ML17342A522

Proposed Tech Specs,Providing Individual Tech Specs for Auxiliary Feedwater Sys & Condensate Storage Tank & Correcting Errors in Valve Numbers in Table 3.16-1

ML17342A522
Person / Time
Site:	Turkey Point
Issue date:	05/07/1986
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17342A521	List: ML17342A520 ML17342A522
References
NUDOCS 8605130308
Download: ML17342A522 (29)

Text

TABLE OF CONTENTS (Continued)

Section Title ~Pa e 3.2 Control Rod and Power Distribution Limits 3.2-1 Control Rod Insertion Limits ,3.2-1 Misaligned Control Rod 3.2-2 Rod Drop Time 3.2-2 Inooerable Control Rods 3.2-2 Control Rod Position Indication 3.2-3 Power Distribution Limits 3.2-3 In-Core Instrumentation 3e2-7 Axial Offset Alarms 3.2-8 3.3 Containment 3.3-1 Engineering Safety Features 3.ti-l Safety Injection and RHR Systems 3.0-1 Emerpency Containment Cooling Systems 3.0-3 Emerpency Containment Filtering System 334 Component Coolinp System 3.0-0a Intake Cooling EVater System 3A-5 Post Accident Containment Vent System 3A-6 Control Room Ventilation 3A-6 3.5 Instrumentation 3.5-1 3.6 Chemical and Volume Control System 3.6-1 3.7 Electrical Systems 3.7-1 3.8 Steam and Power Conversion Systems 3.8-1 3.9 Radioactive Materials Release ~

3.9-1 Liquid EVastes 3.9-1 Gaseous Wastes 3.9-3 Containerized 4Vastes 3.9-5 3.10 Refueling 3.10-1 3.1 l Miscellaneous Radioactive Materials Sources 3.11-1 3.12 Cask Handling 3.12-1 3.13 Snubbers 3.13-1 3.10 Fire Protection Systems 3.10-1 3.15 Overpressure Mitigating System 3.15-1 3.16 Reactor Coolant System Pressure Isolation Valves 3.16-1 3.17 Spent Fuel Storage 3.17-1 3.18 Auxiliary Feedwater System 3.18-1 3.19 Condensate Storage Tanks 3.19-1 0.0 SURVEILLANCE REQUIREMENTS 0.1-1 0.1 Operational Safety Review 0.1-1

~o.2 Reactor Coolant System In Service Insoection 0.2-1 0.3 Reactor Coolant System htegrity 0.3-1 Containment Tests 0A-I Integrated Leakage Rate Test - Post Ooerational i>.0-1 Local Penetration Tests 0.0-2 Isolation Valves 0.0-3 Residual Heat Removal System 0.0-3 Tendon Surveillance OA-0 End Anchorage Concrete Surveillance OA-6 Liner Surveillance V-7 0.5 Safety Injection 0.5-1 0.6 Emerpency Containment Cooling Systems 0.6-1 0.7 . Emergency Containment Filtering, Post Accident Containment Vent Systems and Control Room Ventilation 0.7-1 0.8 Emergency Power System Periodic Tests '>.8-1 0.9 Slain Steam Isolation Valves 0.9-1 8605130308 Amendment Nos. and 05000250 860507'DR ADOCK P 'DR)

TABLE OF CONTENTS (Continued)

Section Title ~Pa e 0.10 Auxiliary Feedwater System 0.10-1 0.11 Reactivity Anomalies 0.11-1 0.12 Environmental Radiation Survey 0.12-1 0.13 Radioactive Materials Sources Surveillance 0.13-1 0.10 Snubbers 0.10-1 0.15 Fire Protection Systems 0.15-1 0.16 Overpressure Mitigating System 0.16-1 0.17 Reactor Coolant System Pressure Isolation Valves 0.17-1 0.18 Safety Related Systems Flowpath 0.18-1 0.19 Reactor Coolant Vent System 0.19-1 0.20 Reactor Materials Surveillance Program 0.20-1 0.21 Condensate Storage Tank 0.21-1 5.0 DESIGN FEATURES 5.1-1 5.1 Site 5.1-1 5.2 Reactor 5.2-1 5.3 Containment 5.3-1

'.0 Fuel Storage 5.0-1 6.0 ADMINISTRATIVECONTROLS 6-1 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Facility Staff Qualifications 6-5 6.0 Training 6-5 6.5 Review and Audit 6-6 6.6 Reportable Event Action 6-10 6.7 Safety Limit Violation 6-10 6.8 Procedures 6-10 6.9 Reporting Requirements 6-16 6.10 Record Retention 6-27 6.1 1 Radiation Protection Program 6-29 6.12 High Radiation Area 6-29 6.13 Post Accident Sampling 6-30 6.10 Systems Integrity 6-30 6.15 Iodine Monitoring 6-30 6.16 Back-up Methods for Determining Subcoolin g Margin 6-'30 6.17 Process Control Program (PCP) 6-31 6.18 Offsite Dose Calculation Manual (ODCM) 6-31 92.1 Bases for Safety Limit, Reactor Core B2.1-1 B2.2 Bases for Safety Limit, Reactor Coolant System Pressure B2.2-1 B2.3 Bases for Limiting Safety System Settings, Protective Instrumentation B2.3-1 B3.0 Bases for Limiting Conditions for Operation, Applicability B3.0-1 B3.1 Bases for Limiting Conditions for Operation, Reactor Coolant System B3.1-1 B3.2 Bases for Limiting Conditions for Operation, Control and Power Distribution Limits B3.2-1 B3.3 Bases for Limiting Conditions for Operation, Containment 83.3-1 Bases for Limiting Conditions for Operation, Engineered Safety Features B3.0-1 Amendment Nos. and

t TABLE OF CONTENTS (Continued)

~

Section Title ~Pa e 83.5 Bases for Limiting Conditions for Operation, Instrumentation 83.5-1 B3.6 Bases for Limiting Conditions for Operation, Chemical and Volume Control System 83.6-1 83.7 Bases for Limiting Conditions for Operation, Electr ical Systems 93.7-1 83e8 Bases for Limiting Conditions for Operation, Steam and Power Conversion Systems 83.8-1 83.9 Bases for Limiting Conditions for Operation, Radioactive Materials Release 83.9-1 83.10 Bases for Limiting Conditions for Operation, Re fueling 83.10-1 83.11 Bases for Limiting Conditions for Operation, Miscellaneous Radioactive Material Sources 83.11-1 83.12 Bases for Limiting Conditions for Operation, Cask Handling 83.1?-1 83.13 Bases for Limiting Conditions for Operation, Snubbers 83.13-1 93.10 Bases for Fire Protection System 83.10-1 83.15 Bases for Limiting Conditions of Operation, Overpressure Mitigating System 83.15-1 83.17 Bases for Limiting Conditions for Operation, 83.17-1 Spent Fuel Storage 83.18 Bases for Auxiliary Feedwater System 83.1$ -1 83.19 Bases for Condensate Storage Tanks 83.l.9-1 90.1 Bases for Operational Safety Review 80.1-1 90.2 Bases for Reactor Coolant System In-Service Inspection 80.2-1 90.3 Bases for Reactor Coolant System integrity 80.3-1 90.0 Bases for Containment. Tests BOA-1 91.5 Bases for Safety Injection Tests 94.5-1 91.6 Bases for Fmergency Containment Cooling System Tests 80.6-1 80.7 Bases for Emergency Containment Filtering and Post'Accident Containment Venting Systems Tests 90.7-1 80.8 Bases for Emergency Power Syste~ Periodic Tests 90.8-1 80.9 Bases for Main Steam Isolation Valve Tests 90.9-1 80.10 Bases for Auxiliary Feedwater System Tests 80.10-1 80.11 Bases for Reactivity Anomalies 84.11-1 80.12 Bases for Environmental Radiation Survey 80.12-1 80.13 Bases for Fire Protection Systems 80.13-1 80.10 Bases for Snubbers 80.10-1 90.15 Bases for Surveillance Requirements, Overpressure Mitigating System 80.15-1 90.18 Bases for System Flow Path Verifications 88.18-1 80.19 Bases for Reactor Coolant Vent System 89.19-1 80.20 Bases for Reactor Materials Surveillance Program M.20-1

-iv- Amendment Nos. and

LIST OF TABLES Table Title 1.1 Operational Modes 3.5-1 Instrument Operating Conditions for Reactor Trip 305 Engineering Safety Features Actuation 3.5-3 Instrument Operating Conditions for Isolation Functions 3.5-4 Engineered Safety Feature Set Points 3.13-1 Safety Related Snubhers 3.10-1 Fire Detection System 3.17-1 Soend Fuel Burnup Requirements, for Storage in Region II of the Spent Fuel

!it 3.18-1 Auxiliary Feedwater System Operability I

'.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument 0.1-2 Minimum Frequencies for Equipment and Sampling Tests 0.1-2 Minimum Frequencies for Equipment and Sampling Tests 0.2-1 Reactor Coolant System In-Service Inspection Schedule 0.2-2 Vlinimum Number of Steam Generators to be Inspected During Inservice Inspection 0.2-3 Steam Generator Tube Inspection 0.12-1 Operational Environmental Radiological Surveillance Program 0.12-2 Operational Environmental Radiological Surveillance Program Types of Analysis 6.2-1 Operating Personnel Amendment Nos. and

Applies to the operating status of the steam and power conversion systems.

To define conditions of the steam-relieving capacity.

l. When the reactor coolant of a nuclear unit is heated above 350oF, the following conditions mustbe met:

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

b. Its main steam stoo valves shall be operable and caoable of closing in 5 seconds or less.

c. System piping, interlocks and valves directly associated with the related components in TS 3.8.l,a, b 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the reactor shall be shutdown and the reactor coolant temperature reduced below 350oF.

Specification 3.0.1 applies.

3.8-l Amendment Nos. and

TABLE 3.16-1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum 4~(b)

System Valve No. Allowable Leaka e High-Head Safety Injection Unit 3 Unit 0 5.0 gpm Loop A, hot leg 3-870 A 0-870A 5.0 gpm cold leg 3-875 A 0-875A 5.0 gpm cold leg 3-873A 0-873A 5.0 gpm Loop B, hot leg 3-870B 0-870B 5.0 porn cold leg 3-875B 0-8758 5.0 gpm cold leg 3-873B 0-873B 5.0 gpm Loop C, cold leg 3-875C 0-875C 5.0 cpm cold leg 3-873 C 0-873C 5.0 gom Residual Heat Removal Looo A, cold leg 3-876A 0-876A 5.0 gpm 0-876E, . 5.0 gpm Loop B, cold leg 3-876B 0-876B 5.0 gpm 3-876D 0-876D 5.0 gpm Loop Cs cold leg 3-876C 0-876C 5.0 gom 3-876E 5.0 gpm

~a~ 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.

Leakage rates greater than 1.0 gom but less than or equal to 5.0 gom are considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50/o or greater.

3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the mawin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50'~0 or greater.

0. Leakage rates greater than 5.0 gpm are considered unacceptable.

~b~ Minimum differential test pressure shall not be less than 150 osid.

3.16-2 Amendment Nos. and

3.18 AUXILIARYFEEDWATER SYSTEM 3.18.1 Two independent auxiliary feedwater trains as specified in Table 3.18-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTION:

1) With one of the two required independent auxiliary feed water trains inoperable, either restore the inoperable train to an OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or place the affected unit(s) in HOT STANDBY within the next 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2) With both required auxiliary feedwater trains 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 both trains to an OPERABLE statug or restore one train to an OPERABLE status and follow ACTION statement 1 above for the other train, or place the affected unit(s) in 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 /> and in HOT SHUTDO>VN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.18-1 Amendment Nos. and

TABLE 3.18-1 AUXILIARYFEEDWATER SYSTEM OPERABILITY Unit Train Steam S 1 Flo ath ~Pum Dischar e Water Flo aths(3)

SG 3C via MOV-3-1005 A or C(2) SG 3A via CV-3-2816 or SG 38 via MOV-3-1000(1) SG 3B via CV-3-2817 SG 3C via CV-3-2818 SG 3A via 'VIOV-3-1403 B or C(2) SG 3A via CV-3-2831 or SG 3B via VlOV-3-1400(I) SG 3B via CV-3-2832 SG 3C via CV-3-2833 SG 0C via MOV-0-1005 A or C(2) SG OA via CV-0-2816 or SG 0B via VOV-0-1000(1) SG OB via CV-0-2817 SG 0C via CV-0-2818 SG OA via MOV-0-1003 B or C(2) SG 0A via CV-0-2831,.

or SG OB via MOV-0-1000(I) SG OB via CV-0-2832 SG OC via CV-0-2833 NOTES (1) Steam admission valves MOV-3-1000 and MOV-0-la00 can be aligned to either train to restore operability in the event ALOV-3-1003 or VOV-3-1005, or MOV-0-1003 or VOV-0-1005 are inoperable.

(2) During single and two unit operation, one pump is required to be ooerable in each train. The standby pump "C" can be aligned to either train to restore operability in the event one of the required pumps is inoperable.

(3) One flow control valve in each train for each unit can be inooerable for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The ACTION required for a single train inoperable shall be followed.

3.18-2 Amendment Nos. and

3.19 CONDENSATE STORAGE TANKS 3.19.1 The Condensate Storage Tanks shall be OPERABLE with a contained water volume of at least 185,000 gallons of water as follows:

3.19.1.l Sin le Unit Prior to Escalatin into Mode 3 a) ONE water supply from either Condensate Storage Tank including flowoath piping and valves.

3.19.1.2 Second Unit Prior to Escalatin into Mode 3 a) ONE water supply from each unit corresponding Condensate Storage Tank including flowpath piping and valves.

APPLICABILITY: MODES 1, 2, and 3.

ACTIO%

Sin le Unit at or Above Mode 3

1) With one water supply from a Condensate Storage Tank inoperable, within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, either realign the other Condensate Storage Tank containing the required water volume to the suction of the Auxiliary Feedwater pumps or restore the inonerable water supply to OPERABLE status or be in HOT STANDBY in the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2) With both water supplies from the Condensate Storage Tanks inoperable, within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> restore the water supply'from either Condensate Storage Tank to Operable status or be in IOT STANDBY 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 HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Both Units at or Above Mode 3

1) With one water supply from a Condensate Storage Tank inoperable, restore the inoperable water supply to OPERABLE status within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> or place one unit in 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 /> and in IOT SHUTDO'AN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. Refer to Single Unit Operation ACTION for single unit at or above MODE 3.

2) With both water supplies from the Condensate Storage Tanks inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore one water supply from a Condensate Storage Tank to OPERABLE status or place one unit in 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 /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. If unable to restore at least one water supply from a Condensate Storage Tank to OPERABLE status within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> from initial declaration of inoperability, the second unit shall be placed in 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 /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.19-1 Amendment Nos. and

CONDENSATE STORAGE TANKS The Condensate Storage Tanks shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying the contained water volume is within its limit when the tank is the supply source for the auxiliary feedwater pumps.

0.21-1 Amendment Nos. and

B3.8 BASES FOR LIMITINGCONDITIONS FOR OPERATION, STEAM AND POWER CONVERSION SYSTEMS The limit on secondary coolant iodine-131'pecific activity is based on a postulated release of secondary coolant equivalent to the contents of three steam generators to the atmosphere due to a net load rejection. %he limiting dose for this case would result from radioactive iodine in the secondary coolant. I-131 is the dominant isotope because of its low VlPC in air and because the other shorter lived iodine isotopes cannot build up to significant concentrations in the secondary coolant under the limits of primary system leak rate and activity. One tenth of the iodine in the secondary coolant is assumed to reach the site boundary making allowance for plate-out and retention in water droplets. 1he inhalation thyroid dose at the site boundary is then; Dose (Rem) = C ~

Y ~ 6 ~ OCF ~

X/Q 0.1

~

Where: C = secondary coolant I-131 specific activity

= 1.34 curies/m3 (pCi/cc) or 0.67 Ci/m3, each unit V = equivalent secondary coolant volume released = 210 m3 B = breathing rate = 3dk7 x 10 < m3 sec.

X/Q = atmospheric dispersion parameter = 1.50 x 10 ~ sec/m3 0.1 = equivalent fraction of activity released DCF =. dose conversion factor, Rem/ci The resultant thyroid dose is less than 1.5 Rem.

In the unlikely event of complete loss of electrical power to the nuclear units, decay heat removal will be assured by the availability of the steam-driven auxiliary feedwateI'umps and steam discharge via the steam generator safety valves and PORVs. (I) 1 FSAR - Section 10.3 B3.8-1 Amendment Nos. and

The OPERABILITY of the Auxiliary Feedwater System ensures that the Reactor Coolant System can be cooled down to less than 350oF from normal operating conditions in the event of a total loss of off-site power. Steam can be supolied to the pump turbines from either or both units through redundant steam headers. Two D.C.

motor operated valves and one A.C. motor operated valve on each unit isolate the three main steam lines from these headers. Both the D.C. and A.C. motor operated valves are powered from safety related sources. Auxiliary feedwater can be supplied through redundant lines to the safety related portions of the main feedwater lines to each of the steam generators. Air operated fail closed flow control valves are provided to modulate the flow to each steam generator. Each steam driven auxiliary feedwater pump has sufficient caoacity for single and two unit operation to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350oF when the Residual Heat Removal System may be placed into operation.

13.18-1 Amendment Nos. and

B3.19 BASES - CONDENSATE STORAGE TANKS There are two (2) seismically designed 250,000 gallon condensate storage tanks. A minimum of 185,000 gallons is maintained in each tank. Ihe OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the Reactor Coolant System at HOT STANDBY conditions for approximately 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> or maintain the Reactor Coolant System at HOT STANDBY conditions for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and then cool down the Reactor Coolant System to below 350oF at which point the Residual Heat Removal System may be placed into operation.

B3.19-1 Amendment Nos. and

PL A-039 ATTACHMENTI SAFETY AND NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION Descri tion of Amendment R uest:

~Pa e 3.$ -1 The proposed amendment would delete the Specifications for the Auxiliary Feedwater (AF'V) System and the Condensate Storage Tank (CST) in current Technical Specification 3.8, Steam and Power Conversion Systems. Requirements for the AF'V System and CST willbe included in new Technical Specifications 3.18 and 3.19.

Pa es 3.18-1 3.18-2 3.19-1 The proposed amendment would add Technical Specification 3.18, Auxiliary Feedwater System, and 3.19, Condensate Storage Tank. These proposed Specifications provide explicit limiting conditions for ooeration (LCO), applicability requirements, and ACTION requirements for operation of the AFW System and CST. The format (i.e., LCO, applicability, action requirements) is that of NUREG-0052, Standard Technical Specifications for 'Vestinghouse Pressurized Water Reactors (WSTS), although the requirements in the proposed Specifications differ from the iVSTS because of the uniqueness of the Turkey Point Plant AFW System design (i.e., shared system, three turbine driven pumps, etc.).

Proposed Specification 3.18 would differ from the current Technical Specification 3.8 as follows:

1) Table 3.18-1 defines the number of independent auxiliary feedwater pumps and their associated flowpaths (steam and water) required to be operable for single and two unit ooeration.

2) The prooosed Specification (LCO) requires that two of the three turbine driven AFW oumps be operable for both single and two unit operation. A single AFiV pump is sized to provide adequate flow to satisfy the minimum APE flow requirements for two unit ooeration. A recent Westinghouse reanalysis of the Loss of Non-Emergency AC Power to the Plant Auxiliaries event is attached. A second operable pump would satisfy the single Active failure criterion. Although all three AFiV pumps would normally be ooerable and aligned to the AFW system, as is required by the current Specification for two unit operation, the proposed Soecification (LCO) is consistent with the current design basis and safety analyses, would permit additional operational flexibility (reducing heatup/cooldown transients on the units), and is consistent with 10 CFR 50.36(c)(2) which states that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

3) The aoplicability of the proposed AFW specification is Modes 1, 2, 3, as defined in the Technical Specifications. This change differs from the current requirements in that the action requirements are applicable in all specified modes, whereas, under the current Technical Specification action is only specified to be taken when a limiting condition is not met during power operation, although the AF'V System is required to be operable when the reactor coolant temperature is above 350oF. Modes for AFW operation are not soecified in the current Technical Specifications.

0) The ACTION requirements in the proposed AFW Specification are consistent with the current Specification except for the following. The proposed Specification would allow one discharge water flowpath (i.e., a flow control valve )to be inoperable in both trains for a period not to exceed 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and allow one train to be inoperable in both units for a 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> period. In both cases, the AFiV System will orovide the minimum required flow through the remaining four operable flowpaths, or through the remaining operable train in each unit, respectively.

T18:7

~ Attachment I PL A-039 Proposed Specification 3.19 would differ from the current Technical Specification 3.8 as follows:

1) The proposed ACTIOiV requirements are more restrictive in that they require action to be taken within 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (consistent with the WSTS) as opposed to 08 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> in the

. current Specification.

Pa e 0.21-1 The proposed amendment would add Technical Specification 0.21, Condensate Storage Tank. This specification provides a surveillance requirement to demonstrate the CST operable by verifying at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that the water volume in the CST is within its limits when the CST is the supply source for the AF~V pumps. There is no similar requirement in the current Specifications.

Pa es B3.8-1 B3.18-1 B3.19-1 The proposed amendment would add separate bases (B3.18 and 83.19) for the AFlV system and the CST. The Bases for the Steam and Power Conversion Systems, B3.8, would be modified accordingly to delete reference to the AF'V System and CST.

In Table 3.16-1, the valve numbers for HHSI Loop C Cold Leg and RHR Loop B Cold Leg shown as 3-875B and 3-876A would be corrected to read 3-875C and 3-876B, respectively, to reflect the correct valve numbers.

Basis for No Significant Hazards Consideration Determination:

The Commission has orovided standards for determining whether a significant hazards consideration exists 10 CFR 50.92(c)]. A proposed amendment to an operating license for the facility involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated, or (2) create the possibility of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety.

Operation of Turkey Point Units 3 and 0 in accordance with the proposed amendments would not:

1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

Technical Specification 3.18 and Table 3.18.1 define the number of independent AFEV pumps and their associated flowpaths (steam and water) required to be operable for single and two unit ooeration. Operation of the system in accordance with this Specification would ensure that adequate core and RCP heat removal is available to prevent water relief out the pressurizer relief or safety valves. This is the basis for the current Technical Specification and consistent with the FSAR safety analyses.

The requirements for CST operation in proposed Technical Specification 3.19 are as restrictive or more restrictive than the requirements in current Technical Specification 3.8.

tT The addition of Specification 0.21 to verify operability of the CSTs further ensures that the limiting conditions for operation for the CSTs willbe met.

The changes to Table 3.16-1 would correct valve designations. Vo changes to the systems were made.

T18:7

Attachment I PL A-039 Based on the above, operation in accordance with the proposed changes would not involve an increase in the probability or consequences of an accident previously evaluated.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated.

The operation of the AFlV System and CSTs is not significantly different from that allowed by the current Technical Specifications, and since the conclusions of the safety analyses remain valid (i.e., adequate core and reactor coolant pump heat removal is available), operation in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Involve' significant reduction in a margin of safety.

As noted in response to (1) and (2) above, the operation of the APV System and CSTs is permitted by the proposed Technical Specification is not significantly different from that allowed by the current Technical Specifications. Adequate heat removal capability is available to remove core and RCP heat and to prevent water relief out the pressurizer relief or safety valves, insuring that the integrity of the core and RCS is not compromised. Also, the addition of CST surveillance requirements further ensures that the LCO for the CSTs will be met. Thus, operation in accordance with the proposed changes will not involve a significant reduction in a margin of safety.

Based on the above discussion, operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated, or create the possibility of a new or different kind of accident from any accident previously evaluated, or involve.a significant reduction in a margin of safety.

Therefore, operation of the facility in accordance with the oroposed amendment would pose no threat to the public health and safety, and would not involve a significant hazards consideration.

14.1.12 LOSS OF NON-t. ERGENCY A-C POWER TO THE PLANT ILIARIES 14.1.12.1 Ident1fication of Causes and Acc1dent Oescrf tion A complete loss of non-emergency AC power may result in the loss 'of all power to the plant auxiliaries, 1.e., the reactor coolant pumps, condensate'pumps, etc. The loss of power may be caused by a complete loss of the offs1te grid accompanied by a turbine generator tr1p at the station, or by a loss of the onsite AC distribution system.

This transient 1s more severe than the turbine tr1p event because for th1s case the decrease 1n heat removal by the secondary system is accompanied by a flow coastdown which further reduces the capacity of the pr1mary coolant to remove heat from the core. The reactor w111 trip:

(1) upon reach1ng one of the trip setpoints 1n the primary and secondary systems as a result of the flow coastdown and decrease in secondary heat removal; or (2) due to loss of power to the control rod drive mechanisms as a result of the loss of power to the plant.

Following a loss of AC power w1th turbine and reactor trips, the sequence described below will occur:

l. Plant vital instruments are suppl1ed from emergency DC power sources.

2. As the steam system pressure r1ses following the tr1p, the steam generator power-operated relief valves may be automatically opened to the atmosphere. The condenser 1s assumed not to be available for steam dump.

If the steam flow rate through the power rel1ef valves 1s not available, the steam generator safety valves may 11ft to diss1pate the sensible heat of the fuel and coolant plus the residual decay heat produced 1n the reactor.

84660:1D/030686 14.1.12-1

3. As the no load t rature 1s approached, the ste~enerator power-operated relief valves (or safety valves, ice power operated relief valves are not available) are used to dissipate the residual decay heat and to ma1nta1n the plant at the hot shutdown condition.

4. Both emergency diesel generators will start on loss of voltage on. both the 4160 volt buses of either un1t. At the same time, these buses w1ll be 1solated from the1r normal supply and the1r motor feed breakers will be opened. Rotor control center t1e breakers w1ll open, separating the vital loads from the others. At th1s time, the generator breakers will close.

All further operat1ons required to p1ck up the emergency loads will be done manually by the operator. These operations will be done automatically in a sequential manner only 1f there has been a coincident safety in)ection s1gnal in the same unit.

The following provides the necessary protection against a loss of AC power:

l. Reactor trip on

a. Low-low water level in any steam generator

b. Steam flow-feedwater flow mismatch coincident with low water level in any steam generator

2. Three turb1ne-driven aux1liary feedwater pumps (shared by units 3 5 4) are started on any of the follow1ng:

a. Low-low water level in any steam generator

b. Any safety in)ection signal

c. Loss of offs1te power

d. Loss of 4 kV bus

e. Trip of all ma1n feedwater pumps

f. manual actuation 8466/:10/030686 14.1.12-2

The steam driven aux ary feedwater pumps are starte on the loss of normal feedwater supply. The turb1ne utilizes steam from the main steam line to drive the feedwater pump to del1ver makeup water to the steam generators. The turbine driver exhausts the steam to the atmosphere., The pumps take suct1on directly from the condensate storage tanks for delivery to the steam generators.

The steam-driven auxiliary feedwater pump can be tested at any t1me by admitting steam to the turbine driver. The auxil1ary feedwater control valves and main steam power rel1ef valves can be operat1onally tested whenever the unit is at hot shutdown and the remain1ng valves 1n the system are operationally tested when the turbine driver and pump are tested.

Upon the loss of power to the reactor coolant pumps, coolant flow necessary for core cooling and the removal of residual heat is maintained by natural circulat1on 1n the reactor coolant loops.

A loss of AC power event, as described above, is a more lim1ting event than the turbine-tr1p 1n1t1ated decrease in secondary heat removal without loss of AC power. However, a loss of AC power to the plant auxil1ar1es as postulated above could result 1n a loss of normal feedwater if the condensate pumps lose their power supply.

Following the reactor coolant pump coastdown caused by the loss of AC power, the natural circulat1on capability of the RCS will remove residual and decay heat from the core, aided by aux111ary feedwater 1n the secondary system. An analys1s 1s presented here to show that the natural circulation Flow in the RCS following a loss of AC power event 1s sufficient to remove residual heat from the core.

Turkey Point Units 3 and 4 share co+son electr1cal and auxiliary feedwater systems. Thus, a loss of non-emergency AC power to the plant auxiliaries could simultaneously affect both un1ts. The auxiliary feedwater system would then be required to prov1de flow to both units.

84660:1D/030686 14. l . 1 2-3

The worst single fai~re in the auxiliary feedwater system could result in ava1lability of online of the three auxiliary feeder pumps. Flow from this pump could be as low as 125 gpm to one of the units until the operator takes action from the control board to realign the flow split to the un1ts.

The analysis 1s performed for one un1t, representing the worst case of the two un1ts.

14.l.l2.2 Anal sis of Effects and Conse uences Hethod of Analysis

'I A detailed analysis using the LOFTRAN Code (Reference 1) 1s perfor'med to obtain the plant transient following a station blackout. The simulat1on describes the plant thermal k1netics, RCS 1ncluding the natural circulat1on, pressurizer, steam generators and feedwater system. The dig1tal program computes pertinent variables 1ncluding the steam generator mass, pressurizer water level, and reactor coolant average temperature.

Assumptions made in the analysis are:

l. The plant is initially operating at 102 percent of the Engineered Safety Features design rating, 2307.4 HMt.

2. Core residual heat generation is based on the 1979 version of ANS-5.1 (Reference 2). ANSI/ANS-5.1-1979 1s a conservative representation of the decay energy release rates.

3. A heat transfer coeffic1ent in the steam generator associated with RCS natural circulation, following the reactor coolant pump coastdown.

4. Reactor tr1p occurs on steam generator low-low water level. No credit is taken for iaeediate release of the control rod drive mechan1sms caused by a loss of offsite power.

8466':19/030686 14.1.12&

5. The worst s1ngle lure:occurs 1n the aux1liary water system. Th1s results 1n the availab111ty of one aux111ary feedwater pump supplying l25 gpm to three steam generators three minutes follow1ng a start signal on low-low steam generator water level. At ten m1nutes follow1ng reactor trip the operator takes action from the control board to 1ncrease the flow to 230 gpm delivered to three steam generators.

6. Secondary system steam rel1ef 1s ach1eved through the steam generator safety valves.

7. The 1nitial reactor coolant average temperature 1s 4'F higher than the nominal value, and 1n1tial pressurizer pressure 1s 30 psi h1gher.than nominal.

8. The pressurizer power-operated rel1ef valves and pressurizer spray system are assumed to operate normally. Th1s results in a conservat1ve transient with respect to peak pressurizer water level. If these control systems d1d not operate the pressurizer safety valves would maintain peak RCS pressure at or below the actuation setpoint (2500 psia) throughout the transient.

9. A control rod drop time to dashpot of 2.4 seconds was assumed, consistent with optimized fuel assemblies (OFA).

The assumptions used in the analysis are essentially 1dent1cal to the loss of normal feedwater flow 1nc1dent (Section 14.l.ll) except that power 1s assumed to be lost to the reactor coolant pumps at the time of reactor trip.

Results The transient response of the RCS follow1ng a loss of AC power is shown in F1gures 14.1.12-l and 14.1.12-2. The calculated sequence of events for this acc1dent is listed 1n Table 14.l.l2-2.

8466Q:10/030686 14.1.12-5

The first few will closely seconds t after the loss of power to the resemble a simulation of the complete loss ctor coolant pumps of flow incident i.e.,

core damage due to rapidly increasing core temperatures is prevented by promptly tripping the reactor.

Upon the loss of power to the reactor coolant pumps, coolant flow nec'essary for core cooling and the removal of residual heat is maintained by natural circulation in the reactor coolant loops. The natural circulation flow was calculated using an analytical method based on the conditions of equilibrium flow and maximum loop flow impedence. The model has given results within 15$

of the measured flow values obtained during natural circulation tests conducted at the Yankee-Rowe plant and has also been confirmed at San Onofre and Connecticut Yankee. The natural circulation flow ratio as a function of reactor power is given in Table 14.1.12-1.

14.1.12.3 Conclusions Analysis of the natural circulation capability of the Reactor Coolant System has demonstrated that sufficient heat removal capability exists following RCP coastdown to prevent fuel or clad damage.

14.1.12.4 References

1. Burnett, T. M. T., et al, "LOFTRAN Code Description,'CAP-7907-P-A (Proprietary), WCAP-7907-A (Non-Proprietary), April 1984.

2. ANSI/ANS-5.1-1979, "American National Standard For Decay Heat Power in Light Mater Reactors," August 1979.

84669:10/030686 14.1.12-6

TABLE 14.1.12-1 NATURAL CIRCULATION REACTOR COOLANT FLOW VS. REACTOR POWER Reactor Power Reactor Coolant Flow Nominal Flow 3.5 4.6 3.0 4.3 2.5 4.0 2.0 3.7 1.5 3.3 1.0 2.9 8466/:1D/0306B6 14.1.12-7

TABLE l4.1.12-2 TINE SE UENCE OF EVENTS FOR LOSS OF NON-EMERGENCY AC POWER Event T1me sec Hain feedwater flow stops 10 Low-low steam generator water level tr1p 64 Rods begin to drop Reactor coolant pumps beg1n to coastdown Flow from one turb1ne driven aux111ary 244 feedwater pump is started Operator realigns system to 1ncrease 664 auxiliary feedwater flow to 230 gpm Feedwater 11nes are purged and cold 1080 auxil1ary feedwater is delivered to three steam generators Peak water level 1n pressurizer occurs 3720 Core decay heat decreases to auxiliary &000 feedwater heat removal capacity 8466/:lD/030686 14.1.12-8

a: 2588.

L) 2488.

g 2588.

gg 2288 '

2188.

~ 2888.

1 F88.

1888.

188 181 182 184 TIME (SEC) 1488.

1288 ~

1888.

CI OC Lal 888.

CK OC La>

688.

t4 OC EA 488.

C/l 4l CC CL 288.

188 181 182 18~

T IME t SEC)

Figure 14.1.12-1. Pressurizer Pressure and Water Volume Transients for Loss of Offsite Power

788.

688.

s 668.

oc 648.

gg 628. HOT LEG

~

I 688.

588.

C7

~~ 568. COLD LEG 548.

528.

188 181 . 182 184 TAHE tSEC) 1288.

Cg 1888.

W 888.

P 688 'K TLUTE u 488.

EK W

~ 288.

8.

188 181 182 185 184 l SEC 1 Figure 14.1.12-2. Loop Temperatures and Steam Generator Pressure for Loss of Offsite Power