Text

Proposed Tech Specs Reflecting Administrative Updates, Removal of Outdated Matl,Minor Changes to Text & Correction of Errors to Achieve Consistency Throughout Tech Specs
	ML17222A473
Person / Time
Site:	Saint Lucie
Issue date:	09/07/1988
From:	; FLORIDA POWER & LIGHT CO.
To:	;
Shared Package
ML17222A472	List: ML17222A471; ML17222A473;
References
	NUDOCS 8809150199
	Download: ML17222A473 (70)
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ATTACHMENT 1 Marked-up St. Lucie Unit 1 Technical Specification Pages:

B 2-4 (plus insert)

B 2-5 3/4 3-4 3/4 3-11 3/4 3-15 3/4 3-19 3/4 7-44 3/4 9-1 3/4 9-3 3/4 9-5 3/4 12-1 3/4 12-11 3/4 12-12 6-12 8809i 5049'9'80907 PDR ADOCN P 05000336

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2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SETPOINTS The Reactor Trip Setpoints specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter. The Trip Values have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits. Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for each trip in the safety analyses.

Manual Reactor Tri The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Level-Hi h The Power Level-High trip provides reactor core protection against reactivity excursions which are too rapid to be protected by a Pressurizer Pressure-High or Thermal Margin/Low Pressure trip.

The Power Level-High trip setposnt is operator adjustable and can be set no higher than 9.61'5 above the indicated THERMAL POWER level.

Operator action is required to increase the trip setpoint as THERMAL POWER is increased. The trip setpoint is automatically decreased as THERMAL POWER decreases. The trip setpoint has a maximum value of 107.0" of RATED THERMAL POWER and a minimum setpoint of 155 of RATED THERMAL POWER. Adding to this maximum value the possible variation in point due to calibration and instrument errors, the maximum. actual 'rip THERMAL POWER level. at which a trip would be actuated is 112>> of RATED POWER, which is consistent with the- value used in the safety 'HERMAL analysis.

Reactor Coolant Flow-Low ST. LUCIE -

O'~

UNIT 4 p I '8 4

's-i~been-made-in-4)e-maa4o~ro4eot-~

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Amendment No. $ 7, 4 8 LRn'l

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r St. Lucie Unit-1 Technical Specifications Insert p. B 2-4 The Reactor Coolant Flow Low trip provides core protection against DNB in the event of a sudden significant decrease in RCS flow. The reactor trip setpoint on low RCS flow is calculated by a relationship between steam generator differential pressure, core inlet temperature, instrument errors and response times. When the calculated RCS flow falls below the trip setpoint an automatic reactor trip signal is initiated. The trip setpoint and allowable values ensure that for a degradation of RCS flow resulting from expected transients, a reactor trip occurs to prevent violation of local power density or DNBR safety limits.

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2.2 LIMITING SAFETY SYSTEH SETTINGS'ASES Dc lc.W hi~

Reactor Coolant Flow-Low (Continued) jns ~f Pllopa>~

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4 Pressurizer Pressure-Hi h The Pressurizer Pressure-High trip, backed up by the pressurizer code .,

safety valves and main steam line safety valves, provides reactor coolant system protection against overpressurization in the event of loss of load without reactor trip. This trip's setpoint is 100 psi below the nominal lift setting (2500 psia) of the pressurizer code safety valves and its con-current operation with the power-operated relief valves avoids <<he undesir-able operation of the pressurizer code safety valves.

Containment Pressure-Hi h ~

The Containment Pressure High trip provides assurance that a reactor trip is initiated concurrently with a safety injection.

Steam Generator Pressure-Low The Steam Generator Pressure-Low trip provides protection against an excessive rate of heat extraction from the steam generators and sub-sequent cooldown of the reactor coolant. The setting of 600 psia is sufficiently below -the full-load operating point -of 800 psig so as not ST. LUCIE - L'NIT 1 B 2-5 Amendment No. g2, P$ , P5, 63

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TABLE 3.3-1 Continued TABLE NOTATION

With the protectiv0 system trip breakers in the closed position and the CEA drive system capable of CEA withdrawal.

'he provisions of Specification 3.0.4 are not applicable.

(a) Trip may be bypassed below 1% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 1% of RATED THERMAL POWER.

(b) Trip may be manually bypassed below 685 psig; bypass shall. be.:

automatically removed at or above 685 psig.

(c) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is > 155 of RATED THERMAL POWER.

(d) Trip may be bypassed bel'ow 10 5 and above 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL power is > 10 l or < 15% of RATED THERMAL POWER.

Oclc.kd (e)

(f) There shall be at least two decades of overlap between the Wide Range Logarithmic Neutron Flux Monitoring Channels and the Power Range Neutron Flux Moni.toring Channels ACTION'TATEMENTS'ith ACTION 1 the number of channels OPERABLE one less than required by the Minimum. Channels. OPERABLE requiregeat, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6'ours.

and/or open the protective system trip breakers.

ACTION 2 - With the number of OPERABLE channels 'one less than the Total Number of Channels, STARTUP and/or POWER OPERATION

'ay proceed provided the following conditions are satisfied:

a 0 The. inoperable channel is placed in either the bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable channel may be- bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . from time of initial loss of OPERABILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

ST. LUG IE - UNIT 1 3/4 3-4 Amendment No. J$ , g7, 45

TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION I MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE m FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

5. CONTAINMENT SUMP RECIRCULATION (RAS)

a. Manual RAS (Trip Buttons) 1,2,3,4 8

b. Refueling Water Tank - Low 1 ~ 2 3 9¹

6. LOSS OF POWER

a. 4.16 kv Emergency Bus Under-voltage (Loss of Yoltage) 2/Bus 2/Bus 1/Bus 1, 2, 3 12

b. 4.16 kv Emergency Bus Under- Dol~j voltage (Degraded Voltage)

Cog (1) Undervoltage Device ¹1, 2/Bus 2/Bus 1/Bus 1 2.3 12 I

(2) Undervoltage Device ¹2 2/Bus 2/Bus 1/Bus 1,2,3 12

c. 480 V Emergency Bus Under-voltage (Degraded Voltage)~ .2/Bus 2/Bus 1/Bus l. 2. 3 12

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) 4/SG 2/SG 4/SG le Zs 11

b. Automatic Actuation Logic 4/SG 2/SG 3/SG 1 I 8

c. SG Level (1A/1B) - Low 4/SG 2/SG 3/SG 1 ~ 2 3 13¹, 14

8. AUXILIARY FEEDWATER ISOLATION O " a.'G 1A - SG 1B Differential

'hogg Pressure 4/SG 2/SG 3/SG 1,2,3 13¹, 14

b. Feedwater Header tA SG lA - SG 1B Differential le Pressure 4/SG 2/SG . 3/SG 1,2,3 13¹, 14 M

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TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE FUNCTIONAL UNIT TRIP VALUE VALUES

6. LOSS OF POWER

a. (1) 4.16 kv bnergenc (Loss of Voltage)

Bus Undervoltage

'900 1 +

+ 29 volts with a

.5 second time delay 2900 + 29 volts with a 1 + .5 second time delay t

b. 4.16 kv Emergency.Bus Undervoltage (Degraded Voltage) gC (1) Uedervettage Oevtce gl~ 3675 + 36 7 + 1 volts with a minute time delay 3675 + 36 7 + 1 volts with a minute time delay (2) Undervoltage Device f2 .3592 + 36 volts with a 3592 + 36 volts with a 18 + 2 second time delay 18 + 2 second time delay

c. 480 volts Emergen Bus Undervoltage (Degraded Voltage) 429 + 5-0 volts with a 429 + 5 -0 volts with a 7 + 1 second time delay 7 + 1 second time delay

7. AUXILIARY FEEDWATER (AFAS)-

a. Manual (Trip Buttons) Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applidab]e

c. SG 1A 8 1B Level Low >29.0X >28.5X

8. AUXILIARY FEEDWATER ISOLATION

a. Steam Generator hP-High <275 psid <281 psid

b. Feedwater Header High hP <150.0 psid <157.5 psid

I TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INS RUMENTATION SURVEILLANCE RE UIREMENTS MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST -

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6. LOSS OF POWER a . 4.16 kv Emergency Bus Un er-voltage (Loss og"- Voltage) S 1, 2, 3

b. 4.16 kv Emergency Bus Under-voltage (Degraded Voltage) D~~~

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(1) Undervoltage Devtce tlag S 1, 2, 3

'a (2) Undervoltage Devtce 22~ S 1, 2, 3 C. 480 V Emergency Bus Under- S 1, 2, 3 voltage (Degraded Voltage)~

7. AUXILIARY FEEDWATER (AFAS)

a. Manual (Trip Buttons) N.A. N.A. 1, 2, 3

b. SG Level (A/8) - Low 1, 2, 3

c. Automatic Actuation Logic N.A. N.A. 1, 2, 3

8. AUXILIARY FEEDWATER ISOLATION

a. .SG Level (A/B) - Low and SG Differential Pressure (BtoA/AtoB) - High N.A. 1, 2, 3

b. SG Level (A/B) - Low and Feedwater Header Differential Pressure (BtoA/AtoB) - High I

N.A. 1,2,3

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TABLE 3.7-3 FIRE HOSE STATIONS A. Hose Stations (Turbfne Bui1dfng)

Operating F1oor (northeast corner)

2. Operatfng Floor (southeast corner)

3. Oper atfng F1oor (mfdd1e east sfde)

8. Hose Stations (Reactor Auxf1fary Hufldfng) 43 ft. 1evel south wall of HVK room

"~" s~~~ N~ Wy Rol4~ ~'"

2. 43 ft. level "b w'~a

3. 43.ft. 1evel southwest corner of'oor.

near

4. 43 ft. 1evel cable spreading room " west wall

5. 19.5 ft. 1evel east end of east-west hall

6. 19.5 ft. level midd1e of east-west hall T. 19.5 ft. 1evel south- end of north-south hall

8. '9.5 ft. 1evel entranci hall on south wall

9. -5 ft. 1evel east end of hall

10. -5 ft. 1evel south wall of hall near AC 182

11. -5 ft. 1evel west end of ha11 ST. LUCIE - UN?T 1 3/4 7-44 Amendment No. 2$ ,

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3/4. 9 REFUELING OPERATIONS BORON CONCENTRATION LIMITING CONDITION FOR OPERATION 3.9.1 With the reactor vessel head unbolted or removed, the boron concentration of all filled portions of the Reactor Coolant System and the refueling cavity shall be maintained uniform and sufficient to ensure that the more restrictive of following reactivity conditions is met:

a. Either a Keff of 0.95 or less, which includes a conservative allowance ior uncertainties, or

b. A boron concentration of > 1720 ppm, which includes a 50 ppm conservative allowance for uncertainties.'PPLICABILITY:

MODE 6*.

ACTION'ith the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at > 40 gpm of 1720 ppm boron or its equivalent until K ++ is reduced to ( 0.95 or the boron concentration is restored to > f720 ppm, whichever is the more restrictive.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.1.1. The more restrictive of the above two reactivity conditions shall be determined prior to:

a. Removing or unbolting the reactor vessel head, and

b. Withdrawal of any full length CEA in excess of 3 feet from its fully inserted position.

4.9.1.2 The boron concentration of the refueling cavity shall be determined by chemical analysis at least 3 times per 7 days with a maximum time interval between samples of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

"The reactor shall be maintained in MODE 6 when the reactor vessel head is unbolted or removed.

ST. LUG IE - UNIT 1 3/4 9-1

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'P BB REFUELING OPERATIONS DECAY TIME LIMITING CONDITION FOR OPERATION 3.9.3 The reactor shall be subcritical for a minimum of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY: B I B pressure vessel.

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With the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movement of irradiated fuel in the reactor pres-sure vessel. The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.3 The reactor shall be determined to have been subcritical for at least .72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.

ST. LUCIE - UNIT 1 3/4 9-3

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REFUELING OPERATIONS COMMUNICATIONS LIMITING CONDITION FOR OPERATION 3.9.5 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During CORE ALTERATIOHS.

ACTION:

When direct conmunications between the control room and personnel at the refueling station cannot be maintained, suspend all CORE ALTERATIONS.

The provisions Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.9.5 Direct copmunications between the control room and personnel at the refueling station shall be demonstrated within one hour prior to the start of and at least once per 1Z hours during CORE ALTERATIONS.

ST. LUCIE - UNIT 1 3/4 9-5

3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3 4.12.1 MONITORING 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 0 With the radiological environmental onitoring program not being con-ducted as specified in Table 3.12- , prepare and submit to the I Commission, in the Annual Radiolo ical Environmental Operating Report required by Specification 6.9.1. a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b With 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 Coamission within 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. When more than one of the radionuclides in Ta'ble 3.12-2 are detected in the sampling medium, this report shall be submitted if:

concentration 1 + concentration 2 + ... > 1.0 reporting level 1 reporting level 2 When radionuclides other than those in Table 3.12-2 are detected and are the result of plant effluents, this report 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 r eport is not required if the measured level of radioactivity was not the r esult of plant effluents; however, in such an event, the condition shall be reported and descr ibed iri the Annual Radiological Environmental Operating Report.

c~ With 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. LUCIE - UNIT 1 3/4 12-1 Amendment No. //4s 69

RADIOLOGICAL ENVIRONMENTAL HOtt ITORING 3/4.12. 2 LAtlD USE. CENSUS LIMITING 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 all times. /

ACTION:

a. With a land use census identifying a ocation(s) that yields a cal-culated dose or dose commitment gre er than the values currently being calculated in Specification .11.2.3, identify the new loca-tion(s) in the next Semiannual Ra oactive Effluent Release Report, pursuant to Specification 6.9.1.

b. With a land use census identifying a location(s) that yields a cal-culated dose or dose coanitment (via the same exposure pathway) MX greater than at a location from which samples ar e currently in accordance with Specification 3.12.1, add the new loca-being'btained tion(s) to the radiological environmental monitoring program withi 30 days. The sampling location(s), excluding the control stat location, having the lowest calculated dose or dose comni t(s).

via the same exposure pathway, may be deleted from this onitoring program after October 31 of the year in which rhis nd use census was conducted. Pursuant to Specification 6.9.1. , identify the new location(s) fn the next Semiannual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the OOCH reflecting the new location(s).

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.12.2 The land use census shall be conducted during th'e growing season at least once per 12 months using that information that will provide the best results, such as by a door-to-door survey, aer ial survey, or by consulting local agriculture authorities. The results of the land use census shall be included in the Annual adiological Environmental 'Operating Report pursuant to Specification 6.9.1.

8

Broad leaf vegetation sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted D/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. $ /S.

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RAOIOLOG I CAL ENVIRONMENTAL MONITORING 3 4.12.3 INTERLABORATORY COMPARISON PROGRAM il LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission.*

APPLICABILITY: At all times.

ACTI OII:

a. With analyses not being performed as requfred above, report he corrective actions to the Commission in the Annual Radiologic Environmental Operating Report pursuant to Specification-6.9;l.

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

SURVEILLANCE RE UIREMENTS 4.12.3 A summary of the results obtained as part of the above required Inter-laboratory Comparison Program shall be included fn the Annual Radiological Environmental'Operating Report pursuant to Specification 6.9.1.

This condition is satisfied by participation fn the Environmental Radioactivity Laboratory Intercomparison Studies Program conducted by the Xn%fronmental Protection Agency (EPA).

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

5pc'cg~PGA4 0 J ADMINISTRATIVE CONTROLS

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=nose 2 9 The CNRB shall report to and advise the xecutive Vice President on areas of responsibility specifed in Specifi ations 6.5.2.7 and 6.5.2.8.

'l; RECOROS ll 6.5.2.10 i Records of CNRB activities shall be pr pared, approved and distrib-i i ~ted as indicated below:

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a. Minutes of each CNRB meeting shall be repared, approved and t

forwarded to the Executive Vice Presi en within 14 days following each meeting.

b. Reports of reviews encompassed by~.'.5.2.? above, shall 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 of the audit by the auditing organization. 'ompletion 6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Coamission shall be notified and a repor t submitted pursuant

o the requirements of Section 50.73 to 10 CFR Part 50, and Each REPORTABLE EVENT shall be reviewed by the FRG, and the results of the review shall be submitted to the CNRB, and the Senior Vice President - Nuclear.

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

a. The NRC Operat'ions Center shall be nogffied by telephone as soon as possible and in all cases within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . The Senior Vice President - Nuclear and the CNBR sh'all be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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.

ST. LUCIE - UNIT 1 6-12 Amendment No. ~~a ~3

E ATTACHMENT 2 Marked-up St. Lucie Unit 2 Technical Specification Pages B 2-1 B 2-6 3/4 3-41 3/4 3-45 (plus insert) 3/4 3-46 3/4 6-1 3/4 7-16 3/4 7-22 3/4 7-32 3/4 7-36 3/4 7-37 3/4 12-1 3/4 12-11 3/4 12-12 6-23

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2.1 SAFETY LIMITS BASES

2. 1. 1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel clad-ding and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel is prevented by maintaining the steady-state peak linear heat rate below the level at which centerline fuel melting wi 11 occur. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coeffi-cient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the CE-1 correlation. The CE-1 DNB correlation has been developed to predict the DNB heat flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as.

the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal opera-tional transients, and anticipated transients is limited to 1.28. This value is derived through a statistical combination of the system parameter probability distribution functions with the CE-1 DNB correlation uncertainty. This value corresponds to a 95K probability at a 95K confidence level that DNB will not occur and is chosen as an appropriate margin to DNB .for all operating conditions.

The curves of Figure 2.1-1 shoQ the loci of points of THERMAL POWER, Reactor Coolant System pressure and maximum cold leg temperature with four Reactor Cool-ant Pumps operating for which the minimum DNBR is no less than 1.28 for the family of axial shapes and corresponding radial peaks shown in Figure B 2.1-1.

The limits in Figure 2.1-1 wer e calculated for reactor coolant inlet temperatures less than or equal to 580'F. The dashed line at 580'F coolant inlet temperature is not a safety limit; however, operation above 580'F is not possible because of the actuation of the main steam line safety valves which limit the maximum value of reactor inlet temperature. Reactor operation at THERMAL POWER levels higher than 112% of RATED THERMAL POWER is prohibited by the high power level trip set-point ~c+fed- in Table 2.2-1. The area of safe operation is below and to the sf'<<t"<c~ 1 o f these lines.

The conditions for the Thermal Margin Safety Limit curves in Figure 2.1-1 to be valid are shown on the figure.

The Thermal Margin/Low Pressure and Local Power Density Trip Systems, in conjunction with Limiting Conditions for Operation, the Yariable Overpower Trip and the Power Dependent Insertion Limits, assure that the Specified Acceptable g i A d~0 Fuel Design Limits on DNB and Fuel Centerline Mel't are not exceeded during normal Operate>al ST. LUCIE-UNIT 2 B 2-1 Amendment No. 8

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SAFETY LIMITS ANO L MITING SAFETY SYSTEM SETTINGS BASES RCP Loss or Com onent- Coolin Water A,loss of component cooling water to the reactor coolant pumps causes a delayed reactor trip. This trip provides protection to the reactor coolant pumps by ensuring that plant operation is not continued without cooling water available. The trip is delayed 10 minutes following a reduction in flow to below the trip setpoint and the trip does not occur if flow is restored before i0 minutes elapses. No credit was taken for this trip in the safety analysis.

rts functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protective System.

Rate of Chan e of Power-Hi h The Rate of Change of Power-High trip is provided to protect the core .

during startup operations and its use serves as a backup to the administra-tively enforced startup rate limit. Its trip setpoint does not correspond to a Safety Limit and no credit was taken in the safety analyses for operation of his trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Rea'ctor Protection System.

Reactor Coolant Flow - Low or Coolant Flow - Low trip provides protection against a reactor coolant pump shear event and a two pump opposite loop flow coas event. A trip is initiated w ressure differential e primary side of either steam generator decreases variable setpoint stays a set amoun limited by a set maximu e

e pressure se rate or a set minimum value.

'tial e setpoint.

T e This unless

'ed setpoint ens a a reactor trip occurs to prevent violation of local nsity or ONBR safety 1imitd'nder the stated conditions.

The Reactox Coolant Flow - Low trip provides core protection against DNB in the event of a sudden significant decrease in RCS flow. The reactor trip setpoint on low RCS flow is calculated by a relationship between steam generatox differential px'essure, core inlet temperature, instrument errors and response times.

When the calculated RCS flow falls below the trip setpoint an automatic reactor trip signal is initiated. The trip setpoint and allowable values ensure that fox a degradation of RCS flow resulting from expected transients, a reactor trip occurs to pxevent violation of local power density or DNBR safety limits.

ST. LUCIE - UNIT 2 8 2-6

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-1O shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.* With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in able 3.3-10, either restore the inoperab'le channel to OPE BLE status within 7 days, or be in HOT SHUTDOWN within the n t 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.* With the number of OPERABLE accident mon toring channels less than the Minimum Channels OPERABLE requirem ts of Table 3.3-10, either restore the inoperable channel(s) to ERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN with'he next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.** With the number of OPERABLE Chan 's ane less than the Total Number F

of Channels shown in Table 3.3- , either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

r io d.~ With the number of OPERABLE Channe less than the Minimum Channels requirements of Table 3.3- , &/ther restore the inoperable if

'PERABLE channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months repairs are feasible without shutting down or:

l. Initiate an alternate method of monitoring the reactor vessel inventory; and 2.: Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and

3. Restore the Channel to OPERABLE status at the next scheduled refueling.

~ e. The provisions of Specification 3.0.4 are not applicable.'ction statements do not apply to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

- Action statements apply only to Reactor Vessel Level Monitoring System.

Containment Sump Water Level (narrow range) and Containment Sump Water .

Level (wide range) instruments.

ST. LUG IE - UNIT 2 3/4 3-41 Amendment No. tK

TABLE 3. 3-11 FIRE O'ETECTION INSTRUMENTS INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS" HEAT SMOKE (x/y) (x/y)

REACTOR AUXILIARY BUILDING ZONE-1A REACTOR AUX. SLOG. EL. 0.50 6/0 ZONE"2A REACTOR AUX. BLDG..KL. 0.50 4/0 ZONE" EACTOR AUX. BLDG. EL. 19.50 6/0 ZON 4 ACTOR AUX. BLDG. EL. 19. 50 5/0 ZONE- REACTOR AUX. BLDG. EL. 19.50 8/0 ZONE-6A REACTOR AUX. BLDG. EL. 43.00 5/0 ZONE-7A REACTOR AUX. BLDG. EL. 43.00 7/0 ZONE"BA REACTOR AUX. BLDG. EL. 62.00 ZONE-9A REACTOR AUX. BLDG. EL. 43.00 (o 6/0 2/0 ZONE-10A REACTOR AUX. BLDG. EL. 43.00 2/0

~ONE-18 REACTOR AUX. BLDG. EL. 0.50 6/0 ZONE-2B REACTOR, AUX. BLDG. EL. 0.50 5/0 ZONE-38 REACTOR AUX. BLDG. EL. 19.50 6/0 ZONE-48 HSCP-1 REC. AUX. BLDG. EL. 43.00 1/0 ZONE-58 REACTOR AUX. BLDG. KL. 19.50 6/0 ZONE-68-REACTOR AUX. BLDG. EL. 43.00 4/0 ZONE-78 REACTOR AUX. BLDG. EL. 43.00 6/0 ZONE-88 REACTOR AUX. BLDG. EL. 62.00 5/0 ZONE-98 REACTOR AUX. BLDG. EL. 43.00 2/0 ZONE-108 REACTOR AUX. SLOG. EL. 43.00 2/0

~~gf OZONE-1F ZONK-2F FAN ROOM KL. 43.00 CABLE LOFT EL. 19.50 0/2 0/26 ZONE-3F IODINE REMOVAL/WASTE GAS/

HALLWAYS EL. 0.50 0/15 ZONE-4F 8 ELECTRICAL PENETRATION ROOM EL. 19.50 0/2 ZONE"5F A ELECTRICAL PENETRATION ROOM EL. 19.50 0/1 ZONE-6F CABLE SPREADING ROOM EL; 43.00 0/9 FUEL HANDLING BUILDING ZONE 20A FUEL HANDLING BLDG. EL. 19.50 I/O ZONE-21A FUEL HANDLING BLDG. EL. 48.00 3/0 ZONE-208 FUEL HANOLING BLDG. EL. 19.50 1/0 ZONE-218 FUEL HANDLING BLDG. EL. 48.00 2/0 DIESEL GENERATOR BUILDING ZONE-22A DIESEL GEN. BLDG./O.o. s~oe4aC <hM 2/2 2/0 ZONE-228 DIESEL GEN. BLDG./g.o, cvoRP6t 'fALL 2l2 2lo (AGe~

ST. LUCIE - UNIT 2 3/4 3-45,

INSERTS FOR TABLE 3.3-11 SMOKE (x/y) gl: Zone-12A Elect. Pen. Room EL. 19.50 3/0 g2: Zone-12B Elect. Pen. Room EL. 19.50 4/0

$ 3: Zone-llA Annulus 1/0 54: Zone-11B Annulus 1/0

TABLE 3. 3" 11 (Continued)

FIRE DETECTION INSTRUMENTS INSTRUMENT LOCATION TOTAL NUMBER OF INSTRUMENTS" HEAT SMOKE

~(x y) (x/y)

SAFETY RELATED PUMPS ZONE-17A COMPONENT COOLING AREA 4/0 ZONE"18A INTAKE COOLING WATER PUMP AREA 1/0 ZONE-]9A STEAM TRESTLE AREA-AUX.

FEEDWATER PUMP 2/0 ZONE"17B COMPONENT COOLING AREA 2/0 ZONE-188 INTAKE COOLING WATER PUMP AREA ZONE"198 STEAM TRESTLE AREA"AUX.

FEEDWATER PUMP 2/0 TURBINE BUILDING/SWITCHGEAR ROOM ZONE-16A TURBINE BLDG. SWITCHGEAR ROOM 3/0 ZONE-168 TURBINF. BLDG. SWITCHGEAR ROOM 3/0 CONTAIHMENT ZONE-13A REACTOR TUNNEL BELOW EL. 18.00 2/0 ZONE-14A REACTOR EL. 18.00 5/0 ZONE"15A REACTOR EL. 45.00 2/0 4/0 (yg~~ONE"138 REACTOR'TUNNEL BELOW EL. 18.00 1/0 ZONE-14B REACTOR EL. 18.00 5/0 ZONE-15B REACTOR EL. 45.00 2/0 5/0 CABLE SPREADIHG ZONE-11A ELECT. PENET. REC. (ANNULUS) 1/0 ZONE-12A ELECT. PENET. REC. AUX. BLDG.

EL. 19.50 3/0 ZONE-llB ELECT. PENET. REC. (ANNULUS) 1/0 ZONE-12B ELECT. PENET. REC. AUX. BLDG.

EL. 19.50 4/0 (x/y): x is number of early warning fire detection and notification only instruments. c-y is number of actuation of fire suppression systems and early warning and notification) instruments.

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

ST. LUCIE - UNIT 2 3/4 3-46

3/4.6 CONTAINMENT SYSTEMS 3/4. 6. 1 PRIl1ARY CONTAINtlENT CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6. l. 1 Primary CONTAINMENT INTEGRITY shall be maintained.

APPLICABILITY: MODES 1", 2', 3, and 4.

ACTION:

Without primary CONTAINMENT INTEGRITY, restore CONTAINMENT INTEGRITY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS

4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstr ated:

a~ At least once per 31 days by verifying that all penetrations"" not capable of being closed by OPERABLE containment automatic isolation valves and required to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in .their p itio s, except as provided in Table 3.6-2 of Specification 3.6

b. By verifying that each containment air lock is in compliance with the requirements of Specification 3.6. 1.3.

C. After each closing of each penetration subject to Type B testing, except containment air locks, if opened following a Type A or 8 test, by leak rate testing the seal with gas at P 43.4 psig and verifying that when the measured leakage rate for these seals is added to the leakage rates determined pursuant to Specifica-tion 4.6. 1.2d. for all other Type 8 and C penetrations, the combined leakage rate is 'less than or equal to 0.60 L .

In MODES 1 and 2, the RCB polar crane shall be rendered inoperable by locking the power supply breaker open.

Except valves, blind flanges, and deactivated automatic valves which are located inside the containment and are locked, sealed or otherwise secured in the closed position. These penetrations shall be verified closed during each COLD SHUTDOWN except that such verification need

- not be performed more often than once per 92 days.

ST. LUCIE - UNIT 2 3/4 6-1

~ ~

PLANTS SYSTEMS 3/4.7.6 FLOOD PROTECTION LIMITING CONDITION FOR OPERATION 3.7.6.1 Flood protection shall be provided for the facility site via stoplogs which shall be installed on the southside of the RAB and the southernmost door on east wall whenever a hurricane warning for the plant is posted.

APPLICABILITY: At all times.

ACTION:

With either a Hurricane Watch or a Hurricane Warning issued for the facility site, perform the St. Lucie Plant Beach Survey Procedure pursuant to Surveil-lance Requirement 4.7.6.1.1 below and ensure the stoplogs are removed from storage and are prepared for installation. The stoplogs shall be installed anytime a hurricane warning is posted.

SURVEILLANCE RE UIREMENTS 4.7.6. 1.1 The St. Lucie Plant Beach Survey Procedure shall be conducted at least once per year between the dates of May 25 and June 7 and within 30 days following the termination of either a Hurricane Watch or a Hurricane Warning for the facility site. A Special Report containing the results of these surveys shall be prepared and submitted to the Commission pursuant to Specifi-cation 6.9.2 within 30 days following the completion of the survey. +he-~

4 4.7.6.1.2 The St. Lucie Mangrove Photographic Survey Procedure shall be conducted at least once per 12 months and shall be a color infrared photo-graph(s), or equivalent, of the mangrove area between the facility and the FPEL east property line. The results of these surveys shall be included in the Annual Operating Report for the period in which the survey was completed.

This report shall include an evaluation of the facility flood protection if the survey indicates deterioration, either man-made or natural, of this mangrove area.

4.7.6. 1.3 Meteorological forecasts shall be obtained from the National Hurricane Center in Miami, Florida at least once per 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months during either a Hurricane Watch or a Hurricane Warning.

ST. LUCIE - UNIT 2 3/4 7-16

~ ~

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued C ~ Refuelin Outa e Ins ecttons 4

At least once per 18 months an inspection shall be performed of all safety related snubbers attached to sections of safety systems piping that have experienced unexpected, potentially damaging transients as determined from a review of operational data and a visual inspection of the systems. In addition to satisfying the visual inspection acceptance criteria, freedom of motion of mechanical snubbers shall be verified using one of the following:

(1) manually induced snubber movement; (2) evaluation of in-place snubber piston setting; (3) stroking the mechanical snubber through its full range of travel.

d. Visual Ins ection Acce tance Criteria Visual inspections shall verify (1) that'there are no visble indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure. Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is'learly established and remedied for that particular snubber and for other snubbers, irrespective of type, that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE per Specification 4.7.9f. When a fluid port of a hydraulic snubber is found to be, uncovered the snubber shall .be declared inoperable and cannot be determined OPERABLE via functional'testing unless the test is started with the piston in the as found setting, extending the piston rod in'he tension mode direction. All snubbers connected to an inoperable common hydraulic fluid reservoir shall be determined to be OPERABLE by visually verifying the required level of oil for operation for each affected'snubber; otherwise declare the snubbers inoperable.

e. Functional Tes.ts /ea.sk Our ing the first refueling shutdown and atl~ once per 18 months thereafter during shutdown, a representative sample of either:

(1) At least 10% of the total of each type of safety related snubber in use in the plant shall be functionally tested either in place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Specification 4.7.9f. an additional 1OX of that type of snubber shall be functionally tested until no more failures are found or until all snubbers of that type have been functionally tested or (2) A representative sample of each type of snubber shall be functionally tested in accordance with Figure 4.7-1.

"C" is the total number of snubbers of a type found not meeting the acceptance requirements of Specification 4.7.9f. The cumulative number of snubbers of a type tested is denoted by "N." At the end of each day's testing, the new values of "N" and "C" (previous day' total plus current day's increments) shall be plotted on Figure 4.7-1.

If at any time the point plotted falls in the "Reject" region, all ST. LUCIE - UNIT 2 3/4 7-22 Amendment No. 22

PLANT SYSTEMS SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION

3. 7. 11. 2 The following sprinkler systems shall be OPERABLE:

1. Fire Zone 8 - Oiesel Generator Building 2A

2. Fire Zone 9 - Oiesel Generator Building 2B

3. Fire Zone 19 - RAB East Hallway and Miscellaneous Equipment Areas&

4. Fire Zone 20 - RAB East-West Common Hallway%

5. Fire Zone 22 - RAB Electr ical Penetration Area~

6. Fire Zone 23 - RAB Electrical Penetration Arear'.

Fire Zone 39 - RAB HVAC Equipment Room+

g.'.P.

Fire Zone 51 - RAB Ceiling and Hallways%

~ Fire Zone 52 - Cable Spreading Room+

APPLICAB ILITY::Whenever equipment protected by the sprinkler system is required to be OPERABLE.

ACTION:

a. With one or more of the above required sprinkler systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

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

SURVEILLANCE REOUIREMENTS 4.7. 11.2 Each of the above required sprinkler systems shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path is in its correct.

position.

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel.

ST. LUCIE " UNIT 2 3/4 7"32

TABLE 3.7-4 FIRE HOSE STATIONS LOCATION/ALE'!4T ION HOSC RACK ~

A. Hose Stations (Turbine 8uilding)

1. Operating Floor (northeast corner) HS-15-4

2. Operating Floor (southeast corner) HS-15-10

3. Operating Floor (middle east side) HS-15-7 Hose Stations (Reactor Auxiliary Building)

1. 62 ft level east wall entrance HS-15-44

2. 62 ft level west wall entrance HS-15-45 62 ft level west wall entrance to HEV

\

3. room HS-15-46

4. 43 ft level 6X. Ce mc HS-15-36

5. 43 ft level south wall of HEV room HS-15-37

6. 43 ft level cable spreading room HS-15-31

7. 43 ft l eve southwest corner of 1

~

g

@~near door HS-15-42

8. 19.5 ft level east end of east-west hall HS-15-38

9. 19.5 ft level middle of east-west hall HS-15-40

10. 19.5 ft level south end of north-south hall HS-15-33

11. 19.5 ft level entrance hall on south wall HS-15-34

12. -0.5 ft level east end of hall HS-15-41

13. -0.5 ft level south wall of hall HS-15-28

14. -0.5 ft level west end of hall HS-15-43 C. Hose Stations (React'or Containment Building)

1. RCB at 23 ft level (near stairway no. 3) HS-15-47

2. RCB at 45 ft level (near stairway no. 1) HS-15-48

3. RCB at 45 ft level (near stairway no. 2) HS-15-54

4. RCB at 62 ft level (near stairway no. 3) HS-15-49

0. Hose Station (Fuel Handling Building) 62 ft level northwest corner HS-15-55 ST. LUCIE - UNIT 2 3/4 7-36

1

~ ~

PLANT SYSTEMS YARO FIRE HYORANTS ANO HYORANT HOSE HOUSES LIMITING CONOITION FOR OPERATION 3.7. 11.4 The yard fire hydrants and associated hydrant hose houses shown in Table 3.7-5 shall be OPERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the yard fire hydrants is required to be OPERABLE.

ACTION:

a. With one or more of the yard fire hydrants or associated hydrant hose houses shown in Table 3.7-5 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient additional lengths of 2 1/2 inch diameter hose located in an adjacent OPERABLE hydrant hose house to provide service to the unprotected. area(s) if the inoperable fire hydrant or associated hydrant hose house is the primary means of fire suppression; otherwise provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. The provisions of Specifications 3.0.3 and 3.0. 4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7. 11.4 Each of the yard fire hydrants and associated hydrant hose houses shown in Table 3.7-5 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

b. At least once per 6 months by visually inspecting each yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.

c~ At least once per 12 months by:

1. Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.

2. Inspecting all the gaskets and replacing any degraded gaskets in the couplings.

3. Performing a flow check of each hydrant to verify its OPERABILITY.

ST. LUCIE - UNIT 2 3/4 7-37

0 3/4. 12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12. 1 MONITORING 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 al 8

1 times.

ACTION:

a. Nith the radiological environmental onitor 'ng .program not being conducted as specified in Table 3. -1, prepare and submit to the Commission, in the Annual Radiolo cal Environmental Operating Report required by Specification 6,9.1. a description of the reasons for not conducting the program as required and the plans for preventing a recurrence.

b. With 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 Comnission within 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. When 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 +

reporting eve reporting eve When radionuclides other than those in Table 3. 12-2 are detected and are the result of plant effluents, this report shall be submitted the potential annual dose to A HEHBER OF THE PUBLIC is equal to or if 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 radio-activity 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. With milk or broadleaf vegetation samples unavailable from one or more of the sample locations required by Table 3.12-1, identjfy locations for obtaining replacement samples and add them to the radiological environmental monitoring program within 30 days. The specific 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. LUCIE - UNIT 2 3/4 12-1 Amendment No. 13

~ ( ~

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12.2 LAND USE CENSUS LIMITING 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 all times. 7 ACTION:

a~ With a land use census ide ifying a location(s) that yields a calculated dose, or .dose c itment greater than the values currently being calculated in Spec fication 4.11.2.3, identify the new location(s) in the next Semiannual adioactive Effluent Release Report, pursuant to Specification 6.9. l.

b. With a land use census identifying a location(s) that yields a calculated dose or dose commitment (via the same exposure pathway) 2'reater than at a location from which samples are currently being obtained in accordance with Specification 3. 12. 1, add the new location(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 commitment(s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which thi land use census was conducted. Pursuant to Specification 6.9.1.

identify the new location(s) in the next Semiannual Radioactive fluent Release Report and also include in the report a revised figur'e(s) and table 'for the ODCM reflecting the new location(s).

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

SURVEILLANCE RE UIREMENTS

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 results, 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.

"Broad leaf vegetation sampling may be performed at the site boundary in each of two different direction sectors with the highest predicted Dt's 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 2 3/4 12-11 Amendment No. I3

~ ( ~

RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4. 12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION

3. l2.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission."

APPLICABILITY: At all times.

ACTION:

a~ With analyses not being performed as required above, report th corrective actions to the Commission in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1.

b. The provisions of Specifications 3. 0. 3 and 3. 0.4 are not applicable.

SURVEILLANCE RE UI REMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1.

This condition is satisfied by participation in the Environmental Radioactivity Laboratory Intercomparison Studies Program conducted by the Environmental Protection Agency (EPA).

ST. LUCIE - UNIT 2 3/4 12-12

I ~ ~

a 4'j v

ADHINIST TIVE CONTROLS

6. 13 PROCESS CONTROL PROGRAM PCP Licensee initiated changes to the PCP:

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 that the change has been reviewed anl found acceptable by the FRG.

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

6. 14 OFFSITE DOSE CALCULATION MANUAL ODCM GQC Pf Licensee initiated changes to the

1. 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:

a. Sufficiently detailed information to totally support the rationale for the change without 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 2 6-23 Amendment No. ~~ '

ATTACHMENT 3 St. Lucie Unit 2 Operating License NPF-16 License Condition 2.C.12

~ ~

Pl

'I

NPF-l6 I I Page 5 I 2. Heav Loads (Section 9. I.4 SSER 3)

Propo sc,J ¹rier~~ HREG=OcrR-gg,/~g~~ 4y FPL We J gr- /I I II as iNS ,

-e~~e-I+ceesees-sheik-have->vade-eemo6tments-eeeeepteQe-te-

~I%E

+4HRE6=9~

l3. Fire Protection (Section 9.5.l.l l(a) and (b) SSER 3)

The licensees shall implement the fire protection program on a Dc le ke J schedule specified in Section 9.5.l.l l(a) and (b) of Supplement No.

3 to the Safety Evaluation Report.

I 4. Emer enc Diesel Generator Modifications (Section 9.5.4. I SER)

Prior to startup following the first refueling outage, the licensees shall a) install and have fully operational the automatic prelube pump and b) relocate instruments and controls located on the diesel engine skid to the floor-mounted panel.

l5. Radioactive Waste Mana ement (Section I l.2 I I.S SER SSER 3)

Within l4 months after core load, the licensees shall a) implement the design modifications to automatically shut off the waste management condensate and boric acid condensate pumps prior to the level reaching the overflow nozzle of the Primary Water Storage Tank and b) implement the design modification to automatically isolate the Lower Pressure Safety Injection pump discharge to the Refueling Water Tank upon receipt of a refueling water tank high water level alarm.

Prior to startup following the first refueling outage, FP&L shall a) install waste concentrator bottom tanks and, b) install a second continuous oxygen analyzer.

16. Initial Test Pro ram (Section l4 SER)

The licensees shall conduct the post-fuel loading initial test program (set forth in Section l4 of the St. Lucie 2 Final Safety Analysis Report, as amended through Amendment l3 and FP&L's

~

letter L-83-207) without making any modifications to this program unless such modifications are in accordance with the provisions of IO CFR Section 50.59. In addition, the licensees shall not make any major modifications to this program unless modifications have been identi fied and have received prio'r'RC approval. Major modifications are defined as:

a. Elimination of any test identified as essential in Section l4 of the Final Safety Analysis Report, as amended through Amendment I 3 and FP&L's letter L-83-207; ~ ~

~ ~ ~

I Ip J'

ATTACHMENT 4 SAFETY ANALYSIS Introduction The proposed amendments represent an administrative update to the St. Lucie Units 1 and 2 Technical Specifications and the St. Lucie Unit 2 Operating License. The proposed changes can be grouped into 3 categories:

1) Category 1 A change proposed to achieve consistency throughout the Technical Specifications or between similar Technical Specifications for both St. Lucie Units 1 and 2.

2) Category 2 A change proposed to correct an error in the Technical Specifications.

3) Category 3 A change proposed to delete an outdated and fully satisfied footnote in the Technical Specifications or a fully satisfied License Condition in the Operating License.

Discussion The changes proposed have been grouped into three categories.

The detailed justification and evaluation of each proposed change for each unit is discussed below.

St. Lucie Unit 1

1) Category 1 A change proposed to achieve consistency:

a) Pages B 2-4 and B 2-5 currently discuss operation in Modes 1 and 2 with only 2 or 3 reactor coolant pumps (RCPs) in operation. The trip provides core protection against Departure from Nucleate Boiling in the event of a sudden significant decrease in reactor coolant system flow. The specified setpoint ensures that a reactor trip occurs to prevent violation of Local Power Density or Departure from Nucleate Boiling Ratio (DNBR) limits under stated conditions. St. Lucie Unit 1 Technical Specif ication Limiting Condition for Operation 3.4.1.1 requires both RCPs in both reactor coolant loops be in operation in Modes 1 and 2. Therefore, Bases discussions which consider other than 4 pump EJW/022.PLA

operation in Modes 1 and 2 are inconsistent with other requirements in the Technical Specifications.

The Bases for this reactor trip setpoint are proposed to be changed to more clearly reflect the basis for this function.

b) Page 3/4 7-44, Table 3.7-3, Fire Hose Stations, is revised with respect its tabular format making this table's format similar to the equivalent table in the St. Lucie Unit 2 Technical Specifications. Items B-2, B-3 and B-4 are also revised to be more specific relative to the locations of these hose stations.

c) Page 3/4 9-1, LIMITING CONDITIONS FOR OPERATION 3.9.1a. lists a conservative allowance for reactivity uncertainty as 14 8 k/k. By proposed license amendment, L-87-121, dated March 17, 1987, FPL proposed to change "4 6 k/k" to "pcm". The NRC issued Amendment 86 to the St. Lucie" Unit 1 Operating License on October 23, 1987, which incorporated this change. This item was overlooked in FPL's March 17, 1987 application: "14 8 k/k" should be changed to "1000 pcm".

d) Page 6-12, in RECORDS, Specification 6.5.2.10b. was written "...Section 6.5.2.7...." This should be changed to "...Specification 6.5.2.7..." as similarly written in Specification 6.5.2.10C.

2) Category 2 A change proposed to correct an error:

a) Page 3/4 3-19 currently has listed two "b." items under Functional Unit, 6. Loss of Power. The second item "b." should be changed to item "c.".

b) Page 3/4 9-3, APPLICABILITY:, currently has a typographical error. " (I) rradicated" should be changed to "irradiated".

c) Page 3/4 9-5, ACTION:, currently has a typographical error. The word "fo" in the last sentence of the ACTION statement should be changed to "of".

d) Pages 3/4 12-1, 3/4 12-11, and 3/4 12-12, reference, in 4 locations, the Annual Radiological Environmental Operating Report as being required by Specification 6.9.1.11 and, in 2 locations, the Semiannual Radioactive Effluent Release Report as being required by Specification 6.9.1.10. In fact, the Annual Radiological Environmental Operating Report is required by Specification 6.9.1.8 and the Semiannual HJW/022. PLA

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Radioactive Effluent Release Report is required by Specification 6.9.1.7.

3) Category 3 A change proposed to delete an outdated and fully satisfied footnote.

a) Page 3/4 3-4, Item (e) states "Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3.". Special Test Exception 3.10.3 was deleted by St. Lucie Unit 1 Amendment No. 4 dated April 16, 1976. Therefore, Item (e) should be deleted.

b) Pages 3/4 3-11, 3/4 3-15 and 3/4 3-19 currently include a footnote:

"*This specification will be effective prior to Cycle 7 restart."

The asterisk (*) is applied in 4 places to Functional Unit, 6. LOSS OF POWER, on Technical Specification pages 3/4 3-11, 3/4 3-15 and 3/4 3-19.

This footnote was added to the St. Lucie Unit 1 Technical Specifications in Amendment No. 58 dated May 3, 1983 to address the criteria and Staff positions pertaining to degraded grid voltage protection. In that Amendment, the staff found FPL's proposed modifications for protecting the Class 1E equipment from sustained degraded voltage under accident or non-accident conditions by initiating automatic disconnection of Class lE equipment from the offsite source, load shedding, diesel generator starting and load sequencing to be acceptable. FPL issued Plant Change/Modification (PC/M) 103-184, "Degraded Grid Voltage Design" to implement the design changes proposed to, and accepted by, the NRC Staff. This PC/M was completed on December 15, 1985 and St. Lucie Unit 1 started Cycle 7 on December 25, 1985. As a result, this footnote may now be deleted.

St. Lucie Unit 2

1) Category 1 A change proposed to achieve consistency Page B 2-6 does not clearly represent the Bases for the reactor flow-low trip function. The Bases for this reactor trip setpoint are proposed to be revised to more clearly describe the trip function basis and to be EJW/022.PLA

consistent with the same proposed revision to the Bases for St. Lucie Unit 1.

2) Category 2 A change proposed to correct an error:

a) Page B 2-1 has two typographical errors "sepcified" should read "specified" and "Oqeration" should read "Operational."

b) Page 3/4 3-41, ACTIONs c. and d. list Table 3.3-11 as the applicable Table for this Specification and ACTION statements. As listed in LIMITING CONDITION FOR OPERATION 3.3.3.6 on the same page and in ACTIONs a. and b., the referenced Table should be TABLE 3.3-10.

c) Pages 3/4 3-45 and 3/4 3-46, Table 3.7-3, Fire Detection Instrumentation; the column of this table headed "Flame" is proposed to be deleted since there are no flame detectors relied upon for safety-related equipment protection at St. Lucie Unit 2.

The St. Lucie Unit 2 Safety Evaluation Report (SER)

(NUREG-0843), dated October 1981, Section 9.5.1.2D.,

Fire Detection S stems, states "The fire detection systems consist. of ionization and fixed temperature/rate-of-rise detectors......" St. Lucie Unit 2 Updated Final Safety Analysis Report (UFSAR)

p. 9.5A-115 lists the type of detectors used at St.

Lucie Unit 2 as either ionization (smoke) or heat; flame detectors are not included. Since flame detectors are not and have never been included in the design of St. Lucie Unit 2 for safety-related fire protection and, since these pages of the Technical Specifications have never been amended, the two entries representing three flame detectors has always been in error. These detectors should be located under the "HEAT" column, two in Zone 4A (see UFSAR Figure 9.5A-16) (discussed further below) and one in Zone 8A (see UFSAR Figure 9.5A-18).

Zones 11A, 11B, 12A and 12B under the INSTRUMENT LOCATION CABLE SPREADING are redistributed to the more proper specific INSTRUMENT LOCATION throughout the remainder of the table.

REACTOR AUXILIARY BUILDING, "ZONE 4 REACTOR AUX BLDG. EL. 19.50", should more exactly be "ZONE 4 A. ~ ~ ~

" (UFSAR Fig. 9.5A-19) ~

The terms "Function A" and "Function B" from the p.3/4 3-46 footnote" *(x/y)" are deleted from this footnote since the footnote is complete, and EJW/022.PLA

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interpretationally clearer, without the additional definition. The "Function" terms are not used in plant operation.

d) Page, 3/4 6-1 SURVEILLANCE REQUIREMENT 4.6.1.1a, currently states that Table 3.6-2 is in Specif ication 3. 6. 4. Table 3. 6-2 is actually included in Specification 3.6.3.

e) Page 3/4 7-16 SURVEILLANCE REQUIREMENT 4.7.6.1.1 states "The Special Report shall include an evaluation of the facility flood protection if, as evidenced by this survey program, the beach dune described in Specification 5.1.3 is lost." This SURVEILLANCE REQUIREMENT is verbatim from SURVEILLANCE REQUIREMENT 4.7.6.1.1 in the St. Lucie Unit, 1 Technical Specifications. However, there is no Specification 5.1.3 in the St. Lucie Unit 2 Technical Specifications. Therefore, the final sentence of SURVEILLANCE REQUIREMENT 4.7.6.1.1 in the St. Lucie Unit 2 Technical Specifications should be deleted.

Page 3/4 7-22, SURVEILLANCE REQUIREMENT 4,.7.9e, Functional Tests, has a typographical error in the first sentence "...at last..." should be changed to

"...at least..."

g) Page 3/4 7-32, LIMITING CONDITION FOR OPERATION 3.7.11.2 currently contains "7. Fire Zone 34-Zone I RAB Electrical Equipment Room*." St. Lucie Unit 2 SER, Supplement 3, dated April 1983, Section 9.5.1.6c.4.,Section III.1.G.2 (Automatic Fine Suppression System) states:

"The applicant has requested deviations from providing automatic fire suppression system in the following areas:

Division A Switchgear Room Division B Switchgear Room and this section continues:

Each of the above areas contains early warning fire detection. Redundant safe shutdown cable and equipment is separated by more than 20 feet or enclosed in a fire rated barrier. The fuel load in each of the above .areas is low with the exception of the Division A and Division B Switchgear Room. The fuel load, in the Division A 6 B Switchgear Room has fire severity of 1.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . In these areas, the E2W/022.PLA

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redundant cables will be enclosed in a 2-hour fire rated conduit barrier.

Because the in-situ fuel loads are now, or redundant systems have either adequate separation, or have been enclosed in fire rated barriers with an hourly rating higher than the fire severity of in-situ combustibles, and early warning fire detection is provided, we have reasonable assurance that after a postulated fire in any of these areas, one train of safe shutdown systems will be free of fire damage.

Based on our evaluation, we conclude that the installation of automatic fire suppression system in the Division B Switchgear Room, Hallway to the Division B Fan Room, Component Cooling Area, Steam Tressel, Intake Structure, Aerated Waste Storage Tank Room and Gas Decay Tank Cubicle 2C would not significantly increase the level of fire safety. We find the deletion of the automatic fire suppressions systems to be an acceptable deviation from Section III.G.2 of Appendix R. Therefore, the fire protection provided for these areas is acceptable."

This section of Supplement 3 to the SER was in response to FPL letter L-82-282, dated July 14, 1982.

FSAR Figure 9.5A.2 shows that Fire Zone 34 is Switchgear Room "B" (and, in fact Electrical Equipment Room Zone II, not Zone I as indicated in the Technical Specifications). Fire Zone 34 has never had automatic fire suppression (an accepted NRC deviation, as indicated above). Therefore, the "Fire Zone 34 Zone

RAB Electrical Equipment Room" sprinkler system should be deleted form LCO 3.7.11.2. The remaining 3 Fire Zones'ine numbers are adjusted accordingly.

h) Page 3/4 7-36, Table 3.7-4, FIRE HOSE STATIONS, is amended to more clearly locate and identify three hose stations. The hose station themselves are unchanged, only the description of the hose station location is clarified.

i) Page 3/4 7-37, SURVEILLANCE REQUIREMENT 4.7.11.4b.

states that OPERABILITY demonstrations for yard fire hydrants and associated hydrant hose houses are to be conducted every six months and additionally stipulates:

"...(once during March, April or May and once during September, October or November)..."

EJW/022.PLA

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Generic Letter No. 83-27, "Surveillance Intervals Standard Technical Specifications" states:

"The 12-month surveillance intervals applicable to contain portions of the fire protection (fire water systems) Technical Specifications were based upon annual climatic conditions..."

St. Lucie Plant is located in a climate characterized by the NRC in the St. Lucie Unit 2 SER (NUREG-0843)

Section 2.3.1 Re ional Climatolo as subtropical marine with mean monthly temperatures varying between 63.5 F and 81.6'F. This variation indicates minimal climatic changes. Additionally, SURVEILLANCE REQUIREMENT 4.7.11.3b in the St. Lucie Unit 1 Technical Specifications parallels SURVEILLANCE REQUIREMENT 4.7.11.4b from the St.Lucie Unit 2 Technical Specifications and does not include the limitation on the particular months in which this surveillance is to be performed. Therefore, the parenthetical statement is proposed to be deleted from the St. Lucie Unit 2 Technical Specifications.

j) Pages 3/4 12-1, in 4 locations, 3/4 12-11, and 3/4 12-12 reference, the Annual Radiological Environmental Operating Report as being required by Specification 6.9.1.11 and, in 2 locations, the Semiannual Radioactive Effluent Release Report as being required by Specification 6.9.1.10. In fact, the Annual Radiological Environmental Operating Report is required by Specification 6.9.1.8 and the Semiannual Radioactive Effluent Release Report is required by Specification 6.9.1.7.

k) Page 6-23 Specification 6.14 OFFSITE DOSE CALCULATION MANUAL ODCM has a typographical error in the first sentence. "Licensee initiated changes to the PCP:" should be changed to "Licensee initiated changes to the ODCM:"

3) Category 3 A change proposed to delete a outdated and fully satisfied, footnote in the Specifications or a fully satisfied License Technical Condition in the Operating License.

a) Page 3/4 7-32 currently includes a footnote: "* The sprinkler systems shall be completely installed and operable prior to exceeding 54 of Rated Thermal Power."

E JW/022. PLA

The asterisk (*) is applied in 7 places in LIMITING CONDITION FOR OPERATION 3.7.11.2. This footnote was part of the Unit 2 Technical Specifications as a start-up condition. The 54 of Rated Thermal Power milestone was passed on June 13, 1983, and FPL notified the NRC by letter L-83-348, dated June 6, 1983, that these conditions had been met. FPL, therefore, proposes to delete this footnote.

b) License Condition 2.C.12. in the St. Lucie Unit, 2 Facility Operating License NPF-16 requires FPL meet the requirements of NUREG-0612. By letter L-85-41, dated January 25, 1985, FPL proposed to delete the following portion of this License Condition:

"Prior to startup following the first refueling outage, the licensees shall conform to the guidelines of Section 5.1.1 of NUREG-0612 and" NRC action on this proposed license amendment is pending. ~

FPL now proposes to delete the balance of this License Condition which reads:

"prior to thirty days of startup following the second refueling outage, the licensees shall have made commitments acceptable to the NRC regarding the guidelines of Section 5.1.2 through 5.1.6 of NUREG-0612."

St. Lucie Unit 2 started up for Cycle 2 operation on November 19, 1984. St. Lucie Plant Administrative Procedure AP001438, Control of Heavy Load Lifts and Transporting of Heavy Loads, implements the guidelines of NUREG-0612.

This Administrative Procedure references the following plant procedures:

I-M-0015 Reactor Vessel Maintenance Sequence of Operation M-0020 Lifting of the Spent Fuel Gate M-0021 Lifting of the Pressurizer Missile Shield M-0022 Handling of Spent Fuel Casks M-0023 Handling of the ISI Tool 2-M-0036 Reactor Vessel Maintenance Sequence of Operations EJW/022.PLA

The NRC reviewed these procedures and found them to be acceptable in meeting the guidelines of Section 5.1.2 through 5.1.6 of NUREG-0612. This conclusion by the NRC was documented in NRC Inspection Report 50-389/88-03 dated March 18, 1988. As a result, FPL now proposes that the balance of this License Condition be deleted.

EGW/022.PLA

ATTACHMENT 5 DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION The standards used to arrive at a determination that, a request for amendment involves no significant hazards consideration are included in the Commission s regulations, 10 CFR 50.92, which states that no significant, hazards considerations are involved if the operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident.

from any accident previously evaluated or (3) involve a significant reduction in a margin of safety. Each standard is discussed as follows:

(1) Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability or consequences of an accident previously evaluated.

For the changes intended to achieve consistency throughout the Technical Specifications, the intent of the Specification will not be changed nor will operating limitations of the Technical Specifications be changed. For the changes which involve correction of errors in the Technical Specifications, the changes proposed are intended to solely correct typographical errors or omissions from the Technical Specifications.

For the changes which propose to delete outdated and fully satisfied footnotes or a License Condition, the NRC Staff's evaluation of the effect of the equipment or procedures required by footnote, and its completion to fully satisfy the Technical Specification requirement or License Condition, were previously evaluated and determined not to significantly affect the probability or consequences of an accident previously evaluated.

This change proposes solely to remove outdated footnotes from the Technical Specifications.

The changes do not affect assumptions contained in plant safety analyses, nor do they affect Technical Specifications that preserve safety analysis assumptions. Therefore, the proposed changes do not affect the probability or consequences of accidents previously analyzed.

EVW/022.PLA

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(2) Use of the modified specification would not create the possibility of a new or different kind of accident from any accident previously evaluated.

The changes being proposed by FPL will not lead to material procedure changes or to physical modifications to St. Lucie Plant. Therefore, the proposed changes do not create the possibility of a new or different kind of accident.

(3) Use of the modified specification would not involve a significant reduction in a margin of safety.

The changes being proposed by FPL do not relate to or modify the safety margins defined in and maintained by the Technical Specifications.

Therefore, the proposed changes should not involve any reduction in a margin of safety.

Based on the above, we have determined that the amendment request does not (1) involve a significant increase in the probability or consequences of an accident previously evaluated, (2) create the probability of a new or different kind of accident from any accident previously evaluated, or (3) involve a significant reduction in a margin of safety; and therefore does not involve a significant hazards consideration.

E8W/022. PLA

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