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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] |
Text
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,,'. C. K. Mr Cey Sevaiera Nedent . ,
, Vice Presdent Opere6ag Con 9any. loc.
%d' Vogtle Projo;t 40 Irwerness Centar Paiway PO Box 179$
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. . Tel7059R7171 Iv 205932 ONG SOUTHERN COMPANY January 16, 1998 #""U " k"' D"' D'#
LCWO849.K Docket Nos. 50-424
$0 425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Gentlemen: ,
VOGTLE ELECTRIC GENERATING PLANT l ADDITIONAL INFORMATION i/UEL STORAGE POOL BORQN DILUTION t
Southern Nuclear letter LCWO849.E dated August 8,1997, requested Technical Specifications changes that allow credit for the soluble boron concentration in the fuel storage pools. Enclosure 6 of that letter included an evaluation of the effects of potential fuel storage pool dilution. This letter includes additional information and analysis of the potential for dilution of the fuel storage pool.
The proposed Technical Specifications are based on analysis demonstrating that the fuel to be stored in the fuel storage pool will remain subcritical without credit for soluble boron. ,
Soluble boron credit is used 1o establich suberiticality margin by maintaining Kaless than the normal design value of 0.95. The criticality analyt.es show that La will be less than 1.0 on a 95/95 basis with the fuel storage pool filled with non borated water. The boron dilution analysis demonstrates that additional administrative requirements are not required to maintain the fuel storage pool suberiticality mergin, because it is not credible that such a dilution would go undetected.;The dilution study lists potentia! sources of pool dilution and provides conservative evaluations of the potential for dilution, and the time to reach the dilution point (concentration where La is F. cater than 0.95).
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The analyses of potential dilution events have been revised to assume that the two fuel storage pools are separated. This ciTectisely reduces the amount of water that is required to dilute a fuel storage pool and adds conservatism to the analysis. For the purposes of the dilution analysis, the boron concentration for maintaining Ke r s 0.95 was increased to 600 ppm from the value of 500 ppm in the dilution analysis submitted in LCV 0849 E. This is conservati : with respect to the criticality analyses for the current Vogtle rart a and is consistent with the c.iticality analyses f( .e planned placement of additional racks ir, the Unit I fuel storage pool. These additional racks were described in SNC letter LCV 0828 B dated November 20,1997. The effects of piping failures on dilution water sources have been revised to be in accordance with NRC Branch Technical Po.sition hiEB 3 1.
In the dilutloa analysis report a change was made to the description in section 2.8 regarding the status of administrative controls. The status description in LCV 0849 E stated that the controls "are in place". Some of the items are presently in effect, but a portion of them are planned to be in effect by implementatior. of the proposed revised Technical Specifications. The description was changed to state the controls "cither are in place or will be la place".
With the exception of the change to section 2.8 of the dilution analysis report, the above-described changes were discussed with the NRC staffin a telephone conference on November 24,1997, The incorporation of these changes into the analyses does not alter the Technical Specification changes requested in LCV 0849 E or in LCV-0828 H, and it decs not change the conclusions of the evaluations submitted in support of those requested changes.
Sincerely, C. K. hiccoy CKht/llWht/RJF Attachments:
- 1. Responses to Questions
- 2. Revised Spent Fuel Pool Boron Dilution Analysis
.- - - ~,
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. t U. S. Nuclear Regulatory Commise8on Page 3 l 4
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. xc: Southern Nuclear !
Mr. J. H. Beasley, Jr.
Mr. M. Sheibani j l
. NORMS U. S. Nuctaar Raoulatorv Commission i Mr. L. A. Reyes,' Regional Administrator
._ Mr. D. H. Jaffe, Senior Project Manager, NRR .
Mr. J. Zeiler, Senior Resident Inspector, Vogtle i
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NTTACHMENT 1 ,
i RESPONSE TO NRC QUESTIONS ON DILUTION ANALYSIS ,
OF Tile SPENT FUEL POOL K$V(XiAl' ASi97VAAIShSOl6060 IX1C i
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. REhPONhi NRC OUESTIONS ON DILUTION ANALYSIS OF THE SPENT r FUEL POOL Question it s The boron dilution analysis (enclosure 6) states that the two spent fuel pools (SFPs) ai t ;
normally connected via the cask loading area, which provides communication between the two SFPs. As a result, a combined volume of 772,000 gallons of water is used in the calculation for the boron dilution times and volumes. Equation 1 in the submittal is used for this calculation. llowever, to use Equation 1, timely through mixing between the two ,
pools is assumed. Provide, justification how the relatively narrow channel through the !
i cask loading area (the channel appears in drawing AX4DD304 to be approximately 1/10th the wi.ith of the pools) supports the assumption of timely, through mixing to allow use of the combined water volume. (Alternatively, the boron dilution analysis may be completed for separate SFPs). 3 Responset The primary source of dilution water considered in enclosure 6 was from postulated broken pipes that spray water onto the surface of the spent fuel pool. Under these conditions the water level in the two pools will rise together until exceeding the elevation of the curb around the individual pools (approximately :levation 220' 6"). At this point <
water will flow along the surface between the two pools (also through or over the gates between the pools) as well as over the curb from both pools onto the floor of the fuel handling building. Thereafter, the primary source of unborated water is on to the surface of a pool and the flow of water fcorn the pools is from the surface of both pools. The flow of water from each pool will be approximately the same. Under these conditions the water flowing from the pools over the curb is expected to contain less soluble bomn than the average mixture in the two pools. Therefore, the assumption of fully mixed flow from the pools is believed to be consen ative.
The dilution analysis has been revised assuming that the gates between the two pools are
. closed. This analysis is attached.
Question 2: ,
. Given the use of a combined volume for the boron dilution analysis, clarify the statement that the SFPs are "normally" connected. When are the SFPs separated 7 Ilow long are they allowed to remain separated 7 What controls are in place to ensure that the SFPs are/ remain connected 7 KAvoO\RF.As\97VAA15Mol6060IX)C p w . - > .
Response
Itased on discussions with reactor engineering personnel, the pools hr,ve remained connected (the gates between the two pools are open) since the filling of the Unit 2 pool.
The Unit 2 pool contains 2098 fuel storage locations and the Unit 1 pool only contains 288 storage locations. This arrangement requires the transfer of fuel between the two pools which can only occur if the gates are open.
The gates would have to be closed during activities that require draining of the cask pit or draining of one of the SFPs. There are no scheduled activities requiring the draining of the cask pit. The placement of additional racks into the Unit 1 pool is expected to take place in 1998. This aedvity will include closing of the gates between the two pools. At some date in the future, shipment of spent fuel in spent fuel casks will occur. That activity may include periods of gate closure. There are no limits on the frequency or duration of gate closure and there are no requirements to maintain the gates open.
If a pool dilution event occurs while the gat:s are closed, the water level in the affected pool will rise until it over flows the gates. At that point water will flow into the other pool until both pools reach the same level. If the addition of water continues, the level of water will eventually reach the height of the curb around the pools and over flow on to the Fuel llandling Building floor from both pools. Nevertheless, as discussed in the response to questicn 1 the dilution analysis has been re-evaluated considering separate pools.
Question 3:
The boron dilution analysis states on Page 20 that the fire protection system (FP) flow rates range from 2l00 to 4800 ppm and that the most limiting scenario includes a FP fl w of 2300 gpm (seismic break of non-seismic piping). Why is 2300 gpm more limiting than 4800 gpm?
Response
The dilution of both pools as described in er. closure 6 requires over 1,070,000 gallons of water The combined fire protection water in both tanks is only 600,000 gallons.
Therefore, extraordinary assumptions are required to produce a dilution event capable of diluting the combined pools. Make up to the fire n atection tanks is available at a rate of about 650 gpm. By evah.ating the additional amount of water needed to dilute the pool, assuming that it is supplied by the make .,p water pumps, and that the fire protection tanks reach emptyjust as the required amount of water is supplied to the fuel storage pool, the optimum flow rate necessary to dilute the pools in the shortest amount of time can be determined. For greater flow rates the fire protection tanks will empty and the pumps will trip prior to supplying the 1,070,000 gallons necessary to dilute the pools.
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Lower flow rates will result in longer dilution times. This optimum flow rate was
. detennined to be about 2300 gpm.
The fire protection system has one electric oriven pump and two diesel driven pumps.
Pump flow at runout conditions is about 4800 gpm. For the analysis assuming dilution of both pools to 500 ppm, as described in enclosure 6,2300 gpm is the optimum flow rate to minimize the time for dilution of the pools. These assumptions were used to calculate limiting dilution times, however, no mechanisms have been identified that result in flows of these magnitudes.
These assumptions do not take credit for actual location or configuration of potential failures in fire protection piping. They do not take credit for the location of fire protection piping in the vicinity of the fuel storage pools.
All of the fire protection lines in the vicinity of the SFPs are seismically analyzed and supported even though they are not designated as seismic category 1 piping. For the purposes of flooding analyses, tiranch Technical Position MEll 3 1 requires the postulation of through wall cracks in moderate energy non seismic lines. There are 3 six inch fire protection lines in the vicinity of the Unit 1 pool and 3 in the vicinity of the Unit 2 pcol. The postulation of a through wall crack in accordance with IlTP. MEll 3 1 would result in postulated flows well below 2300 gpm.
Question 4:
Section 3.3: Evaluation ofinfrequent SFP Configurations. The boron dilution analysis states that the most limiting configuration is when the SFPs are separated. From a seismic event, a dilution could occur in 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. What, if any, controls are in place to assist the operators' " heightened awareness" of a SFP dilution following a seismic event?
Such as, a procedure for checking the plant's physical condition or alarms (other than SFP level)?
Response
Abnonnal Operating Procedure 18036-C "Scismic Event" provides operator response fallowing a seismic event. The procedure requires the initiation of a visual inspection of all accessible areas to determine damage. Following implementation of the revised Technical Specifications this procedure will be revised to include a specific reference to the possibility of dilution of the SFP due to pipe leaks.
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Question 5:
The boron dilution analysis states on page 20 that if offsite power is not available as a result of a seismic event it would take approximately 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to reach the dilution endpoint.1he sources and flow rates to reach this conclusion were not provided. Please provide this information.
Response: -
W8 a loss of electric power, the makeup pumps to the fire protection tanks are unavailable.1his significantly increases the amount of time to reach the dilution point following the emptying of the fire protection tanks.
- Flow rates of systems for broken lines (without offsite power) liire Protection: 4800 gpm (pump runout)-600,000 gal. limit.
Demineralized Water 120 gpm (for gravity fill)-for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Normal Chilled Water: 300 gpm for 30 minutes (total water admitted is 9000 gallons)
Utility Water: 40 gpm for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> The calculation in attachn ent 2 has been revised to consider a single fuel storage pool and pipe leaks in accordance with MEB 3 1.
Question 6:
1he boron dilution analysis states on page 19 that certain pipes are assumed to remain intact during/afler a seismic event. Clarify that all such pipes are seismically qualified.
Response
This statement referred to non seismic piping outside the SFP area that feeds piping in the SFP area. It is necessary for that piping to remain intact in order to conduct water to -
the SFP area. .Thus the analysis inherently assumed that nonseismic piping outside the SFP are. mains intact while it fails in the SFP area.
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