (500 ft from containment) M R SA All

b. Unit 1 Containment Gallery (basemat) M R SA All

c. Unit 1 Containment Operating Floor M R SA All

d. Auxiliary Building Basemat M R SA All

e. Unit 1 Containment Pressurizer Support M R SA 5, 6

2. Triaxial Peak Accelerographs

a. Reactor Coolant Pump Motor (210 ft) N.A. R N.A. 5, 6

b. Steam Generator (185 ft) N.A. R N.A. 5, 6

c. NSCW Piping Outside Aux. Bldg. (220 ft) N.A. R N.A. All

3. Triaxial Seismic Switches Q. Containment Tendon Gallery (basemat) M R SA All

4. Triaxial Response-Spectrum Analyzer Q. Control Room

M R N.A. All

With reactor control room indications.

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l V0GTLE - UNIT 1 3/4 3-44

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TA8LE 4.3-6 (Continued)

RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRtMENTATION SURVEILLANCE REQUIREMENTS m

,g ANALOG CHANNEL q CHANNEL SOURCE CHANNEL , OPERATIONAL MODES FOR WHICH g INSTRUMENT SURVEILLANCE CHECK CHECK CALIBRATION TEST IS REQUIRED

4. Plant Vent

a. Noble Gas Activity Monitor D M R(3) Q(2) a (RE-12442C or RE-12444C)

b. Iodine Sampler W N.A. N.A. N.A.

a (RC-12ii_Ter[RE-124448)

c. Particulate Sampler W N.A. N.A. N.A. a R

(E-1?fi2^. er(RE-12444A)

,y d. Flow Rate Monitor D N.A. R Q a

.g (FI-12442)

{ e. Sampler Flow Rate Monitor D N.A. R Q a (FI-12442 or FI-12444)

4. Parmulate. Annor RE-ia 94a A ')

O ' (l od;ne Ae+tv,'+y Monikc

( RE- 12492 B) 3

3/4.33.8 Ajof- Used Table. 44-3 No t- U.secl t

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REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY N LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the reactor coolant shall be limited to:

a. Less than or equal to'1 microcurie per gram DOSE EQUIVALENT I-131, and

b. Less than or equal to 100/E microcuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than 1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANOBY with T 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and avg less than

b. With the specific activity of the reactor coolant greater than 100/E microcuries per gram of gross radioactivity, be in at least HOT STANDBY with T,yg less than 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4, and 5:

With the specific activity of the reactor coolant greater than I microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E micro-Curies per gram of gross radioactivity, perfom the sampling and analysis requirements of Item 4.a) of Table 4.4-/funtil the specific activity of the reactor coolant is restored to within 1".s limits.

i H

SURVEILLANCE REQUIREMENTS 4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-g 4

With T,yg greater than or equal to 500*F.

V0GTLE - UNIT 1 3/4 4-25

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'l TABLE 4.4-#

6 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE C) AND ANALYSIS PROGRAM E

, TYPE OF MEASUREMENT SAMPLE AND ANALYSIS MODES IN WHICH SAMPLE g AND ANALYSIS FREQUENCY AND ANALYSIS REQUIRED i *1 1. Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . 1, 2, 3, 4 l g Determination *

2. Isotopic Analysis for DOSE EQUIVA- 1 per 14 days. 1 LENT I-131 Concentration

3. Radiochemical for 5 Determination ** 1 per 6 months *** 1

4. Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 1#, 2#, 3#, 4#, 5#

Including I-131, 1-133, and I-135 whenever the specific activity exceeds 1 pCi/ gram DOSE -

I EQUIVALENT I-131 w or 100/E pCi/ gram of

) gross radioactivity, and t b) One sample between 2 1,2,3 4

y and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following a THERMAL POWER change exceeding 15%

of the RATED THERMAL POWER within a 1-hour period.

4 .

I

H TABLE 4.4-J'(Continued)

TABLE NOTATIONS SNM "A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the reactor coolant except for radionuclides with half-lives less than 14 minutes and all radioiodines. The total specific activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta-emitting radionuclides.

- A radiochemical analysis for 5 shall consist of the quantitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 14 minutes and all radioiodines, which are identified in the reactor coolant. The specific activities fgr these individual radio-nuclides sample.

shall be used in the determination of_E for the reactor coolant Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.

- - Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subcritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Until the specific activity of the Reactor Coolant System is restored within its limits.

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REACTOR COOLANT SYSTEM N & REVIEW COPY 3/4.4.9 PRESSURE / TEMPERATURE LIMITS l

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a. A maximum heatup of 100*F in any 1-hour period,

b. A maximum cooldown of 100*F in any 1-hour period, and

c. A max'imum temperature change of less than or equal to 10'F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity uf the Reactor Coolant System; rietermine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANOBY within the .

next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the RCS T,yg and pressure to less than 200'F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4. 1.1 e Reactor Coolant System temperature and pressure shall be det reine to be within the limits at least once per 30 minutes during system hea up ooldown, and inservice leak and hydrostatic testing operations.

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s REACTOR COOLANT SYSTEM PROOF & REVN m 3/4.4.10 STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.10 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with Specification 4.4.10.

APPLICABILITY: All MODES.

ACTION:

a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200*F.

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural l integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

't. 4 10. I bio f &

4.4.10.21In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected per the recommendations of Regulatory Position C.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

. 4.4.10.J3 In addition to the requirements of Specification 4.0.5, the four main steam lines and feedwater lines.from the containment penetration flued head outboard weld, to the upstream weld of the five-way restraint, which is down-stream of the main steam isolation valves, shall be inspected. The extent of the inservice examinations completed during each inspection interval (ASME Code Section XI) shall provide 100 percent volumetric examination of circumferential and longitudinal welds to the extent practical.

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_ _ _ = _ _

ML20213G962

Text

Dear Mr. Denton:

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