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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217M2871999-10-21021 October 1999
Refers to Rev 5 Submitted in May 1999 for Portions of Byron Nuclear Power Station Generating Stations Emergency Plan Site Annex.Informs That NRC Approval Not Required Based on Determination That Plan Effectiveness Not Decreased
ML20217P6171999-10-21021 October 1999
Forwards non-proprietary & Proprietary Versions of HI-982083, Licensing Rept for Byron & Braidwood Nuclear Stations. Proprietary Rept Withheld,Per 10CFR2.790(b)(4)
ML20217M4361999-10-19019 October 1999
Forwards Rev 46 to Braidwood Station Security Plan, IAW 10CFR50.4(b)(4).Description of Changes,Listed.Encl Withheld Per 10CFR73.21
ML20217H4661999-10-18018 October 1999
Forwards Changes to EPIPs IAW 10CFR50.54(q) & 10CFR50,App E, Section V.Details of Changes Encl
ML20217G9791999-10-14014 October 1999
Forwards SE Accepting Relief Requests to Rev 5 of First 10-year Interval Inservice Insp Program for Plant,Units 1 & 2
ML20217F7891999-10-0808 October 1999
Forwards Insp Repts 50-454/99-12 & 50-455/99-12 on 990803- 0916.One Violation Occurred Being Treated as NCV
ML20212L1791999-10-0505 October 1999
Informs That as Result of Staff Review of Util Responses to GL 92-01,rev 1,suppl 1 & Suppl 1 Rai,Staff Revised Info in Rvid & Is Releasing Rvid Version 2
ML20217B6351999-10-0505 October 1999
Forwards for Info,Final Accident Sequence Precursor Analysis of Operational Event at Byron Station,Unit 1,reported in LER 454/98-018 & NRC Responses to Util Specific Comments Provided in
ML20217B2991999-10-0101 October 1999
Forwards Insp Repts 50-454/99-16 & 50-455/99-16 on 990907-10.No Violations Noted.Water Chemisty Program Was Well Implemented,Resulted in Effective Control of Plant Water Chemistry
ML20216J8241999-09-30030 September 1999
Notifies of Removal of NRC Headquarters & Region III Offices from Controlled Copy Distribution of Certain CE Documents. Specific Documents,Associated Controlled Copy Numbers & NRC Location Affected Are Shown on Attachment to Ltr
ML20212J6751999-09-30030 September 1999
Forwards Replacement Pages Eight Through Eleven of Insp Repts 50-454/99-15 & 50-455/99-15.Several Inaccuracies with Docket Numbers & Tracking Numbers Occurred in Repts
ML20217A5821999-09-29029 September 1999
Advises of NRC Plans for Future Insp Activities at Facility for Licensee to Have Opportunity to Prepare for Insps & to Provide NRC with Feedback on Any Planned Insps Which May Conflict with Plant Activities
ML20217A9311999-09-29029 September 1999
Informs That NRC 6-month Review of Braidwood Identified That Performance in Maint Area Warranted Increased NRC Attention. Addl Insps Beyond Core Insp Program Will Be Conducted Over Next 6 Months to Better Understand Causes of Problem
ML20216H4301999-09-23023 September 1999
Informs That Arrangements Made for Administration of Licensing re-take Exams at Braidwood Generating Station for Week of 991108
ML20216F7441999-09-17017 September 1999
Forwards Insp Repts 50-456/99-13 & 50-457/99-13 on 990706-0824.Three Violations Noted & Being Treated as Ncvs. Insp Focused on C/As & Activities Addressing Technical Concerns Identified During Design Insp Completed on 980424
ML20216F8051999-09-17017 September 1999
Forwards Insp Rept 50-454/99-14 & 50-455/99-14 on 990823-27. Security Program Was Effectively Implemented in Areas Inspected.No Violations Were Identified
ML20212A6991999-09-10010 September 1999
Forwards SE Accepting Licensee Second 10-year Interval ISI Program Request for Relief 12R-07 for Plant,Units 1 & 2
ML20211Q9011999-09-0808 September 1999
Advises That Us Postal Service Mailing Address Has Changed for Braidwood Station.New Address Listed
ML20211P1841999-09-0808 September 1999
Forwards Insp Repts 50-454/99-15 & 50-455/99-15 on 990824- 26.No Violations Noted.Objective of Insp to Determine Whether Byron Nuclear Generating Station Emergency Plan Adequate & If Emergency Plan Properly Implemented
ML20211Q6821999-09-0606 September 1999
Informs That NRC Tentatively Scheduled Initial Licensing Exam for Byron Operator Licesne Applicants During Wks of 000619 & 26.Validation of Exam Will Occur at Station During Wk of 000529
ML20211Q6611999-09-0606 September 1999
Informs That NRC Tentatively Scheduled Initial Licensing Exam for Braidwood Operator License Applicants During Wk of 010115 & 22.Validation of Exam Will Occur at Station During Wk of 001218
ML20211P1901999-09-0303 September 1999
Forwards Insp Repts 50-456/99-12 & 50-457/99-12 on 990707-0816.No Violations Noted.Insp Generally Characterized by safety-conscious Operations,Sound Engineering & Maint Practices & Careful Radiological Work Controls
ML20211N5151999-09-0303 September 1999
Ack Receipt of Re Safety Culture & Overtime Practices at Byron Nuclear Power Station.Copy of Recent Ltr from NRC to Commonwealth Edison Re Overtime Practices & Safety Culture Being Provided
ML20211K1081999-09-0202 September 1999
Responds to Request for Addl Info to GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity, for Braidwood,Units 1 & 2 & Byron,Unit 2
ML20211M1371999-09-0202 September 1999
Discusses 990527 Meeting with Ceco & Byron Station Mgt Re Overtime Practices & Conduciveness of Work Environ to Raising Safety Concerns at Byron Station.Insp Rept Assigned for NRC Tracking Purposes.No Insp Rept Encl
ML20211P1761999-09-0202 September 1999
Discusses Licensee Aug 1998 Rev 3K to Portions of Braidwood Nuclear Power Station Generating Stations Emergency Plan Site Annex Submitted Under Provisions of 10CFR50.54(q). NRC Approval Not Required
ML20211G1221999-08-27027 August 1999
Forwards fitness-for-duty Program Performance Data for Each of Comm Ed Nuclear Power Stations & Corporate Support Employees within Scope of Rule for six-month Period Ending 990630,IAW 10CFR26.71(d)
ML20211G4021999-08-25025 August 1999
Forwards Insp Repts 50-454/99-10 & 50-455/99-10 on 990622-0802.No Violations Noted
ML20211B8691999-08-20020 August 1999
Forwards Insp Repts 50-254/99-10,50-265/99-10,50-454/99-09, 50-455/99-09,50-456/99-10 & 50-457/99-10 on 990628-0721. Action Plans Developed to Address Configuration Control Weaknesses Not Totally Effective as Listed
BW990053, Forwards post-outage Summary Rept for ISI Examinations Conducted During Seventh Refueling Outage of Braidwood Station,Unit 21999-08-13013 August 1999
Forwards post-outage Summary Rept for ISI Examinations Conducted During Seventh Refueling Outage of Braidwood Station,Unit 2
BW990052, Informs That RW Clay,License OP-31044,no Longer Requires Operator License at Braidwood Station1999-08-12012 August 1999
Informs That RW Clay,License OP-31044,no Longer Requires Operator License at Braidwood Station
05000454/LER-1998-008, Informs That Licensee Determined That Suppl Rept to LER 98-008 Is Not Warranted.No Addl Info Was Generated Following Completion of Root Cause Investigation of Following Completion of Corrective Actions Stated in Original LER1999-08-12012 August 1999
Informs That Licensee Determined That Suppl Rept to LER 98-008 Is Not Warranted.No Addl Info Was Generated Following Completion of Root Cause Investigation of Following Completion of Corrective Actions Stated in Original LER
ML20210N5651999-08-0606 August 1999
Forwards Rev 8 to Nuclear Generating Stations Emergency Plan, for Plants.With Summary of Changes
ML20210U8031999-08-0404 August 1999
Forwards SER Granting Licensee Relief Requests VR-1,VR-3 & Portion of VR-2 Pursuant to 10CFR50.55a(a)(3)(ii).Relief Request VR-4 Does Not Require Explicit NRC Approval for Second 10-year Inservice Testing Program
BW990049, Informs NRC of Plans to Demonstrate Compliance with 10CFR50.46 Requirements for Fuel Predicted to Experience Fuel Pellet to Rod Cladding Gap Reopening,During Current Cycle1999-08-0404 August 1999
Informs NRC of Plans to Demonstrate Compliance with 10CFR50.46 Requirements for Fuel Predicted to Experience Fuel Pellet to Rod Cladding Gap Reopening,During Current Cycle
ML20210M9131999-08-0202 August 1999
Forwards Response to NRC AL 99-02, Operating Reactor Licensing Action Estimates, for Fys 2000 & 2001 for Comed
ML20210K0771999-07-30030 July 1999
Submits 30-day Rept Re Discovery of ECCS Evaluation Model Error for Byron & Braidwood Stations,As Required by 10CFR50.46
ML20210K9761999-07-30030 July 1999
Forwards SE Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design Basis Capability of Safety-Related Movs, for Plant
ML20210G6291999-07-29029 July 1999
Forwards Insp Repts 50-456/99-11 & 50-457/99-11 on 990525-0706.Two Violations Noted & Being Treated as NCV, Consistent with App C of Enforcement Policy
ML20210J8951999-07-29029 July 1999
Submits Other Actions,As Described,To Be Taken for Valves to Resolve Potential Pressure Locking Concerns,In Light of Extended Period for Valve Bonnet Natural Depressurization,In Response to GL 95-07, Pressure Locking & Thermal..
BW990045, Forwards Errata to 1998 Radioactive Effluent Release Rept. Info Has Been Corrected & Revised Spreadsheets Included in Attachment to Ltr1999-07-28028 July 1999
Forwards Errata to 1998 Radioactive Effluent Release Rept. Info Has Been Corrected & Revised Spreadsheets Included in Attachment to Ltr
ML20210E2151999-07-23023 July 1999
Forwards Byron Unit 1 B1R09 ISI Summary Rept Spring 1999 Outage,980309-990424, in Compliance with Requirements of Article IWA-6000, Records & Repts of Section XI of ASME & P&PV,1989 Edition
ML20216D3781999-07-21021 July 1999
Forwards Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR on ITS Format & W(Z) Function, to Account for Error That W Discovered in Computer Code Used to Calculate PCT During LBLOCA
ML20210C3961999-07-20020 July 1999
Forwards Insp Repts 50-456/99-09 & 50-457/99-09 on 990517-0623.No Violations Noted.Weakness Identified on 990523,when Station Supervisors Identified Individual Sleeping in Cable Tray in RCA
ML20216D7061999-07-19019 July 1999
Forwards Rev 45 to Braidwood Station Security Plan,Iaw 10CFR50.4(b)(4).Plan Includes Listed Changes.Rev Withheld, Per 10CFR73.21
BW990042, Forwards Braidwood Station,Unit 1 Post Accident Monitoring Rept for Reactor Vessel Level Indication Sys (Rvlis),Due to Facility Train B RVLIS Being Restored to Operable Status After 7-day Completion Time,Per TS 3.3.3 & 5.6.71999-07-16016 July 1999
Forwards Braidwood Station,Unit 1 Post Accident Monitoring Rept for Reactor Vessel Level Indication Sys (Rvlis),Due to Facility Train B RVLIS Being Restored to Operable Status After 7-day Completion Time,Per TS 3.3.3 & 5.6.7
ML20210B7071999-07-16016 July 1999
Responds to Requesting Review & Approval of Three Proposed Changes to Ceco QA TR,CE-1A Per 10CFR50.54(a)(3) & 10CFR50.4(b)(7)
ML20210A3151999-07-16016 July 1999
Forwards Insp Repts 50-454/99-08 & 50-455/99-08 on 990511-0621.Three Violations Being Treated as Noncited Violations
ML20209H2991999-07-16016 July 1999
Withdraws 980529 LAR to Credit Automatic PORV Operation for Mitigation of Inadvertent Safety Injection at Power Accident.Response to NRC 990513 RAI Re LAR Encl
BW990040, Forwards Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2.Since Issuance of Rept,It Was Determined That Rt That Occurred on Unit 2 During Startup Was Inadvertently Omitted1999-07-15015 July 1999
Forwards Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2.Since Issuance of Rept,It Was Determined That Rt That Occurred on Unit 2 During Startup Was Inadvertently Omitted
1999-09-08
[Table view]
Category:INCOMING CORRESPONDENCE
MONTHYEARML20217P6171999-10-21021 October 1999
Forwards non-proprietary & Proprietary Versions of HI-982083, Licensing Rept for Byron & Braidwood Nuclear Stations. Proprietary Rept Withheld,Per 10CFR2.790(b)(4)
ML20217M4361999-10-19019 October 1999
Forwards Rev 46 to Braidwood Station Security Plan, IAW 10CFR50.4(b)(4).Description of Changes,Listed.Encl Withheld Per 10CFR73.21
ML20217H4661999-10-18018 October 1999
Forwards Changes to EPIPs IAW 10CFR50.54(q) & 10CFR50,App E, Section V.Details of Changes Encl
ML20216J8241999-09-30030 September 1999
Notifies of Removal of NRC Headquarters & Region III Offices from Controlled Copy Distribution of Certain CE Documents. Specific Documents,Associated Controlled Copy Numbers & NRC Location Affected Are Shown on Attachment to Ltr
ML20211Q9011999-09-0808 September 1999
Advises That Us Postal Service Mailing Address Has Changed for Braidwood Station.New Address Listed
ML20211G1221999-08-27027 August 1999
Forwards fitness-for-duty Program Performance Data for Each of Comm Ed Nuclear Power Stations & Corporate Support Employees within Scope of Rule for six-month Period Ending 990630,IAW 10CFR26.71(d)
BW990053, Forwards post-outage Summary Rept for ISI Examinations Conducted During Seventh Refueling Outage of Braidwood Station,Unit 21999-08-13013 August 1999
Forwards post-outage Summary Rept for ISI Examinations Conducted During Seventh Refueling Outage of Braidwood Station,Unit 2
05000454/LER-1998-008, Informs That Licensee Determined That Suppl Rept to LER 98-008 Is Not Warranted.No Addl Info Was Generated Following Completion of Root Cause Investigation of Following Completion of Corrective Actions Stated in Original LER1999-08-12012 August 1999
Informs That Licensee Determined That Suppl Rept to LER 98-008 Is Not Warranted.No Addl Info Was Generated Following Completion of Root Cause Investigation of Following Completion of Corrective Actions Stated in Original LER
BW990052, Informs That RW Clay,License OP-31044,no Longer Requires Operator License at Braidwood Station1999-08-12012 August 1999
Informs That RW Clay,License OP-31044,no Longer Requires Operator License at Braidwood Station
ML20210N5651999-08-0606 August 1999
Forwards Rev 8 to Nuclear Generating Stations Emergency Plan, for Plants.With Summary of Changes
BW990049, Informs NRC of Plans to Demonstrate Compliance with 10CFR50.46 Requirements for Fuel Predicted to Experience Fuel Pellet to Rod Cladding Gap Reopening,During Current Cycle1999-08-0404 August 1999
Informs NRC of Plans to Demonstrate Compliance with 10CFR50.46 Requirements for Fuel Predicted to Experience Fuel Pellet to Rod Cladding Gap Reopening,During Current Cycle
ML20210M9131999-08-0202 August 1999
Forwards Response to NRC AL 99-02, Operating Reactor Licensing Action Estimates, for Fys 2000 & 2001 for Comed
ML20210K0771999-07-30030 July 1999
Submits 30-day Rept Re Discovery of ECCS Evaluation Model Error for Byron & Braidwood Stations,As Required by 10CFR50.46
ML20210J8951999-07-29029 July 1999
Submits Other Actions,As Described,To Be Taken for Valves to Resolve Potential Pressure Locking Concerns,In Light of Extended Period for Valve Bonnet Natural Depressurization,In Response to GL 95-07, Pressure Locking & Thermal..
BW990045, Forwards Errata to 1998 Radioactive Effluent Release Rept. Info Has Been Corrected & Revised Spreadsheets Included in Attachment to Ltr1999-07-28028 July 1999
Forwards Errata to 1998 Radioactive Effluent Release Rept. Info Has Been Corrected & Revised Spreadsheets Included in Attachment to Ltr
ML20210E2151999-07-23023 July 1999
Forwards Byron Unit 1 B1R09 ISI Summary Rept Spring 1999 Outage,980309-990424, in Compliance with Requirements of Article IWA-6000, Records & Repts of Section XI of ASME & P&PV,1989 Edition
ML20216D3781999-07-21021 July 1999
Forwards Revised NFM9900022, Braidwood Unit 2 Cycle 8 COLR on ITS Format & W(Z) Function, to Account for Error That W Discovered in Computer Code Used to Calculate PCT During LBLOCA
ML20216D7061999-07-19019 July 1999
Forwards Rev 45 to Braidwood Station Security Plan,Iaw 10CFR50.4(b)(4).Plan Includes Listed Changes.Rev Withheld, Per 10CFR73.21
BW990042, Forwards Braidwood Station,Unit 1 Post Accident Monitoring Rept for Reactor Vessel Level Indication Sys (Rvlis),Due to Facility Train B RVLIS Being Restored to Operable Status After 7-day Completion Time,Per TS 3.3.3 & 5.6.71999-07-16016 July 1999
Forwards Braidwood Station,Unit 1 Post Accident Monitoring Rept for Reactor Vessel Level Indication Sys (Rvlis),Due to Facility Train B RVLIS Being Restored to Operable Status After 7-day Completion Time,Per TS 3.3.3 & 5.6.7
ML20209H2991999-07-16016 July 1999
Withdraws 980529 LAR to Credit Automatic PORV Operation for Mitigation of Inadvertent Safety Injection at Power Accident.Response to NRC 990513 RAI Re LAR Encl
BW990040, Forwards Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2.Since Issuance of Rept,It Was Determined That Rt That Occurred on Unit 2 During Startup Was Inadvertently Omitted1999-07-15015 July 1999
Forwards Revised Monthly Operating Repts for May 1999 for Braidwood Station,Units 1 & 2.Since Issuance of Rept,It Was Determined That Rt That Occurred on Unit 2 During Startup Was Inadvertently Omitted
ML20207H7501999-07-12012 July 1999
Forwards Revised Pressure Temp Limits Rept, for Byron Station,Units 1 & 2.Revised Pressurized Thermal Shock Evaluations,Surveillance Capsule Rept & Credibility Repts, Also Encl
ML20209G1391999-07-0909 July 1999
Forwards Results of SG Tube Insps Performed During Byron Station,Unit 1,Cycle 9 Refueling Outage within 12 Months Following Completion of Insps
ML20196J9061999-07-0101 July 1999
Provides Evidence That Util Maintains Guarantee of Payment of Deferred Premiums in Amount of $10 Million for Each of Thirteen Reactors,Per 10CFR140.21
ML20196J9131999-07-0101 July 1999
Submits Status of Nuclear Property Insurance Currently Maintained for Braidwood,Byron,Dresden,Lasalle County,Quad Cities & Zion Nuclear Power Stations,Per 10CFR50.54(w)(3)
ML20209B8241999-06-30030 June 1999
Forwards Five 3.5 Inch Computer Diskettes Containing Revised Annual Dose Repts for 1994 Through 1998 for Individuals Receiving Neutron Dose Not Previously Included in Reported Total Effective Dose Equivalent Values.Without Diskettes
ML20196G2161999-06-25025 June 1999
Forwards for NRC Region III Emergency Preparedness Inspector,Two Copies of Comed Emergency Preparedness Exercise Manual for 1999 Byron Station Annual Exercise. Exercise Is Scheduled for 990825.Without Encls
ML20209D4861999-06-17017 June 1999
Informs That R Heinen,License OP-30953-1 & a Snow,License SOP-30212-3,no Longer Require License at Byron Station
05000456/LER-1998-004, Forwards LER 98-004-01,IAW 10CFR50.73(a)(2)(i)(B). LER 98-004 Included Commitment to Transmit Supplemental Rept by 990628,due to on-going Evaluations1999-06-16016 June 1999
Forwards LER 98-004-01,IAW 10CFR50.73(a)(2)(i)(B). LER 98-004 Included Commitment to Transmit Supplemental Rept by 990628,due to on-going Evaluations
05000457/LER-1998-003, Forwards LER 98-003-00 Re Unit 2 Reactor Trip.Actions & Associated Action Tracking Number That Braidwood Station Is Committed to Implement in Response to LER Described Below1999-06-16016 June 1999
Forwards LER 98-003-00 Re Unit 2 Reactor Trip.Actions & Associated Action Tracking Number That Braidwood Station Is Committed to Implement in Response to LER Described Below
05000456/LER-1999-001, Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(i)(B). Description of Action & Associated Action Request Number That Braidwood Station Is Committed to Implement Is Response to LER Is Listed1999-06-15015 June 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(i)(B). Description of Action & Associated Action Request Number That Braidwood Station Is Committed to Implement Is Response to LER Is Listed
BW990028, Forwards Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function, IAW TS 5.6.51999-06-10010 June 1999
Forwards Braidwood Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function, IAW TS 5.6.5
05000454/LER-1999-002, Forwards LER 99-002-00,IAW 10CFR50.73(a)(2)(i)(b).There Are Two Actions Remaining to Address Cause of Event.Both Actions Are Listed1999-06-0808 June 1999
Forwards LER 99-002-00,IAW 10CFR50.73(a)(2)(i)(b).There Are Two Actions Remaining to Address Cause of Event.Both Actions Are Listed
ML20195E3451999-06-0707 June 1999
Forwards 3.5 Inch Computer Diskette Containing Revised File Format for Annual Dose Rept for 1998,per 990520 Telcon Request from Nrc.Each Station Data Is Preceded by Header Record,Which Provides Info Necessary to Identify Data
ML20195D6351999-06-0404 June 1999
Notifies NRC of Actions That Has Been Taken in Accordance with 10CFR26, Fitness for Duty Programs
ML20211M1611999-05-28028 May 1999
Discusses 990527 Meeting with Comed Re Safety Culture & Overtime Control at Byron Nuclear Plant from Videoconference Location at NRC Headquarters.Requests That Aggressive Actions Be Taken to Ensure That Comed Meets Expectations
ML20207D5261999-05-26026 May 1999
Forwards Response to NRC 990318 RAI Concerning Alleged Chilling Effect at Byron Station.Attachment Contains Responses to NRC 12 Questions
ML20195C7911999-05-25025 May 1999
Forwards Revised COLR for Byron Unit 2,IAW 10CFR50.59.Rev Accounts for Planned Increase of Reactor Coolant Full Power Average Operating Temp from 581 F to 583 F
ML20211M1781999-05-25025 May 1999
Summarizes Concerns with Chilling Effect & Overtime Abuses at Commonwealth Edison,Byron Station.Request That Ltr Be Made Part of Permanent Record of 990527 Meeting
05000457/LER-1999-002, Forwards LER 99-002-00,IAW 10CFR50.73(a)(2)(iv).Commitments Made by Util Are Listed1999-05-21021 May 1999
Forwards LER 99-002-00,IAW 10CFR50.73(a)(2)(iv).Commitments Made by Util Are Listed
05000454/LER-1999-001, Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(i)(B).Required Actions to Address Causes of Event Listed1999-05-21021 May 1999
Forwards LER 99-001-00,IAW 10CFR50.73(a)(2)(i)(B).Required Actions to Address Causes of Event Listed
ML20195B2301999-05-19019 May 1999
Requests Approval of Proposed Changes to QA Topical Rept CE-1-A,rev 66a.Attachment a Describes Changes,Reason for Change & Basis for Concluding That Revised QAP Incorporating Proposed Changes Continues to Satisfy 10CFR50AppB
ML20207E9831999-05-18018 May 1999
Forwards Copy of Commonwealth Edison Co EP Exercise Evaluation Objectives for 1999 Byron Station Annual EP Exercise,Which Will Be Conducted on 990825.Without Encl
ML20206T3351999-05-17017 May 1999
Provides Written follow-up of Request for NOED Re Extension of Shutdown Requirement of TS Limiting Condition for Operation 3.0.3.Page 9 of 9 of Incoming Submittal Not Included
ML20206N7861999-05-14014 May 1999
Forwards 1998 Annual Radiological Environ Operating Rept for Braidwood Station. Rept Contains Info Associated with Stations Radiological Environ & Meteorological Monitoring Programs
ML20206Q8521999-05-13013 May 1999
Submits Rept on Numbers of Tubes Plugged or Repaired During SG Inservice Insp Activities Conducted During Plant Seventh Refueling outage,A2R07,per TS 5.6.9
ML20206N8551999-05-11011 May 1999
Forwards 1998 Annual Radioactive Environ Operating Rept for Byron Station. Rept Includes Summary of Radiological Liquid & Gaseous Effluents & Solid Waste Released from Site
ML20210C7221999-05-0303 May 1999
Forwards Initial License Exam Matls for Review & Approval. Exam Scheduled for Wk of 990607
ML20206U3351999-04-30030 April 1999
Forwards Evaluation of Matter Described in Re Byron Station.Concludes That Use of Overtime at Byron Station Was Controlled IAW Administrative Requirements & Mgt Expectations Established to Meet Overtime Requirement of TS
ML20206F5381999-04-30030 April 1999
Forwards Magnetic Tape Containing Annual Dose Repts for 1998 for Braidwood,Byron,Dresden,Lasalle County,Quad Cities & Zion Nuclear Power Stations,Per 10CFR20.2206(c).Without Magnetic Tape
1999-09-08
[Table view]
Category:UTILITY TO NRC
MONTHYEARML20059K6741990-09-17017 September 1990
Suppls Responses to Violations Noted in Insp Repts 50-454/89-11,50-455/89-13,50-456/89-11 & 50-457/89-11. Corrective Actions:Procedures Changed & Valve Tagging Status Provided
ML20059K5081990-09-14014 September 1990
Forwards Tj Kovach to E Delatorre Re Visit by Soviet Delegation to Braidwood Nuclear Station in May 1990
ML20059L6611990-09-10010 September 1990
Forwards Byron Station Units 1 & 2 Inservice Insp Program
ML20064A3751990-08-24024 August 1990
Forwards Revised Pages to Operating Limits Rept for Cycle 2, Correcting Fxy Portion of Rept,Per Tech Spec 6.9.1.9, Operating Limits Rept
ML20064A3681990-08-24024 August 1990
Forwards Response to 900517 Request for Addl Info Re Design of Containment Hydrogen Monitoring Sys.Util Proposes Alternative Design That Ensures Both Containment Isolation & Hydrogen Monitoring Sys Operability in Event of LOCA
ML20064A0181990-08-16016 August 1990
Submits Supplemental Response to NRC Bulletin 88-008,Suppls 1 & 2.Surveillance Testing Revealed No Leakage,Therefore Charging Pump to Cold Leg Outage Injection Lines Would Not Be Subjected to Excessive Thermal Stresses
ML20059A3991990-08-15015 August 1990
Forwards Response to NRC 900521 Request for Addl Info Re Plant Inservice Insp Program
ML20063Q1051990-08-10010 August 1990
Forwards Monthly Operating Repts for Jul 1990 for Byron Units 1 & 2 & Corrected Monthly Operating Rept for June 1990 for Unit 2
ML20058N0551990-08-0707 August 1990
Provides Supplemental Response to NRC Bulletin 88-008, Suppls 1 & 2.Surveillance Testing Performed Revealed No Leakage,Therefore,Charging Pump to Cold Leg Injection Lines Would Not Be Subjected to Excessive Thermal Stresses
ML20056A3351990-08-0202 August 1990
Responds to NRC Bulletin 88-009 Requesting That Addressees Establish & Implement Insp Program to Periodically Confirm in-core Neutron Power Reactors.All Timble Tubes Used at Plant Inspected & 18 Recorded Evidence of Degradation
ML20055H7631990-07-25025 July 1990
Forwards Financial Info Re Decommissioning of Plants
ML20055J1221990-07-25025 July 1990
Notifies That Plants Current Outage Plannings Will Not Include Removal of Snubbers.Removal of Snubbers Scheduled for Future Outages.Completion of Review by NRC by 900801 No Longer Necessary
ML20055J1261990-07-25025 July 1990
Notifies That Replacement of 13 Snubbers w/8 Seismic Stops on Reactor Coolant Bypass Line Being Deferred Until Later Outage,Per Rl Cloud Assoc Nonlinear Piping Analyses
ML20055H0291990-07-17017 July 1990
Forwards Revised Monthly Performance Rept for Braidwood Unit 2 for June 1990
ML20055G3251990-07-16016 July 1990
Responds to SALP Board Repts 50-454/90-01 & 50-455/90-01 for Reporting Period Nov 1988 - Mar 1990.Effort Will Be Made to Continue High Level of Performance in Areas of Radiological Controls,Plant Operations,Emergency Preparedness & Security
ML20055G4631990-07-13013 July 1990
Responds to NRC Re Violations Noted in Insp Repts 50-456/90-08 & 50-457/90-08.Corrective Actions:Discrepancy Record for Cable Generated & Cable That Had Been Previously Approved for Use on Solenoid Obtained & Installed
ML20044A9621990-07-13013 July 1990
Forwards Rev 0 to Topical Rept NFSR-0081, Comm Ed Topical Rept on Benchmark of PWR Nuclear Design Methods Using PHOENIX-P & Advanced Nodal Code (Anc) Computer Codes, in Support of Implementation of PHOENIX-P & Anc
ML20044B1411990-07-12012 July 1990
Forwards Addl B&W Rept 77-1159832-00 to Facilitate Completion of Reviews & Closeout of Pressurized Thermal Shock Issue,Per NRC Request
ML20044B2141990-07-11011 July 1990
Withdraws 891003 Amend Request to Allow Sufficient Time to Reevaluate Technical Position & Develop Addl Technical Justification
ML20044B2081990-07-11011 July 1990
Responds to Generic Ltr 90-04 Re Status of GSI Resolved W/ Imposition of Requirements or Corrective Actions.Status of GSI Implementation Encl
ML20044A9521990-07-10010 July 1990
Provides Supplemental Response to NRC Bulletin 88-001. Remaining 48 Breakers Inspected During Facility Spring Refueling Outage
ML20044B2871990-07-0909 July 1990
Forwards Brief Description of Calculations Performed in Accordance W/Facility Procedure Used to Make Rod Worth Measurements,Per NUREG-1002 & Util 900629 Original Submittal
ML20044A7991990-06-29029 June 1990
Forwards Description of Change Re Design of Containment Hydrogen Monitoring Sys,Per 900517 Request.Util Proposing Alternative Design Ensuring Containment & Hydrogen Monitoring Sys Operability in Event of Power Loss
ML20055D4811990-06-29029 June 1990
Discusses Revised Schedule for Implementation of Generic Ltr 89-04 Re Frequently Identified Weaknesses of Inservice Testing Programs.All Procedure Revs Have Either Been Approved or Drafted & in Onsite Review & Approval Process
ML20058K3521990-06-22022 June 1990
Requests Withdrawal of 900315 Application for Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-77,changing Tech Specs 3.8.1.1 & 4.8.1.1.2 to Clarify How Gradual Loading of Diesel Generator Applied to Minimize Mechanical Stress on Diesel
ML20055D2951990-06-22022 June 1990
Discusses Results of 900529-0607 Requalification Exam.Based on Results of Exam,Station Removed/Prohibited Both Shift & Staff Teams & JPM Failure from License Duties.Shift Team Placed in Remediation Program from 900611-14
ML20056A0361990-06-15015 June 1990
Responds to NRC Re Violations Noted in Insp Repts 50-456/90-10 & 50-457/90-11.Corrective Action:Valve 2CS021b Returned & Locked in Throttle Position & Out of Svc Form Bwap 330-1T4 Modified
ML20043G5851990-06-0808 June 1990
Forwards Repts Re Valid & Invalid Test Failures Experienced on Diesel Generator (DG) 1DG01KB,1 Valid Test Failure on DG 2DGO1KA & 2 Invalid Test Failures Experienced on DG 2AGO1KB
ML20043D3151990-06-0101 June 1990
Forwards Rev 30 to Security Plan.Rev Withheld (Ref 10CFR73.21)
ML20043D3141990-06-0101 June 1990
Forwards Rev 18 to Security Plan.Rev Withheld (Ref 10CFR73.21)
ML20043E3141990-05-31031 May 1990
Withdraws 880302 Application for Amend to Licenses NPF-37, NPF-66,NPF-72 & NPF-77,changing Tech Spec 4.6.1.6.1.d to Reduce Containment Tendon Design Stresses to Incorporate Addl Design Margin,Due to Insufficient Available Data
ML20043F4731990-05-30030 May 1990
Forwards Suppl to 881130 Application for Amends to Licenses NPF-37,NPF-66,NPF-72 & NPF-77.Changes Requested Per Generic Ltr 87-09,to Remove Unnecessary Restrictions on Operational Mode Changes & Prevent Plant Shutdowns
ML20043C8641990-05-29029 May 1990
Forwards Rept of Local Leakage Rate Test Results for Third Refueling Outage.Leakage Rates of Six Valves Identified as Contributing to Failure of Max Pathway Limit
ML20043B7771990-05-23023 May 1990
Forwards Endorsement 9 to Nelia & Maelu Certificates N-108 & M-108 & Endorsement 8 to Nelia & Maelu Certificates N-115 & M-115
ML20043B7691990-05-23023 May 1990
Forwards Endorsement 11 to Nelia & Maelu Certificates N-93 & M-93 & Endorsement 9 to Nelia & Maelu Certificates N-101 & M-101
ML20043A9161990-05-16016 May 1990
Provides Advanced Notification of Change That Will Be Made to Fire Protection Rept Pages 2.2-18 & 2.3-14
ML20043C2811990-05-15015 May 1990
Responds to NRC 900416 Ltr Re Violations Noted in Insp Repts 50-456/90-09 & 50-457/90-09.Corrective Actions:Gas Partitioners Tested Following Maint During Mar 1990 & Tailgate Training Session Will Be Held
ML20043A6391990-05-11011 May 1990
Submits Revised Schedule for Implementation of Generic Ltr 89-04 Guidance.Rev to Procedures for Check Valve & Stroke Time Testing of power-operated Valves Will Be Completed by 900629
ML20043A2891990-05-10010 May 1990
Forwards Monthly Operating Rept for Apr 1990 & Corrected Rept for Mar 1990 for Byron Nuclear Power Station
ML20042G7111990-05-0707 May 1990
Responds to NRC Questions Re leak-before-break Licensing Submittal for Stainless Steel Piping.Kerotest Valves in Rh Sys Will Be Replaced in Byron Unit 2 During Next Refueling Outage Scheduled to Begin on 900901
ML20042F6851990-05-0404 May 1990
Requests Resolution of Util 870429,880202 & 0921 & 890130 Submittals Re Containment Integrated Leak Rate Testing in Response to Insp Repts 50-454/86-35 & 50-455/86-22 by 900608
ML20042F6771990-05-0303 May 1990
Advises NRC of Util Plans Re Facility Cycle 2 Reload Core. Plant Cycle 2 Reload Design,Including Development of Core Operating Limits Has Been Generated by Util Using NRC Approved Methodology,Per WCAP-9272-P-A
ML20042E9601990-04-30030 April 1990
Forwards Response to NRC 900327 Ltr Re Violations Noted in Insp Repts 50-454/90-09 & 50-455/90-08.Response Withheld (Ref 10CFR73.21)
ML20042G3591990-04-30030 April 1990
Forwards Errata to Radioactive Effluent Rept for Jul-Dec 1989,including Info Re Sr-89,Sr-90 & Fe-55 Analysis for Liquid & Gaseous Effluents Completed by Offsite Vendor
ML20055C5761990-04-30030 April 1990
Forwards Results of Investigation in Response to Allegation RIII-90-A-0011 Re Fitness for Duty.W/O Encl
ML20042E9111990-04-25025 April 1990
Forwards Rev 1 to Nonproprietary & Proprietary, Steam Generator Tube Rupture Analysis for Byron & Braidwood Plants.
ML20042F2681990-04-18018 April 1990
Provides Supplemental Response to Violation Noted in Insp Repts 50-456/89-21 & 50-457/89-21 Re Safeguards Info.Util Request Extension of 891010 Commitment Re Reviews of Plants. List of Corrective Actions Will Be Submitted by 900601
ML20042F0241990-03-28028 March 1990
Forwards Part 3 of 1989 Operating Rept.W/O Rept
ML20012E1081990-03-21021 March 1990
Forwards Calculations Verifying Operability of Facility Dc Battery 111 W/Only 57 of 58 Cells Functional & Onsite Review Notes,Per Request
ML20012D8671990-03-21021 March 1990
Reissued 900216 Ltr,Re Changes to 891214 Rev 1 to Updated Fsar,Correcting Ltr Date
1990-09-17
[Table view] |
Text
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/
.- { [ N /7Commonwealth s
v Edison One Fird Natior.al Plaza, Chicago, IHinois Address Reply to: Post Office Box 767
. (j Chicago, Ininois 60690 0767 July 27, 1988 Mr. Thomas E. Murley, Director Of fice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn Document Control Desk
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 Regulatory Guide 1.97 Compliance NRC Docket Nos. 50-454/455 and 50-456/457 References (a): February 27, 1987 K.A. Ainger letter to H.R. Denton (b): September 1, 1987 S.C. Hunsader letter to T.E. Murley (c): November 6, 1987 L. Olshan letter to L.D. Butterfield (d): December 28, 1987 S.C. Hunsader letter to T.E. Hurley (e): May 5, 1988 L. Olshan letter to H.E. Bliss
Dear Mr. Murley:
Reference (a) provided our preliminary evaluation of the Byron and Braidwood instrumentation for compliance with Regulatory Guide 1.97, Revision 3. Reference (b) provided the results of our final evaluation which were included as Attachments to that letter. Attachment A to reference (b) provided an update of Table 5-1. Attachment B to reference (b) provided the revised Human Factors Engineering Review. These supplemented the information provided in reference (a) and, together, constituted our "Final Report" concerning Regulatory Guide 1.97 compliance.
Reference (c) provided the results of the initial NRC review and evaluation of references (a) and (b). Commonwealth Edison's (Edison) response to these was provided in reference (d). In reference (e), the NRC staff indicated that it had completed its review and had concluded that Edison's justification for the exceptions to Regulatory Guide 1.97, Revision 3 for some items was acceptable. However, for other items, additional justification was required.
I (
8800110146 8 O S4 PDR ADOCK O PNV P
i
T. E. Murley December 28, 1987 In response to the NRC staff's review, this letter provides the additional justification or provides an appropriate response to the NRC staff's concerns. These are included in Attachment "A". Please address any questions regarding this matter to this office.
Very truly yours,
,{. ,- - Y1 .-,
______m ._.
S. C. Huusader Nuclear Licensing Administrator cca L. Olshan (NRR)
S. Seads (NRR)
W. F...ney (NRC RIII)
Byron Resident Inspector Braidwood Resident Inspector 4036K
- e .
'N .
TECHNICAL EVALUATION REPORT ITEM 3.3.5 Residual Heat Removal (RHR) System Flow Regulatory Guide 1.97 recommends environmentally qualified instrumentation for this variable. The instrumentation installed to measure the RHR flow is not environmentally qualified. The licensee states that this instrumentation is back-up instrumentation for the key variables to monitor the operation of the RHR system: ESF status monitoring lights, component status indication, pump motor anneters, and the refueling water storage tank level and the reactor vessel level indications.
The licensee has not verified that these key variables are all monitored by Category 2 instrumentation. Therefore, we do not find the deviation acceptabl e. The licensee should verify that all instrumentation provided for '
the identified key variables used to ascertain RHR system performance are Category 2.
CECO Response:
The ESF status monitoring lights, component status indication, and pump motor anneters, including all associated circuits originate at the ESF electrical switchgear and are classified as safety related. These circuits are located in a mild environment and meet the recommendations of Regulatory Guide 1.97, Revision 3, for Category 2, Type D, instrumentation.
As evidenced by the "Regulatory Guide 1.97 Compliance Preliminary Report" submitted to the NRC on February 27, 1987, the refueliing water storage tank (RWST) and reactor vessel water level instruments are qualified to Category 2, Considering Type D, and Category 1, Type B, reconmendations, respectfully.
the previously identified instrumentation as the key variables, the recommended variable is classified as back-up instrumentation and is in compliance with the Category 3, Type D, reconmendation of Regulatory Guide 1.97, Revision 3.
L_
, l 3.3.6 RHR Heat Exchanger Outlet Temperature Regulatory Guide 1.97 recomends environmentally qualified instrumentation with a range of 40*F to 350*F for this variable. The licensee states that the sensors are not environmentally qualified and that the range of 50 F to 350*F is adequate for the intended monitoring functions.
This range deviation is less than three percent of the maximum recomended range. Considering instrument accuracy, and overall range, we consider this deviation minor and, therefore, acceptable.
Regulatory Guide 1.97 recomends environmentally qualified instrumentation for this variable. The instrumentation installed to measure the RHR heat exchanger outlet temperature is not environmentally qualified.
The licensee states that this instrumentation is back-a instrumentation for the key variables to monitor the operation of the RHR system: ESF status monitoring lights, component status indication, pump motor ameters, and the refueling water storage tank level and the reactor vessel level indications.
The licensee has not verified that these key variables are all monitored by Category 2 instrumentation. Therefore, we do not find the deviation acceptable. The licensee should verify that all instrumentation provided for the identified key variables used to ascertain RHR system perfomance are Category 2.
CECO Response:
The ESF status monitoring lights, component status indication, and pump motor ameters, including all associated circuits, originate at the ESF electrical switchgear and are classified as safety-related. These circuits are located in a mild environment and meet the recomendations of Regulatory Guide 1.97, Revision 3, for Category 2, Type D, instrumentation.
As evidenced by the "Regulatory Guide 1.97 Compliance Preliminary Report" submitted to the NRC on February 27, 1987, the refueling water storage tank (RWST) and reactor vessel water level instruments are qualified to Category 2, Type 0, and Category 1, Type B, recomendations, respectfully. Considering the previously identified instrumentation as the key variables, the recomended variable is classified as %ck-up instrumentation and is in compliance with the Category 3, Type D, recomendation of Regulatory Guide 1.97, Revision 3.
s 3.3.7 Accumulator Tank Level and Pressure Regulatory Guide 1.97 recommends environmentally qualified instrumentation for this variable with a pressure range 0 to 750 psig. The licensee has supplied instrumentation with no environmental qualification.
The provided range of the pressure instrument is 0 to 700 psig. The licensee states that the design pressure for the accumulator tank is 700 psig.
The accumulators are passive devices. Their discharge into the reactor coolant system (RCS) is actuated solely by a decrease in RCS pressure. Shoul d three of the four accumulators discharge, the injection is considered successful. We find that the range of the pressure instrumentation supplied for this variable is adequate to determine that the accumulators have discharged. Therefore, the range of the instrumentation for this variable is acceptabl e. The licensee states that no operator action is specified by the Emergency Operating Procedures based on post-accident accumulator level or pressure.
The existing non-qualified instrumentation is not acceptable. An environmentally qualified instrument is necessary to monitor the status of these tanks. The licensee should designate either level or pressure as the key variable to directly indicate accumulator discharge and provide instrumentation for that variable that meets the requirements of 10 CFR 50.49 and Regulatory Guide 1.97.
CECO Response:
Connonwealth Edison Company will perform an engineering evaluation of the feasibility and cost of upgrading the existing accumulator tank instrumentation. If upgrade modifications are required, either level or pressure will be designated the key variable and modified to meet ?.he applicable Regulatory Guide 1.97 recommendations. The schedule for implementing any required modifications will be developed pending resolution of all Regulatory Guide 1.97 issues.
i 3.3.9 Hign Pressure Safety Injection (HPSI) Flow Regulatory Guide 1.97 recommends environmentally qualified instrumentation for this variable. The instrumentation installed to measure the HPSI flow is not environmentally qualified. The licensee states that this instrumentation is back-up instrumentation for the key variables to monitor the operation of the HPSI sytem: ESF status monitoring lights, component status indication, pump motor ameters, and the refueling water storage tank level and the reactor vessel level indications.
The licensee has not verified that these key variables are all monitored by Category 2 instrumentation. Therefore, we do not find the deviation accepta bl e. The licensee should verify that all instrumentation provided for the identified key variables used to ascertain HPSI system performance are Category 2.
CECO Response:
The ESF status monitoring lights, component status indication, and pump motor ammeters, including all associated circuits originate at the ESF electrical switchgear and are classified as safety-related. These circuits are located in a mild environment and meet the recomendations of Regulatory Guide 1.97, Revision 3, for Category 2, Type D, instrumentation.
As evidenced by the "Replatory Guide 1.97 Compliance Preliminary Report" submitted to the NRC on i wr eary 27, 1967, the refueling water storage tank (RWST) and reactor vessel wuer level instruments are qualified to Category 2, Considering Type D, and Category 1, Type B, recommendations, respectfully.
the previously identified instrumentation as the key variables, the recommended variable is classified as back-up instrumentation and is in compliance with the Category 3, Type 0, recommendation of Regulatory Guide 1.97, Revi sion 3.
'o 3.3.10 Low Pressure Sa'ety Injection (LPSI) Flow Regulatory Guide 1.97 recommends environmentally qualified instrumentation for this variable. The instrumentatif on installed to measure the LPSI flow is not environmentally qualified. The licentee states that this instrumentation is back-up instrumentation for the key variables to monitor the operation of the LPSI system: ESF status monitoring lights, component status indication, pump motor ammeters, and the refueling water storage tank level and the reactor vessel level indicati'ons.
The licensee has not verified that these key variables are all monitored by Category 2 instrumentation. Therefore, we do not find the deviation acceptabl e. The licensee should verify that all instrumentation provided for the identified key variables used to ascertain LPSI system performance are Category 2.
CECO Response:
The ESF status monitoring lights, component status indication, and pump motor anneters, including all associated circuits originate at the ESF electrical switchgear and are classified as safety-related. These circuits are located in a mild environment and meet the recommendations of Regulatory Guide 1.97, Revision 3, for Category 2, Type D, instrumentation.
As evidenced by the "Regulatory Guide 1.97 Compliance Preliminary Report" submitted to the NRC on Februray 27, 1987, the refueling water storage tank (RWST) and reactor vessel water level instruments are qualified to Category 2, Considering Type D, and Category 1 Type B, recommendations, respectfully.
the previously identified instrumentation as the key variables, the recommended variable is classified as back-up instrumentation and is in compliance with the Category 3, Type D, recommendation of Regulatory Guide 1.97, Revision 3.
I
e 3.3.12 Quench Tank Temperature Regulatory Guide 1.97 recommends a range of 50*to 750*F for this vari abl e. The licensee has provided a range of 50 to 300*F. The licensee considers this range adequate since it covers the normal operating design conditions under which the tank is expected to maintain its integrity, and because the temperature indication, while not remaining on scale, will only exceed the range of approximately 6 percent.
The regulatory guide is specific that instrumentation should remain on scale or eliminate the potential of the operator misinterpreting an offscale instrument as a failed instrument. Therefore, the licensee should arrange this instrumentation to include the maximum expected saturation temperature (338'F at the relief disc set pressure of 100 psig), during any accident that is expected to lif t the pressurizer relief valves.
Ceco Response:
Commonwealth Edison Company will perfonn an engineering evaluation of the feasibility and cost of upgrading the existing quench tank instrumentation.
If modifications are required, a schedule for implementation will be developed pending resolution of all Regulatory Guide 1.97 issues.
3.3.15 Safety / Relief Valve Position or Main Steam Flow Regulatory Guide 1.97 recomends Category 2 instrumentation for this variable, that can be either safety relief valve position or main steam fl ow. The licensee has provided qualified valve position indication, but only for the power operated relief valves. The licensee also has main steam flow indication, however, it is not environmentally qualified for all analyzed events. The licensee states that steam genertor pressure is used to determine a faulted steam generator with the main steam isolation valves closed.
While excess pressure can indicate excess flow, it will not establish that a safety relief valve is open. The recomended instrumentation will establish that event. Therefore, the licensee should either add Category 2 instrumentation to monitor the safety valves or upgrade the main steam flow indication to meet the requirements of 10 CFR 50.49 and Regulatory Guida 1.97.
CECO Response:
In the event that reactor core / coolant temperature had risen to a value that caused one (1) or more steam generator safety valves to relieve, the control room operator would be, primarily, concerned with core cooling and would monitor reactor coolant average temperature and the core exit thermocouples (qualified to Category 1, Type A recomendations) to ensure that adequate core cooling was being provided. The key variable used to detennine that the steam generator safeties had lifted would be a stable or decreasing reactor coolant average temperature with a differential temperature present in each reactor coolant loop. Additional indication is available by comparing the reactor coolant average temperature during forced circulation or reactor coolant cold leg temperature during natural circulation with the steam generator saturation temperature derived from steam pressure utilizing the steam tables. The presence of a primary-to-secondary differential temperature veriff es that, at least, one (1) safety valve has lif ted on that steam generator. Additionally, main steam flow is not utilized to assess plant conditions in, either, the Critical Safety Function Status Trees or the Function Recovery Guidelines for the Heat Sink and Core Cooling safety functions.
As evidenced by the "Regulatory Guide 1.97 Compliance Preliminary Report" submitted to the NRC on February 27, 1987, the reactor coolant temperature and steam generator pressure instruments are qualified to Category 1, Type A, recomendations. Based on the previous infonnation, main steam flow instrumentation is classified as back-up instrumentation and is in compliance with the Category 3, Type D, recomendations of Regulatory Guide 1.97, Revision 3.
I a
3.3.16 Auxiliary Feedwater Flow Regulatory Guide 1.97 recommends instrumentation for this variable that meets Category 2 recommendations. Since the licensee has designated this instrumentation as Type A, this instrumentation should be Category 1. The licensee indicates that no recording or computer input is provided. The licensee states that this instrumentation is a back-up for steam generator level and is not otherwise required for any function.
We find the recording deviation unacceptable for Type A variables.
Neither NUREG-0737, nor Regulatory Guide 1.97 require this instrumentation to be Type A. The licensee has determined that this instrumentation is Type A.
Therefore, the licensee should provide Category 1 instrumentation for this va ri abl e. However, from the licensee's description, it is apparent that the licensee has classified this instrumentation as Type A as the result of a contingency action in the emergency operating procedures. If this is the The case, it is classified as Type A in error, and would be acceptable as is.
licensee should evaluate the classification of this instrumentation.
CECO Response:
Regulatory Guide 1.97 identifies Type A variables as "Those variables that provide primary information needed for the control room operating personnel to take specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis accident events." The Byron and Braidwood Station Emergency Operating Procedures do not specify any manual actions based on this variable and critical safety function assessment is provided by steam generator level indication in conjunction with AFW system component status indication. On this basis, the classification of 1 A is unwarranted for this variable.
Auxiliary feedwater flow is utilized to diagnose the proper operation of the Auxiliary Feedwater System following the initial verification of proper Regulatory Guide component actuation in response to the ESF actuation signal.
1.97, Revision 3, states that "Those variables that provide infonnation to indicate the operation of individual safety systems and other systems important to safety" shall be classified as Type D variables and recommends a Category 2 classification for key variables of this type. Auxiliary feedwater flow instrumentation should be re-classified as 20 and is, therefore, in compliance with the recommendations of the regulatory guide.
3.3.17 Condensate Storage Tank Water Level Regulatory Guide 1.97 recommends Category 1 instrumentation for the primary source of auxiliary feedwater. The licensee states that the essential service water system, not the condensate storage tank, is the Category 1 source of auxiliary feedwater. The regulatory guide recognizes this pos si bil i ty.
The licensee should provide information that shows that Category 1 instrumentation is available to show proper operation of the essential service water system. This instrumentation should show that water is available to and being used by the auxiliary feedwater system for post-accident situations.
Ceco Response:
For assessing the overall performance of the essential service water (SX) system, supply and return header temperature indication is provided in the control room and input to the plant computer. Instrumentation available to serve as a back-up for the key variable are ESF monitor lights, component status indication, pump motor ameters, and pump discharge pressure indication for monitoring the perfomance of the SX pumps. All of these instruments and associated circuitry are located in mild environments and comply with the criteria for Category 1, Type 0, variables for the temperature instrumentation and Category 2, Type D, variables for the SX pump monitoring instruments.
The key variable utilized to evaluate the overall perfomance of the Auxiliary Feedwater ( AF) system is the steam generator level indication. This instrumentation maintains a Category 1, Type A, qualification and is backed up by AF system flow instruments which permits the assessment of individual AF system train perfomance. The AF system flow instrumentation was originally classified as Category 1, Type A, and is now considered to be Category 2, Type D, instrumentation. In addition to these variables, low suction pressure alarms, with analog input to the plant computer, are associated with each of the AF pumps to monitor the adequacy of the water supply. This instrumentation is located in a mild environment and complies with the criteria delineated by Regulatory Guide 1.97, Revision 3, for Category 2, Type D, variables.
3.3.18 Containment Atmosphere Temperature Regulatory Guide 1.97 recomends environmentally qualified instrumentation for this variable. The instrumentation installed is not environmentally qualified. The licensee has acknowledged this, and states that containment pressure initiates all ESF containment cooling functions.
The containment atmosphere temperature directly indicates the accomplishment of a safety function (containment cooling) and is, therefore, a key variable. Thus, Category 2 requirements should be met by the licensee for this variable. The licensee should provide ir,strumentation for this varialbe that is environmentally qualified to the requirements of 10 CFR 50.49 and Regulatory Guide 1.97.
CECO Response:
At the Byron and Braidwood Stations, the assessment of the Containment Critical Safety Function, per procedure, is based entirely on containment pressure. As evidenced by the Containment CSF Status Tree and Function Recovery Guideline Index (attached) for these stations, containment atmosphere temperature is not required or utilzed to diagnose the effectiveness Additionally,ofthere the containment cooling systems during accident conditions.
are no operator contingency actions based on containment temperature and the emergency procedures and function recovery guidelines do not The identify use this of variable as an alternate for assessing containment conditions.
containment temperature in lieu of pressure would, therefore, require the recomendation of the TSC and the approval of the shif t supervisor and this variable would have to be converted to pressure, utilizingForthethese steam tables, reasons, to be compatible with the function recovery guidelines.
containment pressure is considered to be the key variable for assessing, both, containment conditions and the effectiveness of the containment cooling systems. Containment atmosphere temperature is considered to be a back-up variable and is in compliance with the Category 3, Type D, recomendations of Regulatory Guide 1.97, Revision 3.
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. ' PRO'CEDURE 4 f(UMBER TITLE NUMBER CRITICAL SAFETY FUNCTION STATUS TREES I BST-1 Su bc ri ticali ty (S) A BST-2 Core Cooling (C) B BST-3 Heat Sink (H)
BST-4 Integri ty (P) 0 BST-5 Containment (Z) E BST-6 Inventory (!) F FUNCTION REC 0VERY GUIDELINES SUBCRITICALITY !!
BFR-S.1 Response to Nuclear Power Generation A BFR-S.2 Response to Loss of Core Shutdown B CORE C0OLING III BFR-C.1 Response to Inadequate Core Cooling A BFR-C.2 Response to Degraded Core Cooling B BFR-C.3 Response to Potential loss of Core Cooling C BFR-C.4 Response to Saturated Core Cooling 0 Conditions HEAT SINK IV BFR-H.1 Response to Loss of Secondary Heat Sink A BFR-H.2 Response to Steam Generator Over Pressure B BFR-H.3 Response to Steam Generator High Level C BFR-H.4 Response to Steam Generator Low Level 0 BFR-H.5 Response to Loss of S/G PORVs E or Steam Dumps
-V ,
INTEGRITY BFR-P.1 Response to Iminent Themal Shock Condition A BFR-P.2 Response to Anticipated Thermal Shock B Condition VI CONTAINMENT A
BFR-2.1 Response to High Contairrtent Pressure BTR-Z.2 Response to High Containrent Sumo level B BFR-Z.3 Response to High Containrent Radiation Level C INVENT TY J CFR-[.1 Restonse to Pressuri:er Flooding A E ?-:.? Pes:or:se to Low System Inventory B
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- re:: mse to 'cic in reacter '!essel C
3.3.19 Containment Sump Water Temperature Regulatory Guide 1.97 recomends Category 2 instrumentation with a range The licensee has not provided of 50 to 250'F to monitor this variable.
instrumentation f or this variable, stating that the containment sump water temperature is not required to be monitored to detect RHR This wasorpreviously containment spray accepted pump cas!tation when in the recirculation mode.
by the NRC.
This is insufficient justificatiion for this exception. The licensee should provide the recomended instrumentation for the functions outlined in Regulatory Guide 1.97 or identify other instruments that provide the same infomation (such as the residual heat removal heat exchanger inlet temperature) and that satisfies the regulatory guide.
CECO Response:
To detect or anticipate RHR and/or Containment Spray Pump Cavitation, utilizing this variable requires calculating the available Net Positive Suction Head (NPSH) for these components considering the effects of the existing containment pressure and component flow rate in addition to the containment sump water temperature.
During an accident, it i; highly unlikely that the control room operator or the shif t technical advisor will be in a position to perfom this calculation Aormorethatdirect the control room conditions will be conductive to it's means of detecting pump cavitation is provided by compl eti on.
the pump motor ameters which are qualified to Category 2, Ty.ne D, l
requirements considering the mild environment titat these instruments reside in. Additionally, these instruments conscitute oro (1) of the keyFor variables this utilized for assessing tne performance of these safety systems.
reason, containment sump water temperature is considered to be an unnecessary variable for diagnosing the performance of safety system compor,ents during the recirculation mode of operation. This deviation was orf qir ~,1y accepted by the NRC on this basis.