Text

Proposed Tech Specs,Incorporating TMI Mods Scheduled for Implementation After 811231,per Generic Ltr 83-37
	ML17346A396
Person / Time
Site:	Turkey Point
Issue date:	06/15/1984
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17346A395	List: L-84-155, Application for Amend to Licenses DPR-31 & 41,changing Tech Specs to Incorporate TMI Mods Scheduled for Implementation After 811231,per Generic Ltr 83-37; ML17346A396;
References
	RTR-NUREG-0737, RTR-NUREG-737 GL-83-37, NUDOCS 8406220117
	Download: ML17346A396 (51)
	v • d • e

~

TABLEOF CONTENTS Section 1.0 1.1 1.2 1.3 1.0

'.5 1.6 1.8 1.9 1.10 1.11 1.12 1.13 1.10 1.15 1.16 1.17 1.18 1.19 1.20 1.21 1.22 1.23 1.20 Title TECHNICAL SPECIFICATIONS DEFINITIONS Safety Limits LimitingSafety System Settings LimitingConditions for Operation Operable Containment Integrity

, Protective Instrumentation Logic Instrumentation Surveillance Shutdown Power Operation Refueling Operation Rated Power Thermal Power Design Power Dose Equivalent I-131 Power Tilt Interim Limits Low Power Physics Tests Engineered Safety Features Reactor Protection System Safety Related Systems and Components Per Annum Reactor Coolant System Pressure Boundary Integrity Coolant Loop E-Average Disintegration Energy

~Pa e 1-1 1-1 1-1 1-1 1-1 1-2 1-2 1-3 1-3 I-0 1-0 I-0 14-I-0 1-5 1-5 1-6 1-6 1-6 1-6 1-6 1-6 1-6 1-7 1-7 2.0 2.1 2.2 2.3 SAFETY LIMITSAND LIMITINGSAFETY SYSTEM SETTINGS Safety Limit, Reactor Core Safety Limit, Reactor Coolant System Pressure Limiting Safety System Setting, Protective Instrumentation 2.1-1 2.1-1 2.2-1 2.3-1 3.0 3.1 3.2 3.3 LIMITINGCONDITIONS FOR OPERATION Reactor Coolant System Operational Components Pressure-Temperature Limits Leakage Maximum Reactor Coolant Activity Reactor Coolant Chemistry DNB Parameters Control Rod and Power Distribution Limits Control Rod Insertion Limits Misaligned Control Rod Rod Drop Time Inoperable Control Rods Control Rod Position Indication Power Distribution Limits In-Core Instrumentation Axial Offset Alarms Containment 3.0-1 3.1-1 3.1-1 3.1-2 3.1-3 3.1-0 3.1-6 3.1-7 3.2-1 3.2-1 312-2 3e2 2 3.2-2 3e2 3 3.2-3 302 7 3.2-8 3.3-1

~ 840b220ii7 840bl.5 PDR ADOCK 05000250 P

PDR Q AMENDMENTNOS.

and

Section 0.10 0.11 0.12 0.13 0.10 0.15 0.16 5.0 5.1 5.2 5.3 5.0 6.0 6.1 6.2 6.3 6.0 6.5 6.6 6.7 6.8 6.9 6.10 6.11 6.12 6.13 6.10 6.15 6.16 B2.1 B2.2 B2.3 B3.1 B3.2 B3.3 TABLEOF CONTENTS (Continued)

Title AuxiliaryFeedwater System Reactivity Anomalies Environmental Radiation Survey Radioactive Materials Sources Surveillance Snubbers Fire Protection Systems Overpressure Mitigating System DESIGN FEATURES Site Reactor Containment Fuel Storage ADMINISTRATIVECONTROLS Responsibility, Organization Facility Staff Qualifications Training Review and Audit Reportable Occurrence Action Safety LimitViolation Procedures Reporting Requirements Record Retention Radiation Protection Program High Radiation Area Post Accident Sampling Systems Integrity Iodine Monitoring Back-up Methods for Determining Subcooling Margin Bases for Safety Limit, Reactor Core Bases for Safety Limit, Reactor Coolant System Pressure Bases for Limiting Safety System Settings, Protective Instrumentation Bases for Limiting Conditions for Operation, Reactor Coolant System Bases for Limiting Conditions for Oper ation, Control and Power Distribution Limits Bases for LimitingConditions for Operation, Containment Bases for Limiting Conditions for Operation, Engineered Safety Features

~Pa e 0.10-1 0.11-1 0.12-1 Oe13-I 0.10-1 0.15-1 0.16-1 5.1-1 5.1-1 5.2-1 5.3-1 5.0-1 6-1

'-1 6-1 6-5 6-5 6-6 6-10 6-10 6-10 6-16 6-27 6-29 6-29 6-30

'6-30 6-30 6-30 B2.1-1 82.2-1 82.3-1 B3.1-1 B3.2-1 B3.3-1 B3.0-1 AMENDMENTNOS.

and

~

Section B3.5 B3.6 B3.7 B3.8 B3.9 B3.10 B3.11 B3.12 B3.13 B3.10 B3.15 BV.I B0.2 B0.3 BQ,Q B4.5 B0.6 B4.7 B0.8 B0.9 B0.10 B0.11 B0.12 B0.13 B0.10 B4.15 B4.18 B1.19 TABLEOF CONTENTS (Continued)

Title Bases for Limiting Conditions for Operation,

= Instrumentation Bases for Limiting Conditions for Operation, Chemical and Volume Control System Bases for Limiting Conditions for Operation, Electrical Systems Bases for Limiting Conditions for Operation, Steam and Power Conversion Systems Bases for Limiting Conditions for Operation, Radioactive Materials Release Bases for Limiting Conditions for Operation, Refueling Bases for Limiting Conditions for Operation, Miscellaneous Radioactive Material Sources Bases for Limiting Conditions for Operation, Cask Handling Bases for Limiting Conditions for Operation, Snubbers Bases for Fire Protection System Bases for Limiting Conditions of Operation, Overpressure Mitigating System Bases for Operational Safety Review Bases for Reactor Coolant System In-Service Inspection Bases for Reactor Coolant System Integrity Bases for Containment Tests Bases for Safety Injection Tests Bases for Emergency Containment Cooling System Tests Bases for Emergency Containment Filtering and Post Accident Containment Venting Systems Tests Bases for Emergency Power System Periodic Tests Bases for Main Steam Isolation Valve Tests Bases for Auxiliary Feedwater System Tests Bases for Reactivity Anomalies Bases for Environmental Radiation Survey Bases for Fire Protection Systems Bases for Snubbers Bases for Surveillance Requirements, Overpressure Mitigating System Bases for System Flow Path Verifications Bases for Reactor Coolant Vent System

~Pa e B3.5-1.

B3.6-1 B3.7-1 B3.8-1 B3.9-1 B3.10-1 B3.11-1 B3.12-1 B3.13-1 B3.10-1 B3.15-1 B0.1-1 B0.2-1 B0.3-1 B0.0-1 B0.5-1 B0.6-'1 B0.7-1

'0.8-1 B0.9-1 B0.10-1 B0.11-1 B0.12-1 B0.13-1 B0.10-1 B0.15-1 80.18-1 B0.19-1 iv AMENDMENTNOS.

and

1.20 K-AVERAGE DISINTEGRATIONENERGY E 'shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95% of the total noniodine activity in the coolant.

1-7 AMENDMENTNOS.

and

3.1 REACTOR COOLANT SYSTEM

~Ob'ective:

To specify those limiting conditions for operation of the Reactor Coolant System which must be met to assure safe reactor operation.

a.

Reactor Coolant Pum s

1.

A minimum of ONE pump shall be in operation when the reactor is in power operation, except during low power physics tests.

2.

A minimum of ONE pump, or ONE Residual Heat Removal

Pump, shall be in operation during reactor coolant boron concentration reduction.

3.

Reactor power shall not exceed 10% of rated power unless at least TWO reactor coolant pumps are in operation.

0.

Reactor power shall not exceed 15% of rated power with only two pumps in operation unless the overtemperature bT trip setpoint, Kl, for two loop operation, has been set at 0.88.

5.

A reactor coolant pump shall not be started when cold leg temperature is <275oF unless steam generator secondary water temperature is less than 50oF above the RCS temperature (including instrument error).

b.

Steam Generators 1.

A minimum of TWO steam generators shall be operable when the average coolant temperature is above 350 F.

3.1-1 Amendment Nos. 66 and 58

r

c.

Pressurizer Saf et Valves 1.

ONE valve shall be operable whenever the head is on the reactor vessel except during hydrostatic test.

2.

THREE valves shall be operable when the reactor coolant average temperature is above 350oF or the reactor is critical.

d.

Pressurizer The pressurizer shall be operable with a steam bubble, and with at least 125 KW of pressurizer heaters capable of being supplied by emergency

power, when the reactor coolant is heated above 350oF.

e.

Relief Valves

l. A power operated relief valve (PORV) and its associated block valve shall be operable when the reactor coolant is heated above 350oF.

2. If the average coolant temperature is greater than 350oF and the conditions of 3.1.1.e.l cannot be met because one or more PORV(s) is inoperable, within 1

hour either restore the PORV(s) to operable status or close the associated block valve(s) and remove power from the block valve(s); otherwise, be in a condition with Keff <0.99 within the next 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 .

3. If the average coolant temperature is greater than 350oF and the conditions of 3.1.1.e.l cannot be met because one or more block valve(s) is inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 a condition with Keff <0;99 within the next 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 .

3.l-la Amendment Nos. 66 and 58

f.

Reactor Coolant S stem Vents

1. At least one reactor coolant system vent path consisting of at least two valves in series powered from emergency busses shall be OPERABLE and closed at each of the following locations when Tavg is greater than 350oF:

a.

Reactor Vessel Head b.

Pressurizer steam space 2.

With one of the above reactor coolant system vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the'noperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT SHUTDOWN 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 .

3. At ower 0 eration.

With both reactor coolant system vent paths inoperable, maintain the inoperable vent paths closed with power removed from the valve actuators of all the 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 SHUTDOWN 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 .

0. At Subcritical Conditions and Tav

>350oF.

With bothreactor coolant system vent paths inoperable, maintain the inoperable vent paths closed with power removed from the valve actuators of all the 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 a condition with Tavg (350oF within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months an'd in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.1-1b Amendment Nos.

and

2.

PRESSURE-TEMPERATURE LIMITS The Reactor Coolant System (except for the pressurizer) pressure and temperature shall be limited during heatup, cooldown, criticality (except for lower power physics tests),

and inser vice leak and hydrostatic testing in accordance with the limit lines shown on Figures 3.1-1a through 3.1-ld (Unit 3) and 3.1-2a through 3.1-2d (Unit 0).

Allowable pressure-temperature combinations are BELOW AND TO THE RIGHT of the lines on the Figures.

Heatup and cooldown rate limits are:

a.

A maximum heatup rate of 100oF in any one hour.

b.

A maximum cooldown rate of 100oF in any one hour.

c.

A maximum temperature change of >5oF in any one hour during hydrostatic testing operation above system design pressure.

The pressurizer pressure and temperature shall be limited in accordance with the following:

d.

The pressurizer shall be limited to a maximum heatup rate of 100oF in any one hour, and a maximum cooldown rate of 200oF in any one hour.

e.

The pressurizer shall be limited to a maximum Reactor'oolant

, System spray water temperature differential of 320oF.

With any of the above limits exceeded, restore the temperature and/or pressure within the limits within 30 minutes, determine that the RCS or pressurizer remains acceptable for continued'operations or, if at power, be in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.1-2 Amendment Nos. 76 and 70

With reactor power less than 70 percent Rated Thermal Power, the moderator temperature coefficient+ shall not be more positive than

+5 x 10 5 bK/K/oF.

When this condition is not met, the reactor shall be made subcritical by an amount equal to or greater than the potential reactivity insertion due to depressurization and cooldown.

With reactor power greater than or equal to 70 percent Rated Thermal Power, the moderator temperature coefficient shall not be more positive than 0 hK/K/oF.

When this condition is not met, the reactor shall be made subcritical by an amount equal.to or greater than the potential reactivity insertion due to depressurization and cooldo wn.

+ These moderator temperature coefficient conditions do not apply to low power physics tests.

3.1-2a Amendment Nos. 76 and 70

3.

LEAKAGE a.

Any reactor coolant system leakage indiciation in excess of 1 gpm shall be the subject of an investigation and evaluation initiated within 0 hours0 days 0 hours 0 weeks 0 months of the indication (ex. water inventory changes, radiation level increases, visual or audible indication).

A leak shall be assumed to exist until it is determined that no unsafe condition exists and that the indicated leak cannot be substantiated.

Leakage of reactor coolant through reactor pump seals and system valves to connecting closed systems from which coolant can be returned to the reactor coolant system shall not be considered as leakage except that such losses shall not exceed 30 gpm.

.b. If a reactor coolant system leakage indication is proven real, and is no't evaluated as safe, or exceeds 10 gpm, reactor shutdown shall be initiated within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months of the initial indication, except as noted in Section 3.1.3.g.

c. If reactor coolant leakage exists through a fault in the system boundary that cannot be isolated (ex. vessels, piping, valve bodies)

" the reactor shall be shutdown, and cool down to cold shutdown shall be initiated within 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

d.

The safety evaluation shall consider the source and magnitude of the leak, rates of change of detection variables, and if shutdown is required, this evaluation shall be used to determine shutdown rates and conditions.

A written log of the action taken shall be made as soon as practicable.

The evaluation shall assure that no potential gross leak is developing and that potential release of activity willbe within the guidelines of 10 CFR 20.

3.1-3 Amendment Nos. 81 and 75

e.

After shutdown, corrective action shall be taken before operation is resumed.

f.

Above 2% of rated power, two leak detection systems of different principles shall be operable one of which is sensitive to radioactivity.

The latter may be out of service for 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided two other systems are operable.

g.

Reactor Coolant System leakage shall be limited to 1 gpm total primary-to-secondary leakage through all steam generators not isolated from the Reactor Coolant System and 500 gallons per day through any one steam generator not isolated from the Reactor Coolant System.

0.

MAXIMUMREACTOR COOLANT ACTIVITY The specific activity of the primary coolant shall be limited to:

a.

Less than or equal to 1.0 microcurie per gram DOSE EQUIVALENTI-131, and b.

Less than or equal to 100/E microcuries per gram.

'Vith the above limits being exceeded, the following actions shall be taken:

1.

When the reactor is critical or average reactor coolant temperature is reater than 500oF:

a.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENTI-131 but within the allowable limit (below and to the left of the line) shown on Figure 3.1-1, operation may continue for up to 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provided that the cumulative operating time under these circumstances does not exceed 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months in any consecutive 12 month period.

With the total cumulative operating time at a primary coolant specific activity greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 exceeding 500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months in 3.1-0 Amendment Nos. 95 and 89

any consecutive 6 month period, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.3 within 30 days indicating the number of hours above this limit.

b.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during one continuous time interval or exceeding the limit line shown on Figure 3.1-1, be in a SHUTDOWN condition with average reactor coolant temperature less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With the specific activity of the primary coolant greater than 100/ E microcuries per gram, be in a SHUTDOWN condition with average reactor coolant temperature less than 500 F

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2.

For all modes of o eration a.

With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENTI-131 or greater than 100/ E microcuries per gram, perform the sampling and analysis requirements of item l.h.l of Table 0.1-2 until the specific activity of the primary coolant is restored to within its limits. A REPORTABLE OCCURRENCE shall be prepared and submitted to the Commission pursuant to Specification 6.9.2.b.

This report shall contain the results of the specific activity analyses together with the following information:

1.

Reactor power history starting 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the first sample in which the limit was exceeded, 2.

Fuel burnup by core region, 3.

Clean-up flow history starting 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 'rior to the first sample in which the limit was exceeded, 3.1-5 Amendment Nos. 95 and 89

0.

History of de-gassing operations, if any, starting 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the primary coolant exceeded 1.0 microcurie per gram DOSE EQUIVALENTI-l31.

3.1-5a Amendment Nos. 95 and 89

5.

REACTOR COOLANT CHEMISTRY a.

The following are r eactor coolant chemistry concentration maximum limits in ppm when coolant is above 250oF.

Oxygen Chloride Fluoride Normal Limit 0.10 0.15 0.15 Transient Limit 1.0 1.5 1.5

b. Corrective action shall be initiated if a normal limit is exceeded or if it is anticipated from trends that normal limits may.be exceeded.

c.

Cold shutdown shall be initiated if the transient limits are

reached, or if the corrective action required in (b) above is ineffective in reducing transient concentrations within 2l hours.

If the maximum concentration of any of the elements listed did not exceed the listed transient value, operation may be resumed after corrective action has been taken.

Otherwise, a safety review shall be made prior to startup.

d.

)Vhen reactor coolant is 250oF or below the limits in (a.) above will be maintained, except coolant oxygen content will reach saturation conditions during refueling operations.

If these limits are

exceeded, the unit will be brought to cold, shutdown and corrective action taken.

e.

Reactor coolant pump operation shall be permitted to ensure mixing during the corrective action phases specified above and shall be permitted at temperatures 50 F above the normal cold shutdown limit when bringing the reactor to cold shutdown or after reaching cold shutdown.

3.1-6

6.

DNB PARAMETERS The following DNB related parameter limits 'shall be maintained during power operation.

a.

Reactor Coolant System Tavg <78.2 F b.

Pressurizer Pressure

>2220 psia+

c.

Reactor Coolant Flow >268,500 gpm With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce thermal power to less than 5% of rated thermal power using normal shutdown procedures.

Compliance with a. and b. is demonstrated by verifying that each of the parameters is within its limits at least once each 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Compliance with c. is demonstrated by verifying that the parameter is within its limits after each refueling cycle.

+ Limit not applicable during either a THERMAL POWER ramp increase in excess of (5%)

RATED THERMALPOWER per minute or a THERMAL POWER step increase in excess of (10%) RATED THERMALPOWER.

This amendment effective as of date of issuance. for Unit 3 and date of Start-up, Cycle 10, for Unit 0.

3.1-7 Amendment Nos. 99 and 93

3.5 INSTRUMENTATION instrumentation systems.

~Ob'ective:

To delineate the conditions of the iristrumentation and safety circuits necessary to ensure reactor safety.

S ecification:

1.

Tables 3.5-1 through 3.5-5 state the minimum instrumentation operation conditions.

Specification 3.0.1 applies to Tables 3.5-1 through 3.5-3.

3.5-1 Amendment Nos.

and

TABLE3.5-5 ACCIDENT MONITORINGINSTRUMENTATION INSTRUMENTATION TOTAL NO.

OF CHANNELS MINIMUMCHANNELS OPERABLE APPLICABLE ACTIONS 1.

Pressurizer Water Level 2.

3.

5.

Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator (Primary Detector)

PORV Block Valve Position Indicator

= 6.

Safety Valve Position Indicator (Primary Detector) 7.

Containment Pressure (Wide Range) 8.

Containment Pressure (Narrow Range) 9.

Containment Water Level (Wide Range) 10.

Containment Water Level (Narrow Range) 11.

Containment High Range Area Radiation 12.

Containment Hydrogen Monitors 13.

High Range - Noble Gas Effluent Monitors a.

Plant Vent Exhaust

b. Unit 3 - Spent Fuel Pit Exhaust c.

Condenser Air Ejectors-d.

Main Steam Lines 10.

Incore Thermocouples (Core Exit Thermocouples) 15.

Reactor Vessel Level Monitoring System 2 per generator 2

1/valve 1/valve 1/valve 2

-2 2

0/core quadrant 2

1 per generator 1

1/valve 1/valve 1/valve 2/core quadrant 1

1,2 1,2 1,2 1,2 3

1,2 3

6,7 5 ~

5 1,2 2

(Note 1)

NOTES:

1.

Compliance willdepend on instrumentation operability.

Am~nriment Nns.

and

TABLE3.5-5 (Continued)

ACTION STATEMENTS ACTION 1 With the number of OPERABLE accident monitoring instrumentation channel(s) less than the Total Number of Channels shown in Table 3.5-5. either restore the inoperable channel(s) to OPERABLE status within 7 days, or be in a condition with Keff <0.99,

% thermal power excluding decay heat equal to

zero, and an average coolant temperature Tavg <350oF within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 With the number of OPERABLE accident monitoring instrumentation channels less than the minimum channels OPERABLE requirements of Table 3.5-5, either restore the inoperable channel(s) to OPERABLE status within 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in a condition with Keff <0.99, % thermal power excluding decay heat equal to zero, and an average coolant temperature Tavg <350oF within.the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 Operation may continue up to. 30 days with less than minimum channels OPERABLE for narrow range instruments.

ACTION 0 Or close the associated block valve and open its circuit breaker.

ACTION 5 With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned alternate method of monitoring the appropriate parameter(s), within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and:

1) eithe restore the inoperable channel(s) to OPERABLE status within 7 days of the event, or ACTION 6 2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.3 within 30 days following the event outlining the action

taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

With one hydrogen monitor inoperable, restore the inoperable monitor to OPERABLF status within 30 days or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 7 With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Amendment Nos.

and

3.8 STEAM AND POWER CONVERSION SYSTEMS

,1 N:

)>>

g systems.

~Ob'ective:

To define conditions of the steam-relieving capacity and auxiliary feedwater system.

following conditions must be met:

a.

TWELVE (12) of its steam generator safety valves shall be operable (except for testing).

b.

Its condensate storage tank shall contain a minimum of 185,000 gallons of water.

c.

Its main steam stop valves shall be operable and capable of closing in 5 seconds or less.

d.

System piping, interlocks and valves directly associated with the related components in TS 3.8.1 a, b, c shall be operable.

2.

The iodine-131 activity on the secondary side of a steam generator shall not exceed 0.67 pCI/gm.

3.

With the reactor coolant system above 350oF, if any of above specifications cannot be met within 08 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , the reactor shall be shutdown and the reactor coolant temperature reduced below 350oF.

Specification 3.0.1 applies.

0.

The following number of independent steam generator auxiliary feedwater trains and their associated flow paths (steam and water) shall be operable when the reactor coolant is heated above 350oF:

3.8-1 Amendment Nos.

and

a.

Sin le Nuclear Unit 0 eration b.

Two independent auxiliary feedwater trains capable of being powered from an operable stean supply.

Dual Nuclear Unit 0 eration Two independent auxili ary feedwater. trains and a third pump capable of being powered from, and supplying water to either train.

C ~

If in accordance with TS 4.10.1, testing is required during start-up of either unit, TS 3.8.4.a or b.,

as applicable, shall apply for an auxili ary feedwater

pump, pumps, or associated flow paths (stean and water) found to be inoperable.

5.

During power operation, if any of the conditions of 3.8.4 cannot be

= met, the reactor shall be shutdown and the reactor coolant temperature reduced below 350'F, unless one of the following conditions can be met:

a ~

For single unit operation with one of the two required independent auxili ary feedwater trains inoperable, restore the inoperable train to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or the reactor shall be shutdown and the reactor coolant temperature reduced below 350'F within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

For dual unit operation, one auxili ary feedwater pump and its associated piping, valves, and interlocks m@

be inoprable provided two independent auxili ary feedwater trains raaain oprable 'for time period not to exceed 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

If the inoperable pump cannot be made operable within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , one reactor shall be shutdown and its reactor coolant temperature reduced below 350'F within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c ~

d.

For dual unit operation, with one independent auxili ary feedwater train inoper able in one reactor, the affected reactor shall be SHUTDOWN and its reactor coolant temperature reduced below 350'F within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

TS 3.8.5.a applies for the single uni t still i n oper ation.

For dual unit operation, with one independent auxili ary feedwater train inoperable in both units, one reactor shall be SHUTDOWN and its reactor coolant temperature reduced below 350'F within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

TS 3.8.5.a applies for the single unit still in operation.

TABLE0.1-1 SHEET 2 Channel Descri tion Check Calibrate Test Remarks 10.

Rod Position Bank Counters 11.

Steam Generator Level 12.

Charging Flow 13.

Residual Heat Removal Pump Flow 10.

Boric Acid Tank Level 15.

Refueling Water Storage Tank Level 16.

Volume Control Tank Level 17A. Containment Pressure - Narrow Range 17B. Containment Pressure - Wide Range 13A. Process Radiation+++

N.A.

N.A; With analog Rod Position 13B. Area Radiation 19.

Boric Acid Control N.A.

N.A.

20.

Containment Sump Level 21.

Accumulator Level and Pressure 22.

Steam Line Pressure N.A.

N.A.

Amendment Nos.

and

TABLE0.1-1 SHEET 3 Channel Descri tion Check Calibrate Test Remarks 23.

Logic Channels 20.

Emer. Portable Survey Instruments 25.

Seismograph N.A.

N.A.

Make trace.

Test battery (change semi-annually) 26.

Auxiliary Feedwater Flow Rate 27.

RCS Subcooling Margin Monitor N.A.

N.A.

28.

29.

PORV Position Indicator (Primary Detector)

PORV Block Valve Position Indicator N.A.

N.A.

Check consists of monitoring indicated position and verifying 30.

Safety Valve Position Indicator N.A.

by observation of related parameters.

31.

Loss of Voltage (both Okv bussess) 32.

Trip of both Main Feedwater Pump Breakers N.A.

N.A.

For AFW actuation at power only For AFW actuation at power only 33.

Containment Water Level (Narrow Range)

N.A.

30.

Containment Water Level. (Wide Range)

N.A.

Amendment Nos.

and

TABLE0.1-1 SHEET 0 Channel Descri tion Check Calibrate Test Remarks 35.

Containment High Range Area Radiation 36.

Containment Hydrogen Monitors 37.

High Range Noble Gas Effluent Monitors a.

Plant Vent Exhaust

b. Unit 3 Spent Fuel Pit Exhaust c.

Condenser Air Ejectors d.

Main Steam Lines Sg R(Note 1)

R M

(1) Channel cahbration us>ng sample gas containing:

a.

One volume percent hydrogen, balance nitrogen.

b.

Four volume percent hydrogen, balance nitrogen.

38.

Incore Thermocouples (Core Exit Thermocouples)

N.A;(See Note 2) 39.

Reactor Vessel Level Monitoring System N.A.(See Note 2)

Amendment Nos.

and

TABLE0.1-1 SHEET 5

+ - Using moveable in-core detector system.

++ - Frequency only

+++ - Applies to containment particulate (Rl 1) and gaseous (R12) monitors only. For effluent monitors, refer to Tables 0.1-3 and 0.1-0.

PR - Prior to each release S - Each Shift W-0/M-B/W-M-

N.A.-

Daily Weekly At least 0 per month at intervals of no greater than 9 days and a minimum of 08 per year Every Two Weeks Monthly Quarterly Prior to each startup if not done previous week Each Refueling Shutdown Annually Not applicable

- N.A. during cold or refueling shutdowns.

The specified tests, however, shall be performed within one surveillance interval prior to startup.

tt

- N.A. during cold or refueling shutdowns.

The specified tests, however, shall be performed within one surveillance interval prior to heatup above 200F.

NOTES:

1) Acceptable criteria for calibration is provided in Table II.F.1-3 of NUREG 0737.

2) Compliance willdepend on instrumentation operability.

Am~ndment Nos.

and

J

2.

POST ACCIDENT CONTAINMENTVENT SYSTEM Operating tests shall be performed during refueling but not longer than 18 months.

The tests shall consist of visual in'spection of the

system, operation of all valves and pressure drop and air flow measurements.

Visual inspection shall include a search for any foreign materials and gasket deterioration in the HEPA filters and charcoal adsorbers.

Less than 6" of water pressure drop at 55 cfm flow shall constitute acceptable performance.

2.

Performance Tests a.

A visual inspection of the sytem shall be made before each DOP test and halogenated hydrocarbon leak test.

At least once per 18 months or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, in-place DOP and halogenated hydrocarbon tests at design flow (55 cfm + 10%) and carbon analysis, or carbon replacement, for the Post Accident Containment Vent filters shall be performed.

In

addition, carbon analysis (or carbon replacement),

DOP and halogenated hydrocarbon tests at design flow (55 cfm + 10%) shall be performed after (1) any structural maintenance on system housings which might have affected filter bank efficiency, (2) after complete or partial replacement of a filter bank or (3) after exposure of the filters to effluents from painting, fire or chemical release.

Removal of

> 99%

DOP and

> 99%

halogenated hydrocarbon shall constitute acceptable performance.

b. Laboratory carbon sample analysis shall show >90% methyl radio-iodine removal or the charcoal shall be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52 (Revision 2).

The sample shall be taken in accordance with position C.6.b of Regulatory Guide 0.7-3 Amendment Nos. 83 and 77

1.52.

Carbon analysis shall be performed in accordance with ANSI N510-1975.

Analysis shall verify the above removal efficiency for radio-iodine within 05 days after removal of the sample.

3.

CONTROL ROOM VENTILATION (EMERGENCY INTERNAL CLEANUP) SYSTEM 1.

A visual inspection shall be made before each in-place DOP test, halogenated hydrocarbon leak test, and airflow distribution test.

The Control Room Ventilation System shall be operated monthly for at least 15 minutes to demonstrate operability.

Auto initiation of the systems operations shall be checked during refueling, but not longer than 18 months.

Pressure drop measurements across the filter bank shall be made annually.

Less than 6" of water pressure drop at designed flow (1,000 cfm + 10%)

across the combined HEPA filter and charcoal adsorbers shall constitute acceptable performance.

A visual inspectio'n shall include a

search for any foreign materials and gasket deterioration in the HEPA filters and charcoal adsorbers.

2.

Performance Tests a.

A visual inspection shall be made before each in-place DOP

test, halogenated hydrocarbon leak test and airflow distribution test.

At least once per 18 months or after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation, in-place DOP and halogenated hydrocarbon tests at design flow (1,000 cfm + 10%) and carbon analysis shall be performed after (1) any structural maintenance on system

housings, which might have affected filter bank efficiency, (2) after complete or partial replacement of a filter bank, or (3) after operational exposure 0.7-0

of the filters to effluents from painting, fire or chemical releases.

Removal of >99%

DOP and

>99%

halogenated hydrocarbon shall constitute acceptable performance.

b.'

charcoal surveillance specimen from one of the charcoal adsorbers shall be removed and analyzed for methyl radio-iodine removal capability.

The results of the laboratory carbon sample analysis shall show >90% methyl radio-iodine removal efficiency.

Samples shall be taken in accordance with position C.6.b of Regulatory Guide 1.52.

Carbon analysis shall be performed in accordance with ANSI N510-1975.

Analysis shall verify the above'removal efficiency for methyl radio-iodine within 45 days after removal of the, sample.

Failing this, the charcoal shall be replaced with charcoal which meets or 'exceeds the criteria of position C.6.a of Regulatory Guide 1.52 (Revision 2).

c.

System flow rate should be verified once each 18 months, following maintenance to HEPA or charcoal housings or fire or chemical release in its ventilation zone while th'e system is operating.

0.7-5 Amendment Nos. 83 and 77

4.19 REACTOR COOLANT VENT SYSTEN

~A>>1i 111:

Appl1 <<1 1

1 1

111 1

vent system.

~0b'ective:,

To verify the operability of the system.

1 1 <<<<1 demonstrated OPERABLE at least once per 18 months by:

a.

Verifying all manual isolation valves in each vent path ar e locked in the open position.

b.

Cycling each valve in the vent path through at least one complete cycle of full travel from the control.

room during COLD SHUTDOWN or REFUELING.

c.

Verifying flow through the reactor coolant system vent paths during COLD SHUTDOWN or REFUELINGe

~

A

63.1 BASES FOR LIMITINGCONDITIONS FOR OPERATION REACTOR COOLANTSYSTEM 1.

0 erational Com onents The specification requires that a sufficient number of reactor coolant pumps be operating to provide coastdown core cooling in the event that a loss of flow occurs.

The flow provided will keep DNBR well above the applicable design limit". When the boron concentration of the Reactor. Coolant System is to be reduced, the process must be uniform to prevent sudden reactivity changes in the reactor.

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

The residual heat removal pump willcirculate the reactor coolant system volume in approximately one half hour.

Each of the pressurizer safety valves is designed to relieve 283,300 lbs. per hour of saturated steam at the valve set point.

Below 350oF and 050 psig in the Reactor Coolant System, the Residual Heat Removal System can remove decay heat and thereby control system temperature and pressure.

If no residual heat were removed by any of the means available, the amount of steam which could be generated at safety valve liftingpressure would be less than the capacity of a single valve.

Also, two safety valves have capacity greater than the maximum surge rate resulting from complex loss of load. ~2)

The 50oF limit on maximum differential between steam generator secondary water temperature and reactor coolant temperature assures that the pressure transient caused by starting a reactor coolant pump when cold leg temperature is

<275oF can be relieved by operation of one Power Operated Relief Valve (PORV).

The 50oF limitincludes instrument error.

The plant is designed to operate with all reactor coolant loops in operation, and maintain DNBR above the applicable design limit+ during all normal operations and anticipated transients.

In power operation with one reactor coolant loop not in operation, this specification requires that the plant be in at least Hot Shutdown within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

+ This amendment effective as of date of issuance for Unit 3 and date of Start-up, Cycle 10, for Unit 0.

In Hot Shutdown, a single reactor coolant loop provides sufficient heat removal capability for removing decay

heat, however, single failure considerations require that two loops be operable.

In Cold Shutdown, a single reactor coolant loop or RHR coolant loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be operable.

Thus, if the reactor coolant loops are not operable, this specification requires two RHR loops to be operable.

The oper ation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.

The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The requirement that at least one residual heat removal (RHR) loop be in operation during Refueling Shutdown ensures that (1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 160 F as required during Refueling Shutdown and (2) sufficient coolant circulation is maintained through the reactor core to minimize the effect of a boron dilution stratification.

The requirement to have two RHR loops operable when there is less than 23 feet of water above the core ensures that a single failure of the operating RHR loop will not result in a complete loss of residual heat removal capability.

With the reactor vessel head removed and 23 feet of water above the core, a large heat sink is available for core cooling.

Thus, in the event of a failure of the operating RHR loop, adequate time is provided to initiate emergency procedures to cool the core.

Reactor Coolant S stem Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling.

The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

The valve redundancy of the reactor coolant system vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring

.that a single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

The function and capabilities of the reactor coolant vent system are consistent with the requirements of Item ILB.l of NUREG-0737, "Clarification of TMI Action Plan Requirements",

November 1980.

B3. I-la Amendment Nos.

and

B0.19 BASES FOR REACTOR COOLANT VENT SYSTEM The surveillance requirements of the reactor coolant vent system are consistent with the requirements of Item II.B.l of NUREG-0737, ",Clarification of TMI Action Plan Requirements", November 1980.

The performance of the specified surveillances will verify the operability of the system.

B0.19-1 Amendment Nos.

and

2.

Pressure/Tem erature Limits All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes.

These cyclic loads are introduced by normal load transients, reactor trips and startup and shutdown operations.

The various categories of load.cycles used for design purposes are provided in Section 4.1.5 of the FSAR.

During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for prevention of brittle fracture.

During heatup, the thermal gradients in the reactor vessel wall produce thermal stresses which vary from compressive at the inner wall to tensile at the outer wall.

These thermal induced compressive stresses tend to alleviate the ten'sile stresses induced by the internal pressure.

Therefore, a pressure-temperature

~ curve based on steady state conditions (i.e., no thermal stresses) represents a

lower bound of all similar curves for finite heatup rates when the inner wall of the vessel is treated as the governing location.

The heatup analysis also covers the determination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The thermal gradients established during heatup produce tensile streses at the outer wall of the vessel.

These stresses are additive to the pressure induced tensile stresses which are already present.

The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be

'defined.

Subsequently, for the cases in which the outer wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

The heatup limit curves are composite curves prepared by determining the most conservative

case, with either the inside or outside wall controlling, for any heatup rate up to 100 F per hour.

The cooldown limit curves are composite curves whi"h were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference, temperature at the end of the service period.

83. -2 Amendment Nos. 95 and 89

4

POST ACCIDENTSAMPLING A program 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.

Provisions for maintenance of sampling and analysis equipment.

SYSTEMS INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

This program shall include the following:

1.

Provisions establishing preventative maintenance and periodic visual

, inspection requirements, and 2.

Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

IODINE MONITORING The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas 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.

I BACKUP METHODS FOR DETERMININGSUBCOOLING MARGIN

'The licensee shall implement a program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin.

This program shall include the following:

1.

Training of personnel, and 2.

Procedures for monitoring.

ML17346A396

Text

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