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Category:CORRESPONDENCE-LETTERS
MONTHYEARL-99-035, Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld1999-10-18018 October 1999 Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld ML20217G0801999-10-0707 October 1999 Informs That on 990930,staff Conducted mid-cycle PPR of Farley & Did Not Identify Any Areas in Which Performance Warranted More than Core Insp Program.Nrc Will Conduct Regional Insps Associated with SG Removal & Installation ML20217P0661999-10-0606 October 1999 Requests Withholding of Proprietary Rept NSD-SAE-ESI-99-389, Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs ML20217B1891999-10-0404 October 1999 Submits Clarification Re Development of Basis for Determining Limiting Internal Pressure Loads Re Review of NRC SE for Cycle 16 Extension Request.Util Intends to Use Guidelines When Evaluating SG Tube Structural Integrity ML20212J8391999-09-30030 September 1999 Forwards RAI Re Request for Amends to Ts.Addl Info Needed to Complete Review to Verify That Proposed TS Are Consistent with & Validate Design Basis Analysis.Request Discussed with H Mahan on 990930.Info Needed within 10 Days of This Ltr ML20212J8801999-09-30030 September 1999 Discusses GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps. Util 980731,990607 & 03 Ltrs Provided Requested Info in Subj Gl.Nrc Considers Subj GL to Be Closed for Unit 1 L-99-032, Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 21999-09-23023 September 1999 Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 2 L-99-034, Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 21999-09-23023 September 1999 Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 2 ML20212F8861999-09-23023 September 1999 Forwards Revised Relief Request Number 32 for NRC Approval. Approval Requested by 991231 to Support Activities to Be Performed During Unit 1 Refueling Outage Scheduled for Spring of 2000 ML20212E7031999-09-23023 September 1999 Responds to GL 98-01, Year 2000 Readiness of Computer Sys at Npps. Util Requested to Submit Plans & Schedules for Resolving Y2K-related Issues ML20212F1111999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals ML20212C2351999-09-16016 September 1999 Submits Corrected Info Concerning Snoc Response to NRC GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal ML20212D0101999-09-15015 September 1999 Informs That Submittal of clean-typed Copy of ITS & ITS Bases Will Be Delayed.Delay Due to Need for Resolution of Two Issues Raised by NRC staff.Clean-typed Copy of ITS Will Be Submitted within 4 Wks Following Resolution of Issues ML20212C4641999-09-13013 September 1999 Forwards Info Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams L-99-031, Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed1999-09-13013 September 1999 Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed ML20212C8041999-09-10010 September 1999 Responds to to D Rathbun Requesting Review of J Sherman Re Y2K Compliance.Latest NRC Status Rept on Y2K Activities Encl ML20212D4581999-09-10010 September 1999 Responds to to D Rathbun,Requesting Review of J Sherman Expressing Concerns That Plant & Other Nuclear Plants Not Yet Y2K Compliant ML20212A6951999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20212A8341999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20211N8041999-09-0808 September 1999 Informs That on 990930 NRC Issued GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Condition, to Holders of Nuclear Plant Operating Licenses ML20211N4301999-09-0808 September 1999 Discusses Proposed Meeting to Discuss Kaowool Fire Barriers. Staff Requesting That Affected Licensees Take Issue on Voluntary Initative & Propose Approach for Resolving Issues ML20212C0071999-09-0202 September 1999 Forwards Insp Repts 50-348/99-05 & 50-364/99-05 on 990627- 0807.No Violations Noted.Licensee Conduct of Activities at Farley Plant Facilities Generally Characterized by safety-conscious Operations & Sound Engineering ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl ML20211K2131999-08-31031 August 1999 Informs That Snoc Has Conducted Review of Reactor Vessel Integrity Database,Version 2 (RVID2) & Conclude That Latest Data Submitted for Farley Units Has Not Been Incorporated Into RVID2 ML20211K4101999-08-31031 August 1999 Resubmits Relief Requests Q1P16-RR-V-5 & Q2P16-RR-V-5 That Seek to Group V661 Valves from Each Unit Into Sample Disassembly & Insp Group,Per 990525 Telcon with NRC L-99-030, Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS1999-08-30030 August 1999 Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS ML20211G6851999-08-26026 August 1999 Informs That During Insp,Technical Issues Associated with Design,Installation & fire-resistive Performance of Kaowool Raceway fire-barriers Installed at Farley Nuclear Plant Were Identified L-99-029, Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 07271999-08-19019 August 1999 Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 0727 ML20211B9431999-08-17017 August 1999 Forwards Fitness for Duty Performance Data for six-month Reporting Period 990101-990630,IAW 10CFR26.71(d).Rept Covers Employees at Jm Farley Nuclear Plant & Southern Nuclear Corporate Headquarters ML20211B9211999-08-17017 August 1999 Responds to NRC Re Violations Noted in Insp Rept 50-348/99-09 & 50-364/99-09.Corrective Actions:Security Response Plan Was Revised to Address Vulnerabilities Identified During NRC Insp ML20210R5101999-08-12012 August 1999 Forwards Revised Page 6 to 990430 LAR to Operate Farley Nuclear Plant,Unit 1,for Cycle 16 Only,Based on risk- Informed Approach for Evaluation of SG Tube Structural Integrity,As Result of Staff Comments ML20212C8141999-08-0909 August 1999 Forwards Correspondence Received from Jm Sherman.Requests Review of Info Re Established Policies & Procedures ML20210T2021999-08-0606 August 1999 Forwards Draft SE Accepting Licensee Proposed Conversion of Plant,Units 1 & 2 Current TSs to Its.Its Based on Listed Documents ML20210Q4641999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Authorized Representative of Facility Must Submit Ltr to La Reyes,As Listed,With List of Individuals to Take exam,30 Days Before Exam Date ML20210J8341999-07-30030 July 1999 Forwards Second Request for Addl Info Re Util 990430 Amend Request to Allow Util to Operate Unit 1,for Cycle 16 Based on risk-informed Probability of SG Tube Rupture & Nominal accident-induced primary-to-second Leakage ML20210G4901999-07-30030 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, Issued 990603.Ltr Contains NRC License Commitment to Utilize ASTM D3803-1989 with Efficiency Acceptance Criteria Utilizing Safety Factor of 2 L-99-028, Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines1999-07-30030 July 1999 Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines L-99-027, Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.51999-07-27027 July 1999 Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.5 ML20210G8181999-07-26026 July 1999 Forwards Insp Repts 50-348/99-04 & 50-364/99-04 on 990516- 0626.One Violation Identified & Being Treated as Noncited Violation IR 05000348/19990091999-07-23023 July 1999 Discusses Insp Repts 50-348/99-09 & 50-364/99-09 on 990308- 10 & Forwards Notice of Violation Re Failure to Intercept Adversary During Drills,Contrary to 10CFR73 & Physical Security Plan Requirements ML20210E4071999-07-22022 July 1999 Responds to NRC 990702 RAI Re Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described in NEI 97-06, SG Program Guidelines L-99-026, Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments1999-07-19019 July 1999 Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments L-99-264, Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 20011999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 ML20209H4721999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 ML20196J6191999-07-0202 July 1999 Forwards Final Dam Audit Rept of 981008 of Category 1 Cooling Water Storage Pond Dam.Requests Response within 120 Days of Date of Ltr 05000364/LER-1999-001, Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed1999-07-0202 July 1999 Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed ML20196J7471999-07-0202 July 1999 Forwards RAI Re Cycle 16 Extension Request.Response Requested within 30 Days of Date of Ltr ML20196J5781999-07-0202 July 1999 Forwards RAI Re 981201 & s Requesting Amend to TS Associated with Replacing Existing Westinghouse Model 51 SG with Westinghouse Model 54F Generators.Respond within 30 Days of Ltr Date ML20196J6571999-07-0202 July 1999 Discusses Closure to TAC MA0543 & MA0544 Re GL 92-01 Rev 1, Suppl 1,RV Structural Integrity.Nrc Has Revised Rvid & Releasing It as Rvid,Version 2 as Result of Review of Responses ML20196J3591999-06-30030 June 1999 Forwards SE of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs 1999-09-09
[Table view] Category:NRC TO UTILITY
MONTHYEARML20062E0451990-11-13013 November 1990 Discusses 901106 Meeting Re Alternative Tube Plugging Criteria for Facility Steam Generators ML20062D6911990-11-0909 November 1990 Forwards SER Re Response to Generic Ltr 88-11 Concerning NRC Position on Radiation Embrittlement of Reactor Vessel. Compliance w/10CFR50,Appendix G,For Min Upper Shelf Energy Could Not Be Verified.Operation for 16 EFPY Acceptable ML20058F7931990-11-0606 November 1990 Forwards Summary of NRC Understanding of Current Status of Unimplemented Generic Safety Issues for Plant,Per Util 900627 Response to Generic Ltr 90-04 ML20058B4021990-10-19019 October 1990 Forwards Summary of 900913 Mgt Meeting Re Emergency Preparedness Program & Onsite Technical Support Ctr ML20058B1391990-10-16016 October 1990 Forwards Insp Repts 50-348/90-28 & 50-364/90-28 on 900911-1012.No Violations or Deviations Noted ML20062B7031990-10-12012 October 1990 Requests Completion of Analysis of Liquid Samples Spiked W/ Radionuclides within 60 Days of Receipt of Ltr ML20058B2581990-10-12012 October 1990 Ack Receipt of 900807 & 0904 Ltrs Transmitting Objectives & Scenario Package for Exercise Scheduled on 901024 ML20059N6721990-10-0909 October 1990 Responds to Util 900702 Request for Authorization to Use ASME Boiler & Pressure Vessel Code Case N-395.Request Acceptable IR 05000348/19900181990-10-0909 October 1990 Ack Receipt of 900926 Response Re Violations Noted in Insp Repts 50-348/90-18 & 50-364/90-18 ML20062C6601990-10-0909 October 1990 Ack Receipt of 900913 Response to Violations Noted in Insp Repts 50-348/90-19 & 50-364/90-19 ML20059N3481990-09-24024 September 1990 Forwards Insp Repts 50-348/90-24 & 50-364/90-24 on 900827-31.No Violations or Deviations Noted ML20059K1311990-08-28028 August 1990 Confirms Mgt Meeting on 900913 to Discuss Timeliness of Activation of Emergency Response Facilities ML20059H0921990-08-23023 August 1990 Forwards Insp Repts 50-348/90-20 & 50-364/90-20 on 900711-0810.No Violations or Deviations Noted ML20059B3421990-08-23023 August 1990 Advises That Util 900131 Response to NRC Bulletin 89-003 Re Potential Loss of Required Shutdown Margin During Refueling Operations,Satisfactory ML20056B3721990-08-21021 August 1990 Forwards Order Imposing Civil Monetary Penalty in Amount of $450,000 Re Violations,Per 870914-18,1102-06 & 16-21 Insps. Violations Include Failure to Demonstrate Splice Qualification & Failure to Maintain Environ Qualification ML20058M9441990-08-0303 August 1990 Forwards Insp Repts 50-348/90-19 & 50-364/90-19 on 900711-13 & Notice of Violation.Notice of Violation Withheld (Ref 10CFR2.790 & 73.21) ML20058P8661990-07-31031 July 1990 Forwards Insp Repts 50-348/90-21 & 50-364/90-21 on 900709-13.No Violations or Deviations Noted ML20055H4251990-07-13013 July 1990 Forwards Insp Repts 50-348/90-16 & 50-364/90-16 on 900611-0710.Violations Noted,But Not Cited ML20055H6501990-07-13013 July 1990 Forwards Insp Repts 50-348/90-17 & 50-364/90-17 on 900618-22.No Violations or Deviations Noted IR 05000348/19900121990-07-10010 July 1990 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-348/90-12 & 50-364/90-12.Part (a) of Violation Agreed with Denial & Withdrew This Portion & Revised Records Accordingly ML20059M8061990-06-13013 June 1990 Forwards NRC Performance Indicators for First Quarter 1990. W/O Encl ML20055C3961990-02-23023 February 1990 Forwards Insp Repts 50-348/90-04 & 50-364/90-04 on 900129-0202.No Violations or Deviations Noted ML20248B0201989-09-18018 September 1989 Forwards Summary of 890912 Mgt Meeting in Region II Ofc Re Configuration Mgt Program & Findings of self-initiated Safety Sys Assessment of Svc Water Sys.List of Attendees & Meeting Agenda Also Encl ML20247K0821989-09-15015 September 1989 Forwards Amend 8 to Indemnity Agreement B-81,reflecting Revs to 10CFR140, Financial Protection Requirements & Indemnity Agreements, Increasing Liability Insurance ML20248C2061989-09-15015 September 1989 Advises of Special Emergency Operating Procedure Insp Program Being Performed at Plant on 891127-1208 to Verify That Procedures Technically Accurate & That Actions Can Be Accomplished W/Equipment,Controls & Instrumentation ML20247G5401989-09-12012 September 1989 Forwards Insp Repts 50-348/89-19 & 50-364/89-19 on 890821-25.No Violations or Deviations Noted.Licensed Operator Performance During Simulator Training Did Not Meet Expectations ML20247G9511989-09-12012 September 1989 Forwards Insp Repts 50-348/89-18 & 50-364/89-18 on 890731-0804 & Notice of Violation.Notice of Violation Withheld (Ref 10CFR2.790 & 73.21) ML20247G6611989-09-0707 September 1989 Confirms 890906 Telcon W/J Garlington & D Verrelli Re Mgt Meeting Scheduled for 890912 in Region II Ofc to Discuss Results of Licensee self-initiated Safety Assessment ML20247F4881989-09-0606 September 1989 Ack Receipt of Responding to Violations Noted in Insp Repts 50-348/89-13 & 50-364/89-13.Implementation of Corrective Actions Will Be Examined During Future Insps ML20246N7041989-08-25025 August 1989 Forwards Insp Repts 50-348/89-17 & 50-364/89-17 on 890724-28.No Violations or Deviations Noted ML20246P3561989-08-17017 August 1989 Forwards Insp Repts 50-348/89-16 & 50-364/89-16 on 890711-31.No Violations or Deviations Noted ML20246G3921989-08-11011 August 1989 Confirms Util Participation in NRC Regulatory Impact Survey on 890911,per 890808 Telcon.Agenda Encl ML20245K5571989-08-0909 August 1989 Advises That 890601 Rev 16 to Emergency Plan Consistent W/ Provisions of 10CFR50.54(p) & Acceptable ML20245H9851989-08-0303 August 1989 Requests Responses to Encl Questions Re Requests for Relief from ASME Code Requirements of Second 10-yr Inservice Insp within 45 Days of Ltr Receipt ML20245H1201989-08-0202 August 1989 Requests That Encl Ref Matls Be Furnished to NRC by 890825 for Requalification Program Exam Scheduled for Wk of 891023 ML20248C8061989-08-0101 August 1989 Forwards Insp Repts 50-348/89-10 & 50-364/89-10 on 890424-27 & 0508-12 & Notice of Violation.Unresolved Item Identified ML20247R9781989-07-28028 July 1989 Forwards Final FEMA Rept on 881102 Emergency Response Exercise.Identified Deficiency Promptly Corrected.Resolution of Weaknesses,Identified by Fema,Should Be Corrected Prior to Next full-scale Emergency Preparedness Exercise ML20245F1971989-07-26026 July 1989 Forwards Insp Repts 50-348/89-14 & 50-364/89-14 on 890611-0710.No Violations or Deviations Noted.Matter Re Apparent Excessive Work Hours for Licensed Operators Will Be Discussed W/Util at 890731 Meeting in Region II Ofc ML20247R5141989-07-26026 July 1989 Forwards Insp Repts 50-348/89-13 & 50-364/89-13 on 890509-10 & Notice of Violation.Fact That Two Independent Actions Involving Unauthorized Use of Radioactive Matls Occurred within Short Time Indicates Disrespect for Radioactivity ML20247E2691989-07-20020 July 1989 Forwards Exam Rept 50-348/OL-89-01 Administered During Wks of 890619 & 26.Concerns Noted Re Operator Performance & Exam Admin.Concerns Will Be Discussed at 890731 Mgt Meeting at Region II Ofc IR 05000348/19890111989-07-18018 July 1989 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-348/89-11 & 50-364/89-11 ML20246L8101989-07-13013 July 1989 Forwards Antitrust Settlement Notice Submitted for Fr Publication ML20246B2391989-06-27027 June 1989 Forwards Insp Repts 50-348/89-12 & 50-364/89-12 on 890511-0610.No Violations or Deviations Noted ML20246B0051989-06-16016 June 1989 Forwards Insp Repts 50-348/89-11 & 50-364/89-11 on 890411- 0510 & Notice of Violation ML20245A8551989-06-13013 June 1989 Forwards SER Supporting Util 831104 & 850422 Responses to Generic Ltr 83-28,Item 4.5.3, Reactor Trip Sys Reliability for All Domestic Operating Reactors ML20245F5031989-06-13013 June 1989 Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-348/89-07 & 50-364/89-07 ML20245G3171989-06-13013 June 1989 Requests Completion of Radiochemical Analyses of Liquid Samples Spiked W/Radionuclide within 60 Days of Receipt of Samples ML20248A1751989-06-0202 June 1989 Ack Receipt of 890407 Response to Violations Noted in Insp Repts 50-348/89-03 & 50-364/89-03.Violation Occurred as Stated in Notice of Violation.Written Statement Describing Corrective Actions Requested.Encl Withheld (Ref 10CFR2.790) ML20247L4481989-05-25025 May 1989 Comments on Licensee 881229 Response to Generic Ltr 88-17 W/Respect to Expeditious Actions for Loss of Dhr.Response Appears to Meet Intent of Ltr But Lacks Details for Listed Items ML20247L6451989-05-25025 May 1989 Requests That Licensee Follow Instructions of 890328 Order & Either Request Hearing or Submit Payment of Civil Penalty Re Operation of Safety Injection Pump W/Gas from Crossover Piping.Hydrogen Accumulation Discovered on 880226 Not 06 1990-09-24
[Table view] Category:OUTGOING CORRESPONDENCE
MONTHYEARML20217G0801999-10-0707 October 1999 Informs That on 990930,staff Conducted mid-cycle PPR of Farley & Did Not Identify Any Areas in Which Performance Warranted More than Core Insp Program.Nrc Will Conduct Regional Insps Associated with SG Removal & Installation ML20212J8801999-09-30030 September 1999 Discusses GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps. Util 980731,990607 & 03 Ltrs Provided Requested Info in Subj Gl.Nrc Considers Subj GL to Be Closed for Unit 1 ML20212J8391999-09-30030 September 1999 Forwards RAI Re Request for Amends to Ts.Addl Info Needed to Complete Review to Verify That Proposed TS Are Consistent with & Validate Design Basis Analysis.Request Discussed with H Mahan on 990930.Info Needed within 10 Days of This Ltr ML20212E7031999-09-23023 September 1999 Responds to GL 98-01, Year 2000 Readiness of Computer Sys at Npps. Util Requested to Submit Plans & Schedules for Resolving Y2K-related Issues ML20212F1111999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals ML20212C8041999-09-10010 September 1999 Responds to to D Rathbun Requesting Review of J Sherman Re Y2K Compliance.Latest NRC Status Rept on Y2K Activities Encl ML20212D4581999-09-10010 September 1999 Responds to to D Rathbun,Requesting Review of J Sherman Expressing Concerns That Plant & Other Nuclear Plants Not Yet Y2K Compliant ML20212A8341999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20212A6951999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20211N8041999-09-0808 September 1999 Informs That on 990930 NRC Issued GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Condition, to Holders of Nuclear Plant Operating Licenses ML20211N4301999-09-0808 September 1999 Discusses Proposed Meeting to Discuss Kaowool Fire Barriers. Staff Requesting That Affected Licensees Take Issue on Voluntary Initative & Propose Approach for Resolving Issues ML20212C0071999-09-0202 September 1999 Forwards Insp Repts 50-348/99-05 & 50-364/99-05 on 990627- 0807.No Violations Noted.Licensee Conduct of Activities at Farley Plant Facilities Generally Characterized by safety-conscious Operations & Sound Engineering ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl ML20211G6851999-08-26026 August 1999 Informs That During Insp,Technical Issues Associated with Design,Installation & fire-resistive Performance of Kaowool Raceway fire-barriers Installed at Farley Nuclear Plant Were Identified ML20210T2021999-08-0606 August 1999 Forwards Draft SE Accepting Licensee Proposed Conversion of Plant,Units 1 & 2 Current TSs to Its.Its Based on Listed Documents ML20210Q4641999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Authorized Representative of Facility Must Submit Ltr to La Reyes,As Listed,With List of Individuals to Take exam,30 Days Before Exam Date ML20210J8341999-07-30030 July 1999 Forwards Second Request for Addl Info Re Util 990430 Amend Request to Allow Util to Operate Unit 1,for Cycle 16 Based on risk-informed Probability of SG Tube Rupture & Nominal accident-induced primary-to-second Leakage ML20210G8181999-07-26026 July 1999 Forwards Insp Repts 50-348/99-04 & 50-364/99-04 on 990516- 0626.One Violation Identified & Being Treated as Noncited Violation IR 05000348/19990091999-07-23023 July 1999 Discusses Insp Repts 50-348/99-09 & 50-364/99-09 on 990308- 10 & Forwards Notice of Violation Re Failure to Intercept Adversary During Drills,Contrary to 10CFR73 & Physical Security Plan Requirements ML20196J5781999-07-0202 July 1999 Forwards RAI Re 981201 & s Requesting Amend to TS Associated with Replacing Existing Westinghouse Model 51 SG with Westinghouse Model 54F Generators.Respond within 30 Days of Ltr Date ML20196J6571999-07-0202 July 1999 Discusses Closure to TAC MA0543 & MA0544 Re GL 92-01 Rev 1, Suppl 1,RV Structural Integrity.Nrc Has Revised Rvid & Releasing It as Rvid,Version 2 as Result of Review of Responses ML20196J6191999-07-0202 July 1999 Forwards Final Dam Audit Rept of 981008 of Category 1 Cooling Water Storage Pond Dam.Requests Response within 120 Days of Date of Ltr ML20196J7471999-07-0202 July 1999 Forwards RAI Re Cycle 16 Extension Request.Response Requested within 30 Days of Date of Ltr ML20196J3591999-06-30030 June 1999 Forwards SE of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs ML20196D1931999-06-22022 June 1999 Discusses Requesting Approval & Issuance of Plant Units 1 & 2 ITS by 990930.New Target Date Agrees with Requested Date ML20196H9801999-06-10010 June 1999 Submits Two RAI Re ITS Section 4.0 That Were Never Sent. Reply to RAI Via e-mail ML20196A3401999-06-10010 June 1999 Forwards Insp Repts 50-348/99-03 & 50-364/99-03 on 990404-0515.No Violations Noted ML20206R4741999-05-13013 May 1999 Informs That Staff Reviewed Draft Operation Insp Rept for Farley Nuclear Station Cooling Water Pond Dam & Concurs with FERC Findings.Any Significant Changes Made Prior to Issuance of Final Rept Should Be Discussed with NRC ML20206M2321999-05-11011 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization, Division of Licensing Project Mgt Created ML20206G7341999-05-0404 May 1999 Forwards Safety Evaluation Re Completion of GL 95-07, Pressure Locking & Thermal Binding of Safety-Related Power-Operated Gate Valves ML20206G2411999-04-30030 April 1999 Forwards Revised RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, Program at Farley.Requests to Be Notified of Rev to Original Target Date of 990521 ML20206R2031999-04-29029 April 1999 Forwards Insp Repts 50-348/99-02 & 50-364/99-02 on 990221-0403.No Violations Noted ML20205T1931999-04-0909 April 1999 Informs That on 990316,J Deavers & Ho Christensen Confirmed Initial Operator Licensing Exam Schedule for Farley NPP for Y2K.Initial Exam Dates Are Wks of 000508 & 22 for Approx 12 Candidates.Chief Examiner Will Be C Ernstes ML20205Q1541999-04-0606 April 1999 Forwards Insp Repts 50-348/99-09 & 50-364/99-09 on 990308-10.One non-cited Violation Identified ML20205M2831999-04-0202 April 1999 Forwards Errata Ltr for Farley Nuclear Power Plant FEMA Exercise Rept.Page 19 of Original Rept Should Be Replaced with Encl Corrected Page 19 ML20196K4881999-03-19019 March 1999 Forwards Insp Repts 50-348/99-01 & 50-364/99-01 on 990110- 0220.One Violation of NRC Requirements Occurred.Violation Being Treated as non-cited Violation Consistent with App C of Enforcement Policy ML20205A2611999-03-19019 March 1999 Advises of NRC Planned Insp Effort Resulting from Farley Plant Performance Review on 980202.Historical Listing of Plant Issues & Details of NRC Insp Plan for Next 8 Months Encl ML20204C9561999-03-17017 March 1999 Forwards RAI Re GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Movs, Program at Plant,Units 1 & 2.Response Requested by 990521 ML20204E6601999-03-11011 March 1999 Discusses Ofc of Investigation Rept 2-1998-024 Re Contract Worker Terminated by General Technical Svc Supervisor for Engaging in Protected Activity.Evidence Did Not Substantiate Allegation & No Further Action Planned ML20207M2181999-03-11011 March 1999 Advises That Info Contained in 990125 Application CAW-99-1318 & Affidavit Will Be Withheld from Public Disclosure,Per 10CFR2.790 ML20207M1991999-03-0808 March 1999 Partially Withheld Info Re Meeting Held on 990129 at Region II Ofc in Atlanta,Ga to Discuss Physical Protection Measure for Svc Water Intake Structure Located at Facility (Ref 10CFR73.21).List of Attendees Encl ML20206U1831999-02-0909 February 1999 Responds to Encl Ltrs, & 1223 Re Generic Implication of part-length CRDM Housing Leak.Review Under TAC Numbers MA1380 & MA1381 Considered Closed ML20203G5541999-02-0505 February 1999 Informs That NRC Plans to Administer Generic Fundamentals Exam Section of Written Operator Licensing Exam on 940407. Representative of Facility Must Submit Either Ltr Indicating No Candidates or Listing of Candidates for Exam ML20203G3921999-02-0202 February 1999 Forwards Insp Repts 50-348/98-08 & 50-364/98-08 on 981129- 990109 ML20199L4451999-01-25025 January 1999 Forwards for Review & Comment Revised Draft Info Notice Re Inservice Testing of A-4 Multimatic Deluge Valve for Farley NPP Units 1 & 2.Informs That Comments Submitted on 981120 Were Reviewed & Incorporated Where Appropriate ML20199K7351999-01-21021 January 1999 Responds to 981123 Request by Providing Copy of Latest Draft of Info Notice Being Prepared Which Discusses Failure of Several Preaction Sprinker Sys Deluge Valves.Requests Submittal of Comments by 981120 ML20199K7231999-01-20020 January 1999 Confirms 990119 Telcon Re Informational Meeting Scheduled for 990129 to Be Held in Atlanta,Ga to Discuss Security Related Issues at Plant.Meeting Will Be Closed to Public,Due to Sensitive Nature of Issues ML20199K1381999-01-12012 January 1999 Informs That on 990117,Region II Implemented Staff Reorganization as Part of agency-wide Streamlining Effort, Due to Staffing Reductions in FY99 Budget.Organization Charts Encl ML20199D8541999-01-12012 January 1999 Forwards SE Accepting Relief Request for Inservice Insp Program for Plant,Units 1 & 2 ML20199C9361999-01-0808 January 1999 Forwards Insp Repts 50-348/98-14 & 50-364/98-14 on 981207- 10.No Violations Noted 1999-09-09
[Table view] |
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September 23, 1986 Docket Nos.: 50-348 DISTRIBUTION *Without enclosures and 50-364 O Docket File, *N. Thompson
- Local PDR *D. Miller Mr. R. P. Mcdonald
Senior Vice President *T. Novak
- Gray File Alabama Power Company *0GC-Bethesda Post Office Box 2641 *E. Jordan Birmingham,. Alabama 35291 *B. Grimes
Dear Mr. Mcdonald:
l
SUBJECT:
ANTICIPATED TRANSIENTS WITHOUT SCRAM - JOSEPH M. FARLEY UNITS 1 AND 2 The Nuclear Regulatory Comission (NRC) staff has completed its review of the Westinghouse Owners' Group (h0G) Topical Report WCAP-10858 "AMSAC Generic l Design Package" submitted in response to 10 CFR 50.62 "Reouirements for Re-
! duction of Risk from Anticipated Transient Without Scram (ATWS) Events for
, Light-Water-Cooled Nuclear Power Plants." Guidance for meeting the require-l ments of 10 CFR 50.62 was provided in the preamble to that rule and was further
[
provided to all licensees in Generic Letter 85-06 " Quality Assurance Guidance i
for ATWS Equipment That is Not Safety Related."
l l The results of the staff's review of the generic design for the ATWS mitiga-tion system actuation circuitry (AMSAC) are contained in the attached Safety Evaluction(SE). The staff has concluded that the generic design is acceptable; however, many plant specific details needed in order to ensure conformance with the rule are not addressed by the WOG generic design. These details needed by the NRC to complete the review are defined in the SE.
We request that you review the SE and provide, within 30 days of receipt of this letter, your schedules for addressing the plant specific design features discussed in Appendix A of the SE, and for implementation following the staff's approval of your plant specific design.
This request for information is covered under OMB clearance number 3150-0011 which expires September 30, 1986.
If you have any questions, please contact me at (301) 492-4782.
Sincerely, y signed W
$$k00kOoS SIOob!48 PDR Edward A. Reeves, Project Manager P PWR Project Directorate #2 Division of PWR Licensing-A
Enclosure:
As Stated cc: See next page t kAM2 PM:7 PD: A #2 D iller EReeves:he LRubenstein 9 86 9 /86 9/g/86
Mr. R. P. Mcdonald Alabama Power Company Joseph M. Farley Nuclear Plant cc:
Mr. W. O. Whitt D. Biard MacGuineas, Esquire Executive Vice President Volpe, Boskey and Lyons Alabama Power Company 918 16th Street, N.W.
Post Office Box 2641 Washington, DC 20006 Birmingham, Alabama 35291 Charles R. Lowman Mr. Louis B. Long, General Manager Alabama Electric Corporation Southern Company Services, Inc. Post Office Box 550 Post Office Box 2625 Andalusia, Alabama 36420' Birmingham, Alabama 35202 .
Chairman . Regional Administrator, Region II Houston County Commission U.S. Nuclear Regulatory Commission Dothan, Alabama 36301 101 Marietta Street, Suite 2900 Atlanta, Georgia 30303 Ernest L. Blake, Jr., Esquire Shaw, Pittman, Potts and Trowbridge Claude Earl Fox, M.D.
1800 M Street, N.W. State Health Officer Washington, DC 20036 State Department of Public Health State Office Building Montgomery, Alabama 36130 Robert A. Buettner, Esquire Balch, Bingham, Baker, Hawthorne, Mr. J. D. Woodard Williams and Ward General Manager - Nuclear Plant Post Offict Box 306 Post Office Box 470 Birmingham, Alabama 3520I Ashford, Alabama 36312 Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 24 - Route 2 Columbia, Alabama 36319
EilCLOSURE
, SAFETY EVALUATION OF TOPICAL REPORT (WCAP-10858)
"AM5AC GENERIC DESIGN PACKAGE"
1.0 INTRODUCTION
In response to 10 CFR 50.62 " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nucl ar Power Plants",
Westinghouse on behalf of the Westinghouse Owner's Group (WOG) has submitted for review WCAP-10858 "AMSAC Generic Design Package." This document details the WOG's proposed generic ATWS Mitigation System Actuation Circuitry (AMSAC) designs for compliance with 10 CFR 50.62.
2.0 BACKGROUND
On July 26, 1984 the Code of Federal Regulations (CFR) was amended to include Section 10 CFR 50.62, " Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants" (known as the "ATWS Rule"). An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system (RTS) to shut down the reactor.
The ATWS rule requires specific improvements in the design and operation of com-mercial nuclear power facilities to reduce the likelihood of failure to shut down
[,thereactorfollowinganticipatedtransients,andtomitigatetheconsequencesof en ATWS event.
i
! 3.0 CRITERIA The basic requirement for Westinghouse plants is specified in paragraph (c)(1) of 10 CFR 50.62, "Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, I
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to automatically initiate the auxiliary (or emergency) feedwater system and ini-tiate a turbine trip under conditions indicative of an ATWS. Thi(equipmentmust be designed to perform its function in a reliable manner and be independent (fror sensor output to the final actuation device) from the existing reactor trip system."
},, The criteria used in evaluating the Westinghouse report include; (1) 10 CFR 50.62, (2) guidance anTEfomation published as the preamble to that Rule, and (3)
Generic Letter 85-06 " Quality Assurance Guidance for ATWS Equipment that is not Safety-Related." The evaluation was done on a generic basis, and the relevant criteria is presented below.
The systems and equipment required by 10 CFR 50.62 do not have to meet all of the stringent requirements nomally applied to safety-related equipment. However, this equipment is part of the broader class of structures, systems, and com-ponents defined in the introduction to 10 CFR 50, Appendix A (General Design Criteria).
, GDC-1 requires that " structures, systems, and components important to safety shall be designed, fabricated, erected,.and tested to quality standards commensurate with the importance of the safety functions to be performed." Generic Letter 05-06
" Quality Guidance for ATWS Equipment that is not Safety-Related" details the quality assurance that must be applied to this equipment.
,- w ,v , - . , , - w. - -- a v
3, In general, the equipment to be installed in accordance with the ATWS rule is requiredtobediversefromtheexistingRTS,andmustbetestab1(I at pcwer.
This equipment is intended to provide needed diversity (where only minimal diversity currently exists) to reduce the potential for cocinon mode failures that could result in an ATWS leading to unacceptable plant conditions.
[ The ATWS mitigation design is not required to be safety-related (e.g., meet IEEE-279). HowTv~e~r, the implementation should incorporate good engineering practice and must be such that the existing protection system continues to meet all applicable safety related criteria. Equipment diversity to the extent reasonable and practicable to minimize the potential for comon cause failures is required fro:n the sensors to, but not including the final actuation device.
All mitigating system instrument channel components (excluding sensors and isola-tion devices) must be diverse from the existing RTS. It is desirable, but not.
required, to use sensors and isolation devices that are not part of the RTS.
The basis for not requiring diverse isolators is that the RTS unavailability and AMSAC availability (without a reactor trip signal) are similar with or without the addition of a diverse isolator. Furthermore, with the addition of a new component (e.g., the diverse isolator) within AMSAC, the probability of not get-ting a reactor trip signal or AMSAC signal will be increased somewhat by the additional failure rate of the diverse isolator. However, if existing RTS sen-sors and isolators are utilized, particular emphasis should be placed on the method (s) used to qualify the isolators for their particular function. This 1
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e should include an analysis and tests which will demonstrate that the existing isolator will function under the maximum worst case fault conditions. The required method for qualifying the isolators is presented in Appehir A.
The capability for test and surveillance at power is required, however, sur-veillance frequencies have not been established at this time. During surveil-lance at power, the mitigating system may be bypassed, however, the bypass condi-
,~ tion must be automatically and continuously indicated in the main control room.
The AMSAC system dbsign may also permit bypass of the mitigating function to allow for maintenance, repair, test, or calibration to prevent inadvertent actua-tion of the protective action at the system level. Where operating requirements necessitate automatic or manual bypass of a mitigating system, the design should be such that the bypass will be removed automatically whenever pemissive conditions are not met.
The use of a maintenance bypass should not involve lifting leads, pulling fuses or tripping breakers or physically blocking relays. A permanently installed by-
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pass switch or similar device should be used.
The design should be such that once the ATWS mitigation system has been initiated, the protective action at the system level shall go to completion. Return te operation should require subsequent deliberate operator action.
Manual initiation capability of the mitigating systems at the system level is desirable but not required. Manual initiation should depend upon the operation
5-of a minimum of equipment. The mitigating system should be designed to provide the operator with accurate, complete and timely infomation pertinent to its own status, f
Displays and controls for manual bypass and initiation of the mitigating syster should be integrated into the main control room through system functional ana-lysis and should conform to good human engineering practices in design and j layout. It is important that the' displays and controls added to the control room as a result Of the ATWS rule not increase the potential for operator error.
A humm factor analysis should be perfomed taking into consideration:
(a) the use of this infomation and equipment by an operator during both nomal and abnomal plant conditions, (b) integration into emergency procedures, (c) integration into operator training, and
- m (d) the presence of other alams during an emergency and need for prioritization of alams.
The power supplies are not required to be safety-related but they must be capable of perfoming safety functions with a loss of offsite power, Logic power must be from an ir.strument power supply independent from the power supplies for the existing reactor trip system. Existing RTS sensor and instrument channel power O
C supplies may be used only if the possibility of common mode failure it prevented.
I The most severe ATWS scenarios were determined (see NUREG-0460 Appendix IV. WCAP-8330 and subsequent Westinghouse submittals) to be those in which there was a complete loss of normal feedwater. These included:
Loss of Normal Feedwater/ATWS Transient (LONF/ATWS) k A complete loss of normal feedwater occurs which results from a malfunction in the feedwater condensate system or its control system from such causes as the simultaneous trip of all condensate pumps, the simultaneous trip of all main feedwater pumps or the simultaneous closure of all main feedwater control, pump discharge or block valves.
Because of a postulated common mode failure in the RPS, the reactor is E
incapable of being automatically tripped when any of several plant pro-cess variables have reached their reactor trip setpoints.
. __ Loss of Load /ATWS Transient (LOL/ATWS)
The most severe plant conditions that could result from a loss of load occur following a turbine trip from full power when the turbine trip is caused by a loss of main condenser vacuum. Because of a common mode failure in the protection system, the reactor is incapable of beina automatically tripped as a result of the turbine trip or as the result of any of several other reactor trip signals that occur later in time when several plant process variables reach their reactor trip setpoints.
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Upon loss of the main condenser vacuum, the main feedwater turbine-driven
, i pumpsthatexhaustintothemaincondenseraretripted,thergbycuttingoff feedwater flow to the steam generators. Not all nuclear plants are subject to this transient since many plants have motor-driven main feedwater purps or they have turbine-driven pumps which do not exhaust into the main con-denser. Since there is a complete loss of normal feedwater during both these transients (LONF/ATWS and LOL/ATWS), both transients assumed auxiliary
'k' feedwater (AFW) flow is started 60 seconds after the initiating event for long tem reactor protection. Also the Complete Loss of Nomal Feedwater transient assumed a turbine trip 30 seconds after the initiating event to maintain short tem RCS pressures below 3200 psig. Nomally these features would be actuated by the Reactor Protection System (RPS) and the Engineered Safety Features Actuation System (ESFAS).
The primary safety concern from these two transients is the potential for high pressure within the RCS. If a common mode failure in the RPS and the ESFAS incapacitates AFW flow initiation and/or turbine trip in addition to
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prohibiting a scram, then an alternate method of providing AFW flow and a turbine trip is required to maintain the RCS pressure below 3200 psig.
The final rule which was approved by the Commissioners on November 11, 1983, requires that Westinghouse designed plants install ATWS Mitigating System Actuation Circuitry (AMSAC) to initiate a turbine trip and actuate AFW flow independent of the RPS (from the sensor output). These twe functions, turbine trip and AFW flow actuation, are provided via the AMSAC. .
e 4.0 DESIGN DESCRIPTION I The Westinghouse Owners Group (WOG) has developed generic designs to meet the requirements of 10 CFR 50.62. Three designs were developed which permits each utility to select the design which best fits a particular plant's needs. Factors that may determine the design utilkzed at a plant range from the current control u
4, and protection system design to the ease and cost of installation. The three designs are as follows:
The first design would actuate a turbine trip and auxiliary feedwater flow upon sensing that the steam generator inventory is below the low-low level setpoint.
This logic senses conditions indicative of an ATWS event when a loss of heat sink has occurred but will not actuate until after the reactor protection signals should have been generated. A turbine trip and start-up of all auxiliary feedwater purps will occur upon receipt of an AMSAC signal.
. The steam generator blowdown isolation and sample isolation valves would be automatically closed in all loops when AMSAC is actuated.
The AMSAC signal will be generated by low water level signals in the steam gen-l erators using existing sensor / transmitter units. For two loop plants, AMSAC will l use two channels per loop with 3/4 coincidence to actuate AMSAC. The AMSAC coin- l cidence logic for three loop plants is 2/3 with one channel per steam generator and the four loop plants coincidence logic is 3/4 with one channel per steam generator. .
-9 The AMSAC signal will be automatically blocked below 70% power since short term protection against high reactor coolant system pressure is not required until 70% of nominal power. ThiswillpreventspuriousAMSACactuationfduringstart-up. To ensure that AMSAC remains armed long enough to perform its function in the event of a turbine trip, a C-20 permissive signal will be maintained for approximately 60 seconds. The AMSAC signal will be delayed by approximately 25 seconds to permit the RPS to respond first.
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The second desig'n mitigates the consequences of an ATVS loss of heat sink event by initiating AMSAC on low main feedwater flow measurements.
Actuation of AMSAC will occur on low main feedwater flow as measured by existing main feedwater flow sensor / transmitters. The setpoint to actuate AMSAC is 50t of nominal main feedwater flow. Although 50% flow is more than ample to protect against overpressure in the event of an ATWS, instrumentation error would become unacceptably large if a substantially lower setpoint were used.
, , . . To avoid inadvertent AMSAC actuation on the loss of one main feedwater pump, AMSAC actuation will be delayed approximately 25 seconds to permit the unfaulted main feedwater pump (s) to automatically increase the flow rate to above the AMSA:
actuation setpoint. Recovery in this circumstance is possible since each main feedwater pump is capable of delivering typically 60% of full load capacity.
A turbine trip and start-up of all auxiliary feedwater pumps will occur upon receipt of an AMSAC signal. The steam generator blowdown isolation and sample
e isolation valves should be automatically closed in all loops when AMSAC is actuated.
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1 The AMSAC signal will be generated by low main feedwater flow to the steam I generators. The AMSAC logic is two channels per loop with 3/4 coincidence logic for two loop plants; one channel per loop with 2/3 coincidence logic for three loop plants; and 3/4 coincidence logic for four loop pl, ants.
4 Asinthefirst5esign,theAMSACsignalwillbeautomaticallyblockedbelow 70% power; the AMSAC signal will be delayed by 25 seconds; removal of the C-20 pemissive signal will be delayed by approximately 60 seconds.
The third design determines that conditions indicative of an ATWS event are present by monitoring the feedwater control and isolation valves and the feedwater pump status.
._ Actuation of AMSAC will occur when it has been determined that all main feedwater
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pumps have been tripped or when main feedwater flow to the steam generators has been blocked due to valve closures.
Failures in the main feedwater system upstream of the main feedwater puros that could result in the loss of main feedwater to the steam generators, e.g., trip-ping of all condensate pumps, will result in automatic main feedwater pump trips on low suction pressure. Therefore, explicit actuation of AMSAC based on fail-ures of components. upstream of the main feedwater pumps is not necessary.
11-Since AMSAC anticipates the plant response due to the loss of main feedwater pumps prior to the reactor protection system detecting an anticipated operational oc-currence, it is desirable to delay AMSAC actuation. A 30 second lay is suffi-cient to allow the reactor protection system to respond.
Either of two different AMSAC concepts may be used, depending upon whether or not the main feedwater flow to the steam generators is split during normal power operation. Plants which contain'D-4 and D-5 steam generators have split flow 4' '
during nomal- power operation. All other plants do not, although all plants with preheaters will have a minimal bypass flow through the feedwater bypass temper-ing valve (FBTV). For preheater plants which have split flow during normal power operation, approximately 10 to 20% of the total feedwater flow is passed through the feedwater preheater bypass valves (FPBV), while most of the remaining flow is passed through the feedwater isolation valve (FIV). If all FIVs were to close simultaneously, the flow through the FPBV would increase substantially and still provide protection against RCS overpressurization in the event of an ATWS.
Therefore the accidental closure of all FIVs is not a factor for plants which contain D-4 or D-5 steam generators. All other plants however must account for the accidental closure of all FIVs as well as the accidental closure of all feed-water control valves (FCVs) and the accidental tripping of all main feedwater pumps.
1 A turbine trip and start-up of all auxiliary feedwater pumps will occur upon receipt of an AMSAC signal. The steam generator blowdown isolation and sample
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4 isolation valves should be automatically closed in all loops when AMSAC is actuated.
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The AMSAC signal will be generated by the simultaneous tripping of all main feedwater pumps or the blocking of all main feedwater lines to the steam gen-erators due to valve malfunctions." The AMSAC coincidence logic is as follows:
!' Coincidence FW Valves FW Pumps Loops Closed Tripped 2 3/4 N/N 3 2/3 N/N 4 3/4 N/N where N is the number of main feedwater pumps.
As in the first two designs, the AMSAC signal will be automatically blocked below 70% power and the removal of the C-20 permissive signal shall be delayed by ap-proximately 60 seconds.
5.0 CONCLUSION
Generic The staff has reviewed the Westinghouse Topical Report WCAP-10858, "AMSAC Gen-eric Design Package" and has concluded that the generic designs presented in WCAP-10858 adequat ly meet the requirements of 10 CFR 50.62 and follow the review :
guidelines that have been discussed previously.
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Plant specific j WCAP-10858 presents a generic design, however many details and interfaces are of a plant specific nature. The staff will review the implementation of plant spe-cific designs to evaluate compliance with ATWS rule requirements. Key elements of the plant specific design reviews are denoted below.
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o Di ve rs i ty - --
The plant specific submittal should indicate the degree of diversity that exists between the AMSAC equipment and the existing Reactor Protection System. Equipment diversity to the extent reasonable and practicable to minimize the potential for cannon cause failures is required from the sen-sors output to, but not including, the final actuation device, e.g., exist-a ing circuit breakers may be used for the auxiliary feedwater initiation.
The sensors need not be of a diverse design or manufacture. Existing protection system instrument-sensing lines, sensors, and sensor power
}{' , supplies may be used. Sensor and instrument sensing lines should be selected such that adverse interactions with existing control systems are avoided.
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o Logic power supplies The plant specific submittal should discuss the logic power hpply design.
According to the rule, the AMSAC logic power supply is not required to be safety-related (Class IE). However, logic power should be from an instrument power supply that is independent from the reactor protec-tion system (RPS) power supplies. Our review of additional information submitted by WOG indicated that power to the logic circuits will utilize
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RPS battertes and inverters. The staff finds this portion of the design unacceptable, therefore, independent power supplies should be provided.
o Safety-related interface The plant spacific submittal should show that the implementation is such
, that the existing protection system continues to meet all applicable 4 - safety criteria.
o Quality assurance
',~ The plant specific submittal should provide information regarding com-pliance with Generic Latter 85-06, " Quality Assurance Guidance for ATW5 Equipment that is not Safety-Related."
o Maintenance bypasses The plant specific submittal should discuss how maintenance at power is accomplished and how good human factors engineering practice is incorporated into the continuous indication of bypass status in the control room.
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o Operating bypasses The plant specific submittal should state that operating byphses are continuously indicated in the control room; provide the basis for the 70% or plant specific operating bypass level; discuss the human factors design aspects of the continuous indication; and discuss the diversity and independence of the C-20 pemissive signal (Defeats the block of
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AMSAC).
o Means for bypassing The plant specific submittal should state that the means for bypassing is accomplished with a permanently installed, human factored, bypass switch or similar device, and verify that disallowed methods mentioned in the guidance are not utilized.
s o Manual initiation The plant specific submittal should discuss how a manual turbine
.[, trip and auxiliary feedwater actuation are accomplished by the operator, o Electrical independence from existing reactor protection system The plant specific submittal should show that electrical independence is achieved. This is required from the sensor output to the final actuation device at which point non-safety-related circuits must be isolated from safety related circuits by qualified Class IE isolators. Use of existing isolators is acceptable. However, each plant specific submittal should pro-vide an analysis and tests which demonstrates that the existing isolator will
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function under the maximum worst case fault conditions. The required method for qualifying either the existing or diverse isolators is presented in Appendix A, o Physical separation from existing reactor protection system Physical separation from existing reactor protection system is not required, unless redundant divisions and channels in the existing reactor trip system
- are not physically separated, The implementation must be such that separa-4,-
tion crtteria-applied to the existing protection system are not violated.
The plant specific submittal should respond to this concern.
o Environmental qualification The plant specific submittal should address the environmental qualification of ATWS equipment for anticipated operational occurrences only, not for s accidents.
o Testability at power Measures are to be established' to test, as appropriate, non safety related ATWS equipment prior to installation and periodically. Testing of AMSAC may be perforned with AMSAC in bypass. Testing of AMSAC outputs through the final actuation devices will be performed with the plant shutdown.
The plant specific submittals should present the test program and state that the output signal is indicated in the control room in a manner con-sistent with plant practices including human factors.
o Completion of mitigative action AMSAC shall be designed so that, once actuated, the completiln of mitigating action shall be consistent with the plant turbine trip and auxiliary feed-water circuitry. Plant specific submittals should verify that the pro-tective action, once initiated, goes to completion, and that the subsequent return to operation requires deliberate operator action.
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o Technicti specifications Technical specification requirements related to AMSAC will have to be addressed by plant specific submittels.
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APPENDIX A AMSAC ISOLATION DEVfCE -
REQUEST FOR ADutil0NAL INFORMATION
. e Each light water cooled nuclear reactor shall be provided with a systen for the !
mitigation of the effects from anticipated transients without scrsm'(ATWS). The Comission approved requirements for the ATWS are defined in the Code of Federal Regulations (CFR) Section 10, paragraph 50.62.
The staff has reviewed the Westinghouse Owner's Group generic fun ional AMSAC designs for compliance with the ATWS Rule. As a result, the staff has deter-mined that the use of isolators within AMSAC will be reviewed on a plant specific basis. The following additional infomation is required to continue and con-plete the plant specific isolator review:
Isolation Devices Please provide the following:
o . -
0,' a. For the type of device used to accomplish electrical isolation, describe the spec.ific-testing perfomed to demonstrate that the device is acceptable for its application (s). This description should include elementary diagrams when necessary to indicate the test configuration and how the maximum credible faults were applied to the devices,
- b. Data to verify that the maximum credible faults applied during the test were the maximum voltage / current to which the device could be exposed, and de-fine how the maximum voltage / current was determined,
- c. Data to verify that the maximum credible fault was applied to the output of the device in the transverse mode (between signal and return) and other faults were considered (i.e., open and short circuits).
- d. Define the pass / fail acceptance criteria for each type of device,
- e. Provide a comitment that the isolation devices comply with the environ-ment qualifications (10 CFR 50.49) and with the seismic qualifications which were the basis for plant. licensing.
-. f. Provide a description of the measures taken to protect the safety systems from electrical interference (i.e., Electrostatic Coupling, EMI, Comon Mode and Crosstalk) that may'be generated by the ATWS circuits.
- g. Provide information to verify that the Class IE isolator is powered from a Class IE source.
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