Text

Amends 118 & 60 to Licenses DPR-67 & NPF-16,respectively. Amend Make Administrative Changes to TS by Removing Outdated Matl,Making Text Changes & Correcting Errors
	ML17227A675
Person / Time
Site:	Saint Lucie
Issue date:	12/24/1992
From:	Berkow H; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17227A676	List: ML17227A675; ML17227A677;
References
	NUDOCS 9301040196
	Download: ML17227A675 (63)
	v • d • e

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 FLORIDA POWER

& LIGHT COMPANY DOCKET NO. 50-335 ST.

LUCIE PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

118 License No.

DPR-67 1.

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

A.

The application for amendment by Florida Power

& Light Company, et al. (the licensee),

dated August 21,

1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

E.

The issuance of this amendment will not be inimic'al to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

930f04019b 92f224 f PDR ADQCK 05000335

2.

Accordingly, Facility Operating License No.

DPR-67 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.(2) to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No..118, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications erbert N. Berk w, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Date of Issuance:

December 24, 1992

ATTACHMENT TO LICENSE AMENDMENT NO. 118 TO FACILITY OPERATING LICENSE NO.

DPR-67 DOCKET NO. 50-335 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Pa es III of the Index 3/4 0-3 3/4 1-12 3/4 3-41 3/4 4-1b 3/4 4-8 3/4 4-14 3/4 6-5 3/4 6-20 3/4 6-21 3/4 8-6b 3/4 11-2 3/4 11-10 3/4 12-10 B 3/4 1-4 B 3/4 7-5 B 3/4 9-3 6-24 Insert Pa es III of the Index 3/4 0-3..'/41-12 3/4 3-41 3/4 4-1b 3/4 4-8 3/4 4-14 3/4 6-5 3/4 6-20 3/4 6-21 3/4 8-6b 3/4 11-2 3/4 11-10 3/4 12-10 B 3/4 1-4 B 3/4 7-5 B 3/4 9-3

'-24

INDEX e

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3 4.0 APPLICABILITY.

3 4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown'argin

- T,, > 200'F Shutdown Margin - T., a 200'F Boron Dilution Moderator Temperature Coefficient Minimum Temperature for Criticality.

3/4. 1. 2 BORATION SYSTEMS Flow Paths

- Shutdown Flow Paths

- Operating Charging Pumps

- Shutdown Charging Pumps

- Operating Boric Acid Pumps

- Shutdown Boric Acid Pumps

- Operating Borated Water Sources

- Shutdown Borated Water Sources

- Operating 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Full Length CEA Position Position Indicator Channels CEA Drop Time Shutdown CEA Insertion Limit Regulating CEA Insertion Limits PAGE 3/4 0-1 3/4 1-1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 1-5 3/4 1-7 3/4 1-8 3/4 1-8 3/4 1-10 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 1-16 3/4 1-4 3/4 1-20 3/4 1-20 3/4 1-24 3/4 1-26 3/4 1-27 3/4 1-28 ST.

LUCIE -

UNIT 1 Amendment No. 77, 118

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE.....................................

3/4 2-1 3/4.2.2 DELETED 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR T

F

~

r 3/4 2-6 3/4 2-9 3/4.2.4 3/4.2. 5 DNB PARAMETERS 3/4 2-13 AZIMUTHAL POWER TILT - T............................

3/4 2-11 q

3/4.3 INSTRljMENTATION 3/4.3.1 REACTOR PROTECTIVE INSTRUMENTATION...................

3/4 3-1 3/4 3-21 3/4 3-21 3/4 3-25 3/4 3-27 3/4 3-30 3/4 3-33 3/4 3-37 3/4 3-41 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitorin9 Incore Detectors................................

Seismic Instrumentation..............................

Meteorological Instrumentation.......................

Remote Shutdown Instrumentation......................

Fire Detection Instrumentation.......................

Accident Monitoring Instrumentation.........

Radioactive Liquid Effluent Monitoring Instrumentation......................................

3/4 3-45 Radioactive Gaseous Effluent Monitorin9 Instrumentation......................................

3/4 3-50 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................................

3/4 3-9 3 4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION........

3/4 4-1 3/4.4.2 SAFETY VALVES -

SHUTDOWN.............................

3/4 4-2 3/4.4.3 SAFETY VALVES - OPERATING................,...........

3/4 4-3 ST.

LUCIE - UNIT I IV Amendment No. g7,

$7,

$5, 57,y p 1<

SEP l 0 Sgt

'APPL ICABILITY SURVEILLANCE RE UIREMENTS Continued 4.0.5 (Continued)

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice inspection and testing activities Weekly Monthly quarterly or every 3 months Semiannually or every 6 months Yearly or annually Required frequencies for performing inservice inspection and testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 366 days c.

The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.

d.

Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.

e.

Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.

ST.

LUCIE UNIT I 3/4 0-3 Amendment No.

dd

REACTIVITY CONTROL SYSTEMS CHARGING PUMPS -

SHUTDOWN LIMITING CONDITION FOR OPERATION

3. 1.2.3 At least one charging pump or high pressure safety injection pump* in the boron injection flow path required OPERABLE pursuant to Specification

3. 1.2. 1 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus.

APPLICABILITY:

MODES 5 and 6.

ACTION:

A With no charging pump or high pressure safety injection pump* OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes until, at least one of the required pumps is restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS

4. 1.2.3 At least one of the above required pumps shall be demonstrated OPERABLE by verifying the charging pump develops a flow rate of greater than or equal to 40 gpm or the high pressure safety injection pump develops a total head of greater than or equal to 2571 ft. when tested pursuant to Specification 4.0.5.

The flow path from the RWT to the RCS via a single HPSI pump shall be established only if:

(a) the RCS pressure boundary does not exist, or (b) no charging pumps are operable.

In this case, all charging pumps shall be disabled and heatup and cooldown rates shall be limited in accordance with Fig. 3.1-1b.

At RCS temperatures below 115'F, any two of the following valves in the operable HPSI header shall be verified closed and have their power removed:

Hi h Pressure Header HCV-3616 HCV-3626 HCV-3636 HCV-3646 Auxiliar Header HCV-3617 HCV-3627 HCV-3637 HCV-3647 ST.

LUCI E - UNIT 1

3/4 1-12 Amendment No. 66, Hi, 56, f84,

INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

a ~

b.

Actions per Table 3.3-11.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

ST LUCIE - UNIT 1 3/4 3-41 Amendment No. 37, 118

TABL 3.3-ACCIDENT ON ORING NSTRUMENTATION TOTAL NO.

0 CHANN LS MINIMUM CHANNELS QPPR~AB

$~0 1.

Pressurizer Water Level 2.

Auxiliary Feedwater Flow Rate 3.

RCS Subcooling Margin Monitor 4.

PORV Position Indicator Acoustic Flow Monitor 5.

PORV Block Valve Position Indicator 6.

Safety Valve Position Indicator 7.

Incore thermocouples 8.

Containment Sump Mater Level (Narrow Range) 9.

Containment Sump Water Level (Wide Range)

10. Reactor Vessel Level Monitoring System

11. Containment Pressure 1/pump 1/valve 1/valve 1/valve 4/core

. quadrant 1

1 1/pump 1

1 1

1/valve 1/valve 1/valve 2/core quadrant 4,

5 4,

5

The non-safety grade containment sump water level instrument may be substituted.

A channel is composed of eight (8) sensors in a probe, of which four (4) sensors must be OPERABLE.

~

-'EACTOR COOLANT SY M

I HOT SHUTDOWN

. LIMITING CONDITION FOR OPERATION 3.4. 1.3 At least two of the loops listed below shall be OPERABLE and at least one reactor coolant or shutdown cooling loop shall be in operation.*

a.

Reactor Coolant Lo'op A and, its associated steam generator and at least one associated reactor coolant

pump, b.

Reactor Coolant Loop B and its associated steam generator and at least one associated reactor coolant

pump, c.

Shutdown Cooling Loop A, d.

Shutdown Cooling Loop B.

APPLICABILITY: MODE 4.

)>~CION:

a

~

With less than the above required reactor coolant or shutdown cooling loops OPERABLE, within one (1) hour initiate corrective action to return the required loops to OPERABLE status. If the remaining OPERABLE loop is a shutdown cooling loop, be in COLD SHUTDOWN within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With no reactor coolant or shutdown cooling 'loop in operation, suspend all operations involving a reduction in boron concentra-tion of the Reactor Coolant System and within one (1) hour initiate corrective action to return the required coolant loop to operation.

All reactor coolant pumps and shutdown cooling pumps may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided (1) no operations are permitted that would cause dilu-tion of the Reactor Coolant System boron concentration, and (2) core outlet tempei ature is maintained at least 10'F below saturation temperature.

ST.

LUCIE - UNIT 1 3/4 4-lb Amendment No.

56

~ 118

REACTOR COOLANT SYSTEN HOT SHUTDOWN SURVEILLANCE RE UIREHENTS 4.4.1.3.1 The required reactor coolant pump(s), if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker al'ign-ments and indicated power availability.

4.4.1.3.2 The required steam generator(s) shall be determined OPERABLE by verifying the secondary side water level to be

> 10% of narrow range indica-tion at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.3.3 At least one reactor coolant or shutdown cooling loop shall be verified to be in operation and circulating reactor coolant at least.

once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST.

LUCIE - UNIT 1

3/4 4-lc Amendment No.

5 g

REACTOR COOLANT SYSTEM SURVEILLANCE RE UIRBlENTS Continued b.

If the inservice inspection of a steam generator conducted. in accordance with Table 4.4-2 requires a third sample inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months.

The reduction in inspection frequency shall apply until a subsequent inspection demonstrates that a third sample inspection is not required.

C ~

Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions.

1.

Primary-to-secondary tubes leaks (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6,2, 2.

A seismic occurrence greater than the Operating Basis Earthquake, 3.

A loss-of-coolant accident requiring actuation of the engineered safeguards, or 4.

A main steam line or feedwater line break.

4.4.5.4 Acce tance Criteria a ~

As used in this Specification:

~lf t

pt t tl df f.fffb or contour of a tube from that required by fabrication drawings or specifications.

Eddy-current testing indications

'elow 20% of the nominal tube wall thickness, if detectable, may be considered as imperfections.

2.

~0dti t -td d

plg, g

wear or general corrosion occurring on either inside or outside of a tube.

g.

~0ddfb tb tif gfp f tf 202 of the nominal wall thickness caused by degradation.

P.

~20 dtf th,d tg fth tb thickness affected or removed by degradation.

ST.

LUCIE - UNIT 1

3/4 4-7

R ACTOR COOLANT SYSTE SURVEILLANCE RE UIREMENTS Continued 5.

Defect means an imperfection of such severity that it exceeds the plugging limit. A tube containing a defect is defective.

Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a

defective tube.

6.

~P1 i Lii<<h ip f ti dph b

d which the tube shall be removed from service because it may become unserviceable prior to the next inspection and is equal to 40% of the nominal tube wall thickness.

7.

Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a

loss-of-coolant

accident, or a steam line or feedwater line break as specified in 4.4.5.3.c, above.

b.

8.

Tube Ins ection means an inspection of the steam generator tube from the point oF entry'(hot leg side) completely around the U-bend to the top support of the cold leg.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5

~Re orts a e b.

Within 15 days following the completion of each inservice inspection of steam generator

tubes, the number of tubes plugged in each steam generator shall be reported to the Commission in a special report pursuant to Specification 6.9.2.

The complete results of the steam generator tube inservice inspection shall be submitted to the Commission in a special report pursuant to Specification 6.9.2 within 12 months following completion of the inspection.

This special report shall include:

Number and extent of tubes inspected.

2.

3.

Location and percent of wall-thickness penetration for each indication of an imperfection.

Identification of tubes plugged.

ST.

LUCIE - UNIT 1 3/4 4-8 Amendment No. W, id',

118

j

~

REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION Continued c.

With the contyinment sump level and flow monitoring system in-

operable, restore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIREMENTS 4.4.6.1 The leakage detection systems shall be demonstrated OPERABLE by:

a.

Containment atmosphere gaseous and particulate monitoring systems-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, and b.

Reactor cavity sump level and flow monitoring system-performance of CHANNEL CALIBRATION TEST at least once per 18 months.

ST.

LUCIE - UNIT 1

3/4 4-13

REACTOR COOLANT SYSTE REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System leakage shall be limited to:

a.

No PRESSURE BOUNDARY LEAKAGE, b.

1 GPM UNIDENTIFIED LEAKAGE, c.

1 GPM total primary-to-secondary leakage through steam generators, d.

10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and e.

Leakage as specified in Table 3.4.6-1 for each Reactor Coolant System Pressure Isolation Valve identified in Table 3.4.6-1.

2>>

IIT:E,d,dd.

~CTI ON:

a.

With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY within 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 .

b.

c ~

With any Reactor Coolant System leakage greater than any one of the above limits, excluding PRESSURE BOUNDARY LEAKAGE and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leak-age rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 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 .

With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit in 3.4.6.2.e above reactor operation may continue provided that at least two valves, including check

valves, in each high pressure line having a non-functional valve are in and remain in the mode corresponding

-to the isolated con-dition.

Motor operated valves shall be placed in the closed posi-

tion, and power supplies deenergized.

(Note, however, that this may lead to ACTION requirements for systems involved.)

Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at-least HOT STANDBY within 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 4.4.6.2 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a.

Monitoring the containment atmosphere gaseous and particulate radioactivity at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST LUCIE - UNIT 1

3/4 4-14

TABLE 3.6-8 t o Makeup Water Station Air Va ve umbe Gate (I-MV-15-1)

Check (I-V-15328)

Globe (I-V-18-794)

Globe (I-V-18-796)

Location t

e t Outside Inside Outside Outside

~e~c Primary Makeup Water Station Air Supply Test

~j~*

Bypass Bypas Instrument Air Gate (I-MV-18-1)

Check (I-V-18195)

Outside Inside Instrument Air Supply Bypass 10 14 23 24 25 26 28 Containment Purge Containment Purge Waste Management Component Cooling Component Cooling Fuel Transfer Tube

, CVCS Sampling Butterfly (I-FCV-25-4)

Butterfly (I-FCV-25-5)

Butterfly (I-FCV-25-3)

Butterfly (I-FCV-25-2)

Globe (V-6741)

Check (V-6779)

Butterfly (I-HCV 14-7)

Butterfly (I-HCV-14-1)

Butterfly (I-HCV-14-6)

Butterfly (I-HCV-14-2)

Double Gasket Flange Globe (V-2515)

Globe (V-2516)

Globe (V-5200)

Globe (V-5203)

Globe (I-FCV-03-1E)

Globe (I-FCV-03-1F)

Inside

=--

Outside Inside Outside Outside Outside Outside Outside Outside Outside Inside Inside Inside Outside Outside Outside Outside Containment Purge Exhaust Containment Purge Supply Nitrogen supply to SI Tanks RC Pump CW Supply RC Pump CW Return Fuel Transfer Letdown Line

'eactor Coolant Sample SI Tank Sample SI Tank Sample Type C

Type C

Bypass Bypass Bypass Bypass Bypass Bypass Bypass

TABLE 3.6-1 Continued In Pl ChI Ch Penetration

~S stem 29 Sampling 29 Sampling 31 Waste Hanagement Valve Ta Number Globe (V-5202)

Globe (V-5205)

Globe (V-5201)

Globe (V-5204)

Gate (V-6554)

Gate (V-6555) 41 42 Waste Management Waste Management Gate (I-LCV-07-11A)

Gate (I-LCV-07-118)

Gate V-6301)

Gate V-6302)

Safety Injection Gate (V-3463)

Tank Test Lines Gate (I-V-07009)

Location to Containment Outside Outside Outside Outside Outside Outsitfe Outside Outside Outside Outside Outside Outside Service Pressurizer Steam Space Sample Pressurizer Surge Line Sample Containment Vent Header Test

~Te*.

Bypass Bypass Bypass Reactor Drain Tank Pump Suction Bypass Safety Injection Tank Bypass Fill and Sampling Reactor Cavity Sump

'ypass Pump Discharge tD t+

~ Q Q

46 48a CVCS Fuel Pool Cleanup Fuel Pool

.Cleanup Sampling Gate (V-2505)

Gate I-SE-Ol-l)

Gate (I-V-07-206)

Gate (I-V-07-189)

Gate

( I-V-07-170)

Gate (I-V-07-188)

Globe (I-FSE-27-1, 2, 3, 4)-

Globe (I-FSE-27-8)

Outside Inside Outside Inside Outside Inside Inside Outside RC Pump Controlled Bleedoff Refueling Cavity Purification Flow Inlet Refueling Cavity Purification Flow Outlet H2 Sampling Bypass Bypass Bypass Type C

CONTAINMENT SYSTEMS SURVEILLANCE PE UIREMENTS (Continued) 4.6.3.1.2 Each isolation valve specified in Table 3.6-" shall be demonstrated OPERABLE dur'ing the COLO SHUTOOWN or REFUELING MODE at least once per 18 months by:

a.

Verifying that on a Containment Isolation test signal, and/or SIAS test signal, each isolation valve actuates to its isolation position.

4.6.3.1.3 The isolation time of each power operated or automatic valve of Table 3.6-2 shall be determined to be within its limits when tested pursuant to Specification 4.0.5.

ST.

LUCIE - UNIT 1 3/4 6-19 Amendment No.

90

Ch

~ Valve Ta Number

~A.

CONTAINMENT ISOLATION COM 1.

I-FCV-25-4 5

~

g I

2.

I-FCV-25-2,3 Hg 3.

I-MV-15-1, 4 ~

I-MV-18-1 5.

V-6741 6.

I-HCV-14-1

& 7 Penetration Number 10 14 23 BLE 3.6-CON AINMENT ISOLAT ON VALVES Function Containment purge air exhaust, CIS Containment purge supply, CIS Primary makeup water, CIS Instrument air supply, CIS Nitrogen supply to safety injection

tanks, CIS Reactor coolant pump cooling water

supply, SIAS Testable During Pla t 0 eration No No Yes No Yes No Isolation Time Sec Ch I

O 0

o88 ft 0

7.

I-HCV-14-6

& 2 8.

V-2515,2516 9.

V-5200,5203 10.

V-5201,5204 11.

V-5202,5205 12.

V-6554,6555 13.

I-LCV-07-11A,11B 14.

V-6301,6302 15.

V-2505 16 '-SE-01-1 24 26 28 29 29 31 42 43 44 44 Reactor coolant pump cooling water

return, SIAS Letdown line, CIS, SIAS Reactor coolant sample, CIS Pressurizer surge line sample, CIS Pressurizer steam space

sample, CIS Containment vent header, CIS Reactor cavity sump pump discharge, CIS Reactor drain tank pump suction, CIS Reactor coolant pump controlled

bleedoff, CIS Reactor coolant pump controlled

bleedoff, CIS No No Yes Yes Yes Yes Yes Yes No No 10

3.6-2 Co t'd V

um Penetration Testable During Isolation g

B.

MANUAL OR REMOTE MANUAL I

1 I-V-18-794 I-V-18-796 I

2.

I-V-25-11,12 3 ~ I V 25 13'4'5, 16 4

V-3463 5 ~

I-V-07009 6.

V-07206, V-07189 7 ~

'V-07170, V-07188 8 ~ I FSE 27 1N2t3~

4,8,11 9 ~ I FSE 27 5N 6g7/

9,10 56 57

& 58 41 41 46 47 48a 48c 5la 5lc Station air supply, Manual Hydrogen purge outside air make-up, Manual (NC)

Hydrogen purge exhaust, Manual (NC)

Safety injection tank test line, Manual (NC)

Refueling cavity purification flow inlet, Manual (NC)

Refueling cavity purification flow outlet, Manual (NC)

Hydrogen sampling line, Remote manual Hydrogen sampling line, Remote manual Yes Yes Yes Yes Yes Yes Yes Yes Yes NA

.A ~

NA NA*

NA*

NA NA NA+

NA*

In m

C Valve Ta Number 10.

I-FCV-26-1 5 2 ll.

I-FCV-26-3 5 4 12.

I-FCV-26-5 5 6 13.

I-V00140 I-V00143 14.

I-V00139 I-V00144 15.

I-V00101 16.

I-FCV-03-1E fh 1F Penetration Number 52a 52b 52c 52d 52e 54 28 TABLE 3.6-2 Continued Function Radiation monitoring Radiation monitoring Radiation monitoring, return ILRT test tap ILRT test tap ILRT pressure connection SI Tank Sample Testable During Plant 0 eration Yes Yes Yes Yes Yes Yes Yes Isolation Time Sec NA NA NA NA NA NA NA~*

Ct O

NA - Manual Va ve-so ation time not applicable.

May be opened on an intermittent basis under administrative control.

Normally closed valves - Isolation time not applicable.

R A

PMR Y

SURV ANC R

R H NT nt'nu 6.

7.

Verifying the diesel generator operates for at least 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s****. During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3960 kWI and during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this.

test, the diesel generator shall be loaded within a load band of 3300 to 3500 kWf. The generator voltage and frequency shall be 4160 t 420 volts and 60 i 1.2 Hz within 10 seconds after the start signal; the steady state generator voltage and frequency shall-be maintained within these limits during this test.

Verifying that the auto-connected loads do not exceed the 2000-hour rating of 3730 kW.

8.

Verifying the diesel generator's capability to:

a)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a simulated restoration of offsite power.

b)

Transfer its loads to the offsite power source, and c)

Be restored to its standby status.

9.

Verifying that with the diesel generator operating in a test mode (connected to its bus),

a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power.

10.

Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines.

11.

Verifying that the automatic load sequence timers are operable with the interval between each load block within i 1 second of its design interval.

At least once per ten years or after any modification which could affect diesel generator independence by starting****the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to approximately 900 rpm in less than or equal to 10 seconds.

OThis band is meant as guidance to avoid routine overloading of the engine.

Variations in load in excess of this band due to changing bus loads shall not invalidate this test.

- - - Thistest may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST.

LUGIE - UNIT 1 3/4 8-6a Amendment No. ~ 112

ELECTRICAL POWER

SYSTEM,

~

SURVEILLANCE RE UIREMENTS Continued g.

At least once per ten years by:

1.

Draining each fuel storage

tank, removing the accumulated sediment and cleaning the tank using an appropriate

. cleaning

compound, and 2.

Performing a pressure test oF those portions of the diesel fuel oil system designed to USAS B31.7 Class 3 requirements at a test pressure equal to 110% of the system design pressure.

4.8.1.1.3

~Re orts

- All diesel generator failures, valid or non-valid, shall be reported to the Commission pursuant to Specification 6.9.2.

Reports of diesel generator failures shall include the.information recommended in Regulatory Position C.3.b of Regulatory Guide 1. 108, Revision 1, August 1977.

If the number of failures in the'ast 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide

1. 108, Revision 1, August 1977.

4.8. 1. 1.4 The Class 1E underground cable system shall be demonstrated OPERABLE within 30 days after the movement of any loads in excess of 80% of the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at least 80% of the duct's inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.

ST.

LUCIE - UNIT 1 3/4 8-6b Amendment No. QS, fff, 118

3l4. 11 RADIOACT;VE EFFLUEHTS 3/4.11,1 LiOUi5 EFFLUENTS CONC H(RATION L.".IT:NG COHOITIOH FOR OPERATION

3. 11. 1. I The concentration of radioactive material released from the site (see Figure S. l-l) shall be limited to the concentrations speci,i ed in 10 CFR Part 20, Appendix 8, Table II, Column 2 foT'adionucli des other than d>ssolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration shall be limited'o 2 x 10-~ microcurieslml total activity.

APPLICABILITY: At all times.

ACTiON:

With the concentration of radioac-ive material released from the site the above limits, immediately res.ore the concen ration to within the 1 0m1 ts.

exceeding above SURVEIL(((iHCE R OUIREHEHTS

4. 11. 1. 1. 1 The radioactivity content of each batch of radioactive liquid waste shall be determined prior to release by sampling and analysis in accordance wi h Table 4. 11-1.

The results of pre-release analyses shall be used with the ca'culational methods in the QOCH to assure that the concen"rstion a-t'h e point of release is maintained within the limits of Soecification 3.

1. 1. 1.

Also, i esul-s of the previous post-release analyses shall be used with the calcu-lational methods in the OOCH to assure that the concen

", ations a

. the point of release were maintained within the limits of Specification

3. 11. l.l.

4. 11. 1. 1. 2 Post-release analyses of. samples composited from batch releases shall be performed in accordance with Table 4. 11-1.

4. 11. 1. 1.3 The radioactivity concentration of liquids discharged from continuous release points shall be determined by collec-ion and analysis of samples in accordance with Table 4.11-1.

The results o

the analyses shall be used with the calculational methods in the OOCN o assure tha the cancan rations at the point of release are maintained within tne limits of Specification

3. 11. l. 1.

ST.

LUCIE - UNIT 1 3t4 ll-l Amendment No.

5 9

RADIOACTIVE LI UID WASTE SAMPLING AND ANALYSIS PROGRAM Liquid Release Type Sampling Frequency Minimum Analysi s Frequency Lower Limit of Detection Type of Activity (LLD1 Analysi s (pCi/ml )

A.

Batch Waste Release Tanks' p

Each Batch Each Batch Principal Gamma 5xlO'mitters'-131 Ix10 p

One Batch/M Dissolved and Entrained Gases (Gamma Emitters) lx10' H

H-3 Each Batch

'omposite" Gross Alpha Ix10 1x10 p

Q Each Batch Composite'r-89, Sr-90 Fe-55 5x10 lxlO B.

ContinuoL%~

Releases Daily 4/M Composite Principal Gamma 5xlO Emitters' 4/H Grab Sample Composite Daily I-131 Dissolved and Entrained Gases (Gamma Emitters) 1xlO Ix10 Daily H

Composite H-3 Gross Alpha 1xlO lx10 Daily Q

Composite Sr-89, Sr-90 Fe-55 5x10 1x10 C.

Settl ipg Basin W

Grab Sample Principal Gamma 5x10 Emitters'-131 1xlO ST.

LUCIE - UNIT 1 3/4 11-2 Amendment No. 59, 118

TABL

4. 11 2

Continued)

TABLE NGTATIQN The LLO is defined, for purposes of these specifica ions, as the smallest concen.ration of radioactive material in a sample that will yield a net

count, above system background, that will be detected with 95~ probability with only 5X probability of falsely concluding that a blank obsess vation represents

~ "real" signal.

For a particular measurement

system, which may include radiochemical separation:

LLO '.66 sb E

~

V

~

2.22

~

Y exp(-Mt)

Mhere:

LLO is the "a pr iori" lower limit of detection as defined above, as picocuries per unit mass or volume, sb is the standard deviation o

the background counting rate or of the counting rate of a blank sample as appropriate, as counts per

minute, E is the courting efficiency, as counts per disintegration, V is the sample size in units of mass or

volume,

2. 22 is the number of di'sintegrations per minute per picocurie, Y. is the frac'.ional radiochemical yield, when applicable, A, is the radioactive decay constant for the particular radionuclide, and Lt for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical values of E, V, Y, and At should be used in the calculation.

ST.

LUCIE - UNIT 1 3/4 11-9 Amendment No. 59

TABLE NOTATION b.

Sampling and analysis shall also be performed following shutdown,

startup, or a

THERMAL POWER change exceeding 15% of RATED THERMAL POWER within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months unless (1) analysis shows that the DOSE E(UIVALENT I-131 concentra-tion in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

C.

Samples shall be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each

shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing if (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has increased more than a

factor of 3; and (2) the noble gas activity monitor'shows that effluent activity has increased by more than a factor of 3.

When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

d.

e.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications

3. 11.2. 1, 3. 11.2.2 and 3. 11.2.3.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-137, Ce-141 and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

ST.

LUCIE - UNIT 1 3/4 11-10 Amendment No. 59, 118

TABLE NOTATION This list does no. mean that only tl;ese nuclides are:o be considered:

Other peaks that are identifiable, together with those of he above

nuclides, shall also be, analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification

6. 9. 1. 11.

Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4. 13.

The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net

count, above sys.em background, that will.be detected with 96.probability with only 5 probabili y of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement

system, which may include radiochemical separation:

LLD 4,66 sb E

~

Y

2. 22

~

Y

~

exp(-4t)

Where:

LLD i's the "a priori" lowe~ limit of detection as defined above, as picocuries per unit mass or volume, s

is the s andard deviation of the background counting rate or of tIIe counting rate of a blank sample as appropriate, as count's per

minute, E is the counti.ng efficiency, as counts per disintegration, V is the sample si2e in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A, is the radioactive decay constant for the particular radionuclide, and 5t for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical values of E, Y, Y, and at should he used in the calculation.

ST.

LUCIE - UNIT I 3/4 12-9 Amendment No.

5 9

TABLE 4.12-1 Continued TABLE NOTATION It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an A posteriori (after the fact) limit for a particular measurement.'nalyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally background fluctuations, unavoidable small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances may render these LLDs unachievable.

In such

cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.8.

I

'LLD for drinking water samples.

If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

'An equilibrium mixture of the parent and daughter isotopes which corresponds to 15 pCi/E of the parent isotope.

ST.

LUCIE - UNIT 1

3/4 12-10 Amendment No. 59, 118

REACT'"(.'~Y "",.'<."-"

YSTEMS

~

I BASES 3/4.1. 2 8'3RATION SYSTEt!S Continued T'e boron addition caoability after the plant has been placed in MODES 5

and 6 requires either 3650 gallons of 2.5 to 3.5 weight percent boric acid solution (4371 to 6119 ppm boron) from the boric acid tanks or 11,900 gallons of 173<

~an borated water from the refueling water tank to makeup for contraction of t~e vari-ary coolant that could occur if the temperature is lowered from 200'F to 140'F.

The restrictions associated with the establishing of the flow path

.ro~ the RWT to the,CS via a single HPSI pump provide assurance that Aopendix G pressure/temperature limits will not be exceeded in the case of any inadvertent pressure transient due to a mass addition to the RCS.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of a CEA ejection accident are limited to acceptable levels'he ACTION statements which permit limited variations from the basic requirements are accompanied by additional restrictions which ensure that the original criteria are met.

The ACTION statements applicable to an imnovable or untrippable CEA and to a large misalignment

(>> 15 inches) of two or more CEAs, require a

prompt shutdown of the reactor since either of these conditions may be indicative of a possible loss of mechanical functional capability of the CEAs and in the event of a stuck or untrippable CEA, the loss of SHUTDOWN MARGIN.

For small nisalignments

(< 15 inches) of the CEAs, there is 1) a small degradation in the peaking factors relative to those assumed in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 2) a small effect on the time dependent long term power distributions relative to those used in generating LCOs and LSSS setpoints for DNBR and linear heat rate, 3) a small effect on the available SHUTDOWN MARGIN, and 4) a small effect on the ejected CEA worth used in the safety analysis.

Therefore, the ACTION statement associated with the small misalignment of a CEA permits a one hour time interval during which attempts may be made to restore the CEA to within its alignment requirements prior to initiating a reduction in THERMAL POWER.

The one hour time limit is sufficient to (1) identify causes of a misaligned

CEA, (2) take appropriate corrective action to realign the CEAs, and (3) minimize the effects of xenon redistribution.

Overpower margin is provided to protect the core in the event of a large misalignment

(> 15 inches) of a CEA.

However, this misalignment would cause distortion of the core power distribution.

This distribution may, in turn, have a significant effect on (1) the available SHUTDOWN MARGIN, (2) the time-dependent long-term power distributions relative ta those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis.

Therefo~e, the ACTION statement associated with the large mis-alignment of the CEA requires a prompt realignment of the misaligned CEA.

ST.

LUCIE - UNIT 1

B 3/4 1-3 Amendment No. g7,7J,St~94

R ACTI CON 0

SYST S

BASES 3 4.

3 MO TRO ASS MBL Co t nued The ACTION statements applicable to misaligned or inoperable CEAs include requirements to align the OPERABLE CEAs in a given group with the inoperable CEA.

Conformance with these alignment requirements brings the core, within a short period of time, to a configuration consistent with that assumed in generating LCO and LSSS setpoints.

However, extended operation with CEAs significantly inserted in the core may lead to perturbations in I) local burnup,

2) peaking factors, and 3) available shutdown margin which are more adverse than the conditions assumed to exist in the safety analyses and LCO and LSSS setpoints determination.

Therefore, time limits have been imposed on operation with inoperable CEAs to preclude such adverse conditions from developing.

The requirement to reduce power in certain time limits, depending upon the previous F', is to eliminate a potential nonconservatism for situations when a

CEA has been declared inoperable.

A worst case analysis has shown that a

DNBR SAFDL violation may occur during the second hour after the CEA mis-alignment if this requirement is not met.

This potential DNBR SAFDL violation is eliminated by limiting the time operation is permitted at FULL POWER before power reductions are required.

These reductions will be necessary once the deviated CEA has been declared inoperable.

The time allowed to continue operation at a reduced power level can be permitted for the following reasons:

1.

The margin calculations that support the Technical Specifications are based on a steady-state radial peak of F' > 1.70.

2.

When the actual F' < 1.70, significant additional margin exists.

3.

This additional margin can be credited to offset the increase in F'ith time that can occur following a CEA misalignment.

4.

This increase in F'is caused by xenon redistribution.

5.

The present analysis can support allowing a misalignment to exist for up to 60 minutes without correction, if the initial F' < 1.67.

Operability of the CEA position indicators (Specification 3.1.3.3) is required to determine CEA positions and thereby ensure compliance with the CEA alignment and insertion limits and ensures proper operation of the rod block circuit.

The CEA "Full In" and "Full Out" limits provide an additional inde-pendent means for determining the CEA positions when the CEAs are at either their fully inserted or fully withdrawn positions.

Therefore, the ACTION statements applicable to inoperable CEA position indicators permit continued operations when the positions of CEAs with inoperable position indicators can be verified by the "Full In" or "Full Out" limits.

ST.

LUCIE - UNIT I B 3/4 1-4 Amendment No. 48, 88, 0, 118

PLANT SYSTEMS BASES tl 3 4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM Continued for operations personnel during and following all credible accident condi-tions.

The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5 rem or less whole body, or its equivalent.

This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A", 10 CFR 50.

3 4.7.8 ECCS AREA VENTILATION SYSTEM The OPERABILITY of the ECCS area ventilation system ensures that radio-active materials leaking from the ECCS equipment following a LOCA are filtered prior to reaching the environment.

The operation of this system and the resultant effect on offsite dosage calculations was assumed in the accident analyses.

3 4.7.9 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the total body or individual organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the probable leakage from the source material.

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.

guantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Parts

30. 11-20 and 70. 19.

Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible body burden for total body irradiation if the source material is inhaled or ingested.

3 4.7.10 SNUBBERS All safety related snubbers are required to be OPERABLE to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems.

Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection.

Inspections performed ST.

LUCIE - UNIT 1 B 3/4 7-5 Amendment No. 44, 57., Q, ll8

PLANT SYSTEMS ASES before that interval has elapsed may be used as a

new reference point to deter ine the next inspection.

However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.

Any inspection whose results require a shor'ter inspection interval will override the previous-schedule.

When the cause of the rejection of a snubber is clearly established and emedied for that snubber and for any other snubber. that may be generically susceptible and verified by inservice functional testing, that snubber may be xempted from being counted as inoperable.

Generically susceptible snubbers re those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or re similarly located or exposed to the same environmental conditions such as emperature, radiation,. and vibration.

When a snubber is found inoperable, an evaluation is performed, in addition o the determination of the snubber mode of failure, in order to determine if ny safety-related component or system has been adversely affected by the inoper-bility of the snubber.

The engineering evaluation shall determine whether or ot the snubber mode of failure has imparted a significant effect or degradation n the supported component or system.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested during plant shut-owns at 18 month intervals.

Observed failures of these sample snubbers shall require functional testing of additional units.

In cases where the cause of failure has been identified, additional snubber having a high probability for the same type failure or that are being used in the same application that caused the failure shall be tested.

This requirement increases the probability of locating inoperable snubbers without testing 100%

f the snubbers.

Hydraulic snubbers and mechanical snubbers may each be treated as a

different entity for the above surveillance programs.

The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal

replaced, spring replaced, in high radiation area, in high temperature

area, etc....).

The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

These records will provide statistical bases for future consideration of snubber service life.

The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

ST.

LUCIE - UNIT 1

B 3/4 7-6 Amendment No. 44 JUL 2 v 1987

REFUELING OPERATIONS BASES 3 4.9. 12 FUEL POOL VENTILATION SYSTEM-FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assump-tions of the accident

analyses, 3 4.9. 13 SPENT FUEL CASK CRANE The maximum load which may be handled by the spent fuel cask crane is limited to a loaded single element cask which is equivalent to approxi-mately 25 tons.

This restriction is provided to ensure the structural integrity of the spent fuel pool in the event of a dropped cask accident.

Structural damage caused by dropping a load in excess of a loaded single element cask could cause keakage from the spent fuel pool in excess of the maximum makeup capability.

3 4.9. 14 DECAY TIME -

STORAGE POOL The minimum requirements for decay of the irradiated fuel assemblies in the entire spent fuel storage pool prior to movement of the spent fuel cask into the fuel cask compartment ensure that sufficient time has elapsed to allow radioactive decay of the fission products.

The decay time of 1180 hours0.0137 days 0.328 hours 0.00195 weeks 4.4899e-4 months is based upon one-third of a core placed in the spent fuel pool each year during refueling until the pool is filled.

The decay time of 1490 hours0.0172 days 0.414 hours 0.00246 weeks 5.66945e-4 months is based upon one-third of a core being placed in the spent fuel pool each year during refueling following which an entire core is placed in the pool to fill it.

The cask drop analysis assumes that all of the irradiated fuel in the filled pool (7 2/3 cores) is ruptured and follows Regulatory Guide 1.25 methodology, except that a Radial Peaking Factor of 1.0 is applied to all irradiated assemblies.

ST.

LUCIE - UNIT 1

B 3/4 9-3 Amendment No. 84, 40, SE. ll8

ADMINISTRATIVE CONTROLS 6.13 PROCESS CONTROL PROGRAM PCP Licensee initiated changes to the PCP:

l.

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:

b.

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

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:.

and found acceptable by the FRG.

'2.

Shall become effective upon review and acceptance by the FRG.

6.14 OFFSITE DOSE CALCULATION MANUAL ODCM Licensee initiated changes to the ODCM:

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 supplemental 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 determina-tions; 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 1

6-23 Amendment No. $/,h9, 86 OCT p q lgsy

ADMINISTRATIVE CONTRO 6.15 MAJOR CHANGES TO RADIOACTIVE LI UID GASEOUS AND SOLID WASTE TREATMENT SYSTEMS*

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 individuals in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the license application and amendments thereto; f.

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 choose to submit the information called for in this Specifica-tion as part of the annual FSAR update.

ST.

LUCIE - UNIT 1

6-24 Amendment No.

69. ll8

y,g REC(g P0 C

a o~

++*++

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 FLORIDA POWER 8E LIGHT COMPANY ORLANDO UTILITIES COMMISSION OF THE CITY OF ORLANDO FLORIDA AND FLORIDA MUNICIPAL POWER AGENCY DOCKET NO. 50-389 ST.

LUCIE PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

6O License No. NPF-16 1.

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

A.

The application for amendment by Florida Power 8 Light Company, et al. (the licensee),

dated August 21,

1992, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, Facility Operating License No. NPF-16 is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and by amending paragraph 2.C.2 to read as follows:

2.

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Amendment No.

60

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

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

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 24, 1992

ATTACHMENT TO LICENSE AMENDMENT NO.

60 TO FACILITY OPERATING LICENSE NO.

NPF-16 DOCKET NO. 50-389 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

The corresponding overleaf pages are also provided to maintain document completeness.

Remove Pa es 2-4 2-5 3/4 2-9 3/4 3-2 3/4 3-13 3/4 6-2 3/4 6-26 3/4 8-7 3/4 9-9 3/4 11-10 3/4 12-10 6-23 Insert Pa es 2-4 2-5 3/4 2-9 3/4 3-2 3/4 3-13 3/4 6-2 3/4 6-26 3/4 8-7 3/4 9-9 3/4 11-10 3/4 12-10 6-23

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TABLE 2.2-1 REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS FUNCTIONAL UNIT

1. Manual Reactor Trip

2. Variable Power Level - High

"'our Reactor Coolant Pumps Operating

3. Pressurizer Pressure

- High

4. Thermal Margin/Low Pressure"'our Reactor Coolant Pumps Operating

5. Containment Pressure

- High 6.

Steam Generator Pressure Low 7.

Steam Generator Pressure"'ifference

- High (Logic in TM/LP Trip Unit) 8.

Steam Generator Level Low TRIP SETPOINT Not Applicable

~ 9.61% above THERMAL POWER, with a minimum setpoint of 15% of RATED THERMAL POWER, and a maximum of ~ 107.0% of RATED THERMAL POWER.

~ 2370 psia Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.

Minimum value of 1900 psia.

< 3.0 psig

~ 626.0 psia (2)

< 120.0 psid

~ 20.5% (3)

ALLOWABLE VALUES Not Applicable

~ 9.61% above THERMAL POWER, and a minimum setpoint of 15% of RATED THERMAL POWER and a maximum of ~ 107.0% of RATED THERMAL POWER.

~ 2374 psia Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-3 and 2.2-4.

Minimum value of 1900 psia.

< 3.1 psig

~ 621.0 psia (2)

~ 132.0 psid

~ 19.5% (3)

TABLE 2.2-1 Continued REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS C

y

~

FUNCTIONAL UNIT o

9.

m I

10.

Local Power Density - High

"'perating Loss of Component Cooling Water to Reactor Coolant Pumps-Low 15.

Loss of Load (Turbine)

Hydraulic Fluid Pressure

- Low'"

11.

Reactor Protection System Logic 12.

Reactor Trip Breakers 13.

Rate of Change of Power -

High"'4; Reactor Coolant Flow -

Low"'RIP SETPOINT Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2,

) 636 gpm"*

Not Applicable Not Applicable

~ 2.49 decades per minute

~ 95.4% of design Reactor Coolant flow with four pumps operating*

~ 800 psig ALLOWABLE VALUES Trip setpoint adjusted to not exceed the limit lines of Figures 2.2-1 and 2.2-2.

> 636 gpm Not Applicable Not Applicable

( 2.49 decades per minute

~ 94.9% of design Reactor Coolant flow with four pumps operating*

~ 800 psig

Design reactor coolant flow with four pumps operating is 363,000 gpm.

- 10-minute time delay after relay actuation.

TABLE 2.2-1 (Continued)

REACTOR PROTECTIVE INSTRUHENTATION TRIP SETPOINT LIHITS TABLE NOTATION (1)

Trip may be manually bypassed below 0.5X of RATED THERHAL POWER during testing pursuant to Special Test Exception

3. 10.3; bypass shall be automatically removed when the THERHAt'OWER is greater than or equal to 0.5X of RATED TNERHAL POWER.

(2)

Trip may be manually bypassed below 705 psig; bypass shall be automatically removed at or above 705 psig.

(3)

X of the narrow range steam generator level indication.

(4)

Trip may be bypassed below 10-~X and above 15K of RATED THERHAL POMER; bypass shall be automatically removed when THERHAL POWER is > 10-~X or

< 15K of RATED THERHAL POWER.

'I (5)

Trip may be bypassed below 15K of RATED THERHAL POMER; bypass shall be automatically removed when THERHAL POWER is greater than or equal to 15K of RATED THERHAL POMER.

POWER DISTRIBUTION L ITS TOTAL INTEGRATED RADIAL PEAKING FACTOR -

F

[.IMITING CONDITION FOR OPERATION 3.2.3 The calculated value of F, shall be limited to ~ 1.70.

APPLICABILITY:

MODE 1*.

ACTION:

With F, > 1.70, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months either:

Be in at least HOT STANDBY, or b.

Reduce THERMAL POWER to bring the combination of THERMAL POWER and F,

to within the limits of Figure 3.2-3 and withdraw the full-length CEAs to or beyond the Long Term Steady State Insertion Limits of Specification

3. 1.3.6.

The THERMAL POWER limit deter-mined from Figure 3.2-3 shall then be used to establish a revised upper THERMAL POWER level limit on Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fraction of RATED THERMAL POWER determined by Figure 3.2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4.

SURVEILLANCE RE UIREMENTS 4.2.3. 1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F,

shall be calculated by the expression F, = F,(l+T,) when F, is calculated with a non-full core power distribution analysis code and shall be calculated as F,

= F, when calculations are performed with a full core power distribution analysis code.

F, shall be determined to be within its limit at the following intervals:

a 0 b.

C.

Prior to operation above 70% of RATED THERMAL POWER after each fuel loading, At least once per 31 days of accumulated operation in MODE 1, and Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if the AZIMUTHAL POWER TILT (T,) is

> 0.03.

See Special Test Exception 3. 10.2.

ST.

LUCIE - UNIT 2 3/4 2-9 Amendment No. 8, 60

POWER OISTRIBUTION LIMITS SURVEILLANCE RE UIREMENTS Continued 4.2.3.3 Fr shall be determined each time a calculation of F is required by using the.incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing reactor coolant pump combination.

4.2.3.4 T

shall be determined each time a calculation of F is made using r

a non-full core power distribution analysis code.

The value of T used to T

~

q determine Fr in this case shall be the measured value of T 0

ST.

LUCIE - UNIT 2 3/4 2-10

3/4.3 INSTRUMENTATION 3/4.3. 1 REACTOR PROTECTIVE INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3. 1 As a minimum, the reactor protective instrumentation channels and bypasses of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE RE UIREMENTS

4. 3. l. 1 Each reactor protective instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3"l.

4.3. 1.2 The logic for the bypasses shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days.

The total bypass function shall be demonstrated OPERABLE at least once per 18 months dut ing CHANNEL CALIBRATION testing of each channel affected by bypass operation.

4. 3. 1. 3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

ST.

LUCIE - UNIT 2 3/4 3-1

TABLE 3.3-1 REACTOR PROTECTIVE INSTRUMENTATION FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Variable Power Level - High 3.

Pressurizer Pressure

- High 4.

Thermal Margin/Low Pressure 5.

Containment Pressure

- High 6.

Steam Generator Pressure

- Low 7.

Steam Generator Pressure Difference - High 8.

Steam Generator Level - Low 9.

Local Power Density - High 10.

Loss of Component Cooling Water to Reactor Coolant Pumps TOTAL NO.

OF CHANNELS 4/SG 4/SG CHANNELS TO TRIP 2

2 2(a)(d) 2(a)(d) 2/SG(b) 2(a)(d) 2/SG 2(c)(d)

MINIMUM CHANNELS OPERABLE 3/SG 3/SG APPLICABLE MODES 1,

2 3*

4*

5*

1, 2

ACTION 1

5 2¹ 2¹

]

2¹ 2¹ 2¹

11. Reactor Protection System Logic 4

12. Reactor Trip Breakers 2(f) 4 1,

2 3*

4*

5*

1, 2

3*

4*

5*

2¹ 5

4 5

13.

Wide Range Logarithmic Neutron Flux Monitor a.

Startup and Operating-Rate of Change of Power-High b.

Shutdown

14. Reactor Coolant Flow -

Low 4/SG 2(e)(9) 0 2/SG(a)(d) 3/SG 1,

2 3, 4, 5

1, 2

2¹ 3

z¹ 15.

Loss of Load (Turbine Hydraulic Fluid Pressure

- Low) 4 2(c) 2¹

FUNCTIONAL UNIT TABLE 3.3-3 Continued ENGINEERED SAFETY EATURES ACTUATION SYS E

INSTRUMENTATION MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE OF CHANNELS TO TRIP OPERABLE MODES ACTION 4.

MAIN STEAM LINE ISOLATION (MSIS) a.

Manual (Trip Buttons) b.

Steam Generator Pressure - Low c.

Containment Pressure High d.

Automatic Actuation Logic 5.

CONTAINMENT SUMP RECIRCULATION (RAS) a.

Manual RAS (Trip Buttons) b.

Refueling Water Storage Tank - Low c.

Automatic Actuation Logic 4/steam generator 2/steam generator 3/steam 1, 2, 3(c) generator 1, 2, 3

1, 2, 3

1, 2, 3, 4

1, 2, 3

1 2

1, 2, 3

16 I

13*,

14 13*,

14 12 12 17 12

I C

m C

FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS CHANNELS TO TRIP HININJH CHANNELS OPERABLE APPLICABLE HODES TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEH INSTRINENTATION ACTION 4s I

6.

LOSS OF PS/ER (LOV) a.

(1) 4.16 kV Eaergency Bus Undervol tage (Loss of Voltage)

(2) 480 V Gaergency Bus Undervoltage (Loss of Voltage) 2/Bus b.

(1) 4.16 kV Eaergency Bus Undervoltage (Oegraded Voltage)

- 3/Bus (2) 480 V Baergency Bus Undervoltage (Degraded Voltage) 3/Bus 1/Bus 2/Bus 2/Bus 2/Bus 1/Bus 2/Bus.

2/Bus 2/Bus 1,2,3 1, 2, 3 1, 2, 3

1, 2, 3

12 12 17 17 7.

AUXILIARYFEESfATER (AFAS)

~

a.

Nanual (Trip Buttons) 4/SG b.

Autoaatic Actuation Logic 4/SG c.

SG Level (2A/20) - Lo~

4/SG 8.

AUXILIARYFEEDHATFR ISOLATION a.

SG 2A-- SG 28 Differential Pressure 4/SG 2/SG 2/SG 2/SG 2/SG 4/SG 3/SG 3/SG3/SG',

2, 3

1, 2, 3 1, 2, 3

1,2,3 12 13*, 14 13*, 14 0

b.

Feedeater Header SG 2A -

SG 2B Dif-ferential Pressure 4/SG 2/SG 3/SG 1,2,3 13*, 14

3/4.6 CONTAINMENT SYSTEMS 3/4. 6.

1 PRIMARY COHTAIHMEHT COHTAIHMEHT INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.1.1 Primary CONTAINMENT 'INTEGRITY shall be maintained.

APPLICABILITY:

MODES 1", 2", 3, and 4.

ACTION:

Without primary CONTAINMENT IHTEGRITY, 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 STAHOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLO SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURYEILLANCE RE UIREMENTS

4. 6.1.1 Primary CONTAINMENT INTEGRITY shall be demonstrated:

a.

At least once per 31 days by verifying that all penetrations""

not capable of being closed by OPERABLE containment automatic isolation valves and requi red to be closed during accident conditions are closed by valves, blind flanges, or deactivated automatic valves secured in their positions, except as provided in Table 3.6-2 of Specification 3.6.3.

b.

By ver'.'Tying 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 Tyoe A or B

test, by leak rate testing the seal with gas at P

, 41.8 psig and verifying that when the measured leakage rate for these seals is a'dded to the leakage rates determined pursuant to Specifica-tion 4.6. 1.2d. for all other Type B and C penetrations, the combined leakage rate is less than or equal to 0.60 L

In MOOES 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 insi'de the containment and are locked, sealed or otherwise secured in the closed position.

These penetrations shall be verified closed during each COLO SHUTOOWN except that such verification need not be performed more often than once per 92 days.

ST.

LUCIE - UNIT 2 3/4 6-1 Amendment !!o.

MN 8 ISO

CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION FOR OPERATION 3.6. 1.2 Containment leakage rates shall be limited to:

a.

An overall integrated leakage rate of:

1.

Less than or equal to L., 0.50 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P., 41.8 psig, or 2.

Less than or equal to L 0.35 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduced pressure of P 20d9 psig.

b.

C.

A combined leakage rate of less than or equal to 0.60 L. for all penetrations and valves subject to Type B and C tests, when pressurized to P..

I A combined bypass leakage rate of less than or equal to 0. 12 L. for all penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P..

II E I,,d,d ACTION:

With either (a) the measured overall integrated containment leakage rate exceeding 0.75 L. or 0.75 L as applicable, or (b) with the measured combined leakage rate for all penetrations and valves subject to Types B and C tests exceeding 0.60 L., or (c) with the combined bypass leakage rate exceeding

0. 12 L restore the overall integrated leakage rate to less than or equal to 0.75 L. or less than or equal to 0.75 L as applicable, and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than or equal to 0.60 L and the bypass leakage rate to less than or equal to 0. 12 L, prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREMENTS 4.6. 1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria specified in Appendix J of 10 CFR 50:

a.

Three Type A tests (Overall Integrated Containment Leakage Rate) shall be conducted at 40 + 10 month intervals during ST.

LUCIE

- UNIT 2 3/4 6-2 Amendment No. 86, 81, 60

CONTAINMENT SYSTEMS ELECTRIC HYDROGEN RECOMBINERS "

W LIMITING CONDITION FOR OPERATION

3. 6. 4. 2 Two independent containment hydrogen r ecombiner systems shall be OPERABLE.

APPLICABILITY:

MODES 1 and 2.

ACTION:

With one hydrogen recombiner system inoperable, restore the inoperable system to OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE RE UIREMENTS 4.6.4.2 Each hydrogen recombiner system shall be demonstrated OPERABLE:

At least once per 6 months by verifying during a recombiner system functional test that the minimum heater sheath temperature increases to greater than or equal to 700'F within 90 minutes.

Upon reaching 700 F, increase the power setting to maximum power for 2 minutes and verify that the power meter reads greater than or equal to 60 kW.

b.

At least once per 18 months by:

1.

Performing a

CHANNEL CALIBRATION of all recombiner instru-mentation and control circuits, 2.

Verifying through a visual examination that there is no evidence of abnormal conditions within the recombiner enclosure (i.e.,

loose wiring or structural connections, deposits of foreign materials, etc.).

Verifying the integrity of the heater electrical circuits by performing a resistance to ground test following the above required functional test.

The resistance to ground for any heater phase shall be greater than or equal to 10,000 ohms.

ST.

LUCIE - UNIT 2 3/4 6"25

CONTAINMENT SYSTEMS 3 4.6.5 VACUUM RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.6.5 The primary containment vessel to annulus vacuum relief valves shall be OPERABLE with an 'actuation setpoint of less than or equal to 9.85

+

l 0.35 inches water gauge.

APPLICABILITY:

MODES I, 2, 3 and 4.

ACTION:

With one primary containment vessel to annulus vacuum relief valve inoperable, restore the valve to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 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 4.6.5 No additional Surveillance Requirements other than those required by Specification 4.0.5.

ST.

LUCIE - UNIT 2 3/4 6-26 Amendment No.

60

'ELECTRICAL POWER SY RS SURVEILLANCE RE UIRENENTS Continued c)

Verifying that all automatic diesel generator. trips, except engine overspeed and generator differential, are automatically bypassed upon loss of voltage on the emergency bus concurrent with a safety injection actuation signal.

7.

Verifying the diesel generator operates for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .****

During the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall be loaded within a load band of 3800 to 3985 kW'nd during the remaining 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be loaded within a load band of 3450 to

'685 kW".

The generator voltage and frequency shall be 4160 2 420 volts and 60

+ 1.2 Hz within 10 seconds after the start signal; the steady-state generator voltage and frequency shall be maintained within these limits during this test.

Within 5 minutes after completing this 24-hour

test, perform Surveillance Requirement 4.8. 1. 1.2e.4.b.

8.

Verifying that the auto-connected loads to each diesel generator do not exceed the 2000-hour rating of 3935 kW.

9.

Verifying that the diesel generator's capability to:

a)

Synchronize with the offsite power source while the generator is loaded with its emergency loads upon a

simulated restoration of offsite power.

b)

Transfer its loads to the offsite power source, and c)

Be restored to its standby status.

10.

Verifying that with the diesel generator operating in a test mode (connected to its bus),

a simulated safety injection signal overrides the test mode by (1) returning the diesel generator to standby operation and (2) automatically energizes the emergency loads with offsite power.

11.

Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the engine-mounted tanks of each diesel via the installed cross connection lines.

O'This band is meant as guidance to avoid routine overloading of the engine.

Variations in load in excess of this band due to changing bus loads shall not invalidate this test.

- - - Thistest may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST.

LUCIE - UNIT 2 3/4 8-7 Amendment No. N

ELECTRICAL POWER

~

TERS SURVEILLANCE RE UIREHENTS Continued 12.. Verifying that the automatic load sequence timers are operable with the interval between each load block within

+'1 second of its design interval.

f.

A't least once per 10 years or after any modifications which could affect diesel generator interdependence by starting***~

the diesel generators simultaneously, during shutdown, and verifying that the diesel generators accelerate to approximately 900 rpm in less than or equal to 10 seconds.

g.

At least once per 10 years by:

l.

Oraining each fuel oil storage tank, removing the accumulated sediment and cleanihg the tank using a

sodium hypochlorite solution, and 2.

Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection NO of the ASIDE Code at a test pressure equal to 110% of the system design pressure.

4.g.'l.l.3

~ge orts - All dtdsel generator fatlnres, valtd or non-valtd, sha11 be reported to the Commission pursuant to Specification 6.9.1.

Reports of diesel generator failures shall include the information recoleended in Regula-tory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is greater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

4.8.1.1.4 The Class 1E underground cable system shall be demonstrated OPERABLE within 30 days after the movement of any loads fn excess of 80% ef the ground surface design basis load over the cable ducts by pulling a mandrel with a diameter of at least 80% of the duct's inside diameter through a duct exposed to the maximum loading (duct nearest the ground's surface) and verifying that the duct has not been damaged.

~"*This test may be conducted in accordance with the manufacturer's recommendations concerning engine prelube period.

ST.

LUGIE - UNIT 2 3/4 8-7a Amendment No.

39 FEB z 1989

'EFUELING OPERATION

'LOW WATER LEVEL LIMITING CONDITION FOR OPERATION

't'c 3.9.8.2 Two independent shutdown c'ooling loops shall be OPERABLE and at least one shutdown cooling loop shall be in operation.

APPLICABILITY:

MODE 6 when the water level above the top of the reactor pressure vessel flange is less than 23 feet.

ACTION:

a ~

b.

With less than the required shutdown cooling loops

OPERABLE, within I hour initiate corrective action to return the required loops to OPERABLE status, or to establish greater than or equal to 23 feet of water above the reactor pressure vessel

flange, as soon as possible.

With no shutdown cooling loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and within I hour initiate corrective action to return the required shutdown cooling loop to operation.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMENTS 4.9.8.2 At least one shutdown cooling loop shall be verified to be in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST.

LUCIE - UNIT 2 3/4 9-9 Amendment No. 48, 60

REFUELING OPERATIONS 3/4.9.9 CONTAINMENT ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The containment isolation system shall be OPERABLE.

APPLICABILITY: Ouring CORE ALTERATIONS or movement of irradiated fuel ~ithin t.

ACTION:

With the containment isolation system inoperable, close each of the containment penetrations providing direct access from the containment atmosphere to the outside atmosphere.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREHENTS 4.9.9 The containment isolation system shall be demonstrated OPERABLE ~ithin 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment isolation occurs on manual initiation and on a high radiation test signal from each of the containment radiation monitoring instrumentation channels.

ST.

LUCIE - UNIT 2 3/4 9-10

TABLE 4. 11-2 Continued TABLE NOTATION a

~

The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net

count, above system background, that will be detected with 95K probability with only 5X probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement

system, which may include radiochemical separation:

LLD =

4.66 sb E

~

V

~

2. 22

~

Y

~

exp(-Aht)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, sb is the standard deviation of the background counting rate or of-the counting rate of a blank sample as appropriate, as counts per

minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A, is the radioactive decay constant for the particular radionuclide, and ht for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical values of E, V, Y, and bt should be used in the calculation.

ST.

LUCIE - UNIT 2 3/4 11-9

TABLE 4. 11-2 Continued TABLE NOTATION b.

Sampling and analysis shall also be performed following shutdown,

startup, or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months unless (1) analysis shows that the DOSE EQUIVALENT I-131 concentration in the primary coolant has not increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has not increased by more than a factor of 3.

c ~

Samples shall be changed at least 4 times a month and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (or after removal from sampler).

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 7 days following each

shutdown, startup or THERMAL POWER change exceeding 15% of RATED THERMAL POWER in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing if (1) analysis shows that the DOSE E(UIVALENT I-131 concentration in the primary coolant has increased more than a factor of 3; and (2) the noble gas activity monitor shows that effluent activity has increased by more than a factor of 3.

When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months are analyzed, the corresponding LLDs may be increased by a factor of 10.

d.

The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with Specifications

3. 11.2. 1, 3. 11.2.2 and 3. 11.2.3.

e.

The principal gamma emitters for which the LLD specification applies exclusively are the following radionuclides:

Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe-138 for gaseous emissions and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs-134, Cs-'137, Ce-141 and Ce-144 for particulate emissions.

This list does not mean that only these nuclides are to be detected and reported.

Other peaks which are measurable and identifiable, together with the above nuclides, shall also be identified and reported.

ST.

LUCIE - UNIT 2 3/4 11-10 Amendment No. N

TABLE 4. 12-1 Continued TABLE NOTATION v!

'I This list does not mean that only these nuclides are to be considered.

Other peaks that are identifiable, together with those of the above

nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1. 11.

Required detection capabilities for thermoluminescent dosimeters used for environmental measurements are given in Regulatory Guide 4. 13.

cThe LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a net

count, above system background, that will be detected with 95K probability with only 5X probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement

system, which may include radiochemical separation:

LLD 66.b E

~

V

~

2. 22

~

Y exp(-Q,t)

Where:

LLD is the "a priori" lower limit of detection as defined above, as picocuries per unit mass or volume, s

is the standard deviation of the background counting rate or of tie counting rate of a blank sample as appropriate, as counts per

minute, E is the counting efficiency, as counts per disintegration, V is the sample size in units of mass or volume, 2.22 is the number of disintegrations per minute per picocurie, Y is the fractional radiochemical yield, when applicable, A, is the radioactive decay constant for the particular radionuclide, and bt for environmental samples is the elapsed time between sample collection, or end of the sample collection period, and time of counting Typical values of E, V, Y, and bt should be used in the calculation.

ST.

LUCIE - UNIT 2 3/4 12-9

TABLE 4.12-1 Continued e

a h

TABLE NOTATION It should be recognized that the LLD is defined as an a griori (before the fact) limit representing the capability of a measurement system and f<<h F )lid f ment.

Analyses shall be performed in such a manner that the stated LLDs will be achieved under routine conditions.

Occasionally background fluctuations, unavoidable small sample sizes, the 'presence of interfering

nuclides, or other uncontrollable circumstances may render these LLDs unachievable.

In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9. 1.8.

LLD for drinking water samples.

If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

'An equilibrium mixture of the parent and daughter isotopes which corresponds to 15 pCi/l of the parent isotope.

ST.

LUCIE - UNIT 2 3/4 12-10 Amendment No.

60

- ADMINISTRATIVE CONT S

6. 13 PROCESS CONTROL PROGRAM PCP 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 support 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 and found acceptable by the FRG.

2.

Shall become effective upon review and acceptance by the FRG.

6.14 OFFSITE DOSE CALCULATION MANUAL ODCM Licensee initiated changes to the ODCM:

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 wail 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. 43, 28, 45a 6o

ADMINISTRATIVE CONTROLS 6.15 MAJOR CHANGES TO RADIOACTIVE LI UIO GASEOUS AND SOLID WASTE TREATMENT SYS MS

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

Shall be >eported 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 pred'icted in the license application and amendments thereto; 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; 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 cnose to submit the information called for in this Specification as part of the annual FSAR update.

ST.

LUCIE - UNIT 2 6

24 Amendment No.

~3

ML17227A675

Text

Attachment:

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