Text

Proposed License Amendment Steam Generator Tube Integrity Pursuant to 10 CFR 50.90
	ML061510346
Person / Time
Site:	Saint Lucie
Issue date:	05/25/2006
From:	Johnston G; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2006-094
	Download: ML061510346 (81)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 May 25, 2006 FPL L-2006-094 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 RE: St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Steam Generator Tube Integrity Pursuant to 10 CFR 50.90, Florida Power and Light Company (FPL) requests to amend Facility Operating License NPF-16 for St. Lucie Unit 2.

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 informational markup of the TS Bases.

FPL requests that the proposed amendment be processed normally 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.

an FPL Group company

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 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 Ai'4% Z$, ZC, Very truly yours, Gordon L. Johnston Acting Vice President St. Lucie Plant GLJ/KWF Attachments cc: Mr. William A. Passetti, Florida Department of Health

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 1 of 11 Steam Generator Tube Integrity Evaluation of Proposed Change

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 2 of 11 Steam Generator Tube Integrity INTRODUCTION Pursuant to 10 CFR 50.90, Florida Power and Light Company (FPL) requests to amend Facility Operating License NPF-16 for St. Lucie Unit 2. This proposed license amendment request (LAR) revises the requirements in the St. Lucie Unit 2 Technical Specification (TS) related to steam generator tube integrity and Reactor Coolant System Operational Leakage.

The change is based on the NRC approved Revision 4 to Industry/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).

St. Lucie Unit 2 steam generators (SGs) are scheduled for replacement in the fall of 2007.

Therefore, this request addresses original and replacement SG tubing material with respect to inspection requirements.

DESCRIPTION OF PROPOSED AMENDMENT Based on the NRC-approved Revision 4 of TSTF-449, the proposed changes include:

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 "Steam Generator Tube Integrity" requirements.

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

Add new TS 6.8.4.1, "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 NRC'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 proposed in the TS Bases associated with TSTF-449, Rev. 4 is taken only when necessary to maintain consistency with the St. Lucie Unit 2 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.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 3 of 11 Steam Generator Tube Integrity The 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 2 TS.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 4 of 11 Steam Generator Tube Integrity 3.4.13d Operational primary-to- <0.3 GPM total through all SGs and 3.4.6.2.c RCS Operational Leakage TS < 150 gallons per day secondary leakage <216 gallons per day through any one through any one SG (room temperature).

SG (accident conditions).

3.4. r3- RCS primary-to- Reduce leakage rate to within limits 3.4.6.2 RCS Operational Leakage, secondary 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 3 n g hours.

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . the 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 4.4.6.2.l.c Relocate information to footnote and revise to state:

determined by water inventory balance. 3.4.6.1, "Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after inventory balance ACTION a. establishment of steady state operation. Not

.......... and b. applicable to primary to secondary leakage." Add an b...... 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 4.4.6.2.1 e 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 3/4.4.6 TS - Contains primary-to-secondary leakage limit.

actions included in the Tech Specs. SG tube integrity requirements and ACTIONS required upon failure to meet performance criteria.

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

criteria. III

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 5 of 11 Steam Generator Tube Integrity 3.4.13d Performance criteria <0.3 GPM total through all SGs and 3.4.6.2.c RCS Operational leakage TS - Operational leakage

<216 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 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 6 of 11 Steam Generator Tube Integrity 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 technioues capable of findin2 the degradation.

5.5.9 Inspection scope From the point of entry (hot leg side) 6.8.4.1 TS 6.8.4.1 Steam 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 around 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 6.8.4.1 TS 6.8.4.1 -Requirements unchanged.

previous approval by the NRC.

Approved methods listed in TS.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 7 of 11 Steam Generator Tube Integriy Plugging and repair report required 15 CFK - Senous S(U tube degradation (i.e., tubing tail to meet the structural integrity or accident induced days after each inservice inspection, 12 leakage criteria) requires reporting in accordance month report documenting inspection with 50.72 or 50.73.

results, and reports in accordance with

§50.72 when the inspection results fall TS 6.9.1.12 - 180 days after the initial entry into into category C-3.

HOT SHUTDOWN after performing a SG inspection Definitions Definitions SG Normal TS definitions (i.e., Definitions Definitions TS 6.8.4.1, TS Bases, Steam Generator Program Terminology 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-I _to-secondary leakage."

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 8 of 11 Steam Generator Tube Intemitv REGULATORY REOUIREMENTS 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 part of the CLIP 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 2 and justify this amendment for the incorporation of the changes to the St. Lucie TS considering the differences described in the precedent section below. 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 2 TS and to maintain consistency with the St. Lucie Unit 2 licensing basis.

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.

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

Plant Name, Unit No. St. Lucie Unit 2 Steam Generator Model(s): Original / CE Model 3410 / AREVA 86/19T (See Table Replacement Note 1)

Approximate Effective Full Power Years (EFPY) of service for currently installed 19.3 EFPY at End of Cycle 15 (April 2006)

SGs Tubing Material: Original / Replacement Alloy 600 Mill Annealed / Alloy 690 Thermally Treated Number of tubes per SG: Original / 8411/8999 Replacement Number and percentage of tubes plugged in 2A 1469 (17.5%)

each Original SG as of Cycle 15 2B 1709 (20.3%)

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 9 of 11 Steam Generator Tube Integrity Number of tubes repaired in each SG None as of Cycle 15 Degradation mechanism(s) identified (See Table Note 2)

Current primary-to-secondary leakage Per SG: 216 gpd through any one SG (operating limits: temp.)

Total: 0.3 gpm total through SGs (operating temp.)

Approved Alternate Tube Repair Criteria: (Ref. 1)

Approved SG Tube Repair Methods (Ref. 2)

Performance criteria for accident leakage 0.3 gpm total through steam generators and 216 gpd through any one SG (operating temp.)

Table Notes:

1. St. Lucie Unit 2 original SGs are scheduled for replacement at end of Cycle 16 in the fall of 2007.

2. At the end of the Cycle 15 inspection, degradation mechanisms included mechanical wear, axial outer diameter stress corrosion cracking (ODSCC),

circumferential ODSCC, axial inner diameter stress corrosion cracking (IDSCC),

circumferential IDSCC and volumetric ODSCC.

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 2 considering the differences described in the precedent section below. Therefore, the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91(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 2 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 2 TS or to maintain consistency with the St. Lucie Unit 2 licensing basis and provide conforming changes necessary to address pending NRC approval of Reference 4:

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 10 of I1 Steam Generator Tube Integrity

1. The current format and terminology used in the St. Lucie Unit 2 TS 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 is 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.

3. Proposed TS LCO 3.4.6.2.c, "Reactor Coolant System operational leakage" limits normal operating primary-to-secondary leakage to 150 gpd through any one SG consistent with TSTF-449, Rev. 4. The leakage limit in the St. Lucie Unit 2 accident analysis, however, has been reduced to 216 gpd through any one SG by license amendment 138. Therefore, St. Lucie Unit 2 will administratively limit primary-to-secondary leakage to half the value assumed in the accident analysis to ensure that the margin is consistent with the Staff's expectations as discussed in the model SE for TSTF-449, Rev. 4. A similar approach to further limit operational leakage administratively was approved by the NRC for Calvert Cliffs Units 1 and 2 (Reference 3 below).

4. The St. Lucie Unit 2 SGs are scheduled for replacement in the fall of 2007. As a result, this LAR includes new TS 6.8.4. I., "Steam Generator Program", which provides requirements for the original and replacement SG designs. This TS requires a Steam Generator Program to be established and implemented to ensure that SG tube integrity is maintained, and to describe SG condition monitoring, performance criteria, repair methods, repair criteria, and inspection intervals that are applicable to the original SG and replacement SG designs. TS 6.8.4.1.1. applies to the replacement SG design. TS 6.8.4.1.2.

applies to the original SGs and contains requirements such as a sleeving repair method, alternate repair criteria and additional inspection requirements, which are unique to the original SG design and can be removed following SG replacement.

5. The original SG surveillance reporting criteria under 4.4.5.5.c is relocated to new TS 6.9.1.13 as a conforming change. This requirement was an NRC commitment made for the original SGs and requires reporting of indications in the tubesheet region of the original SGs within 120 days after the initial entry into HOT SHUTDOWN. TS 6.9.1.13 is only applicable to the original SGs and can be removed following SG replacement.

6. NRC approved FPL license amendments for the original St. Lucie Unit 2 SGs to revise inspection and plugging requirements in the tubesheet region (St. Lucie Unit 2 TS amendment 143) and allow installation of leak limiting sleeves (St. Lucie Unit 2 TS amendment 144). Both requests contain analyses that predict leakage within allowable limits under postulated accident conditions. To ensure adequate margin is maintained between operating leakage and postulated accident leakage limits, FPL will further limit

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 11 of 11 Steam Generator Tube Integrity operating leakage to half of the accident analysis value (as discussed in item 3 above), less the amounts of postulated accident leakage. Postulated leakage includes leakage from sleeves, the tubesheet region and other sources as defined in the operational assessment.

If the resulting value is less than the operating procedure shutdown requirement of 75 gpd through any one SG, the procedure will be revised to the lower value for the subsequent operating period. This additional administrative control will be applicable for the remaining life of the original SGs.

REFERENCES

1. FPL letter L-2004-245 dated November 8, 2004, "St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment Define the Depth of the Required Tube Inspections and Clarify the Plugging Criteria Within the Tubesheet Region of the Original Steam Generators" as approved in TS amendment 143.

2. FPL letter L-2004-233 dated January 6, 2005, "St. Lucie Unit 2 Docket No. 50-389 Proposed License Amendment, Add Steam Generator Repair Method, Westinghouse Electric LLC Alloy 800 Leak Limiting Sleeves" as approved in TS amendment 144.

3. NRC Safety Evaluation Related to Amendment No. 278 and 255 for Calvert Cliffs Nuclear Power Plant Unit Nos. 1 and 2, March 9, 2006.

4. FPL letter L-2005-210 dated October 21, 2005, "St. Lucie Unit 2 Docket No. 50-389, Proposed License Amendment, Reduced Reactor System Coolant Flow With a Reduction in Reactor Operating Power," (as supplemented by FPL letter L-2006-086, dated March 28, 2006).

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 1 of 27 Steam Generator Tube Integrity Technical Specification Markups TS Page VI TS Page XIX TS Page 1-3 TS Page 1-5 TS Page 3/4 4-11 TS Page 3/4 4-12 TS Page 3/4 4-12a TS Page 3/4 4-13 TS Page 3/4 4-14 TS Page 3/4 4-14a TS Page 3/4 4-15 TS Page 3/4 4-16 TS Page 3/4 4-17 TS Page 3/4 4-18 TS Page 3/4 4-19 TS Page 3/4 4-20 TS Page 6-15e TS Page 6-20e

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 2 of 27 Steam Generator Tube Integrity INDEX LIMITING CONDITIONRS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES DELETED .............................................................................................. 3/4 4-7 OPERATING ......................................................................................... 3/4 4-8 3/4.4.3 PRESSURIZER ............................................................................................... 314 4-9 3/4.4.4 PORRV BLOCKVALVES E .. .. .... .................... i ............ 3/4 4-10 314.4.5 STEAM GENERATO S . ....... ....................... 3/4 4-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ...................................................... 3/44-18 OPERATIONAL LEAKAGE ................................................................... 3144-19 3,4.4.7 CHEMISTRY ................................................................................................. 3/4 4-22 3/4.4.8 SPECIFIC ACTMTY ..................................................................................... 3/4 4-25 3,4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM ................................................................. 3/4 4-29 PRESSURIZER HEATUP/COOLDOWN UMITS ............................................ 3/4 4-34 OVERPRESSURE PROTECTION SYSTEMS ............................................... 3/4 4-35 3/4.4.10. REACTOR COOLANT SYSTEM VENTS....................................................... 314 4-38 314A.11 STRUCTURAL INTEGRITY ............................. ................................. 3144-39 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 - Tan > 325"F . 3/45-3 3/4.5.3 ECCS SUBSYSTEMS -TTa - I-. .o.. .... o .o......o......o.... ............... - e.... 3/45-7 .

'JIA

O 3/4.5.4 REFUELING WATER TANK., ... ........................ ,* "1"

.Amendment No. 4Cj4)

ST. LUClE -UNIT 2 V?

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 3 of 27 Steam Generator Tube Intelzrit-INDEX ADMINISTRATIVE CONTROLS PAGE

-9, SECTION 6.6 REPORTABLE EVENT ACTION ........................................................................... 6-13 tii 6.7 SAFETY LIMIT VIOLATION ................................................................................. 6-13 6.8 PROCEDURES AND PROGRAMS.................................................................... 6-13 6,9 REPORTING REQUIREMENTS ......................................................................... 6-16 6.9.1 ROUTINE REPORTS ......................................................................................... 6-16 STARTUP REPORT ............................................................................................ 6-16 ANNUAL REPORTS .............................. .6-16 MONTHLY OPERATING REPORTS ..... ..... ............ . ............. 6-17 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT .................................. 6-18 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT ...... 6-19 CORE OPERATING UMITS REPORT (COLR) ..................................................... 6-20 6.9.2 SPECIAL REPORTS .............. .................................................... 6-20e 6.10 .1 RADIATION PROTECTION PRO M ............................... 6-2

......... . CTO RD IATONPRTE ......... '.:,,. .......................................... :.................... 6-201 ITEAM -ENEATORTUBE MISPEC"ION REPORT 6-20e ST. LUCIE - UNrT 2 Amendment No. 4,3, -. £9.02,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 4 of 27 Steam Generator Tube Integrity DEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcurieslgram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132,1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, pages 192-212, Tables entitled, *Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity (Sv/Bq)."

E-AVERAGE DISINTEGRATION ENERGY 1.11, shall be the average (weighted in proportion to the concentration of each radionudlide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than lodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment Is capable of performing its safety function (Le., the valves travel to their required positions, pump discharge pressures reach their required J values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION 1.13 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined InTable 1.1.

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

IDENTIFIED 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.

S..FLUCIE-UNIT2 -3 " ent No. 4K,1vý

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 5 of 27 Steam Generator Tube Integrity DEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (excep en~erator 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 packaging 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 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.

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

REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power to the CEA drive mechanism is Interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRrlY 1.28 SHIELD BUILDING INTEGRITY shall exist when:

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

b. The shield building ventilation system is in compliance with Specification 3.6.6.1, and

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

ST. LUCIE- UNIT2 1-5 Amendment No. 9, 43.41.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 6 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 4A.5 STEAM ENERATO STUBE UMmNG CONDTON FOR OPERATION 3.4.5 Each stea g 9 L IE.~ ~ iI APPLICABILITY: MODES 1, 2,3 and 4.

an nprable, restore the Inoperable INER B 0,u* t.

]Wiln~~~~~ ptmtr generator(s) to OPERABLE status prior to Increasing laewbg 6L3*:L .

SURVEILLANCE REQUIREMENTS OPERABLE by performance o

. W.here ex.once samIn arnSetso- whnc insect r ache genemtr sail determined OPERscritiareastobe selcting and inspecting at eatthemnmumtnumber ofIsteam generatom spese* criticalar 4.4.5.2 Thteam Gfirsto ample t Inspection size, eseel result dassification, M ~and'the geieaor npetio th-4I oepnding ns s IIbesample tube, minim selectedoarao eac orisecp isc-pect oam acin rekied shall be as specified In Table 4.4-2. The Ins 'ce Inspection ofteam generator tubes shall be performed at the Jfreque 'a~s specified in Spe 'ton4.4.5.3 and the Inspected tubes shall lbe verifle. c-eptable per the a

criteria of Specification 4.4.5.4.

The tubes -d for eachInsev ction shall Include at ]east 3% of the l total numberi tubes in a Isteam on gedfo arand tubes selected for

theIsexcept.

thIese Isetin lb ecl ,

a- Whr neinsmlrpats w - water chemistr a*zr

'*_ i (subse:*quent to the preservice Inspection o tem generator/:

shall inclde:

S eparat"e Action entry is allowed for each SG tube. :

/

ST. LUCIE -UNIT 2 3/4 4-11

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 7 of 27 Steam Generator Tube Integritv 314.4.5 INSERT A SG tube integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired 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 or repaired 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 or repair 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 .

314.4.5 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 or repaired in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 8 of 27 Steam Generator Tube Integri THIS PAGE DELETED REACTOR COOLANT SYSTEM

1. All nonplugged tubes that previously had detectable wall penetrations (greater than 20%).

2. Tubes in those areas where experience has indicated I problems.

3. A tube inspection (pursuant to Specification 4.4.5.4.a.

be performed on each selected tube. If any selected, not permit the passage of the eddy current probe for i Inspection, this shall be recorded and an adjacent 1:90 !shall

%be selected and subjected to a tube Inspection. f C.

4. rvice Leak Limiting Alloy 800 sleeves Ilength during each refueling outage both the tube and the sleeve.

I as the second and i

6mples (ifrequired by I

Leach Inservice insp may be subjected to

1. The tubes of the tubes where one of the Categtor C-.1 '..ess than 5% of the total tubes It are degraded tubes and none of tubes are defective.

One or more tubes, but not more than 1%

the total tubes Inspected are defective, or between 5% end 10% of the total tubes im are degraded tubes.

THIS PAGE DELETED ST. LUCIE.-UNIT 2 S3/4 4-12 Amendment No. 24,48,144

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 9 of 27 Steam Generator Tube Integrity ST. LUCIE -UNIT 2 3/4 4-128 34 No. 24) 412aAMendment ST.L~cI.UNT2

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 10 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM S**RVILLANE REQ IREENTS LContinujed 4.4.6. Inspection Freauencles - The above required inservice Inspections of steam g nrator tubes shall be performed at the following frequencies:

The first Inservice Inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of Initial crit-lmly. Subsequent Inservice inspections shall be performed at I rvals of not less than 2 nor more than24 calendar mon fter the revious Inspection. If two consecutive Inspections foll ng servi under AVT (all volatile treatment) conditions, not I uding the p rvice inspection, result In all inspection results lling Into the category or Iftwo consecutive Inspections monstrate that prevyo observed degradation has not contin and no addi-tional degrad 'on has occurred, the Inspection in rval may be extended to a xlmum of once per 40 months.

b. Iffthe results or the Irvce inspection of a earn generator conducted In accorda with Table 4.4- t 40-month Intervals fall Into Category C-3, the I Ion frequ shall be Increased to at least once per 20 months. The I a In Inspection frequency shall apply until the subseq nt ins ctions satisfy the criteria of Specification 4.4.5.3a.; the It may then be extended to a maximumn of once per 40 C. Additional, unscheduled 1 ce I ions shall be performed on each steam generator In ance the first sample inspection specified In Tablee 4. uring t shut we subsequent to any of the following condii

1. Prim ndary tubes leaks (not in uding leaks originati m tube-to-tube sheet welds) exce s of the limits opecfication 3..6.2.

2- As ismic occurrence greater than the Operating sis

3. A loss-of-coolant accident requiring actuation of thle Engineered Safety Features.,

4. Amain steam line or feedwater line break.

THIS PAGE OELETEt ST. LUCIE - TUNT3 2 314 4-13

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 11 of 27 Steam Generator Tube Intem-itv

~THIS PAGE DELETED/

REACTOR COOLANT SYSTEM/

'*UVEILANE REQUIREMENT fContinuedlj

a. As used Inthis Specification

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings specifications. Eddy-current testing Indications below 200of the nominal tube wall thickness, Ifdetectable, may be sdared as imperfections.

2. radation means a service-induced cracking, tage, wear or eral corrosion occurring on either inside or o ide of a 3.tts rae greater

3. sctons caused by

4. %1Dc ý of the tube wall thickness affec

5. Dhfep f such severity that it exceeds the p fning a defect is defective.

Pltgaing or Re r Lim ns the condition at or beyond which the tube shall be removed fron rvi y plugging or repaired by sleeving using the method In Sp* tion 4*4. 4.a.10 in the affected area. The plugging

. In the n -leeved portion of a e, the plugging or repair limit imperf 'on dept is 40% of then inal wall thickness. This Limit is no pplicable In the portion of the be that is greater than 10 inches below the bottom of theh leg expansion transition or t of th tubesheet (wWichever Is lowr the tube end.

j~e ** lO3 inch )low the bottom of the gradation detected. between 10.3 nh lwte otmo6h hot leg expansion transition or top of the tu heat (whichever is lower) and the bottom of the Wilg expans! naition ortop of the tubesheet (whichever Is higher) shall be plu or repaired on detection.

ii. In the region of the tube sleeved using aWestIntghou ak Unfiting loy 800 sleeve, the tube shall be plugged up etection of any service induced Imperfection, degradation or de the,

"(a) sleeve or () pressure beundary portion of the original tu wall Inthe sieeveltube assembly (i.e., the sleeve-to-tube joint).

lii. All Leak Umiting Alloy 800 Sleeves that have a nickel band shall plugged or removed from service after one cycle in operation.

2 ST. LUCIE - UNT- 4-4 3M4 Amendment No. 144 IQlODAMM.I r11r_1: M-Trcr%

TI I I II I* F/'IU* I,/.LL.F.. 1 IF.I.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 12 of 27 Steam Generator Tube Integrity

\~ ~~~4 RECO-OLATSSE AM MKM- Lisd .9 THIS PAGE DELETED

/

7. Unserviceable describes the condition of a tube if it leaks or contains a defect large enough to affect its structural integrity In the event of an Operating Basis Earthquake, a loss-of-coola accident, or a steam line or feedwater line break as specifie in 4.4.5.3c., above.

Tube Inspection fora tube with no portion of a sleeve e ending below 10.3 Inches from the bottom of the hot leg expansion ansilton or the top of the tubesheet (whichever Is lower) means an ins ction of the steam nerator tube from 10.3 inches below the bottof the hot leg expansion trsitlon or top of the tubesheet (whichever is er) completely around the -bend to the top support of the cold leg. be Inspection for a tube with rtion of a sleeve extending below 1 . inches from the bottom of the hot expansion transition or the top the tubesheet (whichever Is lower) ns an Inspection from the boim of the sleeve completely around the nd to the top support cold leg.

9. Preservc s means an r ection of the full length of each tube In ea team generat performed by eddy current techniques prior to rvice to blish a baseline condition of the tubi This I pection shall be performed after the field hydrosta te and prior to Initial POWER OPERATION using the ipment and techniques expected to be used durng subsequen rvice Inspections.

THIS PAGE DELETED ST. LUCIE -UNIT 2a K44-4414a Amendment No. 144

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 13 of 27 Steam Generator Tube Integrity

R COOLANT SYSTEM REACTOR _/ THIS PAGE DELETED

10. Tube Reneir refers to sleeving with Westinghouse Leak Limiting oy 80o sleeves as described in WCAP-15918-P Revision 2, which are ed to maintain a tube in service. Leak Limiting Alloy 800 Sleeves a applicable only to the original steam generators. The press boundary portion of the original tube wall in the sleeve/tube assemb I.e., the sleeve-to-tube joint) shall be inspected prior to installati of each sleeve.

b. e steam generator shall be determined OPERABLE aft completing th rrespondlng actions (plug or repair all tubes excee g the Plugging or Re **Limit and all tubes containing through-wall cra required by Table 42 4.4.5.5 Reports

a. Within 1.5 da following the completion of inservice inspection of steam gene tubes, the number of tu plugged or repaired In each steam generator all be reported to the mission in a Special Report pursuant to Specif tion 6.9.2.

b. The complete results o steam g rator tube Inservice Inspection shall be submitted to the mmtssi in a Special Report pursuant to Specification 6.9.2 within mon following completion of the Inspection. This Special Re rt Itindude:

1. Number and extent of and sleeves inspected.

MODIFIED AS SHOWN LATER AND 2. Location and perceof wall- aess penetration for each ELOCATEDOTo indication of an i ection.

6.9.1.13 3. Identification oubes plugged or re fred.

... f i I

C. Following each i con and within 120 da after the reactor coolant system reenters MODV4, the following Information co eming indications found in the tubesheet 05n (including the expansion tran n) shall be reported to the Commissioin a special report pursuant to S tion 6.9.2. This Special Repors include:

1. mber of total indications, location of each indic n, orientation of each dication, severity of each indication, and whether Indications initiated from the inside or outside diameter.

The cumulative number of Indications detected In the tu t region as a function of elevation within the tubesheet.

3. Projected end-of-cycle accident inducted leakage from tubesh indications, This leakage shall be combined with the postulated el-of-cycle accident Induced leakage from all other sources. Ifthe prell ary estimated total projected end-of-cycle accident Induced leakage from 11 sources exceeds the leakage limit, the NRC staff shall be notified prior Unittrestart.

ST. LucIE- -UNrT2 314 4-15 Amendment No.-4a. 443.144 THIS PAGE DELETED

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 14 of 27 Steam Generator Tube Integrity z

Lu 6

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 15 of 27 Steam Generator Tube Integrity I

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 16 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 314.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMmNG 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. [lanceIa ACTIONUlr~luirenmt 4.4.6.2.1 c

a. With the required reactor cavity sump in flow itoring sVtem inoperable, perform a RCS water inventory balance t least er 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and restore the sump inlet flow monitoring system to OPERABLE s tus within 30 days; otherwise, be In at least HOT STANDBY within the next ours 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 pies of the containment atmosphere or perform a ROB water Inventory balanto at least once per 24hours, and restore the required radioactivity monitor OPERABLstatus within 30 days; otherwise, be In at least HOT STANDBY within the nd~xt 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.

l SURVEILLANCE REQUIREMENTS d ....

4.4.6.1 The ROB leakage detection instruments shall be demonstrated OPERABLE bv:

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.

I* 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. I ST. LUCIE - UNIT 2 3/4 4-18 Amndmert N."N_ý

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 17 of 27 Steam Generator Tube Integritv REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE gMITMpJ rNnrntTn*ti Po nP:RAnrltJI 3.4.6.2 Reactor Coolant Systemrreage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

c. total riary-to-secondar aq. e througm -'

generatr** and 2"6-gallons per day throug anyone steam genertort

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

e. I gpm leakage (except as noted in Table 3.4-1) at a Reactor Coolant System pressure of 2235 + 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

APPLICABILITY: MODES 1, 2,3, and 4. r with *m -t-so ry leakag not within limit, ACTION:/

a. With any PRESSURE BOUNDARY LEAKAGE, be Inat least HOT STANDBY

__l__t__ Lwithin 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 .

a ~er'otional

b. With any Reactr Coolant Systac leakage greater t0W any one of the rimry-to limits, excudindiPRESSURE BOUNDARY LEAKA(Ond leakage from Reactor leakage, Coolant System Pressure Isolation Valves, 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 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 any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, Isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two dosed manual or deactivated automatic valves, or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With RCS leakage alarmed and confirmed In a flow path with no flow Indication, commence an RCS water inventory balance within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to determine the leak rate.

.qIP1 if I R If, 4.4.62.1 Reactor Coolant Systemleakages shall be demonstrated to be within each of the above limits by:

loseo tionall

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

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 .

ST. LUCIE- UNIT 2 3144-19 A dmdmenlto "L

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 18 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM

.URVEILLANCE REQUIREMENTS (Continued)

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 .

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 .

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve check valve specified In Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit:

a. At least once per 18 months,

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

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

d. Following valve actuation due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

4.4.6.2.3 Each Reactor Coolant System Pressure Isolation Valve motor-operated valve specified In Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within its limit;

a. At least once per 18 months, and

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

The provisions of Specification 4.0.4 are not applicable for entry Into MODE 3 or 4.

Fe Primary-to-econdary leakage shall be verified~150 gallons per 0 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 I0 1state operation. NWt applicable to primary-to-secondary leakogeil1 Nat 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 II state operation.

ST. LUCIE-UNI *172 314 4-20 AmendmentNo.T.(,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 19 of 27 Steam Generator Tube Intearitv ADMINISTRATIVE CONTROLS (continued)

It. Ventilation Filter Testing Prooram (VFTP) (continued)

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 specified below.

ESF Ventilation System Dena p Flowrate

< 7.4* W.G. 2000 +/- 200 cfm Control Room Emergency Air Cleanup The provisions of SR 4.02 and SR 4.0.3 are applicable to the VFTP test frequencies.

FORSERT4.

ST. LUCIE - UNIT 2 6-150 Amendment NoPs"

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 20 of 27 Steam Generator Tube Integrity INSERT FOR 6.8.4.1:

Steam Generator (SG) Program

1. A Replacement Steam Generator Program shall be established and implemented for the replacement SGs to ensure that SG tube integrity is maintained. In addition, the Replacement 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.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 21 of 27 Steam Generator Tube Integrty

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 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 0.3 gpm total through SGs and 216 gallons per day 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

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 22 of 27 Steam Generator Tube Integrity 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.

2. An Original Steam Generator Program shall be established and implemented for the original SGs to ensure that SG tube integrity is maintained. In addition, the Original 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 a SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected, plugged or repaired 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

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 23 of 27 Steam Generator Tube Integrity 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 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 0.3 gpm total through SGs and 216 gallons per day 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 or repaired.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. Plugging or Repair Limit means the condition at or beyond which the tube shall be removed from service by plugging or repaired by sleeving using the method in Specification 6.8.4.1.2.f. 1. in the affected area. The plugging or repair limits are as follows:

i. In the non-sleeved portion of a tube, the plugging or repair limit imperfection depth is 40% of the nominal wall thickness. This Limit is not applicable in the portion of the tube that is greater than 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) to the tube end.

Degradation detected between 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) and the bottom of the hot leg expansion transition or top of the tubesheet (whichever is higher) shall be plugged or repaired on detection.

ii In the region of a tube sleeved using a Westinghouse Leak Limiting Alloy 800 sleeve, the tube shall be plugged upon detection of any service induced imperfection, degradation or defect in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 24 of 27 Steam Generator Tube Integrity iii. All Leak Limiting Alloy 800 Sleeves that have a nickel band shall be plugged or removed from service after one cycle in operation.

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, and that may satisfy the applicable tube repair criteria. In addition to meeting the requirements of Tube Inspection. d.1, d.2, d.3 and d.4 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.

Tube Inspection for a tube with no portion of a sleeve extending below 10.3 inches from the bottom of the hot leg expansion transition or the top of the tubesheet (whichever is lower) means an inspection of the steam generator tube from 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) to the tube-to-tubesheet weld at the tube outlet. Tube Inspection for a tube with a portion of a sleeve extending below 10.3 inches from the bottom of the hot leg expansion transition or the top of the tubesheet (whichever is lower) means an inspection from the bottom of the sleeve to the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not part of the tube.

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 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3. Inspect all Inservice Leak Limiting Alloy 800 sleeves over their full length during each refueling outage. These inspections will include both the tube and the sleeve.

4. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 25 of 27 Steam Generator Tube Integrity 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.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

1. Tube Repair refers to sleeving with Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-15918-P Revision 2 (with range of conditions as revised in Appendix A of WCAP-16489-NP, Revision 0),

which are used to maintain a tube in service. Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint) shall be inspected prior to installation of each sleeve.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 26 of 27 Steam Generator Tube Intemritv ADMINISTRATIVE CONTROLS Wcontinuall' CORE OPERATING LIMITS REPORT (COLRI (continued)

b. (continued)

61. WCAP-41397-P-A, (Proprietary), 'Revised Thermal Design Procedure,"

April1989.

62. WCAP-14565-P-A, (Proprietary), "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.

63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-O1 Code,* May 2003.

64. 30% SGTP PLA Submittal and the SER.

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 RT 6.9. MARGIN, transient analysis limits, and accident analysis limits) of the safety I -nd 6.9.1.13 analysis are met.

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

SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the NRC within the time period specified for each reporL 6.10 DELETED ST. LUCIE - UNIT 2 6-20e STLuI~urr -2eAmendxment No. 405, 449,3 )?

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 27 of 27 Steam Generator Tube Integrity INSERT 6.9.1.12 and 6.9.1.13:

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,

h. The effective plugging percentage for all plugging and tube repairs in each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

6.9.1.13 Following each inspection of the Original Steam Generators performed in accordance with Specification 6.8.4.1., and within 120 days after the initial entry into HOT SHUTDOWN, the following information concerning indications found in the tubesheet region (including the expansion transition) shall be reported to the Commission. The report shall include:

a. Number of total indications, location of each indication, orientation of each indication, severity of each indication, and whether the indications initiated from the inside or outside diameter.

b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. Projected end-of-cycle accident induced leakage from tubesheet indications. This leakage shall be combined with the postulated end-of-cycle accident induced leakage from all other sources. If the preliminary estimated total projected end-of-cycle accident induced leakage from all sources exceeds the leakage limit, the NRC staff shall be notified prior to Unit restart.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 1 of 24 Steam Generator Tube Integrity Word-Processed Technical Specifications TS Page VI TS Page XIX TS Page 1-3 TS Page 1-5 TS Page 3/4 4-11 TS Page 3/4 4-12 TS Page 3/4 4-12a TS Page 3/4 4-13 TS Page 3/4 4-14 TS Page 3/4 4-14a TS Page 3/4 4-15 TS Page 3/4 4-16 TS Page 3/4 4-17 TS Page 3/4 4-18 TS Page 3/4 4-19 TS Page 3/4 4-20 TS Page 6-15e TS Page 6-15f TS Page 6-15g TS Page 6-15h TS Page 6-15i TS Page 6-20e TS Page 6-20f

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 2 of 24 Steam Generator Tube Integritv INDEX

.,MTNG CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 314A.2 SAFETY VALVES DELETED ............................................... 3/44-7 OPERATING ........................................................................................... 3144-8 314.4.3 PRESSURIZER ................................................................................................. 3/4 4-9 3/4.4.4 PORV BLOCK VALVES .................................................................................. 3/4 4-10 314.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY ............ 3/44-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ............ ........................................... 3/44-18 OPERATIONAL LEAKAGE ................................... 3/44-19 314.4.7 CHEMISTRY ............................................................................................... 3/44-22 3/4.4.8 SPECIFIC ACTIVITY ..................................................................................... 3144-25 3/4A.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM ..................................................................... 3144-29 PRESSURIZER HEATUP/COOLDOWN LIMITS .............................................. 3/44-34 OVERPRESSURE PROTECTION SYSTEMS ................................................. 3/4 4-35 3/4A.10 REACTOR COOLANT SYSTEM VENTS ........ ............................................... 3/44-38 314.4.11 STRUCTURAL INTEGRITY ... ............................ ....... I........ 3.......4........

3144-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 314.5.1 SAFETY INJECTION TANKS ............................. ................................. .3/4 5-1 0

314.5.2 ECCS SUBSYSTEMS - Tavg > 325 F ...................... .................. ....................3/4 5-3 314.5.3 ECGS SUBSYSTEMS -wT < 325-F ........................ ..................................... .3145-7 3/4.5.4 REFUELING WATER TANK.. ............................... ..................................... .3/4 5-8 ST. LUCIE -UNIT 2 vi Amendment No. 46,440,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 3 of 24 Steam Generator Tube Integrity IN13EX ADMINISTRATIVE CONTROLS SECTION PACE 6.6 REPORTABLE EVENT ACTION ........................................................................... 6-13 6.7 SAFETY LIMIT VIOLATION .................................................................................. 6-13 6.8 PROCEDURES AND PROGRAMS .................................... 6-13 6.9 REPORTING REQUIREMENTS ........................................................................... 6-16 6.9.1 ROUTINE REPORTS ............................................................................................ 6-16 STARTUP REPORT .............................................................................................. 6-16 ANNUAL REPORTS ............................................................................................ 6-16 MONTHLY OPERATING REPORTS .................................................................... 6-17 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT .................................. 6-18 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT .............. 6-19 CORE OPERATING LIMITS REPORT (COLR) ..................................................... 6-20 STEAM GENERATOR TUBE INSPECTION REPORT ....................................... 6-20e 6.9.2 SPECIAL REPORTS .......................................................................................... 6-20e 6.10 DELETED.... ...................................................................................................... 6-20e 6.11 RADIATION PROTECTION PROGRAM ............................................................... 6-21 ST. LUCIE - UNIT 2 XlX Amendment No.43,6-, 43 , 02.

409, 448,433.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 4 of 24 Steam Generator Tube Integrity DEFINTONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (mlcrocurles/gram) which alone would produce the same thyroid dose as the quantity and Isotopic mixture of 1-131, 1-132, 1-133, 1-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in ICRP-30, Supplement to Part 1, pages 192-212. Tables entitled, "Committed Dose Equivalent In Target Organs or Tissues per Intake of Unit Activity (Sv/Bq)."

- AVERAGE DISINTEGRATION ENERGY

,1.11 f shall be the average (weighted in proportion to the concentration of each radionuclide In the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (inMeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant

.ENGINEERED SAFETY FEATURES RESPONSE TIME 1.12 The ENGINEERED SAFETY FEATURES RESPONSE TIME shall be that time interval from when the monitored parameter exceeds Its ESF actuation setpoint at the channel sensor until the ESF equipment Is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall Include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

FREQUENCY NOTATION 1.13. The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the Intervals defined inTable 1.1.

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

IDENTIFIED 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).

ST. LUCE - UNIT2 1-3 Amendment No. 4-06,43.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 5 of 24 Steam Generator Tube Integrity DEFINITIONS 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 packaging 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 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 RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWtL REACTOR TRIP SYSTEM RESPONSE TIME 1.26 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time Interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until electrical power to the CEA drive mechanism Is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

REPORTABLE EVENT 1.27 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

SHIELD BUILDING INTEGRITY 1.28 SHIELD BUILDING INTEGRITY shall exist when:

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

b. The shield building ventilation system is Incompliance with Specification 3.6.6.1, and

c. The sealing mechanism associated with each penetration (e.g.,

welds, bellows or O-rings) is OPERABLE.

ST. LUCIE - UNIT 2 1-5 Amendment No. O, 4,1, 4-,3,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 6 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 314-4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the SG Program.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:*

a. With one or more SG tubes satisfying the tube repair criteria and not plugged or repaired 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 .

SURVEILLANCE 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 or repaired in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube Inspection.

Separate Action entry Is allowed for each SG tube ST. LUCIE - UNIT 2 3/4 4-11 Amerndment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 7 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNIT 2 3/4 4-12 Amendment No.24, 48,44,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 8 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE-UNrr 2 314 4-12a IAmendment No. 2Z4,448,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 9 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I

,ST. LUCIE - UNIT 2 3M44,-13 Amewndment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 10 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNrT 2 3/44.-14 Amendment No. 444,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 11 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNIT 2 3/4 4-44a Amendment No. 444,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 12 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNrT 2 314 4-15 Amendment No. 43.443,444,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Page 13 of 24 Proposed License Amendment Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNr22 314 4-16 Amendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 14 of 24 Steam Generator Tube Integritv THIS PAGE DELETED ST. LUCIE - UNrT 2 314 4-17 Amendment No. 43,4,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 15 of 24 Steam Generator Tube Integdr REACTOR COOLANT SYSTEM 314.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMING CONDITION 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 4.4.6.2.1.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 perform a RCS water inventory balance per surveillance requirement 4.4.6.2.1.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 Ifat 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.

9T. LUCIE - UNIT 2 3/4 4-18 Amendment No. 84,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 16 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDmON FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE,

b. I 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. 1 gpm leakage (except as noted In Table 3.4-1) at a Reactor Coolant System pressure of 2235 + 20 psig from any Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1.

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 limits, excluding primary-to-secondary leakage, PRESSURE BOUNDARY LEAKAGE, and leakage from Reactor Coolant System Pressure Isolation Valves, 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 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 any Reactor Coolant System Pressure Isolation Valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least two dosed manual or deactivated automatic valves, or be In at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With RCS leakage alarmed and confirmed In a flow path with no flow indication, commence an RCS water Inventory balance within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to determine the leak rate.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 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 monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 .

ST. LUCIE - UNrT 2 3/4 4-19 Amendment No. 4U,3*

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 17 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (CMninued)

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 .

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 .

e. 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 ."

4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve check valve specified in Table 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within Its limit:

a. At least once per 18 months,

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

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

d. Following valve actuation due to automatic or manual action or flow through the valve:

1. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying valve closure, and

2. Within 31 days by verifying leakage rate.

4.4.6.2.3 Each Reactor Coolant System Pressure Isolation Valve motor-operated valve specified InTable 3.4-1 shall be demonstrated OPERABLE by verifying leakage to be within Its limit;

a. At least once per 18 months, and

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

The provisions of Specification 4.0.4 are not applicable for entry Into MODE 3 or 4.

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 - UNrT 2 .3144-20 Amnendmrenit No. ;2,

.St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 18 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continuedD

k. Ventilation Filter Testing Program (VFTP) (continued)

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 specified below.

ESF Ventilation System Delta P Flowrate Control Room Emergency Air Cleanup < 7.4* W.G. 2000 + 200 cfm The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the VFTP test frequencies.

1. Steam Generator (SGI Program I. A Replacement Steam Generator Program shall be established and implemented for the replacement SGs to ensure that SG tube Integrity Is maintained. Inaddition, the Replacement 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 "asfound" 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 Indjuded Inthe design specification) and design basis accidenits. 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 th 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 inaccordance with the design and licensing basis, shalt also be evaluated to determine Ifthe associated lbads 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-o-secondary accident induced leakage rate for any design basis accident, other than SG tube rupture, shall not exceed the leakage rate assumed In the accident analysis Interms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage Is not to exceed 0.3 gpm total through SGs and 216 gallons per day through anyone SG.

3. The operational leakage performance criterion is specified In LCO 3.4.6.2.c,

'Reactor Coolant System Operational Leakage."

ST. LUCIE - UNIT 2 6-15e Amendment No. 4WI,

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1. Steam Generator (SG) Program (continued)

(continued)

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). Ifdefinitive 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(e), then the Indication need not be treated as a crack.

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

2. An Original Steam Generator Program shall be established and Implemented for the original SGs to ensure that SG tube integrity is maintained. In addition, the Original Steam Generator Program shall Include the following provisions:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the *asfound' condition of the tubing with respect to the performance criteria for structural integrity end accident Induced leakage.

The "as found" condition refers to the condition of the tubing during a SG Inspection outage, as determined from the Inservice Inspection results or by other means, prior to the plugging or repair of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are Inspected, plugged or repaired to confirm that the performance criteria are being met.

ST. LUCIE - UNIT 2 6-16f Amendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 20 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS foonflnuedi

1. Steam Generator (SGI Proaram (continued)

2. (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 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 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 0.3 gpm total through SGs and 216 gallons per day 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 or repaired.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1. Plueaina or Repalr Limit means the condition at or beyond which the tube shall be removed from service by plugging or repaired by sleeving using the method In Specification 6.8.4J.2f.1. In the affected area. The plugging or repair limits are as follows:

I In the non-sleeved portion of a tube, the plugging or repair limit Imperfection depth Is 40% of the nominal wall thickness. This Limit is not applicable In the portion of the tube that Is greater than 10.3 Inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) to the tube end. Degradation detected between 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) and the bottom of the hot leg expansion transition or top of the tubesheet (whichever is higher) shall be plugged or repaired on detection.

ST. LUCIE - UNIT 2 6.15a Arnendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 21 of 24 Steam Generator Tube Integrity ADMINISrRATIVE CONTROLS tcontlnued)

L Steam Generator (SGI Program (continued)

2. c. 1. (continued)

I. In the region of a tube sleeved using a Westinghouse Leak Umiting Alloy 800 sleeve, the tube shall be plugged upon detection of any service Induced Imperfection, degradation or defect In the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).

ill. All Leak Limiting Alloy 800 Sleeves that have a nickel band shall be plugged or removed from service after one cycle In operation.

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, and that may satisfy the applicable tube repair criteria. In addition to meeting the requirements of Tube Inspection, d.1, d.2, d.3 and d.4 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.

Tube Inspection for a tube with no portion of a sleeve extending below 10.3 Inches from the bottom of the hot leg expansion transition orthe top of the tubesheet (whichever Is lower) means an Inspection of the steam generator tube from 10.3 Inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever Is lower) to the tube-to-tubesheet weld at the tube outlet.

Tube Inspection for a tube with a portion of a sleeve extending below 10.3 inches from the bottom of the hot leg expansion transition or the top of the tubesheet (whichever Is lower) means an Inspection from the bottom of the sleeve to the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not part of the tube.

1. Inspect 100% of the tubes In each SG during the first refueling outage fbllowing SG replacement

2. Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first Inservice Inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being Inspected.

3. Inspect all Inservice Leak Limiting Alloy 800 sleeves over their full length during each refueling outage. These Inspections will include both the tube and the sleeve.

ST. LUCIE - UNIT 2 6-15h Amendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 22 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

1. Steam Generator (SG) Prooram (continued)

2. d. (continued)

4. 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.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary Integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. ADlacceptable tube repair methods are listed below.

1. Tube Repaer refers to sleeving with Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-15918-P Revision 2 (with range of conditions as revised in Appendix A of WCAP-16489-NP, Revision 0), which are used to maintain a tube In service. Leak Umiting Alloy 800 Sleeves are applicable only to the original steam generators. The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint) shall be Inspected prior to Installation of each sleeve.

ST. LUCIE - UNIT 2 6-16i Amendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 23 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS fcontinued)

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

61. WCAP-1 1397-P-A, (Proprietary), 'Revised Thermal Design Procedure,"

April 1989.

62 WCAP-14565-P-A, (Proprietary), "ViPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis,"

October 1999.

63. WCAP-14565-P-A, Addendum 1, "Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code," May 2003.

64. 30% SGTP PLA Submittal and the SER.

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 cycle revisions or supplements, shall be provided upon issuance for each reload cycle on 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 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,

a. 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,

h. The effective plugging percentage for all plugging and tube repairs In each SG, and

i. Repair method utilized and the number of tubes repaired by each repair method.

81. LUCIE -UNff 2 6-20e Amendment No. 405, 449,433,439,

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 24 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS icentinued)

STEAM GENERATOR TUBE INSPECTION REPORT (continued) 6.9.1.13 Following each inspection of the Original Steam Generators performed in accordance with Specification 6.8A.1., and within 120 days after the Initial entry into HOT SHUTDOWN, the following information concerning indications found in the tubesheet region (including the expansion transition) shall be reported to the Commission. The report shall include:

a. Number of total Indications, location of each indication, orientation of each Indication, severity of each Indication, and whether the indications initiated from the inside or outside diameter.

b. The cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet.

c. Projected end-of-cycle accident induced leakage from tubesheet indications.

This leakage shall be combined with the postulated end-of-cycle accident induced leakage from all other sources. Ifthe preliminary estimated total projected end-of-cycle accident induced leakage from all sources exceeds the leakage limit, the NRC staff shall be notified prior to Unit restart.

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

6.10 DELETED ST. LUCIE- UNIT 2 6-2Mf Amendment No.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 1 of 17 Steam Generator Tube Integrity TS Bases Markups

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 2 of 17 Steam Generator Tube Integrity SECTION NO.: TTLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 2of15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 TABLE OF CONTENTS SECTION PAGE BASES FOR SECTION 314.4 ............................................................................. 3 314A REACTOR COOLANT SYSTEM ........................... 3 BASES ..................................................................................... 3 3/4..1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION ..................... ....................................... 3 3/4.4.2 SAFETY VALVES........................................................ 4 3/4.4.3 PRESSURIZER ........................ E .5 3/444 PORV BLOCK VALVES...........................................6 3/4.4.5 STEAM GENERATO .................................... 6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .......... 8 x

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS ......... 8 3/4.4.62 OPERATIONAL LEAKAGE ..................- 8 3/4A.7 CHEMISTRY ......................................................... 9 3/4A.8 SPECIFIC ACTIVITY ............................................ 10 3/4A.9 PRESSURErTEMPERATURE LIMITS . ........... 11 3/4.4.10 REACTOR COOLANT SYSTEM VENTS...........13 TABLE B 3/4.4-1 REACTOR VESSEL TOUGHNESS ....... 14 3/4.4.11 STRUCTURAL INTEGRITY ............................... 15

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 3 of 17 Steam Generator Tube Integrity SECTION NO.: TM*.: TECHNICAL SPECIFICATIONS PAGE:

314.4 BASES ATTACHMENT 6 OF ADM-25.04 6of 15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT2 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.4 PORV BLOCK VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs In conjunction with a reactor trip on a Pressurizer Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The opening of the PORVs fulfills no safety-related function and no credit is taken for their operation In the safety analysis for MODE 1, 2, or 3.

Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become Inoperable. Since it is impractical and undesirable to actually open the PORVs to demonstrate their reclosing, Itbecomes necessary to verify OPERABILITY of the PORV block valves to ensure capability to Isolate a malfunctioning PORV.

As the PORVs are pilot operated and require some system pressure to operate, it is Impractical to test them with the block valve dosed.

The PORVs are sized to provide low temperature overpressure protection (LTOP). Since both PORVs must be OPERABLE when used for LTOP, both block valves will be open during operation with the ITOP range. As the PORV capacity required to perform the LTOP function is excessive for operation in MODE 1, 2, or 3, it Is necessary that the operation of more than one PORV be precluded during these MODES. Thus, one block valve must be shut during MODES 1, 2, and 3.

A4X. STEAM GEEAOC - 1w)TB NE~n Surveillance Requirements for inspection of the steam generator, S tubes ure that the structural integrity of this portion of teRCS will be maintain program for inservie Inspection of steam generator uactue crtubes Is based oinodification of Regulatory Guide 1.83, Revision 1.r Inservice Inspection of stm generator tubing is essential In order to maitan surveillance ota.cnditions of the tubes In the event that

there Is evidence of mechanica age or progressive degradation due todsin anuatrng errrs,'or I~c conditions that lead to Inevc Inpcino temgnrtrtbing als dkes a means of INSERT characterizing the nature and cause of any tube degrada that 93/4.4.5corrective measures can be taken.

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3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 7of15 7

REVISION NO.. REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) *_ *(sg) T7USE INlTEGlRrM 314.4.5 STEAM GENERATO continued)

" e plant is expected to be operated in a manner such that the secondary coo*,nt will be maintained within those chemistry limits found to result In neglig ae corrosion of the steam generator tubes.. If the secondary coolant *hmistry is not maintained within these limits, localized corrosion.

may likely ult in stress corrosion cracking. The extent of cracking during plant o tion would be Hmited by the limitation of steam generator tube I ge between the primary coolant system and the secondary coolant tem (primary-to-secondary leakage = 1.0 gpm from both steam generators. Cracks having a primary-to-secondary leakage less than this limit during ration will have an adequatermargin of safety to withstand the loads impo d during normal operation and by postulated accidents. Operating plants h demonstrated that primary-to-secondary leakage of 0.5 gpm per steam ge rator can readily be detected by radiation monitor of steam gerate lowdown.: Leakage in excess of this limit will require plant shutdown n unscheduled inspection, during which the leaking tubes will be Ioc dand plugged.

Wastage-ype defects are unlikely with proper emistry treatment of the secondary coolant However, even If a defect sho d develop in service, it will be found during scheduled inservice steam gene tor tube examinations. Plugging will be required of all tuIbe wi mperfections exeein the plgiglmto0 of thete nominal wf*~ticknes Steam geeao ue npcin f operatin plants have donstrted

\the capability to reliably detect degradation that has penetrated %of thej liia tub wall thickness. II

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 5 of 17 Steam Generator Tube Integity SECTION NO.: TmTLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 8 of 15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 34.4.6 REACTOR COOLANT SYSTEM LEAKAGE 314.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, 'Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. The LCO is consistent with NUREG-1432, Revision 1, and Is satisfied when leakage detection monitors of diverse measurement means are OPERABLE in MODES 1, 2, 3, and 4. Monitoring the reactor cavity sump inlet flow rate, In combination with monitoring the containment particulate or gaseous radioactivity, provides an acceptable minimum to assure that unidentified leakage Is detected in time to allow actions to place the plant in a safe condition when such leakage indicates possible pressure boundary degradation.

314.4.6.2 OPERATIONAL LEAKAGE e x ustrydexperience has shown from the RCS, the unidentified a limited that whileportion amount of this leakage can be Is of leakage

reducedthreshold value of less than I gpm. This threshold value is r sufficiently ensreearly detection of additional leakage.

. :The 10 gpm IDENTIFIGE limitation provides allowances for a Slimited amount of leakage frn*lown sources whose presence will not" Interfere with the detection of UNI IF'IFED LEAKAGE by the leakage

!dt em:

systems.*: "detection '.. . .

grs vav alr ndcneun ntersystem LOCA. Lea ae from the

... *1RCS pressure.isolation vatves is IDENTIFIED LEAKAGE and Wi'l.

. [considered as a portion of the allowable limit. .

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 6 of 17 Steam Generator Tube Integrity SECTION NO.:I TI.Ex TECHNICAL SPECIFICATIONS PAGE:

314.4 BASES ATTACHMENT 6 OF ADM-25.04 9 of 15 REV SION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 314.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 314.4.6 REACTOR COOLANT SYSTEM LEAKAGE (continued) 314.4.6.2 OPERATIONAL LEAKAGE (continued

ho.,tl steam generator tube leakage limit of I gpm for all steam T N IZNSERT O : t gone ~~nsures that the dosage contribution from the tube leakage will 83/4.4.6.2 be limited toa fraction oforPart all*utre limits 100line event in theThe of either a a U

[(f01olwsl steam generator steam break. I gpm limit is

[zwsrt for a consistent with the a ons used In the' analysis of these accidents.

3/4.45) The 0.5 gpm leakage limit peearn generator ensures that steam geneIrtrtb ntegrity Is maintat i the event of a main steam line rupture or under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any mag e Is unacceptable since It may be Indicative of an impending gross failure he pressure boundary. Therefore, the presence of any PRESSURE BO ARY LEAKAGE requires the unit to be promptly placed in COLD SH WN.

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

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

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 7 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5

Background

Steam generator (SO) 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 SO 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 the 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 Coolant 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."

SO 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. SO tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and 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 SO performance criteria are used to manage SO tube degradation.

Specification 6.8.4 1., "Steam Generator (SO) 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 SO performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage. The SO 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. Specification 6.8.4 1.has two parts to address the replacement SG and original SG designs. Specification 6.8.41.1. applies to the replacement SG design. TS 6.8.41.2.

applies to the original SGs and contains requirements such as a sleeving repair method, alternate repair criteria and additional inspection requirements, which are all unique to the original SO design and can be removed following SO replacement.

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

Arrlicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basis event for SO tubes and avoiding a SOTR 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 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.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 8 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5 The analysis for design basis accidents and transients other than a SGTR assume the SO 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 0.3 gpm total or 216 gpd through any one SO as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 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 limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(cX2Xii).

Limiting Condition for Operation (LCO)

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

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

In the context of this Specification, a SQ tube for the replacement SGs 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. For the original SGs, a SG tube is defined as the length of the tube, including the tube wall and any repairs made to it, between 10.3 inches below the bottom of the hot leg expansion transition or top of the tubesheet (whichever is lower) and the tube-to-tubesheet weld at the tube outlet If a portion of a tube sleeve extends below 10.3 inches from the bottom of the hot leg expansion transition or the top of the tubesheet (whichever is lower) a SG tube is defined as the length of the tube between the bottom of the sleeve to the tube-to-tubesheet weld at the tube outlet The tube-to-tubesheet weld is not considered part of the tube.

A SO tube has tube integrity when it satisfies the SG performance criteria. The SO performance criteria are defined in Specification 6.8.4 L, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SO performance criteria.

There are three SO 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 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 9 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 ofADM-25.04 - INSERT B3/4.4.5 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 III, 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 0.3 gpm total or 216 gpd 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 SO to 150 gpd at room temnperature. 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 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 10 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5 Applicability SO tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SO 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 during 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 SO tube. This is acceptable because the required ACTIONS provide appropriate compensatory actions for each affected SO tube. Complying with the required ACTIONS may allow for continued operation, and subsequent affected SO tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

a.l anda.2 ACTIONS a.1 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 or repaired in accordance with the Steam Generator Program as required by Surveillance Requirement (SR) 4.4.5.2. An evaluation of SO tube integrity of the affected tube(s) must be made. SG tube integrity is based on meeting the SO performance criteria described in the Steam Generator Program. The SO repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SO performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged or repaired has tube integrity, an evaluation must be completed that demonstrates that the SO performance criteria will continue to be met until the next refueling outage or SO 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.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 11 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5 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 SO inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes.

However, the affected tube(s) must be plugged or repaired prior to entering HOT STANDBY following the next refueling outage or SO 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 determines 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 determines 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 SO are to be inspected) is a fimction 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 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 nerformance criteria will be met between scheduled insvections.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 12 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5 SR 4.4.5.2 During a SG inspection any inspected tube that satisfies the Steam Generator Program repair criteria is repaired or 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.

Steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program (Specification 6.8.41.2.).

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

References I. NEI 97-06, "Steam Generator Program Guidelines"

2. 10 CFR 50 Appendix A, GDC 19

3. 10CFR100

4. ASME Boiler and Pressure Vessel Code, Section HI, 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 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 13 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 Backsrotmd Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS). Componentjoints 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 purpose 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 cannot 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).

Applicable 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 0.3 gpm total through SGs and 216 gpd through any one SO as a result of accident induced conditions. The dose consequences ofthese 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 than or equal to 150 gpd is based on room temperature conditions.

When this value is adjusted for operating conditions, it is essentially the same as the leakage limit assumed in the accident analysis. Therefore, the margin assumed in Reference 4 is not maintained. To ensure that the margin is consistent with the Staff's discussion in the Reference 4, St. Lucie Unit 2 procedures further administratively limit operational leakage to half the value assumed in the accident analysis.

The RCS operational leakage satisfies Criterion 2 of 10 CFR 50.36(cX2X)i).

Lnimtins 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

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 14 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 ofADM-25.04 - INSERT B3/4.4.6.2 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, ifthe 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 97-06 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.

The leakage limit assumed in the accident analysis, however, was reduced to 216 gpd/SG by license amendment 138. In Reference 4 the NRC concludes that the proposed operational leakage LCO of 150 gpd is acceptable because it is significantly less than the conditions assumed in the accident analysis. When operational leakage equal to 150 gpd is adjusted for operating conditions, however, the resulting value is essentially the same as the leakage limit assumed in the accident analysis (i.e.,

approximately 216 gpd/SG). Therefore, the margin assumed in Reference 4 is not maintained. To ensure that the margin is consistent with the Staffs discussion in the Reference 4, St. Lucie Unit 2 procedures further administratively limit operational leakage to half the value assumed in the accident analysis.

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 finction not considered leakage). Violation of this LCO could result in continued degradation of a component or system.

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.

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 15 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 ofADM-25.04 - INSERT B3/4.4.6.2 Aenlicabilit 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 IDENTIFIED 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 manual or deactivated automatic 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-fimctional 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 3.4-1 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 .

d. With RCS leakage alarmed and confirmed in a flow path with no flow indication, commencement of an RCS water inventory balance within is required within I hour to determine the leak rate. This action is not applicable to primary-to-secondary leakage.

The allowed completion times are reasonable, based on operating experience, to reach the required plant conditions from fAll 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 4.4.6.2.1 Verifying Reactor Coolant System leakage to be within the LCO limits ensures the integrity of the 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 verformance or a Reactor Coolant System

St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 16 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 ofADM-25.04 - INSERT B3/4.4.6.2 water inventory balance.

a ad 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 .

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 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 .

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 2. The operational leakage rate limit applies to leakage through any one steam generator. Ifit'is not practical to assign the leakage to an individual steam 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

I St. Lucie Unit 2 L-2006-094 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 17 of 17 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 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).

4.4.6.2.2

a. through d.

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

4.4.6.2.3

a. and b.

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

References I. NEI 97-06, "Steam Generator Program Guidelines"

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

3. UFSAR, Section 15.6.3

4. NRC Federal RegisterNotice 70 FR 10298, March 2,2005 "Notice of Opportunity To Comment on Model Safety Evaluation on Technical Specification Improvement To Modify Requirements Regarding The Addition of LCO 3.4. [17] on Steam Generator Tube Integrity Using the Consolidated Line Item Inmrovement Process."

L-2006-094, Proposed License Amendment Steam Generator Tube Integrity Pursuant to 10 CFR 50.90

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