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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217P3791999-10-21021 October 1999 Forwards NRC Form 396 & NRC Form 398 for Renewal of Licenses SOP-20607-1 & SOP-20610-1.Without Encls ML20217N2521999-10-20020 October 1999 Provides Supplemental Info Re 990405 Containment Insp Program Requests for Relief RR-L-1 & RR-L-2,in Response to 991013 Telcon with NRC ML20217K7541999-10-15015 October 1999 Forwards Rev 1 to Unit 1,Cycle 9 & Unit 2 Cycle 7 Colrs,Iaw Requirements of TS 5.6.5.Figure 5, Axial Flux Difference Limits as Function of Percent of Rated Thermal Power for RAOC, Was Revised for Both Units ML20217G6751999-10-13013 October 1999 Requests Withholding of Proprietary Info Contained in Application for Amend to OLs to Implement Relaxations Allowed by WCAP-14333-P-A,rev 1 ML20217G1071999-10-0707 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Vogtle & Did Not Identify Any Areas in Which Performance Warranted More than Core Insp Program.Nrc Plans to Conduct Core Insps at Facility Over Next Six Months ML20216J9041999-10-0101 October 1999 Forwards Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20216J9161999-10-0101 October 1999 Forwards Response to NRC 990723 RAI Re GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20217B0141999-10-0101 October 1999 Forwards Insp Repts 50-424/99-06 & 50-425/99-06 on 990725- 0904 at Vogtle Units 1 & 2 Reactor Facilities.Determined That One Violation Occurred & Being Treated as non-cited Violation ML20212E8751999-09-20020 September 1999 Forwards Response to NRC GL 99-02, Lab Testing of Nuclear Grade Activated Charcoal. Description of Methods Used to Comply with Std Along with Most Recent Test Results Encl ML20212E7481999-09-20020 September 1999 Requests Approval Per 10CFR50.55a to Use Alternative Method for Determining Qualified Life of Certain BOP Diaphragm Valves than That Specified in Code Case N-31.Proposed Alternative,Encl ML20212C2191999-09-16016 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, Which Is Current Need for NRC Operator Licensing Exams for Years 2000 Through 2003 of Plant Vogtle,Per Administrative Ltr 99-03 ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl ML20211J5291999-08-30030 August 1999 Forwards Snoc Copyright Notice Dtd 990825,re Production of Engineering Drawings Ref in VEGP UFSAR ML20211J5251999-08-30030 August 1999 Forwards Response to NRC 990727 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20211J7381999-08-27027 August 1999 Informs That Licensee Vessel Data Is Different than NRC Database Based on Listed Info,Per 990722 Request to Review Rvid ML20211E9251999-08-23023 August 1999 Forwards fitness-for-duty Performance Data for Jan-June 1999,as Required by 10CFR26.71(d).Data Reflected in Rept Covers Employees at Vogtle Electric Generating Plant ML20210V0881999-08-16016 August 1999 Forwards Insp Repts 50-424/99-05 & 50-425/99-05 on 990620- 0724.No Violations Noted.Vogtle Facility Generally Characterized by safety-conscious Operations,Sound Engineering & Maintenance Practices ML20210Q4611999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006 for Vogtle.Requests Info Re Individuals Who Will Take Exam. Sample Registration Ltr Encl ML20210L2181999-08-0202 August 1999 Forwards NRC Form 396 & Form 398 for Renewal of Listed Licenses,Iaw 10CFR55.57.Without Encl ML20210N1191999-08-0202 August 1999 Discusses 990727 Telcon Between Rs Baldwin & R Brown Re Administration of Licensing Exam at Facility During Wk of 991213 ML20210G3351999-07-27027 July 1999 Forwards Second Request for Addl Info Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20210E0121999-07-23023 July 1999 Forwards Second Request for Addl Info Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20210D9341999-07-22022 July 1999 Discusses Closure of TACs MA0581 & MA0582,response to Requests for Info in GL 92-01,rev 1,suppl 1, Reactor Vessel Structural Integrity ML20210C8011999-07-21021 July 1999 Provides Response to NRC AL 99-02,which Requests That Addressees Submit Info Pertaining to Estimates of Number of Licensing Actions That Will Be Submitted for NRC Review for Upcoming Fy 2000 & 2001 ML20210E0431999-07-15015 July 1999 Forwards Insp Repts 50-424/99-04 & 50-425/99-04 on 990502- 0619.Two Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20209H3881999-07-14014 July 1999 Forwards Revs 1 & 2 to ISI Program Second 10-Year Interval Vogtle Electric Generating Plant Unit 1 & 2 ML20209C4041999-07-0101 July 1999 Forwards Rev 29 to VEGP Units 1 & 2 Emergency Plan.Rev 29 Incorporates Design Change Associated with Consolidation of Er Facilities Computer & Protues Computer.Justifications for Changes & Insertion Instructions Are Encl ML20196H8081999-06-28028 June 1999 Discusses 990528 Meeting Re Results of Periodic PPR for Period of Feb 1997 to Jan 1999.List of Attendees Encl ML20212J2521999-06-21021 June 1999 Responds to NRC RAI Re Yr 2000 Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701 ML20196F9171999-06-21021 June 1999 Forwards Owner Rept for ISI for Vogtle Electric Generating Plant,Unit 1 Eighth Maint/Refueling Outage. Separate Submittal Will Not Be Made to NRC on SG Tubes Inspected During Subj Outage ML20195F8031999-06-11011 June 1999 Forwards Changes to VEGP Unit 1 Emergency Response Data Sys (ERDS) Data Point Library.Changes Were Completed on 990308 While Unit 1 Was SD for Refueling Outage ML20207E7421999-06-0303 June 1999 Refers to from NRC Which Issued Personnel Assignment Ltr to Inform of Lm Padovan Assignment as Project Manager for Farley Npp.Reissues Ltr with Effective Date Corrected to 990525 ML20207F6201999-06-0202 June 1999 Sixth Partial Response to FOIA Request for Documents.Records in App J Encl & Will Be Available in Pdr.App K Records Withheld in Part (Ref FOIA Exemptions 7) & App L Records Completely Withheld (Ref FOIA Exemption 7) ML20207D9861999-05-28028 May 1999 Informs That,Effective 990325,LM Padovan Was Assigned as Project Manager for Plant,Units 1 & 2 ML20207D2701999-05-19019 May 1999 Forwards Insp Repts 50-424/99-03 & 50-425/99-03 on 990321- 0501.One Violation of NRC Requirements Identified & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML20206M5141999-05-11011 May 1999 Informs That NRC Ofc of Nuclear Regulation Reorganized Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Created.Rl Emch Section Chief for Vogtle. Reorganization Chart Encl ML20206U4061999-05-11011 May 1999 Confirms Telcon with J Bailey Re Mgt Meeting Scheduled for 990528 to Discuss Results of Periodic Plant Performance Review for Plan Nuclear Facility Fo Period of Feb 1997 - Jan 1999 05000424/LER-1998-006, Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Exi1999-05-10010 May 1999 Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Existed ML20206D6411999-04-29029 April 1999 Forwards Vogtle Electric Generating Plant Radiological Environ Operating Rept for 1998 & Vogtle Electric Generating Plant Units 1 & 2 1998 Annual Rept Annual Radioactive Effluent Release Rept ML20206D5881999-04-29029 April 1999 Forwards Rept Which Summarizes Effects of Changes & Errors in ECCS Evaluation Models on PCT for 1998,per Requirements of 10CFR50.46(a)(3)(ii).Rept Results Will Be Incorporated Into Next FSAR Update ML20206D6951999-04-28028 April 1999 Provides Update of Plans for VEGP MOV Periodic Verification Program Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20206C2241999-04-21021 April 1999 Forwards Revised Monthly Operating Repts for Mar 1999 for Vogtle Electric Generating Plant,Units 1 & 2.Page E2-2 Was Iandvertently Omitted from Previously Submitted Rept on 990413 ML20206A6371999-04-21021 April 1999 Forwards SE Authorizing Licensee Re Rev 9 to First 10-yr ISI Interval Program Plan & Associated Requests for Relief (RR) 65 from ASME Boiler & Pressure Vessel Code ML20205Q3351999-04-15015 April 1999 Forwards Insp Repts 50-424/99-02 & 50-425/99-02 on 990214-0320.Three Violations Identified & Being Treated as Non-Cited Violations ML20205T2351999-04-0909 April 1999 Informs That on 990317,B Brown & Ho Christensen Confirmed Initial Operator Licensing Exam Scheduled for Y2K.Initial Exam Date Scheduled for Wk of 991213 for Approx 10 Candidates ML20205K7501999-04-0505 April 1999 Informs That Effective 990329,NRC Project Mgt Responsibility for Plant Has Been Transferred from Dh Jaffe to R Assa ML20209A3741999-04-0505 April 1999 Submits Several Requests for Relief for Plant from Code Requirements Pursuant to 10CFR50.55a(a)(3) & (g)(5)(iii).NRC Is Respectfully Requested to Approve Requests Prior to Jan 1,2000 ML20205H3481999-03-31031 March 1999 Forwards Georgia Power Co,Oglethorpe Power Corp,Municipal Electric Authority of Ga & City of Dalton,Ga Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81 ML20205F9091999-03-29029 March 1999 Submits Rept of Number of SG Tubes Plugged During Plant Eighth Maintenance/Refueling Outage (1R8).Inservice Insps Were Completed on SGs 1 & 4 on 990315.No Tubes Were Plugged ML20205G0761999-03-26026 March 1999 Provides Results of Individual Monitoring for 1998.Encl Media Contains All Info Required by Form NRC 5.Without Encl 1999-09-20
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217P3791999-10-21021 October 1999 Forwards NRC Form 396 & NRC Form 398 for Renewal of Licenses SOP-20607-1 & SOP-20610-1.Without Encls ML20217N2521999-10-20020 October 1999 Provides Supplemental Info Re 990405 Containment Insp Program Requests for Relief RR-L-1 & RR-L-2,in Response to 991013 Telcon with NRC ML20217K7541999-10-15015 October 1999 Forwards Rev 1 to Unit 1,Cycle 9 & Unit 2 Cycle 7 Colrs,Iaw Requirements of TS 5.6.5.Figure 5, Axial Flux Difference Limits as Function of Percent of Rated Thermal Power for RAOC, Was Revised for Both Units ML20217G6751999-10-13013 October 1999 Requests Withholding of Proprietary Info Contained in Application for Amend to OLs to Implement Relaxations Allowed by WCAP-14333-P-A,rev 1 ML20216J9161999-10-0101 October 1999 Forwards Response to NRC 990723 RAI Re GL 95-07, Pressure Locking & Thermal Binding of SR Power-Operated Gate Valves ML20216J9041999-10-0101 October 1999 Forwards Response to RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves ML20212E7481999-09-20020 September 1999 Requests Approval Per 10CFR50.55a to Use Alternative Method for Determining Qualified Life of Certain BOP Diaphragm Valves than That Specified in Code Case N-31.Proposed Alternative,Encl ML20212E8751999-09-20020 September 1999 Forwards Response to NRC GL 99-02, Lab Testing of Nuclear Grade Activated Charcoal. Description of Methods Used to Comply with Std Along with Most Recent Test Results Encl ML20212C2191999-09-16016 September 1999 Forwards NRC Form 536, Operator Licensing Exam Data, Which Is Current Need for NRC Operator Licensing Exams for Years 2000 Through 2003 of Plant Vogtle,Per Administrative Ltr 99-03 ML20211J5291999-08-30030 August 1999 Forwards Snoc Copyright Notice Dtd 990825,re Production of Engineering Drawings Ref in VEGP UFSAR ML20211J5251999-08-30030 August 1999 Forwards Response to NRC 990727 RAI Re GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design Basis Accident Conditions ML20211J7381999-08-27027 August 1999 Informs That Licensee Vessel Data Is Different than NRC Database Based on Listed Info,Per 990722 Request to Review Rvid ML20211E9251999-08-23023 August 1999 Forwards fitness-for-duty Performance Data for Jan-June 1999,as Required by 10CFR26.71(d).Data Reflected in Rept Covers Employees at Vogtle Electric Generating Plant ML20210L2181999-08-0202 August 1999 Forwards NRC Form 396 & Form 398 for Renewal of Listed Licenses,Iaw 10CFR55.57.Without Encl ML20210C8011999-07-21021 July 1999 Provides Response to NRC AL 99-02,which Requests That Addressees Submit Info Pertaining to Estimates of Number of Licensing Actions That Will Be Submitted for NRC Review for Upcoming Fy 2000 & 2001 ML20209H3881999-07-14014 July 1999 Forwards Revs 1 & 2 to ISI Program Second 10-Year Interval Vogtle Electric Generating Plant Unit 1 & 2 ML20209C4041999-07-0101 July 1999 Forwards Rev 29 to VEGP Units 1 & 2 Emergency Plan.Rev 29 Incorporates Design Change Associated with Consolidation of Er Facilities Computer & Protues Computer.Justifications for Changes & Insertion Instructions Are Encl ML20196F9171999-06-21021 June 1999 Forwards Owner Rept for ISI for Vogtle Electric Generating Plant,Unit 1 Eighth Maint/Refueling Outage. Separate Submittal Will Not Be Made to NRC on SG Tubes Inspected During Subj Outage ML20212J2521999-06-21021 June 1999 Responds to NRC RAI Re Yr 2000 Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701 ML20195F8031999-06-11011 June 1999 Forwards Changes to VEGP Unit 1 Emergency Response Data Sys (ERDS) Data Point Library.Changes Were Completed on 990308 While Unit 1 Was SD for Refueling Outage 05000424/LER-1998-006, Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Exi1999-05-10010 May 1999 Forwards LER 98-006-03 Re Motor Control Ctr Breaker Buckets Not Being Seismically Qualified.Rev Is Submitted to Document Results of Seismic Testing That Demonstrated That No Condition Outside Design Basis of TS Requirements Existed ML20206D5881999-04-29029 April 1999 Forwards Rept Which Summarizes Effects of Changes & Errors in ECCS Evaluation Models on PCT for 1998,per Requirements of 10CFR50.46(a)(3)(ii).Rept Results Will Be Incorporated Into Next FSAR Update ML20206D6411999-04-29029 April 1999 Forwards Vogtle Electric Generating Plant Radiological Environ Operating Rept for 1998 & Vogtle Electric Generating Plant Units 1 & 2 1998 Annual Rept Annual Radioactive Effluent Release Rept ML20206D6951999-04-28028 April 1999 Provides Update of Plans for VEGP MOV Periodic Verification Program Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs ML20206C2241999-04-21021 April 1999 Forwards Revised Monthly Operating Repts for Mar 1999 for Vogtle Electric Generating Plant,Units 1 & 2.Page E2-2 Was Iandvertently Omitted from Previously Submitted Rept on 990413 ML20209A3741999-04-0505 April 1999 Submits Several Requests for Relief for Plant from Code Requirements Pursuant to 10CFR50.55a(a)(3) & (g)(5)(iii).NRC Is Respectfully Requested to Approve Requests Prior to Jan 1,2000 ML20205H3481999-03-31031 March 1999 Forwards Georgia Power Co,Oglethorpe Power Corp,Municipal Electric Authority of Ga & City of Dalton,Ga Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81 ML20205F9091999-03-29029 March 1999 Submits Rept of Number of SG Tubes Plugged During Plant Eighth Maintenance/Refueling Outage (1R8).Inservice Insps Were Completed on SGs 1 & 4 on 990315.No Tubes Were Plugged ML20205G0761999-03-26026 March 1999 Provides Results of Individual Monitoring for 1998.Encl Media Contains All Info Required by Form NRC 5.Without Encl ML20205H4051999-03-25025 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1) ML20205H3891999-03-25025 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1).Page 2 in Third Amend Power Sales Contract of Incoming Submittal Not Included ML20205A9441999-03-25025 March 1999 Forwards VEGP Unit 1 Cycle 9 Colr,Per TS 5.6.5.d ML20205H3811999-03-24024 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81,as Requested IAW 10CFR50.75(f)(1) ML20205H3621999-03-22022 March 1999 Forwards Info on Status of Decommissioning Funding for Each Reactor or Part of Reactor Owned for OLs NPF-68 & NPF-81, as Requested IAW 10CFR50.75(f)(1) ML20204G4361999-03-18018 March 1999 Forwards Summary Rept of Present Level & Source of on-site Property Damage Insurance Coverage for Vegp,Iaw Requirements of 10CFR50.54(w)(3) ML20204C0591999-03-17017 March 1999 Forwards Rev 0 to WCAP-15160, Evaluation of Pressurized Thermal Shock for Vegp,Unit 2 & Rev 0 to WCAP-15159, Analysis of Capsule X from Vegp,Unit 2 Reactor Vessel Radiation Surveillance Program ML20207K9551999-03-11011 March 1999 Forwards Response to Rai,Pertaining to Positive Alcohol Test of Licensed Operator.Encl Info Provided for NRC Use in Evaluation of Fitness for Duty Occurrence.Encl Withheld,Per 10CFR2.790(a)(6) ML20207L9721999-03-10010 March 1999 Forwards Rev 15 to EPIP 91104-C of Manual Set 6 of Vogtle Epips.Without Encl ML20207B0191999-02-25025 February 1999 Forwards Fitness for Duty Performance Data for six-month Reporting Period 980701-1231,IAW 10CFR26.71(d) 05000424/LER-1998-009, Forwards LER 98-009-00 Re Event in Which Improper Testing Method Resulted in Inadequate Surveillances on 9812291999-01-27027 January 1999 Forwards LER 98-009-00 Re Event in Which Improper Testing Method Resulted in Inadequate Surveillances on 981229 ML20199F7701999-01-13013 January 1999 Submits Revised Response to RAI Re Licensee 980713 Proposed Amend to Ts,Eliminating Periodic Response Time Testing Requirements on Selected Sensors & Protection Channels. Corrected Copy of Table,Encl ML20199F7981999-01-13013 January 1999 Forwards Corrected Pages to VEGP-2 ISI Summary Rept for Spring 1998 Maint/Refueling Outage. Change Bar in Margin of Affected Pages Denotes Changes to Rept ML20199G1381999-01-13013 January 1999 Forwards Copy of Permit Renewal Application Package for NPDES Permit Number GA0026786,per Section 3.2 of VP Environ Protection Plan 05000424/LER-1998-007, Forwards LER 98-007-00,re Inadequate Surveillances Due to Improperly Performed Response Time Testing,On 981215,IAW 10CFR50.731999-01-13013 January 1999 Forwards LER 98-007-00,re Inadequate Surveillances Due to Improperly Performed Response Time Testing,On 981215,IAW 10CFR50.73 ML20198F6131998-12-18018 December 1998 Forwards Revised Certification of Medical Exam Form for License SOP-21147.Licensee Being Treated for Hypertension. Util Requests That Individual License Be Amended to Reflect Change in Status ML20198L6631998-12-18018 December 1998 Forwards Amend 37 to Physical Security & Contingency Plan. Encl 1 Provides Description & Justification for Changes & Encl 2 Contains Actual Amend 37 Pages.Amend Withheld,Per 10CFR73.21 ML20198D9291998-12-16016 December 1998 Forwards Requested Info Re Request to Revise TSs Elimination of Periodic Pressure Sensor Response Time Tests & Elimination of Periodic Protection Channel Response Time Tests ML20198D9991998-12-16016 December 1998 Forwards Responses to 980916 RAI Re Response to GL 97-01, Degradation of Control Rod Drive Mechanism Nozzle & Other Vessel Closure Head Penetrations ML20198D8171998-12-14014 December 1998 Forwards NRC Form 396 & Form 398 for Renewal of License OP-20993.Without Encls ML20206N3051998-12-0808 December 1998 Submits RAI Re Replacement of Nuclear Instrument Sys Source & Intermediate Range Channels & post-accident Neutron Flux Monitoring Sys 1999-09-20
[Table view] |
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- Ger:ga Power Company 40 hverness Cemer Parkway Pod O' ice Got 1295 6;rTrogTiam aim >di!a 35201 IER#cre 205 877-7122 '
m
- c. t uccoy Georgia Power r t.uuaem November 5, 1993 LCV-0199 ,
n Docket Nos. 50-424 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen:
V0GTLE ELECTRIC GENERATING PLANT 10 CFR 50.46 ECCS MODEL SIGNIFICANT CHANGE REPORT Attached is Georgia Power Company's 10 CFR 50.46 emergency core cooling system (ECCS) model significant change report based on WCAP-13451 and in compliance with the reporting requirements of 10 CFR 50.46(a)(3)(ii). It is based on information provided by Westinghouse on October 6, 1993, concerning errors and changes assessed against the Vogtle ECCS models.
The assessed sum of the absolute magnitude of the errors / changes is greater than 500F for the Vogtle small break loss of coolant accident (SBLOCA);
however, the net effect of the errors / changes is a reduction in the SBLOCA peak cladding temperature (PCT) for Vogtle. The SBLOCA PCT results are affected by three issues: safety injection in the broken loop, the. improved condensation model, and the drift flux flow regime. The large break LOCA PCT results are not affected by these issues. The attached report also provides a summary of the plant change evaluations performed under the provisions of 10 CFR 50.59 that were previously assessed. The effects on the LOCA results will be incorporated in a future FSAR update.
Because the NOTRUMP small break ECCS model errors / changes affect many plants, the Westinghouse Owners Group is reviewing this issue and possible :
development of a generic program for resolution. A schedule for resolution and reanalysis will be provided following the completion of the Westinghouse Owners Group review.
If you have any questions regarding this report, please contact this office.
Sincerely, C. K. McCoy h8 cc: (See next page}10003 4 'f i i
9311120262 9f'1105 "i PDR ADOCK 35000424 R PDR
Geoigia Powerk~ n .
J. S. Nuclear Regulatory Commission Page'2 W
CKM/BCA/HWM: gps Attachment '
cc: Georaia Power Company '-
Mr. J. B. Beasley, Jr.
Mr. M. Sheibani NORMS U.S. Nuclear Reaulatory Commission Mr. S. D. Ebneter, Regional Administrator ,
Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle _
LCV-0199 Mcn1 i
ATTACHMENT V0GTLE ELECTRIC GENERATING PLANT 10 CFR 50.46 ECCS MODEL SIGNIFICANT CHANGE REPORT ,
BACKGROUND Provisions in 10 CFR 50.46 require applicants and holders of operating licenses or construction permits to notify the Nuclear Regulatory Commission (NRC) of errors and changes in the emergency core cooling system (ECCS) evaluation models on an annual basis when the errors and changes are ;
not significant, and within 30 days of discovery when the errors and
~
changes are significant. Reference 1 defines a significant error or change as one which results ;n a calculated fuel peak cladding temperature (PCT) different by more than 500F from the temperature calculated for the limiting transient using the last acceptable model, or as a cumulation of changes and errors such that the sum of the absolute magnitudes of the respective temperature changes is greater than 500F.
The following presents an assessment of the effects of the significant errors and changes to the Westinghouse ECCS evaluation models on loss of coolant accident (LOCA) analyses p6rformed for Vogtle Electric Generating Plant (VEGP) Units 1 and 2. The current LOCA analysis results were reported in the 1992 annual report (reference 1) based t the format ;
presented in WCAP-13451 (reference 2). The LOCA analysis, assessments, and safety evaluation results reported herein will be included in a future VEGP Final Safety Analysis Report (FSAR) update. ,
i LARGE BREAK LOCA ECCS EVALUATION MODEL The large break LOCA (LBLOCA) analysis PCT for VEGP Units 1 and 2 remains unchanged since the 1992 annual report. The LBLOCA analysis results are based on the Westinghouse BASH large break ECCS evaluation model (reference
- 3) as approved by the NRC for VEGP-specific application (reference 4). The
- limiting size break analysis assumed the following information important to the LBLOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly o Core Power - 1.02
- 3565 Mwt o Vessel Average Temperature - 571.90F c Steam Generator Plugging Level - 10%
o FQ = 2.50 o F-delta-H = 1.65
1 t%
ATTACHMENT '
Page 2 for VEGP Units 1 and 2, the limiting size break continues to be the double-ended guillotine rupture of the cold leg piping with a discharge coefficient of CD - 0.6. The LBLOCA analysis calculated PCT remains 20250F .
(Reference 1).
The transition core penalty, containment purge, and Tavg uncertainty items continue to be listed separately per the format of WCAP-13451. These items are listed separately because they are not explicitly modeled. The PCT assessment values for these items are 10, 11, and 500F, respectively. ,
The steam generator flow area application and structural metal heat modeling assessments continue to be listed separately per the forrrt of WCAP-13451. These items are prior BASH large break ECCS model asses 'nts, the absolute sum of which do not exceed 500F. The PCT assessment vi 4 for these items are +10 and -250F, respectively.
NEW BASH LARGE BREAK ECCS LOCA MODEL ASSESSMENTS There are no new BASH large break ECCS model assessments since the 1992 annual report. l 10 CFR 50.59 EVALUATION ASSESSMENTS e Two plant changes pursuant to 10 CFR 50.59 which affect the LBLOCA analysis results have been evaluated. The evaluations were for fuel reconstitution and the addition of a permanent radiation shield. ,
Because the VEGP Unit 1 Cycle 5 core design contains one fuel assembly that has been reconstituted with one stainless steel filler rod, an evaluation of the LBLOCA PCT has been performed. The evaluation concluded that a 20F penalty must be assessed to the VEGP Unit 1 PCT results.
A permanent reactor vessel head radiation shield was installed on VEGP Unit !
- 2. The previous radiation shielding was temporary, installed during the refueling outage and removed following the outage. The shield will now become an integral part of the reactor vessel head. Due to the additional heat sink surface area and mass, the vessel flooding rate is reduced, which results in a 10F increase in PCT. Thus, the evaluation concluded that a i 10F penalty must be assessed to the VEGP Unit 2 PCT results. At the next i refueling outage for VEGP Unit 1, the permanent shielding will also be installed, and the same IDF penalty will be assessed to VEGP Unit 1.
l 1
LICENSING BASIS LBLOCA PCT Based on the above discussions concerning the VEGP-specific application of the Westinghouse BASH large break ECCS Evaluation Model, the licensing I basis LBLOCA PCT is as follows:- !
I
- 1 ATTACHMENT Page 3
.A. 1992 Annual Report LBLOCA Analysis of Record (Reference 1)
- 1. BASH Large Break ECCS Model Analysis Result 2025.00F
- 2. Evaluation for Containment Purging + 10.00F
- 3. Evaluation for +/- 60F Uncertainty Bard + 11.00F '
- 4. Evaluation for Transition Cycle Penalty + 50.00F B. BASH Large Break ECCS Model Assessmants Prior to 1993
- 1. Steam Generator Flow Area Application + 10.00F
- 2. Structural Metal Heat Modeling - 25.00F C. 10 CFR 50.59 Evaluations
- 1. Fuel Reconstitution (VEGP Unit 1) + 2.00F
- 2. Permanent Radiation Shield (VEGP Unit 2) + 1.00F D. 199310 CFR 50.46 BASH Large Break LLCS Model /.ssessments None + 0.00F LICENSING BASIS LBLOCA PCT (Unit 1) -
ZD31J0F (Unit 2) - 2082.00F CONCLUSION An evaluation of the effect of errors and changes to the Westinghouse BASH 1arge break ECCS evaluation model was performed on the LBLOCA analysis results. When the effects of the BASH ECCS model errors / changes and safety evaluations were combined with the VEGP LBLOCA analysis results, it was determined that VEGP Units 1 and 2 are in compliance with the requirements of 10 CFR 50.46(b).
SMALL BREAK LOCA ECCS EVALUATION MODEL Significant errors / changes were assessed against the small break LOCA (SBLOCA) analysis PCT for VEGP Units 1 and 2 since the 1992 annual report.
The SBLOCA analysis results are based on the Westinghouse NOTRUMP small break ECCS evaluation model (reference 5) as approved by the NRC for VEGP-specific application (reference 4). The limiting size break analysis assumed the following information important to the SBLOCA analyses:
o 17x17 VANTAGE-5 Fuel Assembly e Core Power - 1.02
- 3565 MWT o Vessel Average Temperature - 571.90F o Steam Generator Plugging Level - 10%
a,
i
.; w ,
ATTACHMENT ,
Page 4 o FQ = 2.48 at 9.5 ft o F-delta-H - 1.70 For VEGP Units 1 and 2, the limiting size small break continues to ha a 3-inch equivalent diameter break in the cold leg. The current SBLOCA analysis calculated PCT is 18090F (reference 1).
The steam ger 'rator lower level tap raiocation and Tava uncertainty items '
continue to bi listed separately per the format of WCAP-13451. These items are listed se;arately because they are not explicitly modeled. The PCT assessment va ues on these items are 15 and 40F, respectively.
The Bessel function error was previously assessed as a 250F penalty (reference 1) to the NOTRUMP small break ECCS model and is' listed separately per the format of WCAP-13451.
NEW NOTRUMP SMALL BREAK ECCS MODEL ASSESSMENIS The following errors and changes to the NOTRUMP small break ECCS evaluation model would affect the SBLOCA PCT analysis results:
Safety In.iection in the Broken Loon Westinghouse has completed an evaluation of a potential issue concerning the modeling of safety injection (SI) flow into the broken reactor coolant ,
system (RCS) loop for the SBLOCA. In previous analyses, Westinghouse assumed that modeling SI flow into the broken RCS loop would result in a lower calculated PCT because additional SI flow would be expected to provide additional core cooling. Therefore, in previous analyses, SI flow into the broken RCS loop was modeled as spilling directly into the containment sump to provide conservative PCT results. Recent evaluations indicate for SBLOCA events using NOTRUMP, that modeling SI flow into the ,
broken RCS loop will actually result in a significant increase in PCT. The increase in PCT occurs as a result of competition between the steam venting Nt the break and the SI to the broken loop both exiting through the break.
Ine competition between the steam and the SI results in higher RCS pressures, therefore, lower delivered SI flow rates to the intact RCS loops and an increase in PCT. The effect of this issue is a 1500F increase in PCT. Therefore, a 1500F penalty has been assessed against the VEGP SBLOCA PCT results.
Imoroved Steam Condensation Model lhe SI in the broken loop issue described above is significant with regard to the effect on calculated SBLOCA PCT results. An offsetting steam condensation benefit has been identified which more than offsets the above 1500F PCT penalty. Improved condensation of the loop steam in the intact loops results in lower RCS pressures and increased SI' flow rates. The
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i increased SI flow rates result in a lower calculated PCT. Thus, the- -
negative effects of SI in the broken loop can'be fully offset by an '
improved SI steam condensation model in the intact RCS loops. Hence, e y 1500F benefit has been assessed against the VEGP SBLOCA PCT results to .
offset the PCT penalty for SI in the broken loop.
The Westinghouse Owners Group is reviewing both issues, SI in the broken loop and the improved SI steam condensation model, for possible development of a generic program for resolution that is acceptable to the NRC. '
Westinghouse is keeping the NRC (Reactor Systems Branch) informed of-the resolution of these two issues.
Drift Flux Flow Reaime Errors Errors were discovered by Westinghouse in both WCAP-100B1-A and related ,
coding in NOTRUMP where the improved TRAC-P1 vertical flow regime map is evaluated. This flow regime map is only used during counter-current flow >
conditions in vertical flow links. The error occurs as a result of an i unbounded parameter which may lead to a discontinuity in the flow regime i map under certain circumstances. The effect of this issue is a 130F decrease in PCT. The error has been correct?d in NOTRUMP. A benefit of 130F has been assessed against the VEGP SBLOCA PCT results.
li0 CFR 50.59 EVALUATION ASSESSMENTS Two plant changes pursuant to 10 CFR 50.59 which affect the SBLOCA analysis results have been evaluated. The evaluations were for fuel reconstitution i
1 and a potential loose part.
Because the VEGP Unit 1 Cycle 5 core design contains one fuel assembly that '
has been reconstituted with one stainless steel filler rod, an' evaluation of the SBLOCA PCT results has been performed. The evaluation concluded ;
that a 10F penalty must be assessed to the VEGP Unit 1 SBLOCA PCT results.
An evaluation was performed to determine the effect of a loose part 'in the 'l RCS. The loose part has been identified as the ring from the quick click '
pin on the fuel handling tool. The ring was lost in the spent fuel pool-and may have been carried to the RCS by a fuel assembly. The evaluation concluded that a 20F PCT penalty must be assessed against the VEGP Unit 1 i SBLOCA PCT results. )
1 LICENSING BASIS SBLOCA PCT' !
Based on the above discussions concerning the VEGP-specific application of the Westinghou:e NOTRUMP small break ECCS evaluation model, the licensing basis SBLOCA PCT is as_ follows:
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A. 1992 Annual Report SBLOCA Analysis of Record (Reference 1)
- 1. NOTRUMP Small Break ECCS Model Analysis Result 1809.00F i
- 2. Evaluation for Steam Generator Lower Level Tap Relocation + 15.00F
- 3. Evaluation for +/- 60F Uncertainty Band -
4.00F B. NOTRUMP Small Break ECCS Model Assessments Prior to 195 Bessel function error + 25.00F C. 10 CFR 50.59 Evaluations
- 1. Fuel Reconstitution (VEGP Unit 1) + 1.00F
- 2. Loose Parts (VEGP Unit 1) + 2.00F .
D. 199310 CFR 50.46 NOTRUMP Small Break ECCS Model Assessments '
- l. SI in Broken Loop + 150.00F
- 2. Improved Condensation Model - 150.00F
- 3. Drift Flux Flow Regime -
13.00F LICENSING BASIS SBLOCA PCT (Unit 1) - 1843.00F (Unit 2) = 1840.00F CONCLtL510E An evaluation of the effect of errors and changes to the Westinghouse NOTRUMP small break ECCS Evaluation Model was performed on the SBLOCA 4 analysis results. When the effects of the NOTRUMP' ECCS model errors / changes and safety evaluations were combined with the VEGP SBLOCA ,
analysis results, it was determined that the sum of the absolute magnitude J of the errors / changes is greater than 500F. Therefore, in compliance with 10 CFR 50.46(a)(3)(ii), a 30-day report to the NRC is required. However, the net effect of the errors / changes is a reduction in SBLOCA PCT results.
Because the NOTRUMP sndll break ECCS errors / changes affect many plants, the '
Westinghouse Owners Group is reviewing this issue and possible development .
of a generic program for resolution. A schedule for resolution and '
reanalysis will be provided following the completion of the Westinghouse i Owners Group review.
REFERENCES
- 1. ELV-05255, "Vogtle Electric Generating Plant,10 CFR 50.46 ECCS Model 1992 Report," letter from C. K. McCoy-(GPC) to USNRC, dated February 23, 1993. :)
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- 2. WCAP-13451, " Westinghouse Methooology for Implementation of 10 CFR 50.46 Reporting," dated October 1992.
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- 3. "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code," WCAP-11524-A (Non-Proprietary), March 1987.
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- 4. ELV-02166, "Vogtle Electric Generating Plant, Request for Technical Specifications Changes VANTAGE-5 Fuel Design," letter from W. G. Hairston, III, to USNRC, dated November 29, 1990.
- 5. " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP ,
Code," WCAP-10081-A (Non-Proprietary).
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