Similar Documents at Zion |
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217J4411999-10-18018 October 1999 Forwards Exemption from Certain Requirements of 10CFR73.55 for Zion Nuclear Power Station,Units 1 & 2 in Response to Application Dtd 990730,to Allow Util to Discontinue Specific Aspects of Security Plan & SER 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 ML20216G9541999-09-27027 September 1999 Forwards Environ Assessment & Finding of No Significant Impact Re 990730 Application for Exemption.Exemption Would Allow Zion Nuclear Power Station to Discontinue Specific Aspects of Security Plan A99025, Forwards Response to NRC 990820 RAI Re Proposed Defueled TS for Zion Station1999-09-15015 September 1999 Forwards Response to NRC 990820 RAI Re Proposed Defueled TS for Zion Station ML20211L1551999-09-0101 September 1999 Forwards Insp Repts 50-295/99-03 & 50-304/99-03 on 990608-0812 at Zion 1 & 2 Reactor Facilities.No Violations Noted.Activities in Areas of Facility Mgt & Control, Decommissioning Support,Sf & Radiological Safety Examined ML20211J9701999-08-31031 August 1999 Discusses Request for Approval of Defueled Station Emergency Plan & Exemption from Certain Requirements of 10CFR50.47, Emergency Plans, for Plant,Units 1 & 2.Exemption & Safety Evaluation Encl ML20211J0311999-08-30030 August 1999 Agrees with Content in NRC 990810 Memo Re Plant Spent Fuel Pool.Memo Contains Accurate Summation of Info That Util Communicated to NRC During 990713 Telcon 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) ML20211G8191999-08-26026 August 1999 Forwards Amend 6 to Fire Protection Rept for Zion Nuclear Power Station,Units 1 & 2, IAW 10CFR50.71(e).Attachment 1 Contains Summary of Changes ML20211F5971999-08-20020 August 1999 Submits Request for Addl Info Re Proposed Defueled TSs for Plant,Per NRC Bulletin 94-001, Potential Fuel Pool Draindown Caused by Inadequate Maint Practices at Dresden Unit 1 ML20211C7401999-08-19019 August 1999 Informs That NRC Plans to Stop Using Office Complex at Zion Nuclear Generating Station Which Commonwealth Edison Provided for NRC Resident Inspectors Under 10CFR50.70(b)(1). NRC-owned Equipment to Be Moved on or About 991001 ML20211J4421999-08-13013 August 1999 Forwards, EA & Finding of No Significant Impact Re Application for Exemption ,as Suppl by 990708 & 19 Ltrs.Proposed Exemption Would Modify Emergency Response Plan Requirements as Listed 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 ML20210J0861999-07-30030 July 1999 Requests NRC Approval of Exemption from Certain Requirements of 10CFR73, Physical Protection of Plants & Matls, by 990921.Exemption Is Needed to Support Scheduled 991001 Implementation of Plant Sf Ni.Encl Withheld,Per 10CFR73.21 ML20210C4201999-07-20020 July 1999 Discusses Investigation Rept 3-98-017 Into Info Reported to Comed on 980224,informing NRC That N Everson Brought Personal Handgun Into Personnel Search Area at Plant.Nrc Determined That Violation Occurred.Nov Encl ML20210C3641999-07-20020 July 1999 Discusses OI Rept 3-98-017 Completed on 980224 Re Contract Security Officer That Inadvertently Brought Personal Handgun Into Search Area at Plant & Forward NOV ML20210C2561999-07-19019 July 1999 Forwards Rev 1 to Calculation 22S-B-110X-0068, Determination of Heat Load for Assemblies F50D & Z33A, as Replacement Pages to Rev 0.Calculation Submitted Re Request for Approval of Proposed Plant Defueled Station EP 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) ML20209E7391999-07-0808 July 1999 Forwards Proposed Revised Znps Defueled Station Emergency Plan & Supporting Calculations.Revs to Proposed Defueled Station Emergency Plan Have Been Reviewed & Approved IAW Comed QA Program 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 ML20196D3301999-06-18018 June 1999 Forwards Insp Repts 50-295/99-02 & 50-304/99-02 on 990121-0608.No Violations or Deviations Were Identified. Weakness Noted in Self Assessment/Corrective Action Program in Which Series of Personnel Errors Not Recorded in PIF Sys A99016, Forwards Revised Monthly Operating Rept for Apr 1999 for Zion Nuclear Power Station,Units 1 & 2.Data Was Revised to Reflect Change in Daylight Savings Time1999-06-15015 June 1999 Forwards Revised Monthly Operating Rept for Apr 1999 for Zion Nuclear Power Station,Units 1 & 2.Data Was Revised to Reflect Change in Daylight Savings Time ML20207H6121999-06-10010 June 1999 Provides Synopsis of OI Rept 3-1998-012 & Summary of Relevant Facts Re Violation Involving Employment Discrimination Against SRO at Zion Station.Enforcement Conference Scheduled for 990707 ML20195E3451999-06-0707 June 1999 Forwards 3.5 Inch Computer Diskette Containing Revised File Format for Annual Dose Rept for 1998,per 990520 Telcon Request from Nrc.Each Station Data Is Preceded by Header Record,Which Provides Info Necessary to Identify Data ML20195D6351999-06-0404 June 1999 Notifies NRC of Actions That Has Been Taken in Accordance with 10CFR26, Fitness for Duty Programs A99018, Responds to NRC Re Violations Noted in IR 3-98- 017.Corrective Actions:Three Security Guards Unescorted Access to All Comm Ed Nuclear Stations Put on Temporary Hold1999-05-27027 May 1999 Responds to NRC Re Violations Noted in IR 3-98- 017.Corrective Actions:Three Security Guards Unescorted Access to All Comm Ed Nuclear Stations Put on Temporary Hold 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 ML20207B5991999-05-14014 May 1999 Informs That NRC Ofc of NRR Reorganized,Effective 990328.As Part of Reorganization,Div of Licensing Project Mgt Was Created.Reorganization Chart Encl ML20206D8041999-04-30030 April 1999 Forwards Zion Nuclear Power Station Annual Radiological Environ Operating Rept & Listing of Commitments 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 ML20206B9371999-04-27027 April 1999 Discusses Investigation Rept 3-98-17 Re Info Reported by Licensee on 980224 That Firearm Detected During Routine Search of Contractor Employee & Failure to Rept Incident to Supervisor.Oi Rept Synopsis Encl ML20206B4051999-04-23023 April 1999 Ack Receipt of Encl FEMA Ltr,Transmitting FEMA Evaluation Rept for 980819 Annual Medical Drill for Zion Nuclear Power Station.No Deficiencies Noted.Two Areas Requiring Corrective Action Noted During Drill 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 ML20205S5811999-04-16016 April 1999 Responds to Which Urged Full Public Hearing to Held on Concerns Raised by R Robarge & E Dienethal Re Zion Npp.Commission Issued 990302 Memo & Order Affirming ASLB Decision to Deny Petition to Intervene A99014, Provides Description of Actions Comed Has Taken to Reinforce Safety Conscious Work Environ at Zion Nuclear Power Station1999-04-15015 April 1999 Provides Description of Actions Comed Has Taken to Reinforce Safety Conscious Work Environ at Zion Nuclear Power Station ML20205Q9971999-04-13013 April 1999 Requests NRC Approval of Plant Defueled Station Emergency Plan in Accordance with 10CFR50.54(q).Ltrs of Agreement Between Util & Offsite Agencies Who Have Agreed to Provide Assistance to Plant Encl.Calculations Also Encl IR 05000295/19980021999-04-0909 April 1999 Discusses Insp Repts 50-295/98-02,50-304/98-02 & OI Rept 3-98-015 Completed on 981112 & Forwards Nov.Violation Involves Zion Supervisor Who Willingly Failed to Protect Unattended SGI from Unauthorized Disclosure 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) ML20206B4331999-03-26026 March 1999 Forwards Rept for Annual Medical Drill Conducted for Zion Nuclear Power Station on 980819.Two Areas Requiring Corrective Action Identified During Drill 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 A99011, Responds to NRC Re Violations Noted Insp Repts 50-295/98-02 & 50-304/98-02.Corrective Actions:Uncontrolled Safeguards Documents Were Immediately Taken Into Possession & Controlled by Plant Security Supervisor1999-03-18018 March 1999 Responds to NRC Re Violations Noted Insp Repts 50-295/98-02 & 50-304/98-02.Corrective Actions:Uncontrolled Safeguards Documents Were Immediately Taken Into Possession & Controlled by Plant Security Supervisor ML17191B2671999-03-12012 March 1999 Forwards Revs to Dresden,Quad Cities,Zion,Lasalle,Byron & Braidwood Station Odcm,Current as of 981231.ODCM Manual & Summary of Changes,Included A99009, Forwards Znps Annual Occupational Exposure Rept, for Exposures Occurring in 1998.Listing of Commitments Contained in Ltr Provided as Attachment B1999-03-0101 March 1999 Forwards Znps Annual Occupational Exposure Rept, for Exposures Occurring in 1998.Listing of Commitments Contained in Ltr Provided as Attachment B 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) ML17191B2331999-02-0909 February 1999 Discusses Ceco 950105 Request for Review & Approval of Four near-site Emergency Operations Facilities (Eofs) & Corporate EOF Into One Single,Central Eof.Concludes That Central EOF Meets All of Functional & Physical Requirements for EOFs ML20202J4381999-02-0303 February 1999 Forwards Zion Fuel Evaluation Info,Per Discussion 1999-09-30
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217H4661999-10-18018 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 A99025, Forwards Response to NRC 990820 RAI Re Proposed Defueled TS for Zion Station1999-09-15015 September 1999 Forwards Response to NRC 990820 RAI Re Proposed Defueled TS for Zion Station ML20211J0311999-08-30030 August 1999 Agrees with Content in NRC 990810 Memo Re Plant Spent Fuel Pool.Memo Contains Accurate Summation of Info That Util Communicated to NRC During 990713 Telcon 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) ML20211G8191999-08-26026 August 1999 Forwards Amend 6 to Fire Protection Rept for Zion Nuclear Power Station,Units 1 & 2, IAW 10CFR50.71(e).Attachment 1 Contains Summary of Changes 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 ML20210J0861999-07-30030 July 1999 Requests NRC Approval of Exemption from Certain Requirements of 10CFR73, Physical Protection of Plants & Matls, by 990921.Exemption Is Needed to Support Scheduled 991001 Implementation of Plant Sf Ni.Encl Withheld,Per 10CFR73.21 ML20210C2561999-07-19019 July 1999 Forwards Rev 1 to Calculation 22S-B-110X-0068, Determination of Heat Load for Assemblies F50D & Z33A, as Replacement Pages to Rev 0.Calculation Submitted Re Request for Approval of Proposed Plant Defueled Station EP ML20209E7391999-07-0808 July 1999 Forwards Proposed Revised Znps Defueled Station Emergency Plan & Supporting Calculations.Revs to Proposed Defueled Station Emergency Plan Have Been Reviewed & Approved IAW Comed QA Program 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 A99016, Forwards Revised Monthly Operating Rept for Apr 1999 for Zion Nuclear Power Station,Units 1 & 2.Data Was Revised to Reflect Change in Daylight Savings Time1999-06-15015 June 1999 Forwards Revised Monthly Operating Rept for Apr 1999 for Zion Nuclear Power Station,Units 1 & 2.Data Was Revised to Reflect Change in Daylight Savings Time ML20195E3451999-06-0707 June 1999 Forwards 3.5 Inch Computer Diskette Containing Revised File Format for Annual Dose Rept for 1998,per 990520 Telcon Request from Nrc.Each Station Data Is Preceded by Header Record,Which Provides Info Necessary to Identify Data ML20195D6351999-06-0404 June 1999 Notifies NRC of Actions That Has Been Taken in Accordance with 10CFR26, Fitness for Duty Programs A99018, Responds to NRC Re Violations Noted in IR 3-98- 017.Corrective Actions:Three Security Guards Unescorted Access to All Comm Ed Nuclear Stations Put on Temporary Hold1999-05-27027 May 1999 Responds to NRC Re Violations Noted in IR 3-98- 017.Corrective Actions:Three Security Guards Unescorted Access to All Comm Ed Nuclear Stations Put on Temporary Hold 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 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 ML20206D8041999-04-30030 April 1999 Forwards Zion Nuclear Power Station Annual Radiological Environ Operating Rept & Listing of Commitments 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 A99014, Provides Description of Actions Comed Has Taken to Reinforce Safety Conscious Work Environ at Zion Nuclear Power Station1999-04-15015 April 1999 Provides Description of Actions Comed Has Taken to Reinforce Safety Conscious Work Environ at Zion Nuclear Power Station ML20205Q9971999-04-13013 April 1999 Requests NRC Approval of Plant Defueled Station Emergency Plan in Accordance with 10CFR50.54(q).Ltrs of Agreement Between Util & Offsite Agencies Who Have Agreed to Provide Assistance to Plant Encl.Calculations Also Encl 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) ML20206B4331999-03-26026 March 1999 Forwards Rept for Annual Medical Drill Conducted for Zion Nuclear Power Station on 980819.Two Areas Requiring Corrective Action Identified During Drill 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 A99011, Responds to NRC Re Violations Noted Insp Repts 50-295/98-02 & 50-304/98-02.Corrective Actions:Uncontrolled Safeguards Documents Were Immediately Taken Into Possession & Controlled by Plant Security Supervisor1999-03-18018 March 1999 Responds to NRC Re Violations Noted Insp Repts 50-295/98-02 & 50-304/98-02.Corrective Actions:Uncontrolled Safeguards Documents Were Immediately Taken Into Possession & Controlled by Plant Security Supervisor ML17191B2671999-03-12012 March 1999 Forwards Revs to Dresden,Quad Cities,Zion,Lasalle,Byron & Braidwood Station Odcm,Current as of 981231.ODCM Manual & Summary of Changes,Included A99009, Forwards Znps Annual Occupational Exposure Rept, for Exposures Occurring in 1998.Listing of Commitments Contained in Ltr Provided as Attachment B1999-03-0101 March 1999 Forwards Znps Annual Occupational Exposure Rept, for Exposures Occurring in 1998.Listing of Commitments Contained in Ltr Provided as Attachment B 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) ML20202J4381999-02-0303 February 1999 Forwards Zion Fuel Evaluation Info,Per Discussion 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 A99001, Forwards Zion 1998 Annual Rept of Revised Regulatory Commitments1999-01-19019 January 1999 Forwards Zion 1998 Annual Rept of Revised Regulatory Commitments A98056, Forwards List of Revised Various Commitments Originally Contained in Util to Nrc.Changes in Commitments Were Satisfactorily Evaluated Using NEI Commitment Change Methodology1998-12-11011 December 1998 Forwards List of Revised Various Commitments Originally Contained in Util to Nrc.Changes in Commitments Were Satisfactorily Evaluated Using NEI Commitment Change Methodology ML20198C2501998-12-11011 December 1998 Forwards Rev 45 to Zion Nuclear Power Station Security Plan, Per Requirements of 10CFR50.4(b)(4).Rev Includes Listed Changes.Licensee Has Concluded That Changes Do Not Decrease Safeguards Effectiveness of Plan.Encl Withheld ML20196B5871998-11-20020 November 1998 Requests That Svc List for All NRC Correspondence Re Any of Six Comed Nuclear Stations Be Modified Per Attached List.All Other Names Previously Listed Should Be Removed A98051, Forwards List of Revised Regulatory Commitments Which Were Evaluated as Satisfactory Using NEI Commitment Change Methodology1998-11-20020 November 1998 Forwards List of Revised Regulatory Commitments Which Were Evaluated as Satisfactory Using NEI Commitment Change Methodology A98055, Forwards List of Revised Regulatory Commitments Which Were Evaluated as Satisfactory Using NEC Commitment Change Methodology1998-11-20020 November 1998 Forwards List of Revised Regulatory Commitments Which Were Evaluated as Satisfactory Using NEC Commitment Change Methodology ML20195E6451998-11-12012 November 1998 Provides Results of drive-in Drill Conducted on 981007,as Well as Augmentation Phone Drill Conducted on 980917 A98059, Forwards Page Change Instruction for Aug 1998 DSAR for Zion Nuclear Power Station,Units 1 & 2.Instruction Was Omitted from 981110 DSAR Due to Administrative Error1998-11-11011 November 1998 Forwards Page Change Instruction for Aug 1998 DSAR for Zion Nuclear Power Station,Units 1 & 2.Instruction Was Omitted from 981110 DSAR Due to Administrative Error A98053, Forwards Change Instructions for Update to Zion UFSAR, Rewritten to Reflect Permanent Defueled Condition of Plant. & DSAR,1998-11-10010 November 1998 Forwards Change Instructions for Update to Zion UFSAR, Rewritten to Reflect Permanent Defueled Condition of Plant. & DSAR, ML20155D2701998-10-27027 October 1998 Forwards Changed Pages from 980423 Submittal Providing Addl Info Marked with Revision Bars & Revised Pages to QA Topical Rept Section 18,for Review ML20198C7401998-10-21021 October 1998 Informs That Encl Correspondence Was Received from Constituent,C Paxton Re Request for Investigation as to Documentation Also Encl.Without Encl ML20154J4951998-10-0707 October 1998 Forwards Revised Security Plans for CE Listed Nuclear Power Stations,Per 10CFR50.4(b)(4).Changes Do Not Decrease Effectiveness of Station Security Plans.Encl Withheld A98050, Forwards Zion Station Special Rept Number 30412398002SR.Rept for Reporting Period of Jan 1996-Dec 1997.Attachment B Provides Listing of Commitments Contained in Submittal1998-10-0505 October 1998 Forwards Zion Station Special Rept Number 30412398002SR.Rept for Reporting Period of Jan 1996-Dec 1997.Attachment B Provides Listing of Commitments Contained in Submittal ML20151Y5101998-09-11011 September 1998 Provides Results of drive-in-drill Conducted on 980804 & Augmentation Phone Drills Conducted Between 980601 & 0831 A98047, Forwards Revised Scope of One Commitment Contained in Response to NOV1998-08-28028 August 1998 Forwards Revised Scope of One Commitment Contained in Response to NOV ML20238F7571998-08-28028 August 1998 Forwards fitness-for-duty Program Performance Data for Each of Util Nuclear Power Stations for Six Month Period Ending 980630 05000295/LER-1997-022-01, Forwards LER 97-022-01,IAW 10CFR50.73(a)(2)(v).Commitments Made within Ltr,Encl1998-08-28028 August 1998 Forwards LER 97-022-01,IAW 10CFR50.73(a)(2)(v).Commitments Made within Ltr,Encl 1999-09-30
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L0341990-09-20020 September 1990 Forwards Addl Station Blackout Info Re Diesel Loading & Alternate Ac Analysis,Per 900404 Meeting.Info Will Be Helpful When Evaluating Unit O Diesel Generator cross-tie Mod ML20059G1611990-09-0606 September 1990 Submits Supplemental Response to Generic Ltr 89-06, Certifying That SPDS Satisfies or Will Be Modified to Satisfy Requirements of Suppl 1 to NUREG-0737.Classroom Instruction on SPDS Incorporated Into Program ML20059G1751990-08-31031 August 1990 Provides Results of Util Reevaluation of Response to NRC Bulletin 85-003.Util Believes That Previous Testing Results & Subsequent Submittals Provide Assurance of Continued Operability of Sys motor-operated Valves ML20028G8741990-08-26026 August 1990 Forwards Supplemental Info to Support 900523 Application for Amends to Licenses DPR-39 & DPR-48 Re Generic Ltr 83-37 ML20028G8721990-08-24024 August 1990 Lists Commitments to Provide Backup Water Source to Charging Oil Coolers,Per NUREG-1150 ML20059D8271990-08-23023 August 1990 Provides Written Communication Re Util Intentions & Commitments,Per 900808 Telcon W/M Wallace.Standing Order 90-24 Issued to Provide Administrative Controls Beyond Tech Spec Requirements to Assure Svc Water Pump Availability ML20059C0781990-08-22022 August 1990 Submits Supplemental Info to Util 900412 Response to Notice of Violation in Insp Repts 50-295/89-40 & 50-304/89-36.Info Regards Requalification program,180 H of Required Training Through SAT Process & Policy for Remediation of Individual ML20056B4331990-08-20020 August 1990 Forwards Errata,To Radioactive Effluent Rept for Jan-June 1990 & Gaseous Effluent Errata Sheet ML20058P7791990-08-13013 August 1990 Provides Supplemental Response to NRC Bulletin 89-001 Re Mechanical Plugs Supplied by Westinghouse,Per .All Mechanical Plugs from Heats NX-3513 & NX-4523 Removed from Hot Legs of Plant Steam Generators ML20058N0651990-08-0707 August 1990 Responds to NRC Bulletin 88-009 Re Incore Neutron Monitoring Sys Thimble Tube Integrity.Thimble Tube Insps Will Be Performed at Every Refueling Outage Until Sufficient Data Accumulated on Thimble Tube Wear ML20056A0271990-07-30030 July 1990 Provides Addl Info Re Resolution of Number of Older Licensing Issues,Including Containment Isolation Valve Position Indication,Per NRC ML20056A0741990-07-27027 July 1990 Forwards Justification for Rev to Installation Schedules for Two Commitments Re Safeguards Inadequacies & Program Concerns.Encl Withheld ML20055J1191990-07-26026 July 1990 Responds to Generic Ltr 90-04, Request for Info Re Status of Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions. Current Status of Issues Encl ML20055J2231990-07-26026 July 1990 Requests Withdrawal of Proposed Amend Re Deletion of Confirmatory Order ML20055H7631990-07-25025 July 1990 Forwards Financial Info Re Decommissioning of Plants ML20055H0871990-07-20020 July 1990 Provides Status Rept on Molded Case Circuit Breakers Replacement,Per NRC Bulletin 88-010.Traceable Replacement Breakers for Units 1 & 2 Being Procured safety-related & Meet Criteria of Bulletin Action 7 ML20044A9511990-07-10010 July 1990 Advises of Westinghouse Reassessment of Portion of Analysis to Address Issue of Rapidly Propagating Fatigue Cracks in Steam Generator Tubes Per Bulletin 88-002 ML20058N6821990-07-0606 July 1990 Suppls 900626 Response to CAL-RIII-90-011 Re Torque Switch Settings,Per NRC Bulletin 85-003.Program to Address Generic Ltr 89-10 Which Expands Scope of Bulletin 85-003 Testing to safety-related motor-operated Valves,Being Developed ML17202L2921990-06-28028 June 1990 Provides Update on Alternative Schedule for Generic Ltr 89-10 Re safety-related motor-operated Valve Testing & Surveillance ML20055D5291990-06-27027 June 1990 Forwards Table Outlining Status of Significant Events Re PWR Reload Safety Analysis Methods for Plant Cycle 13.Requests That Transient Analysis Topical Rept Be Reviewed & SERs Issued Prior to 910830 ML20055D5231990-06-26026 June 1990 Responds to Confirmatory Action Ltr CAL-RIII-90-11 Re Actions Required to Confirm Valve Operability for Unit 1,per IE Bulletin 85-003 ML20055D8461990-06-24024 June 1990 Forwards WCAP-10962,Rev 1, Zion Units 1 & 2 Reactor Vessel Fluence & Ref Temp for PTS Evaluations ML20043H3591990-06-19019 June 1990 Provides Supplemental Response to NRC Bulletin 88-008 & Suppls 1 & 2.Repts Use Inconsistent Terminology When Describing Lines 1/2RC005,1/2RC038,1/2RC071 & 1/2RC079 ML20055D8391990-06-18018 June 1990 Provides Supplemental Info in Response to Generic Ltr 88-17, Loss of Dhr. Permanently Installed Level Transmitter, Independent from Refueling Vessel Level Recorder & Led Readout Will Be Utilized ML20043D3171990-06-0101 June 1990 Forwards Rev 22 to Security Plan.Rev Withheld (Ref 10CFR73.21) ML20043F0421990-05-29029 May 1990 Notifies of Change to 770527 Commitment Re 125 Volt Dc Mimic Panels for Plant ML20043C9161990-05-21021 May 1990 Forwards Response to 880828 Request for Addl Info Re Control Room Ventilation Sys.Appropriate Pages from Fire Protection Plan Encl.W/Four Oversize Drawings ML20043C2751990-05-14014 May 1990 Provides Supplemental Response to Bulletin 89-002 Re Insp of Anchor/Darling Check Valves.Insp of All Check Valves Completed.Five Failures Discovered During Insp.Approval of Proposed Schedule for Completion of Valve Insp Requested ML20043D2201990-05-10010 May 1990 Forwards Request for Addl Info Re Steam Generator Girth Welds.Results of Hardness Tests on Girth Welds & Surrounding Areas & Results of Boat Sample Analyses Will Be Submitted W/Outage Rept ML20043F0341990-05-0707 May 1990 Advises That Use of Incorrect Info on Operation of Plant Resulted in Core Damage Probability Higher than Actual. Requests New Info to Reevaluate Conclusions of Rept ML20042G9701990-05-0303 May 1990 Forwards Corrected Monthly Operating Rept for Mar 1990 for Zion Unit 1.Original Rept Did Not Reflect Correct Forced Outage Duration for Time Period ML20042F3321990-05-0101 May 1990 Forwards Detailed Description of Facility Steam Generator Girth Weld Insp Results.Licensee Concludes After Review of Flaw Evaluation That Indications Acceptable ML20055J4471990-04-26026 April 1990 Responds to NRC Re Violations Noted in Insp Repts 50-295/90-03 & 50-304/90-03.Corrective Actions:Training Rev Request Initiated & Personnel Counseled ML20042F1211990-04-19019 April 1990 Provides List of Tubes Plugged or Sleeved in Steam Generator as Result of Most Recent Insp,Per Tech Spec 4.3.1.B.5.A. Totals Do No Include Replugging of 27 Hot Leg Tube Ends Due to Westinghouse Plug Removals for NRC Bulletin 89-001 ML20034A3431990-04-12012 April 1990 Responds to Violations Noted in Insp Repts 50-295/89-40 & 50-304/89-36.Corrective Actions:Importance of Adhering to Requirements of 10CFR55.53 Discussed W/Personnel Involved & Refueling Procedures Revised ML20034B6901990-04-12012 April 1990 Forwards Rev 58 to CE-1-A, QA Program for Nuclear Generating Stations. Rev Being Made to Reflect Correct Frequency of Cumulative Deviation Reporting & to Clarify Applicability of Other ANSI N45.2 Stds ML20034B5021990-04-11011 April 1990 Responds to NRC Re Violations Noted in Insp Repts 50-295/89-37 & 50-304/89-33.Corrective Action:Future Solidifications of Ethylene Glycol Solutions by Us Ecology Will Be Performed Using Revised Process Control Program ML20034A3371990-04-11011 April 1990 Forwards Analysis of Capsule Y Commonwealth Edison Co,Zion Nuclear Plant Unit 1,Reactor Vessel Matl Surveillance Program. Analysis Summary Listed ML20033H1451990-04-0202 April 1990 Informs That Amend Request Identifying Recirculation Phase Action Requirements Will Be Submitted by 900501 ML20012D7871990-03-21021 March 1990 Forwards Recorded Annual Whole Body External Exposures for CY89 for Listed Plants ML20012D2291990-03-15015 March 1990 Forwards Graded Quizzes of Seventh Licensed Operator Retraining Program Administered in 1990,per Confirmatory Action Ltr CAL-RIII-89-21 ML20012C5931990-03-13013 March 1990 Forwards Response to Request for Addl Info Re Degraded Tube R1C55 in Steam Generator A.Athos 3-D T&H Analysis Used to Obtain Calculated Velocities for Same Test Conditions ML20012B6671990-03-0707 March 1990 Requests Extension of Temporary Waiver of Compliance for Addl 144 H to Permit Reactor to Remain in Hot Shutdown While Troubleshooting/Repairing Diesel Generator O & Performing Special Leak Rate Testing ML20012C3721990-03-0707 March 1990 Confirms Temporary Waiver of Compliance Extension Request Made on 900306 for O Diesel Generator Operability Requirements.Request for Extension of Time & Course of Action Approved ML20033E8461990-03-0202 March 1990 Documents Basis for Request for Temporary Waiver of Compliance W/Diesel Generator Operability Requirements. Results of Investigation Will Be Discussed W/Region III Personnel If Diesel Generator Returned to Operable Status ML20012B0321990-03-0202 March 1990 Responds to 900131 Notice of Violation & Proposed Imposition of Civil Penalty in Amount of $100,000.Corrective Actions: Task Force Initiated & Nuclear Engineering Dept Developed Training Course for Preparation of Evaluations for Mods ML20012A2791990-02-27027 February 1990 Forwards Proprietary & Nonproprietary Rev 0 to NFSR-0069, Transient Analysis Envelope for Zion Units 1 & 2. Proprietary Rept Withheld (Ref 10CFR2.790) ML20012A7241990-02-27027 February 1990 Withdraws 891031 & 881208 Tech Spec Amends.Util Will Be Making New Requests for Tech Spec Amends in Future Utilizing New Guidelines to Assure That Submittals Meet Improved Stds ML20006G2081990-02-23023 February 1990 Forwards Response to Safeguards Insp Repts 50-295/89-41 & 50-304/89-37 on 900102-05.Encl Withheld (Ref 10CFR73.21) ML20055C2921990-02-21021 February 1990 Forwards Errata Data for Radioactive Effluent Rept for Jan- June 1989 & Solid Radwaste Rept for Jul-Dec 1988 1990-09-06
[Table view] |
Text
--
N Commonwealth Edison O
~.)One First N:tioritt Plaza. Chictgo. liknois Addrzss Reply to. Post Offica Box 767 Chicago, filinois 60690 January 9, 1986 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Conunission Washington, DC 20555
Subject:
Zion Nuclear Power Station Units 1 and 2 NUREG-0737, Item II.D.1 NRC Docket Nos. 50-295 and 50-304 References (a):
February 19, 1985 letter from S. A. Varga to D. L. Farrar.
(b):
June 18, 1985 letter from P. C. LeBlond to H. R. Denton.
Dear Mr. Denton:
Reference (a) contained 14 questions concerning Zion's pressurizer safety and relief valves.
Reference (b) transmitted Commonwealth Edison Company's response to question numbers 1 through 11 and 14.
The remaining responses to questions 12 and 13 are enclosed with this submittal. These responses were prepared for Commonwealth Edison by the engineering firm of O'Donnell and Associates, Inc. (ODAI).
If you have any further questions regarding this matter, please contact this office.
Very truly yours, P. C. LeBlond Nuclear Licensing Administrator
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ODAI RESPONSE TO NRC QUESTIONS RELATED TO THE THERMAL l
HYDRAULIC ANALYSIS OF THE INLET AND DISCHARGE PIPING MRC QUESTION 12 The submittal states that a thermal hydraulic analysis of the safety / relief l
valve piping systes has been conducted, but does not present details of the ar.alysis.
To allow for a complete evaluation of the methods used and the results obtained from the thermal hydraulic analysis, provide a discussion on the thermal hydraulic analysis that contain at least the following informations a.
Evidence that the analysis was performed on the fluid transient cases producing the maximum loading on the safety /PORY piping system.
The cases should bound all steam, steam to water, and water flow transient conditions for the safety and PORY valves.
ODAI RESPONSE A generic analysis of the valve inlet fluid conditions for Westinghouse f
plants is given in References 1 and 2.
These studies clearly indicate that the most severe rate of pressurization and the highest pressure result from the locked rotor and loss-of-load events, respectively.
The possibility of liquid being vented through the valves for a feedvater line break, for an
[
extended high-pressure injection event, or for a cold over pressurization transient event are discussed on a generic basis for Westirghouse plants in References 1 and 2.
A Zion plant-unique probabilistic assessment of the possibility of the flow of liquid through the valves was performed in Reference 3.
This study determined that the total probability of liquid flow through the valves for these events is on the order of 1 x 10".
These analyses considered single active falure and single operator error in the determination of the event probability.
The discharge of liquid from safety and relief valves in the Zion plant has been shown to be an entremely j
unlikely event.
The estimated frequencies are based upon conservative date end assumptions,
.A they are sufficiently lov that even order-of-sagnitude errors would not.ffect the qualitative conclusion.
The question of whether the pressurizer liquid level increases because of spray actuation (therefore representing some potential for safety and relief valve liquid discherge) was considered from a qualitative standpoint.
The perspective is that it is extremely unlikely that any spray-induced level increase would be sufficient to actually result in liquid discharge through the safety or relief valves.
The effect of the sprays is to condense steaa in the pressurizer and to thereby, in the majority of cases analyzed, curtail the overpressure transient before the safety or relief valve actuation preneure is recognized.
For any resetning cases in which safety t
T 2
and relief valve actuation cannot be entirely ruled out, the liquid level contribution of the pressurizer sprays is not expected to be severe enough to produce liquid discharge through the safety and relief valves.
It may be desirable to confirm this assessment by a quantitative study; however, it is not anticipated that the renults of this study would reveal any spray-induced liquid discharge scenarios that are more likely to occur than those already analyzed in Reference 3.
The analyses of References 1, 2 and 3 serve as the basis of neglecting the transients that result in liquid flow through the valves.
As a result of the above considerations. it was determined that the loss-of-load or locked rotor event produced the limiting conditions for steam discharge through the safety and relief valves.
NRC. QUESTION 12. PART b A detailed description of the methods used to perform this analysis.
This includes a description of methods used to generate fluid pressures and moments over time and methods used to calculate resulting fluid forces on the system.
Identify the computer progress used for the analysis and how these programs were verified.
1 ODAI RESPONSE The method used to perform tais analysis is based on the method used in the many thermal hydraulic analyses of safety / relief valve piping systems as found in the literature (see References 4, 5, 6, 7, 8 and 9).
The steps used are given below 1.
Develop an ANSYS (Reference 10) finite element structural model of the piping system.
(See details below in ODAI response to NRC question 13).
This is the most logical first step because the ultimate goal of the entire analysis is to verify that the stress levels in the piping system are in compliance with the ASME Boiler and Pressure Yessel Code (Reference 11).
Therefore, the thermal hydraulic model which provides the input to the structural model must be compatible with the structural model.
To this end, the guidelines given in References 12 and 13 were followed.
2.
Develop a RELAPS/ MOD 1 (Reference 14) finite difference model of the piping system following the guidelines of References 12 and 13.
RELAPS/ MOD 1 vos written to investigate the thermal hydraulic reponse of light water reactors to a loss-of-coolant accident (LOCA).
Its original intent van not for the determination of pressure waves in piping systems.
As a result it has capabilities which are not necessary for the solution of pressure surge problems, e.g. RELAPS/ MOD 1 contains internal heat generation and reactor kinetics data which are not needed in relief valve applications.
RELAPS/ MOD 1 does not give reaction forces due to
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Therefore a post-processor such so REPIPE (Reference 15) is needed to accept the RELAPS/N001 thermo hydraulic output and resolve it to forces. The application of RELAPS/N001 to problems similar to the relief valve discharge line probles (the safety / relief valve discharge in pressurized water reactore) has been discussed before (Reference 8).
Control Data Corporation (CDC) esintaine RELAPS/ MOD 1 on its CYBERNET system and has written the post processor REPIPE which determines the fluid forces.
3.
Using the thereal hydraulic output from RELAPS/N001 as input to REPIPE, obtain the force time history which is applied to the ANSYS structural model of Step 1.
REPIPE le a post processor computer code that converto the thermal hydraulic output of RELAPS/N001 into force time histories at desired locations.
These force time histories are generated between two arbitrary Junctions in the thermal hydraulic model.
The conson analytical approach to to perfore structural evaluations with the forces acting along the axis of piping elemente.
Therefore, the RELAPS/N001 thermal hydraulic output veo converted into forces using REPIPE.
The force time histories generated by REPIPE vere composed.of the sus of the wave and blowdown forces.
Reference 15 discusses the calculations of each of these forces.
The REPIPE output consiste of the x, y and z componente of the sus of these forces relative to the absolute coordinate system used in the ANSYS structural model.
MRC QUEST 10M 12. PART c Identification of important parametere used in the thermal hydraulic analyste and rationale for their selection.
These include peak pressure and pressurization rate, valve opening time, and fluid conditions at valve opening.
ODAI RESP 0ESE The generic analysis (References 1 and 2) for the four-loop plant predicted a peak pressure of 2SSS pela for the loss-of-load case.
The Zion plant specific lose-of-load analyste presented in the FSAR determined the peak pressure to be 2332 peia.
The manteve preneurization rate results from a locked rotor event.
For the generic analysis, a pressurization rate of 144 pel/soe to predicted, while the Zion FSAR analysis estimates the pressurination rate to be 80 poi /sec.
These analyses also confirmed that only stese le vented from the pressuriser in these cases.
The actual valve stes position (of the safety valve) wereue time for EPRI/CE Test 917 (Reference 2) le shown in Figure 1.
The time history consiste of two distinct periode, the simmering time period and ' pop' time period.
For Test 917, a staaering time of 0.9077 seconde and a pop time of 0.01475 seconde were sensured.
The valve fully opened upon steam flow after the loop seal water had cleared the valve se a result of the elesering process.
4 l
l The valve opening characteristics employed in the RELAPS/N001 valve ocdel are superimposed over the data in Figure 1.
The valve model used in the analysis employed conservative values of 0.88 second and 0.0145 second for j
the siseer and pop periods.
The fluid conditions in the RELAPS/N001 model were based on the actual plant data as obtained from Reference 16.
These conditions are given la Table 1 and Figure 2.
The loop seal water was modeled with all the water in place upstrees of the valve.
MRC QUESTION 12. PART d An explanation of the method used to treet valve resistances in the analysis.
Report the valve flow rates that correspond to the resistances used.
Because the ASME Code requires derating of the safety valves to 90X of actual flow capacity, the safety valve analysis should be based on flows equal to 111X of the valve flow rating, unless another flow rate can be Justified.
Provide information explaining how dorating of the safety valves was handled and describe methods used to establish flow rates for the safety i
valves and PORVs in the analysis.
ODAI RESPQRSE I
The valves were modeled using the conventional RELAP5/N001 valve component.
2 For this component a full open flow area of 0.025 ft, a valve discharge coefficient (C of 0.8 and the opening time given in Ites 12c above were D
used.
The results of the model gave a steady state stems flow rate of 129.3 lb,/sec which corresponds to 111X of the valve flow rating trating is 420,000 lb,/hr).
The flow rate (e) at any instant of time is detereined by the following equations l
e = AC V2 PAP D
l where A is the flow area, C is the valve discharge coefficient, p is the D
density and AP is the preneure drop through the valve.
By using the values l
of A, C, p an AP calculated by REAPS /N001, one oMains the same value of D
a as calculated by RELAPS/M001.
MRC QUESTION 12. PART e l
A discussion of the sequence of opening of the safety valves that was used i
to produce worst case loading conditions.
ODA! RESEQNSE
[
This has been covered in Items 12b and 12c above, i
i l
~9 MRC QUESTIDES 12. PART d A sketch of the thermal hydraulie model showing the size and number of fluid control valuees.
QDALRESPQRSE See Figures 2 and 3.
MRC QUE$H OS 12. PART a later.
ERC QuRCTION 13 The submittal states that a structeral analysis of the safety /PORY valve
. piping system has been conducted. but does not present details of the unelysis.
To allow for a complete evaluation of the methods used and results obtained from the structural unelysis. please provide reporte containing at least the icllowing informations e.
A detailed description,of the methods used to perfore the e
analysis.
Ide ntify the computer progress used for the analysis and her these progrees were varified.
QDAUlfSPGESE As sentioned in Ites 12b above, an ANSf5 finite element structural model was developed for the safety / relief valve piping systes.
The ANSYS computer progree le a large-scale, general perpees computer progree for the solution of several cissees of engineering analysew.
Analysis capabilities include static and dyneeles elasti; and plastics small and large deflootione; linear and non-linear.
The estrix displacement method of analyses based upon finite element idealization is used.
The library of finite eleopate includess elastia pipe, tee, elbow, bene and shell elementes pleetic pipe elbov, bene and shell elementes substructure (superelementals springlease viesents.
The loading on the structure may be in the form of forces, displeuesente, pressures, temperatures or response spoetra.
ANSYS has been verified and quality soeured for Nuclear Safety l
Nelated analysee.
For this structitral analysis, the straight pipe sections were modeled as elastic pipe eleaente, the pipe teos wase modeled es elagtic pipe tee elemente, the valves and pipe supports were mode 14d as explained below in Ites 13c, the pipe elbove vese modeled as elastic pipe elbow elemente except that the three 12 inch diameter elbows in the header at the expected high streme locations were modeled se superelements which were obtained from a detailed elastic shell element model of the 12 inch diameter elbow.
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NRC QUESTION 13b l
A description of the method used to apply the fluid forces to the structural model. Since the forces acting on a typical pipe segment are composed of a net, or " wave", force and opposing " blowdown" forces, describe the methods for handling both types of forces.
ODAI RESPONSE t
ANSYS has the capability to allow the structural analyst to apply the fluid i
forces directly to the uodal locations of interest (i.e. locations of high stresse levels. ) As sentioned above in Ites 12b and Ites 13a, the guidelines given in Reference 12 and 13 were followed in developing the structural model so that the ANSYS model nodes included the locations of high stress levels.
As mentioned in Ites 12b, REPIPE calculated the wave and blowdown forces for the desired locations and then the force time history was applied to the ANSYS structural model in order to determine the stress levels of the i
l discharge piping system.
MRC QUESTION 13a A description of methods used to model supports, the pressurizer and relief j
tank connections, and the safety valve bonnet assemblies and PORY actuator.
j ODAI RESP 0MSE Standard structural modeling practices were followed in developing the ANSYS i
structural model of the discharge piping system.
These include the j
followings l
1.
Pipe Supports The ease of the support clamps and the mass of the dynamic portion of the support attached to the pipe were modeled as a lumped mass i
and placed on the pipe node at or very near to its physical 1
location.
The values for the masses were obtained from References 17 and 18.
A node at its physical location corresponding to the i
centerline of the pipe was used to represent the end of the support i
attached to the pipe.
A node at its physical location was used to l
represent the end of the support not attached to the pipe.
This l
node was constrained in all degrees of freedom.
An ANSYS spring element was used to connect the two nodes of the support.
The j
values for the spring constants were obtained from References 19 i
and 20.
i The constant force supports were modeled as a lumped mass to represent the pipe clamp and the dynamic portion of the support.
The values of the forces and the masses were obtained fros
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References 17 and 18.
The masses and forces were placed on the pipe nodes at or very near to their physical locations.
2.
Press.urizer and Relief Tank Connections j
The locations of the pressurizer and relief tank connections were represented by pipe nodes at their physical locations corresponding to the centerline of the pipe.
These nodes were constrained in all degrees of freedos.
3.
Valves All of the valves were modeled using three relatively stiff beau elements and a mass element at the valve center of gravity as follows: one bene element running from the node at the valve inlet to the node at the valve outlet, one beam element running from the node at the valve inlet to the node at the valve center of gravity and one beau element running from the node at the valve outlet to the node at the-valve center of gravity.
The values for the locations of the nodes at the center of gravities, inlets and l
outlets and the values for the masses were obtained from References 20 and 21.
4.
Safety Valve Stands The safety valve stand was modeled as a two noda beau element.
One node was attached to the center line of safety valve inlet piping corresponding to its physical location.,
The other node was attached to the center line of safety valve inlet piping corresponding to its physical location.
The other node was at the enchor end of the stand at its physical location and was constrained in all degrees of freedom.
NRC QUESTION 13d An identification of the load combinatic'ns performed in the analysis together viih the allowable stress limits.
Differentiate between load combinations used in the piping upstream and downstream of the valve.
Explain the mathematical methods used to perform the load combinations, and identify the governing codes and standards used to determine pipir.g and support adequacy.
ODAI RESPONSE The previous analyses reported in References 9 and 19 showed that an overstressed condition would result only in the event of the simultaneous opening of all three safety valves (see Ites 12 above for description of this event).
Because this type of event is classified as an occasional load, Article F-1000 of the ASNE Boiler and Pressure Vessel Code (Reference
1 6
i I
- 22) applies.
That is, the load combination consists of the sua of the sustained loads during normal plant operation (except peak pressure is used) and the dynamic load caused by the slug flow.
Because the piping upstream of the safety valves is Seismic Class I piping, the dynamic force produced by a maximum credible earthquake (SSE) must also by included in the load combination.
For this load combination the square root of the sua of the squares (SRSS) sethod is used.
From F-1331.1 of Reference 22, the following standards were used to determine the piping and support adequacy:
1.
The general primary membrane stres intensity P, shall not exceed the lesser of 2.4 S,and 0.7 S, for materials included in Table I-1.2, or 0.7 S, for materials included in Table I-1.1.
2.
The local primary membrane stress intensity P shall not exceed g
150% of the limit for general primary membrane stress intensity P,.
3.
The primary membrane (general or local) plus primary bending stress intensity, P + P, shall be limited in accordance with one of the g
B following provisions:
a.
Stress intensity, Pg + P, shall not exceed 150% of the limit B
for general primary membrane stress intensity P,;
b.
static or equivalent static loads shall not exceed 90% of the limit analysis-collapse load using a yield stress which is the lesser of 2.35,and 0.7S, or 100% of the plastic analysis-collapse load or test collapse load (F-1321.6).
4.
The average primary shear stress across a section loaded in pure l
shear shall not exceed 0.42S.
For the materials under consideraiton in this analysis, values of the ultimate stress (S ), and yield stress (S,), and allowable stress (S ) as a function of temperature are given in Table 2.
g t
MRC QUESTION 13e An evaluation of the results of the strbctural analysis, including identification of over stressed locations and a description of modifications, if any.
l l
l l
9 ODAI RESP 0MSE (Later)
NRC QUESTI0M 13f A sketch of the structural model showing lumped mass locations, pipe sizes, and application points of fluid forces.
ODAI RESPONSE See Figure 4.
NRC QUESTION 13a A copy of the contractors structural analysis report.
ODAI RESPONSE Later.
10 l
TABLE 1 Fluid Initial Conditions Itests)
Conditions 1.
Fluid in pressurizer and fluid Saturated steam at the safety upstream of the loop seals valve set point pressure (2499.7 paia) 2.
Fluid in the loop seals See Figure 2 for temperatures 3.
Fluid downstream of the safety Air-water sixture at 100%
valves and inside the relative humidity at 110'F containment and 14.7 psia 4.
Fluid outside of the containment Air-water mixture at 100%
but not in the relief tank relative humidity at BO'F and 14.7 psia 5.
Fluid inside the relief tank Water at 80'F, air at 80'F
11 TABLE 2 Material Data Nominal Wall Size Pipe Thickness Temperature Location (in.)
Schedule (in.)
Neterial Ranae ('F)
Upstream of Safety / Relief Valves 6
160 0.718 SA-376 TP316 120-668 Upstream of Relief Valves 3
106 0.437 SA-376-TP316 120-668 Downstream of Relief Yalves 3
40 0.216 SA-312 TP304 110 Downstream of Safety Reliet Valves 6
40 0.280 SA-312 TP304 110 Header 12 40 0.406 SA-358 316 80-110 SA-312 TP304 SA-376 TP316 AND SA-358 316 Temp.
S S
S 8
g A
u M
A
(*F) ikg11 (kei).
(kei)
(ksi)
(ksi)
(kei) 100 75.0 30.0 52.501 75.0 30.0 52.501 200 71.0 25.0 49.701 75.0 25.8 52.501 300 66.0 22.5 46.201 73.4 23.3 51.381 400 64.4 20.7 45.081 71.8 21.4 50.261 500 63.5 19.4 44.451 71.8 19.9 47.762 600 63.5 18.2 43.682 71.8 18.8 45.122 650 63.5 17.9 42.962 71.8 18.5 44.402 700 63.5 17.7 42.482 71.8 18.1 43.432 IS = 0.7 S, 2S = 2.4 S A
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TO GURE 4B dD 5
DOWNSTREAM SAFETY VALVE PIPING 6 INCH NOMINAL OD 0.280 INCH WALL
- SUPPORT MASS LOCATION C FLUID FORCE LOCATION
'e FLUID FORCE AND SUPPORT MASS LOCATION s
e VALVE MASS LOCATION H HEADER PER FIGURE 4D R RELIEF PIPING PER FIGURE 4C se i
FIGURE 4A STRUCTURAL MODEL 4
e i
i 4
I J
T
16 O
O Q
c g
i O
UPSTREAM SAFETY VALVE PIPING 6 INCH NOMINAL OD C
0.718 INCH WALL O
- SUPPORT MASS LOCATION O FLUID FORCE LOCATION J
L Q
- e
,f' 1:;~
a.
Y 4
Z x
0 0
0 0
O -
FIGURE 4B STRUCTURAL MODEL
17 M0VB a
3 C
ii y
PORV i'H D
h E
[\\
5 ii s
N 3
z x
i, C(
5
)
A H c
RELIEF PIPING A to C 6 INCH NOMINAL 00, 0.718 INCH WALL C to D 3.5 INCH NOMINAL 00, 8 to L 0.437 INCH WALL E to F 3.5 INCH NOMINAL OD, G to K 0.216 INCH WALL F to J 6 INCH NOMINAL 00, 0.280 INCH WALL G
Ortule FORCE LOCATION FIGURE 4C STRUCTURAL MODEL
(;
18
(>
AgN 5
f a
R N)
S o
ib HEADER PIPING 12 INCH NOMINAL OD 0.406 INCH WALL
- SUPPORT MASS LOCATION O FLUID FORCE LOCATION
@ FLUID FORCE AND SUPPORT MASS LOCATION Y
o g
Z x
00 3 l
0 e
ii' i
FIGURE 4D STRUCTURAL MODEL
19 REFERENCES 1.
Westinghouse Nuclear Energy Systems, " Review of Pressurizer Safety Valve Performance as Observed in the EPRI Safety and Relief Valve Test Program," WCAP-10105, June 1982.
2.
Electric Power Research Institute, " Valve Inlet Fluid Conditions for Presurizer Safety and Relief Valves in Westinghouse-Designed Plants,"
EPRI NP-2296, EPRI Project V102-19, Final Report, December 1982.
3.
Science Applications, Inc., "Probabilistic Evaluation of High Pressure Liquid Challenges to Safety / Relief Valves in the Zion, Byron /Braidwood PWR Plants," June 25, 1982.
4.
House, R.
K.,
et al.,
" Application of RELAPS/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads," EPRI NP-2479, EPRI Project V102-28, Final Report, December 1982.
5.
Motloch, C.
G.,
Van Blaricum, C. H.,
and Narum, R. E.,
"RELAP5/ANSYR/ANSYS Hydrodynamic Force Calculation of the Electric Power Research Institute Safety and Relief Valve Discharge Test (CE Test No. 1027)," EI-83-12, December 1983.
6.
Ca]igas, J.
M.,
" Verification of the RELAP5-FORCE Hydraulic Force Calculation Code," Gilbert Associates, Inc., May 1984.
7.
Semprucci, L. B. and Holbrook, B. P., "The Appiciation of RELAP4/REPIPE to determine Force Time Histories on Relief Valve Discharge Piping,"
ASME, PVP-33, June 1977.
8.
Strong, B.
R.,
Jr. and Beschiere, R.
J.,
" Steam Hammer Design Loads for Safety / Relief Valve Discharge Piping,' ASME, PVP-33, June 1977.
9.
Sargent & Lundy Report SL 4283 dated May 2, 1984, " Evaluation of the Pressurizer Safety and Relief Valve Discharge Piping System - Zion Stations 1 and 2.*
10.
ANSYS Engineering Analysis System, Revision 4.1, Swanson Analysis Systems, Inc., Houston, Pennsylvania.
11.
" Power Piping," ANSI /ASME B31.1, ASME Code for Pressure Piping, The American Society of Mechanical Engineers, 345 East 47th Street, New York, New York, 10017.
12.
Norton, P.
J.,
" User's Manual for Program REPIPE," Utilities Service Center, CDC, Rockville, Maryland.
-l 20 t
13.
" Criteria and Guidelines for the Design of Safety and Relief Valve i
Installation in Westinghouse Pressurized Water Reactor Plants,'
Westinghouse Electric Corporation, NES, PWR Systems Division, October 1972.
14.
Ransos, V.
H.,
et al.,
'RELAP5/M001 Code Manual," Volumes 1-2, NUREG/CR-1826, EGG-2070 Draft, Rev. 2, September 1981.
15.
Norton, P.
J.,
" User's Manual for Program REPIPE,' Utilities Service Center, CDC, Rockville, Maryland.
16.
Graesser, K.
L.,
(Zion Station Superintendent) to Butterfield, L.
D.,
(CECO), Letter November 9, 1982, " Unit 2 Pressurizer Safety Valve Loop Seal Temperatures.*
Sargent & Lundy Reactor Coolant System Support Drawings:
17.
1 I
Hanner No.
Egtt Hanner No.
Enig.
IRC146-FR1 8-25-77 RCH-1008 12-18-72 1RC146-SR1 4-21-77 RCH-1009 1-28-74 1RC147-SR1 4-21-77 RCH-1014 10-27-72 1RC147-SR2 4-21-77 RCRS-1112 11-20-72 l
1RC151-RV1 4-21-77 RCRS-1114 6-02-71 1RC157-RV1 8-25-77 RCRS-1115 11-20-72 1RC157-RV2 4-21-77 RCRS-1119 2-16-73 RCH-1005 10-27-72 RCRV-001 12-21-72 RCH-1007 1-12-73 18.
Stone & Webster Bulletin 79-14 Modification Support Drawings:
llanner No.
Dals.
RCH1006 2-10-81 RCH1010 1-30-81 RCRS1117 1-30-81 RCRS1118 1-30-81 I
RCRS1120 2-04-81 RCRS1121 1-30-81 RCRS1122 1-30-81 RCRS1123 2-04-81 RCRS1117A 7-22-81 RCRS1117B 7-22-81 RCRS1118A 7-22-81 RCS1011 RCS1012 RCS1013 4
r n.-,,------,,-,----,-.----w-
r-21 i
19.
Books 1 through 6, inclusive, of Stone & Webster, " Zion Station Pipe Stress and Support Analysis Report," Number 13430RC - 2, 3, 4, 5, Revision O, dated January 17, 1983, Cossonwealth Edison Job Order 13430.01 for Reactor Coolant (Pressurizer 1RC002 to Pressurizer Relief Tank 1RC003).
20.
Sargent & Lundy Report No. 037064, Project No. 6320-00.
"Dynesic Analysis of Typical Pressurizer Safety and Relief Valve Discharge Piping Due to Valve Actuation," dated August 1982.
21.
Sargent & Lundy Drawing M-418, Pressurizer Piping Analytical Data Isometric, Zion Station Unit 1, Sheet No. 1, Rev. D, Dated July 31, 1979.
20.
ASME Boiler and Pressure Vessel Code,Section III, Division 1, Appendix F, 1983.