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MONTHYEARML20237G3951987-08-17017 August 1987 Forwards Proposed Rev to DPC-NF-2010A, Nuclear Physics Methodology for Reload Methodology, for NRC Review.Rev Involves Change in Computer Codes Used to Generate cross- Sections.Fee Paid Project stage: Other 1987-08-17
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217H5811999-10-15015 October 1999 Forwards 1999 Update to FSAR, for McGuire Nuclear Station.With Instructions,List of Effective Pages for Tables & List of Effective Pages for Figures ML20217F3261999-10-13013 October 1999 Submits Quantity of Tubes Insp from Either Side of SGs A-D & Lists Tubes with Imperfections,Locations & Size.No Tubes Removed from Svc by Plugging.Total of Eleven Tubing Wear Indications Identified at Secondary Side Supports in SGs ML20217F3591999-10-13013 October 1999 Forwards Info Copy of Cycle 14 COLR for McGuire Nuclear Station,Unit 1 ML20217F8011999-10-13013 October 1999 Informs That on 990930,NRC Completed mid-cycle PPR of McGuire Nuclear Station.Areas That Warranted More than Core Insp Program Over Next Five Months,Not Identified.Historical Listing of Plant Issues Encl ML20217G7861999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for McGuire Nuclear Station,Unit 1 & 2 ML20217H0041999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20217F8231999-10-13013 October 1999 Informs That on 990930,NRC Completed mid-cycle PPR of Catawba Nuclear Station.Based on Review,Nrc Did Not Identify Any New Areas That Warranted More than Core Insp Program Over Next Five Months.Historical Listing of Issues,Encl ML20217F1301999-10-0707 October 1999 Forwards Rev 1 to Request for Relief 99-03 from Requirements of ASME B&PV Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting (Vg) Sys ML20217J5091999-10-0606 October 1999 Forwards Revs to Section 16.15-4.8.1.1.2.g of McGuire Selected Licensee Commitments Manual.Section Has Been Revised to Allow Testing of Portions of DG Fuel Oil Sys Every 10 Yrs ML20217C8351999-10-0505 October 1999 Communicates Correction to Info Provided During 990917 Meeting with Duke Energy & NRC Region Ii.Occupational Radiation Safety Performance Indicator Values Should Have Been Presented as 1 Instead of 0 ML20217C4471999-10-0404 October 1999 Forwards Insp Repts 50-369/99-06 & 50-370/99-06 on 990801- 0911.Determined That One Violation Occurred & Being Treated as Non-Cited Violation ML20212J2191999-10-0404 October 1999 Informs That Util 980326 Response to GL 97-06, Degradation of SG Internals Provides Reasonable Assurance That Condition of Steam Generator Internals Are in Compliance with Current Licensing Bases for Facility ML20212J7801999-10-0404 October 1999 Discusses GL 98-01 Issued by NRC on 980511 & DPC Responses for McGuire NPP & 990615.Informs That NRC Reviewed Responses & Concluded That All Requested Info Re Y2K Readiness Provided.Subj GL Considers to Be Closed ML20212J3011999-10-0101 October 1999 Forwards Exemption from Certain Requirements of 10CFR54.17(c) Re Schedule for Submitting Application for Operating License Renewal.Se Also Encl ML20217K2651999-10-0101 October 1999 Forwards Retake Exams Repts 50-413/99-302 & 50-414/99-302 on 990921-23.Two of Three ROs & One SRO Who Received Administrative Section of Exam Passed Retake Exam, Representing 75 Percent Pass Rate 05000414/LER-1999-004, Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments1999-09-27027 September 1999 Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments 05000413/LER-1999-015, Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept1999-09-27027 September 1999 Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept ML20217A7911999-09-24024 September 1999 Forwards Insp Repts 50-413/99-05 & 50-414/99-05 on 990718- 0828 at Catawba Facility.Nine NCVs Identified Involving Inadequate Corrective Actions Associated with Degraded Svc Water Supply Piping to Auxiliary Feedwater Sys ML20212E6471999-09-24024 September 1999 Discusses GL 98-01 Issued by NRC on 980511 & DPC Responses for Catawba NPP & 990615.Informs That NRC Reviewed Response for Catawba & Concluded That All Requested Info Provided.Considers GL 98-01 to Be Closed for Catawba ML20212M1651999-09-23023 September 1999 Refers to 990917 Meeting at Region II Office Re Licensee Presentation of self-assessment of McGuire Nuclear Station Performance.List of Attendees & Licensee Presentation Handouts,Encl ML20212F0941999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals for Cns,Units 1 & 2 05000414/LER-1999-005, Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments1999-09-20020 September 1999 Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments ML20212M2001999-09-20020 September 1999 Confirms 990913 Telcon Between M Purser & R Carroll Re Management Meeting to Be Conducted on 991026 in Atlanta,Ga to Discuss Operator Licensing Issues ML20212D1671999-09-20020 September 1999 Forwards Exemption & SER from Certain Requirements of 10CFR50,App A,General Design Criterion 57 Re Isolation of Main Steam Branch Lines Penetrating Containment.Exemption Related to Licensee Application ML20212B6491999-09-15015 September 1999 Informs That Encl Announcement Re 990913 Application for Amend to Licenses NPF-9 & NPF-7 Forwarded to C Observer in North Carolina,For Publication ML20212D5321999-09-15015 September 1999 Informs That Duke Energy Corp Agrees to Restrict Max Fuel Rod Average Burnup to 60,000 Mwd/Mtu,In Order to Support NRC Final Approval & Issuance of Requested Amend ML20216E8791999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for July 1999 for McGuire Nuclear Station ML20212A4131999-09-14014 September 1999 Informs That TR DPC-NE-2009P Submitted in 990817 Affidavit, Marked Proprietary,Will Be Withheld from Public Disclosure, Pursuant to 10CFR2.709(b) & Section 103(b) of Atomic Energy Act of 1954,as Amended ML20212B4641999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for Catawba Nuclear Station,Units 1 & 2 ML20212M1931999-09-13013 September 1999 Refers to 990909 Meeting Conducted at Region II Office Re Presentation of Licensee self-assessment of Catawba Nuclear Station Performance.List of Attendees & Licensee Presentation Handout Encl ML20212A3751999-09-10010 September 1999 Informs That Postponing Implementation of New Conditions Improved by RG 1.147,rev 12,acceptable Since Evaluation on Relief Based on Implementation Code Case for Duration of Insp Interval ML20212A0501999-09-10010 September 1999 Informs That Postponing Implementation of New Conditions Improved by RG 1.147,rev 12,acceptable Since Evaluation on Relief Based on Implementing Code Case for Duration of Insp Interval ML20212A2631999-09-0909 September 1999 Forwards Rev 25 to McGuire Nuclear Station,Units 1 & 2 Pump & Valve Inservice Testing Program, IAW 10CFR50.55a. Section 8.0 Contains Summary of Changes & Detailed Description of Changes Associated with Rev 25 ML20212A5191999-09-0808 September 1999 Requests NRC Approval for Relief from Requirements of ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,App VI,VI-2430(c) & 2440(b).Approval of 99-GO-002 Is Requested by 000301 05000413/LER-1999-014, Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment1999-09-0101 September 1999 Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment 05000414/LER-1999-003, Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev1999-08-31031 August 1999 Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev 0 of LER ML20211J3671999-08-31031 August 1999 Forwards Public Notice of Application for Amend to License NPF-9 Seeking one-time Extension of Surveillance Frequency for TS SR 3.1.4.2 Beyond 25% Extension Allowed by TS SR 3.0.2 ML20211H1741999-08-30030 August 1999 Forwards Comments on Catawba Nuclear Station Units 1 & 2 & McGuire Nuclear Station,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid.Ltr Dtd 990107,rept ATI-98-012-T005 & Partial marked-up Rept WCAP-14995 Encl ML20211M4451999-08-30030 August 1999 Forwards Summary of Util Conclusions Re Outstanding Compliance Issue Re Staff Interpretation of TS SR 3.0.1,per Insp Repts 50-369/99-03 & 50-370/99-03,as Discussed with NRC During 990618 Meeting ML20211K8831999-08-26026 August 1999 Forwards Insp Repts 50-369/99-05 & 50-370/99-05 on 990620-0731.Two Violations Occurred & Being Treated as NCVs ML20211M8191999-08-25025 August 1999 Confirms 990825 Telcon Between G Gilbert & R Carroll Re Mgt Meeting to Be Held on 990909 in Atlanta,Ga,To Allow Licensee to Present self-assessment of Catawba Nuclear Station Performance 05000413/LER-1999-013, Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER1999-08-25025 August 1999 Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER ML20211G5181999-08-24024 August 1999 Forwards SE Re second-10-yr Interval Inservice Insp Program Plan Request for Relief 98-004 for Plant,Unit 1 ML20211A9641999-08-20020 August 1999 Forwards SE Authorizing Licensee 990118 Request for Approval of Proposed Relief from Volumetric Exam Requirements of ASME B&PV Code,Section XI for Plant,Units 2 ML20211B9471999-08-18018 August 1999 Forwards Request for Relief 99-02,associated with Limited Exam Results for Welds Which Were Inspected During Unit 1 End of Cycle 11 RFO ML20211C1191999-08-18018 August 1999 Forwards ISI Rept Unit 1 Catawba 1999 RFO 11, Providing Results of ISI Effort Associated with End of Cycle 11 ML20211C3651999-08-17017 August 1999 Forwards Rev 25 to Catawba Nuclear Station Units 1 & 2 Pump & Valve Inservice Testing Program, Which Includes Reformatting of Manual & Addl Changes as Noted in Attached Summary of Changes ML20211F2971999-08-17017 August 1999 Forwards non-proprietary & Proprietary Updated Pages for DPC-NE-2009,submitted 980722.Pages Modify Fuel Design & thermal-hydraulic Analysis Sections of DPC-NE-2009. Proprietary Page 2-4 Withheld,Per 10CFR2.790 ML20211B1121999-08-16016 August 1999 Forwards Topical Rept DPC-NE-2012, Dynamic Rod Worth Measurement Using Casmo/Simulate, Describing Results of Six Drwm Benchmark Cycles at Catawba & McGuire & Discusses Qualification to Use Drwm at Catawba & McGuire 05000413/LER-1999-011, Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment1999-08-16016 August 1999 Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217H5811999-10-15015 October 1999 Forwards 1999 Update to FSAR, for McGuire Nuclear Station.With Instructions,List of Effective Pages for Tables & List of Effective Pages for Figures ML20217G7861999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for McGuire Nuclear Station,Unit 1 & 2 ML20217F3591999-10-13013 October 1999 Forwards Info Copy of Cycle 14 COLR for McGuire Nuclear Station,Unit 1 ML20217F3261999-10-13013 October 1999 Submits Quantity of Tubes Insp from Either Side of SGs A-D & Lists Tubes with Imperfections,Locations & Size.No Tubes Removed from Svc by Plugging.Total of Eleven Tubing Wear Indications Identified at Secondary Side Supports in SGs ML20217H0041999-10-13013 October 1999 Forwards MOR for Sept 1999 & Revised MOR for Aug 1999 for Catawba Nuclear Station,Units 1 & 2 ML20217F1301999-10-0707 October 1999 Forwards Rev 1 to Request for Relief 99-03 from Requirements of ASME B&PV Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting (Vg) Sys ML20217J5091999-10-0606 October 1999 Forwards Revs to Section 16.15-4.8.1.1.2.g of McGuire Selected Licensee Commitments Manual.Section Has Been Revised to Allow Testing of Portions of DG Fuel Oil Sys Every 10 Yrs ML20217C8351999-10-0505 October 1999 Communicates Correction to Info Provided During 990917 Meeting with Duke Energy & NRC Region Ii.Occupational Radiation Safety Performance Indicator Values Should Have Been Presented as 1 Instead of 0 05000414/LER-1999-004, Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments1999-09-27027 September 1999 Forwards LER 99-004-01,providing Correction to Info Previously Provided in Rev 0 of Rept.Planned Corrective Actions Contain Commitments 05000413/LER-1999-015, Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept1999-09-27027 September 1999 Forwards LER 99-015-00 Re Inoperability of Auxiliary Bldg Ventilation Sys That Exceeded TS Limits Due to Improperly Positioned Vortex Damper.Commitments Are Contained in Corrective Actions Section of Encl Rept 05000414/LER-1999-005, Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments1999-09-20020 September 1999 Forwards LER 99-005-00 Re Missed EDG TS Surveillance Concerning Verification of Availability of Offsite Power Sources Resulted from Defective Procedure.Planned Corrective Actions Stated in Rept Represent Regulatory Commitments ML20212D5321999-09-15015 September 1999 Informs That Duke Energy Corp Agrees to Restrict Max Fuel Rod Average Burnup to 60,000 Mwd/Mtu,In Order to Support NRC Final Approval & Issuance of Requested Amend ML20216E8791999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for July 1999 for McGuire Nuclear Station ML20212B4641999-09-14014 September 1999 Forwards Monthly Operating Repts for Aug 1999 & Revised Monthly Operating Rept for Catawba Nuclear Station,Units 1 & 2 ML20212A2631999-09-0909 September 1999 Forwards Rev 25 to McGuire Nuclear Station,Units 1 & 2 Pump & Valve Inservice Testing Program, IAW 10CFR50.55a. Section 8.0 Contains Summary of Changes & Detailed Description of Changes Associated with Rev 25 ML20212A5191999-09-0808 September 1999 Requests NRC Approval for Relief from Requirements of ASME Boiler & Pressure Vessel Code,Section XI,1989 Edition,App VI,VI-2430(c) & 2440(b).Approval of 99-GO-002 Is Requested by 000301 05000413/LER-1999-014, Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment1999-09-0101 September 1999 Forwards LER 99-014-00, Missed Surveillance & Operation Prohibited by TS Occurred as Result of Defective Procedures or Program & Inappropriate TS Requirements. Planned Corrective Action Stated in Rept Represents Commitment 05000414/LER-1999-003, Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev1999-08-31031 August 1999 Forwards LER 99-003-01, Unplanned Actuation of ESFAS Due to a SG High Level Caused by Inadequate Procedural Guidance. Suppl Rept Provides Info Re Root Cause & Corrective Actions Associated with Event Developed Subsequent to Rev 0 of LER ML20211H1741999-08-30030 August 1999 Forwards Comments on Catawba Nuclear Station Units 1 & 2 & McGuire Nuclear Station,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid.Ltr Dtd 990107,rept ATI-98-012-T005 & Partial marked-up Rept WCAP-14995 Encl ML20211M4451999-08-30030 August 1999 Forwards Summary of Util Conclusions Re Outstanding Compliance Issue Re Staff Interpretation of TS SR 3.0.1,per Insp Repts 50-369/99-03 & 50-370/99-03,as Discussed with NRC During 990618 Meeting 05000413/LER-1999-013, Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER1999-08-25025 August 1999 Forwards LER 99-013-00,re RHR Heat Exchanger Bypass Valves Not Verified Per TS Surveillance.Surveillance Procedures Have Been Revised & There Are No Further Planned Corrective Actions or Commitments in LER ML20211B9471999-08-18018 August 1999 Forwards Request for Relief 99-02,associated with Limited Exam Results for Welds Which Were Inspected During Unit 1 End of Cycle 11 RFO ML20211C1191999-08-18018 August 1999 Forwards ISI Rept Unit 1 Catawba 1999 RFO 11, Providing Results of ISI Effort Associated with End of Cycle 11 ML20211C3651999-08-17017 August 1999 Forwards Rev 25 to Catawba Nuclear Station Units 1 & 2 Pump & Valve Inservice Testing Program, Which Includes Reformatting of Manual & Addl Changes as Noted in Attached Summary of Changes ML20211F2971999-08-17017 August 1999 Forwards non-proprietary & Proprietary Updated Pages for DPC-NE-2009,submitted 980722.Pages Modify Fuel Design & thermal-hydraulic Analysis Sections of DPC-NE-2009. Proprietary Page 2-4 Withheld,Per 10CFR2.790 ML20211B1121999-08-16016 August 1999 Forwards Topical Rept DPC-NE-2012, Dynamic Rod Worth Measurement Using Casmo/Simulate, Describing Results of Six Drwm Benchmark Cycles at Catawba & McGuire & Discusses Qualification to Use Drwm at Catawba & McGuire 05000413/LER-1999-011, Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment1999-08-16016 August 1999 Forwards LER 99-011-00,re Missed Surveillance on Both Trains of CR Area Ventilation Sys Resulting in TS Violation.Planned Corrective Actions Stated in LER Represent Regulatory Commitment ML20210S2751999-08-12012 August 1999 Forwards Monthly Operating Repts for July 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for June 1999,encl ML20210S2231999-08-12012 August 1999 Forwards Monthly Operating Repts for July 1999 for McGuire Nuclear Station,Units 1 & 2.Revised Rept for June 1999,encl ML20210R0031999-08-10010 August 1999 Forwards Revised TS Bases Pages to NRC for Info & Use. Editorial Changes Were Made to Correct Incorrect UFSAR Ref Number Associated with Certain Reactor Coolant Sys Pressure Isolation Valves ML20210R4311999-08-10010 August 1999 Forwards Summary Rept of Mods,Minor Mods,Procedure Changes & Other Misc Changes Per 10CFR0.59 ML20210T4511999-08-10010 August 1999 Forwards Response to NRC RAI Re 981014 Standby Nuclear Svc Water Pond Dam Audit Conducted by FERC ML20210N9521999-08-0404 August 1999 Forwards Changes to Catawba Nuclear Station Selected Licensee Commitments Manual.Documents Constitutes Chapter 16 of Ufsar.With List of Effective Pages ML20210M6411999-07-29029 July 1999 Forwards Request for Relief 99-03 from Requirements of ASME Boiler & Pressure Vessel Code,In Order to Seek Relief from Performing Individual Valve Testing for Certain Valves in DG Starting Air (Vg) Sys 05000413/LER-1999-010, Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units1999-07-22022 July 1999 Forwards LER 99-010-01 Which Replaces LER 99-002.Rept Number Has Been Changed in Order to Conform to Numbering Convention Specified in NUREG-1022,since Primary Event Involved Both Units 05000413/LER-1999-009, Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept1999-07-19019 July 1999 Forwards LER 99-009-00,re Inoperability of Containment Valve Injection Water Sys Valve in Excess of TS Limits.Root Cause & Corrective Actions Associated with Event Are Being Finalized & Will Be Provided in Supplement to Rept 05000414/LER-1999-001, Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed1999-07-15015 July 1999 Forwards LER 99-001-01 Re Unanalyzed Condition Associated with Relay Failure in Auxiliary Feedwater Sys,Due to Inadequate Single Failure Analysis.Rev Is Being Submitted to Include Results of Failure Analysis Which Was Performed ML20209H4431999-07-14014 July 1999 Forwards Monthly Operating Repts for June 1999 for Catawba Nuclear Station,Units 1 & 2.Revised Rept for May 1999 on Unit Shutdowns Also Encl ML20210A5771999-07-14014 July 1999 Forwards Revsied Catawba Nuclear Station Selected Licensee Commitments Manual, Per 10CFR50.71(e),changing Sections 16.7-5,16.8-5,16.9-1,16.9-3,16.9-5 & 16.11-7.Manual Constitute Chapter 16 of UFSAR ML20216D3941999-07-14014 July 1999 Forwards Revs to Catawba Nuclear Station Selected Licensee Commitments Manual ML20209H1551999-07-14014 July 1999 Forwards Monthly Operating Repts for June 1999 for McGuire Nuclear Station,Units 1 & 2.Revised Rept for May 1999 Also Encl ML20209G5151999-07-0808 July 1999 Forwards Amended Pages to Annual Radioactive Effluent Release Repts, for 1997 & 1998 for McGuire Nuclear Station. Portion of Rept Was Inadvertently Omitted Due to Administrative Error,Which Has Been Corrected 05000413/LER-1999-008, Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach1999-07-0808 July 1999 Forwards LER 99-008-00,re Operation Prohibited by TS 3.5.2. Rev to LER Will Be Submitted by 990812 Which Will Include All Required Info About Ventilation Sys Pressure Boundry Breach ML20196G3721999-06-24024 June 1999 Documents Verbal Info Provided to NRR During Conference Call Re Relief Requests 98-002 & 98-003 ML20196G7461999-06-22022 June 1999 Requests Exemption from Requirements of 10CFR54.17(c) That Application for Renewed Operating License Not Be Submitted to NRC Earlier than 20 Yrs Before Expiration of Operating License Currently in Effect ML20196E9541999-06-18018 June 1999 Forwards SG Tube Insp Conducted During Unit 1 End of Cycle 11 Refueling Outage.Attachments 1,2,3 & 4 Identify Tubes with Imperfections in SGs A,B,C & D,Respectively ML20195K4571999-06-14014 June 1999 Forwards MORs for May 1999 & Revised MORs for Apr 1999 for Catawba Nuclear Station,Units 1 & 2 ML20195K3601999-06-14014 June 1999 Forwards MORs for May 1999 for McGuire Nuclear Station,Units 1 & 2 & Revised MORs for Apr 1999.Line 6 Max Dependable Capacity (Gross Mwe) on Operating Data Rept Should Be Revised to 1114 from Jan 1998 to Apr 1999 ML20195J1691999-06-10010 June 1999 Forwards Written Documentation of Background & Technical Info Supporting Catawba Unit 1,notice of Enforcement Discretion Request Re TS 3.5.2 (ECCS-Operating),TS 3.7.12 (Auxiliary Bldg Filtered Ventilation Exhaust Sys) ML20217G5771999-06-0909 June 1999 Forwards Post Exam Comments & Supporting Reference Matls for Written Exams Administered at Catawba Nuclear Station on 990603 1999-09-09
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L0101990-09-19019 September 1990 Provides Addl Response to Violations Noted in Insp Repts 50-369/90-11 & 50-370/90-11,per .Corrective Action:Operations & Compliance Mgt Further Examined Event & Concluded That Trains Vc/Yc Inoperable Per Special Order ML20059L0091990-09-18018 September 1990 Informs of Delay in Response to Violations Noted in Insp Repts 50-369/90-14 & 50-370/90-14.Delay Due to Ongoing Review ML20059J7121990-09-14014 September 1990 Forwards Monthly Operating Rept for Aug 1990 for McGuire Nuclear Station Units 1 & 2 & Monthly Operating Status Rept for Jul 1990 ML20059L5491990-09-14014 September 1990 Forwards Proprietary Response to Question Re Scope of Review of Topical Rept, Safety Analysis Physics Parameter & Multidimensional Reactor Transients Methodology, Per & 900723 Meeting.Response Withheld ML20059L5521990-09-14014 September 1990 Forwards Response to 18 Questions Re Topical Rept DPC-NE-2004,per NRC 900802 Request for Addl Info.Encl Withheld (Ref 10CFR2.790) ML20059K2021990-09-12012 September 1990 Submits Supplemental Response to Generic Ltr 89-14, Svc Water Sys Problems Affecting Safety-Related Equipment. Intake Structure Insp Program Developed.Procedures for Insp Implemented & Intake Structures Sampled & Analyzed ML20065D4671990-09-11011 September 1990 Informs of Delay in Response to Violation Noted in Insp Repts 50-369/90-11 & 50-370/90-11,per Commitment in ML20064A8031990-09-10010 September 1990 Responds to NRC Ltr Re Violations Noted in Insp Repts 50-369/90-13 & 50-370/90-13.Corrective Actions:Procedures Dealing W/Fuel Handling Will Be Enhanced by Adding Sign Off to Sections Which Ref Fuel Handling in Tech Specs ML20064A8041990-09-0505 September 1990 Notifies NRC of Mod to 890301 Response to Violations Noted in Insp Repts 50-413/86-18-01 & 50-414/86-18-01 Re Valves. All Valve Locking Mechanisms Would Be Installed by End of Unit 2 Refueling Outage (Approx Aug 1990) ML20059G3011990-09-0404 September 1990 Forwards Response to NRC 900327 Request for Addl Info Re BAW-10174, Mark-BW Reload LOCA Analysis for Catawba & Mcguire ML20064A5741990-09-0404 September 1990 Discusses Re Info to Support Util Position Relative to Resolving Issue of Main Steam Line Breaks Inside Ice Condenser Containments & Requests That Info Be Withheld (Ref 10CFR2.790) ML20064A5671990-09-0101 September 1990 Advises That Corrective Actions Re Violation of Insp Rept 50-369/89-01 & 50-370/89-01 Will Be Completed by 910101 ML20059D3461990-08-30030 August 1990 Forwards Semiannual Effluent Release Rept for First Half of CY90 & Revs to Process Control Program ML20059G8321990-08-30030 August 1990 Withdraws 880726 Proposed Tech Spec Change,Clarifying Tech Spec 3/4.7.6 Re Emergency Power Requirements for Control Room Ventilation Sys ML20059D2011990-08-27027 August 1990 Forwards Piedmont Municipal Power Agency , Authorizing Use of Annual Rept for NRC Docket Requirements ML20059D2441990-08-24024 August 1990 Forwards Special Rept PIR-1-C90-0261 on 900725 Re Cathodic Protection Sys Failure to Pass Acceptance Criteria of 60-day Surveillance.Std Work Request Generated to Check Voltage Potential at Test Station TS-36 on Weekly Basis ML20056B5011990-08-22022 August 1990 Responds to Violation Noted in Insp Repts 50-413/90-17 & 50-414/90-17.Corrective Actions:Review Will Be Conducted to Determine Category of Infrequently Run Procedures Needing Addl Verification Controls ML20056B4981990-08-22022 August 1990 Responds to NRC Request for Addl Info Re General Relief Request for Pump Vibration Submitted 900315.Relief Request Changed to Insure Data Taken Over Range That Encompasses All Main Potential Noise Contributors ML20059D1901990-08-22022 August 1990 Responds to Violations Noted in Insp Repts 50-369/90-11 & 50-370/90-11.Corrective Actions:Valves VC-1,VC-2,VC-3,VC-4, VC-9,VC-10,VC-11 & VC-12 Reopened Upon Discovery of VC-1 - VC-4 Being Closed & Procedure IP/0/13/3006/09 Revised ML20059G9141990-08-22022 August 1990 Forwards Public Version of Revised Epips,Including Change 0 to RP/O/A/5700/01,Change 0 to RP/O/A/5700/02 & Change 0 to RP/O/A/5700/03 ML16259A2391990-08-22022 August 1990 Forwards Public Version of Rev 27 to Company Crisis Mgt Implementing Procedure CMIP-2, News Group Plan. W/ Dh Grimsley 900906 Release Memo ML20056B4971990-08-20020 August 1990 Clarifies Info Submitted in 871207 & s Re Steam Generator Tube Rupture Analysis Demonstration Runs. Demonstration Runs Met plant-specific Requirements in Section D to NRC SER on WCAP-10698 ML20059B4451990-08-20020 August 1990 Requests That NRC Approval of Original Relief Request 89-02 Be Extended to Include Changes to Mods MG-22233 & MG-22243 as Described.Hydrostatic Testing Impractical Due to Inadequacy of Test Boundaries ML20059C1201990-08-20020 August 1990 Forwards Rept Summarizing Util Findings Re Three False Negative Blind Performance Urine Drug Screens Which Occurred During Jan & Feb 1990.Recommends That NRC Consider Generic Communication to Clearly State Reporting Requirement ML20059C1591990-08-17017 August 1990 Suppls by Providing Addl Info to Support Util Position Re Anl Confirmatory Analysis of Main Steamline Breaks in Ice Condenser Plants.Encl Withheld ML20059B6581990-08-17017 August 1990 Responds to Violation Noted in Insp Repts 50-413/90-15 & 50-414/90-15.Corrective Actions:Present Methods of Testing Operability of CO2 Fire Protection Sys Will Be Evaluated by 910201 to Determine If Addl Testing Necessary ML20063Q0951990-08-15015 August 1990 Forwards Monthly Operating Rept for Jul 1990 for Catawba Nuclear Station Units 1 & 2 & Revised Rept for June 1990 ML20059C1231990-08-15015 August 1990 Advises That Util Submitting Special Rept Re Valid Failure of Diesel Generator 2B Would Be Delayed Until 880229 Had Incorrect Ltr Date.Date of Ltr Should Have Been 880204 Instead of 880104.Corrected Ltr Encl ML20063Q0901990-08-15015 August 1990 Forwards Monthly Operating Repts for Jul 1990 for McGuire Nuclear Station Units 1 & 2 & Revised Rept for June 1990 ML20063Q2671990-08-14014 August 1990 Forwards Public Version of Revised Crisis Mgt Implementing Procedures,Including Rev 36 to CMIP-1,Rev 32 to CMIP-4,Rev 36 to CMIP-5,Rev 41 to CMIP-6,Rev 40 to CMIP-7,Rev 27 to CMIP-8 & Rev 35 to CMIP-9.W/DH Grimsley 900821 Release Memo ML20059C2211990-08-13013 August 1990 Forwards Revised Chapter 16, Selected Licensee Commitments Manual, to Plant Updated Fsar,Per 10CFR50.4 & 50.71.Manual Contains Commitments Which Require Control But Not Appropriate in Tech Specs ML20063Q0671990-08-10010 August 1990 Submits Revised Response to Violations Noted in Insp Rept 50-413/90-09.Procedure to Verify Test Inputs Modified to Verify Dummy Input Signal to Channel RTD Circuit ML20063Q0261990-08-10010 August 1990 Forwards Rev 0 to Catawba Unit 2 Cycle 4 Core Operating Limits Rept, Per Tech Spec 6.9.1.9 ML20058N0181990-08-0808 August 1990 Forwards Response to Request for Addl Info Re BAW-10174, Mark-BW Reload LOCA Analysis for Catawba & Mcguire ML20063Q0181990-08-0707 August 1990 Confirms That Planned Corrective Actions Listed in Paragraph 3 of Insp Repts 50-369/90-10 & 50-370/90-10,appropriately Described ML20058L1811990-08-0101 August 1990 Forwards Safety Evaluation of Large Break LOCA Event Reanalysis Per 10CFR50.46 Re Peak Clad Temp Increases ML20058L4061990-08-0101 August 1990 Forwards Rev 0 to Inservice Insp Rept Unit 1 McGuire 1990 Refueling Outage 6 ML20059K0291990-08-0101 August 1990 Forwards Comments on Reactor Operator & Senior Reactor Operator Written Exams Administered on 900730 ML20058P3261990-08-0101 August 1990 Forwards Public Version of Rev 26 to Station Directive 3.8.4, Onsite Emergency Organization ML20081E1601990-08-0101 August 1990 Advises of Completion of 900330 Commitment Re Standing Work Request for Insp of Air Flow Monitors & Dampers,Per Violations Noted in Insp Rept 50-413/90-03 & 50-414/90-03 ML20056A9541990-08-0101 August 1990 Provides Final Response to NUREG-0737,Item II.D.1, Performance Testing of Safety Relief Valves. Util Will Change Seal Configuration for Existing Valves to Seal on Steam Medium & Install Continuous Drain for Loop Piping ML20081E0951990-07-27027 July 1990 Forwards Decommissioning Financial Assurance Certification Rept for Duke Power Co,co-owner of Catawba Nuclear Station Units 1 & 2 ML20055H9741990-07-26026 July 1990 Forwards end-of-cycle 3 Steam Generator Insp Rept.Nineteen Tubes Removed from Svc by Plugging W/Rolled Mechanical Plug ML20055H5231990-07-24024 July 1990 Discusses co-licensee Relationship & Obligations Re Decommissioning Financial Assurance for Facilities ML20056A1571990-07-23023 July 1990 Forwards Public Version of Change 0 to RP/0/A/5700/10 & Rev 25 to Section 18.2 of Radiation Protection Manual.W/ Dh Grimsley 900730 Release Memo ML20055H4571990-07-19019 July 1990 Responds to NRC Re Violations Noted in Insp Repts 50-413/90-11 & 50-414/90-11.Corrective actions:I-beams/ Hoists Rolled to Ends of Ice Condenser & Securely Located on Rails to Prevent Any Movement ML20055H1741990-07-18018 July 1990 Withdraws 880527 & 0725 Amends Clarifying Requirements for Containment Pressure Control Sys ML20055J3441990-07-17017 July 1990 Advises That Commitment Re Procedure IP/O/A/3190/01,per Violation in Insp Repts 50-413/90-06 & 50-414/90-06, Completed on 900619 NLS9000259, Responds to Violations Noted in Insp Repts 50-369/90-09 & 50-370/90-09.Corrective actions:in-house Training Will Be Provided on Correct Method for Determining Proper Response to Problem Investigation Processes1990-07-16016 July 1990 Responds to Violations Noted in Insp Repts 50-369/90-09 & 50-370/90-09.Corrective actions:in-house Training Will Be Provided on Correct Method for Determining Proper Response to Problem Investigation Processes ML20055H4131990-07-16016 July 1990 Forwards Public Version of Epips,Including RP/0/A/5000/07 & HP/0/B/1009/04 1990-09-05
[Table view] |
Text
_ _-- . _ _ _ . _ _ _ _ - - _ _ _ - - _ _ _ . _ _ - _ - - _ _ - _ _ - _ _ - _ - - _
f.
.' Duxe POWER GOMPANY P.O. HOX 33189 '
CHARLOTrE, N.O. 28242 HALH. TUCKER TELEPHONE
' vms enesiossrr (704) 07H531 Mvos. nan anonvoru,n August 17, 1987 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555
Subject:
McGuire Nuclear Station Docket Numbers 50-369, -370 Catawba Nuclear Station Docket Numbers 50-413, -414 Reload Design Methodology Topical Report Revision Gentlemen:
Attached for your review is a proposed revision to DPC-NF-2010A, " Nuclear Physics Methodology for Reload Methodology," 'which was approved by the Staff in March of 1985. The revision is. similar to one (DPC-NE-1002A, " Reload Design Methodology II)' approved in October,1985 for the Oconee Nuclear Station, involving a change in computer codes used to generate cross-sections.
Pursuant to 10CFR 170.21, attached is a check in the amount of $150.00 as an application fee.
Very truly yours, Hal B. Tucker W SAG /79/jge Attachment xc: Dr. J. Nelson Grace, Regional Administrator U.S. Nuclear Regulatory Commission - Region II 101~darietta' Street, Suite 2900 Atlanta, Georgia 30s23 Mr. W.T. Orders Mr. P A. Van Doorn NRC Resident Inspector NRC Resident Inspector McGuire Nuclear Station Catawba Nuclear Station s
.Mr. Darl Hood, Project Manager Dr. K.N. Jabbour, Project Manager U.S. Nuclear Regulatory Commission U.S. Nuclear Regulatory Commission \
Office of Nuclear Reactor Regulation Office of Nuclear Reactor Regulation I Washington, D.C. 20555 Washington, D.C. 20555 k
P ft.O (NfC W C='U Y N 0' O
l
, Document Control Desk-
' August 17, 1987 Page 2 l
l bxc: R.H. Clark L.H.-Flores J.H. Randles D.M. Rowan N.A. Rutherford R.O. Sharpe R.L. Gill MC-801.02 CM-801.02 (9) i i
l l
I l
I i
i
,x 12.0. PHYSICS TEST COMPARISONS AND NUCLEAR RELIABILITY FACTORS FOR CASMO METHODOLOGY 12.1. INTRODUCTION AND
SUMMARY
12.1.1. INTRODUCTION In Section 3.0 CASMO-2E/ DUKE was referenced for generation of cross section's for strong absorbers, such as rod cluster control assemblies and burnable poisons, and the baffle and reflector.
EPRI-CELL was used to generate all other cross sections. All EPRI-NODE-P results presented in Section 12.0 are based exclusively on CASMO-2E/ DUKE. CASMO-2E/ DUKE was also used to generate all inputs to 1 PDQ. The discussions of CASMO-2E/ DUKE in this report are applicable to all future versionzfand enhancements of CASMO which might be used by Duke Power Company. Hence, the lattice code is referred to simply as CASMO throughout this document. The use of an all-CASMO setup of EPRI-NODE-P has already been approved in NFS-1002A (Reference 3) for the Oconee plant. EPRI-NGDE-P has buen benchmarked against Catawba 1 Cycle 1 and parts of Catawba 1 Cycle 2 and Catawba 2 Cycle 1.
This benchmark is presented here for approval of the CASMO setup of l l EPRI-NODE-P for reload design of McGuire and Catawba.
The above benchmark study encompassed: deriv, *. ion of measured power distributions for the above cycles, simulations of the above cycles using EPRI-NODE-P, reactivity comparisons at HZP and MFP of the above cycles, and development of a statistical basis for estimating the calculational accuracy of the CASMO setup of EPRI-NODE-P.
12.1.2.
SUMMARY
Section 12.2 presents comparisons of calculated and measured core physics data. Calculated and measured powers were statistically combined in Section 12.3 to derive Statistically Combined Uncertainty Factors (SCUF) for EPRI-NODE-P in the same way as in the SER for DPC-NF-2010A. SCUFs of 1.053 for the radial powers and 1.050 for the peak axial powers were determined. These can be compared to the values based on the EPRI-CELL models of 1.048 and 1.074, respectively.
All best estimate reactivity predictions agreed well with measurements when biases of +68 ppmb were applied to all boron calculations and +4 pcm/ degree F to all ITC calculations. These biases are based on hist.rical data such as core follow and zero power physics tests. These biases are reviewed each cycle to reflect experience with model performance.
12.2. CORE PHYSICS AND REACTIVITY PARAMETERS All calculations were performed with EPRI-NODE-P. The methodology used in all calculations is documented in Sections 10.0 and 11.0.
All measurement techniques are currently those used at the station.
l 12-1 l
l L __-_ _
)
12.2.1. HZP REACTIVITY COMPARISONS i
This section presents comparisons of EPRI-NODE-P calculations cersus measurements for some key core physics parameters at HZP. These physics parameters include HZP critical boron concentrations, HZP i rod worths by boration/ dilution and rod swap, and HZP isotherma'l temperature coefficients.
The HZP measurements were taken at beginning-of-cycle (BOC) during Zero Power Physics testing.
i Table 12-1 shows a comparison of BOC predicted versus measured critical boron con:entrations. The predicted boron concentrations include a bias of 68 ppmb. The mean difference was 4.11 ppmb with a standard deviation of 2.74.
I Table 12-2 shows a comparison of predicted versus measured rod )
worths. These rod worths were obtained by the boration/ dilution 4 method. The mean difference ie 6.89 pcm with a standard deviation l of 24.7. The mean percent error is 2.17% with a standard deviation l 1
of 3.82.
Table 12-3 shows a comparison of predicted versus measured rod worths. These rod worths were obtained by rod swap. The mean difference is -10.06 pcm with a standard deviation of 30.33. The mean percent error is -0.46% with a standard deviation of 5.76. The methodology used in rod swap is documented in Reference 28. The Catawba 2 Cycle 1 rod swap predictions were calculated at a biased j boron concentration of 909 ppmb. The Catawba 1 Cycle 2 rod swap predictions were calculated at the NODE calculated reference bank in boron concentration.
Table 12-4 shows a comparison of predicted and measured ITCs. The predicted ITCs include a bias of +4.0 pcm/ degree F. The mean difference is 0.63 pcm/ degree F with a standard deviation of 1.44.
12.2.2. DEPLETION POWER DISTRIBUTIONS AND REACTIVITY COMPARISONS This section presents comparisons of EPRI-NODE-P calculations to measurements for some key core physics parameters at HFP. These physics parameters include typical depletion power distribution comptrisona and HFP critical boron concentrations. j l
Figures 12-1 through 12-7 show typical power distribution comparisons for Catawba 1 Cycles 1 and 2 and Catawba 2 Cycle 1.
Table 12-5 shows a comparison of predicted and measured HFP boron concentrations. The predictes boron concentrations include a bias of 68 ppmb. The mean difference was -7.1 ppmb with a standard {
I deviation of 24.6.
12-2
12.3. STATISTICAL COMBINED UNCERTAINTY FACTORS SCUFs were calculated using the same methodology as used in Reference 3. Using Catawba Unit 1, Cycle 1 and part of Cycle 2, and Unit 2 part of Cycle 1, the SCUFs for the assembly radial and l assembly peak nodal powers were found to be 1.053 and 1.050, respectively. These factors were calculated using the formulas given in the Safety Evaluation Report of the "McGuire and Catawba Nuclear Physics Methodology for Reload Design," DPC-NF-2010A. The calculations of these factors are shown in Tables 12-6 through 12-
~
- 15. A local pin uncertainty of .022 uas used. This is the same uncertainty used and approved in DPC-NE-1002A (Reference 3a).
i Tables 12-6 through 12-8 show the. state points that were used for l the calculations. They also show the measured and calculated axial !'
offset for each state point.
Table 12-9 shows the results of the normal'.* test for assembly radial powers and assembly peak axial powers. The difference distribution of the assembly radial powers was nearly normal. The i difference distribution of the peak axial powers was normal.
Tables 12-10 and 12-11 show the difference means and standard deviations for assembly radial powers and assembly peak axial powers, respectively.
Tables 12-12 and 12-13 show the percent difference means and standard deviations for assembly radial powers and assembly peak axial powers.
Tables 12-14 and 12-15 show the calculation of SCUFs for assembly radial powers and assembly peak axial powers.
12.4. CONCLUSIONS A CASMO based EPRI-NODE-P model has been developed for Catawba that has shown good agreement with measured reactivity parameters.
Statistically combined uncertainty factors have also been developed that are consistent with the EPRI-CELL SCUFs. Therefore, based on these comparisons, the CASMO based EPRI-NODE-P model has been shown to be an acceptable methodology for reload design for McCuire and Catawba.
l 12-3 l
l 13.0. REFERENCES
- 1. Nuclear Associates International Corp., " Advanced Recycle Methodology Program System Documentation," CCM-3, (EPRI Confidential), September 1977.
- 2. Studsvik Energiteknik AB, "CASMO-2 A Fuel Assembly Burnup Program,"
Studsvik/NR-81/3, 1981. i
- 3. Duke Power Company, "Oconee Nuclear St'ation Reload Design Methodology,"
NFS-1001,-Rev. 4, June 1981.
3a. Duke Power Company, "Oconee Nuclear Station Reload Design Methodology II,"
DPC-NE-1002A, October 1985.
- 4. Bettis Atomic Power Laboratory, C. J. Pfeif fer, "PDQ-7 Feference Manual II," WAPD-TM-947(L), February 1971.
- 5. Rothleder, B. M., Fisher, J. R., "EPRI-NODE-P," EPRI-ARMP System j Documentation,'CCM-3, Part II, Chapter 14, September 1977.
- 6. Verbuk, P., Hoppe, N., "COMETHE-IIIJ A Computer for Predicting !
Mechanical and Thermal Behaviour of a Fuel Pin," BN 7609.1, Belgonucleaire S. A., March 1977.
- 7. TACO 2 -
Fuel Pin Performance Analysis, BAW-10141-PA, Rev. 1, (Proprietary), Babcock & Wilcox, Lynchburg, Virginia, June 1983.
- 8. Cobb, W. R., Eich, W. J., Tivel, D. E., "EPRI-CELL Code Description,"
EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 5, October 1978.
- 9. Edenius, M., Ekberg, K., Haggblom, H., "CASMO - THE DATA LIBRARY,"
Studsvik/K2-81/491, 1981.
- 10. Cobb, W. R., Tivel, D. E., "EPRI-CELL: GAM-THERMOS Library Descriptions," EPRI-ARMP System Documentation, CCM-3, Part II, Chapter i 2, April 1976.
- 11. Rothleder, B. M., Poetschat, G. R., "NUPUNCHER Code Description," EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 8, October 1975.
- 12. Duke Power Company, "MULTIFIT User Documentation," (Proprietary),
February 1983.
- 13. Hebert, M. J., et al., " PROGRAM C-HA-R-T CASMO to HARMONY I hieset ~
Conversion Processor," YAEC-1313P, May 1982.-
- 14. Rothleder, B. M., et al., "PWR Core Modeling Procedures for Advanced Recycle Methodology Program," RP-976-1, August 1979.
13-1
r j
D
- 15. Rothleder, B. M., Poetschat, C. R., "EPRI-FIT Code Description," EPRI-ARMP System Documentation, CCM-3,'Part II, Chapter 10,.0ctober 14, 1975.
- 16. . Rothleder,u B. M., Poetschat,. G. R., "SUPERLINK-P Code Description,"
EPRI-ARMP System Documentation, CCM-3, Part II, Chapter 12, October 22,- !
1975.
- 17. Smith, M. L., "PDQ7V2P7," (Proprietary), Virginia Electric and Power l Company, December 1977. ]
l 1 l
- 18. McGuire Nuclear' Station, Unit's 1 and 2, Final Safety Analysis Report, j l Docket Nos. 50-369, -370. .l
- l
- 19. Catawba Nuclear Station, Units.1 and 2, Final Safety Analysis Report,
Docket Nos. 50-413, ~414,
- 20. Letter, W. O. Parker to H. R. Denton, "Oconee Reload Design Me'hodology c Topical Report," Question 3, Docket Nos. 50-269, -270, -287, November 13, 1980.
.21. Duke Power Company, " Administrative Policy Manual for Nuclear Stations,"
Revision 21, August 1, 1983.
i
- 27. Duke Power Company, "PDQEDIT User Documentation," (Proprietary), March {
1982.
I
- 23. Morita, T., et al., " Topical Report Power Distribution Control and Load )
Following Procedures," WCAP-8385 (Proprietary), Westinghouse Electric I Corporation, September 1974.
J
'24. Duke Power Company, " Computer Code Users Manual for the NODE Utility 1 Code.'(NUC) - Margins," Revision 2, (Proprietary), September 24, 1982.
- 25. U. B. Owen, " Factors for One-Sided Tolerance Limits and for Variables Sampling Plans," SCR-607, Sandia Corporation Monograph, March 1963.
a
- 26. Shanstrom, R. T., et al., " CORE Codes for Operating Reactor Evaluation,"
SNA1617 (Proprietary), Shanstrom Nuclear Associates, April 1982.
- 27. American National Standards Institute, Inc., " Assessment of the Assumption of Normality (Employing Individual Observed Values)," ANSI N15.15-1974, 1974.
- 28. Duke Power Company, " Rod Swap Methodology Report for Startup Physics Testing," DPC-NE-1003A, December 4, 1986.
I 13-2 l
l Table 12-1 CRITICAL BORON CONCENTRATIONS AT HOT ZERO POWER, BOC Critical Baron Conc. PPM Unit / Cycle Calculated
- Measured Difference 975 975 I
-C1/C1 0 C1/C1 915 902 +13 C1/C1 820 810 +10 .
C1/C1 685 686 -1 C1/C1 500 508 -8 C1/C2 1402 1405 -3 C1/C2 1303 1273 +30 C2/C1 975 975 0 C2/C1 909 913 -4 Mean - -
4.11 Standard Deviation - -
2.74 Difference = Calculated - Measured
j
s Table 12-2 ROD WORTHS BY BORATION/ DILUTION Banks Predicted Measured Difference Unit / Cycle Inserted Rod Worths (PCM) Rod Worths (PCM) (PCM) % Error C1/C1 D 773 788 -15 -1.9
'C1/C1 0-C 1214 1203 11 0.9 C1/C1 D-B 1190 1171 19 1.6 C1/Cl D-A 572 548 24 4.4 C1/C1 0-A,SE 508 460 48 10.4 C1/C1 D-A,SE-SD 755 772 -17 -2.2 C1/C1 0-A,SE-SC 1098 1099 -1 -0.1 C2/C1 8 (Ref. Bank) 852 882 -30 -3.5 C1/C2 8 (Ref. Bank) 1041 1014 23 2.7 Mean - - -
6.89 2.17 Standard - - - 24.7 3.82 Deviation Total Worth 8003 7937 66 .8
% Error = Predicted - Measured Measured 1
)
mm_-a_ . - --m---- - - - - - - - - - - - - _ - - - - - - - - . . - _ , . . -
Table 12-3 ROD WORTHS BY R00 SWAP Predicted Inferred Unit / Cycle Bank Worth (PCH) Worth (PCM) Difference % Error 1
C2/C1 D 772 794 -22 -2.7 C2/C1 C 790 849 -59 -7.5 C2/C1 A 249 250 -1 -0.5 C2/C1 SE 377 385 -8 -2.0 C2/C1 SD 497 525 -28 -5.6 C2/C1 SC 497 522 -25 ~5.1 C2/C1 SB 765 834 -69 -9.0 C2/C1 SA 674 706 -32 -4.8 C1/C2 D 477 426 51 12.0 C1/C2 B 723 716 7 1.0 C1/C2 A 421- 420 1 0.2 C1/C2 SE 417 438 -21 -4.8 l C1/C2 320 4.9 SD 305 15 C1/C2 SC 322 307 15 4.9 C1/C2 Sa 772 781 -9 -1.2 C1/C2 SA 350 326 24 7.4 Mean - - -
-10.06 -0.46 Standard - - - 30.33 5.76 Deviation C2/C1 )
Total Worth -
5473 5747 -274 -4.8 i C1/C2 Total Worth - 4843 4737 106 2.2 ,
i h
l Table 12-4 ISOTHERMAL TEMPERATURE COEFFICIENTS Banks Predicted Measured Difference ,
Unit / Cycle Inserted (PCM/ F)* (PCM/'F)* (C-M) !
Cl/Cl ARO -1.38 -1.75 0.37 C1/Cl D -2.69 -2.77 0.08 C1/Cl D-C -9.02 -8.01 -1.01 C1/C2 ARO 4.63 1.71 2.92 C2/C1 ARO -1.04 -1.81 0.77 Mean - - -
0.63 Standard - - -
1.44 Deviation
- These values include a bias of 4.0 pcm/oF. l
FIGURE 12-1 CATAWBA 1 CYCLE 1 - CASH 0 NODE VS. HEASURED RADIAL POWERS 36 EFPD 100 % POWER CONTROL BANX D AT 198 SWD H G F E D C B A
- ,
- ;
- :;,aa;;;;;;;;;;;;;;;;;;;;;;;;, ::::::::::: ,,,;;;;;;;
- 0.00 * -2.68 * -1.60 * -2.80 * -1.89 * -2.17
- 0.00 * -1.80 * .85 * -2.04
- . ; : ;;;;;;;;;;, ;;;;;;;;;,;;;;;;;;;;;;;;;;;,,;;;;;;;;, ;,;;;;;;;;;,;,a
- ,;;,;, ::::::::::::,;;. : ;,;, za;;;;;;;;;; ;, ::, ::
- CASH 0-NODE 13 * .92 * .66
- 1.52 * % ERROR CASH 0-N0DE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00
$ ERROR CORE AVERAGE .02 CASH 0-N0DE MAXIMUM MAGNITUDE IS 1.24 AT ASSEMBLY F - 9 MEASURED MAXIMUM MAGNITUDE IS 1.25 AT ASSEMBLY E - 8 GERROR MAXIMUM MAGNITUDE IS 6.12 AT ASSEMBLY C - 12 PERCENT ERROR BETWEEN THE HAXIMUM VALUES IS .80 AVERAGE ABSOLUTE RELATIVE ERROR 1.59 PERCENT ROOT HEAN SQUARE OF THE RELATIVE ERROR 2.08 PERCENT ROOT HEAN SQUARE OF THE DIFFERENCE 2.06 PERCENT
FIGURE 12-2 CATAWBA 1 CYCLE 1 - CASH 0 N0DE VS. MEASURED RADIAL POWERS 156 EFPD 100 % POWER CONTROL BANK D AT 199 SWD 1
H G F E D C B A
-1.19 * )
f
)
- .98 * -1.47 * % ERROR CASMO-NODE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE .09 I
CASMO-NODE MAXIMUM MAGNITUDE IS 1.20 AT ASSEMBLY D - 12 MEASURED MAXIMUM MAGNITUDE IS 1.19 AT ASSEMBLY D - 12 l ERROR MAXIMUM MAGNITUDE IS -2.86 AT ASSEMBLY F - 8 )
PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS .84 AVERAGE ABSOLUTE RELATIVE ERROR 1.36 PERCENT ROOT MEAN SQUARE OF THE RELATIVE ERROR 1.54 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE lo61 PERCENT
FIGURE 12-3 CATAWBA 1 CYCLE 1 - CASH 0 NODE VS. MEASURED RADIAL POWERS 296 EFPD 100 % POWER . CONTROL BANK D AT 215 SWD H G F E D C B A
- A;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;.;;;;;;;;;;;;
- 0.00 * .88 * -2.00 * -2.61 * -2 08 * -2.33 *
- 0.00 * -1.02 * .87 * -1.00
- &*****;;;;;;;;;;;;;;;;;;;;;;;;;;;k*e
- -1.02 * .88 * -1.00 * .88 * -1.08 * -1.23 *
- 2.99 * % ERROR CASH 0-N0DE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE 11 CASH 0-N0DE MAXIMUM MAGNITUDE IS 1.19 AT ASSEMBLY B - 9 MEASURED MAXIMUM MAGNITUDE IS 1.17 AT ASSEMBLY D - 12 ERROR MAXIMUM MAGNITUDE IS 2.99 AT ASSEMBLY B - 13 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS 1.71 AVERAGE ABSOLUTE RELATIVE ERROR 1.10 PERCENT ROOT HEAN SQUARE OF THE REIATIVE ERROR 1.40 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 1.38 PERCENT l
FIGURE 12-4 t
CATAWBA 1 CYCLE 2 - CASMO NODE VS. HEASURED RADIAL POWERS 51 EFPD 100 % POWER CONTROL BANK D AT 222 SWD H G F E D C B A
8-* .83
- 1.74 * -1.03 * .89 * -2.78 * -1.04 *
- 0.00 * -1.04 * -1.85 * -1.09 *
- .93 * -7.32 * % ERROR CASH 0-N0DE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE .07 CASH 0-NODE MAXIMUM MAGNITUDE IS 1.29 AT ASSEMBLY E - 10 MEASURED MAXIMUM MAGNITUDE IS 1.25 AT ASSEMBLY E - 10 ERROR MAXIMUM MAGNITUDE IS -7.32 AT ASSEMBLY B - 13 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS 3.20 AVERAGE ABSOLUTE RELATIVE ERROR 1.67 PERCENT ROOT MEAN SQUARE OF THE RELATIVE ERROR 2.37 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 1.88 PERCENT
7 - --- -__ l l
f FIGURE 12-5 L
e I
l l ' CATAWBA 1 CYCLE 2 - CASH 0 NODE VS. MEASURED RADIAL POWERS f 157 EFPD 100 % POWER CONTROL BANK D AT 212 SWD G F E D C B A H
1
- ,
- A;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;;
-;%;;. A;;;;;;;;;;;;;;;;;;;;;;;;;
- e CASH 0-NODE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE .03 CASH 0-NODE MAXIMUM MAGNITUDE IS 1.30 AT ASSEMBLY E - 10 MEASURED MAXIMUM MAGNITUDE IS 1.30 AT ASSEMBLY E - 10 ERROR MAXIMUM MAGNITUDE IS -4.65 AT ASSEMBLY B - 13 PERCENT ERROR BETWEEN THE HAXIMUM VALUES IS 0.00 AVERAGE ABSOLUTE RELATIVE ERROR 1.06 PERCENT ROOT MEAN SQUARE OF THE RELATIVE ERROR 1.56 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 1.25 PERCENT
f-e FIGURE 12-6 CATAWBA 2 CYCLE 1 - CASMO NODE VS. MEASURED RADIAL POWERS l
45 EFPD 100 % POWER CONTROL BANK D AT 210 SWD H G F E D C B A' l
1 m; ;. ., ; ; ;, ;. ;, ;, ;. ., ; ; ;, ;, ;, ;. - - ;, ;, ;, ;, ;, ;, ;, ;. :.-;, ;. :. ;. ;, ;, ;, ;, ;. ;. ;, ;. :. ;. :. - - - :. - - - - ;. - - -
- e l '* -1.92 * .87 '* -2.70 * .81 * -2.83 * .95 * -4.30 * -1.18 *
- 0.00 * -1.82 * .85 * -2.04
- 10.
- s-;,;, ;, ;. ;, ; ;, ;, ;. ;, ;, ;, ;. ;, ; ;, ;. ;, ;, ;, ;. ;, ;, ;, - - ;. ;. :. ;, ;, ;, ;. - - - - - ;. ;. ;, ; ;, ;. ;, - - - - ;. ;. ;, ;, ;.
- CASM0-NODE 13 * .93 * .65
- 3.08 * % ERROR CASH 0-NODE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE .05 CASMO-NODE MAXIMUlf MAGNITUDE IS 1.25 AT ASSEMBLY D - 12 MEASURED MAXIMUM MAGNITUDE IS 1.24 AT ASSEMBLY F - 9 ERROR MAXIMUM MAGNITUDE IS 5.00 AT ASSEMBLY C - 12 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS .81 AVERAGE ABSOLUTE RELATIVE ERROR 1.75 PERCENT ROOT MEAN SQUARE OF THE RELATIVE ERROR 2.18 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 2.07 PERCENT
r- -l I
FIGURE 12-7 1
1 CATAWBA 2 CYCLE 1 - CASH 0 N0DE VS. MEASURED RADIAL POWERS 177 EFPD 100 % POWEk CONTROL BANK D AT 215 SWD H G F E D C B A
- -1.01 * .88 * -2.91 * .85 * -1.96 * .90 * -1.08
- 0.00 * -1.94 * .85 * -2.00
- . ::::::: ::::::;;;;;;;;;;;;;4*****
- -1.94 * -1.68 * -2.91 * -1.80
- -2.86 * .85 * -1.01 * .98
- CASMD-NODE 13 * .99 * .65
- 4.62 * % ERROR CASMO-N0DE CORE AVERAGE 1.00 MEASURED CORE AVERAGE 1.00 ERROR CORE AVERAGE .05 CASH 0-N0DE MAXIMUM MAGNITUDE IS 1.22 AT ASSEMBLY D - 12 MEASURED MAXIMUM MAGNITUDE IS 1.21 AT ASSEMBLY D - 12 ERROR MAXIMUM MAGNITUDE IS 4.62 AT ASSEMBLY B - 13 PERCENT ERROR BETWEEN THE MAXIMUM VALUES IS .83 AVEEAGE ABSOLUTE RELATIVE ERROR 1.80 PERCENT ROOT MEAN SQUARE OF THE RELATIVE ERROR 2.25 PERCENT ROOT MEAN SQUARE OF THE DIFFERENCE 2.16 PERCENT j
_ _ _ _ _ _ _ _ - - - - _ - - _ - - - - - - -- )
3; w l
,if6 p LTable-12-5 l .t>
[ [' . DEPLETION CRITICAL BORON CONCENTRATIONS c:
' Unit / Cycle; EFPD' Predicted (PPM)* Difference Measured (PPM)
C1/Cl -36 627 63) '
'C1/C1 52. 614 621 -7 1 ~C1/C1 80 590 Sfi5 5
.C1/C1 110 529 57.7 12
.C1/C1 156 444 - /,37 - 7 C1/C1' '180; 395 383 12 e C1/Cl .220 331 if 3 25 p ,
C1/C1' 266' 193 s!n 13
- . -C1/C1 296 118 120 -2
[. ,
'C1/C1 ~310 117 72 45 C1/C2 27 840 876 -36
'C1/C2 5 11 761 809 -48 C1/C2: 77 671 712 -41 C1/C2 97 606 653 -47 L C1/C2- 126' 514 552 -38 C1/C2 . 157 415 445 -30 C2/Cl -45 621 626 -5
.C2/C1' 88 603 604 -1 C2/Cl 1100 551 575 -24 C2/C1 128 501 506 ,
C2/C1' '149 491 487 4 C2/C1; 177 402 394 8 y
Mean - - -
-7.1 Standard - - -
24.6 Deviation
- These values include a bias of 68 ppm.
l-1 l
l 4
h-_____ ______________mm _
Table 12-6 CATAWBA UNIT 1 CYCLE 1 STATE POINTS Control Bank D Axial Offset Point # EFPD Power (%) Position (Steps) (Meas /Cale)(%)
1 36 100 200 -11.22/-10.51 2 52 100 200 -7.11/-8.19 3 80 9A 195 -3.90/-8.27 4 110 100 193 -6.92/-7.52 5 156 100 199 -4.34/-4.95 6 180 100 215 -1.24/-2.47 7 220 100 223 -1.04/-0.62 8 266 100 217 -2.23/-0.18 9 296 100 215 -2.16/-0.64 10 310 84 216 4.27/4.50
Table 12-7 CATAWBA UNIT 1 CYCLE 2 STATE POINTS
)
Control Bank D . Axial Offset Point # EFPD Power (%) Position (Steps) (Meas / Calc)(%)
1 27 100 214 4.26/1.40 2 51 100 222 1.48/ .98 3 77 100 209 -1.46/-4.06 4 97 100 211 -2.57/-4.23 5 126 100 211 -3.29/-4.77 6 157 100 212 -4.15/-4.76 i
Table 12-8 CATAWBA UNIT 2 CYCLE 1 STATE POINTS Control Bank D Axial Offset Point # EFPD Power (%) Position ~(Steps) (Meas / Calc)(%)
1 45 100 210 -7.39/-9.83 2 77 88 199 -4.78/-7.79 3 100 100 209 -4.51/-7.30 4 128 100 210 -4.72/-5.91 5 149 80 187 -3.89/-5.27 6 177 100 215 -3.66/-3.74
'4 Table 12-9 DIFFERENCE DISTRIB IION NORMALITY TESTS FOR C, M i 1.0 - 5% LEVEL OF SIGNIFICANCE ASSEMBLY RACIAL POWERS q Unit / Cycle N D'(P=.025) D' D'(P=.975) Remarks C1/Cl 168 602.8 595.2 622.3 Nearly Normal C1/C2 109 313.7 310.9 326.4' Nearly Normal C2/Cl 112 326.8 313.0 339.8 Nearly Normal All Combined 389 2138.7 2120.3 2183.6 Nearly Normal ASSEMBLY PEAK AXIAL POWERS Unit / Cycle N D'(P=.025) D' D'(P=.975) Remarks C1/C1 272 1246.8 1259.0 1279.0 Normal C1/C2 162 570.1 582.9 588.9 Normal C2/Cl 166 591.9 594.9 611.2 Normal All Combined 600 4106.0 4132.9 4176.0 Normal I r
l 1
1
I l
Table 12-10 DIFFERENCE MEANS AND STANDARD DEVIATIONS i FOR ASSEMBLY RADIAL POWERS (C, M 2. 1.0) !
l l
Unit / Cycle N D S(D) ABS (D) S(ABS (D))
C1/Cl 168 .004 .020 .021 .013
.C1/C2 109 .004 .017 .013 .012 J C2/Cl 112 .008 .024 .017 .011 ;
All Combined 389 .003 .021 .017 .012 !
l i
i m
- - _ _ - - - _ _ . - - - _ _ - - _ . _ _ . _ .___2
4 Table 12-11 DIFFERENCE MEANS AND STANDARD DEVIATIONS FOR ASSEMBLY PEAK AXIAL POWERS (C, M A'1.0) ycle N E S(D) ABS (D) S(ABS (D))
C1/C1 272 .025 .032 .031 .025 C1/C2 162 .029 .036 .036 .029 C2/C1 166 .017 .038 .031 '.028 All Combined 600 .024 .035 .032 .027 tl
m
.. i j
' l i
Table 12 PERCENT DIFFERENCE MEANS (C, M 2.1.0)' - ASSEMBLY RADIAL POWERS Unit / Cycle Mean % Difference Mean Absolute % Difference C1/C1 .334 1.54 C1/C2 .303 '1.15 C2/C1 .696 1.95 All Combined .260 1.55 l
I Table 12-13 )
I PERCENT DIFFERENCE MEANS j (C, M 1 1.0) - ASSEMBLY PEAK AXIAL POWERS Unit / Cycle Mean % Difference Mean Absolute % Difference C1/Cl 1.99 2.47 C1/C2 2.35 2.91 C2/Cl 1.32 2.40 All Combined 1.90 2.57
s l*
c-Table 12-14 CASMO-N0DE ASSEMBLY RADIAL POWER FN t2 fI 1.111 5 -0.003 S(D) 0.021 N 389
= 1.0 + (D/R) + [(.03) + (K
M FN = 1.0 + (.003/1.111) + [(.03)2 + (.034) + (.022)2)
FN LH
= 1.053
- Since 5 is negative 6 5 term is included because the model consistently underpredicts measured powers.
s p.=
Table 12-15 CASMO-NODE ASSEMBLY PEAK AXIAL POWER F R 1.274 5 0.024 S(D) 0.035 N 600
- = 1.0 + (5/R) + [(.03) + (K
F _
M F = 1.0 + [(.03)2 + (0.033)2 + (.022)2)h F = 1.050 0
- Since 6 is positive the D/T4 term is not included because the model consistently overpredicts measured powers.
j i
l
,