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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. GatlinVice President, Nuclear Operations803.345.4342A SCANA COMPANYOctober 7, 2015Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555
 
==Dear Sir!/ Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1DOCKET NO. 50-395OPERATING LICENSE NO. NPF-1210 CFR 50.59 BIENNIAL REPORTSouth Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Fourth.VCSNSReport pursuant to 10 CFR 50.59(d)(2).This report contains a brief description and summary of the evaluations performed to supportthe 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. GatlinWCM/TDG/tsAttachment -10OCFR50.59 Summary of Evaluations and Changesc: K. B. MarshS. A. ByrneJ. B. ArchieN. S. CamsJ. H. HamiltonJ. W. WilliamsW. M. CherryL. D. WertS. A. WilliamsNRC Resident InspectorK. M. SuttonNSRCRTS (LTD-324)File (81 8.02-8, RR 8450)PRSF (RC-15-0153)-Iii¢zV. C. Summer Nuclear Station. P. O. Box 88
* Jenkinsville, SC. 29065. F (803) 941-9776 Document Control DeskAttachment ILTD 324, RR-8450RC-1 5-0153Page 1 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document Change Description Evaluation SummaryECR-71781,RMA-2 GuidanceECR-71 781 supports removalof Technical SpecificationRelocation (TSR) TSR-I1069operability criteria for RMA-2based on definition of TSTF-513. Operability guidance forRMA-2 has been developedunder design calculationDC00030-058, Revision 1.The calculation serves to:1. More formallydocument the FSARevaluations for thebuildup of reactorbuilding activity due toa one gpm leak usingthe licensing/designbasis methods andRCS source terms.2. Determine limits onthe alarm setpointsthat ensure that theTSTF definition ofoperability is metwhen using licensing/design basisassumptions.ECR-71 781 implements theoperability limits within plantprocedures to facilitateremoval of TSR-i1069,updates the FSAR asappropriate, and entersDC00030-058 into records.The 50.59 Applicability Determinationconcluded that the proposed changesfor ECR-71781 require a 10CFR50.59review. All 50.59 Screen questions wereanswered NO except Question 111.4relating to revising a FSAR describedevaluation methodology. The fullI 0CRE50. 59 evaluation, however,concluded that the revisions do notrepresent a departure from a method ofevaluation described in the ESAR sincetheir use is conservative, thus leading tothe overall conclusion that the proposedchanges for ECR-71781 can beimplemented without prior NRCapproval.
Document Control DeskAttachment ILTD 324, RR-8450RC-1 5-0153Page 2 of 710 CFR 50.59 Summary of Chanties andEvaluationsm IParent Document Change Description Evaluation SummaryCR-1 1-02428, Cladding Cladding stress/strain The 50.59 Applicability DeterminationStress/Strain Methodology, methodology changes concluded that the proposed changes toWCAP 10125-P-A, (existing NRC approval) are the FSAR/FPER would require a 50.59Addendum 1-A, Revision 1- recommended to regain Screen. All 50.59 Screen questionsA. Modify portions of ESAR margin lost due to the were answered NO except Question 111.4section 4.2.1 to incorporate activities performed to regarding revising or replacing an FSARchanges to cladding stress! address a code error described evaluation methodology. Allstrain methodology discovered during the Cycle answers to the 50.59 evaluation were20 reload design. Margin was NO as the FSAR/FPER describedavailable for Cycles 20 and methodology change does not require21, but it is desired to prior NRC approval to implementeliminate possible limit because it is within the limitationsviolations for this parameter. described in WCAP 10125-P-A and theThe vendor code and SER. Changing from one methodanalysis will be performed for described in the FSAR/FPER to anotherthe Cycle 22 and future method is not a departure if that methoddesigns using methods that has been approved by the NRC for theare currently approved by the intended application.NRC, but not consistent withour existing FSAR.CR (NC) 01131, Address non-conforming Technical evaluations that include springNiobium-rich Inclusions in condition observed in the mechanical performance, fuel assemblyTop Nozzle Hold-down Inconel 718 material used to holddown permanent set from fuelSprings manufacture the fuel assembly liftoff, and reactor internalsassembly top nozzle core barrel flange holddown, haveholddown springs. The region shown that an affected spring continuesof fuel that was recently to meet all holddown requirements.delivered to VCS is in the Conservatively assuming that corrosionpopulation of holddown would cause an upper spring failure, thesprings made from the potential for loose parts was evaluated.affected ingot material. The It was shown that a single fractured leafabnormality relates to surface will not be released from a nozzle and itsindications (discoloration) motion would be limited such that it willnoted in a small number of not interfere with RCCA motion ortop nozzle upper springs, handling tool engagement. Additionally,the potential loss of fuel assemblyholddown force is not a concern duringoperation. The fuel assembly upliftmotion is small enough such that thefuel assemblies remain engaged on theupper and lower core plate alignmentpins. Thus, the basic structuralcharacteristics of the fuel assemblies Document Control DeskAttachment!ILTD 324, RR-8450RC-1 5-0153Page 3 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document Change Description Evaluation Summaryremain the same. Insertion of the controlrods will not be affected and horizontalSeismic and/or LOCA loads will continueto be reacted by the core platealignment pins. Additionally, the fuelassembly axial movement is smallenough such that the required structuralgrid overlap is maintained. Finally, thereis no anticipated post-discharge fuelhandling, or wet and dry storage issues.The 50.59 Screen Question 111.2 (changeto an SSC that adversely affects anFSAR/FPER described design function)screened in YES to require the 50.59evaluation. All Evaluation questionswere answered NO. Results fromtechnical evaluation of condition of theholddown springs determined themcapable of meeting their intended designfunction and not require a licenseamendment.ECR-50846D, Weld RepairContingency for ReactorVessel Inspection (RF-21)Perform an overlay weldrepair as detailed in WCAP-15987-P Revision 2-P-A"Technical Basis forEmbedded Flaw for Repair ofReactor Vessel HeadNozzles." VC Summerperformed RV HeadInspections during RefuelOutage 21. During theexamination, Primary WaterStress Corrosion Cracking(PWS CC) indications werefound in the CRDMpenetrations. These flawsmust be repaired prior toentering Mode 5.Since the WCAP process to be used is adeviation from the usual ASME Xlmethodology of repairing these flaws,the 50.59 screening question concerningmethods of analysis (Question 111.4) isanswered YES. This different processmethod requires evaluation. Answers toall other screening questions are NO.Answers to all Evaluation questions areNO.The method of welding described inWCAP-1 5987-P Revision 2-P-A hasbeen approved for use at Westinghouseplants per an NRC SER approved inDecember 2003, provided that the plantfulfills the criteria for use as defined inthe WCAP. There are two applicablecriteria for a plant to use this WCAP.
Document Control DeskAttachment ILTD 324, RR-8450RC-15-0 153Page 4 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document j Change Description Evaluation Summary1. The plant must be ofWestinghouse or CombustionEngineering design.a. VC Summer is a 3-loopWestinghouse NSSS Reactorand thus meets thiscondition.2. FEA Analysis of the found flawsmust support the ability to usethe Westinghouse repairprocess.a. This analysis shall becompleted before enteringMode 5. Analysis shallsupport the use of WCAP-15987-P Revision 2-P-A forrepair. This analysis wascompleted and isdocumented in WCAP-17758-NP "Technical Basisfor Westinghouse EmbeddedFlaw Repair for V.C. SummerUnit 1 Reactor Vessel HeadPenetration Nozzles andAttachment Welds."Because VC Summer met the conditionsand the repair method has beenapproved by NRC, the proposed repairactivity may be implemented withoutobtaining a License Amendment. This isconsistent with the Relief Requestapproved by the NRC for the RF-20repairs. The WOAP repair processbounds flaws found in the J-grooveweld, the CRDM Tube CD, and the J-weld to CRDM Tube OD interface.
Document Control DeskAttachmentILTD 324, RR-8450RC-1 5-0153Page 5 of 710 CFR 50.59 Summary of Changes and EvaluationsParent Document Change Description Evaluation SummaryECR-71888, Cycle 22 CoreReload DesignCycle 22 Reactor CoreDesign is needed to producepower in the reactor past theRF-21 date of 4/4/2014.Results from the ApplicabilityDetermination concluded a Screen wasrequired. Screen Question 111.4(proposed activity involve revising orreplacing a method of evaluationdescribed in the FSAR/FPER) wasscreened YES. All Evaluation questionswere answered NO.The NRC has placed limitations andconditions on the use of the newcorrosion model, as described in Section5.0 of the Safety Evaluation Report(SER) included in Reference 7 (WCAP-1261 0-P-A & CENPD-404-P-A,Addendum 2, "Westinghouse CladCorrosion Model for ZIRLO andOptimized 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 reactionaccumulated duties (TRDs) arerestricted to numberscorresponding to a claddingcorrosion amount of 100 micronsfor licensing applications.* A hydrogen pickup limit of 600ppm is implemented for ZIRLOand Optimized ZIRLO cladding.* The clad oxide thickness andhydrogen pickup limits areneither eliminated nor replaced.* A fuel-duty index (FDI)-basedcorrosion model is not used forlicensing applications.Thus, the limitations and conditionsimposed by the NRC as part of Section5.0 of the SER included in WCAP-Document Control DeskAttachment!ILTD 324, RR-8450RC-1 5-0153Page 6 of 710 CFR 50.59 Summary of Chancies andEvlutinParent Document Change Description JEvaluation Summary12610-P-A are met. There are nochanges to the 10 CFR 50.46acceptance criterion that the maximumlocal oxidation not exceed 17% of thecladding thickness during a Loss ofCoolant Accident (LOCA) as a result ofthe new corrosion modelimplementation.Westinghouse has evaluated V. C.Summer Cycle 22 using the integralform ZIRLO and Optimized ZlRLOcorrosion model approved by the NRC inWCAP-1 2610-P-A and determined thatthe best estimate clad oxide thicknessand clad hydrogen pickup limits are met.The change to the corrosion model asdescribed in WCAP-1 261 0-P-A hasbeen done in accordance with thelimitations and conditions in the SER forWCAP-12610-P-A. Therefore, thechange does not result in a departurefrom a method of evaluation described inthe UFSAR because WCAP-1 2610-P-Ahas been approved by the NRC for theintended application.Based on the above, the V. C. SummerCycle 22 reload core design can beimplemented without prior NRC reviewand approval under 10 CFR 50.59.EIR-82139, Chill WaterTrain functionality with "C"Chiller Racked in but inPull-to-Lock+EIR-82139 documents whyeither Chilled Water (VU)train remains functionalduring confidence runs of anon-functional chiller whilethe swing chiller is racked inon the same train but in Pull-to-Lock. Maintaining thefunctionality of the VU trainprevents the plant fromentering a 72 hour Tech Specaction due to potentiallyManually starting the swing chiller andassociated chilled water pump within areasonable amount of time (<30 min)after a Safety Injection (SI) or Loss ofOffsite Power (LOOP) event will ensurethat the required safety relatedequipment rooms are cooled and thatthe equipment operating in these roomsare maintained below the Tech Specroom temperature limits of each piece ofequipment. Maintaining roomtemperatures below the Tech Spec limits Document Control DeskAttachment ILTD 324, RR-8450RC-15-0 153Page 7 of 710 CFR 50.59 Summary of Changes and EvaluationsParent Document Change Description Evaluation Summaryimpacting Charging/SI pump ensures that each piece of equipmentroom cooling, remains operable. This will allow fortesting of a non-functional chiller that isable to provide cooling water withoutmaking the Charging-SI pumpsinoperable and entering an 72 hr LCOas long as C-chiller and pump areracked in and in Pull-to-Lock, ready tobe manually started if the non-functionalchiller being tested is not restartedautomatically by the sequencer after aSI or LOOP. Manually starting the swingchiller after a SI or LOOP, if required,conforms to the current licensing basisfor the plant and can be implementedwithout obtaining a License Amendment.
Thomas 0. GatlinVice President, Nuclear Operations803.345.4342A SCANA COMPANYOctober 7, 2015Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555
 
==Dear Sir!/ Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1DOCKET NO. 50-395OPERATING LICENSE NO. NPF-1210 CFR 50.59 BIENNIAL REPORTSouth Carolina Electric & Gas Company (SCE&G) hereby submits the Twenty-Fourth.VCSNSReport pursuant to 10 CFR 50.59(d)(2).This report contains a brief description and summary of the evaluations performed to supportthe 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. GatlinWCM/TDG/tsAttachment -10OCFR50.59 Summary of Evaluations and Changesc: K. B. MarshS. A. ByrneJ. B. ArchieN. S. CamsJ. H. HamiltonJ. W. WilliamsW. M. CherryL. D. WertS. A. WilliamsNRC Resident InspectorK. M. SuttonNSRCRTS (LTD-324)File (81 8.02-8, RR 8450)PRSF (RC-15-0153)-Iii&#xa2;zV. C. Summer Nuclear Station. P. O. Box 88
* Jenkinsville, SC. 29065. F (803) 941-9776 Document Control DeskAttachment ILTD 324, RR-8450RC-1 5-0153Page 1 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document Change Description Evaluation SummaryECR-71781,RMA-2 GuidanceECR-71 781 supports removalof Technical SpecificationRelocation (TSR) TSR-I1069operability criteria for RMA-2based on definition of TSTF-513. Operability guidance forRMA-2 has been developedunder design calculationDC00030-058, Revision 1.The calculation serves to:1. More formallydocument the FSARevaluations for thebuildup of reactorbuilding activity due toa one gpm leak usingthe licensing/designbasis methods andRCS source terms.2. Determine limits onthe alarm setpointsthat ensure that theTSTF definition ofoperability is metwhen using licensing/design basisassumptions.ECR-71 781 implements theoperability limits within plantprocedures to facilitateremoval of TSR-i1069,updates the FSAR asappropriate, and entersDC00030-058 into records.The 50.59 Applicability Determinationconcluded that the proposed changesfor ECR-71781 require a 10CFR50.59review. All 50.59 Screen questions wereanswered NO except Question 111.4relating to revising a FSAR describedevaluation methodology. The fullI 0CRE50. 59 evaluation, however,concluded that the revisions do notrepresent a departure from a method ofevaluation described in the ESAR sincetheir use is conservative, thus leading tothe overall conclusion that the proposedchanges for ECR-71781 can beimplemented without prior NRCapproval.
Document Control DeskAttachment ILTD 324, RR-8450RC-1 5-0153Page 2 of 710 CFR 50.59 Summary of Chanties andEvaluationsm IParent Document Change Description Evaluation SummaryCR-1 1-02428, Cladding Cladding stress/strain The 50.59 Applicability DeterminationStress/Strain Methodology, methodology changes concluded that the proposed changes toWCAP 10125-P-A, (existing NRC approval) are the FSAR/FPER would require a 50.59Addendum 1-A, Revision 1- recommended to regain Screen. All 50.59 Screen questionsA. Modify portions of ESAR margin lost due to the were answered NO except Question 111.4section 4.2.1 to incorporate activities performed to regarding revising or replacing an FSARchanges to cladding stress! address a code error described evaluation methodology. Allstrain methodology discovered during the Cycle answers to the 50.59 evaluation were20 reload design. Margin was NO as the FSAR/FPER describedavailable for Cycles 20 and methodology change does not require21, but it is desired to prior NRC approval to implementeliminate possible limit because it is within the limitationsviolations for this parameter. described in WCAP 10125-P-A and theThe vendor code and SER. Changing from one methodanalysis will be performed for described in the FSAR/FPER to anotherthe Cycle 22 and future method is not a departure if that methoddesigns using methods that has been approved by the NRC for theare currently approved by the intended application.NRC, but not consistent withour existing FSAR.CR (NC) 01131, Address non-conforming Technical evaluations that include springNiobium-rich Inclusions in condition observed in the mechanical performance, fuel assemblyTop Nozzle Hold-down Inconel 718 material used to holddown permanent set from fuelSprings manufacture the fuel assembly liftoff, and reactor internalsassembly top nozzle core barrel flange holddown, haveholddown springs. The region shown that an affected spring continuesof fuel that was recently to meet all holddown requirements.delivered to VCS is in the Conservatively assuming that corrosionpopulation of holddown would cause an upper spring failure, thesprings made from the potential for loose parts was evaluated.affected ingot material. The It was shown that a single fractured leafabnormality relates to surface will not be released from a nozzle and itsindications (discoloration) motion would be limited such that it willnoted in a small number of not interfere with RCCA motion ortop nozzle upper springs, handling tool engagement. Additionally,the potential loss of fuel assemblyholddown force is not a concern duringoperation. The fuel assembly upliftmotion is small enough such that thefuel assemblies remain engaged on theupper and lower core plate alignmentpins. Thus, the basic structuralcharacteristics of the fuel assemblies Document Control DeskAttachment!ILTD 324, RR-8450RC-1 5-0153Page 3 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document Change Description Evaluation Summaryremain the same. Insertion of the controlrods will not be affected and horizontalSeismic and/or LOCA loads will continueto be reacted by the core platealignment pins. Additionally, the fuelassembly axial movement is smallenough such that the required structuralgrid overlap is maintained. Finally, thereis no anticipated post-discharge fuelhandling, or wet and dry storage issues.The 50.59 Screen Question 111.2 (changeto an SSC that adversely affects anFSAR/FPER described design function)screened in YES to require the 50.59evaluation. All Evaluation questionswere answered NO. Results fromtechnical evaluation of condition of theholddown springs determined themcapable of meeting their intended designfunction and not require a licenseamendment.ECR-50846D, Weld RepairContingency for ReactorVessel Inspection (RF-21)Perform an overlay weldrepair as detailed in WCAP-15987-P Revision 2-P-A"Technical Basis forEmbedded Flaw for Repair ofReactor Vessel HeadNozzles." VC Summerperformed RV HeadInspections during RefuelOutage 21. During theexamination, Primary WaterStress Corrosion Cracking(PWS CC) indications werefound in the CRDMpenetrations. These flawsmust be repaired prior toentering Mode 5.Since the WCAP process to be used is adeviation from the usual ASME Xlmethodology of repairing these flaws,the 50.59 screening question concerningmethods of analysis (Question 111.4) isanswered YES. This different processmethod requires evaluation. Answers toall other screening questions are NO.Answers to all Evaluation questions areNO.The method of welding described inWCAP-1 5987-P Revision 2-P-A hasbeen approved for use at Westinghouseplants per an NRC SER approved inDecember 2003, provided that the plantfulfills the criteria for use as defined inthe WCAP. There are two applicablecriteria for a plant to use this WCAP.
Document Control DeskAttachment ILTD 324, RR-8450RC-15-0 153Page 4 of 710 CFR 50.59 Summary of Chanqes and EvaluationsParent Document j Change Description Evaluation Summary1. The plant must be ofWestinghouse or CombustionEngineering design.a. VC Summer is a 3-loopWestinghouse NSSS Reactorand thus meets thiscondition.2. FEA Analysis of the found flawsmust support the ability to usethe Westinghouse repairprocess.a. This analysis shall becompleted before enteringMode 5. Analysis shallsupport the use of WCAP-15987-P Revision 2-P-A forrepair. This analysis wascompleted and isdocumented in WCAP-17758-NP "Technical Basisfor Westinghouse EmbeddedFlaw Repair for V.C. SummerUnit 1 Reactor Vessel HeadPenetration Nozzles andAttachment Welds."Because VC Summer met the conditionsand the repair method has beenapproved by NRC, the proposed repairactivity may be implemented withoutobtaining a License Amendment. This isconsistent with the Relief Requestapproved by the NRC for the RF-20repairs. The WOAP repair processbounds flaws found in the J-grooveweld, the CRDM Tube CD, and the J-weld to CRDM Tube OD interface.
Document Control DeskAttachmentILTD 324, RR-8450RC-1 5-0153Page 5 of 710 CFR 50.59 Summary of Changes and EvaluationsParent Document Change Description Evaluation SummaryECR-71888, Cycle 22 CoreReload DesignCycle 22 Reactor CoreDesign is needed to producepower in the reactor past theRF-21 date of 4/4/2014.Results from the ApplicabilityDetermination concluded a Screen wasrequired. Screen Question 111.4(proposed activity involve revising orreplacing a method of evaluationdescribed in the FSAR/FPER) wasscreened YES. All Evaluation questionswere answered NO.The NRC has placed limitations andconditions on the use of the newcorrosion model, as described in Section5.0 of the Safety Evaluation Report(SER) included in Reference 7 (WCAP-1261 0-P-A & CENPD-404-P-A,Addendum 2, "Westinghouse CladCorrosion Model for ZIRLO andOptimized 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 reactionaccumulated duties (TRDs) arerestricted to numberscorresponding to a claddingcorrosion amount of 100 micronsfor licensing applications.* A hydrogen pickup limit of 600ppm is implemented for ZIRLOand Optimized ZIRLO cladding.* The clad oxide thickness andhydrogen pickup limits areneither eliminated nor replaced.* A fuel-duty index (FDI)-basedcorrosion model is not used forlicensing applications.Thus, the limitations and conditionsimposed by the NRC as part of Section5.0 of the SER included in WCAP-Document Control DeskAttachment!ILTD 324, RR-8450RC-1 5-0153Page 6 of 710 CFR 50.59 Summary of Chancies andEvlutinParent Document Change Description JEvaluation Summary12610-P-A are met. There are nochanges to the 10 CFR 50.46acceptance criterion that the maximumlocal oxidation not exceed 17% of thecladding thickness during a Loss ofCoolant Accident (LOCA) as a result ofthe new corrosion modelimplementation.Westinghouse has evaluated V. C.Summer Cycle 22 using the integralform ZIRLO and Optimized ZlRLOcorrosion model approved by the NRC inWCAP-1 2610-P-A and determined thatthe best estimate clad oxide thicknessand clad hydrogen pickup limits are met.The change to the corrosion model asdescribed in WCAP-1 261 0-P-A hasbeen done in accordance with thelimitations and conditions in the SER forWCAP-12610-P-A. Therefore, thechange does not result in a departurefrom a method of evaluation described inthe UFSAR because WCAP-1 2610-P-Ahas been approved by the NRC for theintended application.Based on the above, the V. C. SummerCycle 22 reload core design can beimplemented without prior NRC reviewand approval under 10 CFR 50.59.EIR-82139, Chill WaterTrain functionality with "C"Chiller Racked in but inPull-to-Lock+EIR-82139 documents whyeither Chilled Water (VU)train remains functionalduring confidence runs of anon-functional chiller whilethe swing chiller is racked inon the same train but in Pull-to-Lock. Maintaining thefunctionality of the VU trainprevents the plant fromentering a 72 hour Tech Specaction due to potentiallyManually starting the swing chiller andassociated chilled water pump within areasonable amount of time (<30 min)after a Safety Injection (SI) or Loss ofOffsite Power (LOOP) event will ensurethat the required safety relatedequipment rooms are cooled and thatthe equipment operating in these roomsare maintained below the Tech Specroom temperature limits of each piece ofequipment. Maintaining roomtemperatures below the Tech Spec limits Document Control DeskAttachment ILTD 324, RR-8450RC-15-0 153Page 7 of 710 CFR 50.59 Summary of Changes and EvaluationsParent Document Change Description Evaluation Summaryimpacting Charging/SI pump ensures that each piece of equipmentroom cooling, remains operable. This will allow fortesting of a non-functional chiller that isable to provide cooling water withoutmaking the Charging-SI pumpsinoperable and entering an 72 hr LCOas long as C-chiller and pump areracked in and in Pull-to-Lock, ready tobe manually started if the non-functionalchiller being tested is not restartedautomatically by the sequencer after aSI or LOOP. Manually starting the swingchiller after a SI or LOOP, if required,conforms to the current licensing basisfor the plant and can be implementedwithout obtaining a License Amendment.}}

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