Similar Documents at Ginna |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
IR 05000244/19990081999-10-14014 October 1999
Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
IR 05000244/19992011999-09-24024 September 1999
Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
1999-09-30
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999
Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated
ML17265A7651999-10-0808 October 1999
Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request
ML17265A7631999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000
ML17265A7641999-10-0505 October 1999
Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel
ML20212J3561999-09-30030 September 1999
Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21
ML20212J3801999-09-30030 September 1999
Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21
IR 05000244/19990051999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7571999-09-24024 September 1999
Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs
ML17265A7461999-08-31031 August 1999
Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726
ML17265A7401999-08-26026 August 1999
Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components
ML17265A7411999-08-26026 August 1999
Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded
ML17250B3021999-08-23023 August 1999
Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806
ML17265A7451999-08-23023 August 1999
Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d)
ML17265A7271999-07-30030 July 1999
Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply
ML17265A7151999-07-23023 July 1999
Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions
ML17265A7061999-07-22022 July 1999
Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations
ML17265A7191999-07-21021 July 1999
Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999
ML17265A7141999-07-21021 July 1999
Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future
ML17265A7041999-07-16016 July 1999
Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603
ML17265A6911999-06-30030 June 1999
Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached
ML17265A6871999-06-22022 June 1999
Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl
ML17265A6841999-06-21021 June 1999
Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment
ML17265A6741999-06-15015 June 1999
Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT
ML17265A6751999-06-11011 June 1999
Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification
ML17309A6551999-06-0707 June 1999
Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment
ML17265A6671999-06-0101 June 1999
Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301
ML17265A6641999-05-25025 May 1999
Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested
ML17265A6611999-05-24024 May 1999
Forwards LER 99-007-00 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Rt.Util Is Planning to Submit Suppl to LER by 990730
ML20206U0921999-05-13013 May 1999
Forwards Four Copies of Rev R to Gnpp Security Plan,Per Provisions of 10CFR50.54(p).Rev Clarifies Armed Response Team Assignments & Does Not Degrade Physical Security Effectiveness.Rev Withheld,Per 10CFR73.21
ML17265A6461999-05-12012 May 1999
Forwards 1998 Annual Radioactive Effluent Release Rept & 1998 Annual Radiological Environ Operating Rept, for Re Ginna NPP
ML17265A6411999-05-12012 May 1999
Forwards LER 99-004-00 IAW 10CFR50.73 & 10CFR21.Further Assessment Will Be Provided in Suppl to LER by 990630
ML20206E7221999-04-29029 April 1999
Forwards Four Copies of Rev Q to Re Ginna Nuclear Power Plant Security Plan,Per 10CFR50.54(p).Changes Do Not Degrade Physical Security Effectiveness.Encl Withheld,Per 10CFR73.21
ML20206H6911999-04-22022 April 1999
Forwards Info Requested During Informal Telcon on 980408 Concerning Upcoming Osre at Ginna Station.Info Requested Listed.Without Encls
ML17265A6271999-04-19019 April 1999
Forwards Rev 0 & Rev 1 to Colr,Cycle 28 for Re Ginna NPP, Per TS 5.6.5
ML17309A6501999-04-14014 April 1999
Forwards Revised Ginna Station EOPs & Procedures Index
ML17265A6121999-03-29029 March 1999
Forwards Rept on Status of Decommissioning Funding for Re Ginna Npp,For Which Rg&E Is Sole Owner,Per 10CFR50.75. Data Presented Herein,Current as of 981231
ML17265A6041999-03-24024 March 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(ii)(B) & 10CFR21.Addl Analyses Are Being Performed to Support Future Cycle Operation & Supplemental LER Is Scheduled to Be Submitted by 990618
ML17265A5641999-03-0101 March 1999
Forwards Response to NRC 990217 RAI Concerning Changes to QA Program for Re Ginna Station Operation.Rg&E Is Modifying Changes Requested in 981221 Submittal.Modified QA Program,Encl
ML17265A5671999-03-0101 March 1999
Forwards Application for Amend to License DPR-18,to Revise TSs Battery Cell Parameters Limit for Specific Gravity (SR 3.8.6.3 & SR 3.8.6.6).Supporting Tss,Encl
ML17265A5551999-02-25025 February 1999
Informs That Util Is in Process of Revising fitness-for-duty Program,Developed in Accordance with 10CFR26.Util Will Continue to Use Dept of Health & Human Svcs Certified Test Facility for Majority of Tests During Yr
ML17265A5561999-02-22022 February 1999
Forwards FFD Performance Data Rept for Six Months Ending 981231,per 10CFR26.71(d)
ML17265A5451999-02-12012 February 1999
Forwards Simulator Four Year Certification Rept,Per 10CFR55.45(b)(5)(ii)
ML17309A6491999-02-12012 February 1999
Forwards Ginna Station EOPs
ML17265A5431999-02-0909 February 1999
Supplements 980806 Relief Request with Attached Table.Util Third 10-Yr ISI of Reactor Vessel Being Performed During 1999 Refueling Outage,Beginning on 990301
ML17265A5361999-02-0202 February 1999
Forwards Response to NRC 981203 RAI Re Resolution of Unresolved Safety Issue USI A-46.Util Does Not Agree with NRCs Interpretation.Detailed Bases,Encl
ML17311A0691999-01-25025 January 1999
Forwards Revs to Ginna Station Emergency Plan Implementing Procedures (Epips).Previous Rev Had Incorrect Effective Date
ML17311A0671999-01-14014 January 1999
Forwards Revised Emergency Operating Procedures for Re Ginna NPP
ML17265A5131999-01-12012 January 1999
Forwards Revised Cover Page for Ginna Station Technical Requirements Manual (Trm), Rev 7,correcting Error in Effective Date
ML17265A5111999-01-11011 January 1999
Requests Relief Per 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME Bp&V Code for ISI Program.Relief Requests 37,38 & 39 Encl
ML17265A5101999-01-11011 January 1999
Requests Relief Per to 10CFR50.55a(a)(3)(ii) from Certain Requirements of Section XI of ASME B&PV Code for ISI Program.Relief Request 40 Encl
1999-09-30
[Table view]
Category:RESEARCH INSTITUTION/LABORATORY TO NRC
MONTHYEARML17261A8541989-01-13013 January 1989
Discusses 890110-11 Meetings Re Nozzle Sizing Study.Most Accurate Sizing Obtained by Collecting Data of Edge Diffracted Waves from Geometric Extremities of Flaw
ML20212N6571986-08-19019 August 1986
Forwards EGG-PHY-7350, Technical Evaluation Rept for Proposed Change to Operation of Ginna Fuel Pool Charcoal Filter Sys, Per FIN D-6023, Review of Plant-Specific Licensing Actions for Operating Reactor Issues, Task 4-5
ML17251A7281986-06-0404 June 1986
Forwards Inel EGG-NTA-7266, Assessment of Removal of Raised Floor & Mod to Fire Detection Zones for Plant Process Computer Sys (PPCS) Computer Room at Ginna Plant, Final Informal Rept Per Task 4-1, Review of Plant-Specific....
ML19306A0151984-02-10010 February 1984
Summarizes 840206-07 Visit to Nevada Test Site Hydrogen Burn Facility to Inspect Condition of Equipment & Cable/Splice Samples After Series of Hydrogen Burn Tests Conducted by Epri.No Indication of External Damage to Equipment Noted
ML17256B1011982-06-21021 June 1982
Comments on NUREG-0821, Integrated Plant Safety Assessment, SEP Re Ginna Nuclear Power Plant. Document Is Comprehensive W/Respect to Arguments Leading to Resolution of Various SEP Topics
ML20063C5531982-06-14014 June 1982
Compares Format & Positions of Facilities for Integrity Plant Safety Assessments Per SEP Objectives.Wind & Tornado Loadings Need Improvement.Pipe Break Outside Containment Requires Clarification
ML20065A7311982-03-23023 March 1982
Informs That Info Submitted by Licensee in Response to NUREG-0737,Item III.D.3.4, Control Room Habitability Acceptable
ML17258A5181981-10-30030 October 1981
Forwards Progress Rept Re Open Items for Unit.Remaining Item Is Primary Coolant Pump Casing
ML17258A1941981-09-30030 September 1981
Forwards Rept Addressing Resolution of Open Items in Mechanical Electrical Equipment SEP Re Seismic Integrity,Per 810909 Meeting
ML17258A5201981-09-11011 September 1981
Forwards Info Re Seismic Capacity of Reactor Coolant Pump. Action Proposed to Close Item
ML19345D6991980-11-24024 November 1980
Submits Monthly Mgt Ltr 8 for Oct 1980 Re Containment Analysis Support for Sep.
ML17250A7611980-06-16016 June 1980
Discusses Completed mini-review of Util Safe Shutdown Analysis.Impossible to Confirm Safe Shutdown During or After Fire.Suggests NRC Require Installation of More Conservative Shutdown Sys
ML17261A1701980-01-31031 January 1980
Forwards Input to Fire Protection Design Review & Supplemental Items.Items 3.1.24,24 & 43 Remain Open Due to Lack of Info from Licensee
ML17249A8831980-01-23023 January 1980
Recommends That Licensee Proposal on Fire Protection Review Item 3.2.6 Re Backflow Protection Be Accepted Subj to Review of Four Listed Documents.Also Recommends That Fire Brigade Size Be Increased to Five Men
ML17261A1691980-01-23023 January 1980
Discusses Util Fire Protection Review Items 3.2.6 & 3.2.9. Recommends Accepting Backflow Protection Proposal Pending Submittal of Prints of Drain Sys & Problem Drain Location. Five-man Fire Brigade Size Acceptable
ML17249A2391979-10-30030 October 1979
Informs of Util 790525 Agreement to Increase 3-man Fire Brigade to Five Men as Recommended in J Townley 780710 Rept & Justified in 790214 Safety Evaluation.Item Now Deemed Acceptable & Closed
ML17244A3171978-12-19019 December 1978
Proposed Fire Protection as Presented in SER Adequately Assures Health & Safety of Public.Exception Re Valve Supervision & Cable Protection Must Be Evaluated by Nrc.Fire Protection Milestones & Documents Are Discussed
1989-01-13
[Table view] |
Text
LawrenceLivermore NationalLaboratory September ll,1981SM81-247Docket850-244FINA0415Mr.WilliamT.Russell,BranchChiefSystematic Evaluation ProgramBranchDivisionofLicensing OfficeofNuclearReactorReg.Washington, D.C.20555
DearBill:
Ihaveenclosedinformation regarding theseismiccapacityofareactorcoolantpumpattheGinnaplant.Whilethisopenitemhasnotbeencompletely
- resolved, arathersimpleactionisproposedwhichshouldclosethisitem.Sincerely, TAN/mg0131mEnclosure ThomasA.NelsonStructural Mechanics GroupNuclearTestEngineering Divisionsooaoqoo79 saoiisPDRADOCK05000244IPDRAnEndleaf~tnffyEl~NiVeetyef CaifemiaPQBaX808LimmesCaifemia94550~TelephOne(415)422-1100
~TWX9f0-38&8339 UCLLLLWR ii11IEPP.'I STRUCTURAL IECHAfllCS ASSOClATESACollf.Coro.5160BirchStreet,NewportBeach,Caiif.S2660(714)833-7552SMA12205.20August27,1981Mr.ThomasA.Nelson(L-9Q}LawrenceLivermore Laboratory NuclearTestEngine'ering DivisionP.0.Box08Livermore, California 94550P
Subject:
Resolution ofOpenItemsonGinnaEquipment l
DearTom:
!nmyJune22letter,therewerestilltwoopenitemsforGinnaequipment whichwere:1.ControlRodDrivesand.Supports 2.PrimaryCoolantPumpThisletteristoadviseyouofthecurrentstatusoftheseitems..GeorgeWrobelofRochester Gas&Electric(RGE}CompanyarrangedformetotalkdirectlytoRobertKellyof'Westinghouse, whowasresponsible forproviding muchoftheGinnaequipment seismicdocumentation.
Regarding loadingdocumentation ontheCRDassemblies, R.KellywillsendhisanalysisofCRDhousingandseismicsupportloadingfor'a0.8gstaticcoefficient loading.Aquickreviewoftheanalysisshouldresolvealloutstanding issues.ontheCRDsystem.Alegiblecopyoftheprimarycoolantpumpreportwastransmitted tomefromRGE.Conclusions fromthereviewofthatreportarecontained intheattachment, TomChengsuggested thatIbepresentatanupcomingopenitemsmeetingforGinnainSeptember.
InviewofthecurrentstatusofopenitemsforwhichIhavebeenresponsible, Idon'tfeelitnecessary formetoattend;the=
meeting.Aftermyconversation withR.KellyofWestinghouse, IfeelthattheCRDandprimarycoolantpumpissueswillbe.resolvedvery'quickly andcanbehandledwithoutameeting.Itis.therefore recommended thatthese'temsnotbeatopicofthemeeting.rYerytrulyyours,STRUCTURAL MECHANICS ASSOCIATES, INC.rZWRobertD.CampbellProjectManagerRDC:lcacc:T.Cheng{NRC)
~~ATTACHMENT REACTORCOOLANTPUMPSEISMICDESIGNREVIEMTheoriginalstressreportsubmitted forreview,wasillegible.
Amorelegib1ecopywassubmitted toSMAbyRGEonAugust17,1981,andthefollowing conclusions canbereachedfromreviewofthatsubmittal.
I~1.Thepumpwasanalyzedfora0.8ghorizontal staticcoef-ficientanda0.54g.vertical staticcoefficient.
Asreportedin.NUREG/CR-1821, the7Xdampedpeakspectralacceleration forbothhorizontal directions is0.55g'sresulting inavectorsumof0.78g's.Thus,theequiva-Ilentstaticcoefficient usedintheoriginalanalysisisconservative byasmallmargin.2.'llstressescalculated fdrthe0.8gHand0,54gVstaticcoefficients aiewithinallowables.
designated fortheoriginaldesignbasis.3.Analytical methodsusedinthedesignanalysisarereason-ableexceptintheca'seofpumpnozzles.4.Thepumpnozzlesanalysisisunrealistic andinadequate forthefollowi.ng reasons:a.Onlythestraightpipeportionofthenozzleswereevaluated.
Localmembranestressesinthepumpcasingwerenotcomputed.
ATTACHHENT Continued b.Thederivedpumpnozzleloadshavenoresemblance toactualachievable loads.Thepumpwasassumedtobesupported bythepipingforpurposesofderivingnozzleloads.Thisisprobablyhighlyconservative butnotnecessarily so.Actualpipingreactions areavailable andshouldbeusedinanevaluation ofthepumpcase.Aconversation betweenR.KellyofMestinghouse andR.CampbellofSMArevealedthat:.i1.TheSanOnofrepumpsareverysimilartotheGinnapumpsandthatadetailedfiniteelementanalysiswasconducted fortheSanOnofreUnitsforaspecified setofnozzleloads.2.AloopanalysisofGinnawasconducted byMestinghouse forseismicloading.Actualpumpnozzleloadsareobtainable fromtheanalysis.
Reconmended ActionsRGEshouldhaveVestingfiouse makeacomparison ofGinnavsSanOnofrepumpcasinggeometryandpumpnozzleloadsandscaleresulting stressesfromtheSanOnofrepumpanalysisforGinnanozzleloadingconditions.
Theload/stress comparison andacomparison ofnozzleandcasinggeometryshouldbesufficient todemonstrate seismiccapability oftheGinnaprimarycoolantpumps.
81 p~~r-STRUCTURAL lTIECHAAICS ASSOCIATES ACalif.Corp.5160ErchStreet,NewportBeach,CaN.92BBO(714)833-7552.
SMA12205.20June22,1981Mr.ThomasA.Nelson(L-90)LawrenceLivermore Laboratory NuclearTestEngineering DivisionP.0.Box808Livermore, California 94550
DearTom:
IntheReference 1submittal regarding reviewofopenitemsonRobertE.GinnaNuclearPowerPlant,therewerethree(3)openitemsstillremaining.
1.ControlRodDriveMechanism 2.ReactorCoolantPump3.SteamGenerator TubeSupportsITomChengoftheUSNRCcontacted meonJune16,toinquireifuseofa,sitespecificspectraforGinna,inlieuoftheregulatory guidespectrumanchoredto0.2g,wouldeliminate theoustanding items.'indicated thatfortwoofthethreeoutstanding items,lackofinformation, notmarginalstressconditions, wastheprincipal concern.'s aresultoftheconversa-tion.Ihavereexamined thethreeopenitemsandcaneliminate thesteamgenerator tubesupportoverstress problembyusingLevel0Service(faultedcondition) allowable stressesandscalinguptheoriginalanalysisresultstoapplicable floorspectralaccelerations.
Thecontrolroddriveandprimarycoolantpumpitemsstillremainopenbutshouldbeeasilyresolvedwithsubmittal ofthenecessary information.
Theremaining actionitemsandresolution ofthesteamgenerator tubesupportstressproblemaresum-marizedintheattachment.
Verytrulyyours,STRUCTURAL MECHANICS ASSOCIATES, INC.RobertD.CampbellProjectManagerRDC:lcaAttachment cc:J.Stevenson ATTACHMENT SUMMARYOFOPENITEMSFORROBERTE.GINNANUCLEARPOMRPLANTASPARTOFTHESEPPROGRAMCONTROLRODDRIVEMECHANISM Thesubmittals providedbythelicensee, References 2and3,donotcontainacorrelation betweenloadingusedintheanalysesandaccel-erationsattheRPVsupport.Aconclusion regarding seismicresistance oftheCRDsystemcannotbereachedwithoutsuchacorrelation.
Mesting-houseshouldbeabletosupplythenecessary information toRochester GasandElectric.
REACTORCOOLANTPUMPSTheReactorCoolantPumpStressReportsubmitted forreview,Reference 4,isillegible duetopoorreproduction quality.Alegiblereportneedstobesubmitted forreview.STEAMGENERATOR TUBESUPPORTSSection16ofReference 5documents thestressanalysisofSeries44steamgenerator internals forseismicloading.Aconservative analysisofthetubesupportsresultedinaprimarymembranestressof.7900psiforanequivalent staticcoefficient of0;19ghorizontal.
Thestresswascalculated intheligaments betweenthetubeholesandcircu-latingholesinalocalareaadjacenttoawrapperchannel.Theanalysis'conservatively ignoredaredundant loadpathfromthe.edgeofthetube"supporttothetubeholes.Reference 6,submitted forreview,isanupdateanalysisoftheSeries44steamgenerator butdoesnotaddressthe~1 tubesupportforhorizontal loading.NeitherReference 5or6describethedynamiccharacteristics ofthesteamgenerator anditsinternals noridentifythematerialofconstruction forthetubesupportplates.Reference 7,obtainedduringtheSSMRPprogram,documents agenericdynamicanalysisoftheSeries51steamgenerator forvaryingsupportstiffnesses andlocations.
TheSeries44steamgenerator issimilar,butsmaller.Thefundamental frequency oftheSeries51steamgenerator rangesfrom4.8to9.6Hzdepending uponthesupportstiffness andlocation.
Thefundamental modeispredominantly translation androckingofthesteamgenerator shell.Agenericresponsespectrumthatisflatthroughmostofthisfrequency rangewasusedtocomputeresponseaccelerations andloadfngsinthesteamgenerator.
Areviewwasconducted ofReference 7todetermine thedegreeofcouplingbetweenthesteamgenerator shellandtheinternalstructures andtoestablish thevalidityofconsidering thesteamgenerator asaSDOFsystemforestimating anappropriate equivalent staticcoefficient forevaluation ofthetubesupportplates..Itwasconcluded fromreviewofReference 7thatatthemostcriticalsupportlocation, asdetermined inReference 5,thatthetubesandtubesupportswouldaccelerate asrigidbodieswiththeshell;thus,usingthespectralacceleration fromtheGinnaspectrumforafunda-mentalfrequency ofabout.5Hzisareasonable approximation ofanequiva-lentstaticcoefficient, touseforevaluation ofthetubesupports.
FromtheresponsespectrainReference 8,for7Xdamping,themaximumspectralaccelerations at5Hzare0.58g'sineachoftwoortho-gonaldirections.
Combining thetwodirectional accelerations, theresulting maximumvectoris0.82g.UsingthisvalueandscalingtheligamentstresscomputedinReference 5,theresulting ligamentstressis34,095psi.Reference 7indicates thatSeries51steamgenerator internals areconstructed ofSA285-Grade Ccarbonsteel.Acomparisonoftheallowable stressforthismaterialatthedesign.temperature of556Ftotheallowable stressquotedfortheSeries44steamgenerator tube supportsinReference 5,indicates thattheSeries44tubesupportsarealsoconstructed ofSA285-Grade Corequivalent.
Thismaterialandadesigntemperature of'556Fareusedasabasisforestablishing allowable stressesfortheSafeShutdownEarthquake.
Thetubesupportsareconsidered tobeClass1plateandshelltypecomponent supportsandtheallowable primarymembranestressiscomputedforLevel0ServicefromAppendixFoftheASMECode,Reference 9.Theallowable stressisthegreaterof1.5Smand1.2Sy.ForSA285-Grade Cmaterial, 1.2Sygovernsandtheallowable stressis27840psi.NotethattheoriginaldesigncriterialimitedthetubesupportstresstoS.Comparison ofthecalculated andallowable stressforLevelDService-results ina22%overstress condition.
Ifsitespecificspectraanchoredto0.172gareconsidered inlieuofregulatory guidespectraanchoredto0.2g,thecalculated.
stressdecreases.
Decreasing thecalcu-latedstressbytheratiosofthesitespecificpeakgroundacceleration dividedbythe0.2gpeakgroundacceleration usedtogeneratefloorspectra,theresulting ligamentstressis29,320psi.Thisisstill.about.5'.3XovertheLevelDServiceallowable of27,840psi.Inconsideration oftheconservatism inherentinobtaining thecalculated stress,thecomputed5.3Xoverstress condition isconsidered acceptable forseveralreasons.1.Thesitespecificspectrumenvelopehaslowerspectralaccelerations inthefrequency rangeofthecontainment structure thantheregulatory guidespectrumifbothareanchoredtothesamepeakgroundacceleration.
Conse-quently,thein-structure responsespectrawillbelowerthanthoseinReference 8.'.ThestaticanalysisfromReference 5didnotaccountfor'heredundant loadpathbetweentheoutsidediameterofthetubesupportplateandtheouterrowoftubeholes.'Thedegreeofconservatism couldnotbeevaluated sinceperti-nentdimensions arenotprovidedinReference 5.The'egreeofconservatism iscertainly greaterthan5Xthough.
3.Theevaluation considered theapplicable acceleration tobethevector.sumofthetwoorthogonal directional acceler-ations.Thisassumesthatbothdirectional responses areinphaseandthattheresulting vectorisalignedintheworstdirection.
4-Thein-structure responsespectrawerepeakbroadened
+15%andsmoothedsothattheresulting spectraareessentially flatfrom2-1/2to9Hz,whichcoverstherangeoffunda-mentalfrequency forthesteamgenerators.
Items3and4areconsistent withcurrentregulatory criteriaandareprudentconservatisms tocovermanyoftheuncertainties inthesimplified treatment ofthetubesupport.Recommended acceptance is,therefore, basedontheconservatism ofItems1and2beingsufficient toovercometheestimated 5.3foverstress condition.
Furtheranalysesorsubmittals fromthelicenseearenotconsidered necessary.
REFERENCES 1.SMAletter,R.D.CampbelltoT.R.Nelson,ReviewofOpenItemsResulting fromSeismicReviewoftheRobertE.GinnaNuclearPowerPlantasPartoftheSEPProgram,4May1981.2.HighSpeedControlRodDriveStressAnalysisJune26,1968.3.ControlRodDriveMechanism SeismicFrameCalculations, August13,1968.4.StaticSeismicLoadStressAnalysis, ModelRGEPump(V-11001-BI),
July30,1968.5.VerticalSteamGenerator StressReport,Mestinghouse ElectricCorporation, TampaDivision, April,1969.6.MTO-SM-75-028, 44SeriesSteamGenerator StressReport,ExternalLoadAnalysisUpdate,May,1975.J7.StressReport;51SeriesSteamGenerator, GenericSeismicAnalysis, westinghouse ElectricCorporation, TampaDivision,
- December, 1974..8.NUREG/CR-1821, SeismicReviewoftheRobertE.GinnaNuclearPowerPlantasPartoftheSystematic Evaluation Program,15November1980.9.ASMEBoilerandPressureVesselCode,SectionIII,NuclearPowerPlantComponents, Appendices, 1980.
STRUCTURAL ICCHAAICS ASSOCIATES AC~Ilt.Coro.5160BirchStreet,NewportBeach,Calif:82660(714)833-7552SMA12205.20May4,1981Mr.ThomasA.Nelson(L-90)LawrenceLivermore Laboratory NuclearTestEngineering DivisionP.O.Box808Livermore,California 94550
DearTom:
SMAhasreviewedthepackageofdocuments transmitted withyourApril15letteraddressing openitemsonGinna.Ourcommentsandrecommended actionarecontained intheenclosure.
Verytruly'yours, STRUCTURAL MECHANICS ASSOCIATES, INC.piRobertD.CampbellProjectManagerRDC.mwEnclosurecc:J.Stevenson w/encl.
REVIEWOFOPENITEMSRESULTING FROMSEISMICREYIEMOFTHEROSERTE.GINNANUCLEARPOWERPLANTASPARTOFTHESEPPROGRAMReference 1documents areview.conducted bytheLawrenceLivermore Laboratory andtheirconsultants ontheseismicadequacyoftheRobertE.GinnaNuclearPowerPlant.Conclusions oftheadequacyoftheseveralitemsintheNSSSsystemwerebaseduponsummaryinformation provided; however,thesourcesofthesummaryinforma-tionwerenotavailable forindependent review'.ThoseitemslistedinSection5.4ofthereportascomponents forwhichseismicdesignanalyses.
havenotbeenindependently verifiedare:ReactorControlRodDriveReactorVessel'upports SteamGenerator ReactorCoolantPumps-Pressurizer anditsSupportsReferences 2through7wereprovidedtoSMAinresponsetotheaboveidentified openitems.-Thefollowing summaryandconclusions resultedfromSMA'sreview'of thesubmittals-.
1)ControlRodD~iveMechanism Inreference 7,theallowable bending'oment intheCRDMduetoseismicloadsisdeveloped.
Thisreportdoesnot,however,provideacorrelation betweenbendingmomentsandacceleration levels,Thus,noconclusion canbereachedonthebasisofthesubmittal.
Reference 2containsastressanalysisofthecontrolroddrivesupportstructure.
Theanalysisprovidesstressesas'functionofastaticload"P"inpounds.Thereisnocorre-lationbetweenthisstaticloadandacceleration level.Therefore, aconclusion ontheseismiccapability ofthesupportstructure cannotbereachedbaseduponthesubmittal.
ls5~~SMA'sexperience withtheSSMRPreference plant,whichuses106Afull lengthcontrolroddrive.mechanisms, indicated thatalargemarginofsafetyexistsfora1.15gspectralaccel-erationloadingcondition andwewouldnotanticipate aseismicproblemwiththeGinnaCROM.Recomme'rided action-CROMloadsdocumentation applicable toGinnaneedtobesuppliedtoSMAforfinalresolution.
2JReactor'Vessel SuortsDocume'ntation verifying theseismicadequacyofthereactorlvessel;supportswasnotsubmitted.
BaseduponSMA'sSSMRPexperience fornozzlesupported RPY'stheseismicinducedstressesinthenozzlesandadjacentshellareverysmallandthegoverningelementforRPVsupportistheconcrete,
.shieldwall.Theshieldwallwasconsidered inRef.1tobeadequatetowithstand the0.2gSSE.3lSteamGenerator Reference 6containsa1969staticanalysisoftheSeries44steamgeneator.
Themostcriticalareaduetoseismicload-ingidentified inthisreportisthetubesupportbaffleligaments whicharestressedpastyieldfora0.38ghorizon-talstaticload.TheSEPrevisedspectra'resultinanSRSSresponseof0.85ghorizontal whichwillincreasethestressesbyafactorof2.24.Thus,basedonthesubmittal, thetubesupportbafflesareoverstressed forthe0.2gSSE..Thestaticanalysisatthesetubesupportbaffleswas.donequiteconser-vatively, however,andamorerigorous'~nalysis willmostlikelyresultinlowerstateofstress.In1975,anupdateontheseries44steamgenerator wascon-ducted(Reference 4)butthetubesupportareawasonlyevaluated forverticalseismicloadingandnotforthe PI horizontal seismicloading.Thus,theresultsofReference 4cannotbeusedtoupdatetheresultsofReference 6intheareaofconcern.Recommended Action-Documentation evaluating thetubesupportbaffleligaments forhorizontal seismicloadingshouldbesubmitted forreview,4)ReactorCoolantPumsReference 3summarizes thestressesinducedinthereactorcoolantpumpbya0.8g'orizontal anda0.54gverticalload-ing.ThereportedstressesarebelowtheASMECodeallowables, butSMAisunableto'evaluate themodelortheana1ysisduetopoorreproduction qualityofReference 3.Ifthestaticanalysisofthepumpcanbeshowntobevalid,thenthestressesduetotherevisedGinnaspectraloadingwillbelessthanthosecontained withinReference 3,"andthusaccept-able.Recommended Action--AlegiblecopyofReference.3 shouldbeprovidedtoSMAforreview.5).Pressurizer Reference 5containsa1969stressreportofan1800cubicfootpressurizer.
Allpressurizers from800to1800cubic'eetwithcastand'fabricated headsutilizethesamesupportskirts,thusconservative genericanalysiswasconducted for.theheavier1800cubicfootmodels.Basedonthisreport,theloadsresulting fromthenewGinnaspectrawill.causeanover-stressedcondition withinthesupportf1ange,Thisisaveryconservative conclusion, however,sincetheGinnapressurizer isasmaller800cubicfootmodelplus,theflangeanalysisitselfwasveryconservatively conducted usingabeambendingmodelinlieuofamorerigorousfiniteelementmodel.
The1973pressurizer reportreferredtoinReference 1andthepr'essurizer summarystressreport(reference 45ofRefer-ence1)werenotsuppliedtoSMAforreview;thus,theade-quatecapacityfora6.7gequivalent staticloadportrayed inreference 45ofReference 1cannotbeverified.
Lateranalysis(References 8and9)obtainedintheSSHRPprogram,'for theSeries511800cubicfootcastandfabricated headpressurizers showedthesupportstobeacceptable fora0.96g.horizontal and0.64gverticalloadingcondition.
Thislateranalysisutilizedafiniteelementmodeloftheskirtandflangeasopposedtotheconservative beamtheory.usedinreference 5.Recommended Action-Noactionrequired.
Thepressurizer stressanalysiswithinreferences 8and9andthestatements inReferences 5,8and9thatthesupportskirtsareidentical forthe1800cubicfootandthe800cubicfootdesignssub-=stantiate theadequacyoftheGinnapressurizer fora0.2g.SSE.
