Text

Proposed Tech Specs Re Containment Integrity,Esfas, Radiation Monitoring Instrumentation & ESFAS Instrumentation Trip Setpoints
	ML17345A854
Person / Time
Site:	Turkey Point
Issue date:	05/28/1991
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17345A855	List: L-91-142, Application for Amends to Licenses DPR-31 & DPR-41,revising Tech Specs to Remove Outdated Matl,Incorporate Administrative Clarifications & Correct Typos for Consistency w/NUREG-0452 Re STS for Westinghouse PWRs; ML17345A854;
References
	NUDOCS 9106040012
	Download: ML17345A854 (68)
	v • d • e

ATTACHMENT 3 PROPOSED TECHNICAL SPECIFICATION Marked-up Technical Specifications

Pages, 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 6-15 6-17 6-2g 9106040012 910528 PDR ADOCK 05000250 P

PDR

DEFINITIONS t

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 s cur in their closed positions, except as provided in 53 5.5.:<<3. 6.

6 6

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 1-131 1.10 DOSE EQUIVALENT 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

. 11 E shall be the average (weighted in proportion to the concentration of each adionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (HeV/d) for the radionuclides in the sample

isotopes, other than iodines, with half lives'reater than 30 minutes, making up at least 95 percent of the total non-iodine activity in the coolant.

TURKEY POINT " UNITS 3

8 4.

1= 2.

AMENDMENT-NOS. 1.37AND$'32'

m TABLE 2.2-1 (Continued)

TABLE NOTATIONS (Continued)

CD NOTE 3:

(Continued) lcc 0.00068/'F for T > T" ~~ = 0 for T ~ T" As defined in Note 1, Indicated T

at RATED THERtlAL POWER (Calibration temperature for hT avg instrumentation, c 574.2'F),

f (ai)

As defined in Note 1, and As defined in Note l.

NOTE 4:

CD (This note number is not used.)

m CD cD 8 If no allowable value is specified as indicated by I'llowable value.

'V CD

], the trip set point shall also be the

t

~ ~

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 +

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 deinand 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 Iw s<RT

~a~ a~si ~n v

~ <

>~~

>~o~p P

)gag /%DE.

IMI~4OV t

+.t-

~

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

"*See Special Test Exceptions Specification

3. 10.

TURKEY POINT - UNITS 3 8

4 3/4 1-23 AMENDMENT NOS. 137AND 132

f p~ ~ ~gpJ l

go km o t 's t I P o s +

g g,> q

<<<q"')g J hfA'w.3.f"=-

-Y 9 klan.VA,)

e t $ y~QQ ca~/

fl~rj'(.-

j pf.f.']~

)

C='3

C:

PCm TABLE 3. 3-2 Continued ENGINEERED.SAFETY FEATURES ACTUATION SYSTEH INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP HINIHUH CHANNELS OPERABLE APPLICABLE MODES ACTION f.

Steam Line FlowHigh 2/steam line Coincident with:

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

in any two in any two steam lines steam lines 15 Steam Generator Pressure-"Low or T

Low 2.

Containment Spray 1/steam generator 1/loop eahsAakof'/steam

+Inc 1/steam 1, 2, 3" in any two generator steam in any two earn +H~

'Sew~~4c rs 1/loop n any oop in any 1, 2, 3" two loops two loops 15 15 a.

Automatic Actuation Logic and Actuation Relays 1, 2, 3, 4

14 m

CD b.

Containment Pressure High-High Coincident with:

Containment Pressure--

High 1, 2, 3

1, 2, 3

15 15 3.

Containment Isolation CD CA C3 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

I I

4 0

lp d4

C:

Kl ycm TABLE 3. 3-2 Continued ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH INSTRUHENTATION CI FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP HINIHUM CHANNELS OPERABLE APPLICABLE HODES ACTION 4.

Steam Line Isolation (Continued)

CA CsLI d.

Steam Line FlowHigh Coincident with:

Steam Generator Pressure Low or T

Low 2/steam line 1/steam generator 1/loop steam 1/ team lin 1, 2, 3

+WO I+

C4N~

TWa f,'<q~

S4e~

tN'~eS 1/steam 2.

3 generator generator in any two in any two steam +Res steam Ca8r~4er s 1/loop oop in 1, 2, 3

any two any two loops loops 15 15 15 5.

Feedwater Isolation a.

Automatic Actua-tion Logic and Actuation Relays 1,

2 22 b.

Safety-Injection 6.

Auxi1 iary Feedwater888 See Item l. above for all Safety Injection initiating functions and requirements.

a.

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

20

~\\

+"

kg

'V't Ip

'4 oO4y

~ e g>>

$o

TABLE 3.3-3 (Continued)

ENGINEEREO SAFETY FEATURES ACTUATION SYSTEH

~ FUNCTIONAL UNIT go.

Stem Line Isolation'l'(gfj4nnod) b.

Autoaatic Actuation Log)c and Actuation Relays c.

Contaireent Pressure-High-High Coincident with:

Contaiment Pressure-High Stem Line FlowHigh ALLS/ANCE TA N.A.

1 3

Z S

N.A. N.A.

f ]f ]

]f 3

f 3f ]

TRIP sETpolnT ALLSQBLE VALUE N.A.

N.A.

<6.0 psig

] psig

<A function defined f

]

as follows:

A hp corresponding to 0.64 x 10 lbs/hr at OX load increa-ing linearly to a hp corresponding to 3.84 x 10'bs/hr at full load.

<30.0 psig

] psig Coincident with:

Stem Line Pressure"-Low or Tayg Low g 5.

FeedoateP Isolation 4.0 f 3f 3

2.0 1.0

>600 psig

>5434F

>f 3 psig

)542.5'F 5

8 a.

Autoaatic Actuation Logic and Actuation Relays b.

Safety Injection N.A.

see itea 1 N.A N.A.

N.A.

See Itea 1.

above for all Safety Injection Trip Setpoints and Allowable Values.

FUNCTIONAL UNIT 1.

Containment TABLE 3.3-4 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS WL1 llAUW CHANNELS APPLICABLE ALARM/TRIP TO TRIP/ALARM 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 Al1" Part>culate

<6.1xlOsCPM Gaseous See Note 2.

1,2,3,4 NA.

26 for MODES 1, 2, 3, 4

or 27~

DES 5

AND 6 26 a.

Unit 3 Radioactivity -

1 High Gaseous b.

Unit 4 Radioactivity-1 HighGaseous'5.

5xlO-2 ~Ci 28 CC

<2.8x10-2 pCi 28 CC (SPING) or

<l. Ox106CPM (PRMS)

ED C/l

L

INSTRUHENT

- TABLE 3.3-5 (Continued)

ACCIDENT HONITORING INSTRUHENTATION TOTAL NO.

OF CHANNELS HINIHUH CHANNELS OPERABLE APPLI-CABLE HOOES ACTIONS 14.

15.

16.

17.

18.

19.

20.

21.

22.

In Core Thermocouples (Core Exit Thermo-couples)

Containment High Range Area Radiation Reactor Vessel Level Honitoring System Neutron Flux, Backup NIS (Mide Range)

Containment Hydrogen Honitors High Range-Noble Gas Effluent Honitors a.

Plant Vent Exhaust b.

Unit 3-Spent Fuel Pit Exhaust c.

Condenser Air Ejectors d.

Hain Steam Lines RMST Mater Level Steam Generator Mater Level (Narrow Range)

Containment Isolation Valve Position Indication" 4/core quadrant 2

2(1) 2 2/stm.

gen.

1/valve 2/cor qv.~ra<+

1 1(1) 1/stm.

gen.

1/valve 1, 2, 3

31, 32 1, 2, 3

34 1, 2, 3

37, 38 ALL ALL 1, 2, 3

1, 2, 3

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

31, 32 1,

2 35 TABLE NOTATIONS A channel is eight sensors in a probe.

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

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

1.

2.

>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).

f,

INSTRUMENT LOCATION TABLE 3.3-6 FIRE DETECTION INSTRUMENTS TOTAL NUMBER.

OF INSTRUMENTS FIRE ZONE AREA 4

5 910-ll-12-13 "

14-15-16-19-20-21-22-25-26-27-30-40-45-47-54-55-58-59-60-61-62-63-67-68-70-71-72-73-74-75-76.-

79A-81-82-84-Aux. Bldg. Corridor E.

10'hetn.

Drain/Laundry/Shower Tank=Room Laundry/Chemical Drain Tank Roon Pipeway Unit 3 RHR Heat Exchanger Rooe RHR Pump 3A Roota RHR Pump 3B Roow Unit 4 RHR Heat Exchanger Rood RHR Pump 4A Roow RHR Pump 4B Rooe Unit 3 W Elect Penet Rooa Unit 3 S Elect Penet Rooa Instrument Shop Radioactive Laboratory Aux. Bldg. Elect. Equipat.

Rooa Unit 4 N Elect Penet Roow Unit 4 W Elect Penet Rooa Unit 4 Piping and Valve Rooa Unit 3 Piping and Valve Rooa Unit 4 Charging Pump Rooe Unit 4 Component Cooling Water Area Unit 3 Coiponent Cooling Water Area Unit 3 Charging Puap Rooe Aux Bldg Corridor, El; 18'nit 4 Containment Electrical Penet.

Area~

Unit 3 Containment Electrical Penet.

Area"~

Reactor Control Rod Eqpat Rooe - Unit 4 Computer Rooe Reactor Control Rod Eqpat Rooa - Unit 3 4160V Switchgear 4$

4160V Switchgear 4A 4160V Switchgear 3B 4160V Switchgear'A Diesel Generator 3B Diesel Generator 3A Day Tank Rooe 3B Day Tank Rooa 3A Unit 4 Turbine Lube Oil Reservoir North-South Breezeway Unit 4 Hain Transformer Unit 4 Aux Transforeer Area Unit 3 and 4 Aux Feedwater Puip Area (DC Enclosure Bldg.)

~y)"

xy +

Aoo (2/0)

(0/4)

(5/2)

(0/4)

(4/2)

(0/4)

(0/3)

(IJO)

(0/3)

(1/0)

(IJ1)

(1/1)

(1/0)

(0/63 (1/0)

(1/0) xy +

(2/0)

(1/0)

(ll/0)

(5/0)

(2/0)

(5/0)

(2/0)

(5/0)

(IIJO)

(2/0)

(5/0)

(8/0)

(6/0)

(4/0)

(3/0)

(18/0)

(10/0)

(16/0)

(4/0)

(11/0)

(4/0)

(10/0)

(6/0)

(10/0)

(6/0)

(IJO)

(1/0)

(4/0)

(3/0)

TURKEY POINT - UNITS 3 4 4 3/4 3-48 AMEN~NT NOS.138AND 133

8 anA

TABLE 3.3-6 Continued FIRE DETECTION INSTRUMENTS (1/0)

(2/0)

(1/0)

(2/0)

(1/0)

(16/15)

(1/0)

(2/0)

(17/0)

(3/4)

(4/4)

(1/0)

(5/0)

(2/0)

'(2/0)

(5/0)

(2/0)

(N/A)f TOTAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS FIRE ZONE AREA

~y)"

xy +

xy 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 Roea 98.= Cable Spreading Room 101-RPI Inverter and MG Sets 102-Battery Rack48'1/0) 103-Battery Rack 3A

'1/0) 104-RPI Inverter and HG Sets

--'06-Control Roea (1/0) 108A-Train A Inverters 108B-Train B Invcrters 109-Battery Rack 4A (1/0) 110- Battery Rack 3B (1/0) 113-Unit 4 Feedwater Platfora (2/0) 116-Unit 3 Feedwater Platform (2/0) 119-Unit 4 Intake Cooling Mater Pump Area (4/0) 120- Unit 3 Intake Cooling Mater Pump Area (4/0) 132-Control Roon Electrical Chase 133-Diesel Generator 4B (5/5)

(3/0) 134-4160V Switchgear 3D Rooa 135-Diesel Generator 4B Control Panel Rooa 136-Diesel Generator 48 Fuel Transfer Puap 138-Diesel Generator 4A (5/5)

(3/0) 139-4160V Switchgear 4D.Rooa 140- Diesel Generator 4A Control Panel Rooa 141-Diesel Generator 4h Fuel Transfer Pump N/A - 18'evel of the Turbine Area

~ (N/A)f (N/A)f g~wvEW TABLE NOTATIONS x is nueber 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) instrments.

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

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

TURKEY POINT - UNITS 3 4 4 3/4 3-49 AMENDMENT NOS.1 38 AND 133

Q~

')cfpi 8 d'>"l-(gg

~s~.~w./~

(

w 4

>a-a.>BC!'

TABLE 4.3-5 RADIOACTIVE LI UID EFFLUENT HONITORING INSTRUHENTATION SURVEILLANCE RE UIREHENTS INSTRUHENT CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION ANALOG CHANNEL OPERATIONAL TEST l.

Gross Radioactivity Honitors 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 D(3)

D(3)

N.A.

R(2 ~

Q(1)

R(2)

/<4 Q(l)

~pl e~

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>csee'.~em levels above the Alarm/Trip Setpoi nt.

~CcLs uv'6 4 2

The i itial CHANNEL CALIBRATION shall be performed using one or more o the reference standards certified by t e

) or using stand ds that have been obtained from suppliers that participate in measurement assurance activities with 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.

(3)

CHANNEL CHECK shall consist of verifying indication of flow during Periods of release.

CHANNEL CHECK shall be made at least once 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 made.

T~s4 4<

o4

<4awdav Ja CKW Padkmo f,ag g

t. H TmT ).

a,J Qt fp Q

~l P~

~~

6 p,,

5" y

,l~

qp p

+0 t

l t

l jl I

Il l

~ g

~ Q

C:

m TABLE 3'-8 Continued RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION I

C Qe C7 m

INSTRUMENT a \\

b.

Plant Vent Systea (Include Unit 4's Spent Fuel Pool)

Noble Gas Activity Honitor (SPING or PRHS)

Iodine Saapler c.

Particulate Sampler d.

Effluent Systea Flow Rate Heasuring Device e,

Sampler Flow Rate Heasuring Device a.

Noble Gas Activity Honitor b.

Iodine Saapler c.

Particulate Sampler d.

Sampler Flow Rate Heasuring Device 5.

Unit 3 Spent Fuel Pit Building Vent MINIMUM CHANNELS OPERABLE 1

I APPLICABILITY ACTION 47 48

~AO 47 48 48 46 ED CA Cad C7

Q

~p

TABLE 4.

Continued RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT 3.

Condenser Air Ejector Vent System (Continued) e.

Sample Flow Rate Measuring Device 4.

Plant Vent System (Include Unit 4's Spent Fuel Pool)

CHANNEL CHECK SOURCE CHANNEL CHECK CALIBRATION N.A.

ANALOG CHANNEL OPERATIONAL TEST N.A.

MODES FOR WHICH SURVEILLANCE IS RE UIREO a.

Noble Gas Activity Monitor (SPING or PRMS)

M (3Q) 9(2) b.'.

d.

Iodine Sampler Particulate Sampler Effluent System Flow Rate Measuring Device N.A.

N.A.

N.A.,

N.A.

~

N.A.

w N.A.

'k ~

(.~

N.A.

e.

Sampler Flow Rate Measuring Device 0

N.A.

I f~i 8

/

Ni II

(-

p (Q.)

C m

TABLE 4.3-6 Continued RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS C

CA INSTRUMENT 5.

Unit 3 Spent Fuel Pit Building Vent CHANNEL SOURCE CHANNEL CHECK CHECK CALIBRATION ANALOG CHANNEL OPERATIONAL TEST MODES FOR MHICH SURVEILLANCE IS RE UIRED a 0 Noble Gas Activity Monitor R(3)

S(2)

CARI CA C) b.

Iodine Sampler c.

Particulate Sampler d.

Sampler Flow Rate Measuring Device D

N.A.

N;A.

N.A.

TABLE NOTATION N.A.

N.A.

N.A; N.A.

CD m

ID CA At all times.

During GAS DECAY TANK SYSTEM operation.

Applies during MODE 1, 2, 3 and 4.

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

(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 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 PRHS only) and in the computer room (for SPING only).

(3)

The initial CHANNEL CALIBRATION shall be performed using ne or more of the reference standards certified y

e

) or using stand ds that have been obtained from suppliers that participate in measurement assurance activities with AQUA.

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 calib ation shall be used.

~P~ f~~P,L tWSVlaa~g OF'~AtVC~gy 5 A<1 T< CHW<<<<"f (~XIV

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 5 4 3/4 3-61 NENDNENT NOS 137AND 1

1

TABLE 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER Unit 3 Unit 4 FUNCTION High-Head Safety Injection Check Valves 3"874A 3-875A 3-873A 3-874B 3-875B 3-873B 3-875C 3-873C 4"874A 4-875A 4-873A 4"874B 4"875B 4-873B 4-875C 4-873C Loop A, hot 1 eg cold leg cold leg Loop B, hot leg cold leg cold leg Loop C, cold leg cold leg 87(n A Residual Heat Removal Line Check Valves 3"876B 3"8760 4"876A 4"876E 4-876B 4-8760 Loop A, cold leg Loop B, cold leg 3"876C 4"876C 3"876E MOV3-750 MOV4-750 MOV3-751 MOV4-751 Loop C, cold leg Loo A

hot le to HR oop o

eg to R

R C

ACCEPTABLE LEAKAGE LIMITS 1.

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

2.

Leakage rates greater than 1.0 gpm but less than or equal to 5 '

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 SO% 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 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 4 3/4 4"22 AMENDMENT NOS. 137ANO 132

p$ eb5 4s

AND TABLE 4.4-4 REACTOR COOLANT SPECIFIC ACTIVITY SAHPLE ANO ANALYSIS PROGRAH TYPE OF HEASUREHENT AH% ANALYSIS 1.

Gross Radioactivity Determination SAHPLE ANO ANALYSIS FRE UENCY At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

HOOES IN WHICH SAHPLE ANO ANALYSIS RE UlRED 1, 2, 3, 4

2.

3.

Tritium Activity Determination

'Isotopic Analysis for POSE EQUIVALENT I-131 Concentration 1 per 7 days.

1 per 14 days.

1, 2, 3, 4

m C7 m

Cl CA C7 4.

Radiochemical Isotopic Determination Including Gaseous Activity 5.

Radiochemical for K Determination 6.

Isotopic Analysis for Iodine Including I-131, 1-133, and I-135 Honthly 1 per 6 months".

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 EQUIVALENT 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 THERHAL POWER change exceeding 15K of the RATED THERHAL POWER within a I-hour period.

1, 2, 3, 4

III, 2II, 38, 48, 5N 1, 2, 3

eQ

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued) d.

Assuring that the observed lift-offforce for each tendon exceeds the minimum required lift-offforce.

Required lift-offforces shall

'e 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. 1.6.2 End Anchora es and Ad 'acent Concrete Surfaces.

The structural integrity o the end anchorages o

a tendons 1nspecte pursuant, to Specifi-cation 4.6. 1.6. 1 and the adjacent concrete surfaces shall be demonstrated by determining through visual inspection 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 performed concurrently with the containment tendon surveillance (Technical Specification 4.6. 1.6. 1).

4. 6. 1. 6. 3 Containment Surfaces.

In accordance with 10 CFR 50, Appendix J.

Section V.

a visua inspection of the accessible interior and exterior surfaces of the containment, including the liner plate, shall be performed during the shutdown for (but prior to ach Type A containment leakage rate test (Technical Specification 4.6.1.

The purpose of this inspection shall be to identify any evidence of s ructural 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 utilization 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 8

'4 3/4 6-10 AMENDMENT NOS. 137AND 132

CONTAINMENT SYSTEMS 3/4.6.3'MERGENCY CONTAINMENT FILTERING SYSTEM LIMITING CONDITION FOR OPERATION 3.6.3 Three emergency containment filtering units shall be OPERABLE.

APPLICAEBLITY:

MOOES I, 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 COLD 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 ~

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) 3)

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 of Regulatory Guide 1.52, Rev.

2; and sk li re~J C. l.. ~

Verifying a system flow.rate of 37,500 cfm 110X 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 8( 4 3/4 6-15 AMENDMENT NOS. 137AND 132

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PLANT SYSTEMS 3/4.7 '

FIRE RATED ASSEMBLIES LIMITING CONDITION FOR OPERATION 3.7.9 All f r rated assemblies walls floor/ceilin s fire barrier enetra ion seals and other ire arriers separating safety-re a ed fire areas or separating portions of redundant 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, and ventilation duct penetration seals) shall be OPERABLE.

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APPLICABILITY: At al 1 times.

ACTION:

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.

b.

The provisions of Specification 3.0.3 are not applicable.

C.

SURVEILLANCE RE UIREMENTS 4.7.9.1 At least once per 18 months the above required fire rated assemblies and penetration sealin'g devices shall be verified OPERABLE by performing a

visual inspection of:

a.

b.

C.

The exposed surfaces of each fire rated

assembly, Each fire window/fire damper and'associated

hardware, and 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 10K of each type of sealed penetration shall be made.

This inspection process shall continue until a 10K 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 AMENDMENT NOS. 137AND 132

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REFUELING OPERATIONS 3/4.9. 11 WATER LEVEL -

STORAGE POOL LIMITING CONDITION FOR OPERATION 3.9.11 T

water level shall be mainta'd greater than or equal to elevation 56' 10 the spent fuel storage pool.

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APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

a ~

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.

I II

~he requirements of this specification may be suspended for more than a hours 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 5 4 3/4 9-12 AMENDMENT NOS.137 ANO 132

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'EFUELING 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:

2)

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

'nimum of one thousand five hundred twenty-five (1,525) hours.

~Wc.Ev8 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 e

movement of the spent fuel cask within the spent fuel storage area.

SURVEILLANCE RE UIREMENTS 4.9.12.1 The following required decay times of the spent fuel assemblies shall be determined prior to the movement of a spent fuel cask by verification of date and time the spent fuel assemblies were placed into the spent fuel pit:

a.

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.

b.

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

4. 9. 12.2 Prior,-',to any operations involving spent fuel spent fuel pit;

,'verify only a single element cask will fuel pit.

2.3 The spent fuel cask crane interlock shall be t in 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.

in the spent fuel cask the spent fuel pit.

cask movement into the be moved into the spent demonstrated OPERABLE 7 days (7 days is maximum here) when the crane is TURKEY POINT - UNITS 3 8 4 3/4 9"13 AMENDMENT NOS.137 AND 132

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REFUELING OPERATIONS 3/4.9. 14 SPENT FUEL STORAGE LIMITING CONOITION FOR OPERATION 3.9. 14 The following conditions shall apply to spent fuel storage:

a.

fAAe maximum enrichment loading for the fuel assemblies in the spent fuel racks sha1144e 4.5 wei ht percent of U-235.

Cxc.e b.

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

c.~

Storage in Region II of the Spent Fuel Pit shall be further restricted by burnup and enrichment limits specified in Table 3.9-. 1.

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

e--

Ri fkRv" al Witt@an~ conditionii a,<c ee-d 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.

With 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 8( 4 3/4 9-15 AMENOMENT NOS.137 AND 132

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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.

With the concentration of oxygen in the inservice 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 .

With 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.

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

8. B, E.co 9.3. S "When continuous monitoring capability is inoperable, Table allows the use of grab samples.

TURKEY POINT - UNITS 3 8 4 3/4 11-15 AMENDMENT NOS.137 AND 132

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7'ADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION ACTION C.

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DESIGN FEATURES

5. 6 FUEL STORAGE
5. 6. 1 CRITICALITY 5.6. 1. 1 The spent fuel storage racks are designed 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 k f equivalent to less than or equal to 0.95 when flooded with unbar/ted water, which includes a conservative allowance in region 1
of 0. 97K Dk/k and in region 2 of 1. 96K hk/k for uncertainties for two region fuel storage racks.
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.
e 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 maintained 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 8
4 5-5 AMENOMENT NOS3.37 AND )32
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DESIGN FEATURES
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g 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 analysis 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
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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
8 4
5-6 AMENDMENT NOS, 137AN0132
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ADMINISTRATIVECONTROLS
6. 1 RESPONSIBILITY
6. 1. 1 The Plant Manager " Nuclear shall be responsible for overall unit opera-tion of 5oth units and shall delegate in writing the succession to this responsi&ility 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 0
6.2.1 An operation include t plant.
b.
C.
d.
onsite and an offsite organization shall be established for facility and corporate management.
The onsite and offsite organization shall he positions for activities affecting the safety of the nuclear power Lines of authority, res onsibility and communication shall be established and define from the highest management levels through intermediate levels to including all operating organization posi-tions.
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.
e.
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 )eopardized or in the event of unnecessary personnel radiation exposures.
TURKEY POINT - UNITS 3 & 4 6-1 AMENDMENT NOS.142 AND 1,37
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ADMINISTRATIVE CONTROLS RESPONSIBILITIES Continued) e.
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 b.
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.
WSQg-y Provide written notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Plant Manager-Nuclear, President-Nuclear Division and the Company Nuclea~
w Board, of disagreement between the PNSC and the Plant Manager~Nuclear; however, the Plant Manager - Nuclear shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1.
TURKEY POINT - UNITS 3 4'4 6-7 AMENDMENT NOS.1 42 AND 137
(-'s A
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7 ADMINISTRATIVE CONTROLS PROCEDURES 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 within 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 1ow as practical levels.
The systems include the Safety Injection
System, Chemical and Volume Control System, and the Containment Spray System.
The program shall inc1ude the fo11owing:
(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 personnel, (2)
Procedures for monitoring, and (3)
Provisions for maintenance of sam ling and analysis equipment.
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C he~)'S4 pr am or moni orang o
secon ary wa er 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, h
(2)
Identification of the procedures used to measure the values of the critical variables, TURKEY POINT - UNITS 3 & 4 6-14 AMENDMENT NOSJ37 AND 132
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ADMINI STRATIVE 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 sampl ing
. and anal ys is, 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, Washington, 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 following; (1) receipt ot an Operating Li,cense, (2) amendment to the license involving 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 8
4 6-15 AMENDMENT NOS. 137AND 132
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AOMINISTRATIVE CONTROLS ANNUAL 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 May 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 I, 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 participa-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 conducting the Radiological Environmental Moni-toring Program as required by specification
3. 12. l, and discussion of all deviations from the sampling schedule of'able
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 8
4 6-17 AMENDMENT NOS. 137AND 132
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ADMINISTRATIVE CONTROLS 7t A,Q~g'fiog
'PPM'VE CVtcsA 'pg.~~ gAWL 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 l000 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.
A radiation monitoring device which continuously indicates the radiation dose rate in the area; or b.
A radiation monitoring device which continuously integrates the radiation 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 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 shi ft 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 8L 4 6-22 AMENDMENT NOS. 137AND 132
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