Text

Amends 149 & 144 to Licenses DPR-31 & DPR-41,respectively, Revising TS by Removing Outdated Matl,Incorporating Administrative Clarifications & Correcting Typos
	ML17348B145
Person / Time
Site:	Turkey Point
Issue date:	09/25/1991
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17348B146	List: ML17348B145; ML17348B148;
References
	NUDOCS 9110100148
	Download: ML17348B145 (61)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 FIORIDA POWER AND LIGHT COMPANY DOCKET NO. 50"250 TURKEY POINT PLANT UNIT NO.

3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 149 License.No. DPR-31 1.

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

A.

The application for amendment by Florida Power and I.ight Company (the licensee) dated May 28,

1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9110100148 910925 PDR ADDCH, 05000250 p

PDR.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated ia the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-41 is hereby amended to read as follows (B)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No. 149, are hereby incorporated in the license.

The Environmental Protection Plan contained in Appendix B is hereby incorporated into the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protectioa Plan 3.

This license ameadmeat is effective as of the date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY COMMISSION Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Chaages to the Technical Specifications Date of Issuance:

September 25, 1991

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UNITED STATES NUCLEAR REGULATORY COIVIIVIISSION WASHINGTON. D.C. 20555 FLORIDA POWER AND LIGHT COMPANY DOCKET NO. 50-251 TURKEY POINT PLANT UNIT NO.

4 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 144 License No. DPR-41 1.

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

A.

The application for amendment by Florida Power and Light Company (the licensee) dated May 28,

1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

C.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

1.

1

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B of Facility Operating License No. DPR-31 is hereby amended to read as follows (B)

Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A as revised through Amendment No.

144, are hereby incorporated in the license.

The Environmental Protection Plan contained in Appendix B is hereby incorporated into the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan 3.

This license amendment is effective as of the date of issuance and shall be implemented within 60 days.

FOR THE NUCLEAR REGULATORY CONMISSION Herbert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 25, 1991

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.

149 FACILITY OPERATING LICENSE NO. DPR-31 AMENDMENT NO-144 FACILITY OPERATING.LICENSE NO. DPR-41 DOCKET NOS. 50-250 AND 50-251 Revise Appendix A as follows:

1-2 2-10 3/4 1-23 3/4 3"15 3/4 3-18 3/4 3"26 3/4 3-36 3/4 3-43 3/4 3-48 3/4 3-49 3/4 3-53 3/4 3"56 3/4 3-59 3/4 3-60 3/4 3-61 3/4 4-22 3/4 4-28 3/4 6-10 3/4 6-15 3/4 7-33 3/4 9-12 3/4 9-i3 3/4 9-15 3/4 1'1-15 3/4 12-2 B 3/4 9"4 5-5 5-6 6 6-7 6-14 6-.15.

6-i7 6-22 Insert Pa es 1-2 2-10 3/4 1-23 3/4 3-15 3/4 3-18 3/4 3-26 3/4 3-36 3/4 3-43 3/4 3-48 3/4 3-49 3/4 3-53 3/4 3-56 3/4 3-59 3/4 3-60 3/4 3-61 3/4 4-22 3/4 4-28 3/4 6-10 3/4 6-15 3/4 7-33 3/4 9-12 3/4 9-13 3/4 9-15 3/4 11-15 3/4 12-2 B 3/4 9"4 5"5 5-6 6-1 6-7 6-14 6-i5 6-17 6-22

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DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1)

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 2)

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Specification 3.6.4.

b.

The equipment hatch is closed and sealed, c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3, d.

The containment leakage rates are within the limits of Specification 3.6.1.2, and e.

The sealing mechanism associated with each penetration (e.g., welds,

bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

CORE ALTERATIONS 1.9 CORE ALTERATIONS shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative, position.

DOSE E UIVALENT I-131 1.10 DOSE E(UIVALENT I-131 shall be that concentration of I-131 (microCurie/gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be. those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test-Reactor Sites" or Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

E " AVERAGE DISINTEGRATION ENERGY 1.11 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 (MeV/d) for the radionuclides in the sample isotopes, other than iodines, with half lives greater than 30 minutes, making up at least 95 percent of the total non-iodine activity in the coolant.

TURKEY POINT - UNITS 3 & 4 1-2 AMENDMENT NOS.

149 AND 144

I f11 g 'OTE 3:

(Continued)

I Kg C

CA CrO Qe TABLE 2.2-1 Continued TABLE NOTATIONS Continued 0.00068/

F for T > T" 0 for T<T" As defined in Note 1, Indicated T

at RATED THERHAL POWER (Calibration temperature for 4T instrumentation,

< 574.2 F),

As defined in Note 1, and fg (hI)

=

0 for all dI s

NOTE 4..

C)

The channel's maximum trip setpoint shall not exceed its computed trip setpoint by more than 1.4X of instrument span.

REACTIVITY CONTROL SYSTEMS POSITION INDICATION SYSTEM -

SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 The group step counter demand position indicator shall be OPERABLE and capable of determining within f 2 steps the demand position for each shut-down and control rod not fully inserted.

APPLICABILITY:

MODES 3" "", 4* *", and 5" **

ACTION:

With less than the above required group step counter demand position indica-tor(s)

OPERABLE, open the reactor trip system breakers.

SURVEILLANCE RE UIREMENTS 4.1.3.3. 1 Each of the above required group step counter demand position indi-cator(s) shall be determined to be OPERABLE by movement of the associated control rod at least 10 steps in any one direction at least once per 31 days.

4. 1.3.3.2 OPERABILITY of the group step counter demand position indicator shall be verified in accordance with Table 4.1-1.

"With the Reactor Trip System breakers in the closed position.

"*See Special Test Exceptions Specification 3.10.5.

TURKEY POINT " UNITS 3 8L 4 3/4 1-23 AMENDMENT NOS. 149 AND 144

C m

FUNCTIONAL UNIT TABLE 3. 3-2 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION C

I Qe f.

Steam Line FlowHigh Coincident with:.

Steam Generator Pressure Low 1/steam generator 1/steam generator in any two steam lines 1/steam generator in any two steam lines 1, 2, 3*

2/steam line 1/steam line 1/steam line 1, 2, 3*

in any two in any two steam lines steam lines 15 15 CAR I

4Jl or T

Low 2.

Containment Spray a.

Automatic Actuation Logic and Actuation Relays 1/loop 25 1, 2, 3, 4 14 1/loop in any 1/loop in any 1, 2, 3" two loops two loops b.

Containment Pressure High-High Coincident with:

Containment Pressure High 1, 2, 3

1,2,3 15 15 3.

Containment Isolation a.

Phase "A" Isolation 1)

Hanual Initiation 2

2)

Automatic Actuation 2

,Logic and Actuation Relays 1, 2, 3, 4 17 1, 2, 3, 4 14

TABLE 3.3-2 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION Steam Line Isolation (Continued) d.

Steam Line FlowHigh 2/steam. line Coincident with:

Steam Genel'ator Pressure Low 1/steam line in any two steam lines 1/steam line 1

2 3

in any two steam lines 15 or ow Feedwatel Isolation a.

Automatic Actua-tion Logic and Actuation Relays 1/steam generator 1/loop 1/steam generator in any two steam lines 1/loop in any two loops 1/steam generator in any two steam lines 1/loop in any two loops 1, 2, 3

1,2,3 1,

2

. 15 25 22 b.

Safety-Injection 6.

Auxiliary Feedwaterk@

See Item 1.

above for all Safety Injection initiating functions and requirements.

a.

Autdmatfc Actua-tion Logic and Actuation Relays 1, 2, 3

20

TABLE 3.3-3 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM FUNCTIONAL UNIT 4.

Steam Line Isolation (Continued)

ALLOWANCE TA Z

S TRIP SETPOINT ALLOWABLE VALUES b.

Automatic Actuation Logic and Actuation Relays'.A.

N.A.

c.

Containment Pressure High" High Coincident with:

Containment Pressure-High

21. 3

13. 3 2.7 0.0

<20.0 psig 10.3 0.0

<4.0 psig

<22.6 psig

<4.5 psig d.

Steam Line Flow""High 16.7 2.86 3.9

<A function defined as follows:

A hp corresponding to 40X steam flow at OX load increasing linearly from 20X load to a value corresponding to 120X steam flow at full load.

<A function defined as follows:

A bp corresponding to 42.6X steam flow at OX load increasing linearly from 20X load to a value corresponding to 122.6X steam flow at full load.

Coincident with:

Steam Line Pressure Low or Tavg Low edwater Isolation 5.

Fe ae Autoiatic Actuation Logic and Actuation Relays

13. 0 4.0 N.A.

1.16 2.3

>614 psig 2.0 1.0

>543 F

N.A N.A.

N.A.

>588 psig 542 5oF N.A.

b.

Safety Injection see item 1 See Item l. above for all Safety Injection Trip Setpoints and Allowable Values.

TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS FUNCTIONAL UNIT 1.

Containment MINIMUM CHANNELS CHANNELS APPLICABLE ALARM/TRIP TO TRIP/ALARH OPERABLE MODES SETPOINT ACTION 2.

a.

Containment Atmosphere 1

Radioactivity-High (Particulate or Gaseous (See Note 1.))

b.

RCS Leakage Detection N.A.

Particulate Radio-activity or Gaseous Radioactivity Spent Fuel Storage Pool Areas A11*

Particulate

<6.lxl0sCPM t'aseous See Note 2.

1, 2, 3, 4 N.A.

26 for MODES 1, 2, 3, 4 or 27 for MODES 5 AND 6 26-a.

Unit 3 Radioactivity -

1 High Gaseous b.

Unit 4 Radioactivity-1 HighGaseous'5.5xl0-2 pCi 28 CC

<2.8xl0-2 gCi 28 CC (SPING) or

<1.0xlOsCPM (PRMS)

INSTRUMENT TABLE 3.3-5 (Continued)

ACCIDENT MONITORING INSTRUMENTATION TOTAL NO.

OF CHANNELS MINIMUM CHANNELS OPERABLE APPLI-CABLE NOOES ACTIONS C

Vl CAB Qo EA3 I

14.

In Core Thermocouples (Core Exit Thermo-couples) 15.

Containment High Range Area Radiation 16.

Reactor Vessel Level Monitoring System 17.

Neutron Flux, Backup NIS (Mide Range) 18.

Containment Hydrogen Monitors 19.

High Range-Noble Gas Effluent Monitors a.

Plant Vent Exhaust b.

Unit 3-Spent Fuel Pit Exhaust

'c.

Condenser Air Ejectors

. d.

Hain Steam Lines 20.

RMST Mater Level 21.

Steam Generator Mater Level (Narrow Range) 22.

Containment Isolation Valve Position Indication*

4/core quadrant 2

2(1) 2 2/stm.

gen.

1/valve 2/core quadrant 1

1(1) 1 1

1 1

1 1/stm.

gen.

1/valve 1, 2, 3 1.2 3

1,2,5 1,

2 37, 38 0 31 32 35 ALL ALL 1.2 3

1, 2, 3 1, 2, 3

1, 2, 3 1, 2, 3 34 34 34 34 31, 32 31, 32 39 1, 2, 3

31, 32 CD CD 1.

2.

TABLE NOTATIONS A charm 1 is eight sensors in a probe.

A channel is OPERABLE if a minimum of four sensors are OPERABLE.

Inputs to this. instrlnt are from instrument items 3, 4, 5 and 14 of this Table.

"Applicable for containment isolation valve position indication designated as post-accident monitoring instru-mentation (containment isolation valves which receive containment isolation Phase A, Phase B, or containment ventilation isolation signals).

TABLE 3.3-6 FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION FIRE ZONE AREA 4-5-

9 10 "

ll "

12-13-14-15 "

16-19-20-21-22-25 "

25A-26 "

27-30-40-45-47-54-55-58-59-60-61-62 "

63-67-68 "

70 "

71-72-73-74-75-76-79A-81-82-84-Aux. Bldg. Corridor E.

10'hem.

Drain/Laundry/Shower Tank Room Laundry/Chemical Drain Tank Room Pipeway Unit 3 RHR Heat Exchanger Room RHR Pump 3A Room RHR Pump 3B Room Unit 4 RHR Heat Exchanger Room RHR Pump 4A Room RHR Pump 4B Room Unit 3 W Elect Penet Room Unit 3 S Elect Penet Room Instrument Shop Radioactive Laboratory Aux. Bldg. Elect.

Equipmt.

Room Spare Battery Room Unit 4 N Elect Penet Room Unit 4 W Elect Penet Room Unit 4 Piping and Valve Room Unit 3 Piping and Valve Room Unit 4 Charging Pump Room Unit 4 Component Cooling Water Area Unit 3 Component Cooling Water Area Unit 3 Charging Pump Room Aux Bldg Corridor, El.

18'nit 4 Containment Electrical Penet.

Area""

Unit 3 Containment Electrical Penet.

Area**

Reactor Control Rod Eqpmt Room - Unit 4 Computer Room Reactor Control Rod Eqpmt Room - Unit 3 4160V Switchgear 4B 4160V Switchgear 4A 4160V Switchgear 3B 4160V Switchgear 3A Diesel Generator 3B Diesel Generator 3A Day Tank Room 3B Day Tank Room 3A Unit 4 Turbine Lube Oil Reservoir North-South Breezeway Unit 4 Main Transformer Unit 4 Aux Transformer Area Unit 3 and 4 Aux Feedwater Pump Area (DC Enclosure Bldg.)

(2/o)

(0/4)

(5/2)

(0/4)

(4/2)

(0/4)

(0/3)

(1/1)

(1/0)

(0/6)

(1/0)

TOTAL NUMBER OF INSTRUMENTS

~xy)"

~xlyy"

~x)~y (2/o)

(2/0)

(1/0)

(11/0)

(5/0)

(2/0)

(5/0)

(2/o)

(2/0)

(5/0)

(ll/0)

(2/0)

(6/0)

(8/o)

(6/o)

(4/0)

(3/0)

(18/0)

(10/0)

(16/0)

(4/0)

(11/0)

(4/0)

(10/0)

(6/0)

(10/0)

(6/0)

(1/0)

(4/0)

(3/0)

TURKEY POINT - UNITS 3 8c 4 3/4 3-48 AMENDMENT NOS. 149 AND 144

TABLE 3.3-6 Continued FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS (1/0)

(2/0)

(1/0)

(2/0)

(1/0)

(16/15)

(1/0)

(2/0)

(16/0)

(3/4)

(4/4)

(1/2)

(5/0)

(2/0)

(5/0)

(2/0)

(2/0),

(N/A)¹ (N/A)¹ (N/A)¹ FIRE ZONE AREA

~y)*

~xi'*

~x~y~

87 - Unit 3 Aux Transformer Area (1/0) 93 - 480V Load Center 4A and 4B 94 - 480V Load Center 4C and 4D 95 - 480V Load Center 3A and 3B 96 - 480V Load Center 3C and 3D 97 - Mechanical Equipment Room 98 - Cable Spreading Room 101-RPI Inverter and MG Sets 102-Battery Rack 4B (1/0) 103-Battery Rack 3A (1/0) 104-RPI Inverter and MG Sets 106-Control Room (1/0) 108A-Train A Inverters 108B-Train B Inverters 109-Battery Rack 4A (1/0) 110- Battery Rack 3B (1/0) 113-Unit 4 Feedwater Platform (2/0) 116-Unit 3 Feedwater Platform (2/0) 119-Unit 4 Intake Cooling Water Pump Area (4/0) 120- Unit 3 Intake Cooling Water Pump Area (4/0) 132-Control Room Electrical Chase 133-Diesel Generator 4B (5/5)

(3/0)

- 134-4160V Switchgear 3D Room 135-Diesel Generator 4B Control Panel Room 136-Diesel Generator 4B Fuel Transfer Pump

=

138-Diesel Generator 4A (5/5)'3/0) 139-4160V Switchgear 4D Room 140- Diesel Generator 4A Control Panel Room 141-Diesel Generator 4A Fuel Transfer Pump N/A - 18'evel of the Turbine Area TABLE NOTATIONS x is number of Function A (early warning fire dectection and notification only) instruments.

y is number of Function B (actuation of Fire Suppression 'Systems and early warning fire detection and notification) instruments.

The fire detection instruments located within the containment are not required to be operable during the performance of Type A Containment Leakage Rate Test.

¹ A fire watch patrol shall be established to inspect the 18 foot level of the Turbine Area once each hour.

TURKEY POINT - UNITS 3 8( 4 3/4 3-49 AMENDMENT NOS. 149 AND >44

TABLE 4.3-5 RADIOACTIVE LI UID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT 1.

Gross Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Liquid Radwaste Effluents Line b.

Steam Generator Blowdown Effluent Line 2.

Flow Rate Heasurment Devices a.

Liquid Radwaste Effluent Line b.

Steam Generator Blowdown Effluent Lines CHANNEL CHECK D(3)

D(3)

SOURCE CHECK N.A.

N.A.

CHANNEL CALIBRATION R(2)

R(2)

ANALOG CHANNEL OPERATIONAL TEST qo) q(1)

TABLE NOTATIONS (l)

The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

(2)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBJNTION, sources that have been related to the in>tial calibration shall be used.

(3)

CHANNEL CHECK shall consist shall be made at least once made.

of verifying indication of flow during periods of release.

CHANNEL CHECK per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuous, periodic, or batch releases are

TABLE 3.3-8 Continued RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT Plant Vent System (Include Unit 4's Spent Fuel Pool)

Noble Gas ActivityMonitor (SPING or PRMS)

Iodine Sampler c.

Particulate Sampler d.

Effluent System Flow Rate Measuring Device e.

Sampler Flow Rate Measuring Device 5.

Unit 3 Spent Fuel Pit Building Vent a.

Noble Gas Activity Monitor b.

Iodine Sampler c.

Particulate Sampler d.

Samp'ler Flow Rate Measuring Device MINIMUM CHANNELS OPERABLE APPLICABILITY ACTION 47 46 46

TABlE 4. 3-6 Continued RADIOACTIVE GASEOUS EFFLUENT HONITORING INSTRUHENTATION SURVEILLANCE RE UIREHENTS INSTRUMENT 3.

Condenser Air Ejector Vent System (Continued)

ANALOG CHANNEL MODES.FOR NIGH CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE. IS CHECK CHECK CALIBRATION TEST RE UIREO e.

Sample Flow Rate Measuring Device N.A.

N.A.

4.

Plant Vent System (Include Unit 4's Spent Fuel Pool) aO Noble Gas Activity Monitor (SPING or PRHS)

R(3) q(~)

.b.

Iodine Sampler N.A.

N.A.

C.

Particulate Sampler N.A.

N.A.

d.

Effluent System Flow Rate Measuring Device N.A.

N.A.

e.

Sampler Flow Rate Measuring Device D

N.A.

TABLE 4.3-6 Continued RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL SOURCE CHANNEL INSTRUMENT CHECK CHECK CALIBRATION I

5.

Unit 3 Spent Fuel Pit Building Vent ANALOG CHANNEL MODES FOR WHICH OPERATIONAL SURVEILLANCE IS TEST RE UIRED aO Noble Gas Activity Monitor R(3)

S(2) b.

Iodine Sampler c.

Particulate Sampler d.

Sampler Flow Rate Measuring Device 0

N.A.

TABLE NOTATION At all times.

" During GAS DECAY TANK SYSTEM operation.

Applies during MODE 1, 2, 3 and 4.

H Applies during MODE 1, 2, 3 and 4 when primary to secondary leakage is detected as indicated by condenser air ejector noble gas activity monitor.

(1)

'The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic isolation of this pathway and control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

(2)

The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that if the instrument indicates measured levels above the Alarm Setpoint, alarm annunciation occurs in the control room (for PRMS only) and in the computer room (for SPING only).

(3)

The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Institute of Standards and Technology (NIST) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NIST.

These standards shall permit calibrating the system over its intended range of energy and measurement range.

For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

TABLE 4.3-6 (Continued)

TABLE NOTATIONS Continued (4)

The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal.

a.

One volume percent

hydrogen, balance nitrogen, and b.

Four volume percent

hydrogen, balance nitrogen.

(5)

The CHANNEL CALIBRATION shall include the use of standard gas samples containing a nominal:

a.

One volume percent

oxygen, balance nitrogen, and b.

Four volume percent

oxygen, balance nitrogen.

TURKEY POINT - UNITS 3 & 4 3/4 3-61 AMENDMENT NOS. 149 AND 144

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER Unit 3 Unit 4 3-874A 4"874A 3-875A 4-875A 3-873A 4-873A FUNCTION High-Mead Safety Injection Check Valves Loop A, hot leg cold leg cold leg 3-874B 3-875B 3-873B 3-875C 3-873C 3-876A 3"876B 3-8760 3-876C 3-876E MOV3-750 MOV3-751 4-874B 4"875B 4-873B 4-875C 4-873C 4"876A 4-876E 4-876B 4-876D 4-876C MOV4-750 MOV4-751 Loop B, hot leg cold leg cold leg Loop C, cold leg cold leg Residual Heat Removal Line Check Valves Loop A, cold leg Loop B, cold leg Loop C, cold leg Loop A, hot leg to RHR Loop C, hot leg to RHR ACCEPTABLE LEAKAGE LIMITS 1.

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

2.

3.

Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable provided that the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the margin between previously measured leakage rate and the maximum permis-sible rate of 5.0 gpm by 50X or greater.

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 margin between previously measured leakage rate and the maximum permissible rate of 5.0 gpm by 50K or greater.

4.

Leakage rates greater than 5.0 gpm are considered unacceptable.

TURKEY POINT - UNITS'3 & 4

/

3/4 4-22 AMENDMENT NOS. 149 AND 144

TASLE 4.4"4 REACTOR COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM TYPE OF MEASUREMENT AND ANALYSIS 1.

Gross Radioactivity Determination 2.

Tritium Activity Determination 3.

Isotopic Analysis for DOSE E(UIVAL8iT I-131 Concentration SAMPLE AND ANALYSIS FRE UENCY At least once per

. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

1 per 7 days.

1 per 14 days.

MODES IN WHICH SAMPLE AND ANALYSIS RE UIRED 1, 2, 3, 4 1, 2, 3, 4 5.

Radiochemical Isotopic Determination Including Gaseous Activity Radiochemical for E Determination Monthly 1 per 6 months*

1, 2, 3, 4 6.

Isotopic Analysis for Iodine Including I-131, I-133, and I-135 a)

Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever the specific activity exceeds 1 pCi/gram DOSE E(UIVALENT I-131 or 100/E pCi/gram of gross radioactivity, and b)

One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months follow-ing a THERMAL POWER change exceeding 15K of the RATED THERMAL POWER within a 1-hour period.

lP,2$,38, Q, 5k 1, 2, 3

n pl A

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS (Continued d.

Assuring that the 'observed lift-offforce for each tendon exceeds the min>mum required lift-offforce.

Required lift-offforces shall be calculated individually for each surveillance tendon prior to the beginning of each surveillance, and should consider such factors as:

1)

Prestressing history; 2)

Friction losses; and 3)

Time-dependent losses (creep, shrinkage, relaxation), considering time elapsed from prestressing.

e.

Verifying the OPERABILITY of the sheathing filler grease by:

1)

Minimum grease coverage exists for the different parts of the anchorage

system, and 2)

The chemical properties of the filler material are within the tolerance limits as specified by the manufacturer.

4. 6. l. 6. 2

'End Anchora es and Ad 'acent Concrete Surfaces.

The structural integrity o e en anc orages o

a en ons 1nspec e

pursuant to Specifi- ;

cation 4.6. 1.6. 1 and the adjacent concrete surfaces shall be demonstrated by determining through visual snspection that no unacceptable levels of corrosion exist on the end anchorages and no unacceptable cracking exists in the concrete adjacent to the end anchorages.

Determination of acceptance levels shall be by engineering evaluation of the areas in question.

If unacceptable conditions are found, the tendons inspected during the previous surveillance shall be examined to determine whether the corrosion levels or concrete cracking have increased since the previous surveillance.

Inspection of adjacent concrete surfaces shall be per'formed concurrently with the containment tendon surveillance (Technical Specification 4.6.1.6. 1).

4. 6. l. 6. 3 Containment Surfaces.

In accordance with 10 CFR 50, Appendix J.

Section V.

a vssua snspec son of the accessible interior and exterior surfaces of the containment, includin the liner plate, shall be performed during the shutdown for (but prior to each Type A containment leakage rate test (Technical Specificaticn 4.6.1.2 The purpose of this inspect>on shall be to identify any evidence of structural deterioration which may affect containment structural integrity or leaktightness.

The visual inspection shall be general in nature; its intent shall be to detect gross areas of widespread cracking, spalling, gouging, rust, weld degradation, or grease leakage.

The visual examination may include the utHiza'tion of binoculars or other optical devices.

Corrective actions taken, and recording of structural deterioration'and corrective actions, shall be in accordance with 10 CFR 50, Appendix J, Section V. A.

Records of previous inspections shall be reviewed to verify no apparent changes in appearance.

The first inspection performed will form the baseline for future surveillances.

TURKEY POINT - UNITS 3 8c 4 3/4 6-10 AMENDMENT NOS. 149 AND 144

1 t

CONTAINMENT SYSTEMS 3/4.6:3 EMERGENCY CONTAINMENT FILTERING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3 Three emergency containment filtering units shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

With one emergency containment filtering unit inoperable, restore the inoperable filter to OPERABLE status within 7 days 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 in COLO SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.6.3 Each emergency containment filtering unit,shall-be demonstrated OPERABLE:

a 0 b.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 15 minutes; At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber

housings, or (2) following operational exposure of filters to effluents from painting, fire, or chemical release or (3) after every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation by:

1)

Performance of a visual inspection for foreign material and gasket deterioration, and verifying that the filtering unit satisfies the in-place penetration and bypass leakage testing acceptance criteria of greater than or equal to 99K removal of DOP and halogenated hydrocarbons at the system flow rate of 37,500 cfm %10K; 2)

Verifying within 31 days after removal, that a laboratory analy-sis of a representative carbon sample obtained in accordance with applicable portions of Regulatory Position C.6.b of Regula-tory Guide 1. 52, Revision 2, March 1978, and performed in accordance with ANSI N-510-1975, meets the acceptance criteria of greater than 99.9X removal of elemental iodine; 'and that any charcoal failing to meet this criteria be replaced with charcoal that meets or exceeds the criteria of position C.6.a of Regulatory Guide 1.52, Rev. 2; and 3)

Verifying a system flow rate of 37,500 cfm %10X and a pressure drop across the HEPA and charcoal filters of less than 6 inches water gauge during system. operation when tested in accordance with ANSI N510-1975; TURKEY POINT - UNITS 3 8A 4 3/4 6-15 AMENDMENT NOS. 149 AND 144

~

> PLANT SYSTEMS 3/4.7.9 FIRE RATED ASSEHBLIES LIMITING CONDITION FOR OPERATION 3.7.9 All fire rated assemblies (walls, floor/ceilings, and other fire barriers) separating safety-related fire areas or separating portions of redun-dant systems important to safe shutdown within a fire area and all sealing devices in fire rated assembly penetrations (fire doors, fire windows fire

dampers, cable, piping, fire barrier penetration

seals, and ventilation duct penetration seals) shall be OPERABLE.

APPLICABILITY: At al 1 times.

ACTION:

a.

b.

With one or more of the above required fire rated assemblies and/or sealing devices inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either establish a

continuous fire watch on at least one side of the affected assembly, or verify the OPERABILITY of fire detectors on at least one side of the inoperable assembly and establish an hourly fire watch patrol.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREHENTS 4.7.9. 1 At least once per 18 months the above required fire rated assemblies

.and penetration sealing devices shall be verified OPERABLE by performing a visual inspection of:

a.

The exposed surfaces of each fire rated assembly, b.

Each fire window/fire damper and associated

hardware, and c.

At least 10K of each type of sealed penetration.

If apparent changes in appearance or abnormal degradations are found, a visual inspection of an additional 10'f each type of sealed penetration shall be made.

This inspection process shall continue until a 1(C sample with no apparent changes in appearance or abnormal degradation is found.

Samples shall be selected such that each penetration will be inspected every 15 years.

TURKEY POINT - UNITS 3 8L 4 3/4 7-33 AHENDHENT NOS. 149 AND 144

S a

REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 The water level shall be maintained greater than or equal to elevation 56' 10'he spent fuel storage pool."

APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

With the requirements of the above specification not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

"The requirements of this specification may be suspended for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months hours to perform maintenance provided a safety evaluation is prepared prior to suspension of the above requirement and all movement of fuel assemblies and crane operation with loads in the fuel storage areas are suspended.

If the level is not restored within 7 days, the NRC shall be notified within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

TURKEY POINT - UNITS 3 8L 4 3/4 9-12 AMENDMENT NOS.

149 AND 144

REFUELING OPERATIONS 3/4.9.12 HANDLING OF SPENT FUEL CASK LIMITING CONDITION FOR OPERATION 3.9.12 The handling of spent fuel cask shall be limited to the following conditions:

1)

The spent fuel cask shall not be moved into the spent fuel pit until all the spent fuel in the pit has decayed for a minimum of one thousand five hundred twenty-five (1,525) hours.

2)

Only a single element cask may be moved into the spent fuel pit.

3)

A fuel assembly shall not be removed from the spent fuel pit in a shipping cask until it has decayed for a minimum of one hundred twenty (120) days.

APPLICABILITY: During movement of spent fuel cask in the spent fuel storage area.

ACTION:

With the requirement of the above specification not satisfied, suspend all movement of the spent fuel cask within the spent fuel storage area.

SURVEILLANCE RE UIREMENTS 4.9. 12. 1 The followinq required decay times of the spent fuel assemblies shall be determined prior to the movement of a spent fuel cask by ver'ification of date and time the spent fuel assemblies were placed into the spent fuel pit:

'a 0 b.

1525 hours0.0177 days 0.424 hours 0.00252 weeks 5.802625e-4 months of decay of all spent fuel assemblies in the spent fuel pit for movement of a spent fuel cask into the spent fuel pit.

120 days of decay of the spent fuel assembly in the spent fuel. cask prior to removal of the spent fuel cask from the spent fuel pit.

4.9.12.2 Prior to any operations involving spent fuel cask movement into the spent fuel pit, verify only a single element cask will be moved into the spent fuel pit.

4.9.12.3 The spent fuel cask crane interlock shall be within 7 days of crane operation and at least once per time between tests; specification 4.0.2 does not apply being used to maneuver the spent fuel cask.

demonstrated OPERABLE 7 days (7 days is maximum here) when the crane is TURKEY POINT - UNITS 3 84 4 3/4 9-13 AMENDMENT NOS.

149 AND 144

I

REFUELING OPERATIONS 3/4.9. 14 SPENT FUEL STORAGE LIMITING CONDITION FOR OPERATION 3.9.14 The following conditions shall apply to spent fuel storage:

a.

The maximum enrichment loading for the fuel assemblies in the spent fuel racks shall not exceed 4.5 weight percent of U-235.

b.

The minimum boron concentration in the Spent Fuel Pit shall be 1950 ppm.

c.

Storage in Region II of the Spent Fuel Pit shall be further restricted r

by burnup and enrichment limits specified in Table 3.9-1.

APPLICABILITY: At all times when fuel is stored in the Spent Fuel Pit.

ACTION:

a.

Mith either condition a, or c not satisfied, suspend movement of additional fuel assemblies into the Spent Fuel Pit and restore the spent fuel storage configuration to within the specified conditions.-

b.

Mith boron concentration in the Spent Fuel Pit less than 1950 ppm, suspend movement of spent fuel in the Spent Fuel Pit and initiate action to restore boron concentration to 1950 ppm or greater.

SURVEILLANCE RE UIREMENTS 4.9. 14 The boron concentration of the Spent Fuel Pit shall be verified to be 1950 ppm or greater at least once per month.

TURKEY POINT - UNITS 3 8A 4 3/4'-15 AMENDMENT NOS. 149 AND 144

P, P

RADIOACTIVE EFFLUENTS EXPLOSIVE GAS MIXTURE LIMITING CONDITION FOR OPERATION 3.11.2.5 The concentration of oxygen in the GAS DECAY TANK SYSTEM (as measured in the inservice gas decay tank) shall be limited to less than or equal to 2X by volume whenever the hydrogen concentration exceeds 4X by volume.

APPLICABILITY: At all times.

ACTION:

a.

b.

C.

kith the concentration of oxygen in the inser vice gas decay tank greater than 2X by volume but less than or equal to 4X by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Mith the concentration of oxygen in the inservice gas decay tank greater than 4X by volume and the hydrogen concentration greater than 4X by volume, immediately suspend all additions of waste gases to the gas decay tanks and reduce the concentration of oxygen to less than or equal to 4X by volume, then take ACTION a.,

above.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS

4. 11.2.5 The concentrations of hydrogen and oxygen in the inservice gas decay tanks shall be determined to be within the above limits by continuously*

monitoring the waste gases in the inservice gas decay tank with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-8 of Specification 3.3.3.6.

"Mhen continuous monitoring capability is inoperable, Table 3.3-8 allows the use of grab samples.

TURKEY POINT - UNITS 3 8c 4 3/4 11-15 AMENDMENT NOS. 149 AND 144

LIMITING CONDITION FOR OPERATION ACTION Continued C.

With milk or broad leaf vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCH.

The specific locations from which samples were unavailable may then be deleted from the monitoring program.

Pursuant to Specification

6. 14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure(s) and table for the ODCM reflecting the new location(s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location(s) for obtaining samples.

d.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.12.1 The radiological environmental monitorinq samples shall be collected pursuant to Table 3. 12-1 from the specific locations given in the table and figure(s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3. 12-1 and the detection capabilities required by Table 4. 12-1.

TURKEY POINT - UNITS 3 8( 4 3/4 12-2 AMENDMENT NOS.

149 AND 144

REFUELING OPERATIONS BASES SPENT FUEL STORAGE (Continued) in Region II cell locations, strict controls are employed to evaluate burnup of the spent fuel assembly.

Upon determination that the fuel assembly meets the burnup requirements of Table 3.9-1, placement in a Region II cell is authorized.

These positive controls assure the fuel enrichment limits assumed in the safety analyses will not be exceeded.

TURKEY POINT - UNITS 3 4 4 B 3/4 9-4 AMENDMENT NOS. 149 AND 144

I i

DESIGN FEATURES 5.6 FUEL STORAGE

5. 6. 1 CRITICALITY 5.6.1.1 The spent fuel storage racks are desiqned to provide safe subcritical storage of fuel assemblies by providing sufficient center-to-center spacing or a combination of spacing and poison and shall be maintained with:

a.

A k equivalent to less than or equal to 0.95 when flooded with unbNted water, which includes a conservative a1lowance in region 1

of 0.971 hk/k and in region 2 of 1.96X hk/k for uncertainties for two region fuel storage racks.

b.

A nominal 10.6 inch center-to center distance for Region 1 and 9.0 inch center-to-center distance for Region 2 for two region fuel storage racks.

c.

The maximum enrichment loading for fuel assemblies is 4.5 weight percent of U-235.

5.6.1.2 The racks for new fuel storage are designed to store fuel in a safe subcritical array and shall be mainta>ned with:.

a.

A nominal 21 inch center-to-center spacing to assure k ff equal to or less than 0.98 for optimum moderation conditions and equal to or less than 0.95 for fully flooded conditions.

b.

Fuel assemblies placed in the New Fuel Storage Area shall contain no more than 4.5 weight percent of U-235.

TURKEY POINT - UNITS 3 L 4 5-5 NENDMENT NOS. 149 AND 144

~ J

~

DESIGN FEATURES 5.6.1.3 Credit for burnup is taken in determining placement locations for spent fuel in the two-region spent fuel racks.

Administrative controls are employed to evaluate the burnup of each spent fuel assembly stored in areas where credit for burnup is taken.

The burnup of spent fuel is ascertained by careful analysis of burnup history, prior to placement into the storage loc-cations.

Procedures shall require an independent check of the analysss of suitability for storage.

A complete record of such analysis is kept for the time period that the spent fuel assembly remains in storage onsite.

DRAINAGE 5.6.2 The spent fuel storage pit is designed and shall be maintained to prevent inadvertent draining of the pool below a level of 6 feet above the fuel assemblies in the storage racks.

CAPACITY 5.6.3 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1404 in two region storage racks 5.7 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.7. 1 The components identified in Table 5.7-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.7-1.

TURKEY POINT - UNITS '3 5 4 5-6 AMENDMENT NOS. 149 AND 144

v'

ADMINISTRATIVE CONTROLS

6. 1 RESPONSIBILITY 6.1.1 The Plant Manager - Nuclear shall be responsible for overall unit opera-tion of both units and shall delegate in writing the succession to this responsibility during his absence.

6. 1.2 The Plant Supervisor - Nuclear (or during his absence from the control

room, a designated individual) shall be responsible for the control room com-mand function.

A management directive to this effect, signed by the Site Vice President shall be reissued to all station personnel on an annual basis.

6. 2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATION 6.2. 1 An onsite and an offsite organization shall be established for facility.

operation and corporate management.

The onsite and offsite organization shall'.

include the positions for activities affecting the safety of the nuclear power plant.

aO

. b.

C.

d.

e.

Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to, and including all operating organization positions.

Those relationships shall be documented and updated, as appropriate, in the form of organizational charts.

These organiza-tional charts will be documented in the Topical guality Assurance Report and updated in accordance with 10 CFR 50.54(a)(3).

The President-Nuclear Division shall have corporate responsibility for overall plant nuclear safety, and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety.

The Plant Manager-Nuclear shall be responsible for overall plant safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

Although the individuals who train the operating staff and those who carry out the quality assurance functions may report to the appropriate manager onsite, they shall have sufficient organizational freedom to be independent from operating pressures.

Although health physics individuals may report to any appropriate manager onsite, for matters relating to radiological health and safety of employees and the public, the health physics manager shall have direct access to that onsite individual having responsibility for overall unit management.

Health physics personnel shall have the authority to cease any work activity when worker safety is jeopardized or in the event of unnecessary personnel radiation exposures.

TURKEY POINT - UNITS 3 8c 4 6-1 AMENDMENT NOS249 AND 144

I lip.

"ADMINISTRATIVECONTROLS

. RESPONSIBILITIES Continued) e.

g.

h.

k.

Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the President-Nuclear Division and to the Chairman of the Company Nuclear Review Board; Review of all REPORTABLE EVENTS Review of reports of significant operating abnormalities or deviations from normal and expected performance of plant equipment or systems that affect nuclear safety.

Performance of special reviews, investigations, or analyses and reports thereon as requested by the Plant Manager - Nuclear or the Chairman of the Company Nuclear Review Board; Review of the Emergency Plan and implementing procedures and submittal of recommended changes to the Chairman of the Company Nuclear Review Board; Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL; Review of any accidental, unplanned, or uncontrolled radioactive release including the preparation of reports covering evaluation, recommendations, and disposition of the corrective action to pre-vent recurrence and the forwarding of these reports to the President-Nuclear Division and to the Chairman of the Company Nuclear Review Board.

6.5.1.7 The PNSC shall:

Recommend in writing to the Plant Manager - Nuclear approval or disapproval of items considered under Specification 6.5.1.6a.

through

d. prior to their implementation and items considered under Specifica-tion 6. 5. 1.6i through k.

b.

Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Plant Hanager-Nuclear, President-Nuclear Division and the Company Nuclear Review Board of disagreement between the PNSC and the Plant Manager-Nuclear; however, the Plant Hanager - Nuclear shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.

TURKEY POINT - UNITS 3 8 4 6-7 AMENDMENT NOS.

149 AND 144

E

ADMINISTRATIVE CONTROLS

'ROCEDURES AND PROGRAMS Continued b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Operator license on the unit affected; and c.

The change is documented, reviewed in accordance with Specification 6.5.3 and approved by the Plant Manager-Nuclear or the department head of the responsible department wsthin 14 days of implementation.

6.8.4 The following programs shall be established, implemented, and maintained:

a.

Primar Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels.

The systems include the Safety Injection

System, Chemical and Volume Control System, and the Containment Spray System.

The program shall include the following:

(1)

Preventive maintenance and periodic visual inspection requirements, and (2)

Integrated leak test requirements for each system at refueling cycle intervals or less.

b.

In-Plant Radiation Monitorin 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 per sonnel, (2)

Procedures for monitoring, and (3)

Provisions for maintenance of sampling and analys'.s equipment.

C.

Secondar Water Chemistr A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation.

This program shall include:

(1)

Identification of a sampling schedule for the critical variables and control points for these variables, (2)

Identification of the procedures used to measure the values of the critical variables, TURKEY POINT - UNITS 3 8E 4 6-14 AMENDMENT NOS.,149 AND 144

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS Continued (3)

Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, (4)

Procedures for the recording and management of data, (5)

Procedures defining corrective actions for all off-control point chemistry -conditions, and (6)

A procedure identifying:

(a) the authority responsible for the interpretation of the data, and (b) the sequence and tim-ing of administrative events required to initiate corrective action.

d.

Post-Accident Sam lin 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 acci-dent conditions.

The program shall include the following:

(1)

Training of personnel, (2)

Procedures for sampling and analysis, and (3)

Provisions for maintenance of sampling and analysis equipment.

6. 9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the U.S.

Nuclear Regulatory Commission, Document Control Desk, Mashington, DC pursuant to 10 CFR 50.4.

STARTUP REPORT 6.9.1. 1 A summary report of plant startup and power escalation testing shall be submitted followinq:

(1) receipt of an Operating License, (2) amendment to the license involvang a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the

nuclear, thermal, or hydraulic performance of the unit.

TURKEY POINT - UNITS 3 8a 4 6-15 AMENDMENT NOS. 149 AND

>44

, ADMINISTRATIVE CONTROLS

'NNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT*

6.9.1.3 Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to Hay 1 of the following year.

The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, with operational

controls, as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of the Land Use Census required by Specification

3. 12. 2.

The Annual Radiological Environmental Operating Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation

Manual, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some indivi-.

dual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

The reports shall also include the following:

a summary description of the Radiological Environmental Monitoring Program; at least two legible maps**

covering all sampling locations keyed to a table giving distances and direc-tions from the centerline of one reactor; the results of licensee par ticipa-tion in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Specifi-cation 3. 12.3; reasons for not conductinq the Radiological Environmental Moni-toring Program as required by specificatson 3.12.1, and discussion of all deviations from the sampling schedule of Table 3. 12-1; discussion of environ-mental sample measurements that exceed the reporting levels of Table 3. 12-2 but are not the result of plant effluents, pursuant to ACTION b. of Specifi-cation 3.12.1; and discussion of all analyses in which the LLD required by Table 4. 12-1 was not achievable.

A single submittal may be made for a multiple unit station.

- One map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

TURKEY POINT - UNITS 3 5 4 6-17 AMENDMENT NOS. 149 AND 144

ADMINISTRATIVE CONTROLS RADIATION PROTECTION PROGRAM Continued) maintained, and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA

6. 12.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the "control device" or 'alarm signal" required by paragraph 20.203(c),

each high radiation area, as defined in 10 CFR Part 20, in which'the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

Indi-viduals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protec-tion procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

A radiation monitoring device which continuously indicates the

. radiation dose rate in the area; or b.

A radiation monitoring device which continuously integrates the

'adiation dose rate in the area and alarms when a preset integrated

.dose-is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been estab-lished and personnel have been made knowledgeable of them; or l

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for pro-viding positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Health Physics Shift Supervisor in the RWP.

6. 12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift supervisor on duty and/or health physics supervision.

Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area.

In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection proce-dures to provide positive exposure control over the activities being performed within the area.

TURKEY POINT - UNITS 3 8c 4 6-22 AMENDMENT NOS.

149 AND 144

ML17348B145

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