Text

Reply to Request for Additional Information, Steam Generator Tube Integrity Amendment Request
	ML070250070
Person / Time
Site:	Saint Lucie
Issue date:	01/22/2007
From:	Johnston G; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2007-003, TSTF-449
	Download: ML070250070 (84)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 January 22, 2007 FPL L-2007-003 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."

Interactions between the NRC staff and FPL resulted in the NRC transmitting a request for additional information (RAI) in the NRC letter from Mr. Brendan T. Moroney to Mr. J. A. Stall dated October 24, 2006. 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-2006-094.

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 o 2007.

Veryuly yours, ordon

!L. ohnsto Site Vice President St. Lucie Plant GLJ/KWF Attachments cc: Mr. William A. Passetti, Florida Department of Health A4 an FPL Group company

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 1 of 13 Steam Generator Tube Integrity REQUEST FOR ADDITIONAL INFORMATION ST LUCIE UNIT 2 STEAM GENERATOR TUBE INTEGRITY TECHNICAL SPECIFICATION AMENDMENT DOCKET NO. 50-389

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 2 of 13 Steam Generator Tube Integrity I1. Proposed Limiting Condition for Operation (LCO) 3.4.5 states that "all SG (steam generator) tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the SG Program." The staff is aware that the Saint Lucie Unit 2 (STL-2) original SGs (OSGs) are to be replaced in 2007, therefore, this LCO applies only to the OSGs. Please discuss your plans to modify this LCO to indicate that repair applies only to the OSGs. For the same reason, discuss your plans to modify LCO 3.4.5.a, LCO 3.4.5.a.2, and Surveillance Requirement (SR) 4.4.5.2. In addition, LCO 3.4.5.a.2 on Page 7 of Attachment 2 is not consistent with LCO 3.4.5.a.2 on Page 6 of Attachment 3 (i.e., "...or repair..." omitted from Attachment 3).

Reply to RAI 1 - These specifications have been modified to clarify that "repair" applies only to the original SGs. The omission on Page 6 of Attachment 3 is also corrected.

2. Regarding proposed Action b for Technical Specification (TS) Section 3/4.4.5, please discuss your plans to replace "next" with "following" to make it consistent with the language used in other portions of your TSs or describe your plans to adopt the wording from TSTF-449 (i.e., be in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ).

Reply to RAI 2 - The requested change has been made.

3. On Page 3/4 4-18 of the proposed TSs, you proposed to include the statement "...per Surveillance Requirement 4.4.6.2.1 .c..." in two places. The purpose of adding this statement in this section is not clear since SR 4.4.6.2.1 .c has no additional details that are not already included in the requirements where this statement is added. Please explain the purpose of adding this statement or discuss your plans to remove this statement from Page 3/4 4-18.

Reply to RAI 3 - This statement has been removed from Page 3/4 4-18.

4. Proposed TS Section 4.4.6.2.1 .e, Page 3/4 4-20, appears to have a typographical error (i.e., it is missing a verb). Please discuss your plans to modify this requirement from "verified *150 gallons per day" to "verified to be *150 gallons per day."

Reply to RAI 4 - TS Section 4.4.6.2.1 .e has been modified to read "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 . **". This closely aligns with TSTF-449 and is consistent with the NRC approved design and licensing basis for St. Lucie Unit 2 for SR 4.4.6.2.1.

5. Proposed TS Section 6.8.4.1.1 references a "Replacement Steam Generator Program" and proposed TS Section 6.8.4.1.2 references an "Original Steam Generator Program."

These references are awkward (since it implies that the program is being replaced

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 3 of 13 Steam Generator Tube Integrity rather than the program applies to the replacement steam generators (RSGs)). Please discuss your plans for removing "replacement" and "original" from these sections since the remaining text in these sections clearly indicate that proposed TS Section 6.8.4.1.1 applies to the RSGs and proposed TS Section 6.8.4.1.2 applies to the OSGs.

Reply to RAI 5 - "Replacement" and "Original" have been removed from these sections.

6. There appears to be a couple of typographical errors associated with proposed TS Sections 6.8.4.1.1.b.2 and 6.8.4.1.2.b.2. Please discuss your plans to correct these apparent errors: (a) there should be an "a" between "than" and "SG", (b) "gpm" should be "gallons per minute", and (c) there should be an "all" between "through" and "SGs."

Reply to RAI 6 - These changes have been made to TS 6.8.4.1.1.b.2 and 6.8.4.1.2.b.2.

7. The last sentence of proposed TS Section 6.8.4.I1. .d, Steam Generator Program for the RSGs, states "...the tube may be susceptible..." Please discuss your plans to modify this TS section to state "...the tubes may be susceptible..." to be consistent with TSTF-449. In addition, discuss your plans to make the same modification to TS Section 6.8.4.1.2.d for the Steam Generator Program for the OSGs.

Reply to RAI 7 - These changes have been made to TS 6.8.4.1. .d and TS 6.8.4.1.2.d.

8. Regarding proposed TS Section 6.8.4.1.2.c, please address the following:

a. As proposed, TS Section 6.8.4.1.2.c.l.ii and iii can be applied as an alternative to the 40-percent depth based criteria. However, since these are not alternatives to the 40-percent depth based criteria, discuss your plans to modify your proposal to indicate that these are not alternatives.

b. Discuss your plans to remove reference to "service induced imperfection, degradation, or defect" in proposed TS Section 6.8.4.1.2.c. 1.ii since these terms are no longer defined in your proposed TS. TSTF-449 uses the term "flaw" rather than imperfection, degradation, or defect.

c. Discuss your plans to remove TS Section 6.8.4.1.2.c.1 since this text is no longer needed given the sentences before it. The staff notes that the reference to TS Section 6.8.4.1.2.f. I in this proposed TS can be moved to TS Section 6.8.4.1.2.c (if needed).

d. Discuss your plans to modify TS Section 6.8.4.1.2.c.l.i to simply indicate that the flaws above a specified distance in the tubesheet (i.e., the distance specified

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 4 of 13 Steam Generator Tube Integrity in your current TSs) must be plugged on detection and all flaws below this distance may remain in service.

e. Given that a tube includes the sleeve, discuss your plans to clarify which criteria apply specifically to the non-sleeved region of the tube.

To address these comments, wording such as the following could be considered:

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 of 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.

Reply to RAI 8 - The wording proposed by the Staff to address comments a. through

e. has been adopted for TS 6.8.4.1.2.c. The following words from existing TS

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 5 of 13 Steam Generator Tube Integrity 4.4.5.4.a.6.ii, however, are appended to the Staff's suggested words for TS 6.8.4.1.2.c.4.ii for clarity and to maintain consistency with the current NRC approved design and licensing basis; ".... in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve to tube joint)."

9. Regarding proposed TS Section 6.8.4.1.2.d, please address the following:

a. Discuss your plans to indicate that the length of the tube is "from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet.

The tube-to-tubesheet weld is not part of the tube."

Reply to RAI 9.a - The additional words suggested have been included.

b. Discuss your plans for indicating that: "In tubes repaired by sleeving, the portion of the original tube wall between the sleeve's joints is not an area requiring inspection."

Reply to RAI 9.b - The additional words suggested have been included.

c. It is not clear that the "Tube Inspection" section of this proposed TS is needed since the criteria in the first paragraph of this specification indicates that the inspections shall be performed to detect flaws that may satisfy the applicable tube repair criteria. Since all flaws are acceptable below 10.3 inches from the top of the tubesheet or the bottom of the expansion transition (whichever is lower), no "inspections" are needed. Discuss your plans for removing the "Tube Inspection" paragraph and reference to it in the first paragraph of this proposed TS.

Reply to RAI 9.c - The "Tube Inspection" paragraph has been removed as suggested. For clarity, however, the words "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 hot leg tubesheet or expansion transition, whichever is lower, is excluded" were added to the first paragraph in TS Section 6.8.4.1.2.d.

d. Discuss your plans to clarify proposed TS Section 6.8.4.1.2.d.3 since your inspections may not coincide with refueling outages (and since it implies that the parent tube behind the sleeve would require an inspection). For example, "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)."

Reply to RAI 9.d - The clarification suggested has been included.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 6 of 13 Steam Generator Tube Integrity

10. Regarding proposed TS Section 6.8.4.1.2.f.1, discuss the reason for adding the following phrase: "(with range of conditions as revised in Appendix A of WCAP-16489-NP, Revision 0)." In addition, discuss your plans for removing the second sentence since TS Section 6.8.4.1.2 only applies to the OSGs. For example, "1.

Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-15918-P, Revision 2. Prior to the installation of each sleeve, the location where the sleeve joints are to be established shall be inspected."

Reply to RAI 10 - The wording used in proposed TS 6.8.4.1.2.f.1 was approved by the Staff in license amendment 144, and the phrase "(with range of conditions as revised in Appendix A of WCAP- 16489-NP, Revision 0)" added a necessary conforming change consistent with Staff approval of license amendment 145 to the St. Lucie Unit 2 TS. FPL has no objection to the new alternative wording proposed by the Staff and has made this change. However, the phrase "(with range of conditions as revised in Appendix A of WCAP-16489-NP, Revision 0)" is retained for consistency with the current NRC approved design and licensing basis for St. Lucie Unit 2.

11. As currently proposed, TS Section 6.9.1.12 would apply to both your OSGs and RSGs. Since it applies to the OSGs, it discusses tube repairs (although the staff notes that the reference to repair is missing from items "e" and "f"). Since tube repair is not applicable to the RSGs, please discuss your plans to clarify what reporting requirements are applicable to the RSGs and which are applicable to the OSGs (alternatively, discuss your plans to make TS Section 6.9.1.12 apply to the RSGs and TS Section 6.9.1.13 apply to the OSGs). Regarding proposed TS Section 6.9.1.13, please discuss your plans to specifically reference TS Section 6.8.4.1.2. In addition, discuss your plans to delete the last two sentences of TS Section 6.9.1.13.c since operation is not permitted when the accident induced leakage limit is exceeded.

Reply to RAI 11 - The alternative approach suggested has been incorporated to modify TS 6.9.1.12 so that it applies to the RSGs and modify TS 6.9.1.13 so that it applies to the OSGs. Further, TS 6.9.1.12 now specifically references TS 6.8.4.1.1, and TS 6.9.1.13 now specifically references TS 6.8.4.1.2. In addition, the last two sentences of proposed TS 6.9.1.13.c are deleted as requested.

12. In the first paragraph on Page 8 of Attachment 4, you indicate that your accident analyses assumes the total primary-to-secondary leakage from all SGs is 0.3 gpm total o__r 216 gallons per day (gpd) through any one SG as a result of accident conditions.

As currently written, it appears that your accident analyses could either be based on 0.3 gpm or 216 gpd (i.e., 0.15 gpm). Please discuss if it was your intent to indicate that your accident analyses assumes the primary-to-secondary leakage from all SGs is 0.3 gpm with no more than 216 gpd coming from any one SG. If so, discuss your plans to modify your Bases to indicate this. The same comment applies to similar statements on Pages 9 and 13 of 17 of Attachment 4 (although Page 13 indicates that

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 7 of 13 Steam Generator Tube Integrity both conditions must be met). Also, discuss your plans to indicate that the accident analysis assumes the leakage could increase to these levels (to make your submittal consistent with TSTF-449).

Reply to RAI 12 - The word "or" has been replaced with "and" to indicate that both conditions must be met. These statements have also been modified to indicate that the accident analysis assumes the leakage could increase to these levels.

13. Proposed Bases Section 3/4.4.6 discusses repairs which applies only to the OSGs. In addition, proposed Bases Section 3/4.4.5, Steam Generator Tube Integrity, discusses tube repairs which only applies to the OSGs. Please discuss your plans to modify your Bases to indicate that repair only applies to the OSGs.

Reply to RAI 13 - The suggested changes were made to the proposed Bases Section 3/4.4.5. However, no references to repair or sleeving are present in proposed Bases Section 3/4.4.6.

14. In the first paragraph on Page 9 of Attachment 4, there appears to be a typographical error, "verses" should be "versus". Please discuss your plans to correct this typographical error.

Reply to RAI 14 - The correction has been made.

15. On Page 11 of Attachment 4, you indicate that the affected tubes(s) must be plugged or repaired prior to entering Hot Standby. Please discuss your plans to modify this statement since the corresponding requirement to plug the tube(s) is before entering Hot Shutdown.

Reply to RAI 15 - "HOT STANDBY" has been changed to "HOT SHUTDOWN."

16. You made several changes to Bases Section 3/4.4.6, Reactor Coolant System Leakage, that go beyond TSTF-449. Please confirm that all of the proposed changes are consistent with your current U.S. Nuclear Regulatory Commission (NRC) approved design and licensing basis. If they are not consistent, please provide the technical justification or discuss your plans to remove them.

Reply to RAI 16 - TSTF-449 uses the standard technical specifications (NUREG-1432) as a starting bases document and details a number of changes to Bases Section B3.4.13, "RCS Operational LEAKAGE." In our May 25, 2006 submittal, FPL adopted NUREG-1432 as the starting bases document and modified it to incorporate both the changes introduced by TSTF-449 and those changes necessary to maintain consistency with the current NRC approved design and licensing basis for St. Lucie Unit 2. The proposed bases changes that go beyond TSTF-449 are discussed below.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 8 of 13 Steam Generator Tube Integrity A few changes that are in addition to those proposed in our May 25, 2006 submittal are also discussed below.

" In our May 25, 2006 submittal, the third paragraph in the Background section of NUREG-1432 B3.4.13 was omitted (i.e., beginning with "10CFR50, Appendix, A, GDC 30 (Ref.l)..."). Upon further review it was determined that this paragraph is consistent with the current NRC approved design and licensing basis for St. Lucie Unit 2 and was added in response to this RAI. Conforming changes were also made to add the references for this paragraph.

In our May 25, 2006 submittal, the first, second and third paragraphs in the Applicable Safety Analyses section of NUREG-1432 B3.4.13 were omitted. Upon further review and in response to this RAI, these paragraphs are added to St. Lucie Unit 1 Bases section B3/4.4.6.2. The first paragraph is modified to be consistent with the current NRC approved design and licensing basis. The second paragraph is added without modification. The third paragraph is modified to be consistent with the St. Lucie Unit 2 accident analyses and the current NRC approved design and licensing basis.

" In our May 25, 2006 submittal, the fourth paragraph in the Applicable Safety Analyses of NUREG-1432 B3.4.13 was modified to be consistent with the current NRC approved design and licensing basis, which includes 10 CFR 50.67 (also see the response to RAI #20).

In our May 25, 2006 submittal, the first line in the Limiting Condition for Operation (LCO) of NUREG-1432 B3.4.13, "RCS" was changed to "Reactor Coolant System" to maintain consistency with the format of the current TS.

In our May 25, 2006 submittal, the second paragraph under c. Identified LEAKAGE in the Limiting Condition for Operation (LCO) in NUREG-1432 B3.4.13 (i.e., beginning "LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage.. ."), was substantially modified and relocated to St. Lucie Unit 2 Bases B3/4.4.6.2.e. Bases section B3/4.4.6.2.e is subsequently replaced in this RAI response using the wording in the second paragraph under c. Identified LEAKAGE in NUREG-1432 B3.4.13, and modified slightly to be consistent with the current NRC approved St. Lucie Unit 2 design and licensing basis.

" In our May 25, 2006 submittal, the first paragraph of NUREG-1432 Bases SR 3.4.13.1 is used as the first paragraph for St. Lucie Unit 2 Bases SR 4.4.6.2.1. The remaining paragraphs in NUREG-1432 Bases SR 3.4.13.1 are used in item c under St. Lucie Unit 2 Bases SR 4.4.6.2.1. Items a, b, and d under St. Lucie Unit 2 Bases SR 4.4.6.2.1 were added to address monitoring of containment sump, containment atmosphere radioactivity

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 9 of 13 Steam Generator Tube Integrity levels, and reactor head flange leak-off to maintain consistency with the current NRC approved St. Lucie Unit 2 design and licensing basis.

In our May 25, 2006 submittal, LCO ACTION b did not address the need to be in 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 in the extent unidentified or identified leakage can not be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The staff identified this in RAI #8 in their letter dated August 22, 2006 regarding the St. Lucie Unit 1 request to adopt TSTF-449 (FPL letter L-2006-089, April 24, 2006).

Therefore, a conforming change is made to LCO ACTION b for St. Lucie Unit 2.

" An editorial change is made with this RAI response to the third paragraph under SR 4.4.6.2.1 .c to clarify which leakage detection systems are specified in LCO 3.4.6.1, "Reactor Coolant System Leakage Detection Systems."

An editorial change is made with this RAI response to the first paragraph under SR 4.4.6.2.1 .e to add "Steam Generator Tube Integrity" after LCO 3.4.5. Another editorial change is made to the third paragraph under SR 4.4.6.2.1 .e to change "LEAKAGE" to "leakage" in two places for consistency with the current format of the St. Lucie Unit 2 licensing basis.

In our May 25, 2006 submittal, SR 4.4.6.2.2 was included to address existing requirements for RCS Pressure Isolation Valve check valve integrity.

In our May 25, 2006 submittal, SR 4.4.6.2.3 was included to address existing requirements for RCS Pressure Isolation Valve motor-operated valve integrity.

17. Please provide justification for omitting the following paragraph from proposed Bases Section 3/4.4.6.2, Reactor Coolant System Operational Leakage.

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

Alternatively, discuss your plans to incorporate this information into this bases section.

Reply to RAI 17 - The above information has been incorporated into this bases section.

18. On Page 13 of Attachment 4, proposed Bases Section 3/4.4.6.2 states, "Therefore, monitoring reactor coolant leakage into the containment area is necessary." Please

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 10 of 13 Steam Generator Tube Integrity discuss your plans to modify this Bases section by stating "Therefore, detecting and monitoring reactor coolant leakage into the containment area is necessary."

Reply to RAI 18 - The words "detecting and" have been added.

19. Regarding the Applicable Safety Analyses Section on Page 13 of Attachment 4, please address the following:

a. In the second sentence, the word "all" appears to be missing between "through" and "SGs." Discuss your plans to correct this apparent typographical error.

Reply to RAI 19.a - The word "all" has been added.

b. Confirm that your proposed 150 gpd operating leakage limit (measured at room temperature) is less than or equal to your accident analysis assumption of 216 gpd (measured at operating temperature). If it is not, please discuss your plans to modify your operating leakage limit to be less than (or equal to) your accident induced leakage limit. In addition, discuss your plans to modify/delete the fifth sentence of this section.

Reply to RAI 19.b - The discussion in the fifth sentence has been modified to state that the proposed 150 gpd operating leakage limit is less than or equal to the accident analysis assumption of 216 gpd (measured at operating temperature).

c. Discuss your plans to remove the sixth sentence of this section since it adds no insights into the basis for your TSs.

Reply to RAI 19.c - The sixth sentence has been removed.

d. Discuss your plans to remove the last half of the seventh sentence of this section since it is not clear that simply limiting the operating leakage to half the value assumed in the accident analysis will always be sufficient at ensuring the accident induced leakage limit will be met (as evidenced by operating experience). The staff notes that it will depend, in part, on the nature of the flaw that is leaking.

Reply to RAI 19.d - The last portion of this sentence (i.e., "... to half the value assumed in the accident analysis") has been removed from the seventh sentence.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 11 of 13 Steam Generator Tube Integrity

20. On Page 6 of Attachment 4, your current Bases state that the dosage contribution from the tube leakage will be limited to a small fraction of Title 10 of the Code of Federal Regulations (10 CFR) Part 100 limits. Then, on Page 13 of the same Attachment, your proposed Bases state that the dose consequences are within the limits of 10 CFR Part 50, Appendix A, General Design Criteria (GDC) 19, 10 CFR Part 100, 10 CFR Part 50.67, or the NRC approved licensing basis. Please clarify whether your current NRC approved accident source term is based on Part 100 (which is referenced in your current TS Bases), 10 CFR 50.67, or both. On Page 8 of Attachment 4 in your submittal, the last sentence of the first paragraph currently reads: "...or the NRC approved licensing basis (e.g., a small fraction of these limits)." Then, on Page 13 of the same Attachment, the same sentence appears but the statement included in the parenthesis is missing. Please discuss your plans to add this statement in this section.

Reply to RAI 20 - Our current NRC approved source term is based on both Part 50.67 and Part 100. For steam generator tube rupture, which is approved with the Alternative Source Term (AST) methodology, the source term is based on 10 CFR Part 50.67, whereas for all other dose events, including the steam line break, the source term is based on 10 CFR Part 100 for offsite dose and GDC 19 for control room dose. The proposed Bases statements on pages 8 and 13 refer to all dose events and, therefore, the statements cover all applicable dose limits, which include a small fraction of 10 CFR 100 for other events, such as the feedwater line break event. The statement in the parenthesis on page 8 was added to the same sentence on page 13.

21. Regarding the second paragraph under Item "c" on Page 14 of Attachment 4, please discuss your plans for removing this paragraph since it is (for the most part) redundant to the paragraph in the Applicable Safety Analysis section. In addition, the staff concluded that the 150 gpd operating leakage limit is acceptable since it is an effective measure for limiting the frequency of tube ruptures. Although there is a statement in the Bases associated with TSTF-449 that the 150 gpd leakage limit is less than the conditions assumed in the accident analysis, this is just an observation reflecting that previously the normal operating leakage limit was consistent with the accident-induced leakage limit and now it was "significantly less" than the accident induced leakage limit.

Reply to RAI 21 - The second paragraph under Item "c" on Page 14 of Attachment 4 has been deleted.

22. In Bases Section 3/4.4.6.d on Page 14 of Attachment 4, there appears to be a typographical error in the first sentence. It appears that "with" should be "within."

Reply to RAI 22 - The typographical error has been corrected.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 1 Proposed License Amendment Page 12 of 13 Steam Generator Tube Integrity

23. In Action c on Page 15 of Attachment 4, you have inserted several statements that go beyond the corresponding Action statements in your TSs (e.g., adding check valves).

Please discuss your plans for removing these extra statements/clarifications or demonstrate that these statements/clarifications are consistent with your currently approved design and licensing basis.

Reply to RAI 23 - FPL concurs with the NRC observation and has removed the discussion of check valves in the TS bases section on PIV leakage.

24. On Page 15 of Attachment 4, the first "within" should be deleted from Action d.

Reply to RAI 24 - The requested change has been made.

25. On Page 15 of Attachment 4, there appears to be a typographical error in the last sentence of the proposed Bases for SR 4.4.6.2.1. It appears that "or" should be "of."

Reply to RAI 25 - The typographical error has been corrected.

26. Under SR "a and b." on Page 16 of Attachment 4, you indicate that gaseous or particulate radioactivity monitor and the containment sump level must be monitored.

Since the associated SRs indicate that gaseous and particulate monitors should be monitored, please discuss your plans to make your Bases consistent with the SR. In addition, since the associated SR indicates that containment sump inventory and discharge should be monitored, please discuss your plans to make your Bases consistent with the SR.

Reply to RAI 26 - Both of the requested changes have been made to make the Bases consistent with the SR

27. A discussion was provided for the RCS pressure isolation valve leakage in proposed Bases Section 3/4.4.6.2.e. This discussion does not correspond with RCS pressure isolation valve leakage discussion in the standard TSs. Please discuss your plans to incorporate the standard TS RCS pressure isolation valve leakage into this bases or discuss why it is not appropriate to do so.

Reply to RAI 27 - FPL does not plan to incorporate the ISTS bases for PIVs. Unlike the ISTS, the St. Lucie TSs do not have a stand-alone PIV TS LCO. FPL included a PIV discussion in the proposed TS Bases changes to maintain consistency with the level of detail included in the TSTF Bases for the SG integrity program. However, this LAR made no changes to the existing St. Lucie TS PIV requirements and, therefore, it is not appropriate to incorporate the ISTS bases for PIVs.

St. Lucie Unit 2 L-2007-003 DocketNo. 50-389 Attachment 1 Proposed License Amendment Page 13 of 13 Steam Generator Tube Integrity

28. Please provide justification for omitting the following information from SR 4.4.6.2.1 .c in proposed Bases Section 3/4.4.6.2.

"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."

Alternatively, discuss your plans to incorporate this information into the bases.

Reply to RAI 28 - This wording has been added to the Bases for SR 4.4.6.2.1 .c.

29. Please discuss your plans to modify SR 4.4.6.2.1 .c in proposed Bases Section 3/4.4.6.2 to state, "Steady state operation is required to perform a proper water inventory balance ..."

Reply to RAI 29 - The requested change has been made.

St. Lucie Unit 2 L-2007-003 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-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 2 of 27 Steam Generator Tube Integrity INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES DELETED ............................................................................................ 3/4 4-7 O PERATING ............................................................................................ 3/4 4-8 3/4.4.3 PRESSU RIZER ............................................................................................... 3/4 4-9 3/4.4.4 PORV BLOCK VALVES ..... ...................................... 3/4 4-10 3/4.4.5 STEAM GENERATO ............... ............. . .......... 3/4 4-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ....................................................... 3/44-18 OPERATIONAL LEAKAGE .................................................................... 3/44-19 3/4.4.7 C HEM ISTRY .......... :....................................................................................... 3/4 4-22 3/4.4.8 SPEC IFIC ACTIVITY ...................................................................................... 3/4 4-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM .................................................................... 314 4-29 PRESSURIZER HEATUP/COOLDOWN 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 ............................................................................ 3/4 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ........................................................................... 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 325°F .............................................................. 3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 325°F .............................................................. 3/4 5-7 3/4.5.4 REFUELING WATER TANK ............................................................................. 3/4 5-8 ST. LUCIE - UNIT 2 Vi Amendment No. 4-6,J

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 3 of 27 Steam Generator Tube Integrity INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION ......................................................................... 6-13 6.7 SAFETY LIM IT VIO LATIO N .................................................................................. 6-13 6.8 PROCEDURES AND PROGRAMS ....................................................................... 6-13 6.9 REPORTING REQUIREMENTS ........................................................................... 6-16 6.9.1 RO UTINE REPO RTS ........................................................................................... 6-16 STARTUP REPO RT .............................................................................................. 6-16 ANNUAL REPO RTS ............................................................................................. 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 REPORTS ........ .............................. .................................. 6-20e 1ý 6.10 DELETED.................................................. 6-20e 6.11 RADIATION PROTECTION PROGRAM.............................................. 6-21 ISTEAM GENERATOR~ TUBE INSPECTION REPORT 6-20e 1 ST. LUCIE - UNIT 2 XlX Amendment No. 4-3, 614-, Pg, 2,

4 W0 . -4 U8

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 4 of 27 Steam Generator Tube Integrity DEFINITIONS DOSE EQUIVALENT 1-131 1.10 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (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. -15,..J3 primary-to-secondy leaka e

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 5 of 27 Steam Generator Tube Integrity DEFINITIONS PRESSURE BOUNDARY LEAKAGE i -to-secondarJy 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except eratort 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 when the monitored parameter exceeds its trip setpoint at the channel sensor Y

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 0-rings) is OPERABLE.

ST. LUCIE - UNIT 2 1-5 Amendment No. 9, 4-3,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 6 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 314.4.5 STEAM GENERATOIIS 1(SG) TUBE INTEGRIITY]

LIMITING CONDITION FOR OPERATION 3.4.5 Each steam ge gfPRBLE. INERTIA APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:---

Wi on r inoerable, restore the inoperable generator(s) toOPERABLE status prior to increasing SURVEILLANCE REQUIREMENTS r'14qs.ýenerator shall be demonstrated OPERABLE by performance of~r' the following au ' nevc 4... **teez*r Samp~le Selection annspec on - each steam generator*

thseietermineds OPERABL t doa selecting and inspecting at minimum number of steam generators speci eastthe

I'NSERT_C 3tear Generato ueo ampde eecion pection - The steam generator tube mn mm sample size, inspection result classification, and the cresponding actzion re ired shall be as specified in Table 4.4-2. The insetce inspection of ste'wq generator tubes shall be performed at the frequeh ies specified in Spe~fction 4.4.5.3 and the inspected tubes shall be verifie cceptable per the acc-'kpance crit~eria of Specification 4.4.5.4.

The tubes se-'1ed for each inservicelrvpection shall include at least 3% of the total nmber bftubes in all steam genLr*ktors; the tubes selected for

a. hr x iec nsmlar plants With* i ar water chemistry inicts criti a~reas to be inspected, then a a50% st of the tubes inspected s efro these critical areas. "

b. Thb is sml f ueele o ah inservice inspection

~(subsequent to the preservice inspection) o team generator/.

shall include:

ST. LUCIE -UNIT2 3/4 4-11

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 7 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM SRVEILLANCE REQUIREMENTS (Continue~d' CE REQUIREMENTS (Continued) I J

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

2. Tubes in those areas where experience has indicated potential/

problems.

q A f~h.

-4i-.f Q.;r' 4;r- A aA roil1 V

.. I "t"',/". .

... U I .1 c T tbessel peted nach selcdad thie I any selected t does

,onta perm. t the passageroice eddy ourre nt probe for tube relocated as in'*ection, this shall be recorded and an adjacent be shall shown later in be s elected and subjected to a tube insp)ection./

pr po ed4. 11l Inse ijceLeak Limiting Alloy 800 sle (

TS 6.8.4.1.2.d.3 J

"* their full le th during each refueling out incude toe tube and the sleeve. / ýI

c. The tubes selected a khe second and thi, dsa imples (if required by Table 4.4-2) during eac inservice insp ion partial tube inspection pro 'ded:/ may be subjected to

1. The tubes selected fort s samples include the tubes from those areas of the tube s t array wher e tubes with imperfections were prev us found.

2. The inspections :ions of the tubes where imperfections w(

The results of each sample i ne of the following three categ o.r es:.

C..aatego Inspection Result 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

ST. LUCIE - UNIT 2 3/44-12 Amendment No. 24,_4ý,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 8 of 27 Steam Generator Tube Intearitv Inspection Results Note: (1) In all inspections, p viously degraded tubes must i significant (greatethan 10%) further wall penetratli be included inte above percentage calculations.

THIS PAGE DELETED ST. LUCIE - UNIT 2 314 4-12a Amendment No. 2-4)-

St. Lucie Unit 2 L-LUU / -UU-j Docket No. 50-389 Attachment 2 Proposed License Amendment Page 9 of 27 Steam Generator Tube Integrity 4.4.5.N3 Inspection Frequencies - The above required inservice inspections of /

steam g nerator tubes shall be performed at the following frequencies:/

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

b. If the results of the i ervice inspection of a eam 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 e in inspection frequency shall apply until the subseq nt ins ctions satisfy the criteria of Specification 4.4.5.3a.; the mt may then be extended to a maximum of once per 40 mont

c. Additional, unscheduled in rvice in ections shall be performed on each steam generator in rdance 'h the first sample inspection specified in Table 4.4- uring the shut wn subsequent to any of the following conditio s:

1 Primar-to- condary tubes leaks (not in uding leaks originatin from tube-to-tube sheet welds) excess of the limits o pecification 3.4.6.2.

2. A s ismic occurrence greater than the Operating asis

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

4. A main steam line or feedwater line break.

THIS PAGE D)ELETEr)

ST. LUCIE - UNIT 2 314 4-13

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 10 of 27 Steam Generator Tube Intearity REACTOR COOLANT SYSTEM SRVEILLANCE REQUIREMENTS (Continued)/

4.4.5. Acceptance Criteria As used in this Specification

1. Imperfection means an exception to the dimensions, finish or contour of a tube from that required by fabrication drawings r specifications. Eddy-current testing indications below 200 of the nominal tube wall thickness, if detectable, may be sidered s imperfections.

2. De cdation means a service-induced cracking, stage, wear or gener corrosion occurring on either inside or tside of a tube.

3. De raded be means a tube containing i perfections greater than or equal 20% of the nominal wa thickness caused by degradation.

Modif led and reloca ted as 4. % Degradation mea s the perce ge of the tube wall thickness shown later in affected or removed b degrad n.

propossed 5. Defect means an imperfe 'n of such severity that it exceeds TS 6.18.4 .1.2c the plugging limit. A tube taming a defect is defective.

X 6. fPlugin or Repair Li it means-e condition at or beyond which the tube shall be removed fr service by ugging or repaired by sleeving using the method in Sp ification 4.4.5.4. 10 in the affected area. The plugging or repair limits e as follows:

i. In the on-sleeved portion of a tube, e plugging or repair limit imp ection depth is 40% of the nomi I wall thickness. This Limit is ot applicable in the portion of the tub that is greater than 0.3 inches below the bottom of the hot le expansion transition or top of the tubesheet (whichever is lower) to t" tube end.

Degradation detected between 10.3 inches beaw the bottom of the hot leg expansion transition or top of the tubeshe t (whichever is 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 ction of any service induced imperfection, degradation or defect in t (a) sleeve or (b) pressure boundary portion of the original tube w I in the sleeve/tube assembly (i.e., the sleeve-to-tube joint).

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

ST. LUCIE - UNIT 2 3/4 4-14 Amendment No.',ý _

THIS PAGE DELETED

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page I11 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM RVEILLACE REQUIREMENTS (continued) 7*7. Unserviceable describes the condition of a tube if it leaks or /

contains a defect large enough to affect its structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant/

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

8. ue Inspection for a tube with no portion of a sleeve ex. nding below "

10.3 inches from the bottom of the hot leg expansion transition or the topk Modifed an ' *the tubesheet (whichever is lower) means an insp, tion of the steam igead Modf .erator tube from 10.3 inches below the bottom~d the hot leg expansior!

reloatedas f.-----tran 'bon or top of the tubesheet (Whichever is 11 er) completely around relcaedasthe U- end to the top support of till col leg.. 7be Inspection for a tube shown later in with With a on ofa sleeve extending below 1 .. inches from the bottom of proposed the hot legxpansion transition or the topjf the tubesheet (whichever is/

TS 6.8.4.1.2.d lower) mean an inspection from the bo m of the sleeve completely

,.around the U-b nd to the top support~p the cold leg. ,,

9. Preservice-Inspe n means an i 'pectionof the full length of each tube in each st m gener Yr performed by eddy current techniques prior to serice to ,etablish a baseline condition of the tubing. Ti 'ynspection shall be performed after the field hydrostatic ;St and prior to initial POWER OPERATION using the ~qui ent and techniques expected to be ST. LUCIE - UNIT 2 3/4 4-14a Amendment No.,..44L-

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 12 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM

10. Thin 15 ay fown th oleg with Westinghouse Leak LimitingAlloyn conditions as revised in Appendix A of WCAP-16489-NP, Revisionrt which are used to maintain a tube in service. Leak Limiting Allo600 Modified and Sleeves are applicable only to the original ts steam generators. et S pressure boundary portion of the o1ginal tube wall in the slee /tube Sassembly (i.e., the sleeve-to-tube joint) shall be intpeted . or toe new proposed ndtallation of each sleeve.tio ofsteeesam gegenerator id ieatn esratrbed, shall thenume be determinedof tub8 pluggedto Rvsio OPERABLE afte reaired platendinoac cthesponding actions (plug or repair all tubes exces If the Pliminao Repasi it and all tubes containing through-wall crac required by nTabrestar.4-

a. Within 15 days fo wing the completion of ea inservice inspection steam generator shallxb the number reported of tus to the repaired Report 6mnmissio*n in a Special in each pursuant to Specificati 6.9.2. /

b. The complete results of the team g efirator tube inservice inspection shall be submitted to the Corh iss,in in a Special Report pursuant to Specification 6.9.2 within 12 mn s following completion of the Modiiedand inspection. ThislSpecial Repo* alhlinclude:

Modiiedand 1. Number aS684.~. and extent oube s an sleeves inspected.

fuciolfelvtinwthnthnueset Relocated to 2- Loationand per nt of wall-thickn ss penetration for each 3.Identificatioryf tubes plugged or repar

c. Following eac i~nspection and within 120 days a *r the reactor coolant system reenters MO/E 4, the following information conce, i g indications found in the tubesheet 'egion (including the expansion transition) *all be reported to the Commis. ion in a special report pursuant to SpecifiCatio l6.9.2. This Special R e p o h all in c lu de : _ _ _

1.. Number of total iniatos loctio of Aea ndictomentatNo.

- of3 each$4* -

indication, sevrit oeahidctnndwethertei'ciosntated

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 13 of 27 Steam Generator Tube Inteeritv 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 SGtubes satisfying the tube repair criteria and not plugged (or repaired if original SGs) in accordance with the Steam Generator Program;

. 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.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 14 of 27 Steam Generator Tube Integritv THIS PAGE DELETECt 0I I (UD U) OC ) )

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St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 15 of 27 Steam Generator Tube Inteaitv

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THIS PAGE DELETED

St. Lucie Unit 2 L-2007-003 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 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: t

a. With the required reactor cavity sump inlet flow monitoring s tem 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 21hours, and restore the required radioactivity monitor to OPERABLl status within 30 days; otherwise, be in at least HOT STANDBY within the n~xt 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . n]

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 3/4 4-18 Amendment No."-ý

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 17 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.6.2 Reactor Coolant SystemIeakage shall be limited to: [op ai

a. No PRESSURE BOUNDARY LEAKAGE,

c. m totalprimary-to-secondaryIake throug1 g 110gallons~per d 150 "an~I day *generators and 2-tga lons per ay throug-i* any onesem generator

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. ar with primary-to-secondary leakage or wth pimar-to-econary eakae no witin Imt ACTION: I

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

____atna twithin 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 Systemn leakage greater tba any one of the primary-to-secondary limits, excluding PRESSURE BOUNDARY LEAKAcG,. 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.

SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant Systerleakages shall be demonstrated to be within each of the above limits by: _

a. Monitoring the .i.nment 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 N.'11

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 18 of 27 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS (Continued)

V 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.

c.

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 inthe previous 9 months, Prior to returning the valve to service following maintenance, repair or replacement work on the valve, clb

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 <115 gallons per day through any one steam generator at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . **

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

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

ST. LUCIE - UNIT 2 3/4 4-20 Amendment No

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 19 of 27 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS fcontlnued)

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

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

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

INSERT 6.8.4.1.

ST. LUCIE - UNIT 2 6-15e Amendment No. Vlý

St. Lucie Unit 2 L-2007-003 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) 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 L-2007-003 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 SO replacement.

2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective foll 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 SO 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 SO 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-2007-003 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 thana 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.

I. 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-2007-003 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 of 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 in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve to tube joint).

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 hot leg tubesheet or expansion transition, whichever is lower, is excluded.

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 inservice inspection of the SGs. No SG shall operate for more than 24

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 24 of 27 Steam Generator Tube Integrity 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-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 25 of 27 Steam Generator Tube Intemrity ADMINISTRATIVE CONTROLS (continued)I CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

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

April 1989.

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

October 1999.

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

64. 30% SGTP PLA Submittal and the SER.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SHUTDOWN INSERT 6.9..112 MARGIN, transient analysis limits, and accident analysis limits) of the safety and 6.9.1.13 analysis are met.

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

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

6.10 DELETED ST. LUCIE - UNIT 2 6-20e Amendment No. 405, 449,4.33,

St. Lucie Unit 2 L-2007-003 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. L.ocation, 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 or repaired to date,

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

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

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 2 Proposed License Amendment Page 27 of 27 Steam Generator Tube Integrity L Repair method utilized and the number of tubes 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-2007-003 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-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 2 of 24 Steam Generator Tube Intea-itv INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES DELETED ................................................................................................. 3/4 4-7 OPERATING ............................................................................................ 3/4 4-8 3/4.4.3 PRESSURIZER ................................................................................................. 3/4 4-9 3/4.4.4 PO RV BLOCK VALVES .................................................................................. 3/4 4-10 314.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY .............................................. 3/4 4-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS ........................................................ 3/4 4-18 OPERATIONAL LEAKAGE ..................................................................... 3/4 4-19 3/4.4.7 CHEMISTRY ................................................................................................... 3/4 4-22 3/4.4.8 SPECIFIC ACTIVITY ...................................................................................... 3/4 4-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM ..................................................................... 3/4 4-29 PRESSURIZER HEATUP/COOLDOW N LIMITS .............................................. 3/4 4-34 OVERPRESSURE PROTECTION SYSTEMS ................................................. 3/44-35 3/4.4.10 REACTOR COOLANT SYSTEM VENTS ........................................................ 3/4 4-38 3/4.4.11 STRUCTURA L INTEGRITY ............................................................................ 3/4 4-39 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS ............................................................................ 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 325OF ........................................................... 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 VI Amendment No. 46, 4-9,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 3 of 24 Steam Generator Tube Inteanitv INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.6 REPORTABLE EVENT ACTION ........................................................................... 6-13 6.7 SAFETY LIMIT VIOLATION .................................................................................. 6-13 6.8 PROCEDURES AND PROGRA MS ....................................................................... 6-13 6.9 REPORTING REQUIREMENTS ........................................................................... 6-16 6.9.1 ROUTINE REPORTS ............................................................................................ 6-16 STARTUP REPORT .............................................................................................. 6-16 ANNUAL REPORTS ............................................................................................ 6-16 MONTHLY OPERATING REPORTS ..................................................................... 6-17 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT .................................. 6-18 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT ............... 6-19 CORE OPERATING LIMITS REPORT (COLR) ..................................................... 6-20 STEAM GENERATOR TUBE INSPECTION REPORT ........................................ 6-20e 6.9.2 SPECIAL REPORTS .......................................................................................... 6-20e 6.10 DELETED ............................................................................................................ 6-20e 6.11 RADIATION PROTECTION PROGRA M ............................................................... 6-21 ST. LUCIE - UNIT 2 XlX Amendment No. 43,64,89, 92, 499,448, 433,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 4 of 24 Steam Generator Tube Integrity DEFINITIONS 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 Eshall 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 (primary-to-secondary leakage).

ST. LUCIE - UNIT 2 1-3 Amendment No. ,4@5,42,1W,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 5 of 24 Steam Generator Tube Integrity PEFINITIONS PRESSURE BOUNDARY LEAKAGE 1.22 PRESSURE BOUNDARY LEAKAGE shall be leakage (except primary-to-secondary leakage) through a non-isolable fault in a Reactor Coolant System component body, pipe wall or vessel wall.

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

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

RATED THERMAL POWER 1.25 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2700 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 0-rings) is OPERABLE.

ST. LUCIE - UNIT 2 1-5 Amendment No. 4,4,-15,-1-, 7,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 6 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY LIMITING CONDITION FOR OPERATION 3.4.5 SG tube integrity shall be maintained AND All SG tubes satisfying the tube repair criteria shall be plugged or repaired in accordance with the SG Program. 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-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 7 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-12 Amendment No. 24, 48,444,

St. Lucie Unit 2 L-2007-003 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, 48,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 9 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNIT 2 314 4-13 Amendment No.

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St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 13 of 24 Steam Generator Tube Integrity THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-16 Amendment No.

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 14 of 24 Steam Generator Tube Integ'itv THIS PAGE DELETED ST. LUCIE - UNIT 2 3/4 4-17 Amendment No. 4-,444,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 15 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM 314.4.6 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 REOUIREMENTS 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. 84,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 16 of 24 Steam Generator Tube Integrity REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION 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 4-1 9 Amendment No. 4,39,

St. Lucie Unit 2 L-2007-003 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 - UNIT 2 3/4 4-20 Amendment No. 7-4,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 18 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

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

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

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

I. Steam Generator (SGI Proqram

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. 4-39,

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 19 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

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

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

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 20 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

L 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.L.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 of 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.L.2.c.1.

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

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 21 of 24 Steam Generator Tube Integrity ADMINISTRATIVE CONTROLS (continued)

1. Steam Generator (SG) Pro-gram (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 in the (a) sleeve or (b) pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve to tube joint).

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 hot leg tubesheet or expansion transition, whichever is lower, is excluded. 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 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.

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

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 22 of 24 Steam Generator Tube Integrity ADMINISTRATIVE 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-2007-003 Docket No. 50-389 Attachment 3 Proposed License Amendment Page 23 of 24 Steam Generator Tube Intearity ADMINISTRATIVE CONTROLS (continued)

CORE OPERATING LIMITS REPORT (COLR) (continued)

b. (continued)

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

April 1989.

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

October 1999.

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

64. 30% SGTP PLA Submittal and the SER.

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

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

STEAM GENERATOR TUBE INSPECTION REPORT 6.9.1.12 A report shall be submitted within 180 days after the initial entry into HOT SHUTDOWN following completion of an inspection 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. 405, 448, 433, 438,

St. Lucie Unit 2 L-2007-003 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 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 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.

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-2007-003 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-2007-003 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 2 of 19 Steam Generator Tube Integrity SECTION NO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

3/4.4 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 FOR SECTION 3/4.4 .............................................................................. 3 3/4.4 REACTOR COOLANT SYSTEM ................................................ 3 BA S ES ..................................................................................... 3 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION ................................. 3 3/4.4.2 SAFETY VALVES .................................................... 4 3/4.4.3 PRESSURIZER ...........................

SC7TRFs nTm&rr-OT 5 3/4.4.4 PORV BLOCK VALVES?- ............................. 6 3/4.4.5 STEAM GENERATO. ......................................... 6 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE .........

8 ..

3/4.4.6.1 LEAKAGE DETECTION SYSTEMS..*

3/4.4.6.2 OPERATIONAL LEAKAGE ................... 8*

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 ........ 14 314.4.11 STRUCTURAL INTEGRITY ....................... 15

St. Lucie Unit 2 L-2007-003 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) 314.4.4 PORV BLOCK VALVES The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs in conjunction with a reactor trip on a Pressurizer Pressure-High signal minimizes the undesirable opening of the spring-loaded pressurizer code safety valves. The opening of the PORVs fulfills no safety-related function and no credit is taken for their operation in the safety analysis for MODE 1,2, 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.

314.4.5 STEAM GENERATOR 6) TUBE INTEGRITY c urveillance Requirements for inspection of the steam genera.or tubes u~re that the structural integrity of this portion of the RCS will be Smaintaine fT.Te program for inservica inspection of steam generator t ubes is baseId 0 -amodification of Regulatory Guide 1.83, Revision 1.

Inservice Inspection o team generator tubing is essential in order Ito maintain surveillance of ceonditions of the tubes in the event that ithere is evidence of mechanical as.,ge or progressive degradation due Sto desigmnuatrng errors, or in *ce conditions that lead to corrosion.

I. tInsevc npcino ta eeao ubing also p' ides a means of

\ characterizing the nature and cause of any tube degradatl so that corrective measures can be taken.

St. Lucie Unit 2 L-2007-003 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)

BASES (continued) 1(, TUBE INTEGRITY I 3/4.4.5 STEAM GENERATO continued) e plant Is expected to be operated in a manner such that the secondary coo nt will be maintained within those chemistry limits found to result in neglig e corrosion of the steam generator tubes. Ifthe 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 o ration would be limited by the limitation of steam generator tube le age between the primary coolant system and the secondary coolant tem (primary-to-secondary leakage = 1.0 gpm from both steam generators. Cracks having a primary-to-secondary leakage less than this limit during eration will have an adequate margin of safety to withstand the loads impo d 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 loca d and 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 scheduled inservice steam gene tor tube examinations. Plugging will be required of all tubes wit imperfections exceeding the plugging limit of 40% of the tube nominal w thickness.

Steam generator tube inspections of operating plants have d onstrated the capabilitywall to reliably detect degradation that has penetrated % of the INER oiiatue thickness.

St. Lucie Unit 2 L-2007-003 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 3/4.4 REACTOR COOLANT SYSTEM (continued)

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

314.4.6.2 OPERATIONAL LEAKAGE cosidereds ae s shown that while a limited amount of leakage is

/exp, std from the RCS, the unidentified portion of this leakage can be Sreduced t*,hreshold value of less than 1 gpm. This threshold value is sufficiently lo esre early detection of additional leakage.

The 10 gpm IDENTIFI E"-AKAGE limitation provides allowances for ta s w ho se presence will no edoun oflea ag e ro' -,l ow n source li mi m interfere with the detection of UN IDE TIFIED LEAKAGE by the leakage Idetection systems.

T e S r el a c Re u r m n s f rR CS p re ssu re-" lation va lve s p ro vid e

\ added a ss r n e o a v n e rt h reby reduc~ing the babifity of or gross valve failure and consequent intersystem LOCA. Lea ekfrom the/

IRCS pressure isolation valves is IDENTIFIED LEAKAGE and wilf'e

,considered as a portion of the allowable limit.

St. Lucie Unit 2 L-2007-003 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) 3/4.4.6.2 OPERATIONAL LEAKAGE (continued

/the~oal steam generator tube leakage limit of 1 gpm for all steam

INSERT genera qensures that the dosage contribution from the tube leakage will 83/4.4.6.2 be limited to-mall fraction of Part 100 limits in the event of either a (follows steam generator rupture or steam line break. The 1 gpm limit is Insert for consistent with the ass tions used in the analysis of these accidents.

B3/4.4.5) The 0.5 gpm leakage limit pe eam generator ensures that steam Telitatirubonteacty is Coainta i the event of a main steam lineta rupture or under LOCA Ccnditions. i SPRESSURE BOUNDARY LEAKAGE of any mag Ji e is unacceptable potential fo Reatior oolan imleakag groes failure he pressure bounda Thein aining the chest of any hiePRESSURE BOLimRYi providesudes the unit to be promptly placed in COLD SHU WN.t 3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

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

St. Lucie Unit 2 L-2007-003 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 SO 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 A. 1.2, "Hot Standby," LCO 3A.41.3, "Hot Shutdown," LCO 3.4.1.4.1, "Cold Shutdown - Loops Filled," and LCO 3.4.1.4.2, "Cold Shutdown - Loops Not Filled."

SQ 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. SQ 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 SQ 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 ersure that SG tube integrity is maintained. Pursuant to Specification 6.8.4.1., tube integrity is maintained when the SQ performance criteria are met. There are three SQ performance criteria: structural integrity, accident induced leakage, and operational leakage.

The SG performance criteria are described in Specification 6.8.4.1. Meeting the SQ 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 SQ designs. Specification 6.8.4.1.1. applies to the replacement SQ 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 SQ replacement.

The processes used to meet the SQ 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 SQ 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-2007-003 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 ore SG or is assumed to increase to 0.3 gpm total through I 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 tibe integrity. Tube repair (i.e., sleeving) is applicable only to the original SGs.

In the context of this Specification, a SO tube for the replacement SGs is defined as the entire length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. For the original SGs, a 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 SG tube is defined as the length of the tube between the bottom of the sleeve to the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

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

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

St. Lucie Unit 2 L-2007-003 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 circumfeiential 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 II, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification.

This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures that the primary-to-secondary leakage caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The accident analysis assumes that accident induced leakage does not exceed 0.3 gpm total 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.

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.Applicabilit 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, prmarny-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 SQ tube. Complying With the'required ACTIONS may allow for continued operation, and subsequent affected SQ tubes are governed by subsequent Condition entry and application of associated required ACTIONS.

fLI and a.2 ACTIONS a9. and a2 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., sleev ) is applicable only tothe original SGs. An evaluation ofSG tubeintegrity of the affected tube(s) must beiiiade. SQ tube integrity is based on meeting the SQ performance criteria described in te '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 todetermine if a SQ 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 tlhe time the situation is discovered and the estimated growth of the degradation.

prior to the: next SQ tube inspection. If it is determined that tube integrity is not being I aintained, ACTION 6b ies.

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-2007-003 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-2007-003 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 perfbrmed 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 I. 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

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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 facility and the public.

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

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

Applicable Safety Analyses The safety analysis for an event resulting in steamn 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. 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.

St. Lucie Unit 2 L-2007-003 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. 4) 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 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 of the RCPB. Leakage past seals and gaskets is not PRESSURE BOUNDARY LEAKAGE.

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

c. Primary-to-Secondary Leakage Through Any One Steam Generator The limit of 150 gpd per steam generator is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Re f. 5).

The Steam Generator Program operational leakage performance criterion in NEI 97-06 states, "The RCS operational primary-to-secondary leakage through any one steam generator shall be limited to 150 gallons per day." The limit is based on operating experience with steam generator tube degradation mechanisms that result in tube

St. Lucie Unit 2 L-2007-003 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-2007-003 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 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 ofa Reactor Coolant System water inventory balance.

a andb.

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-2007-003 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 Surwvillance 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 6. 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-2007-003 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.6).

4.4.6.2.2

a. through d.

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

4.4.6.2.3

a. and b.

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

References

1. 10 CFR 50, Appendix A, GDC 30

2. Regulatory Guide 1.45

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

4. UFSAR, Section 15.6.3

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

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

St. Lucie Unit 2 L-2007-003 Docket No. 50-389 Attachment 4 Proposed License Amendment Page 19 of 19 Steam Generator Tube Integrity SECTIONNO.: TITLE: TECHNICAL SPECIFICATIONS PAGE:

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

BASES (continued) 0.3 gpm total primary-to-secondary leakage through 314.4.8 SPECIFIC ACTIVITY SGs and 216 gallons per day through any one SG The limitations on the specific activity of the prima coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will n t exceed an appropriately small fraction of Part 100 limits followin a steam generator tube rupture accident in conjunction with an assumed tead state primary-to-secondary steam generator leakage rate o 1.0 gpm an a concurrent loss of offsite electrical power. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the St. Lucie site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

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

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-2007-003, Reply to Request for Additional Information, Steam Generator Tube Integrity Amendment Request

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