Text

Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.i.4
	ML031070133
Person / Time
Site:	Saint Lucie
Issue date:	04/14/2003
From:	Jefferson W; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2003-083
	Download: ML031070133 (101)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0

FPL April 14, 2003 L-2003-083 10 CFR 50.4 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: St. Lucie Unit 1 Docket No. 50-335 Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.i.4 Pursuant to Technical Specification (TS) 6.8.4.j.4, Florida Power & Light Company is submitting the periodic report of changes made to the St. Lucie Unit I TS Bases without prior NRC approval. The requirement for the periodic report was added by St. Lucie Unit I License Amendment 176 on July 12, 2001, and is required on a frequency consistent with 10 CFR 50.71(e) for UFSAR updates. FPL submits the 10 CFR 50.71(e) reports within six months of the completion of each refueling outage. This first periodic report covers the period from July 12, 2001 to the startup from the fall 2002 Unit 1 refueling outage (SLI-1 8).

Based on discussions with the NRC project manager, FPL is submitting the current revision of ADM.25.04, Appendix A, Attachments 1 through 13, St. Lucie Unit I Technical Specification Bases. Each attachment summarizes the revisions on the attachment cover page.

Please contact George Madden at (772) 467-7155 if there are any questions about this submittal.

Vice President St. Lucie Plant WJ/spt Attachments L

an FPL Group company

REVISION NO.: PROCEDURE TITLE: PAGE:

7 ST. LUCIE PLANT TECHNICAL SPECIFICATIONS BASES PROCEDURE NO.: CONTROL PROGRAM AND TECHNICAL SPECIFICATIONS BASES 8 of 9 ADM-25.04 ST. LUCIE PLANT APPENDIX A ST. LUCIE UNIT I TECHNICAL SPECIFICATION BASES (Page 1 of 1)

Attachment Title Revision BASES for Section 2.0 -

SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 BASES for Sections 3.0 and 4.0 -

2 LIMITING CONDITIONS FOR OPERATION AND 1 -J SURVEILLANCE REQUIREMENTS BASES for Sections 3/4.1 -

REACTIVITY CONTROL SYSTEMS 1 BASES for Sections 3/4.2 -

POWER DISTRIBUTION LIMITS 0 BASES for Sections 3/4.3 -

INSTRUMENTATION 0 6 BASES for Sections 3/4.4 -

REACTOR COOLANT SYSTEM 1 BASES for Sections 3/4.5 -

EMERGENCY CORE COOLING SYSTEMS (ECCS) 0 BASES for Sections 3/4.6 -

8 CONTAINMENT SYSTEMS 1 BASES for Sections 3/4.7 -

PLANT SYSTEMS 0 10 BASES for Sections 3/4.8 -

ELECTRICAL POWER SYSTEMS 1 11 BASES for Sections 3/4.9 -

REFUELING OPERATIONS 2 12 BASES for Sections 3/4.10 - 0 SPECIAL TEST EXCEPTIONS 0 13 BASES for Sections 3/4.11 -

RADIOACTIVE EFFLUENTS 0 END OF APPENDIX A

ST. LUCIE UNIT 1 SectionNo.

TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT I I FPL OF ADM-25.04 Current Revision No.

.Effective Date SAFETY RELATED 09106101

Title:

SAFETY LIMITS AND LIMITING SAFETY SETTINGS Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S I OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN SECTION 2.0 SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHN ICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 2 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 . LUC UNI 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 2.0 ................. 3 2.1 SAFETY LIMITS .. 3 BASES .. 3 2.1.1 REACTOR CORE .3 2.1.2 REACTOR COOLANT SYSTEM PRESSURE . 4 FIGURE B 2.1-1 AXIAL POWER DISTRIBUTION FOR THERMAL MARGIN SAFETY LIMITS .5 2.2 LIMITING SAFETY SYSTEM SETTINGS . . 6 BASES .. 6 2.2.1 REACTOR TRIP SETPOINTS .6 Manual Reactor Trip .6 Power Level-High .6 Reactor Coolant Flow-Low .7 Pressurizer Pressure-High .7 Containment Pressure-High .7 Steam Generator Pressure-Low .7 Steam Generator Water Level-Low .8 Local Power Density-High .8 Thermal Margin/Low Pressure .9 Asymmetric Steam Generator Transient Protective Trip Function (ASGTPTF) .9 Loss of Turbine .10 Rate of Change of Power-High .10

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT I OF ADM-25.04 3 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 BASES FOR SECTION 2.0 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel cladding 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 will 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 coefficient. DNB is not a directly measured parameter during operation and therefore THERMAL POWER, Reactor Coolant Temperature and Pressure have been related to DNB using a DNB correlation developed to predict the Critical Heat Flux (CHF) for DNB. The CHF is the heat flux at a particular core location that would cause DNB. The ratio of the CHF to the actual local heat flux at a particular core location is called the DNB Ratio (DNBR) and is indicative of the margin to DNB.

The minimum allowed value of the DNBR during steady state operation, normal operational transients, and anticipated transients is the DNBR limit from the appropriate DNB correlation. The DNBR limit corresponds to a 95%

probability at a 95% confidence level that DNB will not occur at a particular core location, providing appropriate margin to DNB for all operating conditions. In a core with fuel assemblies of different designs (mixed core),

there may be more than one DNB correlation and associated DNBR limit that defines DNB for the core.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure, and maximum cold leg temperature with four Reactor Coolant Pumps operating for which the DNBR limit corresponding to the Siemens Power Corporation (SPC)XNB DNB correlation is not violated for the following conditions:

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 4 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 2.1 SAFETY LIMITS (continued)

BASES (continued) 2.1.1 REACTOR CORE (continued)

1. reactor coolant inlet temperatures less than or equal to 580 0F,

2. THERMAL POWER less than or equal to 112%,

3. reactor coolant vessel flow of 365,000 gpm, and

4. the axial power shape shown on Figure B2.1-1.

The dashed line at 5800 F coolant inlet temperature is not a safety limit; however, operation above 5800 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 setpoint specified in Table 2.2-1. The area of safe operation is below and to the left of these lines.

The reactor protective system in combination with the Limiting Conditions for Operation is designed to prevent any anticipated combination of transient conditions for reactor coolant system temperature, pressure, and thermal power level that would result in a DNBR of less than the DNBR limit and preclude the existence of flow instabilities. Specific verification of the DNBR limit with an appropriate DNB correlation ensures that the Reactor Core Safety Limit is satisfied.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant components which permits a maximum transient pressure of 110% (2750 psia) of design pressure. The Reactor Coolant System piping, valves and fittings are designed to ANSI B 31.7, Class I which permits a maximum transient pressure of 110% (2750 psia) of component design pressure. The Safety Limit of 2750 psia is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3125 psia to demonstrate integrity prior to initial operation.

SECTIONNO.:ITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 5 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 FIGURE B 2.1-1 AXIAL POWER DISTRIBUTION FOR THERMAL MARGIN SAFETY LIMITS 1.6 1.4 1.2 z

0 0.

w 0.6 0.4 0.2 0

0 10 20 30 40 50 60 70 80 90 100 PERCENT OF ACTIVE CORE LENGTH FROM BOTTOM

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT I OF ADM-25.04 6 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 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 Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Level-High 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 setpoint is operator adjustable and can be set no higher than 9.61% 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 15% of RATED THERMAL POWER. Adding to this maximum value the possible variation in trip point due to calibration and instrument errors, the maximum actual THERMAL POWER level at which a trip would be actuated is 112% of RATED THERMAL POWER, which is consistent with the value used in the safety analysis.

SECTION NO.: MILE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 7 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 2.2 LIMITING SAFETY SYSTEM SETTINGS (continued)

BASES (continued) 2.2.1 REACTOR TRIP SETPOINTS (continued)

Reactor Coolant Flow-Low 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.

Pressurizer Pressure-High 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 concurrent operation with the power-operated relief valves avoids the undesirable operation of the pressurizer code safety valves.

Containment Pressure-High 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 subsequent cooldown of the reactor coolant. The setting of 600 psia is sufficiently below the full-load operating point so as not to interfere with normal operation, but still high enough to provide the required protection in the event of excessively high steam flow. This setting was used with an uncertainty factor of +/- 22 psi in the accident analyses.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 8 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 2.2 LIMITING SAFETY SYSTEM SETTINGS (continued)

BASES (continued) 2.2.1 REACTOR TRIP SETPOINTS (continued)

Steam Generator Water Level-Low The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the design pressure of the reactor coolant system will not be exceeded due to loss of steam generator heat sink. The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to provide sufficient time for any operator action to initiate auxiliary feedwater before reactor coolant system subcooling is lost.

Local Power Density-High The local Power Density-High trip, functioning from AXIAL SHAPE INDEX monitoring, is provided to ensure that the peak local density in the fuel which corresponds to fuel centerline melting will not occur as a consequence of axial power maldistributions. A reactor trip is initiated whenever the AXIAL SHAPE INDEX exceeds the allowable limits of Figure 2.2-2. The AXIAL SHAPE INDEX is calculated from the upper and lower ex-core neutron detector channels. The calculated setpoints are generated as a function of THERMAL POWER level with the allowed CEA group position being inferred from the THERMAL POWER level. The trip is automatically bypassed below 15 percent power.

The maximum AZIMUTHAL POWER TILT and maximum CEA misalignment permitted for continuous operation are assumed in generation of the setpoints. In addition, CEA group sequencing in accordance with the Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 9 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 2.2 LIMITING SAFETY SYSTEM SETTINGS (continued)

BASES (continued) 2.2.1 REACTOR TRIP SETPOINTS (continued)

Thermal MaralnlLow Pressure The Thermal Margin/Low Pressure trip is provided to prevent operation when the DNBR is less than the DNBR limit.

The trip is initiated whenever the reactor coolant system pressure signal drops below either 1887 psia or a computed value as described below, whichever is higher. The computed value is a function of the higher of AT power or neutron power, reactor inlet temperature, the number of reactor coolant pumps operating and the AXIAL SHAPE INDEX. The minimum value of reactor coolant flow rate, the maximum AZIMUTHAL POWER TILT and the maximum CEA deviation permitted for continuous operation are assumed in the generation of this trip function. In addition, CEA group sequencing in accordance with Specifications 3.1.3.5 and 3.1.3.6 is assumed. Finally, the maximum insertion of CEA banks which can occur during any anticipated operational occurrence prior to a Power Level-High trip is assumed.

The Thermal Margin/Low Pressure trip setpoints include appropriate allowances for equipment response time, calculational and measurement uncertainties, and processing error. A further allowance is included to compensate for the time delay associated with providing effective termination of the occurrence that exhibits the most rapid decrease in margin to the DNBR limit.

Asymmetric Steam Generator Transient Protective Trip Function (ASGTPTF)

The ASGTPTF consists of Steam Generator pressure inputs to the TM/LP calculator, which causes a reactor trip when the difference in pressure between the two steam generators exceeds the trip setpoint. The ASGTPTF is designed to provide a reactor trip for those events associated with secondary system malfunctions which result in asymmetric primary loop coolant temperatures. The most limiting event is the loss of load to one steam generator caused by a single main steam isolation valve closure.

The equipment trip setpoint and allowable values are calculated to account for instrument uncertainties, and will ensure a trip at or before reaching the analysis setpoint.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

2.0 BASES ATTACHMENT 1 OF ADM-25.04 10 of 10 REVISION NO.: SAFETY LIMITS AND LIMITING SAFETY SETTINGS 0 ST. LUCIE UNIT 1 2.2 LIMITING SAFETY SYSTEM SETTINGS (continued)

BASES (continued) 2.2.1 REACTOR TRIP SETPOINTS (continued)

Loss of Turbine A Loss of Turbine trip causes a direct reactor trip when operating above 15%

of RATED THERMAL POWER. This trip provides turbine protection, reduces the severity of the ensuing transient and helps avoid the lifting of the main steam line safety valves during the ensuing transient, thus extending the service life of these valves. No credit was taken in the accident analyses for operation of this trip. Its functional capability at the specified trip setting is required to enhance the overall reliability of the Reactor Protection System.

Rate of Chanae of Power-High 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 administratively enforced startup rate limit. The trip is not credited in any design basis accident evaluated in UFSAR Chapter 15; however, the trip is considered in the safety analysis in that the presence of this trip function precluded the need for specific analyses of other events initiated from subcritical conditions.

ST. LUCIE UNIT 1 SectionsNo.

TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 2 2 LOF ADM-25.04 Current Revision No.

FPL I Effective Date SAFETY RELATED 01106103

Title:

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS Responsible Department: Licensing REVISION

SUMMARY

Revision I - Updated TS Bases for TS Amendment No. 186 - missed surveillances.

(Larry Donghia, 01/03/03)

Revision 0- Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S-LOPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Secions3.0&4.0 1 01/03/03 R.E. Rose 01/03/03 SYS Plant General Manager COM COMPLETED ITM 1

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 2 of 12 1 AND SURVEILLANCE REQUIREMENTS IaST. LUCIE UNITP _

TABLE OF CONTENTS SECTION PAGE BASES FOR SECTIONS 3.0 & 4.0 ........................ 3 3/4.0 APPLICABILITY .3 BASES .3 3.0.1 .3 3.0.2 .4 3.0.3 .5 3.0.4 .7 4.0.1 . .......................................................................................... 8 4.0.2 . ........................................................................................ 10 4.0.3 . ........................................................................................ 10 4.0.4 . ........................................................................................ 12 4.0.5 . ........................................................................................ 12

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 3 of 12 1 AND SURVEILLANCE REQUIREMENTS I ST. LUCIE UNIT 1 _

BASES FOR SECTIONS 3.0 & 4.0 314.0 APPLICABILITY BASES The specifications of this section establish the general requirements applicable to Limiting Conditions for Operation. These requirements are based on the requirements for Limiting Conditions for Operation stated in the Code of Federal Regulations, 10 CFR 50.36(c)(2):

"Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility.

When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specification until the condition can be met."

3.0.1 This specification establishes the Applicability statement within each individual specification as the requirement for when (i.e., in which OPERATIONAL MODES or other specified conditions) conformance to the Limiting Conditions for Operation is required for safe operation of the facility. The ACTION requirements establish those remedial measures that must be taken within specified time limits when the requirements of a Limiting Condition for Operation are not met.

There are two basic types of ACTION requirements. The first specifies the remedial measures that permit continued operation of the facility which is not further restricted by the time limits of the ACTION requirements. Inthis case, conformance to the ACTION requirements provides an acceptable level of safety for unlimited continued operation as long as the ACTION requirements continue to be met. The second type of ACTION requirement specifies a time limit in which conformance to the conditions of the Limiting Condition for Operation must be met. This time limit is the allowable outage time to restore an inoperable system or component to OPERABLE status or for restoring parameters within specified limits. If these actions are not completed within the allowable outage time limits, a shutdown is required to place the facility in a MODE or condition in which the specification no longer applies. It is not intended that the shutdown ACTION requirements be used as an operational convenience which permits (routine) voluntary removal of a system(s) or component(s) from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 4 of 12 1 AND SURVEILLANCE REQUIREMENTS ST. LUCIE UNIT 1 _

314.0 APPLICABILITY (continued)

BASES (continued) 3.01 (continued)

The specified time limits of the ACTION requirements are applicable from the point in time it is identified that a Limiting Condition for Operation is not met.

The time limits of the ACTION requirements are also applicable when a system or component is removed from service for surveillance testing or investigation of operational problems. Individual specifications may include a specified time limit for the completion of a Surveillance Requirement when equipment is removed from service. Inthis case, the allowable outage time limits of the ACTIN requirements are applicable when this limit expires if the surveillance has not been completed. When a shutdown is required to comply with ACTION requirements, the plant may have entered a MODE in which a new specification becomes applicable. In this case, the time limits of the ACTION requirements would apply from the point in time that the new specification becomes applicable if the requirements of the Limiting Condition for Operation are not met.

3.0.2 This specification establishes that noncompliance with a specification exists when the requirements of the Limiting Condition for Operation are not met and the associated ACTION requirements have not been implemented within the specified time interval. The purpose of this specification is to clarify that (1) implementation of the ACTION requirements within the specified time interval constitutes compliance with a specification and (2) completion of the remedial measures of the ACTION requirements is not required when compliance with a Limiting Condition for Operation is restored within the time interval specified in the associated ACTION requirements.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 5 of 12 1 AND SURVEILLANCE REQUIREMENTS I ST. LUCIE UNIT I I 3/4.0 APPLICABILITY (continued)

BASES (continued) 3.0.3 This specification establishes the shutdown ACTION requirements that must be implemented when a Limiting Condition for Operation is not met and the condition is not specifically addressed by the associated ACTION requirements.

The purpose of this specification is to delineate the time limits for placing the unit in a safe shutdown MODE when plant operation cannot be maintained within the limits for safe operation defined by the Limiting Conditions for Operation and its ACTION requirements. It is not intended to be used as an operational convenience which permits (routine) voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. One hour is allowed to prepare for an orderly shutdown before initiating a change in plant operation. This time permits the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the cooldown capabilities of the facility assuming only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the primary coolant system and the potential for a plant upset that could challenge safety systems under conditions for which this specification applies.

If remedial measures permitting limited continued operation of the facility under the provisions of the ACTION requirements are completed, the shutdown may be terminated. The time limits of the ACTION requirements are applicable from the point in time there was a failure to meet a Limiting Condition for Operation. Therefore, the shutdown may be terminated if the ACTION requirements have been met or the time limits of the ACTION requirements have not expired, thus providing an allowance for the completion of the required actions.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 6 of 12 1 AND SURVEILLANCE REQUIREMENTS ST. LUCIE UNIT I 3/4.0 APPLICABILITY (continued)

BASES (continued) 3.03 (continued)

The time limits of Specification 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the plant to be in the COLD SHUTDOWN MODE when a shutdown is required during the POWER MODE of operation. If the plant is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE of operation applies. However, if a lower MODE of operation is reached in less time than allowed, the total allowable time to reach COLD SHUTDOWN, or other applicable MODE, is not reduced. For example, if HOT STANDBY is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the time allowed to reach HOT SHUTDOWN is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months because the total time to reach HOT SHUTDOWN is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Therefore, if remedial measures are completed that would permit a return to POWER operation, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

The same principle applies with regard to the allowable outage time limits of the ACTION requirements, if compliance with the ACTION requirements for one specification results in entry into a MODE or condition of operation for another specification in which the requirements of the Limiting Condition for Operation are not met. If the new specification becomes applicable in less time than specified, the difference may be added to the allowable outage time limits of the second specification. However, the allowable outage time limits of ACTION requirements for a higher MODE of operation may not be used to extend the allowable outage time that is applicable when a Limiting Condition for Operation is not met in a lower MODE of operation.

The shutdown requirements of Specification 3.0.3 do not apply in MODES 5 and 6, because the ACTION requirements of individual specifications define the remedial measures to be taken.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 7 of 12 1 AND SURVEILLANCE REQUIREMENTS ST. LUCIE UNIT 1 314.0 APPLICABILITY (continued)

BASES (continued) 3.0.4 This specification establishes limitations on MODE changes when a Limiting Condition for Operation is not met. It precludes placing the facility in a higher MODE of operation when the requirements for a Limiting Condition for Operation are not met and continued noncompliance to these conditions would result in a shutdown to comply with the ACTION requirements if a change in MODES were permitted. The purpose of this specification is to ensure that facility operation is not initiated or that higher MODES of operation are not entered when corrective action is being taken to obtain compliance with a specification by restoring equipment to OPERABLE status or parameters to specified limits. Compliance with ACTION requirements that permit continued operation of the facility for an unlimited period of time provides an accept-able level of safety for continued operation without regard to the status of the plant before or after a MODE change. Therefore, in this case, entry into an OPERATIONAL MODE or other specified condition may be made in accordance with the provisions of the ACTION requirements. The provisions of this specification should not, however, be interpreted as endorsing the failure to exercise good practice in restoring systems or components to OPERABLE status before plant startup.

When a shutdown is required to comply with the ACTION requirements, the provisions of Specification 3.0.4 do not apply because they would delay placing the facility in a lower MODE of operation.

Exceptions to this provision have been provided for a limited number of specifications when startup with inoperable equipment would not affect plant safety. These exceptions are stated in the ACTION statements of the appropriate specifications.

The specifications of this section establish the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3):

"Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met."

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 &4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 8 of 12 1 AND SURVEILLANCE REQUIREMENTS I ST. LUCIE UNIT 1 _

3/4.0 APPLICABILITY (continued)

BASES (continued) 4.0.1 SR 4.0.1 establishes the requirement that Surveillance Requirements (SR) must be met during the MODES or other specified conditions in the applicability for which the requirements of the Limiting Condition for Operation apply, unless otherwise specified in the individual SRs. This Specification is to ensure that SRs are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a SR within the specified frequency, in accordance with SR 4.0.2, constitutes a failure to meet a Limiting Condition for Operation (except as allowed by SR 4.0.3).

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when either

a. the systems or components are known to be inoperable, although still meeting the SRs, or

b. the requirements of the SR(s) are known to be not met between required SR performances.

SRs do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated Limiting Condition for Operation are not applicable, unless otherwise specified. The SRs associated with a SPECIAL TEST EXCEPTION (STE) are only applicable when the STE is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

SRs, including SRs invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. SRs have to be met and performed in accordance with SR 4.0.2, prior to returning equipment to OPERABLE status.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 9 of 12 1 AND SURVEILLANCE REQUIREMENTS ST. LUCIE UNIT 1 3/4.0 APPLICABILITY (continued)

BASES (continued) 4.01 (continued)

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable SRs are not failed and their most recent performance is in accordance with SR 4.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process follow.

a. Auxiliary feedwater (AFW) pump turbine maintenance during refueling that requires testing at steam pressures > 800 psi.

However, if other appropriate testing is satisfactorily completed, the AFW System can be considered OPERABLE.

This allows startup and other necessary testing to proceed until the plant reaches the steam pressure required to perform the testing.

b. High pressure safety injection (HPSI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPSI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 10 of 12 1 AND SURVEILLANCE REQUIREMENTS I ST. LUCIE UNIT 1 _

314.0 APPLICABILITY (continued)

BASES (continued) 4.0.2 This specification establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with an 18-month surveillance interval.

It is not intended that this provision be used repeatedly as a convenience to extend the surveillance intervals beyond that specified for surveillances that are not performed during refueling outages. The limitation of Specification 4.0.2 is based on engineering judgment and the recognition that most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

4.0.3 SR 4.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a SR has not been completed within the specified frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified frequency, whichever is greater, applies from the point in time that it is discovered that the SR has not been performed in accordance with SR 4.0.2, and not at the time that the specified frequency was not met.

This delay period provides adequate time to complete SRs that have been missed. This delay period permits the completion of a SRs requirement before complying with required ACTION(s) or other remedial measures that might preclude completion of the SR.

The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the SR, the safety significance of the delay in completing the required SR, and the recognition that the most probable result of any particular SR being performed is the verification of conformance with the requirements.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 11 of 12 1 AND SURVEILLANCE REQUIREMENTS ST. LUCIE UNIT 1 314.0 APPLICABILITY (continued)

BASES (continued) 4.03 (continued)

When a SR with a frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 4.0.3 allows for the full delay period of up to the specified frequency to perform the SR. However, since there is not a time interval specified, the missed SR should be performed at the first reasonable opportunity.

SR 4.0.3 provides a time limit for, and allowances for the performance of, a SR that becomes applicable as a consequence of MODE changes imposed by required ACTION(s).

Failure to comply with the specified frequency for a SR is expected to be an infrequent occurrence. Use of the delay period established by SR 4.0.3 is a flexibility which is not intended to be used as an operational convenience to extend surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified frequency is provided to perform the missed surveillance, it is expected that the missed SR will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on plant risk (from delaying the surveillance as well as any plant configuration changes required or shutting the plant down to perform the SR) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the SR. This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants. This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutdown. The missed surveillance should be treated as an emergent condition as discussed in the Regulatory Guide.

The risk evaluation may use quantitative, qualitative, or blended methods.

The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed SRs for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the course of action. All cases of a missed SR will be placed in the licensee's Corrective Action Program.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3.0 & 4.0 BASES ATTACHMENT 2 OF ADM-25.04 REVISION NO.: LIMITING CONDITIONS FOR OPERATION 12 of 12 1 AND SURVEILLANCE REQUIREMENTS I ST. LUCIE UNIT 1[

3/4.0 APPLICABILITY (continued)

BASES (continued) 4.03 (continued)

If a SR is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the completion times of the required ACTION(s) for the applicable Limiting Condition for Operation begin immediately upon expiration of the delay period. If a surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the completion times of the required ACTION(s) for the applicable Limiting Condition for Operation begin immediately upon the failure of the surveillance.

Completion of the SR within the delay period allowed by this specification, or within the completion time of the ACTIONS, restores compliance with SR 4.0.1.

4.0.4 This specification establishes the requirement that all applicable surveillances must be met before entry into an OPERATIONAL MODE or other condition or operation specified in the Applicability statement. The purpose of this specification is to ensure that system and component OPERABILITY requirements or parameter limits are met before entry into a MODE or condition for which these systems and components ensure safe operation of the facility.

This provision applies to changes in OPERATIONAL MODES or other specified conditions associated with plant shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a lower MODE of operation.

4.0.5 This specification ensures that inservice inspection of ASME Code Class 1, 2 and 3 components will be performed in accordance with a periodically updated version of Section Xl of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Relief from any of the above requirements has been provided in writing by the Commission and is not part of these Technical Specifications.

ST. LUCIE UNIT 1 I Section No.

314.1 F

Attachment No.

TECHNICAL SPECIFICATION S 3

BASES ATTACHMENT 3 Current Revision No.

FP L OF ADM-25.04 I

Title:

SAFETY RELATED I Effective Date 12/04/01 REACTIVITY CONTROL SYSTEMS 111111 Responsible Department: Licensing REVISION

SUMMARY

Revision I - Changes made to reflect TS Amendment #179. (K. W. Frehafer, 11/30/01)

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S I OPS 0 08/30101 R.G. West 08/30101 DATE Plant General Manager DOCT PROCEODURE Revision FRG Review Date Approved By Approval Date DOCN Sectiori 314.1 1 11/29/01 R.G. West 11/30/01 SYS Plant General Manager COM COMPI 'LETED ITM 1

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 2 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.1 .................... 3 3/4.1 REACTIVITY CONTROL SYSTEMS .3 BASES ........................................................................................ 3 3/4.1.1 BORATION CONTROL .3 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN .3 3/4.1.1.3 BORATION DILUTION AND ADDITION .3 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC) .4 3/4.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY .4 3/4.1.2 BORATION SYSTEMS .5 3/4.1.3 MOVABLE CONTROL ASSEMBLIES .7

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 3 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 lST. LUCIE UNIT 1 BASES FOR SECTION 3/4.1 3/4.1 REACTMITY CONTROL SYSTEMS BASES 314.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS Tavg. The most restrictive condition occurs at EOL, with Tavg, at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown. In the analysis of this accident, a minimum SHUTDOWN MARGIN as specified in the COLR for Specification 3.1.1.1 is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN required by Specification 3.1.1.1 is based upon this limiting condition and is consistent with FSAR accident analysis assumptions. For earlier periods during the fuel cycle, this value is conservative. With Tavg < 2000 F, the reactivity transient resulting from a boron dilution event with a partially drained Reactor Coolant System requires a SHUTDOWN MARGIN as specified in the COLR for Specification 3.1.1.2 and restrictions on charging pump operation to provide adequate protection. This SHUTDOWN MARGIN is 1000 pcm conservative for Mode 5 operation with total RCS volume present, however LCO 3.1.1.2 is written conservatively for simplicity.

3/4.1.1.3 BORATION DILUTION AND ADDITION A minimum flow rate of at least 3000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes will be gradual during boron concentration changes in the Reactor Coolant System. A flow rate of at least 3000 GPM will circulate an equivalent Reactor Coolant System volume of 11,400 cubic feet in approximately 26 minutes. The reactivity change rate associated with boron concentration changes will be within the capability for operator recognition and control.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 4 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 ST. LUCIE UNIT 1 314.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 314.1.1 BORATION CONTROL (continued) 314.1.1.4 MODERATOR TEMPERATURE COEFFICIENT (MTC)

The limiting values of the MTC ensure that the assumptions for the MTC used in the accident and transient analyses remain valid through each fuel cycle. Determination of MTC at the specified conditions ensures that the maximum positive and/or negative values of the MTC will not exceed the limiting values.

314.1.1.5 MINIMUM TEMPERATURE FOR CRITICALITY The MTC is expected to be slightly negative at operating conditions.

However, at the beginning of the fuel cycle, the MTC may be slightly positive at operating conditions and since it will become more positive at lower temperatures, this specification is provided to restrict reactor operation when Tavg is significantly below the normal operating temperature.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 5 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 ST. LUCIE UNIT 1 3/4.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 314.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources,

2) charging pumps, 3) separate flow paths, 4) boric acid pumps, and

5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200 0F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions corresponding to the requirements of Specification 3.1.1.2 after xenon decay and cooldown to 2000 F. The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions. This requirement can be met for a range of boric acid concentrations in the Boric Acid Makeup Tanks (BAMTs) and Refueling Water Tank (RWT). This range is bounded by 5400 gallons of 3.5 weight percent (6119 ppm boron) boric acid from the BAMTs and 17,000 gallons of 1720 ppm borated water from the RWT to 8700 gallons of 2.5 weight percent (4371 ppm boron) boric acid from the BAMTs and 13,000 gallons of 1720 ppm borated water from the RWT.

A minimum of 45,000 gallons of 1720 ppm boron is required from the RWT if it is to be used to borate the RCS alone.

The requirements for a minimum contained volume of 401,800 gallons of borated water in the refueling water tank ensures the capability for borating the RCS to the desired level. The specified quantity of borated water is consistent with the ECCS requirements of Specification 3.5.4.

Therefore, the larger volume of borated water is specified here too.

With the RCS temperature below 200 0F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 6 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS ST. LUCIE UNIT 1 314.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 314.2 BORATION SYSTEMS (continued)

Temperature changes in the RCS impose reactivity changes by means of the moderator temperature coefficient. Plant temperature changes are allowed provided the temperature change is accounted for in the calculated SDM. Small changes in RCS temperature are unavoidable and so long as the required SDM is maintained during these changes, any positive reactivity additions will be limited to acceptable levels.

Introduction of temperature changes must be evaluated to ensure they do not result in a loss of required SDM.

The boron addition capability 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 1720 ppm borated water from the refueling water tank to makeup for contraction of the primary coolant that could occur if the temperature is lowered from 200OF to 140 0F.

The restrictions associated with the establishing of the flow path from the RWT to the RCS via a single HPSI pump provide assurance that 10 CFR 50 Appendix G pressure/temperature limits will not be exceeded in the case of any inadvertent pressure transient due to a mass addition to the RCS. If RCS pressure boundary integrity does not exist as defined in Specification 1.16, these restrictions are not required. Additionally, a limit on the maximum number of operable HPSI pumps is not necessary when the pressurizer manway cover or the reactor vessel head is removed.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 7 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 ST. LUCIE UNIT 1 314.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 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.

The 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 immovable or untrippable CEA and to a large misalignment (L15 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 misalignments (< 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 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 (L 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 to those used in generating LCOs and LSSS setpoints, and (3) the ejected CEA worth used in the safety analysis. Therefore, the ACTION statement associated with the large misalignment of the CEA requires a prompt realignment of the misaligned CEA.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 8 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS ST. LUCIE UNIT 1 3/4.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 314.1.3 MOVABLE CONTROL ASSEMBLIES (continued)

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 1) 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 Ft , 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 CEA misalignment 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 Ft > the limits of Specification 3.2.3.

2. When the actual Ft < the limits of Specification 3.2.3, significant additional margin exists.

3. This additional margin can be credited to offset the increase in FO with time that can occur following a CEA misalignment.

4. This increase in Ft is caused by xenon redistribution.

5. The present analysis can support allowing a misalignment to exist without correction, if the time constraints and Initial Ft limits of COLR Figure 3-1-1a are met.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.1 BASES ATTACHMENT 3 OF ADM-25.04 9 of 9 REVISION NO.: REACTIVITY CONTROL SYSTEMS 1 ST. LUCIE UNIT 1 314.1 REACTIVITY CONTROL SYSTEMS (continued)

BASES (continued) 314.1.3 MOVABLE CONTROL ASSEMBLIES (continued)

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 independent 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 CEA positions and OPERABILITY of the CEA position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable. These verification frequencies are adequate for assuring that the applicable LCOs are satisfied.

The maximum CEA drop time permitted by Specification 3.1.3.4 is the assumed CEA drop time of 3.1 seconds used in the safety analyses.

Measurement with Tavg > 515 0F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

The LSSS setpoints and the power distribution LCOs were generated based upon a core burnup which would be achieved with the core operating in an essentially unrodded configuration. Therefore, the CEA insertion limit specifications require that during MODES 1 and 2, the full length CEAs be nearly fully withdrawn. The amount of CEA Insertion permitted by the Long Term Steady State Insertion Limits of Specification 3.1.3.6 will not have a significant effect upon the unrodded bumup assumption but will still provide sufficient reactivity control. The Power Dependent Insertion Limits of Specification 3.1.3.6 are provided to 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; however, long term operation at these insertion limits could have adverse effects on core power distribution during subsequent operation in an unrodded configuration.

ST. LU CIE UNIT I Sec~tio4nN.2° TECHNICAL SPECIFICATIONS Attachment No.

BASE ATTACHMENT 4 4 B S ATTACHMEN Current Revision No.

FP L OF ADM-25.04 0 Effective Date SAFETY RELATED 09106101

Title:

POWER DISTRIBUTION LIMITS Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - Bases for Technical Specifications. (E. Weinkam, 08130/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.2

.SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.2 BASES ATTACHMENT 4 OF ADM-25.04 2 of 5 REVISION NO.: POWER DISTRIBUTION LIMITS 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.2 ................... 3 3/4.2 POWER DISTRIBUTION LIMITS .3 BASES .. 3 3/4.2.1 LINEAR HEAT RATE .3 3/4.2.2 DELETED .3 3/4.2.3 and 3/4.2.4 TOTAL INTEGRATED RADIAL PEAKING FACTOR - FT AND AZIMUTHAL POWER TILT -Tq .44..............

3/4.2.5 DNB PARAMETERS .5

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.2 BASES ATTACHMENT 4 OF ADM-25.04 3 of 5 REVISION NO.: POWER DISTRIBUTION LIMITS 0 ST. LUCIE UNIT 1 BASES FOR SECTION 3/4.2 314.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 22000F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring System, provides adequate monitoring of the core power distribution and is capable of verifying that the linear heat rate does not exceed its limits.

The Excore Detector Monitoring System performs this function by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits specified in the COLR. In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made: 1) the CEA insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are satisfied, 2) the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL INTEGRATED RADIAL PEAKING FACTOR does not exceed the limits of Specification 3.2.3.

The Incore Detector Monitoring System continuously provides a direct measure of the peaking factors and the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits specified in the COLR. The setpoints for these alarms include allowances, set in conservative directions, for 1) a measurement-calculational uncertainty factor, 2) an engineering uncertainty factor,

3) a THERMAL POWER measurement uncertainty factor.

314.2.2 DELETED

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.2 BASES ATTACHMENT 4 OF ADM-25.04 4 of 5 REVISION NO.: POWER DISTRIBUTION LIMITS 0 ST. LUCIE UNIT 1 3/4.2 POWER DISTRIBUTION LIMITS (continued)

BASES (continued) 3/4.2.3 and 314.2.4 TOTAL INTEGRATED RADIAL PEAKING FACTOR. FT AND AZIMUTHAL POWER TILT - Tq The limitations on FT and Tq are provided to ensure that the assumptions used in the analysis for establishing the Linear Heat Rate and Local Power Density-High LCOs and LSSS setpoints and the DNB Margin LCO, and Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various allowable CEA group insertion limits. If FT or Tq exceed their basic limitations, operation may continue under the additional restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat Rate, Thermal Margin/Low Pressure and Local Power Density - High LCOs and LSSS setpoints remain valid.

An AZIMUTHAL POWER TILT > 0.10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The requirement that the measured value of (1+Tq) be multiplied by the calculated value of Fr to determine FT is applicable only when Fr is calculated with a non-full core power distribution analysis. With a full core power distribution analysis code the azimuthal tilt is explicitly accounted for as part of the radial power distribution used to calculate Fr.

The surveillance requirements for verifying that FT and Tq are within their limits provide assurance that the actual values of FT and Tq do not exceed the assumed values. Verifying FT after each fuel loading prior to exceeding 75% of RATED THERMAL POWER provides additional assurance that the core was properly loaded.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.2 BASES ATTACHMENT 4 OF ADM-25.04 5 of 5 REVISION NO.: POWER DISTRIBUTION LIMITS 0 ST. LUCIE UNIT 1 3/4.2 POWER DISTRIBUTION LIMITS (continued)

BASES (continued) 314.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR greater than or equal to the DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months basis.

ST. LU CIE UNIT 1 Section No.

3,4.3 TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 5 5 OF ADM-25.04 Current Revision No.

FPL OFAM2.40 Effective Date SAFETY RELATED 09/06J01

Title:

INSTRUMENTATION Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S I OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Secton 314.3

. _ SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.3 BASES ATTACHMENT 5 OF ADM-25.04 2 of 4 REVISION NO.: INSTRUMENTATION 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.3 .................... 3 3/4.3 INSTRUMENTATION .3 BASES .3 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION .3 3/4.3.3 MONITORING INSTRUMENTATION .4 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION .4 3/4.3.3.2 Deleted .4 3/4.3.3.3 Deleted .4 3/4.3.3.4 Deleted .4 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION .4

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.3 BASES ATTACHMENT 5 OF ADM-25.04 3 of 4 REVISION NO.: INSTRUMENTATION 0 ST. LUCIE UNIT 1 BASES FOR SECTION 3/4.3 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF)

INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and bypasses ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system functional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient con-ditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests per-formed at the minimum frequencies are sufficient to demonstrate this capability.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, over-lapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The Safety Injection Actuation Signal (SIAS) provides direct actuation of the Containment Isolation Signal (CIS) to ensure containment isolation in the event of a small break LOCA.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.3 BASES ATTACHMENT 5 OF ADM-25.04 4 of 4 REVISION NO.: INSTRUMENTATION 0 ST. LUCIE UNIT 1 3/4.3 INSTRUMENTATION (continued)

BASES (continued) 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that

1) the radiation levels are continually measured in the areas served by the individual channels; and (2)the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3)sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980 and NUREG-0737, "Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.3.2 Deleted 3/4.3.3.3 Deleted 3/4.3.3.4 Deleted 3/4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT SHUTDOWN of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

ST. LUCIE UNIT 1 Section No.

314.4 TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 6 6 FPLOF FP ADM-25.04 OF DM-2.04Effective Date Current Revision No.

SAFETY RELATED 12104/01

Title:

REACTOR COOLANT SYSTEM Responsible Department: Licensing REVISION

SUMMARY

Revision I - Changes made to reflect TS Amendment #179. (K.W. Frehafer, 11/30/01)

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S I OPS 0 08130/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 314.4 1 11/29/01 R.G. West 11/30/01 SYS Plant General Manager COM COMPLETED ITM 1

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 2 of 1(

REVISION NO.: REACTOR COOLANT SYSTEM I ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.4 .................. 3 3/4.4 REACTOR COOLANT SYSTEM . . 3 BASES .. 3 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION .3 3/4.4.2 DELETED .4 3/4.4.3 SAFETY VALVES .4 3/4.4.4 PRESSURIZER .5 3/4.4.5 STEAM GENERATORS .5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE . 7 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS .7 3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE .7 3/4.4.7 CHEMISTRY .8 3/4.4.8 SPECIFIC ACTIVITY .8 3/4.4.9 PRESSURE/TEMPERATURE LIMITS . 9 3/4.4.10 STRUCTURAL INTEGRITY .11 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS .. 12 3/4.4.11 DELETED .15 3/4.4.12 PORV BLOCK VALVES .15 3/4.4.13 POWER OPERATED RELIEF VALVES and . 15 3/4.4.14 REACTOR COOLANT PUMP - STARTING .15 3/4.4.15 REACTOR COOLANT SYSTEM VENTS .16

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 3 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 BASES FOR SECTION 314.4 314.4 REACTOR COOLANT SYSTEM BASES 314.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above the DNBR limit during all normal operations and anticipated transients. In MODES 1 and 2 with one reactor coolant loop not in operation, this specification requires that the plant be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In MODE 3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops (either shutdown cooling or RCS) be OPERABLE. In MODE 5 with reactor coolant loops not filled, a single shutdown cooling loop provides sufficient heat removal capability for removing decay heat; but single failure considerations and the unavailability of the steam generators as a heat removing component, require that at least two shutdown cooling loops be OPERABLE.

The operation of one Reactor Coolant Pump or one shutdown cooling pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reductions will, therefore, be within the capability of operator recognition and control.

If no coolant loops are in operation during shutdown operations, suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1.1 or 3.1.1.2 is required to assure continued safe operation.

Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum SDM or refueling boron concentration. This may result in an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operation. X

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 4 of 16 RtEVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION (continued)

The restrictions on starting a Reactor Coolant Pump are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 300F above each of the Reactor Coolant System cold leg temperatures.

3/4.4.2 DELETED 314.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 2 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any over-pressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient to limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operating at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operated relief valve or steam dump valves.

Surveillance Requirements are specified in the Inservice Testing Program.

Pressurizer code safety valves are to be tested in accordance with the requirements of Section Xl of the ASME Code, which provides the activities and the frequency necessary to satisfy the Surveillance Requirements. No additional requirements are specified.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 5 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.3 SAFETY VALVES (continued)

The pressurizer code safety valve as-found setpoint is 2500 psia +3/-2.5%

for OPERABILITY; however, the valves are reset to 2500 psia +/- 1%

during the Surveillance to allow for drift. The LCO is expressed in units of psig for consistency with implementing procedures.

314.4.4 PRESSURIZER A steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation. The steam bubble also protects the pressurizer code safety valves and power operated relief valve against water relief.

The power operated relief valve and steam bubble function to relieve RCS pressure during all design transients. Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer-Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The required pressurizer heater capacity is capable of maintaining natural circulation sub-cooling. Operability of the heaters, which are powered by a diesel generator bus, ensures ability to maintain pressure control even with loss of offsite power.

314.4.5 STEAM GENERATORS One OPERABLE steam generator provides sufficient heat removal capability to remove decay heat after a reactor shutdown. The requirement for two steam generators capable of removing decay heat, combined with the requirements of Specifications 3.7.1.1, 3.7.1.2 and 3.7.1.3 ensures adequate decay heat removal capabilities for RCS temperatures greater than 3250 F if one steam generator becomes inoperable due to single failure considerations. Below 3250F, decay heat is removed by the shutdown cooling system.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 6 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.5 STEAM GENERATORS (continued)

The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1.83, Revision 1.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or in-service conditions that lead to corrosion. Inservice inspection of Steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 gallon per minute, total). Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 1 gallon per minute can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

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

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 7 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. The LCO is consistent with NUREG-1432, Revision 1, and is satisfied when leakage detection monitors of diverse measurement means are OPERABLE in MODES 1,2, 3, and 4. Monitoring the reactor cavity sump inlet flow rate, in combination with monitoring the containment particulate or gaseous radioactivity, provides an acceptable minimum to assure that unidentified leakage is detected in time to allow actions to place the plant in a safe condition when such leakage indicates possible pressure boundary degradation.

314.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.

The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

The total steam generator tube leakage limit of 1 GPM for all steam generators ensures that the dosage contribution from the tube leakage will be limited to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line break. The 1 GPM limit is consistent with the assumptions used in the analysis of these accidents.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTDOWN.

The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 8 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1STE 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limit time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM and a concurrent loss of offsite electrical power. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the St. Lucie site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 giCi/gram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 9 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.8 SPECIFIC ACTIVITY (continued)

Reducing Tag to < 500OF prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take correction action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4A.9 PRESSUREITEMPERATURE LIMITS All components in the Reactor Coolant System are designed to withstand the effects of cyclic loads due to system temperature and pressure changes. These cyclic loads are introduced by normal load transients, reactor trips, and startup and shutdown operations. The various categories of load cycles used for design purposes are provided in Section 5.2.1 of the FSAR. During startup and shutdown, the rates of temperature and pressure changes are limited so that the maximum specified heatup and cooldown rates are consistent with the design assumptions and satisfy the stress limits for cyclic operation.

During heatup, the thermal gradients through the reactor vessel wall produce thermal stresses which are compressive at the reactor vessel inside surface and are tensile at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside surface and outside surface locations, the total applied stress is greatest at the outside surface location. However, since neutron irradiation damage is larger at the inside surface location than at the outside surface location, the inside surface flaw may be more limiting.

Consequently, for the heatup analysis, both the inside surface and outside surface flaw locations must be analyzed for the specific pressure and thermal loadings to determine which is more limiting.

During cooldown, the thermal gradients through the reactor vessel wall produce thermal stresses which are tensile at the reactor vessel inside surface and are compressive at the reactor vessel outside surface. Since reactor vessel internal pressure always produces tensile stresses at both the inside and outside surface locations, the total applied stress is greatest at the inside surface.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 10 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.9 PRESSURIZER/TEMPERATURE LIMITS (continued)

Since neutron irradiation damage is also greater at the inside surface, the inside surface flaw location is the limiting location during cooldown.

Consequently, only the inside surface flaw must be evaluated for the cooldown analysis.

The heatup and cooldown limit curves (Figures 3.4-2a and 3.4-2b) are composite curves which were prepared by determining the most conservative case, with either the inside or outside wall controlling, for the heatup rate of up to 500 F/hr and for any cooldown rate of up to 1000 F per hour. The heatup and cooldown curves were prepared based upon the most limiting value of the predicted adjusted reference temperature at the end of the applicable service period.

The reactor vessel materials have been tested to determine their initial RTNDT; the results of these tests are shown in Table B 3/4.4-1. Reactor operation and resultant fast neutron (E>1 Mev) irradiation will cause an increase in the RTNDT. Therefore, an adjusted reference temperature can be calculated based upon the fluence. The heatup and cooldown limit curves shown on Figures 3.4-2a and 3.4-2b include predicted adjustments for this shift in RTNDT at the end of the applicable service period, as well as adjustments for pressure differences between the reactor vessel beltline and pressurizer instrument taps.

The actual shift in RTNDT of the vessel material will be established periodically during operation by removing and evaluating, in accordance with ASTM E185-82, reactor vessel material surveillance specimens installed near the inside wall of the reactor vessel in the core area. The capsules are scheduled for removal at times that correspond to key accumulated fluence levels within the vessel through the end of life. Since the neutron spectra at the irradiation samples and vessel inside radius are essentially identical, measured ARTNDT for surveillance samples can be applied with confidence to the corresponding material in the reactor vessel wall. The heatup and cooldown curves must be recalculated when the ARTNDT determined from the surveillance capsule is different from the calculated ARTNDT for the equivalent capsule radiation exposure.

The pressure-temperature limit lines shown on Figures 3.4-2a and 3.4-2b for reactor criticality and for Inservice leak and hydrostatic testing have been provided to assure compliance with the minimum temperature requirements for Appendix G to 10 CFR 50.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 11 of 1 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.9 PRESSURIZER/TEMPERATURE LIMITS (continued)

The maximum RTNDT for all reactor coolant system pressure-retaining materials, with the exception of the reactor pressure vessel, has been established to be 900F. The Lowest Service Temperature limit line shown on Figures 3.4-2a and 3.4-2b is based upon this RTNDT since Article NB-2332 of Section III of the ASME Boiler and Pressure Vessel Code requires the Lowest Service Temperature to be RTNDT + 1000 F for piping, pumps and valves. Below this temperature, the system pressure must be limited to a maximum of 20% of the system's hydrostatic test pressure of 3125 psia.

The limitations imposed on the pressurizer heatup and cooldown rates and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection program for ASME Code Class 1, 2 and 3 components ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. This program is In accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a (g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code 1971 Edition and Addenda through Winter 1972.

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q SECTION NO.: MIT: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 15 of 1 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.11 DELETED 314A.12 PORV BLOCK VALVES The opening of the Power Operating Relief Valves fulfills no safety related function. The electronic controls of the PORVs must be maintained OPERABLE to ensure satisfaction of Specifications 4.5.1.d.1 and 4.5.2.d.1. Since it is impractical and undesirable to actually open the PORVs to demonstrate reclosing, it becomes necessary to verify operability of the PORV Block Valves to ensure the capability to isolate a malfunctioning PORV.

314.4.13 POWER OPERATED RELIEF VALVES and 3/4.4.14 REACTOR COOLANT PUMP - STARTING The low temperature overpressure protection system (LTOP) is designed to prevent RCS overpressurization above the 10 CFR 50 Appendix G operating limit curves (Figures 3.4-2a and 3.4-2b) at RCS temperatures at or below 3040 F during heatup and 281 OF during cooldown. The LTOP system is based on the use of the pressurizer power-operated relief valves (PORVs) and the implementation of administrative and operational controls.

The PORVs aligned to the RCS with the low pressure setpoints of 350 and 530 psia, restrictions on RCP starts, limitations on heatup and cooldown rates, and disabling of non-essential components provide assurance that Appendix G P/T limits will not be exceeded during normal operation or design basis overpressurization events due to mass or energy addition to the RCS. The LTOP system APPLICABILITY, ACTIONS, and SURVEILLANCE REQUIREMENTS are consistent with the resolution of Generic Issue 94, "Additional Low-Temperature Overpressure Protection for Light-Water Reactors," pursuant to Generic Letter 90-06.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 16 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.15 REACTOR COOLANT SYSTEM VENTS Reactor Coolant System vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function. The redundancy design of the Reactor Coolant System vent systems serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant System vent system are consistent with the requirements of Item ll.b.1 of NUREG-0737, TClarification of TMI Action Plan Requirements,' November 1980.

ST. LUCIE UNIT 1 Section No.

3,4.5 TECHNICAL SPECIFICATIONS Attachment No.

BASE ATTACHMENT 7 7.

BASOF ADTM-25.04 Current Revision No.

FPL 0 O REAT4 09Effective Date SAFETY RELATED 09106101

Title:

EMERGENCY CORE COOLING SYSTEMS (ECCS)

Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.5 SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.5 BASES ATTACHMENT 7 OF ADM-25.04 2 of 5 REVISION NO.: EMERGENCY CORE COOLING SYSTEMS (ECCS) 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.5 .................... 3 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) . 3 BASES .3 3/4.5.1 SAFETY INJECTION TANKS .3 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS .4 3/4.5.4 REFUELING WATER TANK (RWT) .5

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.5 BASES ATTACHMENT 7 OF ADM-25.04 3 of 5 REVISION NO.: EMERGENCY CORE COOLING SYSTEMS (ECCS) 0 ST. LUCIE UNIT 1 BASES FOR SECTION 314.5 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

BASES 3/4.5.1 SAFETY INJECTION TANKS The OPERABILITY of each of the RCS safety injection tanks ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the safety injection tanks. This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on safety injection tank volume, boron concentration and pressure ensure that the assumptions used for safety injection tank injection in the accident analysis are met.

The limit of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for operation with an SIT that is inoperable due to boron concentration not within limits, or due to the inability to verify liquid volume or cover-pressure, considers that the volume of the SIT is still available for injection in the event of a LOCA. If one SIT is inoperable for other reasons, the SIT may be unable to perform its safety function and, based on probability risk assessment, operation in this condition is limited to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.5 BASES ATTACHMENT 7 OF ADM-25.04 4 of 5 REVISION NO.: 0EMERGENCY CORE COOLING SYSTEMS (ECCS) 0 ST. LUCIE UNIT 1 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

BASES (continued) 314.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two separate and independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration. Either subsystem operating in conjunction with the safety injection tanks is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides long term core cooling capability in the recirculation mode during the accident recovery period.

TS 3.5.2, ACTION a.1. provides an allowed outage/action completion time (AOT) of up to 7 days from initial discovery of failure to meet the LCO provided the affected ECCS subsystem is inoperable only because its associated LPSI train is inoperable. This 7 day AOT is based on the findings of a deterministic and probabilistic safety analysis and is referred to as a risk-informed" AOT extension. Entry into this ACTION requires that a risk assessment be performed in accordance with the Configuration Risk Management Program (CRMP) which is described in the Administrative Procedure (ADM-17.08) that implements the Maintenance Rule pursuant to 10 CFR 50.65.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensure that at a minimum, the assumptions used in the accident analyses are met and that subsystem OPERABILITY is maintained.

The limitations on HPSI pump operability when the RCS temperature is

< 2700 F and < 2360F, and the associated Surveillance Requirements provide additional administrative assurance that the pressure/temperature limits (Figures 3.4-2a and 3.4-2b) will not be exceeded during a mass addition transient mitigated by a single PORV. A limit on the maximum number of operable HPSI pumps is not necessary when the pressurizer manway cover or the reactor vessel head is removed.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.5 BASES ATTACHMENT 7 OF ADM-25.04 5 of 5 REVISION NO.: EMERGENCY CORE COOLING SYSTEMS (ECCS) 0 ST. LUCIE UNIT 1 314.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) (continued)

BASES (continued) 3/4.5.4 REFUELING WATER TANK (RWT)

The OPERABILITY of the RWT as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on RWT minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWT and the RCS water volumes with all control rods inserted except for the most reactive control assembly. These assumptions are consistent with the LOCA analyses.

I-I Section No.

ST. LUCIE UNIT I 314.6 Attachment No.

TECHNICAL SPECIFICATIONS 8

BASES ATTACHMENT 8 Current Revision No.

OF ADM-25.04 FPL I Effective Date SAFETY RELATED I 06110/02

Title:

CONTAINMENT SYSTEMS Responsible Department: Licensing REVISION

SUMMARY

Revision I - Extended the allowed outage time for the containment vacuum relief lines from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for returning an inoperable containment vacuum relief line to operable status. (M. DiMarco, 06/07/02)

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08130/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Secton 314.6 1 06106/02 D. Rose 06/07/02 SYS Plant General Manager COM COMPLETED ITM 1

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 2 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.6 ..................... 3 3/4.6 CONTAINMENT SYSTEMS .3 BASES .3 3/4.6.1 CONTAINMENT VESSEL . . 3 3/4.6.1.1 CONTAINMENT VESSEL INTEGRITY .3 3/4.6.1.2 CONTAINMENT LEAKAGE . 3 3/4.6.1.3 CONTAINMENT AIR LOCKS . 3 3/4.6.1.4 INTERNAL PRESSURE .4 3/4.6.1.5 AIR TEMPERATURE .4 3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY .4 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS .5 3/4.6.2.1 CONTAINMENT SPRAY AND COOLING SYSTEMS .5 3/4.6.2.2 SPRAY ADDITIVE SYSTEM . 6 3/4.6.2.3 DELETED .6 3/4.6.3 CONTAINMENT ISOLATION VALVES . 6 3/4.6.4 COMBUSTIBLE GAS CONTROL .6 3/4.6.5 VACUUM RELIEF VALVES .7 3/4.6.6 SECONDARY CONTAINMENT .10 3/4.6.6.1 SHIELD BUILDING VENTILATION SYSTEM .10 3/4.6.6.2 SHIELD BUILDING INTEGRITY 10 3/4.6.6.3 SHIELD BUILDING STRUCTURAL INTEGRITY .10

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAW:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 3 of 1C REVISION NO.: CONTAINMENT SYSTEMS

_____ ST. LUCIE UNIT 1 BASES FOR SECTION 314.6 3/4.6 CONTAINMENT SYSTEMS BASES 3/4.6.1 CONTAINMENT VESSEL 3/4.6.1.1 CONTAINMENT VESSEL INTEGRITY CONTAINMENT VESSEL INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with the leakage rate limitation, will limit the site boundary radiation doses to within the limits of 10 CFR Part 100 during accident conditions.

In accordance with Generic Letter 91-08, "Removal of Component Lists from Technical Specifications," the opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3) assuring that environmental conditions will not preclude access to close the valves and this action will prevent the release of radioactivity outside the containment.

314.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that total containment leakage volume will not exceed the value assumed in the accident analyses at the peak accident pressure, Pa (39.6 psig) which results from the limiting design basis loss of coolant accident.

The surveillance testing for measuring leakage rates is performed in accordance with the Containment Leakage Rate Testing Program and is consistent with the requirements of Appendix "J"of 10 CFR 50, Option B and Regulatory Guide 1.163 Rev. 0, as modified by approved exemptions.

3/4.6.1.3 CONTAINMENT AIR LOCKS The limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment leak rate. Surveillance testing of the air lock seals provides assurance that the overall air lock leakage will not become excessive due to seal damage during the Intervals between air lock leakage tests.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 4 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 3/4.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 3/4.6.1 CONTAINMENT VESSEL (continued) 3/4.6.1.4 INTERNAL PRESSURE The limitations on containment internal pressure ensure that 1) the containment structural is prevented from exceeding its design negative pressure differential with respect to the annulus atmosphere of 0.70 psi and 2) the containment peak pressure does not exceed the design pressure of 44 psig during steam line break accident conditions.

The maximum peak pressure obtained from a steam line break accident is 41.6 psig. The limit of 2.4 psig for initial positive containment pressure will limit the total pressure to 44.0 psig which is the design pressure and is consistent with the accident analyses.

3/4.6.1.5 AIR TEMPERATURE The limitation on containment air temperature ensures that the containment vessel temperature does not exceed the design temperature of 2640F during LOCA conditions. The containment temperature limit is consistent with the accident analyses.

3/4.6.1.6 CONTAINMENT VESSEL STRUCTURAL INTEGRITY The limitation ensures that the structural integrity of the containment steel vessel will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the vessel will withstand the maximum pressure of 39.6 psig in the event of the limiting design basis loss of coolant accident. A visual inspection in accordance with the Containment Leakage Rate Testing Program is sufficient to demonstrate this capability.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 5 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 314.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 314.6.2.1 CONTAINMENT SPRAY AND COOLING SYSTEMS The OPERABILITY of the containment spray and cooling systems ensures that depressurization and cooling capability will be available to limit post-accident pressure and temperature in the containment to acceptable values. During a Design Basis Accident (DBA), at least two containment cooling trains or two containment spray trains, or one of each, is capable of maintaining the peak pressure and temperature within design limits. One containment spray train has the capability, in conjunction with the Spray Additive System, to remove iodine from the containment atmosphere and maintain concentrations below those assumed in the safety analyses. To ensure that these conditions can be met considering single-failure criteria, two spray trains and two cooling trains must be OPERABLE.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action interval specified in ACTION l.a and ACTION 1.d, and the 7 day action interval specified in ACTION 1.btake into account the redundant heat removal capability and the iodine removal capability of the remaining operable systems, and the low probability of a DBA occurring during this period. The 10 day constraint for ACTIONS l.a and 1.b is based on coincident entry into two ACTION conditions (specified in ACTION 1.c) coupled with the low probability of an accident occurring during this time. If the system(s) cannot be restored to OPERABLE status within the specified completion time, alternate actions are designed to bring the unit to a mode for which the LCO does not apply. The extended interval (54 hours6.25e-4 days 0.015 hours 8.928571e-5 weeks 2.0547e-5 months ) specified in ACTION 1.a to be in MODE 4 includes 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of additional time for restoration of the inoperable CS train, and takes into consideration the reduced driving force for a release of radioactive material from the RCS when in MODE 3. With two containment spray trains or any combination of three or more containment spray and containment cooling trains inoperable in MODES 1, 2, or Mode 3 with Pressurizer Pressure > 1750 psia, the unit is in a condition outside the accident analyses and LCO 3.0.3 must be entered immediately. In MODE 3 with Pressurizer Pressure < 1750 psia, containment spray is not required.

The specifications and bases for LCO 3.6.2.1 are consistent with NUREG-1432, Revision 0 (9/28/92), Specification 3.6.6A (Containment Spray and Cooling Systems; Credit taken for iodine removal by the Containment Spray System), and the plant safety analyses.

SECTION NO.: TLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 6 of 10 REVISION NO.: CONTAINMENT SYSTEMS ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS (continued) 314.6.2.2 SPRAY ADDITIVE SYSTEM The OPERABILITY of the spray additive system ensures that sufficient NaOH is added to the containment spray in the event of a LOCA. The limits on NaOH volume and concentration ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics. These assumptions are consistent with the iodine removal efficiency assumed in the accident analyses.

314.6.2.3 DELETED 3/4.6.3 CONTAINMENT ISOLATION VALVES The OPERABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. This includes the containment purge inlet and outlet valves.

3/4.6.4 COMBUSTIBLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammable limit during post-LOCA conditions. Either recombiner unit is capable of controlling the expected hydrogen generation associated with 1) zirconium-water reactions, 2) radiolytic decomposition of water and

3) corrosion of metals within containment.

The containment fan coolers are used in a secondary function to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 7 of 1C REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 .

314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 314.6.5 VACUUM RELIEF VALVES BACKGROUND: The vacuum relief valves protect the containment vessel against negative pressure (i.e., a lower pressure inside than outside).

Excessive negative pressure inside containment can occur if there is an inadvertent actuation of the containment cooling system or the containment spray system. Multiple equipment failures or human errors are necessary to have inadvertent actuation.

The containment pressure vessel contains two 100% vacuum relief lines installed in parallel that protect the containment from excessive external loading. The vacuum relief lines are 24-inch penetrations that connect the shield building annulus to the containment. Each vacuum relief line is isolated by a pneumatically operated butterfly valve in series with a check valve located on the containment side of the penetration.

A separate pressure controller that senses the differential pressure between the containment and the annulus actuates each butterfly valve.

Each butterfly valve is provided with an air accumulator that allows the valve to open following a loss of instrument air. The combined pressure drop at rated flow through either vacuum relief line will not exceed the containment pressure vessel design external pressure differential of 0.7 psid with any prevailing atmospheric pressure.

APPLICABLE SAFETY ANALYSES: Design of the vacuum relief lines involves calculating the effect of an inadvertent containment spray actuation that can reduce the atmospheric temperature (and hence pressure) inside containment. Conservative assumptions are used for all the pertinent parameters in the calculation. The resulting containment pressure versus time is calculated, including the effect of the vacuum relief valves opening when their negative pressure setpoint is reached.

It is also assumed that one vacuum relief line fails to open.

The containment was designed for an external pressure load equivalent to 0.7 psig. The inadvertent actuation of the containment spray system was analyzed to determine the resulting reduction in containment pressure.

This resulted in a differential pressure between the inside containment and the annulus of 0.66 psid, which is less than the design load.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 8 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 314.6.5 VACUUM RELIEF VALVES (continued)

The vacuum relief valves must also perform the containment isolation function in a containment high-pressure event. For this reason, the system is designed to take the full containment positive design pressure and the containment design basis accident (DBA) environmental conditions (temperature, pressure, humidity, radiation, chemical attack, etc.) associated with the containment DBA.

The vacuum relief valves satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO: The LCO establishes the minimum equipment required to accomplish the vacuum relief function following the inadvertent actuation of the containment spray system. Two vacuum relief lines are required to be OPERABLE to ensure that at least one is available, assuming one or both valves in the other line fail to open.

APPLICABILITY SAFETY ANALYSES: In MODES 1,2, and 3 with pressurizer pressure equal to or greater than 1750 psia, the containment cooling features, such as the containment spray system, are required to be OPERABLE to mitigate the effects of a DBA. Excessive negative pressure inside containment could occur whenever these systems are OPERABLE due to inadvertent actuation of these systems. In MODES 1, 2, 3, and 4, the containment internal pressure is maintained between specified limits. Therefore, the vacuum relief lines are required to be OPERABLE in MODES 1,2, 3, and 4 to mitigate the effects of inadvertent actuation of the containment spray system or containment cooling system.

In MODES 5 and 6, the probability and consequences of a DBA are reduced due to the pressure and temperature limitations of these MODES.

The containment spray system and containment cooling system are not required to be OPERABLE in MODES 5 and 6. Therefore, maintaining OPERABLE vacuum relief lines is not required in MODE 5 or 6.

SECTION NO.: TMLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 9 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 3/4.6.5 VACUUM RELIEF VALVES (continued)

ACTIONS: With one of the required vacuum relief lines inoperable, the inoperable line must be restored to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The specified time period is consistent with other LCOs for the loss of one train of a system required to mitigate the consequences of a LOCA or other DBA. If the vacuum relief line cannot be restored to OPERABLE status within the required ACTION time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowed ACTION times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE REQUIREMENTS: This SR references the Inservice Testing Program, which establishes the requirement that inservice testing of the ASME Code Class 1,2, and 3 pumps and valves shall be performed in accordance with Section Xl of the ASME Boiler and Pressure Vessel Code and applicable Addenda and approved relief requests.

Therefore, the Inservice Testing Program governs SR interval. The butterfly valve setpoint is 2.25+/-0.25 inches of water gauge differential.

The maximum butterfly valve stroke time is within 8 seconds when tested in accordance with the IST Program.

D

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.6 BASES ATTACHMENT 8 OF ADM-25.04 10 of 10 REVISION NO.: CONTAINMENT SYSTEMS 1 ST. LUCIE UNIT 1 314.6 CONTAINMENT SYSTEMS (continued)

BASES (continued) 314.6.6 SECONDARY CONTAINMENT 3/4.6.6.1 SHIELD BUILDING VENTILATION SYSTEM The OPERABILITY of the shield building ventilation systems ensures that containment vessel leakage occurring during LOCA conditions into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the site boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions.

3/4.6.6.2 SHIELD BUILDING INTEGRITY SHIELD BUILDING INTEGRITY ensures that the release of radioactive materials from the primary containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the accident analyses. This restriction, in conjunction with operation of the shield building ventilation system, will limit the site boundary radiation doses to within the limits of 10 CFR 100 during accident conditions.

3/4.6.6.3 SHIELD BUILDING STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment shield building will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to provide 1) protection for the steel vessel from the external missiles,

2) radiation shielding in the event of a LOCA, and 3) an annulus surrounding the steel vessel that can be maintained at a negative pressure within two minutes after a LOCA.

ST. LUCIE UNIT 1 SecItionNo.

TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 9 9 FPLOF ADM-25.04 Current Revision No.

Effective Date SAFETY RELATED 09106101

Title:

PLANT SYSTEMS Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - ases for Technica Specifcations. (E.Weink~am, 0830/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08/30/01 R.G. West 08/30101 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.7

.SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 2 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.7 ..................... 3 3/4.7 PLANT SYSTEMS .3 BASES .. 3 3/4.7.1 TURBINE CYCLE . .3 3/4.7.1.1 SAFETY VALVES .3 3/4.7.1.2 AUXILIARY FEEDWATER PUMPS .4 3/4.7.1.3 CONDENSATE STORAGE TANKS .4 3/4.7.1.4 ACTIVITY .5 3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES ... 5 3/4.7.1.6 SECONDARY WATER CHEMISTRY .5 3/4.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION .6 3/4.7.3 COMPONENT COOLING WATER SYSTEM . 6 3/4.7.4 INTAKE COOLING WATER SYSTEM . 6 3/4.7.5 ULTIMATE HEAT SINK .6 3/4.7.6 DELETED .7 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM .7 3/4.7.8 ECCS AREA VENTILATION SYSTEM . 7 3/4.7.9 SEALED SOURCE CONTAMINATION . 7 3/4.7.10 SNUBBERS ................ 8

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 3 of 9 REVISION NO.: PLANT SYSTEMS

° ST. LUCIE UNIT 1 BASES FOR SECTION 314.7 3/4.7 PLANT SYSTEMS BASES 314.7.1 TURBINE CYCLE 314.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1000 psia during the most severe anticipated system operational transient. The maximum relieving capacity is associated with a turbine trip from 100% RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no steam bypass to the condenser).

The specified valve lift settings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure Code, 1971 Edition and ASME Code for Pumps and Valves, Class II. The total relieving capacity for all valves on all of the steam lines isl2.38 x 106 lbs/hr which is 102.8 percent the total secondary steam flow of 12.04 x 106 lbs/hr at 100% RATED THERMAL POWER. A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for removing decay heat.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Level-High channels. The reactor trip setpoint reductions are derived on the following bases:

'For two loop operation SP_= (X)_(Y) x (106.5) where:

SP = reduced reactor trip setpoint in percent of RATED THERMAL POWER V = maximum number of inoperable safety valves per steam line

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 4 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 314.7 PLANT SYSTEMS (continued)

BASES (continued) 3/4.7.1 TURBINE CYCLE (continued) 106.5 = Power Level-High Trip Setpoint for two loop operation X = Total relieving capacity of all safety valves per steam line in lbs/hour (6.192 x 106 lbs/hr.)

Y = Maximum relieving capacity of any one safety valve in lbs/hour (7.74 x 106 lbs/hr.)

Surveillance Requirement 4.7.1.1 verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance with the Inservice Testing Program. The MSSV setpoints are 1000 psia

+1/-3% (4 valves each header) and 1040 psia +1/-3% (4 valves each header) for OPERABILITY; however, the valves are reset to 1000 psia

+/-1 % and 1040 psia +/- 1%, respectively, during the Surveillance to allow for drift. The LCO is expressed in units of psig for consistency with implementing procedures.

The provisions of Specification 3.0.4 do not apply. This allows entry into and operation in MODE 3 prior to performing the Surveillance Requirements so that the MSSVs may be tested under hot conditions.

3/4.7.1.2 AUXILIARY FEEDWATER PUMPS The OPERABILITY of the auxiliary feedwater pumps ensures that the Reactor Coolant System can be cooled down to less than 325 0F from normal operating conditions in the event of a total loss of off-site power.

Any two of the three auxiliary feedwater pumps have the required capacity to provide sufficient feedwater flow to remove reactor decay heat and reduce the RCS temperature to 3250 F where the shutdown cooling system may be placed into operation for continued cooldown.

314.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available for cooldown of the Reactor Coolant System to less than 3250 F in the event of a total loss of off-site power. The minimum water volume is sufficient to maintain the RCS at HOT STANDBY conditions for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with steam discharge to atmosphere.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 5 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 3/4.7 PLANT SYSTEMS (continued)

BASES (continued) 314.7.1 TURBINE CYCLE (continued) 314.7.1.4 ACTIVITY The limitations on secondary system specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. The dose calculations for an assumed steam line rupture include the effects of a coincident 1.0 GPM primary to secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power.

These values are consistent with the assumptions used in the accident analyses.

3/4.7.1.5 MAIN STEAM LINE ISOLATION VALVES The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses.

314.7.1.6 SECONDARY WATER CHEMISTRY This section left blank intentionally.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 6 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 314.7 PLANT SYSTEMS (continued)

BASES (continued) 314.7.2 STEAM GENERATOR PRESSURE/TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 700 F and 200-psig are based on a steam generator RTNDT of 500F and are sufficient to prevent brittle fracture.

314.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the component cooling water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident 3/4.7.4 INTAKE COOLING WATER SYSTEM The OPERABILITY of the intake cooling water system ensures that sufficient cooling capacity is available for continued operation of vital components and Engineered Safety Feature equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.5 ULTIMATE HEAT SINK The limitations on the ultimate heat sink level ensure that sufficient cooling capacity is available to either 1) provide normal cooldown of the facility, or

2) to mitigate the effects of accident conditions within acceptable limits.

The limitation on minimum water level is based on providing an adequate cooling water supply to safety related equipment until cooling water can be supplied from Big Mud Creek.

Cooling capacity calculations are based on an ultimate heat sink temperature of 950 F. It has been demonstrated by a temperature survey conducted from March 1976 to May 1981 that the Atlantic Ocean has never risen higher than 860F. Based on this conservatism, no ultimate heat sink temperature limitation is specified.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 7 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 .

3/4.7 PLANT SYSTEMS (continued)

BASES (continued) 3/4.7.6 DELETED 3/4.7.7 CONTROL ROOM EMERGENCY VENTILATION SYSTEM The OPERABILITY of the control room emergency ventilation system ensures that 1) the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system and 2) the control room will remain habitable for operations personnel during and following all credible accident conditions. 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 WAN, 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. Quantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Part 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.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 8 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 314.7 PLANT SYSTEMS (continued)

BASES (continued) 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 before that interval has elapsed may be used as a new reference point to determine 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 shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubber that may be generically susceptible and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are 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 are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

When a snubber is found inoperable, an evaluation is performed, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the inoperablility of the snubber. The engineering evaluation shall determine whether or not the snubber mode of failure has imparted a significant effect or degradation on 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 shutdowns at 18 month intervals. Observed failures of these sample snubbers shall require functional testing of additional units.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.7 BASES ATTACHMENT 9 OF ADM-25.04 9 of 9 REVISION NO.: PLANT SYSTEMS 0 ST. LUCIE UNIT 1 3/4.7 PLANT SYSTEMS (continued)

BASES (continued) 3/4.7.10 SNUBBERS (continued)

In cases where the cause of failure has been identified, additional snubbers 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% of 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 associated 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 Section No.

314.8 ITECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 10 10 Current Revision No.

FPL OF ADM-25.04 Effective Date SAFETY RELATED 12/18101

Title:

ELECTRICAL POWER SYSTEMS Responsible Department: Licensing REVISION

SUMMARY

Revision I - Implemented License Amendment 180. (K.W. Frehafer, 12/17/01)

Revision 0 - Bases for Technical Specifications. (E. Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08130/01 R.G. West 08/30101 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.8 1 12/17/01 R.G. West 12/17/01 SYS Plant General Manager COM COMPLETED ITM 1

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.8 BASES ATTACHMENT 10 OF ADM-25.04 2 of 5 REVISION NO.: ELECTRICAL POWER SYSTEMS 1 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.8 .................................... 3 3/4.8 ELECTRICAL POWER SYSTEMS .................................... 3 BASES ................................... 3

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.8 BASES ATTACHMENT 10 OF ADM-25.04 3 of 5 REVISION NO.: ELECTRICAL POWER SYSTEMS 1 ST. LUCIE UNIT I BASES FOR SECTION 314.8 3/4.8 ELECTRICAL POWER SYSTEMS BASES The OPERABILITY of A.C. and D.C. power sources and associated distribution systems during operation ensures that sufficient power will be available to supply the safety related equipment required for 1) the safe shutdown of the facility and 2) the mitigation and control of accident conditions within the facility. The minimum specified independent and redundant A.C. and D.C. power sources and distribution systems satisfy the requirements of General Design Criteria 17 of Appendix "A"to 10 CFR 50.

The ACTION requirements specified for the levels of degradation of the power sources provide restriction upon continued facility operation commensurate with the level of degradation. The OPERABILITY of the power sources are consistent with the initial condition assumptions of the accident analyses and are based upon maintaining at least one of each of the onsite A.C. and D.C. power sources and associated distribution systems OPERABLE during accident conditions coincident with an assumed loss of offsite power and single failure of the other onsite A.C. source. When one diesel generator is inoperable, there is an additional requirement to check that all required systems, subsystems, trains, components and devices (i.e.,

redundant features) that depend on the remaining OPERABLE diesel generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE. These redundant required features are those that are assumed to function to mitigate an accident, coincident with a loss of offsite power, in the safety analysis, such as the emergency core cooling system and auxiliary feedwater system. Upon discovery of a concurrent inoperability of required redundant features the feature supported by the inoperable EDG is declared inoperable. Thus plant operators will be directed to supported feature TS action requirements for appropriate remedial actions for the Inoperable required features.

The four hour completion time upon discovery that an opposite train required feature is inoperable is to provide assurance that a loss of offsite power, during the period that a EDG is inoperable, does not result in a complete loss of safety function of critical redundant required features. The four hour completion time allows the operator time to evaluate and repair any discovered inoperabilities. This completion time also allows for an exception to the normal utime zero" for beginning the allowed outage time "clock." The four hour completion time only begins on discovery that both an inoperable EDG exists and a required feature on the other train is X

inoperable.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.8 BASES ATTACHMENT 10 OF ADM-25.04 4 of 5 REVISION NO.: ELECTRICAL POWER SYSTEMS 1 ST. LUCIE UNIT 1 3/4.8 ELECTRICAL POWER SYSTEMS (continued)

BASES (continued)

TS 3.8.1.1, ACTION Mb" provides an allowed outage/action completion time (AOT) of up to 14 days toto restore a single inoperable diesel generator to operable status. This AOT is based on the findings of a deterministic and probabilistic safety analysis and is referred to as a "risk-informed" AOT.

Entry into this action requires that a risk assessment be performed in accordance with the Configuration Risk Management Program (CRMP),

which is described in the Administrative Procedure that implements the Maintenance Rule pursuant to 10 CFR 50.65.

All EDG inoperabilities must be investigated for common-cause failures regardless of how long the EDG inoperability persists. When one diesel generator is inoperable, required ACTIONS 3.8.1.1.b and 3.8.1.1.c provide an allowance to avoid unnecessary testing of EDGs. If it can be determined that the cause of the inoperable EDG does not exist on the remaining OPERABLE EDG, then SR 4.8.1.1.2.a.4 does not have to be performed.

Eight (8) hours is reasonable to confirm that the OPERABLE EDG is not affected by the same problem as the inoperable EDG. If it cannot otherwise be determined that the cause of the initial inoperable EDG does not exist on the remaining EDG, then satisfactory performance of SR 4.8.1.1 .2.a.4 suffices to provide assurance of continued OPERABILITY of that EDG.

If the cause of the initial inoperability exists on the remaining OPERABLE EDG, that EDG would also be declared inoperable upon discovery, and ACTION 3.8.1 .1.e would be entered. Once the failure is repaired (on either EDG), the common-cause failure no longer exists.

Ambient conditions are the normal standby conditions for the diesel engines. Any normally running warmup systems should be in service and operating, and manufacturer's recommendations for engine oil and water temperatures and other parameters should be followed.

The OPERABILITY of the minimum specified A.C. and D.C. power sources and associated distribution systems during shutdown and refueling ensures that 1) the facility can be maintained in the shutdown or refueling condition for extended time periods and 2) sufficient instrumentation and control capability is available for monitoring and maintaining the facility status.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.8 BASES ATTACHMENT 10 OF ADM-25.04 5 of 5 REVISION NO.: ELECTRICAL POWER SYSTEMS 1 ST. LUCIE UNIT 1 314.8 ELECTRICAL POWER SYSTEMS (continued)

BASES (continued)

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.108, "Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants," Revision 1, August 1977, Regulatory Guide 1.137, uFuel Oil Systems for Standby Diesel Generators," Revision 1, October 1979, Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," dated July 2, 1984, and NRC staff positions reflected in Amendment No. 48 to Facility Operating License NPF-7 for North Anna Unit 2, dated April 25, 1985; as modified by Generic Letter 93-05, "Line-Item Technical Specifications Improvements to Reduce Surveillance Requirements for Testing During Power Operation," dated September 27, 1993, and Generic Letter 94-01, "Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," dated May 31, 1994.

ST. LUCIE UNIT 1 Section No.

314.9 TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 11 11 OF ADM-25.04 Current Revision No.

F PL 2 Date Effective SAFETY RELATED 09/23102

Title:

REFUELING OPERATIONS Responsible Department: Licensing REVISION

SUMMARY

Revision 2 - Changes made to reflect TS Amendment #184. (M. DiMarco, 09/20/02)

Revision I - Changes made to reflect TS Amendment #179. (K. W. Frehafer, 11/30/01)

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S I OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.9 2 09/19/02 D. Rose 09/20/02 SYS Plant General Manager COM COMPLETED ITM 2

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 2 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.9 .................... 3 3/4.9 REFUELING OPERATIONS .. 3 BASES .. 3 3/4.9.1 BORON CONCENTRATION .3 3/4.9.2 INSTRUMENTATION .3 3/4.9.3 DECAY TIME .3 3/4.9.4 CONTAINMENT PENETRATIONS .4 3/4.9.5 COMMUNICATIONS .6 3/4.9.6 MANIPULATOR CRANE OPERABILITY . 6 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING .6 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION .6 3/4.9.9 CONTAINMENT ISOLATION SYSTEM . 7 3/4.9.4.10 and 3/4.9.11 WATER LEVEL-REACTOR VESSEL AND STORAGE POOL WATER LEVEL . 7 3/4.9.12 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE.......................................................................7 3/4.9.13 SPENT FUEL CASK CRANE ................................8 3/4.9.14 DECAY TIME - STORAGE POOL ................................ 8

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 3 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 BASES FOR SECTION 314.9 314.9 REFUELING OPERATIONS BASES 314.9.1 BORON CONCENTRATION The limitation on minimum boron concentration ensures that 1) the reactor will remain subcritical during CORE ALTERATIONS, and 2) a uniform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. The limitation on Kff is sufficient to prevent reactor criticality with all full length rods (shutdown and regulating) fully withdrawn.

If the boron concentration of any coolant volume in the RCS, the refueling canal, or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately. Operations that individually add limited positive reactivity (e.g., temperature fluctuations from inventory addition or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g., intentional boration) result in overall net negative reactivity addition, are not precluded by this action. Suspension of CORE ALTERATIONS or positive reactivity additions shall not preclude moving a component to a safe position.

314.9.2 INSTRUMENTATION The OPERABILITY of the wide range logarithmic range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.

314.9.3 DECAY TIME The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor pressure vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 4 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 314.9 REFUELING OPERATIONS (continued)

BASES (continued) 314.9.4 CONTAINMENT PENETRATIONS The requirements on containment penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment. The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a recently irradiated fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE. The fuel handling accident analysis assumes a minimum post reactor shutdown decay time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, recently irradiated fuel is defined as fuel that has occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This represents the applicability bases for fuel handling accidents.

Containment closure will have administrative controls in place to assure that a single normal or contingency method to promptly close the primary or secondary containment penetrations will be available. These prompt methods need not completely block the penetrations nor be capable of resisting pressure, but are to enable the ventilation systems to draw the release from the postulated fuel handling accident in the proper direction such that it can be treated and monitored.

In accordance with Generic Letter 91-08, Removal of Component Lists from the Technical Specifications, the opening of locked or sealed closed containment isolation valves on an intermittent basis under administrative control includes the following considerations: (1) stationing an operator, who is in constant communication with the control room, at the valve controls, (2) instructing this operator to close these valves in an accident situation, and (3)assuring that environmental conditions will not preclude access to close the valves and that this action will prevent the release of radioactivity outside the containment.

FPL made the following regulatory commitment, which is consistent with NUMARC 93-01, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Revision 3, Section 11.3.6, Assessment Methods for Shutdown Conditions, subheading 11.3.6.5, Containment - Primary (PWR)ISecondary (BWR).

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 5 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 314.9 REFUELING OPERATIONS (continued)

BASES (continued) 314.9.4 CONTAINMENT PENETRATIONS (continued)

The following guidelines are included in the assessment of systems removed from service during movement of irradiated fuel:

During fuel handling/core alterations, ventilation system and radiation monitor availability(as defined in NUMARC 91-06) should be assessed, with respect to filtration and monitoring of releases from the fuel. Following shutdown, radioactivity in the fuel decays away fairly rapidly. The basis of the Technical Specification operability amendment is the reduction in doses due to such decay. The goal of maintaining ventilation system and radiation monitor availability is to reduce doses even further below that provided by the natural decay and to avoid unmonitored releases.

A single normal or contingency method to promptly close primary or secondary containment penetrations should be developed. Such prompt methods need not completely block the penetration or be capable of resisting pressure. The purpose of the "prompt methods" mentioned above are to enable ventilation systems to draw the release from a postulated fuel handling accident in the proper direction such that it can be treated and monitored.

Availability as defined by NUMARC 91-06, Guidelines for industry Actions to Assess Shutdown Management, December 1991, relies on the definitions of functional, and operable. The NUMARC 91-06 definitions for these three terms follow.

Available (Availability): The status of a system, structure, or component that is in service or can be placed in service in a functional or operable state by immediate manual or automatic actuation.

Functional (Functionality): The ability of a system, structure, or component to perform its intended service with considerations that applicable technical specification requirements or licensing/design basis assumptions may not be maintained.

Operable: The ability of a system to perform its specified function with all applicable TS requirements satisfied.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 6 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 314.9 REFUELING OPERATIONS (continued)

BASES (continued) 314.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during CORE ALTERATIONS.

3/4.9.6 MANIPULATOR CRANE OPERABILITY The OPERABILITY requirements of the cranes used for movement of fuel assemblies ensures that: 1) each crane has sufficient load capacity to lift a fuel element, and 2) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

314.9.7 CRANE TRAVEL - SPENT FUEL STORAGE BUILDING The restriction on movement of loads in excess of the nominal weight of a fuel assembly and CEA over irradiated fuel assemblies ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is consistent with the activity release assumed in the accident analyses.

314.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION The requirement that at least one shutdown cooling loop be in operation ensures that 1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor pressure vessel below 140OF as required during the REFUELING MODE, and 2) sufficient coolant circulation is maintained through the reactor core to minimize the effects of a boron dilution incident and prevent boron stratification.

If SDC loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Suspending positive reactivity additions that could result in failure to meet the minimum boron concentration limit is required to assure continued safe operation.

Introduction of coolant inventory must be from sources that have a boron concentration greater than what would be required in the RCS for minimum refueling boron concentration. This may result In an overall reduction in RCS boron concentration, but provides acceptable margin to maintaining subcritical operations.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 7 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 314.9 REFUELING OPERATIONS (continued)

BASES (continued) 314.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION (continued)

The requirement to have two shutdown cooling loops OPERABLE when there is less than 23 feet of water above the irradiated fuel in the core ensures that a single failure of the operating shutdown cooling loop will not result in a complete loss of decay heat removal capability. With the reactor vessel head removed and 23 feet of water above the irradiated fuel in the core, a large heat sink is available for core cooling, thus in the event of a failure of the operating shutdown cooling loop, adequate time is provided to initiate emergency procedures to cool the core.

314.9.9 CONTAINMENT ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment isolation valves will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material resulting from a fuel handling accident of a recently irradiated fuel assembly from the containment atmosphere to the environment.

Recently irradiated fuel is defined as fuel that has occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

3/4.9.4.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

314.9.12 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE The limitations on the fuel handling building ventilation system ensures that all radioactive material released from a recently 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 assumptions of the fuel handling accident analyses.

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.9 BASES ATTACHMENT 11 OF ADM-25.04 8 of 8 REVISION NO.: REFUELING OPERATIONS 2 ST. LUCIE UNIT 1 314.9 REFUELING OPERATIONS (continued)

BASES (continued) 3/4.9.12 FUEL POOL VENTILATION SYSTEM - FUEL STORAGE (continued)

The fuel handling accident analysis assumes a minimum post reactor shutdown decay time of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Therefore, recently irradiated fuel is defined as fuel that has occupied part of a critical reactor core within the previous 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . This represents the applicability bases for fuel handling accidents. Containment closure will have administrative controls in place to assure that a single normal or contingency method to promptly close the primary or secondary containment penetrations will be available. These prompt methods need not completely block the penetrations nor be capable of resisting pressure, but are to enable the ventilation systems to draw the release from the postulated fuel handling accident in the proper direction such that it can be treated and monitored.

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 approximately 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 leakage from the spent fuel pool in excess of the maximum makeup capability.

314.9.14 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. L UCIE ST. UCI UNIT UNT I1314.10 Section No.

TECHNICAL SPECIFICATIONS Attachment No.

BASES ATTACHMENT 12 12 OF Current Revision No.

FPL 00 Effective Date SAFETY RELATED 09106/01

Title:

SPECIAL TEST EXCEPTIONS Responsible Department: Licensing REVISION

SUMMARY

Revision 0 - Bases for Technical Specifications. (E.Weinkam, 08/30/01)

Revision FRG Review Date Approved By Approval Date S 1 OPS 0 08/30/01 R.G. West 08/30/01 DATE Plant General Manager DOCT PROCEDURE Revision FRG Review Date Approved By Approval Date DOCN Section 3/4.10 SYS Plant General Manager COM COMPLETED ITM 0

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.10 BASES ATTACHMENT 12 OF ADM-25.04 2 of 3 REVISION NO.: SPECIAL TEST EXCEPTIONS 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.10 .................................................... 3 3/4.10 SPECIAL TEST EXCEPTIONS .................................................... 3 BASES ................................................... 3 3/4.10.1 SHUTDOWN MARGIN .................................................... 3 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS .................................................. 3 3/4.10.3 This specification deleted ................................................ 3 3/4.10.4 This specification deleted ................................................ 3 3/4.10.5 CENTER CEA MISALIGNMENT ..................................... 3

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.10 BASES ATTACHMENT 12 OF ADM-25.04 3 of 3 REVISION NO.: SPECIAL TEST EXCEPTIONS 0 ST. LUCIE UNIT 1 BASES FOR SECTION 3/4.10 314.10 SPECIAL TEST EXCEPTIONS BASES 3/4.10.1 SHUTDOWN MARGIN This special test exception provides that a minimum amount of CEA worth is immediately available for reactivity control when tests are performed for CEAs worth measurement. This special test exception is required to permit the periodic verification of the actual versus predicted core reactivity condition occurring as a result of fuel burnup or fuel cycling operations.

314.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS This special test exception permits individual CEAs to be positioned outside of their normal group heights and insertion limits during the performance of such PHYSICS TESTS as those required to 1) measure CEA worth and 2) determine the reactor stability index and damping factor under xenon oscillation conditions.

3/4.110.3 This specification deleted 314.10.4 This specification deleted 314.10.5 CENTER CEA MISALIGNMENT This special test exception permits the center CEA to be misaligned during PHYSICS TESTS required to determine the isothermal temperature coefficient and power coefficient.

ST. LUCIE UNIT I TECHNICAL SPECIFICATIONS BASES ATTACHMENT 13 OF ADM-25.04 FPL SAFETY RELATED RADIOACTIVE EFFLUENTS Revision 0 - Bases for Technical Specifications. (E. Weinkam, 08/30/01)

DATE DOCT PROCEDURE DOCN Section 3/4.11 SYS COM COMPLETED

SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.11 BASES ATTACHMENT 13 OF ADM-25.04 2 of 3 REVISION NO.: RADIOACTIVE EFFLUENTS 0 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.11 .......................................... 3 3/4.11 RADIOACTIVE EFFLUENTS ................... ....................... 3 BASES .......................................... 3 3/4.11.2.5 EXPLOSIVE GAS MIXTURE ..................................... 3 3/4.11.2.6 GAS STORAGE TANKS .3

SECTION NO.: TrTLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.11 BASES ATTACHMENT 13 OF ADM-25.04 3 of 3 REVISION NO.: RADIOACTIVE EFFLUENTS 0 ST. LUCIE UNIT 1 BASES FOR SECTION 3/4.11 314.11 RADIOACTIVE EFFLUENTS BASES Pages B 3/4 11-2 through B 3/4 11-3 (Amendment No. 123) have been deleted from the Technical Specifications. The next page is B 3/4 11-4.

314.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas holdup system is maintained below the flammability limits of hydrogen and oxygen. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be con-trolled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

314.11.2.6 GAS STORAGE TANKS Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 15.7.1, "Waste Gas System Failure."

L-2003-083, Technical Specification Bases Control Program Periodic Report of Bases Changes TS 6.8.4.i.4

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