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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
ENCLOSURE 3 gth oNna Electric & Gas Company Da A. n 1 *D Colum 29218 Nuclear Operations SCE&G a . .co. -.,
November 21, 1986 Mr. Jesse Arildsen License Examinor U.S. Nuclear Regulatory Commission Region II, Suite 2900 101 Marietta Street, N.W.
Atlanta, Georgia 30323
SUBJECT:
V.C. Summer Nuclear Station Docket No. 50/395 License No. NPF-12 Operator License Examinations
Dear Mr. Arildsen:
Enclosed are the utility comments to the NRC Operator and Senior Operator examinat{ons administered at V.C. Summer on November 17, 1986. Your consideration is appreciated.
Very truly you"s, D
. A. Nau RCR: DAN:dgb Enclosures cc: J. G. Connelly, Jr. (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 F. Jagger (w/ attachment)
$0 V
ATTACHMENT
QUESTION: [ 0. 3]
l.16 From the words or phrases in parentheses, choose the one which correctly completes each statement below,
- d. Calculation of Enthalpy Rise Hot Channel factor assumes core power is (uniform /or/not uniform) and flow through each channel is (the same/or/different) throughout the co re . (0.6.)
ANSWER KEY RESPONSE:
- d. Not uniform [0.3] The same [0 3]
REFERENCE:
GP HTFF, pp. 193-198 - Thermal Hydraulic Principles, pp. 13-30 thru 36 CNS Exam Bank SUGGESTED CORRECT RESPONSE:
- d. Not uniform [0.3] Delete second part.
REASON: While it is true that the calculation of enthalpy rise hot channel factor assumes the flow rate through each channel is the same (see attached reference TS-13, Design and Operational Charts Limits on Thermal Parameters of the Reactor Core, p. 29 ) , a four percent penalty is used to account for non-uniform distribution of flow (see same reference as above, p. 30, attached). This penalty implies different flow through core channels.
i , .
T t
point during normal operations. These limits on F also g '
. ensure that cladding temperatures will not exceed the self-
- sustaining zirconium-water reaction temperatures if a loss of T
l coolant accident and DNS conditions occur. F g, however, does l
! not ensure that DNS conditions will not occur during normal plant
) operations. In order to prevent DNS, limits are imposed on the enthalpy rise hot channel factor.
l Departure from nucleate boiling depends not only on the power density of the fuel, but also on the local enthalpy of the coolant l and the local mass flow rate of the coolant. The fuel rod l i ultimately transferrMg the maximum amount of heat to each unit mass of coolant flowing upward through the core is the rod where DNS is most likely to occur.
~
Since power density is not the only factor determining the onset of DN8, an evaluation of the total enthalpy rise along the flow channel must be made to ensure that DN8 conditions have not been met. In order to calculate the total enthalpy rise along a flow channel, the integrated ;ower for the entire length of the fuel
! rod is evaluated with the channel inlet enthalpy and the flow rate j through the channel.
Since power density throughout the core is not uniform, there is a fuel rod in the core with the highest integrated power.
l Assuming the inlet enthalpy of all channels and the flow rate through, each channel is the same, the highest coolant enthalpy rise occurs along this rod. The flow in this channel will be closest to DNS.
l I
The ratio of the maximum integrated rod power to the average rod power is defined as the enthalpy rise nuclear hot channel factor (F )
I 29 OD29V 1
l
i ,.
N maximum intearated rod power
- IAH " average integrated rod power i
Based on the worst case power distribution and manufacturing tolerances allowed during fuel rod fabrication, the limit imposed on F is equal to 1.55. This limit ensures a minimum DNBR of 1.3 at 100 percent power.
F3 " 1 1.55 Similar to the heat flux hot channel factor, F is measured using the incere instrumentation system. A four percent '
penalty is used to account for the fact that the highest power rod in the core may not be measured. Additionally, another four_ percent _
penalty is used to account for non-uniform distribution of flow.
j Using these penalties, the limit on the measured F AH
~
- reduced to i
! N 7
FAH" (measured) = 1 - 1.435 If power is less than 100 percent, the fluid enthalpy rise is lower. With a smaller enthalpy rise, the coolant is further away ,
from a DNS condition and a higher F is allowed. Analyzing I worst case radial and axial flux profiles at various power levels produces the limit ecuation F "Ha 1 1.55 [1 + .2(1 - P)]
where P = fraction of full power The graph of the limit of F as a function of power is shown in Figure FND-HT-101. The shaded area represents the allowed range
- of F a to ensure a DN8R > 1.3.
30 0029V i
I
- .. ,i QUESTION: [1i50]~ 1 2.22 Of the signals used to automatically;close the Feedwater
~
Isolation Valves (1611A, B, C), which two are specifically designed to' prevent water hammer.in the feedwater piping and steam generator inlet connections. Include setpoints and coincidence (logic) where applicable.
ANSWER KEY RESPONSE:
1.- Lo-Lo-Lo S/C level [ 0. 2] -- 2/3 [0.13 --- 5% [0.1].
- 2. The signal which is the coincidence oft [0.1] ;
- a. Lo S/G pressure [0.2] -- 2/3 [0.1] -- 625 psig
[0.1]. x s
- b. Lo feedwater temperature [0.2]'< 225 F [0.13
- c. Lo feed flow [0.2] < 20% [0.1].
REFERENCE:
VCS, TB-7, p.22 SUGGESTED CORRECT RESPONSE:
s Either the above answer or:
- 1. Lo-Lo-Lo S/G level [0.2] -- 2/3 [0.1] -- 5% [0.1].
- 2. The signal which is the coincidence or! [0.1].
- a. Lo feedwater temperature [0.3] < 225*F [0.2]. .
i
- b. Lo feed flow [0 3] < 13% [0.2]. '
REASON: Recent plant changes have removed the lo isteam pressure interlock and changed the lo feed flow setpoint to < 13%.
(See attached B-208-045 electrical drawing, notes ' 10 &
11). Training materials have not been revised to-reflect these changes..
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QUESTION: [2.00]
3 12 a. Describe the instrument coincidence and setpoints necessary to insert the Reactor Coolant low flow trips into the protection system for 1 and 2 loop loss of flow.
~b. Explain how and why the undervoltage and underfrequency low flow reactor trips operate differently. t ANSWER KEY RESPONSE:
- a. 1. Single loop loss of flow will insert if 2/4 power ranges are above 38% (P-8). (0.5).
I
- 2. Two loop loss of flow will insert if 2/4 power ranges
> 105 [0.2] or; [0.1],1/2 impulse pressures > 10%
. [ 0. 2 ] .,
- b. The undervoltage trip is to provide a trip signal on loss-of power to the RCP's. [ 0. 2 ] The reactor will trip and RCP's continue to provide coastdown flow. [0.2] The underfrequency trip provides protection for a grid disturbance. [0.2] The RCP breakers and reactor are tripped [0.2] to prevent deceleration and loss of coastdown flow. [0.2]
REFERENCE:
VCS IC-9, pp. 46, 47 SUGGESTED CORRECTED RESPONSE:
Either the above answer or:
- a. 1. Single loop loss of flow will cause a-trip if 2/3 flow elements in one loop sense less than 90% flow with reactor power above P-8 (38%). [0.5]
- 2. Two ;'op loss of flow will'cause a trip if 2/3 flow .
elementt *n each of two loops sense less than 90%
flow with reactor power above F-10 (10%). [0 5J
REFERENCE:
VCS 10-9, pp. 46 (attached)
^
REASON: Because of the length and wordiness of the question, it lends itself to different interpretations of what is being asked.
- b. While the flow trip provided by the loop flow transmitters supplies the basis protection required when flow is lost in one or two loops, undervoltage and underfrequency sensors associated with each RCP electrical supply provide the more rapid protection ,
required if flow is lost in l l
l
'l
- - .-..s , , . ,.c - .,-c - - . , ,
QUESTION: (3.12b continued) [2.00]
two or more loops. (Reference 10-9, Reactor Protection and Logic, pp. 46, 47, attached ) . These rapid loss of i flow reactor trips are provided to protect against DNB.
Reliance upon the actual loss of flow trip does not provide adequate protection against DNB if flow is lost-i in two or more loops, so the anticipating underfrequency l and undervoltage trips are required. [1.0] (Reference VCS
! FSAR, p. 15.3-6).
l l The above answer is offered as an alternate because the explanation described by question 3.12b could be interpreted as a requirement to explain how and why UV.
and UF reactor trips operate differently from the low flow trips mentioned in part a of the same question.
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- . = . _. _.
- a. ..
Primary Coolant Sysi!em Trips low Reactor Coolant Flow (Sh. 5)
Three redundant differential pressure transmitters are used to measure the flow in the cold leg of each reactor coolant loop. The low-pressure taps for each transmitter penetrate the RCS at the inner piping elbow located downstream of the steam generator primary outlet. The high-pressure taps for each transmitter join at one common line that penetrates the RCS at the outer piping elbow.
Each of the transmitters for a loop has a low-flow bistable associated with it. The bistables are set to trip if the differential pressure being sensed falls to a value that corresponds to less than 90 percent loop flow. If 2 of 3 bistables in a loop trip, a single loop loss of flow signal is sent to the Solid State Protection System. A reactor trip will occur only if 2 of 4 power range channels exceed the P-8 setpoint (38 percent). If a loss of flow con-dition is sensed in more than one loop, a reactor trip will occur if 2 of 4 power range channels exceed the P-7 (P-10 and P-13 ) . Conversely, the reactor will not trip on any condition of low flow when the plant is operating below P-7.
Reactor Coolant Pump (RCP) Undervoltage (Sh. 5)
While the flow trip provided by the loop flow transmitters supplies the basic protection required when flow is lost in_on_e or two loops, undervoltage_sen. sors.
aNociated with each RCP electrical supply provide the more rapid protection required if flow is lost in two or more loops. Each of the RCP electrical buses is supplied with three undervoltage sensing devices. These W sensors are located on the motor side of the RCP circuit breakers to detect a W condition should the breaker (s) open. When all three W sensors associated with one bus sense that pump supply voltage has dropped to 70 percent of normal (4830 volts),
a W signal exists for that bus. Signals from the W sensors are sent through time delay energize devices. This prevents spurious reactor trips caused by !
l short-term voltage perturbations. The W trip is one of the low-power trips.
As such, the trip is automatically blocked when the plant is operating below the P-7 setpoint.
l 46 1155S:4 Rev 2 6/85 l
i i
's ..
_ Reactor Coolant Pump Underfrequency,(Sh. 5)
There is one underfrequency sensor monitoring the frequency on each of the RCP.
The trip is designed to provide fast response to a low-flow condition in two or more loops. A major disturbance on the grid could result in a frequency reduction. Since the RCPs are driven by ac induction motors, pump speed and
[ thus primary flow would be dragged down by such a frequency condition. If the sensor detects an underfrequency of 57.5 Hjz on any of the RCPs it produces
! an underfrequency signal . A reactor trip will occur if reactor power is greater than P-7. '
l The underfrequency condition also causes the RCP breakers to trip. This allows the pumps to achieve the coastdown time duration that is analyzed in the loss of flow analysis. With the plant operating below P-7, the pump bus
~
underfrequency reactor trip is automatically blocked. The signals from the ,
underfrequency sensors pass through time delay energize devices so that short-term frequency dips will not cause a reactor trip.
l
. Overtemperature AT Trip (Sh. 5)
This plant trip prevents the core from approaching the condition of departure from nucleate boiling (DNB). A setpoint is continuously and separately calcu-lated for each of the three loops. The setpoint is in terms of percent of rated full power loop AT. It basically takes into account the factors that affect the margin to DNB. Should the actual loop tenperature difference exceed the present loop setpoint, an OTAT trip signal will exist for that loop. If 2 of 3 loops exceed their setpoint, a reactor trip occurs. The ;
OTAT trip cannot be bypassed.
Overpower E Trip (Sh. 5)
This trip is designed to limit core power and thus minimize fuel tempera-tures . Again, an individual setpoint is continuously calculated for each loop. The setpoint is in terms of percent of full power loop AT. The parameters used to determine the limit are T0 avg? and axial flux difference.
The input from axial flux difference is not presently being used for setpoint 1155S:4 47 Rev 2 6/85. I i l l
15.3.4 COMPLETE IASS OF PORCED REACTOR COOLANT FIAW 15.3.4.1 Identification of Causes and Accident Description A complete loss of forced reactor coolant flow may result from a simultaneous loss of electrical supplies to all reactor coolant pumps. If the reactor is at power at the time of the accident, the_inunediate_effect, of loss of coolant flow is a rapid increase in the coolant _ temptrAtutex This increase could result in DNB with sub_s_equent fuel damage,_i_f_the reactor were not tripped,promptly. The following reactor tr_ip,s_ provide necessary protection against a loss of coolant flow accident:
l
- 1. Reactor coolant pump power supply undervoltage or underfrequency.
i 2. Low reactor coolant loop flow.
The reactor trip on reactor coolant pump undervoltage is provided to protect against conditions which can cause a loss of voltage to all reactor coolant pumps, i.e., station blackout. This function is blocked below approximately 10 percent power (Permissive 7).
The reactor trip on reactor coolant pump underfrequency is provided to trip the reactor for an underfrequency condition, resulting from frequency disturbances on the power grid. Reference (7] provides analyses of grid frequency disturbances and the resulting nuclear steam supply system protection requirements which are generally applicable to current generation Westinghouse plants.
The reactor trip on low primary coolant loop flow is_provided to_ protect, against loss of flow condittons which affect only one reactor coolant loop.
~
This' function is generated ~by two out of three low flow signals per reactor coolant loop. Above approximately 39 percent power (Permissive 8), low flow in any loop will actuate a reactor trip. Between approximately 10 ,
percent power and 39 percent power (Permissive 7 and Permissive 8), low i flow in any two loops will actuate a reactor trip. If the maximum grid frequency decay rate is less than approximacley 2.5 Hz/second this trip function will protect the core from underfrequency events. This effect is fully described in Reference (7].
i Normal power for the reactor coolant pumps is supplied through buses from the unit auxiliary transformer which is connected to the main generator downstream of the main generator breaker. Continuity of power to the pump buses is maintained by backfeed from the main power transformer in case of a trip of the main generator output breaker. On a loss of the main power transformer, the buses are automatically transferred to the emergency auxiliary transformer.
15.3.4.2 Analysis of Effects and Conseeuences J
15.3.4.2.1 Nethod of Analysis This transient is analyzed by four digital computer codes. First the-PHOENIX [8] Code is used to calculate the loop and core flow during the 15.3-8
QUESTION: [0 5]
4.14 During a serious emergency,. operators may be called upon to assist in search and rescue or recovery operations in the-plant.
- a. In such cases, according to Health Physics manual, what dose could your receive:
- 1. To bring an injured worker to safety? [.05]
ANSWER KEY RESPONSE:
- a. 1. 100 REM [0.5]
REFERENCE:
VCS HP Manual, pp. 5-13 SUGGESTED CORRECT RESPONSE:
- a. 1.' 100 REM or.75 REM [0 53 REASON: Emergency Plan Procedure limits on exposure to save a human life differ -from the same exposure limits described in the HP manual.- (Reference EPP-Oll, p. 1 of 5, and EPP-020, p. 2 of.5 -- attached).
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EPP-011 REVISION 4 7/14/86 1.0 PURPOSE To provide guidelines for performing a search and rescue j' operation to find missing individuals.
2.0 REFERENCES
2.1 " Virgil C. Summer Nuclear Station Radiation Emergency Plan."
2.2 NUREG 0654, Criteria for preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants.
l 2.3 EPP-009, "On-Site Medical".
2.4 EPP-020, " Emergency Personnel Exposure Control".
2.5 EPP-008, "Oncite Assembly" -
l 2.6 EPP-012, "Onsite Personnel Accountability and Evacuation" l l
2.7 EPP-016 " Emergency Pacilities Activation and Evacuacion" !
($)' 3.0 CONDITIONS AND PREREQUISITES 3.1 Personnel accountability nas been performed and it has been determined that a person on-site may be missing. This i procedure assumes that a missing person (s) is injured until proven different.
3.2 At least one memoer of the Search and Rescue Team will be qualified in Health Physics monitoring techniques and one i
member will be at least Multi-media First Aid qualified. I 3.3 The leader of the Search and Rescue Tbam should be the "B" Nuclear Reactor Operator or his designee as availaole.
Other members of the Search and Rescue Team should be:
available Auxiliary Operators, Maintenance Personnel, Chemistry Pe'esonnel, etc. , as per EPP-016 Attachment II.
3.4 If necessary, the members of the Search and descue Team mag, be authorized to receive 75 rem to save the life of the missFn~g fifdividdalfsT~iiis per Reference 2.4.
PAGE 1 0F 5
H- -
l EPP-020 REVISION 6 5/27/86 3.2.1 Normal Administrative exposure limits =ay be suspended or modified verbally by the Interim 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 Federal limits of 10 CFR 20 will not knowingly be exceeded, except as described in Section 3.2.4 of <
this procedure.
l 3.23 Maintain exposure ALARA. )
J 3.2.4 Planned exposures of up to 75 Rem to save human life l
~or prevent the serious endangerment of_ human lifA or up to 25 ReE to save or delgate 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 SCE&G employees who are volunteers broadly familiar l with exposure consequences. '
l NOTE: This does not apply to " spontaneous reactions" by on the scene personnel in a threatening situation. ,
O 3.2.5 Non-company employees such as medical support i
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
- i that rotation can not be acomplished. 1
- 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).
PAGE 2 0F 5 ,
I
QUESTION: [0.25]
4.15 c. To what value can the Administrative Guideline for whole body be raised to and who (by title) must approve this limit increase?
A"SWER KEY RESPONSE:
, c. 2 R/qtr.:not to exceed 5 R/yr. [0.25]
i Plant Manager j
REFERENCE:
VCS HP Manual,, pp. 5-13 SUGGESTED CORRECT RESPONSE:
Either the above answer or:
- c. 2.5 R/qtr. not to exceed 5 R/yr. [and 5(N-16)J. [0.25]
} Director, Nuclear Plant Operations i REASON: Health Physics Procedure - 153, Administrative Exposure
- Limits, contain'the 2.5 R/qtr. limits. (See attached HPP-153, p. 3, and Attachment I, p. l'of-1). Also, the title Plant Manager has been changed to Director, Nuclear.
1 Plant Operations. (Attached reference HP-153, Attachment I. p. 1 of 1).
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HPP-153 REVISION 5 10/9/85 d
4.0 GENERAL 4.1 The current quaEterly allowed exposure for each individual is listed on up-to-date printouts available in the Health Physics Lab and at key locations in the plant.
- 4. 2 The limits addressed within this procedure shall be enforced via the use of Self-Reading Dosimeter Cards, as describ ed in HPP-152. These cards reflect current accunulated quarterly exposure received and the remaining ,
( quarterly exposure allowed.
4.3 Additional controls established to ensure that administrative limits are not exceeded include:
, 4.3.1 Review of all current exposure records by Health Physics.
j 4.3.2 Computer printouts (" Daily Dose Printout") of l
individual exposures for personnel to review. .
4.3.3 Reports forwarded to Managers and Supervisors t6 inform them of personnel who are approaching quarterly exposure Ibnits.
1
- , 4.4 Health Physics performs audits of exposure extension i
(s%) requests to identify cases where supervision is requesting
? higher extensions of administrative limits than necessary for the performance of the work. Upon identification of such cases, the responsible supervisor will be contacted to justify these extension requests.
5.0 LB4ITS AND PRECAUTIONS 5.1 Exposure limits for new personnel on site may vary depending on the individuals prior site, current quarter exposure and will be set as described in Reference 2.1.6.
Without an extension, the administrative limit for whole l 5.2 l body exposure received at the V. C. Summer Station is 1.0 ' )
- Rem /qtr. Female employees without a doctors verification t )
- of infertility will be 10mited to 500 mrem /Qtr.
5.3 Quarterly Administrative Limits are reviewed to ensure no, individual will exceed 2500 mRC4 during a quarter arter ,
_ previous site dose is added to the V.C. Summer Station (
allowable Administrative Dose for the quarter. The 1
-l Quarterly Administrative Ltnits are also reviewed to ensure l
- j
!' the annual administrative limits (5000 MREM for SCE&G 1
! personnel-'aiiF1FOT6 iiiREM for contractors) and liretime l
p)
( exposure limits (5 x (age-18)) are not exceeded. ;
'j PAGE 3 0F 5 GENERAL REVISION l
(
e aw e ._ e 4 ese o em
~-"*-"-"****N6" j
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l HPP-153 ATTACHMENT I PAGE 1 0F 1 REVISION 5 AUTHORIZATION TO EXCEED QUARTERLY ADMINISTRATIVE EXPOSURE LIMIT Added Exposure needed by Da t e/ Tim e '
Nrmer SSN:
Company: TLD: DOB(N):
g TM/ MUN/U A Y N Current V.C. Summer Qtr. Limit: mR I
T This request is to extend the individual's whole body exposure tot mR I
A YE S NO b T . .
} O 1 1 I L_J Is this extension or the Administrative Whole Body Ex po sure limits absolutely i
R . . . . necessary?
2 I I I I Has every reasonable effort been made to maintain this individuals exposure to within his current administrative limit?
3 I I I I Has ex posure been distributed am ong workers to the extent possible?
t 4 i i i i Are other qualified individuals with less ex po sure available?
5 I i i t Will this exposure extension minimize collective ex po sure?
Purpose:
Associate Man ag e r/ Supe rv iso r Date/Ilme Total Exposure Cur r ent to: Verified By:
Uste/Ilme
~
This qtr.
L this site mR mR T previous site. mR H To t al mR 2500 mR -
P This year (excluding c'ur r e n t qtr.) mR mR H Previous Ye ar s ~
mR Y Total Lifetime to date mR mR 5 (N -18 ) '
S Of Current Quarter To t al mM rocket Dosimeter Latimate mR TLD I
'C Co uld this exposure result in the individual receiving greater than S 5000 mR for the current year (including this qtr) yes no If the answer is yes, then the initiator will complete and include At t a chm ent II.
S This individual's Whole Body Exposure Limit is extended to mR ;
I C
N s-A In d iv id ual Da te/ line Associate Manag er , He al t h Physics ate /ilme T
U R Supervisor Daterline Group Manager Date/Ilme E
m Associate Manag er/ 5upervisor of Ops Date/line Director, Nuclear Plant 0g3/ _
Date/Ilme
- Manager Da te/ lim e Vice President, Nuclear Operations Da te/ Ilm e w- -- - - , - , - - - - - . ---n , , , , , , - - .
QUESTION: [ 2.0]
~
4.16 According to SOP-401, Reactor Protection and Control System, list four' of the five Instrument Channels that need not be -
placed in the trip mode when they are declared to be ou t-of-se rvice .
ANSWER KEY RESPONSE:
- 1. Source Range Excores
- 2. Intermediate Range Excores
- 3. Containment Pressure (High-3)
- 4. RWST Levels 5 Emergency Feedwater Suction Pressure Low.
[any 4, 0.5 each]
REFERENCE:
VCS, SOP-401, p. 4 SUGGESTED CORRECT RESPONSE:
- 1. Source Range Excores
- 2. Intermediate Range Excores 3 Containment Pressure (High-3)
- 4. RWST Levels 5 Emergency Feedwater Suction Pressure Low
- 6. RCS Flow Protection (Underfrequency)
- 7. RCS Flow Protection (Undervoltage)
- 8. Control Channel THOT RTD 9 Control Channel TCOLD RTD ,
2
- 10. Pressurizer Pressure Control l
[any'4 8 0.5 each]
REASON: (See attached SOP-401 pages). None of the additional responses (6-10) require tripping-bistables.
_ . . . ~ - . . - , , _ ._ _ , _ . __ _ .
- r. _
SOP-401 REVISION 6 2/21/do B. CONTROL CHANNEL HOT LEO RTD FAILS LOW 1.0 SYMPTOMS 1.1 Abnormally low indication on one of tne following indicators:
- a. TI-411A, T-A/G*F.
- b. TI-421A , T-AVJ'F.
c.. TI-431A, T-AVG'F.
1.2 Aonormally low indication on one of the following indicators.
- a. TI-411B, T 4.
- b. TI-4218, T %. ,
- c. TI-4313, T%.
13 dCS TAVG DEV dI/LO annunicator.
1.4 RCS T DEV HI/LO annunicator.
2.0 AJTOMATIC ACTIONS None 3.0 IMMEDIATE CORRECTIVE ACTIONS 31 verify instrument failure by comparing control channel T and Tavs indication with Protection Channel T and Tavs indication for the faulty loop.
PAGE 15 0F b3
SOP-401 REVISION 6 2 /21/do 4.0 FOLLOWGP ACTIONS 4.1 Place dI f AdC INdI8IT switch to the loop wnien nas tne failed instrument.
4.2 Place HI TAVG AJC IddIBIT switch to the loop witn the failed instrument.
NOTS 4.3 RCS T DEV HI/LO annunicator will not clear and the indications will remain abnor.nal until the channel is repaired.
43 Confirm that RCS TAVG DEV HI/LO annunicator clears.
4.4 Have the failed channel repaired and verify proper '
operation prior to returning it to service.
PAGE 16 0F 63
802-401 dd/ISION 6 2 /21/do D. CONTROL CHANNEL COLD LEG RTD FAILS LOW l.0 SYMPTOMS 1.1 Abnormally low indication on one of the followind indicators:
- a. TI-411A, T-AVG *F.
- b. TI-421A, T-AVG *F.
- c. TI-431A, T-AV3*F.
1.2 ADnormally nigh indication on one of the followind indicators:
- a. TI-4113, T%.
- o. PI-421B , T %.
- c. TI-4313, T %.
1.3 Hign readings on one pen of ZR-403, ROD POSITION /dOD INSERTION LIMIT Recorder.
1.4 RCS Tavg DEV HI/LO annunicator.
15 dCS T DEV dI/LO annanicator.
1.6 CRB INSERT LMT LO annunicator.
1.7 CRs IdSERT LAT LO-LO annunicator.
2.0 AUTOHATIO ACTIONS None i
PAGE 20 0F 63 1
S0P-401 RSVISION 6 2/21/86 3.0 IMMSOI ATE C0ddSCTIVE ACTIONS 3.1 Verify instrument failure by comparing Control Cnannel T and Tavs indication with Protection Channel T and Tavs indication for tne faulty loop.
4.0 FOLLOWUP ACTIONS 4.1 Place HI T AUC INHIBIT switch to the loop which has tne failed instrument.
4.2 Place HI TAVG AUC INdIBIT SWITCH to tne loop witn the failed instrument.
NOTE 4.3 RCS T DEV HI/LO annunicator will not clear and
- the indicators will not return to normal until the channel is repaired.
4.3 Confirm that RCS Tavs DEV HI/LO.
4.4 Have the failed enannel repaired and verify proper operation prior to returning to service.
PAGE 210F 63
S0P-401 dEVISI0d 6 2/21/86 I. PRESSURIZER PRESSURE CONTROL CHANNEL PAILS HIGH 1.0 SYMPTOMS 1.1 Abnormally hign indication on one pen of PR-444, PZR PSIG RECORDER.
1.2 Abnormally high indication on one of the following indicators:
- a. PI-444, PZR PRESS CNTRL CHAN.
b.' PI-445, PZR PRESS CMTRL CHAD.
13 PZR CNTRL PRESS dI annunicator.
1.4 PZR PRESS dI/LO annunicator.
1.5 PZR PRESS dI annunicator. .
1.6 PZR PCS HI annunicator.
LO AUTOMATIC ACTIONS 2.1 If channel PT-445 failed, the following valves open if in AUTO:
- a. PCV-445A, P4R RELIEP.
- o. POV-4458, P4R RELISP.
2.2 If channel PT-444 failed, PCV-4443, P4d RELId? opens if in A'JTO .
23 If channel PT-444 failed, tne folic wing spray valves open if in AUTO.
- a. PCV-4440, PZR SPRAY.
- c. POV-4440, PZR SPRAY.
PAJE 34 0F 63 J
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- 'o' SOP-401 dEVISION 6 2/21/66 i
NOTE 3.0 When pressurizer pressure reaches 1985 psig, automatic reset or manual block on high pressurizer pressure (P-11) will shut the power relier valves.
2.4 It channel PT-444 railed, the following heaters turn off:
- a. BU GRP 1.
- b. 80 GRP 2.
- c. CNTdL GRP.
3.0 IMME0 Card C0dddCTIVE ACPIONS 31 Verity the failure of the Control channel oy comparing PI-444, PZR PRESS or PI-445, PZR PRESS To the following:
- a. PI-455, PRESS PSIG.
- b. PI-456, PRESS PSIG.
- c. PI-457, PRESS PSIG.
32 It channel PT-445 nas failed, close the following valves:
l
- a. POV-445A, P4R dELIEP.
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- b. PCV-4453, PWR RELIEP.
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. . . . . *o SOP-401 dEVISION 6 2 /21/d o 1
33 If channel PT-444 nas failed, do the followind;
- a. Close POV-4448, P4R RELIEF.
- o. Set PZR PRESS dASTER CONTROL in MAN, and operate to re-establish Reactor Coolant System pressure at 2235 115 psig using functioning pressure enannels.
4.0 POLLOWUP ACTIONS 4.1 If channel PT-445 has failed, verify that the Reactor Coolant System pressure has increased to 2235 115 psig by automatic action of enannel PT-444.
4.2 If channel PT-444 has failed, verify that the Reactor Coolant System pressure has been re-established and can be maintained by manual operation of PZR PRESS dASTER CONTROL.
4.3 Have the failed channel repaired and verify proper
- operation prior to returning it to service.
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- 4'.' SOP-401 l REVISIOd o 2/21/8o J. PRESSURIZER PRESSURE CONTROL i CHANNEL FAILS LOW l 1.0 -SYMPTOMS 1.1 Abnormally low indication on one pen of PR-444, PZR PSIG RECORDER.
1.2 Abnormally low indication on one of the following indicators:
- a. PI-444, PZd PRESS CNTdL CdAN.
- b. PI-445, PZR PRESS CNTRL CHAN.
13 PZR PRESS dI/LO annunicator.
1.4 PZR PCS LO 80 DTRS ON annunicator.
2.0 AUTOMATIC ACTIONS NOTE 2.1 Wnen channel PT-445 indicates a pressare of 2335 psig, the following valves will open:
- a. POV-445A, PWR RELISP
- b. POV-445B, PWR RELIEP 2.1 If channel PT-444 fails, both control and backup pressurizer heaters will come full on.
2.2 If channel PT-445 fails, no automatic action will tEce place.
PAGE 37 0F 63 l
- .t'o' , SOP-401 RSVISION 6
, 2/21/06 3.0 IMMEDI ATE C0ddECTIVE ACTI0dS 3.1 verify tne failure of tne Control Channel by comparing PI-444, PZR PRESS or PI-445, PZR PRESS to the following:
- a. PI-455, PRSSS PSIG.
- o. PI-456, PRESS PSIG.
- c. PI-457, PRESS PSIG.
3.2 If channel PT-444 nas failed, set PZR PRESS i4ASTSR CONTROL to MAN and operate to re-estaolish Reactor Coolant Systeta pressure at 2235215 psig using functioning pressure channels.
4.0 FOLLOWUP ACTIONS 4.1 If enannel PT-444 has failed, verify that tne Reactor Coolant System pressure of 2235115 psig nas been re-established and can be maintained by manual operation of PC-444A, (PZR PRESS MASTER CONTROL).
4.2 dave the failed channel repaired and verify proper operation prior to returning to service.
PA35 38 OP 63
k SOP-401 ATTACHMENT I PAGE 3 0F I4
, a REVISION 6 i INSTRUMENT F AI LURE REFERENCE MANUAL j
I NSTRUMENT SISTA8tES BlSTABLE SYSTEN FUNCTION NO. ASSOCI ATED LOCATI ON TRIP STATUS LIGHT TECH SPECS STP j 111.F. RCS FLOW FT-426 F8-426A C3-733-GS-1 CHAN 111 RCS LPS FLO LO TABLE 2.2-1 ITEM 12 302.017 i PROTECTION YmLE 3.5-1 I TEM 12 345.013 (LOSS OF FLOW) i TA8tE 3.5-2 ITEN 82 OATC RCS FLOW LOG III.G. FT-434 F8-434A C l-434-85-1 CHAN I RCS LPC FLO LO TA8LE 2.2-1 ITEM 12 302.012 PROTECTION TA8LE 3.5-1 ITEM 12 345.008
, (LOSS OF FLOW)
TA8LE 3.5-2 ITEM 12 04TC l LOG 111.H. RCS FLOW FT-435 F8-435A C2-734-85-1 CHAN 11 RCS LPC FLO LO TA8LE 2.2-1 ITEM I2 302.015 PROTECTION ,
TABLE 3.5-1 ITEM 12 345.011 (LOSS OF FLOW)
TABLE 3.5-2 ITEM 12 OATC LOG III .I . RCS FLOW FT-436 FS-436A C3-73445-1 CHAN lli RCS LPC FLO LO TA8LE 2.2-1 ITEM 12 302.018 4
PROTECTION
' TABLE 3.5-1 1 TEM 12 345.014 (LOSS OF FLOW)
TABLE 3.3-2 ITEM 12 OATC
,A LOG lit.J. RCS FLOW 81 (RCPA) XPN-60ll
/N/A CHAN l IEP SUS UF TABLE 2.2-1 ITEM 16 506.003
, PROTECTI ON 81 (RTS) [ N/A MPN-6012 CHAN ll RCP SUS UF TABLE 3.5-1 ITEM 16 4 506.007 (UM)ER FREQUENCY) 81 (RCPC) N/A XPN-6013 CHAN III fCP SUS UF TABLE 3.5-2 ITEM 16 Ill.K. RCS FLOW 27 (RCPA) M/A XPN-60ll CHAN I RCP BUS UY TABLE 2.2-1 ITEM 15 506.002 PROTECTION 27 (RCPB) NPN-6012
.{ N/A CHAN 11 RCP BUS UV TA8LE 3.3-1 ITEM 15 506.006 (UM)ER VOLTAGE) 27 (RTC) \ 1(/A MPN-6013 CHAN III IEP BUS IN TABLE 3.5-2 ITEM 15 IV.A.
1 I
TAVE TE-412 8/D M2-8-8 C 1-421-85-1 CHAN I LPA (P T TABLE 2.2-1 1 TEMS 7, 8 302.001
. PROTECTI ON 78-412 4 -2 Cl-42145-2 CHAN I W/DRWL ROD 8LCK TASLE 3.3-1 ITEMS 7, 8 345.001 TB-412-C-I C l-421-65-3 CHAN 1 LPA OT T TA8LE 3.5-2 ITEMS 7, 8 TB-412-C-2 Cl-42345-4 CHAN I W/DRWL ROD BLCK TABLE 3.5-4 ITEMS 4.d, i
TB-412-0-1 Cl-422-85-2 CHAN I RCS LPA TAVG LO
! 9.b TB-412-0-2 C l-42245-3 NONE TABLE 3.5-4 ITEMS 4.d, TB-412E Cl-422-GS-1 CHAN I LPA P-12 9.b TABLE 3.5-5 I TEM 8 i .
l d
e
77 ,
.: .ofysjw.
QUESTION:- 6.12L (same as. Question:2.22)
~. QUESTION: 7 16a.1- (same'as Question 4.14a.1)
QUESTION:: 7.17c '(same'as Question.4.15c)-
QUESTION: 7.18' (same as Quescion'4.16)-
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L
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1 1 -
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t l
4
. , , , ,_ . , , . _ . , e, , . . , , - . . . ~ e, ' _ . "