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Category:CORRESPONDENCE-LETTERS
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML18066A6881999-10-19019 October 1999 Forwards Rev 5 to Palisades Nuclear Plant COLR, Per Requirements of TS 6.6.5.Ltr Contains No New Commitments & No Revs to Existing Commitments ML18066A6871999-10-19019 October 1999 Forwards Response to NRC 990908 RAI Re Inservice Insp Program Relief Request 14.Ltr Contains No New Commitments & No Revs to Existing Commitments IR 05000244/19990081999-10-14014 October 1999 Forwards Insp Rept 50-244/99-08 on 990809-0919.Severity Level IV Violation of NRC Requirements Occurred & Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML18066A6741999-10-0202 October 1999 Forwards MOR for Sept 1999 for Palisades Nuclear Plant & Operating Data Rept Sheet for Month of Aug 1999.MOR for Aug 1999 Inadvertently Had Copy of Ref Data Sheet for Apr 1999 Data ML18066A6791999-10-0101 October 1999 Provides Response to RAI Re Draft Rept, Study of Air- Operated Valves in Us Nuclear Power Plants. ML18066A6621999-09-30030 September 1999 Notifies NRC That Util Will Implement ITS at Plant on or Before Oct 31,2000 & Attachments 1 & 2 Contains Request for License Condition Which Relates First Performance of New or Revised Surveillance Requirements to Implementation of ITS ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML18066A6601999-09-29029 September 1999 Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations. IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19992011999-09-24024 September 1999 Forwards Insp Rept 50-244/99-201 (Operational Safeguards Response Evaluation) on 990621-24.No Violations Noted. Primary Purpose of Osre to Assess Licensee Ability to Respond to External Threat.Insp Rept Withheld ML18066A6471999-09-17017 September 1999 Forwards Final Clean Copies of ITS & Bases Pages Which Incorporate All Changes Proposed in Listed Ltrs.Clean Copies Also Incorporate Some Editorial Changes & Bases Clarifications as Result of Ongoing Reviews to LCOs ML18066A6331999-09-0202 September 1999 Forwards Monthly Operating Rept for Aug 1999 & Revised Monthly Operating Rept for Apr 1999 for Palisades Nuclear Plant ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML18066A6261999-08-26026 August 1999 Forwards Addl New Valve Relief Request as Alternative to Code Requirements That Will Provide Acceptable Level of Quality & Safety.Request Would Allow Use of App II, Check Valve Condition Monitoring Program, of ASME OM Code-1995 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML20211D5661999-08-17017 August 1999 Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d).Attachments 1 & 2 Summarize Test Results at Palisades Plant,Big Rock Point Plant & Corporate Ofc ML18066A6111999-08-13013 August 1999 Requests Exemption from Certain Requirements of 10CFR50,App R, Fire Protection Program for Nuclear Power Facilities Operating Prior to 790101. Request Concerns Oil Collection Sys Requirements for PCP Motors ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML18066A5971999-07-30030 July 1999 Forwards Markup of Draft NRC SE Re Util Proposal to Convert to Its.Ltr Contains No New Commitments & No Revs to Existing Commitments ML18066A5921999-07-30030 July 1999 Forwards Results of Review by Consumers Energy of Two NRC Draft Repts Entitled, Evaluation of Air-Operated Valves at Light-Water Reactors & Study of Air-Operated Valves in Us Nuclear Power Plants. ML18066A5881999-07-30030 July 1999 Provides Rev to Instrument Channel Drift Measurement Submitted on 990611,in Response to NRC Comments on Util RAI Response for Sections 3.3,3.5 & 3.6 & Editorial Changes Revs Necessary for Consistency within ITS ML20210G8351999-07-29029 July 1999 Final Response to FOIA Request for Documents.Records in App a Encl & Will Be Available in PDR ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML18066A5651999-07-19019 July 1999 Forwards Corrections to Previously Submitted TS Section 3.7, Plant Systems, Converting to Its,Per NUREG-1432.Licensee Realized That Certain Provisions of CTS Had Been Inappropriately Replaced with Provisions from STS ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML18066A5111999-06-29029 June 1999 Provides Voluntary Confirmation of Facility Readiness as Outlined in GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Disclosure & Response Encl ML20209B2081999-06-29029 June 1999 Discusses Closure of Response to RAI Re GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity. Rvid,Version 2 Issued as Result of Review of Responses.Info Should Be Reviewed & Comments Submitted by 990901 ML20210G8791999-06-23023 June 1999 FOIA Request for All Document Communications Between NRC & Region III Involving R Landsman,B Jorgensen & R Caniano & NRC Staff Under Their Supervision & All Communications in Their Possession to & from Consumers Power Re Plant ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML18066A5061999-06-17017 June 1999 Forwards Responses to NRC Questions for ITS LCOs 3.6.3 & 3.6.6 of 990126 Submittal.One Editorial Change in Addition to Those Made in Response to NRC Comments & Conforming Changes Made to Associated Bases,Encl ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification ML18066A4991999-06-11011 June 1999 Forwards Responses to NRC Comments Re ITS Section 3.3 & Associated Revs to ITS Sections 1.0,3.3,3.4 & 3.9 of 990126 ITS Conversion Submittal.One Technical Change & Several Editorial Changes Unrelated to NRC Comments,Also Provided ML18066A4921999-06-0909 June 1999 Discusses Response of 980226 Violation Re Insp Rept 50-255/97-18 Re Failure to Take Adequate Corrective Action. Ltr Contains New Commitments & No Rev IR 05000255/19970181999-06-0909 June 1999 Discusses Response of 980226 Violation Re Insp Rept 50-255/97-18 Re Failure to Take Adequate Corrective Action. Ltr Contains New Commitments & No Rev ML18068A6011999-06-0808 June 1999 Forwards Description of Recent Changes Made to Palisades Site Emergency Plan,Excluding Minor & Editorial Changes Not Requiring Further Explanation 1999-09-30
[Table view] Category:ENGINEERING/CONSTRUCTION/CONSULTING FIRM TO NRC
MONTHYEARML19327A8691989-09-0707 September 1989 Submits Info Re Alchemie & Anderson County Bank Financing Transaction ML19332F2171989-07-10010 July 1989 FOIA Request for Documents Re Communications Between Ofcs of Edo,Deputy Edo,Ofc of Director,Regional Administrators & Commissioners Ofcs Re Plants During period,890301-0615 ML20246F1311989-06-26026 June 1989 FOIA Request for Minutes of Meeting Ref in 820210 Memo from NRR Re Design & Const Assurance for Upcoming OL Cases ML20245D9851989-06-22022 June 1989 Forwards 21 Insp Rept Executive Summaries,Per NRC Contract NRC-03-87-029,Task Order 037.Individual Quality Evaluations of Insp Repts Also Prepared ML20247P9411989-05-17017 May 1989 FOIA Request for Final Open Item Transmittal Ltrs Per NRC Insp Procedure 94300B for Listed Plants ML20245C1421989-04-0303 April 1989 Forwards Endorsements 75,108,108,96 & 110 to Maelu Policies MF-56,MF-26,MF-58,MF-39 & MF-52,respectively & Endorsements 93,129,127,109 & 122 to Nelia Policies NF-186,NF-76,NF-188, NF-151 & NF-173,respectively ML20247N1551989-03-31031 March 1989 Forwards Revised Proprietary Conformance of HPCS Div to NUMARC 87-00 Alternate AC Criteria, for Review as Result of Comments from 890216 Meeting.Rept Withheld ML20246M7331989-03-15015 March 1989 Responds to NRC Info Notice 88-082, Torus Shells W/ Corrosion & Degraded Coatings in BWR Containments. Summary of Relevant Projects for Various Utils Successfully Employing Underwater Alternative to Draining Vessel Encl ML20246N1281989-02-27027 February 1989 FOIA Request for Jl Smith to NRC Re Spent Fuel Shipment from Brunswick Nuclear Power Station to Harris Plant ML17285A2351989-02-0606 February 1989 Forwards Proprietary Draft Conformance of HPCS Div to NUMARC 87-00 App B Aac Criteria, for 890214 Meeting ML17285A2341989-01-0606 January 1989 Discusses Issues Highlighted at BWR/6 Alternate Ac Task Force Meeting on 881115,including Need for Capability of Div III Sys to Maintain Plant in Safe Shutdown Condition (Hot Shutdown) for Min of 4 H ML20206H0511988-11-14014 November 1988 Urges Relicensing of Pilgrim & Expedited Operation of Seabrook.Newspaper Clipping Encl ML20150D5721988-03-0808 March 1988 Provides Summary of Utils Test Results & Calculations on Emergency Diesel Generators,Including Review of Design of Static Exciter & Voltage Regulator for Emergency Diesel Generators ML20196C1591988-02-0303 February 1988 Forwards Monthly Progress Rept P-C6177-5, Independent Analysis & Assessment, for Period Ending 880131 ML20147G0741988-01-18018 January 1988 FOIA Request for All Documents Re NRC Investigation of Wg Dick Allegations About S&W & Lilco Re Performing NRC Instructions to Bring Facility Up to Fuel Load Stds ML20235A1251987-12-16016 December 1987 Forwards Info Re Resource Technical Svcs,Inc,Including Summary of NRC Contract Work,Nrc Form 26 for Three Existing Contracts,Audit Info,Work History & Lists of Expertise Available for Special Insps & of Current Resource Svcs ML20237B8051987-11-25025 November 1987 FOIA Request That Encls to Listed Documents,Including NRC Forwarding Amend 1 to License NPF-73,be Placed in PDR ML20236S4291987-10-20020 October 1987 FOIA Request for Listed Documents,Including Encls from NRC Requesting Addl Info on Gpu Topical Repts TR-033 & TR-040 & Encl to NRC Meeting Summary Re SPDS ML20236U5221987-10-19019 October 1987 FOIA Request for LERs for Listed Plants,Including All Attachments & Encls from Original Documents ML20235V1321987-08-28028 August 1987 Forwards EGG-NTA-7471, Technical Evaluation Rept,Reactor Trip Sys Reliability Conformance to Item 4.5.2 of Generic Ltr 83-28.... Based on Licensee Responses,Plants Reviewed Conform W/Exceptions Listed in Section 14 ML17342A7741987-07-13013 July 1987 Forwards Technical Evaluation of Rept, Retran Code: Transient Analysis Model Qualification, Dtd Jul 1985. Criteria for Use of Single & Two Loop Plant Models Listed. NRC Audit of Util QA Procedure Recommended ML20235K8731987-07-0909 July 1987 Informs That Tayloe Assoc Cannot Produce Mag or nine-track Tapes of Hearing Transcripts Until NRC Finalizes Arrangements W/Others to Provide Lexis Format,Including Library & File Numbers & Segmentation Info ML20237J2141987-07-0202 July 1987 FOIA Request for Listed Documents Ref in NUREG-1150 & Related Contractor Repts ML20238E3011987-06-29029 June 1987 FOIA Request for All Documents Described in App,Including Listed LERs & Revs for Plants,W/Original Attachments & Encls ML18052B1911987-06-17017 June 1987 Forwards EGG-NTA-7720, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Calvert Cliffs-1 & -2,Millstone-2 & Palisades, Final Informal Rept. Plants Conform to Generic Ltr Item ML20234B6211987-05-12012 May 1987 Requests That Listed Plants Be Added to Encl 870508 FOIA Request Re 94300 Region Input on Plant Readiness ML20234B6571987-05-0808 May 1987 FOIA Request for Placement,In Pdr,Region Input to NRC Headquarters,Nrr Re Status of Listed Plants in Terms of Plant Readiness for OL IE Manual,Chapter 94300 ML20214R4051987-05-0808 May 1987 FOIA Request for Region Input to NRR Re Status of Listed Plants Readiness for Ol,Per IE Manual Chapter 94300 ML18052B1531987-05-0101 May 1987 Forwards Final EGG-NTA-7622, Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components:Palisades, Informal Rept ML18150A1861987-05-0101 May 1987 Forwards EGG-NTA-7612, Conformance to Generic Ltr 83-28, Item 2.2.2 - Vendor Interface Programs for All Other Safety- Related Components,North Anna Units 1 & 2 & Surry Units 1 & 2, Final Informal Rept ML18052B0521987-04-17017 April 1987 Forwards EGG-NTA-7439, Conformance to Generic Ltr 83-28, Item 2.2.1 -- Equipment Classification for All Other Safety- Related Components:Palisades, Final Rept.Facility Conforms to Generic Ltr on Item ML20214Q8801987-04-17017 April 1987 Forwards EGG-NTA-7591, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28 Reactor Trip Sys Vendor Interface,Hatch-1 & 2,Millstone-1, Final Rept.Plants Conform to Item ML20214R0621987-04-17017 April 1987 Forwards EGG-NTA-7613, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface,Arnold, Brunswick-1 & 2, Final Rept.Plants Conform to Item ML18150A1171987-04-14014 April 1987 Forwards Final rept,EGG-NTA-7625, Conformance to Item 2.1 (Part 2) Generic Ltr 83-28,Reactor Trip Sys Vendor Interface,North Anna 1 & 2 & Surry 1 & 2. ML18052B0491987-03-30030 March 1987 Forwards EGG-NTA-7484, Technical Evaluation Rept for Palisades Plant,Response to NRR Generic Ltr 83-37, Interim Rept.Licensee Not in Compliance W/Control Room Habitability Requirement (III.D.3.4).Items Still Under Review Listed ML20214R8611987-03-27027 March 1987 Forwards EGG-NTA-7614, Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Cook-1 & -2,Haddam Neck, Final Informal Rept.Facilities Conform to Generic Ltr ML20214R1361987-03-26026 March 1987 Forwards Conformance to Item 2.1 (Part 2) of Generic Ltr 83-28,Reactor Trip Sys Vendor Interface:Maine Yankee, St Lucie 1 &-2 & Waterford 3, Final Rept.Plants Conform to Generic Ltr ML20214R1861987-03-26026 March 1987 Forwards Conformance to Generic Ltr 83-28,Item 2.2.2 - Vendor Interface Programs for All Other Safety-Related Components,Haddam Neck & Millstone 1,2 & 3, Final Rept ML20211D6631987-02-12012 February 1987 Notifies of 830204 Meeting W/Util,Idvp,Nrc & BNL in San Francisco,Ca to Discuss Status of Containment Annulus Steelwork & Status of Auxiliary Bldg ML20211D7031987-02-12012 February 1987 Notifies of 830209 Meeting in San Francisco,Ca to Discuss Shake Table Tests of Electrical Equipment ML20211D7441987-02-12012 February 1987 Notifies of 830517 Meeting in San Francisco,Ca to Discuss Development of Piping Stress Intensification Factor ML20209A8551987-01-16016 January 1987 FOIA Request for Documents to Be Placed in Pdr,Including NRC Re Calibr of Test Equipment allegation,1986 Inservice Insp Repts for McGuire 1 & Surry 1 & NRC 830307 SALP on Nine Mile Point 2 ML20209F8601987-01-15015 January 1987 Requests Publicly Available Copies of Handouts or Viewgraphs Used at Region III Meeting ML20207K0151986-12-19019 December 1986 FOIA Request That Encls to Insp Rept 50-247/86-26,Byron Semiannual Radioactive Effluent Rept & Millstone 1 & 2 SALP Rept Be Placed in PDR ML20211P2051986-11-24024 November 1986 FOIA Request for La Crosse & Big Rock Point Semiannual Effluent Repts & Turkey Point & St Lucie SALP Repts ML20214R8831986-11-0505 November 1986 FOIA Request for Encls to 860821 SALP Repts ML20213F8841986-10-30030 October 1986 FOIA Request for Encls to NRC 860724 & 31 Requests for Addl Info Re Vermont Yankee Spent Fuel Pool Expansion & Browns Ferry Seismic Reevaluation Program,Respectively ML20209D1691986-10-29029 October 1986 Forwards Rev 3 to EGG-EA-7035, Conformance to Generic Ltr 83-28,Items 3.1.3 & 3.2.3 Braidwood Units 1 & 2,Bryon Station Units 1 & 2,Callaway Plant Unit 1,Indian Point.... Licensees Conform to All Items W/Exception of Trojan ML20214K0231986-10-15015 October 1986 FOIA Request for All Documentation Re Accident Sequence Evaluation Program Repts Re Listed Facilities in Preparation for NUREG-1150 ML20214J9761986-10-15015 October 1986 FOIA Request for Containment Event Trees for Listed Facilities,Technical Repts & Memoranda Re Interpretation & Quantification & Identification of FIN Numbers,Contractors & Investigators Involved in Creation/Analysis of Event Trees 1989-09-07
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML17250B3041999-10-20020 October 1999 Forwards Changes to Tech Specs Bases for Re Ginna Nuclear Power Plant.Change Bars Indicate Those Revs Which Have Been Incorporated ML18066A6881999-10-19019 October 1999 Forwards Rev 5 to Palisades Nuclear Plant COLR, Per Requirements of TS 6.6.5.Ltr Contains No New Commitments & No Revs to Existing Commitments ML18066A6871999-10-19019 October 1999 Forwards Response to NRC 990908 RAI Re Inservice Insp Program Relief Request 14.Ltr Contains No New Commitments & No Revs to Existing Commitments ML17265A7651999-10-0808 October 1999 Forwards Fifteen Relief Requests That Will Be Utilized for Ginna NPP Fourth Interval ISI Program That Will Start on Jan 1,2000.Attachment 1 Includes Summaries & Detailed Description of Each Relief Request ML17265A7631999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 43 to Address Volumetric Examination Limitations (Less than 90%) Associated with a & B RHR Heat Exchanger Outlet Nozzle to Shell Welds.Approval Is Requested by Dec 31,2000 ML17265A7641999-10-0505 October 1999 Requests Approval for Use of Relief Request Number 42 to Address Volumetric Examinations Limitations (Less than 90%) Associated with Eight Class 1 Identified Welds or Areas of Reactor Pressure Vessel ML18066A6741999-10-0202 October 1999 Forwards MOR for Sept 1999 for Palisades Nuclear Plant & Operating Data Rept Sheet for Month of Aug 1999.MOR for Aug 1999 Inadvertently Had Copy of Ref Data Sheet for Apr 1999 Data ML18066A6791999-10-0101 October 1999 Provides Response to RAI Re Draft Rept, Study of Air- Operated Valves in Us Nuclear Power Plants. ML18066A6621999-09-30030 September 1999 Notifies NRC That Util Will Implement ITS at Plant on or Before Oct 31,2000 & Attachments 1 & 2 Contains Request for License Condition Which Relates First Performance of New or Revised Surveillance Requirements to Implementation of ITS ML20212J3561999-09-30030 September 1999 Forwards Four Copies of Re Ginna NPP Training & Qualification Plan for Security Officers, Rev 7,dtd 990930. Synopsis of Changes,Encl.Encl Withheld Per 10CFR73.21 ML20212J3801999-09-30030 September 1999 Forwards Four Copies of Rev to Re Ginna NPP Security Plan.Rev Changes Contingency Weapons Available to Response Force to Those Most Effective in Current Defensive Strategy.Encl Withheld Per 10CFR73.21 ML18066A6601999-09-29029 September 1999 Forwards NRC Form 536,in Response to NRC Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Examinations. ML17265A7571999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs IR 05000244/19990051999-09-24024 September 1999 Forwards More Detailed Response Re Main Steam Check Valve Performance Questions Arising from NRC Insp Rept 50-244/99-05 Following Completion of Independent Assessment Being Performed by Duke Engineering & Svcs ML18066A6471999-09-17017 September 1999 Forwards Final Clean Copies of ITS & Bases Pages Which Incorporate All Changes Proposed in Listed Ltrs.Clean Copies Also Incorporate Some Editorial Changes & Bases Clarifications as Result of Ongoing Reviews to LCOs ML18066A6331999-09-0202 September 1999 Forwards Monthly Operating Rept for Aug 1999 & Revised Monthly Operating Rept for Apr 1999 for Palisades Nuclear Plant ML17265A7461999-08-31031 August 1999 Submits Response to NRC Administrative Ltr 95-03,rev 2, Availability of Reactor Vessel Integrity Database,Version, Dtd 990726 ML17265A7411999-08-26026 August 1999 Forwards LER 99-004-01 Re Plant Being Outside Design Basis Due to Containment Recirculation Fan Moisture Separator Vanes Being Incorrectly Installed.Part 21 Notification of 990512 Is Being Rescinded ML17265A7401999-08-26026 August 1999 Requests Approval for Use of Relief Request Number 35 Re Use of ASME Section XI Code,1995 Edition,1996 Addenda.Code Will Be Used to Develop Plant Fourth 10-year Interval ISI Program on Class 1,2 & 3 Components ML18066A6261999-08-26026 August 1999 Forwards Addl New Valve Relief Request as Alternative to Code Requirements That Will Provide Acceptable Level of Quality & Safety.Request Would Allow Use of App II, Check Valve Condition Monitoring Program, of ASME OM Code-1995 ML17265A7451999-08-23023 August 1999 Forwards fitness-for-duty Performance Data Rept for Six Months Ending 990630,per 10CFR26.71(d) ML17250B3021999-08-23023 August 1999 Informs That Util & NRC Had Conference Call on 990816 to Review Approach in Responding to Questions,As Result of Questions Re Main Steam Check Valve Performance Included in Insp Rept 50-244/99-05,dtd 990806 ML20211D5661999-08-17017 August 1999 Forwards semi-annual Fitness for Duty Program Performance Rept for 990101-990630,IAW 10CFR26.71(d).Attachments 1 & 2 Summarize Test Results at Palisades Plant,Big Rock Point Plant & Corporate Ofc ML18066A6111999-08-13013 August 1999 Requests Exemption from Certain Requirements of 10CFR50,App R, Fire Protection Program for Nuclear Power Facilities Operating Prior to 790101. Request Concerns Oil Collection Sys Requirements for PCP Motors ML18066A5881999-07-30030 July 1999 Provides Rev to Instrument Channel Drift Measurement Submitted on 990611,in Response to NRC Comments on Util RAI Response for Sections 3.3,3.5 & 3.6 & Editorial Changes Revs Necessary for Consistency within ITS ML18066A5921999-07-30030 July 1999 Forwards Results of Review by Consumers Energy of Two NRC Draft Repts Entitled, Evaluation of Air-Operated Valves at Light-Water Reactors & Study of Air-Operated Valves in Us Nuclear Power Plants. ML18066A5971999-07-30030 July 1999 Forwards Markup of Draft NRC SE Re Util Proposal to Convert to Its.Ltr Contains No New Commitments & No Revs to Existing Commitments ML17265A7271999-07-30030 July 1999 Forwards 10CFR21 Interim Rept Per Reporting of Defects & Noncompliance,Section 21 (a) (2).Interim Rept Prepared Because Evaluation Cannot Be Completed within 60 Days from Discovery of Deviation or Failure to Comply ML17265A7151999-07-23023 July 1999 Forwards LER 99-007-01 Re Personnel Error Which Caused Two Channels to Be in Tripped Condition,Resulting in Reactor Trip.Further Investigation of Event Identified Addl Corrective Actions ML17265A7061999-07-22022 July 1999 Forwards LER 98-003-02,re Actuations of CR Emergency Air Treatment Sys.All Creats Actuations,Including Those Originally Believed to Be Valid Actuations,Were,In Fact Invalid Actuations ML17265A7141999-07-21021 July 1999 Withdraws Relief Request 35 for Plant Inservice Insp Program Section XI Requirements,Submitted on 980806.Licensee Plans to Resubmit Relief Request,Which Includes Addl Level of Detail,In Near Future ML17265A7191999-07-21021 July 1999 Forwards Ginna Station ISI Rept for Refueling Outage Conducted in 1999 ML18066A5651999-07-19019 July 1999 Forwards Corrections to Previously Submitted TS Section 3.7, Plant Systems, Converting to Its,Per NUREG-1432.Licensee Realized That Certain Provisions of CTS Had Been Inappropriately Replaced with Provisions from STS ML17265A7041999-07-16016 July 1999 Submits Info Re Specific Licensing Actions Which May Be Expected to Generate Complex Reviews,In Response to Administrative Ltr 99-02,dtd 990603 ML17265A6911999-06-30030 June 1999 Responds to NRC Request for Info Re Y2K Readiness at Nuclear Power Plants.Gl 98-01 Requested Response on Status of Facility Y2K Readiness by 990701.Readiness Disclosure Attached ML18066A5111999-06-29029 June 1999 Provides Voluntary Confirmation of Facility Readiness as Outlined in GL 98-01,Suppl 1, Y2K Readiness of Computer Sys at Npps. Disclosure & Response Encl ML20210G8791999-06-23023 June 1999 FOIA Request for All Document Communications Between NRC & Region III Involving R Landsman,B Jorgensen & R Caniano & NRC Staff Under Their Supervision & All Communications in Their Possession to & from Consumers Power Re Plant ML17265A6871999-06-22022 June 1999 Forwards Response to RAI Made During 990225 Telcon Re GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves. Calculation Encl. Encl ML17265A6841999-06-21021 June 1999 Informs That Util Wishes to Amend Extend of Alternate Exams Provided for Relief Request Re ISI Program ASME Section XI Require Exams for First 10-Yr Interval for Containment ML18066A5061999-06-17017 June 1999 Forwards Responses to NRC Questions for ITS LCOs 3.6.3 & 3.6.6 of 990126 Submittal.One Editorial Change in Addition to Those Made in Response to NRC Comments & Conforming Changes Made to Associated Bases,Encl ML17265A6741999-06-15015 June 1999 Submits Annual ECCS Rept IAW 10CFR50.46(a)(3)(ii) Requirements.No Changes Have Been Made to Large Break LOCA PCT & Small Break LOCA PCT ML18066A4991999-06-11011 June 1999 Forwards Responses to NRC Comments Re ITS Section 3.3 & Associated Revs to ITS Sections 1.0,3.3,3.4 & 3.9 of 990126 ITS Conversion Submittal.One Technical Change & Several Editorial Changes Unrelated to NRC Comments,Also Provided ML17265A6751999-06-11011 June 1999 Responds to NRC RAI Re Licensee GL 96-05 Program.Encl Info Verifies That Util Is Implementing Provisions of JOG Program on MOV Periodic Verification IR 05000255/19970181999-06-0909 June 1999 Discusses Response of 980226 Violation Re Insp Rept 50-255/97-18 Re Failure to Take Adequate Corrective Action. Ltr Contains New Commitments & No Rev ML18066A4921999-06-0909 June 1999 Discusses Response of 980226 Violation Re Insp Rept 50-255/97-18 Re Failure to Take Adequate Corrective Action. Ltr Contains New Commitments & No Rev ML18068A6011999-06-0808 June 1999 Forwards Description of Recent Changes Made to Palisades Site Emergency Plan,Excluding Minor & Editorial Changes Not Requiring Further Explanation ML17309A6551999-06-0707 June 1999 Responds to NRC 990310 RAI Re Verification of Seismic Adequacy of Mechanical & Electrical Equipment ML18066A4881999-06-0404 June 1999 Provides Responses to NRC Questions & Associated Editorial Revs for ITS LCOs 3.6.1,3.6.2,3.6.4,3.6.5 & 3.6.7 of 980126 Submittal.Responses to Comments on Remaining Section 3.6 LCOs Will Be Submitted Separately ML17265A6671999-06-0101 June 1999 Requests Approval of Ginna QA Program for Radioactive Matl Packages,Form 311,approval Number 0019.Ginna QA Program for Station Operation, Was Most Recently Submitted to NRC by Ltr Dtd 981221 & Supplemented on 990301 ML17265A6641999-05-25025 May 1999 Forwards Addl Info on Use of GIP Method a for Re Ginna Nuclear Power Plant.Copy of Re Ginna Station USI A-46 Outlier Resolution Table as Requested 1999-09-30
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I REGULATOR NFORMATION DISTRIBUTION TEM (RIDS)
AOCESSION NBR :8206290270 DOC ~ DATE: 82/06/28 NOTARIZED! NO DOCKET FACIL!50 000 Generic Docket 05000000 50-244 Robert Emmet Ginna Nuclear Plant< Unit 1P Rochester G 0500024/
50-255 Palisades Nuclear PlantP Consumers -Power Co. 05000255 AUTH'AME AUTHOR AFFILIATION BUDKITZ,R,J ~ Future Resources Associatesi Ines RBCIP ~ NAME RECIPIENT AFF IL'IATION RUSS ELL g H ~ T' NRC No Detailed Affiliation Given A v ~ 'v
SUBJECT:
Repts under 'Purchase Order DR 82 0961 re rreviey of two draft 'SEP.integrated asessments for facilities.Ltr compr ises rept on Ginna SEP (draft NUREG"0821).
DISTRIBUTION CODE: AOSSS iCOPIES 'RECEIVED:LTR . ENCL ./ SI ZE:
TITLE: SEP Topics NOTES!NRR/DL/SEP icy. 05000244 NRR/DL/SEP icy 05000255 RECIPIENT COPIES RECIPIENT iCOPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL ORB P5 BC 01 7 7 INTERKAL: NRR/DE/ADMQE 13 1 1 NRR/DE/HGEB 10 2 2 NRR/DL/ORAB 11 1 1 NRR/DL/SEPB 12 3 3 DS /AEB 1 1 NRR/DS I/CSB 07 1 1 04 1 1 RGN1 1 RGN3 1 1 EXTE RKAL: ACRS 14 10 10 LPDR 03 '2 2 NRC PDR 02 1 1 NTIS 5 1 1 NOTtES: 1 1
'TOTAL NUMBER OF COPIES REQUIRED: LTTR 34 ENCL 34
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2OOQCenterSfreet Berkeley, California 94704 416/62S-6111 Koom 415 28 June i 982 Mr. Wil liam T. Russel I U.S. Nuclear Regu'latory Commission Washington, DC 20555 RA:-
Dear Mr. ssel I:
This letter is a report to you under Purchase Order No., DR-82-0961, for which the scope of work ls a review of the two draft SEP Integrated Assessments for the Palisades and Ginna plants. This letter will comprise my report on the Ginna SEP assessment (Draft NUREG-0821, "Integrated Plant Safety Assessment, R.E. Ginna Nuclear Power Plant" ).
I understand that my report will be timely if delivered to you by 28 June, and I am pleased that I have made the deadline.
What I accomplished during this review was li m ited by a couple of constraints.
One of them was that I did not have access to enough detail about the actual Ginna plant design. I found seyeral issues where questions that I had could not be answered by the material at ha'nd .... of course, the character of the review that you asked me to do did not require such in-depth detail, but I found myself handicapped nonetheless, and probably could have provided better comments had I been in possession of, say, the FSAR. A second constraint was that sometimes found myself in need of the full documentation about the NRC I
regulatory position (Reg. Guides, Branch Technical Positions, etc.) in order to comprehend some issues.
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Despite these limitations, I do noi think that the essential issues within have been significantly compromised. Indeed, it is probably more thes'e'omments
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overview comments than detailed comments thai you want from me anyway.
One last introductory comment<>is that I did try to compare the Ginna SEP assessment with the Palisades assessment (NUREG-0820) thai I reviewed in April.
The comparison was useful, mainly because I attempted to see if any of the comments made by myself or others during the Palisades review had made their way into the Ginna report. I can report that some did and some did noi (see below).
A. GENERAL OBSERVATIONS One general observation relates to ihe burdens imposed on all licesees in the period after the TMI accident. I believe that many important a'afety improvements were made in the aftermath of TMI, but that the huge burdens placed on engineering staffs of both ut< lities and thei r contractors were probably too great. Had the TMI-related activities been stretched out more, ihe whole activity could have been more orderly. In that light, the retrofits being imposed within the SEP~program will also be more effective if imposed in an orderly way. I am pleased to report my personal observation that this lesson seems to have been learned, and that the SEP-'mposed changes seem to be quite orderly. The re'gulaiory staff has noi always shown such constraint. Congratulations.
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28 June l982 W.T. Russell page 2 Another comment worth repeating is one made by several reviewers (I among them) of the Palisades report: That is, the "list" of items under review is old and obsolete. Not only are many important issues omitted, but the way some of the issues are cast obviously reflects a pre-TMI approach to safety assurance.
This is unfortunate, and I would like to recommend thai some revisions be made ln the "SEP list" before it is used on the remaining plants. The most glaring examples are the only minor treatment in the SEP of human factors and control systems issues.
A further shortcoming, related to ihe comment just made, is that the "integrated, assessment" is not integrated, because a number of highly significnai items are simply left out, typically because they are being coped with in other regula-tory initiatives (TMI Action Plan, Unresolved Safety Issues, etc.). This is unfortunate. I only hope that not too many changes will be made to Ginna that would have been done differently had a fully integrated assessment been made at this time.
This then leads to a comment, that I must make here again, having made it earlier in my Palis'ades (SEP review. It is probably best simply..to quote my earlier words:
"The fact that NRC is systematically addressing these USI and TMI issues gives me comfort. In my view it is very likely that all of them will be resolved sooner or later, that all of our plants wil.l somehow be safer because of it, and that the safety improvements will be highly cost-effective. Nevertheless, I believe that the draft report I have in front of me is somehow inadequate or insufficient to the extent that it does not highlight this key point. I would feel better if the report had something like the following, up front somewhere, to guide the reader:
'The regulatory staff recognizes thai several of the most important safety issues have not been addressed or resolved in the course of
'his SEP effort, ln each case because they are being addressed through other regulatory efforts: in particular, the Unresolved Safety Issues list and the TMI Action Plan list contain some issues whose safety significance is probably far greater than a majority of the issues dealt with and resolved herein.'ertain aspects of this Ginna assessment are significantly improved over the Palisades assessment. One is the explicit discussion of Rochester Gas and Electric Company (RG+E) management. I am pleased to note this discussion, because It is an important issue to assess. I am even more pleased that the RG+E management seems to be of high competence. The nice language about this on pages I-6/1-7 is important.
A similar issue is the competence of the RG+E in-house engineering staff, whose competence should also be addressed: I note that this issue doesn't seem to have been touched on in ihe report, in contrast to the management issue.
I also found the general discussion (Appendix F) on operating experience to have been a good one. I found it to have been somewhat clearer than the sister analysisof Palisades. Perhaps my favorable response is a reaction to the contents',rather than to the analysis. I wiii discuss the'contents more below.
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28 June l982 -- W.T. Russell page 3 My last general observation is one that I somehow omitted in the Palisades letter, but is equally true of reviews of all older plants. It concerns acCing. I note that the Ginna is reviewed as if al,l 3 its components are new, and there is little or no discussion of whether the age of Ginna has any effect on the assessment. I recognize that this is because NRC's regulations generally contain no recognition of this issue: except in a few special circum-stances, systems are considered as if new throughout their life. I am dis-and hereby recommend this approach.
(.'nalysis, (My personal pre-conception on this issue is that for plants of Ginna's vintage there will be only a very few places where ihe aging issue will have negative safety significance; for most issues aging is probably a neutral consideration; and for quite a large number the plant's age is an affirmative safety advantage, in my view. But this "feeling" is not supported by any analysis.)
B. APPENDIX D (THE PRA ANALYSIS I begin by noting with pleasure that insights from probabilistic risk assessment (PRA) methods have been partially incorporated into the Ginna assessment. I also recognize that there has been no PRA carried out on ihe Ginna plant itself, so that ihe approach to gaining insights from PRA had to be through analyses of other plants. The approach taken was to use the PRA studies previously com-pleted on The Westinghouse-three-loop Surry plant (the WASH-I400 PWR) and the B8W-two-loop Crystal River plant. The analysts concluded that sufficient in-sights into Ginna, a Westinghouse-iwo-loop plant, could be obtained by judiciously combining insights from the Surry and Crystal River studies.
I believe that this is a reasonable approach, given the circumstances. I('also believe that the emphasis on rankin the safet im ortance of s stems in the present analysis is proper: indeed, I believe that to go much further than such a ranking would not be defensible in the absence of more plant-specific analysis.
Thus I concur in the restraint shown by the analysts in extending their PRA conclusions only as far as a rough (high-medium-low) ranking as to safety signi-ficance.
I am on balance even a little skeptica I as to the validity of the Ginna system rankings. Study of significant differences between Surry and Crystal River in the importance of some systems reveals a lot about these two plants, but the very fact that they are so~/different makes me a little wary of our ability to "interpolate", even with good engineering judgment.
Thus I would be Issg i,ng a stern admonition about how even
~ this limited PRA application was going to far, except that in actual fact it hasn'. I have examined the few issues covered (9 (I'n number) and find that nowhere did the PRA analysts seem to ~overstep the bounds of reasonable use of the available information. Congratulations to Sandia !
Also, for some of the PRA applications the study team did plant-specific analysis anyway. The best example of this, in my view, is the analysis of the safety significance of containment penetrations.
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28 June l982 W.T. Russell page 4 I believe that several of the issues that remain difficult within the integrated analysis before me could benefit from some limited PRA-type analysis, which offers a way to "crack" certain problems nicely. Among these are the following, all of which will be touched upon again later in this letter:
(a) The issue of flooding along Deer Creek. Here one could use some PRA-type insights into how important to safety are the several systems that might be compromised by the assumed standard project flood.
(b) The issue of vulnerability to high winds. Here there has been only a very little work done with PRA on ~an reactor, but insights are never-theless possible. For example, a very nice study of winds is incorpo-rated into the recent Indian Point PRA, and gives important vulnerability
.insights'even-though 'the"-quantitative conclustons. are, in my view,
.,highly Qncertain. What could be gained at Ginna is a more systematic understanding of which systems vulnerable to winds comprise which types of safety compromises in which combinations.
(c) The issue of the service water system (Issue ill-5-B, page 4-II). This be discussed further below. 'ill Despite my warning above thai it is dangerous to take PRA analysis too far, I believe that a little too much restraint is shown in Appendix D, specifically in the second long paragraph on page 0-8. Here the view seems to be set forth that PRA's applicability is limited, among other ways, to situations "for which the initiating event frequencies are relatively well-known....". I disagree.
are unknown, there is substantial insight to be gained by studying system dependencies and topologies using the type of thinking that characterizes PRA.
Indeed, it is just such analysi's thai I recommend for studying the vulnera-bility of Ginna from flooding on Deer Creek, and from high winds.
C. APPENDIX F "REVIEW OF OPERATING EXPERIENCE" I found this analysis to be a good one, despite the limitation that the analysis covers only the period through l979. (Irecognize that preliminary examinations were made of operating experience data since then, and that nothing striking pops up except the rather well-known steam generator problem in the, event of 25 January l982.) The discussion of ihe differentiation between "reportable events" and "forced shutdowns" is good, and important because the two categories include quite separate types of events.
The emphasis on trying to ferret out various human errors is appropriate. I myself believe that much more can be learned than has been about how human errors Interact with various hardware and control systems to compromise safety.
In this regard, the fact that such a large percentage (nearly one quarter) of reported events at Ginna can only be assigned an "indeterminate cause" (Page F-72) ls unfortunate. I believe that little effort should now be spent on digging back through Ginna's old log books to clarify these events, but I also think that careful thought should be given to how the large fraction can be reduced in the future by better reporting.
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28 June 1982 W.T. Russell page 5 In the context of the SEP mission, I am quite pleased with the overall message of this analysis. The most striking message is that many important systems have quite low failure rates. Examples include only one turbine trip over the plant lifetime; only one loss of offsite power (a very interesting event below); only three losses of feedwater, all in a short period in l97I;
.'ee zero fuel failures; and zero diesel failures on demand in service,'coupled with a very low diesel failure rate in tests. The observation that since l972 there have been no forced shutdowns associated with "operator error" is another example. All of this testifies to a well-'.run power station, because these low rates don't "just happen" .... they exist because of good practice.
The experience with what are called "recurring failures" is mixed. While RG+E seems to have fixed the earlier troubles with control rod drive mechanisms, the problems with emergency bus breakers are not yei solved, apparently. Again, the report affirmatively shows RG+E's diligence, which should be commended. But there is some intersting irony in the report (page F-77) that in the steam gene-rators, "tube thinning and corrosion problems have not yei been solved" ! The utility's multi-million-dollar problem at present with this issue attests to that The loss-of-offsite-power event of October l973 is interesting on two counts.
Fi,ist, apparently there was no definitive pinning down of the cause to operator error, yet one is suspected. This is a good example of why defense-In-depth engineering is so important: such events with no firmly established "cause" obviously cannot be remedied by changes, and obviously cannot be totally avoided either. Second, the rapid cooldown is clearly an event in the "thermal shock" category, probably at some decently high pressure as well. Has this event been analyzed in the pressurized thermal shock context '?
A final comment about Ginna's overall operating experience concerns the finding that over one quarter of all forced shutdowns were instrumantation and control anomalies (page F-5I). Again, as with human errors, this category deserves ihe most careful attention.
To summarize the analysis of operating experience at Ginna is pleasantly easy.
Not much in the way of failures has occurred until the steam-generator problem of recent vintage, and a well-managed plant emerges from the picture.
D. COMMENTS ON SPECIFIC TECHNICAL ISSUES I offer the following comments on specific technical issues within ihe Ginna SEP report. The order of these comments is not lndicati've of their relative importance.
i) Section 3.3.4.I Containment Isolation S stem (Electrical). I note with puzzlement the comment that "the safety injection reset pushbutton was inade-quately physically protected. The licensee has installed a protective tube to provide further protection against Inadvertent actuation." I have two possible explanation for this item's presence in the SEP report: either the staff happened to find this issue, quite unanticipated, in the course of reviewing other things, or the staff was specifical iy JooktncO for this. if the former, fine (it is obviously useful to make any improvements that one notices if they are easy and significant), but in this case what's the comment doing in the SEP
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28 June l982 -- W.T. Russell page 6 report ? If the latter, I am disturbed by the level of detail of staff review.
Does the sjtaff actually review stuff like that, specifically ? .... I mean, is it in the SRP or some other staff review guidance ? If so, I think we should get stuff at that level of detail out of the review.
This comment does not imply In any way that the improvement made is not useful.
ii) Section 4.24 Batter Monitorin and Annunciation. The issue is whether Ginna should install indications of battery current, charger output current, battery high discharge rate, and perhaps otI)er indications and annunciations in the control room. This should help to increase the likelihood that battery faults will be detected between battery service tests. RG+E has apparently agreed to ihe NRC staff position. However, ihe PRA discussion(page D-78 of Appendix D) contains some very interesting information.
Quoting directly: ."The industry-wide battery failure rate is 8.7 x IO /year based on study of Licensee'Event -Report". Approximately half of these failures were not detected until test or surveillance even though the minimum requirements include instrumentation such as is being proposed as backfii to Ginna .... The change in unavailability of a DC bus is only about a factor of 0.5 due to resolution of this issue. This is because a large fraction of battery degra-dations will remain undetected even with increased instrumentation."
My personal conclusion Is thai the licensee's agreeing to conform to the NRC's requirements on this issue, while obviously in the direction of "improved safety", is also obviously either overkill or underkill. Either the DC bus availability is a safety problem, in which case the gain in unavailibility of only a factor of 2 is insufficient (underkill); or the DC bus issue is not a safety problem, in which case the backfit is noi very important (overkill).
My opinion is that DC bus availability ls probably a general safety concern, although whether it is true at Ginna isn't (known to me. If this is the case, ihe staff's imposition of this change is inadequate: the issue may be "resolved" in the context of the SEP program but a different approach is needed to knock on this ?
This is a good example of an issue where satisfying "the letter of the regu-lations" makes people "feel good" but probably has little to do with actual safety improvement. I say "probably" because I don't know for sure what' going on (or by whom) to study this issue more. If further study is underway, I can only apologize for carrying on in this way.
iii) Section 4.I8 Loose Parts Monitorin . Although Ginna does not have a loose parts monitoring program that meets Regulatory Guide I.I33, backfit is not recommended by the staff. The reasons cited include the observation that no safety-related accidents occurred within a 3I-Incident sample studied recently. The no-backfit decision shows admirable restraint on the part of the staff . Congratulations.
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28 June l982 -- W.T. Russell page 7 iv) Section 4.I4 Pi e Break Outside Containment. The issue is in part that pipe breaks in the service water system (SWS) would trip the plant because several key components depend on service water. As the text states (page 4-II),
"In accordance with current criteria, a pipe break thai results in a reactor or turbine trip causes, in turn, a loss of offsite power." This is a arentl an assum tion for re ulator anal sis ur oses. The loss of SWS would bring down the emergency diesels too, however, because at Ginna these depend on SWS.
Loss of diesels would in turn bring down all electrically-driven auxiliary feedwaier pumps, leaving only the single steam-driven auxiliary feedwater pump, "which is susceptible to a postulated single active failure" (quoting the text).
Based on this analysis, the staff has assigned high priority to upgrading the SWS. This seems reasonable on general grounds, but I believe that this is a good example of the possible inadequacy of the "single failure criterion".
Specifically, I think an Imp'roved regulatory position might result from careful analysis of overall system vulnerability. Such an analysis, on a plant-specific basis, would ~indicate whether the scenario just sketched out is at all likely in ihe context of other safety issues. What is the actual likelihood of losing off-site power together with a pipe break in the SWS ? And what is the expected frequency of SWS pipe breaks '?
While I am not proposing the abandonment of the conservative regulatory approach in favor of probabilisiically-based decision-making, I do firmly believe that we could 'learn at two levels from such analysis: the lower level is insight into how urgent the Ginna backfit is; the higher level is how sensible I's the regulatory approach being applied.
v) Section 4.25.4 Pressure Sensor on Com onent Coolin Water Pum s. While Ginna does not satisfy current regulatory policy on this issue, ihe staff has concluded that backfitting is not required because other means 'exist to detect low flow if there is a failure of ihe single pressure sensor.. Insights from the PRA analysis were used to indicate that 'this issue has low safety significance.
I applaud the staff's restraint.
vi) Section 4.5 et al. Flood Protection from Deer Creek. In my discussion on PRA above, I mentioned my feeling that a good systems vulnerability analysis would be of significant value in resolving this issue. Even without much quantification, such an analysis could reveal the interdependencies and correlations among failed systems. The topologies themselves would reveal a lot, and,in fact the exercise of thinking through the topologies would produce the main insights. I repeat here my recommendation that this could be one way to resolve the difference between RG+E and the NRC staff on Deer Creek flooding.
I would also like to comment that the write-up in the Ginna report has insuffi-cient detail to explain what the real issues are.
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28 June l982 W.T. Russell page 8 vli) Section 4.22 Containment Penetrations. The discussion in the text, though quite long, is,'.impenetrable to the uninitiated. I have read it several times, have tried to compare it with the lengthy discussion and figures in the PRA appendix, and have failed miserably to come up with any feeling about what is right or wrong here. (Hal Lewis'mmortal word "inscruiibleu comes to mind.)
I generally accept the conclusion<of> the PRA analysis that the suggested backflts will make only a modest change in overall containment integrity.
Containment leakage was not found to be an important risk contributor in either the Surry (WASH-l400) or Crystal River analyses. However, I also respect the underlying rationale behind the containment isolation requirements in the General Design Criteria of Part 50. I therefore concur with the staff that the backfits ought to be required unless other considerations (such as very high cost) put a different light on ihe issue.
viii) Section. 4.8 Wind and Tornado Loadin s. The issue here is the apparent vulnerability of several Ginna structures to winds of only modest speeds (and hence of very high expected frequencies of occurrence). The windspeeds are shown in Table 4.2 of the text. Within the SEP, The entire issue is put aside for a later date. Specifically, it has been consolidated with several other issues and will be analyzed on an agreed-upon schedule, with structural upgrading decisions 'de'ferred for now. I concur with this reasonable approach.
On more general gro'unds, however, I would like to state my view that insights gained from a PRA-type vulnerability analysis from winds might be quite valuable. I mentioned this above in my general remarks about PRA, but think it is worth repeating here. A good i'recent example of how PRA thinking gives insights into structural vulnerabilities and their overall role in risk is the wind analysis in the recent Indian Point PRA, which broke new ground in both conclusions from such analyses are highly uncertain with the present state of the art.
E.
SUMMARY
AND CONCLUSIONS My summary and conclusions about the Ginna SEP integrated assessment have been hinted at above in the text. In a nutshell, I believe that the endeavor has been highly successful, has been carried out with admirable restraint vis-a-vis backfitting, and is well balanced. Various reservations about ihe underiakl;ng, stated above, in no way undermine this overall conclusion.
Sincerely yours, h
Robert J. Budnitz
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