Text

Proposed License Amendment Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections
	ML093580093
Person / Time
Site:	Saint Lucie
Issue date:	12/14/2009
From:	Richard Anderson; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2009-262
	Download: ML093580093 (54)
	v • d • e

Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 December 14, 2009 FPL L-2009-262 10 CFR 50.90 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Re:

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Proposed License Amendment Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections Pursuant to 10 CFR 50.90 and 10 CFR 50.91(a)(1), Florida Power and Light Company (FPL) requests approval of a change to St. Lucie Units 1 and 2 Facility Operating Licenses DPR-67 and NPF-1 6, respectively. Attached for Nuclear Regulatory Commission review and approval is a proposed Technical Specification (TS) change to remove the structural integrity requirements contained in TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) and their associated Bases from the St.

Lucie TSs. Removal of the structural integrity TS is consistent with NUREG-1432 in that it does not meet the criteria of 10 CFR 50.36 for inclusion in the TSs. The proposed amendment will also incorporate changes to accident monitoring instrumentation for consistency with NUREG-1432 actions and allowed outage times for conditions that drive a unit to HOT SHUTDOWN.

The proposed amendment also makes several administrative corrections based on obvious typos, previous amendments, or obsolete requirements.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. Attachment I provides an evaluation of the proposed change. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides the proposed TS changes in final typed format.

The proposed change is neither exigent nor emergency. Once approved, the amendment will be implemented within 60 days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated State Official. If you should have any questions regarding this submittal, please contact Ken Frehafer at (772) 467-7748.

an FPL Group company

L-2009-262 Page 2 I declare under penalty of perjury that the foregoing is true and correct.

Executed on the I '

" day of-beCe&*&Ur

, 2009 Very truly yours, Richard L. Anderso Site Vice President St. Lucie Plant Attachments cc:

Mr. William Passetti, Florida Department of Health

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 1 of 15 Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections Analysis of Proposed Technical Specification Change

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Attachment I Page 2 of 15 1.0 Description of Proposed Changes 1.1 Technical Specification (TS) 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2)

The proposed change removes the St. Lucie structural integrity requirements contained in TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) and the associated TS Bases from the TSs. The proposed change is consistent with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0 (Reference 1).

1.2 Reactor Coolant Pump Flywheel Inspection (administrative in nature)

The change also relocates the Unit 2 reactor coolant pump (RCP) flywheel inspection requirements in Surveillance Requirement (SR) 4.4.11 to a new administrative TS program consistent with NUREG-1432.

1.3 Administrative TS 6.4.1 (administrative in nature)

The proposed change replaces the obsolete references to ANSI / ANS-3.1 - 1978 and 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of Enclosure I of the March 28, 1980 NRC letter to all licensees with the current NUREG-1432 Standard Technical Specifications Combustion Engineering Plants, Revision 3.0 (Reference 1) wording for Unit Staff Qualifications.

1.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)

The proposed changes to Unit I TS 3.3.3.8, Table 3.3-11 ACTION 1, 2, 6, and 7 and Unit 2 TS 3.3.3.6 ACTION 'a' and 'b' will revise the required end states and completion times.

1.5 Minor changes (administrative in nature)

" Unit 2 TS 3.1.2.6 - change the action to be logically consistent.

" Unit 2 TS Surveillance Requirement (SR) 4.3.3.2 - correct typo.

" Unit I TS index page VI - correct section heading for CONTAINMENT SYSTEMS

" Unit 2 TS index page XXIV - Table 3.6-1.

Unit 2 TSs 6.8.4.1.2 and 6.9.1.13 pertain to inspection requirements for the St. Lucie Unit 2 original steam generators.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 3 of 15 2.0 Proposed Change 2.1 TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2)

TS Limiting Condition for Operation (LCO) 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2), Structural Integrity, including its associated actions and SR 4.4.10 (Unit 1) and 4.4.11 (Unit 2) would be removed from the St. Lucie TSs and TS Bases.

2.2 RCP Flywheel Inspection The RCP flywheel inspection requirements in Unit 2 SR 4.4.11 would be relocated to a new administrative TS program, 6.8.4.o. The prescribed inspection methods are unchanged and will be reworded to: "Reactor Coolant Pump Flywheel Inspection Program - This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975." This wording is identical to the RCP Flywheel inspection program requirements in NUREG-1432, TS 5.5.7.

Relocating the RCP flywheel inspection requirements to an administrative TS program will not revise any current requirements.

2.3 Administrative TS 6.4.1 Training Requirements The existing text for TS 6.4.1 will be replaced by "Each member of the unit staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 3."

2.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)

The proposed changes to Unit I TS 3.3.3.8, Table 3.3-11 ACTION 1, 2, 6, and 7 and Unit 2 TS 3.3.3.6 ACTION a and b will require the end states to be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

2.5 Minor changes (administrative in nature)

Unit 2 TS 3.1.2.6 - change the action to be logically consistent.

Change the ACTION to read "With no boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3.1.2.2 operable..."

0 Unit 2 TS SR 4.3.3.2 - correct typo.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 4 of 15 Change the SR to read "At least once per 18 months..."

" Unit 1 TS index page VI - correct section heading for CONTAINMENT SYSTEMS Change the section heading to read "3/4.6 CONTAINMENT SYSTEMS..."

" Unit 2 TS index page XXIV - Table 3.6-1.

Delete Table 3.6.1.

" Unit 2 TSs 6.8.4.1.2 and 6.9.1.13 Delete the TSs.

3.0

Background

3.1 TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2)

The purpose of TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2), Structural Integrity, is to specify the requirements for maintaining the structural integrity of ASME Code Class 1, 2 and 3 components. This specification was originally intended to support assurance that structural integrity and operational readiness of these components are maintained at an acceptable level throughout the life of the facility. The specification is applicable in all operational modes. However, the specification does not provide actions for plant shutdown if its LCO is not met.

This is because the specification addresses the passive pressure boundary function of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) Class 1, 2 and 3 components as established by compliance with the Inservice Inspection (ISI) program. The ISI program is required pursuant to 10 CFR 50.55a, Codes and Standards (Reference 2) and SR 4.0.5. This TS does not fulfill any of the criteria of 10 CFR 50.36(c)(2)(ii)

(Reference 3) for retention in the TSs.

Maintaining a program-type requirement within an LCO creates significant interpretation issues for Operations personnel. The structural integrity TS was part of the original TSs and, therefore, no basis history is available regarding its intent. However, TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) appear to have been included to help ensure that plant heatup and startup would not occur until all required portions of applicable systems were verified to meet ISI acceptance

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 5 of 15 criteria following inspections performed during a plant outage (normally performed during refueling outages). Meeting this acceptance criteria helps ensure the integrity of applicable systems during all modes of operation, including accident events. For instance, the RCS pressure boundary is purposely breached during Mode 5 and 6 operations to support plant outage activities and such openings are not historically considered a violation of TS 3/4.4.10 (Unit 1) or TS 3/4.4.11 (Unit 2). Furthermore, TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) contain no actions suggesting they were designed to accommodate integrity concerns once plant heatup has commenced. Structural integrity ISI activities are performed only during plant outages when conditions exist that permit access to the applicable systems and are not monitored or controlled through application of the ISI program during the operational cycle.

For example, other TSs are designed to monitor the structural integrity of the RCS during operation and provide actions to shutdown the unit if compliance is not maintained. RCS heatup and cooldown rates (TSs 3.4.9.1 and 3.4.9.2), and the overpressure mitigation system (TS 3.4.9.3) protect against applying undue stresses as a result of pressure/temperature transients on RCS components and piping. RCS leakage TSs (3.4.6.1 and 3.4.6.2) provide a means of protecting the RCS integrity by detecting and monitoring leakage. Therefore, it is not necessary to apply TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) when integrity issues become evident during plant operation above cold shutdown. Because TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) are redundant to other regulations, it is acceptable to remove TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2) requirements from the TSs.

Removing these specifications does not reduce the controls that are necessary to ensure compliance with the ASME Code. Structural integrity is maintained by compliance with 10 CFR 50.55a as implemented through the St. Lucie ISI program required by TS 4.0.5, as well as by compliance with TSs 3.4.6.1, 3.4.6.2, 3.4.9.1, 3.4.9.2 and 3.4.9.3 for the RCS.

3.2 RCP Flywheel Inspection The Unit 2 RCP flywheel inspection requirements in SR 4.4.11 would be relocated to a new administrative TS program, 6.8.4.o. The prescribed inspection methods are unchanged. With the addition of the existing RCP flywheel inspection requirements as an administrative TS program consistent with NUREG-1432, no surveillance requirements will be revised as a result of the relocation.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 6 of 15 3.3 Administrative TS 6.4.1 Training The St. Lucie Units 1 and 2 UFSARs, Sections 13.2, describe the training program as meeting or exceeding the requirements and recommendations of Section 5.5 of ANSI/ANS-3.1 1978 and 10 CFR Part 55 and the supplemental requirements specified in Sections A and C of enclosure 1 of the March 28, 1980 NRC letter to all licensees as outlined in Section 6.4, Training, of the plant TSs.

3.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)

The primary purpose of the post accident monitoring (PAM) instrumentation is to display plant variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions, for which no automatic control is provided, that are required for safety systems to accomplish their safety functions for design basis events.

3.5 Minor changes (administrative in nature) - the justification for the following administrative issues will be discussed in the next section.

Unit 2 TS 3.1.2.6

" Unit 2 TS SR 4.3.3.2 - correct typo.

Unit 1 TS index page VI - correct section heading for CONTAINMENT SYSTEMS

" Unit 2 TS index page XXIV-Table 3.6-1.

" Unit 2 TSs 6.8.4.1.2 and 6.9.1.13 4.0 Regulatory Analysis 4.1 TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2)

Section 182a of the Atomic Energy Act, as amended (the Act), requires applicants for nuclear power plant operating licenses to incorporate TSs as part of the license. The Commission's regulatory requirements related to the content of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) 50.36. That regulation requires that the TSs include items in five categories, including: (1) safety limits, limiting safety system settings and limiting control settings, (2) limiting conditions for operation, (3) surveillance requirements, (4) design features, and (5) administrative controls.

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 7 of 15 On July 22, 1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36. The Final Policy Statement gave guidance for evaluating the required scope of the TSs and defined the guidance criteria to be used in determining which of the LCOs and associated SRs should remain in the TSs. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TSs, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co.

(Trojan Nuclear Plant), ALAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:

[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicant's safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

By this approach, existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TSs; those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents.

The Commission codified the four criteria in 10 CFR 50.36 (60 FR 36953, July 19, 1995). The four criteria are stated as follows:

(1)

Installed instrumentation that is used to detect, and indicate in a control room, a significant abnormal degradation of the reactor coolant pressure boundary; (2)

A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier;

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 8 of 15 (3)

A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; and (4)

A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As a result, existing LCO requirements that fall within or satisfy any of the criteria in 10 CFR 50.36(c)(2)(ii) must be retained in the TSs while those LCO requirements that do not fall within or satisfy these criteria may be relocated to other licensee-controlled documents.

4.2 RCP Flywheel Inspection The Unit 2 RCP flywheel inspection requirements in SR 4.4.11 would be relocated to a new administrative TS program, 6.8.4.o. The prescribed inspection methods are unchanged. With the addition of the existing RCP flywheel inspection requirements as an administrative TS program consistent with NUREG-1432, no surveillance requirements will be revised as a result of the relocation. Therefore, this change is administrative in nature and will not be evaluated further in this amendment request.

4.3 Administrative TS 6.4.1 Training St. Lucie plant training programs are accredited through the National Nuclear Accrediting Board (NNAB) and have been for over 20 years. The NRC has agreed that this an acceptable alternative to some requirements outlined in 10 CFR 50.120 (regarding training programs) and 10 CFR 55 (regarding operator licensing) where a "Systems Approach to Training (SAT)" is used in lieu of the stated CFR requirements.

On March 19, 1987, Generic Letter (GL) 87-07, "Information Transmittal of Final Rulemaking for Revisions to Operator Licensing - 10 CFR Part 55 and Conforming Amendments," informed facility licensees that they had the option of substituting an accredited, SAT-based program for their operator training program previously approved by the NRC. The GL indicated that this option may be implemented upon written notification to the NRC and that it did not require any staff review. The GL also noted the NRC's expectation that facility licensees would update their licensing basis documents (e.g., their final safety

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Attachment I Page 9 of 15 analysis reports (FSARs) and technical specifications (TSs)), as necessary, to conform with their accredited program status.

As stated in RIS 2001-001, the NRC has not changed its requirements or position with regard to license eligibility for senior reactor operators and reactor operators since 1987. Regulatory Guide (RG) 1.8 (Revision 2 or 3) and the NANT's guidelines for education and experience (those that were in effect in 1987 or those that were issued in January 2000) outline acceptable methods for implementing the Commission's regulations in this area. As stated in the RIS, any required TS changes would be considered administrative in nature.

FPL chose to replace the existing administrative TS training wording with the NUREG-1432 wording for training.

Therefore, this change is administrative in nature and will not be evaluated further in this amendment request.

4.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)

St. Lucie has custom TS, and when an LCO is not met and the required action allowed outage time expires the action statements normally step through all intervening modes to get into a condition where the LCO is not applicable.

However, the St. Lucie accident monitoring instrumentation TSs do not follow this standard. The St. Lucie accident monitoring TS LCOs are applicable in Modes 1, 2, and 3. However, the action statements either drive the end state to Mode 3, HOT STANDBY, or they drive the end state to Mode 4, HOT SHUTDOWN, with no intervening step through Mode 3, HOT STANDBY.

The NUREG 1432, Rev. 3, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0, completion times for post accident monitoring instrumentation (analog) contain a structured way to transition to Mode 4, HOT SHUTDOWN, when the LCO actions and completion times are not satisfied for instrumentation. This TS change will follow the conventions of NUREG-1432 for St. Lucie accident monitoring instrumentation whose actions should drive the unit to Mode 4, HOT SHUTDOWN conditions.

4.5 Minor corrections

" Unit 2 TS 3.1.2.6 - This change corrects an obvious error in the logic of the action statement.

" Unit 2 TS SR 4.3.3.2 - This change corrects an obvious typo in the action statement.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Attachment I Page 10 of 15

" Unit 1 TS index page VI - This change corrects an obvious typo in the TS index section heading for CONTAINMENT SYSTEMS.

" Unit 2 TS index page XXIV - This change deletes Table 3.6-1 from the TS index. This table was removed from the TS in Amendment 88 that implemented 10 CFR 50 Appendix J, Option B (TAC Nos.

M97156/M97157).

" Unit 2 TSs 6.8.4.1.2 and 6.9.1.13 - These TSs pertain to the steam generator integrity program and reporting requirements for the St. Lucie Unit 2 original steam generators and are no longer applicable to the replacement steam generators that were installed in the SL2-17 refueling outage. The replacement steam generator inspection/reporting requirements are unchanged and contained in TSs 6.8.4.1.1 and 6.9.1.12.

Because these changes are administrative in nature, they will not be evaluated further in this amendment request.

5.0 Technical Analysis 5.1 TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2)

The purpose of TS 3/4.4.10 (Unit 1) and TS 3/4.4.11 (Unit 2), Structural Integrity, is to specify the requirements of maintaining the structural integrity of ASME Code Class 1, 2 and 3 components. However, this is redundant to and does not contain the detail of the requirements contained within 10 CFR 50.55a.

10 CFR 50.36(c)(2)(ii) states that a TS LCO of a nuclear reactor must be established for each item meeting one or more of the following criteria:

Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) is not applicable to

installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the RCS.

Structural Integrity TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) does not meet Criterion 1.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 1 Iof 15 that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) is not applicable to a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Although the specification is related to the integrity of applicable systems, compliance is maintained by meeting the requirements of 10 CFR 50.55a through implementation of the St. Lucie ISI program required by TS 4.0.5 and is not specifically monitored or controlled during plant operation. Structural Integrity TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) does not meet Criterion 2.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

No specific TS-related structure, system, or component (SSC) is being revised or removed from the TSs. Each TS SSC must continue to meet the requirements of 10 CFR 50.55a as implemented through the St. Lucie ISI program required by TS 4.0.5. Structural Integrity TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit

2) does not meet Criterion 3.

Criterion 4 A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

As stated above, no specific TS-related structure, system, or component (SSC) is being revised or removed from the TSs. Each TS SSC must continue to meet the requirements of 10 CFR 50.55a as implemented through the St. Lucie ISI program required by TS 4.0.5. Structural Integrity TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit

2) does not meet Criterion 4.

The scope of this specification has been evaluated against the criteria of 10 CFR 50.36(c)(2)(ii) and none of these criteria require that the structural integrity controls specified in TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) are appropriate for retention in the St. Lucie TSs. This conclusion is consistent with

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 12 of 15 NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0.

Based on the above discussion, removal of structural integrity requirements contained in TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the TSs is acceptable.

5.2 Unit 2 RCP Flywheel Inspection - This is an administrative change.

5.3 Administrative TS 6.4.1 -This is an administrative change.

5.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)

The St. Lucie Unit I TS 3.3.3.8 Table 3.3-11 Actions 1 and 2 have a specified end state of HOT STANDBY. The required end state where the LCO is no longer applicable needs to be HOT SHUTDOWN. Consistent with NUREG-1432 TSs for post accident monitoring instrumentation, the proposed change drives the action end state to HOT SHUTDOWN (with a completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) through the intervening state of HOT STANDBY (with a completion time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ). The proposed HOT STANDBY completion times are equal to or more conservative than the existing MODE 3 completion times and are acceptable. The HOT SHUTDOWN end state and completion time are acceptable as they are consistent with NUREG-1432 requirements.

St. Lucie Unit I TS 3.3.3.8, Table 3.3-11 Actions 6 and 7, and St. Lucie TS 3.3.3.6 actions 'a' and 'b' have a specified end state of HOT SHUTDOWN with a completion time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed outage time, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , is consistent with the NUREG-1432 allowed outage time for entry into Mode 4 and is acceptable. FPL proposes to include the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowed outage time to enter the intervening state of HOT STANDBY. This change is consistent with the NUREG-1432 completion time for entry into Mode 3 conditions.

Based on the above, there is no increase with any accident mitigation risk associated with the change. The proposed allowed outage times and the intervening step through HOT STANDBY are consistent with the equivalent NUREG-1432 completion times and actions for post accident instrumentation and are equal to or more conservative than the current TS requirements.

5.5 Minor changes - These changes are all administrative.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Attachment I Page 13 of 15 6.0 Determination of No Significant Hazards Consideration FPL is proposing that the St. Lucie Operating Licenses be amended to revise the TS requirements for structural integrity, accident monitoring instrumentation, and make several administrative corrections based on obvious typos, previous amendments, or obsolete requirements. The proposed changes will remove Structural Integrity TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the TSs. This specification is redundant to ASME Code compliance as required by 10 CFR 50.55a and specified in TS 4.0.5.

Additionally, these proposed changes provide consistency for accident monitoring instrumentation actions and allowed outage times for conditions that drive the unit to HOT SHUTDOWN conditions. The proposed changes are consistent with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0.

FPL has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

No. The proposed change to remove structural integrity controls from the TSs does not impact any mitigation equipment or the ability of the RCS pressure boundary to fulfill any required safety function. The proposed change will continue to ensure the requirements of 10 CFR 50.55a are maintained as specified in TS 4.0.5 and the new administrative TS program for RCP flywheel inspections. The changes to the accident instrumentation actions and allowed outage time have no appreciable effect on accident initiation or mitigation. Since no other accident mitigation or initiators are impacted by this change, no design basis accidents are affected.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of any accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any previously evaluated?

The proposed change will not alter the plant configuration or change the manner in which the plant is operated. Structural integrity will continue to be maintained as required by 10 CFR 50.55a and specified in TS 4.0.5 and the new administrative TS program for RCP flywheel inspections. Accident monitoring instrumentation does not contribute to failure modes. No new failure modes are being introduced by the proposed change.

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 P.age 14 of 15 Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

Does the proposed change involve a significant reduction in the margin of safety?

Removing TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the TSs does not reduce the controls that are required to maintain the structural integrity of ASME Code Class 1, 2, or 3 components. There is no increase with any accident mitigation risk associated with the accident monitoring instrumentation TS changes as the proposed allowed outage times and the intervening step through HOT STANDBY are consistent with the equivalent NUREG-1432 completion times and actions for post accident instrumentation and are equal to or more conservative than the current TS requirements. No other safety margins are impacted due to the proposed change.

Therefore, the proposed change does not involve a significant reduction in the margin of safety.

Based on the above, FPL concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

7.0 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative

.occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Attachment I Page 15 of 15 8.0 Precedence The proposed change to remove TS 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) from the TSs is consistent with NUREG-1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0 and is similar to the amendment issued for Arkansas Nuclear One, Unit No. 2 in Amendment 270 dated March 1, 2007 (ML070570506)and the amendment request currently under review for Turkey Point Units 3 and 4 dated February 16, 2009 (ML090630238).

9.0 References

1.

NUREG 1432, Standard Technical Specifications Combustion Engineering Plants, Revision 3.0

2.

10 CFR 50.55a, Codes and Standards

3.

10 CFR 50.36, Technical Specifications

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 1 of 19 Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections Proposed Technical Specification Changes (mark-up)

Unit 1 Page V Page VI Page 3/4 3-43 Page 3/4 4-26 Page 6-6 Unit 2 Page VI Page XXIV Page 3/4 1-12 Page 3/4 3-24 Page 3/4 3-41 Page 3/4 4-39 Page 6-7 Page 6-15f Page 6-15g Page 6-15h Page 6-15i Page 6-15j Page 6-20f

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 2 of 19 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.4 P R ES S U R IZ E R..................................................................................................

3/4 4-4 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY................................................

3/4 4-5 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE....................................................

3/4 4-12 Leakage Detection System s.............................................................................

3/4 4-12 Reactor Coolant System Leakage....................................................................

3/4 4-14 3/4.4.7 C H E M IS T R Y.....................................................................

............................... 3/4 4-15 3/4.4.8 S PEC IFIC AC TIV ITY........................................................................................

3/4 4-17 3/4.4.9 PRESSURE/TEMPERATURE LIMITS.............................................................

3/4 4-21 Reactor Coolant System.............

........................... 3/4 4-21 Pressurizer.............................

3/4 4-25 3/4.4.10 E:TRUCTURAL ITECRIL........

3/4 4-26 A,M E,od, e e

1. 2. an 3 eu...

voII eI II.................................................... SI 4 2&

3/4.4.11 3/4.4.12 3/4.4.13 3/4.4.14 3/4.4.15 DELETED.........................................................................................................

3/44-56 PORV BLOCK VALVES....................................................................................

3/4 4-58 POW ER O PERATED RELIEF VALVES..................................................

........ 3/4 4-59 REACTO R COO LANT PUM P - STARTING....................................................

3/4 4-60 REACTO R COO LANT SYSTEM VENTS.........................................................

3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SA FETY INJECTIO N TA NKS.............................................................................

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 W ATER TA NK...............................................................................

3/4 5-8 ST. LUCIE - UNIT 1 V

Amendment No. 28, 60, 68, 80, 434, 200

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 3 of 19 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS S C N

PAGE 314,46 CONTAINMENT SYSTEMS CO NTAINM ENT VESSEL...................................................................................

3/4 6-1 Containment Vessel Integrity..............................................................................

3/4 6-1 Containment Leakage.........................................................................................

3/4 6-2 Containment Air Locks......................................................................................

3/4 6-10 Internal Pressure...............................................................................................

3/4 6-12 Air Temperature................................................................................................

3/4 6-13 Containment Vessel Structural Integrity..................................

3/4 6-14 3/4.6.2 DEPRESSURIZATIO N AND COO LING SYSTEMS.........................................

3/4 6-15 Containment Spray and Cooling Systems........................................................

3/4 6-15 Spray Additive System....................................................................................

3/4 6-16a 3/4.6.3 CO NTAINM ENT ISO LATIO N VALVES............................................................

3/46-18 3/4.6.4 DELETED.........................................................................................................

3/4 6-23 DELETED.........................................................................................................

3/4 6-24 DELETED.........................................................................................................

3/4 6-25 3/4.6.5 VACUUM RELIEF VALVES..............................................................................

3/4 6-26 3/4.6.6 SECO NDARY CO NTAINM ENT........................................................................

3/4 6-27 Shield Building Ventilation System...................................................................

3/4 6-27 Shield Building Integrity....................................................................................

3/4 6-30 Shield Building Structural Integrity....................................................................

3/4 6-31 3/4.7 PLANT SYSTEMS 3/.4.7.1 TURBINE CYCLE...............................................................................................

3/47-1 Safety Valves......................................................................................................

3/4 7-1 Auxiliary Feedwater System...............................................................................

3/4 7-4 Condensate Storage Tank..................................................................................

3/4 7-6 Activity.................................................................................................................

3/4 7-7 Main Steam Line Isolation Valves.......................................................................

3/4 7-9 ST. LUCIE - UNIT 1 VI Amendment No. 27, 434-, 4-34, 204

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 4 of 19 TABLE 3.3-11 (continued)

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels less than the Total No. of Channels shown in Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days BF be in HOT STANDBY ";thin the n¢xt 12 hcur*.

ACTION 2 -

With position indication inop/era restore the inoperable indicator to OPERABLE stat or close the associated PORV block valve and remove p r from its operator within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ACTION 3 -

With ny individu valve position indicator inoperable, obtain que ch tank te perature, level and pressure information once p

shift to ermine valve position.

ACTION 4 -

h t number of OPERABLE Channels one less than the Total Nu er of Channels shown in Table 3.3-11, either restore the I perable channel to OPERABLE status within 7 days if repairs or be in HOT are feasible without shutting down or prepare and submit a STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Special Report to the Commission pursuant to the specification and HOT SHUTDOWNAA' 6.9.2 within 30 days following the event outlining the action in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION h the number of OPERABLE Channels less than the Minimum Ch nels OPERABLE requirements of Table 3.3-11, either restore the in erable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repai are feasible without shutting down or:

Initiate n alternate method of monitoring the reactor vessel in ntory; and

2.

repare and bmit a Special Report to the Commission rsuant to Spe ification 6.9.2 within 30 days following th event outlinin he action taken, the cause of the ino rability and the lans and schedule for restoring the s stem to OPE RAE status; and

3.

Restor the Channel to 0 RABLE status at the next schedule refueling.

ACTION 6 Wth the numbe of OPERABLE acci nt monitoring channels less thantheMinimu Channels OPERABL requirements of Table 3.3-11, either restore the i perable channel(s) to PERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months e" --b

.- t a

WO

.~

OW -VehiR the.Re~t !2 he, -F-.

ACTION 7 -

With the number of 0 RABLE accident monitoring channels less than the Minimum Channels PERABLE requirements of Table 3.3-11, either restore the inoperable c nel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of be at least IHOI T HU TDrOlirFWN within the next 12 h.U...

ST. LUCIE - UNIT 1 3/4 3-43 Amendment No.

7, 7, 165

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 5 of 19 340 STRUCTURaL INTEGRITM UMM MMIPY142 WPM H -p lp, H pK q 40 The-str-

(oxoop! steom go.ort Spoifloation 1.1.10.1.

ral intewrit of ASME Code Class 1. 2 and 3 Geompononb ortubes) hchll bs mointainod in aco.rdano.. oth AGT4GN --

a, W.th the truotur.al intogrity of any ASME Code Claco. 1 com.ponent(s) not conforming to the abc..

requirmonts, rootr.*

the.truotur.l intoegity of the affo.ted

... mp.ncnt(.) to Wiin is lim Or

.oelat.

the aff.tod Ssetempneont(r) pr to inoroa°ing tho ReaOtOr Cont'-M System temporalturo moroe than S00Fcabovo tho minimum tomnperaturoF ruirIoI by '19T moncidrtioro.

."t tho tr'-.oturia' :.tcgrfty of cny ASM Codo lccss 2 oemponcnt~)~ not cnr.IormIn. to tno aeovo rq.uiromntes, ro.tor t Etrudural intogry of the affo.t. d oomponont(o) to Yhin ito lI'Mi or iolato the affootodd Gempenont(o) pricF to iner~easing tho ReaotOr Coolant System tomnporaturo above 2009F.

I.t.r.h the otru--tural intogrity of any ASME Codo Cl:sa 3

..o.p.ncnt(.)

not o..nfming to the bo.. roguiromontO, restore the

,truotural integrity of the affeoted oompnonet(s) to Shin RIt limit Or icelate tho affooted.

oomponont(o) fromA orwioo.

Tho provwicino of Spoolfication 3.0.4 crc not applieablt.

ci, SURVEILLANCE REQUIREMENTS 4A41041 No additional Sur.'cllanso Roguiremonts other than these roguirod by Specification 1.0.5.

ST. LUCIE - UNIT 1 3/4 4-26 Amendment No. 6W, 90

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 6 of 19 6.0 ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI / ANS-3.1-1978 for comparable positions, except for:

(1) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have specific training in plant design and plant operating characteristics, including transients and accidents, and any of the following educational requirements:

Bachelor's degree in engineering from an accredited institution; or Professional Engineer's (PE) license obtained by successful completion of the PE examination; or Bachelor's degree in engineering technology from an accredited institution, including course work in the physical, mathematical, or engineering sciences, or Bachelor's degree in physical science from an accredited institution, including course work in the physical, mathematical, or engineering sciences.

(3) the Multi-Discipline Supervisors who shall meet or exceed the following requirements:

a.

Education: Minimum of a high school diploma or equivalent.

b.

Experience: Minimum of four years of related technical experience, which shall include three years power plant experience of which one year is at a nuclear power plant.

c.

Training: Complete the Multi-Discipline Supervisor training program.

For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of 6.3.1, perform the functions described in 10 CFR 50.54(m).

6.4 TRAINING 6.4.1 A rztr4aiRnn and roplc9m..nt tr=anORO pregr.a'.

f9r thz1 unit Skiff hRAl, :R rnaintaiR.* d

-inder the dire*iRn cfthe training rnanagzr and shall meet zr AFz eWed thz rtquirzeFRnt6 an.d-rzccFMmcdatieRsz ef ScctieR the c'upplmneRtal rzqUi-rmentc sp M~aF~h 28, 1980 NSRC lz#AFrtz al" I; 6 ndELEtED aporational xprinec.

6.5 DELETED rz AMCOI ! AkQ q.

4 Q7 Q A in n

O..,,04 r er.0iod in Sestioncs A. And Q Of ER.Gk)curz 1 ef the ST. LUCIE - UNIT 1

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 7 of 19 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES D E L E T E D...................................................................................................

3/4 4-7 O P E R A T IN G..............................................................................................

3/4 4-8 3/4.4.3 P R ES S U R IZ E R..................................................................................................

3/4 4-9 3/4.4.4 PO RV BLO C K VALVES...................................................................................

3/4 4-10 3/4.4.5 STEAM GENERATOR (SG) TUBE INTEGRITY..............................................

3/4 4-11 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS........................................................

3/4 4-18 O PERATIO NAL LEAKAGE......................................................................

3/4 4-19 3/4.4.7 C H E M IS T RY.....................................................................................................

3/4 4-22 3/4.4.8 S PEC IFIC AC TIV ITY........................................................................................

3/4 4-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTO R COO LANT SYSTEM......................................................................

3/4 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 COOL7ING 7SYSTEMS:(E "-2)"ýj'*

3/4.5.1 SAFETY INJECTIO N TANKS.............................................................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 3251F................................................................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 3250F................................................................

3/4 5-7 3/4.5.4 REFUELING W ATER TANK...............................................................................

3/4 5-8 ST. LUCIE - UNIT 2 VI Amendment No. 46, 440, 147

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 8 of 19 INDEX LIST OF TABLES (Continued)

TABLE PAGE 3.3-9 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION................................

3/4 3-39 4.3-6 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE R E Q U IR E M E N T S...........................................................................................

3/4 3-40 3.3-10 ACCIDENT MONITORING INSTRUMENTATION..........................................

3/4 3-42 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE R E Q U IR E M E N T S...........................................................................................

3/4 3-43 3.3-11 DELETED 3.3-12 DELETED 4.3-8 DELETED 3.3-13 DELETED 4.3-9 DELETED 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTIO N...............................................................

3/4 4-16 4.4-2 STEAM GENERATOR TUBE INSPECTION...................................................

3/4 4-17 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES.............. 3/4 4-21 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY...............................................

3/4 4-23 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE R EQ U IR E M E N TS...........................................................................................

3/4 4-24 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS P R O G R A M....................................................................................................

3/4 4-27 4.4-5 DELETED 3.4-3 LOW TEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..... 3/4 4-37a 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP......... 3/4 4-37a 3.6-1 CONTAINMENT LEAKAGE PATI 34 65 3.6-2 CONTAINMENT IS TION VALVES....

...................... 3/4 6-21 ST. LUCIE - UNIT 2 XXIV Amendment No. 8, 53, 54, 73, 86

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 9 of 19 REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s) in Specification 3.1.2.2 is OPERABLE.

Arrlp,-1-'1 ACTION:

mLlI I T:; IV!u'r..LJ 1 4, 3 ad 4-.

operable With no boric acid makeup pump requi d for the boron injection flow path(s) 414 restore e

^

r^ c a, pursuant to Specification 3.1.2.2

, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200°F; restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.6 The above required boric acid makeup pump(s) shall be demonstrated OPERABLE by verifying that the pump(s) develop the specified discharge pressure when tested pursuant to the Inservice Testing Program.

ST. LUCIE - UNIT 2 314 1-12 Amendment No. 8, 25, 40, 94, 405, 136

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 10 of 19 INSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a.

With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-3.

4.3.3.2 At lease once per 18 months, each Control Room Isolation radiation monitoring instrum ntation channel shall be demonstrated OPERABLE by verifying that the respon I

o annel is within limits.

ST. LUCIE - UNIT 2 3/4 3-24 Amendment No. 152

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 11 of 19 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

or be in HOT STANDBY in 6 APPLICABILITY: MODES 1, 2 and 3.

hours and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months a.*

With the number of OPERABLE acciden m itoring channels less than the Required Number of Channel shq in Table 3.3-10, either restore the inoperable channe 0 ERABLE status within b.*

With the number of OPERABLE accid t monitoring channels less than the Minimum Channels OPERABLE jrquirements of Table 3.3-10, either restore the inoperable channels to EFRABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

eFbe in at 'east HOT SHUT-DON wihnthe next 12 hOLurS.

c.*

With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

d.** With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; and

2.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and

3.

Restore the Channel to OPERABLE status at the next scheduled refueling.

e.

The provisions of Specification 3.0.4 are not applicable.

Action statements do not apply to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

- Action statements apply only to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

ST. LUCIE - UNIT 2 3/4 3-41 Amendment No. 49. 45

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 12 of 19 REACTOR COOLANT SYSTEM WA/44.44 STUTRLITEGRITYV LIMnI:NG CONDITION FOR GPERAllON 3.4.11 I no orruowrai ntcgr;ty or ACM11L t..oao Ulasa I, ar-a ~ ocmponor.t3 shall bo mointfflinold in. oordonooMIRMF-wit.h SDR~offioaetion 4.4.11.

APrLICADILITY: ALL MODESE AGT4GW a-:

h t......... intogrity of any ACME Cod. dlass 1 om.ponont(o.

not..~,

eefof,,n to the above reurmns rastora the struetural intgrity of th aff*cted o.n.pon.nt(.)

to within its mit-;s or

.l.o.t.

tho affootod m.pn*..nt(o) prieo to.inor..oing tho R....o Coolant Cystem tomperaturo meoe than 500 F above....

rnif mur tomporoturoe roquirod by NOT oonsidorotiono.

^r-h tho s*Utru.ural intogrity of any ASME Codo Glass 2.. mpon.nt(s) not oor~fOrmnirO to tho obovo roqukoemonts, roStoro toe Struoturol integrit; of the offo.t.d oompon.nt(s) to w..hin its

lm itao CGoolnt Systogm tompor.turo abov' 200F.

... i. the stru.ctural Weg

,37,

of a A

E

,,;e-Gls C

Ch mo;,

3 ^

not confo~rming to the abovo roguiromonts, rostoro tho structuraFl intogrit; of the affootod oomponont to within Ro limit or isolato tho aifootodeoonponoent~fre 07 wne rvso n ot bp.A mysat on 3j.u.A aro not app;oi oe~.

Su1RVEILLArJCF REQUIRE rNT 4.4.-1 In add""ion to tho rog Coolant Pumnp flywheel shall be u

JFrornts of Spoiflt in 4.0.6,

.aoh

Roaeto, dns~eoted pe the, reenmndationa of Rogulatory ooiB;oon I

t.4.

IoguTatoryuGumo

.14,, ev;3;,,n 1, ;xuguot ii;'o.

ST. LUCIE - UNIT 2 3/4 4-39

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 r A AfMItIg.TAT1XIF ANITRflI -Z L-2009-262 Page 13 of 19 6.4 TRAINING 6.4.1 A rotrainR*Ag and rOPlaGOmont training program for tho Unit ctaff thoul be M-aintaind undo the dircation of the training m~arago an tho rn tr ad

!he rqUiro ments and roaommtndotiono of Statin 6. o;f ANsI! ANS; 3. 10475 and 10 CFR Pa~ 65 and the euipplomontal roGu6iFr:monts Sepaifiod in Section A and C Of EnclocuroF 1 of the March; 25, oporational o)xprito.

6.5 DELETED st.",

ST. LUCIE - UNIT 2 6-7 Amendment No. 13, *5*4-2, 433,146

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 14 of 19 ADMINISTRATIVE CONTROLS (continued)

I.

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 dl, 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 A SG Pr tgrom i Shall be established.Rdl implmF.rt"d f. r the

.*i.

S..

t SFur. that SG tube intcgrity is maintained. In addition, the SG PrOgromR shall inoludo the fellowing a-Rfe -iCosfFeiam~i iE~9tFigaesrek.Gndit en mHOn4 tRg sese, Rger~s an eovoluatien BF the "as feunof' eonditionz of the tubing With roopeot to !he perfeFrmonoc oritoria fOr Structural intogrity and aseident induoodl leakage. The asfound" eeneitien FefeFS to the eefnditien of the tubng dudig a SG no~peaobon outage, as deteormined from" the nspe npeet onrFesultsr eF byther moo no, pri1or to the pluffging Or ropair of tubes. ConditionR monitoring assessments shall be eenduoted dur~ing eaeh outoge duFr@n which t19e SG tubes We inopeeted, plugged er roerd to eanfirmA that the neorefor.nae..8riLr.F.Ro boino hP Rt.P ST. LUCIE - UNIT 2 6-15f Amendment No, 147

St. Lucie Units 1 and 2 Docket Nos. 50-33 5 and 50-389 ADMINISTRATIVE CONTROLS (continuedi I-Steam Ccncrter (SG). roqra L-2009-262 Page 15 of 19 Fmteont',nuc) 2-seenlinued)

Porfoa for SC tubo intogFrit.

SC tube intogrity shall be moninltainod by mooeting the parForm~e n~te air.

f8rtube struEtural intogrity, cocident induzod loakago, and opratea.....

l*k...ge.

4 tructural intagrity pctffirmcnc cFritorin: All in seR'iO SG tubas Shall rotafin structurlal integrity aver the full range of normal operatig conditions (including sta.up,

oprati8n in the pawaFr rang:, ht ota ody, ond o.oldE...

n.d oi anticipated trano into inoludod in tho dosign,psoifloation) aRd dooign basis acoidento. Thio inoludoo rotaining a safety factor of 3.0 againet bu'-t undor normal stoady stato full power operation* piFmar to secondary pr..our.

difforontial and a safety faotcr Of 1.4 againat burot applied to tho doosgn basel sooldcA tpFima' to sooon.do prsu ifforontilo. Apo"^

from tho abovo r

.,ui.romonto, additional loading oonditions aoceiatod with tho design ba,-o aedets br eiqaatinR f aeides ti aeOF re..witht4ho dosian and e-lioonsing baice, ohall also be avaluatad t-odctcrmnincif the as,.sLstd led.

ontribote ognifioantly to burst or ocllapso.

In the assessment of tuba intogrity, those loads thatdg..

e s......tl.

af burJ Or eFellapsU shall bo dotormlned anc.

assessed in

..bin.atiOn with tHa lo.ads dua ta praSSura with a suc' ftsr of 1.2 on tho cmb in-d prim.ary loads andI 1.0 eR aa rd...'

loads.

2-Accident Rinduced leakage perfoFrmanecG te;ritrin Tha-primary tal Geaandary asoident induced leakags rats for angy dsaign basis acciden~t, sthsr than a SC tuba ruptura, shall not eaceed the leakag Fat mosýd Rnthe accident anal'ysis in termsG of total leakage rate for all S~s and leakaga rota for an inividAdual 8G. Leakage is not to oxoood 0.3 gallonse par mninute total throughl all GC and 216 gallons par day through any ona SC.

The operational leakage patrf, crm r, eFR i e,, is dee

^`4 in L

'W 3.4.6.2.e-,

"Racator Coolant System Operational Leakage."

Pro-isions for SC tuba repair critoria Tubas fsund by n nest'nts1-ntain* a flawn*a n**

c nslavadr*ag*,en with a depth equaltoO e~eeding 10 poec n~t of the nominal tuba wall thicknRess shall be plugged o, Fepaqrd exept if permiftad t) remain in senyea throuigh application of the alternate tuba repair criteria discuacad in:Fchnia Specification 6.8.4.1.2.e.4.

Tubas foundE by ina.anpatio ta8contai a flaw in (a) acloevearB (b) the pressure boundary pdeticn of the sriginael tubs wall in the slee've to tuba joint shallI be plugged.

.All tubes w.th sleev.es that have a nikel band shall ba plugged aftar one cycle in aporation.

4-The

,C methodology, as daseribed below, may be appliod to the expanded po,"i9n of the tuba in the hot log tubeaheat raglan as an alternativo to the r4 4u peFrc, nt c bpan scscd uraar:e or iacnn:ca W6.:t:ati 4..2.UE.iQ ST. LUCIE - UNIT 2 6-15g Amendment No. 147

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 AnfMINI.TRATI\\II fll'NJTPfl)I A Irnntinull L-2009-262 Page 16 of 19 I,

2-e-

4-(eentqnued}

T~boo with no port eio ofta toworsloov jo int in tho hot log tuhoohoot rogion ohall ho roFpirod Or pluggod upon dotootion of any flaw identified wihn 10.3inhoo bo lewtho bottom of the hot log oxponsein taronition or top of the tuboo*hs t, whiahovor l..ati n it lowor. Flaws l.. atod below thio elovation moy ram.

s i, rgdlesa ef ae.

Tubts whioh haa'o aty portion of a sleeve joint in the hot log tub..hot*

thc hot hog be plugged uponeten dotiao of thy flow thot io lootod bherw theoE)e loweyr T"to tube joint and within 10.3 inohoo bolow the boftom-et tho hot log oxpanolon tranosition or top of the tuiboohoot, wvhiohoevor elovotfienistlowo.

9, Provioiono for SC tube inospoationo. Pariodia SC tuba inopeations ohatl ho par5formad.

The number and poitono of the tubes inspootod and methods of inospotlon sholl b performed with the abjttive of dotetting flawa of ty typo (e.g., valumotria flaws, axial ondI oiroumfo~roniol orookos) that Fnoy ho prooont along tho Ion§4h of tho tuho, fromf tho tube to tuboohoot wo'd ot tho tube Filet to tho tubab to tuhoohoot wo'd ot the tube ot~lot, and that may satisfy the applioablo tube ropar Foritoria. Tho tube to) tuehohot weld it n.t pa eof tho tuba. For tubes with no poion of o Iewr slee; oint in tho hot log tuboohoot rogion, tho portion of theotube below 10.3 inehoc fromn the top at the hot log tuhoshoot or oxpontion taronition,. whiohovor isolowo, OStnoludod whon tho olto rnoto rOp F irta in T6 Station.S... 4.12.e.14 ao oppli d. In tubes ropoirod byolooving, tho portion of theo o*r;ial tuba wall o'*. '-on the sleeve's jointS iS not an aroa roquiring inospeOtti.

ln addfion to metotng theoeq rtguroonttef d.1, d.2, d.3 and d.4 boleaw,,the ioation ooopo, inopootion mothodo, and~ inopootion intonwalo shall ho ouoh ao to ewo hoSSU th ntgrit; s n itaminbind until tho nont SG inopootion.

An a iant dgrdton tholl ho pe~Odrmod to datoFrmina tho typo andI locationof flawo to whioh the tubeo may ho outooptiblo ond, batod ong this as.es.nft, U dotormnoRe whi~h nopostein methods need to be employed and at what eato*ns.

-1 Inopoat 100% of the tubes in oath SC during the firot rofuoling outoga following SC replaoement.

2-Inspoot 100%4; f tho tubat at to.unt ial pariodo of 60 offootivo full powFr months.

Tho trot o.quontiol po"iod ehlbe

..id*^

te begin afcr the frost inoawio inopootion of the sat. No S oshall o*pratr for i.ro than 2.4 offo-tivo full p..W.r a-4.

Ffontnt or onoe Foruohng ourogo twnionovor a te)ovt; u Betno i tog ;nopooroa.

Inopoot 100 peaeet o al nser.eslee anqd sleve u~bo jointo ovor, 21 tifetotivo full powor roantho or ant rofuoling outage (whiehru.Fisless),

If orFackI indicationo art found in any SC tuba, than tha noxi inopoation for eaa SC for the dogradatin moohanim that.au.d the or*o. indl ationg ahall not oxacoed 21 offootivo full pewor montpho OrFn r9ofuoling outaga (whiahovorF a loot). if d.finitivo info.rmation, ouch at from.. ox.amino.tion of a pulled tube, diagnoto non d.... tru...ti.o tottinig, or onginooring ovaluation indlatts that a croak! iko indiottion it not acoociatod with a Oraok(o), thtng tho windiotion oo not ho trootod at a trothl.

ST. LUCIE - UNIT 2 6-15h Amendment No. 147

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 17 of 19 nnfin!!ýril 13&1WM0jQ-LEU3.U-V-F. 1dbd0J-[3W6Q &

6.

Stoom Conorotor (SO) Proarom toontinuod) 2-(eentinued}

I 'rs;'s.onoR. for mor',tor~naorcr'tionoln:

rimo'rv tO 0000, cr',IooI:Cooo

&ý Piro.-osons for SO tube ropoir moethods. 8toomg gonoroator tube ropotF mogthods shall pro.ido tho moans to r..otablioh tho PR0 prosouro boundary intogrity' of Sc tubos without

.omoving tho tube fFrom sob..... For tho pdrpeses ef thPuse S--ifie.tie.., tube plugging is not:a ropoir. All oseeptable tub: ropoir methots ore listed bolew-.

4-W

-stinghou.o Laok Limiting Alloy 800 oleovos os.dosorib. din WOAP 16013 P Revision 2 (with rango of sonditinO soa roeylsod in Appondix A of WCAP 16489 NP, Roisi*; 0). Looak Limiting Alloy 800 Slooves or.

Ippiooblo on*ly to the orFiginal otoamn gonreatars. Prior to installation of oaah sloovo, tho Icoo~tionR whoro theosloovo jontsoroetob6eostablished sholl b:enpotd

m.

Control Room Envelope Habitability Program A Control Room Envelope (ORE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Air Cleanup System (CREACS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREACS, operating at the flow rate required by the VFTP, at a Frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 36 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

ST. LUCIE - UNIT 2 6-15i Amendment No. -447, 153

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 18 of 19 ADMINISTRATIVE CONTROLS (continued)

n.

Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

(i) Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

An API gravity or an absolute specific gravity within limits,

2.

A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3.

A clear and bright appearance with proper color or a water and sediment content within limits; (ii) Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and (iii) Total particulate concentration of the fuel oil is

10 mg/I when tested every 31 days.

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

(*

This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

ST. LUCIE - UNIT 2

&-15j Amendment No. 155

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 ArMIMITPATIVIF f(lflrTRflI C (_nntinm*ditl L-2009-262 Page 19 of 19 STEAM GENERATOR TUBE INSPECTION REPOQRT (continuod)

&.9.4.3 A rcport shall be submitted Yvithin 180 days aftor the in HOT SHUJTDO)WN followin~g oomnplotiori of an iropootie po~ffrmod in aGocodanoc with Speoifloation 6.8.4.1.2. T a-The seepe zf inspeetoens pciorfnsd en eoch SC, bý- Aetivo dogradatoen moohoniongo found, 0,

Nondootruotivo cwa-minatinn tochniguoc utilizod fc mneeheeiste nl of the originol SGo ReODOR Gfra:: :neluno r oaoh doarandatin d,-

Loation, oriCntatiCn (if line.O, and r,, aur.d sizes (if available) of r.io.

i nduood indiootiono, e,

Numbor zf tubes pluggod Cr rclpairod during the inspotien outage for oaoh aetive degreeate datt,,

sn9 f-Tot.l numbr ad p.*tago of tubes Pg BF *..airod to dat.,

,,ThCe rCultC Cf Condilin mnitering, inluding the rult, of tube pulls and in situ 4h, The offootivoe plugging pereentagC for all plugging and tube repairo in Ca~h SC, end fr.

Repair mo.thod utilized and tho number of tubes ropairod by oaoh r.pair moethod.

The following information eonoornling indicationo found in the tuboshoot rogion (ineluding the expansion tranoftion) chatl be includod in this ropeol:

p Numbor of total indiontiono, looction of ocob indication, Orientaltion of ccoh

= M-ml ay e

.h.

h indio.tion, and

.v.thc th i

V the inside or outside dia"et*.*

k7 The oumulativo numnbor of indicotiono dotootod in the tuboohoot rogionc fu nction of cicyation Wthin tho tubecphoot.

4.

Projocted end of cyole acoident ind-o.d leakage ferom tube-he^t indioctiono 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 6-20~f Amendment No. 147

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 1 of 18 Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections Word Processed TS Unit 1 Page V Page VI Page 3/4 3-43 Page 3/4 4-26 Page 6-6 Unit 2 Page VI Page XXIV Page 3/4 1-12 Page 3/4 3-24 Page 3/4 3-41 Page 3/4 4-39 Page 6-7 Page 6-15f Page 6-15g Page 6-15i Page 6-15j Page 6-20f

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 2 of 18 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION 3/4.4.4 3/4.4.5 3/4.4.6 3/4.4.7 3/4.4.8 3/4.4.9 3/4.4.10 3/4.4.11 3/4.4.12 3/4.4.13 3/4.4.14 3/4.4.15 PAGE PRESSURIZER..................................................................................................

3/44-4 STEAM G ENERATOR (SG) TUBE INTEG RITY................................................

3/4 4-5 REACTO R COO LANT SYSTEM LEAKAG E....................................................

3/4 4-12 Leakage Detection Systems............................................................................

3/4 4-12 Reactor Coolant System Leakage....................................................................

3/4 4-14 CHEM ISTRY....................................................

................................................ 3/4 4-15 SPECIFIC ACTIVITY........................................................................................

3/4 4-17 PRESSURE/TEM PERATURE LIM ITS.............................................................

3/4 4-21 Reactor Coolant System...................................................................................

3/4 4-21 Pressurizer........................................................................................................

3/4 4-25 DELETED.........................................................................................................

3/4 4-26 DELETED.........................................................................................................

3/4 4-56 PORV BLOCK VALVES....................................................................................

3/4 4-58 POW ER O PERATED RELIEF VALVES...........................................................

3/4 4-59 REACTO R COOLANT PUM P - STARTING....................................................

3/4 4-60 REACTO R COO LANT SYSTEM VENTS.........................................................

3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTIO N TA NKS.............................................................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - Tavg > 3256F.................................................................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tavg < 325 0F.................................................................

3/4 5-7 3/4.5.4 REFUELING W ATER TANK...............................................................................

3/4 5-8 ST. LUCIE - UNIT 1 V

Amendment No. 28. 60, 68, 80, 434,200,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 3 of 18 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CO NTAINM ENT VESSEL...................................................................................

3/4 6-1 Containm ent Vessel Integrity..............................................................................

3/4 6-1 Containment Leakage.........................................................................................

3/4 6-2 Containm ent Air Locks......................................................................................

3/4 6-10 Internal Pressure...............................................................................................

314 6-12 Air Tem perature................................................................................................

3/4 6-13 Containment Vessel Structural Integrity............................................................

3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.........................................

3/4 6-15 Containment Spray and Cooling System s........................................................

3/4 6-15 Spray Additive System....................................................................................

3/4 6-16a 3/4.6.3

'CO NTAINM ENT ISO LATIO N VALVES............................................................

3/4 6-18 3/4.6.4 DELETED.........................................................................................................

3/4 6-23 DELETED.........................................................................................................

3/4 6-24 DELETED.........................................................................................................

3/4 6-25 3/4.6.5 VAC UUM RELIEF VALVES..............................................................................

3/4 6-26 3/4.6.6 SECO NDARY CO NTAINM ENT........................................................................

3/4 6-27 Shield Building Ventilation System...................................................................

3/4 6-27 Shield Building Integrity....................................................................................

3/4 6-30 Shield Building Structural Integrity....................................................................

3/4 6-31 3/4.7 PLANT SYSTEMS 3/.4.7.1 TURBINE CYCLE...............................................................................................

3/4 7-1 Safety Valves......................................................................................................

3/4 7-1 Auxiliary Feedwater System...............................................................................

3/4 7-4 Condensate Storage Tank..................................................................................

3/4 7-6 Activity.................................................................................................................

3/4 7-7 M ain Steam Line Isolation Valves.......................................................................

3/4 7-9 ST. LUCIE - UNIT 1 VI Amendment No. 27, 434-, 4-34, 24,

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 4 of 18 TABLE 3.3-11 (continued)

ACTION STATEMENTS ACTION 1 -

With the number of OPERABLE channels less than the Total No. of Channels shown in Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 30 days or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 2 With position indication inoperable, restore the inoperable indicator to OPERABLE status or close the associated PORV block valve and remove power from its operator within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 3 -

With any individual valve position indicator inoperable, obtain quench tank temperature, level and pressure information once per shift to determine valve position.

ACTION 4 With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-11, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to the specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

ACTION 5 -

With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; and

2.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status; and

3.

Restore the Channel to OPERABLE status at the next scheduled refueling.

ACTION 6 -

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ACTION 7 -

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-11, either restore the inoperable channel(s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

ST. LUCIE - UNIT I 3/4 3-43 Amendment No. 3-, 79, 465,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 5 of 18 DELETED ST. LUCIE - UNIT 1 314 4-26 Amendment No. 69, 90,

I,.'

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 6 of 18 6.0 ADMINISTRATIVE CONTROLS 6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI / ANS-3.1-1978 for comparable positions, except for:

(1) the radiation protection manager who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor who shall have specific training in plant design and plant operating characteristics, including transients and accidents, and any of the following educational requirements:

Bachelor's degree in engineering from an accredited institution; or Professional Engineer's (PE) license obtained by successful corhpletion of the PE examination; or Bachelor's degree in engineering technology from an accredited institution, including course work in the physical, mathematical, or engineering sciences, or Bachelor's degree in physical science from an accredited institution, including course work in the physical, mathematical, or engineering sciences.

(3) the Multi-Discipline Supervisors who shall meet or exceed the following requirements:

a.

Education: Minimum of a high school diploma or equivalent.

b.

Experience: Minimum of four years of related technical experience, which shall include three years power plant experience of which one year is at a nuclear power plant.

c.

Training: Complete the Multi-Discipline Supervisor training program.

For the purpose of 10 CFR 55.4, a licensed senior reactor operator and a licensed reactor operator are those individuals who, in addition to meeting the requirements of 6.3.1, perform the functions described in 10 CFR 50.54(m).

6.4 TRAINING 6.4.1 Each member of the unit staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 3.

6.5 DELETED ST. LUCIE - UNIT 1 6-6 Amendment No. 25, 37, 69,

4_26, 464, 4-73, 4 89, 4 99,

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 7 of 18 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES D E LET E D...................................................................................................

3/4 4-7 O P E R A T IN G..............................................................................................

3/4 4-8 3/4.4.3 P R ES S U R IZ E R..................................................................................................

3/4 4-9 3/4.4.4 PO RV BLO C K VALV ES....................................................................................

3/4 4-10 3/4.4.5 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 C H E M IS T RY.....................................................................................................

3/4 4-22 3/4.4.8 S P EC IFIC AC T IV ITY........................................................................................

3/4 4-25 3/4.4.9 PRESSURE/TEMPERATURE LIMITS REACTOR COOLANT SYSTEM......................................................................

3/4 4-29 PRESSURIZER HEATUP/COOLDOWN 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 D E LET E D.........................................................................................................

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 > 3251F................................................................

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - Tayg <325OF................................................................

3/4 5-7 3/4.5.4 REFUELING W ATER TA NK...............................................................................

3/4 5-8 ST. LUCIE - UNIT 2 V1 Amendment No. 46, 440, 1-47,

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 8 of 18 INDEX LIST OF TABLES (Continued)

TABLE PAGE 3.3-9 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION................................

3/4 3-39 4.3-6 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION SURVEILLANCE R E Q U IR E M E N T S............................................................................................

3/4 3-40 3.3-10 ACCIDENT MONITORING INSTRUMENTATION...........................................

3/4 3-42 4.3-7 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE R E Q U IR E M E N T S............................................................................................

3/4 3-43 3.3-11 DELETED 3.3-12 DELETED 4.3-8 DELETED 3.3-13 DELETED 4.3-9 DELETED 4.4-1 MINIMUM NUMBER OF STEAM GENERATORS TO BE INSPECTED DURING INSERVICE INSPECTIO N...............................................................

3/4 4-16 4.4-2 STEAM GENERATOR TUBE INSPECTION...................................................

3/4 4-17 3.4-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES............. 3/4 4-21 3.4-2 REACTOR COOLANT SYSTEM CHEMISTRY...............................................

3/4 4-23 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE R E Q U IR E M E N T S............................................................................................

3/4 4-24 4.4-4 PRIMARY COOLANT SPECIFIC ACTIVITY SAMPLE AND ANALYSIS P R O G R A M......................................................................................................

3/4 4-2 7 4.4-5 DELETED 3.4-3 LOWTEMPERATURE RCS OVERPRESSURE PROTECTION RANGE..... 3/4 4-37a 3.4-4 MINIMUM COLD LEG TEMPERATURE FOR PORV USE FOR LTOP........ 3/4 4-37a 3.6-1 DELETED 3.6-2 CONTAINMENT ISOLATION VALVES...........................................................

3/4 6-21 ST. LUCIE - UNIT 2 XXIV Amendment No. 8, 53, 54, 73,86,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 9 of 18 REACTIVITY CONTROL SYSTEMS BORIC ACID MAKEUP PUMPS - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 At least the boric acid makeup pump(s) in the boron injection flow path(s) required OPERABLE pursuant to Specification 3.1.2.2 shall be OPERABLE and capable of being powered from an OPERABLE emergency bus if the flow path through the boric acid pump(s) in Specification 3.1.2.2 is OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With no boric acid makeup pump required for the boron injection flow path(s) pursuant to Specification 3.1.2.2 operable, restore the boric acid makeup pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 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 borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 200°F; restore the above required boric acid makeup pump(s) to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.6 The above required boric acid makeup pump(s) shall be demonstrated OPERABLE by verifying that the pump(s) develop the specified discharge pressure when tested pursuant to the Inservice Testing Program.

ST. LUCIE - UNIT 2 3/4 1-12 Amendment No.,

25, 40, 64, 405,

436,

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 10 of 18 INSTRUMENTATION 314.3.3 MONITORING INSTRUMENTATION RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.1 The radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarrnitrip setpoints within the specified limits.

APPLICABILITY: As shown in Table 3.3-6.

ACTION:

a.

With a radiation monitoring channel alarm/trip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable.

b.

With one or more radiation monitoring channels inoperable, take the ACTION shown in Table 3.3-6.

c.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.3.3.1 Each radiation monitoring instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations forthe MODES and at the frequencies shown in Table 4.3-3.

4.3.3.2 At least once per 18 months, each Control Room Isolation radiation monitoring instrumentation channel shall be demonstrated OPERABLE by verifying that the response time of the channel is within limits.

ST. LUCIE - UNIT 2 a/4 3-24 Amendment No. 452,

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 11 of 18 INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a.*

With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days, or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.*

With the number of OPERABLE accident monitoring channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c.** With the number of OPERABLE Channels one less than the Total Number of Channels shown in Table 3.3-10, either restore the inoperable channel to OPERABLE status within 7 days if repairs are feasible without shutting down or prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

d.** With the number of OPERABLE Channels less than the Minimum Channels OPERABLE requirements of Table 3.3-10, either restore the inoperable channel(s) to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months if repairs are feasible without shutting down or:

1.

Initiate an alternate method of monitoring the reactor vessel inventory; and

2.

Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days following the event out-lining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status, and

3.

Restore the Channel to OPERABLE status at the next scheduled refueling.

e.

The provisions of Specification 3.0.4 are not applicable.

Action statements do not apply to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

Action statements apply only to Reactor Vessel Level Monitoring System, Containment Sump Water Level (narrow range) and Containment Sump Water Level (wide range) instruments.

ST. LUCIE - UNIT 2 3/4 3-41 Amendment No. 49, 46,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 12 of 18 DELETED ST. LUCIE - UNIT 2 3/4 4-39 Amendment No.

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 13 of 18 6.0 ADMINISTRATIVE CONTROLS 6.4 TRAINING 6.4.1 Each member of the unit staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Revision 3.

6.5 DELETED ST. LUCIE - UNIT 2 6-7 Amendment No. 4-8, 66,402,443,446,

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 14 of 18 ADMINISTRATIVE CONTROLS (continued_

I.

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.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

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

2.

Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50%

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

3.

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

e.

Provisions for monitoring operational primary-to-secondary leakage ST. LUCIE - UNIT 2 6-15f Amendment No. 447,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 15 of 18 ADMINISTRATIVE CONTROLS (continuedi PAGES 6-15g AND 6-15h HAVE BEEN DELETED.

THE NEXT PAGE IS 6-15i.

ST. LUCIE - UNIT 2 6-15g Amendment No. 447,

St. Lucie Units I and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 16 of 18 ADMINISTRATIVE CONTROLS (continued)

m.

Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Air Cleanup System (CREACS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c.

Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors,' Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREACS, operating at the flow rate required by the VFTP, at a Frequency of 36 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 36 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

ST. LUCIE - UNIT 2 6-15i Amendment No. 447, 4-53,

St. Lucie Units 1 and 2 L-2009-262 Docket Nos. 50-335 and 50-389 Page 17 of 18 ADMINISTRATIVE CONTROLS (continued)

n.

Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

(i) Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:

1.

An API gravity or an absolute specific gravity within limits,

2.

A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3.

A clear and bright appearance with proper color or a water and sediment content within limits; (ii) Other properties for ASTM 2D fuel oil are within limits within 31 days following sampling and addition to storage tanks; and (iii) Total particulate concentration of the fuel oil is 5 10 mgfl when tested every 31 days.

The provisions of SR 4.0.2 and SR 4.0.3 are applicable to the Diesel Fuel Oil Testing Program test frequencies.

o.

Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendation of Regulatory position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

ST. LUCIE - UNIT 2 6-15j Amendment No. 4-65,

St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2009-262 Page 18 of 18 ADMINISTRATIVE CONTROLS (continued)

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. 447,

v t e Letter Sequence Request
TAC:ME3489, (Approved, Closed)
Initiation Request, Request Acceptance Administration Questions Answers Results Approval Other:
MONTHYEAR2010-12-28 [Table View]

L-2009-262, Proposed License Amendment Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections

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