|
---|
Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217J3141999-10-15015 October 1999 Requests Emergency Publication of Document Entitled South Carolina Electric & Gas Co;Vc Summer Nuclear Station,Environ Assessment Transmitted on 991015 to Ofc of Fr for Publication ML20217J3281999-10-15015 October 1999 Forwards Copy of Environ Assessment & Finding of No Significant Impact Re Application for Exemption from Requiremets of 10CFR50,Section 50.60(a) for VC Summer Nuclear Station ML20217F8851999-10-0808 October 1999 Forwards Insp Rept 50-395/99-06 on 990801-0911.One Violation Occurred Being Treated as NCV RC-99-0192, Forwards Updated 1999 ECCS Evaluation Model Revs Rept for VC Summer Nuclear Station.Rept Is Being Submitted Pursuant to 10CFR50.46,which Requires Licensees to Notify NRC of Corrections to or Changes in ECCS Evaluation Models1999-09-28028 September 1999 Forwards Updated 1999 ECCS Evaluation Model Revs Rept for VC Summer Nuclear Station.Rept Is Being Submitted Pursuant to 10CFR50.46,which Requires Licensees to Notify NRC of Corrections to or Changes in ECCS Evaluation Models RC-99-0181, Forwards Anticipated Schedule for Operator Licensing Examinations.Sce&G Requests That NRC Prepare Examinations Stated on Attachment1999-09-21021 September 1999 Forwards Anticipated Schedule for Operator Licensing Examinations.Sce&G Requests That NRC Prepare Examinations Stated on Attachment ML20212C5091999-09-15015 September 1999 Forwards Anticipated Schedule for Operator Licensing Exams for Sce&G.Util Requests That NRC Prepare Exams on Encl RC-99-0184, Submits Seven Requests for Using Alternatives to Requirements of ASME Code,Section XI Re Subsection IWE & Iwl Insps to Be Performed at Vsns.Proposed Alternatives Will Provide Acceptable Level of Quality & Safety1999-09-15015 September 1999 Submits Seven Requests for Using Alternatives to Requirements of ASME Code,Section XI Re Subsection IWE & Iwl Insps to Be Performed at Vsns.Proposed Alternatives Will Provide Acceptable Level of Quality & Safety ML20212A8341999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20212A6951999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20211N4301999-09-0808 September 1999 Discusses Proposed Meeting to Discuss Kaowool Fire Barriers. Staff Requesting That Affected Licensees Take Issue on Voluntary Initative & Propose Approach for Resolving Issues ML20211Q8911999-09-0101 September 1999 Sumbits Summary of Training Managers Conference on Recent Changes to Operator Licensing Program.Meeting Covered Changes to Regulations,Exam Stds,New Insp Program & Other Training Issues.List of Attendees Encl RC-99-0177, Forwards Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, IAW Section 6.9.1.111999-08-31031 August 1999 Forwards Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, IAW Section 6.9.1.11 RC-99-0173, Requests That Info Listed in Rvid,Version 2,be Amended to Reflect Date for VC Summer Nuclear Station,As Marked in Encl to Ltr1999-08-31031 August 1999 Requests That Info Listed in Rvid,Version 2,be Amended to Reflect Date for VC Summer Nuclear Station,As Marked in Encl to Ltr ML20211L5181999-08-30030 August 1999 Forwards Insp Rept 50-395/99-05 on 990620-0731.One Violation Identified & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML20211H2481999-08-25025 August 1999 Forwards Four Controlled Copies of Amend 43 to Physcial Security Plan. Summary of Plan Changes, Are Included as Part of Each Controlled Copy.Encls Withheld Per 10CFR73.21 05000395/LER-1999-004, Submits Suppl 1 to LER 99-004-00 Re Discovery of Several Fuel Assembly Top Nozzle Holdown Screws Which Had Failed. Root Cause Will Not Be Completed by 990829,as Committed.W Analysis Will Be Issued After Fall Outages Are Complet1999-08-24024 August 1999 Submits Suppl 1 to LER 99-004-00 Re Discovery of Several Fuel Assembly Top Nozzle Holdown Screws Which Had Failed. Root Cause Will Not Be Completed by 990829,as Committed.W Analysis Will Be Issued After Fall Outages Are Completed RC-99-0171, Notifies NRC of Intent Re Submittal of Application to Renew OL of Vcs.Preparatory Work Has Begun to Develop Application for License Renewal to Be Submitted After 020806 Contingent Upon Final Approval of Board of Directors1999-08-23023 August 1999 Notifies NRC of Intent Re Submittal of Application to Renew OL of Vcs.Preparatory Work Has Begun to Develop Application for License Renewal to Be Submitted After 020806 Contingent Upon Final Approval of Board of Directors RC-99-0152, Seeks Exemption Under 10CFR0.12a(2)ii from 10CFR50,App G Requirements to Establish pressure-temperature Limits Curves Using Methodology Presented in 1989 ASME Section Xi,App G1999-08-19019 August 1999 Seeks Exemption Under 10CFR0.12a(2)ii from 10CFR50,App G Requirements to Establish pressure-temperature Limits Curves Using Methodology Presented in 1989 ASME Section Xi,App G RC-99-0164, Forwards semi-annual Fitness for Duty Rept from 990101 to 990630 for VC Summer Nuclear Station,Iaw 10CFR26.71(d)1999-08-17017 August 1999 Forwards semi-annual Fitness for Duty Rept from 990101 to 990630 for VC Summer Nuclear Station,Iaw 10CFR26.71(d) ML20210Q4851999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006 at VC Summer.Requests Info Re Individuals Who Will Take Exam,Personnel Who Will Have Access to Exam.Sample Registration Ltr Encl ML20210R5501999-08-0505 August 1999 Ack Receipt of 990707 Response to NCVs Identified on 990607 Re Activities Conducted at VC Summer.Informs That After Consideration of Basis for Denial of NCV 50-395/99-03, Concluded,For Reasons Stated,That NCV Occurred RC-99-0156, Forwards Rev 1 to VC Summer Nuclear Station COLR for Cycle 12, IAW TS Section 6.9.1.11.Sections 2.1 & 3.0 Were Added to Include Beacon Tsm1999-08-0404 August 1999 Forwards Rev 1 to VC Summer Nuclear Station COLR for Cycle 12, IAW TS Section 6.9.1.11.Sections 2.1 & 3.0 Were Added to Include Beacon Tsm RC-99-0147, Submits Attached Request for Relief from Performing SG PORV Strike Time Testing to Acceptance Criteria of Asme/Ansi OMa-19881999-07-26026 July 1999 Submits Attached Request for Relief from Performing SG PORV Strike Time Testing to Acceptance Criteria of Asme/Ansi OMa-1988 ML20210B7451999-07-22022 July 1999 Informs That as Result of Staff Review of Licensee Responses to GL 92-01,rev 1 & Rev 1,suppl 1,staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 ML20210E3771999-07-16016 July 1999 Forwards Insp Rept 50-395/99-04 on 990509-0619.One Violation Being Treated as Noncited Violation RC-99-0127, Estimates Submittal of Eleven Licensing Actions in Fy 2000. Based on Statistical Estimates of Past Licensing Actions, Number of Licensing Actions in Fy 2001 Should Be Approx Ten, in Response to AL 99-021999-07-0707 July 1999 Estimates Submittal of Eleven Licensing Actions in Fy 2000. Based on Statistical Estimates of Past Licensing Actions, Number of Licensing Actions in Fy 2001 Should Be Approx Ten, in Response to AL 99-02 RC-99-0129, Provides Response to non-cited Violations Noted in Insp Rept 50-395/99-03.C/As:concluded That Cask Loading Pit Inaccessible & Duration of Dose Rates on Operating Floor of Fhb So Short That High Radiation Area Did Not Exist1999-07-0707 July 1999 Provides Response to non-cited Violations Noted in Insp Rept 50-395/99-03.C/As:concluded That Cask Loading Pit Inaccessible & Duration of Dose Rates on Operating Floor of Fhb So Short That High Radiation Area Did Not Exist RC-99-0131, Forwards Rev 9 to VC Summer Nuclear Station Safeguards Contingency Plan,Per 10CFR50.54(p).Encl Withheld1999-07-0707 July 1999 Forwards Rev 9 to VC Summer Nuclear Station Safeguards Contingency Plan,Per 10CFR50.54(p).Encl Withheld ML20210B7111999-07-0606 July 1999 Provides Summary of 990701 Meeting with Sce&G in Atlanta, Georgia Re Recent Virgil C Summer Refueling Outage & Other Items of Interest.List of Meeting Attendees & Licensee Presentation Handouts Encl RC-99-0114, Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps, Under Oath or Affirmation1999-06-30030 June 1999 Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps, Under Oath or Affirmation ML20195H5861999-06-0707 June 1999 Confirms 990604 Telcon Between J Proper & R Haag Re Meeting Scheduled for 990701 in Atlanta,Ga,To Discuss Plant Refueling Outage & Items of Interest ML20207H5241999-06-0707 June 1999 Forwards Insp Rept 50-395/99-03 on 990328-0508.Six Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20207D1881999-05-28028 May 1999 Informs That Effective 990524,K Cotton Assigned as Project Manager,Project Directorate II-1,for Virgil C Summer Nuclear Station 05000395/LER-1999-006, Forwards LER 99-006-00,describing Identified Safety Hazard with GE 7.2kV Magne-Blast Circuit Breakers.Event Is Being Reported Per 10CFR21.21a(1)1999-05-17017 May 1999 Forwards LER 99-006-00,describing Identified Safety Hazard with GE 7.2kV Magne-Blast Circuit Breakers.Event Is Being Reported Per 10CFR21.21a(1) RC-99-0104, Forwards Amend 17 to Training & Qualification Plan, Under Provisions of 10CFR50.54(p).Summary of Plan Changes Is Included as Part of Controlled Copy1999-05-13013 May 1999 Forwards Amend 17 to Training & Qualification Plan, Under Provisions of 10CFR50.54(p).Summary of Plan Changes Is Included as Part of Controlled Copy RC-99-0105, Forwards Copy of Sce&G Co 1998 Annual Financial Rept & Sc Public Service Authority 1998 Annual Financial Rept, for VC Summer Nuclear Station1999-05-13013 May 1999 Forwards Copy of Sce&G Co 1998 Annual Financial Rept & Sc Public Service Authority 1998 Annual Financial Rept, for VC Summer Nuclear Station 05000395/LER-1999-005, Forwards LER 99-005-00 for VC Summer Nuclear Station.Rept Describes Potential Condition for Exceeding Vsns Plant Design Basis Due to Submergence Qualification Issues for Certain ESF Components1999-05-12012 May 1999 Forwards LER 99-005-00 for VC Summer Nuclear Station.Rept Describes Potential Condition for Exceeding Vsns Plant Design Basis Due to Submergence Qualification Issues for Certain ESF Components ML20206L5121999-05-11011 May 1999 Informs That NRC Reorganized,Effective 990328.Reorganization Chart Encl ML20206P5771999-05-0707 May 1999 Informs That During 980519 Telcon Between T Matlosz & G Hopper,Arrangements Were Made for Administration of Licensing Exam at Virgil C Summer Nuclear Station During Wk of 990927 RC-99-0080, Submits Supplemental Info Re 970128 Response to NRC GL 96-06 Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Addl Analysis & Manpower Expenditure Involved Not Cost Effective1999-05-0606 May 1999 Submits Supplemental Info Re 970128 Response to NRC GL 96-06 Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Addl Analysis & Manpower Expenditure Involved Not Cost Effective RC-99-0097, Forwards Sce&G Cycle 12 COLR, IAW TS Section 6.9.1.111999-05-0606 May 1999 Forwards Sce&G Cycle 12 COLR, IAW TS Section 6.9.1.11 RC-99-0092, Informs That Util Has Reviewed Proposed Notice of Rulemaking & Fully Endorse Comments Prepared & Submitted on Behalf of Commercial Nuclear Power Industry by NEI1999-05-0303 May 1999 Informs That Util Has Reviewed Proposed Notice of Rulemaking & Fully Endorse Comments Prepared & Submitted on Behalf of Commercial Nuclear Power Industry by NEI RC-99-0090, Submits Special Rept (Spr 1999-003) Re Completion of ISI of SG Tubes,Indicating Number of Tubes Plugged or Repaired in Each Generator,Per TS 4.4.5.5.a & Section 4.4.5.5.b1999-04-29029 April 1999 Submits Special Rept (Spr 1999-003) Re Completion of ISI of SG Tubes,Indicating Number of Tubes Plugged or Repaired in Each Generator,Per TS 4.4.5.5.a & Section 4.4.5.5.b ML20206E1681999-04-29029 April 1999 Informs That FERC & NRC Will Conduct Category I Svc Water Pond (Swp) Dam Insp at Facility on 990610 ML20206P5021999-04-26026 April 1999 Forwards Insp Rept 50-395/99-02 on 990214-0327.One Violation of NRC Requirements Occurred & Being Treated as non-cited Violation,Consistent with App C of Enforcement Policy ML20205M0431999-04-13013 April 1999 Eighth Partial Response to FOIA Request for Records.App Q & R Records Encl & Being Made Available in PDR 05000395/LER-1999-002, Forwards LER 99-002-00 Re Condition for Exceeding Vsns Design Basis During Surveillance Testing Utilizing Certain ECCS Valves.Simplified Flow Diagram Included to Identify Configurations Discussed by Rept Encl1999-04-12012 April 1999 Forwards LER 99-002-00 Re Condition for Exceeding Vsns Design Basis During Surveillance Testing Utilizing Certain ECCS Valves.Simplified Flow Diagram Included to Identify Configurations Discussed by Rept Encl ML20205T2311999-04-0909 April 1999 Informs That on 990318,A Koon & Ho Christensen Confirmed Initial Operator Licensing Exam Scheduled for Y2K.Initial Exam Date Schedules for Wk of 000807 for Approx Eight Candidates ML20205G4181999-04-0101 April 1999 Advises That 970725 Application & Affidavit Which Submitted, WCAP-14932, Probabilistic & Economic Evaluation of Reactor Vessel Closure Head Penetration Integrity for Plant, Will Be Withheld from Public Disclosure,Per 10CFR2.790(a)(4) RC-99-0078, Submits Summary of Present Levels of Property Insurance & Cash Flow Statement for VC Summer Nuclear Station,Per 10CFR50.54(w)(3) & 10CFR140.21(e)1999-04-0101 April 1999 Submits Summary of Present Levels of Property Insurance & Cash Flow Statement for VC Summer Nuclear Station,Per 10CFR50.54(w)(3) & 10CFR140.21(e) 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARRC-99-0192, Forwards Updated 1999 ECCS Evaluation Model Revs Rept for VC Summer Nuclear Station.Rept Is Being Submitted Pursuant to 10CFR50.46,which Requires Licensees to Notify NRC of Corrections to or Changes in ECCS Evaluation Models1999-09-28028 September 1999 Forwards Updated 1999 ECCS Evaluation Model Revs Rept for VC Summer Nuclear Station.Rept Is Being Submitted Pursuant to 10CFR50.46,which Requires Licensees to Notify NRC of Corrections to or Changes in ECCS Evaluation Models RC-99-0181, Forwards Anticipated Schedule for Operator Licensing Examinations.Sce&G Requests That NRC Prepare Examinations Stated on Attachment1999-09-21021 September 1999 Forwards Anticipated Schedule for Operator Licensing Examinations.Sce&G Requests That NRC Prepare Examinations Stated on Attachment ML20212C5091999-09-15015 September 1999 Forwards Anticipated Schedule for Operator Licensing Exams for Sce&G.Util Requests That NRC Prepare Exams on Encl RC-99-0184, Submits Seven Requests for Using Alternatives to Requirements of ASME Code,Section XI Re Subsection IWE & Iwl Insps to Be Performed at Vsns.Proposed Alternatives Will Provide Acceptable Level of Quality & Safety1999-09-15015 September 1999 Submits Seven Requests for Using Alternatives to Requirements of ASME Code,Section XI Re Subsection IWE & Iwl Insps to Be Performed at Vsns.Proposed Alternatives Will Provide Acceptable Level of Quality & Safety RC-99-0177, Forwards Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, IAW Section 6.9.1.111999-08-31031 August 1999 Forwards Rev 2 to VC Summer Nuclear Station,Colr for Cycle 12, IAW Section 6.9.1.11 RC-99-0173, Requests That Info Listed in Rvid,Version 2,be Amended to Reflect Date for VC Summer Nuclear Station,As Marked in Encl to Ltr1999-08-31031 August 1999 Requests That Info Listed in Rvid,Version 2,be Amended to Reflect Date for VC Summer Nuclear Station,As Marked in Encl to Ltr ML20211H2481999-08-25025 August 1999 Forwards Four Controlled Copies of Amend 43 to Physcial Security Plan. Summary of Plan Changes, Are Included as Part of Each Controlled Copy.Encls Withheld Per 10CFR73.21 05000395/LER-1999-004, Submits Suppl 1 to LER 99-004-00 Re Discovery of Several Fuel Assembly Top Nozzle Holdown Screws Which Had Failed. Root Cause Will Not Be Completed by 990829,as Committed.W Analysis Will Be Issued After Fall Outages Are Complet1999-08-24024 August 1999 Submits Suppl 1 to LER 99-004-00 Re Discovery of Several Fuel Assembly Top Nozzle Holdown Screws Which Had Failed. Root Cause Will Not Be Completed by 990829,as Committed.W Analysis Will Be Issued After Fall Outages Are Completed RC-99-0171, Notifies NRC of Intent Re Submittal of Application to Renew OL of Vcs.Preparatory Work Has Begun to Develop Application for License Renewal to Be Submitted After 020806 Contingent Upon Final Approval of Board of Directors1999-08-23023 August 1999 Notifies NRC of Intent Re Submittal of Application to Renew OL of Vcs.Preparatory Work Has Begun to Develop Application for License Renewal to Be Submitted After 020806 Contingent Upon Final Approval of Board of Directors RC-99-0152, Seeks Exemption Under 10CFR0.12a(2)ii from 10CFR50,App G Requirements to Establish pressure-temperature Limits Curves Using Methodology Presented in 1989 ASME Section Xi,App G1999-08-19019 August 1999 Seeks Exemption Under 10CFR0.12a(2)ii from 10CFR50,App G Requirements to Establish pressure-temperature Limits Curves Using Methodology Presented in 1989 ASME Section Xi,App G RC-99-0164, Forwards semi-annual Fitness for Duty Rept from 990101 to 990630 for VC Summer Nuclear Station,Iaw 10CFR26.71(d)1999-08-17017 August 1999 Forwards semi-annual Fitness for Duty Rept from 990101 to 990630 for VC Summer Nuclear Station,Iaw 10CFR26.71(d) RC-99-0156, Forwards Rev 1 to VC Summer Nuclear Station COLR for Cycle 12, IAW TS Section 6.9.1.11.Sections 2.1 & 3.0 Were Added to Include Beacon Tsm1999-08-0404 August 1999 Forwards Rev 1 to VC Summer Nuclear Station COLR for Cycle 12, IAW TS Section 6.9.1.11.Sections 2.1 & 3.0 Were Added to Include Beacon Tsm RC-99-0147, Submits Attached Request for Relief from Performing SG PORV Strike Time Testing to Acceptance Criteria of Asme/Ansi OMa-19881999-07-26026 July 1999 Submits Attached Request for Relief from Performing SG PORV Strike Time Testing to Acceptance Criteria of Asme/Ansi OMa-1988 RC-99-0129, Provides Response to non-cited Violations Noted in Insp Rept 50-395/99-03.C/As:concluded That Cask Loading Pit Inaccessible & Duration of Dose Rates on Operating Floor of Fhb So Short That High Radiation Area Did Not Exist1999-07-0707 July 1999 Provides Response to non-cited Violations Noted in Insp Rept 50-395/99-03.C/As:concluded That Cask Loading Pit Inaccessible & Duration of Dose Rates on Operating Floor of Fhb So Short That High Radiation Area Did Not Exist RC-99-0131, Forwards Rev 9 to VC Summer Nuclear Station Safeguards Contingency Plan,Per 10CFR50.54(p).Encl Withheld1999-07-0707 July 1999 Forwards Rev 9 to VC Summer Nuclear Station Safeguards Contingency Plan,Per 10CFR50.54(p).Encl Withheld RC-99-0127, Estimates Submittal of Eleven Licensing Actions in Fy 2000. Based on Statistical Estimates of Past Licensing Actions, Number of Licensing Actions in Fy 2001 Should Be Approx Ten, in Response to AL 99-021999-07-0707 July 1999 Estimates Submittal of Eleven Licensing Actions in Fy 2000. Based on Statistical Estimates of Past Licensing Actions, Number of Licensing Actions in Fy 2001 Should Be Approx Ten, in Response to AL 99-02 RC-99-0114, Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps, Under Oath or Affirmation1999-06-30030 June 1999 Submits Response to GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps, Under Oath or Affirmation 05000395/LER-1999-006, Forwards LER 99-006-00,describing Identified Safety Hazard with GE 7.2kV Magne-Blast Circuit Breakers.Event Is Being Reported Per 10CFR21.21a(1)1999-05-17017 May 1999 Forwards LER 99-006-00,describing Identified Safety Hazard with GE 7.2kV Magne-Blast Circuit Breakers.Event Is Being Reported Per 10CFR21.21a(1) RC-99-0104, Forwards Amend 17 to Training & Qualification Plan, Under Provisions of 10CFR50.54(p).Summary of Plan Changes Is Included as Part of Controlled Copy1999-05-13013 May 1999 Forwards Amend 17 to Training & Qualification Plan, Under Provisions of 10CFR50.54(p).Summary of Plan Changes Is Included as Part of Controlled Copy RC-99-0105, Forwards Copy of Sce&G Co 1998 Annual Financial Rept & Sc Public Service Authority 1998 Annual Financial Rept, for VC Summer Nuclear Station1999-05-13013 May 1999 Forwards Copy of Sce&G Co 1998 Annual Financial Rept & Sc Public Service Authority 1998 Annual Financial Rept, for VC Summer Nuclear Station 05000395/LER-1999-005, Forwards LER 99-005-00 for VC Summer Nuclear Station.Rept Describes Potential Condition for Exceeding Vsns Plant Design Basis Due to Submergence Qualification Issues for Certain ESF Components1999-05-12012 May 1999 Forwards LER 99-005-00 for VC Summer Nuclear Station.Rept Describes Potential Condition for Exceeding Vsns Plant Design Basis Due to Submergence Qualification Issues for Certain ESF Components RC-99-0097, Forwards Sce&G Cycle 12 COLR, IAW TS Section 6.9.1.111999-05-0606 May 1999 Forwards Sce&G Cycle 12 COLR, IAW TS Section 6.9.1.11 RC-99-0080, Submits Supplemental Info Re 970128 Response to NRC GL 96-06 Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Addl Analysis & Manpower Expenditure Involved Not Cost Effective1999-05-0606 May 1999 Submits Supplemental Info Re 970128 Response to NRC GL 96-06 Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions. Addl Analysis & Manpower Expenditure Involved Not Cost Effective RC-99-0092, Informs That Util Has Reviewed Proposed Notice of Rulemaking & Fully Endorse Comments Prepared & Submitted on Behalf of Commercial Nuclear Power Industry by NEI1999-05-0303 May 1999 Informs That Util Has Reviewed Proposed Notice of Rulemaking & Fully Endorse Comments Prepared & Submitted on Behalf of Commercial Nuclear Power Industry by NEI RC-99-0090, Submits Special Rept (Spr 1999-003) Re Completion of ISI of SG Tubes,Indicating Number of Tubes Plugged or Repaired in Each Generator,Per TS 4.4.5.5.a & Section 4.4.5.5.b1999-04-29029 April 1999 Submits Special Rept (Spr 1999-003) Re Completion of ISI of SG Tubes,Indicating Number of Tubes Plugged or Repaired in Each Generator,Per TS 4.4.5.5.a & Section 4.4.5.5.b 05000395/LER-1999-002, Forwards LER 99-002-00 Re Condition for Exceeding Vsns Design Basis During Surveillance Testing Utilizing Certain ECCS Valves.Simplified Flow Diagram Included to Identify Configurations Discussed by Rept Encl1999-04-12012 April 1999 Forwards LER 99-002-00 Re Condition for Exceeding Vsns Design Basis During Surveillance Testing Utilizing Certain ECCS Valves.Simplified Flow Diagram Included to Identify Configurations Discussed by Rept Encl RC-99-0078, Submits Summary of Present Levels of Property Insurance & Cash Flow Statement for VC Summer Nuclear Station,Per 10CFR50.54(w)(3) & 10CFR140.21(e)1999-04-0101 April 1999 Submits Summary of Present Levels of Property Insurance & Cash Flow Statement for VC Summer Nuclear Station,Per 10CFR50.54(w)(3) & 10CFR140.21(e) RC-99-0066, Submits Rept of Status of Decommissioning Funding (RR-1950), for Vsns Per 10CFR50.751999-03-31031 March 1999 Submits Rept of Status of Decommissioning Funding (RR-1950), for Vsns Per 10CFR50.75 ML20205B9981999-03-29029 March 1999 Informs That Authority & Sce&G Has Ownership Interests of one-third & two-thirds,respectively in VC Summer Nuclear Station.Operating License Scheduled to Expire in 2022.Rept Addresses Decommissioning Cost Estimates & Financing RC-99-0054, Forwards Rev 2 to VC Summer Nuclear Station Training Simulator Quadrennial Certification Rept,1996-99, Per 10CFR55.45(b)(5)(ii)1999-03-22022 March 1999 Forwards Rev 2 to VC Summer Nuclear Station Training Simulator Quadrennial Certification Rept,1996-99, Per 10CFR55.45(b)(5)(ii) RC-99-0053, Requests That Implementation Date of Proposed TS Change Request Re Best Estimate Analyzer for Core Operations - Nuclear (Beacon) Be Extended. Util Requests 120 Day Time Frame to Perform Initial Beacon Calibrs During Cycle 121999-03-22022 March 1999 Requests That Implementation Date of Proposed TS Change Request Re Best Estimate Analyzer for Core Operations - Nuclear (Beacon) Be Extended. Util Requests 120 Day Time Frame to Perform Initial Beacon Calibrs During Cycle 12 RC-99-0048, Informs That Util Has Implemented Policy That Requires All Personnel Granted Unescorted Access to Vsns Satisfactorily Complete Test on Site Specific Info1999-03-10010 March 1999 Informs That Util Has Implemented Policy That Requires All Personnel Granted Unescorted Access to Vsns Satisfactorily Complete Test on Site Specific Info ML20207J5661999-02-16016 February 1999 Requests That Proprietary Rev 1 to WCAP-14932 Re Rv Closure Head Penetrations Integrity for VC Summer Nuclear Plant,Be Withheld from Public Disclosure,Per 10CFR2.790(b)(4) RC-99-0026, Provides Response to NRC RAI Re TS Change Request Re Best Estimate Analyzer for Core Operations - Nuclear1999-02-0505 February 1999 Provides Response to NRC RAI Re TS Change Request Re Best Estimate Analyzer for Core Operations - Nuclear RC-99-0023, Informs That in Response to GL 97-06,SCE&G Informed NRC of Plan to Perform Secondary Side Examination Scheduled for Refueling Outage RF-11.SCE&G Has Decided to Defer Secondary Side Insp of Sg.Reasons for Change of Plan Listed1999-02-0101 February 1999 Informs That in Response to GL 97-06,SCE&G Informed NRC of Plan to Perform Secondary Side Examination Scheduled for Refueling Outage RF-11.SCE&G Has Decided to Defer Secondary Side Insp of Sg.Reasons for Change of Plan Listed 05000395/LER-1998-009, Forwards LER 98-009-01 for VC Summer Nuclear Station.Rept Describes Unanalyzed Condition for non-safety Related Component for Which All Failure Mechanisms Had Not Been Evaluated1999-01-28028 January 1999 Forwards LER 98-009-01 for VC Summer Nuclear Station.Rept Describes Unanalyzed Condition for non-safety Related Component for Which All Failure Mechanisms Had Not Been Evaluated RC-99-0015, Forwards Amend 16 to Training & Qualification Plan,Per 10CFR50.54(p).Summary of Changes,Encl1999-01-22022 January 1999 Forwards Amend 16 to Training & Qualification Plan,Per 10CFR50.54(p).Summary of Changes,Encl RC-99-0005, Responds to 980908 RAI Re GL 97-01, Degradation of Control Rod Drive Mechanism Nozzle & Other Vessel Closure Head Penetrations1999-01-15015 January 1999 Responds to 980908 RAI Re GL 97-01, Degradation of Control Rod Drive Mechanism Nozzle & Other Vessel Closure Head Penetrations RC-98-0225, Forwards Rev 41 to EP-100, Radiation Emergency Plan. List of Changes by Page Number Affected by Rev 41 Also Encl1998-12-14014 December 1998 Forwards Rev 41 to EP-100, Radiation Emergency Plan. List of Changes by Page Number Affected by Rev 41 Also Encl RC-98-0226, Forwards Amend 42 to Psp.Changes Do Not Degrade Safeguards Effectiveness in PSP or Safeguards Contingency Plan,As Described in 10CFR50.54(p).Without Encl1998-12-14014 December 1998 Forwards Amend 42 to Psp.Changes Do Not Degrade Safeguards Effectiveness in PSP or Safeguards Contingency Plan,As Described in 10CFR50.54(p).Without Encl RC-98-0216, Requests Extension of Response Period to 990115 to Respond to NRC 980908 RAI Re GL 97-01, Degradation of CRDM Nozzle & Other Vessel Closure Head Penetrations. Util Intends to Utilize Industry Generic RAI Response1998-12-0404 December 1998 Requests Extension of Response Period to 990115 to Respond to NRC 980908 RAI Re GL 97-01, Degradation of CRDM Nozzle & Other Vessel Closure Head Penetrations. Util Intends to Utilize Industry Generic RAI Response RC-98-0189, Provides Assessment Results of GL 98-02, Loss of Rc Inventory & Associated Potential for Loss of Emergency Mitigation Functions While in Shutdown Condition, Per 10CFR50.54f1998-11-24024 November 1998 Provides Assessment Results of GL 98-02, Loss of Rc Inventory & Associated Potential for Loss of Emergency Mitigation Functions While in Shutdown Condition, Per 10CFR50.54f RC-98-0207, Forwards 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment1998-11-11011 November 1998 Forwards 120-day Response to NRC GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment RC-98-0177, Informs That Sce&G Will Classify as Moderate Any Stratification Condition That Results in Total Cuf,Based on Design Basis Values Plus Any Contribution from Stratification,Of Between 0.1 & 0.71998-11-0909 November 1998 Informs That Sce&G Will Classify as Moderate Any Stratification Condition That Results in Total Cuf,Based on Design Basis Values Plus Any Contribution from Stratification,Of Between 0.1 & 0.7 RC-98-0194, Provides Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of SR Movs1998-11-0202 November 1998 Provides Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of SR Movs RC-98-0202, Forwards Response to RAI Re Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions1998-10-30030 October 1998 Forwards Response to RAI Re Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions RC-98-0186, Expresses Appreciation for Opportunity to Present Topical Rept TR-104965, On-Line Monitoring of Instrument Channel Performance1998-10-26026 October 1998 Expresses Appreciation for Opportunity to Present Topical Rept TR-104965, On-Line Monitoring of Instrument Channel Performance RC-98-0185, Forwards non-proprietary Trs,Including Rev 0 to WCAP-15101, Analysis of Capsule W from Sce&G VC Summer Unit 1 Rv Radiation Surveillance Program & Rev 0 to WCAP-15103, Evaluation of PTS for VC Summer Unit 11998-10-0909 October 1998 Forwards non-proprietary Trs,Including Rev 0 to WCAP-15101, Analysis of Capsule W from Sce&G VC Summer Unit 1 Rv Radiation Surveillance Program & Rev 0 to WCAP-15103, Evaluation of PTS for VC Summer Unit 1 RC-98-0182, Responds to 980402 RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs1998-10-0808 October 1998 Responds to 980402 RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs RC-98-0178, Provides Comments on SALP Insp Rept 50-395/98-99.Util Ack That Station Can Enhance Future Performance Further with More Focus & Attention on Change Mgt Practices Re Plant & Procedure Changes1998-10-0505 October 1998 Provides Comments on SALP Insp Rept 50-395/98-99.Util Ack That Station Can Enhance Future Performance Further with More Focus & Attention on Change Mgt Practices Re Plant & Procedure Changes 1999-09-28
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059K4071990-09-17017 September 1990 Advises That Author Succeeded Os Bradham as vice-president, Nuclear Operations,Effective 900915.All Correspondence to Util Should Be Sent to Listed Address ML20059J9141990-09-14014 September 1990 Responds to Violations Noted in Insp Rept 50-395/90-21. Violation Occured,However Description of Violation in Error. Corrective Action:Valve XVT02803B Placed in Proper Position & Verification of Emergency Feedwater Sys Lineup Completed ML20059E2821990-08-30030 August 1990 Advises That Programmed Enhancements Per Generic Ltr 88-17 Re Loss of DHR Implemented & Mods Operable ML20059D6911990-08-29029 August 1990 Forwards Technical Rept 90-02, Seismic Activity Near VC Summer Nuclear Station for Apr-June 1990 ML20059B9541990-08-28028 August 1990 Forwards Semiannual Effluent & Waste Disposal Rept for Jan-June 1990, Per 10CFR50.36a & Sections 6.9.1.8 & 6.9.1.9 of Tech Spec ML20056B4901990-08-22022 August 1990 Discusses NRC Bulletin 88-004, Potential Safety-Related Pump Loss. Decision on Future Insp Frequency &/Or Sys Mods Should Be Deferred Until Internal Components Could Be Inspected ML20059A2041990-08-16016 August 1990 Forwards First Semiannual fitness-for-duty Rept from 900103- 0630.Util Pleased W/Program,However,Concerned About Incorrect Test Results for Blind Performance Specimens Received from Roche Labs ML20058P6581990-08-15015 August 1990 Forwards Monthly Operating Rept for Jul 1990 & Rev 13 to ODCM for Virgil C Summer Nuclear Station.Rev Implements Changes Necessary to Permit Removal of Radiological Effluent Specs from Tech Specs as Recommended by Generic Ltr 89-01 ML20058P7681990-08-13013 August 1990 Forwards Mods to 900516 Tech Spec Change Request Re Pressurizer Code Safety Valve Setpoint Tolerance & Mode 3 Exception,Per 900713 Telcon.Changes Do Not Affect Technical Intent of 900516 Submittal Nor Alter Safety Evaluation ML20058N6961990-08-10010 August 1990 Responds to Violations Noted in Insp Rept 50-395/90-18. Corrective Action:Personnel Counseled on Procedural Compliance & Importance of Independent Verification in Maintaining Proper Sys Alignment ML20056A3441990-08-0101 August 1990 Responds to NRC Request for Further Justification Re Relocating Emergency Operations Facility to Corporate Headquarters ML20059A3691990-07-30030 July 1990 Ack Receipt of SALP Rept 50-395/90-11 & Forwards Comments on Rept ML20056A0201990-07-27027 July 1990 Provides Notification That All Actions Re Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment, Completed ML20055H9821990-07-20020 July 1990 Forwards Amend 28 to Physical Security Plan.Amend Withheld (Ref 10CFR73.21) ML17305B7681990-07-19019 July 1990 Responds to NRC NRC Bulletin 90-001, Loss of Fill Oil in Transmitters Mfg by Rosemount. No Transmitters Identified in Item 1 from Suspect Lots Identified by Rosemount ML20044B0241990-07-11011 July 1990 Responds to Violations Noted in Insp Rept 50-395/90-15. Corrective Actions:Relay Rewired & Verified Against Design Drawings ML20055F3631990-07-10010 July 1990 Forwards Rev 1 to Plant Core Operating Limits Rept Cycle 6, as Result of Typo ML20055E0341990-07-0505 July 1990 Forwards Technical Rept 90-1, Seismic Activity Near VC Summer Nuclear Station for Period Jan-Mar 1990 ML20055D4661990-07-0303 July 1990 Forwards Amend 4 to Updated Virgil C Summer Nuclear Station Fire Protection Evaluation Rept, Effective 900301 ML20058K4421990-06-29029 June 1990 Forwards Response to Generic Ltr 90-04 Re Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions ML20043F6621990-06-0707 June 1990 Requests for Change in QA Program 10CFR50.54 & FSAR Biennial Reviews of Plant Procedures Based on Justification Contained in Proposed Change to FSAR.SAP-139 Will Be Revised Immediately Following Approval of Request ML20043D5861990-06-0101 June 1990 Forwards LERs 90-004 & 90-006 Which Respond to Violation Noted in Insp Rept 50-395/90-12.Corrective Actions:Training Conducted & Tech Spec Instruments Evaluated for Adequate Testing ML20043C5001990-05-29029 May 1990 Forwards marked-up Pages to Util 900410 Tech Spec Change Request Indicating Location of Incorrect Std Refs,Per 900419 Telcon W/Jj Hayes ML20043B7351990-05-23023 May 1990 Forwards Rev 0 to Core Operating Limits Rept,Cycle 6. ML20043A9181990-05-17017 May 1990 Forwards List Detailing Tubes in Which F* Criteria Applied During Steam Generator Tube Insp Subsequent to Fifth Inservice Eddy Current Exam.Location of Degradation Measured from Tube End on Hot/Cold Leg Up to Degradation ML20043G4621990-05-0303 May 1990 Advises That Util Will Control Operations to Abide by More Restrictive Required Shutdown Margin Curve While Awaiting NRC Approval to Place Revised Curve in Tech Specs ML20042G4741990-05-0101 May 1990 Special Rept Spr 90-003 Listing Number of Steam Generator Tubes Plugged or Repaired During Fifth Refueling Outage ML20042E1931990-04-11011 April 1990 Forwards Seismic Activity Near VC Summer Nuclear Station, Oct-Dec 1989. Several Seismic Monitoring Stations Inoperable During Reporting Period & Also During First Quarter 1990 ML20012F3551990-04-0303 April 1990 Withdraws 900321 Request for Relief from Testing RHR Containment Isolation Valves Xvg 08701 A/B at Frequency Specified in ASME Section XI Code.Meeting Requested to Discuss Util Interpretation of Testing in 10CFR50,App J ML20012E9061990-03-23023 March 1990 Submits Supplemental Response to Station Blackout Re Proper Documentation & Consistent Implementation of NUMARC 87-00 Guidance.Plant Currently Maintains Targeted Emergency Diesel Generator Reliability by Ensuring Compliance W/Tech Specs ML20012E0431990-03-23023 March 1990 Forwards Corrected Pages to Rev 26 to Radiation Emergency Plan. ML20012D9691990-03-23023 March 1990 Forwards, Sante Cooper 1989 Annual Financial Rept, South Carolina Gas & Electric 1989 Annual Financial Rept Summary of Present Levels of Property Insurance & Cash Flow Statement for VC Summer Nuclear Plant ML20012E5881990-03-21021 March 1990 Forwards Corrected Slide to Clarify Util Current Plans Re Steam Generator Insp Plan,Per 900312 Meeting W/Nrc ML20012D8271990-03-21021 March 1990 Requests Relief from Testing RHR Containment Isolation Valves Xvg 08701 A/B at Frequency Specified in ASME Section XI Code.Valves Will Be Tested at Frequency Specified for Type a Valves in 10CFR50,App J ML20012D0371990-03-19019 March 1990 Responds to Generic Ltr 89-19 Re Design of Steam Generator Overfill Protection.Util Does Not Plan to Implement Physical Mods or Administrative Changes Since Overfill Protection Sys Meets or Exceeds Guidance in Generic Ltr ML20012D4781990-03-16016 March 1990 Forwards First Annual ECCS Evaluation Model Revs Rept,Per 881017 Rev to 10CFR50.46.Rept Identifies Several Mods to Large & Small Break LOCA Evaluation Models Used at Facility & Provides Estimated Effects of Changes on ECCS Analyses ML20012C0661990-03-0909 March 1990 Forwards Addl Info Re Natural Circulation Evaluation Program Rept,Per 900208 Telcon Request ML20012A0921990-03-0101 March 1990 Forwards Response to Generic Ltr 90-01,consisting of Completed NRC Regulatory Impact Survey Questionnaire Sheets Containing Estimates of Time Spent by Managers on Insps & Audits ML20006E3411990-02-0606 February 1990 Forwards Proprietary Addl Info,Per NRC 900104 Request,Re Util Tech Spec Change Request for Elimination of Resistant Temp Detector Bypass Manifold sys,marked-up Tech Spec Pages & Block Diagrams.Encls Withheld ML20006D1241990-02-0101 February 1990 Requests Approval,Per 10CCFR50.55a,for Use of Alloy 690 Matl in Fabricating/Use of Steam Generator Plugs During Upcoming Refueling Outage.Alloy 690 Currently code-approved Matl for Steam Generator Tubing Based on Corrosion Resistance ML20011E1211990-01-31031 January 1990 Forwards Response to Generic Ltr 89-13, Svc Water Sys Problems Affecting Safety-Related Equipment. Util Currently Performs Visual Insps & Dredgings of Svc Water & Circulating Water Intake Structures Once Each Refueling Cycle ML20006D2951990-01-26026 January 1990 Forwards WCAP-12464, VC Summer Nuclear Station Natural Circulation Evaluation Program Rept. Rept Provides Info to Resolve Branch Technical Position Rsb 5-1, Design Requirements for RHR Sys for Facility,Per 890719 Telcon ML20006B1211990-01-24024 January 1990 Forwards Technical Rept 89-3, Seismic Activity Near VC Summer Nuclear Station for Period Jul-Sept 1989. ML20006A7191990-01-19019 January 1990 Provides Update of long-term Corrective Action for 890711 Loss of Offsite Power.Installation of Voltage Regulator on 230/7.2 Kv Emergency Auxiliary Transformer Neither Required Nor Any Appreciable Benefit Derived from Installation ML20006A5521990-01-16016 January 1990 Advises That Util Will Validate Adequacy of Any Sys Used for Periodic Inservice Insp & Will Upgrade,As Required,Eddy Current Test Methods Used as Better Methods Developed & Validated for Commercial Use,Per 900111 Discussion ML20005H0791990-01-12012 January 1990 Advises That Response to NRC Request for Addl Info Re L* Implementation Will Be Submitted by 900801 Due to Refueling Outage Schedule ML20005F1191990-01-0505 January 1990 Forwards Description of F* Application at Plant & Sample Eddy Current Lissajous Figures for 12 Steam Generator Tubes, Per 891218 Request for Extension of Application of F* Tube Plugging Criterion for Life of Steam Generators ML20005G4921990-01-0505 January 1990 Responds to NRC Bulletin 89-003, Potential Loss of Required Shutdown Margin During Refueling Operations. Bulletin & Core Shuffle Procedure Will Be Discussed Among Core Engineering Personnel in Documented Training Session ML20005F4531990-01-0404 January 1990 Forwards Application for Approval to Incinerate Oil Contaminated W/Very Low Levels of Licensed Radioactive Matls within Site Boundary of Facility ML20005F0601990-01-0303 January 1990 Forwards Justification for Continued Operation Re Pressurizer Surge Line Thermal Stratification,Per NRC Bulletin 88-011.Util Believes That Plant Can Continue Operation for at Least 20 Addl Heatup/Cooldown Cycles 1990-09-17
[Table view] |
Text
k
.l SourH CAROLINA ELECTRIC a GAS COMPANY 705T CF FICE BOR 764 COLuweiA, south CAROLINA 29ais T. c. Nicwo ts, J a February 23, 1981 wet Passentwf aut G40w' [EECVNE 9tAttam CrpthateOm5 Mr. Harold R.'Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 FSAR Question 421.83 - Q-List
Dear Mr. Denton:
South Carolina Electric & Cas Company (SCE6G) acting for itself and as agent for the South Carolina Public Service Authority, hereby submits forty-five (45) copies of the response to FSAR question 421.83 concerning the Q-list. This question was incorrectly numbered as 420.83 in a previous NRC letter to SCE&G.
The response will be incorporated into the FSAR in amendment 24.
If you have any questions, please let us know.
Very truly yours, V $/
c sce T. C. Nichols, Jr.
NEC:TCN:glb Enclosure ec: V. C. Summer w/o enclosure
- G. H. Fischer w/o enclosure -
T. C. Nichols, Jr. w/o enclosure E. H. Crews, Jr.
O. W. Dixon, Jr.
C. A. Price D. A. Nauman W. A. Williams, Jr.
R. B. Clary q,.
~~'
A. R. Koon , ,
A. A. Smith H. N. Cyrus J. B. Knotts, Jr. s[ 21 ild /f533EG J. L. Skolds B. A. Bursey ..
O. S. Bradham . . l'rt 32 f f : '^
ISEG N!WlEld U5! 022 :
PRS NPCF/Whitaker File 'E103 02 0
4 421.83 Section 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identification of safety-related structures, systems, and components (Q-list) controlled by the QA program. You are requested to supplement
~
and clarify the Q-list in Tables 3. 2-1, 3. 2-2, 3.10-1, and .3.10-2 o f the FSAR in accordance with the following:
- a. The following items do not appear on the Q-list. Add these items or justify not doing so.
- 1) Biological shielding within reactor building, control building, auxiliary building, fuel handling building, and intermediate building.
- 2) Missile barriers protecting safety-related equipment within the reactor building, control building, auxiliary building, fuel handling building, intermediate building, diesel generator building pump house, anu around air intakes, vent stacks, and other outside structures as applicable.
- 3) Pressurizer spray nozzles.
- 4) Steam generator steam flow restrictors.
- 5) Fuel handling building crane.
- 6) Meterological data collection programs.
- 7) Steel liner of the reactor building.
- 8) Supports for containment heat removal system.
- 9) Combustible gas control system a) H purge supply and exhaust system
. b) H{2monitoringsystem c) H2 analyzer d) Supports .
- 10) Cont _1.iment and emergency cooling system a) Cooling coils b) Ductwork dampers and supports
- 11) Onsite power systems (Class IE) a) Diesel generator packages including auxiliaries (i.e. , governor, voltage regulator, excitation system) b) 4160 volt switchgear c) 480V load centers d) 480V motor control centers e) Instrumentation, control, and power cables (including underground cable system, cable splices, connectors, and ter=inal blocks)
[ .: ..
! L f) Conduit and cable trays and their supports flote--Raceway installations containing Class lE cables and other raceway installations required to meet seismic
- category 1 requirements (those whose failure during --
a seismic event may result in damage to any Class lE or other safety-related system or components) should be included.
g) Transformers h) Valve operators i) Protective relays and control panels j) AC control power inverters k) 120 AC vital bus distribution equipment
- 1) Containment electrical penetration assemblies m) Other cable penetrations (fire stops)
- 12) DC power systems (Class 1E) a) l?5V battery, battery chargers, and distribution equipment b) Cables c) Conduit and cable. trays and their supports tiote--Raceway installations containing Class IE cables and other raceway installations required to meet seismic - - - -
category 1 requirements (those whose failure during a seismic event may result in damage to any Class lE or other safety-related system or components) should be included.
d) Battery racks e) ' Protective relays and controi panels
- 13) Reactor building charcoal cleanup plenums.
- 14) React'or building purge exhaust plenum.
- 15) Controlled access area charcoal exhaust plenum.
- 16) Process and effluent radiation monitoring systems.
- 17) Auxiliary. building charcoal filter system plbnum.
18)~ Buried service' water piping systems.
- 19) Safety-related masonry walls (see IE Bulletin tio. 80-11).
, -20) Expendable and consumable items' necessary for the functional perform-ance of CSSC (i.e., weld rod, fuel oil, boric acid, snubber oil, etc.).
+ 4 4 g _
- 21) Measuring and test equipment., -
- 22): ; teak detection system (per FSAR Section 5.2.7).
l23) Pressuri er safety and relief ~ valves,-block valves, and associat.ed l actuators.
- 24) Steamline safety valves and PORVs'.
, l
- 25) Pressurizer discharge line to pressurizer drain tank.
- 26) Containment sump, sump screen, and sump vortex suppression devices.
' ' ~
- 27) Radiation monitoring (fixed and portable).
- 28) Radioactivity monitoring (fixed and portable).
- 29) Radioactivity sampling (air, surfaces, liquids).
- 30) Radioactive contamination measurement and analysis.
Personnel monitoring internal (e.g., whole body counter) and !
31) external.(e.g., TLD system).
- 32) Instrument storage, calibration, and maintenance. l l
- 33) Decontamination (facilities, personnel, and equipment).
l
- 34) Respiratory protection, including testing. !
- 35) Contamination control.
- b. Enclosure 2 of NUREG-0737, " Clarification of TMI Action Plan Requirements" t (November 1980) identified numerous items that are safety-related and appro-priate for OL application. and therefore should be on the Q-list for the l Virgil C. Summer Nuclear Station. These items are listed below. Add these l
~
items to the Q-list and/or indicate where on the Q-list they can be found. Otherwise justify not doing so.
NUREG-0737
. (Enclosure 2)
Clarification Item l
Plant-safety-parameter display console. I.D.2 l
- 1) ,
1 2)
Reactor coolant system vents. II.B.1 Plant shielding. II.B.2
- 3) _
- 4) Post' accident sampling. II.B.3
- 5) . Valve position indication. II.D.3
- 6) Auxiliary feedwater system. II.E.1.1
- 7) - Auxiliary feedwater system initiation II.E.1.2 -
and~ flow. l
- 8) . Em$rgency power for pressurizer heaters. II.E.3.1 .
Dedicated hydrogen- penetrations. II.E.4.1 9)
~%* < . . - . .- ._ 9
l i
- 10) Containment isolation dependability. II.E.4.2 ;
11 ) Accident monitoring instrumentation. II.F.1 ;
I
- 12) Instrumentation for detection of inadequate II.F.2 -
core-cooling. j t
- 13) Power supplies for pressurizer relief II.G.1 !'
valves, block valves, and level indicators.
i
- 14) Automatic PORV isolation. II.K.3(1)
- 15) Automatic trip of reactor coolant pumps. II .K.3( 5) t II.K.3(9)
- 16) PID controller.
- 17) Anticipatory reator trip on turbine trip. II .K.3(12) ;
~
- 18) Power on pump seals. II.K.3(25)
- 19) Emergency plans. III.A.l.1/III.A.2 '
- 20) Emergency suoport facilities. III.A.l.2 21') Inplant 12 radiation monitoring. III.D.3.3 ,
- 22) Control-room habitability. III.D.3.4
- c. The instrumentation and control systems and components must be identified on the Q-list to the same scope and level of detail provided,_in Chap,ter_7 of the FSAR. (Tables 3.10-1 and 3.10-2.are too general .) Include safety-related display instrumentation such as source range neutron flux monitors.
Also include applicable items listed in Section 3.11.
, , ,w >%ww. ..we=** *- ,
d I .
4 e
Response
Section 17.1.2.2 has been revised to include Tables 3.10-1 and 3.10-2 as a part of the Q-list. Tables 3.11-0 and 3.11-Oa, as shown in Section 17.1.2.2 have been a part of the Q-list. These tables, along with Tables 3.2-1 and 3.2-2, have been revised as necessary to include items requested by this '
question as discussed below. The te'rm safety related in this response is defined in 10CFR 50 Appendix B as these systems, structures and components i that prevent or mitigate the consequences of potential accidents that could cause undue risk to the health and safety of the public. j a.1) Included in revised Table 3.2-2. :
a.2) Included in revised Table 3.2-2.
a.3) The pressurizer spray nozzles are safety related and are an integral part of the pressurizer. The pressurizer is included in Table 3.2-1. ,
a.4) The steam generator steam flow restrictors are safety related and are an integral part of the steam generators. The steam generators are included in Table 3.2-1.
a.5) The fuel handling building crane is not safety related and is not included as an item on the Q-list.
a.6) The equipment associated with meteorological data collection is not safety related. Refer to Section 2.3. Also, see Note 1 of this response.
a.7) Included in revised Table 3.2-2. ,
a.8) Included in revised Table 3.2-1. !
a.9a) These items !
H, areremoval includedisinbv
~
use of3.2-1, Tables internal H$recombiners.
3.10 2, and 3.11-0. !
Included in revised Table 3.2-1. I a.9b) a.9c) Included in Tables 3.10-1 and 3.11-0a.
a.9d) Included in revised Table 3.2-1. ,
a.10a) Included in Table 3.2-1. Refer to Note 17 of Table 3.2-1.
a.10b) Included in revised Table 3.2-1.
a.11a) Included in Tables 3.10-1 and 3.11-0a.
a.11b) 7200 volt switchgear is used at the Virgil C. Summer Nuclear Station in lieu of 4160 volt switchgear. 7200 volt switchgear
. is included in Tables 3.10-1 and 3.11-0a. ;
a.11c) Included in Table 3.11-Oa. - I a.11d) Included in Tables 3.10-1 and 3.11-0a. .
a.lle) Included in Table 3.11-Or..
a.11f) Cable trays and supports are included in Table 3.10-1 and 3.11-0a.
Conduit is not safety related and is not included on any of these tables. The conduit is, however, seismically supported for !
safety related circuits and those non safety related circuits !
over safety related quipment.
a.11g) The 480 volt vital system transformers are included in Tables .
3.10-1 and 3.11-Oa. The 7200 volt to 480 volt transformers are ,
a part of the 7200 volt switchgear included on Tables 3.10-1 and
3.11-0a.
-a.11h) Included in Tables 3.10-1, 3.11-0 and 3.11-0a.
a.111) Included in Tables 3.10-1 and 3.11-0a. -
a.11j) . Included in Tables 3.10-2 and 3.11-0. ,
1 r
I a.11k) The safety related 120 volt vital AC equipment is included in Table 3.11-0a. ,
a.111) Included in Tables 3.10-1 and 3.11-0a.
- a.11m) Included in revised Table 3.2-2.
a.12a) Included in Tables 3.10-1 and 3.11-0a.
a.12b) Included in Table 3.ll-0a. ~
a.12c) Cable trays and supports ar'e included in Table 3.10-1 and 3.11-0a.
Conduit is not safety related and is not included on any of these tables. The conduit is, however, seismically supported for safety j
related circuits and those non safety related circuits over safety related equipment, a.12d) Included in Table 3.11-0a.
- a.12e) Included in Table 3.11-0a.
a.13) Reactor building charcoal cleanup plenum is not safety related and l
is not included on any of the tables. Refer to FSAR Section 9.4.
a.14)_ Reactor building purge exhaust plenum.is not safety related an,d is not included on any of the tables. Refer to FSAR Section 9.4.
- a.15) Controlled access area charcoal er.haust plenum is not safety
' related and is not included on any of the tables. Refer to FSAR I Section 9.4.
a.16) Radiation monitors which are safety related are included in Table 3.10-1. Also refer to the response to items a.27. a.28, a.29, a.30,
[ a.31, a.33, a.34, a.35 and b.411.
Auxiliary building charcoal filter system plenum is not safety l
- j. a.17) related and is not included in any of these tables. Refer to FSAR l
- Section 9.4.
a.18) Inclyded in Table 3.2-1. Buried pipe is not excluded.
- a.19) There are no safety related masonry walls at the Virgil C. Summer Nuclear Station.
- a.20) Consumable products are addressed in revised notes to Table 3.2-1.
- a.21) Measuring and test equipment is not safety related and is not includet in any of the Q-list Tables. The calibration program and control of equipment used in safety related activities are under the aspects 1of the quality assurance program.
- a.22) Leak detection system consisting of the following elements (flow switches,' level switches, temperature sensors, pressure transmitters 1
' and radiation monitors) meets the requirements of regulatory guide ,
, 1.45 as described in Appendix 3A. The elements which are safety
.related are' included in revised Table 3.10-1 and 3.10-2. - *
! a.23). Included in revised Table 3.2-1.
a.24) Since the PORV's and Safety Valves are in between the steam -
. generators and the main steam isolation valves, they are included in Table 3.2-1.
,a.25) ' Included in revised Table 3.2-1.
a.26) The reactor building sumps and sump screens are safety related and _
are a part of the " reactor building. interior structures"found on Table 3-2-2.'
The equipment involved is not safety related per ANS 18.2. The
. a.27) programs which control _these activities are in compliance with 110CFR 50 Appendix B and are administered to meet the requirements of-10CFR 20..
a.28) See response to a.27).
a.29) .See response to a.27).
' p+e* +e-+-- -=r =3r y Tw*P?T--*e - ' "r W *
. ll l
l a.30) See response to a.27).
a.31) See response to a.27).
a.32) See response to item a.21).
a.33) See response to item a.27).
a.34) See response to item a.27).
a.35) See response to item a.27). -
b.1) This item will be addressed'when NURIG 0696 Revision 1.is issued.
b.2) Included as a part of the reactor coolant system in Table 3.2-1.
b.3) Included in revised Table 3.2-1.
b.4) Post accident sampling equipment is not safety related and is not included in any of the Q-list Tables. Refer to Chapter 11.
b.5) Limit switches are included in Tables 3.11-0 and 3.11-0a. Critical system leak monitoring system is included in Table 3.10-2.
b.6) The emerger.cy feedwater system is included in Table 3.2-1.
b.7) Included in Tables 3.10-1, 3.10-2, 3.11-0 and 3.11-0a.
b.8) Backup heaters for the pressurizer are powered from the 7200 ~
volt switchgear which is included in Table 3.10-1.
b.9) Since SCE&G uses internal H2 recombiners, no dedicated H2 penetra-tions are required.
b.10) The equipment involved in containment isolation is safety related and is included in Tables 3.2-1, 3.11-0 and 3.11-0a. The program for containment isolation dependability is under the aspects of the quality assurance program and is performed as a part of the pre-operational test program of the plant.
b.11) Table 7.5-1 provides a list of accident monitoring instrumentation. ,
' These instruments are included in Tables 3.11-0 and 3.11-0a.
b.12) Subcooled monitor and reactor vessel level instrument are included -
in revised Table 3.10-2.
b.13) All AC and DC power supplies for the pressurizer relief valves, block valves and level indicators are included in Tables 3.10-1, !
3.10-2, 3.11-0 and 3.11-0a.
b.14) Automatic PORV isolation is determined not to be necessary and is not included in the current design.
b.15) Automatic trip of reactor coolant pumps is determined not to be necessary and is not included in the current design, b.16) 'The PID c'ontroller is turned off. Refer to Figure 7.7-4. 1
-b.17) 'That portion of the anticipatory reactor trip on turbine trip which is safety related is on the solid state protection system which~is-included in Table 3.10-2. .
- b.18) -Refer to the response to Question 211.123. Devices used are included
.in Table 3.11-Oa.
b.19)~ " Emergency Plans" are not a piece'of safety related equipment and is therefore not. included.on any of the Q-list Tables. The emergency plans and applicable implementing procedures are under the aspect ,
of the quality assurance procedure. I L b. 20) This item will.be discussed when NUREG 0696 Revision 1 is issued.
'b.21) The equipment used for inplant2I radiation monitoring is not aafety ,
'related and'is not included-in any of the Q-list Tables. The programs 1 Lwhich control.the monitoring activities are-in compliance with 10CFR
50 Appendix R and are administered to' meet the requirements of 10CFR 20.
b.22)' Control. room ventilation equipment is included in Table 3.2-1.
c.)' Included in Tables 3.11-0 and 3.11-0a.
a w
w
_ _ _ _ _ ___q 1 . !
[
j i i
i NOTES: l r
t
- 1. Procedural coverage of some or all aspects of this equipment is covered in the quality assurance program per 10CFR 50 Appendix B.
i !
I
. 1
+
8 i 3
- - t
} I i, l t
I '
- + i i
I l I
! l' 1 L
)
l i
l' i, ,
4 1 , t
! i t
s 6 1 t 1
4 i
i I-
, c
. F L
S O
3 1
I i
1
?
I '
- 1. t i
h I.
L ,. ,.
Q)+~.6,--..,~,._,...._.-,r.m.,..--. , . . . . - . _ - - . . _ . . . - - , , - . - , _ _ - . _ . . . _ , _ , _ _ _ _ _ . . _ . . . . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ , _ . . . . _ . _ _ _ _ _ , . . _
_m . . _. _ _ . _ _ . _ _
TABLE 3.2-1 24 AMENDMENT 38 APRIL, 1979 MECllANICAL EQUIPMENT CLASSIFICATION ANS Code Seismic QA Component Scope Sa fety Class Code Class Category Notes Clas_s MECHANICAL COMPONENTS (by systeJn)
REACTOR COOLANT SYSTEM Reactor Vessel NSSS 1 AStE III 1 I 1 1 Ful1. Length CRDM llousing (48)' NSSS .1 ASME III 1 I 1 2 Part Length CRDM llousing (5). NSSS 1 ASME III 1 I 1 2 Reactor Coolant Pump Assemblies (3) ~ NSSS 1 ASME III 1 I 1 -
Reactor Coolant Pump Casings (3) NSSS 1 ASME III 1 'I I 2 Reactor Coolant Pump Internals (3). NSSS 1 ASME III 1 I 1 2 Reactor Coolant Pump Motors (3) NSSS 2b NEMA 20 -
I 1 2 Steam Generator, tube side (3) NSSS 1 ASME III 1 I 1 2 y Steam Cenerator, shell side (3). NSSS 2a ASME III 1 I ,
1 2,3 t;3 Pressurizer- NSSS 1 ASME III I I 1 2 u . Reactor. Coolant Thermowell NSSS ASME III 1 1 I 1 4 Reactor. Coolant Piping and Fittings NSSS 1 ASME III 1 I 1 1,5 Surge Pipe and Fittings NSSS 1 ASME III 1 I 1 1 Loop B fpass Line NSSS 1 ASME III 1 I 1 1 Bypass. Manifolds NSSS 1 ASME III 1 I 1 4 Relief Valves (3),Sifety WJver(8),81===W/**8(S) NSSS 1 ASME III 1 I _I 1
. Valves to Reactor Coolant System NSSS/ BOP 1 ASME III 6 Boundary 1 I 1 Q ,'
Piping to Reactor Coolant System NSSS/ BOP 1 ASME III 1 1 1 5,6 pr.,, , . nharye INe to PAT 80? #dS * *#/* ##Y ~
I i ~
if Pressurizer Relief Tank (,I'Rd NSSS -
NNS ASME VIII - - -
4 CRDM liesd Adapter Plugs ' NSSS 1 ASME III I I 1 4 Fuel Assemblies NSSS NA - -
I 1 - >
CHEMICAL AND VOLUME CONTROL SYSTEM
- Regenerative IIcat Exchanger NSSS 2a ASME III 2 I 1 1 Letdown IIcat Exchanger, tube side NSSS 2a ASME III 2 I 1 1 Letdown l{ca t Exchanger, shell side NSSS 2b ASME III 3 I 1 1,7 Mixed Bed Demineralizers (2) NSSS 3 ASME III 3 I 1 4
- . - ~ __- _. ..
W:.
O n Q .
q) O.a e.
-:c TABLE - 3. 2-1 (Continued) 29 AMENDMENT 4 N, 1HR MEC11ANICAL EQUIPMENT CLASSIFICATION M eck,#1FO ANS Code Seismic QA
-Component Scope Safety Class Code Class Category Class Notes r
NUCLEAR SAMPLING SYSTEM Residual Heat Removal Sample Cooler ' BOP 3 ASME III 3 I 1 -
ASME III 3 I 1 -
- ' Pressurizer Sample Cooler BOP 3
. Pressurizer Sample Vessel ' BOP NNS ASME VIII - - - -
3 ASME III I 1 -
Reactor Coolant Sample Coolers (2) B0P _
3 l4
. Reactor Coolant Sample Cylinders (2) BOP NNS ASME VIII - - - -
3- ASME III 3 I 1 -
Steam Generator Blowdown Sample BOP t
Coolers.(3)
Volume Control Tank Gas Space Sample BOP NNS ASME VIII - - - -
w NNS - - - - -
Sample Sink BOP ,
g 1 5 Steam Generator Blowdown Sample BOP 3 ASME III 3 I 4'
w Piping l Steam Generator Blowdown Sample BOP 3- ASME III- 3 I 1 5
-Valves 3 I 1 5 Nuclear Sampling System Piping BOP 3 ASME III 3 ASME III 3 I 1 5 Nuclear Sampling System Valves BOP Reactor Coolant Sampling Delay BOP 2a ASME III 2 I 1 -
Coils-(2) 3 ASME III 3 I I -
CVCS Sampling Delay Coils (2) BOP e.ad - Sample
[ 28 POST ACCIDENT HYDROGEN REMOVAIASYSTEMS
- s. 1 -
2a ASME III 2 I
' Electric 11ydrogen Recombiner NSSS - -
Sample Vessel BOP. NNS ASME VIII -
BOP- 2a/NNS ASME III/- 2/- I/- 1/- 5 2't Piping jand- Valves a sJ S.pferTs ,
. REACTOR MAKEUP WATER SYSTEM .
2b ASME III 3 I 1 -
Reactor Makeup Water Storage Tank BOP 1
Reactor Makeup Water Pumps (,2) BOP 2b ASME III 3 I 9
_ .,w-. -- - _ .._,m ,
TABLE 3. 2-1 (Continued) AMENDMENT
. MECllANICAL EQUIPMENT CLASSIFICATION ANS Code Seismic QA-JComp~onent Scope Safety Class Code Class Category ~ Class Notes
' VENTILATION EQUIPMENT Reactor Building Cooling Units (4)[thI BOP 2b ASME III 3 I 1 16,17 II s Reactor Building Charcoal Cleanup BOP NNS - -
I 1 18 System Fans (2)
Reactor Building Purge Valves BOP 2a ASME III 2 I 1 -
Auxiliary, Building Charcoal Exhaust BOP -NNS - - - -
13 System Fans (4)
Fuel.llandling Building Exhaust Fans BOP 3 - -
I 1 la Spent Fuel Pool Supply Fan BOP NNS - - - -
13
.." Residual licat Removal / Spray Pump _ BOP. 2b ASME III 3- I 1 17,18 7 Room Cooling-Unit-os ! Charging Pump Room Cooling Unit-BOP 2b ASME III 3 I 1 17,18 l3 Auxiliary Building Charcoal Filter BOP NNS~ - - -
13 System Plenum (2)
. Fuel'Ilandling Building Charcoal BOP 3 - -
1 1 13 Filter Plenum (3);
Iritermediate Building. Pump Area BOP 2b ASME III 3 I 1 17,18 Cooling Units (2)
Control Room Emergency Filtering BOP 2b - - I 1 18 f3 System Fans Control Room Normal Supply Units BOP. 2b ASME III 3 I 1 17,18
-Relay Room Cooling System Units BOP 2b ASME III 3 I 1 17,18 Battery Room Air Supply Fans BOP 2b - -
I 1 13 Battery Room Exhaust Fans BOP 2b - - I 1 18 IIVAC Water Chillers . BOP 2b ASME III 3 1 1 -
IIVAC Chilled Water Pumps. BOP 2b ASME III 3 I ,1 -
9 4
e n a o @ @ O
~
TABLE 3.2-1 (Continued) AMEND!ENT 3 E 8/
W M Arch,1991 ffEC11ANICAL EQUIPMENT CLA9SIFICATION ANS Code Seismic. QA Componen t - Scope Sa fety Class Code- Class Ca te go ry Class Notes Service Water Pumphouse Supply- BOP 2b'- - -
I 1 18 Fanc (2)
ESF Switchgear Room Cooling BOP 2b ASME III 3 I 1 17,18 Units (2) ,
3 Speed Switch Room Cooling BOP 2b AStfE III 3 1 1 17,18 Units (2)
Motor Con trol. Center 12-28 BOP 2b ASME III 3 I 1 17,18 Cooling Units (2)
- Switchgear 63-01 ~ Cooling BOP 2b ASifE III 3 I 1 17,18 Unit (1) p>. S=rwar,IA-res, sw% av g --- -
I .1 z'/
y s f ry Rat rd W~r'il.n' .-
3n. symm ,
s e
4 b e 4
- - -- - ---~
s TAELE 3.2-1 (Continued) b
)
,?
PECHANICAL EQUIPMENT CLASSIFICATION NOTES TO TABLE 3.2-1 .
4 Certain pressure retaining parts exist within a system which do not fall within the normal definition of pipes, pumps or valves. In scme cases these are not ccmmercially available f abricated f rom materials conforming (J to ASTM specifications allowed in ANSI B31.1, ASME III or ASME VIII; in addition, instrumentation is specifically excluded from ASME III. Exam- ,
ples of these items are strainers, sight glasses, level switches, pressure transmitters, thermowells, etc. These items are specified and procured -
in a manner which ensures that these components are comparable with the rcmainder of the systems to preclude their structural failure under operating, accident or test conditions. W*i c *' " c ' ' ' * -
- a -l"w~5 aJ m'e ry rs k rsd . Gmt l" 6 f H < sc i n ~ s u r<- wel.f evd, e )
w , n f, ~.inb ol, belc d.esst lsta nd c y ,'y- s m re c o -s id s esco's), 1,M e %,y sten 2< , e re,
- 1. Meets " Quality Control System Requirements," Westinghouse OCS-1, c.
- ;
which satisfies the requirements of 10 CFR 50, Appendix B.
- 2. Meets the quality assurance program of one of the Westinghouse NES Manufacturing Divisions, and is in accordance with 10 CFR 50',
Appendix B. .
- 3. As permitted by Paragraph NA-2134 of the ASME Code,Section III, this component is upgraded from the minimum required Code Class 2 to Code Class 1.
4 Meets " Quality Requirements for Manufacture of Nuclear Plant
-1 Equipment," Westinghouse QCS-?, which satisfies the requirements of 10 CFR 50, Appendix B.
f "L/
- 5. The classifications shown are for the predc=inant portion of the system. There may be portions that are classified higher or lower.
Safety class boundaries are shown on applicable system diagrams.
j Seismic category, code and QA classes for other safety classes are consistent with those other safety classes.
Aneadnear 24 bi s ec k 1961 3.2-20
e TABLE 3.2-1 (Continued)
==
MECHANICAL EQUIPMENT CLASSIFICATION
- 15. Main steam and feedwater piping, excluding branch lines, between -the l 12 -
C. e associated isolation valves and the wall between the intermediate building and turbine building satisfies all requirements, except for 12 stamping, of the ASME Code,Section III, Code Class 2.
- 16. No code. The fans and motors are specifically designed for opera-tion in the containment atmosphere under both normal operating and -
post LOCA conditions.
- 17. Code and Code Class apply to unit coils.
- 18. No code. TheAfans are designed and manufactured in accordance with the intent of ANS Safety Class 2b.
C.. 19.'No code. Ductwork is designed to withstand expected pressures and .
shocks' for the section of the plant in which it is located.
- 20. Any' reactor vessel internal, the single failure of which could cause release of a mechanical' piece having potential for direct damage (as to 'the vessel cladding) or flow blockages, sha.11 be classified to a minimum of Safety Class 2a.
e *
- 21. Failure can cause no nuclear safety problem, although an economic loss may result.
- 22. Portions which transmit loading from CRDM seismic supports are Safety Class 2a.
(
~
- 23. Applicable code is Crane Manufacturers Association of America, Specifi-cation No. 70 for Electric _ Overhead Traveling Cranes.
13
~
'24. Equipment is not' ANS Safety Class but is safety related.
25, 5 pers f.e re.crwr .bvilA**y coef Jry w#!7s are safey related. l19 3.2 AMENDMENT g ty must, ines 64 Arsh ,196l
e o n O 3 O O .
TABl.E 3.2-2 AMENDNENT & l'/
assese, w CLASSIFICATION OF STRUCTURES / I9 I Seismic Hon-Seismic Method of Tornado Tornado Missile Caterory i Category Nissile Protection } Crade to 30 Feet Above 30 Feet Licer i ReactorBuilding."PenetrationsandItatches(I)[b) X A. C 1,2,3,4,5.6.7 3.5,6.7 l?
Reactor Building interior Structures X Reactor Building Reactor Building Reactor Building Control Building UI(h X A,B.C,D E 1,2,3,4.5.6,7 3.5,6.7 Aux 11'f ary Building (S)(h X A.B.D,E 1,2.3,4,5,6,7 3.5,6.7
-Fuel llandling Building (5) X
- a. Ceneral structures for fuel pools and protective barriers for equipment (see Figures 3.8-58. 3.8-59, 3.8-60) A, B, C. D, E 1.2.3,4.5.6,7 3. 5, 6, 7 l7
- b. Steel superstructure ( '(see Figures 3.8-58, 3.8-59, 3.8-60) (4) (4) (4)
Intermediate Building I X A,B,C.D,E 1.2,3,4.5,6,7 3.5.6.7
. Diesel Cenerator Building X A,B.C.D,E 1,2,3.4,5,6.7 3,5.6.7 7
Service Water intake Purphouse and Discharge Structuren )(b X A, B C 1.2.3.4.5.6.7 Not applicable Service Veter Fond Damn X Not applicable Not applicable Not a.aplicable
^
Supports' for Safety Class Components X A, B. C. D. E 1,2,3.4,5,6,7 3, 5. 6 T l7 Turbine Building X Substation Structure and Control liouse X Water Treatment Building X Circulating Water Intake and Discharge Structure X Service Building X Aux 111ery Boiler llouse X Warehoune(s) X X F-Guard llouse(s)
Mnnticelin Reservoir Dams X c==
Jetty X Sanitary Waste Facility X Industrial Weste Tacility X .
e C' TAELE 3.2-2 (Continued)
CLASSIFICATION OF STRUCTURES l
3 C.;_ ., f NOTES:
i
- 1. Method of tornado missile protection is as follows:
A. Reinforced concrete valls.
B. Reinforced concrete slabs, i C. Reinforced concrete barriers.
D. Orientation. f E. Probability studies, probability criteria < 10' . f
- 2. Numbers correspond to. tornado missiles identified in Table 3.5-5.
- 3. Refer to discussion of probability study, response to Question 010.7. 2
. .)-
- 4. Steel fra=e is designed to maintain its integrity under tornado
- nissile impaet. ;
i 5 ' Sio\ogical shiddin3 a.nd missi\c barries- s h ch rol e peerdt 2tf j reta.w % ca.t4. p 3 I st ru.ctares idad:sta 1 ram 3.t _2.
- are c.tosst ha as saf ety - re.tohl . ,
(p, Fire stoft f*e e.lecric.I ysHerranb W
t c'
) .
3.2-23a RENDMENT 2 f/
. . , X:
M a sk,Itti r
1 i i O
TABLE 3.10-2 (Continued)
O IDENTIFICATION OF NUCLEAR STEAM SUPPLY SYSTEM l
SEISMIC CATEGORY I INSTRUMENTATION, ELECTRICAL EQUIPMENT AND SUPPORTS
' . ~?
4 Item Method 0' 12. Core Subcooling Monitor- Not yet qualified; qualification to be in accordance with methods i
~
described in Section 3.10.
- 13. Critical System Leak. ,
Not yet qualified; qualification i- Monitoring System to be in accordance with methods i described in Section 3.10.
Y ty, M M A L I S&G & A if g
...,,n.*
g , ::, . ; : ,.
~
- +
- . , -s . . , . _. , . . . . . /., .' _
,.. , * . . .. , - (; 1
~
x: .: , l _
. . . O i:~ b . ,.
....a c.' ., .
-;
~
A.
s l'.
- $. ;.., .
n
. AMENDMENT W 3.10-15a. : 2 7-
^
.msrb ,t9 %i I
D
-- .- . - - .- . -~ . -- -~~- , - ~ -. -
e I
17.2.1.6 Supporting Companies, Vendors or Contrac+or Organizations SCE6C may utilize the services of other companies to provide natcrials or. service s and' augment an t'. aupport its staf f in, selec ted plant opera- ,
tions, or' modification and maintenance projects. . To qualify for '
4 safety-related work, supporting companies must implement an approved QA program or work under the requirements of the SCE6C Operational QA Pro-gram. Use of the SCE6G Operational QA Program is limited to those cases where management functions are not performed. After purchase order or -
contract award, the vendor, supplier, or contractor shall conduct all
! quality related activities for safety-related structures, systems or components, whether at the plant or other locations, in accordance with the appropriate approved QA program.
l
! 17.2.2 QUALITY ASSURANCE PROGRAM The. Operational QA Program for the Virgil C. Sumer Nuclear Station
. consists of managerial and adminstrative controls 'by involved SCE6G .
'l ~ organizations, combined with surveillance and audits by the QA Department.
?~ As described generally in this section, and following sections of this chapter, the organizations responsible for implementing safety-related action are . clearly identified. _ At the discretion of SCE6G management, certain non-safety related activities may be conducted utilizing control 7
techniq'ues specified for' safety-related activities. These activities I are outside the scope of required regulatory compliance and are dis-i
' 12 criminated by the term " Quality Related". All " safety-related" activity '
l is ;"qualitiy : related, but not all '.' quality related" activity is -
"sa f et y-re la ted" . . L 0 4 i
17.2.2.1 Applicability The.0perational'QA program is applicable to startup, operation, mainte--
/
- ' nance, andLmodification of the structures, systems and components of Su=mer Nuclear Station classified aas The
.the Virgil C.'
mlsafet,y-related.
r+rve mt safety-related designations for= those Mechanical, site =s are listed in
, ped 1.t-t s}f8
- Tables 3.2-11 The list of Class 1E equipment required to function during -
D ,13 - _ ,
t go.1 3,jo,1, .
s and/or ' subsequent to design basis accidents is included in Tab 1'esA3 .11-0
.and 3.11-0a.; The. Operational _QA Program, as detailed'in the Operational
.QA Plan, shall be AMESDMENT S O .
17~2-1' W?!*
. J.L . . , . - - - z.