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{{Adams
#REDIRECT [[L-2010-122, Proposed License Amendment, Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections]]
| number = ML102150172
| issue date = 07/30/2010
| title = St. Lucie, Units 1 & 2, Proposed License Amendment, Application to Delete Structural Integrity Technical Specifications, Update Accident Monitoring Instrumentation Requirements, and Minor Corrections
| author name = Anderson R L
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000335, 05000389
| license number = DPR-067, NPF-016
| contact person =
| case reference number = L-2010-122
| document type = Letter, License-Application for Facility Operating License (Amend/Renewal) DKT 50
| page count = 60
| project =
| stage = Request
}}
 
=Text=
{{#Wiki_filter:Florida Power & Light Company, 6501 S. Ocean Drive, Jensen Beach, FL 34957 0 July 30, 2010 FPL L-2010-122 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 Via Florida Power and Light Company (FPL) letter L-2009-262 dated December 14, 2009, FPL requested approval of a change to St. Lucie Units 1 and 2 Facility Operating Licenses DPR-67 and NPF-16, respectively.
The amendment request's purpose was 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, and also made several administrative corrections based on obvious typos, previous amendments, or obsolete requirements.
One of the non-technical changes involved administrative TS 6.4 for the training program.Subsequent to the December 2009 submittal, the St. Lucie training department determined that the proposed TS change was inconsistent with the St. Lucie training program. This submittal provides changes to the administrative TS training requirements and supersedes the changes contained in L-2009-262 for TS 6.3 and 6.4.This change remains bounded by the original no significant hazards consideration contained in the original submittal.
Attachment 1 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-2010-122 Page 2 I declare under penalty of perjury that the foregoing is true and correct.Executed on the 3Ct" day of ,5k 1.Very truly yours, Richard L. Anderson Site Vice President St. Lucie Plant Attachments
,2010 cc: Mr. William Passetti, Florida Department of Health St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment I 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 I and 2 L-2010-122 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 Training (administrative in nature)The proposed change deletes TS 6.4, Training Program in its entirety, and relocates the definition of licensed operators from TS 6.3.1 to new TS 6.3.2.This change eliminates the obsolete training references to 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 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-2010-122 Docket Nos. 50-335 and 50-389 Attachment I 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 Training Requirements The existing text for TS 6.4, Training, will be eliminated.
The last sentence in TS 6.3.1 will be removed from 6.3.1 and renumbered as step 6.3.2.2.4 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 hours and HOT SHUTDOWN in 12 hours.2.5 Minor changes (administrative in nature)" Unit 2 TS 3.1.2.6 -change the action to be 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..."" Unit 2 TS SR 4.3.3.2 -correct typo.Change the SR to read "At least once per 18 months..."
St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 4 of 15 0 Unit 1 TS index page VI -correct section heading for CONTAINMENT SYSTEMS Change the section heading to read "3/4.6 CONTAINMENT SYSTEMS..." 0 Unit 2 TS index page XXIV -Table 3.6-1.Delete Table 3.6.1.0 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 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 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment I Page 5 of 15 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 TSs, 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.
 
===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 I of the March 28, 1980 NRC letter to all licensees as outlined in Section 6.4, Training, of the plant TSs.
St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment I Page 6 of 15 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 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 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.On July 22, 1993, the Commission issued its Final Policy Statement on Technical Specifications Improvements, 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 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment I Page 7 of 15 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.However 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;(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.
St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 8 of 15 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 is an acceptable alternative to some requirements outlined in 10 CFR 50.120 (regarding training programs) and 10 CFR Part 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 analysis reports (FSARs) and technical specifications (TSs)), as necessary, to conform to 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), or 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 eliminate the administrative TS for training, as it contains obsolete references that are more restrictive than those required by the current TS qualification requirements of ANSI ANS 3.1-1978 and the NANT guidelines.
The last sentence in 6.3.1 will now be numbered as 6.3.2. The Staff has accepted ANSI ANS 3.1-1978 qualification requirements for the St. Lucie St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment I Page 9 of 15 facilities, and this component of training qualifications will remain unchanged from the current obligation.
The format and content in TS 6.3.1 will now conform to NUREG-1432 with respect to unit staff qualifications and training.Because the St. Lucie St. Lucie operator licensing training programs are SAT based and are accredited through the NNAB, this proposed TS 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 action statement." Unit 2 TS SR 4.3.3.2 -This change corrects an obvious typo in the action statement." Unit I 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 St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 10 of 15 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 I 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 TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) do 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 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 IS] program required by TS 4.0.5 and is not specifically monitored or controlled during St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 11 of 15 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 TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) do 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 TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) do 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 TSs 3/4.4.10 (Unit 1) and 3/4.4.11 (Unit 2) be maintained in the St. Lucie TSs. This conclusion is consistent with 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 for Training and Staff Qualification
-These are administrative changes.5.4 TS 3.3.3.8, Table 3.3-11 (Unit 1) and TS 3.3.3.6 (Unit 2)
St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 12 of 15 The St. Lucie Unit 1 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 hours) through the intervening state of HOT STANDBY (with a completion time of 6 hours). 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 hours. The allowed outage time, 12 hours, is consistent with the NUREG-1432 allowed outage time for entry into Mode 4 and is acceptable.
FPL proposes to include the 6 hour 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 I and 2 L-2010-122 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 I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 1 Page 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-2010-122 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 I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 1 of 22 Attachment 2 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 XIV Page 3/4 3-43 Page 3/4 4-26 Page 6-6 Unit 2 Page VI Page XVIII Page XXIV Page 3/4 1-12 Page 3/4 3-24 Page 3/4 3-41 Page 3/4 4-39 Page 6-6 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 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 2 of 22 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVI:ILLANCE REOUIREMPINTS 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 P EC IFIC A C T IV 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 TRUJCTU,'RAL INTECRI ... ...... ..................
............................
3/4 4-26 A3ME Cede e',, i, 2, 2 nd 3 e. .., oe, .t .....................................................
344 2C 3/4 .4 .11 D E LET E D .........................................................................................................
3/4 4-56 3/4.4.12 PO RV BLOC K VA LV ES ....................................................................................
3/4 4-58 3/4.4.13 POWER OPERATED RELIEF VALVES ...........................................................
3/4 4-59 3/4.4.14 REACTOR COOLANT PUMP -STARTING ....................................................
3/4 4-60 3/4.4.15 REACTOR COOLANT SYSTEM VENTS .........................................................
3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3/4.5.1 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&deg;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 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 3 of 22 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS S CMON PAGE 314,4.6 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 Tem perature ................................................................................................
3/416-13 Containment Vessel Structural Integrity
............................................................
3/4 6-14 3/4.6.2 DEPRESSURIZATIO N AND COO LING SYSTEM S .........................................
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 LATION 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 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 M ain Steam Line Isolation Valves .......................................................................
3/4 7-9 ST. LUCIE -UNIT 1 VI Amendment No. 27, 4-3--, 4-34, 204 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 4 of 22 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 .1 R E S PO N S IB ILIT Y ......................................................................................................
6-1 6.2 ORGANIZATION
 
====6.2.1 ONSITE====
AND OFFSITE ORGANIZATION
................................................................
6-1 6 .2 .2 U N IT S T A F F .............................................................................................................
6 -2 6.2.3 SHIFT TECHNICAL ADVISOR FUNCTION ...............................................................
6-5 6.3 UN IT STA FF Q UA LIFIC ATIO NS ................................................................................
6-6 6 .4 T R ,A, 'W ..................................................................................................................
6 -6 6.5 DELETED DELETED ST. LUCIE -UNIT 1 xIV Amendment No. 69, 93, 434, 49G9,49 St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 5 of 22 TABLE 3.3-11 (continued)
ACTION STATEMENTS ACTION 1 -Wrth 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 8r be in HOT STANDBY ,Dv-hin the ncxt 12 hUFc'.ACTION 2 -W'th position indication inopera restore the inoperable indicator to OPERABLE stat or close the associated PORV block valve and remove p:r from its operator within 48 hours 6I a- M rA kiIOSJ ..,L:.. ,u......,.4 0 L...or be in STAND and HO in 12 hc 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 -ith t number of OPERABLE Channels one less than the Total Nu er of Channels shown in Table 3.3-11, either restore the i f perable channel to OPERABLE status within 7 days if repairs DHOT are feasible without shutting down or prepare and submit a BY in 6 hours Special Report to the Commission pursuant to the specification T SHUTDOWON
 
====6.9.2 within====
30 days following the event outlining the action.urs ..) taken, the cause of the inoperability and the plans and schedule xX. for restoring the system to OPERABLE status.i the number of OPERABLE Channels less than the Minimum ,nels OPERABLE requirements of Table 3.3-11, either restore nsiserable channel(s) to OPERABLE status within 48 hours ppairare feasible without shutting down or: alternate method of monitoring the reactor Vntory; and 2. &#xfd;repareand Tbmit a Special Report to the Commission 5rsuant to Spe ification
 
====6.9.2 within====
30 days following th event action taken, the cause of the ino rability and the"lans and schedule for restoring the s stem to OPERA E status; and 3. Restor the Channel to 0 RABLE status at the next schedule refueling.
ACTION 6 -With the numbe of OPERABLE acci nt monitoring channels less than the Minimu Channels OPERABL requirements of Table 3.3-11, either restore the i perable channel(s) to PERABLE status within 48 hours ,U,- I T.
-:fhdR the next 12 h--r.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 hours eo_V ST. LUCIE -UNIT 1 3/4 3-43 Amendment No. 37, 7-, 165 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 6 of 22 REACTOR COOLANT SYSTEM 3-4'" STUThRAL INV WGIRA fl ASME CODE CLASS 1. 2. AND 3 C 'MPONENTS LIMING CONDITION FOR flflIflAPltMI
&.4.0.4 The stutrlinort f ASME Codo Clacss1, 2 and 3GOcmponcntS (cxccpt stcam gcr.crator tubes) shall bo maintained in aee.dn. rt..W .th Spcsifica.tion 4.4.A0.1.APPLICAB1.LIWy:
Ad! MODES.AGT4G a, Wit&#xb6;h the otruotural intOgrity of any ASME Code 1 oompancnt(s) not confO.rming to tho abo. .r..uircm .nts, roctoro thc StruouriGIa integrity Of tho affeoted oornponcnt(z) to within its imitA or icolate the cifootod-ocmponont(o) prior to inro~aoirng the RoarAor Cooant Syotom tomporaturo mMcO than 502F abo'o the minimum t.mp..rturo ro ..uiro by NOT onidorationo.
With tho otruoturol kintogiy of ony ASME Cote Cleso 2 oomponont(o) not rfoonforin~g to tho abovo roguiromento, rootoro tho otruoturol intogrit; of the affootod ocmponont(o) to v~hin its lirmi Or isolato tho affested.om..ponent(s) pFrio to inracing the Re-"tor Coo'lat Systom temporoturo above 2001F.Wh t49he Mstrucural int.. rity of any ASMAE CodeClass 3 oemponcct(s}
not oorfefori!Rg to the obove reguircmcnto, restore tho struoturel integrity Of the affooted oomponont(o) to within its limint or icebato the affeotod oornponont(s) fromn sory~oc.Tho roMV0iesin of Soooifioation 3.0.4 arce not oppic &#xfd;abic.r-SURVEILLANCE REQUIREMENTS 44I)INe addit'SpcTifiaCtIE-N N.0..ST. LUCIE -UNIT 1 Onai u ,i,,,iianoo -guir,,,n, onor man.. t ,noc ey 3/4 4-26 Amendment No. 69, 90 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 7 of 22 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 TRANIN th.4 .a .A ' .... MA ....... .;..-epefe fS9~~IB18 h 6.5 DELETED ST. LUCIE -UNIT 1 6-6 Amendment No. 25, 37, 69, 4-28, 4454, 4-7-, 4 89, 4.99 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 8 of 22 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 E S S U R IZ E R ..................................................................................................
3/4 4-9 3/4.4.4 PO RV BLO C K V ALV 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 O PERATIO NAL LEAKAG E ......................................................................
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 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 .................................................
3/4 4-35 3/4.4.10 REACTOR COOLANT SYSTEM VENTS ........................................................
3/4 4-38 3/4.4.11 ,3TFRuCTURAL I NTEGRITY..
.................
....... ............
3/4 4-39 3/4.5 EMERGENCY CORE COOLING S7YST7EkMS ( E ;&#xfd;3/4.5.1 SAFETY INJECTIO N TANKS .............................................................................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS
-Tavg > 325&deg;F ................................................................
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-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 9 of 22 INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY
......................................................................................................
6-1 6.2 ORGANIZATION
.................................................................................................
6-1 6.2.1 ONSITE AND OFFSITE ORGANIZATION
.................................................................
6-1 6.2.2 UNIT STAFF ..............................................................................................................
6-2 6.2.3 SHIFT TECHNICAL ADVISOR FUNCTION ...............................................................
6-6 6.3 UNIT STAFF QUALIFICATIONS
................................................................................
6-6 6 .4 T R ,A I NI G ..................................................................................................................
6 -7 6 .5 D E L E T E D ...... ...........................................................................................................
6 -7 DELETED ST. LUCIE -UNIT 2 XVIII Amendment No. 29, 7-3, 89, 43, 446 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 10 of 22 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 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 CON'TAINMENT LEAKACE PATHC .................................................................
3/4 65 3.6-2 CONTAINMENT IS TION VALVES....
.. ...................
......................
3/4 6-21 ST. LUCIE -UNIT 2 XyIV Amendment No. 8, 53, 54, 7, 886 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 1 I of 22 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.operable ACTION: With no boric acid makeup pump requi d for the boron injection flow path(s)pursuant to Specification 3.1.2.2 inopea", restore the boric acid makeup pump to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and borated to a SHUTDOWN MARGIN equivalent to its COLR limit at 2001F; 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 hours.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. 8, 25, 40, 94, 405, 136 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 12 of 22 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 alarmltrip setpoints within the specified limits.APPLICABILITY:
As shown in Table 3.3-6.ACTION: a. With a radiation monitoring channel alarrmtrip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours 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 CHANN EL 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 instruentation channel shall be demonstrated OPERABLE by verifying that the respon 't of t channel is within limits.ST. LUCIE -UNIT 2 3/4 3-24 Amendment No. 152 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 13 of 22 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.
1, 2 nd 3 or be in HOT STANDBY in 6 APPLICABILITY:
MODES 1, 2 and 3. hours and HOT SHUTDOWN in 12 hours ACTION: a.* With the number of OPERABLE acciden m itoring channels less than the Required Number of Channel sshhwin Table 3.3-10, either restore the inoperable channe 0 ERABLE status within b. With the number of OPERABLE accidt monitoring channels less than the Minimum Channels OPERABLE r quirements of Table 3.3-10, either restore the inoperable channels to ERABLE status within 48 hours OFbe int leat HO'T- 6SHUTDOWVN
.-itho thez next 12 hours.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 hours 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 1 and 2 Docket Nos. 50-335 and 50-389 REACT-OR COOLANT SYSTEM 314.4.44 TUTRL ERT L-2010-122 Attachment 2 Page 14 of 22 LIMMIlNC CONDIllON FOR OPERATION shall be mnar.taned On aacardcme APPLICABILITY:
ALL MODES AGiGW. .o of AGME Co.. G.as 1 2. and.......t.
I e with 4.4.11.ft-XAMh the ;tfuett.,e
'.1 teg. t, of ., ACME Code Class 1 eeopmponcn" integrity of the effoated eenmparnet(s) tC Mihqmint-.
fimit...Ae izolate th: affeated eemponort(G) prior to inoroaoirng the Roactor Coolant Systerm tempeaur 0 Uo e m then 60OF above the mgnimumn tenperaturC reguired by NDT eonskdelroticnG.
WV~h the Struotural integrity of any ACME Code Class 2 comnponent".gt efOF..l .n0t theaobove reguirzmcR~tS, rasterC the Struetuzl integrity of the affootod Camnponont(s) to vwithin ito tm&#xfd; ar Coolant Systorn tcmgporature obavo 200 2 F.Wtth the 0 hotrural integrt; of any ACME Coda Class 300 enpamnat(s) not aornforirnig to the abovo roquiramareIt5, rGStOro the atruatural C7 flteg~t e the-effeete
+- The-mfovie
; Inc CTc~IC3 CCrnpanot tC '-,rn no e:mit Cr :501510.d CCmpnont from.. se..i. .a: ....o G o ofc t n 3'_ _& _&__r ..0. ar not IppmeabI SUF 4.4.11 in satr:or~ to Inc ~ .L. .A ......flk... r~r gus. asor ~csacr Coolant Pum.p flywheel shel be inspooted per the roommrendation, of Regulatory Poastin C.4b of Regulatory Guide 1.14, Revision 1, August 1075,.ST. LUCIE -UNIT 2 3/4 4-39 St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 15 of 22 6.0 ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR FUNCTION An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.6.3 UNIT STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI I 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.D 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).ST. LUCIE -UNIT 2 6-6 Amendment No. 5, 49, 65, 69, 402,443, 444 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 16 of 22 6.0 ADMINISTRATIVE CONTROLS 6.4 T.AI.U4Q 6.5#440a+Nfte ord rz-n-ozsmzt training F.c ;ar;m fc; th uinit ctaff shall be ,iRtair, d Ur-dcr eF BRd ShOll Fne 1 eF e)(eeed the Fequ'Ferments and F Iq,, R'i "R"A st'LR, "Fi 9- 1 49:78 and Q GFR RaFt 66 and the.... ... n &#xfd;antwnn A -2Rd Q Of RR61091 -WO I of the MaFGh 29, 4 &#xfd;Nb- wRG- a Rd 11-11&#xfd; iR.11 &#xfd;19 faMili A FIORWR 'A 4th FRIVIRRt "FAFY a PeFafig-Ra-R-*lp DELETED DELETED ST. LUCIE -UNIT 2 6-7 Amendment No. 43, 66, -02, 433, 446 St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 17 of 22 ADMINISTRATIVE CONTROLS (continued)
: 1. 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 2- A SC Ptrccram shall bc cstablishcd and implcmentcd far thc criiRanl SCc te encurz that 8G tube intcgrity is manitained.
IR additiOR, the SG PrcGramA Shall Ir-.clude the fcllc:.-ing a- PFE)YizEcn98f8Frzndit 09 mcF9tOF Ag assesment&.
Gadte RROii se~fei mcneen an ezvpluat OR of the -"a fcund" eendition of the tubin19 W419 roopeet to h perfOFrm@nee 6riteria for StrUcturcl integrity and acoident indused leackogo.
Th 11 aofHE"e~ 11 eeSt h edte f the t~b~igg d~rin a GC Fiopeeton outage, as Eletorminod fROM the Fiseory io inopeetien Fesults BF by othor H9n, ro to the plugging Or ropai Fof tubeso. Condition FRnORitring assess&R.11 osholl ho conductod during eaeoh ebtoge dufrng whieb the S GG llbe Weisee plugged or rocoiFed to BaMfiM that the Beorfforrnae~ ,iteF ai We beina Fet.ST. LUCIE -UNIT 2 6-15f Amendment No. 147 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 ADMINISTRATIVE CONTROLS Icontinued) 2- (ee~n~ued) b, Pzz , armar, ,Ladtuda for SC tube intzgri Fooeting 'th pztzrm.F;aty Orqrf.Lretudb lzakage, anid operationIga izakage.-1-- Structural ntz'-ty pzizrmonoz98 6on L-2010-122 Attachment 2 Page 18 of 22 bb~l tbflO irazant' Snail Eo mnaFRaainzeayn St. --te^.dOR tural intognit,, azoiden~tidud iAll 'R zznizz SCG lbLoc 81h@il roton otlartup, eporotion in the powe FRgrn 0 , hat J tab aRE d dodwn ond oill ant ipaptod taro n~tS .lue Wtd the deinse f eaie) nd9E dzoigni boo ozoidento.
This rotaoning a Safety fato1 r of 3 0 againot  Rnormoc, steady state full p.w. preatin pimar` te a.nd' pr... II uI dliffz~rntiai and a safety' fooor) Of 1. 1 gainst buar Iapplied to the dzoign bai ozoldon~t primary to azzondo~ ,.SS~ url iffeeq asi A Il fromF the obovo'roguiromonta, additional loading end'tien ass'eated wit the dooign baci aooidonto, or ozmbinatiz" of aeed rts +R aseeF de lb with te dlsiRn aond llonsing baoso, shall alc be eyaluated te dtmene if the a iselated lde contrib- zig&#xfd;nifizantly to barct or ociapse, In 4- ......nt.. of tube ingrity, those loads that do Signifoonltiy offzzt burd zr oollapoz shall be dztzrmnzFid cno assessed in ozEMbinatizn with the ieads duo to p~erooUro with a cafot faztoro 1 +2 on the aombinzd primoy ieads and 1.0 enaia seedayleds
-9 Azzidan-t induzod leakage pto9Frmanzz zritcnizn The Primor,' to ocondar,'acoidont induzd ieakage rFat for any design basci a-.idont, other than a SC tubz rupl.u , sha .... e.... .... .........F .... ............
.......in. t... .cf tOtal lcakage rFtc fo, r c' 3C and leakage rate for an 0Rindiviual 8C. Loahago io not to ocozoed 0.3 gailanO per FHR6ine total throuHgh aii 3Cc and 216 gallon" pzr day through any .n. SC.2, The opatij lekg p r',',n F ....d n L.. "Roaetr Coolant Syctem Opzratiznal Leakage." e, Prov, icina for SC tube ropair Oritoria 4- Tuboo found by.. nispotion to ..n.c n a flaw mn a ..... Feg4e.with a dEpth gquol tz or9F eox ding 10 p.r...n of t1hz nom.inal tube wali hlbe' shall be plugged or repa rFd exeept f .rnAttd t F..... ; in sarvice throughp appltiaton o"f thz tube ropir or+'itric Edi.ousod in T-eohni;, Spoifictogn 6.8.4.1.2.e.4. Tubes fzund by' nonize ie nspzztzan tezontainoa flow ini(a) aozozvzzrO (b) the pracsuro loundar' portion of the original tuba wall in the alov'; to tube joint be plugged.4, All tubes with oloovo that havo a niekel band ohall bo plugged aftor ono yot inoeperation.
J 4, The CA m.thodology, cc dce ..bed may bo appli.d to 1hz o.p.nd, pot of. ,h tubae n th hot leg tuboohetl rg.ian ao on allrnativ" to the 10 poeroondept pbasoed oritzrio zf Tzohnkzcl Gpeefiection 6.....o.ST. LUCIE -UNIT 2 6-15g Amendment No. 147 St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 19 of 22 ADMINISTRATIVE CONTROLS (continued) 2- e {eentttRued)
I : .boo Wth e i ot in tf ..... th..ht log tubohoeot ro gi.n bhl bo 8 rpio or plugg Hgod upon dotooti. of cry.. ifel .. Rtifi.d within .... ,J-n betwe tuhe bottom of. (h hot ieg oexpctieo trco3itioo or top of the tuboohoot, whiohovorF olevation is lowor. Plaws Izzatod tolvN i, Tuboo whieh hayo coyt portico of C oloovo joiot R toe hot log tubeohoot rzgion ohall be pluggod upon dotootioog of co~y flow that s0 looctod below tho .aWor oloo.. to tube joint and within 10 G)3 bolow tho bottom of tho hot l e en t e, eF r p of tho tubohot0 whio.h.vor d 7-Prov'iciro for 8G lube inopoctions.
Poriodic SC lube Reopeat ion ohoall be porfeformd.
Tho numbor cod. poioo. of tho tu .. r.pe.te. nRd mothodo of ioopo;tior ohall t p ....rod with the .. .tiv. of do.. ting flews of cry type (e.g., volum.tri.
flaws, cxicl Codoit6OUoRznticl o truo) hth oy t otg , ro f nt lngr the l'-tnh of the tutze, fro tho tubo to tuboohoot wold at tho tube olot to thoe tube t tubao.h.ot t ohe d the totube outlot, orEI that moay satiotj 1he applooablo tubo ropaiF aritoria.
The tube to tuboohoot weld not padt of toe tubo, For tubes with 00 pc~too of a lowor oloove o (Cit io 1tho hot log tutozhoot rogior, tho portion of tho tube below 10.3 inohoo fro~m the top of toe hot log, tubesee O Fd ex., si20 taronitcOR, whegee iSlow., is xeWtld wlel h Ilrotz~rocroritoric in TS Sozbzon 6.3. 1.1.2.e. crWe appliod. in taboo ropoirod by sleeoving, 1hz prio of the oriqql~a lut wall bohozoeR the Szlccozo jo tSin 10 nt an amea roguir4og iropeotion.
IRoadditien to mooetirgthe roguiromonte of d.l, d.2, d.3 orb d.14 below, the inopeOtiOn zoopo, iRopeotion Mothodo, orEb inopoolionR inteolo oholl ho zuoh ao to onouro ltha SG tube iologryis 10a ciotciood utin tho ocot SC inopoctioo.
An aoocoomont of doegrodotion ohatl ho porfformod to dotorminoR the typ and ooctio of tlwo to wigohl tho tubes Fay ho ouoeopt blo Cod, tcood 00n this c,,...,.,,,,,t-,l,.
doloFrmino whiah inopeolion methods Rood ta o oe mpiyod andJ at what Izoationo.
t- Inopoot 1004; ofthe tubes in zach SC during tho fin~t rFeuoling outage followin 2- Inopoot 100% of the tubes at zoguonliol Poriodo of 60 offootlvo full power mon)tho.The flint ooquooictl poriod shall ho coozsidorod to bogkq cftor (ho tro~t ioooeio inopeoofon of tho gso. No SG ohall oporoto for moro~ than 21 offootivo full powore FnethseFeig Ffuli Rg otgo (whiohovor lesso) without beiiq inspee,, 2lz ffoctivz fatl pewrn ~nthS rF rn fueli g oulogo (wehi 'e sls)4.- If oroob irdiootians are found iron~y Sc tube, thon tho root in peat'o rfF roa SG for (ho do Fdato Fflehoolm that oousod theorcoki, Adeie rshallno ouzood 21 offeet NefullIpwe Fnenhs en Fefuelin outage (hiheF less). if dofiniti~v 48Ff rm iG, oaeh asofromooeamination of a pulled tube Eiogncztiz Oz, de~F~y tes I o , Or zoginooriog zvoluatioR Rdis1 l toat rot ho Iroctod as r I~e(ST. LUCIE -UNIT 2 6-16h Amendment No. 147 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 ADMINISTRATIVE CONTROLS frnntinmadl L-2010-122 Attachment 2 Page 20 of 22 lv onoom ~sor.ororor iw: I TOO~Ofl ioontinuooi 2- (eontiuedl)
I 'rov;4;onS ttS, niF F 'tcrRing oporotio'nci pr'mor to oooonoioa Provisions for SC tue ropo F ethodos. Stoom go..o..tor tub ropo;r ..othodo oholl providoeto moono to rooctoblio!he tRCS pro..uro boundary of SC tubo-o*th.,out , omov, .. the tubeb from sooieo. For the pu rposes of thoso Speeofio etleni, tube plugging i not a r.. .p " elir. Al pt ,bl, tubo r.p. ir mo-thodo .ro lioted below -.4-Rovio.on 2 (with r.ngo of ondit FIon ao reo^od in Appondi. A  IP, Rovioon 90). Look UIitig Alloy 800 Slo... ..ppl 'ab" o "ly to tho or0 igina otoam gonoratoms.
PriorF to inotollotioen a! eah seo, tho loootior whoro ,them, sleeve ,,"rts t. bestaJ.blishod shall be. ipootod..
..m+ Control Room Envelope Habitability Prooram A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that GRE 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 ORE 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 ORE 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 ORE boundary into the GRE 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 GRE 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 ORE pressure relative to all external areas adjacent to the GRE 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 ORE boundary.e. The quantitative limits on unfiltered air inleakage into the GRE. 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 ORE 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 ORE habitability, determining CRE unfiltered inleakage, and measuring GRE 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-2010-122 Docket Nos. 50-335 and 50-389 Attachment 2 Page 21 of 22 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 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 6-15i Amendment No. 155 St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 2 Page 22 of 22 ADMINISTRATIVE CONTROLS (continued')
STEAM GENERATOR TUBE INSPECTION REPORT (continued) 6.9.-A o Arptoshall be 5u HOT LSHUTrDOW pForFrmed in asoeF b-. Aeti, vdegted 07 Nendestruet'-
mneehan'sem-d- 1eeebien-
: i. ie e, PN1umFbe IFe#-till setive degFed The results ef b, The ea.d R epei*-m~ethe The follow~ing infOrr brrittcd wvithin 180 days after the initial cntr; into fellowing eompiction of an inospeetion of the originol CGo anee wim Soeemeation 6.8.4.1.2.
+M Fevout shall inoiude&#xfd;rFiopcctvno Penamcio ac u otior. mooehonwrio found.oxamnatin tonniguc utnucca fcr coon acgroaaancn ntation (if Iine r), .,d ..BS~e size (of ovoilobic) of eve:cc pluggcd or ropairod during the inopcelion outage for eaeh stion r..che .L I. .-and poroontoaci of tuboo plucjcjd or rciaoired to daett'condition including thc .coult. of tubc pullo and in situ plugging pcroontagc for all plugging and tubc ropairo in cach SO, d uti-cd and the numbcr of t.'-bcS rpaird by c t-; .........;;.-;.nd;a t no found in the tubes (including thc cxar(oin tronof.bo) chall be j, Numbcr of 1totl indicationo, location e Indioticn, socvcrlit' of coch indic:tion, the inside or outoide don oeter.k7 The oumnulative numbor of indicotiono f includcd in thic report: each crint dctcte.d in the tub;ch.aoh rcpair moethod.heet fegien etien of eeeh tiono initicted fromn)et Fegflen a eet fldeeetwens.
of.
Wth'in,: theC t'--eshe, t.I-Irrcjcctea enai or cycle oeciaenr inaucea iecoogc tromn tuccon 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. 147 St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 1 of 21 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 XIV Page 3/4 3-43 Page 3/4 4-26 Page 6-6 Unit 2 Page VI Page XVIII Page XXIV Page 3/4 1-12 Page 3/4 3-24 Page 3/4 3-41 Page 3/4 4-39 Page 6-6 Page 6-7 Page 6-15f Page 6-15g Page 6-15i Page 6-15j Page 6-20f St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 2 of 21 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
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3/4 4-4 STEAM G ENERATOR (SG) TUBE INTEGRITY
................................................
314 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 ........................................................................................................
3/44-56 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 LA NT SYSTEM VENTS .........................................................
3/4 4-61 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)3/4.5.1 SAFETY INJECTION TANKS .............................................................................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS
-Tavg > 325OF .................................................................
3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS
-Tvg < 325 0 F .................................................................
3/4 5-7 3/4.5.4 REFUELING WATER TANK ....................................
.....................................
3/4 5-8 ST. LUCIE -UNIT 1 V Amendment No. 28, 60, 68, 80, 434, 200, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 3 of 2l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 CO NTAINM ENT VESSEL ...................................................................................
3/46-1 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 DEPRESSURIZATION AND COOLING 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 ............................................................
3/4 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 VACUUM RELIEF VALVES ..............................................................................
3/46-26 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
....................................................................
3/46-31 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 A ctiv ity .................................................................................................................
3 /4 7 -7 M ain Steam Line Isolation Valves .......................................................................
3/4 7-9 ST. LUCIE -UNIT 1 VI Amendment No. 27, 43-, 4-34, 204, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 4 of 2l INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 .1 R E S PO N S IB ILIT Y ........................................................................................................
6-1 6.2 ORGANIZATION 6.2.1 O NSITE AND OFFSITE O RGANIZATIO N ...................................................................
6-1 6 .2 .2 U N IT S T A F F .................................................................................................................
6 -2 6.2.3 SHIFT TECHNICAL ADVISO R FUNCTION .................................................................
6-5 6.3 UNIT STAFF Q UALIFICATIO NS ..................................................................................
6-6 6.4 DELETED 6.5 DELETED ST. LUCIE -UNIT 1 XlV Amendment No. 69, 93, 4-34,489,499, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 5 of 2l 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 hours and HOT SHUTDOWN in 12 hours.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 hours or be in HOT STANDBY in 6 hours and HOT SHUTDOWN in 12 hours.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 hours 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 hours or be in HOT STANDBY in 6 hours and HOT SHUTDOWN in 12 hours.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 hours or be in HOT STANDBY in 6 hours and HOT SHUTDOWN in 12 hours.ST. LUCIE -UNIT 1 314 3-43 Amendment No. -37, 7-9, 465, St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 6 of 21 DELETED ST. LUCIE -UNIT 1 3/4 4-26 Amendment No. 69, 99, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 7 of 2l 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.6.3.2 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 DELETED 6.5 DELETED ST. LUCIE -UNIT 1 6-6 Amendment No. 26. 37, 69,-26, 654, 489, 499, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 8 of 21 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4.2 SAFETY VALVES D E L E T E D ..................................................
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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 VA LV 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 O PERATIO NAL LEAKAG E ......................................................................
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 INJECTIO N TA NKS .............................................................................
3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS
-Tavg > 3250F F...............................4.....
3/45-3 3/4.5.3 ECCS SUBSYSTEMS
-Tavg < 3250F ................................................................
3/4 5-7 3/4.5.4 REFUELING W ATER TA NK ...............................................................................
3/4 5-8 ST. LUCIE -UNIT 2 VI Amendment No. 4-6, 440, 4-47, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 9 of 2l INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6 .1 R E S PO N S IB ILITY ........................................................................................................
6-1 6 .2 O R G A N IZ A T IO N ...........................................................................................................
6-1 6.2.1 ONSITE AND OFFSITE ORGANIZATION
...................................................................
6-1 6 .2 .2 U N IT S T A F F .................................................................................................................
6 -2 6.2.3 SHIFT TECHNICAL ADVISOR FUNCTION .................................................................
6-6 6.3 UNIT STAFF QUALIFICATIONS
..................................................................................
6-6 6 .4 D E L E T E D .....................................................................................................................
6 -7 6 .5 D E L E T E D ......................................................................................................................
6 -7 ST. LUCIE -UNIT 2 xviII Amendment No. 29, 73, 89, 43, 446, St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 10 of 2l INDEX LIST OF TABLES tContinued)
TABLE PAGE 3.3-9 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION
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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
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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 INSPECTION
...............................................................
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 .....................................................
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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 DELETED 3.6-2 CONTAINMENT ISOLATION VALVES ...........................................................
3/4 6-21 ST. LUCIE -UNIT 2)(XIV Amendment No. 8, 63, 64,-73, 86, St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 11 of 21 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 hours or be in at least HOT STANDBY within the next 6 hours 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 hours.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. 8, 26, 40, 94,  436, St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 12 of 21 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 alarmntrip setpoint exceeding the value shown in Table 3.3-6, adjust the setpoint to within the limit within 4 hours 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 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 3/4 3-24 Amendment No. 462, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 13 of 2l 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 hours and HOT SHUTDOWN in 12 hours.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 hours or be in HOT STANDBY in 6 hours and HOT SHUTDOWN in 12 hours.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 hours 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 2 314 3-41 Amendment No. -14R 46, St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 14 of 21 DELETED ST. LUCIE -UNIT 2 3/4 4-39 Amendment No.
St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 15 of 21 6.0 ADMINISTRATIVE CONTROLS 6.2.3 SHIFT TECHNICAL ADVISOR FUNCTION An individual shall provide advisory technical support to the unit operations shift crew in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. This individual shall meet the qualifications specified by the Commission Policy Statement on Engineering Expertise on Shift.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.6.3.2 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).ST. LUCIE -UNIT 2 6-6 Amendment No. 6, 49, 66, 69, 4-02, 44,3, 446, St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 16 of 21 6.0 ADMINISTRATIVE CONTROLS 6.4 DELETED 6.5 DELETED ST. LUCIE -UNIT 2 6-7 Amendment No. -V3, 66, 4-., 4-43, 4-46, St. Lucie Units I and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 17 of 21 ADMINISTRATIVE CONTROLS leontinuedl
: 1. 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-2010-122 Attachment 3 Page 18 of 21 ADMINISTRATIVE CONTROLS lcontinued)
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-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 19 of 21 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 ORE 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 ORE 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.The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining ORE unfiltered inleakage, and measuring GRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
ST. LUC]IE -UNIT 2 6-15i Amendment No. 447, 463, St. Lucie Units 1 and 2 L-2010-122 Docket Nos. 50-335 and 50-389 Attachment 3 Page 20 of 21 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 mgA 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. 465, St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 L-2010-122 Attachment 3 Page 21 of 21 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. 4-47,}}

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