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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217F9941999-10-15015 October 1999 Discusses FPC 970819 Request for Temporary Relief from ASME Code Section XI Requirements to Repair ASME Class 3 Nuclear Service & Decay Heat Sea Water System Piping.Forwards SE Containing Results of Staff Review ML20217J5171999-10-13013 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Plant,Unit 3 & Did Not Identify Any New Areas That Warranted More than Core Insp Program.Previously Planned Regional Initiative Insp of safety-related Mod Will Be Performed 3F1099-14, Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed1999-10-13013 October 1999 Requests Copy of NRC Radtrad Code & Copy of User Instructions.Conditions for Receiving Code Listed 3F1099-11, Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made1999-10-0404 October 1999 Provides Info on Requested Minor Permit Mod of Encl NPDES Permit.No New Regulatory Commitments Are Made ML20212L0771999-10-0404 October 1999 Forwards SER Accepting Licensee Relief Requests 98-012 Through 98-018 Involving Containment Insps at Crystal River Unit 3 Pursuant to 10CFR50.55a(a)(3)(i) & 10CFR50.55a(a)(3)(ii) ML20217D6551999-10-0101 October 1999 Requests That Natl Communication Sys Arrange for Licensee Participation in Government Emergency Telecommunications Service,Per NRC Info Notice 99-025 ML20212J8481999-10-0101 October 1999 Forwards Safety Evaluation Re Second 10 Yr Interval ISI Program Requests for Relief 98-009-II.Reliefs Granted for 98-009-II,Parts B & C & 98-010-II & 98-011-II 3F0999-03, Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment1999-09-27027 September 1999 Notifies of Approved Change to NPDES Permit Applicable to Crystal River Unit 3 IAW Section 3.2.3 of Epp.Proposed Change Was Approved on 990914 by State of Fl & Provided in Attachment 3F0999-18, Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 0003311999-09-27027 September 1999 Notifies NRC That Due Date for Commitment Common to Ltrs 980115 & 980209 Will Be Extended.Revised Completion Date for Cable Ampacity Project Is 000331 ML20212F7251999-09-23023 September 1999 Discusses Staff Review of Util 980330 Response,As Suppl on 990514,to GL 97-06, Degradation of SG Internals. Staff Concludes That Licensee Responses to GL Provide Reasonable Assurance That Condition of SG Internals Acceptable ML20212F7331999-09-23023 September 1999 Discusses Util Licensing Action for GL 98-01, Year 2000 Readiness of Computer Systems at Nuclear Power Plants. NRC Ack Efforts Util Completed to Date in Preparing Crystal River,Unit 3 for Y2K Transition 3F0999-20, Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-461999-09-21021 September 1999 Forwards Summary Re Justification to Defer USI A-46 Commitment,Per Work Needed to Resolve GL 87-03, Verification of Seismic Adequacy of Mechanical & Electrical Equipment in Operating Reactors,Usi A-46 ML20212E6741999-09-21021 September 1999 Forwards Safety Evaluation Accepting Proposed EAL Changes Submitted by ,As Supplemented by 981120,990713 & 0831 Ltrs,Incorporating Guidance in NUMARC/NESP-007,Rev 2, Methodology for Development of Eals 3F0999-01, Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv)1999-09-17017 September 1999 Forwards FPC Crystal River Unit 3 Plant Reference Simulator Four-Year Simulator Certification Rept Sept 1995-Sept 1999, Per 10CFR55.45(b)(5)(ii) & 10CFR55.45(b)(5)(iv) 3F0999-19, Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief1999-09-15015 September 1999 Provides Clarification of Minor Inconsistency Identified During Review of NRC SE for Plant Third 10-year Interval Inservice Insp Program Plan & Associated Requests for Relief ML20212F3141999-09-13013 September 1999 Forwards Insp Rept 50-302/99-05 on 990704-0814.Violations Noted,But Being Treated as non-cited Violations ML20211L9081999-09-0303 September 1999 Informs of Completion of Licensing Action for GL 92-08, Thermo-Lag 330-1 Fire Barriers, Dtd 921217,for Crystal River Unit 3 ML20211Q7581999-09-0101 September 1999 Forwards Summary of 990812-13 Training Managers Conference in Atlanta,Georgia Re Recent Changes to Operator Licensing Program.List Conference Attendees,Copy of Presentation Slides & List of Participant Questions Encl 3F0899-23, Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals1999-08-31031 August 1999 Provides Addl Info in Response to Several NRC Staff Questions Needed to Complete Review of Request to Adopt NEI 97-03,Draft Final Rev 3, Methodology for Development of Eals ML20211G7111999-08-30030 August 1999 Modifies Approval of 980521 Request for Exception to 10CFR50.4(b)(6) & Grants Util Approval to Submit Copies of Future Updates to FSAR as Listed ML20211G7031999-08-30030 August 1999 Informs of Approval of Util 980521 Request for Exception to 10CFR50.4(b)(6),allowing Util to Submit Updates to Plant Ufsar.Ltr Modifies That Approval & Grants Util Approval 3F0899-07, Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 20021999-08-27027 August 1999 Provides Formal Notification to NRC of FPC Plans Relative to Renewal of Crystal River Unit 3,FOL DPR-72.FPC Plans to Submit Application for License Renewal by End of 2002 ML20212C1351999-08-27027 August 1999 Requests Withholding of Proprietary Version of Enhanced Spent Fuel Storage Project Engineering Input 3F0899-20, Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.711999-08-26026 August 1999 Forwards six-month fitness-for-duty Program Performance Data for Period 990101-990630,IAW 10CFR26.71 3F0899-05, Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 31999-08-20020 August 1999 Forwards Response to NRC 990716 RAI Re Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in Crystal River Unit 3 3F0899-17, Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-051999-08-19019 August 1999 Submits Relief Request 99-0001-RR,seeking NRC Approval for Evaluation Performed by Util on through-wall Flaw in Nuclear Svc & Decay Heat Sea Water (RW) Sys,Per Guidance of GL 90-05 3F0899-16, Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal1999-08-19019 August 1999 Informs That Licensee Is Requesting State of Fl Dept of Environ Protection to Make Changes in Plant NPDES Permit to Modify Conditions on Use of Biocide in Instrument Air Compressor Sys.No New Commitments Are Made in Submittal 3F0899-02, Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 21999-08-16016 August 1999 Forwards Rev 2 to Cycle 11 COLR IAW Plant TS Section 5.6.2.18.Rev 1 of Cycle 11 COLR Was Not Submitted Due to Administrative Error.Changes Made in Rev 1 Listed & Incorporated in Encl Rev 2 3F0899-06, Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value1999-08-13013 August 1999 Forwards Monthly Operating Rept for July 1999 for Crystal River,Unit 3,per ITS 5.7.1.2.Revised Repts for Apr,May & June 1999,also Encl.Data on Line Item 6 Updated to Agree with More Accurate Computer Point That Measures Value 05000302/LER-1997-038, Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented1999-08-13013 August 1999 Forwards LER 97-038-01,IAW 10CFR50.73(c).Submittal Also Provides Notification That Commitment Common to LER 97-038-00 & Reply to NOV 50-302/97-16 Has Been Revised & Revised Commitment Has Been Implemented ML20210Q4511999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006 ML20210P0741999-08-0505 August 1999 Forwards SE Accepting Licensee 980416 & 1130 Ltrs Re Third 10-year Interval ISI Program Plan & Associated Requests for Relief for Plant,Unit 3 3F0799-30, Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 9906031999-07-29029 July 1999 Forwards List of Licensing Actions Currently Estimated for Fys 2000 & 2001,in Response to Administrative Ltr 99-02,dtd 990603 ML20210G8551999-07-27027 July 1999 Forwards Insp Rept 50-302/99-04 on 990523-0703.One Violation Identified & Being Treated as Noncited Violation 3F0799-09, Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments1999-07-19019 July 1999 Provides Response to NRC 990625 Telcon RAI Re Util Use of Relief Request 98-009-II for Plant ASME Section XI, Inservice Insp Second Interval.Ltr Established No New Regulatory Commitments ML20209H5211999-07-16016 July 1999 Forwards Request for Addl Info Re Licensee Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in CR-3 once-through Steam Generators in Order to Complete Review ML20209G3231999-07-15015 July 1999 Forwards Biological Opinion Issued by Natl Marine Fisheries (NMFS) of Dept of Commerce.Nmfs Concluded That Operation of Cw Intake Sys of Crystal River Not Likely to Jeopardize Existence of Species Listed in Biological Opinion ML20209G3481999-07-15015 July 1999 Transmits Natl Marine Fisheries Svc (NMFS) Biological Opinion Based on Review of Continued Use of Cw Intake Sys at Crystal River Energy Complex.Concludes That Continued Use of Cw Intake Sys Not Likely to Adversely Affect Gulf Sturgeon 3F0799-21, Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl1999-07-14014 July 1999 Forwards Copy of Revised NPDES Permit IAW Section 3.2.3 of Unit 3 Environ Protection Plan,Per 990430 Request to Allow Use of Biocide in Station Air Compressor Cooling Sys. Wastewater Permit FL0000159 Issued 990630 Also Encl 3F0799-05, Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl1999-07-14014 July 1999 Requests Exemption from 10CFR70.51, Matl Balance,Inventory & Records Requirements, as It Relates to 10CFR70.51(d) Re Physical Inventory of SNM for Crystal River Unit 3.Detailed Justification for Request,Encl 3F0799-25, Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl1999-07-14014 July 1999 Forwards License Renewal Applications for Four Individuals, IAW 10CFR55.57.Without Encl 3F0799-26, Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 9907301999-07-14014 July 1999 Provides Notice of Change in Status for Senior Operator,Iaw 10CFR50.74(a).RD Demontfort,License Number SOP 20528-2,has Been Reassigned & No Longer Requires License Effective 990730 3F0799-22, Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held1999-07-13013 July 1999 Provides Update & Rev to Submittal Made by Util Ltr with Regard to EAL Classification Methodology for Unit 3.Reponses to NRC Staff Questions Provided as Attachment D to Ltr & Reflects Discussions Held 3F0799-03, Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Forwards Rev 5-0 to Safeguards Contingency Plan,Replacing Current Rev to Safeguards Contingency Plan,Rev 4,in Entirety.Rev Withheld,Per 10CFR73.21 3F0799-02, Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.211999-07-0808 July 1999 Submits Rev 7-3 to Physical Security Plan,Replacing Current Rev to CR-3 Physical Security Plan,Rev 7-2,in Entirety.Rev Withheld,Per 10CFR73.21 ML20196L1261999-07-0707 July 1999 Discusses Closeout of TAC MA0538 Re License Response to RAI Re GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity, Issued on 950519 to Plant,Unit 3 3F0799-10, Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 31999-07-0707 July 1999 Submits Copy of Historical NPDES Permit Rev That Was Made in 1997 Re Use of Biocide at Crystal River Unit 3 ML20196J4991999-07-0101 July 1999 Advises That Info Contained in ,Which Included TR BAW-2346P,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20209C0811999-06-25025 June 1999 Forwards Overdue Controlled Document Transmittals for Listed Documents 3F0699-06, Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl1999-06-23023 June 1999 Submits Final Response to GL 98-01,Suppl 1 Re Year 2000 Readiness of Nuclear Power Plants.Year 2000 Readiness Disclosure for Crystal River,Unit 3,encl 1999-09-03
[Table view] Category:NRC TO UTILITY
MONTHYEARML20062E6821990-11-14014 November 1990 Forwards Insp Rept 50-302/90-30 on 900924-28.No Violations or Deviations Noted IR 05000302/19900261990-11-13013 November 1990 Ack Receipt of Responding to Violations Noted in Insp Rept 50-302/90-26.Response Meets Requirements of 10CFR2.201 IR 05000302/19900321990-11-0707 November 1990 Discusses Insp Rept 50-302/90-32 on 900918-1005 & Forwards Notice of Violation ML20062E0781990-11-0202 November 1990 Forwards Requalification Exam Rept 50-302/OL-90-01 Administered During Wk of 900917.Attention Should Be Given to Concern Over Exam Matl Development ML20058F6371990-10-31031 October 1990 Forwards Insp Rept 50-302/90-29 on 900908-1005.No Violations or Deviations Noted ML20058D6621990-10-30030 October 1990 Informs That Util Response to Bulletin 90-003, Relaxation of Staff Position in Generic Ltr 83-28, Acceptable ML20062C1891990-10-24024 October 1990 Forwards Evaluation Rept Re Implementation of B&W Owners Group Safety & Performance Improvement Program at Plant ML20058F1491990-10-23023 October 1990 Confirms Arrangements for Enforcement Conference to Be Conducted on 901031 at Region II Re Containment Integrity & Design Requirements for Maintaining Containment Integrity ML20058B0811990-10-19019 October 1990 Forwards Insp Rept 50-302/90-32 on 900918-1005.Violations Noted ML20062B5701990-10-15015 October 1990 Advises of Suspension of Action on Tech Spec Change Request 170 Re Pressure/Temp Limits Rept Until Suitable Solution Can Be Worked Out within Framework of Tech Spec Improvement Program,Per 890726 & 1031 Ltrs ML20058B1561990-10-12012 October 1990 Ack Receipt of 900928 Response Re Violations Noted in Insp Rept 50-302/90-24 ML20062B6691990-10-12012 October 1990 Requests Analyses of Liquid Samples Spiked W/Radionuclides Be Completed within 60 Days of Sample Receipt ML20058A9621990-10-11011 October 1990 Forwards Insp Rept 50-302/90-31.No Violations or Deviations Noted ML20059N4111990-10-0404 October 1990 Forwards Insp Rept 50-302/90-27 on 900917-21.No Violations or Deviations Noted.Rept Withheld (Ref 10CFR2.790 & 73.21) ML20059L9501990-09-24024 September 1990 Forwards Insp Rept 50-302/90-25 on 900812-17.No Violations or Deviations Noted ML20059M1791990-09-11011 September 1990 Advises That Objectives & Scenario Package for Emergency Response Plan Exercise,Scheduled for 900927,acceptable ML20059G7391990-09-0606 September 1990 Accepts Util 890601 & 1218 Responses to Item 1.b of Bulletin 88-11, Pressurizer Surge Line Thermal Stratification. Certain Issues Must Be Resolved Before Plant Meets All Appropriate Code Limits for 40-yr Plant Life ML20059G5941990-08-31031 August 1990 Forwards Insp Rept 50-302/90-24 on 900707-0810 & Notice of Violation ML20059H1421990-08-28028 August 1990 Requests Approved Ref Matl Listed in Encl 1,by 900926 for Reactor Operator & Senior Operator Licensing Exams Scheduled for 901126 ML20056B5321990-08-23023 August 1990 Forwards Safety Evaluation Re Util 890417 & 900330 Responses to Station Blackout Rule.Plant Not in Conformance W/Station Blackout Rule.Revised Response Should Be Submitted within 60 Days of Ltr Date.Rev 1 to SAIC-89/1150 Also Encl ML20056B2651990-08-22022 August 1990 Advises That Review of Auxiliary Feedwater Pump Conceptual Design,Submitted in ,Continuing.Nrc Offers No Comment Presently ML20058L4241990-07-31031 July 1990 Forwards Request for Addl Info Re Spent Fuel Pool Rerack Amend.Response to Items Should Be Submitted by 900820 ML20058N2901990-07-31031 July 1990 Advises That 900613 Rev 5 to Physical Security Plan Consistent W/Provisions of 10CFR50.54(p) & Acceptable ML20058Q2001990-07-31031 July 1990 Requests Supplemental Response Describing Util Corrective Actions & Date of Full Compliance,Per Insp Rept 50-302/89-18 ML20056A9191990-07-27027 July 1990 Forwards Insp Rept 50-302/90-21 on 900602-0706.No Violations or Deviations Noted ML20055H3561990-07-11011 July 1990 Advises That 900620 Response to Notice of Violation of Insp Rept 50-302/90-09 Meets Requirements of 10CFR2.201 ML20055H3701990-07-10010 July 1990 Forwards Insp Rept 50-302/90-17 on 900501-03.Noncited Violations Noted.Corrective Measures to Assure Randomness of Chemical Testing Program & to Improve Reporting of Chemical Test Results to NRC Will Be Reviewed in Future Insps ML20055H5751990-07-10010 July 1990 Forwards Insp Rept 50-302/90-22 on 900612-16.Violations Noted But Not Cited ML20055H4361990-07-0505 July 1990 Advises That Requalification Program Evaluation Visit Scheduled for Wk of 900917.Licensee Requested to Furnish Approved Items Listed in Encl Ref Matl Requirements by 900716.NRC Rules & Guidance for Examinees Also Encl ML20055E5121990-07-0202 July 1990 Forwards Safety Evaluation Re Util 831104 & 840731 Responses to Generic Ltr 83-28,Item 2.2.1 Re Equipment Classification Programs for All safety-related Components ML20055D6551990-07-0202 July 1990 Ack Receipt of & Check for $100,000 in Payment of Civil Penalty Imposed by NRC Order .Corrective Actions Will Be Examined During Future Insp IR 05000302/19900051990-06-29029 June 1990 Forwards Meeting Summary & Slides from Presentation on 900523 Re SALP 50-302/90-05.No Formal Changes to SALP Rept Needed,Therefore 900511 Rept Constitutes Final Rept.W/O Encl ML20059M9381990-06-13013 June 1990 Forwards NRC Performance Indicators for First Quarter 1990. W/O Encl ML20055C5981990-05-24024 May 1990 Forwards Order Imposing Civil Monetary Penalty in Amount of $100,000,per Violations Noted in Insp Rept 50-302/89-09. Penalty Proposed to Emphasize Importance of Implementing Adequate Mgt & Programmatic Controls ML20248D7351989-09-28028 September 1989 Requests That Impell Rept 03-0920-1186,Rev 0, Pipe Rupture Analysis Criteria Outside Reactor Bldg, Be Revised & Modified to Include Listed Conditions,Per NRC Approval to Calculate High Energy Line Breaks,Per ANSI B31.1 1967 ML20248E4811989-09-25025 September 1989 Forwards Staff Position on Preoperational Testing of Upgraded Emergency Diesel Generators to Be Implemented During Refuel 7.Type Qualification Tests Not Required Provided Preoperational Test Program Successfully Completed ML20248E7881989-09-25025 September 1989 Advises That Encl Notice of Consideration of Issuance of Amend to License & Proposed NSHC & Opportunity for Hearing on 890809 Request to Extend Surveillance Period for Diesel Generator full-load Test Sent to Ofc of Fr for Publication ML20248B7501989-09-22022 September 1989 Confirms 890919 Telcon Re Rescheduling & Change in Topics for Enforcement Conference on 890928,per Insp Repts 50-302/89-200 & 50-302/89-24 Concerning Insp Issues & Improper Impeller Installed on Raw Water Pump ML20248D1381989-09-22022 September 1989 Ack Receipt of 890908 Response to Violations Noted in Insp Rept 50-302/89-17 IR 05000302/19890151989-09-19019 September 1989 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-302/89-15 ML20248C8381989-09-18018 September 1989 Forwards Insp Rept 50-302/89-24 on 890907-08.No Notice of Violations Issued for Violations Described in Rept. Enforcement Conference to Discuss Violations Scheduled for 890928 ML20247K5711989-09-18018 September 1989 Advises That Voltage Levels Discussed in LER-86-015-1 & LER-87-007 & in Util 880229 & 0902 Ltrs Considered within Design Basis of Plant.Util Should Confirm Values for 480 & 120 Voltages by Testing During Next Refueling Outage IR 05000302/19892001989-09-18018 September 1989 Confirms 890912 Telcon Re Enforcement Conference to Be Conducted on 890928 to Discuss Issues from Insp Repts 50-302/89-200 & 50-302/89-24 & Items Reported by Plant Staff on 890907 & 08 ML20247K5911989-09-18018 September 1989 Advises That Plant PRA Considered Fair Evaluation of Most Likely Paths to Core Damage at Plant.W/Stated Exceptions, Review of PRA Did Not Uncover Any Reason Why PRA Could Not Be Used as Aid to Evaluate Plant Mods or for Decisions ML20247L0841989-09-18018 September 1989 Forwards Amend 8 to Indemnity Agreement B-54.Amend Increases Primary Layer of Nuclear Energy Liability Insurance Provided by ANI & Maelu ML20247K5531989-09-15015 September 1989 Advises That If Util Revises FSAR to Account for Revised Steam Generator Tube Rupture Mitigation Emergency Operating Procedure 10CFR50.59,NRC Will Review Evaluation as Part of Periodic Review of 10CFR50.59 Changes IR 05000302/19890091989-09-13013 September 1989 Discusses Insp Rept 50-302/89-09 on 890424-28 & Forwards Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $100,000.NRC Concerns Re Insp Findings Were Discussed During 890628 Enforcement Conference ML20247H0731989-09-12012 September 1989 Confirms 890914 Conference in Region II Ofc to Discuss Plans to Resolve Electrical Issue Prior to Restart ML20247H4451989-09-12012 September 1989 Ack Receipt of Informing NRC of Steps Taken to Correct Violation Noted in Insp Rept 50-302/89-16.Severity Level Reduced from IV to V,Based on Review of Response ML20247C2251989-09-0707 September 1989 Advises That Util 890608 Response to NRC Bulletin 89-001, Failure of Westinghouse Steam Generator Tube Mechanical Plugs, Acceptable.Acceptance Based on Implicit Assumption of Util Commitment to Perform Action of Item 2.D 1990-09-06
[Table view] Category:OUTGOING CORRESPONDENCE
MONTHYEARML20217F9941999-10-15015 October 1999 Discusses FPC 970819 Request for Temporary Relief from ASME Code Section XI Requirements to Repair ASME Class 3 Nuclear Service & Decay Heat Sea Water System Piping.Forwards SE Containing Results of Staff Review ML20217J5171999-10-13013 October 1999 Informs That on 990930,NRC Staff Completed mid-cycle PPR of Plant,Unit 3 & Did Not Identify Any New Areas That Warranted More than Core Insp Program.Previously Planned Regional Initiative Insp of safety-related Mod Will Be Performed ML20212L0771999-10-0404 October 1999 Forwards SER Accepting Licensee Relief Requests 98-012 Through 98-018 Involving Containment Insps at Crystal River Unit 3 Pursuant to 10CFR50.55a(a)(3)(i) & 10CFR50.55a(a)(3)(ii) ML20212J8481999-10-0101 October 1999 Forwards Safety Evaluation Re Second 10 Yr Interval ISI Program Requests for Relief 98-009-II.Reliefs Granted for 98-009-II,Parts B & C & 98-010-II & 98-011-II ML20212F7251999-09-23023 September 1999 Discusses Staff Review of Util 980330 Response,As Suppl on 990514,to GL 97-06, Degradation of SG Internals. Staff Concludes That Licensee Responses to GL Provide Reasonable Assurance That Condition of SG Internals Acceptable ML20212F7331999-09-23023 September 1999 Discusses Util Licensing Action for GL 98-01, Year 2000 Readiness of Computer Systems at Nuclear Power Plants. NRC Ack Efforts Util Completed to Date in Preparing Crystal River,Unit 3 for Y2K Transition ML20212E6741999-09-21021 September 1999 Forwards Safety Evaluation Accepting Proposed EAL Changes Submitted by ,As Supplemented by 981120,990713 & 0831 Ltrs,Incorporating Guidance in NUMARC/NESP-007,Rev 2, Methodology for Development of Eals ML20212F3141999-09-13013 September 1999 Forwards Insp Rept 50-302/99-05 on 990704-0814.Violations Noted,But Being Treated as non-cited Violations ML20211L9081999-09-0303 September 1999 Informs of Completion of Licensing Action for GL 92-08, Thermo-Lag 330-1 Fire Barriers, Dtd 921217,for Crystal River Unit 3 ML20211Q7581999-09-0101 September 1999 Forwards Summary of 990812-13 Training Managers Conference in Atlanta,Georgia Re Recent Changes to Operator Licensing Program.List Conference Attendees,Copy of Presentation Slides & List of Participant Questions Encl ML20211G7031999-08-30030 August 1999 Informs of Approval of Util 980521 Request for Exception to 10CFR50.4(b)(6),allowing Util to Submit Updates to Plant Ufsar.Ltr Modifies That Approval & Grants Util Approval ML20211G7111999-08-30030 August 1999 Modifies Approval of 980521 Request for Exception to 10CFR50.4(b)(6) & Grants Util Approval to Submit Copies of Future Updates to FSAR as Listed ML20210Q4511999-08-0505 August 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 991006 ML20210P0741999-08-0505 August 1999 Forwards SE Accepting Licensee 980416 & 1130 Ltrs Re Third 10-year Interval ISI Program Plan & Associated Requests for Relief for Plant,Unit 3 ML20210G8551999-07-27027 July 1999 Forwards Insp Rept 50-302/99-04 on 990523-0703.One Violation Identified & Being Treated as Noncited Violation ML20209H5211999-07-16016 July 1999 Forwards Request for Addl Info Re Licensee Proposed Alternate Repair Criteria for Axial Tube End crack-like Indications in CR-3 once-through Steam Generators in Order to Complete Review ML20209G3231999-07-15015 July 1999 Forwards Biological Opinion Issued by Natl Marine Fisheries (NMFS) of Dept of Commerce.Nmfs Concluded That Operation of Cw Intake Sys of Crystal River Not Likely to Jeopardize Existence of Species Listed in Biological Opinion ML20196L1261999-07-0707 July 1999 Discusses Closeout of TAC MA0538 Re License Response to RAI Re GL 92-01,Rev 1,Suppl 1, Rv Structural Integrity, Issued on 950519 to Plant,Unit 3 ML20196J4991999-07-0101 July 1999 Advises That Info Contained in ,Which Included TR BAW-2346P,will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) ML20207G3081999-06-0808 June 1999 Discusses Providing NRC Biological Assessment of Impact to Sea Turtles at Plant.Forwards Comments on Draft Biological Opinion Re Impact to Sea Turtles at Plant ML20195F3071999-06-0404 June 1999 Forwards Insp Rept 50-302/99-02 on 990411-0522.One Violation of NRC Requirements Occurred & Being Treated as non-cited Violation,Consistent with App C of Enforcement Policy ML20206N5511999-05-13013 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Created.Sr Peterson Will Be Section Chief for Crystal River Npp.Organization Chart Encl ML20206P5411999-05-0606 May 1999 Forwards Insp Rept 50-302/99-02 on 990228-0410.No Violations Noted.Three Violations of NRC Requirements Occurred & Being Treated as non-cited Violations,Consistent with App C of Enforcement Policy ML20206G1971999-05-0404 May 1999 Forwards Summary of 990412 & 14 Telcon with FPC Representatives Re NRC Questions Concerning Util 980729 & 1120 Ltrs Requesting Approval to Adopt NEI 97-03 Draft Final Rev 3, Methodology for Development of Eal ML20206K4461999-04-27027 April 1999 Forwards Notice of Withdrawal in Response to Util 990412 Request for Withdrawal of Application & Suppl by for Amend Re Normal Standby Position of Dh Removal Sys Valves DHV-34 & DHV-35 ML20206B9141999-04-20020 April 1999 Refers to Open Mgt Meeting Conducted at Licensee Request in Atlanta,Ga on 990414 Re Recent Crystal River 3 Plant Performance.List of Attendees & Copy of Presentation Handout Encl ML20205S4741999-04-20020 April 1999 Informs That NRC Determined FTI Topical Rept BAW-2342P,Rev 0,entitled OTSG Repair Roll Qualification Rept,Addendum a, Marked as Proprietary Will Be Withheld from Public Disclosure,Per 10CFR2.790(b)(5) & Section 103(b) ML20205T3241999-04-0909 April 1999 Informs That on 990402,K Mccall & Ho Christensen Confirmed Initial Operator Licensing Exam Schedule for Plant for Y2K Initial Exam Dates Are Wks of 000828 & 0911 for Approx Nine Candidates ML20205K6001999-04-0808 April 1999 Forwards Draft Biological Opinion from National Marine Fisheries Service Re Impact on Endangered Sea Turtles of Operation of Crystal River Energy Complex for Review & Comment ML20205K6111999-04-0202 April 1999 Forwards Request for Addl Info Re Util 980416 & Revised 981130 Submittals of Third 10-year Inservice Insp Program for Crystal River Unit 3 ML20205Q4871999-03-29029 March 1999 Forwards Insp Rept 50-302/99-01 on 990117-0227.No Violations Noted ML20205D5101999-03-26026 March 1999 Discusses FPC 961008 & 970228 Responses to NRC RAI Re Postulated Failure of RCP Seal Area Heat Exchanger Which Could Cause Overpressurization of Nuclear Svcs Closed Cycle Cooling Sys.Determined That in-depth Study Not Needed ML20205D3391999-03-22022 March 1999 Advises of NRC Planned Insp Effort Resulting from Plant PPR Completed on 990203,to Develop Integrated Understanding of Safety Performance.Historical Listing of Plant Issues & Details of Insp Plan for Next Eight Months Encl ML20204J5601999-03-22022 March 1999 Confirms 990309 Telcon Between R Mclaughlin of Licensee Staff & Ninh of NRC Re Mgt Meeting to Be Held at Licensee Request & Scheduled for 990313 in Atlanta,Ga to Discuss Recent Crystal River 3 Performance ML20207L3501999-03-0404 March 1999 Forwards Request for Addl Info Re 980730 License Amend Request 222 on Control Room Emergency Ventilation Sys IR 05000302/19982011999-02-16016 February 1999 Forwards Operational Safeguards Insp Rept 50-302/98-201 on 980504-06.No Violations Noted.Briefings,Observations,Drills & Exercises Indicated That Modified Protection Strategy Generally Sound.Without Encl ML20203G4891999-02-0505 February 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 990407. Representative of Facility Must Submit Either Ltr Indicating No Candidates or Listing of Candidates for Exam ML20203A4301999-02-0303 February 1999 Forwards SE Re 3 EAL Changes.Nrc Has Concluded That Changes Meet Requirements of 10CFR50.47(b)(4) & App E to 10CFR50 & Therefore Acceptable ML20206S4191999-01-26026 January 1999 Forwards FEMA Final Exercise Rept for 981014,emergency Response Exercise.No Deficiencies or Areas Requiring Corrective Action Were Identified During Exercise ML20199J8701999-01-14014 January 1999 Informs That on 990117,Region II Will Implement Staff Reorganization as Part of agency-wide Streamlining Effort. Copy of Organization Charts Encl for Info ML20199F4971999-01-11011 January 1999 Confirms FPC Informing That Request for USNRC Review of Topical Rept Boron Dilution by RCS Hot Leg Injection Submitted on 970227 Was Being Withdrawn.Informs That All Remaining Review Effort for Request,Canceled ML20199E3441999-01-11011 January 1999 Confirms Licensee Informing NRC That Request for Exemption from Requirements of 10CFR70.24(a) for Crystal River Unit 3,submitted by Util by Ltr Being Withdrawn ML20199D6071999-01-0404 January 1999 Forwards Insp Rept 50-302/98-10 on 981025-1205 & Notice of Violation.Nrc Concluded That Info Re Reason for Violation, C/A Taken & Plan to Correct Violation & Prevent Recurrence Already Adequately Addressed ML20199E3991998-12-23023 December 1998 Discusses Training Managers Conference Conducted at RB Russell Bldg on 981105.Agenda Used for Conference,List of Attendees,Slide Presentation & Preliminary Schedule for FY99 & FY00 Encl ML20198N9051998-12-16016 December 1998 Forwards Insp Rept 50-302/98-13 on 981116-20.No Violations Noted.Insp Consisted of Selective Exams of Procedures & Representative Records,Interviews with Personnel & Observations of Activities in Progress ML20196H2511998-12-0101 December 1998 Forwards Insp Rept 50-302/98-12 on 981013-16.No Violations Noted.Insp Team Observed Selected Portions of Emergency Organizations Response in Key Facilities During EP Plume Exposure Exercise ML20196C2931998-11-24024 November 1998 Forwards Notice of Withdrawal of 971031 Application for Amend to License ML20196A5321998-11-23023 November 1998 Ack Receipt of That Informed of Clarifications Needed Re Time Necessary to Start Control Complex Chillers at Plant,After Loss of Offsite Power.Changes to Start Times Do Not Alter Conclusion in SE Supporting Amend 163 to TSs ML20196C6291998-11-20020 November 1998 Advises of Planned Insp Effort Resulting from Insp Planning Meeting Held on 981102.Insp Plan for Next 4 Months & Historical Listing of Plant Issues,Called Plant Issues Matrix, Encl ML20195G4461998-11-17017 November 1998 Discusses FPC 980518 Submittal of Operational Assessment of Once Through SG Tube Degradation at CR-3 in Accordance with License Commitment Submitted on 971207 1999-09-03
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Docket No. 50-302 Mr. W. S. Wilgus Vice President, Nuclear Operations Florida Power Corporation ATTN: Manager, Nuclear Licensing P. O. Box 219 Crystal River, Florida 32629
Dear Mr. Wilgus:
SUBJECT:
CRYSTAL RIVER UNIT 3 - 10 CFR 50.62 (ATWS RULE) CONCEPTUAL DESIGN REVIEW AND REQUEST FOR INFORMATION (TAC NO. 59085)
The ATWS Rule (10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram [ATWS] Events for Light-Water-Cooled Nuclear Power Plants") requires improvements in the design and operation of commercial nuclear power facilities to reduce the likelihood of failure to shut down the reactor following anticipated transients and to mitigate the consequences of an ATWS event.
The requirements for Babcock & Wilcox (B&W) plants, such as Crystal River Unit 3 (CR-3), are to provide a diverse scram system (DSS) and diverse ATWS Mitigation System Actuation Circuitry (AMSAC). Paragraph (c)(6) of the Rule requires that information sufficient to demonstrate compliance with these requirements be submitted to the Office of Nuclear Reactor Regulation (NRR).
By letter to you dated June 30, 1988, the staff requested that within 90 days you address the design requirements in that letter and in the generic Safety Evaluation enclosed therewith, and that you proceed immediately with implementa-tion of ATWS modifications at CR-3. The staff's letter of September 7,1988, to L. C. Stalter summarized the meeting with the B&W ATWS Owners Group on August 17, 1988. It noted that the staff would review each plant-specific conceptual design package, and subsequently would prepare a Safety Evaluation upon receipt of a more detailed design package. Issuance of the Safety Evaluation would not necessarily precede implementation of ATWS modifications. The Septenter 7,1988 letter also stated that after receipt of approval of the conceptual design, each plant should install the ATWS equipment during its next refueling outage. Your submittal of September 28, 1988, constituted the conceptual design package.
Based on our review of your October 9,1985 and September 28, 1988 submittals, and on subsequent clarifying discussions, additional information, as discussed in the enclosed "Request for %dditional Information," is required to allow us to determine full compliance with the Commission's regulations, g p m%: 4 P
W. S. Wilgus November 18, 1988 The principal remaining concern is that feeding AMSAC signals into the Emer-gency Feedwater Initiation and Control (EFIC) system may not satisfy the requirements of 10 CFR 50.62 and may alter the design basis of the EFIC system.
During the meeting scheduled for November 29, 1988 we hope to clarify this and other matters through a discussion of the details of your proposed design, the function of EFIC, its interfaces with AMSAC and the reactor trip system, and to the extent possible the remaining additional information requested in the enclosure.
Our review did not encompass all interfaces of the required ATWS equipment with other plant systems to assure that the original design is not compromised. As part of your efforts on ATWS. you should perform such a review in accordance with the requirements of 10 CFR 50.59.
Subject to resolution of our concerns in the near future, it is our understanding that you will proceed with implementation of the ATWS redifications and complete them at the next refueling outage in 1989 in accordance with 10 CFR 50.62. We request that the final design package for DSS and AMSAC, including the requested additional information, be submitted to tha flRC staff within 90 days of the date of this letter.
The reporting and/or recordkeeping requirements contained in this letter affect
! fewer than 10 respondents; therefore, OMB clearance is not required under P.L.96-511.
Sincerely, Original signed by ETourigny for Harley Silver, Project Manager Project Directorate 11-2 Division of Reactor Projects-I/II Office of fluclear Reactor Regulation
Enclosure:
As stated cc w/ enclosure:
See next page DISTRIBUTI0ft 4
pocket ttle- HSilver NRC & Local PDRs OGC (for info only)
PDII-2 Rdg. EJordan, 3302 Mf:BB SVarga, 14/E/4 BGrines,9/A/2 Glaines,14/H/3 ACRS(10)
DMiller AThadani, 7/E/4 GHolahan,13/E/4 HBerkow
[ TAC HO. 59085/LTR.]
- See previous concurrence I LA:PDII-2* PH:PDII-2* D:PDII-2* SAD /r (A J pS W-DHiller HSilver:bd HBerkow GHo a ori 1
j 11/17/88 11/17/88 11/17/88 /Mffdah'i 11/ /188 11////88 A
4 GUa as 11/ /88
Mr. W. S. Wilgus Crystal River Unit No. 3 thclear Florida Power Corporation Generating Plant cc:
Mr. R. W. Neiser State Planning and Development Senior Vice President Clearinghouse and General Counsel Office of Planning and Budget Florida Power Corporation Executive Office of the Governor '
P. O. Box 14042 The Capitol Building St. Petersburg, Florida 33733 Tallahassee, Florida 32301 ftr. P. F. McKee Mr. F. Alex Griffin, Chairman Director, Nuclear Plant Operations Board of County Comissioners Florida Power Corporation Citrus County P. O. Box 219 110 North Apopka Avenue Crystal River, Florida 32629 Inverness, Florida 36250 Mr. Robert B. Borsum Mr. E. C. Simpson Babcock & Wilcox Director, Nuclear Site Nuclear Power Generation Division Florida Power Corporation Support 1700 Rockville Pike, Suite 525 P.O. Box 219 Rockville, Maryland 20852 Crystal River, Florida 32629 Resident Inspector ,
U.S. Nuclear Regulatory Comission 15760 West Powerline Street Crystal River, Florida 32629 Regional Administrator, Region 11 V.S. Nuclear Regulatory Comission 101 Marietta Street fl.W., Suite 2900 Atlanta, Georgia 30323 Jacob Daniel Nash Office of Radiation Control Departrnent. of Health and Rehabilitative Services 1317 Winewood Blvd.
Tallahassee, Florida 32399-0700 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304 '
s ENCLOSURE CRYSTAL RIVER UNIT 3 10 CFR 50.62 (ATWS RULE)
REQUEST FOR ADDITIONAL INFORMATION Introduction and Discussion On July 26, 1984, the Code of Federal Regulations (CFR) was vnended to include the ATWS Rule (Section 10 CFR 50.62, "Requirements for Reduction of Risk from Anticipated Transients Without Scram [ATWS] Events for Light-Water-Cooled Nuclear Power Plants"). An ATWS is an expected operatior.41 transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power), which is c : companied by a failure of the Reactor Trip Cystem (RTS) to shut down the reac;or. The ATWS Rule requires specific improvements in the design and ope ation of commercial nuclear power facilities to reduce the likelihood of f'.tlure to shut down the reactor fullowing anticipated transients and to 1'tigate the consequences of an ATWS event.
Paragraph (c)(6) of the Rule requires that information suf#icient to demon-strate compliance with the requirements of the Rule be submitted to the Direc-tor, Office of Nuclear Reactor Regulation. The ATWS Rule requirements for Babcock & Wilcox (B&W) plants, such as Crystal River Unit 3 (CR-3), are to provide a diverse scram system (OSS), and diverse (from the existing reactor trip system) ATWS n.itigation system actuation circuitry (AMSAC).
Based on review of the information provided with FPC letters dated October 9, 1985 and September 28. 1988 and in subsequent clarifying discussions, this request for adoitional information is needed to allow the staff to determine whether the CR-3 design fully cortplies with the A1WS Fule requirements of hardware diversity, electrical independence, and reliability and testability at power. This information should include simplified block diagrams showing OSS and AMSAC circuit components with a description of manufacturer, model, principle of operation (e.g., electro-mechanical, solid state, etc.), mode of operation (e.g., energize or de-energize to trip, etc.), power supplies (e.g.,
AC or DC, operating voltages, etc.), and identification and location of all Class 1E/non-Class 1E system interfaces.
The principal function of the DSS at CR-3 is to prevent an ATWS by tripping the reactor if, for any reason, the rods fail to drop in response to a Reactor Protection System (RPS) trip. The DSS must function to provide a reactor trip, diverse from the existing Reactor Trip Systen (RTS), for all ATWS tran-sients that require a reactor trip (in addition to AHSAC actions) to prevent the potential for oamage to, or overpressurization of, the Reactor Coolant System (RCS).
The AMSAC must function to actuate emergency feedwater (EFW) and trip the turbine on ATWS transients, where required, to prevent serious RCS overpres-surization, to maintain fuel integrity, and to meet 10 CFR radioactivity release requirements. Considerations for avoidance of inadvertent actuation dictate that there be at
,' ~
2 least two channels, powered from separate sources and coupled with appropriate coincidence capability. The ATWS transients of concern for the BWOG plants have been shown to be a complete loss of main feedwater (LMFW) and the loss of offsite power (LOOP) leading to LMFW,
.he following discussion and associated questions are applicable to the FPC "conceptual design" for the DSS and AliSAC at CR-3.
Diversity from the Existino RPS In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and AMSAC diversity are such that the "primary input signals will be ;
diverse from existing protection systems from the sensor output." Also, the 1 logic "system shall be diverse from existing protection systems," except that "certain plent-specific configurations may require enabling signals and power supply interconnections with existing protection systems." The output of the DSS "will degate SCRs using relays different from RPS SCR degate relays." The AMSAC "dctuation devices will be shared with existing systems."
For the DSS, equipment diversity to the extent reasonable and practicable to minimize the potential for common cause (mode) failures is required from the sensors to and including the components used to interrupt control rod power.
For the A!! SAC, equipment diversity to the extent reasonable and practicable to minimize the potential for comon cause (mode) failures is required from the sensors to, but not including, the final actuation device.
It is the staff's understanding that FPC's "conceptual design" for the DSS at CR-3 will use the RG 1.97 Rosemount pressure transmitters with Foxboro Model Spec 200 signal conditioning equipment, while the RTS uses Rosemount pressure transmitters with Bailey liodel 880 signal conditioning equipment. Since diversity of sensors is not required, the use of Rosemount transmitters for both the DSS and RTS appears to be acceptable. Also, the use of Foxboro signal conditioning for the DSS and Bailey signal conditioning for the RTS '
appears to provide adequate diversity.
It is also the staff's understanding that FPC's "conceptual design" for the i AliSAC at CR-3 will use fission chambers for indication of reactor power levels and Rosemount transmitters with Bailey Meter Co. Model 820 buffer modules to detect a complete loss of main feedwater flov, while the RTS uses ion chambers j to determine reactor power and does not use main feedwater flow as an input '
signal. Therefore, diversity of sensors and signal conditioning equipment appears to be adequate for the CR-3 AMSAC "conceptual design."
It is the staff's understanding that FPC's "conceptual design" for the DSS and AMSAC logic circuitry at CR-3 will use Bailey Meter Co. Model 820 equipment, while the RTS logic circuitry is Bailey Metc.r Co. Model 880. The FPC "concep- 7 tual design" for CR-3 indicates the Model 820 equipment operates in a -10 to i
+10 volt signal range while the Model 880 equipment operates in a 0 to +10 ,
volt range. This differentiation alone does not ensure that diversity is I achieved between the RTS and the DSS /AMSAC for the CR-3 logic circuitry.
Adequate diversity is best achieved by the use of components from different manufacturers / manufacturing processes, the use of mechanical versus electronic ,
devices, AC versus DC equir'nent, or equipment employing different princiales ,
of operation. Therefore, FPC must further determine how diversity is aciieved between the /odel 820 and Model 880 equipment and include this information in the CR-3 final plant-specific submittal.
. 3 It is the staff's understanding that the type of relays used to interrupt power to the control rod drives for the DSS have not been determined, but will be from a manufacturer different than Potter and Brumfield which manufactures the relays used in the RTS. It is also the staff's understanding that FPC's "conceptual design" for the AMSAC will use Potter and Brumfield relays to provide a turbine trip, while the existing RTS uses G.E. Model HEA relays.
For the final plant-specific CR-3 design FPC should not only consider dif-ferent relay manufacturers, but should also consider the other methods pre- !
vicusly stated in this section for ensuring diversity for the DSS end AMSAC equipment.
The FPC "conceptual design" for CR-3 states that "the RTS and AMSAC logic trip inputs will both use the circuitry resident in the EFIC system cabinett, to start the Emergency Feedwater Pumps and Control Emergency Feedwater flow Dy modulating the Emergency Feedwater control valves." This is in conflict with the requirement for AHSAC diversity and independence up to the final actuation device (i.e., the circuit breakers for emergency feedwater pumps and the motor contactors for emergency feedwater control valves represent the final actuation devices). Therefore, the final plant-specific CR-3 design provided by FPC should address this conflict and describe in detail how diversity and independence is achieved between the RTS and AMSAC if portions of the Emergency Feedwater Initiation and Control (EFIC) system are shared.
Electrical Independence from the Existing RPS In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and AMSAC electrical independence are such that "the system will be ,
electrically independent frem existing protection systems, except fer power supplies and certain enabling signals which will be appropriately isubted."
Electrical independence of the DSS from the existing RTS should be provided from the sensor output up to and including the final actuation device.
Electrical independence of the AMSAC systems from the existing RTS should be provided from the sensor output up to, but not including, the final actuation device. I r
It is the staff's understanding that FPC's "conceptual design" for the DSS /APSAC at CR-3 shows power being supplied via a 480-volt bus with its own independent (i.e., not associated with the RTS) non-Class 1E battery, rectifier, and !
charger providing 120 VAC to the ATWS circuitry. It is also the staff's '
understanding that the "conceptual design" for the CR-3 AMSAC ties the AMSAC to the EFIC system upstream of the final actuation device and that the EFIC and .
the RTS both share the same vital AC power. Thus, portions of the AMSAC do share vital AC power with the RTS. This design appears to be in conflict with Option 1 as described in the September 7, 1988 letter from G. Holahan (NRC) to L. C. Stalter (BWOG) and clearly falls into the Option 2 criteria "if EFIC is powered through 120 VAC RTS buses." Consequently, FPC must identify all DSS and AMSAC system components for CR-3 that %;ive power from sources that ere also used to provide power to the existint '
If RTS power supplies are used, information , ;t be provided to demonstrate that faults within the DSS or AMSAC circuits cannot degrade the reliability /
integrity of the existing RTS below an acceptable level. Information must also be provided to demonstrate that the potential for a connon mode failure affecting the RTS power distribution system that could compromise both the RTS and ATWS prevention / mitigation functions simultaneously is highly unlikely.
e 4 This information should include an analysis that addresses both degraded voltage and frequency conditions (e.g., overvoltage and undervoltage; the effects of degraded voltage / frequency conditions over time must be considered if such conditions can go undetected).
If alarms are relied upon to provide early detection of degraded voltage /
frequency conditions, the infnrmation provided shculd include the specific alarm (s) and the setpoint value(s). Information thould also be included regarding the limiting voltage / frequency values for which the affected circuits / components have been analyze'/ 'emonstrated to still be capable of performing their intended functions. A discussion of the periodic surveillance / testing performed to verify operability of the alarm circuits should also be provided.
In addition, even tr. . J c i Class 1E to non-Class 1E isolators used in the DSS and AMSAC at CR-3 have seen previously reviewed and accepted by the staff for use in the EFIC, RG 1 *7, and SPDS equipment, FPC must make a determination that the new DSS /AMSAC al/ W ations are bounded by the previously documented testing and so state in the final CR-3 plant-specific submittal.
The FFC "conceptv1 design" for the DSS /AMSAC electrical independence require-ments at CR-3 arrears to be in accordance with the ATWS Rule requirements except in the area of the sharing of power supplies. Thus, for the staff to make a final determination of gomplete acceptance, the CR-3 plant-specific information requested above must be provided for staff review.
Physical Separ,ation from Existing RpS In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and Ah AC are such that "channel separation shall be provided in accordance wie.i plant-specific requirements for routing non-safety signals."
To allow the staff to determine if this part of the CR-3 design complies fully with the current approved plant design requirements, specific details on component location and physical separation should be supplied in the plant-specific submittal.
Environmental Qualification (EQ) and Quality Assurance (QA) for Testing 2 Maintenance, and Surveillance, In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and AMSAC are such that "equipment shall be qualified for a normal environment in accordance with plant-specific EQ program requirements; quality assurance measures are to be provided on a plant-specific basis equal to or better than the requirements promulgated by HRC's Generic QA guidance letter, GL 85-06."
It is the staff's understanding that the CR-3 "conceptual design" provides environmental specifications for the locations of ATWS equipment and that this
, equipment should remain within the control complex boundaries and should be subject to essentially the same environment as that specified for Zone #43 Elevation 108 feet. It is also the staff's undarstanding that the ATWS System will be controlled in accordance with the general requirements of the FPC l Quality Program in a manner similar to that currently used for other nonsafety-related systens and equipment. Testing, maintenance, and any specified surveillances will be conducted and controlled in accordance with I
, 5 approved procedures. Collectively, the controls applied to the ATWS System will meet or exceed the "Quality Assurance Guidance For ATWS Equipenent That is Not Safety-Related," as set forth in Gcneric Letter 85-06.
The approach to the EQ and QA requirements identified above appears to be aCCep;able. It should be noted that the EQ and QA programs will be audited periodically during NRC regional inspections to assure continued compliance.
During the life of commercial light-water-cooled nuclear power plant, many components reach their end of life and must be replaced, including components installed in the RTS, DSS, and AMSAC. In its plant-specific submittal FPC should provide a description of the measures / programs impleinented for CR-3 to assure that the equipment diversity provided in accordance with the ATWS Pule will be maintained during component repair, replacement, and modifications and/or design changes, etc. throughout the life of the plant.
Safety-Related (1E) Power Supplies In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and AMSAC are such that a "safety-related power supply is not required." However, "operability during Loss of Offsite Power is required."
Although the use of safety-related (1E) power supplies is not required for the DSS and AMSAC systems, the logic and actuation device power for the DSS and logic power for the AMSAC designs must be from an instrument power supply independent from the power supplies for the existing RPS. In this regard, it is the staff's understanding that the CR-3 "conceptual design" for DSS and A!! SAC provides for power to be supplied b off-site power with a battery, inverter, and charger to provide 120 VAC power to the ATWS logic circuitry.
However, for the staff to make a final determination of acceptance, the exceptions noted in the diversity and independence sections concerning the use of EFIC equipment in the ATWS designs trust be addressed in detail.
Testability at Power Ire accordance with B&W Document 47-1159091-00, the ger.eric design requirements for DSS and AMSAC testability at power are such that "the system shall be testable at power, at power tests shall be performed at 6-month intervals with the complete system test being performd every refueling. The folicwing exceptions exist. The DSS input sensors and the ANSAC input sensors and final actuation devices will only be tested at refueling outages."
To ensure that the DSS circuits perform their safety functicns in a reliable manner, the circuits should be maintained and periodically tested at power in accordance with Technical Specification operability and surveillance require-ments or equivalent means.
It is the stoff's understanding that the portion of the CR-3 "conceptual design" that will allow testability at power is provided by the design of the DSS and AliSAC systems. These systems are designed so that both are 2 out of 2 logic actuated systems, and provisions are incorporsted which disable the second channel when a channel is placed in the test mude.
, 6 This approach appears to be in accordance with the above mentioned design requirements. However, the plant-specific submittal should also address the frequency of each test and time limits associated with channel testing, disabling of channels, actions to be taken if one channel has failed, etc.
Inadvertent Actuation In accordance with B&W Document 47-1159091-00, the generic design requirements for 053 and AMSAC to prevent inadvertent actuation are such that "the system shall be designed to minimize challenges to other safety systems by usinq at least two channels with appropriate coincidence logic; the use of two channels concurrent with the energize-to-trip design should minimize the number of inadvertent actuations."
To avoid the potential for inadvertent actuations of this nonsafety system, the system shall be designed so as not to revert to a 1-out-of-1 status during channel test. For systems designed using the minimum two-channel logic, this dictates that the system shall become inoperable during channel test.
It is the staff's understanding that the CR-3 "conceptual design" is in accordance with the above rnentioned design requirements.
Maintenance Bypasses, Operating Bypasses, Indication of Bypasses, and Means for Bypassing In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS and AMSAC bypassing are such that "the system shall incorporate a channel test capability; the test function should simultaneously test an input and output channel togetner from sensor to final actuation device."
System status during testing should be annunciated in the control room. The system should be designed to provide output to the control room inounted alarms for input channel trip conditions, output chMnel trip conditions, system trip, and test status. The system status will also be annunciated when the system is in the startup bypass mode / condition and reactor power *is less than 25% of rated full power.
It is the staff's understanding that the Cp-3 "conceptual design" concerning bypassing provides for disabling of a channel for maintenance, testing, repair or calibration by placing the other channel in test. Administrative controls will be provided to require pl:cing a DSS or AMSAC channel in test in order to provide control room annunciation any time work is to be performed which would disable operation of the other channel. These administrative controls will also prohibit personnel from working on more than one CSS or AMSAC channel simultaneously.
In addition, this design provides for automatically bypassing the AMSAC logic below a reactor power level of 25%. The DSS system does not require an operational bypass, and none is provided. The "conceptual design" also provides for indication of DSS and AMSAC status, including maintenance bypasses in the control room, with inputs to the sequence of events recorder. The sequence of events recorder displays alarm status to the operators via a color CRT mounted above the center of the main control board. The "conceptual design" provides these bypass capabilities for maintenance and test by using installed test modules. However, for the staff to make a determination of complete acceptance, FPC should assure that all items of concern addressed in the testability section of this report that are applicable to bypassing at CR-3 are discussed in the plant-specific submittal.
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7 Completion of Protective Action In accordance with B&W Document 47-1159091-00, the generic design requirements for DSS (AMSAC is not addressed) to assure completion of the protective action once it has been initiated are such that "the DSS shall ine' " ate a trip lockup with manual reset capability in the output channele Ntivated by a true DSS trip." Therefore, the plant-specific OSS desit be such that, once initiated, the protective action at the system leve ;u completion.
Return to operation should require subsequent deliberate p rator a: tion, e.g.,
mdnual reset of the tripped Circuits.
It is the staff's understanding that the FPC "conceptual design" for the 055 at CR-3 provides lockup of the DSS trip function such that reset of the 055 trip function requires manual operator action using Reset Switch S-1. It is also the staff's understanding that the AMSAC does not currently use a lockup trip function but that the plant-specific updated AMSAC design will incorporate trip lockup requiring manual operator action for reset. However, FPC should provide specific information which confirms that both the DSS and AMSAC at CR-3 are designed such that, upon receipt of a trip signal, the protective action goes to completion and deliberate operator action is required to reset the systems in order to comply with the ATWS Rule. In addition to the specific information on the system's design, FPC should include a discussion of any required operator actions.
Information Readout Although this item is not specifically addressed in B&W Document 47-1159091-00, it is the staff's understanding that FPC's "conceptual design" for CR-3 provides for indication of DSS and AMSAC system status both remotely (by means of the sequence of events recorder) and locally on the DSS and AMSAC test modules (by means of iiidicating lights). This type of design appears to be acceptable.
However, in the FPC plant-specific submittal for CR-3, more detailed information relating to how the operator is provided with accurate, complete, and timely information (i.e., what actuates or deactuates alarms, annunciators, lights, and what functions are performed by specific switches, etc.) pertinent to system status should be provided. In addition, FPC should provide a discussion of how good human factors engineering practices are incorporated into the design of ATWS prevention / mitigation system components located in the control room. The coordination of displays used to provide the status of ATWS syst, ms/
equipment to the operator with existing displays should be addressed specifically.
Safety-Related Interfaces In accordance with B&W Document 47-1159091-00, the generic design requirements for safety-related components / interfaces are such that "the DSS and AMSAC are not required to be safety related nor designed to meet IEEE 279, however riust be designed and engineered for high reliability to preclude unnecessary challenges to existing safety systems."
It is the staff's understanding that the FPC "conceptual design" for the CR-3 AMSAC does include interfaces with the existing reactor protection systems, as noted in the previous discussions of this report. These concerns (i.e.,
the sharing of power supplies via EFIC/AMSAC, and the adequacy of isolation devices used in these systems) must be adequately addressed in the plant-specific submittal in order for the staff to make a final decision regarding compliance of the CR-3 safety-related interfaces between the existing protection system and the DSS and AMSAC.
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. 8 I Technical Specifiestions i i
The staff, in its Technical Specifications Improvement Program, is presently evaluating the need for Technical Specification operability and surveillance ,
rzquirements. This evaluation includes those actions considered to be oppropriate to ensure that equipment installed per the ATWS Rule will be '
maintained in an operable condition when operability requirements cannot be met (i.e., limiting conditions for operation). In its Interim Comission
. Policy Statement on Technical Specification Improvements for Nuclear Power Plants [52 Federal Register 3778, February 6,1987), the Comission j established a specific set of objective criteria for determining which regulatory requirements and operating restrictions should be included in Technical Specifications. This aspect of the staff's review of the CR-3
- design for compliance with the ATWS Rule remains open pending completion of t
the staff's review to determine whether and to what extent Technical
, Specifications are appropriate. The staff will provide guidance regarding the Technical Specification requirements for DSS and AMSAC at a later date.
Installation of ATWS prevention / mitigation system equipment should not be delayed pending the development or staff approval of operability and surveillance requirements for ATWS equipment.
i Suma ry In order for the staff to make a final determination on the CR-3 compliance
- with the ATWS Rule and issue a Safety Evaluation, the additional information requested in the body of this document will be required. This additional information, as well as pertinent previously supplied information, should be
! submitted to the staff in a single ATWS system final design package. This will
] assure that on1 ' the most complete and up to date information is reviewed.
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