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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217J3341999-10-19019 October 1999 Forwards Request for Addl Info Re Sale of Portion of Land Part of Oyster Creek Nuclear Generating Station Site Including Portion of Exclusion Area ML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20212J6721999-09-30030 September 1999 Informs of Completion of mid-cycle PPR of Oyster Creek Nuclear Generating Station on 990913.No Areas Identified in Which Licensee Performance Warranted Addl Insp Beyond Core Insp Program.Historical Listing of Plant Issues Encl ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20217B2531999-09-24024 September 1999 Informs That on 980903,Region I Field Ofc of NRC Ofc of Investigations Initiated Investigation to Determine Whether Crane Operator Qualification/Training Records Had Been Falsified at Oyster Creek Nuclear Generating Station ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212A7921999-09-13013 September 1999 Forwards Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, Issued on 950817 to Plant ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J9831999-09-0202 September 1999 Discusses 990804 Telcon Re Sale of Portion of Oyster Creek Nuclear Generating Station Land.Requests Info Re Location of All Areas within Property to Be Released Where Licensed Radioactive Matl Present & Disposition of Radioactive Matl ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211C0161999-08-19019 August 1999 Advises That Info Submitted by Ltr,Dtd 990618, Licensing Rept for Storage Capacity Expansion of Oyster Creek Spent Fuel Pool, Holtec Rept HI-981983,rev 4,will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210U4341999-08-17017 August 1999 Responds to to Chairman Dicus of NRC on Behalf of Fm Massari Concern About Oyster Creek Nuclear Generating Station Not Yet Being Fully Y2K Compliant ML20210Q7331999-08-12012 August 1999 Responds to Re TS Change Request (TSCR)264 from Oyster Creek Nuclear Generating Station.Questions Re Proposed Sale of Property within Site Boundary & Exclusion Area ML20210L6311999-08-0606 August 1999 Discusses Licensee Response to GL 92-01,Rev1,Suppl 1, Rv Structural Integrity, for Plant.Staff Has Revised Info in Rv Integrity Database & Releasing as Rvid Version 2 ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195G6541999-06-0707 June 1999 Discusses 981204 Initiation to Investigate Whether Contract Valve Technician,Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20195G6631999-06-0707 June 1999 Discusses 981204 Intiation to Investigate Whether Contract Valve Technician Was Discriminated Against for Raising Concern Re Use of Untrained/Unqualified Workers Performing Valve Repairs.Technician Was Not Discriminated Against ML20209B0561999-06-0404 June 1999 Informs That NRR Has Reorganized,Effective 990328.Forwards Organizational Chart ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P5381999-04-14014 April 1999 Ack Receipt of Re Request for Exception to App J. Intended Correction Would Need to Be Submitted as Change to TS as Exceptions to RG 1.163 Must Be Listed in Ts,Per 10CFR50,App J ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record ML20205P0651999-04-0909 April 1999 Discusses 990225 PPR & Forwards Plant Issues Matrix & Insp Plan.Results of PPR Used by NRC Mgt to Facilitate Planning & Allocation of Insp Resources 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205J3281999-04-0101 April 1999 Discusses Arrangements Made on 990323 for NRC to Inspect Licensed Operator Requalification Program at Oyster Creek Nuclear Generating Station During Week of 990524 ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207F0331999-03-0404 March 1999 Forwards Insp Rept 50-219/98-12 During Periods 981214-18, 990106-07 & 20-22.Areas Examined During Insp Included Implementation of GL 89-10 & GL 96-05.No Violations Noted ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217E0181999-10-0606 October 1999 Provides Nj Dept of Environ Protection Comments on Oyster Creek Nuclear Generating Station TS Change Request 267 Re Clarifications to Several TS Sections 05000219/LER-1998-011, Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts1999-09-30030 September 1999 Forwards LER 98-011-02, Three Small Bore Pipe Lines Did Not Meet Design Bases for Siesmic & Thermal Allowables. Engineering Std Will Not Be Completed Until End of 4th Quarter of 1999 Due to Scheduling Conflicts ML20216J7591999-09-30030 September 1999 Informs NRC That Remediation Efforts for Software Sys Etude & Rem/Aacs/Cico Have Been Completed According to Schedule & Now Y2K Ready ML20216K1421999-09-29029 September 1999 Provides NRC with Name of Single Point of Contact for Purpose of Accessing Y2K Early Warning Sys,As Requested by NRC Info Notice 99-025 ML20217D1661999-09-27027 September 1999 Forwards Proprietary Completed NRC Forms 396 & 398,in Support of License Renewal Applications for Listed Individuals,Per 10CFR55.57.Encl Withheld ML20212E1971999-09-16016 September 1999 Forwards Rev 11 of Gpu Nuclear Operational QAP, Reflecting Organizational Change in Which Functions & Responsibilities of Nuclear Safety & Technical Support Div Were Assigned to Other Divisions ML20212B5571999-09-10010 September 1999 Forwards Rev 11 to Oyster Creek Emergency Dose Calculation Manual, IAW 10CFR50,App E,Section V ML20211N2941999-09-0303 September 1999 Responds to NRC 990802 Telcon Request for Environ Impact Assessment of TS Change Request 251 Concerning Movement of Loads Up to 45 Tons with RB Crane During Power Operations ML20211J6771999-08-30030 August 1999 Submits Response to NRC 990802 Telcon Request for Gpu to Provide Environ Impact Assessment for Tscr 251 ML20211K2391999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Nj ML20211B9011999-08-18018 August 1999 Forwards Rev 0 to EPIP 1820-IMP-1720.01, Emergency Public Info Implementing Procedure ML20210D2801999-07-22022 July 1999 Submits Response to Administrative Ltr 99-02 Operating Reactor Licensing Action Estimates. Estimate of Licensing Actions Projected for Fy 2000 Encl.No Projection Provided for Fy 2001 ML20209H5001999-07-14014 July 1999 Forwards Revised TS Pages 3.1-15 & 3.1-17 Which Include Ref to Note (Aa) & Approved Wording of Note H of Table 3.1.1, Respectively ML20210U4411999-07-12012 July 1999 Forwards Article from Asbury Park Press of 990708 Faxed to Legislative Officer by Mutual Constitute Fm Massari Indicating That Oyster Creek Nuclear Generating State Not Fully Y2K Compliant ML20209G1451999-07-0909 July 1999 Forwards Rev 1 to 2000-PLN-1300.01, Oyster Creek Generating Station Emergency Plan. Attachment 1 Contains Brief Summary of Changes,Which Became Effective on 990702 ML20209E0821999-07-0707 July 1999 Forwards TS Change Request 269 for License DPR-16,changing Component Surveillance Frequencies to Indicate Frequency of Once Per Three Months ML20209B7501999-07-0101 July 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Generic Ltr 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Licensee Y2K Readiness Disclosure for Ocngs,Encl ML20196G1361999-06-23023 June 1999 Provides Status of Corrective Actions Proposed in in Response to Insp Rept 50-219/98-80 & Revised Schedule for Completion of Actions Which Are Not Yet Complete ML20196E6421999-06-22022 June 1999 Forwards Revised Pages of TS Change Request 261,dtd 990618. Replacement Requested Due to Several Dates Being Omitted on Certain Pages ML20195D0551999-06-0303 June 1999 Forwards TS Change Request 226 to License DPR-16,permitting Operation with Three Recirculation Loops.Certificate of Svc & Tss,Encl ML20195C5511999-05-25025 May 1999 Forwards Book of Controlled Drawings Currently Ref But Not Contained in Plant Ufsar.Drawings Were Current at Time of Submittal ML20206N7711999-05-11011 May 1999 Forwards Rev 0 to Oyster Creek Emergency Plan, IAW 10CFR50.47(b) & 10CFR50.54(q).Changes Became Effective on 990413 ML20206H9441999-04-28028 April 1999 Forwards Application for Amend to License DPR-16,requesting Approval to Handle Loads Up to & Including 45 Tons Using Reactor Bldg Crane During Power Operations,Per NRC Bulletin 96-002 ML20206B6991999-04-26026 April 1999 Forwards Copy of Rev 11 to UFSAR & Rev 10 to Oyster Creek Fire Hazards Analysis Rept. Without Fire Hazard Analysis ML20206D3801999-04-26026 April 1999 Forwards Rev 11 to UFSAR, & Rev 10 to Fire Hazards Analysis Rept, for Oyster Creek Nuclear Generating Station, Per 10CFR50.712(e) ML20206C8261999-04-22022 April 1999 Submits Financial Info IAW Requirements of 10CFR50.71(b) & 10CFR140.21 ML20206A9931999-04-22022 April 1999 Forwards Number of Personnel & Person Rems by Work & Job Function Rept for Period Jan-Dec 1998. Included in Rept Is Listing of Number of Station,Util & Contractor Personnel as Well as Diskette Reporting 1998 Occupational Radiation ML20205P8411999-04-15015 April 1999 Forwards TS Change Request 267 to License DPR-16,modifying Items in Sections 2 & 3 of Ts,Expanding Two Definitions in Section 1 & Modifying Bases Statements in Sections 2,3 & 4. Certificate of Svc Encl ML20205P9401999-04-12012 April 1999 Informs NRC That Gpu Nuclear Is Modifying Oyster Creek FSAR to Reflect Temp Gradient of 60 F & to Correct Historical Record 05000219/LER-1998-015, Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page1999-04-0505 April 1999 Forwards LER 98-015-01,as Original Submittal on 981028 Inadvertently Indicated That Suppl Would Be Submitted.Suppl Should Not Have Been Required as Only Change Is on Cover Page ML20205H1081999-03-31031 March 1999 Forwards Current Funding Status for Decommissioning Funds Established for OCNPP,TMI-1,TMI-2 & SNEC ML20205F0611999-03-25025 March 1999 Submits Info on Sources & Levels of Property Insurance Coverage Maintained & Currently in Effect for Oyster Creek Nuclear Generating Station,Iaw 10CFR50.54(w)(3) ML20205E1171999-03-24024 March 1999 Forwards Rev 39 to Oyster Creek Security Plan & Summary of Changes,Iaw 10CFR50.54(p).Rev Withheld ML20207K2471999-02-25025 February 1999 Forwards Fitness for Duty Performance Data Repts for TMI, Oyster Creek & Corporate Headquarters Located in Parsippany, Ny ML20206S2541999-01-20020 January 1999 Confirms Resolution of Thermo-Lag Fire Barriers in Fire Zones OB-FZ-6A & OB-FZ-6B (480 Switchgear Rooms) IAW Previous Commitments Contained in Gpuns Ltrs to NRC & 971001 ML20199J2631999-01-18018 January 1999 Requests That Listed Changes Be Made to Correspondence Distribution List for Oyster Creek Generating Station ML20199D0271999-01-11011 January 1999 Requests Listed Addl Info in Order to Effectively Review TS Change Request 264 Re Ownership of Property within Exclusion Area ML20199A6521999-01-0707 January 1999 Notifies That Reactor Operators G Scienski,License SOP-11319 & D Mcmillan,License SOP-3919-4 Have Terminated Licenses at Oyster Creek Nuclear Generating Station, Effective 990101 ML20198T1061999-01-0606 January 1999 Forwards Rev 15 to Gpu Nuclear Corporate Emergency Plan for TMI & Oyster Creek Nuclear Station. with Summary of Changes Which Reflect Use of EALs Approved in NRC Ltr to Gpun on 980908 & Other Changes Not Related to Use of New EALs ML20198K0331998-12-23023 December 1998 Forwards Change Request 268 for Amend to License DPR-16. Amend Would Change TS to Specify Surveillance Frequency of Once Per Three Months ML20198H0181998-12-22022 December 1998 Forwards Attachment Addressing New Info & Modifying 980505 Submittal Re Request for Change to Licensing Bases for ECCS Overpressure,In Response to NRC Bulletin 96-03, Potential Plugging of ECCS by Debris in Bwrs ML20198H8521998-12-16016 December 1998 Dockets Completion of Physical Inventory Performed in July 1997,as Addl Info to Nuclear Matl Balance Rept Submitted on 980416 ML20196H4461998-12-0202 December 1998 Provides Final Response to NRC GL 96-01, Testing of Safety-Related Logic Circuits ML20196B4471998-11-23023 November 1998 Provides Required Response 2 to NRC Bulletin 96-003, Potential Plugging of ECC Suction Strainers in Bwrs. During Recently Completed 17R Refueling Outage,New Strainers Were Installed ML20195J8451998-11-12012 November 1998 Forwards Rev 11 to 1000-PLN-7200.01, Gpu Nuclear Operational QA Plan, as Change Previously Made Without Appropriate Notification to NRC ML20195C7201998-11-11011 November 1998 Forwards 120-day Required Response to GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Construction & Protective Coating Deficiencies & Foreign Matl in Containment, ML20195E1221998-11-10010 November 1998 Notifies NRC of First Time Usage of Code Case N-504 & Inclusion Into OCNGS ISI Program,As Accepted by RG 1.147, Inservice Insp Code Case Acceptability ML20155J6851998-11-0505 November 1998 Forwards TS Change Request 266,to Modify Safety Limits & Surveillances of LPRM & APRM Sys & Related Bases to Ensure APRM Channels Respond within Necessary Range & Accuracy & Verify Channel Operability ML20155H5641998-11-0202 November 1998 Informs That Bne Has No Comments on Proposed Change 259 to Ts,Correcting Required Water Level in Condensate Storage Tank So That Design Basis Is Correctly Implemented ML20155G3741998-10-29029 October 1998 Forwards Response to NRC 980619 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L2551990-09-14014 September 1990 Advises of Preparation for Final Refueling Outage to Complete Second 10-yr Interval for Inservice Insps ML20059F5121990-09-0505 September 1990 Requests Exemption from Filing Requirement of 10CFR55.45(b)(2)(iii) to Allow Submittal of NRC Form 474, Simulator Facility Certification, After 910326 Deadline & to Allow Administering of Simulator Portion of Test ML20059F7441990-08-31031 August 1990 Forwards Util Review of NRC Backfit Analysis for Hardened Wetwell Vent.Nrc Analysis Does Not Support Conclusion That Hardening Existing Vent Is cost-beneficial Mod for Plant ML20059E9061990-08-30030 August 1990 Forwards Response to 900808 Request for Addl Info Re NRC Bulletin 90-002, Loss of Thermal Margin Caused by Fuel Channel Box Bow ML20059G1841990-08-29029 August 1990 Ack NRC Request to Perform Type C Testing During Unscheduled Outage,As Plant Conditions Will Allow.Type C Exemptions Should Remain in Effect Until New Outage Start Date ML20059C8231990-08-27027 August 1990 Advises That SPDS Enhancements Described in Completed,Per 900628 Request.Offline & Online Testing Completed & Enhancements Considered to Be Operational ML20059C8571990-08-24024 August 1990 Provides Results of Evaluation of Ability to Meet Acceptance Criteria for Eccs,In Response to 900804 Notice of Violation. Plant Meets Acceptance Criteria Contained in 10CFR50.46 W/ Valve Logic Design Deficiency in Containment Spray Sys ML20058N0781990-08-0909 August 1990 Submits Info Re pressure-temp Operating Limits for Facility, Per Generic Ltr 88-11.Util Recalculated Adjusted Ref Temp for Each Belt Line Matl as Result of New Displacement Per Atom Values ML20063P9521990-08-0909 August 1990 Advises That Response to NRC 900523 Request for Assessment of Hazardous Matl Shipment Will Be Sent by 910531 ML20058L9521990-08-0303 August 1990 Forwards Rev 2 to Security Contingency Plan.Rev Withheld (Ref 10CFR73.21) ML20058L9551990-08-0303 August 1990 Responds to SALP Rept 50-219/88-99.Although Minor,Several Factual Errors Noted.Dialogue Promotes & Identifies Areas Where Improvements Should Be Made ML20056A2071990-07-30030 July 1990 Forwards Response to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Record Review Performed & Sys Walkdowns Completed to Assemble Requisite List ML20055H7961990-07-20020 July 1990 Advises of Change to Preventive Maint Program for Electromatic Relief Valves.Rebuild Schedule Will Be Modified to Require Rebuilding Two or Three Valves During Refueling Outage & Remaining Valves During Next Refueling Outage ML20058N9911990-07-20020 July 1990 Partially Withheld Response to NRC Bulletin 90-002 Re Loss of Thermal Margin Caused by Box Bow (Ref 10CFR2.790(b)(1)) ML20055J0481990-07-19019 July 1990 Requests 2-wk Extension for Submittal of Response to Re Installation of Hardened Wetwell Vent W/ Appropriate Extension Period to Be Decided Pending Outcome of 900724 Meeting Discussion W/Bwr Owners Group ML20064A1221990-07-11011 July 1990 Discusses 900710 Telcon W/Nrc Re Util Corrective Actions in Response to NRC Finding That Operator Received Passing Grade on Administered Requalification Exam in 1989 Should Have Received Failing Grade.Corrective Actions Listed ML20055F8491990-07-10010 July 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 188,reducing Low Condenser Vacuum Scram Setpoint ML20043H7471990-06-21021 June 1990 Confirms Telcon W/A Dromerick Re Util Plans to Inspect CRD Hydraulic Control Units During Plant Walkdown to Address USI A-46, Seismic Qualification of Equipment in Operating Nuclear Power Plants. Walkdown Planned in Oct 1992 ML20043H2301990-06-14014 June 1990 Documents Licensee Commitment to Improve Seismic Restraints for Diesel Generator Switchgear Encls,Per 900613 Telcon W/ Nrc.Engineering Will Be Finalized & Mods Completed Prior to 900622 ML20043F7581990-06-0707 June 1990 Responds to Request for Info Re Util Compliance W/Generic Ltr 88-01 & Insp Plans for Upcoming 13R Outage.Frequency of Insp of Welds Classified as IGSCC Categories C,D & E Will Not Be Reduced During 13R Outage ML20043C5801990-05-25025 May 1990 Provides Descriptions & Conclusions of Three Remaining Issues of SEP Topic III-7B.Issues Include,Evaluation of Drywell Concrete Subj to High Temp & Thermal Transients ML20043C2461990-05-25025 May 1990 Forwards Rev 7 to EPIP 9473-IMP-1300.06 & Rev 4 to Radiological Controls Policy & Procedure Manual 9300-ADM-4010.03, Emergency Dose Calculation Manual. ML20043B2981990-05-21021 May 1990 Responds to NRC 900420 Ltr Re Violations Noted in Insp Rept 50-219/90-06.Corrective Actions:Incident Critique Rept Incorporated as Required Reading for Appropriate Operations Personnel & Change Made to Procedure 201.1 ML20043D0701990-05-17017 May 1990 Provides NRC W/Addl Info Re SPDS & Responds to Concerns Raised During 900117 & 18 SPDS Audit Documented in 900130 Ltr ML20043B3901990-05-0909 May 1990 Responds to NRC 900408 Ltr Re Violations Noted in Insp of License DPR-16.Corrective Actions:Two Narrow Range Drywell Pressure Monitoring Instruments to Be Provided During Cycle 14R Refueling Outage,Per Reg Guide 1.97,Category 1 ML20042G7071990-05-0808 May 1990 Forwards Summary of Initiatives & Accomplishments Re SALP, Per 891031 Commitment at mid-SALP Meeting.Plant Div Responsibilities Now Include Conduct of Maint Outages & Emergency Operating Procedure Training Conducted ML20042G2291990-05-0707 May 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 180,revising Tech Specs Re Fuel cycle-specific Parameters ML20042G2601990-05-0404 May 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 187,revising Tech Specs to Accommodate Implementation of 24-month Plant Refueling Cycle ML20042E9431990-04-20020 April 1990 Forwards Revised Epips,Consisting of Rev 7 to 9473-IMP-1300.01,Rev 4 to 9473-1300.11 & Rev 2 to 9473-ADM-1319.04.Deleted EPIPs Listed,Including Rev 3 to 9473-1300.19,Rev 2 to 9473-1300.21 & Rev 5 to 9473.1300.24 ML20042E6371990-04-16016 April 1990 Informs of Plans to Install Safety Grade Check Valve in Supply Line Inside Emergency Diesel Generator Fuel Tank Room Coincident W/Replacement or Repair of Emergency Diesel Generator Fuel Oil Tank ML20042E5001990-04-13013 April 1990 Forwards Rev 1 to Topical Rept 028, Oyster Creek Response to NRC Reg Guide 1.97. ML20012E8711990-03-28028 March 1990 Lists Property Insurance Coverage,Effective 900401,per 10CFR50.54(w)(2) ML20012D4391990-03-19019 March 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 186,allowing Idle Recirculation Loop to Be Isolated During Power Operation by Closing Suction,Discharge & Bypass Valves ML20012B6781990-03-0202 March 1990 Requests Exemption of Specified Local Leak Rate Test Intervals to Include Next Plant Refueling Outage Scheduled for Jan 1991,per 10CFR50,App J ML20012A1501990-02-23023 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 184,removing 3.25 Limit on Extending Surveillance Intervals,Per Generic Ltr 89-14 ML20011F2571990-02-21021 February 1990 Advises That 891003 Request for Appropriate Tech Specs for Chlorine Detection Re Control Room Habitability,Not Warranted ML20006G0101990-02-21021 February 1990 Discusses 900110 Meeting W/Nrr Re 13R Insp Criteria for RWCU Welds Outboard of Second Containment Isolation Valve. All Welds Required 100% Radiography Based on Review of Piping Spec.Response to Generic Ltr 88-01 Will Be Revised ML20006F5931990-02-20020 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 177.Amend Changes Tech Spec 4.7.B to Include Battery Svc Test Every Refueling Outage & Mod of Frequency of Existing Battery Performance Test ML20011F6641990-02-20020 February 1990 Responds to NRC 900122 Notice of Violation & Forwards Payment of Civil Penalty in Amount of $25,000.Corrective Actions:Change Made to Sys Component Lineup Sheets in 125- Volt Dc Operating Procedure to Include Selector Switches ML20006F9181990-02-15015 February 1990 Forwards Application for Amend to License DPR-16,consisting of Tech Spec Change Request 183,permitting No Limitation on Number of Inoperable Position Indicators for 16 ASME Code Safety Valves During Power Operation ML20006D2551990-01-30030 January 1990 Forwards Response to Generic Ltr 89-13 Re Plant Svc Water Sys.Insp Program for Intake Structure at Plant Implemented During Past Two Refueling Outages & Emergency Svc Water Currently Chlorinated to Prevent Biofouling ML19354E8571990-01-24024 January 1990 Forwards Omitted Pages of 900116 Ltr Re State of Nj DEP Comments on Draft full-term OL SER & Clarification of Page 10,fourth Paragraph on New Seismic Floor Response Spectra ML20006B2171990-01-23023 January 1990 Responds to Unresolved Items & Weaknesses Identified in Insp Rept 50-219/89-80.Corrective Actions:Procedure Re Containment Spray sys-diagnostic & Restoration Actions Revised to Stand Alone Re Installation of Jumpers ML19354E3891990-01-19019 January 1990 Responds to Violations Noted in Insp Rept 50-219/89-27. Corrective Actions:Procedure 108 Revised to Allow Temporarily Lifting of Temporary Variation ML19354E8441990-01-19019 January 1990 Forwards Revised Tech Spec Table 4.13-1, Accident Monitoring Instrumentation Surveillance Requirements, in Support of Licensee 890630 Tech Spec Change Request 179,per NRC Project Manager Request ML20006B2881990-01-18018 January 1990 Forwards Results from Feedwater Nozzle Exam,In Accordance w/NUREG-0619 Insp Format ML20005G8161990-01-16016 January 1990 Provides Assessment of State of Nj Concerns Re full-term OL Plant,Per NRC 891222 Request.Comments Did Not Raise Any Concerns That Refute Conclusions Reached by NRC That Facility Will Continue to Operate W/O Endangering Safety ML19354D8281990-01-15015 January 1990 Responds to Violation Noted in Insp Rept 50-219/89-21. Corrective Action:Procedure A000-WMS-1220.08, Mcf Job Order Revised to Provide Detailed Guidance for Performance of Immediate Maint ML20042D4891989-12-28028 December 1989 Forwards Response to Generic Ltr 89-10, Safety-Related Motor-Operated Valve Testing & Surveillance, Fulfilling 6-month Reporting Requirement ML20005E1401989-12-22022 December 1989 Forwards Integrated Schedule Semiannual Update for Dec 1989 1990-09-05
[Table view] |
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.
- GPU Nuclear Corporation Nuclear :: = = = 388 Forked River, New Jersey 08731-0388 ,
609 971-4000 Writer's Direct Dial Number:
June 13, 1984 Dennis M. Crutchfield, Chief Operating Reactors Branch #5 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555
Dear Mr. Crutchfield:
Subject:
Oyster Creek Nuclear Generating Station Docket No. 50-219 Spent Fuel Pool Expansion - Additional Information Cnclosed are responses to questions forwarded to me by your letter of June 1,1984 concerning GPU Nuclear's request to expand the capacity of the spent fuel pool.
Very truly yours, .
G Pete edler Vice President & Director '
Oyster Cfeek l PBF:SD: dam l Enclosure l cc: Dr. Thomas E. Murley, Administrator Region I U.S. Nuclear Regulatory Commission 631- Park Avenue King of Prussia, PA 19406 l
l NRC Resident Inspector l Oyster Creek Nuclear Generating Station Forked River, NJ 08731 -
I 8406190197 840613 PDR ADOCK 05000219 .
P PDR ;
j GPU Nuclear Corporation is a subsidiary of the General Public Utilities Corporation I
ENCLOSURE OYSTER CREEK NUCLEAR GENERATING STATION Additional information regarding the following is requested.
A. High-Density Spent Fuel Rack Dynamic Structural Analysis based on the review of the recently submitted responses by Joseph Oat Corporation (consultant to Oyster Creek) [1, 2] and the information pre-sented in a meeting with Joseph Oat at Franklin Research Center on May 7, 1984 [3], the following questions are prepared for additional information and/or clarification.
- 1. In the referenced meeting [3] and responses submitted on May 10 [2],
Joseph Oat indicated that the equivalent gap was developed to take into account the hydrodynamic effects on all four sides of the rack. The following reasons are given for this concept of equiva-lent gap:
- a. -The seismic loading should be equal to zero when taking the average of the complete seismic input time history.
- b. The hydrodynamic masa, according to Fritz [4], is related to the gap (g) by the following expression:
Hydrodynamic mass = MH = Constant g
Please respond.to the following questions:
a.- Provide the technical basis as to why the seismic loading will be zero; even if this is true, it is not clear how this would be used to justify the proposed approach.
- b. ' Provide a discussion of why, at the instant the motion starts, the rack is assumed to be at an artificial center position in-stead of its actual position.
Resnouse to Question #1 Since the seismic motion is essentially vibratory, the excitation imparted to the structures is one characterized by a large number of harmonic inputs. Harmonic inputs, by definition, have a zero net integral. A statement to this effect was made at the above ref-erenced meeting to allay the reviewer's concern that the rack may move towards the most proximate rack and impact it before the mathe-matical model, predicated on geg, would predict the incidence of impact. In our response, we point out that the mathematical model for the rack seeks to replace a highly coupled system of vibrating bodies, to one wherein the rack vibrates inside a fictitious con-tainer attached to the ground. The coupling and hydrodynamic masses
- .- . - - - .. .~
^5 _
[*? EE 4
are estimated by centering this rack in that container. 'This as-sumption is merely consistent with the rest; not in itself neces - '
sarily conservative. .The conservatism, perhaps an excessive amount,- '
< is introduced at earlier steps in the model wherein a complete
-coherence of motion of all fuel assemblies at all times during the ;
earthquake is enforced, all fluid drag ef fects are neglected, etc.
- Recent research on the non-linear hydraulic coupling effect between f
-vibrating bodies subject to harmonic excitations confirms the in- j
~tuitive result that the smaller gap provides more of an inertial t resistance to its closing-than a larger gap. . Consequently, an !
eccentrically placed body will tend to center itself if subjected to l harmonic excitation of sufficient magnitude. !
These concepts have provided the rationale for the model used by Oat cin' analyzing high density racks. i
- 2. In the referenced ameting [3], Jost ph Ost Corporation indicated ~ that the seismic loading in three directions was applied simultaneously [
to'the model. Please indicate whether these three input time his-tories are statistically independent as specified by Regulatory l Guide 1.92 [5].
- Response to Question #2 ;
This rack was designed on the basis of one horizontal ~ seismic motion l
.and one. vertical seismic motion. The time histories associated with !
- these two motions are statistically independent as specified by
-Regulatory Guide 1.92. In order to study what we feel is the most
, ' severe condition (i.e., a 3-D seismic input), the one specifiea horizontal component was broken into two components acting along the U rack x ands y directions, respectively. ,
L
- 3. With respect to the influence of coefficient of friction to the rack i f' displacement, the following table is prepared tased on the outputs given in Joseph Oat's submittal [6):
Loading Coefficient Maximum Case Rack Condition of Friction X-Displacement l
- i. E Full load 0.8 0.125 F Full load 0.8 1.298 i -
ii E Full load 0.8 0.125 1
- E Full load 0.2 0.655 ,
iii. F Full load 0.8 1.298 F Full load 0.2 0.535 l
I
_____U.__..-________.__._________._______
n-
,- a With reference to the above table, please respond to the following questions:
- a. For case i, racks E' and F are very similar with the exception that E is higher that F. Explain possible reasons why F has higher maximum displacement.
- b. For cases ii and iii, please provide possible reasons why a high coefficient of friction in case ii produces smaller maximum displacement and high coef ficient of friction in case iii pro-duces higher maximum displacement.
- c. Please provide disulacement and base support reaction time histories for case i (both racks E and F) with the following coefficients of friction: 0.2, 0.4, 0.6, and 0.8. If this information is not available, it is strongly recommended, as a minimum, that outputs for coef ficients of friction 0.2 and 0.8 should be generated for review.
Response to Question #3 While it is true that racks E and F are very similar in total mass and shape, there are some differences, which, when coupled with the highly non-linear nature of the problem, can account for the dif-ferences in maximum displacement.
- a. Rack F has somewhat less stiffness for bending in the x direction.
- b. Rack F has shorter legs which leads to higher vertical stiff-nesses in these members.
- c. The gaps, simulating the spacing between adjacent structure used to calculate the virtual fluid masses, are somewhat different.
- d. The location of the support feet (in the x y plane) is somewhat ,
dif ferent for each rack type.
Graphs showing the reaction force history in each of the legs for rack E (run 10) and for rack F (run 51) have been provided to FRC.
The results are for full racks, C0F=.8, and are listed as case i in the FRC table. We can see that there is a substantial dif ference in behavior. Note that a zero value for any foot load implies that the foot has lifted off. The feet labelled "1,2", are in the positive x half of the base, while the feet labelled "3,4" are in the negative x half of the base plane.
Up to about 4.5 seconds into the event, both racks show essentially the same foot behavior. That is, only one foot loses contact.
Subsequent to this point, the behavior dif fers markedly. For rack F, we see considerable rocking, where 2 feet are off the ground. We m
r -
a cannot ascertain whether sliding is occurring from this data, but certainly at about 13.5 seconds into the event, a third foot has a relatively low compressive load. Rack E shows most of its rocking motion between 2 seconds and 8 seconds, while the rocking of rack F ,
is carried through the entire event. Note from the graphical re-suits that the peak compresrive load on a single foot generally 1 correlates well in time with the occurrence of the maximum displace- i ment. We conclude that the differences in maximum displacement are to be expected given the differences in the rack rocking behavior under the seismic event. ;
- 4. With respect to Response No. 3 of Reference 1, please confirm whether the following information is true:
- a. For a coefficient of friction of 0.8, maximum displacement always occurs at the instant three support legs are lifted off the pool floor and the rack never gets into the sliding mode.
- h. For a coefficient of friction of 0.2, the maximum displacement always occurs in the sliding mode and the rack never lifts off the pool floor.
Also, please respond to the following question: ,
For the case where three support legs are lif ted off the pool floor, please indicate a typical number of time steps during which they are ,
off the floor. It is noted that this is a completely unstable con-figuration in which the rack itself loses all of its res istir.g capacity against the motion along the horizontal directi ...c, and the chance that the two horizontal components of seismic inp t sotion would form a stable balance (no rotation of the rack) is cemote.
Please address this concern.
Response to Question #4 We cannot confirm the generality of the FRC statement. From the detailed data presented here, what we are willing to say is that the
, maximum displacement under a specified 3-D earthquake probably ,
occurs when either three feet are of f the ground and the rack is pivoting about the fourth foot, or when two feet are of f the ground and a third foot is sliding.
We are not clear on the meaning of the FRC statement concerning unstable configurations. It is precisely because we have recognized the possibility of pivoting about a single foot that we have gone to the detail of a 3-D model. Our model admits the possibility of gross rotation about the z axis; when and if this occurs equilibrium is still satisfied since the moment due to the seismic input is balanced by the moment due to the inertia forces.
=
f
- 5. With respect to Response 2 of Reference 1 Joseph Oat Corp, ration indicated'that the fluid coupling term represents the hydrodynamic
> mass contribution due to motion of the plane of symmetry in anti-symmetric motion. Please respond to the following questions:
Since the analysis was carried out for cne rack at a time, indi-cate whether the model has the capability to identify symmetric or antisymmetric motion. For symmetric motion, please confirm whether this plane of symmetry exists and how the gap is treated.
According to the Joseph Oat Corporation approach, the planes of symmetry around a rack, in ef fect, will form four rigid walls around the rack and have the motion of the pool floor. Provide a technical basis to validate this approach. In reality, it is most likely that the fluid will cross these planes of symmetry, and.there should be= free interaction between racks.
Response to Question #5 The response to this question is best answered by reference to Figure 1 attached. FRC is correct in their ascertion that in general all-racks will move independently. However, the extremes are as shown on the figure. In the antisymmetric mode, we can define hypothetical planes of symmetry and use Fritz's relation for virtual fluid mass based on the nominal gap. In this case, inter-rack impact potential is most severe. In the symmetric mode, no such symmetry plane can be defined, except at infinity. The fluid virtual mass expression used here is the value for an isolated rack. Note that in the symmetric mode, postulated, inter-rack im-pact is precluded and rack stress levels are the only considera-l-. tion. Because of the infinite number of possibilities, we have chosen to' study only the two extreme cases; these extremes are applied to both horizontal directions in any specific run.
B. Spent *.uel Pool Analysis
- 1. The Licensee stated that different finite element models were used for static and dynamic (seismic) analysis of the fuel pool slab, and that the results of the two' analyses are compared to determine the dynamic load factors. The resulting small value of 0.005 (Page 8-7
[7]) of the seismic multiplying factor does not seem to confirm the conservative nature of this approach.
A clarification of this comparison and the unusually small value of dynamic amplification factor is requested.
Please provide information on how the ef fect of a 40-f t column of water was included, and on the lumping of the distributed masa to 9 master degree of freedom.
r .
Information is also ' requested describing how the effects of the !
wall hydrostatic and hydrodynamic pressures on the slab were considered.
Response to Question #1
- a. Y small value of dynamic amplification. factor of .005 for the response of the pool floor itself (and the column of water) '
reflects the fact that the vertical seismic acceleration itself is relatively low and that the structural 9 DOF model for the !
, floor uses 7% damping reflecting the predominate concrete ;
structure. ,
- b. Wall hydrostatic and hydrodynamic effects are not included in the model. To permit disregard of these effects, we have assumed simply supported boundary conditions.co exist on the
'three edges not abutting the reactor walls. The reactor wall was assumed to be completely fixed against rotation. -The weight of the walls was included, however, so that we could. approximate the correct column reactions. ,
- c. The effect of the 40-ft. head is accounted for by defining an l Leffective density for the concrete slab. Thus, the water weight
.is accounted for in the. dead weight ANSYS analysis,-and its mass
~
i is also included in the ANSYS eigenvalue analysis. The effect of the racks and the fuel assemblies are accounted for in the j dynamic foot loadings. '
- d. A clarification of the determination of emplifi' cation factors is-
(1) The ANSYS model of the floor is used to determine the
+
behavior of the floor under the. dead weight of the floor plus the 40' of water. ;
(2) The ANSYS model is also used to determine the behavior of 7 the floor under the dead weight of the racks plus as- [
semblies. This case is analyzed under the assumption that the loading is a concentrated loading applied at 9 interior points and a series of points around the edge of the slab. ,
i' The location of the concentrated load points approximates ;
i the location of the rack feet groups (from 2,3, or 4 adjacent racks).
(3) The output from the above static runs are floor moments and floor displacements. Of importance to the derivation of amplification factors are the displacements at the 9 chosen interior points.
k i
4- >
f i >
F .3 '
V-
. y. . .
(4) We next use the ANSYS model for an eigenvalue-eigenvector analysis. .'We choose the lateral deformation of the 9
. locations mentioned above as the master degrees of free- '
dom. . We need not make any judgment on. mass lumping since
' ANSYS does the lumping based on our specification of the degrees of freedom. With reference to p. 8-3 of the .
-Licensing Document,'the output from the ANSYS analysis, the 'I
' ectual seismic time history -in the vertical direction, and the output of the floor load time history from our indi-vidual rack analysis, provides us with all of the informa-
, tion to write the 9 equations t
x Zn (t) + w n Zn (t) = Gn(t) n = 1,2. 9 where wn are the natural frequencies of the reduced system (obtained from ANSYS) and nG (t) are known func-tions. Thus, by using the ANSYS model, we are able to i derive the information necessary to develop a 9 DOF floor model which can be used to generate dynamic displacements of-the floor slab.
(5) The' Joseph Oat dynamic program DYNAHIS'is used for the ,
above 9 DOF model to study the slab dynamics first under the vertical seismic event alone, and then under the rack r support loads (applied at each of the nine locations simul- '
taneously). The output from the dynamic analysis (maximum !
displacement at each location) is compared with the static i displacement obtained from ANSYS and an average amplifica- ,
tion factor is obtained. These factors are then used to amplify the static ' floor moments which are then compared to r the ACI allowables.
- 2. The Licensee has stated (Section 8.2.2) that the stiffness and i strength of concrete are based on complete. cracking of concrete.
Please provide information whether'the section capacities listed in Tables 8.2 and 8.3 are also based on the same assumption. ,
Please provide information on whether properties of the slab were ;
calculated on an orthotropic or isotropic basis, and how the varia- '
tioa of reinforcement on the two faces of slab and in dif ferent directions was accounted for.
Response to Question #2 {
The section capacities listed in Tables 8.2, 8.3 are based on com- ;
plete cracking of the concrete. The slab properties were calculated !
on an orthotropic basis for use in the ANSYS model. That is, the 7 actual reinforcement in the two orthogonal directions was used to ',
. compuce uffective properties. The effective properties of the beam l d
i L i
I i
i
)
o 2 <-m- -sw- - e- - . - - , , - ~,, , , . - . . , . . , , ,,,-.,,,.,%,_,..-,e,.,.,,,,, .,.,,e,,,-we, -%n- -, .~%m.,---.,..,,,m., - , - . - - , , , - - , . .
elements were computed using the actual locations of the rein-forcements as ascertained for the GPU' drawings. For the ANSYS analysis, the properties based on the water side being in tension were used (as opposed to the rack side of the slab in tension).
- 3. The thermal loading has been based on a 21*F temperature differen-
'tial across the slab depth. The thermal moment due to a temperature gradient is calculated by a formula given on page 8-6 [7] which uses
~
the. ef fective Young's modulus, E*, for a homogeneous slab. Please-provide full details on the calculation of E* and the conservatism of using cracked . sections in this calculation, if it was based on this assumption.
Response to Question #3 The calculation of the effective E* is based on a standard strength of materials approach which calculates a new neutral axis reflecting the number and location of reinforcing rods, accing in tension. A balance of forces evolves the location of the new neutral axis; a calculation of the moment curvature relation about the new neutral axis then gives M
- g=EI which defines E* based on I = H 3 /12 (1-v 2), H being the slab thickness.
Since the thermal. moment gives compressive stresses in the rack side, the E* used is based on tension in the water side concrete.
An E* ef fective is used to be consistent with the remainder of our analysis. -Since we compute an ultimate carrying capacity based on cracked concrete, it is consistent to use the same assumption to compute the thermal moment. Use of the actual E in the thermal moment equation simply predicts that the concrete must crack, thus relieving the moment.
4.. The floor slab moment capacity from Table 8.2 [7] (Mu = 48,350 lb-in/in) seems quite low in comparison to the other values. Please confirm the correct value.
Response to Question #4 The floor slab moment capacity of 48,350 in.#/in. is in fact low in t.his area because the imbedment length of the reinforcement in this area (water side of the slab in compression) is less than called for by ACI. Therefore, in this area, a reinforcement maximum stress of 6000 psi was 'used, per ACI, to account for the less than fully effective reinforcement rods.
REFERENCES n- 1. . Joseph Oat Corporation, Preliminary Responses to FRC's List of
- Questions, May 7,-1984
- 2. Joseph Oat Corporation, Additional Responses to FRC's List of Questions, May 11, 1984
- 3. Technical Meeting with Joseph Oat at FRC on May 7, 1984
- 4. R. J. Fritz, "The Ef fects of. Liquids on the Dynamic Motions of Immersed
-Solids," Journal of Engineering for Industry, Trans. ASME, February 1972, pp. 167-172 5.- U.S. Nuclear Regulatory Commission, Regulatory Guide'1.92, " Combining Modal Responses and Spatial Components in Seismic Response Analysis,"
' February 1976
- 6. ,A. I. Soler, " Seismic Analysis of High-Density Fuel Racks for GPU Oyster Creek Nuclear Station," TM Report No.'678, April 24, 1984
- 7. GPU Nuclear, " Licensing Report on High-Density Spent Fuel Racks for Oyster Creek Nuclear Generating Station," August 1983 4
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