ML20107G291

Amends 100 & 99 to Licenses DPR-32 & DPR-37,respectively, Changing Tech Spec Re TMI Items for Relief Valves,Reactor Vessel Head Vents & Chlorine Room Detection Sys

ML20107G291
Person / Time
Site:	Surry
Issue date:	10/15/1984
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20107G293	List: ML20107G291
References
NUDOCS 8411070305
Download: ML20107G291 (25)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION

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WASHmCTON. O. C. 20555 '

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VIRGINIA ELECTRIC AND POWER COMPANY 00CKET NO. 50-280 SURRY POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 100 License No. OPR-32 1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated March 31,1983, June 16,1983, and February 9,1984 (as supplemented February 14 and 21,1984),

complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's

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rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulationst 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 3.8 of Facility Operating Licenst.

No. DPR-32 is hereby amended to read as follows:

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8411070305 841015 l

PDR ADOCK 05000200 P

PDR L

T io 2-(B) Technical Specifications The Technical Specifications contained.in Appendix A, as revised through Amendment No.100, 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 NUCLE R REGULATORY COMMISSION f

)A1AfLkkJ m -... m.44ri a,\\

f Operating Reactor anch v1 Division of Licen

Attachment:

Changes to the Technical Specifications Date of Issuance: October 15, 1984 9

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UNITED STATES e

• [ T.- f i NUCLEAR REGULATORY COMMISSION 1 g*g(1, j W ASHING TON. C. C. 20555

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VIRGINIA ELECTRIC M D POWER COMPANY DOCKET NO. 50-281 SURRY POWER STATION, UNIT NO. 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE Anendment No. 99 License No. OPR-37

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1.

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

A.

The application for amendment by Virginia Electric and Power Company (the licensee) dated March 31, 1984, June 16, 1983, and February 9,1984 (as supplemented February 14 and 21,1984),

complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the 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 1

Specifications as indicated in the attachment to this license amendment, and paragraph 3.8 of Facility Operating License No. OPR-37 is hereby amended to read as follows:

D

-2 (C) Technical-Specifications-The Technical Specifications contained in Appendix A, as revised through Amendment No.- 99, 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 and shall be implemented prior to startup'from the 1986 refueling -outage.

FOR'THE NUCLEAR RE TORY COMMISSION t----

e$en A. Varga,s _

Operating Reactors Br ~ c #1

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Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: October 15, 1984 e

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. ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

100 FACILITY OPERATING LICENSE'NO. DPR-32 AMENDMENT NO.

99 FACILITY OPERATING LICENSE NO. DPR-37 DOCKET NOS. 50-280 AND 50-281 Revise Appendix A as follows:

Remove Pages~

Insert Pages 3.1-5 3.1-5 3.1-Sa 3.1-Sa 3.1-5b 3.1-5b 3.1-Sc 3.1-Sc 3.7-2a 3.7-2a 3.7-2b 3.7-9 3.7-9

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3.7-9a 3.7-9b 3.7-9c 3.7-20 3.7-20 3.7-21 3.7-21 3.7-22 4.1-1 4.1-1 4.1-9 4.1-9 4.1-9a 4.1-9a 4.1-9d 4.1-9d 6.4-7 6.4-7 l

i 1

e-

-TS 3.1-5

c. With the pressurizer otherwise inoperable, be in at least hot shutdown

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with the reactor trip-breakers-open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and the reactor.

coolant system temperature and pressure-less than 350*F and 450 psig, respectively, within the following 12. hours.

6.

Relief Valves,

a. Two' power operated relief valves (PORVs) and their associated block valves shall be operable whenever the reactor keff is 2 0.99.

b. With one or more PORVs inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to operable status or close the associated block valve (s) and

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remove power from the block valve (s); otherwise, be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

c. With one or more block valve (s) inoperable, within.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block. valve (s) to operable status or close the block valve (s) and remove power from the block valve (s); otherwise, be in at least hot shutdown within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within a

the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

7.

Reactor Vessel Head Vents

a. At least two Reactor Vessel Head vent paths consisting of two isolation valves in series powered from emergency buses shall be operable and closed whenever RCS temperature and pressure are > 350*F and 450 psig.

Amendment Nos.100 and Nos. 99

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ej TS 3.1-Sa.

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b. With one. Reactor Vessel Head vent' path inoperable; startup'and/or.

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-power operation may continue provided the inoperable vent path is maintained closed with power _ removed from the valve actuator of both isolation valves in the ino},erable vent path.

c.:With two Reactor Vessel Head vent paths inoperable; 4

maintain the inoperable vent path _ closed with power removedifrom the; valve actuator of all isolation valves in the inoperable vent paths, and restore.at least one of the vent paths to operable status within 30 days or be-in hot standby within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in cold shutdown within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

t Basis Specification 3.1.A-1 requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling flow in the event of a loss of reactor coolant flow accident.

This provided flow will maintain'the I

DNBR above 1.30.

Heat transfer analyses also show that reac' tor heat equiva-lent to approximately 10% of rated power can be removed with natursl circula-i tion; however, the plant is not designed for critical operation with natural I

circulation or one loop operation and will not be operated under these conditions.

I I

When the boron concentration of the Reactor Coolant System is to be reduced l

the process must be uniform to prevent sudden reactivity changes in the l

reactor.

Mixing of the reactor coolant will be sufficient to maintain a uniform ' concentration if at least one reactor coolant pump or one residual heat removal pump is running while the change is taking place. The residual i

heat removal pump will circulate the equivalent of the reactor coolant system l

I volume in approximately one half hour.

Amendment Nos. 100 and Nos. 99 ;

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TS 3.1-5b One steam generator capable of performing its heat transfer function will after a provide sufficient heat-removal capability to remove core decay heat normal reactor shutdown. The requirement for redundant coolant loops ensures the capability to remove core decay heat when the reactor coolant system average temperature is less than or equa'l to 350*F.- Because of the low-low l

steam generator water level reactor trip, normal reactor criticality cannot be achieved without water in the steam generators in reactor coolant loops with open loop stop valves.

The requirement for two operable steam generators, combined with the requirements of Specification 3.6, ensure adequate heat removal capabilities for reactor coolant system temperatures of greater than-350*F.

Each of the pressurizer safety valves is designed to relieve 295,000 lbs. per-

' hr. of saturated steam at the valve setpoint. Below 350*F and 450 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure.

There are no credible accidents which could occur when the Reactor' Coolint System is connected to the Residual Heat Removal System which could give a surge rate exceeding the capacity of one pressurizer safety valve.

Also, two safety valves have a i

capacity greater than the maximum surge race resulting from complete loss of load.

The limitation specified in item 4 above on reactor coolant loop isolation will prevent an accidental isolation of all the loops which would eliminate the capability of dissipating core decay heat when the Reactor Coolant System is not connected to the Residual Heat Removal System.

kr.:rdent Nos.100and Nos. 99

I TS 3.1-5c The requirement for' steam bubb1'e formation in_the pressurizer when the reactor

- passes 1% subcriticality will ensure that the Reactor Coolant System will not be_ solid when criticality is achieved.

The requirement that 125 Kw of pressurizer heaters and

• heir associated controls be capable of being supplied electrical power from an emergency bus provides assurance that - these heaters can be energized during a loss of offsite power condition to maintain natural. circulation at hot shutdown.

The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leakage path.

The accumulation of non-condensable gases in the heactor Coolant System may result from sudden depressurization, accumulator discharges and/or inadequate core cooling conditions. The function of the Reactor Vessel Head Vent is to i

remove non-condensable gases from the reactor vessel head. The Reactor Vessel i

Head Vent is designed with redundant safety grade vent paths.

Venting of

- non-condensable gases from the pressurizer steam space is provided primarily through the Pressurizer PORVs. The pressurizer is, however, equipped with a steam space vent designed with redundant safety grade vent paths.

References:

1 (1) FSAR Section 14.2.9 (2) FSAR Section 14.2.10 Anendment Nos. 100 and Nos. 99 L

m p.

-j TS 3.7-2a.

3.

~The requirements of Specification 3.0.1 and 6.6.2 are not

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applicable.

-F.

The accident monitoring instrumentation.for its associated operable components _ listed in.TS Table 3.7-6 shall be operable'in accordance with the following:

1.

With the number of operable accident monitoring instrumentation channels less'than the total numb'er of channels shown in TS f

Table 3.7-6, either restore the inoperable channel (s) to operable status within 7 days or be in 'at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

2.

With the number _of operable accident monitoring instrumentation channelslessthantheminimumchannelsoierablerequirementof' TS Table 3.7-6, either restore the inoperable channel (s) to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least hot shutdown within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

G.

The Main Control Room Chlorine Detection System shall be operable at all times. The number of operable channels, alarm / trip setpoint, and required operator actions shall be as

.specified in Table 3.7-7.

This capability shall be i

demonstrated by the surveillance requirements specified in Table 4.1-1.

Amendment Nos.100 and Nos. 99 e

TS 3.7-2b

[hecontainmenthydrogenanalyzersandassociatedsupportequipment H.

.shall be operable in accordance with the following:

1.=

^~ reactor shall not be made critical nor'be operated at power

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without two independent. containment hydrogen analyzers

. operable..

2.

During power operation or return to criticality from hot shutdown conditions, the following restrictions apply:

With'one hydrogen analyzer inoperable, restore the a.

inoperable analyzer to operable status within 30 days or be in at least. hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.

^ With both hydrogen analyzers inoperable, restore at least one analyzer to operable status within 7 days or be in at-- '

least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Note: Operability of the hydrogen analyzers includes proper operation of the respective Heat Tracing System.

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Amendment Nos. 100 and Nos. 99

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.TS 3.7-9 4.h '

" monitor' indication.- The pressurizer safety valves utilize an acoustic monitor'

-channel and a downstream high. temperature' indication channel. 'This capability.

'is consistent,with th'e: recommendations of Regulatory Guide'1.97, 4

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'" Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditons.During;and Following an' Accident" December'1975, and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term-Recommendations". Potential accident effluent release paths are equipped with radiation monitors to detect-and measure. concentrations of noble gas' fission-products in plant gaseous effluents during and following.an accident.- ~The-effluent release paths monitored are the Process Vent Stack, Ventilation Vent--

Stack, Main. Steam Safety Valve-and Atmospheric Dump Valve discharge and the Auxiliary Feedwater Pump Turbine Exhaust. These mon'. tors meet the requirements of NUREG 0737.

Radioactive Liquid Effluent Monitoring Instrumentation

-The radioactive liquid effluent instrumentation is provided to. monitor and control, as applicable, the releases of radioactive materials in liquid-effluents.during actual or potential releases of liquid effluents. The alarm /

trip setpoints for these instruments shall be calculated. and adjusted in accordance with the procedures in the ODCM to ensure that the alarm / trip will i

occur prior to exceeding the limits of 10 CFR Part 20.

The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 of Appendix A to 10 CFR Part 50.' The purpose of tank level indicating devices is to assure the detection and control of leaks that if not controlled could potentially result in the transport of

= radioactive materials to unrestricted areas.

Radioactive Gaseous Effluent Monitoring Instrumentation j

The radioactive gaseous effluent instrumentation is provided to monitor and Amendment Nos.100 and Nos. 99

L TS 3.7-9a o

~ control, as applicable. the. releases of radioactive materials in gaseous effluents during actual or potential releases of gaseous effluents. The alarm / trip setpoints for chese ' instruments shall be calculated and adjusted in accordance with the procetures in the ODCM to ensure that the' alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. This instrumentation also includes provisions for monitoring (and controlling) the concentrations of potentially explosive gas mixtures in the waste gas holdup system. The operability'and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63 and 64 or Appendix A to 10 CFR Part 50.

Containment Hydrogen Analyzers Continuous indication of hydrogen concentration in the containment atmosphere ~

is provided in the control room over the range of 0 to 10 percent hydrogen concentration.

These redundant, qualified hydrogen analyzers are shared by Units 1 and 2 with the capability of measuring containment hydrogen concentration for the range of 0 to 10 percent and the installation of instrumentation to indicate and record this measurement.

A transfer switch with control circuitry is provided for the capability of Unit I to utilize both analyzers or for Unit 2 to utilize both analyzers.

Each unit's hydrogen analyzer will receive a transferable power supply from Unit 1 and Unit 2.

This will ensure redundancy for each unit.

Indication of Unit 1 and Unit 2 hydrogen concentration is provided on Unit 1 PAMC panel and Unit 2 PAMC panel. Hydrogen concentration is also recorded on qualified recorders.

In addition, each hydrogen analyzer is provided with an alarm for trouble /high hydrogen content. These alarms are located in the Amendment Nos. 100 and Nos. 99

n1 i

p E-TS 3.7-9b control room.

!The supply lines installed from-the containment pen'etrations to'the hydrogen

' analyzers have Category I Class IE heat tracing applied. The. heat tracing isystem receives the same transferable emergency power as is provided to the containment hydrogen analyzers. The' heat trace system is de-energized during normal. system operation. Upon receipt of a safety injection signal (Train A or Train B), the system is automatically started, after a preset time delay, to bring the piping process temperature to 250*F + 10*F within 20 minutes.

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Each heat trace circuit is equipped with an RTD to provide individual circuit readout, over temperature alarm and cycles the circuit to maintain the process 9

~temperature via the solid state control modules.

The hydrogen analyzer heat trace system is equipped with high temperature, locs of D.-C. power, loss of A. C. power, loss of control power and failure of automatic initiation alarms.

I I

Control Rocm Chlorine Detection System j

i The operability of the chlorine detection system ensures that sufficient i

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capability is available to promptly detect and automatically initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel, and is consistent with the recommendations of Regulatory Guide 1.95, " Protection of huclear Power Plant Control Room Operators Against an Accidental Chlorine Release".

February 1975.

Aiemdient Nos.

100 and Nos. 99 a

p, _..

TS 3.7-9c References (1) FSAR - Section 7.5 (2) FSAR - Section 14.5 (3) FSAR - Section 14.3.2 (4) FSAR - Section 11.3.3

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e AT&didit Nos. 100 and Nos. 99

TABLE 3.7-5 AUTOMATIC FUNCTIONS OPERATED FROM RADIATION MONITORS ALARM AUTOMATIC FUNCTION MONITORINC ALARM SETPOINT MONITOR CHANNEL AT ALARM CONDITIONS REQUIREMENTS pCI/cc 1.

Process vent particulate Stops discharge f rom contain-See Specifications Particulat,g64x10~

and gas monitors ment vacuum systems and waste 3.11 and 4.9 Cast 9x10 (RM-CW-101 & RM-CW-102) gas decay tanks (shuts Valve Nos. RCV-CW-160. FCV-CW-260 FCV-CW-101) 2.

Component cooling water Shuts surge tank vent valve See Specifications Twice Background-radiation monitors HCV-CC-100

.3.13 and 4.9

~3 3.

Liquid waste disposal Shuts efiluent discharge See Specification

&l.5x10 radiation monitors valves.FCV-LW-104A and 3.11 and 4.9 (RM-LW-108)

FCV-LW-104B 4.

Condenser air ejector Diverts flow to the contain-See Specification 61.3 radiation monitors ment of the affected unit 3.11 and 4.9' i

(RM-SV-Ill & RM-SV-211)

(0 pens TV-SV-102 and shuts, TV-SV-103 or opens TV-SV-202 p

and shuts TV-SV-203)

Particulate s 9x10 '

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5.

Containment particulate Trips affected unit's purge See Specifications g

and gas monitors supply and exhaust fans, 3.10 and 4.0 Cas s lx10 (RM-RMS-159 & RM-RMS-160, closes affected unit's f

RM-RMS-259 & RM-RMS-260) purge air butterfly valves (MOV-VS-100A, B, C & D or I

MOV-VS-200A, B, C & D) 8 6.

Manipulator crane area Trips affected unit's purge

.See Specifications s 50 mren/hr monitors (RM-RMS-162 &

supply and exhaust fans, 3.10 and 4.9 i

RM-RMS-262) closes affected unit's k

purge air butterfly valves (MOV-VS-100A, B C & D or f

110V-VS-200A, B, C & D)

~2 I

~rrocess vent n9rmal and Stops discharge from contain-See Specifications Cas s 9x10 7.

high range effluent monitors ment vacuum system and waste 3.11 and 4.9 e*

(RM-CW-130-1 and RM-GW-gas decay tanks (shuts valves 130-2)

FCV-CW-160. FCV-GW-260, and FCV-CV-101)

tABLt J.7-6 ACdibENT HONITORINC INSTRUMENTAttDN TOTAL NO.

MININUM CHANNELS OF CilANNELS OPERABLE INSTRUMENT 1.

Auxiliary Fecdwater Flow Rate I per S/G 1 per S/G 2

1 2

2.

Reactor Coolant System Subcooling Margin Monitor 3.

PORV Position Indicator (Primary Detector) 1/ valve 1/ valve 4.

PORV Position Indicator (Backup Detector) 1/ valve 0

5.

PORV Block Valve Position Indicator 1/ valve 1/ valve 6.

Safety Valve Position Indicator (Primary Detector) 1/ valve 1/ valve 7.

Safety Valve Position Indicator (Backup Detector) 1/ valve 0

2 1

8.

Reactor Vessel Coolant Level Monitor 2

1 9.

Containment Pressure 2

1 10.

Containment Water Level (Narrow Range) 2 1

11: Containment Water Level (Wide Range) 2 1 (Note 1 b and c only) g 12.

Containment High Range Radiation Monitor 2

2 (Note 1, a, b, and c) kll 13.

Process Vent High Range Effluent Moditor k

2' 2 (Note 1, a, b, and c)

Er

14. Ventilation Vent High Range Effluent Monitor 3

3 (Note 1, a, b, and c)

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15. Main Steam High Range Radiation Monitors l

(Units 1 and 2) is 4

c) 16.

Aux. Feed Pump Steam Turbine Exhaust Radiation

~

l (Note 1, a, b, and c)

C I

Monitor L

With the number of operable channels less than required by the Minimum Channels Operable requirements 4

/

4 Note 1:

Initiate the preplanned alternate method of monitoring the appropriate parameter (s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> g

Either restore the inoperable channel to operable status within 7 days of the event, a.

or m

b.

Prepare and submit a Special Report to the commission pursuant.to specification 6.6 within 14 days following the event outlining the action taken, the cause of the inoperability and the' c.

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plans and schedule for restoring the system to operable.

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TABLE 3.7 -7 MAIN CONTROL 50(36 CNI.ORINE DETECTION SYSTEN 4 '

g 2

3 i

Operator Action if Condition in Total No.

No.

Functional Unit of Channela Alarm / Trip Setpoint Column 2 Cannot be Met 2

s 5 pra chlorine with one channel inoperable, re,-

1.

Chlorine Detector store the inoperable channel l

within seven dayss or within the ne't 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate and main-u tain operation of the control I

room eastgency wantilation sys-i i

tem.

h

)

With two channels inoperable.

within one hour initiate and maintain operation of the control room emergency ventilation eye-

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TS 4.1-1 4.1 OPERATIONAL SAFETY REVIEW Applicability Applies to items directly related to safety limits and limiting conditions for operation.

Objective To specify the minimum frequency and type of surveillance to be applied to unit equipment and conditions.

Specification A.

Calibration,, testing, and checking of instrumentation channels shall be performed as detailed in Table 4.1-1 and 4.a-2.

B.

Equipment tests shall be conducted as detailed below and in Table 4.1-2A.

1.

Each Pressurizer PORV shall be demonstrated operable At least once per 31 days by performance of a channel functional a.

test, excluding valve operation, and b.

At least once per 18 months by performance of a channel calibration.

2.

Each Pressurizer PORV block valve shall be demonstrated operable at least once per 92 days by operating the valve through one complete cycle of full travel.

ATs digit Nos.

100 and Nos. 99

(7 ;,n c.

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L.-.

TS 4.1-la E.'

3.:

The pressuriser' water volume shall be determined to be within its j,

-limit as defined in Specification 2.3.A.3.a at least once per.12l hours whenever' the reactor' is 'not. subcritical by at least 1% ak/k.

J

. 4.-

Each' Reactor Vessel Head vent path isolation valve not required.to

-be closed by Specification 3.1.A.7a or 3.1.A.7.b shall be demonstrated operable at each cold shutdown but not more often than once per 92 days by operating the valve through one complete cycle of fu11Ttravel from the' control room.'

-5.

Each. Reactor Vessel Head vent path shall be demonstrated operable following each refueling by:

a.'

. Verifying that the upstream manual isolation valve in each vent path is locked in the open position.

b.

Cycling'each. isolation valve through at Itact one complete cycle of full travel from the control room.

c.

Verifying flow through the reactot vessel head vent systou vent-paths.

i C.

Sampling tests shall be conducted as detailed in Table 4.1-2B.

D.

Whenever containment integrity is not required, only the asterisked items in Table 4.1-1 and 4.1-2A and 4.1-2B are applicable.

E.

Flushing of sensitized stainless steel pipe sections shall be conducted as detailed in TS Table 4.1-3A and 4.1-35.

Anandment Nos. 100 and Nos.- 99

$ +

E

.TS 4.1-1b F.

The outside containment purge and vent isolation valves and the isolation valve in the steam jet air ejector suction line outside containment shall be determined. locked, sealed,' or otherwise secured in the closed. position at least once per 31 days.

G.

The inside containment purge and vent isolation valves shall be determined 1ccked, seal, or otherwise secured in the closed position each cold shutdown but not more of ten than once per 92 days.

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i 1

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4 licr.rdist Nos. 100 and Nos. 99 l

+.

TAtl.E 4.1-1 (Coatfneed) a.

Remarks 1

Channel Check Calibrate Test

,*i Description l

34. Loss of Power N.A.

R M

i 4.16 KV Emergency Bus s.

i undervoltage (1.oes of voltage)

N.A.

R M

J b.

4.16 EY Emergency Bus undervoltage (Degraded voltage) 5 R

H

]

35. Control Room Chlorine Detectors a

I 4

.I J

t 4

k i

l l

5 6

rt h

1 P

a*

n O

Y.

,I 2

)

_ ;[

.);?"

1 i

a m

TABLE 4.1-2 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS, CHANNEL CHANNEL-INSTRUMENT CHECK CALIBRATION P

R 1.

Auxiliary Feedwater Flow Rate M

R 2.

iteactor Coolant System Subcooling Margin Monitor M

R 3.

PORY Position Indicator (Primary Detector)

M R

4.

PORV Position Indicator (Backup Detector)

M'

-R 5.

PORV Block Valve Position Indicator M

R~

6.

Safety Valve Position Indicator M

R 7.

Safety Valve Position Indicator'(Backup Detector)

M R

8.

Reactor Vessel Coolant Level Monitor M

R 9.

Containment Pressure 2l*

M R

l

10. Containment Water Level (Narrow Range)

EF M

R g

11.

Containment Water Level (Wide Range)

.=

5 O

-t k

g

.4 1

TABl.E 4.1-2A (CONTINUED)

-) '

MINIMtM FREQUENCY FOR EQUIPMENT TESTS FSAR SECTION -

DESCRIPTION TEST FREQUENCY REFERENCE 18.

Primary Coolant System Functional

1. Periodic leakage testing on each valve listed in Specification 3.1.C.7a shall be accomplished prior to entering power operation condition after every time the plant is placed in the cold shutdown condition for refueling, af ter each time the plant is placed in-cold shutdown condition

.for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomp-11shed in the preceeding 9 months, and prior to returning the valve to service-after maintenance.-repair or replace-ment work is performed.

Functional Semi-Annual (Unit at power or shutdown)

19. Containment Purge MOV Leakage if purge valves are operated during interval (c) 20.

Containment Hydrogen Analyzers

a. Channel Check Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

b. Channel Functional Test once per 31 days

c. Channel Calibration using sample gas Once per 92 days on containing:

staggered basis-1.*0ne volume percent (1 0.25%)

hydrogen, balance nitrogen

2. Four volume percent (1 0.25%).

f hydrogen. balance nitrogen

3. Channel Calibration test will include startup and operation of 8

the Heat Tracing System 4

To satisfy AIARA - requirements, leakage may be measured indirectly (as from the performance of pressure k

tim indicators) if accomplished in accordance with approved procedures and supported by computations showing that f

method is capable of demonstrating valve compliance with the leakage criteria.

{

o.

111nimum differential test pressure shall not be below 150 psid.'

(c) Refer to Section 4.4 for acceptance criteria.

• kee Rnerification 4.1.D.

i TS 6.4-7 L.

Iodine Monitoring The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital area under accident conditions. This program shall include the following:

1.

Training of personnel, 2.

Procedures for monitoring, and 3.

Provisions for maintenance of sampling and analysis equipment.

M.

Post-Accident Sampling A program shall be established, implemented and maintained which will ensure the capability to obtain and analyze reactor coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions. The program shall include the following:

1.

Training of personnel, 2.

Procedures for sampling and analysis, 3.

Procedures for maintenance of sampling and analysis equipment.

Amendment Nos. 100 and Nos. 99