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MONTHYEARIR 05000344/19870271987-07-0909 July 1987 Insp Rept 50-344/87-27 on 870622-26.No Violations or Deviations Noted.Major Areas Inspected:Usi A-26 & Followup & Closure of Previous Insp Findings Project stage: Request ML20235K2511987-07-0909 July 1987 Requests Addl Info to Assess Safe Operation of Facility Following 1987 Refueling Outage.Concerns Expressed Re Supports Inadequate to Meet Design Loads & Original Design Loads Derived Inappropriately.Submittal of Fee Not Required Project stage: Other IR 05000344/19870241987-07-10010 July 1987 Insp Rept 50-344/87-24 on 870524-0725.Violations Noted.Major Areas Inspected:Operational Safety Verification,Maint, Surveillance,Piping Support & Restraint Sys & Followup on Previously Identified Items Project stage: Request ML20235J3961987-07-10010 July 1987 Responds to NRC on Concerns Re safety-related Supports at Plant,Including Original Design,Derivation of Original Design Loads & Original Design Calculation for Main Steam Sys Designed for Plant Project stage: Other ML20235R8801987-07-15015 July 1987 Forwards Revised Pages to Util 870710 Response to NRC Re safety-related Piping Supports at Facility.Changes Do Not Affect Results or Conclusions of Verification Program in Project stage: Other ML20236F1071987-07-27027 July 1987 Forwards Addl Info Requested During NRC 870721-23 Review of Util Support Design Verification Program.Util Response to NRC Concerns Re Inadequate Supports,Original Design Loads & Documentation of Design Calculations Discussed Project stage: Other ML20236H9721987-07-31031 July 1987 Forwards Addl Info Clarifying Util 870727 Response to NRC Concerns Re Design Adequacy of Piping Supports.Util Concludes That Heatup Commencement & Return to Power Operation Safe Although All Activities Not Complete Project stage: Other ML20237H2081987-08-10010 August 1987 Rev 0 to 01-0300-1629, Independent QA Audit & Technical Quality Review of Trojan Nuclear Power Plant Pipe Support Verification Program Project stage: Other ML20237H1881987-08-18018 August 1987 Forwards Info Including Independent Audit Re Scope of Pipe Support Design Verification Program,Per 870710,15,27 & 31 Submittals.Supports Verified & short-term Program Complete. Util Will Provide Details of long-term Program by 870930 Project stage: Other ML20237G6261987-08-21021 August 1987 Informs That Based on Info Provided,Nrc Finds That Util Adequately Evaluated Design Adequacy of Pipe Supports & Supports Will Fulfill Requirement,As Stated in Updated Fsar. NRC Concludes That Power Operation May Be Resumed Project stage: Other ML20238C6691987-09-0404 September 1987 Forwards Revised Pages of Util Re Pipe Support Design Verification Program.Ltr Revised Due to Typos Re Std Deviations in Min Rock Bolt Depth Assumption & Revised Chi- Squared Values Project stage: Other ML20235G8061987-09-25025 September 1987 Summary of 870715 Meeting W/Util,Bechtel & Gao Re Main Steam Support Adequacy & Main Feedwater Pipe Thinning.List of Attendees & Util Presentation Encl Project stage: Meeting ML20235Q8051987-09-30030 September 1987 Forwards Pipe Support Design Verification long-term Plan,Per 870818 short-term Plan,For safety-related,large-bore Piping. Encls Include Responses to Remaining short-term Concerns, Evaluation Plan for Ae Design Activity & Ae Audit Response Project stage: Other ML20236A8971987-10-13013 October 1987 Forwards Safety Evaluation Re Pipe Support Design Verification.Issues & Details of Pipe Support Design Verification long-term Plan Discussed in Will Be Addressed in Future Correspondence Project stage: Approval 1987-07-09
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Category:CORRESPONDENCE-LETTERS
MONTHYEARML20217E9711999-10-13013 October 1999 Submits Notification of Major Changes to Trojan Liquid Radioactive Waste Treatment Sys,Iaw PGE-1201.Detailed Description of Change Provided ML20217C8171999-10-0606 October 1999 Forwards Notice of Receipt of Availability for Comment & Meeting to Discuss License Termination Plan,Per 990805 Application ML20216F7621999-09-23023 September 1999 Forwards Corrected Response to Request 2 Contained in NRC 990920 RAI Re Application of Pacificorp for Transfer of License NPF-1.Response 2 Should Have Stated That Na General Partnership Is Partnership Formed in Nv ML20216F2871999-09-20020 September 1999 Informs NRC of Developments That Have Occurred Since 990524 Application Was Filed Re Pacificorp Transfer of License of FOL NPF-1.NRC Is Urged to Act & Approve Transaction Expeditiously by 990930.Supporting Documentation Encl ML20211Q3281999-09-0909 September 1999 Forwards Insp Rept 50-344/99-06 on 990630-0701,21 & 0408-08. No Violations Noted.Insp Conducted to Review Decommissioning Activities Underway at Trojan Site & to Accompany Shipment of Reactor Vessel to Hanford,Washington for Burial ML20211J2101999-08-30030 August 1999 Forwards Request for Addl Info Re Application for Approval of Proposed Corporate Merger of Pacificorp & Scottishpower ML20211B6611999-08-16016 August 1999 Forwards fitness-for-duty Program Performance Data Rept for Period of 990101-0630,IAW 10CFR26.71(d) ML20211B4091999-08-16016 August 1999 Forwards Environ Assessment & Finding No Significant Impact to Application for an Exemption & License Amend Dated 980129.Proposed Exemption & License Amend Would Delete Security Plan Requirements of 10CFR50.54(p) & 10CFR73.55 ML20211A7131999-08-16016 August 1999 Forwards Environ Assessment & Finding of No Significant Impact to Application for Exemption & License Amend Dtd 980827.Proposed Exemption & License Amend Would Delete EP Requirements of 10CFR50.54(q),10CFR50.47(b) & 10CFR50,app E ML20210R7691999-08-11011 August 1999 Forwards Proposed Rev 23 to PGE-8010, Trojan Nuclear QAP, in Response to NRC 990708 RAI Re Relocation of TS ACs to Qap.Revised QAP Will Be Made Effective Concurrently with Implementation of License Change Application Lca 245 ML20210H5971999-07-27027 July 1999 Forwards Notice of Consideration of Approval of Application Re Merger & Opportunity for Hearing.Notice Being Forwarded to Ofc of Fr for Publication ML20216D6611999-07-23023 July 1999 Submits Summary of Proprietary Submittals for Transtor Part 71 & Part 72 & Trojan ISFSI Applications ML20210F8601999-07-22022 July 1999 Forwards Rev 1 to PGE-1076, Trojan Reactor Vessel Package Sar. Changes to Rept Contained in Rev 1 Received NRC Approval by Ltr ML20210A6401999-07-19019 July 1999 Corrects Ref in Item 4 of Which Constitutes Rev 2 of Authorization from Wf Kane, for Trojan Reactor Vessel Package as Approved Package for Shipment Under General License,Subj to Listed Conditions ML20210B4481999-07-12012 July 1999 Forwards Rept Describing Effects of Earthquake That Occurred on 990702 Near Satsop,Wa,Iaw Trojan Nuclear Plant Defueled Sar,Section 4.1.3.1 ML20209D6231999-07-0808 July 1999 Forwards RAI Re Licensee 980827 Request for Amend That Would Delete Number of License Conditions & TS Requirements That Would Be Implemented After All Sf Has Been Removed from 10CFR50 Licensed Area.Response Requested within 30 Days ML20209C6481999-07-0606 July 1999 Forwards Rev 8 to Defueled Sar,Including Changes Since Last Submittal.Attachment Includes Brief Description of Each Change Included in Rev ML20209B7821999-07-0101 July 1999 Responds to NRC 990609 RAI Re License Change Application 244 & Accompanying Request for Exemption.Detailed Info Supports Estimation of Remaining Radioactive Matl Previously Provided by Licensee ML20212J3281999-06-15015 June 1999 Forwards Amend 22 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan, IAW 10CFR50.48(f).Amend Reflects Revs Made During Decommissioning Activities & Does Not Reduce Effectiveness of Fire Protection ML20195J0111999-06-0909 June 1999 Responds to Requesting License & Exemption Re Emergency Preparedness ML20207D3861999-06-0101 June 1999 Forwards Rev 1 to PGE-1077, Trojan Nuclear Plant Reactor Vessel & Internals Removal Project Transportation Safety Plan ML20196L1251999-05-24024 May 1999 Forwards Application for Amend to License NPF-1 for Indirect Transfer of License,To Extent That Such Approval Required Solely to Reflect Change in Upstream Economic Ownership of Pacificorp ML20207A2751999-05-14014 May 1999 Informs That NRC Office of Nuclear Reactor Regulation Reorganized Effective 990328.As Part of Reorganization, Division of Licensing Project Mgt Created.Organization Chart Encl ML20206N9411999-05-11011 May 1999 Forwards Revised Epips,Including Rev 7 to EPIP 3, Response Organization Checklists & Rev 9 to EPIP 5, Emergency Preparedness Test Propgram. Changes to EPIPs 3 & 5 Ref New Owners of on-site Railroad Line,Portland & Western Railroad ML20206J8681999-05-0707 May 1999 Forwards Insp Repts 50-344/99-05 & 72-0017/99-04 on 990419-22.No Violations Noted.Insp Observed Work Activities Associated with Lifting of Reactor Vessel in Preparation for Removal & Shipment to Hanford Reservation for Burial ML20206J7931999-05-0707 May 1999 Forwards Insp Repts 50-344/99-04 & 72-0017/99-02 on 990322- 25 & 29-0408.One Violations Identified & Being Treated as non-cited Violation,Consistent with App C of Enforcement Policy ML20206H4331999-05-0505 May 1999 Forwards Amend 201 to License NPF-1 & Se.Amend Revises PDTSs by Deleting ISFSI Area,Revises Subsection 4.1.1,replaces Figure 4.1-1 with New Figure 4.1-1 & Adds New Page to Figure 4.1-1 to Reflect Access Control (ISFSI) Area ML20206N1571999-05-0404 May 1999 Forwards Util Quarterly Decommissioning Status Rept for First Quarter of 1999,IAW State of or Energy Facility Siting Council Order Approving Trojan Decommissioning Plan, ML20206E1731999-04-29029 April 1999 Informs That NRC Staff Has Performed an Acceptance Review of Trojan Nuclear Plant License Termination Plan,Submitted by ,To Determine Whether LTP Provides Adequate Info to Allow Staff to Conduct Detailed Review ML20206E0211999-04-28028 April 1999 Forwards Copy of Environ Assessment & Finding of No Significant Impact Re 970212 Application for Amend.Proposed Amend Would Revise Trojan Permanently Defueled TS to Delete ISFSI Area ML20206C9591999-04-23023 April 1999 Forwards Amend 200 to License NPF-1 & Safety Evaluation. Amend Changes License NPF-1 by Revising License Condition 2.C.(10), Loading of Fuel Into Casks in Fuel Building ML20206C9221999-04-23023 April 1999 Forwards Amend 199 to License NPF-1 & Safety Evaluation. Amend Changes License NPF-1 by Adding New License Condition Entitled, Loading of Fuel Into Casks in Fuel Building ML20205S9761999-04-21021 April 1999 Forwards Trojan Nuclear Plant,Radiological Environ Monitoring Rept for CY98. Rept Submitted in Accordance with Trojan Permanently Defueled TS 5.8.1.2 & Sections IV.B.2, IV.B.3 & Iv.C of App I to Title 10CFR50 ML20205T5471999-04-20020 April 1999 Forwards Insp Repts 50-344/99-03 & 72-0017/99-03 on 990301-04,15-18 & 22-25.No Violations Noted ML20205N9061999-04-13013 April 1999 Forwards Insp Rept 50-344/99-02 on 990329-0401.No Violations Noted.Inspectors Examined Portions of Physical Security, Access Authorization & FFD Programs ML20205P7301999-04-0808 April 1999 Forwards PGE-1009-98, Operational Ecological Monitoring Program for TNP,Jan-Dec 1998, Including All Existing non-radiological Effluents ML20205B3661999-03-25025 March 1999 Transmits Completed Application for Renewal of NPDES Permit for Trojan Nuclear Plant,Iaw License NPF-1,App B,Epp,Section 3.2 ML20204G1781999-03-18018 March 1999 Forwards Rev 4 to PGE-1063, Suppl to Applicants Environ Rept - Post Operating License Stage ML20204B7551999-03-18018 March 1999 Forwards Updated TS 5.6 Re High Radiation Area,Per Telcons with Nrc.Justification for TS Was Provided Previously with Util Ltr Dtd 990317,but Has Been Updated & Is Included as Encl 2 ML20204C5151999-03-17017 March 1999 Forwards Licensee Comments on NRC Preliminary SER & License Re Trojan Isfsi.Encl Includes Justification for Inclusion in ISFSI TS of Alternative Method to 10CFR20.1601(c) for Controlling Access to High Radiation Areas ML20207J3721999-03-10010 March 1999 Forwards License Amend Application 247 Requesting Amend to License NPF-1 to Add License Condition Denoting NRC Approval of PGE-1078, Trojan Nuclear Plant License Termination Plan, Also Encl.With Certificate of Svc ML20207L0011999-03-0808 March 1999 Transmits Tnp co-owners Annual Rept of Status of Decommissioning Funding for Tnp.Rept Is Based on Most Recent Analysis of Tnp Decommissioning Estimate & Funding Plan,Per Rev 6 to Pge, Tnp Decommissioning Plan ML20207G9691999-03-0303 March 1999 Forwards Rev 6 to PGE-1061, Trojan Nuclear Plant Decommissioning Plan. Summary of Changes,Attached.Revised Portions Denoted by Side Bars ML20207D7751999-03-0202 March 1999 Forwards Amend 21 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan, Per 10CFR50.48(f).Amend Reflects Revs Made During Decommissioning Activities ML20207B0441999-02-24024 February 1999 Forwards Endorsements 139 to Nelia Policy NF-0225 & 2 to Nelia Policy NW-0602 ML20207J0701999-02-11011 February 1999 Forwards Proposed Ts,Update to ISFSI SAR & Revised Calculation,Per Application for Trojan ISFSI License ML20202G4291999-02-0202 February 1999 Forwards Rev 0 to PGE-1076, Trojan Reactor Vessel Package Sar. Approval,With Certain Conditions,For one-time Shipment of Trojan Reactor Vessel Package Granted by Commission Via Ltr ML20202G0931999-01-26026 January 1999 Submits Following Info That Will Be Needed for NRC Staff to Complete Review & Issue Trojan ISFSI License,As Result of 990121 Meeting with NRC Following Insp & Observation of ISFSI Preoperational Testing During Wk of 990118 ML20202F2551999-01-25025 January 1999 Forwards Fitness for Duty Program Performance Data Rept for July-Dec 1998 ML20202C1621999-01-21021 January 1999 Forwards Insp Repts 50-344/98-04 & 72-0017/98-01 & NOV Re Inadequate Actions Taken by Radiation Protection Technician to Ensure That Radiological Conditions Safe Prior to Removing Warning Signs for Airborne Radioactivity Area 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARML20217E9711999-10-13013 October 1999 Submits Notification of Major Changes to Trojan Liquid Radioactive Waste Treatment Sys,Iaw PGE-1201.Detailed Description of Change Provided ML20216F7621999-09-23023 September 1999 Forwards Corrected Response to Request 2 Contained in NRC 990920 RAI Re Application of Pacificorp for Transfer of License NPF-1.Response 2 Should Have Stated That Na General Partnership Is Partnership Formed in Nv ML20216F2871999-09-20020 September 1999 Informs NRC of Developments That Have Occurred Since 990524 Application Was Filed Re Pacificorp Transfer of License of FOL NPF-1.NRC Is Urged to Act & Approve Transaction Expeditiously by 990930.Supporting Documentation Encl ML20211B6611999-08-16016 August 1999 Forwards fitness-for-duty Program Performance Data Rept for Period of 990101-0630,IAW 10CFR26.71(d) ML20210R7691999-08-11011 August 1999 Forwards Proposed Rev 23 to PGE-8010, Trojan Nuclear QAP, in Response to NRC 990708 RAI Re Relocation of TS ACs to Qap.Revised QAP Will Be Made Effective Concurrently with Implementation of License Change Application Lca 245 ML20216D6611999-07-23023 July 1999 Submits Summary of Proprietary Submittals for Transtor Part 71 & Part 72 & Trojan ISFSI Applications ML20210F8601999-07-22022 July 1999 Forwards Rev 1 to PGE-1076, Trojan Reactor Vessel Package Sar. Changes to Rept Contained in Rev 1 Received NRC Approval by Ltr ML20210B4481999-07-12012 July 1999 Forwards Rept Describing Effects of Earthquake That Occurred on 990702 Near Satsop,Wa,Iaw Trojan Nuclear Plant Defueled Sar,Section 4.1.3.1 ML20209C6481999-07-0606 July 1999 Forwards Rev 8 to Defueled Sar,Including Changes Since Last Submittal.Attachment Includes Brief Description of Each Change Included in Rev ML20209B7821999-07-0101 July 1999 Responds to NRC 990609 RAI Re License Change Application 244 & Accompanying Request for Exemption.Detailed Info Supports Estimation of Remaining Radioactive Matl Previously Provided by Licensee ML20212J3281999-06-15015 June 1999 Forwards Amend 22 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan, IAW 10CFR50.48(f).Amend Reflects Revs Made During Decommissioning Activities & Does Not Reduce Effectiveness of Fire Protection ML20207D3861999-06-0101 June 1999 Forwards Rev 1 to PGE-1077, Trojan Nuclear Plant Reactor Vessel & Internals Removal Project Transportation Safety Plan ML20196L1251999-05-24024 May 1999 Forwards Application for Amend to License NPF-1 for Indirect Transfer of License,To Extent That Such Approval Required Solely to Reflect Change in Upstream Economic Ownership of Pacificorp ML20206N9411999-05-11011 May 1999 Forwards Revised Epips,Including Rev 7 to EPIP 3, Response Organization Checklists & Rev 9 to EPIP 5, Emergency Preparedness Test Propgram. Changes to EPIPs 3 & 5 Ref New Owners of on-site Railroad Line,Portland & Western Railroad ML20205S9761999-04-21021 April 1999 Forwards Trojan Nuclear Plant,Radiological Environ Monitoring Rept for CY98. Rept Submitted in Accordance with Trojan Permanently Defueled TS 5.8.1.2 & Sections IV.B.2, IV.B.3 & Iv.C of App I to Title 10CFR50 ML20205P7301999-04-0808 April 1999 Forwards PGE-1009-98, Operational Ecological Monitoring Program for TNP,Jan-Dec 1998, Including All Existing non-radiological Effluents ML20205B3661999-03-25025 March 1999 Transmits Completed Application for Renewal of NPDES Permit for Trojan Nuclear Plant,Iaw License NPF-1,App B,Epp,Section 3.2 ML20204G1781999-03-18018 March 1999 Forwards Rev 4 to PGE-1063, Suppl to Applicants Environ Rept - Post Operating License Stage ML20204B7551999-03-18018 March 1999 Forwards Updated TS 5.6 Re High Radiation Area,Per Telcons with Nrc.Justification for TS Was Provided Previously with Util Ltr Dtd 990317,but Has Been Updated & Is Included as Encl 2 ML20204C5151999-03-17017 March 1999 Forwards Licensee Comments on NRC Preliminary SER & License Re Trojan Isfsi.Encl Includes Justification for Inclusion in ISFSI TS of Alternative Method to 10CFR20.1601(c) for Controlling Access to High Radiation Areas ML20207J3721999-03-10010 March 1999 Forwards License Amend Application 247 Requesting Amend to License NPF-1 to Add License Condition Denoting NRC Approval of PGE-1078, Trojan Nuclear Plant License Termination Plan, Also Encl.With Certificate of Svc ML20207L0011999-03-0808 March 1999 Transmits Tnp co-owners Annual Rept of Status of Decommissioning Funding for Tnp.Rept Is Based on Most Recent Analysis of Tnp Decommissioning Estimate & Funding Plan,Per Rev 6 to Pge, Tnp Decommissioning Plan ML20207G9691999-03-0303 March 1999 Forwards Rev 6 to PGE-1061, Trojan Nuclear Plant Decommissioning Plan. Summary of Changes,Attached.Revised Portions Denoted by Side Bars ML20207D7751999-03-0202 March 1999 Forwards Amend 21 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan, Per 10CFR50.48(f).Amend Reflects Revs Made During Decommissioning Activities ML20207B0441999-02-24024 February 1999 Forwards Endorsements 139 to Nelia Policy NF-0225 & 2 to Nelia Policy NW-0602 ML20207J0701999-02-11011 February 1999 Forwards Proposed Ts,Update to ISFSI SAR & Revised Calculation,Per Application for Trojan ISFSI License ML20202G4291999-02-0202 February 1999 Forwards Rev 0 to PGE-1076, Trojan Reactor Vessel Package Sar. Approval,With Certain Conditions,For one-time Shipment of Trojan Reactor Vessel Package Granted by Commission Via Ltr ML20202G0931999-01-26026 January 1999 Submits Following Info That Will Be Needed for NRC Staff to Complete Review & Issue Trojan ISFSI License,As Result of 990121 Meeting with NRC Following Insp & Observation of ISFSI Preoperational Testing During Wk of 990118 ML20202F2551999-01-25025 January 1999 Forwards Fitness for Duty Program Performance Data Rept for July-Dec 1998 ML20199J1661999-01-14014 January 1999 Submits Summary of ISFSI Lid Welding Demonstration to Be Held 990118-25.Dates & Days Are Tentative & Can Be Adjusted as Necesssary or as Required ML20199J0801999-01-12012 January 1999 Forwards Schedule for Trojan Nuclear Plant ISFSI pre-op Test,Per Request ML20206Q0121999-01-0808 January 1999 Forwards Rev 1 to PGE-1073 Tnp ISFSI Security Plan,As Result of 981208 Telcon with Nrc.Encl Withheld,Per 10CFR73.21 ML20206P5991999-01-0404 January 1999 Requests That Svc List for Delivery of Documents from NRC Be Updated,Per 981230 Telcon.Substitute for Listings of HR Pate & J Westvold,Submitted ML20198T1651999-01-0404 January 1999 Forwards Rev 5 to PGE-1061, Trojan Nuclear Plant Decommissioning Plan. Table 2.2-5 Is Revised to Depict Major Components Removed from Plant During 1998.Revs Are Denoted by Side Bars ML20198E3981998-12-17017 December 1998 Forwards Updated ISFSI Emergency Plan,Per 981123 & Discussion in Meeting with NRC on 981119.Description of Changes Encl ML20196E2851998-11-25025 November 1998 Forwards Final Rept of Matl Properties of Neutron Shielding Matl to Be Used in Trojan ISFSI & Test Repts That Verified Matl Properties ML20196D9831998-11-24024 November 1998 Forwards Listed ISFSI Calculation Refs,Per 981110 Telcon ML20196B2321998-11-23023 November 1998 Informs That,Per Commitments Made in 981119 Meeting with Nrc,Rev to PGE-1075 Will Be Submitted No Later than 981218 ML20195J2321998-11-17017 November 1998 Forwards Rev 7 to Trojan Nuclear Plant Defueled Sar. Attachment to Ltr Includes Brief Description of Each Change Included in Rev ML20195E4811998-11-10010 November 1998 Forwards Rev 51 to Trojan Nuclear Plant Security Plan.Rev Withheld,Per 10CFR73.21 ML20155E0331998-10-27027 October 1998 Forwards Amend 7 to PGE-1052, Quality-Related List Classification for Tnp. Amend Removes Ref to Several Clean Up Sys,Deletes Exemptions Allowed by Reg Guide 1.143 & Deletes Figures No Longer Required to Support Discussion ML20154M4151998-10-14014 October 1998 Requests That Further Review of License Change Application LCA-238 Be Suspended as Result of Progress Made in Decommissioning,Installation of Natural Gas Pipe Line ML20154A4751998-09-28028 September 1998 Forwards Amend 20 to PGE-1012, Trojan Nuclear Plant Fire Protection Plan, Reflecting Revisions to Fire Protection Program Made During Decommissioning Activities ML20153D3021998-09-21021 September 1998 Forwards Rev 0 to PGE-1077, Tnp Reactor Vessel & Internals Removal Project Transportation Safety Plan, Per Request That Util Submit Description of Measures to Be Taken to Ensure Safe Transport of Tnp Rv & Internals Package ML20151X7171998-09-10010 September 1998 Forwards Rev 3 to PGE-1063, Trojan Nuclear Power Station, Suppl to Applicant Environ Rept. Section 4.5 Revised to Delete Ref to Selected Room Coolers ML20151X8571998-09-10010 September 1998 Provides Rev 2 to Poge, Trojan NPP Permanently Defueled TS Bases. Changes Identified in Revised Text by Side Bars ML20153B0001998-09-0909 September 1998 Forwards Figures & Tables Identifying Nodes Assigned for Trojan ISFSI Transfer Station Analysis,TI-056,Rev 1 ML20237E9421998-08-27027 August 1998 Requests Exemption of Emergency Plan Requirements of 10CFR50.54(q),10CFR50.47(b) & 10CFR50,App E Following Transfer of Spent Nuclear Fuel to ISFSI ML20237E1811998-08-25025 August 1998 Forwards Description of Review Methodology & Summary of Results,Including List of Identified Variations from Guidance Contained in Std Review Plan,Per NUREG-1536 ML20151W5381998-08-25025 August 1998 Forwards Rev 22 to PGE-8010, Poge Nuclear QA Program for Trojan Nuclear Plant. Rev Reflects Administrative Changes That Continue to Meet Requirements of 10CFR50,App B 1999-09-23
[Table view] Category:UTILITY TO NRC
MONTHYEARML20059L6651990-09-21021 September 1990 Forwards Application for Amend to License NPF-1,consisting of License Change Application 203,Rev 1.Rev Reduces Scope of Previous Changes Re Tech Spec Table 3.6-1, Containment Isolation Valves ML20059J9841990-09-14014 September 1990 Responds to Re Violations Noted in Insp Rept 50-344/90-21.Corrective Action:Individual Employee Terminated on Day of Event ML20059J9691990-09-14014 September 1990 Forwards Estimates of Facility Operator Licensing, Requalification & Generic Fundamentals Exam Requirements for FY91-94,per Generic Ltr 90-07 ML20059J0321990-09-0707 September 1990 Requests That Six Listed Candidates Take PWR & Generic Fundamentals Exam to Be Administered on 901010 ML20059D8031990-08-31031 August 1990 Forwards Update on Actions for NRC Bulletin 88-008, Thermal Stresses in Piping Connected to Rcs ML20059D2401990-08-30030 August 1990 Extends Commitment Date from 900831 to 901031 to Amend FSAR Per Util 900404 Response to Notice of Violation ML20059D0831990-08-30030 August 1990 Revises Response to Violations Noted in Insp Rept 50-344/90-02.License Document Change Request Correcting FSAR Discrepancy Noted in App A,Violation C Re Request for Design Change Initiated & Will Be Incorporated Into Amend 14 ML20059C1741990-08-28028 August 1990 Forwards Rev 31 to Security Plan.Rev Withheld (Ref 10CFR73.21) ML20059D5601990-08-28028 August 1990 Forwards Fitness for Duty Performance Rept for 900103-0630. Overall Positive Rate for All Testing Conducted at Plant from 900103-0630 Is 0.71% ML20059D3201990-08-28028 August 1990 Responds to Deficiencies Identified by FEMA During 891115 Emergency Preparedness Exercise at Plant.Corrective Actions: Procedures Revised to Prioritize Roadblocks & Clearly Define Roadblocks for 5-mile & 10-mile Locations ML20058M7861990-08-0303 August 1990 Forwards Monthly Operating Rept for Jul 1990.W/o Encl ML20056A1551990-07-31031 July 1990 Forwards Updated Response to Address Maint/Surveillance Area of SALP for Period from Jan 1989 - Mar 1990.Overall Plant Performance During Period Acceptable & Found to Be Directed Toward Safe Operation ML20055J3421990-07-26026 July 1990 Forwards Decommissioning Financial Assurance Certification Rept for Plant,Per 10CFR50.75.Certifies That Finanical Assurance Utilizing Methodology as Listed Will Be Available to Decommission Plant ML20055G7131990-07-20020 July 1990 Comments on Operator Licensing Exams Administered on 900717 ML20055G7141990-07-19019 July 1990 Responds to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount. Plant Has Four Rosemount Model 1153DD6 Differential Pressure Transmitters Installed to Associated Logics for Safety Injection ML20055F8791990-07-16016 July 1990 Forwards 1989 Annual Repts for Portland General Corp,Eugene Water & Electric Board & Pacific Power & Light Co ML20055F8161990-07-13013 July 1990 Advises of New Addressee for Nrc/Util Correspondence ML20055F6911990-07-13013 July 1990 Responds to Generic Ltr 90-04, Request for Info on Status of Implementation of Generic Safety Issues Resolved W/ Imposition of Requirements or Corrective Actions ML20055F2441990-07-12012 July 1990 Forwards Response to SALP Evaluation for Jan 1989 - Mar 1990 Re Maint/Surveillance.Util Plans to Improve Performance in Areas of Operation Through Commitment of Support from Nuclear Div Mgt & Personal Involvement of Corporate Mgt ML20055F2961990-07-0606 July 1990 Forwards Addl Info Re ATWS Mitigating Sys Actuation Circuitry ML20055E0241990-07-0606 July 1990 Responds to NRC Re Violations Noted in Insp Rept 50-344/90-11.Corrective Actions:Generic Issues That Contributed to Violation Will Be Discussed During Mgt Meetings & Quality Insp Supervisor Counseled ML20055E0281990-07-0606 July 1990 Informs That Actions Committed to in Util Re Replacement of Breaker During 1990 Refueling Outage & Provision of Rept Documenting Replacement of Breaker, Complete,Per NRC Bulletin 88-010 ML20055E3401990-07-0606 July 1990 Forwards Application for Renewal of Plant NPDES Permit ML20055E1751990-07-0606 July 1990 Provides Update to 891130 Response to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment. Rev to Maint Procedure 12-4, Valve Air Actuators, to Include Maint Insp Completed in Apr 1990 ML20058K4891990-06-30030 June 1990 Discusses Reduced Min Measured RCS Flow for Plant.Encl Figure 3.2-3 & Evaluation Support Operation W/No Tech Spec Setpoint Changes ML20055D0441990-06-29029 June 1990 Advises That Response to Generic Ltr 90-04 Re Status of Generic Safety Issues at Facility Delayed from 900629 to 900713 ML20055D0411990-06-29029 June 1990 Discusses Design Verification Program Re Main Steam Support Structure.Extension of Completion Date for Project Justified Since Documentation,Not Mod of Main Steam Support Structure Is Reason for Delay ML20055D7091990-06-26026 June 1990 Forwards Steam Generator Tube Plugging Rept,Per Inservice Insp ML20055C2691990-02-20020 February 1990 Responds to NRC Re Violations Noted in Insp Rept 50-344/89-30.Corrective Actions:Administrative Order 5-8 Temporary Mods to Be Evaluated for Need to Improve Control Over Temporary Mods Issued for Both Trains of Sys ML17311A1041989-10-30030 October 1989 Advises That Commitment Date for Performance of Surveillance Moved to 891130 to Allow for Incorporation in Biennial Audit Per Rev 1 to LER 89-08.NRC Resident Inspector Informed of Change on 891013 ML20248G7631989-10-0606 October 1989 Forwards Description of Alternate Compensatory Measures Implemented During Upgrade Activities.W/O Encl ML20248G3131989-10-0303 October 1989 Forwards Response to Generic Ltr 89-04, Guidance on Developing Acceptable Inservice Testing Programs. Revised Topical Rept PGE-1048 Will Be Submitted by 891117 ML20248E2391989-09-29029 September 1989 Forwards Summary of Action Plan to Resolve Fuel Bldg Overstress Conditions ML20248D3711989-09-29029 September 1989 Forwards Rept of Util Plan for Improving Performance of Nuclear Div Staff.New Mgt Described in Plan Will Serve as Catalyst for Decisive Performance Changes ML20248G9001989-09-29029 September 1989 Forwards Rev 1 to PGE-1049, Inservice Insp Program for Second 10-Yr Interval from 860520-960519 ML20248D6161989-09-28028 September 1989 Requests 14-day Extension Until 890928 to Respond to Violations Noted in Insp Rept 50-344/89-09.Individuals Involved in Providing Input to Response Currently Assigned to Solving Problems During Current Forced Outage ML20248G1071989-09-28028 September 1989 Advises That Util Will Submit Final Closeout of NRC Bulletin 88-004, Potential Safety-Related Pump Losss, by 900115 05000344/LER-1989-007, Informs That Rev 1 to LER 89-07 Will Be Submitted by 891013, Due to Forced Outage & Difficulties1989-09-28028 September 1989 Informs That Rev 1 to LER 89-07 Will Be Submitted by 891013, Due to Forced Outage & Difficulties ML20248D4211989-09-27027 September 1989 Suppls 881004 Response to NRC Bulletin 88-008, Thermal Stresses in Piping Connected to Rcs. NDE Confirmed That No Existing Flaws Exist in Critical Piping Locations Re Thermal Stratification ML20247G9011989-09-15015 September 1989 Responds to Generic Ltr 89-06 Re SPDS Requirements,Per NUREG-0737,Suppl 1 as Clarified by NUREG-1342.Concludes That Isolators Used Were Acceptable for Interfacing SPDS W/Safety Sys,Therefore Util Does Not Intend to Make Mods to SPDS ML20247C2801989-09-0808 September 1989 Responds to NRC Re Violations Noted in Insp Rept 50-344/89-18.Corrective Action:Reevaluation of OAR 87-29, Involving Mods to Hydrogen Gas Supply Sys Implemented as Part of 1989 Refueling Outage 05000344/LER-1989-012, Advises That Rev 1 to LER 89-012 Will Be Submitted by 890920 Due to Extension of 1989 Refueling Outage & Plant Trip of 8908091989-09-0808 September 1989 Advises That Rev 1 to LER 89-012 Will Be Submitted by 890920 Due to Extension of 1989 Refueling Outage & Plant Trip of 890809 ML20247B9281989-09-0808 September 1989 Responds to NRC Re Violations Noted in Insp Rept 50-344/89-12 & Forwards Review of Closeout of First 10-yr Inservice Insp Interval.Concurs That Primary Reason for Individual Problems Was Lack of Insp Contractor Activities ML20247E9371989-09-0707 September 1989 Discusses Nuclear Div Improvement Plan,Per Recent Event Re Containment Recirculation Sump.Util Corrective Actions, Including Organizational Changes,Encl.Plant Improvement Plan Will Be Submitted by 891001 ML20247A6171989-09-0505 September 1989 Informs That Intake Structure Mod Completed as Scheduled & Security Plan Rev Will Be Submitted by 891015 ML20246M5181989-08-31031 August 1989 Forwards Update on Action Plan Developed in Response to Generic Ltr 88-14, Instrument Air Supply Sys Problems Affecting Safety-Related Equipment ML20246M5771989-08-31031 August 1989 Forwards Status of Progress in Achieving Improvement Goals Identified in 890515 SALP Response Re Safety Assessment/ Quality Verification.Addl Resources Applied to Qualify commercial-grade Suppliers ML20246M5651989-08-31031 August 1989 Discusses long-term Pipe Support Design Verification Program for Plant.Evaluation of Fuel Bldg Confirmed & Quantified Code Overstresses Existing Under Seismic & Tornado Loading Conditions.Details of Action Plan Will Be Sent by 890929 1990-09-07
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=_= Portland General ElectricCompany David W. Cockfield Vice President. Nuclear July 10, 1987
/
Trojan Nucicar Plant Docket 50-344 License NPF-1 l
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555
Dear Sir:
Support Design Verification Pursuant to the Nuclear Regulatory Commission (NRC) Ictter of July 9, 1987 and in accordance with Part 50.54(f) of Title 10 of the Code of Federal Regulations, this letter is provided to address the NRC concerns regarding the safety-related supports at the Trojan Nuclear Plant. The NRC letter expressed concerns regarding the original design, the derivation of the original design loads, and original design calculations for Main Steam )
System supports designed by the Architect-Engineer (A-E) for Trojan. The j NRC concerns also extend to other safety-related system supports designed j by Trojan's A-E. Portland Conoral Electric Company (PGE) had similar concerns and has directed a thorough review of this issue. A description .
of this issue, the scope of the review performed, and the results of the review are provided below.
During the 1987 refueling outago, a discrepancy was noted when main steam line support EDB-1-1-SS-81 (SS-81) was inspected. The discrepancy con-sisted of separation betwoon the baseplato and the baseplato grout, and betwoon the grout and the concrete wall in several locations. The dis-crepancy was ovaluated by the PCE Civil Engineering Branch of Nuclear Plant Engineering. During their ovaluation of the condition, design criteria for the support were reviewed and it was determined that the design of the support anchorage was inadequate for the specified dynamic load.
This problem has been identified, investigated, and is being resolved. An action plan was developed by PGE and the Trojan A-E to verify the design of affected supports. A description of this action plan is provided in Attachment 1. A total of 479 supports have been reviewed and evaluated (the verification of 15 pipe anchors is still in progress). It was determined 10 supports, for which the A-E Civil Engineering Group was responsible, were inadequately designed for the originally speciflod bounding dynamic loads (loads due to turbino trip from 100 porcent power).
These supports are all on the Main Steam System and were all designed and installed late in construction in 1975 to account for the turbine trip dynamic loads. This design problem is limited to the A-E Civil Engincoring Group and specifically to the Main Steam System designs performed in 1975.
Although these supports did not meet the original design criteria., they have functioned acceptably under numerous actual loading conditions from turbino trips from full power. k 8707150530 870710 k \
l PDR ADOCK 05000344 k P PDR l 121 S w. damon weet mrta% Oregon 97204
Portland General ElectricCorpsy Document Control Desk July 10, 1987 Page 2 To ensure proper implementation of the verification action plan, a PGE
- management team was sont to the A-E's offices to perform a quality 1 assurance surveillance. The team was hesded by the Panager of the Nuclear -l Quality Assurance Department and included the Branch Manager of Mechanical !
Engineering and the Branch Manager of Nuclear Regulation. The objectives of the surveillance were to verify the A-E had adequately defined the problem, had identified and was taking appropriate corrective action, and that the documentation for the corrective action program was in place and was being complied with. The team concluded these objectives were being met.
On June 22 and 23, PGE and the A-E met with three NRC reviewers at the A-E I offices. The NRC reviewers raised 10 concerns to which the A-E and PGE l have responded. These concerns and responses are provided as' Attachment 2.
I The investigation of this issue has been comprehensive end thorough. PGE l
has been carefully monitoring the A-E's progress and results and is con-vinced the support verification plan has been thorough and has been cor-rectly performed. The verification plan constitutes a short-term action to demonstrate the acceptability of supports to perform their safety-related function. In the long-term, PGE intends to verify designs for other. safety-related supports. The scope of work and schedule for completing-this review is under evaluation and will be determined by September 1, 1987. l The design verification for 15 pipe anchors in Containment is not expected i to be complete until July 14, 1987. Plant startup will not commence until- l these 15 verifications are completed and are determined to be satisfactory. !
Upon completion of this verification action plan, PGE has reasonable f assurance that safety-related supports will be properly designed and the Trojan Nuclear Plant will be safe to return to operation.
Sincernly, f 4 Attachments c: Mr. John B. Martin .;
Regional Administrator, Region V l U.S. Nuclear Regulatory Commission Mr. R. C. Barr NRC Resider.t Inspector .
Trojan Nucioar Plant Mr. David Kish, Director 1
tate of Oregon j . k. . .artment of Energy l l
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- v. C N Subscribed and sworn to before me this,10th day of July 1987.
I b " : dri fe / ? X J' '
b f' Notary Public of Ore'gon '
63r tv My Commission Expires: 4fhaf y
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i Trojan Nucicar plant Document Control Desk July 10, 1987 j Docket 50-344 Attachment 1 l License NpF-1 page 1 of 3 l l
j SUpp0RT VERIFICATION ACTION PLAN i
i j
To investigate and resolve the design load deficiency for SS-81, the j Architect-Engineer (A-E) was contacted. The A-E could not determine if the j turbine trip dynamic load of 83.2 kips had been properly included in the l
l sup- port design. The A-E's investigation determined the design of this t support was the responsibility of their Civil Engineering Group, l t'
D, l
The A-E, under direction from pGE, took action to identify the root cause l )
of the problem and to develop an action plan to determine the extent of thia j problem. The A-E determined the deficiency was isolated to their Civil '
Engineering Group and proposed an actinn plan to review all the supports 1
for which the Civil Engineering Group was responsible in addition to the Main Steam System safety-related supports. The review. included the u t
64 safety-related supports in the Main Steam System, 33 safety-related pipe I anchors, 265 safety-related pipe snubber anchorages and structural members l end five non-safety-related pipe support structures containing aptotal of l 41 individual supports. While the original design documentation (for these supporth was incomplete, the verification plan developed documen'tstion j demonstkwting the adequacy of the as-built configuration based on the support design criteria. <
, \
Since the deficiency for SS-81 was a failure to account for the dynamic l load in the support design, PGE expanded the scope of the program to include an analysis of all 76 sa gty-related supports.for other cystems j for which dynamic load analysesi had been performed. These supports were These additional supports were reviewed l l ,dysigned by a pCE subcontractor.
. itol confirm that the deficiency was limited to the Civil Engineering Group.
Thusf the plan encompassed a total of 479 supports. The results of the l
l verification plan to date are described below:
s
- a. Main Steam System Supports. l Of the 64 safety-related main steam supports, 34 were designed entirely ,]
by a pGE subcontractor, and 30 were jointly designed by the XI subcontractor (hardware) andL 3 ha A-E (anchorage and structure) .
Fifty-fourofthesesupportqhereverifiedtobeacceptabicas-is'.
l
' The other 10 supports were determined to be inadequately designed for. !
the originally specified dydanic loads. The original loads were based l on a conservatively applied pressure ramp taken from a turbine genara-l l
' tor similar to Trojan's. This method of deriving the original dynamic I loads wds unnecessarily conservative, but was standard practice in the \ ,d
)
early 1970s. During the verification effort, the A-E reanalyzed the s dynamic loads for the Main Steam System supports insids Containment l based on actual turbine trip data from Trojan startup testing, and the .
supports were reevaluated using the new dynamic loads. Three of the l
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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 Licanse NPF-1 Attachment 1 Page 2 of 3 l 10 supports were verified to be acceptable without modification based on the revised dynamic loads. Seven of the supports were modified by PGE utilizing the revised dynamic loads in the design process. It is likely the original design of these supports could have performed satisfactorily for the revised loads but PGE felt insufficient design j margin would have existed. In conclusion, the Main Steam System j supports meet the design bases for the system,
- b. Civil-DesiRned Pipe Anchors. J The A-E evaluated the design of all safety-related civil-designed pipe anchors (33). These supports are on the Safety Injection, Chemical and Volume Control, Reactor Coolant Loop Vent / Drains, Pressurizer Spray, Residual Heat Removal, and Reactor Coolant Pump Seal Water Systems. To date, 18 of these supports have been verified to be satisfactory. The i verification of the design of the remaining 15 is still in progress.
- c. CivC Designed Pipe Snubber Anchorages.
The A-E verified the design of cll safety-related civil-designed pipe I snubber anchorages (265).
These supports are on the chemical and
l Volume Control, Component Cooling Water, Pressurizer Relitf and Snfety, Pressurizer Spray, Reactor Coolant Instrumentation, Reactor Coolant Pump Seal Water, Reactor Coolant System Drain, Realdual Heat Removal, Safety Injection, and Steam Generator Blowdown and Ssmple Systems.
These 265 supports were verified to be satisfactorily designed.
- d. Civil-Designed Pipe Support Structures The A-E verified the designs for five non-safety-related civil-designed pipe support structures. Two of these structures provide support for mois< tre separator reheater relief valve discharge lines, one for the main steam bypass line, and two for main steam stop valves. These five structures contain a total of 41 individual pipe supports and have been verified to be acceptable,
- e. Supports With Dynamic Load Calculations (not des 1Rned by Civil.
Engineering Group).
The A-E identified 229 supports on systems for which dynamic load cal-culations had been performed. Of these, 153 are not safety-related and ~,6 are safety-related. The designs for all 76 safety-related sup-ports were verified to be satisfactory. These supports are on the pressurizer relief valve discharge line and the main steam lines to the turbine-driven auxiliary feodwater pump.
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Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 License NPF-1 Attachment 2 page 1 of 7 NRC CONCERNS ON SUpp0RT DESIGN ISSUE
- 1. Description of Concern What is considered as "dcoign verification" by the Architect-Engineer (A-E) for 50 safety-related main steam supports designed by the subcontractor. ]
Licensee's Response The group of 50 safety-related main steam supports included designs solely by the subcontractor, combination designs by the A-E civil and the subcontractor, and designs by the A-E' civil only. The 50 safety-related main steam supports were originally designed based upon loads calculated by the A-E.
The A-E's current " design verification" effort (1987) of these supports consisted of a review and reconciliation. New dynamic loads had been calculated in June 1987 based upon actual turbine trip test data, more recent regulatory guidance, and improved modeling techniques. The new design loads for the supports were compared with the original design loads. If the new design load was less than or equal to original load, and if the structure appeared adequate for the new design load, the support was considered acceptable. If the new design load was larger than the original design load, a structural evaluation was performed by the A-E to qualify the support. This evaluation consisted of engineering judgement for small load increases or a detailed calculation for larger load increases. The evaluations addressed the following items:
- a. Bolts,
- b. Base plates,
- c. Wolds,
- d. Member sizos, and
- e. Components. )
- 2. Description of Concern What is considered as " design verification" by the A-E for 14 safety-related main steam supports inside Containment designed by the A-E Civil Croup.
Trojan Nuclear plant Document Control Desk i Docket 50-344 July 10, 1987
.j License NpF-1 Attachment 2 l page 2 of 7
,icensee's L Response 1
In 1975, 14 safety-related main steam line supports inside_Contain-ment were added to the original design for the dynamic load result- 1 ing from a turbine trip. These 14 supports were added after the design of the other 50 main cteam line supports. These 14 supports were designed by the A-E Civil Group and were apparently based upon )
design calculations that had not been performed (see Concern No. 3). )
1 Design verification for these 14 safety-related main steam line !
supports inside the Containment consisted of the following:
- a. New design loads were calculated for the turbine-trip event (see response to Concerns No.1 and 6) . 1
- b. New calculations were performed based upon the new design loads I to evaluate the structural adequacy of the supports. The !
calculations considered the following items:
)
1 (1) Bolts, J l
(2) Base plates, (3) Welds, (4) Member sizes.
- c. The support calculations will be documented, checked, and approved in accordance with the existing A-E procedures and the A-E Quality Assurance program. !
l
- 3. Description ef Concern j l
Why did the original calculations of supports designed by the A-E Civil Group contain only stiffness calculations but not stress l (strength) calculations?
Licensee's Response l The investigation,has not revealed why only stiffness calculations can be found for certain of the 14 main steam line supports designed f by the A-E Civil Group. Due to the time interval since the stiff-l ness calculations were performed, it is not expected it will ever be known why the structural calculations were not performed.
However, the A-E is performing a review of pipe-support anchorages and structural members to verify design adequacy and to assist in l'
1
1 1
-Trojan Nuclear plant Document Control Desk l Docket 50-344 July 10, 1987 )
License NPF-1 Attachment 2 )
page 3 of 7 )
)
q root cause determination. The review is being conducted on the following designs l l
- a. All safety-related main steam line pipe supports (" design ]
verification"), o
- b. All of the civil-designed pipe anchors inside Containment
(" design verification").
1
- c. All of the civil-designed pipe snubber anchorage and structurel membero located inside Containment. l
- d. A sample of the civil-designed pipe support structures in the i ma.4.n steam and main steam relief systems.in the Turbine Building.
- e. A sample of all pipe support calculations with dynamic loadings.
Calculations in progress to date do not indicate that any ., ember would have failed. For the main steam line supports this coaclusion is supported by satisfactory performance under actual load {
conditions.
- 4. Description of Concern Why do the four main steam supports need design modification? Any l changes in loads or acceptance criteria? What are the modifications? {
Licensco's Response The four main steam supports are EBB-1-1-SS-81 (B loop),
EBB-1-1-SS-86 (A loop), EBB-1-2-SS-88 (C loop), and EBB-1-2-SS-92 I (D loop). TF;. four structures are used to support the large hydraulic sr.ibi..rs for the main steam lines.
Initially, it was determined that the four main steam supports q required modifications to meet the original design loads. However, 4 upon review of the original design loeds, it was determined that they were inappropriate and overly conservative and new design loads were calculated (sco responses to Concerns 1, 2, and 6). The four main steam supports were not able to meet the newly calculated i design loads within the original criteria limits. Therefore, I modifications were made to restore the intended margins.
1 The design loads were changed by recent calculations performed in 1987. The dynamic loads were reduced by a factor of approximately i 3.1 for SS-81 and SS-86, and by a factor of approximately 1.6 for j SS-88 and SS-92, as a result of the dynamic analysis performed in i l
l l
I
5 f
. I q
Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987-d 1
Licence NpF-1 Attachment 2 page 4 of 7 I
June 1987. Although the seismic loads showed a small general l increase from a seismic reanalysis performed earlier in the year, I l
the largest changes (reductions) in support loads were still attributable to the dynamic. effects due to the turbine trip transient analysis, l
The acceptance criteria for load combinations and allowable stress limits in structural steel were not changed. Due to its availabil-ity and the fact that tha old drill holes could be used, a dif-ferent type of anchor bolt was selected for the support redesign.
ASTM A-193 grouted threaded rods were used in the support anchorage redesign in lieu of the grouted rock anchors that were originally q used. The acceptance criteria for the grouted threaded rods were j based on current capacity development guidance for this type of anchorage.
The following is a specific description of the support modifications:
1
- a. Supports SS-81, SS-88, and SS-92. l The anchor bolts were changed from grouted 1 inch diameter rock anchors to grouted 1-1/4 inch diameter ASTM A-193 threaded rods ;
in order to rostore anchorage capacity margins. The 1 inch j thick steel base plate was replaced with a 1-3/4 inch thick base l plate to increase stiffness and capacity. The sinubber support l was changed from a gusset plate stiffened 12 UF 36 bracket assembly to a 10 x 10 x 1/2 inch thick structural tube braced frame to provide overall increased capacity and reduce localized l stresses. New grout was placed between the base plate and concrete wall,
- b. Support SS-86.
A new frame made up of a 4 x 12 x 1/2 inch structural tube column and beam war. connected to the existing stiffened 12 WF 36 snubber support to redistribute a portion of the load from the
( existing support to the new frame, thereby substantially reducing the loads on the existing anchcr bolts. The new frame also significantly increased torsional resistance of the existing snubber support beam. The new column 3/4 it.ch thick steel base plate was anchored with 3/4 inch diameter ASTM A-193 grouted threaded rods. New base plate grout was placed where necessary.
- 5. Description of Concern
- Describe status and findings of the A-E verification review of 50 safety-related main steam supports designed by the subcontractor.
Document Control Desk .j Trojan Nuclear Plant July 10, 1987 i Docket 50-344 License NpF-1, Attachment 2 l
)
page 5 of 7 Licensee's Response I The verification, of the 50 safety-related main steam line supports is complete; no discrepancies or findings were noted. All supports were found to be acceptable. (See response to Concern No. 1).
- 6. Description of Concern Mcw were the dynamic loads related to turbine trip defined in the original design of he main steam lines. What is the AE's.conclu-sion on verification of this load and its effects on support design.
Licensee's Response The original design turbine trip dynamic loads, for the design of l the main steam piping and the 14 supports added in 1975, resulted l from a force-time history analysis. This analysis, which was com-pleted in 1975, utilized the pressure transient data from a test cf a similar General Electric turbine at another nucinar power plant.
The analysis (original dynamic load) was validated by testing at Trojan and found to be conservative (eg, for main steam support SR-86, the original design load was 83.2 kips and the measured value was 22.4 kips). The differences between the original calculated design loads and the test data were attributed to the differences between the nuclear power plant from which the original test data was obtained (a Boiling Water Reactor) and Trojan (a pressurized Water Reactor). The test data were always less than the calculated loads and the amount of conservatism was never questioned.
Subsequently, it has been determined that this analysis was overly l
conservative. The main steam piping inside Containment was recently I reanalyzed to calculate more realistic design loads. The new cal-culation was based upon the data from the Trojan 100 percent power turbine trip tests, more recent modeling techniques, and regulatory guidelines. The predicted turbine trip dynamic loads on all main steam supports inside Containment were reduced by this reanalysis.
- 7. Description of concern The A-E analysis for support SS-38 used a nonlinear interaction formula for combining shear and tension. The manufacturer recommends a linear interaction formula. Why did the A-E use I rather than T + V s 1?
I T I +I V i1 T (VALL, TALL VALL GALL /
1 Trojan Nuclear Plant Document Control Desk-Docket 50-344 July 10, 1987 ,
License NPF-1 Attachment 2 l Page 6 of 7 .
l I
Licensee's Response ]
The nonlinear interaction formula is an appropriate representation of shear-tension effects as demonstrated by test data. This formula was utilized for the evaluation resulting from IE Bulletin 79-02, regarding Seismic' Category I pipe support base plates which use f expansion anchors. A report of this evaluation (using the formula '
in question on Page 4) was transmitted to the Nuclear Regulatory Commission via a letter from Mr. D. J. Brochl, PGE, to Mr. R. H. j Engelken, NRC dated November 21, 1979. The NRC closed-out PGE action on IE Bulletin 79-02 in Inspection Report 86-05, dated j April 14, 1986. .
- 8. Description of Concern i PGE or the A-E must determine what the allowable loads are for the Williams' groutable (Hollow-Core) rock bolts.
Licensee's Response j l
The criteria for the design of the grouted-in rock anchors are:
- a. Calculate the maximum tension load to be resisted by the rock anchor. ]
l
- b. Determine the embedment required to develop the tension load {
determined in Item a. above, based on the ultimate pull-out ]
capacity of concrete calculated, based on Paragraphs 1504 and )
1707 of ACI Code 318-63. l l
- c. Select not-cross sectional area of rock bolt by limiting the j stress on the net section of 0.9 fy. ]
i
- 9. Description of Concern 1
Snubber angularity, due to tolerance in installation, should be i considered b the design calculations for supports. !
Need to provide justification for not including.
Licensee's Response Y
The Trojan Nuclear Plant pipe support design criteria (DC-11760-003, Rev. 0) limits strut and snubber offsets to the lesser of 5' or 2 inches. The design practice of the industry during the era of Trojan Nuclear Plant design did not consider these effects. The 5*
[
tolerance would result in a maximum potential out of plane loading I
i e
l
.- j I
Trojan Nuclear Plant Document Control Desk Docket 50-344 July 10, 1987 License RPF-1 Attachment 2 Page 7 of'7 of 8 percent of the design load, and generally has a negligible effect on the structure.
- 10. Description of Concern 1
Calculations for the 265 snubbers inside Containment did not always .l contain enough information for a reviewer to determine'what the engineer / checker was accepting by comparing to a given standard j{
(example, a structure was accepted without any calculations or ]
discussion when lengths were longer and the member's sizes were smr11er). l Licensee's Response Further documentation will be performed of the engineering judgement made during review of the 265 snubber supports inside Containment.
This documentation will be added to the'A-E's calculation files.
The following were considered in the review:
- a. Supports which are similar to one of the generic supports but i differ in some way (ie, member sizes, anchorage, welds, loads, lengths) are evaluated in the following fashion: i (1) Differences between the subject support configuration / load and the similar generic support are stated.
(2) The reviewer determined the effects of the differences based on judgement or based on a calculation. The reviewer then documented his conclusion,
- b. pipe supports that are not similar to the generic type supports are evaluated in the following manner:
(1) If the load is small and the support configuration is simple, the support is reviewed by judgement after a review of the ana,horage, base plate, and structural members. The conclusion is documented by a statement.
(2) If a load is substantia) or support configuration is complex, a calculation is performed on critical structural items.
TDW/kal 1426P.687
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