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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
. _ _ _ _ - _ _ _ _ - _ _ _ - _ _ _-
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ENCLOSURE 3 goumyn.
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4 September 19, 1986 .
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Mr. Jesse Arildsen License Examiner U.S. Nuclear Regulatory Commission
,. Region.II, Suite 2900 101 Marietta Street, N.W.
Atlanta, Georgia 30323-
Subject:
Virgil C. Summer Nuclear Station Docket No. 50/395 License No. NPF Operator License Examinations
Dear Mr. Arildsen:
Enclosed are South Carolina Electric & Gas Company comments on the NRC Operator'and Senior Operator examinations administered at the Virgil C.
Summer Nuclear Station on September 12, 1986. Your consideration is appreciated.
rytrtly[urs.
N a a W
RCR/ DAN:jez Enclosures c: J. G. Connelly w/o attachment O. S. Bradham w/o attachment M. B. Williams w/o attachment K. W. Woodward w/o attachment
.A. M. Paglia w/o attachment NPCF File 8710060224 861113 5 PDR ADOCK 050
- i. .
Question 5.01 b.3: ,
Match the items in column "A" to their respective core locations ..
in column "B". Assume full power.
"A" "B"
- 3. Maximum actual a. bottom heat flux b. between bottom and middle
- c. middle
- d. between middle and top
- e. top d
Answer key response: c i
l
REFERENCE:
WJE 237, HT&FF Manual, pp. 227-229 VCS, Hot Channel Factors p.4, Figure 2 4
Suggested correct response: either b or c Reason: When operating with a negative 6I as we most frequently do (Cycle I was negative the entire duration), maximum actual heat flux will be between the bottom and middle. _
See Enclosure I, The Nuclear Design of the Virgil C.
Summer Power Plant Unit 1.
Question 5.13: (1.0)
Why are Rod Insertion Limits established?
Answer key response:
Rod Insertion Limits are established to insure that the reactor core can be shut down at any time under any condition or to insure the minimum shutdown margin is maintained.
REFERENCE:
V. C. Summer Reactor Theory p.I5.50, TS p.3/4 1-3 001/000 KS.04 (4.3/4.7)
Suggest correct response:
The above answer and/or the following:
o Minimize the reactivity consequences of an ejected rod (rod misalignment).
o Maintain uniform power distribution limits.
Reason: Technical Specification 3/4.1.3 Bases contain tht -
reasons for rod insertion limits. See Enclosure
~,
QUESTION 6.06. (1.0)
LIST 4 means by which crystallization of the Boron Injection
. Tank (BIT) is prevented during normal plant operation.
ANSWER 6.06 (1.0) '
(1) Electrical heaters on BIT L (2) Trace-heat all piping, valves, and pumps-(3) Continuous recirculation flow (4 8 0.25 each)
(4) Sparger inlet of BIT I
REFERENCE:
SU: AB-10, p 22-23, 4/10 K4.02 (3.2/3.8)
Suggestion action: Delete Reason: MRP #20482 removed the BIT. Training materials have not incorporated these plant changes. See Enclosure III.
I 1
1 1
i I
QUESTION 6.14 (2.0)
State the specific back-up source of emergency feedwater for
. each of the- emergency feedwater pumps, and the condition under which the back-up supply suction line isolation valves automatically open.
ANSWER 6.14 (2.0) turbine-driven pump ------ service water loops A or B motor-driven pump A ------ service water loop A motor-driven pump B ------ service water loop B The back-up supply suction line isolation valves automatically open on a low suction pressure condition in the combined emergency feedwater pur- header from the condensate storage tank. (The action occurt 2 seconds af ter 2 of 4 pressure switches decrease to 10.4 psig.)
REFERENCE:
SU: IB-3 p 10 Suggested correct response:
turbine-driven pump ------ service water [ loops A or B]
motor-driven pump A ------ service water [ loop A]
motor-driven pump B ------ service water [ loop B]
The back-up supply suction line isolation valves automatically open on a low suction pressure condition in the combined emergency feedwater pump header from the condensate storage tank. (The action occurs 2 seconds after 2 of 4 pressure switches decrease to 10.4 psig.)
Reason: The specific backup source of emergency feedwater for each emergency feedwater pump is service water. To designate loops is assumed and superfluous.
w _ _-_
QUESTION 7.11 (2.50)
A Safety Injection (SI) Signal has just occurred and it appears to have been spurious. Prior to resetting the SI and recovering ESF components:
- a. List three containment building parameters that must be normal. (1.50)
ANSWER -7.11 (2 50)
- a. Reactor building temp. </=120 F Reactor building sump levels indicate "0" Reactor building radiation levels normal (any 3 0 0.50 ea)
Reactor building pressure indicates 0-1.5 psig
REFERENCE:
E0P 1.0 p.12 Suggested correct response:
Reactor building temp. [</ = 120 F]
Reactor building sump levels [0]
Reactor building radiation levels normal Reactor building pressure [0-1.5 psig]
Reason: The question asked for the parameters, not parameters and associated values.
QUESTION 7.20 (1.0)
V. C. Summer has guidelines for radiation exposure during emergencies. In what situation can a person receive a radiation exposure of 100 REM, whole body?
ANSWER For life saving operations (0.50) such as search and rescue for KNOWN missing persons. (0.50)
REFERENCE:
VCS, Radiation Protection Fundamentals, p 5 5-14 Suggested correct answer: For life saving operations.
Reason: Emergency Plan Procedure, EPP-020, Emergency Personnel Exposure Control, addresses radiation exposure guidelines during emergencies. No reference is made to " search and rescue for known missing persons." See Enclosure IV. Although EPP-020 allows for planned exposures of up to 75 Rem for lifesaving operations, the 100 Rem exposure addressed in the question is covered by the note associated with step 3 2.4 that disallows the 75 Rem for " spontaneous reactions."
{
_4_
J e, QUESTION f
7.22 (1.0)'
When is control rod cooling air required to be operating? l ANSWER 7 22 (1.0)
When the reactor is at hot shutdown temp. (> 200 F) or at power.
REFERENCE:
VCS, 10-5, Rod Control p 15 Suggested correct answer:
Accept also as correct response " anytime the control rods are energized (reactor trip breakers closed) and/or prior to exceeding 350 F RCS temperature."
Reason: The system operating procedure for Reactor Building Ventilation, SOP-ll4, states the above suggested answer as a precaution for control rod drive mechanism cooling. See Enclosure V.
QUESTION 7.23 (1.0)
Answer the following questions concerning the requirements associated with doing a reactor startup:
- b. If the startup is delayed, when must the Estimated Critical Condition Calculation be reviewed? (0.50)
ANSWER
- b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality
REFERENCE:
VCS, 00P-3, pp 9 & 11 Suggested correct answer:
- b. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to criticality ~or within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to
~~
criticality.
Reason: Reactor engineering procedure REP-109.001, Estimated Critical Conditions Calculation, notes recalculation /
reverification of ECC if criticality does not occur within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of criticality. See Enclosure VI. The GOP-3 pages 9 & 11, 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> requirements are Technical Specification 4.1.1.1.1.b requirements.
l I
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, QUESTION. Ic14
.See SRO question 5.13 comments QUESTION 2.09 '
. 'See SRO. question 6.06 comments
-QUESTION 2.16
.See SRO question 6.14 comments i QUESTION 3.09 (1.50)
Excluding various logic test panel testing, list three of the four remaining faults or actions which will generate a Solid
. State Protection System general warning alarm.
ANSWER
- 1) Loss of either 48-V DC power supply-
- 2) Loss of either 15-V DC power supply
- 3) A protection train's bypass breaker is racked in (inserted) and closed.
- 4) Any solid state circuit card not properly inserted (3 of 4 8 0.50 each)
REFERENCE:
SU: IC-9 p 34, 35 Suggested correct answer:
Accept " ground return fuse open (blown)" as a correct response.
Reason: The Westinghouse Tech Manual for Solid State Protection System references a blown ground return fuse as a possible cause for a general warning. See Enclosure VII.
QUESTION 4.22 See SRO question 7.22 comments QUESTION 4.23 See SRO question 7 23 comments i
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WESTINGHOUSE PriCP:<!ETAHY CLAS3 2
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- s THE NUCLEAR DESIGN OF THE VIRGIL C. SUMbiER POWER PLANT UNIT 1 R. J. Fabean R. E. Radcliffe R. H. Pituiski March, 1980 APPROVED: bk.h. APPROVE A ^A M c/'n P. K. Doshi, Manager h.k.Iudwiczak,YrojectManager NE Product Group C NFD Fuel Projects Work Performed Under Shop Order CGWF-11001 l
eeun w co no uwA This document contains information proprietary to Westinghouse Electric
-WESTINGHOUSE-PROPRIETARYTATA-7*f/f'*/{.7"I",,f"("]
f Corporation; it is submitted in confidence and is to be used solely for
' the purpose for which it is furnished and returned upon request. This document and such information is not to be reproduced, transmitted, disclosed or used otherwise in whole or in part without authorization
, of Westinghouse Electric Corporation, Nuclear Fuel Division.
WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230 i (AC d$4 VR p,f
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Ifuel'loadin9 figure identifies the assemblywise average and peak power, the value location of. the pea,k power rod, the assemblywise burnup and BP fract1.-
,erating remaining, the regionwise power and burnup sharing, the critical boron-concentration, and cycle burnup.
flowthe rela tive 3.3 BURNABLE POISON DEPLET10N The depletion of Boron-10 in the burn'able poison rods as a function of 2 and 3) core depletion is shown in Figure 3.9. The burnable poison rods are used 3r than in gegions 2 and 3'only. Both the regionwise and core average depletion of Baron-10 is shown in Figure 3.9. The burnable poison worth including structural material and moderator displacement effects is 7.0%4p at BOL ttern and reduces to about 1.3%ao at E0L.
.2 and 2.5 AXIAL POWER DISTRIBUTIONS !
3.4 .
Core average axial power distributions were calculated using the one-al boron =
dimensional APOLLO code.
.rnup, discrete The axial power distribution calculated at hot zero power conditions at the beginning-of-life is shown in Figure 3.10. Since the core contains all fresh fuel and the moderator temperature is constant axially, the power u
~
"9 distribution is approximately symmetrical about the core center. Figure 1%
n as 3.11 shows the HFP axial powdr distribution at BOL and EOL. Since the~ core t li fe-has atnegativectemperature coefficient:and the coolant temperature increases 1ths from the ' bottom.of. the: core to . the top of. the core. the axial' power;. dis-x tribution,is' shifted stoward - the core . inlet at- BOLT ^ As, the. core depletes .
the higher production in the- center of- the;. core;results. in a .non-uniform axial burnup: distribution and consequently the~ flattened power shape at Em : E0L'as!shownlin Figure 3.11. The ' axial' offset changes fromL -9.8%iat' BOL
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DkCb6hMVt NEACTIVITY CONTROL' SYSTEMS BASES BORATION SYSTEMS (Continued)
MARGIN from expected operating conditions of 1.77% delta k/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires 12475 ~
gallons of 7000 ppe borated water from the boric acid storage tanks or 64,040 gallons of 2000 ppa borated water from the refueiing water storage tank.
With the RCS temperature below 200'F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.
The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 275'F provides assurance that a mass. addition pressure transient can be relieved by the operation o,f a . ~
single P.RV. O The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 2 percent delta k/k after xenon' decay and cooldown from 200'F to 140'F. This condition requires.either 2000' gallons of 7000 ppa borated water from the boric acid storage tanks or 9690 gallons of 2000 ppm borated water from the refueling water storage tank.
The cont [ained water volume limits include allowance for water not available because of discharge line lo. cation and other physical characteristics.
The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.
The OPERABILITY of one boron injection system during REFUELING ensures that this system is available .for reactivity control while in MODE 6.
l
~
3/4.li3" MDVABLE' CONTROL ASSEMBLIES k
The specifications of-this section ensure that'(1)' acceptable power distribution' limits.are aintained,L(2)'the m minimum SHUTDOWN MARGIN is, main-tained,;and.(3) limit the potential effects'of rod misalignment on associated accident analyses ~. OPERABILITY of tric control rod position indicators is~
required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.
SUMMER - UNIT 1 B 3/4 1-3
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ClAC.CC)QY&
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~ MRF # Z()4,p 1
- SOUTH CAROLINA ELECTRIC & GAS COMPANY MCN REV.' # j VIRGIL C, SUMMER NUCLEAR STATION I
' MODIFICATION CHANGE NOTICE syggeig, 7[
PROBLEM DESCRIPTION g g ,,,, , gffgy e j gj,,,
A P bes/f/ 4 0 nWI f Joe /~c, t- ?.NE f f / ,~ f kg --ff/,
PROPOSED CHANGE:
f a,,, f fh /];f f f,y) ogg f' / m & i . yngp ,[,y,
S/f.;T;</U197 ct.J l/7 k f i m $'r.ffn) & l h /0 W> y J hSh / 2-l/-0$ n'l/ S/MC /Z-///d
// INITIATOR DATE ASSOCIATE MANAGER DATE ORALLY REQUESTED BY ORALLY APPROVED BY DATE
_ _ _ _ TIME DOCUMENTS AFFECTED ~ ~"~
OTHER GROUPS AFFECTED FSAR: [ ] YES { I NO SIGNATURES SECTION SIGNATURES GROUPS (AS REQ'D.) GROUPS ESSENTIAL DRAWINGS: (AS REQ'D.)
[JC72. d'7/ rev fg NE/ELEC. SEC11tITY g Jod 7.5/ revf // NE/I&C TRAINING NE/STRUCT.- HEALTH PHYSICS CIVIL (SITE)
OTHER DRAWINGS AND NE/ MECH. MAINTENANCE DOCUMENTS: NE/ FIRE PRCTI. OPERATIONS p jej NUC. ANALYSIS TECH. SUPPORT Q,/ NUC. SYSTEMS COMPUTER SERVICES CORP. HEALTH PEYSICS AND ENV. PROGRAMS DISPOSITION: P u i,i,b M,*u /n ysi. i ,,., D,ps yt ;, / ,/,,, , , u r HAVE RESULTSjoF THE 10CFR50.59 EVALUATION ANALYSIS SHEET FOR THE BASE MRF CHANGEDY
{ } YES {V) NO IF YES, PROVIDE REVISED 10CFR50.59 EVALUATION ANALYSIS SHEET.
DOES THIS MCN AFFECT THE BASE MRF FINAL EVALUATION? [ ] YES [(,,f NO DID THE EVALUATION qESTIONS FOR THIS MCN RESULT IN "YES" ANSWERS? ( l YES (rINO I REVIEWSPERFORMEDINTHEDJSFOSITIONOFTHISMCN7(,)YESIFEITHERQU [ j NO (r N/A IF NO, PROVIDE EXPLANATION M iV jf(A/ j fefie/p //p snb;tw/ yy)/diing7w sy a sj'rnf A-m L /
draanMos.
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DOES THIS CHANGE INVOLVE: [ () UMENT REVISION 7 [ ] PHYSICAL PLANT CHANGE? [ ] BOTH ?
INTERFACE REVIEW REQUIRED? [ YES , { l NO VERIFICATION REQUIRED? [ ] YES [hi NO VERIFICATION COMPLETE? [d YES [ ] NO
]'YES ( NO RECORD IF NO, JUSTIFY FORM Afn PR0i'I,DF,Dchoiss.p,/inknrgL
?,,fpj[ ihn A'/$
/ /V/9 kial' ba nr Awd hom i,fdd/ba se d A w ),, ga, VERIFIER DATE DIS SITION'APPR VAL ' '
MANAG ENT A OV I C/B;r ,r7 / d-d/*d?S ef f 4.+ s: .14d / 2 f//$ k O'GG1NEER/ LEAD ENGINEER DATE ASSOCIATE MANAGER DATE h hC. O N Vf. ~
k 13 2-E-I-1-- --
Revision 3
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(?nC.]dS'4V4 f f<I
'*h' SOUTH CAROLINA ELECTRIC AND GAS COMPANY 6
VIRGIL C. SUMMER NUCLEAN STATION
't NUCL$AROPERATIONS l' )
EMERGENCY PLAN PROCEDURE 4
EPP-020 EMERGENCY PERSONNEL EXPOSURE CONTROL REVISION 6 1 1
MAY 27, 1986 SAFETY RELATED G
'Date DISCI {LINESUPEfWISOR d2A$ d h APPROVAL AUTHORITY
& -3 0-&4 Date RECORD OF CHANGES CHANGE TYFE DATE DATE CHANGE TYPE DATE DATE NO. CHANGE APPROVED CANCELLED NO. CHANGE APPROVED CANCELLED-1 0
4
- _ - - _ - - - _ - _ - _ a
jl7-/_.
s hhCf054 rf
. .. - EPP-020 REVISIOH 6 5/27/86 3.2.1 Normal Administrative exposure limits may be suspended or modified verbally by the I.terim Emergency Director / Emergency Director, Radiological Assessment Supervisor, or Shift Supervisor. Verbal exposure extensions should be followed-up with the appropriate documentation as defined by reference 2.4.
3 2.2 Pederal limits of 10 CFR 20 will not knowingly be exceeded, except as described in Section 3.2.4 of this procedure.
3.2 3 Maintain exposure ALARA.
3.2.4 Planned exposures'of up to 75 Rem to save. human life l or prevent the serious endangennent of human life, or up to 25 Rem to save or mitigate significant damage to vital equipment and/or facilities must be approved by the Interim Emergency Director / Emergency Director or Radiological Assessment Supervisor.
Persons performing the planned actions should be 30E&O employees who are volunteers broadly familiar with exposure consequences.
l NL'TE: 'Thiefdoes.not apply to " spontaneous' reactions" by on the scene personnel in a threatening situation.
3.2 5 Non-company employees such as medical support personnel should be limited to:
- a. 3 Rem if there is an adequate number of personnel so that rotation may be accomplished.
b 5 Rem if the number of personnel is limited such that rotation can not be accomplished.
- c. 25 Rem to save a life.
3.2.6 If an individual receives an annual dose equivalent in excess of twice the Annual Dose Equivalent Limit, the case should be referred to a medical physician for review (as per ICRP Report 26).
hk C fo$ ur-R ;f1 PAGE 2 0F 5 i
______________U
bt c$u,rR M --
. .. 30P-114
~
REVI SION 11 1/03/86 I. PURPOSE This procedure outlines the steps involved in operating the It also includes l
Raactor Buildt.ng Ventilation System.
Off-Normal Operations.
II. PREC AUTIONS
- 1. For PI-8254, RB NR PRESS to accurately indicate pressure, the following valves must be open:
- a. SVX-6054-HR, RB NR PRESS CNTMT ISOL.
LOOP A (IRB) .
- b. SVX-6050A-HR, POST ACCID H2 2.
The concrete around the vessel supports and nozzles must be limited to less than 150*F.
The Control Rod Drive Mechanism cooling f ans must lessmaintain g
3 the temperature around the Control Rod Mechanisms at n than 170*F. j e
4.
The Digital Rod Position Indication Data Cabinet Cooling '
System must operate continuously to limit the space }
temperature in the cabinet area to less than 95 P. i l
5 The Control" Rod Drive Mechanism Cooling' Unit must be operating when one or more full length rods are energized and/or. prior to exceeding 350*F' Reactor Cooling System
' t empe rature. -
i 1
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h OENERAL REVISION PAGE 1 of 37 LL Eve osuve !
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- REP-109.001 ATTACHMENT I I
PAGE 4 0F 4 )
REVISION 0 !
l l
Summary of Estimated Critical Conditions I
E.1 Bank Position (0.1.1 or D.3.2150 steps)
E.1.1 Minimum: Bank at Steps E.1.2 Maximum: Bank at Steps E.2 Boron Concentration (C.2.5 or D.1.1) opm NOTE:,, : Recalculate: or reverify. the ECC. if criticality
<willJnot:occuriwithin;2LhoursLof;the: estimated
~ '
'timelof~ criticality. a E.3 Time (A.1 1 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
E.3 1 Earliest: Date Time _
E.3.2 Latest: Date Time E.4 The Bank Position in E.1 is above the Rod Insertion Limits (Figure II-16) and. below Rod Withdrawl Limits (Figure II-19)
INITIALS Calculated By Checked By Shift. Supervisor
- 9. Actual Criticality Data F.1 Bank Position: Bank at Steps P.2 Boron Concentration opm Tave 'F Time P.3 Reason if Bank Position was not within the desired band for Criticality:
Determined By:
[5htfoSurf 17~f~
&~ &sure l/IL , p. I l The remaining three input circuits, each having a separate ED p.ator, require constantly grounded inputs in order to prevent Q17 from turned on and causing a general alam. The inputs are grounded through
- h lly closed in positions where a general
' crie3 of contacts t at are norma da'* is not required. The breaking of the ground path, from a Test switch I from its nomal position, bypass breaker auxiliary contact open (i.e.
asM I
set) or the Ground Return Fuse blown or removed will cause a general geaker alarm.
hhen these three inputs are grounded, the 30.1K input resistors which are tied to +48V will insure breakdown of any switch contact film and These
,.aus.e the cathodes of diodes CR100, CR102, and CR104 to be at ground.
jiodes will then be fo m ard biased, producing approximately 0.6V at the anodes of the LED diodes (CR101, CR103, CR105). The ED diodes will be back-biased and non-indicating because the anode voltages are below the required junction voltages of the base-emitter of Q17, CR99, and the ED. If the ground is broken to one of the inputs, its ED and CR99 will be fomard biased and Q17 will be turned on because of the current path from 48V through the 7.5K pull-up resistor. This will cause a general alarm.
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+48V CARD INTERLOCK @ R40 < g Row No. 2 14 7 g' ' y i___p ____ L $'h4
+4 By CARD tNTERLOCg C #9,5 g l 'h"Di
@- - -+48v R0W NO 5 Ret ; RS7 s4 7N '
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' CR98 R0W NO 4 R 4 7 .'
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30IRl ' D7 SK GND FROW GNO RET URN FUSE . (SFU S)
CRioO a
CRIOe p y (g0 "tNPut ERROR iNMiBiT" i SWITCH (5509) CRiO61I +4ev
+40V RIO 3 m! ! SM ADoitioNAL output GND FROM g2 MW : FOR GENERAL
- LOGIC ANDA"*WVLTiptExtR SwifCM (5501)v Q4 '
LED WARNING INoiCatt0N TEST
C8,99 , /
GNO f ROM GNO ; RSt 0 21 A605 <
T RETURN f uSE 458 U 9)
">ERW155;vtS[, $mitCM tSS05) ,, 4 CRIOS l 332M ! ,
2 MEMORIES SWITCH (5506), ' l QUTPuf REL AY ' MODE LED SELECTOR SWITCH 15802). Q g
AND tvPA55 BRE AnER AUILlU ARY CONT ACT j I (OPEN wMEN SET) I Figure 2-33, General Warning Circuits i
2-88 acy, 1, af77
_ _ _ . . .. ..