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#REDIRECT [[RC-15-0153, Submittal of 10 CFR 50.59 Biennial Report]]
| number = ML15286A055
| issue date = 10/07/2015
| title = V.C. Summer, Unit 1 - Submittal of 10 CFR 50.59 Biennial Report
| author name = Gatlin T D
| author affiliation = South Carolina Electric & Gas Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000395
| license number = NPF-012
| contact person =
| case reference number = 818.02-8, LTD-324, RC-15-0153, RR 8450
| document type = Letter, Report, Miscellaneous
| page count = 8
}}
 
=Text=
{{#Wiki_filter:Thomas 0. Gatlin Vice President, Nuclear Operations 803.345.4342 A SCANA COMPANY October 7, 2015 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
 
==Dear Sir!/ Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 10 CFR 50.59 BIENNIAL REPORT South Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Fourth.VCSNS Report pursuant to 10 CFR 50.59(d)(2).
This report contains a brief description and summary of the evaluations performed to support the changes and modifications made to the facility in accordance with 10 CER 50.59(c)(Attachment).
This report covers the time frame from October 1, 2013 to September 30, 2015.If you have any questions or require additional information, please contact Bruce Thompson at (803) 931-5042.Very truly yours, Thomas D. Gatlin WCM/TDG/ts Attachment
-10OCFR50.59 Summary of Evaluations and Changes c: K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton J. W. Williams W. M. Cherry L. D. Wert S. A. Williams NRC Resident Inspector K. M. Sutton NSRC RTS (LTD-324)File (81 8.02-8, RR 8450)PRSF (RC-15-0153)-Iii¢z V. C. Summer Nuclear Station. P. O. Box 88
* Jenkinsville, SC. 29065. F (803) 941-9776 Document Control Desk Attachment I LTD 324, RR-8450 RC-1 5-0153 Page 1 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document Change Description Evaluation Summary ECR-71781, RMA-2 Guidance ECR-71 781 supports removal of Technical Specification Relocation (TSR) TSR-I1069 operability criteria for RMA-2 based on definition of TSTF-513. Operability guidance for RMA-2 has been developed under design calculation DC00030-058, Revision 1.The calculation serves to: 1. More formally document the FSAR evaluations for the buildup of reactor building activity due to a one gpm leak using the licensing/design basis methods and RCS source terms.2. Determine limits on the alarm setpoints that ensure that the TSTF definition of operability is met when using licensing/design basis assumptions.
ECR-71 781 implements the operability limits within plant procedures to facilitate removal of TSR-i1069, updates the FSAR as appropriate, and enters DC00030-058 into records.The 50.59 Applicability Determination concluded that the proposed changes for ECR-71781 require a 10CFR50.59 review. All 50.59 Screen questions were answered NO except Question 111.4 relating to revising a FSAR described evaluation methodology.
The full I 0CRE50. 59 evaluation, however, concluded that the revisions do not represent a departure from a method of evaluation described in the ESAR since their use is conservative, thus leading to the overall conclusion that the proposed changes for ECR-71781 can be implemented without prior NRC approval.
Document Control Desk Attachment I LTD 324, RR-8450 RC-1 5-0153 Page 2 of 7 10 CFR 50.59 Summary of Chanties and Evaluations m I Parent Document Change Description Evaluation Summary CR-1 1-02428, Cladding Cladding stress/strain The 50.59 Applicability Determination Stress/Strain Methodology, methodology changes concluded that the proposed changes to WCAP 10125-P-A, (existing NRC approval) are the FSAR/FPER would require a 50.59 Addendum 1-A, Revision 1- recommended to regain Screen. All 50.59 Screen questions A. Modify portions of ESAR margin lost due to the were answered NO except Question 111.4 section 4.2.1 to incorporate activities performed to regarding revising or replacing an FSAR changes to cladding stress! address a code error described evaluation methodology.
All strain methodology discovered during the Cycle answers to the 50.59 evaluation were 20 reload design. Margin was NO as the FSAR/FPER described available for Cycles 20 and methodology change does not require 21, but it is desired to prior NRC approval to implement eliminate possible limit because it is within the limitations violations for this parameter.
described in WCAP 10125-P-A and the The vendor code and SER. Changing from one method analysis will be performed for described in the FSAR/FPER to another the Cycle 22 and future method is not a departure if that method designs using methods that has been approved by the NRC for the are currently approved by the intended application.
NRC, but not consistent with our existing FSAR.CR (NC) 01131, Address non-conforming Technical evaluations that include spring Niobium-rich Inclusions in condition observed in the mechanical performance, fuel assembly Top Nozzle Hold-down Inconel 718 material used to holddown permanent set from fuel Springs manufacture the fuel assembly liftoff, and reactor internals assembly top nozzle core barrel flange holddown, have holddown springs. The region shown that an affected spring continues of fuel that was recently to meet all holddown requirements.
delivered to VCS is in the Conservatively assuming that corrosion population of holddown would cause an upper spring failure, the springs made from the potential for loose parts was evaluated.
affected ingot material.
The It was shown that a single fractured leaf abnormality relates to surface will not be released from a nozzle and its indications (discoloration) motion would be limited such that it will noted in a small number of not interfere with RCCA motion or top nozzle upper springs, handling tool engagement.
Additionally, the potential loss of fuel assembly holddown force is not a concern during operation.
The fuel assembly uplift motion is small enough such that the fuel assemblies remain engaged on the upper and lower core plate alignment pins. Thus, the basic structural characteristics of the fuel assemblies Document Control Desk Attachment!I LTD 324, RR-8450 RC-1 5-0153 Page 3 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document Change Description Evaluation Summary remain the same. Insertion of the control rods will not be affected and horizontal Seismic and/or LOCA loads will continue to be reacted by the core plate alignment pins. Additionally, the fuel assembly axial movement is small enough such that the required structural grid overlap is maintained.
Finally, there is no anticipated post-discharge fuel handling, or wet and dry storage issues.The 50.59 Screen Question 111.2 (change to an SSC that adversely affects an FSAR/FPER described design function)screened in YES to require the 50.59 evaluation.
All Evaluation questions were answered NO. Results from technical evaluation of condition of the holddown springs determined them capable of meeting their intended design function and not require a license amendment.
ECR-50846D, Weld Repair Contingency for Reactor Vessel Inspection (RF-21)Perform an overlay weld repair as detailed in WCAP-15987-P Revision 2-P-A"Technical Basis for Embedded Flaw for Repair of Reactor Vessel Head Nozzles." VC Summer performed RV Head Inspections during Refuel Outage 21. During the examination, Primary Water Stress Corrosion Cracking (PWS CC) indications were found in the CRDM penetrations.
These flaws must be repaired prior to entering Mode 5.Since the WCAP process to be used is a deviation from the usual ASME Xl methodology of repairing these flaws, the 50.59 screening question concerning methods of analysis (Question 111.4) is answered YES. This different process method requires evaluation.
Answers to all other screening questions are NO.Answers to all Evaluation questions are NO.The method of welding described in WCAP-1 5987-P Revision 2-P-A has been approved for use at Westinghouse plants per an NRC SER approved in December 2003, provided that the plant fulfills the criteria for use as defined in the WCAP. There are two applicable criteria for a plant to use this WCAP.
Document Control Desk Attachment I LTD 324, RR-8450 RC-15-0 153 Page 4 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document j Change Description Evaluation Summary 1. The plant must be of Westinghouse or Combustion Engineering design.a. VC Summer is a 3-loop Westinghouse NSSS Reactor and thus meets this condition.
: 2. FEA Analysis of the found flaws must support the ability to use the Westinghouse repair process.a. This analysis shall be completed before entering Mode 5. Analysis shall support the use of WCAP-15987-P Revision 2-P-A for repair. This analysis was completed and is documented in WCAP-17758-NP "Technical Basis for Westinghouse Embedded Flaw Repair for V.C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles and Attachment Welds." Because VC Summer met the conditions and the repair method has been approved by NRC, the proposed repair activity may be implemented without obtaining a License Amendment.
This is consistent with the Relief Request approved by the NRC for the RF-20 repairs. The WOAP repair process bounds flaws found in the J-groove weld, the CRDM Tube CD, and the J-weld to CRDM Tube OD interface.
Document Control Desk AttachmentI LTD 324, RR-8450 RC-1 5-0153 Page 5 of 7 10 CFR 50.59 Summary of Changes and Evaluations Parent Document Change Description Evaluation Summary ECR-71888, Cycle 22 Core Reload Design Cycle 22 Reactor Core Design is needed to produce power in the reactor past the RF-21 date of 4/4/2014.Results from the Applicability Determination concluded a Screen was required.
Screen Question 111.4 (proposed activity involve revising or replacing a method of evaluation described in the FSAR/FPER) was screened YES. All Evaluation questions were answered NO.The NRC has placed limitations and conditions on the use of the new corrosion model, as described in Section 5.0 of the Safety Evaluation Report (SER) included in Reference 7 (WCAP-1261 0-P-A & CENPD-404-P-A, Addendum 2, "Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO, October 2013;hereafter referred to as WCAP-1 261 0-P-A). These conditions, described below, are met for V. C. Summer Cycle 22.*The maximum thermal reaction accumulated duties (TRDs) are restricted to numbers corresponding to a cladding corrosion amount of 100 microns for licensing applications.
* A hydrogen pickup limit of 600 ppm is implemented for ZIRLO and Optimized ZIRLO cladding.* The clad oxide thickness and hydrogen pickup limits are neither eliminated nor replaced.* A fuel-duty index (FDI)-based corrosion model is not used for licensing applications.
Thus, the limitations and conditions imposed by the NRC as part of Section 5.0 of the SER included in WCAP-Document Control Desk Attachment!I LTD 324, RR-8450 RC-1 5-0153 Page 6 of 7 10 CFR 50.59 Summary of Chancies and Evlutin Parent Document Change Description JEvaluation Summary 12610-P-A are met. There are no changes to the 10 CFR 50.46 acceptance criterion that the maximum local oxidation not exceed 17% of the cladding thickness during a Loss of Coolant Accident (LOCA) as a result of the new corrosion model implementation.
Westinghouse has evaluated V. C.Summer Cycle 22 using the integral form ZIRLO and Optimized ZlRLO corrosion model approved by the NRC in WCAP-1 2610-P-A and determined that the best estimate clad oxide thickness and clad hydrogen pickup limits are met.The change to the corrosion model as described in WCAP-1 261 0-P-A has been done in accordance with the limitations and conditions in the SER for WCAP-12610-P-A.
Therefore, the change does not result in a departure from a method of evaluation described in the UFSAR because WCAP-1 2610-P-A has been approved by the NRC for the intended application.
Based on the above, the V. C. Summer Cycle 22 reload core design can be implemented without prior NRC review and approval under 10 CFR 50.59.EIR-82139, Chill Water Train functionality with "C" Chiller Racked in but in Pull-to-Lock
+EIR-82139 documents why either Chilled Water (VU)train remains functional during confidence runs of a non-functional chiller while the swing chiller is racked in on the same train but in Pull-to-Lock. Maintaining the functionality of the VU train prevents the plant from entering a 72 hour Tech Spec action due to potentially Manually starting the swing chiller and associated chilled water pump within a reasonable amount of time (<30 min)after a Safety Injection (SI) or Loss of Offsite Power (LOOP) event will ensure that the required safety related equipment rooms are cooled and that the equipment operating in these rooms are maintained below the Tech Spec room temperature limits of each piece of equipment.
Maintaining room temperatures below the Tech Spec limits Document Control Desk Attachment I LTD 324, RR-8450 RC-15-0 153 Page 7 of 7 10 CFR 50.59 Summary of Changes and Evaluations Parent Document Change Description Evaluation Summary impacting Charging/SI pump ensures that each piece of equipment room cooling, remains operable.
This will allow for testing of a non-functional chiller that is able to provide cooling water without making the Charging-SI pumps inoperable and entering an 72 hr LCO as long as C-chiller and pump are racked in and in Pull-to-Lock, ready to be manually started if the non-functional chiller being tested is not restarted automatically by the sequencer after a SI or LOOP. Manually starting the swing chiller after a SI or LOOP, if required, conforms to the current licensing basis for the plant and can be implemented without obtaining a License Amendment.
Thomas 0. Gatlin Vice President, Nuclear Operations 803.345.4342 A SCANA COMPANY October 7, 2015 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555
 
==Dear Sir!/ Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 10 CFR 50.59 BIENNIAL REPORT South Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Fourth.VCSNS Report pursuant to 10 CFR 50.59(d)(2).
This report contains a brief description and summary of the evaluations performed to support the changes and modifications made to the facility in accordance with 10 CER 50.59(c)(Attachment).
This report covers the time frame from October 1, 2013 to September 30, 2015.If you have any questions or require additional information, please contact Bruce Thompson at (803) 931-5042.Very truly yours, Thomas D. Gatlin WCM/TDG/ts Attachment
-10OCFR50.59 Summary of Evaluations and Changes c: K. B. Marsh S. A. Byrne J. B. Archie N. S. Cams J. H. Hamilton J. W. Williams W. M. Cherry L. D. Wert S. A. Williams NRC Resident Inspector K. M. Sutton NSRC RTS (LTD-324)File (81 8.02-8, RR 8450)PRSF (RC-15-0153)-Iii&#xa2;z V. C. Summer Nuclear Station. P. O. Box 88
* Jenkinsville, SC. 29065. F (803) 941-9776 Document Control Desk Attachment I LTD 324, RR-8450 RC-1 5-0153 Page 1 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document Change Description Evaluation Summary ECR-71781, RMA-2 Guidance ECR-71 781 supports removal of Technical Specification Relocation (TSR) TSR-I1069 operability criteria for RMA-2 based on definition of TSTF-513. Operability guidance for RMA-2 has been developed under design calculation DC00030-058, Revision 1.The calculation serves to: 1. More formally document the FSAR evaluations for the buildup of reactor building activity due to a one gpm leak using the licensing/design basis methods and RCS source terms.2. Determine limits on the alarm setpoints that ensure that the TSTF definition of operability is met when using licensing/design basis assumptions.
ECR-71 781 implements the operability limits within plant procedures to facilitate removal of TSR-i1069, updates the FSAR as appropriate, and enters DC00030-058 into records.The 50.59 Applicability Determination concluded that the proposed changes for ECR-71781 require a 10CFR50.59 review. All 50.59 Screen questions were answered NO except Question 111.4 relating to revising a FSAR described evaluation methodology.
The full I 0CRE50. 59 evaluation, however, concluded that the revisions do not represent a departure from a method of evaluation described in the ESAR since their use is conservative, thus leading to the overall conclusion that the proposed changes for ECR-71781 can be implemented without prior NRC approval.
Document Control Desk Attachment I LTD 324, RR-8450 RC-1 5-0153 Page 2 of 7 10 CFR 50.59 Summary of Chanties and Evaluations m I Parent Document Change Description Evaluation Summary CR-1 1-02428, Cladding Cladding stress/strain The 50.59 Applicability Determination Stress/Strain Methodology, methodology changes concluded that the proposed changes to WCAP 10125-P-A, (existing NRC approval) are the FSAR/FPER would require a 50.59 Addendum 1-A, Revision 1- recommended to regain Screen. All 50.59 Screen questions A. Modify portions of ESAR margin lost due to the were answered NO except Question 111.4 section 4.2.1 to incorporate activities performed to regarding revising or replacing an FSAR changes to cladding stress! address a code error described evaluation methodology.
All strain methodology discovered during the Cycle answers to the 50.59 evaluation were 20 reload design. Margin was NO as the FSAR/FPER described available for Cycles 20 and methodology change does not require 21, but it is desired to prior NRC approval to implement eliminate possible limit because it is within the limitations violations for this parameter.
described in WCAP 10125-P-A and the The vendor code and SER. Changing from one method analysis will be performed for described in the FSAR/FPER to another the Cycle 22 and future method is not a departure if that method designs using methods that has been approved by the NRC for the are currently approved by the intended application.
NRC, but not consistent with our existing FSAR.CR (NC) 01131, Address non-conforming Technical evaluations that include spring Niobium-rich Inclusions in condition observed in the mechanical performance, fuel assembly Top Nozzle Hold-down Inconel 718 material used to holddown permanent set from fuel Springs manufacture the fuel assembly liftoff, and reactor internals assembly top nozzle core barrel flange holddown, have holddown springs. The region shown that an affected spring continues of fuel that was recently to meet all holddown requirements.
delivered to VCS is in the Conservatively assuming that corrosion population of holddown would cause an upper spring failure, the springs made from the potential for loose parts was evaluated.
affected ingot material.
The It was shown that a single fractured leaf abnormality relates to surface will not be released from a nozzle and its indications (discoloration) motion would be limited such that it will noted in a small number of not interfere with RCCA motion or top nozzle upper springs, handling tool engagement.
Additionally, the potential loss of fuel assembly holddown force is not a concern during operation.
The fuel assembly uplift motion is small enough such that the fuel assemblies remain engaged on the upper and lower core plate alignment pins. Thus, the basic structural characteristics of the fuel assemblies Document Control Desk Attachment!I LTD 324, RR-8450 RC-1 5-0153 Page 3 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document Change Description Evaluation Summary remain the same. Insertion of the control rods will not be affected and horizontal Seismic and/or LOCA loads will continue to be reacted by the core plate alignment pins. Additionally, the fuel assembly axial movement is small enough such that the required structural grid overlap is maintained.
Finally, there is no anticipated post-discharge fuel handling, or wet and dry storage issues.The 50.59 Screen Question 111.2 (change to an SSC that adversely affects an FSAR/FPER described design function)screened in YES to require the 50.59 evaluation.
All Evaluation questions were answered NO. Results from technical evaluation of condition of the holddown springs determined them capable of meeting their intended design function and not require a license amendment.
ECR-50846D, Weld Repair Contingency for Reactor Vessel Inspection (RF-21)Perform an overlay weld repair as detailed in WCAP-15987-P Revision 2-P-A"Technical Basis for Embedded Flaw for Repair of Reactor Vessel Head Nozzles." VC Summer performed RV Head Inspections during Refuel Outage 21. During the examination, Primary Water Stress Corrosion Cracking (PWS CC) indications were found in the CRDM penetrations.
These flaws must be repaired prior to entering Mode 5.Since the WCAP process to be used is a deviation from the usual ASME Xl methodology of repairing these flaws, the 50.59 screening question concerning methods of analysis (Question 111.4) is answered YES. This different process method requires evaluation.
Answers to all other screening questions are NO.Answers to all Evaluation questions are NO.The method of welding described in WCAP-1 5987-P Revision 2-P-A has been approved for use at Westinghouse plants per an NRC SER approved in December 2003, provided that the plant fulfills the criteria for use as defined in the WCAP. There are two applicable criteria for a plant to use this WCAP.
Document Control Desk Attachment I LTD 324, RR-8450 RC-15-0 153 Page 4 of 7 10 CFR 50.59 Summary of Chanqes and Evaluations Parent Document j Change Description Evaluation Summary 1. The plant must be of Westinghouse or Combustion Engineering design.a. VC Summer is a 3-loop Westinghouse NSSS Reactor and thus meets this condition.
: 2. FEA Analysis of the found flaws must support the ability to use the Westinghouse repair process.a. This analysis shall be completed before entering Mode 5. Analysis shall support the use of WCAP-15987-P Revision 2-P-A for repair. This analysis was completed and is documented in WCAP-17758-NP "Technical Basis for Westinghouse Embedded Flaw Repair for V.C. Summer Unit 1 Reactor Vessel Head Penetration Nozzles and Attachment Welds." Because VC Summer met the conditions and the repair method has been approved by NRC, the proposed repair activity may be implemented without obtaining a License Amendment.
This is consistent with the Relief Request approved by the NRC for the RF-20 repairs. The WOAP repair process bounds flaws found in the J-groove weld, the CRDM Tube CD, and the J-weld to CRDM Tube OD interface.
Document Control Desk AttachmentI LTD 324, RR-8450 RC-1 5-0153 Page 5 of 7 10 CFR 50.59 Summary of Changes and Evaluations Parent Document Change Description Evaluation Summary ECR-71888, Cycle 22 Core Reload Design Cycle 22 Reactor Core Design is needed to produce power in the reactor past the RF-21 date of 4/4/2014.Results from the Applicability Determination concluded a Screen was required.
Screen Question 111.4 (proposed activity involve revising or replacing a method of evaluation described in the FSAR/FPER) was screened YES. All Evaluation questions were answered NO.The NRC has placed limitations and conditions on the use of the new corrosion model, as described in Section 5.0 of the Safety Evaluation Report (SER) included in Reference 7 (WCAP-1261 0-P-A & CENPD-404-P-A, Addendum 2, "Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO, October 2013;hereafter referred to as WCAP-1 261 0-P-A). These conditions, described below, are met for V. C. Summer Cycle 22.*The maximum thermal reaction accumulated duties (TRDs) are restricted to numbers corresponding to a cladding corrosion amount of 100 microns for licensing applications.
* A hydrogen pickup limit of 600 ppm is implemented for ZIRLO and Optimized ZIRLO cladding.* The clad oxide thickness and hydrogen pickup limits are neither eliminated nor replaced.* A fuel-duty index (FDI)-based corrosion model is not used for licensing applications.
Thus, the limitations and conditions imposed by the NRC as part of Section 5.0 of the SER included in WCAP-Document Control Desk Attachment!I LTD 324, RR-8450 RC-1 5-0153 Page 6 of 7 10 CFR 50.59 Summary of Chancies and Evlutin Parent Document Change Description JEvaluation Summary 12610-P-A are met. There are no changes to the 10 CFR 50.46 acceptance criterion that the maximum local oxidation not exceed 17% of the cladding thickness during a Loss of Coolant Accident (LOCA) as a result of the new corrosion model implementation.
Westinghouse has evaluated V. C.Summer Cycle 22 using the integral form ZIRLO and Optimized ZlRLO corrosion model approved by the NRC in WCAP-1 2610-P-A and determined that the best estimate clad oxide thickness and clad hydrogen pickup limits are met.The change to the corrosion model as described in WCAP-1 261 0-P-A has been done in accordance with the limitations and conditions in the SER for WCAP-12610-P-A.
Therefore, the change does not result in a departure from a method of evaluation described in the UFSAR because WCAP-1 2610-P-A has been approved by the NRC for the intended application.
Based on the above, the V. C. Summer Cycle 22 reload core design can be implemented without prior NRC review and approval under 10 CFR 50.59.EIR-82139, Chill Water Train functionality with "C" Chiller Racked in but in Pull-to-Lock
+EIR-82139 documents why either Chilled Water (VU)train remains functional during confidence runs of a non-functional chiller while the swing chiller is racked in on the same train but in Pull-to-Lock. Maintaining the functionality of the VU train prevents the plant from entering a 72 hour Tech Spec action due to potentially Manually starting the swing chiller and associated chilled water pump within a reasonable amount of time (<30 min)after a Safety Injection (SI) or Loss of Offsite Power (LOOP) event will ensure that the required safety related equipment rooms are cooled and that the equipment operating in these rooms are maintained below the Tech Spec room temperature limits of each piece of equipment.
Maintaining room temperatures below the Tech Spec limits Document Control Desk Attachment I LTD 324, RR-8450 RC-15-0 153 Page 7 of 7 10 CFR 50.59 Summary of Changes and Evaluations Parent Document Change Description Evaluation Summary impacting Charging/SI pump ensures that each piece of equipment room cooling, remains operable.
This will allow for testing of a non-functional chiller that is able to provide cooling water without making the Charging-SI pumps inoperable and entering an 72 hr LCO as long as C-chiller and pump are racked in and in Pull-to-Lock, ready to be manually started if the non-functional chiller being tested is not restarted automatically by the sequencer after a SI or LOOP. Manually starting the swing chiller after a SI or LOOP, if required, conforms to the current licensing basis for the plant and can be implemented without obtaining a License Amendment.}}

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