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Category:CORRESPONDENCE-LETTERS
MONTHYEARSVP-99-181, Forwards Biennial Update to Quad Cities Ufsar,Iaw 10CFR50.71(e).Update Includes Changes to Facility & Procedures & Are Current Through 9906301999-10-20020 October 1999 Forwards Biennial Update to Quad Cities Ufsar,Iaw 10CFR50.71(e).Update Includes Changes to Facility & Procedures & Are Current Through 990630 ML20217H4661999-10-18018 October 1999 Forwards Changes to EPIPs IAW 10CFR50.54(q) & 10CFR50,App E, Section V.Details of Changes Encl ML20217K7011999-10-13013 October 1999 Provides Response to Questions Related to Request for License Amend,Per 10CFR50.90, Credit for Containment Overpressure. Supporting Calculations Encl 05000254/LER-1999-004, Forwards LER 99-004-00,IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Is Committed to Listed Action1999-10-12012 October 1999 Forwards LER 99-004-00,IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Is Committed to Listed Action ML20217F6321999-10-0707 October 1999 Forwards Insp Repts 50-254/99-01 & 50-265/99-01 on 990721- 0908.No Violations 05000254/LER-1999-003, Forwards LER 99-003-00 IAW 10CFR73(a)(2)(v)(A).Commitment Made by Util,Listed.Any Other Actions Described in LER Represent Intended or Planned Actions by Licensee1999-10-0707 October 1999 Forwards LER 99-003-00 IAW 10CFR73(a)(2)(v)(A).Commitment Made by Util,Listed.Any Other Actions Described in LER Represent Intended or Planned Actions by Licensee ML20212K9421999-10-0505 October 1999 Informs That NRC Accepts 990513 Inservice Inspection Relief Request CR-31 for Quad Cities Nuclear Power Station,Units 1 & 2 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 SVP-99-189, Confirms Completion of Actions Identified During Review of NRC Safety Evaluation of Boiling Water Reactor Owners Group Rept, Utility Resolution Guidance of ECCS Suction Strainer Blockage1999-09-22022 September 1999 Confirms Completion of Actions Identified During Review of NRC Safety Evaluation of Boiling Water Reactor Owners Group Rept, Utility Resolution Guidance of ECCS Suction Strainer Blockage ML20212J0451999-09-21021 September 1999 Forwards Safety Evaluation of Licensee USI A-46 Program at Quad Cities Nuclear Power Station,Units 1 & 2,established in Response to GL 87-02 Through 10CFR50.54(f) Ltr ML20212D8231999-09-20020 September 1999 Informs That Effectieve 991101,NRC Region III Will Be Conducting Safety System Design & Performance Capability Pilot Insp at Quad Cities Nuclear Power Station.Insp Will Be Performed IAW NRC Pilot Insp Procedure 71111-21 ML20212C6961999-09-15015 September 1999 Forwards Insp Repts 50-254/99-17 & 50-265/99-17 on 990823- 0827.No Violations Noted SVP-99-190, Clarifies Statement Contained in NRC Insp Repts 50-254/99-12 & 50-265/99-12,dtd 990813,paragraph 4OA1.3B, Observation & Findings, Re Sys Mod to DG Air Start Sys1999-09-13013 September 1999 Clarifies Statement Contained in NRC Insp Repts 50-254/99-12 & 50-265/99-12,dtd 990813,paragraph 4OA1.3B, Observation & Findings, Re Sys Mod to DG Air Start Sys ML20211Q7961999-09-0909 September 1999 Forwards Correction to Administrative Error on Page 8 of NRC Insp Repts 50-254/99-16 & 50-265/99-16,transmitted by Ltr, ML20217H5661999-09-0909 September 1999 Discusses 990907 Pilot Plan Mgt Meeting Re Results to-date of Pilot Implementation of NRC Revised Reactor Oversight Process at Prairie Island & Quad Cities.Agenda & Handouts Provided by Utils Encl ML20211Q6511999-09-0808 September 1999 Informs That NRC Tentatively Scheduled Initial Licensing Exam for Quad Cities Operator License Applicants During Wk of 000327.Validation of Exam Will Occur at Station During Wk of 000306 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) ML20211F8251999-08-25025 August 1999 Forwards Insp Repts 50-254/99-15 & 50-265/99-15 on 990816-20.No Violations Noted.Insp Evaluated Effectiveness of Maint Rule Program & Review Periodic Evaluation Specifically Required for 10CFR50.65 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 ML20211D1491999-08-19019 August 1999 Forwards Insp Repts 50-254/99-16 & 50-265/99-16 on 990719-22.Staff Identified Major Discrepancy Re Accuracy of Data Submitted to NRC for Protected Area Security Equipment Performance ML20211C7601999-08-19019 August 1999 Confirms NRC Intent to Meet with NSP & Ceco on 990807 in Lisle,Il to Discuss with Region III Pilot Plants,Any Observations,Feedback,Lessons Learned & Recommendations Relative to Implementation of Pilot Program ML20210R7451999-08-13013 August 1999 Forwards Insp Repts 50-254/99-11 & 50-265/99-11 on 990601-0720.NRC Identified Several Issues Which Were Categorized as Being of Low Risk Significance.Two Issues Involved NCVs of Regulatory Requirements SVP-99-147, Notifies of Change to Bases of TS to Licenses DPR-29 & DPR- DPR-30.Change to TS Section 3/4.9 Provides Clarity & Consistency with Sys Design Description in Ufsar,Sections 8.3.2.1 & 8.3.2.2.TS Bases Page,Encl1999-08-13013 August 1999 Notifies of Change to Bases of TS to Licenses DPR-29 & DPR- DPR-30.Change to TS Section 3/4.9 Provides Clarity & Consistency with Sys Design Description in Ufsar,Sections 8.3.2.1 & 8.3.2.2.TS Bases Page,Encl SVP-99-170, Forwards Relief Requests CR-25,CR-26,CR-27,CR-28,PR-11, PR-12 & PR-13,on Basis That Compliance with Specified Requirements Would Result in Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality & Safety1999-08-13013 August 1999 Forwards Relief Requests CR-25,CR-26,CR-27,CR-28,PR-11, PR-12 & PR-13,on Basis That Compliance with Specified Requirements Would Result in Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality & Safety ML20210T9941999-08-13013 August 1999 Forwards Insp Repts 50-254/99-12 & 50-265/99-12 on 990628-0716.Violations Noted SVP-99-154, Notifies NRC That M Price,License SOP-31389,is No Longer Required to Maintain Operator License.Price Was Removed from Licensed Duty on 990729 & Comm Ed Requests License Be Terminated1999-08-13013 August 1999 Notifies NRC That M Price,License SOP-31389,is No Longer Required to Maintain Operator License.Price Was Removed from Licensed Duty on 990729 & Comm Ed Requests License Be Terminated ML20210R9541999-08-10010 August 1999 Informs That During 990804 Telcon Between J Bartlet & M Bielby,Arrangements Were Made for NRC to Insp License Operator Requalification Program at Plant.Insp Planned for Wks of 991018 & 25 ML20210N5651999-08-0606 August 1999 Forwards Rev 8 to Nuclear Generating Stations Emergency Plan, for Plants.With Summary of Changes ML20210M5461999-08-0606 August 1999 Discusses 990804 Telcon Between J Bartlet & M Bielby,Where Arrangements Were Made for NRC to Inspect Licensed Operator Requalification Program at Plant.Insp Planned for Wks of 991018 & 25 ML20210M9131999-08-0202 August 1999 Forwards Response to NRC AL 99-02, Operating Reactor Licensing Action Estimates, for Fys 2000 & 2001 for Comed ML20210L8371999-08-0202 August 1999 Forwards SE Accepting Licensee 60-day Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety Related Motor-Operated Valves ML20210M4691999-07-30030 July 1999 Forwards Insp Repts 50-254/99-14 & 50-265/99-14 on 990713-15.One NCV Was Identified & Discussed in Encl Insp ML20210H4661999-07-29029 July 1999 Forwards Insp Repts 50-254/99-13 & 50-265/99-13 on 990628-0702.No Violations Noted.Insp Consisted of Selective Examination of Procedures & Representative Records, Observations of Activities & Interviews with Personnel 05000254/LER-1999-002, Forwards LER 99-002-01,IAW 10CFR50.73(a)(2)(iv).Rev Includes Cause Codes & Energy Industry Identification Sys Identifiers Which Were Erroneously Omitted in Original Ler.Commitments Listed1999-07-29029 July 1999 Forwards LER 99-002-01,IAW 10CFR50.73(a)(2)(iv).Rev Includes Cause Codes & Energy Industry Identification Sys Identifiers Which Were Erroneously Omitted in Original Ler.Commitments Listed SVP-99-151, Responds to NRC 990701 Telcon RAI Re Licensee Amend Request Re Use of Containment Overpressure to Support NPSH Available for ECCS at Quad Cities Nuclear Power Station,Units 1 & 21999-07-23023 July 1999 Responds to NRC 990701 Telcon RAI Re Licensee Amend Request Re Use of Containment Overpressure to Support NPSH Available for ECCS at Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-150, Forwards Responses to 990520 RAI Re Annual 10CFR50.46 Rept1999-07-23023 July 1999 Forwards Responses to 990520 RAI Re Annual 10CFR50.46 Rept SVP-99-146, Informs of Termination of License SOP-31132-1,for G Green, as Required by 10CFR50.74(b).Individual Was Removed from License Duty on 9906251999-07-21021 July 1999 Informs of Termination of License SOP-31132-1,for G Green, as Required by 10CFR50.74(b).Individual Was Removed from License Duty on 990625 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) 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 SVP-99-139, Forwards Revised Action 2 & Associated TS Bases Changes, Describing Specific Alternate Method for Determining Drywell Floor Drain Sump Leakage1999-06-30030 June 1999 Forwards Revised Action 2 & Associated TS Bases Changes, Describing Specific Alternate Method for Determining Drywell Floor Drain Sump Leakage ML20209B2081999-06-29029 June 1999 Discusses Closure of Response to RAI Re GL 92-01,Rev 1,Suppl 1, Reactor Vessel Structural Integrity. Rvid,Version 2 Issued as Result of Review of Responses.Info Should Be Reviewed & Comments Submitted by 990901 05000265/LER-1999-002, Forwards LER 99-002-00 for Quad Cities Nuclear Power Station IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Commits to One Listed Action1999-06-25025 June 1999 Forwards LER 99-002-00 for Quad Cities Nuclear Power Station IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Commits to One Listed Action SVP-99-122, Forwards Regulatory Commitment Change Summary Rept, Containing Summary Info from 980601 Through 9906011999-06-25025 June 1999 Forwards Regulatory Commitment Change Summary Rept, Containing Summary Info from 980601 Through 990601 SVP-99-066, Informs That J Reed,License SOP-31034-1,was Removed from License Duty in 990618.Termination of License Is Requested1999-06-25025 June 1999 Informs That J Reed,License SOP-31034-1,was Removed from License Duty in 990618.Termination of License Is Requested SVP-99-103, Informs NRC of Results of Subject Evaluation as Committed to in .Evaluation Consisted of Analytically Demonstrating That Cfu Factor for Rv Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period1999-06-25025 June 1999 Informs NRC of Results of Subject Evaluation as Committed to in .Evaluation Consisted of Analytically Demonstrating That Cfu Factor for Rv Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period ML20196F7921999-06-24024 June 1999 Forwards Meeting Summary,Nrc Meeting Handout & Licensee Handout from 990608 Meeting ML20196E7131999-06-23023 June 1999 Forwards Insp Repts 50-254/99-09 & 50-265/99-09 on 990421-0531.One Violation of NRC Requirements Occurred & Being Treated as non-cited Violation,Consistent with App C of Enforcement Policy ML20196E4821999-06-21021 June 1999 Discusses 990617 Meeting by Region III Senior Reactor Analysts (SRA) in Cordova,Il to Meet with PRA Staff to Discuss Initiatives in Risk Area & to Establish Dialog Between SRAs & PRA Staff 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARSVP-99-181, Forwards Biennial Update to Quad Cities Ufsar,Iaw 10CFR50.71(e).Update Includes Changes to Facility & Procedures & Are Current Through 9906301999-10-20020 October 1999 Forwards Biennial Update to Quad Cities Ufsar,Iaw 10CFR50.71(e).Update Includes Changes to Facility & Procedures & Are Current Through 990630 ML20217H4661999-10-18018 October 1999 Forwards Changes to EPIPs IAW 10CFR50.54(q) & 10CFR50,App E, Section V.Details of Changes Encl ML20217K7011999-10-13013 October 1999 Provides Response to Questions Related to Request for License Amend,Per 10CFR50.90, Credit for Containment Overpressure. Supporting Calculations Encl 05000254/LER-1999-004, Forwards LER 99-004-00,IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Is Committed to Listed Action1999-10-12012 October 1999 Forwards LER 99-004-00,IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Is Committed to Listed Action 05000254/LER-1999-003, Forwards LER 99-003-00 IAW 10CFR73(a)(2)(v)(A).Commitment Made by Util,Listed.Any Other Actions Described in LER Represent Intended or Planned Actions by Licensee1999-10-0707 October 1999 Forwards LER 99-003-00 IAW 10CFR73(a)(2)(v)(A).Commitment Made by Util,Listed.Any Other Actions Described in LER Represent Intended or Planned Actions by Licensee 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 SVP-99-189, Confirms Completion of Actions Identified During Review of NRC Safety Evaluation of Boiling Water Reactor Owners Group Rept, Utility Resolution Guidance of ECCS Suction Strainer Blockage1999-09-22022 September 1999 Confirms Completion of Actions Identified During Review of NRC Safety Evaluation of Boiling Water Reactor Owners Group Rept, Utility Resolution Guidance of ECCS Suction Strainer Blockage SVP-99-190, Clarifies Statement Contained in NRC Insp Repts 50-254/99-12 & 50-265/99-12,dtd 990813,paragraph 4OA1.3B, Observation & Findings, Re Sys Mod to DG Air Start Sys1999-09-13013 September 1999 Clarifies Statement Contained in NRC Insp Repts 50-254/99-12 & 50-265/99-12,dtd 990813,paragraph 4OA1.3B, Observation & Findings, Re Sys Mod to DG Air Start Sys 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) SVP-99-154, Notifies NRC That M Price,License SOP-31389,is No Longer Required to Maintain Operator License.Price Was Removed from Licensed Duty on 990729 & Comm Ed Requests License Be Terminated1999-08-13013 August 1999 Notifies NRC That M Price,License SOP-31389,is No Longer Required to Maintain Operator License.Price Was Removed from Licensed Duty on 990729 & Comm Ed Requests License Be Terminated SVP-99-170, Forwards Relief Requests CR-25,CR-26,CR-27,CR-28,PR-11, PR-12 & PR-13,on Basis That Compliance with Specified Requirements Would Result in Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality & Safety1999-08-13013 August 1999 Forwards Relief Requests CR-25,CR-26,CR-27,CR-28,PR-11, PR-12 & PR-13,on Basis That Compliance with Specified Requirements Would Result in Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality & Safety SVP-99-147, Notifies of Change to Bases of TS to Licenses DPR-29 & DPR- DPR-30.Change to TS Section 3/4.9 Provides Clarity & Consistency with Sys Design Description in Ufsar,Sections 8.3.2.1 & 8.3.2.2.TS Bases Page,Encl1999-08-13013 August 1999 Notifies of Change to Bases of TS to Licenses DPR-29 & DPR- DPR-30.Change to TS Section 3/4.9 Provides Clarity & Consistency with Sys Design Description in Ufsar,Sections 8.3.2.1 & 8.3.2.2.TS Bases Page,Encl 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 05000254/LER-1999-002, Forwards LER 99-002-01,IAW 10CFR50.73(a)(2)(iv).Rev Includes Cause Codes & Energy Industry Identification Sys Identifiers Which Were Erroneously Omitted in Original Ler.Commitments Listed1999-07-29029 July 1999 Forwards LER 99-002-01,IAW 10CFR50.73(a)(2)(iv).Rev Includes Cause Codes & Energy Industry Identification Sys Identifiers Which Were Erroneously Omitted in Original Ler.Commitments Listed SVP-99-150, Forwards Responses to 990520 RAI Re Annual 10CFR50.46 Rept1999-07-23023 July 1999 Forwards Responses to 990520 RAI Re Annual 10CFR50.46 Rept SVP-99-151, Responds to NRC 990701 Telcon RAI Re Licensee Amend Request Re Use of Containment Overpressure to Support NPSH Available for ECCS at Quad Cities Nuclear Power Station,Units 1 & 21999-07-23023 July 1999 Responds to NRC 990701 Telcon RAI Re Licensee Amend Request Re Use of Containment Overpressure to Support NPSH Available for ECCS at Quad Cities Nuclear Power Station,Units 1 & 2 SVP-99-146, Informs of Termination of License SOP-31132-1,for G Green, as Required by 10CFR50.74(b).Individual Was Removed from License Duty on 9906251999-07-21021 July 1999 Informs of Termination of License SOP-31132-1,for G Green, as Required by 10CFR50.74(b).Individual Was Removed from License Duty on 990625 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 SVP-99-139, Forwards Revised Action 2 & Associated TS Bases Changes, Describing Specific Alternate Method for Determining Drywell Floor Drain Sump Leakage1999-06-30030 June 1999 Forwards Revised Action 2 & Associated TS Bases Changes, Describing Specific Alternate Method for Determining Drywell Floor Drain Sump Leakage SVP-99-103, Informs NRC of Results of Subject Evaluation as Committed to in .Evaluation Consisted of Analytically Demonstrating That Cfu Factor for Rv Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period1999-06-25025 June 1999 Informs NRC of Results of Subject Evaluation as Committed to in .Evaluation Consisted of Analytically Demonstrating That Cfu Factor for Rv Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period SVP-99-122, Forwards Regulatory Commitment Change Summary Rept, Containing Summary Info from 980601 Through 9906011999-06-25025 June 1999 Forwards Regulatory Commitment Change Summary Rept, Containing Summary Info from 980601 Through 990601 05000265/LER-1999-002, Forwards LER 99-002-00 for Quad Cities Nuclear Power Station IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Commits to One Listed Action1999-06-25025 June 1999 Forwards LER 99-002-00 for Quad Cities Nuclear Power Station IAW Requirements of 10CFR50.73(a)(2)(i)(B).Util Commits to One Listed Action SVP-99-066, Informs That J Reed,License SOP-31034-1,was Removed from License Duty in 990618.Termination of License Is Requested1999-06-25025 June 1999 Informs That J Reed,License SOP-31034-1,was Removed from License Duty in 990618.Termination of License Is Requested SVP-99-125, Forwards Technical Info Re ECCS Suction Strainers at Quad Cities Nuclear Power Station Units 1 & 2,to Support Review of 990129 Lar.Rev 0 to TR-VQ1500-02, Clean ECCS Suction Strainer Head Loss Test Rept, Encl1999-06-15015 June 1999 Forwards Technical Info Re ECCS Suction Strainers at Quad Cities Nuclear Power Station Units 1 & 2,to Support Review of 990129 Lar.Rev 0 to TR-VQ1500-02, Clean ECCS Suction Strainer Head Loss Test Rept, Encl ML20195E3491999-06-0707 June 1999 Withdraws Util Requesting License Change for Plant Security Plan Rev.Licensee Will re-evaluate Situation & May Request Approval of Change in Future ML20207G1451999-06-0707 June 1999 Forwards Rev 45 to Comed Quad Cities Nuclear Power Station Security Plan.Rev Includes Changes Listed.Security Plan Is Withheld from Public Disclosure Per 10CFR73.21 ML20195D6351999-06-0404 June 1999 Notifies NRC of Actions That Has Been Taken in Accordance with 10CFR26, Fitness for Duty Programs SVP-99-105, Informs NRC of Rev to Schedule for Completing Setpoint/ Uncertainty Calculations & Procedure Changes,Originally Planned for Completion on 9905291999-05-20020 May 1999 Informs NRC of Rev to Schedule for Completing Setpoint/ Uncertainty Calculations & Procedure Changes,Originally Planned for Completion on 990529 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 SVP-99-111, Informs NRC of Current Status of Actions on Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions1999-05-17017 May 1999 Informs NRC of Current Status of Actions on Response to GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Conditions SVP-99-098, Fulfills Thirty Day & Annual Reporting Requirements of 10CFR50.46 for Plant.Eccs Evaluation Change Rept Transmitted in Entirety,Fulfilling Thirty Day & Annual Reporting Requirements Specified in 10CFR50.46(a)(3)(i)1999-05-17017 May 1999 Fulfills Thirty Day & Annual Reporting Requirements of 10CFR50.46 for Plant.Eccs Evaluation Change Rept Transmitted in Entirety,Fulfilling Thirty Day & Annual Reporting Requirements Specified in 10CFR50.46(a)(3)(i) SVP-99-099, Requests Relief from Requirements of 10CFR50.55a(g) Re Submittal of Relief Requests for Those Welds for Which Examinations of Greater than 90% of Weld Vol Was Not Acheived During 2nd ISI Program Interval1999-05-13013 May 1999 Requests Relief from Requirements of 10CFR50.55a(g) Re Submittal of Relief Requests for Those Welds for Which Examinations of Greater than 90% of Weld Vol Was Not Acheived During 2nd ISI Program Interval SVP-99-096, Provides Suppl Response to Violations Noted in Insp Repts 50-254/98-20 & 50-265/98-23.Corrective Actions:Listed Multidiscipline Team Will Perform self-assessment IAW Station Program for self-assessments in May 19991999-05-12012 May 1999 Provides Suppl Response to Violations Noted in Insp Repts 50-254/98-20 & 50-265/98-23.Corrective Actions:Listed Multidiscipline Team Will Perform self-assessment IAW Station Program for self-assessments in May 1999 05000254/LER-1999-001, Forwards LER 99-001-00 for Quad Cities Nuclear Power Station.Licensee Shall Rept Any Operation or Condition Prohibited by Plant Tech Specs.Util Committing to Listed Actions1999-05-12012 May 1999 Forwards LER 99-001-00 for Quad Cities Nuclear Power Station.Licensee Shall Rept Any Operation or Condition Prohibited by Plant Tech Specs.Util Committing to Listed Actions 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 SVP-99-108, Forwards Quad Cities 1998 Radiological Environ Operating Rept, IAW Plant TS 6.9.A.3.Rept Contains Results of Radiological Environ & Meteorological Monitoring Programs. Radioactive Effluent Release Rept Was Submitted 9903301999-04-30030 April 1999 Forwards Quad Cities 1998 Radiological Environ Operating Rept, IAW Plant TS 6.9.A.3.Rept Contains Results of Radiological Environ & Meteorological Monitoring Programs. Radioactive Effluent Release Rept Was Submitted 990330 SVP-99-036, Forwards Reg Guide 1.16 Rept Number of Personnel-Rem by Work & Job Function for 1998. Associated Collective Deep Dose Equivalent Reported According to Work & Job Functions1999-04-29029 April 1999 Forwards Reg Guide 1.16 Rept Number of Personnel-Rem by Work & Job Function for 1998. Associated Collective Deep Dose Equivalent Reported According to Work & Job Functions SVP-99-088, Informs That Util Is Withdrawing IST Relief Rquests RV-02B & RV-03B1999-04-29029 April 1999 Informs That Util Is Withdrawing IST Relief Rquests RV-02B & RV-03B ML20205T1141999-04-22022 April 1999 Provides Comments from Technical Review of Draft Info Notice Re Unanticipated Reactor Water Draindown at Quad Cities Nuclear Power Station,Unit 2,ANO,Unit 2 & JAFNPP 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 SVP-99-065, Requests That License SOP-31516,for Jf Graham,Be Terminated, Per 10CFR50.74(b).Individual Was Removed from License Duty on 990319 & No Longer Requires Operator License1999-04-14014 April 1999 Requests That License SOP-31516,for Jf Graham,Be Terminated, Per 10CFR50.74(b).Individual Was Removed from License Duty on 990319 & No Longer Requires Operator License SVP-99-058, Submits Plant Specific ECCS Evaluation Changes,Per Annual Reporting Requirements of 10CFR50.46.Attachments Include Current Assessment Data Re PCT Info Limiting LOCA Evaluations1999-04-14014 April 1999 Submits Plant Specific ECCS Evaluation Changes,Per Annual Reporting Requirements of 10CFR50.46.Attachments Include Current Assessment Data Re PCT Info Limiting LOCA Evaluations SVP-99-063, Responds to NRC Re Violations Noted in Insp Repts 50-254/98-21 & 50-265/98-21.Corrective Actions:Revised Site ISI Procedure Qcap 0410-06, ISI Plan Implementation for Third Ten Year Insp Interval1999-04-0909 April 1999 Responds to NRC Re Violations Noted in Insp Repts 50-254/98-21 & 50-265/98-21.Corrective Actions:Revised Site ISI Procedure Qcap 0410-06, ISI Plan Implementation for Third Ten Year Insp Interval JAFP-99-0129, Submits Comments on Technical Review of Draft Info Notice Describing Unanticipated Reactor Water Draindown at Quad Cities Unit 2,ANO & FitzPatrick1999-04-0909 April 1999 Submits Comments on Technical Review of Draft Info Notice Describing Unanticipated Reactor Water Draindown at Quad Cities Unit 2,ANO & FitzPatrick SVP-99-057, Notifies of Change to Bases for TSs Section 3/4.5, ECCS, Re1999-04-0505 April 1999 Notifies of Change to Bases for TSs Section 3/4.5, ECCS, Re 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) SVP-99-062, Informs NRC of Rev to Schedule for Analytically Demonstrating That Cumulative Usage Factor for Reactor Vessel Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period1999-03-31031 March 1999 Informs NRC of Rev to Schedule for Analytically Demonstrating That Cumulative Usage Factor for Reactor Vessel Head Closure Studs Will Remain Below 1.0 for Remainder of Current License Period 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20065D0511990-09-17017 September 1990 Forwards Objectives & Scope of 901205 Emergency Plan Exercise ML20064A7091990-09-14014 September 1990 Forwards Endorsement 133 to Nelia Policy NF-187 & Endorsement 116 to Maelu Policy MF-54 ML20059F4891990-09-0404 September 1990 Forwards Listing of Changes,Tests & Experiments Completed During Month of Aug 1990 for Plant ML20059B9721990-08-28028 August 1990 Forwards Reactor Head & Upper Shell Insp Plan,Per 900419 Meeting.Insp Plan Does Not Encompass Uppermost shell-to- Shell Weld Due to Technological Limitations ML20059F0311990-08-27027 August 1990 Provides Schedule for Completion of Installation of Mods to Plants Reactor Water Level Instrumentation,Per Generic Ltr 84-23.Penetrations Will Be Installed During Outage 13 for Dresden & During Outage 12 for Quad-Cities ML20059E9531990-08-27027 August 1990 Forwards Summary of Fabrication History for Upper Reactor Vessel,Per 900419 Technical Meeting.Summary Indicates That Fabrication Mismatches,Considered to Be Significant for Development of Insp Plan,Identified at head-to-flange Weld ML20059C7201990-08-23023 August 1990 Forwards Effluent & Waste Disposal Semiannual Rept,Jan-June 1990 Gaseous Effluents-Summation of All Releases & Rev 8 to Quad-Cities Station Process Control Program for Processing of Radioactive Wet Waste ML20058P3481990-08-0909 August 1990 Forwards Summary of Fuel Performance,End of Cycle 10,May 1990. No Leakage or Fuel Failure Noted ML20058M8221990-08-0707 August 1990 Forwards Response to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issues Resolved W/Imposition of Requirements or Corrective Actions ML20058M8041990-08-0606 August 1990 Advises That W/Completion of Operator Training Program,Plant SPDS Meets Requirements Delineated in NUREG-0737,Suppl 1 ML20058M8591990-08-0606 August 1990 Forwards Rept of Metallurgical Exam That Revealed No Evidence of Defects,Porosity or Slag in Weld Overlay. Rept Responds to IGSCC Insp Performed on Facility IGSCC Susceptible Piping ML20058M4101990-08-0101 August 1990 Forwards Listing of Changes,Tests & Experiments Completed During Month of Jul 1990 for Plant ML20058M8291990-07-31031 July 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Licensee Implementation of Generic Safety Issue Resolved W/Imposition of Requirements of Corrective Actions. Status of Implementation of Generic Safety Issues Encl ML20055J1631990-07-26026 July 1990 Forwards Reactor Containment Bldg Integrated Leak Rate Test Quad-Cities Nuclear Power Station Unit 2,900427-28, & Related Apps Describing Type a Test,Per 10CFR50,App J, Section V.B.1.Next Test Scheduled for Fall 1991 ML20055H7631990-07-25025 July 1990 Forwards Financial Info Re Decommissioning of Plants ML20055G6331990-07-18018 July 1990 Responds to Generic Ltr 89-06 Re SPDS to Meet Requirements of Suppl 1 to NUREG-0737.SPDS Lesson Plan Incorporated Into Initial License Class Training Program ML17202L2861990-07-0202 July 1990 Forwards Dresden II Upper Vessel Contract Variation Review, La Salle II Upper Vessel Fabrication Summary & Quad-Cities II Upper Vessel Fabrication Summary. ML20055D4741990-06-29029 June 1990 Forwards Annual FSAR Update for Quad-Cities Station ML20055D4341990-06-29029 June 1990 Forwards Comm Ed Rept on Evaluation of Cracking in Quad- Cities Unit 2 Reactor Head, Per Commitment Made at 900419 Meeting W/Nrr.Rept Concludes That Cracks Caused by Interdendritic Stress Corrosion Cracking Mechanism ML20055C8551990-06-15015 June 1990 Forwards Special Neutron Attenuation Test for High Density Spent Fuel Racks (Wet), Final Rept.Rept Provides Results of Neutron Radioassay Measurement Program Conducted During Fall,1989 Refueling Outage ML20043D7661990-06-0404 June 1990 Responds to J Lieberman 900501 Ltr Re Rl Dickherber. Confidence in Dickherber Performance in Future for Nonlicensed Duties Can Be Based Upon Demonstrated Record of Good Past Performance ML20043D7691990-06-0404 June 1990 Responds to 900501 Ltr Re Work Hours for Dickherber.During Outage,Dickherber Worked Extended Hours Traditionally Associated W/Refueling Activities ML20043G4251990-06-0202 June 1990 Forwards Listing of Changes,Tests & Experiments Completed During May 1990 ML20043D3201990-06-0101 June 1990 Forwards Rev 24 to Security Plan.Rev Withheld (Ref 10CFR73.21) ML20043B6681990-05-22022 May 1990 Forwards Proposed Changes to SER Re Hot Shutdown Repairs in Event of Fire,Per 10CFR50,App R Section Iii.G Covering Spurious Operations & High Impedance Faults & Electrical Isolation Deficiency ML20043A4681990-05-10010 May 1990 Forwards Proposed Changes to 880721 SER Re App R Section Iii.G Exemption for Fire Zones 1.1.1.1S & 1.1.1.2,southern & Northern Torus Level in Unit 1 Reactor Bldg Column & Unit 1 Reactor Bldg Elevations 623 Ft & 647 Ft ML20042H0011990-05-0303 May 1990 Forwards Listing of Changes,Tests, & Experiments Completed During Apr 1990 ML20042G3501990-05-0202 May 1990 Responds to NRC 900404 Ltr Re Violations Noted in Insp Repts 50-254/90-02 & 50-265/90-02.Corrective Actions:Continuous Fire Watch Initiated & Training Conducted on Procedure Rev ML20042F1181990-05-0101 May 1990 Advises of Listed Value for Secondary Containment,Per NRC Request for Addl Info Re LER 50-254/87-025.Value Based on Info Contained in Plant FSAR ML20042F0691990-05-0101 May 1990 Responds to Generic Ltr 83-28,Item 4.5.3 Re Reactor Protection Sys on-line Functional Test Intervals.Endorses Two BWR Owners Group Topical Repts NEDC-30844 & NEDC-30851P Generic Evaluations ML20042F1221990-05-0101 May 1990 Forwards Preliminary Rept of IGSCC Insp Results.Flaw Indication Detected in Weld Overlay Matl of Weld 02J-S3 & Removed by Boat Sample & Std Weld Overlay Thickness Restored.Final Rept Will Be Forwarded within 30 Days ML20042E4491990-04-11011 April 1990 Forwards Request for Rev to Previous NRC Exemption Approval on 860625 Re Combustible Load Values ML20042F0351990-03-23023 March 1990 Forwards Part 3 of 1989 Operating Rept.W/O Rept ML19330D5161990-03-14014 March 1990 Advises That Revs to Inservice Testing Program & Implementation Procedures Will Be Completed by 900629,per Generic Ltr 89-04 ML20012C0721990-03-0808 March 1990 Comments on SALP Board Repts 50-254/89-01 & 50-265/89-01 for Oct 1988 to Nov 1989.Util Appreciates NRC Recognition of Overall Improvements in Areas of Operation & Emergency Preparedness & Good Performance in Area of Security ML20012B5921990-03-0202 March 1990 Forwards Listing of Changes,Tests & Experiments Computed During Month of Feb 1990 for Plant ML20006F3361990-02-0808 February 1990 Responds to NRC Ltr 900110 Ltr Re Violations Noted in Insp Repts 50-254/89-25 & 50-265/89-25.Corrective Actions:Safety Evaluations Submitted Via 900116 Ltr & Table of Content Will Be Completed for 1989 FSAR Update to Be Submitted by 900630 ML20012A9551990-02-0808 February 1990 Responds to Violations Noted in Insp Repts 50-254/89-26 & 50-265/89-26.Corrective Action:Procedure Qis 47-1 Revised to Include Requirement That Equalizing Valve Be Open During Isolation of Transmitter ML20011E7131990-02-0606 February 1990 Forwards Reactor Containment Bldg Integrated Leak Rate Test,Quad Cities Nuclear Power Station,Unit 1,891114-15. Next Type a Test Scheduled for Fall 1990 ML20006E1721990-02-0202 February 1990 Forwards Listing of Changes,Tests & Experiments Completed During Jan 1990,including Items Completed in 1989. Interlocks Installed on Refuel Bridge Fuel Handling Machine to Prevent Raising Hoist While Hoist Loaded ML20006C5071990-01-30030 January 1990 Identifies Schedular Change for Completion of Corrective Actions Associated W/Human Engineering Deficiencies 159,187 & 489 Re Escutcheon Plates for Control Switches Which Need Replacement.Plates Will Be Replaced During Outages ML20006C7401990-01-22022 January 1990 Advises of Receipt of Accreditation Renewal by INPO in Sept 1989 for Operator Requalification Training Program,Per Generic Ltr 87-07 Requirements & Informs That Programs Developed Using Systematic Approach to Training ML19354E8591990-01-16016 January 1990 Responds to NRC 891128 Ltr Re Violations Noted in Insp Repts 50-254/89-17 & 50-265/89-17.Corrective Actions:Procedure NSWP-E-01, Electrical Cable Installation Insp, Will Be Revised to Enhance Human Factor Aspect ML19354D8131990-01-11011 January 1990 Forwards Corrected App C to Monthly Operating Rept for Dec 1989 for Quad Cities Units 1 & 2 ML20005F6441990-01-0303 January 1990 Forwards Listing of Changes,Tests & Experiments Completed During Dec 1989.Summary of Safety Evaluations Being Reported in Compliance w/10CFR50.59 & 10CFR50.71(e) Also Encl ML20005E1691989-12-22022 December 1989 Forwards Rev 22 to Security Plan,Reflecting Administrative Changes in Mgt Structure at Facility.Rev Withheld (Ref 10CFR73.21) ML20043A5741989-12-21021 December 1989 Responds to NRC 891124 Ltr Re Violations Noted in Insp Repts 50-254/89-23 & 50-265/89-23.Corrective Actions:Compressed Gas Cylinder Bottles Secured W/Chain & Fire Marshall Will Increase Tours of Plant Re Transient Combustible Matl ML20005E1211989-12-18018 December 1989 Forwards Final Rept of Fall 1989 IGSCC Insp Plan,Discussing Items Such as Overlay Repair on Weld 02G-S4,mechanical Stress Improvement & Piping Mods ML19332G3401989-12-0808 December 1989 Forwards Response to Generic Ltr 89-21, Implementation Status of USI Requirements. Actions to Resolve USI A-9 Re ATWS Will Be Completed in June 1990 & USI A-42 Re Pipe Cracks in BWRs Will Be Completed in Dec 1990 ML19332F9091989-12-0101 December 1989 Forwards Listing of Changes,Tests & Experiments Completed During Nov 1989 1990-09-04
[Table view] |
Text
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- -Commonwealth Edison'=
1 Quad Cities Nuclear Power Station -
22710 206 Avenue North .
4:e; . Coreove, Illinois 612424
-g ' Telephone 309/654-2241 J g!l p , - '
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RAR-89-69 u
October 2,.1989L
[b Director of: Nuclear _ Reactor Regulations U. S. Nuclear Regulatory Commission
. Mail' Station PI-137-3 .- -Washington, D. C. 20555 i
=Enclased please. find'a listing of-those changes, tests, and experiments
. completed during the month of September,'1989, for Quad-Cities Station
. Units ~l'-and 2.-DPR-29 and DPR-30. A summary of the safety evaluations are: being -reported.'in compliance with 10CFR50.59 and 10CFR50.71(e).
E -? ~~ Thirty-nine copies are provided for your use.
g
. Respectfully,z
. COMMONWEALTH EDISON COMPANY
. QUAD-C1 TIES NUCLEAR POWER-STATION
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R..A. Rober .
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Technical: Superintendent .
RAR/LFD/vmk .
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. Enclosure cc: - R. Stols-H - T. Watts /J. Galligan u
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'8910170233 891002 ]
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- Procedure Change-QOSf2300-1 4 1
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F. :This' revision provides clarification of IST requirements, system:startup+
j isteps and provides for-shutdown of the drywell-torus differential pressure: control-
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, system during HPCI-testing;if,drywell pressure becomes' excessive.
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l..-TheLprobability of an occurrence or' the consequence of an accident, j
he t j' -or. malfunction of equipment.important.to safety as previously; evaluated ,
s In in-the Final Safety Analysis' Report is not increased because-this revision. is I -should decrease the< probability of:an accident by clarifying-certain. :!
hi; steps in the-test procedure'. .Also,.this change should ensure that highi 4:
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drywellLpressuresLare avoided during HPCI: testing by allowing shutdown
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of the.drywell-torus dp control system.
"* 2.- The possibility for-an accident or malfunction'of a different-type than any'previously evaluated in the Final Safety Analysis' Report is not 1.
7 created because the basic method of system' testing remains unchanged,
- .therefore =no new possibility of an accident or malfunction is created.
3 '. Th'e margin of safety, as defined in the basis for any Technical Speci
,fication', is not. reduced'hecause operation of the llPCI system and drywell--
f torusfcontrol: system remains.within the requirements of Technical Speci-ifications, therefore,ethe margin of safety is notLreduced-.
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k # o Safety Evaluation 189-335; Technical Specification-Proposed l Change,:Section 3.6/4.6
,1
.. .. This change: adjusts lthe' pressure-temperature _ operating limita for-Quad Cities o _ Unit 1 and)2 reactor vessels by updating Figure 3.6-1 and make the Jimits' valid h through_16 effective full power years. .This is necessary to comply with' Reg.
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Guide 1.99. Revision 2-(NRC Generic letter 88-11).
' Removes the limitation that the reactor' vessel be vented unless the reactor
> . vessel' temperature is equal to or greater than the minimum reactor pressurization L . temperature curve (Figure 3.6-2, DPR-29, and Figure 3.6-1. DPR-30). Additionally, F 'these' figures will be removed from the Technical Specifications.
An administrative change to correct the reactor vessel speciman withdrawal-
' dates in table 4.6-2.
.1. The probability of an occurrence or the consequence of an accident, or-malfunction of-equipment important to safety as previously evaluated 6 g ;in the-Final Safety Analysis Report is not increased because the pressure-
~
temperature operating limits are adjusted to incorporate the initial fracture toughness conservatism present when the reactor vessel was new._ GE's analysis (NED0-21778-A) shows.that for a control rod drop
. accident transient in the conditions identified for venting, that no operator actions are needed to alter the vessel conditions. For water levels-as great 780 inches above the vessel bottom, a maximum vessel y pressure rise of 15.8 psi was calculated. The' venting requirement was j a result of a postulated pressure spike of sufficient magnitude that i would place the vessel in a condition that violates 10CFR50 Appendix G.
GE's analysis shows that this requirement was overly conservative and restrictive.Early withdrawal of the specimens simply provided irradiation E cffects at a lower fluence level. There remains sufficient vessel specimens-g; to_ support the requirements of Appendix H.
- 2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not j g
' created because the new pressure-temperature operating limits are merely
~
an update of the old limits, no physical changes are being made. The ;
removal of the venting requirement is only an adjustment to an overly !
restrictive limit which has been shown not to be needed (NEDO-21778-A). i The maximum pressure spike was calculated to be 15.8 psi and as a result f GE states that no operator action is needed to alter vessel conditions l such as opening the vessel head vent. No new or different kind of accident i is created as a result of removing the reactor vessel specimen early.
The. vessel specimen was_ subjected to a slightly lower fluence level but provides information on irradiation effects of the vessel material. t There are sufficient vessel specimens remaining in the vessel to satisfy _l the requirements of Appendix H. j s
- 3. The margin of safety, as defined in the basis for any Technical Speci- j fication, is not reduced because the new pressure-temperature operating !
p limits are actually restoring the margin of safety to a level similar l to when the reactor vessel was new and the fracture toughness slightly j greater. Removing the venting requirements still includes an adequate i l margin of safety as-shown by GE's analysis (NED0-21778-A). The calculated ;
maximum pressure rise as a result of a CRDA was 15.8 psi and thus, GE states no operation action to alter vessel conditions is needed. The margin of safety is not reduced by the early removal of the reactor vessel specimen. The specimen was used to determine the irradiation effects on the reactor vessel material. There are still enough specimens l remaining to support the requirements of Appendix H.
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Safety Evaluations (!89-432 and #89-439 Reactor Recirculation and Reactor Water Cleanup System Decontamination During the Unit 1 Refuel Outage decontamination of piping associated with the Reactor vessel was performed, the Reactor Water Cleanup Piping was performed with fuel in the vessel and the vessel head removed. The decontamination chemicals did not enter the vessel during this process.
The Recirculation Pump Suction and Disharge Piping was also decontaminated.
This was done with the fuel removed from the vessel. The vessel head was in place but not tensioned. Water icvel in the vessel was maintained below the core area of the vessel. The decontamination chemicals were flushed from the vessel prior to reloading fuel.
- 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased since metallurgy effects are minimal because the solvent corrosion rates are less than the original allowances. 304 stain 1 css steel coupons were placed in the decontamination flow path and analyzed upon completion of the project for assurance of the actual corrosion rates. Water purity effects are minimal because the reactor coolant were returned to a conductivity and a TOC level that is acceptable to station chemistry.
- 2. The possibility for an accident or malfunction of a different type than any prev Husly evaluated in the Final Safety Analysis Peport is not created 'oecause the ef fects of residual solvent in the system was determined to be negligible. Reactor Coolant is c1 caned and returned to a conductivity and a TOC level which is acceptable to the station chemistry steff. Station radiation protection procedures were followed throughout the decontamination. During resin transfer to the soJidifi-cation truck, the affected areas of the reactor building was evacuated.
Access into the drywell during the process was strictly controlled by station health physicists. The level of the solvent in the recircu-lation system risers and annulus was contir.uously monitored. Since SMAD has reviewed the material / solvent interface for materials within the core and has accepted the solvent for use, the consequences of a failure in the level controla causing a spill into the core are negligibic.
- 3. The margin of safety, as defined in the basis for any Technical Speci-fication is not reduced because the decontamination project was performed in accordance with the existing Technical Specifications.
The reactor was maintained in the shutdown or refuel rode with all interlocks in the shutdown or refuel position.
b>
'N Safety Evaluation #89-437 Process Control Program for CNSI Cement Solidifcation This provides the Process Control Program for CNSI to process LOMI decon solution on bead resin using Formula II.
- 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the solidi-fication of decon spent resins does not involve plant systems and will not increase the probability of an occurrence or the consequence of an accident or malfunction of equipment important to safety as previously evaluated in the FSAR.
- 2. The possibility for an accident or malfunction of a different type thwn
!' any previously evaluated in the Final Safety Analysis Report is not
[
created because this procedure does not contradict FSAR Section 9.
This procedure assures that the solidification is donc according to a pre-approved Process Control Program.
I
- 3. The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because this is in accordance with Tech Spec 6.9 and ensures this margin of safety is incorporated.
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k * , Safety Evaluation'#89-445 i (j)p\ "
<:e FSAR Correction. '
, so 'i "
Change FSAR-Table 5.2.5 to better describe power to elose 1601-21, 22, 23
,f24, 56 and 60 frca spring to air.-
l'. LThe probabilityfof_an occurrence or the consequence _of an accident.
cnf,p -s
, or. malfunction of equipmant important to safety;as.previously ovaluated. '
1 1n the Final Safety Analysis.Raport-is not increased because this safety :i evaluation is for a FSAR correction and does not involve any equipment. :
?. ' procedure : design function or operating method changes. *
+
g 2.- The' possibility for an accident or malfunction of a different' type than_ '
any previously evaluated in the FinallSafety Analysis ReportLis not j
-created because the probability of human error due to misinterpretation :
9- , -of the FSAR is reduced.
- 3. The margin of' safety, as defined in the basis for any Technical Speci-
'fication, is not reduced because Technical. Specifications are not affected.
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Reduce the number of temperature switches from 16 to 4-and change the trip i Lsetpoint from 185'T to 155'F on the Unit One RCIC and HPCI Turbine Area High-
. Temperature Isolation system. .
1.The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because reducing 1 the number of temperaure elements,will not degrade:the integrity of. ,
the leak detection system. Decreasing the trip level setting will reduce i
,' response time and maintain radiation releases within acceptable limits.' i
.Therefore, the probability of an occurrence or consequence of an accident ..-
is not increased.
2.. .The possibility for an accident or malfunction of=a different type-than .
any previously evaluated in the Final Safety Analysis Report is not !
created because the modified system will still maintain one-out-of-two !
taken twice trip logic and separation criteria for electrical power supplies. The system will still ensure isolation in the event of an l
< actual steam line break but should preclude spurious isolations due i to smallL1ocalized' steam leaks. Therefore, there is no possibility 1
- for,an-accident or malfunction created.
-3. The margin of safety, as defined in the basis for any Technical Speci- [
fication, is not reduced because this change requires a revision to ;
. Technical' Specifications. However. the change will not reduce the effectiveness of the steam leak detection system. The modified system '
should increase the reliability of RCIC and HPCI by reducing the.
probability of sporadic isolations. Therefore, the margin of safety 'l has not been reduced.
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f n R s M-4-1(2)-84-21A and B f Safety Evaluation #89-468
'HpCI and RCIC Area High Temperature l' Isolation System i e This modification involves decreasing the number of tempereture elements from 16 to_4 and reducing the trip level setting from 1200'F to 1170'F. The o current system consists of four groupa of switches at four different locations.
Each group of four switches at one location is arranged in a.one-out-of-two
! taken twice trip logic. This has resulted in spurious system isolations due f to minor steam leaks at the turbine bearings. The modified system will consist of two groups of switches at two different locations. The four switches will then be arranged in a one-cut-of-two taken twice trip logic. The trip level setting will be' reduced'to maintain system response time and limit radiation release in the event of-a steam line break.
!['
- 1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated l
in the Final Safety _-Analyais Report is not increased because an analysis
} ;.s performed by General Electric and a calculation performed by Impe11
!_ to-evaluate the HPCI and RCIC area temperature monitoring systems-and P -proposed modification. The calculations determined that reducing the number of-temperature elements will not degrade the integrity of the leak detection system. Decreasing the trip level setting vill reduce response time'and mr.intain radiation releases within acceptable limits.
L Therefore, the probability of an occurrence or consequence of an accident is not increased.
t
- 2. The possibility for an accident or malfunction of a different type than any:previously. evaluated in the Final Safety Analysis Report is not created because the modified system will still maintain one-out-of-two taken twice trip logic and separation criteria for electrical power supplies. The system will still ensure isolation in the event of_an actual steam line break but should preclude spurious isolations due to small localized steam leaks. Therefore, there is_no possibility for an accident or malfunction created.
- 3. The margin of safety, as defined in the basis for any Technical Spect-fication, is not reduced because the modification requires a change to Technical Specifications. However, the change.will not reduce the i effectiveness of the steam leak detection tystem. The modified system should increase the reliability of HpCI and RCIC by reducing the number
( of sporadic isolations. Therefore, the margin of safety has not been reduced.
YETY '
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i bJ' gm . Modifications M-4-1-84-027A, B, C and D w ',
s
+ Description General Electric. identified that a potential existing of a 'b' contact b bounce: problem;in their BGA relays during a seismic event. After review,'certain
'important to safety HGA relays were exchanged for HFA relays. In some cases 7
- the wiring was moved from an HFA to an HGA,to free up the HFA for use. No H circuit. logic was altered - function and operation of the system was unaffected.
p1 H . The modification covers the HPCI, RCIC and Core Spray systems.
b
- p. Evaluation-
.p s ~1. :The: probability of an occurrence or the consequence of an accident, (J of malfunction,of equipmer.t important to safety as previously evaluated zin;the Final Safety Analysis Report is not increased because the HFA type relays, which have a higher seismic rating than the HGA relays,
- ( will now be used in place of HGA relayo in safety circuits. Thus h., reliability is increased and the probability of a malfunction is
? reduced.
- 2. The possibility for an accident or malfunction of a different type p .than any previously evaluated in the Final Safety Analysis Report is
, not created because this is a one-for one exchange of the function
'a performed by. existing relays. Therefore, no new malfunction is created.
,- _ 3.- The margin of safety. as defined in the basis-for any Technical Speci-ffcation, is not reduced because since the seismic rating of the replacement HFA relay exceeds original equipment ratings, the margin
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,' Modification M-4-1-85-26 m
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c L Description x
h The existing General Electric CFD Diesel Generator differential current [
I. -protection relay was replaced with a new seismically qualified Westinghouse L
SA-1 type differential relay in order to satisfy OPEX 84-75S1. The new relay ;
is in the'same physical location (at the 4KV switchgear) r.s before. The relay !
continues to. provide a trip signal to the' lockout relay to disconnect an ;
internally faulted Diesel Generator from its 4KV switchgear bus. .
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Evaluation j i
T l' . .The probability.of an occurrence or the consequence of an accident, or. malfunction,of equipment important to safety as previously evaluated !
g in the Final Safety Analysis Report is not increased because existing ;
differential relays are being replaced with seismically qualified relays. -j therefore the' probability of.an occurrence or an accident, or malfunction !
of equipment important to safety as evaluated in FSAR is not increased. '
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2.' The possibility for an accident or malfunction of a different type than .
l L any previously evaluated in the Final Safety Analysis Report is not created ,
f because new relays have the identical system interfaces as the existing .
f ,
relays therefore the possibility for an accident or malfunction of a i different type than previously evaluated in the FEAR does no: exist.
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t f 3. .'The margin of safety, as defined in the basis for any Technical Speci- t
& fication, is not reduced because new relays will provide improved reliability during seismic.eventb. therefore the margin of safety as defined in the basis'of Quad Cities Technical. Specification is not reduced. The presently installed relays are non-seismic. ,
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i Modification M-4-1-87-074A and 74B o"- .
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L .These modifications replaced restricting orifice 1-3241-53A on the 'A' !
L and 1-3241-53B on the 'B' feedwater flush line with spectacle. flanges. The ;
spectacle flange consists of a blank plate and a large bore orifice. Blank r plates will be inctalled during normal operation. _The large bore orifice will t be used for flushing operations._ The original restrictive orifice was sized extremely small to-form a pressure barrier between the feedwater piping and the condenser since flushing was originally designed to be done using a feedwater ,
h pump.; However, the restrictive orifice did not allow enough flow to provide ;
' adequate flushing of the system. !
E>aluation i
~1. The probability _of an occurrence or the consequence of an accident. '
.or malfunction of. equipment important to safety as previously_ evaluated- .
in the Final Safety Analysis Report is not increased because the feed- !
water flush lines are not mentioned in Section 11 of the FSAR which .
deals with the feedwater system. Since the original conditions and '
assumptions made in the FSAR have not been changed, the probability of an occurrence or the consequence of an accident, or malfunction ;
of equipment important to safety as previously evaluated in the FSAR '
is not. increased. !
i
'2. The, possibility for an accident or malfunction of a different type
[g than any previously, evaluated in the Final Safety Analysis Report is '
L not created because this modification does not interfere with any safety-related equipment and would not fa11'outside any singic failure i event or design basis accident which has already been analyzed in the FSAR. !
- 3. The margin of_ safety, as defined in the basis for any Technical Speci- ';
fication, is not reduced because feedwater flush lines do not interact with any systems described in the Technical Specifications. Therefore, the margin of safety is not reduced.
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i Modification M-4-0-89-064 Safety Evaluation #89-467 l install RACS Video Capture System Modification l
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.A RACS (Redundant Access Control System) Video Capture System will be installed to further enhance the security perimeter intrusion assessment by CCTV (Closed Circuit Television). The potential exists for a CAS (Central Alarm Station)
,s console operator to miss an intruder on CCTV, when an intrusion alarm comes up. l This RACS Video Capture System can freeze a video frame inside one second of the
-intrusion. The video frame, through a pair of dedicated monitors, provides the i CAS console operator the reaction time capability of the existing electronic devices.
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[ i b A video printer will provide =a hard copy of that video frame to assess and-document I any human intrusions. .
b The RACS Video Capture System consist of six devices (two monitors, one video printer, two video digitalizer bontds, one video switch board) and two manual !
E switches (for transferring the communications lines). The first three devices '
L mentioned operate at 120V and will be connected to the existing security system UPS.
b- 1. 'the probability.of an occurrence or the consequence of an accident. I or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the reliability p: ~ of.the CCTV and the~ Perimeter Intrusion Detection System will be enhanced !
L by the addition of.this new RACS Video Capture System. However this l
[, wou3d have no bearing on the probability or consequence of an accident or malfunction of equipment important to safety, since analyses take Q. no credit for this security system.
- 2. . The possibility for an accident or malfunction of a different type than .i any previously evaluated in the Final Safety Analysis Report is not .J created because this modification does not alter the description of i N- .any equipment or systems imp,rtant to safety as previously evaluated a
'in the FSAR/UFSAR. Installation of the new PACS VCS involves non-safety- :
related equipment which will be located remote from any safety-related '
system.
y 3. The. margin of safety, as defined in the basis for any Technical Speci- ,
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fication, is not reduced because this modification does not alter or affect any equipment described in the Technical Specification. Therefore, the margin-of safety will not be reduced.
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i y ag Modificatioti M-4-1(2)-89-152 h4H , Safety Evaluations #89-519 and #89-520 p
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[ Description Fm
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This modification is being installedDas a corrective action'per Potential H s Significant. Event Report PSE-89-006, titled "New Fuel' Bundle Drop While in Fuel IL . Pool". .-The PSE' occurred on September 21, 1989 at Quad Cities Unit 1. See PSE -
89-006 for' details 2 L .The modification will install an additional electrical interlock that will
$(;O; prevent raising the hoist on the fuel moving eschine while the hoist is loaded a unless the grapple is fully closed and in the engage position.
The modification will be contained in the G.E. fuel moving panel located-.
,on the refuel bridge .
n ' Evaluation' N .
[ 1.--The probability of an occurrence or the consequence of an accident, 9 or malfunction of. equipment important to safety as previously evaluated-L '
in'the Final Safety Analysis' Report is not increased because the modi-l , fication will add an additional feature to the interlock system to
[. , , enhance the safe movement of fuel, h 2. The possibility for an accident or malfunction of a different_ type s , ;, - t .than any previously evaluated in the-Final Safety Analysis Report is not created ~because this modification will add an additional interlock ~
protection to an evaluation condition.
e' L3. 'The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because this modification will increase the L,
margin of. safety while moving fuel.
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