Similar Documents at Byron |
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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
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Text
'
. N C mmonw:alth Edison
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c' . 1400 Opus Place Downers Grove, lihnois 60515 October 4,1994 Mr. William Russell, Director OfIice of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington D.C. 20555 Attention: Document Control Desk
Subject:
GL 92-01, " Reactor Vessel Structural Integrity" Data Update Byron Station Units 1 and 2 NPF-37/66; NRC Docket Nos. 50-454/455
References:
- 1. J. Bauer letter to W. Russell transmitting Commonwealth Edison Company response to NRC inquiry to GL 92-01 dated July 22,1994
- 2. R. Assa letter to D. Farrar dated June 24,1994, requesting verification of data pertaining to GL 92-01,
" Reactor Vessel Structural Integrity"
Dear Mr. Russell:
By letters dated July 2,1992, and November 19,1993, Commonwealth Edison Company (Comed) provided a response to GL 92-01, Revision 1. In the above reference letter 2, the Staff requested the Comed verify the previously supplied information by July 24,1994, as this data will be entered into a Reactor Vessel Integrity Database.
In reference letter 1, Comed requests that the Byron data verification / update be delayed until September,1994 due to completion of Capsule W results. The attachment includes Byron's response to the reference 2 correspondence. In the summary tables, the changes from the data transmitted in reference 2 are clearly indicated.
Please address any comments or questions regarding this matter to this office.
Sincerely, hDenise n M./ff b Shecomando Nuclear Licensing Administrator ec: G. Dick, Byron Project Manager, NRR II. Peterson, Senior Resident Inspector - Byron B. Clayton, Branch Chief- Regina III Office of Nuclear Facility Safety - IDNS m:nla\ byron \g19201\1 j
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m s 9410110?s2 941004 PDR h, 'I ADOCK 05000454 1-F' PDR l
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Attachment Byron Units 1 and 2 !
Response to NRC Generic Letter 92-01, Revision 1 e
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Introduction This report provides a response to the Generic Letter 92-01, Revision 1, closure letter recently issued by the NRC for Commonwealth Edison Company's Byron Units 1 and 2. The fbilowing is the full Data Summary Tables for Pressurized Thermal Shock and Upper-Shelf Energy. Those values that are unchanged are shown in the shaded boxes. Revised values are indicated in the unshaded boxes. >
i 1
1 1
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Table 2-1. Braidwood Unit 1 -- Data Surary for Pressurized Thermal Shock Calculation ,
IS Neutron Method of Method of Beltline Fluence at IR'Im Determin. Chemistry Determin.
Heat No. 32 EFPY F IRTm Factor CF %Cu %Ni Material 1 Lower SP-7016 6 . 8 2 E +18 +10
- Nozzle Specific Table 2 Belt Forging Upper 49C344-1-1/ 3 . 0 3E +19 -30* Plant 31 RG1.99 0.05* 0.73"'I Shell 49D383-1-1 Specific Table 2 Forging
-20* , Plant 20 RG1.99 0.03* 0,73
Shell 49C813-1-1 Specific Table 2 Forging WF-645 H4498 6 . 8 2 E+18 -30* Plant 41 RG1.99 0.03* 0.50
- Upper Specific Table 1 Cire. Weld WF-562 442G11 3 . 0 3E + 19 - +40* Plant 41 RG1.99 0.03* 0.65
- Middle Specific Table 1 Cire. Weld WF-653 31401 < 1. 0 0E + 17 '** -40* Plant 150.8 RG1.99 0.19* 0.56 "
Lower Specific Table 1 cire. Weld NOTES:
- a. Fluence data are from WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Surveillance Program," August 1990.
- b. Chemical compositions and initial RT m data for all materials are from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 3 V-- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _
Table 2-2. Braidwood Unit 1 -- Data Sumr.ary for Uroer-Shelf Enercy Calculation 1/4T Method of 1/4T USE Neutron Determin.
Beltline Material at 32 Flttence at UniirOd. Unirrad.
Material Heat No. Type EFPY 32 EFPY USE USE Lower 5P-7016 A 508-2 13 7 4 . 0 9E +18 'd' 16 2 Direct Nozzle Belt Forging ,
Upper 49C344-1-1/ A 508-3 9 3 '*' 1.66E+19'd' 118"' Direct Shell 49D383-1-1 Forging Lower 49D867-1-1/ A 509-3 10 7 1.66E+19 488 13 6 Direct-Shell 49C813-1-1 Forging WF-645 M4498 Linde 80, 7 5 4 . 0 9 E + 18 '*' 87(*1 Direct Upper SAW Circ. Weld WF-562 442011 Linda 80, 5 5 'b' 1.66E+19 i d' 7 0 Direct Middle SAW Cire. Weld WF-653 31401 Linde 80, ---' cl.00E+17 88) 7 9
5 Direct Lower SAW Circ. Weld Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 4
NOTES FOR TABLE 2-2: ,
- a. EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.1% copper for base metal.
- b. EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.05% copper for welds. l
- c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revisian 2.
- d. Fluence data are from WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Surveillance Program," August 1990.
- e. UUSE data are from the November 19,1993 letter from T. W. Simpkin to T. E. Murley, Braidwood Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
- f. UUSE data for forging 49C344-1-1/49D383-1-1 is from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
Date: 7/1/94 77-1234176-00 Prepared By:. M. J. DeVan Date: 7/1/94 Page S Reviewed By: L. B. Gross
Table 2-3. Braidwood Unit 2 -- Data Summarv for Pressurized Themal Shock Calculation IS Neutron Method of Method of Beltline Fluence at IRTm Determin. Chemistry Determin.
F Factor CF %Cu %Ni Material Heat No. 32 EFPY IRTc Lower SP-7056 6 . 82E+18 +30*' Plant 26 RG1.99 0.04*' O.90*'
Nozzle Specific Table 2 Belt Forging Upper 49D963-1-1/ 3 . 0 3E + 19 -30*' Plant 20 RG1.99 0.03*' O.71*'
Shell 49C904-1-1 Specific Table 2 Forging Lower 50D102-1-1/ 3 . 0 3 E+ 19 -30*' Plant 37 RG1.99 0.06*' O.75*'
Shell 50C97-1-1 Specific Table 2 Forging WF-645 H4498 6 . 82 E + 18 '*i -30*' Plant 41 RG1.99 0.03*' O.50*'
Upper Specific Table 1 Cire. Weld WF-562 442011 3 . 0 3E+19 +40*' Plant- 41 RG1.99 0.03*' O.65*'
Middle Specific Table 1 Cire. Weld 1084-18 < 1. 0 0 E+ 17 (*' -16*' Plant 54 RG1.99 0 . 0 4 *' O.60*'
hF-696 Lower Specific Table 1 Cire. Weld NOTES:
- a. Fluence data are from WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.
- b. Chemical compositions and initial RTm data for all materials are from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
M. J. DeVan Date: 7/1/94 77-1234176-00 Prepared By: Page 6 Reviewed By: L. B. Gross Date: 7/1/94 l
Table 2-4, Braidwood Unit 2 -- Data Sumarv for Upper-Shelf Enerov Calculation .
1/4T Method of
- 1/4T USE Neutron Determin.
Beltline Material at 32 Fluence at Unirrad. Unirrad.
Material Heat No. Type EFPY 32 EFPY USE USE Lower 5P-7056 A 508-2 10 9 '*) 4.09E+18* 128* Direct Nozzle Belt Forging Upper 49D963-1-1/ A 508-3 9 4 1.66E+19'd' 119'" Direct shell 49C904-1-1 j Forging Lower 50D102-1-1/ A 508-3 118 1.66E+19'*) 15 0 Direct Shell 50C97-1-1 Forging WF-645 H4498 Linde 80, 75
- Direct Upper SAW Cire. Weld WF-562 442011 Linde 80, 55
- Direct-Middle SAW Circ. Weld WF-696 1084-18 Linde 80, --- * <1.00E+17* 78
- Direct Lower SAW-Circ. Weld l
l Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 7
NOTES FOR TABLE 2-4:
- a. EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.1% copper for base metal, ,
- b. EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.05% copper for welds.
- c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revision 2.
- d. Fluence data are from WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.
- e. UUSE data are from the November 19,1993 letter from T. W. Simpkin to T. E. Murley, Braidwood Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
- f. UUSE data for forging 49D963-1-1/49C904-1-1 is from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,
Subject:
Braidwood Station, Units 1 and 2.
l l
l 7/1/94 77-1234176-00
' Prepared By: M. J. DeVan Date:
Date: 7/1/94 Page 8 f Reviewed By: L. B. Gross l
l
s Table 2-5. Byron Unit 1 -- Data Summary for Pressurized Thermal Shock Calculation .
l l
IS Neutrc n Method of Method of j
Beltline Fluence at IRTm Determin. Chemistry Determin.
F IRTm Factor CF %Cu %Ni I Material Heat No. 32 EFPY l
Lower. 123J218 4 . 8 6 E+18 +30* Plant 31 RG1.99 0.05
- 0.72'*3 Specific Table 2 Nozzle Belt Forging Upper 5P-5933 2.159E+19 +40* Plant 19.14 Calculated 0.05
Shell Specific Forging Lcwer SP-5951 2.159E+19 +10* Plant 26 RG1.99 0.04* 0.64
- Shell Specific Table 2 Forging 4 . 8 6 E +18 0* Generic 41 RG1.99 0.03* 0 . 6 3 'b' WF-501 442011 Table 1 Upper Circ. Weld 442002 2 .15 9 E+19 -30* Plant 22.33 Calculated 0.03
- WF-336 Middle Specific Cire. Weld
< 1. 0 0 E+ 17 +10" Plant 164.65 RG1.99 0.23* 0.57 "
WF-472 31401 Table 1 Lower Specific Cire. Weld NOTES:
- a. Fluence data are from WCAP-13880, " Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Surveillance Program," January 1994.
Chemical compositions and initial RTmy data for all materials are from the July 2, 1992 letter from b.
M. A. Jackson to T. E. Murley,
Subject:
Byron S'.ation, Units 1 and 2.
f.
7/1/94 77-1234176-00 Prepared By: M. J. DeVan Date:
Date: 7/1/94 Page 9 Reviewed By: L. B. Gross
Table 2-6. Byron Unit 1 -- Data Summary for Urner-Shelf Enerav Calculation .
1/4T Method of 1/4T USE Neutron Determin.
Beltline Material at 32 Fluence at Unirrad. Unirrad.
Material Heat No. Type EFPY 32 E7PY USE USE Lower 123J218 A 508-2 119 '*' 2.92E+18" 13 8 Direct Nozzle Belt Forging Upper SP-5933 A 508-2 111 1.179E+19'd' 138" Direct Shell Forging Lower SP-5951 A 508-2 12 0 1.179E+19' 15 0 Direct Shell I Forging WF-501 442011 Linde-80, 63
- 2 . 9 2 E+ 18 '8' 73'S' Direct Upper SAW Circ. Weld WF-336 442002 Linde 80, 60
- 1.17 9 E + 19 '*' 74'S' Direct Middle SAW Circ. Weld WF-472 314Ol' Linde 80, -- cl.00E+17'd' 72M' . Direct Lower SAW Circ. Weld Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 10
. w 5
. i NOTES FOR TABLE 2-6:
- a. EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.1% copper for base metal. ,
f b. EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.05% copper for welds.
,' i
! c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, l I Revision 2.
i
- d. Fluence data are from WCAP-13380, " Analysis of Capsule X from the Commonwealth Edison Company l
]
Byron Unit 1 Reactor Vessel Surveillance Program," January 1994.
i
- e. UUSE data for forgings 123J218 and SP-5951 are from WCAP-11651, " Analysis of Capsule U from the l
l Commonwealth Edison Co. Byron Unit 1 Reactor Vessel Radiation Surveillance Program," November-1987.
i
- f. UTTSE data for forging 5P-5933 is from the July 2, 1992. letter from M. A. Jackson to T. E.
2 Murley,
Subject:
Byron Station, Units 1 and 2.
- g. UUSE data for the welds are from the November 19,199s letter from T. W. Simpkin to T. E. Murley, Byron Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
i i
t M. J. DeVan Date: 7/1/94 77-1234176-00 Prepared By: Page 11 .
Reviewed By: L. B. Gross Date: 7/1/94' 1
4 ".
~
Table 2-7. Byron Unit 2 -- Data Summarv for Pressurized Thermal Shock Calculation ,
a IS Neutron Method of Method of Fluence at IRTm Determin. Checistry Determin.
Beltline Factcr CF tCu tNi Material Heat No. 32 EFPY F IRTm Plant 31 RG1.99 0 . 0 5 - 0.74'*8 Lower 4P-6107 6 . 8 2 E+18 + 10
Specific Table 2 Nozzle Belt Forging
- 2 0 Plant 20 RG1.99. 0.01'*8 0 . 7 0
Upper 49D329-1-1/_ 2,192 E+19 '*' Specific Table 2 Shell 49C297-1-1 Forging 31 RGl'.99 ' O . 0 5 - 0.73 *) 8 Lower 49D330-1-1/ 2.192E+19 - 2 0 - Plant .
Table 2 49C298-1-1 Specific Shell Forging RG1.99 0.03'*8 0. 6 5 tel WF-562 442011 6 . 8 2 E+ 18 '*8 +40 '*8 ! . Plant 41 Specific Table 1 Upper Circ. Wald Plant 68 RG1.99 0. 0 5 '*8 -
'O.62*8 8 WF-447 442002 2.192E+19 + 10 -
Specific Table 1 Middle Cire. Weld
+ 4 0 Plant 144.4 RG1.99 0 .18 . ' O . 5 4 '*8 WF-614 31401 < 1. 0 0 E + 17
Specific Table 1 Lower j Circ. Weld NOTES:
- a. Fluence data are from WCAP-12431, "Arnalysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Surveillance Program," October 1989.
- b. Fluence data for weld WF-614 is from the January 17, 1986 letter from G. L. Alexander to H. R.
Denton,
Subject:
Zion Station Units 1 and 2; Byron Station Units 1 and 2; Braidwood Station Units 1 and 2; Pressurized Thermal Shock.
1992 letter from
- c. Chemical compositions and initial RTm data for all materials are from the July 2, M. A. Jackson to T. E. Murley, S"bject: Byron Station, Units 1 and 2.
77-1234176-00 Prepared By: M. J. DeVan Date: 7/1/94 Page 12 Reviewed By: L. B. Gross Date: 7/1/94 5
Table 2-8. Byron Unit 2 -- Data Summary for Upper-Shelf Energy Calculation .
1/4T Method of 1/4T USE Neutron Determin.
Beltline Material at 32 Fluence at Unirrad. Unirrad.
Material Heat Nc. Type EFPY 32 EFPY USE USE Lower 4P-6107 A 508-2 131 '* ' 4 . 0 9E+18 '*' 155"' Direct Nozzle Belt Forging Upper 49D329-1-1/ A 508-3 117 '** 1.66E+19'** 149'" Direct Shell 49C297-1-1 Forging Lower 49D330-1-1/ A 508-3 9 9 1. 6 6 E + 19 127'55 Direct Shell 49C298-1-1 Forging WF-562 442011 Linde 80, 60
- Direct Upper SAW Cire. Weld WF-447 442002 Linde 80, 53
- Direct Middle SAW Circ. Weld WF-614 31401 Linde 80i --- < 1. 0 0 E+17 74
- Direct Lower SAW Circ. Weld Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 13
b
- a. EOL USE values for the forgings were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.1% copper for base metal. .
- b. EOL USE values for the welds were calculated using Regulatory Guide 1.99, Revision 2, Figure 2, assuming the lower limiting value of 0.05% copper for welds.
- c. EOL fluence is below the limits of the Figure 2 curves defined in Regulatory Guide 1.99, Revision 2.
- d. Fluence data are from WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Surveillance Program," March 1991.
I e. Fluence data for weld WF-614 is from the January 17, 1986 letter from G. L. Alexander to H. R. ;
Denton,
Subject:
Zion Station Units 1 and 2; Byron Station Units 1 and 2; Braidwood Station Units 1 and 2; Pressurized Thermal Shock. l
- f. UUSE data for forgings 4P-6107 and 49D329-1-1/49C297-1-1 are from WCAP-12431, " Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation i Surveillance Program," October 1989.
- g. UUSE data for forging 49D330-1-1/49C298-1-1 is from the July 2, 1992 letter from M. A. Jackson to T. E. Murley,
Subject:
Byron Station, Units 1 and 2.
- h. UUSE data for the welds are from the November 19,1993 letter from T. W. Simpkin to T. E. Murley, Braidwood Station Units 1 and 2, Response to Request for Additional Information Regarding NRC Generic Letter 92-01.
I f
Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 ,
Reviewed By: L. B. Gross Date: 7/1/94 Page 14 M
9
- - - . . _ _ . - - -.__.__.-.-----__-__-_-_-_--.__.--_-----_.-a-- - - - - . - - - - - _ _ _ - - - -
- \
PRESSURIZED THERMAL SHOCK AND USE TABLES FOR ALL PWR PLANTS NOMENCLATURE Pressurized Thermal Shock Table !
Column 1: Beltline material location identification. !
I Column 2: Beltline material heat 2. umber; some welds that a single-wire or tandem-wire process has been reported, (s) indicates single wire was used in the SAW process, (T) indicates ,
tandem wire was used in the SAW process. l Column 3: End-of-life (EOL) neutron fluence at vessel inner wall; cited directly from inner diameter (ID) value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 4: Unirradiated reference temperature.
Column 5: Method of determining unirradiated reference temperature (IRT).
Plant-Specific This indicates that the IRT was determined from tests on material removed from the same heat of the beltline material.
Generic This indicates that the unirradiated reference temperature was determined from the mean value of tests on material of I similar types.
l Column 6: Chemistry factor for irradiated reference temperature evaluation.
Column 7: Method of determining chemistry factor.
RG1,99 Table 1 or 2 This indicates that the chemistry factor was determined from i the chemistry factor tables in Regulatory Guide 1.99, !
Revision 2.
Calculated This indicates that the chemistry factor was determined from surveillance data via procedures described in Regulatory Guide 1.99, Revision 2. .
Column 8: Copper content; cited directly from licensee value except when more than one value was reported. (Staff used the average value in the latter case.)
Column 9: Nickel content; cited directly from licensee value except 4 when more than one value was reported. (Staff used the l average value in the latter case.) l Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 15 1
Upper-$helf Energy Table Column 1: Beltline material location identification.
Column 2: Beltline material heat number; some welds that a single-wire or tandem-wire process has been reported, (s) indicates single wire was used in the SAW process, (T) indicates tandem wire was used in the SAW process.
Column 3: Material type; plate types include A 533B-1, A 302B, A 302B Mod.; forging types include A 508-2 and A508-3; weld types include SAW welds using Linde 80, 0091, 124, 1092, ARCOS-B5 flux, Rotterdam welds using Grau Lo, SMIT 89, LW 320, and SAF 89 flux, and SMAW welds using no flux.
Column 4: EOL upper-shelf energy (USE) at T/4; calculated by using the EOL fluence and either the copper value or the surveillance data. (Both methods are described in Regulatory Guide 1.99, Revision 2.)
Column 5: EOL neutron fluence at T/4 from vessel inner wall; cited directly from T/4 value or calculated by using Regulatory Guide 1.99, Revision 2, neutron fluence attenuation methodology from the ID value reported in the latest submittal (GL 92-01, PTS, or P/T limits submittals).
Column 6: Unirradiated USE Column 7: Method of determining unirradiated USE.
Direct For forgings, this indicates that the unirradiated USE was from specimens oriented in the weak direction. For welds, this indicates that the unirradiated USE was from test data.
l l
Prepared By: M. J. DeVan Date: 7/1/94 77-1234176-00 Reviewed By: L. B. Gross Date: 7/1/94 Page 16 w