Similar Documents at Byron |
---|
Category:CORRESPONDENCE-LETTERS
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) ML20217M2871999-10-21021 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 ML20217H4661999-10-18018 October 1999 Forwards Changes to EPIPs IAW 10CFR50.54(q) & 10CFR50,App E, Section V.Details of Changes Encl 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 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 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 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 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 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 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 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 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 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.. 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 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 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 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) 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 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) 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 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 ML20196K0161999-06-30030 June 1999 Discusses 990622 Meeting at Byron Nuclear Power Station in Byron,Il.Purpose of Visit Was to Meet with PRA Staff to Discuss Ceco Initiatives in Risk Area & to Establish Dialog Between SRAs & PRA Staff 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 ML20212H8241999-06-24024 June 1999 Informs That Effective 990531 NRC Project Mgt Responsibility for Byron & Braidwood Stations Was Transferred to Gf Dick 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 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 ML20207G0601999-06-0707 June 1999 Provides Updated Info Re Number of Failures Associated with Initial Operator License Exam Administered from 980914-0918. NRC Will Review Progress Wrt Corrective Actions During Future Insps ML20207G0421999-06-0404 June 1999 Forwards Insp Repts 50-454/99-04 & 50-455/99-04 on 990330-0510.Violations Identified & Being Treated as non-cited Violations ML20195D6351999-06-0404 June 1999 Notifies NRC of Actions That Has Been Taken in Accordance with 10CFR26, Fitness for Duty Programs ML20207E5451999-05-28028 May 1999 Forwards Insp Repts 50-454/99-07 & 50-455/99-07 on 990517-20.No Violations Noted.Fire Protection Program Was Effective 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 ML20207B6361999-05-25025 May 1999 Forwards SE Accepting Revised SG Tube Rupture (SGTR) Analysis for Bryon & Braidwood Stations.Revised Analysis Was Submitted to Support SG Replacement at Unit 1 of Each Station 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 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 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 ML20206U3471999-05-20020 May 1999 Forwards Insp Rept 50-454/99-05 on 990401-22.No Violations Noted.Insp Reviewed Activities Associated with ISI Efforts Including Selective Exam of SG Maint & Exam Records, Calculations,Observation of Exam Performance & Interviews ML20207A2151999-05-19019 May 1999 Forwards Insp Repts 50-454/99-06 & 50-455/99-06 on 990419-23.No Violations Noted.Insp Consisted of Review of Liquid & Gaseous Effluent Program,Radiological Environmental Monitoring Program,Auditing Program & Outage Activities 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) 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 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) 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 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.. 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 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 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 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 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 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 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 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 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 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 ML20206C7901999-04-23023 April 1999 Provides Suppl Info Re Use of W Dynamic Rod Worth Measurement Technique,As Requested During 990413 Telcon.Rev Bars in right-hand Margin Identify Changes from Info Submitted by ML20206E7521999-04-22022 April 1999 Submits Rept on Number of Tubes Plugged or Repaired During Inservice Insp Activities Conducted at Plant During Cycle 9 Refueling Outage,Per TS 5.6.9 ML20206A7431999-04-22022 April 1999 Forwards Comments Generated Based on Review of NRC Ltr Re Preliminary Accident Sequence Precursor Analysis for Byron Station,Unit 1 ML20206B3941999-04-21021 April 1999 Forwards Annual & 30-Day Rept of ECCS Evaluation Model Changes & Errors, for Byron & Braidwood Stations.Updated Info Re PCT for Limiting Small Break & Large Break LOCA Analysis Evaluations & Detailed Description of Errors ML20206B2471999-04-20020 April 1999 Informs That SE Kuczynski Has Been Transferred to Position No Longer Requiring SRO License.Cancel License SOP-31030-1, Effective 990412 ML20205S9621999-04-20020 April 1999 Responds to 981203 RAI Telcon Re SG Tube Rupture Analysis for Byron Station,Unit 2 & Braidwood Station,Unit 2.Addl Info & Subsequent Resolution of Issues Discussed During 990211 Telcon Are Documented in Encl ML20206A8141999-04-20020 April 1999 Advises NRC of Review of Cycle 10 Reload Under Provisions of 10CFR50.59 & to Transmit COLR for Upcoming Cycle ML20205T3901999-04-13013 April 1999 Forwards Byron Station 1998 Occupational Radiation Exposure Rept, Which Is Tabulation of Station,Utility & Other Personnel Receiving Annual Deep Dose Equivalent of Less than 100 Mrem ML20196K6661999-03-31031 March 1999 Forwards Byron Nuclear Power Station 10CFR50.59 Summary Rept, Consisting of Descriptions & SE Summaries of Changes, Tests & Experiments.Rept Includes Changes Made to Features Fire Protection Program,Not Previously Presented to NRC ML20205K5841999-03-31031 March 1999 Submits Rept on Status of Decommissioning Funding for Reactors Owned by Comm Ed.Attachment 1 Contains Amount of Decommissioning Funds Estimated to Be Required Pursuant to 10CFR50.75(b) & (C) ML20205B4241999-03-23023 March 1999 Provides Results of drive-in Drill Conducted on 990208,as Well as Augmentation Phone Drills Conducted Since 981015,as Committed to in Util ML20207K0351999-03-0404 March 1999 Forwards Util Which Transmitted Corrected Pages to SG Replacement Outage Startup Rept.Subject Ltr Was Inadvertently Not Sent to NRC Dcd,As Required by 10CFR50.4 ML20205C6861999-03-0404 March 1999 Provides Notification That Byron Station Implemented ITS on 990205 & Braidwood Station Implemented ITS on 990219 ML20207D6831999-03-0101 March 1999 Forwards fitness-for-duty Program Performance Data for Each Comed Nuclear Power Station & Corporate Support Employees for Six Month Period Ending 981231,per 10CFR26.71(d) ML20207D4301999-02-26026 February 1999 Informs NRC That Supplemental Info for Byron & Braidwood Stations Will Be Delayed.All Mod Work Described in Ltr Is on Schedule,Per GL 96-06 ML20207B8971999-02-25025 February 1999 Expresses Concern That Low Staffing Levels & Excessive Staff Overtime May Present Serious Safety Hazard at Some Commercial Nuclear Plants in Us ML20203C7001999-02-0202 February 1999 Informs That Mhb Technical Associates No Longer Wishes to Receive Us Region III Docket Info Re Comed Nuclear Facilities.Please Remove Following Listing from Service List ML20202F5911999-01-29029 January 1999 Forwards Byron Unit 1 Cycle 9 COLR in ITS Format & W(Z) Function & Byron Unit 2 Cycle 8 COLR in ITS Format & W(Z) Function. New COLR Format Has Addl Info Requirements ML20199E1611999-01-15015 January 1999 Forwards Response to 980902 RAI Re GL 97-01, Degradation of Crdm/Cedm Nozzle & Other Vessel Closure Head Penetrations. CE Endorses Industry Response to RAI as Submitted by NEI ML20199B7511999-01-0808 January 1999 Forwards Proprietary Versions of Epips,Including Rev 52 to Bzp 600-A1 & Rev 48 to Bzp 600-A4 & non-proprietary Version of Rev 52 to Bzp 600-A1 & Index.Proprietary Info Withheld 1999-09-30
[Table view] |
Text
. -
- Commonwealth Edison Comtuny Dyron Gencrating Station 6150 North German Church Road ny ron. n. nioitm9 6 Tcl His-U M 6 i t December 15,1995 LTR: BYRON 95-0399 FILE: 1.10.0101 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission l Washington, D.C. 20555 Attn: Document Control Desk
Subject:
Byron Station Unit i Cycle 7 Mid-Cycle Inspection Outage Steam Generator Interim Plugging Criteria Summary Report Docket No. 50454
Reference:
- 1) September 2,1995, Letter from D. Saccomando to U.S. NRC, regarding Commonwealth Edison Steam Generator Initial Inspection Plan Used to Verify Load Path Necessary to Support 3.0 Volt Interim Plugging Criteria.
- 2) September 20,1995, Letter from H. Pontious to U.S. NRC, regarding the September 13,1995 Teleconference between Comed l and NRC Concerning the increase in Interim Plugging Criteria.
- 3) October 7,1995, Letter from D. Saccomando to U.S. NRC, regarding the October 3 and 4,1995 Teleconference between ,
Comed and NRC Concerning the Increase in Interim Plugging Criteria Byron Station Technical Specification 4.4.5.5 d specifies the reporting requirements for the implementation of the voltage-based repair criteria (IPC) for steam generator tube support plate intersections. This specification requires that the Staffis to be notified prior to returning the steam generator to service should any of the following conditions arise:
9512200006 951215 PDR ADOCK 05000454 y Q PDR (pA95byritrs\950399. doc \l21595) 3 A Unicom Company
Document Control Desk Page 2
- 1. If the estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of cycle) voltage distribution exceeds the leak limit (determined by from the licensing basis dose calculation for the postulated main steam line break) for the next operating cycle.
- 2. If circumferential crack-like indications are detected at the tube support plate intersections.
- 3. Ifindications are identified that extend beyond the confines of the tube support plate.
- 4. Ifindications are identified at the tube support plate elevations that are attributable to primary water stress corrosion cracking.
- 5. If the calculated conditional burst probability based on the projected end of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1x10E-2.
- 6. Ifindications detrimental to the integrity of the load path necessary to support the 3.0 volt IPC are found during a steam generator internals inspection.
It should be noted that Byron Station did not encounter any of the above conditions for Staff notification during the B1P02 inspection. This summary of the IPC inspection and implementation is being submitted for information.
In addition, References 2 and 3 require a summary of the steam generator internal inspection be submitted to the Staff prior to entering Mode 4. Attachments 1 and 2 contain the results of the Bymn Unit 1 Cycle 7 mid-cycle outage (BlP02) steam generator IPC implementation and eddy current inspection to satisfy these reporting requirements. Mode 4 entry is currently scheduled for December 22,1995.
The complete report of the IPC implementation and steam generator intemal inspection will be submitted to the Staff within 90 days of plant startup, consistent with NRC Generic Letter 95-05 and References 2 and 3.
(p:\95byritrs\950399. doc \l21595)
Document Control Desk Page 3 Please direct any questions regarding this submittal to Mr. Jay Smith, Byron Site Engineering, at (815) 234-5441, extension 2604.
Sincerely,
. /
K.L. Koto Station Manager Byron Nuclear Power Station ec: H. Peterson, Senior Resident Inspector - Byron G. Dick, Byron Project Manager - NRR H.' Miller, Regional Administrator - RIII Office ofNuclear Safety -IDNS 1
I (p:\95 byritts\950399. doc \ l 21595)
ATTACilMENT 1 l BYRON UNIT 1 CYCLE 7 MID-CYCLE OUTAGE INTERIM PLUGGING CRITERIA
SUMMARY
4 Introduction As a re:dt ofimplementing a 1.0 volt Interim Plugging Criteria (IPC) for axial oriented stress corrosion cracking at tube support plate intersections during the previous refuel outage, Byron Unit I committed to perform a mid-cycle steam generator inspection outage (BlP02) in the Fall of 1995. During this outage, a complete eddy current inspection was to be performed in accordance with the Byron Technical Specifications and the requirements for IPC as specified in Generic Letter 95-05, " Voltage-based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking" Prior to BlP02, a Technical Specification Amendment Request was submitted to increase the hot-leg IPC voltage repair limit from 1.0 volt to 3.0 volts applicable to BlP02 and through Cycle 8. The basis for this increase was a reliance on the presence of the tube support plates to eliminate the probability of tube burst and the limited displacement of the plates during transient conditions. The structural integrity of the steam generator (SG) internal components relied upon to support the basis for 3.0 volt IPC were also required to be verified through visual inspection and eddy current inspection techniques during B1P02. Additional steam generator inspections were required during BlP02 in response to Generic Letter (GL) 95-03, "Circumferential Cracking of Steam Generator Tubes".
This report provides a brief summary of the IPC eddy current inspections, IPC leakage and tube burst probability assessments, and the results of the SG structuralintegrity inspection. Attachment 2 provides a detailed report of the IPC inspections and the MSLB leakage and tube burst assessments. I i
l Summarv of the SG Eddy Current Inspection Scope 1 During BlP02, the following SG eddy current examinations were performed per Technical l Specification 4.4.5, Generic Letter 95-05, and Generic Letter 95-03:
- 1. 100% full length bobbin coil inspection of the inservice tubes in all 4 SGs.
- 2. 100% Plus-Point and Rotating Pancake Coil (RPC) of hot-leg top of tubesheet roll transition regions in all 4 SGs (GL 95-03).
- 3. 20% RPC of Row I and 2 U-Bends in all 4 SGs (GL 95-03).
- 4. 20% RPC of Pre-heater Bafile Plate Expansion Areas in 1 SG (GL 95-03).
- 5. RPC of support plate intersections to support IPC, which includes:
- a. All indications that exceed the upper voltage repair limit.
- b. 20% ofintersections that are between 1.0 - 3.0 volts.
11 l
i
- c. All dented TSP intersections that exceed 5.0 volts.
- d. 20% of dented TSP intersections that are between 2.5 and 5,0 Volts.
- e. Allintersections that contain large mixed residuals.
- f. Allindications that are susceptible to deformation during a i LOCA+SSE event (i.e., wedge locations).
- g. All intersections that contain interfering copper signals.
All inspections were performed in accordance with the Byron /Braidwood Eddy Current Inspection Guidelines. The bobbin coilinspections utilized IPC transfer standards that are 1 calibrated to a laboratory standard used in the development ofIPC. Probe wear was
! monitored and controlled to 15% through the use of probe wear standards. A 0.610 inch l diameter probe was used on all tubes where IPC was implemented. Data analysts were trained and tested on the site specific guidelines and IPC implementation.
The results from the inspections required by NRC GL 95-03 will be submitted under a separate report in accordance with Technical Specification 4.4.5.5.b.
l
~
Summary of SG Eddy Current IPC Inspection Results A total of 5005 indications were identified by the bobbin coil inspection that were indicative of outer diameter stress corrosion cracking at the tube support plate intersections. One hot-leg indication exceeded the 3.0 volt IPC voltage repair limit and the voltage of this indication was 3,17 volts. Of the 5005 indications,8 were located at cold-leg intersections and ranged from 0.15 volts to 1.05 volts. All cold-leg indications were RPC inspected and none were confirmed as flaws. A total of 774 support plate
- indications were identified that were between 1.0 volt and 3.0 volts. The remaining 4230 support plate indications were 1,0 volt and lower.
]
~
Rotating pancake coil inspections were performed on 219 indications, which included all indications greater than 3.0 volts,20% of the indications between 1.0 volt and 3.0 volts 4 and a sample ofindications 1.0 volt and less. The number of these indications that were confirmed as flaws by RPC was 161. Rotating pancake coilinspections were also d
performed on 89 dented intersections that were greater than 5.0 volts and 25 dented intersections between 2.5 and 5.0 volts. No flaw indications were found. Eight (8)large mixed residual signals that could mask a 1.0 volt indication were RPC inspected and no j flaws were confirmed No circumferential indications or PWSCC was found at the tube I support plate intersections. All ODSCC indications that were RPC inspected were confined within the thickness of the tube support plate. The RPC inspection scope is summarized in Table 1.
l 1
) 12 l
j, -
1 l Cycle Length and Voltage Growth
' The operating period from the previous cycle 6 refuel outage to the start of the current cycle 7 mid cycle inspection outage is designated as Cycle 7A. Cycle 7A was 0.869 i effective full power years (EFPY). The remaining operating period of cycle 7 is designated as Cycle 7B and is estimated to be 0.318 EFPY when the end-of cycle 7 refuel i outage is scheduled to begin. This information is presented in Table 2.
1 The average growth rate of the tube support plate indications during Cycle 7A was 0.235
, volts per EFPY The largest growth rate for a single indication was 2.5 volts per EFPY.
d
- MSLB Leakage and Burst Probability Assessment i
Postulated MSLB primary to secondary leakage and conditional probability of tube burst were evaluated for the as-found actual measured voltage distribution of the tube support plate indications. Since the bases for 3.0 volt IPC takes credit for limited tube support plate (TSP) displacement (locked assumption) such that tube burst is eliminated, the
- contribution to the total tube burst probability of the hot-leg indications is negligible.
4 Also, the MSLB leakage contains the effects ofindications restricted from burst (IRB) i due to the constraining effects of the TSP for the locked assumption. However, for comparison, MSLB leakage and tube burst probability assessments were performed for
. both the free span assumption (TSPs are allowed to displace) and for the locked
, assumption.
i
- For the locked assumption in the worst case SG, the maximum MSLB leakage for the
! actual measured voltage distribution was 0.075 gpm. This is well below the site specific
- MSLB leakage limit of 36.5 gpm for a primary coolant dose equivalent Iodine-131 limit of 03 i micro-curies per gram. This MSLB leak rate is the same for the freespan assumption.
l The conditional probability of burst for the locked assumption in each SG for the actual ;
measured voltage distribution was less than 4x10E-6. This is also well below the l
! Technical Specification limit of lx10E-2. For the freespan assumption, the worst case l conditional probability of burst was calculated to be 1.30x10E-3, which is also below the Technical Specification limit. I The projected end-of-cycle 7B MSLB leakage and conditional probability of burst assessments will be submitted in a final report to be submitted within 90 days of start-up.
i 1
5
- l-3
,.-_.-.._-_.l
y- _.;_ _ _ . . _ _ _ _ __ _ _ _ _ _ _ _ . _ . _ _ _ _ _ _ .
e l
Tube Support Plate Integrity Verifications l The structural integrity of steam generator internal components that are important to the bases of 3.0 volt IPC were inspected during BIP02 in accordance with the "SG Structural i Integrity Plan in Support of Braidwood-l and Byron-13.0 volt IPC"(Inspection Plan). ,
This Inspection Plan involved the performance of visualinspections and enhanced eddy l current examinations of key components.
A visual inspection was performed in the ID SG on seven (7) stayrod nuts located at the top support plate. All areas inspected were cleaned with a high pressure water lance prior to inspection. Proper lighting and resolution was verified to meet ASME VT-1 )
requirements. Each stayrod nut was verified to be in contact with the support plate and I properly welded to the support plate at two locations. Each stayrod nut also was verified to contain a weld that joined the nut to the stayrod. Degradation of the welds was not found in any location. Confirmation of proper nut to plate contact and integrity of the welds ensures that the stayrods serve their intended function as assumed by the bases for 1 the 3.0 volt IPC.
Enhanced eddy current examinations were performed in the areas of the three anti rotation I devices in each SG using the EPRI developed technique. The focus of this inspection was to verify the integrity of the tube support plate. The enhanced technique involved acquiring data with a bobbin coil probe at a reduced pull speed of 12 inches per second or less. Anomalies found were compared to defect signals from laboratory support plates fabricated and tested by EPRI. Filly (50) intersections were inspected at each anti-rotational device. Due to SG symmetry,75 tubes were inspected to encompass the 50 intersections per anti rotation device. Data was collected for the entire tube and each support plate was evaluated. No anomalies indicative of degradation were detected in any SG.
The presence of each tube support plate was verified for all SG tubes. This was performed as part of the normal eddy current analysis of each tube.
l-4
Table 1 RPC Inspection Results
- Category-' ~'
pNo.1 3No:
' Inspected-
'Confirnied'
> 3.0 volts 1 1 1.0 - 3.0 volts 155 137
<= 1.0 volt 63 23 Dents > 5.0 volts 89 0 Dents 2.5-5.0 volts 25 0 Large Mixed 8 0 Residuals Table 2 Cycle Length L BOC (EFPY) 1EOC (EFPI); Cycle Lengtli?
Cycle 7A 6.949 7.818 0.869 Cycle 78 7.818 8.136 (est.) 0.318(est.)
15
1 ,
ATTACIIMENT 2 WESTINGIIOUSE REPORT NSD-SGD-1187 Byron Unit 1 Interim Plugging Criteria Return to Power Report December 1995
-. . - - ._. -