Text

Proposed Tech Specs,Taking Current Requirement for Safety Injection Accumulators & Consolidating Requirements Into One Limiting Condition for Operation Section & One Surveillance Section
	ML17342B122
Person / Time
Site:	Turkey Point
Issue date:	02/03/1988
From:	; FLORIDA POWER & LIGHT CO.
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ML17342B121	List: ML17342B120; ML17342B122;
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	NUDOCS 8802090530
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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO. FACILITY OPERATING LICENSE NO. DPR-41 DOCKET NO. 50-250 AND 50-251 Revise Appendix A as follows:

1-7 1-7 1-8 1-8 1-9 1-9 Table 1.1 Table 1.1 3.4 -1 3.4 -1

3. 4-2 3. 4-2

3. 4-2a 3.4-2a 3.4-2b Table 4.1-1, Sheet 2 Table 4.1-1, Sheet 2 Table 4.1-2, Sheet 2 Table 4.1-2, Sheet 2 4.5-2 4. 5-2

4. 5-3

83. 4-1 83. 4-1

83. 4-2 83. 4-2

84. 5-1 84.5-1; Sa020905a0

'.PDR aSoaoa ADO'5000250 PDR

TABLE OP CONTENTS Section Title ~Pa e TECHNICAL SPECIPICATIONS 1.0 DEPINITIONS 1.1 Safet Limits 1.2 Limiting Safety System Settings 1.3 Limiting Conditions for Operation 1-1 1.4 Operable 1-1 1.5 Containment Integrity 1-2 1.6 Protective Instrumentation Logic 1-2 1.7 Instrumentation Surveillance 1"3 1.8 Reportable Event 1-3 1.9 Action 1-4 1.10 Core Alteration 1-4 1;11 Rated Power 1-4 1.12 Thermal Power 1-4 1.13 Design Power 1-4 1.14 Dose Equivalent I-131 1-5 1.15 Power Tilt 1-5 1.16 Interim Limits 1-6

. 1.17 Low Power Physics Tests 1-6 1.18 Engineered Safety Peatures 1-6 1.19 ..Reactor Protection System ,1-6 1.20 Safety Related. Systems and,Components 1-6 1.21 Per Annum 1-6 1.22 Reactor Coolant System Pressure Boundary Integrity 1-6 1.23 Coolant Loop 1-7 1.24 E-Average Disintegration Energy 1-7 1.25 Gas Decay Tank System 1-7 1.26 Ventilation Exhaust Treatment System 1-7 1.27 Process Control Program (PCP) 1-7 1.28 Offsite Dose Calculation Manual (ODCM) 1-7 1.29 Dose Equivalent I-131 1-8 1.30 Purge-Purging 1-8 1.31 Venting 1-8 1.32 Site Boundary 1-8 1.33 Unrestricted Area 1-8 1.34 Member(s) of the Public 1-8 1.35 Heavy Loads 1-9 1.36 Operational Modes 1-9 1.37 Staggered Test Basis 1-9 1.38 Analog Channel Operational Test 1-9 2.0 SAPETY LIMITS'ANDLIMITINGSAPETY SYSTEM SETTI NGS 2.1-1 2.1 Safety Limit, Reactor Core 2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure 2.2-1 2.3 Limiting Safety System Setting, Protective Instrumentation 2.3-1 3.0 LIMITINGCONDITIONS POR OPERATION 3.0-1 3.1 Reactor Coolant System 3.1-1 Operational Components 3.1-1 Pressure-Temperature Limits 301-2 Leakage 301-3 Maximum Reactor Coolant Activity 3.1-4 Reactor Coolant Chemistry 3.1-6 DNB Parameters 3.1-7 Amendment Nos. and

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1.24 I?-AVERAGE DISINTEGRATION ENERGY B 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~halfilives.greater:than 40-.minutes;.making<<upiat~least<<9596..of, the; ..~~<<:. ~

total noniodine activity in the coolant.

1.25 GAS DECAY TANK SYSTEM The GAS DECAY TANK SYSTEM is designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

1.26 VI? NTILATIONEXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TRI?ATMBNT SYSTEM is any system designed and installed to reduce radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through HEPA filters for the purpose of removing particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety. Feature.(ESF).atmospheric cleanup systems are not

,,-;considered to,be. ventilation exhaust-.treatment system components.

1.27 PROCESS CONTROL PROGRAM PCP The PROCESS CONTROL PROGRAM shall contain the provisions, based on full scale testing, to asure that dewatering of spent bead resins results in a waste form with the properties that meet the requirements of 10CFR61 (as implemented by 10CFR20) and of the low level radioactive waste disposal site at the time of disposal.

1.28 OPFSITB DOSE CALCULATIONMANUAL ODCM The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents and in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints.

Amendment Nos. and

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DOSE E UIVALENTI-131 The DOSE EQUIVALENT I-131 shall be that concentration of 1-131 (microcurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, 1-132, I-133, I-134. and I-135.actually present. The

="thyroid'"dose conversion"factors"used"-for'his-calculation=shall-be those. listed in Table III of TID-14844, "Calculation of Distance Pactors for Power and Test Reactor Sites", or in NRC Regulatory Guide 1.109, Rev. 1, October, 1977.

PURGE - PURGING PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is required to purify the confinement.

VENTING VENTING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required

'during VENTING. Vent, used in system names, does not imply a VENTING process.

'ITE BOUNDARY The SITE, BOUNDARY shall be that line beyond which the land is neither owned, leased nor otherwise controlled by the licensee.

UNRI?STRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation from radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional and/or recreational purposes.

MI?MBER S OP THE PUBLIC MEMBER(S) OP THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, vendors or members of the Armed Porces using property located within the SITE BOUNDARY. Also excluded from this category are persons

.who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

1-8 Amendment Nos. and

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HEAVY LOADS'ny load in excess of the nominal weight of a fuel and control rod assembly and associated handling tool. Por the purpose of this specification, HEAVY LOADS will be defined as loads in excess of 2000 pounds.

OPERATIONAL MODE - MODE An OPERATIONAL MODE (i.e., MODE) shall correspond to any one inclusive combination of core reactivity condition, power level, and average reactor coolant temperature specified in Table 1.1.

STAGGI?RED TEST BASIS A STAGGERED TEST BASIS shall consist of:

a. A test schedule for (n) systems, subsystems, trains, or other designated components obtained by dividing the specified test interval into (n) equal subintervals, and

b. The testing of one system, subsystem, train, or other designated component at

'the beginning of each subinterval.

ANALOG CHANNEL OPERATIONAL TI?ST An ANALOG CHANNEL OPERATIONAL TEST shall. be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, interlock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

1-9 Amendment Nos. and

TAHLF. 1.1 OPERATIONAL MODL<'S ***

Reactivity % Rated Average Coolant Power Operation ) 5% &350'F

2. Start-up a 0.99 s5% % 350'F

3. Hot Standby < 0.99 > 350'F

4. Hot Shutdown < 0.99 350'F ) Tavg )200'F

5. Cold Shutdown < 0.99 %200'F

6. Refueling ** < 0.90 < 140'F Excluding decay heaL F<uel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

This table shall only be applicable to those specifications that have been modified to reflect Operational Modes in the Applicability section of the LCOs, except as specified in Section 3.0.1 (Note).

Amendment Nos. and

3.4 ENGINEERED SAPETY PEATURES A licabilit: Applies to the operating status of the Engineered Safety Peatures.

~l ective:,

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.<<~; 3'oAefine. those~limiting.conditions.for'operation.that>>are<necessary:

~Ob .(1). to, remove decay heat from the core in emergency or normal shutdown situations, (2) to remove heat from containment in normal operating and emergency situations, and (3) to remove airborne iodine from the containment atmosphere in the event of a Maximum Hypothetical Accident.

S ecification: 1. SAPETY INJECTION AND RESIDUAL HEAT REMOVAL SYSTEMS

a. The reactor shall not be made critical, except for low power physics tests, unless the following conditions are met:

1. The refueling water tank shall contain not less than

.. 320,000.,gal..of.waterwith a boron. concentration of at least 1950-ppm.

20 The boron injection tank shall contain not less than 900 gal. of a 20,000 to 22,500 ppm boron solution. The solution in the tank, and in isolated portions of the inlet and outlet piping, shall be maintained at a temperature of at least 145P. TWO channels of heat tracing shall be operable for the flow path.*

3. POUR safety injection pumps shall be operable.

See reference (11) on Page B.3.4-2 3.4-1 Amendment Nos. and

A

4. TWO residual heat removal pumps shall be operable. l

5. TWO residual heat exchangers shall be operable.

I

6. All valves, interlocks and piping associated with the above components and required for post accident operation, shall be operable except valves that are positioned and locked. Valves 862-A and B; 863-A and B; 864-A and 864-B and 866-A and B shall have power removed from their motor operators by locking open the circuit breakers at the Motor Control Centers. The air supply to valve 758 shall be shut off to the valve operator.

b. During power operation, the requirements of 3.4.1a may be modified to allow one of the following components to be inoperable (including associated valves and piping) at any one time except for the cases stated in 3.4.1.b.2. If the system is not restored to meet the requirements of 3.4.1a within the time period specified, the reactor shall be placed in the hot shutdown condition. If the requirements of 3.4.1a are not satisfied within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the reactor shall be placed, in the cold shutdown condition. Specification 3.0.1 applies to 3.4.1.b.

ONE of POUR safety injection pumps may be out of service for 30 days. A second safety injection pump may be out of service, provided the pump is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . TWO of the POUR safety injection pumps shall be tested to demonstrate operability before initiating maintenance of the inoperable pumps.

2. ONE channel of heat tracing on the flow path may be out of service for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .*

3. ONE residual heat removal pump may be out of service, (

provided the pump is restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In addition the other residual heat removal pump shall be tested to demonstrate operability prior to initiating maintenance of the inoperable pump.

See reference (ll) on Page B.3.4-2 3.4-2 Amendment Nos. and

4. ONE residual heat exchanger may be out of service for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

5. Any valve in the system may be inoperable provided repairs are completed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Prior to initiating maintenance, all valves that provide the duplicate function shall be tested to demonstrate operability.

6. To permit temporary operation of the valve, e.g., for surveillance of valve operability, for the purpose of valve maintenance, etc., the valves specified in 3.4.1.a.6 may be unlocked and may have supplied air or electric power restored for a period not to exceed 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

C. During power operation three Reactor Coolant Loops shall be in operation.

1. With less than three Reactor Coolant Loops in operation, the reactor must be in hot shutdown within one hour.

d. In.hot.shutdown. at least two Reactor Coolant Loops shall be operable and at least one Reactor Coolant Loop shall be in operation.*

With less than two Reactor Coolant Loops operable, restore the required Coolant Loops to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to reduce Tavg to less than or equal to 350P within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2. With no Reactor Coolant Loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Coolant Loop to operation.

e. With average coolant temperature less than 350P, at least two Coolant Loops shall be operable or immediate corrective action must be taken to return two Coolant Loops to operable as soon as possible. One of these Coolant Loops shall be in operation.*

With no Coolant Loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required Coolant Loop to operation.

All reactor coolant pumps and residual heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause dilution of the reactor coolant'system boron concentration, .and 2) core outlet temperature is maintained as last 10P below staturation temperature.

3.4-2.a Amendment Nos. and

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3.4.1.f ACCUMULATORS LIMITINGCONDITION POR OPERATION

, '3.4;1.f Each Reactor Coolant System (RCS) accumulator shall be OPERABLE with:

1. The isolation valve open and its circuit breaker locked open,

2. A contained borated water volume of 6545 to 6665'allons,

3. A boron concentration of 1950 to 2350 ppm, and

4. A nitrogen cover-pressure of 600 to 675 psig.

APPLICABILITY: MODES 1, 2 and 3*.

ACTION:

a. With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce presssurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Pressurizer pressure above 1000 psig 3.4-2.b Amendment Nos. and

TABLE 4.1-1 SHEET 2 Channel Descri tion Check Calibrate Test Remarks

10. Rod Position Bank Counters S4 N/A N/A With analog Rod Position Steam Generator Level S> R Mt

12. Charging Flow N/A R N/A

13. Residual Heat Removal Pump Flow N/A R N/A

14. Boric Acid Tank Level W R ¹/A

15. Refueling Water Storage Tank Level W1 R N./A

16. Volume Control Tank Level N/A R N/A 17A. Containment Pressure - Narrow Range M) j')

R N/A 17B. Containment Pressure - Wide Range ) R N/A 18A. Process Radiation**" D N/A M 18B. A'rea Radiation D A M

19. Boric Acid Control N/A N/A R

20. Containment Sump Level N/A R N/A

21. Deleted

22. Steam Line Pressure S> MI Amendment Nos. and

TABLE 4.1-2 (Sheet 2 of 3)

MINIMUMPRI? UENCIBS POR I? UIPMBNT AND SAMPLING TESTS Max. Time Check ~recruenc r Between Tests Co'ntrol Rods (cont'd) 'artial full mo'vement'"of length rods

"'Biw'eekly while critical (Days)

" ""'"20'"

6. Pressurizer Safety Valves Set Point Bach refueling N/A shutdown

7. Main Steam Safety Valves Set Point Bach refueling N/A shutdown

8. Containment Isolation Trip Punctioning Each refueling N/A shutdown

9. Refueling System Interlocks Punctioning Prior to each N/A refueling

10. Deleted 11.. -.Reactor Leakage Coolant. System ,Evaluate Daily N/A

12. Diesel Puel Supply Puel Inventory Weekly 10

13. Spent Puel Pit Boron Concentration Monthly 45

14. Pire Protection Pump and Operable Monthly Power Supply

15. Turbine Stop and Control Closure Monthly*

Valves, Reheater Stop and Intercept Valves

16. LP Turbine Rotor Inspector V, MT,PT Every 5 years 6 years (w/o rotor disassembly)

17. Spent Puel Cask Crane Punctioning Within 7 days 7 days when crane is being used to manuever spent fuel cask.

Amendment Nos. and

2. Pumps shall start and reach required head for normal and recirculation flow, whichever is applicable to the operating condition; the instruments and visual observations shall indicate

-.proper"functioning. Test operation shall"be for at-least 15 minutes.

b. Valves

1. The boron injection tank isolation valves receiving a Safety Injection signal shall be cycled monthly.tt

2. The containment recirculation sump suction valves shall be cycled monthly.t

3. The refueling water storage tank outlet valves shall be tested in performing the respective pump tests.

N/A during cold or refueling shutdowns. The specified tests, however, shall be performed within one surveillance interval prior to reactor startup.

N/A during cold or refueling shutdowns. The specified tests, however, shall be performed within one surveillance interval prior to heatup above 200 P.

See reference (11) on Page B.3.4-2.

4.5-2 Amendment Nos. and

4.5.3 Each accumulator shall be demonstrated OPERABLE:

- ~, ..i. a.",,~;:At,least once'-per shift"or"12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;".whichever is more limiting, by:

1) Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and

2) Verify that each accumulator isolation valve is open by control room indication (power maybe restored to the valve operator to perform this surveillance if the redundant indicator is inoperable).

b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of greater than or equal to 196 of tank volume by verifying the boron concentration of the accumulator solution; and Co At least once per.31 days. when the RCS pressure is above,1000 psig, by verifying that the power to the isolation-valve operator is disconnected by a locked open breaker.

d~ During each refueling shutdown, accumulator check valves shall be checked for oper ability.

4.5.4 Each accumulator water level and pressure channel shall be demonstrated OPERABLE:

a. "

At least once per 31 days by the performance of an ANALOG CHANNEL OPERATIONAL TEST, and

b. During each refueling by the performance of a CHANNEL CALIBRATION.

4.5-3 Amendment Nos. and

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BASES POR LIMITINGCONDITIONS POR OPERATION RNGINPRRRRD SAPPTY PI? ATURBS

1. Safet In ection and Residual Heat Removal S stems a.l The requirements for refueling water tank storage meet the safety analysis.( )

a+2 The boron injection tank contains sufficient solution to meet the steam line break accident analysis.( )( )(

a+3 Any two safety injection pumps meet the requirements of the MHA analysis and the steam line break accident an alys I s (2) (4) (5) a.4 a.5 A single residual heat removal pump and heat exchanger meets the MHA analysis requirements.( )( )

. b.1, See a.3.above.

b.2 b.3 See a.4 above.

, c.l ACCUMULATORS The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core provides the initial cooling

,mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.

Because the accumulator isolation valves (865A, B, and C)

, fail to meet. single failure criteria, removal of power to the

, valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened the full capability of one accumulator is not available and prompt action is required to place the reactor in a mode where this capability is not required.

B3.4-1 Amendment Nos. and

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B3.4 BASES FOR LIMITINGCONDITIONS FOR OPERATION RNGINRRRRED SAFETY FEATURES continued

2. Rmer enc Containment Coolin S stems Either two of the three emergency containment cooling units or one of the two spray pumps has the cooling capability requried to meet the MHA analysis.(6)(7)( )

3. Emer enc Containment Pilterin S stems Two of three filter units have capacity to meet the MHA analysis. (7)( )

4. Com onent Coolin S stem One pump and two heat exchangers meet the requirement of the MHA analysis.<

5. Intake Coolin Water S stem One pump meets the requirements of the MHA analysis.( )

References:

(1) PSAR 6.2.2 (2) PSAR 14.2.5 (3) PSAR 14.3.2 (4) PSAR 14.1.9 (5) PSAR 6.2.3 (6) 'SAR '14.3.4 (7) PSAR 6.3 (8) PSAR 14.3.5 (9) PSAR 6.4 (10) PSAR 9.3 (11) The requirement for use of the BIT tanks for Mitigation of the Main Steam Line Break accident has been removed following installation of the Model 44P Steam Generators. The required supporting analyses can be found in L (502), dated 11/30/81. The temperature requirement above 145 P is no longer applicable. Therefore, the heat tracing requirement is not necessary. There is no Boron Concentration Requirement in the BIT.

B3.4-2 Amendment Nos. and

BASES FOR SAFETY INJECTION TESTS The Safety Injection system is not'perated "during reactor operation and system test must be run with the reactor shut down. The tests will demonstrate proper operation of controls and components. The test intervals have been selected based on experience with similar equipment.

84.5-1 Amendment Nos. and

ML17342B122

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