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#REDIRECT [[L-2017-100, License Amendment Request (LAR) 248, Changes to Technical Specifications Regarding Inservice Testing Program and Inservice Inspection Program Requirements and Surveillance Frequency Control Program ...]]
| number = ML17180A447
| issue date = 06/28/2017
| title = Turkey Point Nuclear Plant, Units 3 and 4 - License Amendment Request (LAR) 248, Changes to Technical Specifications Regarding Inservice Testing Program and Inservice Inspection Program Requirements and Surveillance Frequency Control Progra
| author name = Summers T
| author affiliation = Florida Power & Light Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000250, 05000251
| license number = DPR-031, DPR-041
| contact person =
| case reference number = L-2017-100
| document type = Letter, Technical Specification, Amendment
| page count = 43
| project = CAC:MD5407, CAC:MD5408
}}
 
=Text=
{{#Wiki_filter:June 28, 2017  L-2017-100 10 CFR 50.90  Florida Power & Light Company  9760 SW 344 St., Florida City, FL  33035      U. S. Nuclear Regulatory Commission  Attn: Document Control Desk  Washington D C 20555-0001 RE:  Turkey Point Nuclear Plant, Units 3 and 4 Docket Nos. 50-250 and 50-251 Renewed Facility Operating Licenses DRR-31 and DPR-41  License Amendment Request (LAR) 248, Changes to Technical Specifications Regarding Inservice Testing Program and Inservice Inspection Program Requirements and Surveillance Frequency Control Program Applicability  Pursuant to 10 CFR Part 50.90, Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant (Turkey Point) Units 3 and 4, respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by relocating to licensee controlled documents select acceptance criteria specified in TS surveillance requirements (SRs) credited for satisfying Inservice Testing (IST) Program and Inservice Inspection (ISI) Program requirements, deleting the SRs for the American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components, replacing references to the Surveillance Frequency Control Program (SFCP) with reference to the Turkey Point IST Program where appropriate, establishing a Reactor Coolant Pump (RCP) Flywheel Inspection Program, and related editorial changes. The enclosure to this letter provides FPL's evaluation of the proposed changes. Attachment 1 to the enclosure provides the existing TS pages marked up to show the proposed changes. Attachment 2 provides the retyped (clean copy) TS pages with revision bars to identify the proposed changes. Attachment 3 provides existing TS Bases pages marked up to show the proposed changes. The TS Bases changes are provided for information only and will be incorporated in accordance with the TS Bases Control Program upon implementation of the approved amendment. FPL has determined that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with the change. The Turkey Point Onsite Review Group (ORG) has reviewed the proposed license amendment. In accordance with 10 CFR 50.91(b)(1), a copy of the proposed amendment is being forwarded to the designee for the State of Florida. This letter contains no new regulatory commitments. FPL requests approval and issuance of the proposed license amendment by June 30, 2018. Once approval is received, the amendment will be implemented within 120 days.
Should you have any questions regarding this submittal, please contact Mr. Mitch Guth, Turkey Point Licensing Manager, at (305)246-6698.   
 
Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 1 of 41    Enclosure  Evaluation of the Proposed Changes  Turkey Point Units 3 and 4 License Amendment Request (LAR) No. 248 Changes to Technical Specifications Regarding  Inservice Testing Program and Inservice Inspection Program  Requirements and Surveillance Frequency Control Program Applicability 1.0 SUMMARY DESCRIPTION  2.0 DESCRIPTION OF PROPOSED CHANGES  3.0 TECHNICAL EVALUATION  3.1 Deletion of SR 4.0.5 and Relocation of RCP Flywheel Inspection Requirements  3.2 Relocation of Acceptance Criteria in Select SRs Credited by the IST Program  3.3 Replacement of References to the SFCP in SRs Credited by the IST Program  4.0 REGULATORY EVALUATION  4.1 Applicable Regulatory Requirements/Criteria  4.2 No Significant Hazards Consideration Determination 4.3 Conclusions 5.0 ENVIRONMENTAL CONSIDERATION  6.0 REFERENCES  ---------------.. Attachment 1 - Proposed Technical Specification Pages (markup) Attachment 2 - Proposed Technical Specification Pages (clean copy) Attachment 3 - Proposed Technical Specification Bases Pages (markup), Information Only Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 2 of 41  Enclosure  1.0 SUMMARY DESCRIPTION  Florida Power & Light Company (FPL) hereby requests amendments to Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant (Turkey Point) Units 3 and 4, respectively. The proposed license amendments revise the Turkey Point Technical Specifications (TS) by relocating to licensee controlled documents select acceptance criteria specified in TS surveillance requirements (SRs) credited for satisfying Inservice Testing (IST) Program and Inservice Inspection (ISI) Program requirements, deleting the SRs for American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components, replacing references to the Surveillance Frequency Control Program (SFCP) with reference to the Turkey Point IST Program where appropriate, and establishing a Reactor Coolant Pump (RCP) Flywheel Inspection Program. The proposed license amendments also delete a duplicative SR regarding returning Pressure Isolation Valves (PIV) to service following maintenance and a footnote previously applicable to Unit 3 Cycle 26. The proposed changes serve to remove duplicative ASME Code examination and testing requirements from the TS and to ensure that the TS surveillances credited for satisfying IST Program and ISI Program requirements are aligned with the provisions of 10 CFR 50.55a.
2.0 DESCRIPTION OF PROPOSED CHANGES  2.1 Deletion of SR 4.0.5 and Relocation of RCP Flywheel Inspection Requirements  SR 4.0.5.a through SR 4.0.5.f. describe the requirements for the Inservice Inspection and the Inservice Testing of ASME Code Class 1, 2, and 3 components and the Reactor Coolant Pump (RCP) flywheel inspection program. The proposed change deletes SR 4.0.5.a through SR 4.0.5.f. and relocates the requirements of SR 4.0.5.f to ISI Program procedures and to a new paragraph in TS 6.8.4, Procedures and Programs, titled Reactor Coolant Pump Flywheel Inspection Program. 2.2 Relocation of Acceptance Criteria in Select SRs Credited by the IST Program  SR 4.5.2.c, Figure 3.5-1, Residual Heat Removal (RHR) Pump Curve, and SR 4.6.2.1.b specify acceptance criteria related to design parameters for Safety Injection (SI), RHR and Containment Spray (CS) pump inservice testing. The proposed change relocates the acceptance criteria specified in SR 4.5.2.c, Figure 3.5-1 and SR 4.6.2.1.b for the SI, RHR and CS pumps, to IST Program procedures by deleting Figure 3.5-1 and replacing the acceptance criteria in SR 4.5.2.c and SR 4.6.2.1.b with statements that each pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. 2.3 Replacement of References to the SFCP in SRs Credited by the IST Program SR 4.4.4 establishes that in accordance with the SFCP, each block valve associated with the Pressurizer power-operated relief valves (PORVs) shall be demonstrated operable by operating the valve though one complete cycle of full travel except when required to be closed. The proposed change replaces the reference to the SFCP in SR 4.4.4 with reference to testing in accordance with the INSERVICE TESTING PROGRAM. SR 4.4.6.2.2.a establishes that each Reactor Coolant System (RCS) Pressure Isolation Valve (PIV) shall be demonstrated OPERABLE by verifying leakage to be within its limit in accordance with the SFCP. The proposed change replaces the reference to the SFCP in SR 4.4.6.2.2.a with reference to testing in accordance with the INSERVICE TESTING Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 3 of 41  PROGRAM. The proposed change truncates SR 4.4.6.2.2.b by ending the sentence with a period and deletes SR 4.4.6.2.2.c regarding returning PIVs to service following maintenance, repair or replacement. SR 4.4.11 establishes that each RCS vent path shall be demonstrated OPERABLE in accordance with the SFCP. SR 4.4.11.a establishes the requirements for manual RCS vent valve position verification. SR 4.4.11.b establishes the requirements for RCS vent valve full-travel cycling. SR 4.4.11.c establishes the requirements for RCS vent valve flow verification during RCS venting. The proposed change relocates the reference to SFCP applicability from SR 4.4.11 to SR 4.4.11.a and SR 4.4.11.c, and modifies SR 4.1.11.b to reference testing in accordance with the INSERVICE TESTING PROGRAM. SR 4.5.1.1.d establishes that each SI Accumulator Discharge check valve shall be demonstrated OPERABLE in accordance with the SFCP. The proposed change replaces the reference to the SFCP in SR 4.5.1.1.d with reference to testing in accordance with the INSERVICE TESTING PROGRAM. In addition, the proposed change deletes the footnote for SR 4.5.1.1.d denoted by an asterisk (*) regarding the maximum allowed surveillance test interval during Unit 3 Cycle 26. 3.0 TECHNICAL EVALUATION  3.1 Deletion of SR 4.0.5 and Relocation of RCP Flywheel Inspection Program  The proposed change deletes SR 4.0.5.a through SR 4.0.5.f and relocates the RCP Flywheel non-destructive examination requirements of SR 4.0.5.f to ISI Program procedures and a new paragraph within TS 6.8.4, Procedures and Programs, titled Reactor Coolant Pump Flywheel Inspection Program. SR 4.0.5.a specifies that the inservice inspection and inservice testing of ASME Code Class 1, 2, and 3 components shall be performed in accordance with the ASME Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda (ASME Section XI) and the ASME OM Code and applicable Addenda (ASME OM Code). The Turkey Point IST and ISI programs are developed and implemented in accordance with the ASME OM Code and the ASME Section XI Code requirements as required by 10 CFR 50.55a(f) and 10 CFR 50.55a(g), respectively. Hence, SR 4.0.5.a is duplicative of the 10 CFR 50.55a inservice inspection and inservice testing requirements and thereby its deletion is reasonable. SR 4.0.5.b defines surveillance interval terminology for the inservice inspection and testing activities. The Turkey Point ISI and IST programs are developed and implemented in accordance with the requirements of ASME Section XI and the ASME OM Code, which adequately define inservice inspection and inservice testing frequencies. Consistent with the model Safety Evaluation (SE) for Technical Specification Task Force (TSTF) Traveler 545, Revision 3 (Reference 6.1), deleting SR 4.0.5.b is reasonable since it is redundant with the surveillance interval terminology specified in 10 CFR 50.55a and its deletion from the TS resolves any inconsistencies between the two. SR 4.0.5.c establishes that the 25% surveillance frequency extension allowed by SR 4.0.2 for TS surveillances applies to inservice inspection and inservice testing activities. Consistent with the model SE for TSTF-545 (Reference 6.1), the deletion of SR 4.0.5.c is reasonable since the TS cannot supersede regulations and the inservice inspection and inservice testing surveillance frequencies established in ASME Section XI and the ASME OM Code cannot be modified except in accordance with 10 CFR 50.55a.
Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 4 of 41    SR 4.0.5.d establishes that the performance of inservice inspection and testing activities shall be in addition to other TS specified surveillance requirements. The deletion of SR 4.0.5.d is reasonable since SR 4.0.5.d is redundant with SR 4.0.1 which states that the SRs shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation (LCO) unless otherwise stated in an individual SR, and with 10 CFR 50.55a which requires inservice inspection and testing at the applicable ASME Section XI and ASME OM Code specified frequencies. SR 4.0.5.e states that nothing in the ASME Boiler and Pressure Vessel Code or the ASME OM Code shall be construed to supersede the requirements of any TS. Consistent with the model SE for TSTF-545 (Reference 6.1), the elimination of SR 4.0.5.e is reasonable since the specification conflicts with the provisions of 10 CFR 50.55a which requires licensees to amend the TS if the TS conflicts with 10 CFR 50.55a requirements. SR 4.0.5.f specifies the periodicity, scope and methodology for non-destructive examination (NDE) of the RCP flywheels. The proposed change deletes SR 4.0.5.f and relocates the RCP flywheel inspection requirements to ISI Program procedures where future changes are subject to the provisions of 10 CFR 50.59 and/or 10 CFR 50.55(a)(g), as appropriate. In addition, the proposed change establishes as TS 6.8.4.n, the Turkey Point Reactor Coolant Pump Flywheel Inspection Program. Relocation of the RCP flywheel inspection requirements to ISI Program procedures and the establishment of TS 6.8.4.n, Reactor Coolant Pump Flywheel Inspection Program, is consistent with Section 5.5 of NUREG-1431, Revision 4.0, Standard Technical Specifications - Westinghouse Plants, (NUREG-1431) (Reference 6.2) for plants that satisfy the requirements in WCAP-15666, Extension of Reactor Coolant Pump Motor Flywheel Examination (Reference 6.3). Hence, the deletion of SR 4.0.5.f, the relocation of the RCP flywheel inspection requirements to ISI Program procedures and the establishment of TS 6.8.4.n as the Reactor Coolant Pump Flywheel Inspection Program are reasonable since they are consistent with NUREG-1431, the RCP flywheel inspection requirements are unchanged, Moreover, any future changes to the RCP flywheel inspection requirements specified in ISI Program procedures will be subject to the provisions of 10 CFR 50.59 and/or 10 CFR 50.55(a)(g), as appropriate. Based upon the foregoing, the deletion of SR 4.0.5.a through SR 4.0.5.f and the relocation of the RCP flywheel inspection requirements to TS 6.8.4 are reasonable. 3.2 Replacement of Acceptance Criteria in Select SRs Credited by the IST Program  The proposed change relocates the acceptance criteria specified in SR 4.5.2.c, Figure 3.5-1 and SR 4.6.2.1.b for the SI, RHR and CS pumps, to IST Program procedures by deleting Figure 3.5-1 and replacing the acceptance criteria in SR 4.5.2.c and SR 4.6.2.1.b with statements that each pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. SR 4.5.2.c.1 specifies tested in the normal and in the cross-tie alignment (the SI pumps can be aligned to either Unit). Consistent with SR 3.5.2.4 of NUREG-1431 (Reference 6.2), the proposed change replaces the acceptance criteria specified in SR 4.5.2.c.1 with a statement that each SI pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. SI pump acceptance criteria account for available margin for the test to Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 5 of 41  assure that the design basis minimums are met. These acceptance criteria are incorporated in the IST Program and procedures. To ensure that pump performance does not degrade below that assumed in plant safety analyses, IST Program procedures specify the minimum allowable pump differential pressure () as the more restrictive of the required, i.e. design basis minimum, and the minimum allowed by ASME OM Code. The deletion of the TS the methodology in which SI pump test acceptance criteria are established nor the manner in which the test results are evaluated. Moreover, any future changes to SI pump inservice testing requirements will be subject to the provisions of 10 CFR 50.59 and/or 10 CFR 50.55(a)(f), as appropriate. Hence, replacing the SI pump acceptance criteria in SR 4.5.2.c.1 with a statement that each SI pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM is reasonable. SR 4.5.2.c.2 specifies that the RHR pumps' indicated applicable to the operating conditions fall within the Acceptable Range of the RHR pump curve specified in Figure 3.5-1. Consistent with SR 3.5.2.4 of NUREG-1431 (Reference 6.2), the proposed change deletes the RHR pump curve in Figure 3.5-1 and replaces the reference to the pump curve in SR 4.5.2.c.2 with a statement that each RHR pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTNG PROGRAM. RHR pump acceptance criteria account for available margin for the test to assure that the design basis minimums are met. These acceptance criteria are incorporated in the IST Program and procedures. To ensure that pump performance does not degrade below that assumed in plant safety analyses, IST Program wed by ASME OM Code. The deletion of Figure 3.5-1 from the TS neither changes the methodology in which RHR pump test acceptance criteria are established nor the manner in which the test results are evaluated. Moreover, any future changes to RHR pump inservice testing requirements will be subject to the provisions of 10 CFR 50.59 and/or 10 CFR 50.55(a)(f), as appropriate. Hence, removing Figure 3.5-1 and replacing its reference in SR 4.5.2.c.2 with a statement that each RHR pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM is reasonable. SR 4.6.2.1.b. specifies the CS pumps' minimum allowable  the specified conditions for pump recirculation flow. Consistent with SR 3.6.6A.4 of NUREG-1431 (Reference 6.2), the proposed change replaces the specified in SR 4.6.2.1.b. with a statement that each pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM. CS pump acceptance criteria account for available margin for the test to assure that the design basis minimums are met. These acceptance criteria are incorporated in the IST Program and procedures. Currently, CS pump flow must be established through the pump mini-recirculation piping or otherwise by substantially throttling flow through full-flow recirculation piping in order to satisfy the current test conditions specified in SR 4.6.2.1.b. However since installation of full-flow recirculation piping in 2012, quarterly CS pump testing is required under mini-recirculation conditions in order to satisfy SR 4.6.2.1.b and full-flow recirculation conditions to fulfill IST Program requirements. The proposed change alleviates duplicative quarterly CS pump testing by eliminating the mini-recirculation testing and continuing the full-flow recirculation testing in accordance with the INSERVICE TESTING PROGRAM, which provides for a more reliable assessment of pump performance and operational readiness. To ensure that CS pump performance does not degrade below that assumed in plant safety analyses, IST Program procedures specify the minimum Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 6 of 41    ASME OM Code. Moreover, any future changes to CS pump inservice testing requirements will be subject to the provisions of 10 CFR 50.59 and/or 10 CFR 50.55(a)(f), as appropriate. Hence, replacing the CS pump mini-recirculation test requirement and minimum allowable  in SR 4.6.2.1.b with a statement that each CS pumps' developed head at the test flow point shall be greater than or equal to the required developed head when tested in accordance with the INSERVICE TESTING PROGRAM is reasonable.      3.3 Replacement of References to the SFCP in SRs Credited by the IST Program The proposed change replaces the reference to the SFCP in SR 4.4.4 for each block valves associated with a PORV, with reference to testing in accordance with the INSERVICE TESTNG PROGRAM. The proposed change also relocates the reference to SFCP applicability for the RCS vent valves from SR 4.4.11 to SR 4.4.11.a and SR 4.4.11.c, and modifies SR 4.1.11.b to explicitly reference testing in accordance with the INSERVICE TESTING PROGRAM. The proposed change also replaces the reference to the SFCP in SR(s) 4.4.6.2.2.a and 4.5.1.1.d for the Reactor Coolant System PIV and SI Accumulator Discharge Check Valves, respectively, with reference to testing in accordance with the INSERVICE TESTING PROGRAM. SR 4.4.4, SR 4.4.11.b, SR 4.4.6.2.2.a, and SR 4.5.1.1.d are credited for satisfying IST Program requirements. Replacing the statement of SFCP applicability with reference to testing in accordance with the INSERVICE TESTING PRORAM is reasonable since the SFCP cannot modify the inservice testing frequencies established in 10 CFR 50.55a. The proposed change also truncates SR 4.4.6.2.2.b by ending the sentence with a period and deletes SR 4.4.6.2.2.c regarding returning PIV(s) to service following maintenance, repair or replacement. Truncating SR 4.4.6.2.2.b with a period and deleting SR 4.4.6.2.2.c is reasonable since SR 4.4.6.2.2.c is redundant with  the requirements of 10 CFR 50.55, Appendix B, Section XIV, Inspection, Test, and Operating Status, Section XV, Nonconforming Materials, Parts, or Components, and Section XVI, Corrective Action. In addition, the proposed change deletes the footnote denoted by an asterisk (*) for SR 4.5.1.1 regarding the Unit 3 Cycle 26 maximum allowed surveillance test interval. Deleting the footnote is reasonable since Unit 3 Cycle 26 was completed and the maximum allowed surveillance interval established by the footnote no longer applies.
4.0 REGULATORY SAFETY ANALYSIS  4.1 Applicable Regulatory Requirements/Criteria  10 CFR 50.36(c) states that the Technical Specifications will include items in the following categories:  3) Surveillance requirements; surveillance requirements are requirements        relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. 10 CFR 50.55(a)(f) states that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this [10 CFR 50.55(a)(f)] paragraph.
Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 7 of 41    10 CFR 50.55(a)(g) states that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this [10 CFR 50.55(a)(g)] paragraph. 10 CFR 50.59(c)(1) states that a licensee may make changes in the facility as described in the final safety analysis report (as updated), make changes in the procedures as described in the final safety analysis report (as updated), and conduct tests or experiments not described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to Sec. 50.90 only if:  (i) A change to the technical specifications incorporated in the license is not required, and  (ii) The change, test, or experiment does not meet any of the criteria in paragraph (c)(2) of this [10 CFR 50.59(c)] section. The proposed changes comply with the requirements of 10 CFR 50.36(c), 10 CFR 50.55(a)(f), 10 CFR 50.55(a)(g) and 10 CFR 50.59 and do not alter the manner in which inservice inspection and testing is performed consistent with Turkey Point TS and ASME Code requirements. All applicable regulatory requirements will continue to be satisfied as a result of the proposed changes. 4.2 No Significant Hazards Consideration  The proposed license amendments revise the Turkey Point Technical Specifications (TS) by relocating to licensee controlled documents select acceptance criteria specified in TS surveillance requirements (SRs) credited for satisfying Inservice Testing (IST) Program and Inservice Inspection (ISI) requirements, deleting the SRs for American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components, replacing references to the Surveillance Frequency Control Program (SFCP) with reference to the Turkey Point IST Program where appropriate, and establishing a Reactor Coolant Pump (RCP) Flywheel Inspection Program. The proposed license amendments also delete a duplicative SR regarding returning Pressure Isolation Valves (PIV) to service following maintenance and a footnote previously applicable to Unit 3 Cycle 26. As required by 10 CFR 50.91(a), Florida Power & Light Company (FPL) has evaluated the proposed changes using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration. An analysis of the issue of no significant hazards consideration is presented below:  (1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?  Response: No  The proposed changes provide added assurance that inservice testing will be performed in the manner and within the timeframes established by 10 CFR 50.55(a). The deletion of SR 4.0.5 and the deletion of IST acceptance criteria from SR 4.5.2.c and SR 4.6.2.1.b neither affects the conduct nor the periodicity of testing which demonstrates the operational readiness of safety-related pumps and valves. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative in nature and can neither initiate nor exacerbate any accident previously evaluated. Similarly, the deletion of SR 4.0.5 Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 8 of 41  and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot affect the likelihood or outcome of any accident previously evaluated. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and related editorial changes are administrative changes in nature and do not alter any plant equipment or the results of any accident analyses. Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.  (2) Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?  Response: No  The deletion of IST acceptance criteria from the TS does not affect the manner in which any SSC is maintained or operated and does not introduce new SSCs or new methods for maintaining existing plant SSCs. Inservice testing will continue in the manner and periodicity specified in the IST program and hence no new or different kind of accident can result. The addition of references to the IST Program in SR(s) where applicable and the deletion of references to the SFCP in SR testing credited by the IST Program are administrative changes and cannot affect the manner in which any SSC is maintained or operated. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes and cannot be an initiator of a new or different kind of accident. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and other editorial changes are administrative changes in nature and do not introduce any new plant equipment, failure modes or accident analyses postulated outcomes.
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 amendment involve a significant reduction in a margin of safety?  Response: No  The proposed changes do not involve changes to any safety analyses assumptions, safety limits, or limiting safety system settings nor do they adversely impact plant operating margins or the reliability of equipment credited in the safety analyses. The reliability of credited equipment is enhanced through added assurance that inservice inspection and inservice testing will be performed in the manner and within the timeframes established by the ASME Code requirements of 10 CFR 50.55(a)(g) and 10 CFR 50.55(a)(f), respectively. The deletion of SR 4.0.5 and the relocation of the RCP flywheel inspection requirements within the TS are administrative changes with no impact on the margin of safety currently described the Updated Final Safety Analysis Report. Deletion of the SR 4.4.6.2.2.c requirement regarding returning PIV(s) to service following maintenance, repair or replacement, deletion of a SR 4.5.1.1.d footnote previously applicable during Unit 3 Cycle 26, and other editorial changes are administrative changes in nature with no impact on nuclear safety.
Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                          Page 9 of 41    Therefore, the proposed change does not involve a significant reduction in a margin of safety. Based upon the above analysis, FPL concludes that the proposed amendment does not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92, "Issuance of Amendment," and accordingly, a finding of "no significant hazards consideration" is justified. 4.3 Conclusion  In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. 5.0 ENVIRONMENTAL CONSIDERATION  The proposed amendment modifies a regulatory requirement with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or changes an inspection or surveillance requirement. However, 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 effluents 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. 6.0 REFERENCES 6.1 FR Notice of Availability RE: Model SE for Traveler TSTF 545, Rev. 3, "TS Inservice Testing Program Deletion & Clarify SR Usage Rule Application to Section 5.5 Testing" using the CLIIP, dated March 18, 2016 (ML15181A045)6.2 NUREG-1431, Standard Technical Specifications - Westinghouse Plants, Revision 4.0, Volume 1, Specifications (Accession No. ML12100A222)  6.3 Safety Evaluation of Topical Report WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination" dated May 5, 2003 (Accession No. ML031250595)
Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                        Page 10 of 41 ATTACHMENT 1 PROPOSED TECHNICAL SPECIFICATION PAGES (MARKUP)  (11 pages follow)
 
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TURKEY POINT - UNITS 3 & 4 3/4 5-2 AMENDMENT NOS. 263 AND 258 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.In accordance with the Surveillance Frequency Control Program and within 6 hours after eachsolution volume increase of greater than or equal to 1% of tank volume by verifying the boronconcentration of the solution in the water-filled accumulator is between 2300 and 2600 ppm;c.In accordance with the Surveillance Frequency Control Program, when the RCS pressure isabove 1000 psig, by verifying that the power to the isolation valve operator is disconnected by alocked open breaker.d.In accordance with the Surveillance Frequency Control Program*, each accumulator check valveshall be checked for operability. *During Unit 3 Cycle 26 only, in lieu of the Technical Specification specified 18 month refueling frequency and4.5 month grace period allowance, the maximum allowed surveillance test interval will be extended to no more than 24.5 months. ByverifyingtheOPERABILITYofwhentestedinaccordancewiththeInserviceTestingProgramdischarge
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TURKEY POINT - UNITS 3 & 4 3/4 5-6 AMENDMENT NOS. 138 AND 133 Figure 3.5-1 RHR Pump Curve TOTAL DEVELOPED HEAD (FT)
TURKEY POINT - UNITS 3 & 4 3/4 6-12 AMENDMENT NOS. 24 AND 29 CONTAINMENT SYSTEMS 3/4.6.2  DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1  Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a.With one Containment Spray System inoperable restore the inoperable Spray System toOPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within the following 30 hours.b.With two Containment Spray Systems inoperable restore at least one Spray System toOPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within the following 30 hours. Restore both Spray Systems to OPERABLE status within 72 hours of initial loss or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within thefollowing 30 hours.SURVEILLANCE REQUIREMENTS 4.6.2.1  Each Containment Spray System shall be demonstrated OPERABLE: a.In accordance with the Surveillance Frequency Control Program by verifying that each valve(manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwisesecured in position, is in its correct position* and that power is available to flow path componentsthat require power for operation;b.By verifying that on recirculation flow, each pump develops the indicated differential pressure,when testedContainment Spray Pump 241.6 psid while aligned in recirculation mode.c.In accordance with the Surveillance Frequency Control Program by verifying containment spraylocations susceptible to gas accumulation are sufficiently filled with water.*Not required to be met for system vent flow paths opened under administrative control.
TURKEY POINT - UNITS 3 & 4 6-14 AMENDMENT NOS. 272 AND 267  ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) I. Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met: a.The Surveillance Frequency Control Program shall contain a list of frequencies of thoseSurveillance Requirements for which the frequency is controlled by the program.b.Changes to the frequencies listed in the Surveillance Frequency Control Program shallbe made in accordance with NEI 04-10, " Risk-Informed Method for Control ofSurveillance Frequencies," Revision 1.c.The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to thefrequencies established in the Surveillance Frequency Control Program.m.Snubber Testing ProgramThis program conforms to the examination, testing and service life monitoring for dynamicrestraints (snubbers) in accordance with 10 CFR 50.55a inservice inspection (lSI) requirementsfor supports. The program shall be in accordance with the following:a.This program shall meet 10 CFR 50.55a(g) lSI requirements for supports.b.The program shall meet the requirements for lSI of supports set forth in subsequenteditions of the Code of Record and addenda of the American Society of MechanicalEngineers (ASME) Boiler and Pressure Vessel (BPV) Code and the ASME Code forOperation and Maintenance of Nuclear Power Plants (OM Code) that are incorporated byreference in 10 CFR 50.55a(a) subject to the use and conditions on the use of standardslisted in 10 CFR 50.55a(b) and subject to Commission approval.c.The program shall, as required by 10 CFR 50.55a(b)(3)(v), meet Subsection ISTA,"General Requirements" and Subsection ISTD, "Preservice and lnservice Examinationand Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear PowerPlants".d.The 120-month program updates shall be made in accordance with 10 CFR 50.55a(g)(4),10 CFR 50.55a(g)(3)(v) and 10 CFR 50.55a(b) (including 10 CFR 50.55a(b)(3)(v))subject to the conditions listed therein.6.8.5  DELETED Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                        Page 22 of 41 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION PAGES (CLEAN COPY)  (11 pages follow)
INDEXTURKEY POINT - UNITS 3 & 4 viii AMENDMENT NOS. 260 AND 255 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTIONPAGE3/4.4.9  PRESSURE/TEMPERATURE LIMITS Reactor Coolant System................................................................................. 3/4 4-23 FIGURE 3.4-2 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS (Heatup Rates of 60 and 100°F/hr) APPLICABLE FOR 48 EFPY (Without Margins for Instrumentation Errors)................................................. 3/4 4-24 FIGURE 3.4-3 TURKEY POINT UNITS 3&4 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS  (Cooldown Rates of 0, 20, 40, 60 and100F/hr) APPLICABLE  FOR 48 EFPY(Without Margins for Instrumentation Errors).......................... 3/4 4-25 Pressurizer ................................................................................................... 3/4 4-26 Overpressure Mitigating Systems................................................................... 3/4 4-27 3/4.4.10 DELETED 3/4.4.11 REACTOR COOLANT SYSTEM VENTS....................................................... 3/4 4-29 3/4.5  EMERGENCY CORE COOLING SYSTEMS3/4.5.1  ACCUMULATORS......................................................................................... 3/4 5-1 3/4/5.2 ECCS SUBSYSTEMS - Tavg GREATER THAN OR EQUAL TO 350F.......... 3/4 5-3 FIGURE 3.5-1 ....................................................................................... 3/4 5-6 3/4.5.3 ECCS SUBSYSTEMS - Tavg LESS THAN 350F........................................... 3/4 5-9 3/4.5.4 REFUELING WATER STORAGE TANK........................................................ 3/4 5-10 TURKEY POINT - UNITS 3 & 4 3/4 0-3 AMENDMENT NOS. AND 2 APPLICABILITYSURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance  Requirement. Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the OPERABILITY requirements for a Limiting Condition for Operation. Surveillance Requirements do not have to be performed on inoperable equipment. 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval. If an ACTION item requires periodic performance on a "once per . . ." basis, the above frequency extension applies to each performance after the initial performance. Exceptions to this Specification are stated in the individual Specifications. 4.0.3 If it is discovered that a Surveillance was not performed within its specified frequency, then compliance with the requirement to declare the Limiting Condition of Operation not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed. If the surveillance is not performed within the delay period, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the Limiting Condition of Operation must immediately be declared not met, and the applicable ACTION(s) must be entered. 4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be made unless the Surveillance Requirement(s) associated with a Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to OPERATIONAL MODES as required to comply with ACTION requirements. 4.0.5 a.
TURKEY POINT - UNITS 3 & 4 3/4 0-4 AMENDMENT NOS. 2 AND 2 APPLICABILITYSURVEILLANCE REQUIREMENTS (CONTINUED) b.c.d.e.f.4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifications or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes or body of the requirement.
TURKEY POINT - UNITS 3 & 4 3/4 4-11 AMENDMENT NOS. 263 AND 258  REACTOR COOLANT SYSTEM RELIEF VALVES SURVEILLANCE REQUIREMENTS 4.4.4  In accordance with the  each block valve shall be demonstrated OPERABLE by operating the valve through one complete cycle of full travel unless the block valve is closed with power removed in order to meet the requirements of Specification 3.4.4 or is closed to provide an isolation function.
TURKEY POINT - UNITS 3 & 4 3/4 4-15 AMENDMENT NOS. 263 AND 258  REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION (Continued) 2.The leakage* from the remaining isolating valves in each high pressure line having avalve not meeting the criteria of Table 3.4-1, as listed in Table 3.4-1, shall be determinedand recorded daily. The positions of the other valves located in the high pressure linehaving the leaking valve shall be recorded daily unless they are manual valves locatedinside containment.Otherwise be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. d.With any Reactor Coolant System Pressure Isolation Valve leakage greater than 5 gpm, reduceleakage to below 5 gpm within 1 hour, or be in at least HOT STANDBY within the next 6 hoursand in COLD SHUTDOWN within the following 30 hours.SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System operational leakages shall be demonstrated to be within each of the  above limits by: a.Monitoring the containment atmosphere gaseous or particulate radioactivity monitor inaccordance with the Surveillance Frequency Control Program.b.Monitoring the containment sump level in accordance with the Surveillance Frequency ControlProgram.c.** Performance of a Reactor Coolant System water inventory balance in accordance with theSurveillance Frequency Control Program***; andd.Monitoring the Reactor Head Flange Leakoff System in accordance with the SurveillanceFrequency Control Program; ande.Verifying primary-to-secondary leakage is < 150 gallons per day through any one SG inaccordance with the Surveillance Frequency Control Program***.4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-1 shall bedemonstrated OPERABLE by verifying leakage* to be within its limit:a.n accordance with the b.Prior to entering MODE 2 whenever the plant has been in COLD SHUTDOWN for 7 days or moreand if leakage testing has not been performed in the previous 9 monthsc.*To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressureindicators) if accomplished in accordance with approved procedures and supported by computations showingthat the method is capable of demonstrating valve compliance with the leakage criteria.** Not applicable to primary-to-secondary leakage.*** Not required to be performed until 12 hours after establishment of steady state operation.
 
TURKEY POINT - UNITS 3 & 4 3/4 5-2 AMENDMENT NOS. 263 AND 258 EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.In accordance with the Surveillance Frequency Control Program and within 6 hours after eachsolution volume increase of greater than or equal to 1% of tank volume by verifying the boronconcentration of the solution in the water-filled accumulator is between 2300 and 2600 ppm;c.In accordance with the Surveillance Frequency Control Program, when the RCS pressure isabove 1000 psig, by verifying that the power to the isolation valve operator is disconnected by a locked open breaker.d.each accumulator check valve TURKEY POINT - UNITS 3 & 4 3/4 5-5 AMENDMENT NOS. 2 AND 2  EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.2 Each ECCS component and flow path shall be demonstrated OPERABLE: a.In accordance with the Surveillance Frequency Control Program by verifying by control roomindication that the following valves are in the indicated positions with power to the valve operatorsremoved:Valve Number Valve FunctionValve Position864A and B Supply from RWST to ECCS Open 862A and B RWST Supply to RHR pumps  Open 863A and B RHR Recirculation Closed 866A and B H.H.S.I. to Hot Legs ClosedHCV-758*RHR HX Outlet Open To permit positive valve position indication for surveillance or maintenance purposes in the event that continuous valve position indication is unavailable in the control room, power may be restored to these valves for a period not to exceed 1 hour. b.In accordance with the Surveillance Frequency Control Program by:1)Verifying ECCS locations susceptible to gas accumulation are sufficiently filled with water, and2)Verifying that each valve (manual, power-operated, or automatic) in the flow path that is notlocked, sealed, or otherwise secured in position, is in its correct position.**c.By verifying that each SI and RHR  when tested *Air Supply to HCV-758 shall be verified shut off and sealed closed in accordance with the SurveillanceFrequency Control Program. **Not required to be met for system vent flow paths opened under administrative control.
TURKEY POINT - UNITS 3 & 4 3/4 5-6 AMENDMENT NOS. 138 AND 133 Figure 3.5-1  TOTAL DEVELOPED HEAD (FT)
TURKEY POINT - UNITS 3 & 4 3/4 6-12 AMENDMENT NOS. 24 AND 29 CONTAINMENT SYSTEMS 3/4.6.2  DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1  Two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System. APPLICABILITY: MODES 1, 2, 3, and 4. ACTION: a.With one Containment Spray System inoperable restore the inoperable Spray System toOPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within the following 30 hours.b.With two Containment Spray Systems inoperable restore at least one Spray System toOPERABLE status within 1 hour or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within the following 30 hours. Restore both Spray Systems to OPERABLE status within 72 hours of initial loss or be in at least HOT STANDBY within the next 6 hours and inCOLD SHUTDOWN within thefollowing 30 hours.SURVEILLANCE REQUIREMENTS 4.6.2.1  Each Containment Spray System shall be demonstrated OPERABLE: a.In accordance with the Surveillance Frequency Control Program by verifying that each valve(manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwisesecured in position, is in its correct position* and that power is available to flow path componentsthat require power for operation;b.c.In accordance with the Surveillance Frequency Control Program by verifying containment spraylocations susceptible to gas accumulation are sufficiently filled with water.*Not required to be met for system vent flow paths opened under administrative control..
TURKEY POINT - UNITS 3 & 4 6-14 AMENDMENT NOS. 272 AND 267  ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS (Continued) I. Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met: a.The Surveillance Frequency Control Program shall contain a list of frequencies of thoseSurveillance Requirements for which the frequency is controlled by the program.b.Changes to the frequencies listed in the Surveillance Frequency Control Program shallbe made in accordance with NEI 04-10, " Risk-Informed Method for Control ofSurveillance Frequencies," Revision 1.c.The provisions of Surveillance Requirements 4.0.2 and 4.0.3 are applicable to thefrequencies established in the Surveillance Frequency Control Program.m.Snubber Testing ProgramThis program conforms to the examination, testing and service life monitoring for dynamicrestraints (snubbers) in accordance with 10 CFR 50.55a inservice inspection (lSI) requirementsfor supports. The program shall be in accordance with the following:a.This program shall meet 10 CFR 50.55a(g) lSI requirements for supports.b.The program shall meet the requirements for lSI of supports set forth in subsequenteditions of the Code of Record and addenda of the American Society of MechanicalEngineers (ASME) Boiler and Pressure Vessel (BPV) Code and the ASME Code forOperation and Maintenance of Nuclear Power Plants (OM Code) that are incorporated byreference in 10 CFR 50.55a(a) subject to the use and conditions on the use of standardslisted in 10 CFR 50.55a(b) and subject to Commission approval.c.The program shall, as required by 10 CFR 50.55a(b)(3)(v), meet Subsection ISTA,"General Requirements" and Subsection ISTD, "Preservice and lnservice Examinationand Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear PowerPlants".d.The 120-month program updates shall be made in accordance with 10 CFR 50.55a(g)(4),10 CFR 50.55a(g)(3)(v) and 10 CFR 50.55a(b) (including 10 CFR 50.55a(b)(3)(v))subject to the conditions listed therein.6.8.5  DELETED Turkey Point Nuclear Plant                                                                                                          L-2017-100 Docket Nos. 50-250 and 50-251                          Enclosure                                        Page 34 of 41 
 
ATTACHMENT 3  PROPOSED TECHNICAL SPECIFICATION BASES PAGES (MARKUP)  (7 pages follow)
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 39 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 22 of 194)    3/4.0 (Continued) Specification 4.0.1 through 4.0.5 establishes the general requirements applicable to Surveillance Requirements. These requirements are based on the Surveillance Requirements stated in the Code of Federal Regulations, 10 CFR 50.36(c)(3): Surveillance requirements are requirements relating to test, calibration, or inspection to ensure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions of operation will be met. Specification 4.0.1 establishes the requirement that surveillances must be performed during the OPERATIONAL MODES or other conditions for which the requirements of the Limiting Conditions for Operation apply unless otherwise stated in an individual Surveillance Requirement. The purpose of this specification is to ensure that surveillances are performed to verify the operational status of systems and components and that parameters are within specified limits to ensure safe operation of the facility when the plant is in a MODE or other specified condition for which the associated Limiting Conditions for Operation are applicable. Surveillance Requirements do NOT have to be performed when the facility is in an OPERATIONAL MODE for which the requirements of the associated Limiting Condition for operation do NOT apply unless otherwise specified. The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception to the requirements of a specification.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 45 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 28 of 194)    3/4.0 (Continued) Under the provisions of this specification, the applicable Surveillance Requirements must be performed within the specified surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage. When a shutdown is required to comply with ACTION requirements, the provisions of Specification 4.0.4 do NOT apply because this would delay placing the facility in a lower MODE of operation. Specification 4.0.5 establishes the requirement that inservice inspection of ASME Code Class 1, 2, and 3 components shall be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50.55a. Inservice testing of ASME Code Class 1, 2 and 3 pumps and valves shall be performed in accordance with the ASME Code of Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a. This Specification includes a clarification of the frequencies for performing the inservice inspection and testing activities required by Section XI of the ASME Boiler and Pressure Vessel Code or the ASME OM Code and applicable Addenda. This clarification is provided to ensure consistency in surveillance intervals throughout the Technical Specifications and to remove any ambiguities relative to the frequencies for performing the required inservice inspection and testing activities.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 46 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 29 of 194)    3/4.0 (Continued) Under the terms of this Specification, the more restrictive requirements of the Technical Specifications take precedence over the ASME Boiler and Pressure Vessel Code and applicable Addenda. The requirements of Specification 4.0.4 to perform surveillance activities before entry into an OPERATIONAL MODE or other specified condition takes precedence over the ASME Boiler and Pressure.
Vessel Code provision which allows pumps and valves to be tested up to one week after return to normal operation. The Technical Specification definition of OPERABLE does NOT allow a grace period before a component that is NOT capable of performing its specified function, is declared inoperable and takes precedence over the ASME Boiler and Pressure Vessel Code provision which allows a valve to be incapable of performing its specified function for up to 24 hours before being declared inoperable. The Westinghouse Owners Group submitted Topical Report (TR) WCAP-15666, "Extension of Reactor Coolant Pump Motor Flywheel Examination," dated July 2001 for NRC staff review. The NRC approved the TR by a Safety Evaluation dated May 5, 2003 revising the reactor coolant pump flywheel inspection interval to a maximum of 20 years. NO extension of this interval is allowed under Specification 4.0.2. Specification 4.0.6 delineates the applicability of the surveillance activities to Unit 3 operations.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 88 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 71 of 194)    3/4.4.4 (Continued) The OPERABILITY of the PORVs and block valves is determined on the basis of their being capable of performing the following functions: a. Manual control of PORVs to Control Reactor Coolant System pressure. This is a function that is used as a back-up for the steam generator tube rupture and to support plant shutdown in the event of an Appendix R fire. These functions are considered to be Important-To-Safety, or Quality Related per the FPL Quality Assurance program. b. Maintaining the integrity of the reactor coolant pressure boundary. This is a function that is related to controlling identified leakage and ensuring the ability to detect unidentified reactor coolant pressure boundary leakage. c. Manual control of the block valve to: (1) Unblock an isolated PORV to allow it to be used for manual control of reactor coolant system pressure, and (2) Isolate a PORV with excessive leakage. d. Manual control of a block valve to isolate a stuck-open PORV. e. Ability to open or close the valves, consistent with the required function of the valves. The PORVs are also used to provide automatic pressure control in order to reduce the challenges to the RCS code safety valves for overpressurization events.  (The PORVs are NOT credited in the overpressure accident analyses as noted above.) Surveillance Requirements provide the assurance that the PORVs and block valves can perform their functions. Specification 4.0.5. is applicable to PORVs and block valves. Specification 4.4.4. also addresses block valves. The block valves are exempt from the surveillance requirements to cycle the valves when they have been closed to comply with the ACTION requirements. This precludes the need to cycle the valves with full system differential pressure, or when maintenance is being performed to restore an inoperable PORV to OPERABLE status.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 133 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 116 of 194)    3/4.5.2 & 3/4.5.3 (Continued) Surveillance 4.5.2.b.2) is modified by a Note which exempts system vent flow paths opened under administrative control. The administrative control should be proceduralized and include stationing a dedicated individual at the system vent flow path who is in continuous communication with the operators in the Control Room. This individual will have a method to rapidly close the system vent path if directed. In the RHR test, differential head is specified in feet. This criteria will allow for compensation of test data with water density due to varying temperature. ECCS pump testing for the SI and RHR pumps accounts for possible underfrequency conditions, i.e., the results of pump testing performed at 60 Hz is then adjusted to reflect possible degraded grid conditions (60+/-0.6) to the lower limit (59.4 Hz). Technical Specifications Surveillance Requirement 4.5.2.e.3 requires that each ECCS component and flow path be demonstrated OPERABLE in accordance with the Surveillance Frequency Control Program by visual inspection which verifies sump components (trash racks, screens, etc.) show NO evidence of structural distress or abnormal corrosion. The strainer modules are rigid enough to provide both functions as trash racks and screens without losing their structural integrity and particle efficiency. Therefore, strainer modules are functionally equivalent to trash racks and screens. Accordingly, the categorical description, sump components, is broad enough to require inspection of the strainer modules.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 137 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT    ATTACHMENT 2 Technical Specification Bases (Page 120 of 194)    3/4.6.1.2 Containment Leakage The limitations on containment leakage rates ensure that the total containment leakage volume will NOT exceed the value assumed in the safety analyses at the peak accident pressure, Pa. The measured As Found overall integrated leakage rate is limited to less than or equal to 1.0 La during the performance of the periodic test. As an added conservatism, the measured overall As Left integrated leakage rate is further limited to less than or equal to 0.75 La to account for possible degradation of the containment leakage barriers between leakage tests. The surveillance testing for measuring leakage rates is in compliance with the requirements of Appendix J of 10 CFR Part 50, Option B [as modified by approved exemptions], and consistent with the guidance of Regulatory Guide 1.163, dated September 1995.
REVISION NO.: PROCEDURE TITLE: PAGE: 22 TECHNICAL SPECIFICATION BASES CONTROL PROGRAM 157 of 211 PROCEDURE NO.: 0-ADM-536 TURKEY POINT PLANT ATTACHMENT 2 Technical Specification Bases (Page 140 of 194) 3/4.7.1.2 (Continued) The specified flow rate acceptance criteria conservatively bounds the limiting AFW flow rate modeled in the single unit Loss of Normal Feedwater analysis. Dual unit events such as a two unit Loss of Offsite Power require a higher pump flow rate, but it is NOT practical to test both units simultaneously. The flow surveillance test specified in 4.7.1.2.1a.1 is considered to be a general performance test for the AFW system and does NOT represent the limiting flow requirement for AFW. Check valves in the AFW system that require full stroke testing under limiting flow conditions are tested under Technical Specification 4.0.5. The testing of the Auxiliary Feedwater Pumps will verify their OPERABILITY. Proper functioning of the turbine admission valve and the operation of the pumps will demonstrate the integrity of the system. Verification of correct operation will be made both from instrumentation within the Control Room and direct visual observation of the pumps. The frequencies of surveillance requirements 4.7.1.2.1a and 4.7.1.2.1b are controlled under the Surveillance Frequency Control Program. 
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