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Category:CORRESPONDENCE-LETTERS
MONTHYEAR1CAN109906, Forwards Framatome Technologies,Inc non-proprietary TR BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheet of Once-Through Sgs, Rev 11999-10-19019 October 1999 Forwards Framatome Technologies,Inc non-proprietary TR BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheet of Once-Through Sgs, Rev 1 ML20217J4971999-10-18018 October 1999 Requests Addl Info Re Results of Util Most Recent Steam Generator Insp at ANO-2 & Util Methodology Used to Predict Future Performance of SG Tubes ML20217J3871999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through SG, Rev 0 Marked as Proprietary Will Be Withheld from Public Disclosure 2CAN109902, Submits Withdrawal of Code Case N-593 for ANO-2 Replacement SGs1999-10-15015 October 1999 Submits Withdrawal of Code Case N-593 for ANO-2 Replacement SGs ML20217J3601999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Integranular Attack in Tubesheets of Once-Through SG, Rev 1 Marked as Proprietary Will Be Withheld from Public Disclosure 2CAN109903, Forwards Response to RAI Re Proposed Tech Specs Change for Special SG Insp1999-10-14014 October 1999 Forwards Response to RAI Re Proposed Tech Specs Change for Special SG Insp ML20217D1721999-10-0808 October 1999 Forwards RAI Re 990729 Request for Amend to TSs Allowing Special SG Insp for Plant,Unit 2.Questions Re Proposed Insp Scope for Axial Cracking Degradation in Eggcrate Support Region Submitted.Response Requested by 991015 1CAN109905, Discusses Insp of Once Through SG Tubing Surveillance Performed During 1R15 Scheduled RFO on 990910.Category C-3 Results,Included1999-10-0404 October 1999 Discusses Insp of Once Through SG Tubing Surveillance Performed During 1R15 Scheduled RFO on 990910.Category C-3 Results,Included ML20212L0621999-10-0101 October 1999 Forwards Safety Evaluation & Exemption from Certain Requirements of 10CFR50,App R,Section III.G.2, Fire Protection of Safe Shutdown Capability 1CAN099908, Withdraws 990919 Exigent TS Change Request to Allow Continued Installation of re-rolls for One Cycle of Operation Through End of Cycle 16 in Conjunction with Addl Insp Criteria1999-09-30030 September 1999 Withdraws 990919 Exigent TS Change Request to Allow Continued Installation of re-rolls for One Cycle of Operation Through End of Cycle 16 in Conjunction with Addl Insp Criteria 2CAN099902, Requests That NRC Assign CENPD-132,Suppl 4-P, Calculative Methods for Abb Cenp Large Break LOCA Evaluation Model, Review Priority So That Approval Will Be Granted No Later than Oct 31,20001999-09-29029 September 1999 Requests That NRC Assign CENPD-132,Suppl 4-P, Calculative Methods for Abb Cenp Large Break LOCA Evaluation Model, Review Priority So That Approval Will Be Granted No Later than Oct 31,2000 1CAN099903, Forwards Rev 0 to COLR for ANO-1 Cycle 16, IAW TS 6.12.31999-09-27027 September 1999 Forwards Rev 0 to COLR for ANO-1 Cycle 16, IAW TS 6.12.3 1CAN099907, Requests That Alternative Be Allowed in Accordance with 10CFR50.55a(a)(3)(i) & (II) as Discussed in Encl 1.Encl 2 & 3 Stress Analysis & Flaw Evaluation Summaries Ref in Encl Alternative1999-09-26026 September 1999 Requests That Alternative Be Allowed in Accordance with 10CFR50.55a(a)(3)(i) & (II) as Discussed in Encl 1.Encl 2 & 3 Stress Analysis & Flaw Evaluation Summaries Ref in Encl Alternative 1CAN099906, Forwards 1R15 Growth Data Obtained & Analyzed Through 990922 & Includes Plus Point Voltages,Axial Extent & Circumferential Extent Patches,As Well as Preliminary Growth Conclusions Based on Analysis of Data1999-09-24024 September 1999 Forwards 1R15 Growth Data Obtained & Analyzed Through 990922 & Includes Plus Point Voltages,Axial Extent & Circumferential Extent Patches,As Well as Preliminary Growth Conclusions Based on Analysis of Data 2CAN099901, Informs That G Kendrick,License SOP-43658,no Longer Has Need to Maintain Operating License on Ano,Unit 2.Entergy Requests That License for Individual Be Withdrawn,Due to Resignation, Effective 9908271999-09-24024 September 1999 Informs That G Kendrick,License SOP-43658,no Longer Has Need to Maintain Operating License on Ano,Unit 2.Entergy Requests That License for Individual Be Withdrawn,Due to Resignation, Effective 990827 2CAN099904, Forwards Ano,Unit 2 10CFR50.59 Rept for Time Period Ending 990225.Rept Contains Brief Description of Changes in Procedures & in Facility as Described in Sar,Tests & Experiments Conducted & Other Changes to SAR1999-09-23023 September 1999 Forwards Ano,Unit 2 10CFR50.59 Rept for Time Period Ending 990225.Rept Contains Brief Description of Changes in Procedures & in Facility as Described in Sar,Tests & Experiments Conducted & Other Changes to SAR ML20212F5031999-09-22022 September 1999 Forwards SER Granting Relief Requests 1-98-001 & 1-98-002 Which Would Require Design Mods to Comply with Code Requirements,Which Would Impose Significant Burden Pursuant to 10CFR50.55a(g)(6)(i) 1CAN099905, Submits Supplemental Info in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria.Proposed TS Rev & Info Related to Use of Alternate Repair Discussed in Attachments1999-09-17017 September 1999 Submits Supplemental Info in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria.Proposed TS Rev & Info Related to Use of Alternate Repair Discussed in Attachments ML20212D9961999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of Arkansas Nuclear One.Nrc Plan to Conduct Core Insps at Facility Over Next 7 Months.Details of Insp Plan Through March 2000 Encl 1CAN099902, Forwards Proprietary Rev 1 to Topical Rept BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs, in Response to 990831 Rai.Proprietary Encl Withheld1999-09-15015 September 1999 Forwards Proprietary Rev 1 to Topical Rept BAW-10235P, Mgt Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through Sgs, in Response to 990831 Rai.Proprietary Encl Withheld 2CAN099905, Informs That Jk Caery,License OP-42589 & as Bates,License OP-42506,no Longer Need to Maintain Operating License at Ano,Unit 2.Withdrawal of Licenses Is Requested1999-09-0909 September 1999 Informs That Jk Caery,License OP-42589 & as Bates,License OP-42506,no Longer Need to Maintain Operating License at Ano,Unit 2.Withdrawal of Licenses Is Requested 1CAN099901, Forwards Responses to 990831 RAI Containing follow-up Questions Discussed on 990823-26,in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria. Revs to Proposed TSs Included in Attachments1999-09-0707 September 1999 Forwards Responses to 990831 RAI Containing follow-up Questions Discussed on 990823-26,in Support of SG Outer Diameter Intergranular Attack Alternate Repair Criteria. Revs to Proposed TSs Included in Attachments ML20211P4121999-09-0707 September 1999 Requests NRC Staff Review & Approval of Integrated Nuclear Security Plan (Insp) & Integrated Security Training & Qualification Plan (Ist&Q), for Use by All Entergy Operations,Inc.Encl Withheld,Per 10CFR2.790(d) 0CAN099906, Forwards Comments on Ano,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid,Version 2,in Response to NRC 990708 & 0715 Ltrs1999-09-0101 September 1999 Forwards Comments on Ano,Units 1 & 2 Specific Reactor Vessel Info Contained in Rvid,Version 2,in Response to NRC 990708 & 0715 Ltrs ML20211L4901999-09-0101 September 1999 Forwards Insp Repts 50-313/99-12 & 50-368/99-12 on 990711- 0821.No Violations Noted ML20211J2351999-08-31031 August 1999 Forwards Request for Addl Info Re SG Outer Diameter Intergranular Attack Alternate Repair Criteria for Plant, Unit 1 ML20211E6161999-08-25025 August 1999 Forwards Amend 15 to ANO Unit 2,USAR,per 10CFR50.71(e) & 10CFR50.4(b)(6).Summary of 10CFR50.59 Evaluations Associated with Amend 15 of ANO Unit 2 SAR Will Be Provided Under Separate Cover Ltr with 30 Days 0CAN089905, Forwards Arkansas Nuclear One Units 1 & 2 FFD Program Performance Data for Period Jan-June 19991999-08-25025 August 1999 Forwards Arkansas Nuclear One Units 1 & 2 FFD Program Performance Data for Period Jan-June 1999 ML20211F4181999-08-25025 August 1999 Forwards SE Accepting Licensee 980603 & 990517 Requests for Approval of risk-informed Alternative to 1992 Edition of ASME BPV Code Section Xi,Insp Requirements for Class 1, Category B-J Piping Welds ML20211G0731999-08-19019 August 1999 Forwards Applications for Renewal of Operating License for Kw Canitz & Aj South.Without Encls 1CAN089904, Forwards Addl Info in Support of SG Tube End Cracking Alternate Repair Criteria,In Response to NRC 990728 Rai. Proposed TS Changes Encl1999-08-19019 August 1999 Forwards Addl Info in Support of SG Tube End Cracking Alternate Repair Criteria,In Response to NRC 990728 Rai. Proposed TS Changes Encl ML20211C5101999-08-19019 August 1999 Forwards Certified Copies of Liability Insurance Policy Endorsements Issued in First Half of 1999 for Each Entergy Operations,Inc Nuclear Unit,Per 10CFR140.15 ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl 0CAN089903, Submits Addl Response to NRC Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Gate Valves1999-08-12012 August 1999 Submits Addl Response to NRC Second RAI Re GL 95-07, Pressure Locking & Thermal Binding of Gate Valves IR 05000368/19990111999-08-12012 August 1999 Forwards Insp Repts 50-313//99-11 & 50-368/99-11 on 990719-23.No Violations Noted.Insp Focused on Review of Licensed Operator Requalification Program & Observation of Requalification Exam Activities at Unit 1 2CAN089901, Forwards Description of Planned Scope & Expansion Criteria for Special SG Tube Insp,In Support of Proposed ANO-2 TS Amend for 2P99 Special SG Insp Submitted on 9907291999-08-0606 August 1999 Forwards Description of Planned Scope & Expansion Criteria for Special SG Tube Insp,In Support of Proposed ANO-2 TS Amend for 2P99 Special SG Insp Submitted on 990729 1CAN089902, Requests NRC Input on Encl Proposed Draft Format for ANO-1 License Renewal Application,Which Will Provide Option to Continue Operating Plant for Addl Twenty Years Beyond End of Current Operating License1999-08-0505 August 1999 Requests NRC Input on Encl Proposed Draft Format for ANO-1 License Renewal Application,Which Will Provide Option to Continue Operating Plant for Addl Twenty Years Beyond End of Current Operating License 2CAN089902, Informs That Tl Russell,License SOP-43587-1 & Jk Fancher, License OP-42300-1,no Longer Have Need to Maintain Operating License at ANO-2.Withdrawal of Licenses Requested1999-08-0404 August 1999 Informs That Tl Russell,License SOP-43587-1 & Jk Fancher, License OP-42300-1,no Longer Have Need to Maintain Operating License at ANO-2.Withdrawal of Licenses Requested ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams 0CAN089902, Submits 60 Day Response to GL 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal. Proposed Actions That Will Be Taken on ANO Unit 1 RB Purge Filtration Sys & Unit 2 Containment Purge & Exhaust Sys,Clarified1999-08-0202 August 1999 Submits 60 Day Response to GL 99-02, Laboratory Testing of Nuclear Grade Activated Charcoal. Proposed Actions That Will Be Taken on ANO Unit 1 RB Purge Filtration Sys & Unit 2 Containment Purge & Exhaust Sys,Clarified 0CAN089901, Forwards Info Re Estimate of licensee-originated Licensing Actions for ANO-1 & ANO-2,in Response to Administrative Ltr 99-02,dtd 9906031999-08-0202 August 1999 Forwards Info Re Estimate of licensee-originated Licensing Actions for ANO-1 & ANO-2,in Response to Administrative Ltr 99-02,dtd 990603 ML20210L3581999-07-29029 July 1999 Ltr Contract,Task Order 43, Arkansas Nuclear One Safety System Engineering Insp (Ssei), Under Contract NRC-03-98-021 1CAN079903, Forwards non-proprietary Addendum to Rev 0 of Topical Rept BAW-2346P,in Support of Proposed TS Changes Revising SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Tube End Cracks1999-07-29029 July 1999 Forwards non-proprietary Addendum to Rev 0 of Topical Rept BAW-2346P,in Support of Proposed TS Changes Revising SG Tubing Surveillance Requirements to Provide Alternate Repair Criteria for Tube End Cracks ML20216D8131999-07-28028 July 1999 Forwards Request for Addl Info Re SG Tube End Cracking Alternate Repair Criteria for Plant,Unit 1 ML20216D3561999-07-23023 July 1999 Discusses non-cited Violation Identified in Insp Rept 50-313/98-21,involving Failure to Have Acceptable Alternative Shutdown Capability for ANO-1 ML20210C2191999-07-21021 July 1999 Forwards Insp Repts 50-313/99-08 & 50-368/99-08 on 990530-0710 at Arkansas Nuclear One,Units 1 & 2,reactor Facility.No Violations Noted.Conduct of Activities at Plant Generally Characterized by safety-conscious Operations ML20209H5251999-07-15015 July 1999 Informs That as Result of NRC Review of Licensee 980701 & 990311 Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 RAI, Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 1CAN079901, Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages1999-07-14014 July 1999 Forwards Proposed Changes to Current Util 990409 Submittal Re Rev to RB Structural Integrity Requirements Contained in Plant Ts.Proposed Revs Affect ACs & Applicable Bases Re ISI Reporting for Containment Structures,Tendons & Anchorages 0CAN079902, Responds to NRC Telcon RAI Re Proposed Administrative Controls TS Changes.Revised TS Pages Which Replaces Pages Previously Provided in 981124 Submittal,Encl1999-07-14014 July 1999 Responds to NRC Telcon RAI Re Proposed Administrative Controls TS Changes.Revised TS Pages Which Replaces Pages Previously Provided in 981124 Submittal,Encl ML20209E5551999-07-0808 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1,suppl 1,staff Revised Info in Rv Integrity Database & Releasing Database as Rvid Version 2 1999-09-09
[Table view] Category:NRC TO UTILITY
MONTHYEARML20062F0481990-11-19019 November 1990 Advises That SR Peterson Replacing C Poslusny as Project Manager for Facility,Effective 901105 ML20216K0001990-11-13013 November 1990 Forwards Insp Repts 50-313/90-30 & 50-368/90-30 on 900905- 1016 & Notice of Violation IR 05000313/19900321990-11-0808 November 1990 Ack Receipt of Outlining Steps to Provide Addl Technical Support for Radiation Protection Manager.Planned Actions Acceptable as Documented in Insp Repts 50-313/90-32 & 50-368/90-32 ML20058G8471990-11-0808 November 1990 Forwards Insp Repts 50-313/90-32 & 50-368/90-32 on 900924-29.No Violations or Deviations Noted ML20058G1631990-11-0707 November 1990 Forwards Summary of Current Status of Unimplemented Generic Safety Issues at Plant ML20058E2881990-11-0101 November 1990 Rejects 900302 Request for Amend to Tech Specs to Revise Power Calibr Requirements for Linear Power Level & Core Protection Calculator delta-T Power & Nuclear Power Signals ML20062D5631990-10-31031 October 1990 Forwards Insp Repts 50-313/90-37 & 50-368/90-37 on 901015-19.No Violations Noted ML20062D5761990-10-31031 October 1990 Forwards Insp Repts 50-313/90-40 & 50-368/90-40 on 901015-19.No Violations Noted ML20062D5511990-10-31031 October 1990 Forwards Insp Repts 50-313/90-31 & 50-368/90-31 on 901001-04.No Violations Noted ML20059P0441990-10-16016 October 1990 Authorizes Use of Inconel 690 (I-690) as Alternate to I-600 in Steam Generator Tube Sleeves/Plugs Per 10CFR50.55a(a)(3) ML20058A5211990-10-16016 October 1990 Forwards Insp Repts 50-313/90-38 & 50-368/90-38 on 901001- 05.Violation Considered for Escalated Enforcement Action ML20058A0811990-10-15015 October 1990 Forwards Questions Re 900808 License Amend Request to Increase Reactor Power to 100% for Response ML20062A5511990-10-10010 October 1990 Forwards SER Re Util 890403,13 & 0717 Responses to Station Blackout Rule.Issue of Conformance Still Open IR 05000313/19900011990-10-0404 October 1990 Ack Receipt of 900420 & 0914 Ltrs Re Validation of Nonlicensed Operator Staffing Per Insp Repts 50-313/90-01 & 50-368/90-01 ML20059M0441990-09-26026 September 1990 Approves 900809 Request to Withhold 86-1179795-01, ANO-1 HPI Flow Rate Requirements (Ref 10CFR2.790(b)(5)) ML20059K3851990-09-14014 September 1990 Forwards Insp Repts 50-313/90-28 & 50-368/90-28 on 900827-31.No Violations or Deviations Noted.Some Weaknesses Identified in Areas of Alternate Safe Shutdown Procedure & Associated Training for Unit 2 ML20059J2431990-09-14014 September 1990 Discusses Programmed Enhancements for Generic Ltr 88-17, Loss of Dhr. Changes in Completion Schedule Should Be Submitted to NRC ML20059J9901990-09-13013 September 1990 Forwards Info Re Generic Fundamentals Exam of Operator Licensing Written Exam to Be Administered on 901010.W/o Encl ML20059G2951990-09-0606 September 1990 Advises That Rev 10 to Emergency Plan,Contained in ,Acceptable.Rev Consists of Changes Resulting from Reorganization,Relocation of Personnel from Little Rock Ofc & Improvements from Annual Emergency Preparedness Exercise ML20059E8621990-08-31031 August 1990 Forwards Insp Repts 50-313/90-27 & 50-368/90-27 on 900813-17.No Citations Issued for Violation ML20059D1551990-08-30030 August 1990 Forwards Request for Addl Info to Continue Review of 891019 Application for Amend Extending Insp Frequency of Spent Fuel Pool from Once Per 18 Months to Once Per 60 Months.Response Requested within 45 Days ML20059E5901990-08-29029 August 1990 Forwards Summary of 900823 Quarterly Performance Meeting at Plant Re NRC Authorized Activities.Meeting Provided Better Understanding of Current Implementation Status of Program Changes at Plant.List of Attendees & Viewgraphs Encl IR 05000313/19900161990-08-29029 August 1990 Discusses 900823 Meeting Re Unresolved Item Concerning Missed Surveillance Tests,Per Insp Repts 50-313/90-16 & 50-368/90-16.Violation of Tech Spec Requirements Noted But Not Cited Due to Listed Reasons ML20059F3441990-08-29029 August 1990 Forwards Insp Repts 50-313/90-23 & 50-368/90-23 on 900716-20.No Violations or Deviations Noted ML20059D6381990-08-29029 August 1990 Ack Receipt of Re Proposed Changes to Unit 1 Tech Spec 6.12.2.6(b) & Unit 2 Tech Spec 6.9.3.1.Proposed Changes Appear to Improve Quality of Semiannual Radioactive Effluent Release Repts ML20059B8461990-08-23023 August 1990 Responds to Re Violations Noted in Insp Repts 50-313/89-33 & 50-368/89-33.Violations Remain Applicable ML20056B4921990-08-22022 August 1990 Forwards Request for Addl Info Re Util 900618 Response to NRC Questions on Condensate Storage Tank Seismic Qualification.Response Should Be Provided within 45 Days to Facilitate Completion of NRC Effort ML20056B4131990-08-21021 August 1990 Forwards Summary of 900718 Meeting Re Exercise Weaknesses Noted in Insp Repts 50-313/90-08 & 50-368/90-08 ML20056B4641990-08-21021 August 1990 Ack Receipt of Advising NRC of Current Status of Security Perimeter Improvements,Per Insp Repts 50-313/87-31 & 50-368/87-31.Implementation of Design Change DCP 90-2001 for Perimeter Lighting Will Be Monitored for Adequacy Later ML20059A7011990-08-17017 August 1990 Forwards Sser Concluding That Rochester Instrument Sys Model SC-1302 Isolation Device Acceptable for Use at Plant for Interfacing SPDS W/Class IE Circuits ML20058P2601990-08-13013 August 1990 Forwards Insp Repts 50-313/90-26 & 50-368/90-26 on 900730-0803.No Violations or Deviations Noted.Rept Does Not Include Specific Insp Followup for Any Diagnostic Evaluation Team Findings ML20058N9741990-08-10010 August 1990 Forwards Insp Repts 50-313/90-19 & 50-368/90-19 on 900601- 0715 & Notice of Violation.Util Should Respond to Failure to Adequately Implement Surveillance Test Required by Tech Specs IR 05000313/19900041990-08-0909 August 1990 Ack Receipt of 900611 & 0731 Ltrs Re Steps Taken to Correct Violations Noted in Insp Repts 50-313/90-04 & 50-368/90-04 ML20056A7491990-08-0707 August 1990 Forwards Safety Evaluation Accepting Licensee Fire Barrier Penetration Seal Program & Commitment to Complete 100% Review of All Tech Spec Fire Penetration Seals by 911231 ML20055J3571990-07-31031 July 1990 Forwards Review of C-E Topical Rept Cen 387-P, C-E Owners Group Pressurizer Surge Line Flow Stratification Evaluation, Per NRC Bulletin 88-011.Adequate Basis Not Provided for Meeting Pressurizer Surge Line Code Limits ML20055J3981990-07-31031 July 1990 Discusses Util 890602 Response to Item 1.b of NRC Bulletin 88-11, Pressurizer Surge Line Thermal Stratification. Sufficient Info Provided to Justify Continued Plant Operation Until Final Rept for Unit 1 Completed ML20056A0211990-07-30030 July 1990 Ack Receipt of 890307 & s Informing NRC of Steps Taken to Correct Violation Noted in Insp Repts 50-313/88-47 & 50-368/88-47 ML20055J1201990-07-24024 July 1990 Advises That Operational Safety Team Insps 50-313/90-24 & 50-368/90-24 Scheduled at Plant Site on 900910-21 ML20055H9511990-07-23023 July 1990 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-313/90-01 & 50-368/90-01 ML20055H9471990-07-20020 July 1990 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-313/89-02 & 50-368/89-02 ML20058P7261990-07-19019 July 1990 Forwards Generic Fundamentals Exam Section of Written Operator Licensing Exam Administered on 900606 ML20055G1801990-07-17017 July 1990 Confirms 900718 Mgt Meeting in Region IV Ofc Re Exercise Weaknesses Noted During Mar 1990 Emergency Exercise ML20055F8011990-07-13013 July 1990 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-313/89-27 & 50-368/89-27.Security Officer Training Will Be Evaluated During Future Insps ML20055E5171990-07-0909 July 1990 Advises That 900607 Control Element Assembly Failure at Maine Yankee Not Applicable to Facility.Nrc Understands That Util Does Not Plan to Use Any old-style Control Element Assemblies in Future ML20055D5081990-06-29029 June 1990 Forwards Insp Repts 50-313/90-18 & 50-368/90-18 on 900521-25.No Violations or Deviations Identified.Two Open Items in Areas of Procedures & Personnel Dosimetry Noted 1990-09-06
[Table view] Category:OUTGOING CORRESPONDENCE
MONTHYEARML20217J4971999-10-18018 October 1999 Requests Addl Info Re Results of Util Most Recent Steam Generator Insp at ANO-2 & Util Methodology Used to Predict Future Performance of SG Tubes ML20217J3871999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Intergranular Attack in Tubesheets of Once-Through SG, Rev 0 Marked as Proprietary Will Be Withheld from Public Disclosure ML20217J3601999-10-15015 October 1999 Informs That Topical Rept BAW-10235P, Management Program for Volumetric Outer Diameter Integranular Attack in Tubesheets of Once-Through SG, Rev 1 Marked as Proprietary Will Be Withheld from Public Disclosure ML20217D1721999-10-0808 October 1999 Forwards RAI Re 990729 Request for Amend to TSs Allowing Special SG Insp for Plant,Unit 2.Questions Re Proposed Insp Scope for Axial Cracking Degradation in Eggcrate Support Region Submitted.Response Requested by 991015 ML20212L0621999-10-0101 October 1999 Forwards Safety Evaluation & Exemption from Certain Requirements of 10CFR50,App R,Section III.G.2, Fire Protection of Safe Shutdown Capability ML20212F5031999-09-22022 September 1999 Forwards SER Granting Relief Requests 1-98-001 & 1-98-002 Which Would Require Design Mods to Comply with Code Requirements,Which Would Impose Significant Burden Pursuant to 10CFR50.55a(g)(6)(i) ML20212D9961999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of Arkansas Nuclear One.Nrc Plan to Conduct Core Insps at Facility Over Next 7 Months.Details of Insp Plan Through March 2000 Encl ML20211L4901999-09-0101 September 1999 Forwards Insp Repts 50-313/99-12 & 50-368/99-12 on 990711- 0821.No Violations Noted ML20211J2351999-08-31031 August 1999 Forwards Request for Addl Info Re SG Outer Diameter Intergranular Attack Alternate Repair Criteria for Plant, Unit 1 ML20211F4181999-08-25025 August 1999 Forwards SE Accepting Licensee 980603 & 990517 Requests for Approval of risk-informed Alternative to 1992 Edition of ASME BPV Code Section Xi,Insp Requirements for Class 1, Category B-J Piping Welds IR 05000368/19990111999-08-12012 August 1999 Forwards Insp Repts 50-313//99-11 & 50-368/99-11 on 990719-23.No Violations Noted.Insp Focused on Review of Licensed Operator Requalification Program & Observation of Requalification Exam Activities at Unit 1 ML20211A9501999-08-12012 August 1999 Discusses 990720-21 Workshop Conducted in Region IV Ofc,Re Exchange of Info in Area of Use of Risk Insights in Regulatory Activities.List of Attendees,Summary of Topic & Issues,Agenda & Copies of Handouts Encl ML20210L1461999-08-0303 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Requests Submittal of Ltr Identifying Individuals Taking Exam,Personnel Allowed Access to Exams & Mailing Address for Exams ML20210L3581999-07-29029 July 1999 Ltr Contract,Task Order 43, Arkansas Nuclear One Safety System Engineering Insp (Ssei), Under Contract NRC-03-98-021 ML20216D8131999-07-28028 July 1999 Forwards Request for Addl Info Re SG Tube End Cracking Alternate Repair Criteria for Plant,Unit 1 ML20210C2191999-07-21021 July 1999 Forwards Insp Repts 50-313/99-08 & 50-368/99-08 on 990530-0710 at Arkansas Nuclear One,Units 1 & 2,reactor Facility.No Violations Noted.Conduct of Activities at Plant Generally Characterized by safety-conscious Operations ML20209H5251999-07-15015 July 1999 Informs That as Result of NRC Review of Licensee 980701 & 990311 Responses to GL 92-01,rev 1 & Suppl 1 & Suppl 1 RAI, Staff Revised Info in Reactor Vessel Integrity Database & Releasing Database as Rvid Version 2 ML20209E5551999-07-0808 July 1999 Informs That as Result of NRC Review of Util Responses to GL 92-01,rev 1,suppl 1,staff Revised Info in Rv Integrity Database & Releasing Database as Rvid Version 2 ML20209D8521999-07-0707 July 1999 Responds to Util 990706 Request That NRC Exercise Discretion Not to Enforce Compliance with Actions Required by TS 3.7.2, Auxiliary Electrical Sys. NOED Warranted & Approval Granted for Extension of Allowed Outage Time to 14 Days ML20209A8561999-06-25025 June 1999 Refers to Investigation Rept A4-1998-042 Re Potential Falsification of Training Record by Senior Licensed Operator at Arkansas Nuclear One Facility.Nrc Concluded That Training Attendance Record Falsified IR 05000313/19990071999-06-21021 June 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-313/99-07 & 50-368/99-07 Issued on 990514.Adequacy of Min Staffing Levels May Be Reviewed During Future Insps ML20196D4241999-06-21021 June 1999 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp of License SOP-43716 Issued on 990325.Believes That NRC Concerns Have Been Adequately Addressed at Present ML20207H3551999-06-10010 June 1999 Forwards Insp Repts 50-313/99-05 & 50-368/99-05 on 990411-0529.No Violations Noted ML20195G3481999-06-0909 June 1999 Ack Receipt of ,Transmitting Changes to Facility Emergency Plan,Rev 25,under Provisions of 10CFR50,App E, Section V IR 05000313/19993011999-06-0909 June 1999 Discusses Arrangements for Administration of Licensing Exam During Wk of 991213,per Telcon of 990602.As Agreed,Exams Repts 50-313/99-301 & 50-368/99-301 Will Be Prepared Based on Guidelines in Rev 8 of NUREG-1021 ML20195F1631999-06-0808 June 1999 Forwards Insp Repts 50-313/99-06 & 50-368/99-06 on 990524-28.Violation Identified & Being Treated as Noncited Violation ML20207G3111999-06-0707 June 1999 Ack Receipt of Changes to ANO EP Implementing Prcoedure 1903.010,Emergency Action Level Classification,Rev 34 PC-2, Received on 981218,under 10CFR50,App E,Section V Provisions. No Violations Identified ML20207G7951999-06-0707 June 1999 Forwards Notice of Violation Re Investigation Rept A4-1998-042 Re Apparent Violation Involving Initialing Record to Indicate Attendance at Required Reactor Simulator Training Session Not Attended ML20207E7131999-06-0202 June 1999 Discusses EOI 990401 Request for Alternative to Requirements of Iwl for Arkansas Nuclear One,Pursuant to 10CFR50.55a(g)(6)(ii)(B) & ASME BPV Code Section XI & Forwards Safety Evaluation Accepting Proposed Alternative ML20207B9521999-05-26026 May 1999 Discusses GL 98-04, Potential for Degradation of ECCS & CSS After LOCA Because of Const & Protective Coating Deficiencies & Foreign Matl in Containment. Staff Will Conduct Limited Survey in to Identify Sampling ML20207B4171999-05-24024 May 1999 Forwards Corrected Cover Ltr to Insp Repts 50-313/99-07 & 50-368/99-07 Issued 990514 with Incorrect Insp Closing Date ML20207A7771999-05-24024 May 1999 Forwards Insp Repts 50-313/98-21 & 50-368/98-21 on 981116-990406.One Violation of NRC Requirements Occurred & Being Treated as Noncited Violation,Consistent with App C of Enforcement Policy ML20206U4541999-05-17017 May 1999 Discusses Util & Suppl Re Changes to License NPF-06,App a TSs Bases Section.Staff Offers No Objection to These Bases Changes.Affected Bases Pages,B 202, B 2-4,B 2-7,B 3/4 2-1,B 3/4 2-3 & B 3/4 6-4,encl ML20206S4721999-05-14014 May 1999 Forwards Insp Repts 50-313/99-07 & 50-368/99-07 on 990426- 30.No Violations Noted.However,Nrc Requests That Util Provide Evaluation of Licensee Provisions to Maintain Adequate Level of Response Force Personnel on-site ML20207B4271999-05-14014 May 1999 Corrected Ltr Forwarding Insp Repts 50-313/99-07 & 50-368/99-07 on 990426-30.No Violations Noted.Areas Examined During Insp Included Portions of Physical Security Program ML20206R4741999-05-13013 May 1999 Informs That Staff Reviewed Draft Operation Insp Rept for Farley Nuclear Station Cooling Water Pond Dam & Concurs with FERC Findings.Any Significant Changes Made Prior to Issuance of Final Rept Should Be Discussed with NRC ML20206N7011999-05-12012 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization, Division of Licensing Project Management Created ML20206M7581999-05-11011 May 1999 Forwards SE Accepting Relief Request from ASME Code Section XI Requirements for Plant,Units 1 & 2 ML20206S1761999-05-11011 May 1999 Responds to Informing of Changes in Medical Condition & Recommending License Restriction for Senior Reactor Operator.No Change Was Determined in Current License Conditions for Individual ML20206N4161999-05-11011 May 1999 Discusses Completion of Licensing Action for GL 96-01, Testing of Safety-related Logic Circuits, for Plant,Units 1 & 2 ML20206S4211999-05-10010 May 1999 Forwards Insp Repts 50-313/99-04 & 50-368/99-04 on 990228- 0410.Four Violations of NRC Requirements Identified & Being Treated as Noncited Violations Consistent with App C of Enforcement Policy ML20206H1031999-05-0606 May 1999 Forwards Results of Gfes of Written Operator Licensing Exam, Administered on 990407,to Nominated Employees of Facility. Requests That Training Dept Forward Individual Answer Sheet & Results to Appropriate Individuals.Without Encl ML20206F0611999-04-29029 April 1999 Forwards SE Accepting Licensee Re ISI Plan for Third 10-year Interval & Associated Requests for Alternatives for Plant,Unit 1 ML20205R6331999-04-20020 April 1999 Ack Receipt of Which Transmitted Rev 39 to ANO Industrial Security Plan,Submitted Under Provisions of 10CFR50.54(p).No NRC Approval Is Required,Since Util Determined Changes Do Not Decrease Effectiveness of Plan ML20205P4641999-04-15015 April 1999 Forwards for Review & Comment Draft Info Notice That Describes Unanticipated Reactor Water Draindown at Quad Cities Nuclear Power Station Unit 2,Arkansas Nuclear One Unit 2 & Ja Fitzpatrick NPP ML20205N7251999-04-13013 April 1999 Forwards Summary of 990408 Meeting with EOI in Jackson, Mississippi Re EOI Annual Performance Assessment of Facilities & Other Issues of Mutual Interest.List of Meeting Attendees & Licensee Presentation Slides Encl ML20205M6881999-04-12012 April 1999 Forwards Safety Evaluation on Second 10-year Interval Inservice Insp Request Relief 96-005 ML20205L7711999-04-0909 April 1999 Forwards Insp Repts 50-313/99-03 & 50-368/99-03 on 990202- 17.No Violations Noted ML20205K7681999-04-0606 April 1999 Forwards RAI Re risk-informed Alternative to Certain Requirements of ASME Code 11,table IWB-2500-1 ML20205G8871999-04-0202 April 1999 Forwards RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, for Plant, Units 1 & 2.Response Requested within 60 Days of Date of Ltr 1999-09-22
[Table view] |
Text
,9 ,,, a UNITED STATES )
- NUCLEAR REGULATORY COMMISSION
)
. . 5 t W ASHING TON, 0. C. 20655
- g August 24, 1988 i Docket No. 50-313 Mr. T. Gene Campbell i Vice President. Nuclear Operations ;
Arkansas Power and Light Company P. O. Box 551
- Little Rock. Arkansas 72203
Dear Mr. Campbell:
SUBJECT:
ARKANSAS NUCLEAR ONE. UNIT 1 (ANO-1) - RESOLUTION OF GENERIC ISSUE (GI) 124 - AUXILIARY FEEDWATER SYSTEM RELIABILITY (TAC NO. 68188)
Anauxiliaryfeedwater(AFW)systemreviewhasbeenconductedtoassessthe overall reliability for each of seven plants with a two train AFW system under l
- GI-124. Auxiliary Feedwater System Reliaoility. This effort includes a plant- '
specific review and an en-site audit of the AFW system, and calculated estimates
- of the reliability of the AFW system given various initiating events. The staff utilized this approach to resolve GI-124 rather than a strictly analytical l approach because it believed that a first-hand audit of the AFW system design and operation more directly addressed the root causes of AFW system unrvailability l 1
and unreliability.
In general, the resolution approach adopted by the staff relied on an a,iit of several parameters that affect the availability and reliability of the ka'
, 1 i
system. These parameters include design configurations; maintenance, surveil- !
lance, and testing procedures and practices; operating procedures; personnel 1 training; system layout; operating experience; instrumentation art control; and environment and accessibility for operator recovery actions following potential malfunctions. The Standar reliabilitycriterion(10"gReviewPlan(SRP)Section10.4.9AFWsystemnumerical to 10'S per demand) served as the bisis for concluding that the AFW system in the seven plants of concern was acceptably reliable.
Because the SRP criterion specifies consideration of compensating factors such as other reliable decay heat removal methods to ,iustify a larger AFW system
, unavailabilii;y the Gff evaluated compensatory features as part of its
- effort.
A detailed review of malatenance, procedures and training was not conducted for ANO-1 since licensee programs and practices in these areas are the same as those for ANO-2 which was previously evaluated in detail for GI-124 resolution.
The licensee satisfactorily addressed the issues in these areas during the !
ANO-2 GI-124 review. Further, a detailed review of AFW instrumentation and !
control was not conducted because of past staff reviews of the emergency !
feedwater initiation and control system at Crystal River and Rancho Seco wtiich are very similar to that at ANO-1.
I h
B'300100 000024 '
p ADOCK 05000313 It PNV l
'- I Mr. T. Gene Campbell !
Eased on its review, the staff concludes that the ANO-1 AFW system, in conjunc-tion with the Startup feedwater pump as a corpensatory decay heat removal l!
feature, provides sufficient reliability to meet the unavailability criterion of SRP Section 10.4.9 for the more frequently occurring transients such as loss of :
main feedwater, and therefore, this issue is considered resolved for ANO-1. !
The report docunenting the staff review under GI-124 is enclosed for your !-
infortation-Sincerely, j
\
\T !t l C. Craig Harbuck, Project Manager l Project Directorate - IV l
! Division of Reactor Projects - III, [
IV, V and Special Projects
Enclosure:
As stated j cc w/encicsure: I See next page l
i I
i DISiRIBUTION !
-ITocket File !
NRC F0R !
Local PCR PD4 Reading f
L. Rubenstein ,
J. Cabo !
P. Noonan ;
C. Harbuck OGC-Rockville l E. Jordan >
B. Grires i ACRS(10) !
PD4 Plant File ;
1 FD4/L4ht PD4/FM M PD4/D TN i FNcenin tuck:sr JCalvo Cty{lt8 CHarhlE8 08$ G8/Lyl88 l
e Mr. T. Gene Carpbell ,
Eased on its review, the staff concludes that the ANO-1 AFW system, in conjunc-tion with the startup feedwater purp as a conpensatory decay heat removal 4 feature, provides sufficient reliability to meet the unavailability criterion of SRP Section 10.4.9 for the more frequently occurring transients such as loss of r.ain feedwater, and therefore, this issue is considered resolved for ANO-1.
The report documenting the staff review under GI-124 is enclosed for your inferr.a tion.
Sincerely.
G 1 C. Craig Harbuck, Project Manager Project Directorate - IV Division of Reactor Projects - !!!,
IV, Y and Special Projects
Enclosure:
As stated cc W/ enclosure:
See next page !
. l l
l l
1 I
i 1
e i
i
Mr. T. Gene Campbell Arkansas Power & Light Ccepany Arkansas Nuclear One Unit 1 -
CC: t Mr. Dan R. Howard, Manager Licensing Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 Mr. James M. Levine, Executive Director Nuclear Oserations Arkansas Nuclear One P. O. Box 608 Russellville, Arkansas 72801 Mr. Nicholas S. Reynolds ,
Bishop. Cook, Purcell & Reynolds i 1400 L Street, N.W.
Washington, D.C. 20005-3502 Mr. Robert B. Eersum Babcock & Wilcox Nuclear Power Generation Division ,
1700 Rockville Pike, Suite 525 Rockville, Maryland 20852 ,
Resident inspector ,
U.S. Nuclear Regulatory Comission i 1 Nuclear Plant Road Russellville Arkansas 72801 ci Re$.onal U. Administrator.
Nuclear Region IV Regulatory Comission Office of Executive Director for Operations 611 Ryan Plaza Drive Suite 1000 l Arlington, Texas 76011 j l
Mr. Frank Wilson, Director :
Division of Environmental Health i Protection De>artment of Health Ar(ansas Department of Pealth 4815 West Harb> treet Little Rock, v is 72201 Honorable V iiw sernathy County Juon P 4 County Pope County .ious e Russellville, Arkansas 72801 l
)
ARKANSAS NUCLEAR ONE, UNIT - 1 AUXILIARY FEE 0 WATER SYSTEM RELIABILITY ASSESSFENT A. Sumary and Conclusions This repo.t contains the staff's assessment of the overall reliability of the auxiliary feedwater system (AFWS) for Arkansas Nuclear One, Unit 1 (ANO-1).
This review was perfomed in ccnnection with resolution of Generic Issue (GI) 124. "Auxiliary Feedwater System Reliability " which addresses AFWS reliability in certain plants.
AFWS reliability analyses indicated that many plants fell in the high reliability range as defined by the staff in the Standard Review Plan, however, several plants fell in the lower reliability range. While these plants met applicable licensino requirements for the AFWS, their system reliability was still in question. Som licensees for this latter grcup of plants irplemented modificatters to increast AFWS reliability to an acceptable range. However, AFWS reliability for seven plants rerM ned questionable. The plants in this category are Ah0-1 and 2, Crystal River, Ft. Calhoun, Prairie Island Units I and 2, and Rancho Seco.
The objective of the review under GI-124 is to evaluate the AFWS reliability for these seven plants and to document any recomendations for further licensee BCtions.
The resolution apprcach adopted by the staff in its review of ANO-1 relied on an audit of several plant features that affect the availability and reliability of the AFW system in addition tc an assessment of numerical unavailability.
Thesevariablesincludedesignconfiguraticts;raintenance,surveillanceand testing procedures and practices; crerating precedures; personnel training; operating experierce; instrumentation and controit and environment and acces-sibility for operator recovery actions fo The AFWS numerical reliability criterion (10'}1owing~gotential to 10 per demand) malfunctions.
given in Section 10,4.9 of the Standard Review Plan (SRP) served as the basis for evaluating the AFWS in the seven plants of concern. The SRP criterion specifica stat "An acceptable AFWS should have an unavailability in the range of 10~}']to 10'gs:
per demand based on an analysis usi.9 rethods and data : resented in h0 REG-0611 and NUREG-0635. Compensating factors such as other met 1ods of accomplishing the safety functions of the AFKS or other reliable rethods for cooling ti; aactor core during abnomal conditions may be considered te justify a larger uaavail-ability of the AFVS." For the plants under consideration in GI-124, the focus of the cencern for adequate AFWS reliability was on the more frequently occurring challenges to the the system such as loss of main feedwater. Further, because the SRP criterion specifies censideration of compensating factors such as the availability of other reliable decay heat removal methods to justify a larger AFVS unavailability, an evaluation of compensatory features was also conducted.
When detemining whether or not to give credit for corpensatory decay heat removal features, the staff position has been and continues to be that only features which relate to secondary side decay htat removal capability (e.g. a startup feedwater pump. AFV purp discharge crossconnections between units, or a third AFW pump) can be considered acceptable for satisfying the SRP criterien.
While the staff recognizes the capability to remove decay heat in the "feed-and-bleed" mode utilizing the primary system safety / relief valves and high pressure
injection purps, such a method involves large uncertainties in operator respense.
Therefore, it is considered to be a suitable backup to the AFWS in emergency procedures as a last resort for decay heat removal, but is not sufficiently reliable to justify it as a compensatory feature in order to rett the SRP goal for AFWS reliability.
The staff did not undertake a atailed review of maintenance, operations (emergency procedures) and training at ANO-1. The licensee's practfees in these areas for ANO-1 are essentially the same as those previously covered in the review of GI-124 for ANO-2. 86cause the outstanding issues in these areas were resolved as discussed in the G1-124 report for ANO-2, no further discussion is included in this report.
It should also be noted that a separate section is not included on instrurentation and control. This is because the emergency feedwater initiation and control (EF10) system at ANO-1 is very similar to the same syster provided at Crystal River and Rancho Seco which have already received extensive detailed staff review. EFIC is a fully safety-related system for automatic AFW system initiation and control and meets staff criteria under item II.E.1.2 of NUREG-0737.
The licensee indicated to the staff during discussions regarding resolution of GI-124 that substantial modifications and improveu nts have been made to .
the AFW syt, tem since 1980 (see Fi vre 3). These include provision for a back-up suction supply from the service water system, replacement of the AFW pump turbine driver and associated turbine centrol improtements, installation of the safety related emergency feedwater initiation and control (EFIC) systen and associated 0TSG instrumentation, installation of new AFW suction and ,
discharge piping and valves, installation of safety related flow indication, i and installation of the "Q" (safety-related) condensate storage tank. These improvements were r4de as part of the post-TMl upgrades to the system imposed by the staff under items II.E.1.1 and II.E.1.2 of NUREG-0737, Clarification of TIM Action Plant Requirements. Some of the miodifications were not a result of specific NRC requirements but resulted from the licensee's reccgnition of the importance cf reliable AFW capability. The staff concurs with tha licensee that the redifications have improved AFW system reliability.
However, as discussed subsequently in this report, despite the above modifica-tions, the licensee concluded that the auxiliary feedwater system numerical unavai-a review of 1 ability will not compe.nsating meet for features thedecay explicit heatSRP criterion.
removal Consequently,licenset was performed. The indicated that the startup feedwater pump provides an additional means to deliver water to the steam generators in the event of a loss of main and Nr.iliary (emergency)feedwater. Use of this purp is clearly discussed in ;. lent emergeticy operating procedure No. 1202-01. The licensee also noted that the high hesi.
safetyinjectionpupsprovideacapabilitytorecovesufficientdecayheatin the "feed-and-bleed mode. Based on staff review of the startup feedwater pump, the staff concludes that it serves as a sufficiently reliable carpensatory decay heat removal feature and, when considered in conjunction with the AFW system, adequately reliable secondary side decay heat removal capability is derenstrated. Therefore, GI-174 is resolved for ANO-1. Additional discussion of the startup feedwater pump is provided in Section F. of this report.
- - - - ~
1
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. 1 I
The staff also finds that the AFU system design and operation adequately I consider other staff generic concerns raised within GI-124 (i.e., GI-68 with respect to environmental qualifications of the motor driven AFW pump, GI-93 l with respect to steam binding of the AFW pumps, GI-122.1.a,b, and c with respect to isolation valve failure, and interruption and recovery of AFW flow, GI-122.2 with respect to initiation of "feed-and-bleed," and GI-125.II.1.b with respect to single failure protection).
B. Introduction This report discusses the staff's assessment of the Auxiliary Feedwater System (AFWS) reliability for Arkansas Nuclear One, Unit 1 (ANO-1). This review was done in connection with the resolution of of Generic Issue (GI) 124 GI-124 "Auxiliary Feedwater System Reliability," addresses the reliability of the AFWS in certain plants. Reliability analyses
- for AFWSs indicated that many plants fell in the high reliability range as defined in the Standard Review Plan, NUREG-0800 However, several plants fell in the lower reliability ranges. Licensees for some of these plants implemented sufficient modifications to increase their AFUS reliability to an acceptable range. However, the reliability of the AFWS for seven plants, including ANO-1, remained questionable. The six other plants are ANC-2, Crystal River, Ft. Calhoun, Prairie Island, Units 1 and 2, and Rancho Seco.
The objective o' this task is to determine whether the AFWS of each of the subject seven plants is sufficiently reliable and to docurrent any recomrren-dations for further licensee or staff actions.
This report presents the issue resolution approach and evaluation philosophy in Section C, and detailed evaluations in Sections 0,E and F. The summary and conclusions are presented in Section A of this report.
C. Resolution Approach The staff believes that a high degree of availability and reliability for the AFWS can only be achieved if such a system is adequately designed, properly maintained and well operated. Proper maintenance and operating practices help reduce component failures. These practices are enhanced by good training programs for the maintenance and operations personnel. Good training programs also help the operations personnel understand the system's capabilities and its importance to safety. System understanding reduces failure due to maloperation of equipment and improves the likelihood of recovery in case of unanticipated component failures.
As indicated previously, detailed reviews of maintenance, emergency procedures and training as they relate to the ANO-1 AFP system were not conducted as part
- NUREG-0611, and NUREG-0635, Generic Evaluation of Feedwater Transients and Small Break LOCAs in Westinghouse and CE Designed Plants, respectively, and NRC rremoranda from A. Thadani to 0. Parr dated October 17, 1983, October 23, 1983, and November 9, 1984
of the specific review ' r ANO-1 because of the applicability of the plant practices and staff re.abw in these areas previously completed under the GI-124 review for ANO-2. Thus, no discussion is provided in this report related to these aspects of the AFW system. Specific discussion is provided on the systemdesignandconfiguration(SectionD),systemwalkdown(SectionE),and operating experience and reliability analysis (Section F).
The approach to resolution adopted by the staff as previously indicated is based on a deterministic assessment of the AFW system design, maintenance and operation in order to ensure its optimum availablity and performance. The sysym numet(cal unavailability is then compared against the SRP criterion (10 to 10 ~ per demand), and con.;ideration of appropriate reliable compensa-tory decay heat removal features is included as necessary. The specifics of this review are provided in subsequent sections of this report. ;
D. Design and Configuration The staff conducted a review of the design and configuration of the ANO-1 AFW system. The staff met with the licensee to discuss the ANO-1 AFW system design and its compliance with the criteria of Standard Review Plan Section 10.4.9.
A walk-down of the AFW system was also conducted by the staff to verify that the as-built configuration was in accordance with the design.
1 ANO-1 is a Babcock and Wilcox designed reactor, with two once-through steam i generators, two FFW trains (each with a turbine-driven pump), and three l motor-driven condensate pumps. In addition, a startup feedwater pump is ;
provided for use during normal plant startup and shutdown. The reactor is 1 located in a large dry reinforced concrete containment. The plant is provided l with two 100% capacity diesel generators for power to shutdown cooling systems if offsite power is lost.
The ANO-1 AFW system is a two-train system (refer Figure 4 for a system design summary). One train contains a centrifugal pump driven by an electric motor ;
(P78) and the other train contains a steam turbine driven pump (P7A). Diversity l in pump drivers eliminates common mode failures in the AFWS motive power. The l AFW system configuration and turbine steam supply is shown in Figures land 2 l respectively. The pumps, P7A and P78, are identical. At rated flow, each pump is capable of providing a minimum of 720 gpm which is sufficient for removing i decay neat loads in excess of 3 percent of rated thermal power. The plant is i also equipped with a steam bypass system. Each of the is equipped with an atmospheric steam dump valve (ADV) plant's upstream two of the steam main lines steam isolationvalve(MSIV). This arrangement makes the dump valves operable even if the MSIVs are closed. The electric motor driver is capable of being powered from the B emergency diesel generator. The valves in the turbine driven train to each OTSG are de powered to ensure AFW flow in the event of a loss of all ac power.
The steam turbine driver for P7A is a single stage, solid wheel, non condensing, horizontal, split cast Terry turbine unit. It is designed for variable speed operation and is equipped with an electrohydraulic actuator for speed control, an overspeed trip mechanism, and an integral trip throttle valve. It is also
. - designed for rapid starting and will operate with steam generator pressures ranging from 1,100 psia to 60 psia. An electronic speed control system with a ramp feature and step open feature on the steam admission valves are provided to reduce the possibility of overspeeding the turbine. The steam admission valves are located outside the AFW pump room, and therefore, the steam supply line is not normally pressurized thereby precluding the need for high energy line break protection. Steam can be supplied to the turbine driver from either or both steam headers. Steam traps are installed on the turbine steam supply lines to continucusly remove any condensate. The turbine exhausts to the atmosphere.
Cooling water for the turbine lube oil cooler is piped from the pump suction.
Pump and motor bearings do not require auxiliary cooling.
The AFWS is not used for normal plant startup or shutdown, but is on standby for emergency conditions. Suction for the AFW pumps is provided by the seismic category I "Q" condensate storage tank through redundant locked open manual valves. A backup supply is also available from the nonseismic condensate storage tank through a locked closed manual valve and from each loop of the seismic Category I service water system through redundant normally closed motor operated valves. A minimum of 160,000 gallons of water in the "Q" CST is required to be available by technical specifications. In addition, the "Q" CST is prcvided with a tornado missile barrier wall which protects e minimum of 30 minutes of AFV supply to permit time for the operator to transfer manually to the service water backup in the event of CST failure in a tornado. A pump minimum flow recirculation line is provided with an orifice which returns to the CST for pump protection. In addition, a full flow recirculation test line to the CST is also provided. The valves in the test line are normally closed and close on receipt of an AFW initiation signal.
The AFWS discharge piping and valving arrangement is designed to allow either pump to supply water to either or both steam generators. The discharge line valves are normally open. Each line to each steam generator is provided with a de powered solenoid operated ficw control valve and an AC powered motor opera-ted valve to ensure isolation of a faulted steam generator, and feed flow to the intact steam generator as required during emergency operation following a postu-lated main steam or feedwater line break. The valves are powered from redundant supplies thus ensuring flow in the event of a single failure. Backleakage of steam / hot water in the AFW lines is prevented by three check valves in series.
The AFW discharge lines are also checked for high temperature each shift to ensure tha; unacceptable backleakage is not occurring and potential steam binding of the pumps is avoided.
Automatic initiation and control of AFWS is provided by the emergency feedwater initiation and control (EFIC) system (refer to Figures 5, 6 and 7). EFIC is a safety related control system which ensures initiation and continuous control of auxiliary feedwater flow to an intact steam generator in the event of loss of main feedwater, loss of reactor coolant pumps (loss of offsite power) or low level in an OTSG. EFIC also controls the steam exhaust path by modulating the atmospheric dump valves. The ANO-1 EFIC system is very similar to that provided at Crystal River and Rancho Seco. Details on the EFIC design are provided in the Crystal River GI-124 report, and the Rancho Seco restart safety evaluation, NUREG-1286.
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. . In r.he event of a complete loss of the AFW system, other methods are available ,
to remove decay heat, including (1) use of the startup feedwater pump, (2) i condensate pumps and (3) "feed-and-bleed." Use of these alternate decay heat .
removal methods is addressed in the plant emergency procedure. I l
On the basis of this evaluation, the staff concludes that the AFW system at l ANO-1 complies with the applicable criteria of Section 10.4.9 of the Standard l Review Plan, including the guidelines of NUREG-0737, Item II.E.1.1. The staff notes that use of the startup feedwater pump, condensate pump, and "feed-and-bleed" capability are effective means of decay heat removal and enhance the plant's overall capability of decay heat removal. Use of the startup feedwater pump as a compensatory decay heat removal feature has been considered and is discussed further in this report.
1 E. Syste.m Walkdown As part of the staff's review, a site visit and AFW system walkdown was conducted.
The walkdcwn afforded the staff the opportunity to examine the as-built system configuration, specific components, and potential for undesirable system inter-actions. The system walkdown had two main objectives. One was to confirm that the installed system conformed to the staff's understanding of the system design basis as identified in previous evaluations, and to determine if the system may be subject to conotn node failure nechanisms or harards (e.g. , flooding, fire, 1 missiles,suctionstrainers,etc.). The other main objective was to examine the ease of operator access to equipment for performing potential recovery actions. ,
This includes assessment of local emergency lighting, connunications, and other I factors (e.g., cleanliness, equipment labeling, use of locking devices, posting of simple instructions at equipment locations, etc).
The walkdown covered the piping and component layout from the condensate storage tanks, through the pumps to the containment penetration and included the turbine I driven AFW pump steam supply lines, switchgear, and the instrumentation ar.d con- l trol provided in the control room. Based on the walkdown, the staff identified no areas of concern regarding the as-built AFW system configuration, common mode failure potential, or ease of access for recovery actions.
F. Operating Experience _ and_ Reliability _ Analysis
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As part of the staff review under GI-124 for ANO-1, the staff discussed AFW component failure history, feedwater transient experience and AFW system numerical reliability evaluation with the licensee.
Since 1984, only two AFW component failures have occurred, both in 1985 (refer l to Figure 9). Prior to that time, five random component failures occurred in l 1983 and six in 1981. Four of the six failures involved the turbine driven pump, but these were corrected with the installation of a new turbine driver in 1982. The recent experience indicates that the licensee prograns for ensuring AFW component operability are effective as no failures have occurred I since 1986.
A review of unanticipated reactor trip experience for ANO-1 since 1981 indicated '
a high of 8 per year in 1983 and 1985, but only two per year in 1986 and 1987, and one in the first six months of 1988 (refer to Figure 10). Further, with
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. . the exception of 1985, an average of less than two trips per year relate to main feedwater system upsets. The six main feedwater upsets which occurred in 1985 are attributable to startup difficulties experienced with installation and tuning of the EFIC system. The licensee indicated their consnitment to improve main feedwater system performance and reduce reactor trips as part of their efforts during implementation of the B&W Owners Group Safety Performance Improve-ment Program reconsendations. The staff concludes that the most recent trip experience and continued licensee efforts have/will contribute to reduced chal-lenges to the AFW system.
As was stated previously, the licensee indicated that a update of the relia-bilityanalysisforthecurrentAN0-1AFWsystemaccountingfgrtherecent modifications resulted in a numerical unavailability of 4X10 per demand for a loss of main feedwater using the NUREG-0611 data base and methodology as specified in the SRP criter This value exceeds the acceptancecriterionof10jon(refertoFigure8).
per demand and thus necessitated consideration of other compensatory decay heat removal features. To address compensatory features, the licensee pointed out that the existing Emergency Procedure 1202.01 section dealing with overheating identifies a hierarcy for decay heat removal capability following transients and accidents (refer to Figure 11). Included in this pro-cedure as the first means following a loss of main feedwater is the startup feed-water pump. This pump provides full AFW flow at above the normal secondary side pressure when operated in series with a concensate pump. It serves as the normal means of plant startup and shutdown. This capability is available in addition to use of a condensate pump alone upon OTSG depressurization, use of a service water pump for service water addition to the OTSG upon depressuriza-tien, and "feed and bleed" cooling. "Feed and bleed" cooling requires only a single high pressure injection pump since it has the capability to lift the primary system safety valves.
Based on the above, the staff concludes that the startup feedwater pump serves as a suitably reliable compensatory feature for decay heat removal to justify a calculi.ed r AFW system unavailability lower than the SRP acceptance criterion, and therefore, the AFW system supplemented by the startup pump meets the SRP numerical reliability criterion for the more frequently occurring transients such as loss of main feedwater. The staff, therefore, considers GI-124 resolved for ANO-1.
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