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Category:CORRESPONDENCE-LETTERS
MONTHYEARL-99-035, Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld1999-10-18018 October 1999 Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld ML20217G0801999-10-0707 October 1999 Informs That on 990930,staff Conducted mid-cycle PPR of Farley & Did Not Identify Any Areas in Which Performance Warranted More than Core Insp Program.Nrc Will Conduct Regional Insps Associated with SG Removal & Installation ML20217P0661999-10-0606 October 1999 Requests Withholding of Proprietary Rept NSD-SAE-ESI-99-389, Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs ML20217B1891999-10-0404 October 1999 Submits Clarification Re Development of Basis for Determining Limiting Internal Pressure Loads Re Review of NRC SE for Cycle 16 Extension Request.Util Intends to Use Guidelines When Evaluating SG Tube Structural Integrity ML20212J8391999-09-30030 September 1999 Forwards RAI Re Request for Amends to Ts.Addl Info Needed to Complete Review to Verify That Proposed TS Are Consistent with & Validate Design Basis Analysis.Request Discussed with H Mahan on 990930.Info Needed within 10 Days of This Ltr ML20212J8801999-09-30030 September 1999 Discusses GL 98-01,suppl 1, Y2K Readiness of Computer Sys at Npps. Util 980731,990607 & 03 Ltrs Provided Requested Info in Subj Gl.Nrc Considers Subj GL to Be Closed for Unit 1 L-99-032, Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 21999-09-23023 September 1999 Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 2 L-99-034, Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 21999-09-23023 September 1999 Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 2 ML20212F8861999-09-23023 September 1999 Forwards Revised Relief Request Number 32 for NRC Approval. Approval Requested by 991231 to Support Activities to Be Performed During Unit 1 Refueling Outage Scheduled for Spring of 2000 ML20212E7031999-09-23023 September 1999 Responds to GL 98-01, Year 2000 Readiness of Computer Sys at Npps. Util Requested to Submit Plans & Schedules for Resolving Y2K-related Issues ML20212F1111999-09-21021 September 1999 Discusses Closeout of GL 97-06, Degradation of Steam Generator Internals ML20212C2351999-09-16016 September 1999 Submits Corrected Info Concerning Snoc Response to NRC GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal ML20212D0101999-09-15015 September 1999 Informs That Submittal of clean-typed Copy of ITS & ITS Bases Will Be Delayed.Delay Due to Need for Resolution of Two Issues Raised by NRC staff.Clean-typed Copy of ITS Will Be Submitted within 4 Wks Following Resolution of Issues ML20212C4641999-09-13013 September 1999 Forwards Info Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams L-99-031, Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed1999-09-13013 September 1999 Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed ML20212C8041999-09-10010 September 1999 Responds to to D Rathbun Requesting Review of J Sherman Re Y2K Compliance.Latest NRC Status Rept on Y2K Activities Encl ML20212D4581999-09-10010 September 1999 Responds to to D Rathbun,Requesting Review of J Sherman Expressing Concerns That Plant & Other Nuclear Plants Not Yet Y2K Compliant ML20212A6951999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20212A8341999-09-0909 September 1999 Requests That Licensees Affected by Kaowool Fire Barriers Take Issue on Voluntary Initiative & Propose Approach for Resolving Subj Issues.Staff Plans to Meet with Licensees to Discuss Listed Topics ML20211N8041999-09-0808 September 1999 Informs That on 990930 NRC Issued GL 96-06, Assurance of Equipment Operability & Containment Integrity During Design-Basis Accident Condition, to Holders of Nuclear Plant Operating Licenses ML20211N4301999-09-0808 September 1999 Discusses Proposed Meeting to Discuss Kaowool Fire Barriers. Staff Requesting That Affected Licensees Take Issue on Voluntary Initative & Propose Approach for Resolving Issues ML20212C0071999-09-0202 September 1999 Forwards Insp Repts 50-348/99-05 & 50-364/99-05 on 990627- 0807.No Violations Noted.Licensee Conduct of Activities at Farley Plant Facilities Generally Characterized by safety-conscious Operations & Sound Engineering ML20211Q4801999-09-0101 September 1999 Informs That on 990812-13,Region II Hosted Training Managers Conference on Recent Changes to Operator Licensing Program. List of Attendees,Copy of Slide Presentations & List of Questions Received from Participants Encl ML20211K2131999-08-31031 August 1999 Informs That Snoc Has Conducted Review of Reactor Vessel Integrity Database,Version 2 (RVID2) & Conclude That Latest Data Submitted for Farley Units Has Not Been Incorporated Into RVID2 ML20211K4101999-08-31031 August 1999 Resubmits Relief Requests Q1P16-RR-V-5 & Q2P16-RR-V-5 That Seek to Group V661 Valves from Each Unit Into Sample Disassembly & Insp Group,Per 990525 Telcon with NRC L-99-030, Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS1999-08-30030 August 1999 Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS ML20211G6851999-08-26026 August 1999 Informs That During Insp,Technical Issues Associated with Design,Installation & fire-resistive Performance of Kaowool Raceway fire-barriers Installed at Farley Nuclear Plant Were Identified L-99-029, Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 07271999-08-19019 August 1999 Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 0727 ML20211B9431999-08-17017 August 1999 Forwards Fitness for Duty Performance Data for six-month Reporting Period 990101-990630,IAW 10CFR26.71(d).Rept Covers Employees at Jm Farley Nuclear Plant & Southern Nuclear Corporate Headquarters ML20211B9211999-08-17017 August 1999 Responds to NRC Re Violations Noted in Insp Rept 50-348/99-09 & 50-364/99-09.Corrective Actions:Security Response Plan Was Revised to Address Vulnerabilities Identified During NRC Insp ML20210R5101999-08-12012 August 1999 Forwards Revised Page 6 to 990430 LAR to Operate Farley Nuclear Plant,Unit 1,for Cycle 16 Only,Based on risk- Informed Approach for Evaluation of SG Tube Structural Integrity,As Result of Staff Comments ML20212C8141999-08-0909 August 1999 Forwards Correspondence Received from Jm Sherman.Requests Review of Info Re Established Policies & Procedures ML20210T2021999-08-0606 August 1999 Forwards Draft SE Accepting Licensee Proposed Conversion of Plant,Units 1 & 2 Current TSs to Its.Its Based on Listed Documents ML20210Q4641999-08-0505 August 1999 Informs That NRC Plans to Administer Gfes of Written Operator Licensing Exam on 991006.Authorized Representative of Facility Must Submit Ltr to La Reyes,As Listed,With List of Individuals to Take exam,30 Days Before Exam Date ML20210J8341999-07-30030 July 1999 Forwards Second Request for Addl Info Re Util 990430 Amend Request to Allow Util to Operate Unit 1,for Cycle 16 Based on risk-informed Probability of SG Tube Rupture & Nominal accident-induced primary-to-second Leakage ML20210G4901999-07-30030 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, Issued 990603.Ltr Contains NRC License Commitment to Utilize ASTM D3803-1989 with Efficiency Acceptance Criteria Utilizing Safety Factor of 2 L-99-028, Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines1999-07-30030 July 1999 Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines L-99-027, Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.51999-07-27027 July 1999 Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.5 ML20210G8181999-07-26026 July 1999 Forwards Insp Repts 50-348/99-04 & 50-364/99-04 on 990516- 0626.One Violation Identified & Being Treated as Noncited Violation IR 05000348/19990091999-07-23023 July 1999 Discusses Insp Repts 50-348/99-09 & 50-364/99-09 on 990308- 10 & Forwards Notice of Violation Re Failure to Intercept Adversary During Drills,Contrary to 10CFR73 & Physical Security Plan Requirements ML20210E4071999-07-22022 July 1999 Responds to NRC 990702 RAI Re Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described in NEI 97-06, SG Program Guidelines L-99-026, Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments1999-07-19019 July 1999 Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments L-99-264, Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 20011999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 ML20209H4721999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 ML20196J6191999-07-0202 July 1999 Forwards Final Dam Audit Rept of 981008 of Category 1 Cooling Water Storage Pond Dam.Requests Response within 120 Days of Date of Ltr 05000364/LER-1999-001, Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed1999-07-0202 July 1999 Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed ML20196J7471999-07-0202 July 1999 Forwards RAI Re Cycle 16 Extension Request.Response Requested within 30 Days of Date of Ltr ML20196J5781999-07-0202 July 1999 Forwards RAI Re 981201 & s Requesting Amend to TS Associated with Replacing Existing Westinghouse Model 51 SG with Westinghouse Model 54F Generators.Respond within 30 Days of Ltr Date ML20196J6571999-07-0202 July 1999 Discusses Closure to TAC MA0543 & MA0544 Re GL 92-01 Rev 1, Suppl 1,RV Structural Integrity.Nrc Has Revised Rvid & Releasing It as Rvid,Version 2 as Result of Review of Responses ML20196J3591999-06-30030 June 1999 Forwards SE of TR WCAP-14750, RCS Flow Verification Using Elbow Taps at Westinghouse 3-Loop Pwrs 1999-09-09
[Table view] Category:INCOMING CORRESPONDENCE
MONTHYEARL-99-035, Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld1999-10-18018 October 1999 Forwards non-proprietary & Proprietary Versions of Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs, Used to Support SG Replacement Project.Proprietary Encl Withheld ML20217P0661999-10-0606 October 1999 Requests Withholding of Proprietary Rept NSD-SAE-ESI-99-389, Farley Units 1 & 2 LBB Calculation Results Due to SG Replacement & SG Snubber Elimination Programs ML20217B1891999-10-0404 October 1999 Submits Clarification Re Development of Basis for Determining Limiting Internal Pressure Loads Re Review of NRC SE for Cycle 16 Extension Request.Util Intends to Use Guidelines When Evaluating SG Tube Structural Integrity L-99-034, Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 21999-09-23023 September 1999 Forwards Comments on Draft Current Tech Specs Discussion of Change Tables for Jm Farley Nuclear Plant.Units 1 & 2 L-99-032, Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 21999-09-23023 September 1999 Responds to NRC Re Adequacy of Kaowool Fire Retardant Fire Barriers in Use at Jfnp,Units 1 & 2 ML20212F8861999-09-23023 September 1999 Forwards Revised Relief Request Number 32 for NRC Approval. Approval Requested by 991231 to Support Activities to Be Performed During Unit 1 Refueling Outage Scheduled for Spring of 2000 ML20212C2351999-09-16016 September 1999 Submits Corrected Info Concerning Snoc Response to NRC GL 99-02, Lab Testing of Nuclear-Grade Activated Charcoal ML20212D0101999-09-15015 September 1999 Informs That Submittal of clean-typed Copy of ITS & ITS Bases Will Be Delayed.Delay Due to Need for Resolution of Two Issues Raised by NRC staff.Clean-typed Copy of ITS Will Be Submitted within 4 Wks Following Resolution of Issues L-99-031, Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed1999-09-13013 September 1999 Informs NRC That Review of MOV Testing Frequency & Changes Made to Frequency of MOV Testing Has Been Completed ML20212C4641999-09-13013 September 1999 Forwards Info Requested in Administrative Ltr 99-03, Preparation & Scheduling of Operator Licensing Exams ML20211K2131999-08-31031 August 1999 Informs That Snoc Has Conducted Review of Reactor Vessel Integrity Database,Version 2 (RVID2) & Conclude That Latest Data Submitted for Farley Units Has Not Been Incorporated Into RVID2 ML20211K4101999-08-31031 August 1999 Resubmits Relief Requests Q1P16-RR-V-5 & Q2P16-RR-V-5 That Seek to Group V661 Valves from Each Unit Into Sample Disassembly & Insp Group,Per 990525 Telcon with NRC L-99-030, Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS1999-08-30030 August 1999 Forwards SNC Review Comments on Draft SE & marked-up Copy of Draft SE Incorporating SNC Comments Re Proposed Conversion to ITS L-99-029, Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 07271999-08-19019 August 1999 Forwards Revised Response to Chapter 3.1 RAI Requested in 990726 Conference Call,Rai Response Related to Beyond Scope Issue for Chapter 3.5 Requested by Conference Call on 990805 & RAI Response to Chapter 3.8 Requested on 990615 & 0727 ML20211B9431999-08-17017 August 1999 Forwards Fitness for Duty Performance Data for six-month Reporting Period 990101-990630,IAW 10CFR26.71(d).Rept Covers Employees at Jm Farley Nuclear Plant & Southern Nuclear Corporate Headquarters ML20211B9211999-08-17017 August 1999 Responds to NRC Re Violations Noted in Insp Rept 50-348/99-09 & 50-364/99-09.Corrective Actions:Security Response Plan Was Revised to Address Vulnerabilities Identified During NRC Insp ML20210R5101999-08-12012 August 1999 Forwards Revised Page 6 to 990430 LAR to Operate Farley Nuclear Plant,Unit 1,for Cycle 16 Only,Based on risk- Informed Approach for Evaluation of SG Tube Structural Integrity,As Result of Staff Comments ML20212C8141999-08-0909 August 1999 Forwards Correspondence Received from Jm Sherman.Requests Review of Info Re Established Policies & Procedures ML20210G4901999-07-30030 July 1999 Responds to GL 99-02, Laboratory Testing of Nuclear-Grade Activated Charcoal, Issued 990603.Ltr Contains NRC License Commitment to Utilize ASTM D3803-1989 with Efficiency Acceptance Criteria Utilizing Safety Factor of 2 L-99-028, Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines1999-07-30030 July 1999 Responds to NRC 990730 RAI Re 990423 OL Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described by NEI 97-06, SG Program Guidelines L-99-027, Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.51999-07-27027 July 1999 Addresses Clarifications to Selected Responses to Chapter 3.8 RAI Requested in NRC Conference Call on 990624, Resolution of Open Issue Related to Containment Purge in Chapter 3.6 & Response Related to Chapter 3.5 ML20210E4071999-07-22022 July 1999 Responds to NRC 990702 RAI Re Change Request to Allow for Risk Informed Approach for Evaluation of SG Tube Structural Integrity as Described in NEI 97-06, SG Program Guidelines L-99-026, Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments1999-07-19019 July 1999 Forwards Response to NRC 990702 RAI Re SG Replacement Related TS Change Request Submitted 981201.Ltr Contains No New Commitments L-99-264, Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 20011999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 ML20209H4721999-07-13013 July 1999 Responds to NRC 990603 Administrative Ltr 99-02, Operating Licensing Action Estimates, for Fy 2000 & 2001 05000364/LER-1999-001, Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed1999-07-0202 July 1999 Forwards LER 99-001-00 Re Reactor Trip Due to Loss of Condenser Vacuum Steam Dump Drain Line Failure.Commitments Made by Licensee,Listed L-99-024, Responds to NRC RAI Re Conversion to ITS for Chapters 3.4, 3.5,3.6,3.7,3.9 & 5.0,per 990419-20 Meetings with NRC1999-06-30030 June 1999 Responds to NRC RAI Re Conversion to ITS for Chapters 3.4, 3.5,3.6,3.7,3.9 & 5.0,per 990419-20 Meetings with NRC L-99-025, Forwards Rev 2 to Jfnp Security plan,FNP-0-M-99,IAW 10CFR50.4(b)(4).Attachment 1 Contains Summary of Changes & Amended Security Plan Pages.Encl Withheld from Public Disclosure Per 10CFR73.211999-06-30030 June 1999 Forwards Rev 2 to Jfnp Security plan,FNP-0-M-99,IAW 10CFR50.4(b)(4).Attachment 1 Contains Summary of Changes & Amended Security Plan Pages.Encl Withheld from Public Disclosure Per 10CFR73.21 ML20196J8631999-06-30030 June 1999 Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA L-99-249, Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA1999-06-30030 June 1999 Submits Correction to Errors Contained in to NRC Re TS Changes Re Control Room,Penetration Room & Containment Purge Filtration Systems & Radiation Monitoring Instrumentation.Errors Do Not Require Rev of SA L-99-224, Submits Rev to Unit 2 SG Tube voltage-based Repair Criteria Data Rept.Ltr Contains No Commitments1999-06-0707 June 1999 Submits Rev to Unit 2 SG Tube voltage-based Repair Criteria Data Rept.Ltr Contains No Commitments ML20195F1731999-06-0707 June 1999 Forwards Proprietary & non-proprietary Responses to NRC RAIs Re W TR WCAP-14750, RCS Flow Verification Using Elbow Taps at W 3-Loop Pwrs. W Proprietary Notice,Affidavit & Copyright Notice,Encl.Proprietary Info Withheld L-99-217, Forwards Proprietary & non-proprietary Responses to NRC RAIs Re W TR WCAP-14750, RCS Flow Verification Using Elbow Taps at W 3-Loop Pwrs. W Proprietary Notice,Affidavit & Copyright Notice,Encl.Proprietary Info Withheld1999-06-0707 June 1999 Forwards Proprietary & non-proprietary Responses to NRC RAIs Re W TR WCAP-14750, RCS Flow Verification Using Elbow Taps at W 3-Loop Pwrs. W Proprietary Notice,Affidavit & Copyright Notice,Encl.Proprietary Info Withheld L-99-225, Responds to GL 98-01, Yr 2000 Readiness of Computer Sys at Nuclear Power Plants1999-06-0707 June 1999 Responds to GL 98-01, Yr 2000 Readiness of Computer Sys at Nuclear Power Plants ML20195F0621999-06-0707 June 1999 Submits Rev to Unit 2 SG Tube voltage-based Repair Criteria Data Rept.Ltr Contains No Commitments ML20195E9581999-06-0707 June 1999 Responds to GL 98-01, Yr 2000 Readiness of Computer Sys at Nuclear Power Plants ML20195C6941999-05-28028 May 1999 Forwards Response to NRC RAI Re GL 96-05 for Farley Nuclear Plant.Farley Is Committing to Implement Phase 3 of JOG Program L-99-021, Forwards Response to RAI Re Conversion to ITSs for Chapter 3.3.Attachment II Includes Proposed Revs to Previously Submitted LAR Re Rais,Grouped by RAI number.Clean-typed Copies of Affected ITS Pages Not Included1999-05-28028 May 1999 Forwards Response to RAI Re Conversion to ITSs for Chapter 3.3.Attachment II Includes Proposed Revs to Previously Submitted LAR Re Rais,Grouped by RAI number.Clean-typed Copies of Affected ITS Pages Not Included L-99-203, Forwards Response to NRC RAI Re GL 96-05 for Farley Nuclear Plant.Farley Is Committing to Implement Phase 3 of JOG Program1999-05-28028 May 1999 Forwards Response to NRC RAI Re GL 96-05 for Farley Nuclear Plant.Farley Is Committing to Implement Phase 3 of JOG Program ML20195F2101999-05-24024 May 1999 Requests That Farley Nuclear Plant Proprietary Responses to NRC RAI Re W WCAP-14750, RCS Flow Verification Using Elbow Taps at W 3-Loop Pwrs, Be Withheld from Public Disclosure Per 10CFR2.790 L-99-180, Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI1999-04-30030 April 1999 Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI ML20206F4321999-04-30030 April 1999 Forwards Responses to NRC RAI Questions for Chapter 3.8 of Ts.Proposed Revs to TS Previously Submitted with LAR Related to RAI L-99-017, Forwards Responses to NRC RAI Questions for Chapters 3.1, 3.2,3.5,3.7 & 3.9 of Ts.Attached Pages Include Proposed Revs Previously Submitted LAR to Rais,Grouped by Chapters & RAI Numbers1999-04-30030 April 1999 Forwards Responses to NRC RAI Questions for Chapters 3.1, 3.2,3.5,3.7 & 3.9 of Ts.Attached Pages Include Proposed Revs Previously Submitted LAR to Rais,Grouped by Chapters & RAI Numbers ML20206C8021999-04-26026 April 1999 Forwards 1998 Annual Rept, for Alabama Power Co.Encls Contain Financial Statements for 1998,unaudited Financial Statements for Quarter Ending 990331 & Cash Flow Projections for 990101-991231 05000348/LER-1998-007, Forwards SG-99-04-001, Farley-1:Final Cycle 16 Freespan ODSCC Operational Assessment, as Committed to in Licensee & LER 98-007-00.Util Is Revising Plant Administrative SG Operating Leakage Requirements as Listed1999-04-23023 April 1999 Forwards SG-99-04-001, Farley-1:Final Cycle 16 Freespan ODSCC Operational Assessment, as Committed to in Licensee & LER 98-007-00.Util Is Revising Plant Administrative SG Operating Leakage Requirements as Listed L-99-015, Forwards Rev 1 to Jfnp Security plan,FNP-O-M-99,resulting from Implementation of Biometrics Sys.Changes Incorporate Changes Previously Submitted to NRC as Rev 28 by Licensee .Encl Withheld,Per 10CFR73.211999-04-21021 April 1999 Forwards Rev 1 to Jfnp Security plan,FNP-O-M-99,resulting from Implementation of Biometrics Sys.Changes Incorporate Changes Previously Submitted to NRC as Rev 28 by Licensee .Encl Withheld,Per 10CFR73.21 ML20206B4391999-04-21021 April 1999 Forwards Corrected ITS Markup Pages to Replace Pages in 981201 License Amend Requests for SG Replacement L-99-172, Forwards FNP Annual Radioactive Effluent Release Rept for 1998, IAW TSs Sections 6.9.1.8 & 6.9.1.9.Changes to ODCM Revs 16,17 & 18 Are Encl,Iaw TS Section 6.14.21999-04-21021 April 1999 Forwards FNP Annual Radioactive Effluent Release Rept for 1998, IAW TSs Sections 6.9.1.8 & 6.9.1.9.Changes to ODCM Revs 16,17 & 18 Are Encl,Iaw TS Section 6.14.2 ML20205S9501999-04-21021 April 1999 Forwards FNP Annual Radioactive Effluent Release Rept for 1998, IAW TSs Sections 6.9.1.8 & 6.9.1.9.Changes to ODCM Revs 16,17 & 18 Are Encl,Iaw TS Section 6.14.2 ML20205R0431999-04-13013 April 1999 Forwards Correction to 960212 GL 95-07 180 Day Response. Level 3 Evaluation for Pressure Locking Utilized Analytical Models.Encl Page Has Been Amended to Correct Error 1999-09-23
[Table view] Category:UTILITY TO NRC
MONTHYEARML20064A7131990-09-17017 September 1990 Advises That Due to Reassignment,Jj Clark No Longer Needs to Maintain Senior Reactor Operator Licenses ML20059J2811990-09-14014 September 1990 Forwards List of Key Radiation Monitors Which Will Be Used as Inputs to Top Level Radioactivity Status Bar Re Spds.List Identifies Monitors Which Would Provide Concise & Meaningful Info About Radioactivity During Accidents ML20065D5961990-09-13013 September 1990 Responds to Violations Noted in Insp Repts 50-348/90-19 & 50-364/90-19.Response Withheld ML20059J1661990-09-13013 September 1990 Forwards Monthly Operating Rept for Aug 1990 for Jm Farley Nuclear Plant & Rev 10 to ODCM ML20059L0751990-09-12012 September 1990 Forwards Revised Pages to Rev 3 to, Second 10-Yr Interval Inservice Insp Program for ASME Code Class 1,2 & 3 Components ML20059J2911990-09-12012 September 1990 Forwards Operator Licensing Natl Exam Schedules for FY91 Through FY94,per Generic Ltr 90-07.Requalification Schedules & Estimated Number of Candidates Expected to Participate in Generic Fundamental Exam,Also Encl ML20064A7111990-09-12012 September 1990 Forwards Rev 1 to Relief Request RR-1, Second 10-Yr Interval Inservice Insp Program for ASME Code Class 1,2 & 3 Components ML20059J2891990-09-12012 September 1990 Confirms Rescheduling of Response to Fitness for Duty Program Notice of Violation 90-18-02,per 900907 Telcon ML20065D6621990-09-12012 September 1990 Forwards NPDES Permit AL0024619 Effective 900901.Limits for Temp & Residual Chlorine Appealed & Stayed ML20064A3431990-08-28028 August 1990 Forwards Corrected Insertion Instructions to Rev 8 to Updated FSAR for Jm Farley Nuclear Plant ML20059D4711990-08-22022 August 1990 Forwards Fitness for Duty Performance Data for Jan-June 1990 ML20059B5101990-08-22022 August 1990 Forwards Semiannual Radioactive Effluent Release Rept for Jan-June 1990.No Changes to Process Control Program for First Semiannual Period of 1990 Exists ML20056B2751990-08-20020 August 1990 Forwards Relief Requests from Second 10-yr Interval Inservice Testing Program for Class 1,2 & 3 Pumps & Valves. Request Incorporates Commitments in 891222 Response to Notice of Violation ML20056B2741990-08-20020 August 1990 Forwards Rev 2 to Unit Inservice Testing Program,For Review & Approval.Rev Incorporates Commitments Addressed in Util 891222 Response to Notice of Violation & Other Editorial & Technical Changes ML20058Q1481990-08-15015 August 1990 Forwards Rev 3 to FNP-1-M-043, Jm Farley Nuclear Plant Unit 1 Second 10-Yr Inservice Insp Program,Asme Code Class 1,2 & 3 Components ML20058P6201990-08-15015 August 1990 Forwards Rev 1 to FNP-2-M-068, Ten-Yr Inservice Insp Program for ASME Code Class 1,2 & 3 Components, Per 891207 & 900412 Responses to NRC Request for Addl Info ML20055G7701990-07-18018 July 1990 Updates 900713 Response to NRC Bulletin 90-001, Loss of Fill-Oil in Transmitters Mfg by Rosemount ML20055F7411990-07-11011 July 1990 Forwards Monthly Operating Rept for June 1990 & Corrected Monthly Operating Repts for Nov 1989 Through May 1990.Repts Revised to Correct Typo on Value of Cumulative Number of Hours Reactor Critical ML20055F3781990-07-10010 July 1990 Submits Final Response to Generic Ltr 83-28,Items 4.2.3 & 4.2.4.Util Position That Procedures Currently Utilized by Plant Constitute Acceptable Ongoing Life Testing Program for Reactor Trip Breakers & Components ML20055D4861990-07-0202 July 1990 Requests Authorization to Use Encl ASME Boiler & Pressure Vessel Code Case N-395 Re Laser Welding for Sleeving Process Described by Oct 1990,per 10CFR50.55a,footnote 6 ML20055D1001990-06-26026 June 1990 Responds to Violations Noted in Insp Repts 50-348/90-12 & 50-364/90-12 on 900411-0510.Corrective Actions:Electrolyte Level Raised in Lights Identified by Inspector to Have Low Electrolyte Level ML20044A6191990-06-26026 June 1990 Suppls 900530 Ltr Containing Results of SPDS Audit,Per Suppl 1 to NUREG-0737.One SPDS Console,Located in Control Room,Will Be Modified So That Only SPDS Info Can Be Displayed by Monitor.Console Will Be Reconfigured ML20043G4741990-06-11011 June 1990 Submits Addl Info Re 900219 Worker Respiratory Protection Apparatus Exemption Rev Request.Proposed Exemption Rev Involves Features Located Entirely within Restricted Area as Defined in 10CFR20 ML20043C1851990-05-29029 May 1990 Forwards Proposed Schedules for Submission & Requested Approval of Licensing Items ML20043B5941990-05-25025 May 1990 Provides Rept of Unsatisfactory Performance Testing,Per 10CFR26,App A.Error Caused by Olympus Analyzer Which Allowed Same Barcode to Be Assigned to Two Different Samples. Smithkline Taken Action to Prevent Recurrence of Scan Error ML20042G7461990-05-10010 May 1990 Certifies That Plant Licensed Operator Requalification Program Accredited & Based Upon Sys Approach to Training,Per Generic Ltr 87-07.Program in Effect Since 890109 ML20042F0831990-05-0101 May 1990 Forwards Rev 18 to Security Plan.Rev Withheld ML20042G3081990-04-25025 April 1990 Forwards Alabama Power Co Annual Rept 1989, Unaudited Financial Statements for Quarter Ending 900331 & Cash Flow Projections for 1990 ML20042E4121990-04-12012 April 1990 Provides Addl Info Re Review of Second 10-yr Inservice Insp Program,Per NRC 890803 Request.Relief Request RR-30 Requested Reduced Holding Time for Hydrostatically Testing Steam Generator Secondary Side ML20012E9571990-03-27027 March 1990 Forwards Annual Diesel Generator Reliability Data Rept,Per Tech Spec 6.9.1.12.Rept Provides Number of Tests (Valid or Invalid),Number of Failures for Each Diesel Generator at Plant for 1989 & Info Identified in Reg Guide 1.108 ML20012D9661990-03-22022 March 1990 Forwards Annual ECCS Evaluation Model Changes Rept,Per Revised 10CFR50.46.Info Includes Effect of ECCS Evaluation Model Mods on Peak Cladding Temp Results & Summary of Plant Change Safety Evaluations ML20012D8901990-03-20020 March 1990 Clarifies 891130 Response to Generic Ltr 83-28,Item 2.2.1 Re Use of Q-List at Plant,Per NRC Request.Fnpims Data Base Utilized as Aid for Procurement,Maint,Operations & Daily Planning ML20012C4701990-03-15015 March 1990 Responds to NRC 900201 Ltr Re Emergency Planning Weaknesses Identified in Insp Repts 50-348/89-32 & 50-364/89-21. Corrective Actions:Cited Procedures Revised.Direct Line Network Notification to State Agencies Being Implemented ML20012C6241990-03-14014 March 1990 Informs of Resolution of USI A-47,per Generic Ltr 89-19 ML20012C4651990-03-13013 March 1990 Provides Verification of Nuclear Insurance Reporting Requirements Specified in 10CFR50.54 w(2) ML20012C2051990-03-0505 March 1990 Forwards SPDS Critical Function Status Trees,Per G West Request During 900206 SPDS Audit at Plant.W/O Encl ML20012A1621990-03-0202 March 1990 Forwards Addl Info Inadvertently Omitted from Jul-Dec 1989 Semiannual Radioactive Effluent Release Rept,Including Changes to Process Control Program ML20012A1301990-03-0101 March 1990 Responds to Generic Ltr 90-01 Re Request for Voluntary Participation in NRC Regulatory Impact Survey.Completed Questionnaire Encl ML20043A7481990-02-0202 February 1990 Forwards Util Exam Rept for Licensed Operator Requalification Written Exams on 900131 ML20006D2311990-01-31031 January 1990 Responds to NRC Bulletin 89-003 Re Potential Loss of Required Shutdown Margin During Refueling Operations. Refueling Procedures Will Be Revised to Incorporate Guidance That Will Preclude Inadvertent Loss of Shutdown ML20006A9091990-01-23023 January 1990 Forwards Response to Generic Ltr 89-13 Re Svc Water Sys Problems Affecting safety-related Equipment.Util Has Program to Perform Visual Insps & Cleanings of Plant Svc Water Intake Structure by Means of Scuba Divers ML20005E4931989-12-28028 December 1989 Provides Certification That fitness-for-duty Program Meets 10CFR26 Requirements.Testing Panel & cut-off Levels in Program Listed in Encl ML20005E3681989-12-28028 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-28 & 50-364/89-28 on 891002-06.Corrective Actions:All Piping Preparation for Inservice Insp Work in Containment Stopped & All Participants Assembled to Gather Facts on Incident ML20005E1971989-12-27027 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-22 & 50-364/89-22 on 890911-1010.Corrective Actions:Steam Generator Atmospheric Relief Valve Closed & Core Operations Suspended.Shift Supervisor Involved in Event Counseled ML20011D5041989-12-22022 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-26 & 50-364/89-26.Corrective Actions:Personnel Involved in Preparation of Inservice Test Procedures Counseled. Violation B Re Opening of Pressurizer PORV Denied ML19332F2111989-12-0707 December 1989 Forwards Final Response to NRC 890803 Request for Addl Info Re Review of Updated Inservice Insp Program,Summarizing Results of Addl Reviews & Providing Exam Listing Info ML19332F0791989-12-0707 December 1989 Responds to Violations Noted in Insp Repts 50-348/89-22 & 50-364/89-22.Corrective Actions:All Managers Retrained on Intent of Overtime Procedures & Sys Established to Provide Independent Check of All Time Sheets Each Pay Period ML19332F1141989-12-0707 December 1989 Forwards Description of Instrumentation Sys Selected in Response to Generic Ltr 88-17, Loss of DHR, Per Licensee 890127 Commitment.Hardware Changes Will Be Implemented During Unit 1 Tenth & Unit 2 Seventh Refueling Outages ML19332F1241989-12-0707 December 1989 Forwards Response to NRC 890803 Request for Addl Info Re Review of Second 10-yr Inservice Insp Program,Per 891005 Ltr ML19353B0071989-12-0606 December 1989 Forwards Rev 1 to Safeguards Security Contingency Plan.Rev Withheld 1990-09-17
[Table view] |
Text
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June 6,19 79 ,',
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Docke t Nos . 50-348 ca 50-364 NRC IE Bulletin No. 79-02
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Mr. James P. O'Reilly -
U. S. Nuclear Regulatory Commission Region II 101 Mariet ta S tree t, N. W.
Suite 3100 Atlanta, Georgia 30303
Dear Mr. O'Reilly:
In response to IE Bulletin 79-02, Pipe Support Base Plate Design Using Concrete Expansion Anchor Bolts, dated March 8,1979, Alabama Power Company submits the following response for Farley Units 1 and 2.
Yours very truly,
. . lay n, Jr .
FLCJr/ KAP /mmb Enclosure cc: Mr. R. A. Thomas Mr. G. F. Trowbridge
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., ~ mo987 OPPCIll COPY
Docket No. 50-364 July 6,1979 NRC IE Bulletin No. 79-02 Mr. James P. O'Reilly
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Page bc: Messrs. J . T . Yo ung R. P. Mcdonald H. O. Ihrash
- 0. Batum Don Crowe W. G. Hairston, III K. A. Pikett Ken McCracken T. N. Epps Ron George C. Biddinger J. A. Mooney R. R. Todd J. D. Jones W. B. Shipman A. A. Vizzi
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t This report is in response to I.E. Bulletin No. 79-02 concerning pipe support base plate designs using concrete expansion anchor bolts. In response to this bulletin, Alabama Power has initiated a testing, verification, design review and repair program for concrete anchor bolts to ensure adequacy of installation. The specific responses to the bulletin are provided below:
Response to Item 1:
Originally, flexibility of the base plate was not specifically taken into account in detennining the concrete anchor bolt loads. Alabama Power Company is in the process of performing a design review that takes base plate flexibility into account in determining the concrete anchor bcit loads. This design review is described below.
Grinnell, Southern Company Services, Inc. (SCS) and Bechtel Power Corporation (as appropriate) are utilizing the calculated Westinghouse /Bechtel piping system hanger / seismic restraint design loads and the ICES STRUDL Program to develop design loading conditions (forces and moments) at the centroid of each attachment to the hanger / seismic restraint base plates. For simple cases the forces and moments are obtained by hand calculations. Bechtel then utilizes this information in conjunction with the inspection and test data for analyses of all base plate anchor bolts to determine if the existing t'ase plate anchorage is adequate to meet the design loads with the prescribed safety factor or if corrective action is necessary. This determination is performed in accordance with FNP-1-ETP-123 (a Farley Nuclear Plant Engineering Technical Procedure) which has been reviewed by NRC, I&E Region II Staff.
More specifically, a summary of the evaluation of base plate design by Bechtel is as follows:
- 1. The method of analysis is based on an empirical-analytic technique developed by Bechtel which takes into account design parameters such as flexibility of the base plate and concrete anchor stiffness (based on actual pre-loaded load-displacement curves furnished by the manufacturer). This method has been verified with appropriate finite element analytical solutions. Description of this empirical-analytic technique is provided in Attachment I.
A computer program for the empirical-analytical technique has been implemented for determinino the anchor bolt loads for the majority of applications. For other cases -der to Item 3 below. This program requires plate dimensions, number of .solts, bolt size, bolt spacing, bolt stiffness, the applied forces and tk allowable bolt shear and tension loads as inputs.
TL allowable loads for a given bolt are determined based on the concrete edge distance, bolt spacing, embedment length, shear cone overlapping, manufacturer's ultimate capacity, and safety factor.
The program computes the forces on the bolt and calculates a shear-tension interaction based on allowable loads. An interaction value greater than the allowable is accepted as failure of the bolt (safety factor less than required).
Unit 1 shear-tension interaction analyses are computed utilizing a linear rel a tion. Even though a subsequent squared interaction formula is acceptable and its use has been justified by Bechtel in representing the shear-tension interactiori, Alabama Power has chosen to continue with the use of linear relation ship recognizing that the results from this technique are more conservative.
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The empirical-analytic method does not consider prying action for the follow-
!ng reasons:
- a. Where the anchorage system capacity is governed by the concrete sSear cone, the prying action would result in an application of an external compressive load on the cone and would not affect the anchorage capacity,
- b. Where the bolt pull out determines the anchorage capacity, the additional load carried by the bolt due to the prying action will be self-limiting since the bolt stiffness decreases with increasing load. At higher loads the bolt extensions will be such that the corners of the base plate will
- separate from the concrete and the prying action will be relieved. This phenomena has been found to occur even when the bolt stiffnesses in the finite element analysis were varied from a high to a low value correspondin to both typical initial stiffnesses and to values beyond the allowable design load.
- 2. Calculated boit loads are used to check stresses in tho support base plate to ensure they are less than the allowable stress as specified by the American Institute of Steel Construction (AISC) code.
- 3. For special cases where the design of the support plate does not lend itself to this method, standard engineering analytical techniques with conservative assumptiot.s are being employed.
All anchor bolts within the scope of this program shall be evaluated by Bechtel in a cordance with the bolt acceptance criteria, current "as built" drawings, and the bolt design loads to determine if corrective action is required.
If any bolt on a base plate fails the acceptance criteria described above, one or more of the fnllowing actions are being taken:
- a. Re-analyze the base piate assuming that the bolt is failed (bolts carries zero load).
- o. Re-analyze the base plate incorporating bolt replacement as corrective action.
- c. In those instances where repair corrective actions result in a piping support modification, Bechtel/ Westinghouse (as appropriate) will analyze the effect of such modifications on the analysis of the piping system.
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. Response to Item 2-In general, the currtnt industry approach concerning the use of safety factors for various design loading conditions are described below:
- 1. Factors of safety (i.e. ratio of bolt ultimate capacity to design load) of four for wedge type and shell type anchor bolts, for service (operating) load cases,are used.
- 2. For factored loadings (which include accident / extreme environmental loads) safety factors of 1.2 and 3.0 are used comensurate with the provisions of Section B.7.2 of the Proposed Adcition to Code Requirements for Nuclear Safety Related Concrete Structures ( ACI-349-76) August,1978. The factors of safety are consistent with the ultimate strength design method. A factor of safety of 1.2 is used if the failure mechanisrd for the anchor is controlled by the bolt material . If the failure mechanism is controlled by concrete shear cone action, a factor of safety of 3.0 is used. The utilization of sampling and quality control methods used are integral to selecting the factor of safety of 3.0.
- 3. For general structural design in steel, the AISC Specification has an approxi-mate factor of safety of 1.7 for services loading (for example, column buckling).
For factored accident / extreme environmental loads, a factor of safety of 1.1 is used on nuclear structures for both ductile (yielding) and non-ductile (column buckling) failures. In concrete design for factored loads, a factor of safety of 1.1 is used for flexural and tension action and 1.2 for shear action.
It can be observed that a higher factor of safety is assigned to the expansion
- anchor only if its capacity is governed by the shear cone.
Based on the aboveinteraction of design parameters and on the following addi-tional factors, Alabama Power Company has concluded that a safety factor of 2 is sufficient to ensure operability of Seismic Category I piping system in the event of a seismic event:
- a. 1007, verification testing program with total Quality Control coverage of scoped systems (described in question 4) which minimizes installation uncertainties (e.g. verification of torque, embedment depth, nut engagement, plate configuration, expansion of shell, etc.)
- b. Verification that plates are not overstressed by bolt loadings (e.g. con-sideration of minimum edge distance and proper bolt spacing).
I Response to Item 3:
In the original design of the piping systems Bechtel/ Westinghouse considered deadweight, thermal stresses, seismic loads, and dynamic loads (e.g. certain rapid valve openings and closings) in the generation of the static equivalent pipe support design loads. -
The safety factors used for concrete expansion anchors, installed on supports for safety related piping systems, were not increased for loads which are cyclic in nature. The use of the same safety factor for cyclic and static loads is based on the Fast Flux Test Facility (FFTF) Tests *. The test results indicate:
- 1. The expansion anchors successfully withstood two million cycles of long term fatigue loading at a maximum intensity of 0.20 of the static ultimate ca pa ci ty. When the maximum load intensity was steadily increased beyond the aforementioned value and cycled for 2,000 times at each load step, the observed failure load was about the same as the static ultimate capacity.
- 2. The dynamic load capacity of the expansion anchors, under simulated seismic loading, was about the same as their corresponding static ultimate capacities.
- Drilled - In Expansion Bolts Under Static and Alternating Loads, Report No. BR-5853-C-4, Rev.1, by Bechtel Power Corporation, October 1976.
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I Response to Item 4:
Since existing Q.C. documentation is not adequate to document the installa-
- 7. ion parameters associated with each anchor bolt, the following programs have been undertaken: -
Test Program Alabama Power Company initiated a program to randomly select and test a sample of anchor bolts installed in Seismic Category I, Safety Related, 2h inch and greater piping systems. Initial results of that program revealed that statistical sampling would not be sufficient to provide a 95% confidence level in anchor bolt reliability.
As a result, the anchor bolt testing program was expanded to include 100% verifi-cation of anchorages associated with pipe hangers for those systems or portions of systems required to meet design basis accidents and those required to bring the plant to cold shutdown condition. These piping systems included in the program are:
- a. Seismic Category I; Safety Related 2h inches and above.
- b. Seismic Category I; Safety Related ASME Section III, Class 1 piping, under 25 inch.
- c. Seismic Category I; Safety Related of other classes for which the designer performed detailed analysis,
- d. All piping through containment penetrations.
The scope of this program given above has been reviewed and approved by the NRC I&E Region II Staff.
The specific systems involved in this testing program are listed in LER 79-21/0lT Anchor bolts on hangers within the scope of this program are tested for the following parameters:
- a. embedment - Actual embedment depth is determined.
- b. grout - The presence of grout and levelin nuts is determined to ensure proper torque test.
- c. type of bolts - Verification is made that installed bolts are in accordance with design bill of material.
- d. number of bolts - Verification is made that the installed number of tolts is in accordance with design bill of enerial.
- e. bolt dimensional measurements - Dimensional measurements are taken to determine the degree of compliance with the manufacturers' recomended bolt installation requirements.
- f. torque - Bolts are torqued to a level such that the resultant tensile load on the anchor is equal to k of the manufacturers' published pull-out load. For sc. ell type bolt torque tests to be considered valid, the shell shoulder must nc t touch the base plate.
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NOTE: A torque / tension relationship r s developed for Hilti wedge type anchors
( based on tests performed at Farley. Torque / tension relationships were developed for Phillips shell type anchors under the direction of Bechtel Corporatic.r vWh tech.ical censultation from ITT-Phillips Drill Division at Plant Hatch. Since these relationships were completed and the majority of anchor bolt field veri-fication was performed prior to I&E Bulletin 79-02 Revision 1 issuance, no site specific testing for the shell type anchors was performed. Torque requirements for Wej-it wedge type anchors were obtained from vendor data.
9 base plate dimensional measurements - Dimensional measurements of base plate parameters which could affect bolt loading or capacity (e.g. bolt spacing, edge distance) are taken.
Based on the results of the test program and the empirical-analytic evaluation, anchors are being repaired according to the following criteria:
- i. Repair individual base plate anchorages not having a safety factor of at least 2.0.
ii. Repairs are done so that all repaired bolts have a safety factor of at least 4.0 and all base plate anchorages have a safety factor s ' at least 2.0.
iii. All repairs are done in accordance with written procedures and quality control checks.
The failure to test inaccessible anchor bolts will be justified by analysis which substantiates operability of the affected systams without assuming integrity of the anchorages which are not tested.
Preloading Available test data indicates that it is not necessary that the bolt preload should be equal to or greater than the bolt design load because pipe supports and anchors are subjected to both static and dynamic loads. The dynamic loads such as seismic loads are short duration cyclic loads and are not fatigue type loads, therefore the amount of preload on the bolts will not greatly affect the perform-ence of the anchorage. The initial installation torque on the bolt accomplishes the purpose of setting the anchor, but the ultimate capacity of the bolt is not affected by the amount of preload present in the bolt at the time of cyclic loading. For vibratory loads, the expansion anchors have successfully withstood long term fatigue conditions as discussed in the previous section (FFTF tests).
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Response to Item 5:
The Alabama Power Company testing, analysis, and repair program will not be completed by July 6,1979; however, Farley Nuclear Plant Unit 1 is currently shutdown during the present critical power deraand period to complete the above crogram. The testing, analysis and repair program described in Item 4 will be completed prior to return to power generation.
Documentation of the program will be maintained on site.
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Response to Item 6:
A similar program for the verification of Unit 2 anchorages will be developed as the result of experience gained from Unit 1 activities.
A full description of this program will be transmitted to NRC oy'a supp'ement to this bulletin response. Such verification program will be completed prior to initial criticality. Currently, the construction activities associated with Unit 2 are temporarily suspended due to the Company's financial condition.
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ATTAGMENT I 4
DETERMINATION OF EXPANSION ANGOR BOLT LOADS IN PIPE SUPPORT BASE PLATES
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Summarl This report deals with the determination of anchor bolt loads in steel base plates supporting Soismic Category I piping systema.
The anchora in question are of the expansion type. The loads are applied to the base plate through some type of httachme~nts, usually concentric with the base plate, and could comprise of moments and foreca in three directions. A review of the typical base plates used in supporting the subject piping systems indicate that the ms.iority of them have either a 4,'6 or 8 bolt connection. The plate tiicknesses usually vary from 1/2" to 1 l/ f and are not generally stif fened. The present fornu-lation will, therefore, be devoted to base plate anchorage systems with afore-nentioned physical characteristics.
From a purely analytical standpoint the load distribution in a base plate anchorage system is f airly compicx and it is necessary, therefore, that certain simplifying sasmptions be made to arrive at conservative yet practical solu-tions. However, such assutnptions should take into consideration the following parameters which might aff ect the load distribution in the anchorage system.
- a. Plexibility of the base plates considering the bending effects.
- b. Eolt stiffness: to be based on actual preloaded load displacement curves as furni,shed by the manufacturer.
- e. Prying a(tion .
For expansion anchor bolts prying act. ion will not be critical for,the following reasons: .
- a. Where the anchorage system capacity is governed by the concrete shear cone, the prying action would result in an application of an external cocipressive load on the cone and would not therefore af fect the anchorage capacity,
- b. Where the bolt pull out determines the anchorage capacity, the additional load carried by the. bolt due to the prying action will be self-liciting since the bolt stif fness decreases with increasing load. At higher loads the bolt extension will be such that the corners of the bsse plate will lift off and the prying action vill be relieved. This pher.onena has been found to occur when the bolt stiffnesses in the Finite Element Analysia were varied from a high to a low valus.
Method of _ Analysis for Anchor Bolt 1.ceds_:
In gc -ral, the Finite Element Method of Analysis may be used to analyze the base plates under consideration. However, such an approach will be both time consuming and expensive considering the number of base plates involved. A quasi analytical approach has been formulated taking into account the base plate flexibility and the bolt stif fness. The results of the analytical solu-
' tion have been verified with appropriate Finite Elsaent solutions and have shown good correlation for the typical cases studiesl, ,
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INTRODUCTION:
THE PURPOSE OF THis STUDY WAS TO DEVELOP AN ANALYTICAL u METHOD FOR DETERMINING TENSION LOADS ON EXPANSION ANCHURS USED AS ANCHORS FOR PIPE SUPPORT BASF. PL AT ES. .
FINITE ELEMENT ANALYSES (REF-1) SERVED AS A DATA B~ASE FOR DEVELOPING LESS EXPENSIVE. AND LESS TIME CONSUMING AN ALYTIC AL, METHODS. THE METHOD WHICH 15 PRESENTED AS A RESULT OF THIS ' STUDY USES PLATE FLEXIBILITY AND BOLT STIFFNESS AS THE PRIMARY PARAMETERS. THIS METHOD WILL BE COMPUTERIZED FOR 4,648-BOLT PA7 T E R NS.
__ ANALYSIS :
IN THE QUASI ANALYTICAL MODEL PRESENTED HERE.T14E PLATE :
15 PRIMARILY TRE ATED AS A BE AM ON ELASTIC SPRINGS.
BASE PL ATES WITH THREE DIFFERENT BOLT CONFIGUR ATIONS HAVE BEEN CONSIDERED.
. ASSUMPTIONS-(o) SYMMETRICAL BOLT PATTERNS (b) CENTRO!DAL LOADING (c) ATTACHMENT DIMENSIONS SMALL COMPARED TO THE PLATE DIMEN SIONS (d) UNITS FOR ALL VARIABLES:
FORCES KIPS LENGTH INCHES l l
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j (3) 4-BOLT PATTERN - MOMENT AND TENSION LOADING CASES t-j GIVE.N A PLA7E WITH A 4-BOLT PATTERN AND A MOMENT l
ABOUT ONE AXISI THIS PLATE WILL BE MODELED AS A '
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COMPRESSIVE STRESS j BLOCK (KIP) d 1 c T (x) c(x) = M i, a
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THE BE AM WILL BE IDE ALIZED AS BEING SUPPORTED AT THE LOCATION OF THE COMPRE6SIVE FORCE RESULTANT. THEREFORE, IF THE COMPRESSION CENTROID CAN BE LOCATED,TBECOMES ..
KNOWW AND"T' CAN BE CALCULATED.
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FOR A 4-BOLT PATTERtJ LOADED CENTROIDALLY
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CQMPRESSION (IN.)
t * '3L AT E 7HICKNESS (IN.)
d= 0lST ANCE FROM EDGE OF ATTACHMENT TO THE EDGE OF THE PL ATE (IN) r7 EsOLT SilFFilESS (K/IN.)
btSED Cli A NUMBER OF FINITE ELEMENT ANALYSIS RESULTS (u . V A f,YING T,d a K3 ), THE FOLLOWING EMPIRIC AL REL ATIONSHIP IYAL DE R!VE D :
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L: S.S [(h)$ (%3)](d)
WHERE L&d ONCE L IS CALCULATED, TOTAL TENSION (T) AND BOLT LOAD (f i) CAN BE FOUND:
M (2) l'h 4 h 4L M
(3)
F1 'I '54b42L 2 FOR CENTROIDALLY LOADED 4-BOLT PATTERNS ONLY TH!S METHOD CAN BE EXTRAPOLATED FOR USE WITH COMEINED LOADING CASES.
e
*'7 i 14
(
I FOR BIAXIAL BENDING:
~
D (4)
CRITICAL F) = g,g g, + 4 + 4 2W F; } g ^nf 7 o C a b J l@J J U[/b ,
FORIDJsB111ED_ BENDING AND TENSION:
M T CRITICAL Fe g,n
SINCE L VARIES WITH t,d r4 K, THE METHOD FOR FINDING L C AN BE USED FOR MANY PLATE AND BOLT PATTERNS. ONCE L 15 KNOWN
' THE PLATE CAN BE MODELED AS A BE AM ON SPRINGS. THE BE AM CAN BE SOLVED BY VARIOUS METHODS AND THE TOTAL T EN SION FORCE FOR AN.Y ROW OF BOLT 5 CAN BE CALCUL ATED.
THIS WILL BE DEMONSTRATED FOR SIX AND EIGHT ' BOLT PAlTERNS IN THE FOLLOWING DETAILS.
Q) B-BOLT PATT ERN - MOMENT LOADING CASE _
- .. .. N. -. w
+ + + BOLT ROW 'A*
[
m Y
+ -
+ BOLT ROW "B' f
m.
._E f ~ +< +e +h BOLT ROW *C" .
~
~
n-
8 j
(
i 2
- J /.1/ M G 'O E L :
j
~
E j
9IU)l@l7 ]p.
j
~
M .L Ofu J (, uj g' g( gg {,
9i i
_ . _ . _ . - .C- '
A,j B
-i
? K, fK 2 e m D7
! (COMPRESSION CENTROID ;
9
. . . _ . . _5 I
ib .
Kas BOLT STIFFNESS j I
12 i
i T Hi. RL A'.TIONS FOR THIS INDETERMINATE BEAM MODEL CAN BE S01 VCD USING VIRT UAL WORK PRINCIPLE. THE FOLLOWING _
E QllATIONS WERE DERIVED FOR B-BOLT PATTERNS:
WHERE L IS DETERM!NED FROM EQ (I) l E$ k+L EIi 411 W t * (KIP IN*)
11 R'.DUNDANTS ARE TAKEN AT Y:
El 6c,= E k Krh g, .
3 .._QES 3
(6) t 1
WHERE 6ce, IS THE DEFLECTION AT 'C' DUE ONLY TO 'M":
- i EI + yE [1+ S] (7)
EI Scc
- s'N,Kg [. k S + 2 K,2 S * (K;* kg)E',,
i f
WHE RE Scc 15 THE DEFl.EETlDN DUE TO A l* FORCE APPLIED AT
'* (B)
REA Tl0N AT C = Rc = -
.'. R 4= -[M -2 (Re)] ; R g = Re- R A q 3- '
J
-o 3 (
l-s ..
.' (
At 1Hl. PLAT E GETS WIDER AND E BECOMES SMALL COMPARED T(. Y, T HE TWO MIDDLE BOLTS CANNOT BE LUMPED TOGETHER Ai ONE SUPPORT WITH K2= 2Kg. Kz WILL BE SOMETHING LESS T H/,N 2Kg. THE FOLLOWING EXPRESSION FOR K2 YlELDED Rl"Sul.TS WHICH WERE IN GOOD AGREEMENT WITH FEM RESULTS: ,
(9)
K=2Kg(})*6.2Kg 2
FC' PL ATE SIZES GENERALLY USED IN PIPE SUPPORTS, THIS Wil TH EFFEET WILL HAVE NEGLIGlBLE EFFECT ON ROW "A' l.e.
THL !.Til FNESSES OF THE AREE BOLTS CAN STILL BE LUl6[i TOGE THER IN THE BEAM MODEL.
l Hi. fii3 (. T IC NS IN THE BEAM MODEL ARE NOW kNOWN. THE RI: AC TION AT ANY ONE SUPPORT IS THE TOTAL TENSION IN THAT F OW OF BOLTS. TO DISTRIBUTE THE LOAD TO THE 80l'S:
F C:8 R06 "B' FROM SYMMETRY, Tl N510h PER BOLT = TF ;FT e
FCS R3W "A",THE RELATIVE STIFFNESS OF THE PL ATE AND 1 H i. BCLTS AND THE BOLT DIST ANCE FROM THE All A.C HME NT WILL AFFECT THE LOAD DlSTRIBUTION BET WEEN THE MIDDLE AND THE CORNER BOLTS.
I EVIDENT LY THE BOLT CLOSEST TO THE ATTACHMENT WILL CARFsY MORE LOAD AND IF THE ATTACHMENT SIZE IS SMALL, BOLT TO THE ATTACHMENT DISTANCE MAY BE SUBSTITUTED BY THE DISTANCE OF THE BOLT TO THI, CENTER LINE OF THE 'PL ATE. THUS TENSION IN THI lDDLE BOLT '9:
(11)
Fu q (R4) h E. bN* EN_ he_
WHERE: Lm= DISTANCE FROM PLATE CENTER TO BOLT 9 '
Le s DISTANCE FROM PLATE CENTER TO BOLTSiin*c'
$$[3) d W D D d[i A=5*E i
e(. CONSTANT p
[> ib: aoj ta r ,- ,
cq w UO ] 7]L Ub J
. l .". 3. . .. . . . . . . . . ,
10
(
M eel, ON SEVERAL FEM ANALYSES THE FOLLOWING EXPRESSION OF F 1 t. Vs'A5 . ARRIVED AT: . , -
bA} (
IT E . A ( AR ) ' bs y t
Ps'ITil T HE LIMITS 0.333 ch 41.0 CORRESPONDING T O VER.Y RIGID AND VERY FLEXIBLE PLATES.
T l'.N E lCI ' IN THE CORNER BOLTS IS GIVEN BY :
F3 ,s f h
At;: Fy Tg:Fih i *O M F C I. E :/- Y!AL E>ENDING , THE RESULT ANT BOLT FORCES WILL EI DETI RMINED BY SUPERPOSITION.
T (E j .f. Cj.,7 P_ATTERN- MOMENT LOADING C ASE L X
o
+ + +- T -
+ + + o
' i 5y
. b, SY .. - ._ , ,_
BY p.. . _ _ _ _ _ . _ _
THE 6-BOLT PATTERN CAN BE SOLVED BY USING A COMBIN ATION -
~
01 THE'. EQUATIONS FOR 4 BOLT AND 8-BOLT PATTERNS. .
onTn Maj o
To~
W '3 o
1 g s) 70 a J1) d.]
_ _ iL l-10
11
(
FOR MOMENT ABOUT THE X-X Axis:
(A) USE ' EQUATIONS (1) AND (2) TO 60LVE FOR TOTAL TENSION .
(B) USE THE B BOLT DISTRIBUTION EQUATIONS (12) AND (13)
FOR SOLVING THE BOLT LOADS h7TH 2,=y + E 4 EI 2417 B yt';
FOR t[OMENT ABOUT THE Y-Y AXIS:
(A) USE EQUATIONS (6),(7) AND (B) TO SOLVE FOR REACTIONS 8
WIT H Kr
- 2 6K (j)'; S= Sy ; Ya y ; EI 241T 6xt (B) D! VIDE THE RE ACT!QNS CORRESPONDING TO EACH BOLT ROW BY 2 TO OBTAIN INDIVIDUAL BOLT LOADS.
(D) 6 AND_8 BOLT PAT TERNS , TENSION LfgD8NG C ASES:
UNLikE THE 4-EOLT PATTERN, FOR THE 6 r,6 BOLT CASES THE EENTRALLY APPLIED TENSION CANNOT BE DISTRIBUTED E QU At.t.Y 10 ALL THE BOLTS DUE TO THE INTERPLAY OF BOLT AND PL ATE STIF F NESS E S AND THE REL ATIVE DISTANCES OF THE BOLT 5 FROM THE POINT OF APPLICATION OF THE LOAD.
BASED ON THE MOMENT CASE IT WILL BE ASSUMED THAT THE PARAMETRIC VARI A'3LES AFFECTING T HE LOAD DISTRIBUTION WILL BE OF THE SAME FORM AS IN THE MOMENT EASE. THE CONSTANT $ FOR THE DISTRIBUTION FACTORS DFM, AND DFMy WAS OBTAINED FROM FINITE ELEMENT ANALYSIS RESULTS.
I e -
r3 "i l-11
12
,(
8-BOLT PATTERNS- TENSION LOADING C ASE:
L ! l
' '?[FlQ['i/?[*
Ul. . ! d O uu ird b-j'
+- -+* $ 4 .
Y
%w -
)
f
-4 0 z
,-,-x x
\ T,/
v>
j T TENSION LOAD U
1 -g 40 4h Fi= LO AD PER BOLT i ~ '
i i CALCUL ATE:
8 Sy Sy EIr 2417 8xt EIz 2417 Byt 8 0
.......Y._._...
. M. EI, 8'
25y Ky,EI 2Sx gr T*' T ; Ty Mx
,k3 4ky ,
Le 5 (S 3)* * (Sy)* 4 ,
k(8 } 6 DFMx & I,00 DFMx= { I 7 EI
- -- 1 Lc-1 e Ks(25,)r 4 3
,g DFMy=3 EI 2 I 7 6 DF M y 41.00 g 4 Lg K
NOTE: FOR PLATE STIFFNESS VARY 1NG FROM INFINITELY RIGlD TO EXTREMELY FLEXIBLE: i e
$ 6 DFM4i SINCE A ' RIGID' PL ATE. DOES NOT EXIST, $ 7 WAS USED AS A LIMIT (
l-l t
13
. (
i I FgsFTe ' _DIMY ,
;t ]{ (\ F Ljb 'b F7 a = F~ t e = [DFMx]_T f[ ,.
FT . FTe} FTt FTh' 4 .
=
IT BY AGOVE EQUATIONS7 F pFr OR F Tb ' f T. . SET F tp F7. OR Fw.Fw AS LIMITING VALUES FOR RECTANGULAR PLATES 6 BC)l.T PATTERN-TENSION LOADING CASE :
y
-+" $ +' k 4 .
l . ._ 'rx a
w g- #
[ d EI,c 2417 8xt' 4 4e 4 EI 2= 2417 6yt'
.} . _____.__....___.J_.._..______.._ ,
. ... 5.Y.
SY U K=Q 3
By
. ...._... ... .._ .-...... ..._.._ .
- EI; 5x IT* .dY. T DFMy= { 8 (S *
, f3 ,4 2ig_
akAND&l.OO
_ E I r ,,
8 W HF. RE Le= A)*
z d8J3 )'
F Tb" FT e = [DFMy][T 4]
'(
Fro Fc-FTd 7
- fif
- 4 BASED ON THE ABOVE EQUATION, IF FTa(= FTc ' F Td: Ip)>FT T b (* I Te),
AS MAY BE THE CASE WHERE 5xz 25y, THEN F oT aFTe=Fi g.FTP- ,.
FTb Fte = {
G
~O l-13
(- (IS$ COMPARISON _OF RE SULTS:
FINITE- E LEME NT METHOD VS BECHTEL MODEL >
5 SKETCNES OF BASE PL ATES ANALYSED: ,
+
(A)'4 - B.O_LT PAT T E R N -)
D
' ^
U Ib'-l ll i /] I y, l) .
1
. JiiUby/[j!
.' + + _ - _
Y J :
7
- .*3; , .
ul
/ -
z/
i 8
4,,
. + , I 1 N Il 2' - 12". . - .. . . . . w 2'
- =
Il t ks LOADING l b' 41_ _ _ f,4 > i l 8 k " .
p A/ _ 44 Mi 16kildvux~
3 h* 44 Ma'IBk'.Fa*4M' 4 7# 44 Mr lB K' 5 fi l50 tAs=l8n" 6 I'/ 300 Mr lBk' Kg BOLT STIFFNESS (k/IN)
- t. PL ATE THICKNESS 3
.; o -
k
(~li
(-
_4_ .8' ___ _ 14 _4 ~____ - ..w.dd
. -l- y !-
Q g- [D')Il ff0 "i i) h. I j .y b
2 /t . . _ . ,
e i
+ 14 + ,
L_. wt
& t, Kg LOADING
' M y 217.Si' Ks= BOLT STif fliE55(k/Ild
-f V 44 44
~~
M v : 2 47.5 x" t = PLATE THICKt4ESS B 2" '~
9 k"~
44
~
$Nh'
~ ~ '
FROM TELEDYNE ENGINEERING REPORT (REFERENCE 0) .__
(b)(,-BOLT PATTERN:
_.31 ~
Sy __ w ___SY _-
t Kg= BOLT STif f NE SS(K/Ill)
-J
, .g 5 .. t= PL ATE THICKNESS
/- ~
o n-
.g s ,I LLJ +
y
+ + o Sy In 3Y LOADiffG lt t Ks Sr .'
- 1 I 44 12 8 16 _
f6 20 20 Ma*36"'
Mr: 3 6 "'_
(40 12 8 2 _l* 4 25.5 12 Fz.=10" 3 l' 44 22.5 4 25.5 12 fr.io" 2* 44 22.5 4
16 fr l0" s
5 T~ 44 ,_
l2 _ __6 16 16 F = ct a ' , __ ,
G 6 l' 44 P _6 16
, .[. er . . _ _ . _
U i
" ' iJ (C) B- BOLT PAT T E R N r,p Ep :
BY
- hLUlijUUu'//jlllglIh'lO J
. - =- -
Sy g Sy ___
, i
. +o +b +c g i i
~
i Y :
J Wi I j u .
Yd /
# Ye cI p l z# :
y b--m.l g.
I
't i- + +- -
i 1
& t kg 53 sy Bx BY b I.0ADING l IV 44 12 12 28 28 6 M x = 180 "
._R IV 440 12 12 2B 2B [~1,s l 8p' 3 I" 300 8 8_ 20 20 !4 M x t 90"'
_[2.,_,, 28 2B ; 6_ F _i 16 5 4 -. 81. 15 0_ 12 .
5 _lS' 44 12 _ l2 28 28 6 F1 = 8" i
,,b., , _ l ' _d4 i 6 10 1 16 24 f .10" Ke BOLT STIFFNESS (KIP /IN) t PL ATE THICKt4E55 i <
b I
TAbt!! t.TE D RESULT S:
4-E011 P. alt E RN :
U(
5t ' '
~
. LOAD PER BOLT (M)
~" ANALYS65 BECHTEL I*
L1koD FINITE ANALYTICAL l E LE ME.NT MODEL DlF F E ret'CE PLATE I A O.15 0.75 0
.._- (1)
A (2) 2.08 2.25 + 8. 2 1.75 + P. 3 l j A (3) i.71 o.6 B < 6.3
[_A (4) 0.64 _
A (5) 0.75 0.78 44.0 i
i A (6) 0.78 0.B 4
~
9.19 + 0. 8
!A (1) 9.12 6.12 6.45 + 5. 4 I A. iid 49.4 A (9) 16.61 1 18.17 _
(.- 6011_ PAT T E RN :
o TENSILE LOAD PER BOLT (k) f bcLT 6 6oLT BOLT 5 BOLT DIFFERENCE a t. c J og y
._\ AdT6'5 FINITE BECHTEL BOLTS BOLT MLW ' AN ALT T tC AL a LC b ELEMENT Moott PLATE N B (1) o.65 f.8 4 0.64 1.72 - 1. 5 - 6. 5 (2) 0.61 1.96 0.72 1.86 +18.0 - 5.1 B '
B (3) 1.6B l.64 1.61 f.67 - 0.7
- 1.5 B (4) 1.67 IA6 1.67 1.67 0 + 0.2 1.55 1.B9 1.67 l.67 + 7. 2 -13.5 -
B (5) ' "
1.45 1.59 1.5 1.5 + 3.2 - 6.1 B (6)
,f- "Y
.. M1 _.-..._.......:.._
DPTD 2
!a S [L
/ '
fiD ~lj: Fn] 'li V /i'~'l r 1
r Ec.L1 PAT't E Rt4 : ' ,
O J b NU //O _,
s
.. ~
; TEN 51LE LOAO PER BOLT (H) p
'Goti ~ E~o'ET T b^r Ct' 3 6CT loTT-J.-} bogT DIF F E RE NC E A.... . . .. .s . _ _ . . . _ .
' . .4 bECHTEL BOLT BOLT BOLT
M. ., NE . - . . . .
FINIT E g g Atl ALYTIC AL ,
ELEMEf4T MODEL i i F i f.l ENN; + 2.69 + 2.5 + 17.0
; i.59 2.64 ' O.75 1.94 2.70 oj2 ,
! (. (I) 1.47 - 2.3 + 0.7
$.26 f.46 1.56 5.14 11.9 C (2) ..-__ : 1.55 - 3.0 3.23 O.85 + B.2 - 2.6 3.32 o.88 - l. 32 j,( ( 3) ,_, ,,
I.22 _
2.92 1.46 0 0 0_
C (4) 1.08_ 2.92 ... . 1,46 1.08 ...
-3.5
+ 3. .6 . . . -. 2.6
....._. 0.57
- 0.59 0.86 1.14 . . . . . . . . , -
0.83.. . 1.17. . . . _ _ _ .. . . _ _ .
C . (. 5_) .
- 3.1 + 4.4 - 5.2 0.99 1.95 1.06 0.96 2.04 1.01 C (6) 1 e i 1
- i
./
-4 i Th e'
\-t s . .......................... .
. PE
$[ @
L tu! ,. ._a M
O lei O
[P
#" '#j [[ "[ '
REFETENC S
~
Ng
-- f 1,ANSYS" ENGINEERING ANALYSIS SYSTEM, DEVELOPED f3 ifj BY SWANSON ANALYSIS SYSTEM,lNC. %
- 2. DILUNA, L.J. AND FL AHERTY, J. A.,"AN_ ASSE SSMENT jj OF THE AFFECT OF PL ATE FLEXIBILITY ON THE j}
DE SIGN OF MOMENT- RE SIST ANT BASE PLATES';
'j TELEDYNE ENGINEERING SERVICE S (SU3MITTED ;j TO ASME FOR PUBLIC ATION) b i
*i
- l4 e
9 I }
n e - .
o g
I-l9
..}}