ML17215A400

Proposed Tech Specs Reflecting Changes in Offsite Organization & Updating 10CFR50.72 & 50.73 Reporting Requirements

ML17215A400
Person / Time
Site:	Saint Lucie
Issue date:	05/15/1984
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17215A399	List: ML17215A398 ML17215A400
References
NUDOCS 8405230189
Download: ML17215A400 (71)

Text

INO EX DEFINITIONS SECTION PAGE

1. 0 DEFINITIONS .-

.1 ACTIONe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ e ~ ~ ~ ~ ~ ~ ~ ~ e e ~ ~ ~ ~ ~ 1" 1 1.2 AXIAL SHAPE INDEX......................................... 1-1 1.3 AZINJTHAL PSKR TILT............................;........... l-l

1. 4 CHANNEL CALIBRATION............................-............. l-l 1 .5 CHANNEL CHECKe ~ ~ ~ ~

~ ~ ~ ~ ~

'e

~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ l-l

1. 6 CHANNEL FUNCTIONAL .TEST..................................... 1-2

l. 7 CONTAINMENT VESSEL INTEGRITY...................-............. '-2

l. 8 CONTROLLED LEAKAGE..................................... '..... 1-2 1.9 CORE ALTERATION............................................. 1-2 1.10 DOSE EQUIVALENT I-131....................................... 1-3 1.11 K-AVERAGE DISINTEGRATION ENERGY............................. 1-3 1.12 ENGINEERED'SAFETY FEATURES RESPONSE TINE.................... 1-3 1.13 FREQUENCY NOTATION.:........................................ 1-3 1.14 GASEOUS RNNASTE TREATMENT SYSTEM........................... 1-3 1.15 IDENTIFIED LEAKAGE.............;............................ 1-3 1.15A LOAD FOLLOll OPERATION ...................................-...... 1-3a 1.16 UN TFHPERATURE RCS OVERPRESSURE PROTECTION RANGE........... 1-4 1.17 KEMBER(S) OF THE PUBLIC..................................... 1-4 1.18. OFFSITE DOSE CALCULATION NANUAL (ODN)....'... ~ ~ ~ ~ 1-4 ~

1.19 OPERABLE - OPERABILITY.......................

1.20 OPERATIONAL NDE - ISDE..................................... 1-4 1.21 PHYSICS TESTS......................................'......... 1-4 1.22 PRESSURE BOUNDARY LEDGE................................... 1-5 1.23 PROCESS CONTROL PROGRN..s.................................. 1-5

.24 PURGE e JRGeINGe ~ o o o ~ ~ ~ ~ ~ r ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ s ~ ~ o ~ ~ ~ ~ ~ ~ ~ s o

~ ~ ~ ~ ~ ~ ~ ~ 1-5 lo25 RATED THERMAL ~ ~ oo ~ oo ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ s ~ ~ o ~ ~ o ~ ~ ~ ~ o ~ ~ o ~

%Re ~ 1-5 1.26 REACTOR TRIP SYSTEM RESPONSE TINE....... ~ so ~ ooo ~ ~ ~ ~ ~ ~ o oo ~ ~ ~ ~ 1-5

'l.27 REPORTABLE EVENT ..... - -. ~ ~ ~ ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ o o ~ ~ s ~ 1-5 1.28 SHIELD BUILDING INTEGRITY......................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-5 lo29 SHUTDOO ORGINe ~ ~ ~ o ~ ~ ~ o ~ ~ oe ~ ~ 'o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ 1-6

1. 30 SITE BOUNDARY........................... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 8405230i89 8405i5 PDR ADOCK NCIE - NIT 1 05000335'T.

INDEX DEFINITIONS Continued SECTION PAGE DEFINITIONS (Continued) 1.31 SOURCE CHECK..................................... 1-6 1.32 STAGGERED TEST BASIS........................ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6

1. 33 THERMAL POMER................................ 1-6

1. 34 UNIDENTIFIED LEAKAGE........................................ 1-6 1.35 UNRESTRICTED AREA........................... ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-6 1.36 UNRODDED INTEGRATED RADIAL PEAKING FACTOR- F ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 1-7 1.37 UNRODDED PLANAR RADIAL PEAKING FACTOR - F. 1-7

1. 38 VENTILATION EXHAUST TREATMENT SYSTEM........ 1-7 ST. LUCIE - UNIT 1

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE RESPONSI BI LITYo ~ ~ ~ ~ ~ ~ ~ ~ o p ~ ~ o ~ ~ ~ ~ ~ ~ ~ o' ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-1

6. 2 ORGANIZATION.................................'.. 6-1 6 .2. 1 OFFSITE.........,......................................... 6-1 6 . 2. 2 UNIT STAFF........................*...... 6-1 6.2. 4 SHIFT TECHNICAL ADVISOR................................... 6-6

6. 3 UNIT STAFF UALIFICATIONS.......;........................... 6-6 6~4 TRAININGo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ o ~ o ~ ~ ~ ~ ~ o ~ 6-7 6.5 REVIEW AND AUDIT............... ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ os ~ ~ 6"7 6.5.1 FACILITY REVIEW GROUP. o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ o '6-7

=

FUNCTION.............. ~ o ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ oo ~ ~ ~ ~ ~ o ~ ~ ~ o ~ oo ~ ~ ~ 6-7 COMPOSITION..;............................................ 6-7 A LTERNATES................................;............... 6-7 MEETING FREQUENCY..... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o ~ ~ o ~ ~ ~ o ~ ~ 6-8 QUORUM.......'......... ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 RESPONSIBILITIES............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-8 I

AUTHO R TYo ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ 'o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9 R ECORDSo ~ ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9 ST. LUCIE - UNIT I XIV

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.5.2 COMPANY NUCLEAR REVIEW BOARD. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-9 FUNCTION ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ s ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o a 6-9 COMPOSITION.................. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ e ~ ~ 'e ~ a ~ e ~ ~ ~ e ~ 6-10 A LTERNATES................................................ 6-10 CO NSULTANTSe ~ ~ ~ ~ ~ ~ e o ~ ~ ~ ~ ~ e ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ o ~ ~ ~ ~ o o ~ ~ ~ ~ ~ ~ ~ a ~ 6-10 MEETING FRE/UENCY................................ 6-10 i/0RUMe ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o a o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o o 6-10 REV I9/e ~ ~ e ~ ~ e ~ ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ s o ~ ~ ~ ~ o ~ o e ~ ~ ~ ~ ~ ~ ~ 6-11 AUDITSo ~ e ~ ~ es ~ ~ ~ ~ ~ e a ~ ~ e ~ ~ ~ e ~ ~ o ~ ~ ~ ~ ~ ~ ~ ~ o ~ ~ o o ~ o a ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 6-11 AUTHORITYe ~ ~ o ~ ~ ~ ~ ~ o ~ s ~ ~ ~ ~ o ~ ~ ~ ~ ~ o ~ ~ ~ ~ ~ ~ o ~ e ~ ~ o ~ e ~ o ~ o ~ ~ ~ ~ ~ ~ ~ ~ 6-12 RE CQRDSea ~ as ~ a ~ ~ ~ oooo oooo ~ ~ sac a ~ ~ e ~ o ~ o ~ ~ ~ ~ as ~ ~ ~ ~ o ~ ~ e ~ ~ ~ o ~ ~ 6-12 6.6 REPORTABLE EVENT ACTION .... -.-..-.....---<<-'.e....os...o*. 6-13

6. 7 SAFETY LIMIT VIOLATION. 6-13

6. 8 PROCEDURES AND PROGRAMS.............. ~ ~ ~ ~ o ~ O o ~ \~~o~~~~e ~ ~ ~ ~ ~ 6-13 6.9 REPORTING RE UIREMENTS.............. o o............. 6-16 6.9.1 ROUTINE REPORTS ............................................ 6-16 STARTUP REPORT.....;...................................... 6-16 ANNUAL REPORTS............................................ 6-16 NNTHLY OPERATING REPORTS................................. 6-17 SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT............ 6-20 ANNUAL.RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT........ 6-21 6.9.2 SPECIAL REPORTS............... -. -... -. - .. - - - - - - - - -.... -. --

6. 10 RECORD RETENTIONe ~ ~ ~ ~ ~ ~ ~ ~ ~ o ~ aso ~ ~ ~ ~ ~ ~ os ~ ~ ~ os ~ o ~ ~ os ~ ~ ~ ~ ~ ~ ~ a o 6-22

6. 11 RADIATION PROTECTION PROGRAM............................... 6-23 6.12 HIGH RADIATION AREA.......... - - .. -. - - - ..- 6-24 ST NCIE - OIIT I XV

INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.13 PROCESS CONTROL PROGRAM.........................;.......... 6-25 6.14 OFFSITE DOSE CALCULATION MANUAL............................ 6-25 6.15 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS AND SOLID WASTE TREATMENT SYSTEMS.................................... 6-26 ST. LUCIE - UNIT I XVI

1. 0 DEFINITIONS .

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.

ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.

AXIAL .SHAPE INDEX 1.2 The AXIAL SHAPE INOEX (Y ) is the power level detected by the lower excore nuclear instrument detectors L) less the power level detected by the upper excore nuclear instrument detectors (U) divided by the sum of these power levels. The AXIAL SHAPE INDEX- (Y ) used for the trip and pretrip signals in the reactor protection system is the above value (Y ) modified by an appropriate multiplier (A) and a constant (B) to determine the (rue core axial power distribution for that channel.

YE

= L"U YI = AYE + B

~+

AZIMUTHAL POWER TILT - T H

1.3 AZIMUTHAL POWER TILT shall be the maximum difference between the power generated in any core quadrant (upper or lower) and the average power of all quadrants in that half (upper or lower) of the core divided by the average power of all quadrants in that half (upper or lower) of the core.

Azimuthal Power in an core uadrant u er or lower)

Power Tilt quadrants 'pper or ower MAX ~

verage power of a CHANNEL CALIBRATION 1.4 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire cha'nnel is calibrated.

CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent, instrument channels measuring the same parameter.

ST. LUCIE - UNIT I

DEFINITIONS CHANNEL FUNCTIONAL TEST 1.6- A CHANNEL FUNCTIONAL TEST shall be the injection of a'simulated signal into the channel as close .to the primary sensor as practicable to verify OPERABILITY including alarm and/or trip functions.

CONTAINMENT VESSEL INTEGRITY 1.7 CONTAINMENT VESSEL INTEGRITY shall exist when:

a. All containment vessel'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 Table 3.6-2 of Specification 3.6.3.

b. All containment vessel equipment hatches are closed and sealed,

c. Each containment vessel 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 CONTROLLED LEAKAGE 1.8 CONTROLLED LEAKAGE shall be the seal water flow supplied from the reactor coolant pump seals.

CORE ALTERATION 1.9 CORE ALTERATION 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 ALTERATION shall not preclude completion of movement of a component to a safe conservative position.

ST. LUCIE - UNIT 1 F2

DEFINITIONS DOSE E UIVALENT I-131 1.10 DOSE E(UIVALENT I-131 shall be that concentration of I-131 (microcuries/

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."

E - AVERAGE DISINTEGRATION ENERGY

l. 11 E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolaat at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95K of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times'shall include diesel generator starting and sequence loading delays where applicable.

FRE UENCY NOTATION

1. 13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table l. l.

GASEOUS RADWASTE TREATMENT SYSTEM 1.14 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and .installed to reduce radioactive gaseous effluents by 'collecting primary coolant system offgases from .the primary system and providing for delay or holdup for the purpose of reducing the tothl radioactivity prior to release to the environment.

IDENTIFIED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into cl'osed systems, 'such as pump seal or valve packing leaks that are captured, and conducted to

.a sump or collecting tank, or

b. Leakage into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be PRESSURE BOUNDARY LEAKAGE, or

c. Reactor Coolant System leakage through a steam generator to the secondary system.

ST. LUCIE - UNIT I 1-3

DEFINITIONS 1.16 'LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE The LOW TEMPERATURE."RCS OVERPRESSURE PROTECTIVE RANGE is that operating condition when (1) the cold leg temperature is ~275oF and (2) the Reactor Coolant System has pressure boundary integrity. The Reactor Coolant System does not have pressure boundary integrity when the Reactor Coolant System is open to containment and the minimum area of the Reactor Coolant System opening is greater than 1.75 square inches.

MEMBER S OF THE PUBLIC 1.17 MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category 'does include persons who use portions of the site for recreational, occupational or other purposes not associated with the plant.

OFFSITE DOSE CALCULATION MANUAL ODCM 1.18 The OFFSITE DOSE CALCUL'ATION MANUAL shall contain the current methodology and parameters, used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and shall include the Radiological'Environmental Monitoring Sample point locations.

OPERABLE - OPERABILITY

,1. 19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s),

and when all necessary attendant instrumentation, controls, electrical power, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

OPERATIONAL MODE - MODE 1.20 An OPERATIONAL NODE (i.e. MODE) shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.2.

PHYSICS TESTS 1.21 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and (1) described in Chapter 14.0 of the FSAR, (2) authorized under the provisions of 10 CFR 50.59, or (3) otherwise approved by the Commission.

ST. LUCIE - UNIT I 1-a

DEFINITIONS PRESSURE BOUNDARY LEAKAGE

>.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except steam generator tube leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM PCP 1.23 The PROCESS CONTROL PROGRAM shall contain the provisions, based on full scale testing, to assure that dewatering of spent bead resins results in a waste form with the properties that meet the requirements of 10 CFR Part 61 (as implemented by 10 CFR Part 20) and of the low level radioactive waste disposal site at the time of disposal.

PURGE - PURGING 1.24 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manne'r that replacement air or gas is required to purify the confinement.

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 Aft.

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power is interrupted to the CEA dr ive mechanism.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to IO CFR Part 50.

SHIELD BUILDING INTEGRITY 1.28 SHIELD BUILDING INTEGRITY shall exist when:

a. Each door is closed except when the access opening is being used for normal transit entry and exit;

b. The shield building'entilation system is in compliance with Specification 3.6.6.1, and ST. UJCIE - iNIT I

DEFINITIONS SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming all full-length control element assemblies (shutdown and regulating) are fully inserted except for 'the single assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY 1.30 The SITE BOUNDARY shall be that line beyond which the land is neither owned, leased, nor otherwise controlled by the licensee.

SOURCE CHECK 1.31 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

STAGGERED TEST BASIS 1.32 A STAGGERED TEST BASIS shall consist of:

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

b. The testing of one system, subsystem, train or other designated component at the beginning of each subinterval.

THERMAL POWER 1.33 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

UNIDENTIFIED LEAKAGE 1.34 UNIDENTIFIED LEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE or CONTROLLED LEAKAGE.

UNRESTRICTED AREA 1.35 An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

ST. LUCIE -, UNIT 'I 1-6

'DEFINITIONS UNRODDED INTEGRATED RADIAL PEAKING FACTOR - F 1.36 The"UNRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of the peak pin power to the average pin power in the unrodded core, excluding tilt.

UNRODDED PLANAR RADIAL PEAKING FACTOR - F 1.37 The UNRODDED PLANAR RADIAL PEAKING FACTOR is the maximum ratio of the peak to average power density of the individual fuel rods. in any of the unrodded horizontal planes, excluding tilt.

VENTILATION EXHAUST TREATMENT SYSTEM 1.38 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form .in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or partic-ulates from the gaseous exhaust stream prior to the release to the environment (such a system is not considered to have any effect on noble gas effluents).

ST. LUCIE - UNIT I 1-7

TABLE 1.1 FRE UENCY NOTATION 7

NOTATION ~FRE UENCY S. At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

At least once per 7 days.

4/M" At least 4 per month at intervals of no greater than 9 days and a minimum of 48 per year.

At least once per 31 days.

. At least once per 92 days.

SA At least once per 184'ays.

At least once per 18 months.

S/U Prior to each reactor startup.

Completed prior to each release.

N.A. Not applicable.

'AA f

For Radi oacti ve Ef 1 uent Sampl i ng For Radioactive Batch Releases only.'

ST. LUCIE - UNIT I 1-,8

TABLE 1.2 OPERATIONAL MODES REACTIVITY X OF RATED AVERAGE COOLANT OPERATIONAL MODE'.

CONDITION K ff THERMAL POWER" TEMPERATURE POWER OPERATION > 0.99 > 5X > 3250F

2. STARTUP > 0.99 < 5X 325oF

3. HOT STANDBY < 0 99 =0 > 325 F

4. HOT SHUTDOWN ~

< 0.99 325 F> T >200 F avg

5. COLD SHUTDOWN < 0.99 < 200'F

6. REFUELING"" < 0.95 0 < 140F Excluding decay heat.

Fuel in the reactor vessel with the vessel head closure bolts less than fully tensioned or with the head removed.

'ST. LUCIE - UNIT I 1" 9

iNSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum. the fire detection instrumentation for each fire detection zone shown in Table 3.3-10 shall be OPERABLE*.

APPLICABILiTY: Whenever equipment in that fire detection zone is required to be OPERABLE.

ACTION:

With the number of OPERABLE fire detection instruments less than required by the minimum instruments OPERABLE requirement of Table 3.3-10:

a. Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> establish a fire watch patrol to inspect the zone(s) with the inoperable instrument(s) at least once per hour (unless the detectors are located inside the annulus, zone ll, then inspect the zone at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />); or.

monitor the containment air temperature at least once per hour

=

at the locations listed in Specification 4.6.1.5 if the inoperable instruments are located inside the containment (zone 13/14/15A 8 B).

b. Restore the inoperable instrument(s) to OPERABLE status within 14 days or prepare and submit a Special Report to the Coamission pursuant to Specification 6.9.2 within the next 30 days outlining-the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument(s) to OPERABLE status.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not appl icabl e.

SURVEILLANCE RE UIREMENTS 4.3.3.7.1 Each of the above required fire detection instruments which are accessible during operation shall be demonstrated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST, exc'ept for- thermal detectors which shall be demonstrated OPERABLE in accordance with Specifica-tion 4.3.3.7.2. Fire detection instruments .which are not accessible during operation shall be demonstrated OPERABLE by performanc'e of a CHANNEL FUNCTIONAL TEST during each COLD SHUTDOWN exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> except that such demonstra-tion need not be performed more often than once per 6 months.

• The emergency power source may .be. inoperable in Modes 5 or 6.

I ST. LUCIE UNIT 1 3/4 3-37 Amendment Nn. g$

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INSTRUMENTATION RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.10 The radioactive gaseous effluent monitoring instrumentation channels shown in Table 3.3-13 shall be OPERABLE Qith their alarm/trip setpoints set to ensure that the limits of Specification 3.11.2.1 are not exceeded. The alarm/

trip setpoints af these channels shall be determined in accordance with the methodology and parameters in the ODCM.

APPLICABILITY: As shown in Table 3.3-13.

ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel alarm/trip setpoint less conservative than required by the above Specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel or declare the channel inoperable.

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. If the inoperable instruments are not returned to operable status wi thin 30 days, explain in the next Semiannaul Radioactive Effluent Release Report why the inopera-bility was not corrected in a timely manner.

c The provisions of Specifications 3.O.3 ana 3.O.'4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.10 Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE'by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-9.

ST. LUCIE - UNIT 1 3/4 3-So Amendment No.

~O 0

~ ~~

REACTOR COOLANT SYSTEN SURVEILLANCE RE UIRENENTS Cant)nued ST. LUCIE - UNIT 1 3/4 4-9

TABLE 4A-2 r STEAM GENERATOR TUBE iNSPECTION n

m 1ST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION Sample Size Result ~ Action Required Result gction Required Result Action Required A mlnlmum of C-1 None N/A N/A N/A N/A S Tubes per gyG.

C-2 Plug defective tubes N/A N/A and inspect additional None'lug defective tubes 2S tubes in this S. G. C-2 and Inspect additional Plug defective tubes 4S tubes ln this S. 6.

Perform action for IC-3 C-3 result of first sample Perform ection for C-3 C-3 result of first N/A N/A sample C-3 Inspect all tubes ln All other this S. G., plug de. S. G.s are N/A N/A factlvo tubas and C-I inspect 2S tubas ln Soma S. C,s Partorm action for each otler 8, G,

• C-2 but no C-2 result of'second N/A NIA additional '.

sample G. are C-3

'g Additional Inspect ill tubes ln 8 S. G. Is C-3 each S. G. and plug dafecthte.tuhtL rt N/A N/A N N ls the number of steam generators In the unit, and n ls the number of steam generators inspected S ~ 3 > Share

- during an inspection e The requirement to Inspect all tubes may be relaxed for Cycle 5 Refuel)ng since an engineering evaluation has shown that the cond1tion(s) has been adequately bounded by inspect)on.

REACTOR COOLANT SYSTEM SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.8 The specific activity of the primary coolant shall be limited to:

a. <<1.0 pCi/gram DOSE E(UIVALENT I-131, and

b. << 100/f pCi/gram.

APPLICABILITY: MODES 1, 2. 3, 4 and 5.

ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT I-131 but within the allowable limit {below and to the left of the line) shown, on Figure 3.4-1, operation may continue for up to 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />'provided that operation under these circumstances shall not exceed 10 percent. of the unit's total yearly operating time. The provisions of Specification 3.0.4 are not applicable.

b. With the specific activity of the primary coolant > 1.0 pCi/gram DOSE EQUIVALENT I-131 for more than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> during one con-tinuous time interval or exceeding the limit line shown on Figure 3.4-1, be in HOT STANDBY with T 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

avg c~ With the specific activity of the primary coolant > 100/E pCi/gram, be in HOT STANDBY with T 500'F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

MODES 1, 2, 3, 4 and 5:

d. With the specific activity of the primary coolant > 1.0 qCi/gram DOSE EQUIVALENT I-131 or > 100/E gCi/gram, perform the sampling and analysis requirements of item 4 a) of Table 4.4-4 until the specific activity of the primary coolant is restored to within its 1 imi ts.

With T > 500'F.

v ST. LUCIE - UNIT 1 3/4 4-17

REACTOR COOLANT SYSTEM ACTION: Continued SURVEILLANCE RE UIREMENTS 4.4.8 The specific activity of the primary coolant shall be determined to be within the limits by performance of the sampling and analysis pro-gram of Table 4.4-4.

ST. LUCIE - UNIT 1 3/4 4-18

REACTOR COOLANT SYSTEM 3.4.10 STRUCTURAL INTEGRITY SAFETY CLASS 1 .COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.1 The structural integrity of components (except steam generator tubes) identified in Section 3.2.2 of the FSAR as Safety Class 1 com-ponents shall be maintained at a level consistent with the acceptance .

criteria in Specification 4.4.10.1.

APPLICABILITY: NODES 1, 2, 3 and 4.

ACTION:

Mith the structural integrity of any of the above components not conform-ing to the above requirements, restore the structural integrity of the affected component to within its limit or isolated the affected component prior to increasing the Reactor Coolant System temperature aare than 50'F above the minimum temperature required by NDT considerations. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREI'RENTS 4.4.10.1 The following inspection program shall be performed during shutdown:

a. Inservice Inspections The structural integrity of the Safety G ass components shall be demonstrated by verifying their acceptability when inspected per the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Minter 1972, as outlined by the inspection program shown in Table 4.4-6.

ST. LUCIE - UNIT 1 3/4 4-26

1 T REACTOR COOLANT SYSTEM SAFETY CLASS 2 COMPONENTS LIMITING CONDITION FOR OPERATION 3.4.10.2 The structural integtity of components identified in Section 3.2.2 of the FSAR as Safety Class 2 components shall be maintained at a level consistent with the acceptance criteria in Specification 4.4.10.2.

~ 0 APPLICABILITY: NODES 1, 2, 3, and 4.

ACTION:

With the structural integrity of any of the above components not con-forming to the above requirements, restore the structural integrity. of the affected component to within its limit or isolate the affected component prior to increasing the Reactor Coolant System above 200'F.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREHENTS t

4.4.10.2 The fo1lowing inspection program shall be performed during shutdown:

a. Inservice Ins ections The structural integrity ef the Safety ass components shall be demonstrated by verifying their acceptability when inspected per the applicable requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition, and Addenda through Winter 1972, as .outlined by t:he inspection, program shown in Table 4.4-7.

ST. LUCIE - UNIT 1 3/4 4-37

REACTOR COOLANT SYSTEM SAFETY CLASS 3 COMPONENTS LIMITING CONDITION FOR OPERATION 3,4.1D.3 The structural integrity of components identified in Section 3.2.2 of the FSAR as Safety Class 3 components shall be maintained at a level consistent with the acceptance criteria in Specification 4.4.10.3.

APPLI CAB Il ITY: ALL MODES.

ACTION:

Mith the structural integrity of .any of the above components not con-forming to the above requirements, restore the structural integrity of the component to within its limit or isolate the affected components from service. The provisions of Specification 3.D.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.10.3 The following inspection program shall be performed:

a. Inservice Ins ections The structural integrity of the Safety ass 3 components shall be demonstrated at least once per 40

.months during periods of zormal reactor operation or during

=-

system performance testing by verifying via visual inspections, as outlined by the inspection program shown in Table 4.4-8, that there is no evidence of unanticipated component leakage, structural distress, or corrosion.

ST. j.UCIE -'UNIT 1 3(4 4-53

CONTAINMENT SYSTEMS CONTAINMENT VESSEL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.6 The structural integrity of the containment vessel shall be maintained at a level consistent with the acceptance criteria in Speci-fication 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the structural integrity of the containment vessel not conforming to the above requirements, restore the structural integrity to within the limits prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREHENTS 4.6.1.6 The structural integrity of he containment vessel shall be determined during the shutdown for each Type A containment leakage rate.

test (reference Specification 4.6.1.2) by .a visual inspection of the accessible interior and exterior surfaces of the vessel and verifying no apparent changes in appearance of the surfaces or other abnormal degra-dation.

ST. LUCIE - UNIT 1 3/4 &14

CONTAINMENT SYSTEMS SHIELD BUILDING STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.6.3 The structural integrity of the shield building shall be main-tained at a level consistent with the acceptance criteria in Specification 4.6.6.3.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION with the structural integrity of the shield building not conforming to the above requirements, r estore the str uctural integrity to within the limits prior to increasing the Reactor Coolant 5ystem temperature above 2DO F.

SURVEILLANCE RE UIREMENT5 4.6.6.3 The structural integrity of the shield building shall be deter-mined during the shutdown for each Type A containment leakage rate test (reference Specification 4.6.1.2) by a visual inspection of the accessible interior and exterior surfaces of the shield building and verifying'no apparent changes in appearance of the concrete surfaces or other abnorma1 degradation.

ST. LUCIE - UNIT 1 3/4 6-31

PLANT SYSTEMS 3 4.7.11 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION MATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.11.1 The fire suppression water system shall be OPERABLE* with:

a. Two high pressure pumps . each with a capacity of 2350 gpm, with their discharge aligned to the fire suppression header,

b. Separate water supplies. each with a minimum contained volume of 300,000 gallons, and

c. An OPERABLE flowpath capable of taking suction from city water storage tank 1A and city water storage tank 1B and transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrants and the first valve ahead of each hose standpipe system riser required to be OPERABLE per Specification 3.7.11.2.

APPLICABILITY: At all times.

ACTION:

a. itith one pump and/or one water supply inoperable. restore the inoperable equipment to OPERABLE status within 7 days or, prepare and submit a Special Report to the Comission pursuant to Speci-fication 6.9.2 within the next 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in this system. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b. Mith the fire suppression water system otherwise inoperable:

l. Establish a backup fire suppression water system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and

2. Submit a Special Report in accordance with Specification 6.9.2; a) By telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b) Confirmed by telegraph, mailgram or facsimile transmis-sion no later than the first working day following the event. and The emergency power source say be inoperable in Hodes 5 or 6.

ST. LUCIE - UNIT 1 3/4 7-40 Amendment No. g$ , 40

RAD'.CACTI'lE EF. LUENTS Pin 5 LIMITING CONDITION FOR OPERATION

3. 11. 1.2 The dose or dose commitment to a MEMBER OF THE PUSLIC from radioac~iye materials in liquid effluents released, from each reactor unit, to UNRESTRICTED .AREAS (see Figure 5. 1-1) shall, be limited:

During any calendar quarter to less than or equal to 1.5 mrems to the total body and to less than or equal to 5 mrems to any organ, and a

b. During any calendar year to less than or equal to 3 mrems to the total body and to less than or equal to 10 mrems to any organ.

APPLICABILITY: At all times.

ACiiON:

Mith the calculated dose from tt~e reTease of radioactive materials in liquid effluen s exceeding any of the above limits prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases and radioactive materials in liquid effluents during the remainder of the current calendar quarter and during the subsequent three calendar quarters, so that the cumulative dose or dose commitment to an individual from these releases is within 3 mrems to the total body and 10 mrems to any organ.

b. The provisions of Specifications 3.0.3 and 3.0.4 are noi applicable.

SURYEILLANCE RE UIREMENTS 4.11.1. 2 Dose Calculations. Cumulative dose contributions from liquid efflu'ents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the OOCM at least once per 51 days.

ST. LUCIE - UNIT 1 3/4 11-5 Amendment No. ~

RADIOACTIVE EFFLUEHTS LIOUIO WASTE TREATMEHT LIMITIHG CONDITION FOR OP RAT'ION 3.11.1.3 The liquid radwas e treatment system shall be OPERABLE. The appro-priate portions of the system shall be used to reduce the radioactive materials in liquid wastes prior to their discharge when the projected doses due .o the liquid ef,luent from the site to UNRESTRICTED AREAS (see Figure 5. 1-1) when averaged over 31 days, ~ould exceed 0.06 mrem to the total body or 0.2 mrem to any organ.

APPLICABILITY: At al 1 tines.

ACTION:

Nth the liquid radwaste treatment system inoperable for nore han 31 days or with radioactive liquid waste being discharged and in excess of 'the above linits prepare and submit to the without'reatment Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following'information:

l. Identifica ion of the inoperable equipment or subsystems and the reason for inoperability, I

2. Action(s) taken to restore the inoperable equipnent to OPERABLE status, and Sumnary description of action(s) taken to preven a recurrence.

b. The provisions of Specifications 3. 0. 3 and 3. 0. 4 are not applicable.

SURVEILLANCE REOUI REMEHTS 4.11.1.3.1 Doses due to liquid reIeases to UNRESTRICTED AREAS shall be projected at leas once per 31 days, in accordance with the OOCM unless the liquid radwas e treatment I

systen is being used.

4. 11. 1.3.2 The Tiquid radwaste treatment sys em shall be demonstrated GPER"BLE by operating the liquid radwaste treatment system equipment for at lees 30 minutes at least once per 92 days unless the liquid radwas e system has been utilized to process radioactive liquid effluents Curing the previous o2 days..

ST. LUCIE - UNIT 1 3/4 11-6 Amendnent No. <

9

RAO OACTIV K. FLU NTS DOSE - HCSLE GASES LIMITING COHO.T ON FOR OPERATION 3.11.2.2 The air dose due to noble gases released in gaseous effluents, from each reactor unit, to areas at and beyond the SITE BOUNOARY (see Figure 5.~=1) shall be limited to the. following:

>a. Ouring any calendar quarter: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrads for beta radi,ation, and,,

b. During any calendar year: Less than or equal to 10 mrads for gamma radiation and less 'than or equal to 20 mrads for beta radiation.

APPLiCABILITY: At all times.

ACTION

a. Mith the calcula ed air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases wi 11 be in compliance with the above limits.

b. Tne provisions of Specifications 3.0.3 and 3.0.4 are not app1icable.

SURVEILLANCE R OUIREHEHTS

.11.2.2 Oose Calculations. Cumulative dose contributions for the current alendar quar er and.current calendar year shall be determined in accordance ith the methodolgy and parameters in the OOCN at least once per 31 days.

ST. LUCIE - UNIT I 3/4 Il-ll Amendment No. ~

RADIOACTIVE EFFLUENTS DOSE -. IODINE-131. IODINE-i33. TRITIUM AND RADIONUCLIDES IN PARTICUlLATE FORM L .".IT:NG CONDITION FOR OPERATiON 3.11.2.3 The dose to a MEMBER OF THE PUBLIC from iodine-131, iodine-133, tritium and all radionuclides in particulate form with half-lives grea er than 8 days'n gaseous effluents released, from each reactor unit to areas .at a'nd beyond he SITE BOUNOARY, (see Figure 5. 1-1) shall be limited to he ollowing:

a. During any calendar quarter: Less than or equal to 7. 5 mrems to any organ and,

b. During any calendar year: Less than or equal to 15 mrems to any or gan.

APPLICABILITY: At all times.

ACTION:

Mi~h the calculated dose from the= release of iodine-131, iodine-133, tri ium, and radionuclides in particulate form with half-lives, greater than 8 days, in gaseous effluents'exceeding any of the above limits prep8re and submit to the Commission within 30 days, pursuant to Specifica-tion 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit and defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

b. The provisions oI Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS

4. 11. 2. 3 Cumulative dose contributions for the current calendar quarter and current calendar year for iodine-131, iodine-l33, tritium, and radionuclides in particulate form with 'half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

ST. LUCIE .- UNIT 1 3/4 11-12 Amendment No. 5 9

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The GASEOUS RADWASTE TREATMENT SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM shall be OPERABLE. The appropriate portions of the GASEOUS RADWASTE TREATMENT SYSTEH shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases from the site to UNRESTRICTED AREAS (see Figure 5.1-1), when averaged over 31 days, would exceed 0'.2 mrad for galena radiation and 0.4 mrad for beta radiation. The appropriate portions of the VENTILATION EXHAUST TREATMENT SYSTEM shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases from the site (see Figure 5.1-1) when averaged'ver 31 days would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the GASEOUS RADWASTE TREATMENT SYSTEM and/or the VENTILATION EXHAUST TREATMENT SYSTEM inoperable for more than 31 days or with gaseous waste being discharged without treatment and in excess of the above limits prepare and submit'to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

l. Identification of the inoperable equipment or subsystems and the reason for the inoperability,

2. Action(s) taken.to restore the inoperable equipment to OPERABLE status, and

3. Sugary description of action(s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREHEHTS 4.11.2.4.1 Doses due to gaseous releases from the site to UNRESTRICTED AREAS shall be projected at least once per 31 days, in accordance with the ODCH unless the GASEOUS RADWASTE TREATMENT SYSTEH is being used.

4.11.2.4.2 The GASEOUS RADWASTE TREATMENT SYSTEfl and VENTILATION EXHAUST TREATMENT SYSTBi shall be demonstrated OPERABLE by operating the GASEOUS RAD-WASTE TREATMENT SYSTEM equipment and VENTILATION EXHAUST TREATMENT SYSTEH equip-ment for at least 30 minutes, at least once per 92 days unless the appropriate system has been utilized to process radioactive gaseous effluents during the previous 92 days.

ST. LUCIE - UNIT 1 3/4 11-13 Amendment No. 5 9

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RADIOACTIVE EFFLUENTS 3 4.11.4 TOTAL DOSE LIMITING CONDITION FOR OPERATION 3.11.4 The annual, (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrems to the total body or any organ, except the thyroid, which shall be limited -to less than or equal to 75 mrems.

APPLICABILITY: At all times.

ACTION:

a. lith the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of Specifi-cation 3.11.1.2a, 3.11.1.2b, 3.11.2.2a, 3.11.2.2b, 3.11.2.3a or 3;11.2.3b, calculations shall be made including direct radiation contributions from the reactor units and from outside storage tanks to determine whether the above limits of Specification 3.11.4 have been exceeded. If such is the case, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that defines the corrective action to be take'n to reduce subse-quent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits.

This Special Report, as defined in 10 CFR Part 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a tlEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release(s) covered by this report. It shall also describe levels of radiation and concentration of radioactive material involved, and the cause of the exposure levels of concentrations. If the estimated dose(s) the above limits, and if the release condition resulting in violation of 40 CFR Part 190 has not already been corrected, the Special

'xceeds Report shall include a request for a variance in accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely request, and a variance is granted unti 1 staff action on the request is 'complete.

b. The provisions'of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.11.4.1 Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with Surveillance Requir ements 4.11.1.2, 4.11.2.2, and 4.11.2.3, and in accordance with the methodology and parameters in the ODCM.

4.11.4.2 Cumulative dose contributions from direct radiation from the reactor units and from radwaste storage tanks shall be determined in accordance with the methodology and parameters in the ODCH. This Surveillance Requirement shall be required only in the event the above Action a. requires the applicable alculations.

ST. LUCIE - UNIT 1 3/4 11-17 Amendment No. 59

3 4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 4.12.1'ONITORING PROGRAM LIMITING CONDITION FOR OPERATION 3.12.1 The radiological environmental monitoring program shall be conducted as specified in Table 3.12-1.

APPLICABILITY: At all times.

ACTION:

a. Mith the radiological environmental monitor-',ng program not being con-ducted as'pecified in Table 3.12-1 prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Specification 6.9.1 11, a description of the reasons for not conducting

~

the program as required and the plans for preventing a recurrence.

b. Mith the confirmed* level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter prepare and submit to the Commission with 30 days, pursuant to Specification 6.9.2,"a* Special Report that identifies the cause(s) for exceeding the limit(s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3. 11. 1.2, 3. 11.2.2, and 3. 11.2.3. Mhen more than one of the radionuclides in Table 3. 12-2 are detected in the sampling medium, this report shall be submitted if:

concentration 1 + concentration 2 + ... s 1.0 reporting level 1 reporting 1 evel 2 Mhen radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents. this r eport shall be submitted if the potential annual dose to a MEMBER OF THE PUBLIC is equal to or greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2 and 3.11.2.3. This report shall include the methodology for calculating the cumulative potential dose contributions for the calendar year from radionuclides detected in environmental samples and can be determined in accordance with the methodology and parameters in the ODCM. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report.

c. Mith milk or broadleaf vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identify locations

• A confirmatory reanalysis of the original, a duplicate, or a new sample may be desirable, as appropriate. The results of the confirmatory analysis shall be .

completed at the earliest time consistent with the analysis but in any case

.within 30 days.

ST. LVCIE - UNIT 1 3/4 12-1 Amendment No. g 9

RADIOLOGICAL ENV IRONMENTAL MOilI TOR ING LIMITING CONDITION FOR OPERATION Continued ACTIOti: (Continued) for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific loca-tions from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.9. 1. 10, identify the cause of the unavailability of samples and identify the new location(s) for obtaining replacement samples in the next Semiannual Radioactive

. Effluent Release Report and also include in the report a revised figure(s) and table for the ODCH reflecting the new location(s).

d; The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.12.1 The radiological environmental monitoring 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.

ST. LUCIE - UNIT 1 3/4 12-2 Amendment No. <

j

'I

RADIOLOGICAL ENV IRONHENTAL MONITORING 3 4.12.2 L'ANO USE-CENSUS LIHITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, the nearest residence and the nearest garden* of greater than '50 m~(500 ft~) producing broad leaf vegetation.

APPLICABILITY: At al 1 times.

ACTION:

a. With a land use census identifying a location(s) that yields a cal-.

culated dose or dose commitment greater than the values currently being'calculated in Specification 4.11.2-3 identify the new location(s) in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.10.

b. Mith a land use census identifying a location(s) that yields a cal-

'culated dose or dose commitment (via the same exposure pathway) 20$

greater than at a location from which samples. are currently being obtained in accordance with Specification 3.12.1, add the new loca-tion(s) to the radiological environmental monitoring program within 30 days. The sampling location(s), excluding the control station location, having the lowest calculated dose or dose comnitment(s),

via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this land use census was conducted. Pursuant to Specification 6.9.1.10, identify'the new location(s) in the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCN re flecting the new 1 ocati on (s ) .

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREHENTS 4.12.2 The land use census shall be conducted during the growing season at least once per 12 months using that information that will provide the best r esults, such as by a door-to-door survey, aerial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in .the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.11 .

• Broad leaf vegetation sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted 0/gs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 3.12-1.4b shall be followed, including analysis of control samples.

ST. LUCIE - UNIT 1 3/4 12-11 Amendment No. 5 9

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS Continued The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage ~

1 gallon per minute, total). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant. However, even if a defect of similar type should develop in service, it will'e found during scheduled inservice steam generator tube examinations. Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 40K of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

ST. LUCIE - UNIT 1 B 3/4 4-3

ADMINISTRATIVE CONTROLS

6. 1 RESPONSIBILITY

6. l. 1 The Plant Manager shall be respons'ible for overall .unit operation and shall delegate in writing. the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor, or during his absence from the control room, a designated individual, shall be responsible for the control room command function. A management directive to this effect, signed by the Vice President Nuclear Operations shall be reissued to all station personnel on an annual basis.

6. 2 ORGANIZATION OFF SITE 6.2.1 The offsite organization for unit management and technical support shall be as "shown in Figure 6.2-1.'NIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

a4 Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b. At least one licensed Reactor Operator shall be in the control room when fuel is in the reactor. In addition, while the reactor is in MODE 1, 2, 3, or 4, at least one licensed Senior Reactor Operator shall be in the control room.

C. A health physics technician shall be on site when fuel is in the reactor.

d. All CORE ALTERATIONS shall be observed by a licensed operator and supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. The SRO in charge of fuel handling normally supervises from the control room and has the flexi-bility to directly supervise at either the refueling deck or the spent fuel pool.

e. A site Fire Brigade of at least five members shall be maintained onsite at all times. The Fire Brigade shall not include the Shift Supervisor, the STA, nor the two other members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

Administrative procedures shall be developed and implemented to limit the working hours of senior reactor operators and reactor operators.

The health physics technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence, provided immediate action is taken to fill the required positions.

ST. LUCIE - UNIT I 6-1

ADMINISTRATIVE CONTROLS UNIT STAFF (Continued)

Adequate shift coverage shall be maintained without routine heavy use of overtime. The objective shall be to have operating personnel work a normal 8-hour day, 40-hour week while the plant is operating.

However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifica-tion, on a temporary basis the following guidelines shall be followed:

a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight, excluding shift turnover time.

b. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in any 7-day period, all excluding shift turnover time.

c. .A break of at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> shoul'd be allowed between work periods, including shift turnover time.

d. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the Plant Manager or his deputy, or higher levels of management, in accordance with established procedures and with documentation of the basis for granting the deviation. Controls shall be included in the procedures such that, individual overtime shall be reviewed monthly by the Plant Manager or his designee to assure that excessive hours have not been assigned. Routine deviation from the above guidelines is not authorized.

ST. LUCIE - UNIT I 6"2

EKECVTIVE VICE PRESIDENT GROUP CONPANY NUCLEAR VICE PRESIDENT VICE PRESIDENT QUALITY VICE PRESIDENT REVIEN BOARD ADVAIN:ED SYSTENS ENGINEERING ASSURANCE NUCLEAR ENERGY AND TECHNOLOGY PRVJECT NANAGEIIENT COltIITTEE AND CONSTRICTIUN NNAGER VICE PRESIDENT DIRECTOR DIRECTOR VF PLANNING AND CONTROL NUCLEAR OPERATIONS NIKLEAR LICENSING QUALITY ASSUHAICE IIANAGER QUALITY ASSURANCE.

S ITE SUPERI NTEIRIENT ST>> LUCIE PROJECTS QA VICE PRESIDENT VICE PRESIDENT TECHNICAL NNAGER NINTENANCE NNAGER TRAINI% SVPERVlSVR ST>> LUCIE

• TURKEY POINT NUCLEAR NUCLEAR NUCLEAR NUCLEAR PLANT NAHAGER PLANT NNAGER NUCLEAR ENERGY NUCLEAR SERVICES STAFF FIGURE 6 2-1 OFFSITE ORGANIZATION FOR FACILITY NNAGENENT ANO TECHNICAL SUPPORT

• The position of Vice President St. Lucie has been approved by Florida Power 8 Light and will be implemented as discussed with the Region I1, Regional Administrator.

I' 'I I'

'i I I' 's I ~ I e. ~ I

Table 6.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH TWO SEPARATE CONTROL ROOMS WITH UNIT 2 IN MODE 5 OR 6 OR DEFUELED POSITION NUMBER OF INDIVIDUALS RE(UIRED TO FILL POSITION MODE1,2,3, or4 MODE 5 or 6 1a 'a SS (SRO) 1 SRO 1 None RO 2 1 AO 2 2b STA 1 None WITH UNIT 2- IN MODE 1, 2, 3 OR 4 POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION MODE1,2,3, or4 MODE 5 or 6 SS (SRO) la 1 a

SRO 1 None RO: 2 1 AO 2 1 STA 1 None SS Shift Supervisor with a Senior Reactor Operator's License on Unit 1 SRO Individual with a Senior Reactor Operator's, License on Unit 1 RO Individual with a Reactor Operator's License on Unit 1 AO Auxiliary Operator STA Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew Composition to within the minimum requirements of Table 6.2-1. This provision does not permit any shift crew position to be unmanned upon shift change due to an oncoming shift crewman being late or absent.

During any absence of the Shift Superviso~ from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Control Room command function. During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid SRO or RO license shall be designated to assume the Control Room command function.

a/ Individual may fill the same position on Unit 2; b/ One of the two required individuals may fill the same position on Unit 2 .

ST. LUCIE - UNIT 6-5

NNINISTRATI VE CONTROLS 6.2.4 SHIFT TECHNICAL ADVISOR The Shift Technical Advisor function is to provide on shift advisory technical support in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit.

6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifica-tions of ANSI/ANS-3.1-1978 as endorsed by Regulatory Guide 1.8, September (reissued Nay 1977), except for the (1) Health Physics Supervisor who shall1975 meet,

• Not responsible for sign-off function.

ST. LUCIE - UNIT I 6-6 Amendment No.

ADMINISTRATIVE CONTROLS

6. 3 UNIT STAFF UALIFICATIONS (Continued) or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design and plant operating characteristics, including transients and accidents.

6. 4 TRAINING 6.4. 1 A retraining and replacement training program for the unit staff shall be maintained under the direction of the Training Supervispr and shall sect or exceed the requirements and recommendations of Section 5.5 of ANSI/ANS 3.1-1978 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience.

6.4.2 Fire drills will be conducted at least quarterly for all Fire Brigade personnel.

6.5 REVIEM AND AUDIT 6.5.1 FACILITY REVIEM GROUP FRG FUNCTION 6.5. 1. 1 The Facility Review Group shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSITION 6.5.1.2 The Facili ty Review Group shall be composed of the:

Member: Plant Manager Member: Operations Superintendent Member: Operations Supervisor Member: Maintenance Superintendent Member: Instrument 8 Control Supervisor Member: Reactor Supervisor Member: Health Physics Supervisor Member: Technical Supervisor Member: Chemistry Supervisor Member: guality Control Supervisor Member: Assistant Plant Supt. Mechanical Member: Assistant Plant Supt. Electrical The Chairman shall be a member of the FRG and shall be designated in writing.

ALTERNATES 6.5. 1.3 All alternate members shall be appointed in writing by the FRG Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in FRG activities at any one time.

ST. LUCIE - UNIT I 6-7

NNINISTRATIVE CONTROLS MEETING FRE UENCY 6.5.1.4 The FRG shall meet at least once per calendar month and as convened by the ERG Chafrman or hfs designated alternate.

UORUM 6.5.1.5 The quorum of the FRG necessary for the performance of the FRG responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates.

RESPONSIBILITIES 6.5.1.6 The Facflfty Review Group shall be responsfble for:

a. Review of (1) all procedures required by Specification 6.8 and changes thereto, (2) all programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

c Review of all proposed changes to Appendix A Technical Specifications.

d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety.

e. Investigation of all violations of the Technfca~ Specffications, including the preparation and forwarding of reports covering evalua-tion and recoaaaendations to prevent recurrence to the Vice President Nuclear Operations, Group Vice President Nuclear Energy, and,to the Chairman of the Company Nuclear Review Board.

f. Review of all REPORTABLE EVENTS.

g. Review of unit operations to detect potential nuclear safety hazards.

h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Hanager or the Company Nuclear Review Board.

Review of the Security Plan, and fapleaentfng procedures and submittal of recommended changes to the Company Nuclear Review Board.

5. Review of the Emergency Plan and faplementfng procedures and submittal of recoaaaended changes to the Company Nuclear Review Board.

ADMINISTRATIVE CONTROLS RESPONS IB ILIT I ES (conti nued)

k. Review of every unpl armed on-site release of radioactive materi al to the environs including the preparation of reports coveri'ng evaluation, disposition'of recommendations and the corrective action to prevent r'ecurrence and the forwarding of these reports to the Vice President Nuclear Operations and to the Company Nuclear Review Board.

1. Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL and RADWASTE TREATMENT SYSTEMS.

AUTHORITY 6.5.1.7 The Facility Revi,ew Group shall:

a. Recommend in writing to the Plant Manager, approval or disapproval of items considered under Specifications 6.5.1.6a. through d. and m. above.

b. Render determinations in writing, with regard to whether or not each item considered under Specifications 6.5.1.6a. through e. above constitutes an unrevi ewed s af ety questi on.

c~ Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Vice President Nuclear Operations, the Group Vice President Nuclear Energy, and the Company Nuclear Review Board of disagreement between the FRG and the Plant Manager; however, the Plant Manager shall have responsibility for resolution of such disagreements pursuant to Specification 6.1.1 above.

RECORDS 6.5.1.8 The Facility Review Group shall maintain written minutes of each FRG meeting that, at a minimum, document the results of all FRG activities performed under the responsibility and authority provisions of these techni al specifications. Copies shall be provided to the Vice President Nuclear Operations, the Group Vice President Nuclear Energy, and the Chairman of the Company Nuclear Review Board.

6.5.2 COMPANY NUCLEAR REVIEW BOARD CNRB FUNCTION 6.5.2.1 The Company Nuclear Review Board shall function to provide independent review and audit of designed activities in the areas of:

a. nuclear power pl ant operations

b. nuclear engineering

c. chemi stry and radiochemi stry

d. met al l urgy ST. LUCIE - UNIT 1 6-9

ADMINISTRATIVE CONTROLS FUNCTION (Centinved)

e. instrumentation and control radiological safety mechanical and electrical engineering quality assurance practices COMPOSITION 6.5.2.2 The CNRB shall be composed of the following members:

Member: Vice President, Advanced Systems and Technology Member: Chief Engineer, Power Plant Engineering Hember: 6 oup Vice President, Nuclear. Energy Member: Vice President, Nuclear Operations Hember: Director of Qual.i,ty Assurance Hember: Hanager, Nuclear Fuel Hember: Power Plant Engineering Principal Engineer ~

Hember: Power Plant Engineering Senior Project Manager The Chairman shall be a member of the CNRB and shall be designated in writing.

ALTERNATES 6.5.2e3 All alternate members shall be appointed in writing by the CNRB Chairman to serve on a temporary basis; however, .no more than two alternates shall participate as voting members in CNRB activities at any one time.

CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the CNRB Chairman to provide expert advice to the CNRB.

MEETING FRE UENCY 6.5.2.5 The CNRB shall meet at least once per calendar quarter during the initial year of unit operation following fuel loading and at least once per 6 months thereafter and as convened by the CNBR Chairman or his designated alternate.

UORUM

6. 5.2. 6 The quorum of the CNRB necessary for the performance of the CNRB review and audit functions of these Technical Specifications shall consist of the'Chairman or his designated alternate and at least four CNRB members including alternates. No more. than a minority of the quorum shall have line responsibility for operation of the unit.

ST. LUCIE - UNIT I 6-10

ADMINISTRATIVE CONTROLS REVIEW 6.5.2.7 The CNRB shall review:

a. The safety evaluations for (1) changes to procedures, equipment, or systems and (2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.

b. Proposed changes to procedures, equipment, or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

C. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

d. Proposed changes to Technical Specifications or this Operating License.

e. Violations of codes, regulations, orders, Technical, Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety.

All REPORTABLE EVENTS.

h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety.

Reports and meetings ainutes of the Facility Review Group.

AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the CHRB. These audits shall encompass:

a. The conformance of unit operation to provisions contained within the Technical Specifications and applicable license conditions at least once per 12 eonths.

b. The performance, training and qualifications of the entire unit staff at least once per 12 oonths.

c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systels, or aethod of operation that affect nuclear safety at least onceper 6 months.

ST. LUCIE - lNIT I 6-D

ADMINISTRATIVE CONTROLS AUDITS (Continued)

d. The performance of activities required by the guality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months.

e. Any other area of unit operation considered appropriate by the CNRB or the Executive, Vice President.

The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee gA personnel.

The fire protection equipment and program implementation at least once per. 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used at least every third year.

h. The radiological environmental monitoring program and the results thereof at least once per 12 months.

The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at least once per 24 months.

The PROCESS CONTROL PROGRAM and implementing procedures for dewatering of radioactive bead resin at least once per 24 months.

AUTHORITY 6.5.2.9 The CNRB shall report to and advise the Executive Vice President on those areas of responsibility specified in Specifications 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2. 10 Records of CNRB activities shall be prepared, approved, and distributed as indicated below:

ae Minutes of each CNRB meeting shall be prepared, approved, and forwarded to the Executive Vice President within 14 days following each meeting.

b. Reports of reviews encompassed by Specification 6.5.2.7 above sh'all be prepared, approved, and forwarded to the Executive Vice President within 14 days following completion of the review.

C. Audit reports encompassed by Specification 6.5.2.8 above shall be forwarded to the Executive Vice President and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

ST. LUCIE " UNIT ',I, 6-12

ADRINISTRATIVE CONTROLS

, 6.6 REPORTABLE EVENT 6.6.1 The following actions shall be taken for REPORTABLE EVENTS

a. The Commission shall be notified and a report submitted pursuant*to the requirements of Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the FRG, and the results of this review shall be submitted to the CNRB,. tne Vice President Nuclear Operations, and the Group Vice President Nuclear Energy.

6. 7 SAFETY LIHIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The Vice President Nuclear Operations and the CNRB,shall be notified within 24 .hours.

b. A Safety Limit Violation Report shall be prepared. The report shall be reviewed by the FRG. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation

'upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence.

c. The Safety Limit Violation Report shall be submitted to the Commission, the CNRB, the Vice President Nuclear Operations, and the Group Vice President Nuclear Energy.

d. Critical operation of the unit shall not be resumed until authorized by the Coaeission.

6.8 PROCEDURES AND PROGRANS 6.8.1 Mitten procedures shall be established, implemented and maintained covering the activities referenced below:

a, The applicable procedures recoaeended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February 1978, and those required for implementing the requirements of NUREG 0737.

b. Refueling operations.

c. Surveillance and test activities of safety-related equipment.

d. Security Plan implementation.

e. Emergency Plan implementation.

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued)

f. Fire Protection Program implementation.

g. PROCESS CONTROL PROGRAM implementation

h. OFFSITE DOSE CALCULATION MANUAL implementation.

i. guality Control Program for effluent monitoring, using the guidance in Regulatory Guide 1.21, Revision 1, June 1974 Control Program for environmental monitoring using the

'uality

~

guidance in Regulatory Guide 4. 1, Revision 1, April 1975.

6.8.2 Each procedure of Specification 6.8. la. through i. above, and changes thereto, shall be reviewed by the FRG and shall be approved by the Plant Manager prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of Specification 6.8. la. through i.

above may be made provided:

a. The intent of the original procedure is not altered.

b. The change is approved by two members of the plant management staff, at least one of whom -holds a Senior Reactor Operator's License on the unit affected.

c. The change is documented, reviewed by the FRG and approved by the Plant Manager within 14 days of implementation.

6.8.4 The following programs shall be established, implemented, maintained, and shall be audited under the cognizance of the CNRB at least once per 24 months:

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 Shutdown Cooling System, High Pressure Safety Injection System, Containment Spray System, and RCS Sampling. The program shall include the following:

(i) Preventive maintenance and periodic visual inspecti.on requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less.

b. In-Plant Radioiodine 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:

(i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment.

ST. LUCIE " UNIT 6-14

ADMINISTRATI VE CONTROLS PROCEDURES AND PROGRAMS (Continued)

c. Secondar Water Chemistr A program for monitoring of secondary water chemistry to inhibit steam generator tube degradation. This program shaTl include:

(i) Identification of a sampling schedule for the critical variables and control points for these variables, (ii) Identification of the procedures used to measure the values of the critical variables, (iii) Identification of process sampling points, which shall include monitoring the discharge of the condensate pumps for evidence of condenser in-leakage, (iv) Procedures for the recording and management of data, (v) Procedures defining corrective actions for all off-control point chemistry conditions, and (vi) A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective action.

d. Backu Method for Determinin Subcoolin Mar in A program which will ensure the capability to accurately monitor the Reactor Coolant System subcooling margin. This program shall include the following:

(i) Training of personnel, and (ii) Procedures for monitoring.

e. 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 accident conditions. The program shall include the following:

(i) 'raining of personnel, (ii) Procedures for sampling and analysis, and (iii) Provisions for maintenance of sampling and analysis equipment.

ST. LUCIE - UNIT I 6-15

ADHINISTRATIVE CONTROLS 6.9 REPORTING RE UIREHENTS 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 Regional Administrator of the Regional Office of the NRC unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (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 plant; 6.9. 1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design. predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.

6.9. 1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial. power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS-6.9. 1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to Harch 1 of each year. The initial report shall be submitted prior to Harch 1 of the year following initial criticality.

6.9. 1.5 Reports required on an annual basis shall include:,

a. A tabulation on an annual basis of the number of station, utility,

~

and other personnel (including contractors) receiving exposures t

should combine those sections that are common to all units at the station.

ST. LUCIE - UNIT 1 6-16

NNINISTRATIVE CONTROLS ANNUAL REPORTS (Continued) greater than 100 mrems/yr and their associated man-rem exposure 2/ e.g., reactor operations and according to work and job functions, surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), waste processing, and refueling.

The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20K of the individual total dose need not be

~

accounted for. In the aggregate, at least SO% of the total whole dose received from external sources should be assigned to specific major work functions.

N)NTHLY OPERATING REPORTS 6.9.1.6 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, shall be submitted on a monthly basis to the Director, Office of Resource Management, U. 5. Nuclear Regulatory Commission, Washington; .D.C. 20555, with a copy to the Regional Administrator of the Regional Office of the NRC, no later than the 15th of each month following the calendar Ionth covered by the report.

This tabulation supplements the requirements of $ 20.407 of 10 CFR Part 20.

ST. NCIE - INIT 1 6-17

XHISTRATIVE CONTROLS This page left intentionally blank

'T. LUCIE - LNIT 1

ADMINISTRATIVE CONTROLS This page left intentionally blank ST. LUCIE - LNIT 1

ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT" 6.9.1.10 Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Heasuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents form Light"Water-Cooled Nuclear Power Plants,"

Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix 8 thereof.

The Radioactive Effluent Release Report to be submitted within 60 days after January 1 of each year shall include an annual summary of hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmos-pheric stability."" Thi.s same report shall include an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar. year. This same report shall also include an assessment of the radiation doses from radioactive liquid and gaseous effluents to MEMBERS OF THE PUBLIC due to their activities inside the SITE BOUNDARY (Figure 5.1-1) during the report period. All assump-tions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological condi-tions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance .with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

Every 2 years using the previous 6 months release history for isotopes and historical meteorolgical data determine the controlling age group for both liquid and gaseous pathways. If changed from current submit change to ODCM to reflect new tables for these groups and use the new groups in subsequent dose calculations.

The Radioactive Effluent Release Report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed HEHBER OF THE PUBLIC from reactor releases for the previous A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

In lieu of submission with the first half year Radioactive Effluent Release Report, the licensee has the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.

ST. LUCIE - UNIT I 6-20

ADMINISTRATIVE CONTROLS SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT (Continued) calendar year. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, March 1976.

The Radioactive Effluent Release Reports shall include the following information for each class of solid waste (as defined by 10 CFR Part 61) shipped offsite during the report period:

Volume,

b. Total curie quantity (specify whether determined by measurement or estimate),

C. Principal radionuclides (specify whether determined by measurement or estimate),

d. Type of waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms)

e. Type of container (e. g., LSA, Type A, Type B, Large guantity), and Solidification agent or absorbent (e.g., cement, urea formaldehyde).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATION MANUAL (ODCM), as well as a listing of new locations for dose calculations andlor environmental monitoring identified by the land use census pursuant to Specification 3.12.2.

ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT" 6.9. l. 11 Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall, be submitted prior to May 1 of the year following initial criticality.

The. Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and information based on trend analysis of the results of the radiological environmental surveillance activities for the report period, including a comparison, as appropriate, with preoperational studies, with operational controls 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 land use censuses required by Specification 3.&.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 du~ing the period pursuant to the A single submittal may be made for a multiple unit station.'T.

LUCIE - UNIT 1 6-21

NNINISTRATIYE COhTROLS ANNUAL RADIOLOGICAL ENYIROHMENTAL OPERATING REPORT (Continued) locations specified in the Table and Figures in the DDT, 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 individual 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 sumaary 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 center line of one reactor; the results the Interlaboratory Comparison Program, required by Specification 3.12.3: discussion of all .

deviations from the sampling schedule of Table 3.12-1; and discussion of all analyses in which the LLD required by Table 4. 12-1 was not achievable.

6.9.1.12 At least once every 5 years, an estimate of the actual population within 10 miles of the plant shall be prepared and submitted to the Regional Administrator of the Regional Office of the NRC.

6.9.1.13 At least once every 10 years, an estimate of the actual population within 50 miles of the plant shall be prepared and submitted to the Regional Administrator of the Regional Office of the NRC.

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the linis'eriod indicated.

6.10.1 The following records shall be retained for at least 5 years:

a. Records and logs of unit operation covering time interval at each power level.

b. .

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

C. All REPORTABLE EVENTS

d. Records of surveillance activities; inspections and calibrations required by these Technical Specifications.

Records of changes made to the procedures required by Specification 6.8.1.

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

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'DMINISTRATIVE CONTROLS RECORD RETENTION (Continued)

Records of radioactive shipments.

Records of sealed source and fission detector leak tests and results.

h. Records of annual physical inventory of all sealed source material of record.

6.10.2 The following records shall be retained for the duration of the unit Operating License:

a. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the F'inal Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories.

c4 Records of reactor tests and experiments.

Records of radiation exposure for all individuals entering radiation control areas.

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or .operational cycles for those unit components identified in Table 5.7-1.

g. Records of training and qualification for current members of the unit staff.

h. Records of inservice inspections performed pursuant to these Technical Specifications.

Records of quality assurance activities required by the QA Manual.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

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k. Records, of meetings of the FRG and the CNRB.

Records of the service lives of all snubbers listed in Tables 3.7-4a

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and 3.7-4b including the date at which the service life commences and associated installation and .maintenance records.

Records of secondary water sampling and water quality.

n. Annual Radiological Environmental Operating Reports; and records of analyses transmitted to the licensee which are used to prepare the Annual Radiological Environmental Monitoring Report.

o. Meteorological data, summarized and reported in a format consistent with the recommendations of Regulatory Guides 1.21 and 1.23.

p. Records of audits performed under the requirements of Specifications 6.5.2.8 and 6.8.4.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained, and adhered to for all operations involving personnel radiation exposure.

ST. LUCIE - UNIT 6-23

ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA

6. 12. 1 In lieu of the "control device" or "alarm signal" required by para-graph 20.203(c)(2) of 10 CFR Part 20, each high radiation area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr 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)". 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.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a.preset integr ated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them.

C. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose greater than 1000 mrem shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative con"rol of the Shift Foreman on duty and/or health physics super-vision. 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 area and the maximum allowable stay time for individuals in that area. For individual areas accessible to personnel with radiation levels such that a major portion of the body could receive in one hour a dose in excess of 1000 mrem"" that are located within large areas, such as PWR containment, where no enclosure exists -for purposes of locking, and no enclosure can be reason-ably constructed around the individual areas, then that area shall be roped off, conspicuously posted and a flashing light shall be activated as a warning device. In lieu of the stay time specification of the RWP, direct or remote (such as use of closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities within the area.

Health Physics personnel or personnel escorted by Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas.

""Measurement made at 18 inches from source of radioactivity.

ST. LUCIE - UNIT I 6-24

ADHINIST ATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM PCP 6.13.1 The PCP shall be approved by the Commission prior to implementation.

6.13.2 Licensee initiated changes to the PCP:

1. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was made. This submittal shall contain:

a. Sufficiently detailed information to totally suport the rationale for the change without benefit of additional or supplemental information;

b. A determination that the change did not reduce the overall conformance of the dewatered bead resin to existing criteria for radioactive wastes; and

c. Documentation of the fact thatthe change has been reviewed and found acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM 6.14.1 The ODCH shall be approved by the FRG prior to implementation.

6.14'.2 Licensee initiated changes to the ODCH:

l. Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change(s) was

. made effective. This submittal shall contain:

aO Sufficiently detailed information to totally support the rationale for the change withbut benefit of additional or supple-mental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);

b. A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and

c. Documentation of the fact that the change has been reviewed and found acceptable by the FRG.

2. Shall become effective. upon review and acceptance by the FRG.

ST. LUCIE - UNIT I 6-25

ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS AND SOLID WASTE TREATMENT 5 STENS

6. 15. 1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

1. Shall be reported to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the evaluation was reviewed by the Facility Review Group. The discussion of each shall contain:

a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59.

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systems;

d. An evaluation of- the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto;

e. An evaluation of the change which shows the expected maximum exposures to individual in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; A comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

g. An estimate of the exposure to plant operating personnel as a result of the change; and

h. Documentation of the fact that the change was reviewed and found acceptable by the FRG.

2. Shall become effective upon review and acceptance by the FRG.

Licensees may chose to submit the information called for in this Specification as part of the annual FSAR update.

ST. LUCIE - UNIT 1 6-26

STATE OF FLORIDA )

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COUNTY OF DADE C. 0. Woody, being first duly sworn, deposes and says:

That he is Vice President Nuclear Operations of Florida Power 8 Light Company, the licensee herein; That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.

C. 0., ody Subscribed and sworn to before me this day of , 1984.

NOTARY PUBLIC, in and for the County of Dade, State of Florida.

Hy commission expires:

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