Letter Sequence Other |
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MONTHYEARML20058M4021993-12-0808 December 1993 Forwards TIA,93TIA006,interpretation of Reporting Requirements Re Multiple Failures of safety-related Components Project stage: Other ML20058L6931993-12-0808 December 1993 Forwards TIA 93TIA006,interpretation of Reporting Requirements Re Multiple Failures of safety-related Components Project stage: Other ML20058L0091993-12-0808 December 1993 Forwards NRR 931102 Response to Region IV 930413 Request for Interpretation of Reporting Requirements Re Multiple Failures of safety-related Components Identified During Performance of Surveillance Testing Project stage: Other ML20058K9811993-12-0808 December 1993 Forwards NRR 931102 Response to Region IV Request for Interpretation of Reporting Requirements Re Multiple Failures of safety-related Components Identified During Performance of Surveillance Testing for Info Project stage: Other ML20058K6001993-12-0808 December 1993 Forwards Tia,Interpretation of Reporting Requirements, 93TIA006 Re Multiple Failures of safety-related Components Project stage: Other ML20058K5961993-12-0808 December 1993 Forwards Tia,Interpretation of Reporting Requirements - 93TIA006,re Multiple Failures of safety-related Components Project stage: Other ML20058K5801993-12-0808 December 1993 Forwards Tia,Interpretation of Reporting Requirements - 93TIA006,re Multiple Failures of safety-related Components Project stage: Other ML20058K5611993-12-0808 December 1993 Forwards Tia:Interpretation of Reporting Requirements, 93TIA006.Guidance Provided in Response to Region 5 Request for Interpretation of Reporting Requirements for safety- Related Components Project stage: Other 1993-12-08
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Category:CORRESPONDENCE-LETTERS
MONTHYEARTXX-9924, Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span1999-10-22022 October 1999 Forwards Responses to Questions by NRC Re Application for Amends to Licenses NPF-87 & NPF-89,by Incorporating Changes Increasing RWST low-level Setpoint from Greater than But Equal to 40% to Greater than But Equal to 45% of Span ML20217M5711999-10-20020 October 1999 Forwards Insp Repts 50-445/99-15 & 50-446/99-15 on 990822- 1002.Two Severity Level IV Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy TXX-9923, Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred1999-10-15015 October 1999 Forwards Monthly Operating Repts for Sept 1999 for CPSES, Units 1 & 2,per Plant TS 5.6.4.No Failures of Challenges to PORVs of SV for Units Occurred ML20217E7951999-10-12012 October 1999 Forwards COLR for Unit 1,Cycle 8,per TS 5.6.5 ML20212L2891999-10-0101 October 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals. Purpose of GL Was to Obtain Info That Would Enable NRC to Verify That Condition of Licensee SG Internals Comply with Current Licensing Bases TXX-9922, Forwards Revised COLR, for Cycle 5 for Unit 21999-10-0101 October 1999 Forwards Revised COLR, for Cycle 5 for Unit 2 ML20216J5571999-10-0101 October 1999 Provides Final Response to GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps ML20212G0721999-09-24024 September 1999 Forwards Rev 4 to Augmented Inservice Insp Plan for CPSES, Unit 1. Future Changes & Revs to Unit 1 Augmented Inservice Insp Plan Will Be Available on Site ML20212H0461999-09-24024 September 1999 Forwards Rev 6 to CPSES Glen Rose,Tx ASME Section XI ISI Program Plan for 1st Interval on 990820 ML20212F7481999-09-24024 September 1999 Forwards SER Authorizing Relief from Exam Requirement of 1986 Edition ASME Code,Section XI Pursuant to 10CFR50.55a(a)(3)(ii) for Relief Request A-3 & 10CFR50.55a(g)(6)(i) for Relief Requests B15,16,17 & C-4 ML20212F1041999-09-23023 September 1999 Requests That NRC Be Informed of Any Changes in Scope of Y2K System Deficiencies Listed or Util Projected Completion Schedule for Comanche Peak Steam Electric Station,Units 1 & 2 ML20212E6661999-09-21021 September 1999 Advises That Info Contained in Application & Affidavit, (CAW-99-1342) Re WCAP-15009,Rev 0, Comache Peak Unit 1 Evaluation for Tube Vibration Induced Fatigue, Will Be Withheld from Public Disclosure ML20212D9111999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of CPSES & Did Not Identify Any Areas in Which Performance Warranted Insp Beyond Core Insp Program.Core Insp Plan at Facility Over Next 7 Months.Insp Plan Through March 2000 Encl ML20212A7601999-09-14014 September 1999 Forwards Insp Repts 50-445/99-14 & 50-446/99-14 on 990707-0821.Four Violations Occurred & Being Treated as Ncvs.Conduct of Activities Was Generally Characterized by safety-conscious Operations & Sound Radiological Controls TXX-9921, Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC1999-09-10010 September 1999 Suppls 981221 LAR 98-010 to Licenses NPF-87 & NPF-89, Clarfying Conditions of Use Re Analytical Methods Used to Determine Core Operating Limits,Per Telcon with NRC ML20211P3761999-09-0707 September 1999 Ack Receipt of Ltr Dtd 990615,transmitting Rev 30 to Physical Security Plan,Per 10CFR50.54(p).No NRC Approval Is Required ML20211L9871999-09-0303 September 1999 Forwards Rev 31 to Technical Requirements Manual. All Changes Applicable to Plants Have Been Reviewed Under Util 10CFR50.59 Process & Found Not to Include Any USQs TXX-9915, Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl1999-09-0303 September 1999 Responds to 990701 & 0825 RAI Telcons Re Spent Fuel Pool Temp,Per LAR 98-008,which Requested Increase in Spent Fuel Storage capacity.Marked-up Page 4-1 of CPSES Fuel Storage Licensing Rept, Encl ML20211K2231999-08-31031 August 1999 Forwards Txu Electric Comments of Rvid,Version 2 ML20211J3801999-08-27027 August 1999 Forwards Corrected TS Page 3.8-26 to Amend 66 to Licenses NPF-87 & NPF-89,respectively.Footnote on TS Page 3.8-26 Incorrectly Deleted ML20211G7301999-08-26026 August 1999 Forwards Revs 29 & 30 to CPSES Technical Requirements Manual (Trm). Attachments 1 & 2 Contain Description of Changes for Revs 29 & 30 Respectively ML20211G1081999-08-26026 August 1999 Responds to NRR Staff RAI Re Util Mar 1999 Submittal for NRC Review & Approval of Changes to CPSES Emergency Classification Procedure ML20211G3441999-08-25025 August 1999 Forwards Response to NRC RAI on LAR 98-010 for Cpses,Units 1 & 2.Communication Contains No New Licensing Commitments Re Cpses,Units 1 & 2 ML20211B2861999-08-18018 August 1999 Forwards Insp Repts 50-445/99-13 & 50-446/99-13 on 990720- 23.No Violations Noted.Insp Included Implementation of Licensee Emergency Plan & Procedures During Util Biennial Emergency Preparedness Exercise ML20211C4661999-08-18018 August 1999 Discusses Proprietary Info Re Thermo-Lag.NRC Treated Bisco Test Rept 748-105 as Proprietary & Withheld It from Public Disclosure,Iaw 10CFR2.790 ML20210U3981999-08-17017 August 1999 Forwards Monthly Operating Repts for July 1999 for CPSES, Units 1 & 2,per TS 6.9.1.5.No Failures or Challenges to PORVs or SVs for Plant Occurred ML20211C0991999-08-17017 August 1999 Forwards Rev 3 to ASME Section XI ISI Program Plan,Unit 2 - 1st Interval, Replacing Rev 2 in Entirety ML20211C4571999-08-16016 August 1999 Forwards Omitted Subj Page of Contractor TER TXX-9919, Forwards Relief Request A-3,Rev 1 to Unit 1 ISI Program,Per Conversations Between NRC & Txu Electric on 9908021999-08-16016 August 1999 Forwards Relief Request A-3,Rev 1 to Unit 1 ISI Program,Per Conversations Between NRC & Txu Electric on 990802 ML20210R6561999-08-13013 August 1999 Forwards Response to NRR 990805 Telcon RAI Re License Amend Request 98-010,to Increase Power for Operation of CPSES Unit 2 to 3445 Mwth & Incorporating Addl Changes Into Units 1 & 2 TS 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 ML20210S6411999-08-12012 August 1999 Informs That Wg Guldemond,License SOP-43780,is No Longer Performing Licensed Duties.Discontinuation of License Is Requested ML20210R2221999-08-12012 August 1999 Forwards Insp Repts 50-445/99-10 & 50-446/99-10 on 990510-0628.Violations Noted & Being Treated as Ncvs, Consistent with App C of Enforcement Policy ML20210N1101999-08-0404 August 1999 Provides Supplemental Info to Util 990623 License Amend Request 99-005 Re Bypassing DG Trips.Info Replaces Info Contained in Subject Submittal in Attachment 2,Section II, Description of TS Change Request 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 ML20210J2301999-08-0202 August 1999 Forwards Amend 96 to CPSES Ufsar.Replacement of FSAR Figures with Plant Process Flow Diagrams Meets Intent & Requirements of NRC Reg Guide 1.70,Rev 2 ML20210J6071999-08-0202 August 1999 Forwards line-by-line Descriptions of Changes in Amend 96 to CPSES UFSAR Transmitted by Util Ltr TXX-99166,dtd 990802. Replacment of FSAR Figures with Plant Process Flow Diagrams Meets Intent & Requirements of NRC Reg Guide 1.70,rev 2 TXX-9916, Notifies NRC That CPSES Units 1 & 2,improved TS Implemented on 9907271999-08-0202 August 1999 Notifies NRC That CPSES Units 1 & 2,improved TS Implemented on 990727 TXX-9918, Forwards CPSES 10CFR50.59 Evaluation Summary Rept 0008,for 970802-990201 & CPSES Commitment Matl Change Evaluation Rept 0003,for 970802-9906301999-08-0202 August 1999 Forwards CPSES 10CFR50.59 Evaluation Summary Rept 0008,for 970802-990201 & CPSES Commitment Matl Change Evaluation Rept 0003,for 970802-990630 ML20210K2321999-07-29029 July 1999 Forwards Insp Repts 50-445/99-12 & 50-446/99-12 on 990530-0710.No Violations Noted ML20210G5861999-07-29029 July 1999 Forwards fitness-for-duty Program Performance Data for Six Month Period of Jan-June 1999 ML20210J0121999-07-27027 July 1999 Forwards Summary of Methodology for Determination of NDE Measurement Uncertainty,In Response to Recent Discussions with NRC Re LAR 98-006 Concerning Rev to SG Tube Plugging Criteria TXX-9917, Provides Info Re Augmented Inservice Insp Plan,Which Requires Periodic Insp of Rv Head & Internals Lifting Devices at CPSES1999-07-26026 July 1999 Provides Info Re Augmented Inservice Insp Plan,Which Requires Periodic Insp of Rv Head & Internals Lifting Devices at CPSES ML20210F3121999-07-26026 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, ML20210D8231999-07-23023 July 1999 Forwards Safety Evaluation of Relief Requests Re Use of 1998 Edition of Subsections IWE & Iwl of ASME Code for Containment Insp ML20210D3211999-07-21021 July 1999 Provides List of Estimates of Licensing Actions,In Response to Administrative Ltr 99-02,dtd 990603 ML20210C2931999-07-21021 July 1999 Supplements 880323 Response to NRC Bulletin 88-02, Rapidly Propagating...Sg Tubes, Non-proprietary WCAP-15010 & Proprietary Rev 0 to WCAP-15009, CP Unit 1 Evaluation for Tube Vibration... Encl.Proprietary Rept Withheld ML20209H0111999-07-16016 July 1999 Forwards Relief Request C-4 to CPSES Unit 2 ISI Program for Approval ML20210C3331999-07-16016 July 1999 Forwards Exam Repts 50-445/99-301 & 50-446/99-301 on 990618- 24.Exam Included Evaluation of Six Applicants for Senior Operator Licenses ML20209H2551999-07-16016 July 1999 Forwards ISI Summary Rept for Fourth Refueling Outage of CPSES Unit 2 & Containment ISI Summary Rept for Fourth Refueling Outage of CPSES Unit 2,per ASME Boiler & Pressure Vessel Code,Section Xi,Paragraph IWA-6230 1999-09-07
[Table view] Category:OUTGOING CORRESPONDENCE
MONTHYEARML20217M5711999-10-20020 October 1999 Forwards Insp Repts 50-445/99-15 & 50-446/99-15 on 990822- 1002.Two Severity Level IV Violations of NRC Requirements Identified & Being Treated as non-cited Violations Consistent with App C of Enforcement Policy ML20212L2891999-10-0101 October 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals. Purpose of GL Was to Obtain Info That Would Enable NRC to Verify That Condition of Licensee SG Internals Comply with Current Licensing Bases ML20212F7481999-09-24024 September 1999 Forwards SER Authorizing Relief from Exam Requirement of 1986 Edition ASME Code,Section XI Pursuant to 10CFR50.55a(a)(3)(ii) for Relief Request A-3 & 10CFR50.55a(g)(6)(i) for Relief Requests B15,16,17 & C-4 ML20212F1041999-09-23023 September 1999 Requests That NRC Be Informed of Any Changes in Scope of Y2K System Deficiencies Listed or Util Projected Completion Schedule for Comanche Peak Steam Electric Station,Units 1 & 2 ML20212E6661999-09-21021 September 1999 Advises That Info Contained in Application & Affidavit, (CAW-99-1342) Re WCAP-15009,Rev 0, Comache Peak Unit 1 Evaluation for Tube Vibration Induced Fatigue, Will Be Withheld from Public Disclosure ML20212D9111999-09-16016 September 1999 Informs That on 990818,NRC Completed Midcycle PPR of CPSES & Did Not Identify Any Areas in Which Performance Warranted Insp Beyond Core Insp Program.Core Insp Plan at Facility Over Next 7 Months.Insp Plan Through March 2000 Encl ML20212A7601999-09-14014 September 1999 Forwards Insp Repts 50-445/99-14 & 50-446/99-14 on 990707-0821.Four Violations Occurred & Being Treated as Ncvs.Conduct of Activities Was Generally Characterized by safety-conscious Operations & Sound Radiological Controls ML20211P3761999-09-0707 September 1999 Ack Receipt of Ltr Dtd 990615,transmitting Rev 30 to Physical Security Plan,Per 10CFR50.54(p).No NRC Approval Is Required ML20211J3801999-08-27027 August 1999 Forwards Corrected TS Page 3.8-26 to Amend 66 to Licenses NPF-87 & NPF-89,respectively.Footnote on TS Page 3.8-26 Incorrectly Deleted ML20211B2861999-08-18018 August 1999 Forwards Insp Repts 50-445/99-13 & 50-446/99-13 on 990720- 23.No Violations Noted.Insp Included Implementation of Licensee Emergency Plan & Procedures During Util Biennial Emergency Preparedness Exercise ML20211C4661999-08-18018 August 1999 Discusses Proprietary Info Re Thermo-Lag.NRC Treated Bisco Test Rept 748-105 as Proprietary & Withheld It from Public Disclosure,Iaw 10CFR2.790 ML20211C4571999-08-16016 August 1999 Forwards Omitted Subj Page of Contractor TER ML20210R2221999-08-12012 August 1999 Forwards Insp Repts 50-445/99-10 & 50-446/99-10 on 990510-0628.Violations Noted & Being Treated as Ncvs, Consistent with App C of Enforcement Policy 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 ML20210K2321999-07-29029 July 1999 Forwards Insp Repts 50-445/99-12 & 50-446/99-12 on 990530-0710.No Violations Noted ML20210D8231999-07-23023 July 1999 Forwards Safety Evaluation of Relief Requests Re Use of 1998 Edition of Subsections IWE & Iwl of ASME Code for Containment Insp ML20210C3331999-07-16016 July 1999 Forwards Exam Repts 50-445/99-301 & 50-446/99-301 on 990618- 24.Exam Included Evaluation of Six Applicants for Senior Operator Licenses ML20209H7501999-07-15015 July 1999 Forwards Safety Evaluation on GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves, for Comanche Peak Steam Electric Station,Units 1 & 2 ML20209G7421999-07-0808 July 1999 Forwards SER Concluding That Licensee Individual Plant Exam of External Events Process Capable of Identifying Most Likely Severe Accidents & Severe Accident Vulnerabilities & IPEEE Met Intent of Supp 4 to GL 88-20 ML20196L0121999-07-0808 July 1999 Forwards Safety Evaluation Granting First 10-Year Interval Inservice Insp Requests for Relief B-6 (Rev 2),B-7 (Rev 2), B-12,B-13,B-14 & C-9,pursuant to Tile 10CFR50.55a(g)(6)(i) ML20196K6771999-07-0202 July 1999 Ack Receipt of & Encl Scenario for Comanche Peak Steam Electric Station Emergency Plan Exercise Scheduled for 990721-22.Determined That Exercise Scenario Sufficient to Meet Emergency Plan Requirements & Exercise Objectives ML20196J4881999-06-29029 June 1999 Informs That as Result of Staff Review of Util Responses to GL 92-01,rev 1,suppl 1,NRC Revised Info in Rvid & Releasing Rvid as Version 2 ML20196J0401999-06-29029 June 1999 Forwards Safety Evaluation Re Plant,Units 1 & 2 Proposed Changes to Emergency Plan ML20196E6641999-06-22022 June 1999 Forwards Insp Repts 50-445/99-11 & 50-446/99-11 on 990418- 0529.No Violations Noted.Licensee Conduct of Activities Generally Characterized by safety-conscious Operations,Sound Engineering & Maintenance & Acceptable Radiological Control ML20207H3801999-06-0909 June 1999 Forwards Insp Repts 50-445/99-08 & 50-446/99-08 on 990503-11.Violations Identified & Being Treated as Noncited Violations ML20195G3771999-06-0909 June 1999 Ack Receipt of Ltr & Encl Objectives for Comanche Peak Steam Electric Station Emergency Plan Exercise Scheduled for 990721.Based on Review,Nrc Determined That Exercise Objectives,Appropriate to Meet Plan Requirements ML20207G3291999-06-0707 June 1999 Ack Receipt of Which Transmitted Rev 27 to Comanche Peak Steam Electric Station EP Under Provisions of 10CFR50.54(q).Based on Determination That Changes Do Not Decrease Effectiveness of EP No NRC Approval Required ML20207E9291999-06-0202 June 1999 Discusses 990526 Request That USNRC Exercise Discretion Not to Enforce Compliance with TS 4.8.2.1e Re Performance of Battery Performance Discharge Test,In Lieu of Battery Svc Test.Concludes Action Satisfactory & Discretion Exercised ML20207D7111999-05-28028 May 1999 Advises That Info Contained in Licensee 990514 Submittal Re License Amend Request 98-01-0 Will Be Withheld from Public Disclosure,Per 10CFR2.790. 10CFR2.790 ML20207D7011999-05-27027 May 1999 Advises That Info Contained in TU Electric 990514 Submittal (TXX-99115) Re License Amend Request 98-010 Will Be Withheld from Public Disclosure (Ref 10CFR2.790),per 990511 Application & Affidavit ML20207B7241999-05-25025 May 1999 Advises That Info Contained in Application & Affidavit 990507 (CAW-99-1333),submitting WCAP-15004,dtd Dec 1997,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20206Q0031999-05-14014 May 1999 Forwards Safety Evaluation Accepting Licensee Response to GL 92-08, Thermo-Lag 330-1 Fire Barriers, Dtd 921217,for Comanche Peak Steam Electric Station,Unit 1 ML20206P5961999-05-12012 May 1999 Forwards Insp Repts 50-445/99-09 & 50-446/99-09 on 990419- 23.No Violations Noted.Nrc Determined That Releases of Radioactive Waste Effluents Controlled,Monitored & Quantified Well ML20206N7061999-05-12012 May 1999 Informs That NRC Ofc of NRR Reorganized,Effective 990328. Reorganization Chart Encl ML20206S5841999-05-11011 May 1999 Forwards Insp Repts 50-445/99-07 & 50-446/99-07 on 990307-0417.No Violations Noted.Insp Generally Characterized by safety-conscious Operations,Sound Engineering & Maint & Acceptable Radiological Controls ML20206K0311999-05-0707 May 1999 Informs That on 990407,NRC Administered Gfes of Written Operator Licensing Exam.Licensee Facility Did Not Participate in Exam,However,Copy of Master Exam with Answer Key Encl for Info,Without Encl ML20206H1701999-05-0606 May 1999 Forwards Copy of Exemption & Safety Evaluation Supporting Requirement in App K to 10CFR50.Proposal Will Use New Feedwater Flow Measurement Sys to Allow More Accurate Measurement to Thermal Power ML20206K3931999-05-0505 May 1999 Ltr Contract,Task Order 41, Comanche Peak Safety System Engineering Insp, Under Contract NRC-03-98-021 ML20206G1221999-05-0303 May 1999 Discusses 981215 Request That Document, Responses & Further Clarifications to NRC Questions from 980929 Meeting, Be Withheld from Public Disclosure.Determined Information to Be Proprietary & Will Be Withheld from Public Disclosure ML20206F5711999-04-30030 April 1999 Forwards Insp Repts 50-445/99-06 & 50-446/99-06 on 990329-0402.No Violations Noted.Insp Re Focus on Radiation Protection Program Activities During Unit 2 Refueling Outage ML20206E5211999-04-27027 April 1999 Discusses GL 96-01 Issued on 960110 & TU Responses, ,970102 & 980502 for Cpses,Units 1 & 2.Determined That Submittals Provided Both Info Requested & Responses Required by GL 96-01 ML20206B3831999-04-23023 April 1999 Forwards FEMA Final Rept for 990311,Comanche Peak Steam Electric Station Medical Drill.No Deficiencies or Areas Requiring Corrective Actions Identified ML20206B5321999-04-22022 April 1999 Ack Receipt of Ltrs Dtd 970407,09 & 0204,which Transmitted Revs 6 & 7 to Safeguards Continency Plan,Rev 10 to Security Training & Qualification Plan & Rev 29 to Physical Security Plan Submitted Under Provisions of 10CFR50.54(p) ML20206A2301999-04-14014 April 1999 Refers to Public Meeting Conducted on 990329 in Glen Rose, Tx Re Results of Plant Performance Review Completed on 990211 & Transmitted to Licensee on 990319.List of Attendees Encl ML20205L8711999-04-0707 April 1999 Forwards Insp Repts 50-445/99-03 & 50-446/99-03 on 990124-0306.No Violations Were Identified.Review of Operability Evaluation Re MOVs Disclosed That Licensee Failed to Include Info About Degraded ECCS Performance ML20205L1051999-04-0606 April 1999 Informs of Completion of Review of Tuec 980312 Submittal Re GL 97-05, SG Tube Insp Techniques. No Concerns Identified with SG Insp Techniques Employed at Cpses,Units 1 & 2,that Would Indicate Noncompliance with Current Licensing Basis ML20205F9141999-04-0101 April 1999 Informs That as of 990329 Dh Jaffe Has Been Assigned as Senior Project Manager for Plant IR 05000446/19920491999-03-24024 March 1999 Discusses Concern That Postulated Fire in CR Could Create Single Hot Short in Control Circuitry of MOVs Resulting in Spurious Operation.Required Hardware Mods Implemented to Control Circuits of Affected Mov,Per Insp Rept 50-446/92-49 ML20204F3311999-03-23023 March 1999 Forwards Discussion Items for 990323 Telcon 1999-09-07
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,j nEcioN av 611 RYAN PLAZA DRIVE, SulTE 400 l
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% ~s [ AR LINGT ON, T E XAS 76011-8064 1 DEC 8 1993 Dockets: 50-445 50-446 Licenses: NPF-87 '
NPF-89 TU Electric ATTN: W. J. Cahill, Jr., Group Vice President Nuclear Engineering and Operations ,
Skyway Tower !
400 North Olive Street, L.B. 81 Dallas, Texas 75201
SUBJECT:
TASK INTERFACE AGREEMENT: INTERPRETATION OF REPORTING ,
REQUIREMENTS - 93TIA006 (TAC NO. M86339) ,
The purpose of this letter is to provide for your information a copy of the guidance recently issued by the Office of Nuclear Reactor Regulation in their .
Memorandum of November 2, 1993, to Region IV (see enclosed). This guidance -
was provided in response to a Region IV request for interpretation of' reporting requirements related to multiple failures of safety-related components that are identified during the performance of surveillance testing. ,
We plan to implement this guidance during out future inspections at your facilities. Should you have questions regarding this matter, please contact ,
Tom Westerman of my staff at 817-860-8145. .;
1
.ue .i irector Division of Reactor Safety ,
Enclosure:
(as noted) cc w/ enclosure:
TU Electric ,
ATIN: Roger D. Walker, Manager of Regulatory Affairs for Nuclear Engineering Organization Skyway Tower 400 North Olive Street, L.B. 81 Dallas, Texas 75201 Juanita Ellis President - CASE 14?6 South Polk Street '
Dallas, Texas 75224 L
9312150307 931208 C '
PDR ADOCK 05000445 M P PDR g
TV Electric i GOS Associates, Inc.
Suite 720 ;
1850 Parkway Place Marietta, Georgia 30067-8237 :
TU Electric Bethesda Licensing ;
3 Metro Center, Suite 610 ,
Bethesda, Maryland 20814 ,
Jorden, Schulte, and Burchette ATTN: William A. Burchette, Esq.
Counsel for Tex-La Electric ,
Cooperative of Texas 1025 Thomas Jefferson St., N.W. t Washington, D.C. 20007 Newman & Holtzinger, P.C.
ATTN: Jack R. Newman, Esq.
1615 L. Street, N.W.
Suite 1000 Washington, D.C. 20036 Texas Department of Licensing & Regulation ATIN: G. R. Bynog, Program Manager /
Chief Inspector Boiler Division P.O. Bcx 12157, Capitol Station ,
Austin, Texas 78711 Honorable Dale McPherson County Judge ,
P.O. Box 851 Glen Rose, Texas 76043 Texas Radiation Control Program Director 1100 West 49th Street Austin, Texas 78756 a
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TO Electric i bec to DMB (IE51) bec distrib. by RIV:
J. L. Milhoan Resident Inspector (2)
Section Chief (DRP/B) Lisa Shea, RM/ALF, MS.: HN8B 4503 MIS System DRSS-FIPS l RIV File Project Engineer (DRP/B) i Section Chief (DRP/TSS) E. Adensan, NMSS 4 E4 i
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bec distrib. by RIV: ,
J. L. Milhoan Resident Inspector (2) '
Section Chief (DRP/B) Lisa Shea, RM/ALF, MS: MNBB 4503 MIS System DRSS-FIPS RIV File Project Engineer (DRP/B)
Section Chief (DRP/TSS) E. Adensan, NMSS 4 E4 W. Reckley, NRR 13 HIS I
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f* "% UNITED STATES h,
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- ki f WASHINGTON, D.C. 2955bM
%, # November 2. 1993 MEMORANDUM FOR: Samuel J. Collins, Director :
Division of Reactor Safety Region IV FROM: Elinor G. Adensam, Assistant Director for Regions IV and V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
SUBJECT:
TASK INTERFACE AGREEMENT: INTERPRETATION OF REPORTING REQUIREMENTS - 93TIA006 (TAC N0. M86339) ,
In response t3 your request dated April 13, 1993, we have reviewed the available guidance associated with the reporting requirements related to multiple failures of safety-related components that are identified during the performance of surveillance procedures. The specific examples cited in your questions regarded the outage surveillances related to primary or secondary i safety relief valves and the discovery that the as-found setpoints were outside the allowable technical specification setpoint tolerances. Please note that the Public Document Room (PDR) has been included on the distribution -
for this response.
Licensees were stated to have presented interpretations of the reporting rules (10 CFR 50.72/50.73) and the related guidance provided in'NUREG-1022, which !
supported the conclusion that the discovery of safety valve setpoint drift was not reportable. Specifically, question 2.3 of NUREG-1022, Supplement 1, had been used to argue that the condition was not reportable, because the condition could be assumed to have occurred at the time of discovery. Another argument presented by licensees was stated to involve analyses or evaluations which determined that the degraded setpoints did not result in the plant operating outside its design basis, and therefore supported a conclusion that the condition was not reportable.
A review of 50.72 and 50.73 identifies several reporting criteria which might -
be relevant to the discovery of safety valves outside the setpoint tolerances given in the Technical Specifications. These criteria and a discussion of their applicability is provided in Enclosure 1.
r The assessment can be summarized as follows: ;
- The use of question 2.3 to NUREG-1022, Supplement 1, is not appropriate to justify a decision to not report many conditions found during ,
refueling outage surveillances. Other guidance in Supplement 1 is clear that if conditions are discovered during an outage, but are believed to have existed during operation, they are reportable so long as an applicable threshold for reporting is reached.
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l Samuel J. Collins k ,cber 2, 1993
- A licensee may determine that a condition such as safety valve setpoint drift, does not constitute operation outside the design basis of the plant, and therefore not report such events in accordance with those ,
criteria in 50.72 and 50.73. However, as discussed below, the condition may be reportable as a result of other criteria.
- 50.73(a)(2)(vii) is deemed the most relevant criterion for the reporting of primary or secondary safety valves found to be outside the acceptable setpoint tolerance. This is due to the fact that this criterion is based on the train or channel level and does not require the loss of a safety function but only the inoperability of multiple channels of a safety system. Some latitude might be given in light of the number of secondary safety valves; but, for most instances of setpoint drift, this criterion would result in the conditions being reportable. ;
- Note that we currently expect to include guidance along these lines in :
the forthcoming Revision 1 to NUREG-1022; if so, that specific guidance should be consulted in the future in determining reportability.
Elinor G. Adensam, Assistant Director '
for Regions IV and V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation
Enclosure:
i Criteria cc w/ enclosure: i W. Hodges, Region I A. Gibson, Region 11 G. Grant, Region III K. Perkins, Region V ;
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. i ENCLOSURE ASSESSMENT OF VARIOUS REPORTING REQUIREMENTS FOR APPLICABILITY TO !
PRIMARY OR SECONDARY SAFETY VALVES FOUND OUTSIDE TECHNICAL SPECIFICATION i ACCEPTABLE SETPOINT TOLERANCE BAND ,
50.72(b)(1)(ii) Any event or condition during operation that results in the :
50.73(a)(2)(ii) condition of the nuclear power plant, including its principal safety barriers, being seriously degraded; or results in the nuclear power plant being:
(A) In an unanalyzed condition that significantly !
compromises plant safety; (B) In a condition that is outside the design basis of l the plant; or -r (C) in a condition not covered by the plant's operating l and emergency procedures.
Discussion: The applicability of these criteria is determined by an -
evaluation of the situation by the licensee. Upon determining that the setpoints were outside the allowable range of the technical specifications, t!.e licensee would be ;
expected to follow the required actions of the technical !
specifications and assess the plant condition in regards to equipment operability and required corrective actions.
Guidance related to the evaluation of degraded and .
nonconforming conditions is provided by Generic Letter }
91-18. As stated in the second draft of NUREG-1022, >
Revision 1 it is expected that licensees may use engineering judgement and experience in determining whether .
a condition meets these reporting criteria. The ability of l a licensee to justify that a given condition is neither ,
unanalyzed nor outside the design basis is dependent on the as-found condition of the equipment and the degree of analyses performed.
50.72(b)(2)(i) Any event. found while the reactor is shut down, that, had it been found while the reactor was in operation, would have resulted in the nuclear power plant, including its principal safety barriers, being seriously degraded or being in an ,
unanalyzed condition that significantly compromises plant I' safety.
Discussion: The arguments are very similar to those above and again can support either a reportable or non-reportable conclusion based on the licensee's assessment of the significance of !
the condition. However. this criterion was intended to capture potential problems which might be discovered'only :
during refueling outage surveillances. Question 7.10 in NUREG-1022, Supplement 1, is considered relevant guidance in .
regard to the reportability of equipment found to be !
inoperable during outage surveillances. .
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P Question 2.3 of NUREG-1022, Supplement 1, and the second draft of NUREG-1022, Revision 1, state that failures should .
be assumed to occur at the time of discovery unless there is '
firm evidence to believe otherwise. It seems appropriate to classify setpoint drift as a mechanism which would occur '
some time (usually indeterminable) during the period between calibration and subsequent surveillance unless some factor, !
such as an extended outage or testing conditions, could be !
identified as a likely cause. If testing conditions or :
other causes are identified such that reporting is deemed j unnecessary, the licensee would still be expected, under !
other programs and regulatory requirements, to evaluate the !
adequacy of the surveillance program to ensure that the activity is ensuring the operability cf the safety valves or '
other components. A voluntary report may still be useful'as !
a means of distributing the information related to the ;
problem and its cause to the industry. Please note that t although question 2.3 may be deemed an insufficient reason -
to determine safety valve drift is not reportable, the licensee may determine that the significance (see above) of ,
the condition does not satisfy the reporting threshold. -
50.72(b)(2)(iii) Any event or condition that alone could have prevented the 50.73(a)(2)(v) fulfillment of a safety function of structures or systems ;
that are needed to* !
(A) Shut down the reactor and maintain it in a safe shutdown condition,-
(B) Remove residual heat, (C) Control the release of radioactive material, or ;
(D) Mitigate the consequences of an accident.
Discussion: The second draft of NUREG-1022, Revision 1, provides safety valve drift as an example of a common mode problem which may be reportable under this criterion. The example was added to the case described in Information Notice 85-27 which dealt with multiple inoperable control rods. Although certain occurrences of multiple safety valve drift problems should be determined to be reportable under this criterion, ;
it should not be assumed that all cases of' one or more '
safety valves exceeding the technical specification tolerance band need be reportable in accordance with this ,
criterion. As in the previously discussed reporting criteria, the licensee's engineering judgement should determine if the condition could have prevented the t fulfillment of a safety function. Candidates for reporting include those cases in which the setpoints of multiple safety valves could have resulted in exceeding the ;
associated system's design pressure. If experience or engineering judgement can reasonably estimate the maximum .
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l drift which might occur and determine that the safety function would be maintained, the licensee can determine that the condition is not reportable. ..
Although discussed in the various drafts and revisions of !
NUREG-1022, it warrants repeating that the primary motivation behind evaluating plant conditions such as safety valve drift should be to ensure safety and only secondarily
- to determine reportability. If engineering assessments ,
identify a problem and determine that plant equipment was -
not and reasonably could not be rendered inoperable by a-phenomenon such as setpoint drift, the licensee can then also justify a determination that the condition is not reportable. Voluntary reports are appreciated if the licensee feels tiie information might be helpful to others. !
The staff should, as always, be cautious in recommending that a licensee make a " voluntary" report.
50.73(a)(2)(vi) Events covered in paragraph (a)(2)(v) of this section may include one or more procedural errors, equipment failures, ,
and/or_ discovery of design, analysis, fabrication, i construction, and/or procedural inadequacies. However, ;
individual component failures need not be reported pursuant ;
to this paragraph if redundant equipment in the same system was operable and available to perform the required safety function.
Discussion: (See above) 50.73(a)(2)(vii) Any event where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to ,
become inoperable in a single system designed to:
(A) Shut down the reactor and maintain it in a safe ;
shutdown condition.
(B) Remove residual heat, (C) Control the release of radioactive material, or (D) Mitigate the consequences of an accident.
Discussion: This criterion may be the most relevant to the specific example of safety valves found outside the technical specification tolerance band. As stated in the second draft '
of NUREG-1022, Revision 1, the reporting threshold for this ,
part of 10 CFR 50.73 is lower than for other parts since it is at the train or channel level rather than the system and function levels. Valves found outside the technical specification setpoint tolerance band can reasonably be considered to have been inoperable during operation unless a licensee determines that testing is not representative of
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conditions during operation (see item 50.72(b)(2)(1)). Dis ;
1 criterion was developed with general consideration given to the normal two train design level of redundancy. Given that most plants can satisfy pressure relief requirements with several main steam safety valves unavailable, a rigid interpretation of this criterion regarding the secondary safety valves (i.e., any case with more than one safety valve outside the tolerance band) may be overly conservative. However, the licensees are considered to have the weakest argument if they determine that this criterion ,
is not applicable, and therefore the condition is not reportable, when finding multiple safety valves outside the acceptable range.
50.73(a)(2)(i.B) Any operation or condition prohibited by the plant's technical specifications.
Discussion: Available guidance regarding operability and technical '
specification requirements generally have licensees enter the allowed outage time and associated action statements upon discovery of equipment inoperability unless a definite time of inoperability can be established. Technical specifications are considered satisfied provided the allowed outage time and associated action statements are-satisfied.
Therefore, provided that licensees restore compliance prior to returning to power operation, reporting of safety valve drift in accordance with this criterion would not be necessary. However, it is expected that upon identification of a problem such as safety valve setpoint drift, licensees t should take actions to prevent recurrence or pursue a change in the technical specification requirements (such as increasing the acceptable tolerance range of the setpoints).
If a licensee determines, through industry experience, information from a vendor, or self assessments, that a component may be inoperable during operation, appropriate actions should be taken in accordance with the technical specifications (reduce power or shutdown). This reporting criterion may be applicable if a licensee fails to satisfy the required action or can determine that a limiting conoition of operation had not been satisfied for longer than the allowed outage time following a specific cause for i a component becoming inoperable.
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