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Category:CORRESPONDENCE-LETTERS
MONTHYEARNOC-AE-000675, Forwards Clarification on Items Included in 990531 Response to RAI Re Proposed License Amend Associated with Operator Action for Sbloca,As Requested1999-10-21021 October 1999 Forwards Clarification on Items Included in 990531 Response to RAI Re Proposed License Amend Associated with Operator Action for Sbloca,As Requested NOC-AE-000680, Forwards Rev 5 to 0PGP03-ZV-0001, Severe Weather Plan1999-10-20020 October 1999 Forwards Rev 5 to 0PGP03-ZV-0001, Severe Weather Plan NOC-AE-000683, Forwards 30-day Rept Concerning Significant Changes to Accepted Large Break Loss of Coolant Accident ECCS Evaluation Model for South Tx Project,Units 1 & 2,IAW 10CFR50.46(a)(3)(ii)1999-10-19019 October 1999 Forwards 30-day Rept Concerning Significant Changes to Accepted Large Break Loss of Coolant Accident ECCS Evaluation Model for South Tx Project,Units 1 & 2,IAW 10CFR50.46(a)(3)(ii) ML20217K9341999-10-15015 October 1999 Forwards SER Accepting Util 990609 Relief Request RR-ENG-2-4 for Relief from ASME Code,Section XI, Nondestructive Exam Requirements Applicable to Stp,Units 1 & 2,reactor Vessel Closure Head Nuts ML20217K9091999-10-15015 October 1999 Forwards SER Accepting Util 990609 Relief Request RR-ENG-2-3 from ASME Code,Section Xi,Nondestructive Exam Requirements Applicable to South Texas Project,Units 1 & 2, Pressurizer Support Attachment Welds 05000498/LER-1999-008, Forwards LER 99-008-00 Re Turbine Trip That Occurred While Performing Main Turbine Emergency Trip Test.Commitments Made by Licensee Are Listed in Corrective Actions Section of LER1999-10-12012 October 1999 Forwards LER 99-008-00 Re Turbine Trip That Occurred While Performing Main Turbine Emergency Trip Test.Commitments Made by Licensee Are Listed in Corrective Actions Section of LER NOC-AE-000674, Forwards Requested Estimates of Needs for Operator Licensing Exams,Per AL-99-03, Operator Licensing National Exam Schedule1999-10-12012 October 1999 Forwards Requested Estimates of Needs for Operator Licensing Exams,Per AL-99-03, Operator Licensing National Exam Schedule NOC-AE-000625, Requests Partial Relief from ASME Section XI Visual Exam Requirements of IWA-5242(a).Relief Request Is Based on Provisions of Draft ASME Section XI Code Case N-616,which Is Expected to Be Published in Near Future1999-10-0707 October 1999 Requests Partial Relief from ASME Section XI Visual Exam Requirements of IWA-5242(a).Relief Request Is Based on Provisions of Draft ASME Section XI Code Case N-616,which Is Expected to Be Published in Near Future NOC-AE-000610, Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of IWA-5250(a)(2) for Second Inservice Insp Interval,Per Provisions of 10CFR50.55a(3)(i)1999-10-0707 October 1999 Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of IWA-5250(a)(2) for Second Inservice Insp Interval,Per Provisions of 10CFR50.55a(3)(i) NOC-AE-000653, Requests Relief from ASME Section XI Code Requirements of Table IWE-2500-1 for VT-3 Visual Exam of Seals & Gaskets on Airlocks,Hatches & Other Devices Required to Assure Containment leak-tight Integrity,Per 10CFR50.55a(a)(3)(i)1999-10-0707 October 1999 Requests Relief from ASME Section XI Code Requirements of Table IWE-2500-1 for VT-3 Visual Exam of Seals & Gaskets on Airlocks,Hatches & Other Devices Required to Assure Containment leak-tight Integrity,Per 10CFR50.55a(a)(3)(i) ML20217C3221999-10-0707 October 1999 Forwards Insp Repts 50-498/99-16 & 50-499/99-16 on 990808-0918.No Violations Noted.Insp Generally Characterized by safety-conscious Operations,Sound Engineering & Maint Practices & Careful Radiological Work Controls 05000499/LER-1999-006, Forwards LER 99-006-00,re Entry Into TS 3.0.3.Licensee Commitments Listed in Corrective Actions Section of Attachment1999-09-30030 September 1999 Forwards LER 99-006-00,re Entry Into TS 3.0.3.Licensee Commitments Listed in Corrective Actions Section of Attachment ML20212L1651999-09-30030 September 1999 Responds to STP Nuclear Operating Co 981012 & s Which Provided Update to TS Bases Pages B 3/4 8-14 Through B 3/4 8-17.NRC Staff Found Change Consistent with TS 3/4.8.2 DC Sources. Staff Found & Deleted Typographical Error NOC-AE-000664, Forwards Rev 1 to STP Electric Generating Station Unit 1 Cycle 9 COLR & Rev 1 to STP Electric Generating Station Unit 2 Cycle 7 Colr1999-09-30030 September 1999 Forwards Rev 1 to STP Electric Generating Station Unit 1 Cycle 9 COLR & Rev 1 to STP Electric Generating Station Unit 2 Cycle 7 Colr ML20212J7141999-09-29029 September 1999 Forwards Insp Repts 50-498/99-15 & 50-499/99-15 on 990920-24 at South Texas Project Electric Generating Station.No Violations Noted.Insp Covered Requalification Training Program & Observation of Requalification Activities NOC-AE-000646, Informs NRC That STP Nuclear Operating Co Is Y2K Ready IAW Nei/Nusmg 97-07 Guidelines & Also Provides Response to NRC Ltr1999-09-28028 September 1999 Informs NRC That STP Nuclear Operating Co Is Y2K Ready IAW Nei/Nusmg 97-07 Guidelines & Also Provides Response to NRC Ltr ML20212J0651999-09-27027 September 1999 Discusses Licensee 980330 Response to GL 97-06, Degradation of SG Internals. Concludes That Response to GL Provides Reasonable Assurance That Condition of SG Internals in Compliance with Current Licensing Bases for Facility ML20212F1791999-09-24024 September 1999 Discusses 990923 Meeting Conducted in Region IV Ofc Re Status of Activities to Support Confirmatory Order, ,modifying OL & to Introduce New Director,Safety Quality Concerns Program.List of Attendees Encl ML20212E9091999-09-23023 September 1999 Discusses GL 98-01, Year 2000 Readiness of Computer Sys at Npps, Supplement 1 & STP Nuclear Operating Co Response for STP Dtd 990629.Understands That at Least One Sys or Component Listed May Have Potential to Cause Transient ML20212F2111999-09-22022 September 1999 Forwards Review of SG 90-day Rept, South Texas Unit-2 Cycle 7 Voltage-Based Repair Criteria Rept, Submitted by Util on 990119 NOC-AE-000633, Forwards Rev 3 to SG-99-04-005, STP 1RE08 Outage Condition Monitoring Rept & Final Operational Assessment. Rept Satisfies Reporting Requirements of NEI 97-06,dtd Dec 19971999-09-21021 September 1999 Forwards Rev 3 to SG-99-04-005, STP 1RE08 Outage Condition Monitoring Rept & Final Operational Assessment. Rept Satisfies Reporting Requirements of NEI 97-06,dtd Dec 1997 NOC-AE-000634, Forwards Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Povs. MOV design-basis Review Checklist,Encl1999-09-21021 September 1999 Forwards Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Povs. MOV design-basis Review Checklist,Encl NOC-AE-000649, Forwards Current Annual Financial Data for STP Electric Generating Station Per 10CFR50.71(b),acting on Behalf of Central Power & Light Co,City of Austin,Tx,City Public Svc Board of San Antonio & Hl&P1999-09-21021 September 1999 Forwards Current Annual Financial Data for STP Electric Generating Station Per 10CFR50.71(b),acting on Behalf of Central Power & Light Co,City of Austin,Tx,City Public Svc Board of San Antonio & Hl&P 05000499/LER-1999-005, Forwards LER 99-005-00,re Esfa Following Loss of Power to Standby Transformer 2 Due to Electrical Fault.Licensee Commitments Are Listed in Corrective Actions Section of LER1999-09-20020 September 1999 Forwards LER 99-005-00,re Esfa Following Loss of Power to Standby Transformer 2 Due to Electrical Fault.Licensee Commitments Are Listed in Corrective Actions Section of LER ML20212D9171999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of South Texas Project & Identified No Areas in Which Performance Warranted Insp Beyond Core Insp Program.Details of Insp Plan Through Mar 2000 & Historical Listing of Plant Issues,Encl ML20216F5471999-09-15015 September 1999 Discusses 990914 Meeting Conducted at Region Iv.Meeting Was Requested by Staff to Introduce New Management Organization to Region IV & to Discuss General Plant Performance & Mgt Challenges IR 05000498/19990121999-09-14014 September 1999 Forwards Insp Repts 50-498/99-12 & 50-499/99-12 on 990816-19.Three Violations Occurred & Being Treated as Ncvs. Areas Examined During Insp Included Portions of Access Authorization & Physical Security Programs 05000498/LER-1999-007, Forwards LER 99-007-00 Re Train 'B' CR Makeup & Cleanup Filtration Sys Being Inoperable for Greater than Aot.Util Intends to Append Addl Info Section of LER with Brief Description of Test Results,Rather than Submit Separate LER1999-09-13013 September 1999 Forwards LER 99-007-00 Re Train 'B' CR Makeup & Cleanup Filtration Sys Being Inoperable for Greater than Aot.Util Intends to Append Addl Info Section of LER with Brief Description of Test Results,Rather than Submit Separate LER ML20211P8201999-09-0909 September 1999 Forwards SE Authorizing 990224 Submittal of First 10-year Interval ISI Program Plan - Relief Request RR-ENG-24,from ASME Section XI Code,Table IWC-2500-1 NOC-AE-000638, Forwards License Renewal Applications & Certifications of Medical Exam for Seven Listed Licensed Operators at Stp,Per 10CFR55.57.Encl Withheld,Per 10CFR2.790(a)(6)1999-09-0909 September 1999 Forwards License Renewal Applications & Certifications of Medical Exam for Seven Listed Licensed Operators at Stp,Per 10CFR55.57.Encl Withheld,Per 10CFR2.790(a)(6) ML20211P7671999-09-0909 September 1999 Forwards SER Authorizing Licensee 990517 Alternative Proposed in Relief Request RR-ENG-2-8 to Code Case N-491-2 for Second 10-year Insp Interval of South Texas Project, Units 1 & 2,pursuant to 10CFR50.55a(a)(3)(i) ML20211P7871999-09-0909 September 1999 Forwards Safety Evaluation Re First 10-yr Interval Inservice Insp Program Plan Request for Relief RR-ENG-31 IR 05000498/19990141999-09-0303 September 1999 Forwards Insp Repts 50-498/99-14 & 50-499/99-14 on 990627-0807.Apparent Violations Identified & Being Treated as Noncited Violations Consistent with App C of Enforcement Policy NOC-AE-000562, Requests Relief from Contruction Code non-destructive Exam Requirements for Repair/Replacement Activities During Second Inservice Insp Interval of Units 1 & 2,IAW Provisions of 10CFR50.55a(a)(3)(i)1999-08-31031 August 1999 Requests Relief from Contruction Code non-destructive Exam Requirements for Repair/Replacement Activities During Second Inservice Insp Interval of Units 1 & 2,IAW Provisions of 10CFR50.55a(a)(3)(i) ML20212A4351999-08-27027 August 1999 Discusses Investigation Rept OI-4-1999-009 Re Activites at South Texas Project.Oi Investigation Initiated in Response to Alleged Employment Discrimination Complaint. Allegation Not Substantiated.No Further Action Planned NOC-AE-000617, Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d)1999-08-26026 August 1999 Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d) ML20211J2511999-08-26026 August 1999 Discusses Proposed TS Change on Replacement SG Water Level Trip Setpoint for Plant,Units 1 & 2 NOC-AE-000585, Provides Notification That South Texas Project Has Completed Integrity Evaluation of Units 1 & 2 Reactor Coolant Pump Casings Required by Paragraph (D) of Code Case N-4811999-08-25025 August 1999 Provides Notification That South Texas Project Has Completed Integrity Evaluation of Units 1 & 2 Reactor Coolant Pump Casings Required by Paragraph (D) of Code Case N-481 ML20211F4421999-08-24024 August 1999 Forwards SE Authorizing Licensee 990513 Request for Relief RR-ENG-2-13,seeking Relief from ASME B&PV Code Section Xi,Exam Vessel shell-to-flange Welds for Second ISI Intervals ML20211F5031999-08-23023 August 1999 Forwards SE Authorizing Licensee 990315 Request for Relief RR-ENG-30,seeking Relief from ASME B&PV Code,Section Xi,Nde Requirements Applicable to Stp,Unit 2 SG Welds ML20212A4391999-08-17017 August 1999 Discusses Investigation Rept OI-4-1999-023 Re Activities at South Texas Project.Oi Investigation Initiated in Response to Alleged Employment Discrimination for Initiating Condition Report to Document Unauthorized Work Practices ML20210U1271999-08-16016 August 1999 Forwards Insp Repts 50-498/99-08 & 50-499/99-08 on 990517-21 & 0607-10.No Violations Noted.Corrective Action Program Was Reviewed ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML20211B7881999-08-10010 August 1999 Transmits Summary of Two Meetings with Risk-Informed TS Task Force in Rockville,Md on 990514 & 0714 ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams NOC-AE-000603, Informs of Addition of Restriction to SRO License 42658, for KM Espinoza,Effective 990721,per 10CFR50.74.Encl Info Withheld,Per 10CFR2.790(a)(6)1999-07-29029 July 1999 Informs of Addition of Restriction to SRO License 42658, for KM Espinoza,Effective 990721,per 10CFR50.74.Encl Info Withheld,Per 10CFR2.790(a)(6) NOC-AE-000470, Forwards Amend 12 to STP Fire Hazards Analysis Rept. Summary of Changes Made Under Provision of 10CFR50.59 Also Encl1999-07-28028 July 1999 Forwards Amend 12 to STP Fire Hazards Analysis Rept. Summary of Changes Made Under Provision of 10CFR50.59 Also Encl NOC-AE-000599, Forwards STP Unit 1,Cycle 9 Startup Testing Summary Rept. No New Licensing Commitments Contained in Ltr1999-07-28028 July 1999 Forwards STP Unit 1,Cycle 9 Startup Testing Summary Rept. No New Licensing Commitments Contained in Ltr NOC-AE-000589, Forwards Rev to 1RE08 ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests,Corecting Date of End of Insp Interval Provided in Item 9 of from NIS-1 from 09/24/99 to 09/24/20001999-07-26026 July 1999 Forwards Rev to 1RE08 ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests,Corecting Date of End of Insp Interval Provided in Item 9 of from NIS-1 from 09/24/99 to 09/24/2000 ML20210F3851999-07-26026 July 1999 Forwards Exam Repts 50-498/99-301 & 50-499/99-301 on 990706- 15.Exam Included Evaluation of 9 Applicants for SO Licenses & 8 Applicants for RO Licenses 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARNOC-AE-000675, Forwards Clarification on Items Included in 990531 Response to RAI Re Proposed License Amend Associated with Operator Action for Sbloca,As Requested1999-10-21021 October 1999 Forwards Clarification on Items Included in 990531 Response to RAI Re Proposed License Amend Associated with Operator Action for Sbloca,As Requested NOC-AE-000680, Forwards Rev 5 to 0PGP03-ZV-0001, Severe Weather Plan1999-10-20020 October 1999 Forwards Rev 5 to 0PGP03-ZV-0001, Severe Weather Plan NOC-AE-000683, Forwards 30-day Rept Concerning Significant Changes to Accepted Large Break Loss of Coolant Accident ECCS Evaluation Model for South Tx Project,Units 1 & 2,IAW 10CFR50.46(a)(3)(ii)1999-10-19019 October 1999 Forwards 30-day Rept Concerning Significant Changes to Accepted Large Break Loss of Coolant Accident ECCS Evaluation Model for South Tx Project,Units 1 & 2,IAW 10CFR50.46(a)(3)(ii) NOC-AE-000674, Forwards Requested Estimates of Needs for Operator Licensing Exams,Per AL-99-03, Operator Licensing National Exam Schedule1999-10-12012 October 1999 Forwards Requested Estimates of Needs for Operator Licensing Exams,Per AL-99-03, Operator Licensing National Exam Schedule 05000498/LER-1999-008, Forwards LER 99-008-00 Re Turbine Trip That Occurred While Performing Main Turbine Emergency Trip Test.Commitments Made by Licensee Are Listed in Corrective Actions Section of LER1999-10-12012 October 1999 Forwards LER 99-008-00 Re Turbine Trip That Occurred While Performing Main Turbine Emergency Trip Test.Commitments Made by Licensee Are Listed in Corrective Actions Section of LER NOC-AE-000625, Requests Partial Relief from ASME Section XI Visual Exam Requirements of IWA-5242(a).Relief Request Is Based on Provisions of Draft ASME Section XI Code Case N-616,which Is Expected to Be Published in Near Future1999-10-0707 October 1999 Requests Partial Relief from ASME Section XI Visual Exam Requirements of IWA-5242(a).Relief Request Is Based on Provisions of Draft ASME Section XI Code Case N-616,which Is Expected to Be Published in Near Future NOC-AE-000610, Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of IWA-5250(a)(2) for Second Inservice Insp Interval,Per Provisions of 10CFR50.55a(3)(i)1999-10-0707 October 1999 Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of IWA-5250(a)(2) for Second Inservice Insp Interval,Per Provisions of 10CFR50.55a(3)(i) NOC-AE-000653, Requests Relief from ASME Section XI Code Requirements of Table IWE-2500-1 for VT-3 Visual Exam of Seals & Gaskets on Airlocks,Hatches & Other Devices Required to Assure Containment leak-tight Integrity,Per 10CFR50.55a(a)(3)(i)1999-10-0707 October 1999 Requests Relief from ASME Section XI Code Requirements of Table IWE-2500-1 for VT-3 Visual Exam of Seals & Gaskets on Airlocks,Hatches & Other Devices Required to Assure Containment leak-tight Integrity,Per 10CFR50.55a(a)(3)(i) 05000499/LER-1999-006, Forwards LER 99-006-00,re Entry Into TS 3.0.3.Licensee Commitments Listed in Corrective Actions Section of Attachment1999-09-30030 September 1999 Forwards LER 99-006-00,re Entry Into TS 3.0.3.Licensee Commitments Listed in Corrective Actions Section of Attachment NOC-AE-000664, Forwards Rev 1 to STP Electric Generating Station Unit 1 Cycle 9 COLR & Rev 1 to STP Electric Generating Station Unit 2 Cycle 7 Colr1999-09-30030 September 1999 Forwards Rev 1 to STP Electric Generating Station Unit 1 Cycle 9 COLR & Rev 1 to STP Electric Generating Station Unit 2 Cycle 7 Colr NOC-AE-000646, Informs NRC That STP Nuclear Operating Co Is Y2K Ready IAW Nei/Nusmg 97-07 Guidelines & Also Provides Response to NRC Ltr1999-09-28028 September 1999 Informs NRC That STP Nuclear Operating Co Is Y2K Ready IAW Nei/Nusmg 97-07 Guidelines & Also Provides Response to NRC Ltr NOC-AE-000633, Forwards Rev 3 to SG-99-04-005, STP 1RE08 Outage Condition Monitoring Rept & Final Operational Assessment. Rept Satisfies Reporting Requirements of NEI 97-06,dtd Dec 19971999-09-21021 September 1999 Forwards Rev 3 to SG-99-04-005, STP 1RE08 Outage Condition Monitoring Rept & Final Operational Assessment. Rept Satisfies Reporting Requirements of NEI 97-06,dtd Dec 1997 NOC-AE-000634, Forwards Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Povs. MOV design-basis Review Checklist,Encl1999-09-21021 September 1999 Forwards Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Povs. MOV design-basis Review Checklist,Encl NOC-AE-000649, Forwards Current Annual Financial Data for STP Electric Generating Station Per 10CFR50.71(b),acting on Behalf of Central Power & Light Co,City of Austin,Tx,City Public Svc Board of San Antonio & Hl&P1999-09-21021 September 1999 Forwards Current Annual Financial Data for STP Electric Generating Station Per 10CFR50.71(b),acting on Behalf of Central Power & Light Co,City of Austin,Tx,City Public Svc Board of San Antonio & Hl&P 05000499/LER-1999-005, Forwards LER 99-005-00,re Esfa Following Loss of Power to Standby Transformer 2 Due to Electrical Fault.Licensee Commitments Are Listed in Corrective Actions Section of LER1999-09-20020 September 1999 Forwards LER 99-005-00,re Esfa Following Loss of Power to Standby Transformer 2 Due to Electrical Fault.Licensee Commitments Are Listed in Corrective Actions Section of LER 05000498/LER-1999-007, Forwards LER 99-007-00 Re Train 'B' CR Makeup & Cleanup Filtration Sys Being Inoperable for Greater than Aot.Util Intends to Append Addl Info Section of LER with Brief Description of Test Results,Rather than Submit Separate LER1999-09-13013 September 1999 Forwards LER 99-007-00 Re Train 'B' CR Makeup & Cleanup Filtration Sys Being Inoperable for Greater than Aot.Util Intends to Append Addl Info Section of LER with Brief Description of Test Results,Rather than Submit Separate LER NOC-AE-000638, Forwards License Renewal Applications & Certifications of Medical Exam for Seven Listed Licensed Operators at Stp,Per 10CFR55.57.Encl Withheld,Per 10CFR2.790(a)(6)1999-09-0909 September 1999 Forwards License Renewal Applications & Certifications of Medical Exam for Seven Listed Licensed Operators at Stp,Per 10CFR55.57.Encl Withheld,Per 10CFR2.790(a)(6) NOC-AE-000562, Requests Relief from Contruction Code non-destructive Exam Requirements for Repair/Replacement Activities During Second Inservice Insp Interval of Units 1 & 2,IAW Provisions of 10CFR50.55a(a)(3)(i)1999-08-31031 August 1999 Requests Relief from Contruction Code non-destructive Exam Requirements for Repair/Replacement Activities During Second Inservice Insp Interval of Units 1 & 2,IAW Provisions of 10CFR50.55a(a)(3)(i) NOC-AE-000617, Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d)1999-08-26026 August 1999 Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d) NOC-AE-000585, Provides Notification That South Texas Project Has Completed Integrity Evaluation of Units 1 & 2 Reactor Coolant Pump Casings Required by Paragraph (D) of Code Case N-4811999-08-25025 August 1999 Provides Notification That South Texas Project Has Completed Integrity Evaluation of Units 1 & 2 Reactor Coolant Pump Casings Required by Paragraph (D) of Code Case N-481 NOC-AE-000603, Informs of Addition of Restriction to SRO License 42658, for KM Espinoza,Effective 990721,per 10CFR50.74.Encl Info Withheld,Per 10CFR2.790(a)(6)1999-07-29029 July 1999 Informs of Addition of Restriction to SRO License 42658, for KM Espinoza,Effective 990721,per 10CFR50.74.Encl Info Withheld,Per 10CFR2.790(a)(6) NOC-AE-000599, Forwards STP Unit 1,Cycle 9 Startup Testing Summary Rept. No New Licensing Commitments Contained in Ltr1999-07-28028 July 1999 Forwards STP Unit 1,Cycle 9 Startup Testing Summary Rept. No New Licensing Commitments Contained in Ltr NOC-AE-000470, Forwards Amend 12 to STP Fire Hazards Analysis Rept. Summary of Changes Made Under Provision of 10CFR50.59 Also Encl1999-07-28028 July 1999 Forwards Amend 12 to STP Fire Hazards Analysis Rept. Summary of Changes Made Under Provision of 10CFR50.59 Also Encl 05000498/LER-1999-006, Forwards LER 99-006-00 Re Automatic Reactor Trip Due to over-temp delta-temp Actuation.Licensee Commitments Are Listed in Corrective Actions Section of LER1999-07-26026 July 1999 Forwards LER 99-006-00 Re Automatic Reactor Trip Due to over-temp delta-temp Actuation.Licensee Commitments Are Listed in Corrective Actions Section of LER NOC-AE-000589, Forwards Rev to 1RE08 ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests,Corecting Date of End of Insp Interval Provided in Item 9 of from NIS-1 from 09/24/99 to 09/24/20001999-07-26026 July 1999 Forwards Rev to 1RE08 ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests,Corecting Date of End of Insp Interval Provided in Item 9 of from NIS-1 from 09/24/99 to 09/24/2000 NOC-AE-000582, Forwards 1RE08 ISI Summary Rept for Welds & Component Supports of STP Electric Generating Station,Unit 1. Summary Rept Satisfies Reporting Requirements of IWA-6000 of Section XI for Welds & Component Supports1999-07-26026 July 1999 Forwards 1RE08 ISI Summary Rept for Welds & Component Supports of STP Electric Generating Station,Unit 1. Summary Rept Satisfies Reporting Requirements of IWA-6000 of Section XI for Welds & Component Supports NOC-AE-000597, Forwards voltage-based Criteria 90-day Rept for SG Tube Exam Performed Under NRC GL 95-05 During Refueling Outage 1RE08. Rept Contains Info Required by Section 6.b of Attachment 2 to GL 95-051999-07-23023 July 1999 Forwards voltage-based Criteria 90-day Rept for SG Tube Exam Performed Under NRC GL 95-05 During Refueling Outage 1RE08. Rept Contains Info Required by Section 6.b of Attachment 2 to GL 95-05 NOC-AE-000598, Forwards Four Copies of 1RE08 Refueling Outage ISI Summary Rept for Steam Generator Tubing1999-07-23023 July 1999 Forwards Four Copies of 1RE08 Refueling Outage ISI Summary Rept for Steam Generator Tubing NOC-AE-00586, Forwards Results of Control Rod Testing,In Response to NRC Bulletin 96-01, Control Rod Insertion Problems, Dtd 960308.Core Map Provided to Assist in Understanding Test Data1999-07-21021 July 1999 Forwards Results of Control Rod Testing,In Response to NRC Bulletin 96-01, Control Rod Insertion Problems, Dtd 960308.Core Map Provided to Assist in Understanding Test Data NOC-AE-000595, Forwards Chapters 1.0 & 16.0 to Operations QA Plan for South Texas Project.Rev Is Strictly Administrative & All Content Was Previously Submitted to NRC on 990503 & 9906151999-07-21021 July 1999 Forwards Chapters 1.0 & 16.0 to Operations QA Plan for South Texas Project.Rev Is Strictly Administrative & All Content Was Previously Submitted to NRC on 990503 & 990615 NOC-AE-000518, Requests Exemption from Various Special Treatment Requirements of 10CFR50,as Described in Encls to Ltr.Stp Believes That Pilot Application Will Assist NRC in Development & Implementation of risk-informed 10CFR501999-07-13013 July 1999 Requests Exemption from Various Special Treatment Requirements of 10CFR50,as Described in Encls to Ltr.Stp Believes That Pilot Application Will Assist NRC in Development & Implementation of risk-informed 10CFR50 NOC-AE-000536, Submits Request for Exemption from Requirements of 10CFR50.34(b)(11),10CFR50,App A,Gdc 2 & 10CFR100,App a, Section VI(a)(3) Re Maint of Seismic Instrumentation.Revised Page to Procedure OERP01-ZV-IN01 Included1999-07-13013 July 1999 Submits Request for Exemption from Requirements of 10CFR50.34(b)(11),10CFR50,App A,Gdc 2 & 10CFR100,App a, Section VI(a)(3) Re Maint of Seismic Instrumentation.Revised Page to Procedure OERP01-ZV-IN01 Included NOC-AE-000580, Forwards Response to NRC 990415 RAI Re Implementation of Commitments Related to GL 89-10, Safety-Related MOV Testing & Surveillance & GL 96-05, Periodic Verification of Design Basis Capability of Safety-Related Movs1999-07-13013 July 1999 Forwards Response to NRC 990415 RAI Re Implementation of Commitments Related to GL 89-10, Safety-Related MOV Testing & Surveillance & GL 96-05, Periodic Verification of Design Basis Capability of Safety-Related Movs NOC-AE-000574, Forwards ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests Performed Between 971004 & Completion of Eighth RO on 9904281999-07-0606 July 1999 Forwards ISI Summary Repts for Repairs & Replacements & for Sys Pressure Tests Performed Between 971004 & Completion of Eighth RO on 990428 NOC-AE-000557, Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of App III,III-3410 for Second ISI Interval.Proposed Alternatives for Ultrasonic Exam of Piping Sys Welds,Attached1999-07-0606 July 1999 Requests Relief from ASME Section XI Code Nondestructive Exam Requirements of App III,III-3410 for Second ISI Interval.Proposed Alternatives for Ultrasonic Exam of Piping Sys Welds,Attached NOC-AE-000498, Requests Relief from ASME Section XI Code Nondestructive Exam Requirements Applicable to SG Main Steam Nozzle inside- Radius Sections.Attachment Includes Discussion of Basis & Justification for Request & Implementation Schedule1999-07-0606 July 1999 Requests Relief from ASME Section XI Code Nondestructive Exam Requirements Applicable to SG Main Steam Nozzle inside- Radius Sections.Attachment Includes Discussion of Basis & Justification for Request & Implementation Schedule NOC-AE-000573, Requests Relief from Requirements of ASME Section XI Code Case N-498,exempting Isolated Class 1 Reactor Vessel Head Vent Atmospheric Vent Piping & Valve from Being Tested at Full RCS Pressure1999-07-0606 July 1999 Requests Relief from Requirements of ASME Section XI Code Case N-498,exempting Isolated Class 1 Reactor Vessel Head Vent Atmospheric Vent Piping & Valve from Being Tested at Full RCS Pressure NOC-AE-000541, Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Systems at Npps. Readiness Disclosure for STP, Encl1999-06-29029 June 1999 Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Systems at Npps. Readiness Disclosure for STP, Encl NOC-AE-000571, Forwards Final Operating Exam Matls for STP Exam Scheduled for 990705.Revised Operating Exam Outline & post-validation Change Summary Has Been Included.Without Encls1999-06-24024 June 1999 Forwards Final Operating Exam Matls for STP Exam Scheduled for 990705.Revised Operating Exam Outline & post-validation Change Summary Has Been Included.Without Encls NOC-AE-000512, Responds to NRC 981201 Telcon Re Jco 93-0004,per Revised MSLB Analysis1999-06-23023 June 1999 Responds to NRC 981201 Telcon Re Jco 93-0004,per Revised MSLB Analysis NOC-AE-000560, Forwards LER 99-S02-00,re Failure to Maintain Positive Control of Vital Area Security Key.Licensee Commitments Are Found in Corrective Action Section of LER1999-06-23023 June 1999 Forwards LER 99-S02-00,re Failure to Maintain Positive Control of Vital Area Security Key.Licensee Commitments Are Found in Corrective Action Section of LER 05000498/LER-1999-005, Forwards LER 99-005-00,re Failure to Meet Requirements of TS Surveillance 3.7.1.2 Action B for Auxiliary FW Sys.Only Commitments Contained in Ltr Are Located in Corrective Action Section of LER1999-06-17017 June 1999 Forwards LER 99-005-00,re Failure to Meet Requirements of TS Surveillance 3.7.1.2 Action B for Auxiliary FW Sys.Only Commitments Contained in Ltr Are Located in Corrective Action Section of LER NOC-AE-000565, Forwards Amended Pages for Insertion Into South Texas Project Nuclear Operating Co Previously Submitted Response to NRC Rai.New Pages Include Expanded Answer to Question 4.b1999-06-16016 June 1999 Forwards Amended Pages for Insertion Into South Texas Project Nuclear Operating Co Previously Submitted Response to NRC Rai.New Pages Include Expanded Answer to Question 4.b NOC-AE-000548, Forwards Response to RAI Re Proposed Amends on Replacement SG Water Level Trip Setpoint Differences for Stp,Units 1 & 2.Nothing Contained in Response Should Be Considered Commitment Unless So Specified in Separate Correspondence1999-06-16016 June 1999 Forwards Response to RAI Re Proposed Amends on Replacement SG Water Level Trip Setpoint Differences for Stp,Units 1 & 2.Nothing Contained in Response Should Be Considered Commitment Unless So Specified in Separate Correspondence NOC-AE-000561, Forwards Change QA-042 to Operations QAP, Rev 13, Reflecting Current Organizational Alignment for STP & Culminating Organizational Realigment That Has Been Taking Place During Past Several Months1999-06-15015 June 1999 Forwards Change QA-042 to Operations QAP, Rev 13, Reflecting Current Organizational Alignment for STP & Culminating Organizational Realigment That Has Been Taking Place During Past Several Months NOC-AE-0559, Forwards STP Commitment Change Summary Rept for Period 981209-990610.Rept Lists Each Commitment for Which Change Was Made During Reporting Period & Provides Basis for Each Change1999-06-15015 June 1999 Forwards STP Commitment Change Summary Rept for Period 981209-990610.Rept Lists Each Commitment for Which Change Was Made During Reporting Period & Provides Basis for Each Change NOC-AE-000499, Forwards Relief Request RR-ENG-2-3,proposing to Perform Alternative Ultrasonic Examination from Outside Surface of Skirt Attachment Weld as Described in Encl,In Lieu of Surface Examination from Inside Pressurizer Skirt1999-06-0909 June 1999 Forwards Relief Request RR-ENG-2-3,proposing to Perform Alternative Ultrasonic Examination from Outside Surface of Skirt Attachment Weld as Described in Encl,In Lieu of Surface Examination from Inside Pressurizer Skirt NOC-AE-000502, Forwards Relief Request RR-ENG-2-6,proposing That Boroscopic VT-1 Visual Examination Be Allowed as Alternative to Section XI Surface Examination of Pump Casing Welds,Or Portions of Welds within Pits1999-06-0909 June 1999 Forwards Relief Request RR-ENG-2-6,proposing That Boroscopic VT-1 Visual Examination Be Allowed as Alternative to Section XI Surface Examination of Pump Casing Welds,Or Portions of Welds within Pits NOC-AE-000500, Forwards Relief Request RR-ENG-2-4,proposing to Perform Alternative Ultrasonic Examination from Outside & End Surfaces of Reactor Vessel Closure Head Nuts,As Described in Encl in Lieu of Surface Examination of Threaded Region1999-06-0909 June 1999 Forwards Relief Request RR-ENG-2-4,proposing to Perform Alternative Ultrasonic Examination from Outside & End Surfaces of Reactor Vessel Closure Head Nuts,As Described in Encl in Lieu of Surface Examination of Threaded Region NOC-AE-000545, Forwards Response to NRC 990416 RAI Re Util Proposed Amend on Operator Action for Small Break Loca, .Draft EOP Re Small Break Loca,Encl to Aid Discussion of Proposed Amend1999-05-31031 May 1999 Forwards Response to NRC 990416 RAI Re Util Proposed Amend on Operator Action for Small Break Loca, .Draft EOP Re Small Break Loca,Encl to Aid Discussion of Proposed Amend 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARST-HL-AE-3578, Forwards 1RE02 Inservice Insp Summary Rept for Welds & Component Supports..., Describing Exams Performed During Period of 900329-0621,per 1983 Edition of ASME Code,Section XI & Summer 1983 Addenda1990-09-20020 September 1990 Forwards 1RE02 Inservice Insp Summary Rept for Welds & Component Supports..., Describing Exams Performed During Period of 900329-0621,per 1983 Edition of ASME Code,Section XI & Summer 1983 Addenda ST-HL-AE-3577, Responds to Generic Ltr 90-07, Operator Licensing Natl Exam Schedule1990-09-18018 September 1990 Responds to Generic Ltr 90-07, Operator Licensing Natl Exam Schedule ST-HL-AE-3567, Requests NRC Review of Proposed Rev to Schedule for Inservice Insp Exam of Class 1,Category B-D Vessel Nozzles1990-09-14014 September 1990 Requests NRC Review of Proposed Rev to Schedule for Inservice Insp Exam of Class 1,Category B-D Vessel Nozzles ST-HL-AE-3570, Forwards Rev 5 to, South Texas Project Unit 1 Pump & Valve Inservice Test Plan1990-09-14014 September 1990 Forwards Rev 5 to, South Texas Project Unit 1 Pump & Valve Inservice Test Plan ST-HL-AE-3553, Forwards WCAP-12629, Analysis of Capsule U from South Texas Unit 1 Reactor Vessel Radiation Surveillance Program. Pressure-temp Curves Currently in Use in Facility Tech Specs Are More Conservative than Presented in App a of Rept1990-09-10010 September 1990 Forwards WCAP-12629, Analysis of Capsule U from South Texas Unit 1 Reactor Vessel Radiation Surveillance Program. Pressure-temp Curves Currently in Use in Facility Tech Specs Are More Conservative than Presented in App a of Rept ST-HL-AE-3565, Forwards Rev 8 to Operations QA Plan. Plan Revised in Order to Include More Detailed Criteria of Chapter 17.2 of Updated Fsar.Approval Requested1990-09-10010 September 1990 Forwards Rev 8 to Operations QA Plan. Plan Revised in Order to Include More Detailed Criteria of Chapter 17.2 of Updated Fsar.Approval Requested ST-HL-AE-3546, Forwards Corrected Semiannual Radioactive Effluent Release Rept for Second Half of 19891990-08-28028 August 1990 Forwards Corrected Semiannual Radioactive Effluent Release Rept for Second Half of 1989 ST-HL-AE-3550, Forwards Semiannual fitness-for-duty Program Performance Rept for Jan-June 1990,per 10CFR26.71(d)1990-08-28028 August 1990 Forwards Semiannual fitness-for-duty Program Performance Rept for Jan-June 1990,per 10CFR26.71(d) ST-HL-AE-3540, Provides Schedule Under Which Facility Turbine Components Inspected for Functional Integrity.Required Insp Intervals Calculated to Maintain Probability of Missile Generation for Each Low Pressure Rotor1990-08-28028 August 1990 Provides Schedule Under Which Facility Turbine Components Inspected for Functional Integrity.Required Insp Intervals Calculated to Maintain Probability of Missile Generation for Each Low Pressure Rotor ST-HL-AE-3551, Forwards Responses to NRC 900807 Request for Addl Info Re Probabilistic Safety Assessment Human Reliability Analysis. Paper on Quantification of Human Error Rates Using slim-based Approach Encl1990-08-26026 August 1990 Forwards Responses to NRC 900807 Request for Addl Info Re Probabilistic Safety Assessment Human Reliability Analysis. Paper on Quantification of Human Error Rates Using slim-based Approach Encl ML20043H6191990-06-21021 June 1990 Forwards 1989 Annual Financial Repts for Licensees for Plant ML20043H5971990-06-19019 June 1990 Forwards Responses to Open Items Resulting from Sandia Draft Rept on Probabilistic Safety Assessment.Dominant Sequence Model Encl,Per NRC Reviewers Request ML20043H7781990-06-18018 June 1990 Forwards Rev 1 to SER Commitment Status for Plant,Per NUREG-0781.List of Action Items Completed But Not Incorporated Into Sser & List of Items for Actions Not Completed Also Encl ML20043F6261990-06-11011 June 1990 Forwards Rev 0 to Unit 1 Cycle 3 Core Operating Limits Rept. ML20043F3191990-06-0404 June 1990 Responds to NRC 900504 Ltr Re Violations Noted in Insp Repts 50-498/90-01 & 50-499/90-01.Corrective Actions:Compressed Gas Cylinders Removed from Power Block & Nashua 357 Tape Returned to Nuclear Purchasing Matl Mgt Co ML20043D3241990-06-0101 June 1990 Forwards Rev 10 to Safeguards Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20043B1651990-05-21021 May 1990 Advises That Qualified Display Processing Sys on Line Parameter Update Mod Will Be Completed During Next Refueling Outages,Per 900312 Ltr ML20043A9951990-05-16016 May 1990 Discusses Actions Taken Re Prompt Notification Sys.Util Found Autodialer Sys Offer Acceptable Alternative to Replacing Majority of Tone Alert Radios ML20043H6531990-05-16016 May 1990 Forwards Plant Owner Draft Decommissioning Certificate & Util & City Public Svc Board of San Antonio Decommissioning Master Trust Agreements for South Texas Project ML20043A6081990-05-16016 May 1990 Forwards Rev 16 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21) ML20043A8641990-05-14014 May 1990 Clarifies Operation of Telephone Autodialer Sys,As Part of Prompt Notification Sys.Autodialer Sys,As Currently Configured,Can Autodial & Deliver Prerecorded Message to Residents at Rate of Approx 20 Calls Per Minute ML20043A3271990-05-10010 May 1990 Forwards Endorsements 7 & 6 to Maelu Certificates M-113 & M-116,respectively & Endorsement 6 to Nelia Certificates N-113 & N-116 ML20042G8641990-05-0909 May 1990 Comments on SALP Repts 50-498/90-06 & 50-499/90-06 for Jan 1989 to Jan 1990.Util Working on Areas Identified During SALP Rept & Mgt Meeting on 900425 ML20042F9571990-05-0707 May 1990 Responds to NRC 900405 Ltr Re Violations Noted in Insp Repts 50-498/90-05 & 50-499/90-05.Corrective Actions:Surveillances of Emergency Response Equipment in Technical Support Ctrs Performed to Ensure That Emergency Requirements Satisfied ML20042F1171990-05-0101 May 1990 Submits Special Rept Re Evaluation of third-yr Containment Tendon Surveillance.Tendons Which Had Voids Have Been Filled & No Evidence of Grease Leakage from Sheathing Exists ML20042F3631990-04-30030 April 1990 Provides Summary of Expected Sequence of Events for Updates to Prompt Notification Sys ML20042E6361990-04-20020 April 1990 Forwards Revised Organization Chart,Correcting Postion Titles Reflected in 900326 Submittal ML20042E5421990-04-12012 April 1990 Responds to 900316 Notice of Violation for Insp Repts 50-498/90-08 & 50-499/90-08.Violation Addressed in LER 90-003 Re Failure to Perform Tech Spec Required Surveillance Due to Deficient Procedure ML20042E1591990-04-0505 April 1990 Provides Listed Guidelines for Development of Operating Procedures Re Ac Power Restoration to Respond to Station Blackout Event,Per 10CFR50.63, Loss of All AC Power. ML20012F2881990-04-0202 April 1990 Provides Rept of Nuclear Insurance Protection,Per 10CFR50.54(w)(2).NEIL-II Decontamination Liability & Excess Property Policy Increased Effective 891115 ML20012F2831990-04-0202 April 1990 Informs of Deferral of Facility Mods to Install Permanent RHR Pump Motor Current Indication.Mod Will Be Completed Before Next Reduced RCS Inventory Conditions on Unit ML20012D8731990-03-19019 March 1990 Forwards Revised Correspondence Distribution List of Designated Recipients ML20012C6121990-03-16016 March 1990 Forwards NRC Regulatory Impact Survey Questionnaire Sheets in Response to Generic Ltr 90-01, Request for Voluntary Participation in NRC Regulatory Impact Survey. Brief Summary Table of Questionnaire Data Also Encl ML20012C6171990-03-16016 March 1990 Forwards Status of Actions Committed to Re NRC Bulletin 88-004 in Response to G Dick Request ML20012C2811990-03-12012 March 1990 Forwards Suppl 3 to Qualified Display Processing Sys (Qdps) Verification & Validation Process Final Rept & Summary of Qualified Display Processing Sys (Qdps) Recurring Component Failure Data. ML20012C0641990-03-12012 March 1990 Forwards Rev 15 to Physical Security Plan.Rev Withheld (Ref 10CFR73.21) ML20012B7651990-03-0909 March 1990 Responds to Generic Ltr 88-17, Loss of Dhr. Util Will Revise Appropriate Procedures to Require Entering Reduced Inventory Operation at 3 Ft Below Reactor Vessel Flange ML20042D6701990-03-0808 March 1990 Responds to NRC Generic Ltr 89-19, Resolution of USI A-47,Safety Implication of Control Sys in LWR Nuclear Power Plants. Plant Design Meets Criteria of Generic Ltr 89-19 for Automatic Steam Generator Overfill Protection ML20012A1291990-03-0101 March 1990 Forwards Responses to Questions Raised by Sandia During Review of Plant PRA Covering Steam Generator Dryout ML20012A3001990-02-28028 February 1990 Forwards Nonproprietary & Proprietary Rev 1,Suppl 2 to WCAP-12087 & WCAP-12067, Reconciliation of Fatigue Crack Growth Results for South Texas Project Unit 1 Surge..., Per NRC Bulletin 88-011.WCAP-12067 Withheld (Ref 10CFR2.790) ML20011F1921990-02-22022 February 1990 Responds to NRC 900123 Ltr Re Violations Noted in Insp Repts 50-498/89-47 & 50-499/89-47.Corrective Actions:Heat Trace Circuit Temp Controllers Calibr & Analog Indication Checked & Found within Tolerance on All But Three Channels ML20011E9061990-02-16016 February 1990 Responds to NRC IE Bulletin 89-003, Potential Loss of Required Shutdown Margin During Refueling Operations. Util Concluded That Sufficient Precautions Will Be in Place to Ensure Against Loss of Required Shutdown Margin ML20006D9121990-02-0707 February 1990 Forwards Updated Schedule of Responses to NRC Generic Ltrs 83-28,88-17,88-20,89-05,89-06,89-07,89-08,89-10,89-12,89-13, 89-17,89-19 & 89-21 & Bulletins 88-001,88-008,88-010,88-011, 89-001,89-002,89-003,88-008,Suppl 1 & 88-010,Suppl 1 ML20011E8251990-02-0505 February 1990 Requests Consideration of Scenario Manual for 900404 Graded Emergency Preparedness Exercise as Proprietary Info Until After Graded Exercise ML20006B8801990-01-31031 January 1990 Forwards Comparison of Instrusion Detection Sys Proposed for Various Physical Security Upgrade.Encl Withheld (10CFR73.21) ML20011E1831990-01-30030 January 1990 Requests Approval of Schedular Exemption from 10CFR50,App J, Type C Local Leak Rate Testing Requirements by 900301,based on Interval Between Completion of Unit 1 First Refueling Outage & Second Refueling Outage Start of Only 6 Months ML20011E1361990-01-29029 January 1990 Responds to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Three Bays of Essential Cooling Water Intake Structure Will Be Inspected Once Every 18 Months for Macroscopic Biological Fouling Organisms ML20006A9791990-01-26026 January 1990 Provides Addl Info Re Surge Line Stratification at Facility, Per NRC Bulletin 88-011.W/exception of One Heatup Cycle, Stratification Observed in Surge Line Determined to Be within Bounds of WCAP-12067,Rev 1 ML20006A7981990-01-26026 January 1990 Forwards Response to Violations Noted in Insp Repts 50-498/89-39 & 50-499/89-39.Response Withheld ML20006B3551990-01-23023 January 1990 Forwards Rev 4 to Pump & Valve Inservice Test Plan. Rev Includes Addition of Component Cooling Water Valves FV-0864,FV-0862 & FV-0863 & Containment Sys Valves FV-1025, FV-1026,FV-1027 & FV-1028 1990-09-20
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47
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The Light man Company n .m su iin,,g u,,m m M,,x nm ih- m.SL omsmn hoch Rb r 1986 File No.: G9.18 Mr. Vincent S. Noonan, Project Director PWR Project Directorate #5 U. S. Nuclear Regulatory Commission Washington, DC 20555 South Texas Project Units 1 and 2 Docket Nos. STN 50-498, STN 50-499-d k SER Open Item #16 and Confirmatory Item #13; Long Term Cooling Referecee: Letter ST-HL-AE-1724 dated August 20, 1986; M. R. Wisenburg to V. S. Noonan
Dear Mr. Noonan:
This letter transmits information with respect to South Texas Project Safety Evaluation Report (SER) open item #16 and confirmatory item #13. These items involve the use of the Westinghouse (W) TREAT computer program to evaluate STP's long term cooling capabilities in the event of a small break loss-of-coolant accident (LOCA), a non-isolable LOCA and a secondary break.
The analysis has been completed and demonstrates that the STP emergency core cooling system (ECCS) meets the requirements of 10CFR50.46. The analy::is, which was discussed with the NRC staff in August 1986 (meeting minutes were provided in the reference), is provided in Enclosure 1.
Enclosure 2 includes associated revisions to the FSAR which will be incorporated in a future FSAR amendment.
Also provided is a comparison of the TREAT and NOTRUMP Westinghouse programs which was used to benchmark TREAT against applicable 10CFR50.46 criteria. Enclosure 3 contains the following:
- 1. 5 copies of WCAP-ll232, " Comparison of the TREAT and NOTRUMP Small Break LOCA Transient Results." (Proprietary)
- 2. 5 copies of WCAP-ll297, " Comparison of the TREAT and NOTRUMP Small Break LOCA Transient Results." (Non-proprietary)
Also enclosed is a Westinghouse authorization letter (CAW-86-084), Proprietary Information Notice, and accompanying affidavit.
l L1/NRC/na 610060980 860%O PDR E
ADOCK 0500049e PDR 03 03 J
8 b 1 Houston Lighting & Power Company Page 2 As item (1) of Enclosure 3 contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the_ owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.790 of the Commission's regulations.
Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10CFR Section 2.790 of the Commission's regulations. Correspondence with respect to the proprietary aspects of the Application for Withholding or the supporting Westinghouse affidavit should reference CAW-86-084 and should be addressed to R. A. Wiesemann, Manager, Regulatory and Legislative Affairs, Westinghouse Electric Corporation, P. O. Box 355, Pittsburgh, Pennsylvania 15230.
Based on the information provided in the enclosures Houston Lighting &
Power Company considers the aforementioned items to be " closed". If you should have any questions on this matter, please contact Mr. J. S. Phelps at (713) 993-1367.
Very truly yours,
/
M. R. Wis urg Manager, N clear Lice is ng JSP/yd Enclosures 1: Long Term Cooling Analysis for South Texas Project 2: FSAR revisions 3: Comparison of the TREAT and NOTRUMP Small Break Transient Results Five copies each of WCAP-ll232 (Proprietary) and WCAP-ll297 (Non proprietary)
Ll/NRC/na
Houston Lighting & Power Company ST-ML-AE-1767 File No.: G9.18 Page 3
Hugh L. Thompson, Jr. , Director Brian E. Berwick, Esquire Division of PWR Licensing - A Assistant Attorney General for Office of Nuclear Reactor Regulation the State of Texas U.S. Nuclear Regulatory Commission P.O. Box 12548, Capitol Station Washington, DC 20555 Austin, TX 78711 Robert D. Martin Lanny A. Sinkin Regional Administrator, Region IV Christic Institute Nuclear Regulatory Commission 1324 North Capitol Street 611 Ryan Plaza Drive, Suite 1000 Washington, D.C. 20002 Arlington, TX 76011 Oreste R. Pirfo, Esquire N. Prasad Kadambi, Project Manager Hearing Attorney U.S. Nuclear Regulatory Commission Office of the Executive Legal Director 7920 Norfolk Avenue U.S. Nuclear Regulatory Commission Bethesda, MD 20814 Washington, DC 20555 Claude E. Johnson Charles Bechhoefer, Esquire Senior Resident Inspector /STP Chairman, Atomic Safety &
c/o U.S. Nuclear Regulatory Licensing Board Commission U.S. Nuclear Regulatory Commission P.O. Box 910 Washington, DC 20555 Bay City, TX 77414 Dr. James C. Lamb, III M.D. Schwarz, Jr., Esquire 313 Woodhaven Road Baker & Botts Chapel Hill, NC 27514 One Shell Plaza Houston, TX 77002 Judge Frederick J. Jhon Atomic, Safety and Licensing Board J.R. Newman, Esquire U.S. Nuclear Re'gulatory Commission Newman & Holtzinger, P.C. Washington, DC 20555 1615 L Street, N.W.
Washington, DC 20036 Citizens for Equitable Utilities, Inc.
c/o Ms. Peggy Buchorn Director, Office of Inspection Route 1, Box 1684 and Enforcement Brazoria, TX 77422 U.S. Nuclear Regulatory Commission Washington, DC 20555 Docketing & Service Section Office of the Secretary T.V. Shockley/R.L. Range U.S. Nuclear Regulatory Commission Central Power & Light Company Washingten, DC 20555 P.O. Box 2121 (3 Copier)
Corpus Christi, TX 78403 Advisory Committee on Reactor Safeguards H.L. Peterson/G. Pokorny U.S. Nuclear Regulatory Commission City of Austin 1717 H Street P.O. Box 1088 Washington, DC 20555 Austin, TX 78767 J.B. Poston/A. vonRosenberg City Public Service Board P.O. Box 1771 San Antonio, TX 78296
- Enclosures 1 & 2 Only Ll/NRC/na Revised 5/22/86
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ATTACHMENT oL ST HL AE- me'7 PAGE I 0F id, STP FSAR monitoring. To accommodate this additional RCS inventory, letdown may be dis-charged from the reactor vessel head vent line with letdown routed to the pressurizer relief tank. The head vent throttling valves are controlled through the Qualified Display Processing System (QDPS), described in Section 53 7.5.6.
The safety-related cooldown is accomplished by increasing the steam release from the SG PORVs to attain a rate of primary side cooling of approximately 25*F/hr. The SC PORVs are also controlled through the QDPS. In conjunction l53 with this portion of the cooldown, the charging pumps are used to deliver water to make up for primary system contraction due to cooling. Makeup is also required for inventory control in the event the reactor vessel head vent ,.
path is periodically opened to provide head cooling. ^- citern tive te head > '
errling er tr previde en 9 Hr.-soak peried* fell:uing--RCS :::Id:r end prier 3-D u RCS d:p::::: i: tier S: : f:ty g: d: .SST h:2 d:qu t: ::pecity te ecce- L 54 r:dete the identif!:d :::P p:ried. Upon approaching the end of this phase of cooldownr(RCS temperature of approximately 350'F), the RCS is depressurized to approximately 350 psig by venting the pressurizer through the safety related pressurizer PORVs.
To ensure that the accumulators do not repressurize the RCS, the accumulator discharge valves are closed prior to the RCS pressure dropping below the accumulator discharge pressure. Each accumulator is provided with a Class lE solenoid actuated valve to ensure that the accumulator may be vented through the nitrogen supply header should the accumulator discharge isolation valve fail. A branch line inside the Containment with a parallel set of Class IE 3g valves allows venting the nitrogen header to Containment atmosphere.
i Actuation of the SIS is precluded by use of the pressurizer low pressure and excessive cooldown signal blocks.
When the reactor coolant temperature and pressure are reduced to approximately 350* and 350 psig, respectively, the second phase of cooldown starts with the RHRS being placed in operation. Since loss of the non safety-grade instrument l53 air system results in a loss of the air supply to the flow control valves that are normally used to limit the initial RHRS cooldown rate, the operator may choose to use only one of the RHR subsystems as a means to control cooldown rate. Should a single failure occur, such as that of an RHRS component, precluding operation of one of the RHR subsystems, the operator could elect to use a fully operational RHR train. Cooldown would continue using the fully operational RHR train (s), until the failed equipment or component could be made available. A failure mode and effects analysis for cold shutdown operations is provided in Table 5.4.A-2.
Cooldown of the RCS is continued using available RHR trains and following cooldown rate limits. The time required to reach the cold shutdown conditions (see definition in Technical Specifications) depends upon the number of RHR trains available, and the CCW and ECW temperatures.
5.4.A.3 Amendment 53
f k
Tantt 5.8.a 1 (ContInve<f)
CtpIPLIAeCE Conraelsos witu semeCN IFCNWICAL POSIflou R$s 5 1 Destsn tevastrements Procese eruf ISystes of stP ess 5-1 Possible Solution for Reccommewted laptemaritetton Degree of STP or C m _t)_ J ult Comptinnee _ _ fa _r Class _2y lants' _Crup_t lanc,e**
W. Test requirement Run tests and confire- Camellence reepairevf. peeets the Intent of inq onelysis to meet meet RC 1.68 for Pues, rep;Irement. DC 1.68 Test date test plus onetysis for and analysis for e cooldows unclar natural plant similer in circulation to confirm design to STP will erfsgaste mining verify adaquete and cooldown within miairg and cooldonat timits specified in under naturet circule-Faergency operotIng t Ion conrfi t f ens Procedsres. ,
(Sectlert 14.2).
VI. Operationet procedsre Develop procedJree and Comptlance reepstred. Generic Procedures as Information from tests developed by the West-sneet et 1.33. For Pues, and onetysis.
include specific proce- in, house Cuners croie dJres eruf informatiert for will be used as the cooldows smder natural basis for plant specific circulation. procedsres. 38 VII. Atatillery Feedseter Emergency feedseter Frome tests and analyels Compliance vill not be Sipply , sigpt y obtain conservetive reepsired if it is shown The AFSt usable cepecify of $25,000 sets is ode-y estimate of sumiliary that on odeepsete etternate .o Selenic Category I feedseter sagety to quote to steport 4 hrs et stely for munittery Seismic Category I source hot stenchy and a 10 hr meet requirments and le ovellebte. en fe
- ter for et croldoem to Rue cut In tesst four hours provide Seismic Cate-cerufltlerig)ot towed t'y 54 4 et het " - ^, *- 5 g gory I st9 ply, our sont period plus cooldown to prior to actueLast int-residant heet removal M tinti with a mergin for cut in tagerature con engencies. The AFSt f rollowed by en 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ) meets Seismic Category I
( seat period /besed ori requirements (Section
' longest time for only 10.4.7) onsite or only of fsite T Ch >
pc-er and ess=md j my single failure.
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- The i=sstementation for Cless 2 plants dnes viet result in e assine imaet ashite providing arsfiticast capahitity en on to entet weedmen.
the map <tr smart resist ts f raun the rc<psirement for safety relate.f st em etes, wat wes.
- Sie tat t s i . th en the cate<pory nf Claes 2 riant es dafinerf tv *:er t im #. *f mf aman' at im." af Ne w 4 ter %*si r at e - i t . , eta ' t, %*-+'== ?
ATTACHMENT A ST HL AE l%?
PAGE 3 0F j A STP FSAR The A WS is also designed for the following normal plant operations. lh 10.4.9.1.1 Plant Cold Startup: The AWS is designed to back up the main W system during plant startup in the event the main W system and/or the startup SGFP is unavailable.
- 10.4.9.1.2 Plant Hot Shutdown: The AWS is designed to back up the main W system during plant hot shutdown (or hot standby) in the event the main W system and/or the startup SGFP is unavailable. The AWS can be used as a means of continuous W supply even if this condition is maintained for ,,
extended periods. W is continuously supplied from the AFST, which during normal operation receives required makeup from the demineralized water storage tank (D',:ST) . The DST in turn is supplied by water from wells through the
. demineralizers, as shown on Figures 9.2.3-1 and 9.2.6-1, 10.4.9.1.3 Plant Cold Shutdevn: The AWS is designed to back up the main W system when achieving plant cold shutdown.
10.4.9.2 System Description. One AWS is provided for each unit. The piping diagram is shown on Figure 10.4.9 1. The system includes an adequate l39 water storage, redundant pumping capacity to supply the SGs, associated piping, valves, and instru=entation.
~
The A WS supplies water to the SGs, where it is converted into steam by the heat transferred from the primary coolant that removes decay heat from the reactor core and heat generated in the primary coolant loop by the reactor C. coolant pumps. 39 The AFST provides water to the AW pumps. It is a concrete, stainless steel lined, tank wich a usable capacity of 525,000 gallons based on the following i plus a margin for contingencies; 1'5 31 e maintaining the plant in hot standby for four hours, then e
cooling down the primary system to 350'F,j m ' * " gO
, 2 ** - > t. . ., . . u u . . . . . t. . . -. 3*-6.~~~ . . ~'"
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The cooldown rate is 50'F/hr with one RCP operating or 25'F/hr with natural circulation. During normal cooldown the rate is limited to 100*F/hr due to l 3,-
structural limits of the RCS components.
Four AW pu:tps, each with independent motive power supplies, are p*ovided to l4(
comply with redundancy requirements of the safety standards, both for equipment and power supplies. Pump ch'aracteristics are given in Table 10.1 1.
l39 Three horizontal, centrifugal, multistage, electric motor driven pumps supply one SG each. Each pump motor is supplied power from a separate engineered safety bus, and the power supply is separated throughout, ,
1 l
l l
10.4 29 Amend. ment 54
ATTACHMENT JL ST HL AE FHo ?
PAGE 4 OFl3-STP FSAR
- 7. Uncoupling of Human Errors This study assumes that test and maintenance activities are staggered. Thst is, redundant AFVS components are not tested by the same personnel on the same shift, but in general, tests and/or maintenance of redundant components involve time and/or personnel changes (e.g., different personnel and shifts, or the -
same personnel on a different day, etc.) In addition, a double check procedure is assumed to assure the correct status of locked open valves after test and maintenance. This signif-icantly reduces the probability of human error in two or more trains simultaneously. Given that test and maintenance activi-ties are staggered and the use of a double check procedure, it
- . is reasonable to assume that human errors for test and mainte-nance are uncoupled.
For the above reasons, the evaluation does not consider con-current disabling of multiple trains because of human error in -
conjunction with test or maintenance to be a credible failure scenario.
1
- 8. Technical Specification
~~
The auxiliary feedwater system design is evaluated in accor-dance with the STP Technical Specifications (Ref. 7).
- Train A Availability is assumed to be degraded since there is 54 no Technical Specification requirement on Train A.
Trains B, - Operable except for the scenarios
- C, and D illustrated in the fault trees in Section 10A.3.2.
- 9. HVAC Support The motor driven auxiliary feedwater pump rooms are cooled by safety related HVAC units powered by their respective trains.
The turbine driven pump room is cooled by a Train A HVAC unit, however, the turbine driven pump is qualified for operation following the loss of all HVAC. Consistent with KUREG 0611 methodology, HVAC support to the pumps is not considered in this evaluation.
- 10. Auxiliary Feedwater Storage Tank
! The AFVST capacity is sufficient to allow the RCS to remain at c'
')y hot standby im o. for 4athours cre' peri:d whichfo116wed by aRCS point further 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown cooldown>' 15 per!d r :
formed by the residual heat removal system. If additional quantities are needed, water can be provided to the AFL'ST from the demineralized water storage tank, the condenser hot well, t or an alternate onsite source. The AFVST has level instrumen-4 tation with control room indication and annunciation'to warn
- operators of low AFVST water inventory.
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10A 4 Amendment 54
l ATTACHMENT A ST-HL AE 11(oq PAGE 5 OF l A uidl o.- pe kf3- VM q STP FSAR CtTD . M
- 4. Auxiliary Fe dwater Storage Tank The Seismic Category I auxiliary feedwater storage tank pro-vides water to the AW pumps. It is a concrete, stainless steel line 510,000 ;;11;s tank d.id has sufficient capacity to allow t RCS to remain at hot standby for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> followed by a 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown " :n ? ' __ :: d 7:ri:87 which point furtherRCScooldownisperformedbytheresidualheatremoval system.
The AWST is designed to withstand environmental design condi-tions, including floods, earthquakes, hurricanes, tornado load-ings, and tornado missiles. The AWST is designed so that no single active failure will preclude the ability to provide water to the AW system. Each train has a dedicated suction line from the AWST to the AW pumps. The water level in the AWST is indicated in the control room as well as at the auxiliary shutdown panel. A low level alarm is also provided in the control room.
10A.2.3 Emergency Operation The AWS is designed for automatic actuation in an emergency. Any of the following conditions automatically starts the three Class lE motor-driven Pumps:
54
. 1. Two out of four channels showing low low water level in any steam generator
- 2. Safety injection signal
- 3. 4.16 kV bus undervoltage. The AW pump is started in conjunction with diesel generator starting and load sequencing.
Water is not automatically fed to the steam generator until condition 1 or 2 above exists.
The turbine driven auxiliary feedwater pump starts automatically on any of the following signals:
- 1. Two out of four channels showing low low water level in any steam generator
- 2. Safety injection signal A one inch bypass line with a normally closed solenoid operated valve (WO143) and orifice is provided around the steam inlet valve (MS0143). This bypass valve (WO143) opens upon receipt of either of the above signals to supply steam to the turbine and allow the turbine to reach governor control speed.
After a time delay to allow governor control speed to be reached, the steam inlet valve is opened which allows rated steam flow to the turbine. This arrangement precludes an overspeed trip due to excessive steam flow prior to governor warmup. This bypass line is not dependent upon AC power to operate.
10A-7 Amendment 54
ATTACHMENT 3 ST HL-AE IPo1 STP FSAR PAGE 6 0F l 3.
Question 440.30N With regard to the information in Appendix 5.4A " Cold Shutdown Capability" identify the most limiting single failure with regard to cooldown capability and verify that the statement of Table 5.4A-1 that the auxiliary feedwater storage tank (AFST) " capacity of 500,000 gallons is adequate to support 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at hot standby conditions followed by 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> cooldown to RHR cut in condition with a margin for contingencies" considers this failure.
Response gq The most limiting failure regarding cooldown time is the loss of "A" train AC power, which results in the loss of two steam generator PORVs. RHR cut-in conditions can be achieved with this failure -ee-hours after reactor trip based on maintaining hot standby for four hours followed by a ten hour natural circulation cooldown,;nd th:n :: :!-ht hr2r :::h p:risdi' Approximately 360 -6#7,000 gallons of water would be added to the effective steam generators 54 during this period. ,
and a ld Specifically the AFST sizing considers: 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> at hot standbyA 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> natural circulation cooldown. " hom. .eek peried?' It also considers possible level instrument error, water delivered to a faulted steam generator, water lost through the turbine lube oil cooler, various small system water losses (ie.,
flangeorpumpsealleakage)landamarginagainstvortexformation. The usable volume in the AFST above the suction nozzles is $25,000 gallons.
A *
- Ak PSV ed ( q be g L ~ s,*.g - ,3 y ,
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Vol.2 Q&R 5.4 5N Amendment 54 I
ATTACHMENT 9L. I ST HL AE l?67 STP FSAR PAGE 7 0F 14-Question 440.38N
- o. Demonstrate that the STP ECCS meets 10 CFR Part 50.46 criteria for long term decay heat removal in the event of a small break LOCA of a size such that recirculation would be required but the RCS pressure either remains above the low-head safety injection (LHSI) pump shutoff head or recovers after loss of the secondary heat sink. An examination of Figures 6.3-1 through 6.3-5 does not indicate that the STP ECCS is designed for,high-head recirculation combined with decay heat removal by the RHR heat exchangers, i.e., there are no apparent provisions for routing recircula-tion flow from the RHR heat exchangers to the HHSI pumps. Also, as described in Appendix 5.4. A " Cold Shutdown Capability," the steam genera-tors have a limited supply of safety grade secondary water supply, since there is not a safety grade backup to the auxiliary feedwater storage tank (AFST). Therefore, provide long term analyses for a spectrum of small break LOCAs that demonstrate that decay heat can be adequately removed and the RCS depressurized using only safety grade equipment and water sources, assuming loss of offsite power and the most severe single failure. If credit is taken for operator actions, the STP emergency response guideline (ERG) sequence of operator actions should be followed.
Justify the timing of operator actions if they are less conservative than those recommended in ANSI N 660 for a condition IV event.
- b. In a conference call held on March 6, 1985, the applicant indicated to NRC that for small break LOCAs the combined heat sink capacity of the RWST and the steam generators would provide core cooling for approxi-mately 18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />, after which the reactor containment fan coolers (RCFCs) would provide an adequate heat sink for decay heat removal. No credit is {
taken for heat removal by the RHR heat exchangers. Provide a detailed explanation of the mechanism of energy removal from the RCS after loss of the secondary heat sink and supporting analyses that demonstrate that energy can be adequately removed to meet the acceptance criteria of 10 CFR Part 50.46. We are concerned that for very small break LOCAs (e.g.,
l~ l inch) energy would not be adequately removed from the RCS for a consi-derable period of time after the accident. Thus, WCAP 9600, " Report on Small Break Accidents for Westinghouse NSSS System" June 1979, indicates that for 1 inch breaks the break can remove all the decay heat only after about 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and that prior to that time, auxiliary feedwater is required to maintain the heat sink.
Rosponse ciWL usponse-stM=be providedEMa71Tter'EMmthwent=
.,ds. s- LAAAS X -
i i
Vol.2 Q&R 6.3 8N Amendment 49 f
ATTACHMENT 3-.
ST44L AE 1967 PAGE 8 OF # p, Insert X Large Break LOCA For large break LOCA (breaks greater than 1 sq. ft.) the break will cause a significant Reactor Coolant System (RCS) depressurization. Breaks of this size are not isolable so the sump is used for long term cooling and makeup.
Breaks considered large breaks will have sufficient energy removal through the break to sump flow path to remove decay heat energy. Sufficient make-up capability to keep the core adequately cooled and to meet 10CFR Part 50.46 (b)
(5) requirements is provided. Containment heat removal will be provided in the STP design by both containment fan coolers and low head safety injection (LHSI) recirculation flow which is cooled by the RHR heat exchangers.
Equipment relied upon is fully qualified for the environmental conditions that prevail during the accident.
Small Break LOCA As result of the accident at Three Mile Island Unit 2, Westinghouse performed extensive analyses that focused on the behavior of small break loss of coolant accidents (SBLOCA) for the Westinghouse NSSS. The purpose of the analyses was to demonstrate adequacy of the Westinghouse NSSS design in mitigation and long term recovery from a range of breaks classified as small breaks (less than 1 sq. ft. area).
The results of the analyses were reported in WCAP-9600, " Report on Small Break Accident for Westinghouse NSSS System," dated June 1979. The "Small Break Evaluation Model" at that time consisted of the WFLASH thermal-hydraulic code and the LOCTA fuel rod model. The analyses were performed for generic application using a standard 4-loop Westinghouse designed, a standard 3 loop and standard 2 loop depending on the nature of the study and which plant type was expected to be bounding. The conclusions are applicable for all Westinghouse designs, including STP with exceptions as described in the following.
STP SBLOCA Design Features STP has a three train low pressure SI system consisting of three high head SI (HHSI) pumps, three LHSI pumps, and three accumulators. Each train is aligned to a separate RCS loop. The pressure ranges for the SI pumps follow:
HHSI: 0 - 1445 psig LHSI: 0 283 psig L1/NRC/na
ATTACHMENT L, ST HL AE F767 fA,GE 9 OF lt.
Insert X (Continued)
For recirculation, the UlSI and HHSI pumps take suction directly from the sump.
The UlSI pump flow passes through the RHR heat exchanger and is cooled before entering the RCS.
The plant has three motor driven auxiliary feedwater (AW) pumps and one turbine driven auxiliary feedwater pump. The normal system alignment connects each AfW pump directly to one steam generator. The steam does not have a common header, but cross connections exist in the AW lines. The valves in the cross connections are normally closed and fail closed. Two motor driven AW pumps and the turbine driven AW pump are required operable by the Technical Specifications.
The Auxiliary Feedwater Storage Tank (AFST) has a useable capacity of
. 525,000 gallons.
Ncn-safety grade sources of condensate grade make-up to the AFST are:
o Demineralized Water Storage Tank - One 1,000,000 gallon storage tank shared between units, o Secondary Hake up Tank - One 300,000 gallon storage tank per unit o Condenser Hotwell - about 100,000 gallons per unit While these tank volumes are not covered by Technical Specifications or other administrative controls, it would be very improbable to have less than 500,000 gallons of condensate grade water available for each unit.
The limiting single failure for the STP design will result in the loss of one train of safety injection (1 UlSI and HHSI pump) and one AW pump.
Since one A W pump is allowed out-of-service for maintenance, this will result in the ability to feed two steam generators.
The STP design provides means to remove energy through the steam generators (AFW and atmospheric relief valves), through Containment steam condensation (fan coolers) and through the RHR heat exchangers (UISI pumps and RHR heat exchane,ers). In this way energy is removed from containment sump water (RHR herc exchangers) so that relatively cool water will be continued to be supplied as make-up and for decay heat removal.
For all break sizes, heat is removed from the core by the break and steam generators. AW is required for secondary inventory and heat removal until the break is able to remove all the decay heat or the RHR System is placed in operation. The break removes energy from the RCS because the makeup water from the RWST is relatively cold and can absorb energy before exiting the RCS.
The WCAP-9600 analyses with consideration of STP design features and STP analyses of long term cooling discussed in the report titled "Long Term Cooling Analysis for South Texas Project" demonstrate decay heat removal capability for SBLOCA. The Long Term Cooling Report was transmitted in HL&P letter ST HL-AE-1767 dated September 30, 1986.
Ll/NRC/na
ATTACHMENT 2.,
ST HL AE I%q PAGElOOF : A Insert X (Continued)
SBLOCA Response The initiating event is the break. If the break is 3/8" or less equivalent diameter and the charging system and feedwater system are available, the event is classified as a leak since normal charging flow would be sufficient to keep up with leak flow without a significant RCS depressurization. There would not be an automatic reactor trip or safety injection signal.
For breaks larger than 3/8", automatic reactor trip and safety injection will occur due to RCS depressurization caused by the loss of primary inventory. After reactor trip and safety injection initiation, safety injection pump flow provides makeup to the RCS and maximum peak clad temperature will remain below 10CFR50.46 Appendix K criteria.
.For breaks greater than 3/8" and less than 1.5", SI flow can match break flow so no significant RCS depressurization or core uncovery will occur. At the point where SI flow matches break flow, the mitigation phase of the accident ends and a long term decay heat removal phase begins. The operator will cool down and depressurize to below the shutoff head pressure of the GSI pumps (283 psig). This will be accomplished using the steam generator PORVs for -
cooldown and pressurizer PORVs in combination with HHSI flow termination for depressurization. The detailed actions will be provided in the STP Emergency Procedures which are based on the WOG Emergency Response Guidelines. The RHRS will be available to provide heat removal at RCS pressures below 350 psig and temperatures below 350 F. Adequate long term decay heat removal will be provided by UlSI pump flow through an RHR heat exchanger in addition to RHRS operation.
For breaks from 1.5" to 4", the operator will cool down and depressurize the RCS to a pressure below the shutoff head pressure of the GSI. The combined heat sink capacity of the Refueling Water Storage Tank and the steam generators would provide core cooling until the containment fan coolers and the RHR heat exchangers via GSI pumps provide an adequate heat sink for decay removal.
For breaks greater than 4", the decay heat will be removed by the break and the containment fan coolers and the RHR heat exchangers via GSI pumps. No operator action is required.
For isolable breaks, the operator will cool down and depressurize the RCS via a sufficient, quantity of auxiliary feedwater to RHRS cut-in conditions of 350 psig and 350 F. Adequate long term decay heat removal will then be provided via the Residual Heat Removal System.
L1/NRC/na Revised 5/22/86
ATTACHMENT A.
ST HL AE l%9 PAGE II 0F ;L STP FSAR Question 440.39N
- a. It is stated in 10 CFR Part 50.46(b)(5) that, for long term cooling. "the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long term radioactivity remaining in the core." In order to assure this, heat removal for this extended period must* utilize equipment that is fully qualified for the environmental conditions that prevail during the accident. Please demonstrate that decay heat can be removed from the STP core with qualified equipment only, following all sizes of LOCAs, including all LOCAs which could be subsequently isolated by the operator. Include consideration of the post-LOCA cooldown period in your response, and the fact that for isolated LOCAs, the sump would not be available for lon5 term cooling,
- b. Discuss whether the RHR pumps are qualified for the environmental effects of the large and small break LOCAs and steam line breaks. If the RER pamps are not qualified discuss how long term mitigation of these acci-dents would be accomplished.
Response
m.s+-wH u w iv -
- N m dment.
- a. See the response to NRC Question 440.038N.
- b. The RHR pumps are qualified for the containment environment following a DBA, including small LOCA and secondary breaks. The NRC was notified regarding the qualification of the RHR pumps via ST-HL-AE-1684 dated June 17, 1986.
Vol.2 Q&R 6.3 9N Amendment 49
ATTACHMENT GL ST HL AE 1967 PAGEla OF19-EMERGENCY CORE COOLING SYSTEMS i
3/4.5.6 RHR LIMITING CONDITION FOR OPERATION 3.5.6 Three independent Residual Heat Removal (RHR) loops shall be OPERABLE with each loop comprised of:
- a. One OPERABLE RHR pump,
- b. one OPERABLE RHR heat exchanger, and
- c. one OPERABLE flowpath capable of taking suction from its associated RCS, hot leg and discharging to its associated RCS cold leg.
APPLICABILITY: Modes 1, 2 and 3 ACTION:
- a. With one RHR loop inoperable, restore the required loop to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- b. With two RHR loops inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
1 c. With three RHR loops inoperable, immediately initiate corrective action to restore at least one RHR loop to OPERABLE status as soon as possible.
SURVEILLANCE REQUIREMENTS 3
4.5.6 Each RHR loop shall be demonstrated OPERABLE pursuant to the requirements of specification 4.0.5 l
South Texas Project 3/4 5 11 Sept. 30, 1986
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Enclosure 1 I
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