Text

Proposed License Amendment Steam Generator Tube Integrity
	ML061170619
Person / Time
Site:	Saint Lucie
Issue date:	04/24/2006
From:	Johnston G; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2006-089
	Download: ML061170619 (66)
	v • d • e

0 Ocean Drive, Jensen Beach, FL '14957 Florida Power & Light Company, 6501 S.

April 24, 2006 FPI..

L-2006-0F9 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attr: Document Control Desk Washington, DC 20555 RE: St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Steam Generator Tube Integrity Purmuant to 10 CFR 50.90, Florida Power and Light Company (FPL) requests to amend Facility Operating License DPR-67 for St. Lucie Unit 1.

The proposed amendment would revise the Technical Specification (TS) requirements related to steam generator tube integrity. The change is based on NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF -

449, "Steam Generator Tube Integrity." The availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP).

Attachment 1 provides a description of the proposed change and confirmation of applicability.

Attachment 2 provides the existing TS pages marked-up to show the proposed changes.

Attachment 3 provides the word-processed TS pages. Attachment 4 provides an information only markup of the TS Bases.

FPL requests that the proposed amendment be processed by January 2007 to support SL1-21 April 2007 refueling outage inspections and that the amendment be effective on the date of issuance with implementation within 90 days.

The license amendment proposed by FPL has been reviewed by the St. Lucie Plant Facility Review Group and the FPL Company Nuclear Review Board. In accordance with 10 CFR 50.91 (b)(1), a copy of this proposed license amendment is being forwarded to the State Designee for the State of Florida.

7qdO an FPL Group company

St. Lucie Unit I L-2006-089 Docket No. 50-335 Page 2 Proposed License Amendment Steam Generator Tube Integrity Please contact Ken Frehafer at 772-467-7748 if there are any questions about this submittal.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on + 2. C...

Very truly yours, Gordon L. Johnston Acting Vice President St. Lucie Plant GLJIKWF Attachments cc: Mr. William A. Passetti, Florida Department of Health

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 1 of 8 Steam Generator Tube Integritv Evaluation of Proposed Change

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 1 Proposed License Amendment Page 2 of 8 Steam Generator Tube Integritv INTRODUCTION Pursuant to 10 CFR 50.90, Florida Power and Light Company (FPL) requests to amend Facility Operating License DPR-67 for St. Lucie Unit 1. This proposed license amendment recuest (LAR) revises the requirements in the St. Lucie Unit I Technical Specification (TS) rel ated to steam generator tube integrity and Reactor Coolant System Operational Leakage.

Th. change is based on the NRC approved Revision 4 to industry/Technical Specification Ta3k Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity." The availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP).

DESCRIPTION OF PROPOSED AMENDMENT Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed changes:

Revise the TS definitions of Identified Leakage (1.15c) and Pressure Boundary Leakage (1.22).

Replace the existing TS 3/4.4.5, "Steam Generators," requirements with the new "Stearr.

Generator Tube Integrity" requirements.

Revise TS 3/4.4.6, "Reactor Coolant System Leakage."

Add new TS 6.8.4 1.(lima), "Steam Generator Program."

Add new TS 6.9.1.12, "Steam Generator Tube Inspection Report."

Proposed revisions to the TS Bases are also included with this LAR. As discussed in the NR C's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Rev. 4 is an integral part of implementing this TS improvement. Departure from the wording prcposed in the TS Bases associated with TSTF-449, Rev. 4 is taken only when necessary to maintain consistency with the St. Lucie Unit I licensing basis. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

BACKGROUND The background for this LAR is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

Thc table below provides a summary of the proposed changes. It also identifies the Improved Standard Technical Specifications (ITS) sections based on TSTF-449, Rev. 4 and the corresponding sections in the St. Lucie Unit 1 TS.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 3 of 8 Steam Generator Tube Intearity TSTF-449:

TSTF-4* Condition or_- Current Lice nsing Basis :St. Lucie TS Location'.

1TS Section:..

requirement -_ . .==--& Proposed Change 3.4.13d Operational primary-to- <1 GPM total through all SGs and <500 3.4.6.2c RCS Operational Leakage TS < 150 gallons per day secondary leakage gallons per day through any one SG through any one SG (room temperature).

(accident conditions).

3.4.13 RCS primary-to- Reduce leakage rate to within limits 3.4.6.2 RCS Operational Leakage, sccondary leakage within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT ACTION a. - be in at least HOT STANDBY within through any one SG not STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within within limits in COLD SHUTDOWN within the the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3.4.13.1 RCS leakage RCS leakage is determined by water 4.4.6.2c Relocate extemporaneous information to footnote determined by water inventory balance. 3.4.6.1, and revise to state: "Not required to be performed inventory balance ACTION a. until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state and b. operation. Not applicable to primary to secondary leakage." Add conforming changes to other affected specifications.

3.4.13.2 SG Tube integrity Sample and analysis program requires 4.4.6.2 Add new RCS Operational Leakage TS 4.4.6.2.g to verification Gross Radioactivity Determination verify primary-to-secondary leakage within LCO every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . limit at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Add Note stating "Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation."

3.4.13, ACTIONS Performance Criteria not defined. 3.4.6.2, RCS Operational Leakage TS and SG Tube Integrity 3.4.20 Primary to secondary leakage limit and TS - Contains primary-to-secondary leakage limit.

actions included in the Tech Specs. 3/4.4.6 SG tube integrity requirements and ACTIONS g or r required upon failure to meet performance criteria.

criteria. 3/4.4.5 Plug or repair tubes satisfying repair criteria.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 4 of 8 Steam Generator Tube Integrity TSTF-449 ITS Secton Conditionor :'; : :Current Licensing Basis

_ S. Luicie TS Location Requirement & Proposed Change 3.4.13d Performance criteria Operational leakage < I gpm total or 3.4.6.2c RCS Operational leakage TS - Operational leakage

< 500 gallons per day through any one < 150 gallons per day through any one SG (room SG (accident conditions). temperature).

3.4.20 No criteria specified for structural 3/4.4.5 SG Tube Integrity TS 3/4.4.5 - Requires that tube integrity or accident induced leakage. integrity be maintained.

5.5.9 6.8.4 1 TS 6.8.4 1- Defines structural integrity and accident induced leakage performance criteria, which are dependent on design basis limits. Provides provisions for condition monitoring assessment to verify compliance.

5.5.9 Frequency of 6 to 40 months depending on SG 6.8.4 1 SG Tube Integrity TS - Requires Surveillance verification of tube category defined by previous inspection Frequency in accordance with TS 6.8.4 1,Steam integrity results. Generator Program. Frequency is dependent on tubing material and the previous inspection results and the anticipated defect growth rate.

Steam Generator Program - Establishes maximum inspection intervals.

5.5.9 Tube sample selection Based on SG Category, industry 6.8.4 1 Steam Generator Program and implementing experience, random selection, existing procedures - Dependent on a pre-outage evaluation indications, and results of the initial of actual degradation locations and mechanisms, and sample set -3% times the number of operating experience - 20% of all tubes as a SGs at the plant as a minimum. minimum.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 5 of 8 Steam Generator Tube Integrity TSTF-449 A _--

TSt i--- Condition or ITS Section-Current Licensing Basis St. Lucie TS Location.

-Requirement---- ........

....-. & Proposed Change 5.5.9 Inspection techniques Not specified 6.8.4 1 SG Tube Integrity TS - SR 4.4.5.1 requires that tube integrity be verified in accordance with the Steam Generator Program.

TS 6.8.4 1 Steam Generator Program and implementing procedures - Establishes requirements for qualifying NDE techniques. Requires use of qualified techniques in SG inspections. Requires a pre-outage evaluation of potential tube degradation morphologies and locations and an identification of NDE techniques capable of finding the degradation.

5.5.9 Inspection scope From the point of entry (hot leg side) 6.8.4 1 TS 6.8.4 1Steam Generator Program procedures completely around the U-bend to the - Inspection scope is defined by the degradation top support of the cold leg, or from the assessment that considers existing and potential point of entry (cold leg side) completely degradation morphologies and locations. Explicitly round the U-bend and to the bottom of requires consideration of entire length of tube from the hot leg. tube-sheet weld to tubesheet weld.

5.5.9 Repair criteria Plug tubes with imperfections extending 6.8.4 1 TS 6.8.4 1- Criteria unchanged.

_>40% through wall.

5.5.9 Repair methods Methods (except plugging) require prior 6.8.4 1 TS 6.8.4 1-Requirements unchanged.

approval by the NRC. Approved methods listed in TS.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment I Proposed License Amendment Page 6 of 8 Steam Generator Tube Integrity TSTF-449-:i: --i- : ::

lTSTF-49 :Condition-or -

ITS Section

----_Current Licensing Basis- St. Lucie TS Locationm

-7 Reu ment- 77- &PPr sed'Change' 5.6.9 Reporting requirements Plugging and repair report required 15 6.9.1.12 CFR - Serious SG tube degradation (i.e., tubing fails days after each inservice inspection, 12 to meet the structural integrity or accident induced month report documenting inspection leakage criteria) requires reporting in accordance results, and reports in accordance with with 50.72 or 50.73.

§50.72 when the inspection results fall TS 6.9.1.12 - 180 days after the initial entry into into category C-3. MODE 4 after performing a SG inspection Definitions Definitions SG Normnal TS definitions (i.e., Definitions Definitions TS 6.8.4 1,TS Bases, Steam Generator Program Tcrminology Section) did not address SG Program procedures - Includes Steam Generator Program issues. The Definitions Section uses the terminology applicable only to SGs. TS Definitions term "SG leakage." 1.15c and 1.22 are revised to use the term "primary-to-secondary leakage."

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 7 of 8 Steam Generator Tube Integrity REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this LAR are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

TECHNICAL ANALYSIS FPL has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298) as pait of the CLIIP Notice for Comment. This included the NRC staffs SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449. FPL has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to St. Lucie Unit 1 and justify this amendment for the incorporation of the changes to the St. Lucie TS considering the differences described in this application. These differences from the TS changes described in TSTF-449, Revision 4 are necessary due to the non-standard format of the St. Lucie Unit 1 TS.

REGULATORY ANALYSIS A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

SuloDorting Information The following information is provided to support the NRC staff's review of this LAR:

Plant Name, Unit No. St. Lucie Unit 1 Steam Generator Model(s): B&W Canada Replacement for CE Series-67 Approximate Effective Full Power Years REFPY) of service for currently installed 7.3 EFPY as of the October 2005 refueling SGs outage.

Tubing Material Alloy 690 Thermally Treated Number of tubes per SG 8523 Number and percentage of tubes plugged in IA 14 (0.17%)

each SG as of 7/2005 1B 0 (0.0%)

Number of tubes repaired in each SG None

'Degradation mechanism(s) identified Mechanical Wear Current primary-to-secondary leakage Per SG: Not specified.

l imits: Total: 1 gallon per minute (operating temp.)

St. Lucie Unit I L-2006-089 Docket No. 50-335 Attachment 1 Proposed License Amendment Page 8 of 8 Steam Generator Tube Integrity Approved Alternate Tube Repair Criteria: None Approved SG Tube Repair Methods None =

Performance criteria for accident leakage 1 gpm total (operating temp.)

NO SIGNIFICANT HAZARDS CONSIDERATION FPL has reviewed the no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FPL has concluded that the proposed determination presented in the notice is applicable to St. Lucie Unit 1 considering the differences described in this application. Therefore, the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.9 1(a).

ENVIRONMENTAL EVALUATION FPL has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. FPL has concluded that the NRC staff's findings presented in that evaluation are applicable to St. Lucie Unit 1 and the evaluation is hereby incorporated by reference for this application.

PRECEDENT This application is being made in accordance with the CLIIP. FPL is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2, 2005 (70 FR 10298). The following differences from the TS changes described in TSTF-449, Revision 4 are necessary due to the non-standard format of the St. Lucie Unit 1 TS:

1. The current format and terminology used in the St. Lucie Unit 1TS is retained to maintain consistency with the current specifications. For example:

The general format and numbering convention associated with the current TS for Limiting Conditions for Operation (LCOs), Actions, Surveillance Requirements (SRs) and Notes are retained.

Terminology used in the current TS Actions is maintained. For example, HOT STANDBY, HOT SHUTDOWN and COLD SHUTDOWN are used in lieu of MODE 3, MODE 4 and MODE 5, respectively.

2. Necessary conforming changes regarding the proper timing and conditions for performing the RCS water inventory balance were made to Specifications 3.4.6.1 ACTIONS a. and b

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 1 of 21 St,-am Generator Tube Integrity Technical Specification Markups TS Page V TS Page XV TS Page 1-4 TS Page 1-5 TS Page 3/4 4-5 TS Page 3/4 4-6 TS Page 3/4 4-7 TS Page 3/4 4-8 TS Page 3/4 4- 10 TS Page 3/4 4-11 TS Page 3/4 4-12 TS Page 3/4 4-14 TS Page 3/4 4-14a TS Page 6-15d TS Page 6-19c

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 2 of 21 Steam Generator Tube Intetgitv INDEX LIMMNG CONDMONS FOR OPERA AND SURVEJLLANCE REQUIREMENTS SECTION PAGE (56) TUBE INTEGRITY 3/4.4A PRESSURIZER .'/4.. ...... ....................................... 314 ,,,,,

4 3/4.4.5 STEAM GENERATO ................................................................................. 3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .3/4 4-12 Leakage Detection Systems .3/4 4-12 Reactor Coolant System Leakage .3/4 4-14 3/4.4.7 CHEMISTRY .3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY .3/4 4-17 3/4A.9 PRESSURE/TEMPERATURE LIMITS .3/4 4-21 Reactor Coolant System .3/4 4-21 Pressurizer .3/4 4-25 3/4A.10 STRUCTURAL INTEGRITY .3/4 4-26 ASME Code Class 1, 2, and 3 Components ........... ................... 3/4 4-26 3/4.4.11 DELETED .............................. 3/4 4-56 3/4.4.12 PORV BLOCK VALVES .............................. 3/4 4-58 3/4.4.13 POWER OPERATED REUEF VALVES ........ ...................... 3/44-59 3/4A.14 REACTOR COOLANT PUMP - STARTING................................................... 3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS .............................. 3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS' 3/4.5.1 SAFETY INJECTION TANKS ......................... 3/45-1 3/4.5.2 ECCS SUBSYSTEMS - Tv > 3251F ......................... 3/4 5-3 0

3/4.5.3 ECCS SUBSYSTEMS -T,, < 325 F ......................... 3M 5-7 3/4.5.4 REFUELING WATER TANK ......................... 3/4 5-8 ST. LUCIE - UNIT 1 V Amendment No. 28.60.88, 80,

St. Lucie Unit I L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 3 of 21 Steam Generator Tube Integrit INDEX ADMINISTRATIVE COHTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION ......................................... 6-12 6.7 SAFETY LIMIT VIOLATION ......................................... 6-12 6.8 PROCEDURES AND PROGRAMS ......................................... 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS ......................................... 6-15d Startup Report ......................................... 6-15d Annual Reports .................................... 6-10 Monthly Operating Reports .................................... 6-16a Annual Radioactive Effluent Release Repcrt .................................... 6-17 Annual Radiological Environmental Operating Report .................................... 6-18 Core Operating Umits Report (COLR) ..................................... 6-19 6.9.2 SPECIAL REPORTS........... ........ \6-19c 6.10 DELETED .................. .. 6-20 9-6.11 RADIATION PROTECTION PROGRAM............................................................ 6-21 6.12 HIGH RADIATION AREA ................................... 6-22 6.13 PROCESS CONTROL PROGRAMI .................. .............................. 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL ................................... 6-23 ISteom Generator Tube Inspection Rcport 6-19c I ST. LUCIE UNIT 1 X\' Amendment No. 2;, 2;.

69. 69,43.434.640.44,..

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 4 of 21 Steam Generator Tube Integrity DEFINITIONS IDENTIFED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) Into dosed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

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

c. Reactor Coolant System leakage through a steam generator to the secondary syste _rleake)

LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE 1.16 The LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE is that operating condition when (1)the cold leg temperature Is < 304IF during heatup or

< 281IF during cooldown and (2) the Reactor Coolant System has pressure boundary Integrity. The Reactor Coolant System does not have pressure boundary integrity when the Reactor Coolant System is open to containment and the minimum area of the Reactor Coolant System opening is greater than 1.75 square Inches.

MEMBER(S) OF THE PUBLIC 1.17 MEMBER OF THE PUBLIC means an indiidual in a controlled or unrestricted area. However, an Individual is not a member of the public during any period in which the Individual receives an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 THE OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used In the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and In the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1)the Radioactive Effluent Controls and Radiological Environmental Monitor-Ing Programs required by Section 6.8.4 and (2) descriptions of the information that should be Included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

ST. LUCIE - UNrT 1 1-4 Amendment No. 19, 60, rig, 4.404.423,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 5 of 21 St-eam Generator Tube Integrity 3EF1NMONS OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when It Is capable of performing its specified function(s),

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

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

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

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except(steamgerator tubel leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packing of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

PURGE - PURGING 1.24 PURGE or PURGING is the controlled process of discharging airorgas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, In such a manner that replacement air or gas is required to purify the confinement ST. LUCIE - UNIT I 1-5 Amendment No. 27, gk 23-

St. Lucie Unit I L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 6 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM STEAM GENERATOOl(sj) TUBE DMGRl LIMMNG CQONDION FOR OPER iO APPLICABILITY: MODES 1. 2.3 and 4.

ACTION:E With one or more steam generators Inoperable, restore the InoperabW lI generator(s) to OPERABLE status prior to Increasing Tsvf above 2000 SURVEILLANCE REQUIREMENTS

[generator shall be determined OPERABLE during shutdown by selecting and easttheminmumnumer of steam generators specified in X inpecingat 4.4.5.2 m Generator Tube Sample Selecticn and g enerator tube minimum sample sor Inspetion result classificatione e,

ond t c snding action required shall be as specified in Table 4.4-2.

h a Winservicercton of steam generator tubes shall be performed at the frequencies s tryeIneS cificaion 4.4.5.3 and theinspected then tubes shall be veriie aetble per the acceptance criteria of Specification 4.4.15.4. Tlheeuls seleced for each inservice inspectbon __

shall include at least 3% of the-lbtal number of tubes in all steam generators; the tubes selected for tseInspections shall be selected on a rar dom basis except: \1

}Weeexperience in ilahs r with similar water istry indicates critclaeste nptdhn at lea 50% of the tubes inspeced sh from these critical Sd. l

b. The first Inse specton (subsequent to the pr

l Inspection) of each sI eneralor shall Include:

\ 1. All pgedt I tbes t it ously had detectable walll a ubs. nthsearasw ere eprnc dicated

Separate Action entry is allowed for each SG tube.

o)l ST/UI NTI3445Mnmn ST. LUCIE - UNIT I 3144,5 Amen dment No.'e

St. Lucie Unit 1 L-2006-089 DocketNo. 50-335 Attachment 2.

Proposed License Amendment Page 7 of 21 Steam Generator Tube Integrity 3/4.4.5 INSERT A SG tube integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the SG Program.

3/4.4.5 INSERT B

a. With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program;

1. Within 7 days verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube inspection.

b. With the requirements and associated allowable outage time of Action a above not met or SG tube integrity not maintained, be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

3/4.4.5 INSERT C Verify SG tube integrity in accordance with the Steam Generator Program.

3/4.4.5 INSERT D Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Przposed License Amendment Page 8 of 21 Stcam Generator Tube IntegritV

\REACTOR COOLANT SYSTEM/

sURVLLANE REQUIREMENnSG{C jOnfflu;D?

er t oThe second and third inservice inspea f ions ay be less than a

\full tube Inspecion by concentrating (selecting at least 50%

\ ofhe tubes to iinspein the 5% on those areasce of the tube sheet array and on those portions of the tubes

\where tubes with Imperfecions were previously found.v The results of ehsample inspectin shall be classified into one of/

the following thOe otebu mrhes: /

t Inspection Resufti Less than 5% of totalfhe tubes

\-3 oaredegraded tubes and none speed

\ tbsare defective. /

2 or more t but mbes han 1% o f total tubes Ins or eiti t 1 c 5% and I the total tubes in te are deg alubes.

.Cre 3 ne Mron10%

eubes total inspected o e e o rared h e at or more than 1% of the inspected tps are defective.

Note: In all Inspections prev sa bed tubes must exhibit significant (> 1 0 fute iQpenetrations to be included in theyov e nt ge Iaculations.

4.4.5.3 Inspection Frequencies -ch above required i vic nspections of steam generator tubes shall be ptrformed at the following f encies:

a. me first hnservt Inspection shall be promdabe 6 Effective Ful ower Months but within 24 calendar ths of Initial crit iy. Subsequent inservice inspections sha be perform at Intervals of not less than 12 nor more than 4 calend months after the previous Inspection. If two cons live I pecons following service under AVT conditions, not In ding the preservice Inspection, result In all Inspection ults falling Into the C-1 category or if two consecutive nspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the Inspection interval may be extended to a maximum of once per 40 months.

THIS PAGE DELETED\

ST. LUCIE - UNIT I a144-s

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 9 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM uRV,~~ N EREQUIREMENTS SlL mContinuedm

b. If the Inservice Inspection of a steam generator conducted In accordance with Table 4.4-2 requires a third sample Inspection whose results fall in Category C-3, the inspection frequency shall be reduced to at least once per 20 months. The reduction in Inspection frequency shall apply until a subsequent inspection demonstrates that a third sample Inspection Is not required.

c. itional, unscheduled inservice Inspections shall be perfo on e steam generator in accordance with the first sampl Inspe n specified In Table 4.4-2 during the shutdown s equent to any o he following conditions.

1. Primary secondary tubes leaks (not Includi eaks originati mmtube-to-tube sheet welds) in cess of the limits cification 3.4.6.2,

2. A seismic occu nce greater than the rating Basis Earthquake,

3. A loss-of-coolant acc nt reqJiria actuation of the engineered safeguards,

4. A main steam line or f or line break.

4.4.5.4 Acceptance Criteria

a. As used In this Specitica

1. Imperfection means an exception to the imensions, finish or contour of ube from that required by brication drawings or=pecifications. Eddy-current te ing Indications below 200/ of the nominal tube wall thickness if detectable, may be idered as Imperfections.

2. rdalo means a service-induced cracking, wage.

or general corrosion occurring on either inside ide of a tube.

3 Dearaded Tube means a tube containing Imperfections > o of the nominal wall thickness caused by degradation.

4. % Degradation means the percentage of the tube wall thickness affected or removed by degradation.

THIS PAGE DELETEiD ST. LUCIE - UNIT I 3/4 4-7

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 10 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM oLLAICE REQUIIREiMENEa (Continu e _

5. Defect means an imperfection of such severity that It exceeds the plugging limit. A tube containing a defect Is defective. Any tube which does not permit the passage of the eddy-current inspection probe shall be deemed a defective tube.

Pluging Limit means the Imperfection depth at or beyond which the tube shall be removed from service because it y me unserviceable prior to the next Inspection and a! to 40% of the nominal tube wall thickness.

7. Unse ceable describes the condition of a tube f leaks or cont s a defec large enough to affect Its ctural Integrity I he event of an Operating Basis quake, a loss-of-coo t accident, or a steam line or er line break as s ed In 4.4.5.3.c, above.

8. Tube Inspection; ans an inspectin fthe steam generator tube from the point entry (hot leg de) completely around the U-bend tome top sup rt of the cold leg.

b. The steam generator shall be ed OPERABLE after completing note the corresponding actions (plug tubes exceeding the plugging limit and all tubes containing -wall cracks) required by 0o Table 4.4-2.

4.4.5.5 Reports

a. Within 15 days followg the completion of e inservice inspection of stea generator tubes, the numbr of tubes plugged in each steam g erator shall be reported to the mmission In a special report rsuant to Specificzition 6.9.2.

b. The compl e results of the steam generator tube Inse ce Inspecio shall be submitted to the Commission In a sp at report rsuant to Specification 6.9.2 within 12 months fol wing com tion of the inspection. This special report shall inclu I Number and extent of tubes Inspected.

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

3. Identification of tubes plugged.

THIS PAGE bELETEI)

ST. LUCIE - LINT 1 3a44-8 Amendment No. n, 408,'4f

I (Dr CDW0 r CD CD.

El 0 2 -'

TABLE 4.4-1 o C MINIMUM NUMBER OF STEAM GENERATORS TO BE LAt INSPECTED DURING INSERVICE INSPECTION " W 3 CD Preservice Inspection \No / Yesl

=S 0 No. of Steam Generators per U Ki Two lThree l ou / o Tre Fu , R First Inservice Inspection \All / one Two Two Second & Subsequent Inservice Inspectins \One1 One One 2 One3 Table Notation:

1. The inservice inspection may be lmited to one steam gener a rotatng schedule encompassing 3 N % of the tubes (where N Is the number of steam generators In the plant he results the first or previous inspections indicate that all steam generators are performing In a like manner ote that under so crcumstances, the operating conditions in one or more steam generators may be found to more severe than those in er steam generators. Under such circum-stances the sample sequence shall be mdf to inspect the most severe condi .

2. The other steam generator not inspect during the first inservice inspection shall be in ced. The third and subsequent inspections should follow the Instrun ns described In 1 above.

3. Each of the other two steam perators not inspected during the first inservice inspections shall nspected during the second and third insp The fourth and subsequent inspections shall follow the instructions de in b ST. LUCIE - UNIT I 3/44-10 0 C OCD e?

00 t W oo

(D O8U CD in TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION /

n - ° I SAMPLE INSPECTION 2nd SAMPLE INSPECTION I 3rd SAMPLE I PECTION ;3 0, L Sample Size esutl Action Required Result Action Required Result [ ctlon Required

.- +

CDl i5 On A minimum of S Tubes per CN None N/A NIA N/A rI NIA S.G. 7, C-2 Plug fecive tubes C-1 None N/A NMA and mnspt additional 2S tubes in S.G. C-2 Plug defecive t es C-1 None and inpdiinal C 4S tubes this S.G. C-2 Plug defective tubes C-3 Perform action for C-3 result of first sample 3 Perform action for N/A N/A C-3 result of first I esamp!e.I C-3 Aljplter N None NMA N/A Inspect all tubes in this S.G., plug de- ;er3.s are fective tubes and C-1 inspec 2S tubes I Some S.G.s Perfom aon for N/A N/A C-2 but no G-2 result of nd additional sample S.G. are C-3 'I-, _

Additional Inspect all tubes In N/A S.G. Is C-3 each S.G. and plug 7-1, defective tubes.

S=3N% W e N is the number of steam generators In the unit, and n Is the number of steam generators Inspe n uring an inspection.

requirement to inspect all tubes may be relaxed for Cycle 5 Refueling since an ngineerlng evaluation has shown that the condition(s) has been adequately bounded by Inspection.

ST. LUCIE - UNIT 1 3M44-11 Amendment No. 4Z8 0 (

tOk OCD 7 k) 00

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 2 Proposed License Amendment Page 13 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 314.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMING CONDITION FOR DPERAl1ON 3.4.6.1 The following RCS leakage detection systems will be OPERABLE:

a. The reactor cavity sump inlet flow monitoring system; and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

APPLICABILITY: MODES 1, 2,3, and 4. per Surveillance Requirement 4.4.6.2.C1 ACTION: E\

a. With the required reactor cavity sump I et fI onitoring inoperable, Istem perform a RCS water Inventory bala at least o e per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and restore the sump Inlet flow monitoring system to OPERABL tatus within 30 days; otherwise, be in at least HOT STANDBY within the ne hours and In COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the required radioactivity monitor inoperable, analyze grab ples of the containment atmosphere or perform a RCS water inventory bela at least once per 241hours, and restore the required radioactivity monitor to OPERABLE~status within 30 days; otherwise, be In at least HOT STANDBY within the ne~t 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . M

c. With all required monitors Inoperable, enter LCO 3.0.3 immediately.

d. The provisions of Specification 3.0.4 are not applicable If at least one of the required monitors is OPERABLE.

S E A REQUIREME 4.4.6.1 The RCS leakage detection Instruments shall be demonstrated OPERABLE by:

a. Performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor at the frequencies specified In Table 4.3-3.

b. Performance of the CHANNEL CALIBRATION of the required reactor cavity sump Inlet flow monitoring system at least once per 18 months.

[Not required to be performed until 12hours after establishment of steady state operation.

ST. LUCIE - UNIT I 3144-12 Amendment No. 144

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 14 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITING CONDITION FOR OPERATIO 3.4.6.2 Reactor Coolant System t ieakage shall be limited to: operatioall

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, 1I

c. Pmary-to-secondary leakage throughteam generator 150 gallons per day d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, and

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

APPLICABILITY: MODES 1, 2,3 and 4.

ACTiON: or with prinr-to-seondary leakage not within limit,

a. With any PRESSURE BOUNDARY LEAKAGEle in at least HOT STANDBY wIthin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Io-rtiotakag graeIhnayoeo

b. With any Reactor Coolant Systenl'leakage greater than any one of prinmory-to-secondoy the above limits, excluding'PRESSURE BOUNDARY LEAKAGFnd Reacor eCoolant System Pressure Isolation Valve leakage, reduce the lea-age rate to within Elmits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and In COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit in 3.4.6.2.e above reactor operation may continue provided that at least two valves, Including check valves, In each high pressure line having a non-functional valve are In and remain In the mode corresponding to the isolated con-dition. Motor operated valves shall be placed in the closed posi-tion, and power supplies deenergzeed. (Note, however, that this may lead to ACTION requirements for systems involved.) Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be In at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and In COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant Syste leakages shall be demonstrated to be within each of the above limits by.

lorcrtionl

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

ST. LUCIE -UNIT 1 3144-14 r~dor daed knowI Amendment No.4e

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 15 of 21 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE SURVEILLANCER UIREMENTS (QCntinued)

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , C. Performance of a Reactor (te ntoory balance at least once per 72 hou teady stat operation xcept when operating in the shutdown cooling mode,

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

e. Verifying each Reactor Coolant System Pressure Isolation Valve leakage (Table 3.4.6-1) to be within limits:

1. Prior to entering MODE 2 after refueling,

2. Prior to entering MODE 2, whenever the plant has been In COLD SHUTDOWN for 7 days or more if leakage testing has not been performed In the previous 9 months, 5

3. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

4. The provision of Specification 4.0.4 is not applicable for entry into MODE 3 or 4.

f. Whenever integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other salve located in each high pressure line having a leaking valve shall be recorded dais

9. Primary-to-secondary leakage shall be verified

150 gallons per iday through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . **

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady 1state operation. Not applicable to primary-to-secondary leakage.

- Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady jstate operation.

ST.LCIE -UNT I 144-14a Amendment Nobk3-

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 16 of 21 Steam Generator Tube 1nteZrity ADMINISTRATIVE CQNTRRQLS (Continued} _

k. Ventlation Filter Testina Procram (VFTP)

A program shall be established to implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified In Regulatory Guide 1.52, Revision 2.

1. Demonstrate for each of the ESF systems that an Inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested In accordance with ANSI N510-1975 at the system flowrate specified below.

ESF Ventilation System Flowrate Control Room Emergency Ventilation 2000 +/- 200 cfn

2. Demonstrate for each of the ESF systems that an Inplace test of the charcoal adsorber shows a penetration end system bypass s 0.05% when tested Inaccordance with ANSI N510-1975 at the system flowrate specified below.

ESF Ventilation System Flowrate Control Room Emergency Ventilation 2000 + 200 cfn

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber. when obtained as described InRegulatory Gutde 1.52. Revision 2.

shows the methyl Iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 301C and the relative humidity specified below.

ESF Ventilation System Penetration RH Control Room Emergency Ventilat on < 2.5% 70%

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and charcoal adsorbers is less than the value specified below when tested at the system flowrate specafied below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Ventilaton <4.15 W.G. 2000 +/- 200 cfm The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

6.9 REPORTING REQUIREMENTS INSERT 6.8.4 L.

ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10. Code of Federal Regulations, the following reports shall be submitted to the NRC.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license. (2) amendment of the license Involving a planned Increase In power level, (3) Installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have signlficantly altered the nuclear, thermal or hydraulic performance of the plant ST. WCIE - UNIT 1 6-15d Amendment No. 4J6. l

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 17 of 21 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS - INSERT 6.8.4.1 (lima).

1. Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.

1. Structural integrity performance criterion: All in-service SG tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cooldown and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 18 of 21 Steam Generator Tube Integrity in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm total through all SGs and 0.5 gpm through any one SG.

3. The operational leakage performance criterion is specified in LCO 3.4.6.2.c, "Reactor Coolant System Operational Leakage."

c. Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tube may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outages nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 PrDposed License Amendment Page 19 of 21 Steam Generator Tube Integritv or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 20 of 21 Stam Generator Tube Integritv ADlMINISTRATIVE CQNTRQLS CORE OPERATING LIMITS REPORT (continued)

C. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency r Alm__o--In- Xm-is na_ -u._ as-)u IKEVP4 orColing Systems (EC)limits, nuclear limits such as SHU TWvWN INSERT 6.9.1.12 MARGIN, transient analysis limits, and accident analysis limits) of the safet analysis are met.

d. The COLRP Including any mid cycle revisions or supplements, shall be proa ided upon issuance for each reload cycle to the NRC.

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

ST. LUCIE - UNIT I D110DC Amendment No. 5/

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 2 Proposed License Amendment Page 21 of 21 Steam Generator Tube Integritv INSERT 6.9.1.12 (new)

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection performed in accordance with Specification 6.8.4.1, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 1 of 18 Steam Generator Tube Integrity Word-Processed Technical Specifications TS Page V TS Page XV TS Page 1-4 TS Page 1-5 TS Page 3/4 4-5 TS Page 3/4 4-6 TS Page 3/4 4-7 TS Page 3/4 4-8 TS Page 3/4 4-10 TS Page 3/4 4-11 TS Page 3/4 4-12 TS Page 3/4 4-14 TS Page 3/4 4-14a TS Page 6-15d TS Page 6-15e TS Page 6-15f TS Page 6-19c

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 2 of 18 Steam Generator Tube Integritv INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS_

SECTION PAGE 3/4.4.4 PRESSURIZER ................................. 3/4 4-4 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY ................................. 3/44-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ................................. 3/4 4-12 Leakage Detecion Systems ................................. 3/4 4-12 Reactor Coolant System Leakage ................................. 3/4 4-14 3/4.4.7 CHEMISTRY ................................. 3/4 4-15 3/4.4.8 SPECIFIC ACTIVITY ................................. 3/4 4-17 3/4.4.9 PRESSURE/TEMPERATURE LIMITS ................................. 3/4 4-21 Reactor Coolant System ................................. 3/4 4-21 Pressurizer ................................. 3/4 4-25 3/4.4.10 STRUCTURAL INTEGRITY ................................. 3/4 4-26 ASME Code Class 1, 2, and 3 Components ................................. 3/4 4-26 3/4.4.11 DELETED ................................. 3/4 4-56 3/4.4.12 PORV BLOCK VALVES ................................. 3/4 4-58 3/4.4.13 POWER OPERATED RELIEF VALVES ................................. 3/4 4-59 3/4.4.14 REACTOR COOLANT PUMP - STARTING ................................. 3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS ................................. 3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ......................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T., 9 > 3251F ......................... 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tag < 325F ......................... 3/4 5-7 3/4.5.4 REFUELING WATER TANK ......................... 3/4 5-8 ST. LUCrE - UNIT V Amendment No. 24,6a, 68. a0 434,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 3 of 18 Steam Generator Tube Intemitv INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION ............................................ 6-12 6.7 SAFETY LIMIT \VOLATION .6-12 6.8 PROCEDURES AND PROGRAMS .6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS .6-15d Startup Report .6-15d Annual Reports .6-16 Monthly Operating Reports .6-1 6a Annual Radioactive Effluent Release Report .6-17 Annual Radiological Environmental Operating Report ............................................ 6-18 Core Operating LImits Report (COLR) .6-19 Steam Generator Tube Inspection Report .6-19c 6.9.2 SPECIAL REPORTS .6-19c 6.10 DELETED ............................... 6-20 6.11 RADIATION PROTECTION PROGRAM ....... ........................ 6-21 6.12 HIGH RADIATION AREA ............................... 6-22 6.13 PROCESS CONTROL PROGRAM ............................... 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL ........ ....................... 6-23 ST. LUCIE - UNIT I XV Amendment No. 2;.37, 59, 6a=,*S. 434. ,4;4.4; 4,480,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 4 of 18 Steam Generator Tube Integrity DEINITIONS IDENTIFED LEAKAGE 1.15 IDENTIFIED LEAKAGE shall be:

a. Leakage (except CONTROLLED LEAKAGE) into closed systems, such as pump seal or valve packing leaks that are captured, and conducted to a sump or collecting tank, or

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

c. Reactor Coolant System leakage through a steam generator to the secondary system (Primary-to-secondary leakage).

LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE 1.16 The LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE is that operating condition when (1)the cold leg temperature is < 3041F during heatup or

Si281OF during cooldown and (2)the Reactor Coolant System has pressure boundary integrity. The Reactor Coolant System does not have pressure boundary integrity when the Reactor Coolant System Is open to containment and the minimum area of the Reactor Coolant System opening Isgreater than 1.75 square inches.

MEMBER(S) OF THE PUBLIC 1.17 MEMBER OF THE PUBLIC means an individual in a controlled or unrestricted area. However, an individual is not a member of the public during any period in which the individual receives an occupational dose.

OFFSITE DOSE CALCULATION MANUAL (ODCM) 1.18 THE OFFSITE DOSE CALCULATION MANUAL (ODCM) shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, Inthe calculation of gaseous and liquid effluent monitoring Alarm/Trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1)the Radioactive Effluent Controls and Radiological Environmental Monitor-ing Programs required by Section 6.8.4 and (2)descriptions of the information that should be included in the Annual Radiological Environmental Operating and Annual Radioactive Effluent Release Reports required by Specifications 6.9.1.7 and 6.9.1.8.

ST. LUCIE - UNIT 1 14 Amendment No. 69. 60, 69, 84,404, 423, 425.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 5 of 18 Steam Generator Tube Integrity DEFINITIQNS OPERABLE - OPERABILITY 1.19 A system, subsystem, train, component or device shall be OPERABLE or have OPERABILITY when It Is capable of performing its specified function(s),

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

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

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

PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault In a Reactor Coolant System component body, pipe wall or vessel wall.

PROCESS CONTROL PROGRAM (PCP) 1.23 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling. analyses, test, and determinations to be made to ensure that processing and packing of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements govemning the disposal of solid radioactive waste.

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

ST. LUCIE - UNIT I

1-5 Amendment No. 27, 69. 423.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 6 of 18 Steam Generator Tube InteZrity REACTOR COOLANT SYSTEM STEAM GENERATOR (SG) TUBE INTEGRITY LIM_ NG-CONDMON FOR OPERATION 3.4.5 SG tube Integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the SG Program.

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

I

a. With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generelor Program;

1. Within 7 days verify tube Integrity of the affected tube(s) Is maintained until the next refueling outage or SG tube Inspection, and

2. Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or SG tube Inspection.

b. With the requirements and associated allowable outage time of Action a above not met or SG tube Integrity not maintained, be In HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and In COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

BMICE REQUIREMENTS 4.4.5.1 Verify SG tube Integrity In accordance with the Steam Generator Program.

4.4.5.2 Verify that each inspected SG tube that satisfies the tube repair criteria Is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube Inspecthon.

Separate Action entry is allowed for each SG tube.l ST. LUCIE - UNIT I 3144 5 SAnendment No. 4A.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 7 of 18 Steam Generator Tube IntezriW THIS PAGE DELETED I ST. LUCIE - UNr I4 3/4 446 Amendment No.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 8 of 18 Steam Generator Tube Integritv THIS PAGE DELETED ST. LUCIE - UNT I4 3144-7 Amendment No.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 9 of 18 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNIT I 3W4 4- Amendment No. 73, 4X8, 448,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 10 of 18 Steam Generator Tube Integritv THIS PAGE DELETED I ST. LUCIE - UNIT 1 3/4 4-10 Amendment No.

St. Lucie Unit I L-2006-089 Dccket No. 50-335 Attachment 3 Proposed License Amendment Page 11 of 18 Stvam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE

UNIT I 3/4 4.1 1 Amendment No. 47, 69,

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 3 Proposed License Amendment Page 12 of 18 Steam Generator Tube Intearity REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDMON FOR OPERATION 3.4.6.1 The following RCS leakage detection systems will be OPERABLE:

a. The reactor cavity sump inlet flow monitoring system: and

b. One containment atmosphere radioactivity monitor (gaseous or particulate).

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

ACTION:

a. With the required reactor cavity sump inlet flow monitoring system Inoperable, perform a RCS water Inventory balance per Surveillance Requirement 4.4.6.2.c at least once per 24- hours and restore the sump Inlet flow monitoring system to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the required radioactivity monitor inoperable, analyze grab samples of the containment atmosphere or perfomr a RCS water inventory balance per Surveillance Requirement 4.4.6.2.c at least once per 24' hours, and restore the required radioactivity monitor to OPERABLE status within 30 days; otherwise.

be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With all required monitors Inoperable, enter LCO 3.0.3 immediately.

d. The provisions of Specification 3.0.4 are not applicable If at least one of the required monitors is OPERABLE.

SURVEILLANCE REQUIREMENTS 4.4.6.1 The RCS leakage detection instruments shall be demonstrated OPERABLE by

a. Performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, and CHANNEL CALIBRATION of the required containment atmosphere radioactivity monitor at the frequencies specified in Table 4.3-3.

b. Performance of the CHANNEL CALIBRATION of the required reactor cavity sump inlet flow monitoring system at least once per 18 months.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

STr.LUCIE - UNIT I 3/4 4-12 Amendment No. 444,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 13 of 18 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE LIMITiNr CONDAON FOR OPERA 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE,

c. 150 gallons per day primary-to-secondary leakage through any one steam generator (SG),

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

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

APPLICABILITY: MODES 1, 2,3 and 4.

ACTION:

a. With any PRESSURE BOUNDARY LEAKAGE, or with primary-to-secondary leakage not within limit, be In at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With any Reactor Coolant System operational leakage greater than any one of the above limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and Reactor Coolant System Pressure Isolation Valve leakage, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be In at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c. With any Reactor Coolant System Pressure Isolation Valve leakage greater than the limit In 3.4.6.2.e above reactor operation may continue provided that at least two valves, including check valves, in each high pressure line having a non-functional valve are In and remain In the mode corresponding to the Isolated con-dition. Motor operated valves shall be placed in the closed posi-tion, and power supplies deenergized. (Note, however, that this may lead to ACTION requirements for systems involved.) Otherwise, reduce the leakage rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SUREILLANCE REQUIREMENTS 4.4.6.2 Reactor Coolant System operational leakages shall be demonstrated to be within each of the above limits by:

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

ST. LUCIE - UNIT 1 3/4 4-14 QRde-dared-4/20/81 Amendment No. 448,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 14 of 18 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM REACTOR COOLANT SYSTEM LEAKAGE SURVEILLANCE REQUIREMENTS (Continued)

b. Monitoring the containment sump inventory and discharge at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

c. 'Performance of a Reactor Coolant System water Inventory balance at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months except when operating in the shutdown cooling mode,

d. Monitoring the reactor head flange leakoff system at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

e. Verifying each Reactor Coolant System Pressure Isolation Valve leakage (Table 3.4.6-1) to be within limits:

1. Prior to entering MODE 2 after refueling,

2. Prior to entering MODE 2, whenever the plant has been in COLD SHUTDOWN for 7 days or more if leakage testing has not been performed In the previous 9 months,

3. Prior to returning the valve to service following maintenance, repair or replacement work on the valve.

4. The provision of Specification 4.0.4 Is not applicable for entry into MODE 3 or 4.

f. Whenever Integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in each high pressure line having a leaking valve shall be recorded daily; and

g. Primary-to-secondary leakage shall be verified *150 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ."

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Not applicable to primary-to-secondary leakage.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

ST. LUCIE - UNIT I 3/4 4-14a Amendment No. 433.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 15 of 18 Steam Generator Tube Integrity V NTROLS continued)

k. Ventilation Filter Testing Program (VFTP)

A program shall be established to Implement the following required testing of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in Regulatory Guide 1.52, Revision 2.

1. Demonstrate for each of the ESF systems that an Inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass E 0.05% when tested In accordance with ANSI N510-1975 at the system flowrate specified below.

ESF Ventilation System Flowrate Control Room Emergency Ventilation 2000 +/- 200 cfm

2. Demonstrate for each of the ESF systems that an Inplace test of the charcoal adsorber shows a penetration and system bypass <0.05% when tested In accordance with ANSI N510-1975 at the system flowrate specified below.

ESF Ventilation System Flowrate Control Room Emergency Ventilation 2000 + 200 cfm

3. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, shows the methyl Iodide penetration less than the value specified below when tested In accordance with ASTM D3803-1989 at a temperature of 30°C and the relative humidity specified below.

ESF Ventilation System Penetration RH Control Room Emergency Ventilation 2.5% 70%

4. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA fiMters and charcoal adsorbers is less than the value specified below when tested at the system tlowrate specified below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Ventilation <4.15' W.G. 2000 + 200 cfm The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

I. Steam Generator (SG) Program A Steam Generator Program shall be established and Implemented to ensure that SG tube Integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the 'as found' condition of the tubing with respect to the performance criteria for structural Irtegrity and accident Induced leakage. The 'as found' condition refers to the condition of tie tubing during an SG Inspection outage, as determined from the Inservice Inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are Inspacted or plugged to confirm that the performance criteria are being met.

ST. LUCIE - UNrT 1 6-15d Amendment No. 476, 407,

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 3 Proposed License Amendment Page 16 of 18 Steam Generator Tube Integritv ADMINSIRATIVE CGO ROLS ltonfinuedl I. Steam Generator (SG) Program (continued)

b. Performance criteria for SG tube Integrity. SG tube Integrity shall be maintained by meeting the performance criteria for tube structural Integrity, accident induced leakage, and operational leakage.

1. Structural Integrity performance criteriont All in-service SG tubes shall retain structural Integrity over the full range of normal operating conditions (Including startup, operation In the power range, hot standby, and cooldown and all anticipated transients Included In the design specification) and design basis accidents. This Includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-t-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents In accordance with the design and licensing basis, shall also be evaluated to determine If the associated loads contribute significantly to burst or collapse. In the assessment of tube Integrity, those loads that do significantly affect burst or collapse shall be determined and assessed In combination with the loads due to pressure with a safety factor of 12 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident Induced leakage performance criterion: The primary-to-secondary accident Induced leakage rata for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed In the accident analysis In terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage Is not to exceed I gpm total through all SGs and 0.5 gpm through any one SG.

3. The operational leakage performance criterion Isspecified In LCO 3.4.6.2.c, Reactor Coolant System Operational Leakage.t

c. Provisions for SG tube repair criteria. Tubes found by Inservice Inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube Inspections. Periodic SG tube Inspections shall be performed.

The number and portons of the tubes Inspected and methods of Inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube Inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld Isnot part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, Inspection methods, and Inspection intervals shall be such as to ensure that SG tube Integrity Is maintained until the next SG Inspection.

An assessment of degradation shall be performed to determine the type and location of flaws to which the tube may be susceptible and, based on this assessment.

to determine which Inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes In each SG during the first refueling outage following SG replacement ST. LUCIE - UNIT I C-15e Amendment No.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 17 of 18 Steam Generator Tube Integrity I. Steam Generator (SG) Program (continued)

d. (continued)

2. Inspect 100% of the tubes at sequential periods of 144,108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, Inspect 50% of the tubes by tie refueling outage nearest the midpoint of the period and the remaining 50%, by the refueling outages nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever Is less) without being Inspected.

3. If crack Indications are found in any SG tube, then the next Inspection for each SG for the degradation mechanism that caused the crack Indication shall not exceed 24 effective full power months or one refueling outage (whichever Is less). If definitive Information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation Indicates that a crack-like Indication Is not associated with a crack(s), then the Indication need not be treated as a crack.

e. Provisions for monitoring operational primary-to-secondary leakage.

6.9 REPORTING REQUIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the NRC.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment of the license involving a planned Increase In power level, (3) Installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantiy altered the nuclear, thermal or hydraulic performance of the plant ST. LUCIE - UNIT 1 E-1 51 Amnendmrent No.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 3 Proposed License Amendment Page 18 of 18 Steam Generator Tube Jntelrity ADMINISTRAiV CONTROLS CORE OPERATING LIMITS REPORT (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN MARGIN, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any mid cycla revisions or supplements, shall be provided upon Issuance for each reload cycle to the NRC.

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the Initial entry Into HOT SHUTDOWN following completion of an inspection performed Inaccordance with Specification 6.8.4.1, Steam Generator (SG) Program. The report shall include:

a. The scope of Inspections performed on each SG.

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service Induced indications,

e. Number of tubes plugged during the Inspection outage for each active degradation mechanism.

f. Total number and percentage of tubes plugged to date,

g. The results of condition monitoring, Including the results of tube pulls and In-situ testing, and

h. The effective plugging percentage for all plugging in each SG.

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

ST. LUCIE - UNIT 1 r-1 PC Amendment No. 474,

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 1 of 17 Steam Generator Tube Integrity TS Bases Markups

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 2 of 17 Steam Generator Tube Integrity SECTIONNO.: rrrLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 2 of 16 REVISION NO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 3/4.4 ................... 3 314.4 REACTOR COOLANT SYSTEM .3 BASES .3 3/4A.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . . 3 3/4.4.2 DELETED . . 4 ISTUB NEm; 3/4.4.3 SAFETY VALVES .................. T...........................T....4 3/4A.4 PRESSURIZER .......... .............................. 5 3/4.4.5 STEAM GENERATO S....................................... 5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .....

314.4.6.1 LEAKAGE DETECTION SYSTEMS ....

3/4.4.6.2 REACTOR COOLANT SYSTEM LEAKAGE..............................................

314A.7 CHEMISTRY .8 3/4.4.8 SPECIFIC ACTIVITY .8 314.4.9 PRESSURE/TEMPERATURE LIMITS . 9 314.4.10 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 . 15 314.4.14 REACTOR COOLANT PUMP - STARTING . 15 314.4.15 REACTOR COOLANT SYSTEM VENTS . 16

St. Lucie Unit I L-20016-U89 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 3 of 17 Steam Generator Tube Integrity SECTIONNO.: T.TLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 5 of 16 REVISIONNO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.3 SAFETY VALVES (continued)

The pressurizer code safety valve as-found setpoint Is2500 psla +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.

3/4.4.4 PRESSURIZER A steam bubble In the pressurizer ensures that the RCS Is not a hydraulically solid system and Iscapable 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 undes rable opening of the spring-loaded pressurizer code safety valves. The required pressurizer heater capacity Is capable of maintaining natural d-culation 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 GENERINTE II K ne FA LE steam generator provides sufficient heat removal capaiiyt e decay heat after a reactor shutdown. The requirmn for tw ta e e pable of removing decay heat, combined with the requirements of Specifel 3..1.1, 3.7.1.2 and 3.7.1.3 ensures adequate decay hea remoa pe for RCS temperatures greater than 3251F if one steam genrator beco p rable due to single failure considerations. Below 325°F, decay heat isrem~ytesudw cooling system.\

St. Lucie Unit I L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 4 of 17 Steam Generator Tube Integrity SECTIONNO.: TrTLE: TECHNICAL SPECIFICATIONS PAGE:

3/4A. BASES ATTACHMENT 6 OF ADM-25.04 6 of 16 REVISIONNO.: REACTOR COOLANT SYSTEM 1 ST. LUCIE UNIT 1 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) l 314.4.5 STEAM GENERATOR co The Surveillance Requirements for inspection of the steam genera oh mubas ensure tha the naturan cutegrity of this portion of the RCS wil mInandTh program for Inservice inspection of steam generator tubes n based oe a modpfcation of Regulatory Guide 1.83tRevision 1.d Innegii corsion of steam generator tubing is essential In order to maintain surveillance that corcived of the conditions of the tubes in the event that there meaurestrcabetuaken.

is e coroesice mechanical iof damage or progressive degradation due to designt _ uac turing errors or in-service conditions that lead to corroso ton of Steam generator tubing also provides a means lm during the nature and cause of any tube degradation so that anethe loas Ipscan be taken.

The pints Opertiganbe operated in a manner such p that he secondary coolankag e within those parameter limits found to result in negligible corrosion s f thteam generator tubes. If the secondary p

coolant chemistre shudw a ined within these parameter limits c localized corrosion may likely suit in stress corrosion cracking. The extent of cracking during pla In would be limited by the limitation of steam generator unlikely en the primary coolant system and the secondary coolantHowev (if a secondary leakage = 1gallon per minute, total). Crcshvn ~ ary-to-secondary leakage less than this limit during operation Wlill hae an adequate margin of safety to withstand the loads imposed during norI operation and by postulated accidents. Operating plants have,d MO laed that primary-to-secondary leakage of I gallon per minute can readily bydetected by radiation monitors of steam generator blovziown. Leakg in excess of this limit will require plant shutdown and an unschedul ~s ection, during which the leaking tubes will be located and piugged.\

Wastage-type defects are unlikely with the all vo aikreatment (AVT) of secondary coolant. However, even if a defect of sim~i type should develop in service, it will be found during scheduled Insr ce steam generator tube examinations. Plugging will be required o Iftubes with Imperfections exceeding the plugging limit which, by the de itlon of Specification 4.4.5.4.a. is 40% of the tube nominal wall tikes ta tue ispetios o oprating plants have demntad h geneato

+ cpabityto eliblydetct egrdation that has penetrae 0& h

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 5 of 17 Steam Generator Tube Integrity INSERT B 3/4.4.5 Ba :kground Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant:

through the primary to secondary heat exchangers. The SG tubes have a number of important:

safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory.

The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as -he heat transfer surface between the primary and secondary systems to remove heat from.

the primary system. This Specification addresses only the RCPB integrity function of the SG.

The SG heat removal function is addressed by LCO 3.4.1.1, "Reactor Coolant Loops and CoDlant Circulation, Startup and Power Operation," LCO 3.4.1.2, "Hot Standby," LCO 3.4.1.3, "Hot Shutdown," LCO 3.4.1.4.1, "Cold Shutdown - Loops Filled," and LCO 3.4.1.4.2, "Cold Shutdown - Loops Not Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

SG tubing is subject to a variety of degradation mechanisms. SG tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, anE.

stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 6.8.4.1, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.1, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SG performance criteria are described in Specification 6.8.4.1. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

Anplicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding a SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to the operational leakage rate

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 6 of 17 Steam Generator Tube Integrity limits in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage," plus the leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is released via the main steam safety valves and/or atmospheric dump valves. The majority of the activity released to the atmosphere results from the tube rupture.

Th. analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of I gpm and 0.5 gpm through any one SG as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-13 1 is assumed to be equal to the limits in LCO 3.4.8, "Reactor Coolant System Specific Activity." For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3),

10 CFR 50.67 (Ref. 7) or the NRC approved licensing basis (e.g., a small fraction of these linrits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO)

Thee LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair cril eria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In ihe context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A 'SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 6.8.4.1, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. Failure to meet any one of these criteria is considered failure to meet the LCO.

St. Lucie Unit I L-2006-089 Dccket No. 50-335 Attachment 4 Proposed License Amendment Page 7 of 17 Steam Generator Tube Integrity The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load verses displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code, Section 111, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5)..

The accident induced leakage performance criterion ensures that the primary-to-secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 1 gpin total from all SGs and 0.5 gpm through any one SG. The accident induced leakage rate includes any primary-to-secondary leakage existing prior to the accident in addition to primary-to-secondary leakage induced during the accident.

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in LCO 3.4.6.2, "Reactor Coolant System operational leakage," and limits primary-to-secondary leakage through any one SG to 150 gpd at room temperature. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 8 of 17 Steam Generator Tube Integrity Applicabilitv SG tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in POWER OPERATION, START UP, HOT STANDBY and HOT SHUTDOWN.

RCS conditions are far less challenging in COLD SHUTDOWN and REFUELING than dufing POWER OPERATION, START UP, HOT STANDBY and HOT SHUTDOWN. In COLD SHUTDOWN and REFUELING, primary-to-secondary differential pressure is low, resulting in lower stresses and reduced potential for leakage.

Actions The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the required ACTIONS provide appropriate compensatory actions for each affected SG tube. Complying with the required ACTIONS may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

a. I and a.2 ACTIONS a. l and a.2 apply if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by Surveillance Requirement (SR) 4.4.5.2. An evaluation of SG tube integrity of the affected tube(s) must be made. SG tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, ACTION b applies.

An allowable completion time of seven days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, ACTION a.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 9 of 17 Steam Generator Tube Integrity that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering HOT STANDBY following the next refueling outage or SG inspection. This allowable completion time is acceptable since operation until the next inspection is supported by the operational assessment.

b.

If the requirements and associated allowable completion time of ACTION a are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The allowable completion times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Surveillance Requirements SR 4.4.5.1 During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, "Steam Generator Program Guidelines" (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment detenmines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program detennines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 4 Proposed License Amendment Page 10 of 17 Stcam Generator Tube Integrity performance criteria at the next scheduled inspection. In addition, Specification 6.8.4.1 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SR 4.4.5.2 During a SG inspection any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. The tube repair criteria delineated in Specification 6.8.4.1 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference I provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The frequency of prior to entering HOT SHUTDOWN following a SG tube inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary-to-secondary pressure differential.

Re.erences

1. NEI 97-06, "Steam Generator Program Guidelines"

2. 10 CFR 50 Appendix A, GDC 19

3. 10 CFR 100

4. ASME Boiler and Pressure Vessel Code,Section III, Subsection NB

5. Draft Regulatory Guide 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," August 1976

6. EPRI "Pressurized Water Reactor Steam Generator Examination Guidelines"

7. 10 CFR 50.67

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 11 of 17 Steam Generator Tube Integrity SECTIONNO.: TM.E: TECHNICAL SPECIFICATIONS PAGE:

314.4 BASES ATTACHMENT 6 OF ADM-25.04 7 of 16 REVISIONNO.: REACTOR COOLANT SYSTEM I ST. LUCIE UNIT 1 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314A.6 REACTOR COOLANT SYSTEM LEAKAGE 3A4..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 ndicates Possible pressure boundary degradation. tEiP7AIDNAL 3/4.4.62 REACTOR COOLANT SYSTEM LEAKAGE In try experience has shown that while a limited amount of leakage expe d from the RCS, the unidentified portion of this leakage can be reduce a threshold value of less than 1 GPM. This threshold value is sufficently to ensure early detection of additional leakage.

The 10 GPM ID IFIED LEAKAGE limitation provides allowance for a limited amount of I kage from known sources whose presence will not Interfere with the dete ion of UNIDENTIFIED LEAKAGE by the leakage detection systems.

rNSERT The total steam generator leakage limit of I GPM for all steam 83/4.4.6.2 generators ensures that the do e contribution from the tube leakage will (follows be limited to a small fraction of Pa 100 limits in the event of either a Disert for steam generator tube rupture or stea line break. The I GPM limit is R3/4.4.5) consistent with the assumptions used Ine analysis of these accidents.

PRESSURE BOUNDARY LEAKAGE of any gnltude is unacceptable since It may be Indicative of an impending gros ailure of the pressure boundary. Therefore, the presence of any PRES RE BOUNDARY LEAKAGE requires the unit to be promptly placed In LD SHUTDOWN.

The Surveillance Requirements for RCS Pressure Isolatia Valves provide added assurance of valve Integrity thereby reducing the prob lity of gross valve failure and consequent Intersystem LOCA.

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 12 of 17 Steam Generator Tube Integrity INSERT B3/4.4.6.2 Ba -kground Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS). Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration. The puipose of the RCS operational leakage LCO is to limit system operation in the presence of leakage from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of leakage.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, monitoring reactor coolant leakage into the containment area is necessary. Quickly separating the IDENTIFIED LEAKAGE from the UNIDENTIFIED LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

A limited amount of leakage inside containment is expected from auxiliary systems that carnot be made 100% leaktight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

Anplicable Safety Analyses Primary-to-secondary leakage contaminates the secondary fluid. The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary-to-secondary leakage from all steam generators is I gpm and 0.5 gpm through any one SG as a result of accident induced conditions. The dose consequences of these events are within the limits of GDC 19, 10 CFR 100, 10 CFR 50.67 or the NRC approved licensing basis. The LCO requirement to limit primary-to-secondary leakage through any one steam generator to less thaa or equal to 150 gpd is significantly less than the conditions assumed in the safety analysis.

The RCS operational leakage satisfies Criterion 2 of 10 CFR 50.3 6(c)(2)(ii).

St. Lucie Unit 1 L-2006-089 Dccket No. 50-335 Attachment 4 Proposed License Amendment Page 13 of 17 Steam Generator Tube Integrity Limiting Condition for Operation (LCO)

Reactor Coolant System operational leakage shall be limited to:

a. PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation of the RCPB. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

b. UNIDENTIFED LEAKAGE One gallon per minute (gpm) of UNIDENTIFED LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the leakage is from the pressure boundary.

c. Primary-to-Secondary Leakage Through Any One Steam Generator The limit of 150 gpd per steam generator is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 1). The Steam Generator Program operational leakage performance criterion in NEI 9706 states, "The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gallons per day." The limit is based on operating experience with steam generator tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion is conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

d. IDENTIFIED LEAKAGE Up to 10 gpm of IDENTIFIED LEAKAGE is considered allowable because leakage is from known sources that do not interfere with detection of UNIDENTIFED LEAKAGE and is well with the capability of the Reactor Coolant System Makeup System.

IDENTIFIED LEAKAGE includes leakage to the containment from specifically known and located sources, but does not include PRESSURE BOUNDARY LEAKAGE or controlled reactor coolant pump seal leakoff (a normal function not considered leakage).

Violation of this LCO could result in continued degradation of a component or system.

St. Lucie Unit 1 L-2006-089 DocketNo. 50-335 Attachment 4 Proposed License Amendment Page 14 of 17 Steam Generator Tube Integrity

e. Reactor Coolant System Pressure Isolation Valve Leakage RCS pressure isolation valve leakage is IDENTIFIED LEAKAGE into closed systems connected to the RCS. Isolation valve leakage is usually on the order of drops per minute. Leakage that increases significantly, suggests that something is operationally wrong and corrective action must be taken.

The specified leakage limits for the RCS pressure isolation valves are sufficiently low to ensure early detection of possible in-series check valve failure.

Applicability In POWER OPERATION, START UP, HOT STANDBY and HOT SHUTDOWN, the potential for PRESSURE BOUNDARY LEAKAGE is greatest when the RCS is pressurized.

In COLD SHUTDOWN and REFUELING, leakage limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for leakage.

Actions

a. If any PRESSURE BOUNDARY LEAKAGE exists, or primary-to-secondary leakage is not within limit, the reactor must be brought to HOT STANDBY with 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This ACTION reduces the leakage and also reduces the factors that tend to degrade the pressure boundary.

b. UNIDENTIFIED LEAKAGE or IDENTIFIEI) LEAKAGE in excess of the LCO limits must be reduced to within the limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This allows time to verify leakage rates and either identify UNIDENTIFIED LEAKAGE or reduce leakage to within limits before the reactor must be shut down. This ACTION is necessary to prevent further deterioration of the Reactor Coolant Pressure Boundary.

c. The leakage from any RCS Pressure Isolation Valve is sufficiently low to ensure early detection of possible in-series valve failure. It is apparent that when pressure isolation is provided by two in-series valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. With one or more RCS Pressure Isolation Valves with leakage greater than that allowed by Specification 3.4.6.2.e, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , at least two valves, including check valves, in each high pressure line having a non-functional valve must be closed and remain closed to isolate the affected line(s). In addition, the ACTION statement for the affected system must be followed and the leakage from the remaining Pressure Isolation Valves in each high pressure line having a valve not meeting the criteria of Table 3.4.6-1 shall be recorded daily. If these requirements are not met, the reactor must be brought to at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 15 of 17 Steam Generator Tube Integrity The allowed completion 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. In COLD SHUTDOWN, the pressure stresses acting on the Reactor Coolant Pressure Boundary are much lower, and further deterioration is much less likely.

Surveillance Requirements 4.4.6.2 Verifying Reactor Coolant System leakage to be within the LCO limits ensures the integrity of ihe Reactor Coolant Pressure Boundary is maintained. PRESSURE BOUNDARY LEAKAGE would at first appear as UNIDENTIFIED LEAKAGE and can only be positively identified by inspection. It should be noted that leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE. UNIDENTIFIED LEAKAGE and IDENTIFIED LEAKAGE are determined by performance or a Reactor Coolant System water inventory balance.

a and b.

These SRs demonstrate that the RCS operational leakage is within the LCO limits by monitoring the containment atmosphere gaseous or particulate radioactivity monitor and the containment sump level at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and reactor coolant pump seal injection and return flows). The Surveillance is modified by a note that states that this Surveillance Requirement is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Steady state operations is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS operational leakage determination by water inventory balance, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and Reactor Coolant Pump seal injection and return flows. An early warning of PRESSURE BOUNDARY LEAKAGE or UNIDENTIFIED LEAKAGE is provided by the automatic systems that monitor containment atmosphere radioactivity, containment normal sump inventory and discharge, and reactor head flange leakoff. It should be noted that leakage past seals and gaskets is

St. Lucie Unit 1 L-2006-089 Docket No. 50-335 Attachment 4 Proposed License Amendment Page 16 of 17 Steam Generator Tube Integrity not PRESSURE BOUNDARY LEAKAGE. These leakage detection systems are specified in LCO 3.4.6.1, "Reactor Coolant System Leakage Detection Systems."

The note also states that this SR is not applicable to primary-to-secondary leakage because leakage of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72-hour frequency is a reasonable interval to trend leakage and recognizes the importance of early leakage detection in the prevention of accidents.

d.

This SR demonstrates that the RCS operational leakage is within the LCO limits by monitoring the Reactor Head Flange Leakoff System at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e. end f.

This Surveillance Requirement verifies RCS Pressure Isolation Valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Leakage from the RCS pressure isolation valve is IDENTIFIED LEAKAGE and will be considered as a portion of the allowed limit.

It is apparent that when pressure isolation is provided by two in-series check valves and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity is required. Since these valves are important in preventing overpressurization and rupture of the ECCS low pressure piping, which could result in a LOCA that bypasses containment, these valves should be tested periodically to ensure low probability of gross failure.

Whenever integrity of a pressure isolation valve listed in Table 3.4.6-1 cannot be demonstrated the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily. In addition, the position of one other valve located in each high pressure line having a leaking valve shall be recorded daily.

This Surveillance Requirement verifies that primary-to-secondary leakage is less than or equal to 150 gpd through any one steam generator. Satisfying the primary-to-secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this Surveillance Requirement is not met, compliance with LCO 3.4.5 should be evaluated. The 150-gpd limit is measured at room temperature as described in Reference 1. The operational leakage rate limit applies to leakage through any one steam generator. If it is not practical to assign the leakage to an individual steam

St. Lucie Unit 1 L-2006-089 Dc'cketNo. 50-335 Attachment 4 Proposed License Amendment Page 17 of 17 Steam Generator Tube Integrity generator, all the primary-to-secondary leakage should be conservatively assumed to be from one steam generator.

The Surveillance Requirement is modified by a note, which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

For Reactor Coolant System primary-to-secondary leakage determination, steady state is defined as stable Reactor Coolant System pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and reactor coolant pump seal injection and return flows.

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary-to-secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref.2).

Re:Ferences

1. NEI 97-06, "Steam Generator Program Guidelines"

2. EPRI "PWR Primary-to-Secondary Leak Guidelines"

3. UFSAR, Section 15.4.4

L-2006-089, Proposed License Amendment Steam Generator Tube Integrity

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