Text

Reply to Request for Additional Information Re Steam Generator Tube Integrity Amendment Request
	ML071210531
Person / Time
Site:	Saint Lucie
Issue date:	04/16/2007
From:	Johnston G; Florida Power & Light Co
To:	; Document Control Desk, NRC/NRR/ADRO
References
	L-2007-071
	Download: ML071210531 (75)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 April 16, 2007 FPLt L-2007-071 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 Reply to Request for Additional Information Steam Generator Tube Integrity Amendment Request Via letter L-2006-094 dated May 25, 2006, Florida Power and Light Company (FPL) requested to amend Facility Operating License NPF- 16 for St. Lucie Unit 2 to change the Technical Specification (TS) requirements related to steam generator tube integrity. The change was based on NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF - 449, "Steam Generator Tube Integrity." FPL responded to the first NRC request for additional information (RAI) in FPL letter L-2007-003 dated January 22, 2007. Further interactions between the NRC staff and FPL resulted in another RAI, and this letter forwards FPL's reply to the RAI.

Attachment 1 provides the RAI reply. Attachment 2 provides marked-up TS pages to support the RAI reply, and Attachment 3 provides the word-processed TS pages. Attachment 4 provides an information only markup of the TS Bases to support the RAI reply. Attachments 2, 3, and 4 of this letter are complete replacements for the word-processed TSs and markups of the TS and TS Bases provided in FPL letter L-2007-003.

The results of the no significance hazards evaluation in the original submittal remain unaffected by the RAI reply. If there are any questions on this submittal, please contact Mr. Ken Frehafer at (772) 467-7748.

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

Executed on the -- day of ,Z , ,2007.

Ve yruly yours, rdonL. Johnst n Site Vice Preside St. Lucie Plant GLJ/KWF Attachments cc: Mr. William A. Passetti, Florida Department of Health an FPL Group company

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment I Proposed License Amendment Page 1 of 4 Steam Generator Tube Intearitv REQUEST FOR ADDITIONAL INFORMATION ST LUCIE UNIT 2 STEAM GENERATOR TUBE INTEGRITY TECHNICAL SPECIFICATION AMENDMENT DOCKET NO. 50-389

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 2 of 4 Steam Generator Tube Integrity By letter dated May 25, 2006 (ML061510346), Florida Power & Light Company (the licensee) submitted a license amendment request (LAR) regarding Saint Lucie Unit 2 steam generator (SG) tube integrity technical specifications (TS). The proposed amendment would revise the SG tube integrity TSs to be consistent with the U.S. Nuclear Regulatory Commission's approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4 (ML0510902003). The licensee submitted additional information regarding this LAR in letter dated January 22, 2007 (ML070250070). After reviewing the information provided by the licensee, the staff has determined that the following information is needed to complete the review.

1. Please discuss your plans to modify TS Section 6.8.4.1.2.c.3 to be consistent with your current TS. "All tubes with sleeves that have a nickel band shall be plugged after one cycle in operation." The staff is aware that the staff's previous suggestion for this TS section had a typographical error.

Reply to #1 - The requested change has been made as shown in the enclosures.

2. Proposed TS Section 6.8.4.1.2.c.4.ii does not appear to be complete, since it assumes the repair criteria is independent of the location of the lower sleeve joint within the tubesheet. For example, if there is a lower sleeve joint at 10 inches from the top of tubesheet, the currently proposed TS would allow flaws below this region to remain in service. It is not clear that this is technically acceptable. As a result, discuss your plans to modify this TS section to address the repair criteria for the various locations where the sleeve joint could exist.

Reply to #2 - The alternative wording below for TS 6.8.4.1.2.c.4.ii was discussed with the Staff and found acceptable. It requires plugging of any flaw within the tubesheet if it is located below the sleeve to tube joint and within the C* distance (i.e., within 10.3 inches .... ). The alternative wording is included in the enclosures provided with this response.

ii. Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw that is located below the lower sleeve to tube joint and within 10.3 inches below the bottom of the hot-leg expansion transition or top of the tubesheet, whichever elevation is lower.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment I Proposed License Amendment Page 3 of 4 Steam Generator Tube Integrity

3. Please discuss your plans to modify proposed TS Section 6.8.4.1.2.d to clearly define when the inspection exclusion is applied. For example, "For tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region, the portion of the tube below 10.3 inches from the top of the hot leg tubesheet or expansion transition, whichever is lower, is excluded when the alternate repair criteria in TS Section 6.8.4.1.2.c.4 are applied."

In addition, the second sentence in the fifth paragraph on Page 8 of Attachment 4 should be modified to read, "For the original SGs, when the alternate repair criteria in TS Section 6.8.4.1.2.c.4 are applied, a SG tube is defined..."

Reply to #3 - The clarifications are included in the enclosures provided with this response.

4. Given that the original SGs have an approved SG tube repair method (i.e., sleeving),

discuss your plans to modify TS Section 6.9.1.13.e to read as follows, "Number of tubes plugged or repaired during the inspection outage for each active degradation mechanism."

Reply to #4 - The change is included-in-the enclosures.provided with this response..

5. On Page 13 of Attachment 4 it is stated, "To ensure that the margin is consistent with the Staff's discussion in the Reference 3, St. Lucie Unit 2 procedures further administratively limit operational leakage." This sentence appears to imply that meeting the margins defined in Reference 3 will ensure you always meet your accident induced leakage limit. That was not the intent of the discussion in Reference 3.

Please discuss your plans to modify this sentence to read as follows, "St. Lucie Unit 2 procedures further administratively limit operational leakage to ensure that the accident induced leakage limits are not exceeded."

Reply to #5 - The alternative wording below was discussed with the Staff and found to be acceptable. It is included in the enclosures provided with this response.

"St. Lucie Unit 2 procedures further administratively limit operational leakage with the intent that the accident induced leakage limits will not be exceeded."

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 4 of 4 Steam Generator Tube Integrity

6. On Page 19 of Attachment 4, you indicate, in part that site boundary doses for a SG tube rupture (SGTR) will not exceed an appropriately small fraction of Title 10 of the Code of Federal Regulations (10 CFR), Part 100. Based on your response to Request for Additional Information 20, the staff was under the impression that the SGTR source term is based on 10 CFR, Part 50.67. Please confirm whether Part 100 is the appropriate reference on Page 19. If it isn't, please correct.

Reply to #6 - FPL agrees that the SGTR dose limit is based on 10 CFR Part 50.67 rather than 10 CFR Part 100. The reference to 10 CFR Part 50.67 was added to our current TS Bases after our May 25, 2006 submittal (based on approval of AST methodology for SGTR event in Amendment # 138). The proposed wording, as marked in the May 25, 2006 submittal is to clarify the primary-to-secondary leakage only. A subsequent RAI letter dated 10-24-2006 did not affect Page 19 of Attachment 4 and our RAI response dated January 22, 2007 (FPL letter L-2007-003) did not update this page. The current Bases wording as approved in Amendment # 138 referencing 10 CFR 50.67 remains unchanged and is included in the enclosures provided with this response

St. Lucie Unit 2 L-2007-071 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-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 2 of 27 Steam Generator Tube Intearitv INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES DELETED ............................................................................................... 314 4-7 O PERATING ............................................................................................ 3/4 4-8 3/4.4.3 PRESSURIZER ............................................................................................... 3/4 4-9 314.4.4 PORV BLOCK VALVES ........ 314 4-10 3/4.4.5 STEAM GENERATO.I y- ............. 3/44-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ...................................................... 314 4-18 OPERATIONAL LEAKAGE ........................ 314 4-19 3/4.4.7 CHEMISTRY .......... ................ . . . . . . . . . .......... 3/44-22.

3/4.4.8 SPECIFIC ACTIVITY ...................................................................................... 3/4 4-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM .................................................................... 3/44-29 PRESSURIZER HEATUPICOOLDOWN LIMITS ............................................ 3/44-34 OVERPRESSURE PROTECTION SYSTEMS ................................................ 3/44-35 314.4.10 REACTOR COOLANT SYSTEM VENTS ........................................................ 3/44-38 3/4.4.11 STRUCTURAL INTEGRITY .......................................................................... 314 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3a4.5.1 SAFETY INJECTION TANKS ........................................................................... 3/4 5-1 314.5.2 ECCS SUBSYSTEMS - Tavg > 325"F .............................................................. 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tan < 325"F .............................................................. 314 5-7 314.5.4 REFUELING WATER TANK ............................................................................. 3/4 5-8 ST. LUCIE - UNIT 2 VI Amendment No, 4-6.1,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 3 of 27 Steam Generator Tube Integrity INDEX ADMIN ISTRATIVE CONTROLS

SECTION PAGE 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 6.9.2 SPECIAL RREPORTST ........................................................................ 6-20eSC 6.10 DELETED .................... .............. ................ 6-20e 6.11 RADIATION PROTEC-TION PROGRAM.............................................. 6-21 6-20e 1 isTEAM GENERATOR~ TUBE INSPECTIQN R~EPORT ST. LUCIE - UNIT 2 X1X Amendment No. 43, ", 80, 02, 409, 448,"fpO .

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 4 of 27 Steam Generator Tube Integrity -.

DEEMNBONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/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 E 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 (in MeV) 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 in Table 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.

ST. LUCIE- UNIT 2 \ 1-3 Amendment No. 4-05,1 riary -to-seconda leako e

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 5 of 27 Steam Generator Tube Integrity DEFINITIONS PRESSURE BOUNDARY LEAKAGE lir-lo-secondlor 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except erator t--e 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 Y 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 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 - UNIT 2 1-5 Amendment No. 9, 43, 84y)1'

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 6 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATO1 SI1 1(5') TUBE "rNTE6ITY LIMfNG CONDITION FOR OPERATION 3,4.5 Each steam ge I oAP\L. INSET AI~i APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:"--]

fwHTn,

inoeable, restore the Inoperable F)~

4geeraor(s) to OPERABLE status prior to increasingi LU2O SURVEILLANGEREOUIREMENTS r -- s--eei ý n~erator shall be demonstrated OPERABLE by performance of 4.4.5.1 St -n Sample Selection a nspe on - steam generator shall be determined OPERAB b selecting and inspecting at minimum number of steam generators speie ion-NRe TsteIa enerator tubteg uier *sample adinsation, s*lcto an thesemm l

be minirr*Km sa, ple sze, inspection result classifiaonadth thesponding action e sl be as specified in Table 4.4-2. The inse, ce inspection of stenar generator tubes shall be performed at the frequehles specified in Spefirtion 4.4.5.3 and the inspected tubes shall I be verifie coeptable per the acc ance crteriasof Specification 4.4.5.4.

These*ed ube or achinsrvie "heto shall include at least 3% of the total num r tubes in all steam gen tlors; the tubes selected for these inspections sh11beselected ona rand ai xcept:

a. hr x 'neI iilar plants with amrwater chemistry iniatsciti alreas to be inspecte, then a 50% of the stte tubes inspected s fro~fm thes citical areas. *(

b. h is sml ftbe ee freach inservice inspection ,..

(subsequent o the preservice inspetono earn generator*

~shall include:

I*Separate Action entry is allowed for each SG tube.I ST. LUCIE - UNIT 2 3/4 4-11

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 7 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM IREILLANCE REQUIREMENTS (Contlnuedl

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

2. Tubes in those areas where experience has Indicated potential problems.

3., Atube inspection (pursuant to Specification 4.4.5.4.a.8), all nd be performed permit the on each selected passage tube.current of the eddy If anyprobe selected for tIe tubedoes Modified a 2.

inon, this shall be recorded and an adjacent be relocated shall shown late .rin be lected and subjected to a tube inspection.

propsed4. Al Inse *ce Leak Limiting Alloy 800 sleeve TS 6.8.4.1.2.d.3 their full le during each refueling outag(

I include both e tube and the sleeve. /

c. The tubes selected a;he second and amples (if required by Table 4.4-2} during eac nservice Insp i may be subjected to partial tube inspection pro.'ded: /

1. The tubes selected for t pies include the tubes from those areas of the tube ay where tubes with imperfections were pre'

2. The inspections ions of the tubes where imperfections we The results of each sample ne of the following three categ1ories:

Ca_

iteso Inspection Results C-1 Less than 5% of the total tubes inspect are degraded tubes and none of the insp ed tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

THIS PAGE DELETED A q-ST. LUCIE - UNIT 2 3/4 4-12 Amendment No. 24,,4e-?47

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 8 of 27 Steam Generator Tube Integrity Inspection Results More than 10% of the total tubes inspected are degraded tubes or more than 1% of the inspected tubes are defective.

Note: (1) In all inspections, piously degraded tubes must, significant (greatge ethan 10%) further wall penetratii be included in It above percentage calculations.

THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-12a Amendment No. 24,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 9 of 27 Steam Generator Tube Integrity 4.4.5.* Inspection Frequencies -The above required inservice inspections of /

Steam gne~retor tubes shall be performed at the following frequencies:/

a*The first inservice inspection shall be performed after 6 Effective Full Power Months but within 24 calendar months of initial crit-lity. Subsequent inservice inspections shall be performed at I rvals of not less than 12 nor more than 24 calendar month after the evious inspection. If two consecutive inspections foll ng servi under AVT (all volatile treatment) conditions, not i uding the pre rvice inspection, result in all inspection results I ling into the category or iftwo consecutive inspectiS emonstrate that previo observed degradation has not contin ed and no addi-tional degrad *on has occurred, the inspection i may be extended to a ximum of once per 40 months

b. If the results of the 1'ervice inspection of a earn generator conducted in accorda with Table 4.4- t 40-month intervals fall into Category C-3, the i pection frequ cy shall be increased to at least once per 20 months. The incre in inspection frequency shall apply until the b eq nt Ins ctions satisfy the criteria of Specification 44.531.; the i may then be extended to a maximum of once per 40 montl

c. Additional, unscheduled in rvice i ections shall be performed on each steam generator in cordance t the first sample inspection specified in Table 4.4- uring the shut subsequent to any of the following condito s:

1. Primary-to- condarv tubes leaks (not in udino leaks i tube-to-tube sheet wek ;ess of the cation 3.4.6.2.

2. A smic occurrence greater than the Operating'la I/rhquake. N

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

4. A main steam line or feedwater line break.

THIS PAGE DELETED ST. LUCIE - UNIT 2 314 4-13

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 10 of 27 Steam Generator Tube Integrity (Contlnued'*

~REILLANCE REACTOR COOLANTREQUIREMENTS SYSTEM 4.4.5. Acceptance Criterian As used in this Specification S1. 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 200 of the nominal tube wall thickness, if detectable, may be

"*sximperfections. sidered "pret

2. D

dation means a service-induced cracking, bng occurring stage, wear or gene 'adedcorrosion on either inside or tside of a tube

3. De raded be means a tube containing perfections greater than or equal 20% of the nominal wa hickness causedreby degradation.

[7-7; ......

MOOiTlea ano relocated as 4. cer*Sge of the tube wall thickness shown later in i,$on.

proposed 5. of such severity that it exceeds TS 6.8.4.1.2.c taining a defect is defective.

6. .ing or Repair UAit means' e condition at or beyond which the tube be removed f service by ugging or repaired by sleeving using nethod in Sp!cation 4.4.5.4. 10 in the affected area. The plugging pair limits e as follows:

In the on-sleeved portion of a tube, e plugging or repair limit im ction depth Is 40% of the nomi I wall thickness. This Umit is applicable in the portion of the tu that is greater than

.3 inches below the bt of the hot le xpansion transition or top of the tubesheet (whichever is lower) to I tube end.

Degradation detected between 10.3 inches be w the bottom of the hot leg expansion transition or top of the tubes (whichever is r lower) and the bottom of the hot leg expansion tra ition or top of the tubesheet (whichever is higher) shall be plugged r repaired on

/ detection.

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

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

ST. LUCIE - UNIT 2 3/44-14 Amendment No.,"Y-THIS PAGE DELETED

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 11 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM RVEILLANCE REQUIREMENTS QContinued) z

7. Unserviceable describes the condition of a tube if it leaks or l contains a defect large enough to affect its structural integrity /

in the event of an Operating Basis Earthquake, a loss-of-coolant accident, a steam line or feedwater or above.

in4.4.5.3c., .. line break as specifiedo ube Inspection for a tube with no portion of a sleeve extending bo below transition 10.3 inches from the bottom of the hot leg expansion or the topt(

insppecionis lower) means an the tubesheet (whichever of the steam 9geerator tube from 10.3 inches below the boto1C the hot leg expanstor Stran'tion or top of the tubesheet (whiclhever Ts lo r) completely around the U- nd to the top support of the cold leg. TJbe Inspection for a tube w Itap on of a leeve etnding below 1,0.3 inches from the bottom of ho lgxpanston

,h tvransition or the tope the tubesheet (whichever is//

loer man'san inspection from the bo tm of the sleeve completely

arund the U-!nd to the top support the cold leg.

9. Prserice Ins n means an i ction of the full length of each tube in each stem gene r performed by eddy current techniques prior to se 'c to tablish a baseline condition of the tubing. i nspection shall be performed after the field hydrostatic t and prior to initial POWER OPERATION using the qui ent and techniques expected to be used during subsequ t inse inspections.

THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-14a Amendment No.,Wp-

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 12 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM

10. Tbh15 ay refers to sleeving with Westinghouse Leak Limiting Alloy sleeves as described in WCAP-15918-P Revision 2 (with range of st gneditions as revised in Appendix A of WCAP-1 i9-NP, Revision b Thewhich are used th maintain a tube in service. Leak Limiting Alloy Modified and Shll ves are applicableueo original steam generators. uT pressure boundary portion1olginalof the tube wall in the slee/tube assembly (i.e., the sleeve-to-tube joint) shall be inspected toregor proposed aewnstallation of each sleeve . t se 6.8.4.1.2.f.1 iTS generatorsam shall be determined OPERABLE afte mpleting ctheacpondig actions (plug or repair all tubes exces"Igthe Plugging or Repair4.4- it and all tubes containing through-wall crac Table required by 4.4.5.5 Reports /

a. Within 15 days foe eing the completion of ea inservice inspection of steam generator bes, the number of tu ~s .plugged or repaired in each generator shal reported to the in a Special Report pursuant to pcfca1692

b. The complete results of th team g erator tube inservice inspection shall be submitted to the Co l~iss'~n in aSpecial-Report pursuant, to.

Specification 6.9.2 within 12 mn s following completion of the inspection. ThsSeilRepo! 11a include:

Modif ied and 1 Nm4 and Amsn dmented nepoosd 2. Location and percnt ofTRelocated H.SNPAe s penetration for each wall-thickn ne poosd indication of an' nperfection. "

IT5 6.9.1.13 3. Identificati~o tubes plugged or repa

c. Following, eac *inspection and within 120 days ,,r the reactor coolant system rreenters M( E4, the following information concer i g indications found in the t~ubesheet/ ein (including the expansion transition) *all be reported to the Com~mis on in a special report pursuant to Specificatio R6.9.2. This Special Re phall include:

I 'Numbe otoainiainlcto eahndaion, rientation of each7

, J !indication, seve~rityo~futeadchdindication' and . ter the in "'cains initiated

/~~3Psiae rojected oa end-of-cycle accident inducted leakage tublle from

/ C sources exceeds the leakage limitL the NRC staff shall be notified prior to

/ ",,*~nit restart. * ,

ST. LUCIE - UNIT 2 314 4-15 Amendment No. -13.443, 44,.14-JJ-THIS PAGE D)ELETED)

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 13 of 27 Steam Generator Tube Integritv 3/4.4.5 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. Repair applies only to the original SGs.

3/4.4.5 INSERT B

a. With one or more SG tubes satisfying the tube repair criteria and not plugged (or repaired if original SGs) 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. Repair applies only to the original SGs.

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

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

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

0*

z CD CD CD 00 TABLE 4,4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTION

""ýz Preservice Inspection No. of Steam Generators per Unit First Inservice Inspection Second & Subsequent Inservice Inspections Table Notation:

1. The inservice inspection may be limited to one steam gen or o rotating schedule encompassing 3 N % of the tubes -4 X

(where N Is the number of steam generators in the pla rf the results the first or previous inspections indicate that all steam generators are performing in a like mann . Note that under so circumstances, the operating conditions in one or more steam generators may be found t e more severe than those in er steam generators. Under such circum-stances the sample sequence shall be m ed to Inspect the most severe condi s. '-I

2. The other steam generator not ins d during the first inservice inspection shall be insp ed. The third and subsequent M inspections should follow the

tions described In 1 above.

m-

3. Each of the other two m generators not inspected during the first inservice inspections shall be in cted during the second and third I ctions. The fourth and subsequent inspections shall follow the instructions describ in 1 above.

CD ST. LUCIE - UNIT 2 3144-16

0*

Co 0

CD Pz;1 CD -

CD.

CD CD CDC TABLE 4.4-2 STEAM GENERATOR TUBE INSPECTION CD5 z

1st S IPLE INSPECTION 2nd SAMPLE INSPECTION 3rd SAMP LEINSPECTION Sample Size ResN, Action Required Result Action Required Result Action Required A minimum of C-1 None N/A N/A N/A S Tubes per S.G.

G-2 Plug or rep defective C-1 None N/AA Ln tubes and lnsp additional _ __.

2S tubes in this S.C0-2 tubesorapdinspect Plug repdefective C-I None addnal 4S tubes in C-2 Plug or repair defective ads S.G.n tubes m C-3 Perform action for C7 M 0-3 result of first r" sample

-3 Perform action for N/A N/A M7 "I'-3result of first

,K sahe'Rle.

C-3 Inspect all tubes In All other one N/A WA this tube or re ir S.G., plug defective S.G.s are defective tubess*dd -C-11_____________

inspect 2S t .In Some S.G.s Perform action fo N/A N/A each 0th .G. C-2 but no C-2 result of second additional sample.

S.G. are C-3 _________

Additional Inspect all tubes In N/A NA S.G. is 0-3 each S.G. and plug or

,_ _ _ _ _ repair defective tubes.

S = 3 AWhere N is the number of steam generators in the unit, and n is the number of steam generators inspected during an Inspection.

oCD t 0 ST. LUCIE - UNIT 2 3144-17 Amendment No. " t,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 16 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITNG CONDITON 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 slm inoperable, perform a RCS water inventory balance at least once per 24hours 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 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and restore the required radioactivity monitor to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the nAxt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . []

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

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

SURVEILLANCE REQUAREJYfENTS 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.

ST. LUCIE - UNIT 2 3/4 4-18 Amendment No.'*

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 17 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE I IMITINir (rnmnnTIAN FAQn ripARATIA 3.4.6.2 Reactor Coolant Systemleakage shall be limited to: [ optal

a. No PRESSURE BOUNDARY LEAKAGE,

c. m totaM mary-to-secondary lea e througl,. Jm.

[150 gallons pr day generators and -t-ga ons per da-ythroug .n one steam generatorJ

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. or with priary-to-secondary leakage not within limit, ACTION: a. With any PRESSURE BOUNDARY LEAKAGE, /b 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.hours.-

b With any Reactor Coolant Systenm leakage greater than any one of the pritnary-to-secondary limits, excludingdPRESSURE BOUNDARY LEAKA(G nd 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 closed 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.

lIQViFILLAM(CF R*fO[IIRF*MFNTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

operational

a. Monitoring the containmnt 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 3/4 4-19 Amendment Noý

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 18 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM SURVEILLAN RE-UIREMENTS (Continued' 0V - 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 ifleakage 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.

e. Verifying primary-to-secondary leakage is 1_150 gallons per day through anry one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . **II 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. Not applicable to primary-to-secondary leakage.

SL** 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.J ST. LUCIE - UNIT 2 3/4 4-20 Amendment No.N1g..,.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 19 of 27 Steam Generator Tube Integrity

k. Ventilation Filter Testing Promram (VFTPI (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 frequenccles.

ST. LUCIE -UNIT 2 6-15e Amendment No.y2*

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 20 of 27 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS - INSERT 6.8.4.1.

1. Steam Generator (SG) Proeram

1. A SG Program shall be established and implemented for the replacement SGs to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following provisions:

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

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

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

2. Accident induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a 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 gallons per minute total through all 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."

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 21 of 27 Steam Generator Tube Integrity

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 tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, 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. A SG Program shall be established and implemented for the original SGs to ensure that SG tube integrity is maintained. In addition, the SG 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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 22 of 27 Steam Generator Tube Integrity 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 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 a 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 gallons per minute total through all 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.

1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40-percent of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate tube repair criteria discussed in Technical Specification 6.8.4.1.2.c.4.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 23 of 27 Steam Generator Tube Integrity

2. Tubes found by inservice inspection to contain a flaw in (a) a sleeve or (b) the pressure boundary portion of the original tube wall in the sleeve to tube joint shall be plugged.

3. All tubes with sleeves that have a nickel band shall be plugged after one cycle in operation.

4. The C* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg tubesheet region as an alternative to the 40-percent depth based criteria of Technical Specification 6.8.4.1.2.c. 1.

i. Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 10.3 inches below the bottom of the hot-leg expansion transition or top of the tubesheet, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.

ii. Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw that is located below the lower sleeve to tube joint and within 10.3 inches below the bottom of the hot-leg expansion transition or top of the tubesheet, whichever elevation is lower.

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

The number and portions of the tubes inspected and methods of inspeetion 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. For tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region, the portion of the tube below 10.3 inches from the top of the hot leg tubesheet or expansion transition, whichever is lower, is excluded when the alternate repair criteria in TS Section 6.8.4.1.2.c.4 are applied. In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring inspection. In addition to meeting the requirements of 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 tubes 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 60 effective full power months. The first sequential period shall be considered to begin after the first

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 24 of 27 Steam Generator Tube Integrity 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 100-percent of all inservice sleeves and sleeve-to-tube joints every 24 effective full power months or one refueling outage (whichever is less).

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. All acceptable tube repair methods are listed below.

1. 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). Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. Prior to installation of each sleeve, the location where the sleeve joints are to be established shall be inspected.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 25 of 27 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS icontinued')

CORE OPERATING LIMITS REPORT (COLRI (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 INSERT 6.9 MARGIN, transient analysis limits, and accident analysis limits) of the safety and 6.9.1.13 analysis are met.

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 report.

6.10 DELETED ST. LUCIE - UNIT 2 6-20e Amendment No. 4=$, 4U48, 4-333,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 26 of 27 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS - 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 of the replacement SGs performed in accordance with Specification 6.8.4.1.1. 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 in each SG, and 6.9.1.13 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection of the original SGs performed in accordance with Specification 6.8.4.1.2. 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 or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired 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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 27 of 27 Steam Generator Tube Integrity

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

The following information concerning indications found in the tubesheet region (including the expansion transition) shall be included in this report:

j. 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.

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

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

St. Lucie Unit 2 L-2007-071 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-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 2 of 24 Steam Generator Tube Integrity INDEX LIMITING CONDmONS FOR OPERATION AND SURVEILLANCE REQUIREMEFN.TS SECTION PAGE 3/4.4.2 SAFETY VALVES DELETED ................................................................................................. 3/4 4-7 OPERATING ............................................................................................ 3/4 4-8 3/4.4.3 PRESSURIZER ................................................................................................. 3/4 4-9 3/4.4.4 PORV BLOCK VALVES ................................................................................. 3/4 4-10 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY .............................................. 314 4-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ........................................................ 3/4 4-18 OPERATIONAL LEAKAGE ................................................................... 3/44-19 3/4.4.7 CHEMISTRY ................................................................................................... 3/44-22 3/4.4.8 SPECIFIC ACTIVITY...................................................................................... 3/44-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM ..................................................................... 314 4-29 PRESSURIZER HEATUP/COOLDOW N LIMITS .............................................. 3/4 4-34 OVERPRESSURE PROTECTION SYSTEMS ................................................. 3/4 4-35 3/4.4.10 REACTOR COOLANT SYSTEM VENTS ...................................................... 3/4 4-38 3/4.4.11 STRUCTURAL INTEGRITY ............................................................................ 314 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS............................................................................ 314 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 325°F ............................................................. 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 3250 F ............................................................... 3/4 5-7 3/4.5.4 REFUELING WATER TANK............................................................................. 3/4 5-8 ST. LUCIE - UNIT 2 V1 Amendment No. 46,4440,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 3 of 24 Steam Generator Tube Integrity INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 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 REPO RT .............................................................................................. 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 XIX Amendment No. 4:3, ", 89, 92.

499, 4U0. 4033

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 4 of 24 Steam Generator Tube Integrity pEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcudes/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 R 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 (in MeV) 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-ismeasured.-.

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 in Table 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. LUCIE - UNIT 2 1-3 Amendment No. 406E,441t7,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 5 of 24 Steam Generator Tube Integrity D9EINITIONS 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, insuch 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 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 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 - UNrT 2 1-5 Amendment No. 9, -$3,64-, 4V.

St. Lucie Unit 2 L-2007-071 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 (SGI TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3A.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. Repair applies only to the original SGs.

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 if original SGs) 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. Repair applies only to the original SGs.

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 following 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. Repair applies only to the original SGs.

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 7 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNrT 2 3144-12 Amendment No. 24,44,4.44,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 8 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-12a Amendment No. 24,44,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 9 of 24 Steam Generator Tube Integritv THIS PAGE DELETED ST. LUCIE - UNIT 2 3(4 4-13 Amendment No,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 10 of 24 Steam Generator Tube Integrity THIS PAGE DELETED I ST. LUCIE - UNIT 2 3/4 4-14 Amnendmnent No. 4144,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 11 of 24 Steam Generator Tube Inte2ritv THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-14a Amendment No. 444,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 12 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-15 Amendment No. 4,,43.444,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 13 of 24 Steam Generator Tube Interitv THIS PAGE DELETED ST. LUCIE -UNIT 2 3/4 4-16 Amendment No.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 14 of 24 Steam Generator Tube Integjity THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4.17 Amendment No. 4-3,444,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 15 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING 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 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 at least once per 24* hours, and restore the required radioactivity monitor to OPERABLE status within 30 days; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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

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

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

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

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

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

ST. LUCIE - UNIT 2 3144-18 Amendment No. 84.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 16 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE 1MITI1NGCONDITION FOR OPERATION 3.4.6.2 Reactor Coolant System operational leakage shall be limited to:

a. No PRESSURE BOUNDARY LEAKAGE.

b. 1 gpm UNIDENTIFIED LEAKAGE,

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

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

e. 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 closed 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 - UNIT 2 3/4 4-`19 Amendment No. 438,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 17 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued')

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. Verifying primary-to-secondary leakage is _<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 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.

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 - UNIT2 3/4 4-2D Armendment No. 7-3,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 18 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS bcontunued*

k. Ventilation Filter Testing Program fVFTP) (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 (SG) Program

1. A SG Program shall be established and implemented for the replacement SGs to ensure that SG tube integrity is maintained. In addition, the SG Program shall include the following provisions:

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

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

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

2. Accident Induced leakage performance criterion: The primary-to-secondary accident induced leakage rate for any design basis accident, other than a 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 gallons per minute total through all 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."

ST. LUCIE - UNIT 2 6-15e Amendment No. Q0,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 19 of 24 Steam Generator Tube Integ-ity ADMINISTRATIVE CONTROLS (continud)

1. Steam Generator (SGI Proqram (continued)

1. (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 tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. inspect 100% of the tubes In each SG during the first refueling outage following SG replacement

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter,-,

60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50%

of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outages nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, 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 A SG Program shall be established and implemented for the original SGs to ensure that SG tube integrity is maintained. In addition, the SG 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 foundr 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 6-15f Amendment No.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 20 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continuedl

1. Steam Generator (SG) Program (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 a 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 gallons per minute total through all 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

1. Tubes found by inservice inspection to contain a flaw in a non-sleeved region with a depth equal to or exceeding 40-percent of the nominal tube wall thickness shall be plugged or repaired except if permitted to remain in service through application of the alternate tube repair criteria discussed in Technical Specification 6.8.4.1.2.c.4.

2. Tubes found by inservice inspection to contain a flaw in (a) a sleeve or (b) the pressure boundary portion of the original tube wall in the sleeve to tube joint shall be plugged.

3. All tubes with sleeves that have a nickel band shall be plugged after one cycle in operation.

4. The C* methodology, as described below, may be applied to the expanded portion of the tube in the hot-leg tubesheet region as an alternative to the 40-percent depth based criteria of Technical Specification 6.8.4.1,2.c.1.

ST. LUCIE - UNIT 2 6-159 Amendment No.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 21 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS icontinued)

1. Steam Generator (SGI Prooram (continued)

2. c, 4. (continued)

i. Tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region shall be repaired or plugged upon detection of any flaw identified within 10.3 inches below the bottom of the hot-leg expansion transition or top of the tubesheet, whichever elevation is lower. Flaws located below this elevation may remain in service regardless of size.

ii. Tubes which have any portion of a sleeve joint in the hot-leg tubesheet region shall be plugged upon detection of any flaw that is located below the lower sleeve to tube joint and within 10.3 inches below the bottom of the hot-leg expansion transition or top of the tubesheet, whichever elevation is lower,

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. For tubes with no portion of a lower sleeve joint in the hot-leg tubesheet region, the portion of the tube below 10.3 inches from the top of the hot leg tubesheet or expansion transition, whichever is lower, is excluded when the alternate repair criteria-in TS Section 6.8,4.1.2.c.4 are applied.-In tubes repaired bysleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring inspection. In addition to meeting the requirements of 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 tubes 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 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 100-percent of all inservice sleeves and sleeve-to-tube joints every 24 effective full power months or one refueling outage (whichever is less).

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 he treated as a crack.

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 22 of 24 Steam Generator Tube Integrity AfMINISTRATIVE CONTROLS (continued)

I. Steam Generator (SG) Program (continued)

2. (continued)

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. Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-1 5918-P Revision 2 (with range of conditions as revised in Appendix A of WCAP-1 6489-NP, Revision 0). Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. Prior to installation of each sleeve, the location where the sleeve joints are to be established shall be inspected.

ST. LUCIE - UNIT 2 6-151 Amendment No.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 23 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

61. WCAP-11397-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 of the replacement SGs performed in accordance with Specification 6.8.4.1.1. 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 in each SG.

ST. LUCIE - UNIT 2 6-20e Amendment No. 406, 448, 433, 438,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 24 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

STEAM GENERATOR TUBE INSPECTION REPORT (continued) 6.9.1.13 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection of the original SGs performed in accordance with Specification 6.8.4.1.2. 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 or repaired during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged or repaired 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.

The following information conceming indications found in the tubesheet region (including the expansion transition) shall be included in this report:

j. 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.

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

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

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 6-20f Amendment No,

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 1 of 19 Steam Generator Tube Integrity TS Bases Markups

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 2 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

314.4 BASES ATTACHMENT 6 OF ADM-25.04 2 of 15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 TABLE OF CONTENTS SECTION PAGE BASES FO R SECTIO N 3/4.4 .............................................................................. 3 3/4.4 REACTOR COOLANT SYSTEM ............................................... 3 BAS E S ..................................................................................... .. 3 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT C IRCULATIO N ....................................................... 3 3/4.4.2 SAFETY VALVES .................................................... 4 3/4.4.3 PRESSURIZER 56 TUBE INTEGRITY 3/4.4.4 PORV BLOCK VALVES ... ............................... 6

......................................... 6 3/4.4.5 STEAM GENERATO 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE ........... 8 x

314.4.6.1 LEAKAGE DETECTION SYSTEMS ..... 8 3/4.4.6.2 OPERATIONAL LEAKAGE ................... i8 3/4.4.7 CHEM ISTRY ........................................................... 9 3/4.4.8 SPECIFIC ACTIVITY ............................................. 10 3/4.4.9 PRESSURE/TEMPERATURE 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-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 3 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

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

BASES (continued) 3/4.4.4 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, or3.

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, it becomes 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 closed.

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 LTOP 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.

3/4.4.5 STEAM GENERATOIý O}-'I(sc, TUBETNTEZ7TY [

Surveillance Requirements for inspection of the steam generator-'

tubes urerznat the structural integrity of this portion of the RCS will be omrntaivne mor e program inservice inspection of steam generator tubes is based o0....modification of Regulatory Guide 1.83, Revision 1.

Inservice Inspection o ear generator tubing is essential in order to maintain --urvela"Ance or conditions of the tubes in the event that there is evidence of mechanicar *.age or progressive degradation due to desgmnuatrng errors, or in ice conditions that lead to corrosion.

Inevc inpcio fsta enrtr tubing also p ies a means of I EP characterizing the nature and cause of any tube degrada iso that R3/.45 orrective measures can be taken. -

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 4 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 7 of 15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 3/4.4 REACTOR COOLANT SYSTEM (continued)

STEAM GENERATOlý ýontinued).y 3/4.4.5 e plant is expected to be operated in a manner such that the secondary co ant ill be maintained within those chemistry limits found to result in neglig e corrosion of the steam generator tubes. If the secondary coolant c mistry is not maintained within these limits, localized corrosion may likely r'ult in stress corrosion cracking. The extent of cracking during plant oiration would be limited by the limitation of steam generator tube le aage between the primary coolant system and the secondary coolant stem (primary-to-secondary leakage = 1.0 gpm from both steam generators. Cracks having a primary-to-secondary leakage less than this limit during eration will have an adequate margin of safety to withstand the loads impo during normal operation and by postulated accidents... Operating plants ha, demonstrated.that-primary-to-secondary leakage of 0.5 gpm per steam ge !rator can readily be detected by radiation monitors of steam generat blowdown. Leakage in excess of this limit will require plant shutdown an ,n unscheduled inspection, during which the leaking tubes will be Ioca d nd plugged.

Wastage-type 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 examinations. scheduled Plugging will be inservice steam gene tor tube required of all tubes wit mperfections exedn h lgiglmto 0 of the tube nominal w thickness.

Stea eeaor tube inspections of operating plants have d ~onstrated ,

the capability to reliably detect degradation \penetrated that has % of the?

riia ue wall thickness.

INSEPT 133/4.4.5

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 5 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: 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) 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary. These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems," May 1973. The LCO is consistent with NUREG-1 432, 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 whensuch leakage indicates possible pressure. boundary...

degradation.

3/4.4.6.2 OPERATIONAL LEAKAGE dustry experience has shown that while a limited amount of leakage is exp d from the RCS, the unidentified portion of this leakage can be reduced threshold value of less than I gpm. This threshold value is sufficiently lo ensure early detection of additional leakage.

The 10 gpm IDENTIFI AKAGE limitation provides allowances for a limited amount of leakage fro nown sources whose presence will not interfere with the detection of UNI TIFIED LEAKAGE by the leakage detection systems.

The Surveillance Requirements for RCS pressure lation valves provide added assurance of valve integrity thereby reducing th obability of gross valve failure and consequent intersystem LOCA. Lea e from the RCS pressure isolation valves is IDENTIFIED LEAKAGE and wil considered as a portion of the allowable limit.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 6 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 9 of 15 REVISION NO.: REACTOR COOLANT SYSTEM 2 ST. LUCIE UNIT 2 3/4.4 REACTOR COOLANT SYSTEM (continued)

BASES (continued) 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE (continued)

"N for all steama 3/4.4.6.2 OPERATIONAL LEAKAGE Iube leakage limit ofe!1 gpm (continuedd

/?ltal steam generato a

fraction of Part 100 limits in the event of either INSERT limited to-"msall genera be ensures that the dosage contribution from The tube limit the1 gpm is will leakage B3/4.4.6.2 r': or steam upture line break.

accidents.\

'(follow s steam generator used in the analysis of these tions Insert for consistent with the ass'en limit pe em generator ensures that steam The 0.5 gpm leakage B3/4.4.5) ons. is unacceptable under LO C A c onditi or any mag "e Srupture LEAKAGE of BOUNDARY heprcessure PRESSURE ndn g gros fma ilur it mayBeO indicAtv ofL EAnitKmpGE s

ince - /

ay Thrfrte pe e c o n RESSURE BO 'SHU ARY N.*

b on the unit to be prom ptly placed in COLD

W L A A E r q ies 314.4.7 CHEMISTRY ensure that Coolant System chemistry The limitations on Reactor System is minimized and reduces the corrosion of the Reactor Coolant due to stress System leakage or failure potential for Reactor Coolant the Steady State Limits chemistry within corrosion. Maintaining the the structural integrity of adequate corrosion protection to ensure associated provides The over the life of the plant.

the Reactor Coolant System chloride and fluoride limits are time and effects of exceeding the oxygen, show that operation may be Corrosion studies temperature dependent. concentration levels in excess of the Steady continued with contaminant Limits, for the specified limited time State Limits, up to the Transient integrity of the without having a significant effect on the structural intervals continued operation The time interval permitting Reactor Coolant System. Transient Limits provides time for taking within the restrictions of the to within the the contaminant concentrations corrective actions to restore Steady State Limits.

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 7 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5

Background

Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. SG tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system.

In addition, as part of the RCPB, the SG tubes are unique in that they act as 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."

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

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

Specification 6.8.4.1., "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.1., tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational leakage.

The SG performance criteria are described in Specification 6.8.4.1. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions. Specification 6.8.4.1. has two parts to address the replacement SG and original SG designs. Specification 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 apply only to the original SG design and can be removed following SG replacement.

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

Applicable Safety Analyses The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding a SGTR is the basis for this Specification. The analysis of a SGTR event assumes a bounding primary-to-secondary leakage rate equal to the operational leakage rate 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-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 8 of 19 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 SG tubes retain their structural integrity (i.e., they are assumed not to rupture). In these analyses the steam discharge to the atmosphere is based on the total primary-to-secondary leakage from all SGs of 0.3 gpm total and 216 gpd through any one SG or is assumed to increase to 0.3 gpm total through all SGs and 216 gpd through any one SG as a result of accident induced conditions. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 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(c)(2)(ii).

Limiting Condition for Operation (LCO)

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

During a SG 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. Tube repair (i.e., sleeving) is applicable only to the original SGs.

In the context of this Specification, a SG 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 S-s, when the alternate repair criteria in ^IS Section 6.8.4.1.2.c.4 are applied a SO 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 SO 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 SG tube has tube integrity when it satisfies the SG performance criteria. The SO performance criteria are defined in Specification 6.8.4.1., "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 9 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-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 versus 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 toWal and 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 SG to 150 gpd at room temperature. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of leakage is due to more than one crack, the cracks are very small, and the above assumption is conservative.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 10 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.5 Applicability SG tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in POWER OPERATION, START UP, HOT STANDBY and HOT SHUTDOWN.

RCS conditions are far less challenging in COLD SHUTDOWN and REFUELING than 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 SG tube. This is acceptable because the required ACTIONS provide appropriate compensatory actions for each affected SG tube. Complying with the required ACTIONS may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

a1 and a.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. Tube repair (i.e., sleeving) is applicable only to the original SGs. An evaluation of SG tube integrity of the affected tube(s) must be made. SG tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program.

The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged or repaired has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, ACTION b applies.

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 11 of 19 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 SG 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 SHUTDOWN following the next refueling outage or SG inspection. This allowable completion time is acceptable since operation until the next inspection is supported by the operational assessment.

b.

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

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

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment 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 SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.4.5.1. The frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection. In addition, Specification 6.8.4.1.

contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 12 of 19 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 1 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 6.8.4.1.2.). Tube repair (i.e., sleeving) is applicable only to the original SGs.

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

References

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

2. 10 CFR 50 Appendix A, GDC 19

3. 10 CFR 100

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

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

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

7. 10 CFR 50.67

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 13 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2

Background

Components that contain or transport the coolant to or from the reactor core make up the reactor coolant system (RCS). Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant leakage, through either normal operational wear or mechanical deterioration. The 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.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the sources of reactor coolant leakage. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS leakage varies widely depending on its source, rate, and duration. Therefore, detecting and 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 facilityand 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 The safety analysis for an event resulting in steam discharge to the atmosphere assumes that primary to secondary leakage from all steam generators (SGs) is 0.3 gpm total through all SGs and 216 gpd through any one SG or is assumed to increase to 0.3 gpm total through all SGs and 216 gpd through any one SG as a result of accident induced conditions. 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 less than or equal to the leakage limit of 216 gpd (measured at operating temperature) through any one SG assumed in the accident analysis. St. Lucie Unit 2 procedures further administratively limit operational leakage with the intent that the accident induced leakage limits will not be exceeded.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 14 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 Primary to secondary leakage is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident. To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The FSAR (Ref. 3) analysis for SGTR assumes the contaminated secondary fluid is released mainly via the safety valves or atmospheric dump valves and only briefly steamed to the condenser. The 0.3 gpm total through all SGs and 216 gpd through any one SG primary to secondary leakage safety analysis assumption is relatively inconsequential.

The SLB is more limiting for site radiation releases. The safety analysis for the SLB accident assumes a value greater than 0.15 gpm primary to secondary leakage through each generator as an initial condition. The dose consequences resulting from the SLB accident are well within the limits defined in GDC 19, 10 CFR 100, 10 CFR 50.67 or the staff approved licensing basis (i.e., a small fraction of these limits).

The RCS operational leakage satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

Limiting Condition for Operation (LCO)

Reactor Coolant System operational leakage shall be limited to:

a. PRESSURE BOUNDARY LEAKAGE No PRESSURE BOUNDARY LEAKAGE is allowed, being indicative of material deterioration. Leakage of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher leakage. Violation of this LCO could result in continued degradation ofthe 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. 4).

The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, "MThe 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 steanm generator tube degradation mechanisms that result in tube

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 15 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 leakage. The operational leakage rate criterion is conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

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

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

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

e. Reactor Coolant System Pressure Isolation Valve Leakage Leakage is measured through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS Leakage when the other is leaktight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable IDENTIFIED LEAKAGE.

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

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

ACTIONS

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

b. UNIDENTIFIED LEAKAGE or 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. Otherwise, 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 following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . This ACTION is necessary to prevent further deterioration of the Reactor Coolant Pressure Boundary.

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 16 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2

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 in each high pressure line having a non-functional valve must be closed and remain closed to isolate the affected line(s). In addition, the ACTION statement for the affected system must be followed and the leakage from the remaining Pressure Isolation Valves in each high pressure line having a valve not meeting the criteria of Table 3.4-1 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 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 full power conditions in an orderly manner and without challenging plant systems. In COLD SHUTDOWN, the pressure stresses acting on the Reactor Coolant Pressure Boundary are much lower, and further deterioration is much less likely.

Surveillance Requirements 4.4.6.2.1 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 performance of a Reactor Coolant System water inventory balance.

aand b.

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

c.

The RCS water inventory balance must be performed with the reactor at steady state operating conditions (stable temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and reactor coolant pump seal injection and return flows). The

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 17 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 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. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and process all necessary data after stable plant conditions are established.

Steady state operation is required to perform a proper water 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. The reactor cavity (containment) sump and containment atmosphere radioactivity leakage detection systems are specified in LCO 3.4.6.1, "Reactor Coolant System Leakage Detection Systems."

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

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

d.

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

e.

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, "Steam Generator Tube Integrity" should be evaluated. The 150-gpd limit is measured at room temperature as described in Reference 5. The operational leakage rate limit applies to leakage through any one steam generator. If it is not practical to assign the leakage to an individual steam generator, all the primary-to-secondary leakage should be conservatively assumed to be from one steam generator.

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

For Reactor Coolant System primary-to-secondary leakage determination, steady state is

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 18 of 19 Steam Generator Tube Integrity Technical Specification Bases Attachment 6 of ADM-25.04 - INSERT B3/4.4.6.2 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 5).

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 ofthe allowed limit.

References

1. 10 CFR 50, Appendix A, GDC 30

2. Regulatory Guide 1.45

3. UFSAR, Section 15.6.3

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

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

St. Lucie Unit 2 L-2007-071 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 19 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 BASES ATTACHMENT 6 OF ADM-25.04 10 of 15 REVISION NO.: REACTOR COOLANT SYSTEM 4 ST. LUCIE UNIT 2 314.4 REACTOR COOLANT SYSTEM (continued) 3/4.4.8 BASES SPECIFIC (continued)

ACTIVITY-J Itotal prima-y-to-secondary leakage through al SG' /

SECFICACIVTYlan 26 alons per day through any one 517 3/44.

The limitationso e specific activity of the primary coolant ensure that the resulting 2 ho doses at the site boundary will not exceed 10CFR50.67 limits following a am generator tube rupture accident in conjunction with an assume eady state primary-to-secondary steam generator leakage rate of 0.3 gpm nd a loss of offsite electrical power. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the St. Lucie site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcurie/gram DOSE EQUIVALENT 1-131, but within the allowable limit shown on Figure 314-1, accommodates possible- iodine; spiking-phenomenonr which may occur following changes in THERMAL POWER.

Reducing Tavg to less than 500OF prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take correction action. Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

v t e Letter Sequence Response to RAI
TAC:MD0105, Steam Generator Tube Integrity (Open)
Initiation Acceptance... Administration Questions Answers 22 January 2007 16 April 2007 Results Approval... Other: ML072080045
MONTHYEAR2007-04-16 [Table View]

L-2007-071, Reply to Request for Additional Information Re Steam Generator Tube Integrity Amendment Request

Text

Background

Background

Navigation menu

L-2007-071, Reply to Request for Additional Information Re Steam Generator Tube Integrity Amendment Request

Text

Background

Background

Navigation menu

Search